2011 Research Portfolio / All Nuclear - EPRImydocs.epri.com/docs/Portfolio/PDF/2011_Portfolio...Electric Power Research Institute 2011 Research Portfolio being implemented into new

Nuclear

Developing safe, reliable, economical, and environmentally responsible technologies that enable the long-term operation of existing nuclear plants and the

deployment of advanced nuclear power plants

Advanced Nuclear Technology

Chemistry, Low-Level Waste and Radiation Management

Equipment Reliability

Fuel Reliability

Instrumentation and Control

Long-Term Operations

Materials Degradation/Aging

Nondestructive Evaluation and Material Characterization

Risk and Safety Management

Used Fuel and High-Level Waste Management


Incorporate plant operating experience and research results into tools and guidance that increase confidence and reduce risks associated with

deploying new nuclear power plants.

Program Advanced Nuclear Technology


Program Overview

Program Description New nuclear power plants incorporating advanced light water reactor (ALWR) technology must overcome a number of regulatory, economic, technical, and social challenges prior to licensing, construction, and successful startup. Many of these challenges can be addressed through application of focused Electric Power Research Institute (EPRI) technical products and targeted deployment tools that minimize deployment risks. The EPRI Advanced Nuclear Technology (ANT) Program complements—and helps accelerate—industry activities aimed at enabling and building confidence in new nuclear plant deployment through coordinated work on cross-cutting issues. By building upon past industry operating experience and previous research and development (R&D) results, new nuclear plants can realize multiple benefits. These benefits include improved designs for long-term operation, implementation of optimized fabrication and construction practices, and more effective overall deployment of an inherently high-risk project.

Research Value Research results from the ANT Program increase confidence and reduce risks associated with developing next-generation nuclear plant designs by incorporating current plant operating experience and results from focused research and development. The research addresses the issues that could impact the ability to license, construct, start up, and efficiently operate advanced light water reactors worldwide. Advanced Nuclear Technology participants gain access to the following: Materials Management Matrices for advanced nuclear plant designs that can serve as living documents

for managing life-cycle material issues. New Plant Deployment Program Model, which enables users to systematically analyze various licensing,

technical, and plant deployment issues, thereby improving the decision-making process. Equipment reliability knowledge and tools that can be incorporated into new plant designs to increase the

potential for meeting business goals related to such issues as plant availability and capacity factors. Refined methodologies for applying risk-informed pre-service and in-service inspection techniques to

advanced nuclear plant designs. Research supporting design and demonstration of next-generation nuclear plants, including high-

temperature gas reactors for process heat, hydrogen production, and cogeneration applications.

Approach The Advanced Nuclear Technology Program heavily engages the utility and vendor communities to collaboratively identify and overcome the technical challenges confronting new plant deployment. The Program leverages EPRI expertise in various disciplines to resolve common issues. The ANT Program focuses on four core elements: Facilitating standardization across the new fleet Nuclear plant developers around the world are working to ensure standardization is factored into all aspects of new project development. However, while the designs may be standardized, many of the supporting systems will not be, including startup testing, balance-of-plant components, equipment reliability programs, operational procedures, and configuration management procedures. Transferring technology to new plant designs Lessons learned from existing plants and from EPRI’s 30+ years of research and development results are being incorporated into new plant designs to drive overall improved performance. Technology advances and lessons learned in materials, chemistry, equipment reliability, nondestructive evaluation (NDE), and fuel performance are

Electric Power Research Institute 2011 Research Portfolio

being implemented into new plant designs. EPRI will continue reviewing available information with subject matter experts, designers, and utility representatives to define and prioritize requirements, guidelines, and assessments. Ensuring top plant performance from start of operations Nuclear plant performance is a balancing act of equipment selection, material selection, design, operation, maintenance, management, and many other factors. Current financial models for evaluating new nuclear power plants are based on availability factors reflecting the fleet of existing nuclear plants. EPRI’s ANT Program provides guidance allowing utilities to maintain high availability factors from new plant startup. Reducing the overall deployment risk and uncertainty Constructing, starting up, and working through initial operations of new nuclear power plants present many large, first-of-a-kind challenges. These challenges establish a deployment risk and uncertainty that affects the ability of utilities to get plants sited, approved, financed, and licensed. The ANT Program will provide research to decrease deployment risk.

Accomplishments EPRI’s ANT Program helps accelerate industry activities aimed at enabling and building confidence in new nuclear plant deployment. Recent accomplishments include the following: Secured membership from 21 U.S. and international utilities and critical nuclear industry vendors. Completed materials management matrices for four of the six advanced nuclear designs: Westinghouse

AP1000, GE-Hitachi ESBWR and ABWR, and Toshiba ABWR. These tools assist industry in identifying and considering materials issues and mitigation and management opportunities through design, component fabrication, plant construction, and initial operations and maintenance.

Evaluated EPRI's Fuel Reliability Guidelines in the context of their applicability to new plant designs. Determined that utilities should apply the recommendations in nearly identical fashion, potentially with some unique considerations for each design.

Benchmarked three companies to investigate the methods used to perform tests and inspections on construction modules. Benchmarking results will help the nuclear industry define and deploy best practices as use of modules increases for new plant construction.

Summarized industry efforts to capture equipment reliability lessons learned during the design phase of a new plant project. The report compiles non-mandatory recommendations that reflect industry best practices.

Identified critical welding and fabrication attributes for specific materials, assessed their effects on potential degradation mechanisms, and identified welding and fabrication process enhancements that can improve long-term asset management of new nuclear plant components.

Initiated the following: risk management projects focused in critical areas such as proactive materials degradation management, a change to the American Society of Mechanical Engineers (ASME) code to assist in materials performance, and industry leadership on the seismic source characterization database for nuclear facilities.

Current Year Activities Advanced Nuclear Technology Program research and development for 2011 will continue its focus on proactive, risk mitigation/management projects for new plants, while expanding the program scope to include construction and startup activities. Specific efforts will include the following: Complete Materials Management Matrices for all six of the new plant designs, and address common

issues identified from various projects Continue ASME code case development and technical justification to allow for fitness-for-purpose under

ASME Section III for Pre-Service Inspection (PSI)

Advanced Nuclear Technology - Program 41.08.01 p. 2


Evaluate the applicability of EPRI's Water Chemistry Guidelines to each of the new plant designs Continue expanding international participation to increase collaboration with those utilities and vendors

currently constructing new nuclear plants worldwide Selected ANT program activities may be conducted in whole or in part in accordance with Title 10, Code of Federal Regulations, Part 50, (10CFR50), Appendix B, and may invoke 10CFR21 at the discretion of ANT member utilities or EPRI, when such action is deemed appropriate.

Estimated 2011 Program Funding $6.5 million

Program Manager Tom Mulford, 650-855-2298, [email protected]

Summary of Projects

Project Number Project Title Description

P41.08.01.01 New Nuclear Plant Materials (supplemental)

Materials management matrices will be produced for each of the advanced light water reactor designs, including ESBWR (completed), AP1000 (completed), ABWR (completed), EPR, APWR, and APR1400. These matrices provide critical materials-related information for each of the components in the nuclear steam supply system and become living documents for managing life-cycle material issues. This project also identifies best practices for welding plant components and for fabricating new plants.

P41.08.01.02 New Nuclear Plant Reliability (supplemental)

EPRI has a substantial portfolio of equipment reliability knowledge that has been effectively deployed across much of the current industry. This project helps deploy existing equipment reliability tools into new plant designs to support high performance and increase the potential for meeting business goals.

P41.08.01.03 New Nuclear Plant NDE (supplemental)

EPRI is testing and refining a risk-informed in-service inspection methodology for advanced plants. The goal is to develop a single pre-service/in-service inspection program for each design (not site-specific) or provide guidance where there may be site-specific requirements. EPRI also is working with ASME to provide the technical basis to accept benign welding flaws by structural analysis. This project will establish and operate the ultrasonic (UT) examination qualification programs necessary to assure accurate, reliable construction inspections of primary pressure boundary components and welds.

P41.08.01.04 Security and Seismic (supplemental)

EPRI is updating a seismic source characterization model for the Central and Eastern United States (CEUS) that will facilitate new plant siting and respond to regulatory concerns. An updated generic CEUS seismic source characterization model will benefit those companies pursuing new nuclear plant development and those nuclear power plant owners that must respond to issues resulting from Nuclear Regulatory Commission (NRC) Generic Issue 199 (GI-199).





P41.08.01.05 Achieving Virtual Power Plant Configuration Management (supplemental)

EPRI is developing a Standard Configuration Management Reference Model and Implementing Guideline to provide a common framework to help manage an automated, integrated, and interoperable configuration management program in a consistent and connected way. The model will be an XML toolkit consisting of configuration management relationship taxonomy, supporting schemas, and supporting Design and Licensing Basis Rule Set. The Implementing Guideline will be a best practices document providing “how to” details for the use and implementation of the Standard Configuration Management Reference Model.

P41.08.01.06 Advanced Nuclear Plant Research (base)

This project supports the development of advanced nuclear plants, including small modular reactor designs and high-temperature gas reactors

New Nuclear Plant Materials (supplemental) (066891)

Key Research Question Material degradation issues in new nuclear power plants must be actively managed to minimize operational impacts. Increased awareness of materials issues related to advanced light water reactors can limit plant deployment risk and enable the incorporation of lessons learned and materials research advances from the existing fleet into new designs. One of the primary factors affecting materials degradation is residual stresses left during the fabrication process. These tensile residual stresses can cause stress corrosion cracking and other failure mechanisms in nuclear plant piping. Proper selection and control of welding and fabrication processes can reduce or eliminate tensile stresses on the inside diameter surface of piping components.

Approach A materials management matrix (MMM) will be produced for each of the advanced light water reactor designs being considered by Advanced Nuclear Technology members (AP1000, EPR, ABWR, APWR, ESBWR, and APR1400). The MMMs will list critical materials-related information for each of the components in the nuclear steam supply system (ASME Class 1 equivalent). These matrices will be living documents that can be used to manage material issues over the operating lives of the reactors. Other materials-related projects including welding best practices will be conducted as needed. A group of welding and materials engineers, fabricators, and other industry metalworking experts will be consulted to develop recommendations and document best practices for welding plant components and for fabricating new plants.

Impact Potential benefits from this project include the following: Assist in evaluating materials-related issues that can significantly affect operating costs in existing reactor

designs Ensure new plants are economically competitive over their operating life through more effective materials

management Minimize tensile residual stresses through application of welding best practices Simplify management and evaluation of materials degradation and flaws Extend initiation time of many degradation mechanisms by eliminating or minimizing the tensile residual

stresses in welds


How to Apply Results The MMMs function as an ongoing roadmap to material issues for the life of the plant. They will be maintained by EPRI and updated periodically with assistance and feedback from vendors and plant owners. Members will use them to assist with purchase decisions and as an ongoing tool for optimizing inspection schedules and mitigation strategies.

New Nuclear Plant Reliability (supplemental) (066888)

Key Research Question Equipment reliability at nuclear plants starts with design and procurement and continues through construction and startup. When the plant begins operating, the level of success and cost associated with equipment reliability programs are directly related to the foundation established in the early stages of the plant’s life. While equipment vendors can focus on first costs, the owner/operator needs to consider first cost, operating cost, and plant revenue when making decisions. Limited actions have been taken to factor lessons learned from current equipment reliability programs into new plant projects. Also, new plants will apply advanced instrumentation and control (I&C) and communications technologies, including higher-frequency components and wireless, far more extensively than the current nuclear fleet, giving rise to new technical and regulatory concerns.

Approach This project will develop numerous products to support enhanced plant reliability programs. Products may address component specifications, commodity standards, and new I&C technology. As one example, a practical, clear, common understanding of the concerns associated with electromagnetic compatibility management and qualification requirements associated with higher-frequency components will be needed among utilities, regulators, and equipment suppliers. Therefore, the current guidance needs to be updated and expanded.

Impact Potential benefits from this project include the following: Establish a foundation for a highly effective operating plant equipment reliability program, leading to high-

capacity factors and contained operations and maintenance costs Incorporate owner expectations for equipment monitoring into initial project plan Minimize impact of electromagnetic interference through spectrum management planning Incorporate experience on nuclear power plant startup from the international community

How to Apply Results Plant staff will apply lessons learned, guidelines, and recommendations from existing plants into plant programs for new plants. With respect to advanced I&C and communications technologies, utilities will be able to make the electromagnetic interference information available to their suppliers as needed to ensure that electromagnetic compatibility requirements are met. The project will result in new plants that are much less likely to experience electromagnetic interference problems, with corresponding improvements in plant reliability, safety, and operating costs.



New Nuclear Plant NDE (supplemental) (066885)

Key Research Question New nuclear power plants must comply with requirements for pre-service inspection (PSI) and in-service inspection (ISI) of pressure boundary components and supports in accordance with regulatory-accepted codes and standards (for example, American Society of Mechanical Engineers [ASME] Section XI). Although these requirements are typically deterministically based, about 90% of existing plants have transitioned to risk-informed methods. Nuclear power plants also must comply with ASME Section III requirements for construction, pre-service, and in-service volumetric inspection of the primary pressure boundary. The construction volumetric inspection ensures that the components and welds meet applicable codes. Changes to ASME Section III could eliminate unnecessary repairs by allowing acceptance of fabrication flaws that are not structurally significant. This will require industry and regulatory confidence in the volumetric nondestructive evaluation (NDE) methods that are used to detect and size fabrication flaws.

Approach The Electric Power Research Institute's (EPRI’s) risk-informed in-service inspection methodology will be tested and refined for advanced plant designs. The goal would be to develop a single PSI/ISI program for each design (not site-specific) or define any issues and provide guidance where there may be site-specific requirements. The project also will establish and operate the ultrasonic (UT) examination qualification programs necessary to ensure accurate, reliable construction inspections of primary pressure boundary components and welds.

Impact Results from this project could support timely and cost-effective compliance with regulatory requirements for construction inspection, including the following: Regulatory acceptance of fitness-for-service philosophy, eliminating unnecessary repairs

Cost savings and elimination of delays during construction Improved resistance to stress corrosion cracking during plant operation

Reduced construction, inspection, and acceptance costs Reduced operating costs Shortened construction and turnover schedules by focusing resources on more important systems and

components and incurring fewer regulatory actions

How to Apply Results Members will have access to a regulatory-approved ISI methodology for use in new plant license applications. This also will support future operations and maintenance activities over the life of the plant. Members will use the UT Qualification products in specifying certification levels and specialized qualifications for UT personnel in construction procurements. As a result of these products, the necessary qualified UT personnel will be available to support new plant construction.



Security and Seismic (supplemental) (067606)

Key Research Question The underlying generic seismic source model for the Central and Eastern United States is more than 18 years old, and the Nuclear Regulatory Commission is on record as indicating that the model should be updated every 10 years. Also, recent NRC Requests for Additional Information are challenging the previously developed seismic source characterization continuing the NRC trend away from its use of EPRI’s baseline seismic source characterization model.

Approach EPRI is developing a new seismic source characterization model for the Central and Eastern United States (CEUS) that will facilitate new plant siting and respond to regulatory concerns. The new generic CEUS seismic source characterization model will benefit several industry participants: a) nuclear power plant owners and developers that have submitted an early site permit (ESP) or combined operating license application (COLA) for NRC review in 2007 (or 2008) or will submit a COLA for a second unit in or after 2009; b) nuclear power plant owners and developers that will submit an ESP or COLA for NRC review in or after 2009; and c) operating nuclear power plant owners that must respond to issues resulting from NRC Generic Issue 199 (GI-199).

Impact Potential benefits from this project include the following: Realize significant time and resource savings through standardization and partnering Achieve stability in the seismic design of new plants Inform decisions regarding the current state of knowledge in a non-regulatory environment using the

approved Senior Seismic Hazard Analysis Committee process Eliminate potential for multiple, possibly conflicting, interpretations of seismic sources by different

consultants Avoid challenges to the EPRI-Seismic Owners Group (1989) seismic source characterizations by NRC

staff, its consultants, or interveners during the hearing process Minimize potential for duplication of effort because seismic hazard models for many CEUS sites have

significant geographic overlap Reduce the opportunity for delays due to more conservative interpretations

How to Apply Results Members will use project results in developing early site permits, combined operating license applications, and in responding to regulatory concerns. This work replaces a study performed approximately 20 years ago. Since that study was completed, substantial work has been done to improve the understanding of seismic sources and their characterization in the Central and Eastern United States.

Achieving Virtual Power Plant Configuration Management (supplemental) (068802)

Key Research Question New plants will be designed using advanced 3-D computer-aided design software, producing detailed models that can be integrated with plant design, operations, and maintenance databases. Associated software suites supporting the 3-D models also may support documentation and characterization of facility system, structure, and component data and attributes. An automated, integrated, and interoperable configuration management program must be established to maintain consistency between the design requirements, physical configuration, and facility configuration documentation to ensure the ability to document and maintain compliance with the license basis.



Approach This project consists of three sub-tasks: Develop a Standard Configuration Management Reference Model and Implementing Guideline to provide

a common framework to help manage an automated, integrated, and interoperable configuration management program in a consistent and connected way.

Develop New Nuclear Plant XML Equipment Schemas for a set of critical components to help facilitate information interoperability throughout the full plant life cycle.

Develo p a New Nuclear Plant Information Handover Guide providing a full plant life-cycle information strategy establishing the methodology for defining the information requirements and the “how to” for developing and implementing an information handover plan.

Impact Configuration management programs implemented using the Standard Configuration Management Reference Model and Implementing Guideline can achieve the following benefits and avoid past problems:

Clear design basis Plant physical configurations match documents/records and design requirements Operating procedures integrated with design constraints and limits Complete set of records to support maintenance (for example, Q-lists and bill of materials) Designs containing instrumentation for monitoring the physical plant to ensure equipment

configuration is maintained consistently with design requirements Avoid long-term plant shutdowns caused in part by configuration management problems

How to Apply Results The 3-D models will assist plant owners in managing and improving the return on their new build investment. The information developed will be in the form of guidelines and new nuclear plant XML equipment schemas that the plant owner can directly implement. Much of the benefit will be from the industry adoption of data interoperability standards developed in this project.

Advanced Nuclear Plant Research (base) (052492)

Key Research Question New sources of energy are being developed to alleviate two key challenges facing all industrialized and developing countries: increased energy security to improve national and global security and reduced carbon footprint in all economic sectors to address manmade contributions to global climate change. One such energy source is the high-temperature gas reactor, which can extend the use of improved nuclear technologies into energy sectors beyond electricity generation.

Approach This project covers multiple activities associated with various advanced nuclear plant designs: Supporting activities around developing a public/private partnership to determine the feasibility and

practicality of using advanced high-temperature gas reactor systems for process heat, hydrogen production, and cogeneration.

Studying sonoluminescing bubbles and nanomaterials synthesis. Sonoluminescence is the light emission associated with catastrophic bubble collapse of a gas bubble oscillating under an ultrasonic field. Sonoluminescence can be applied to synthesize nanomaterials, which can be used for the direct catalyzed decomposition of methane into hydrogen—a potential process for hydrogen production in high-temperature gas reactors.




Support industry activities related to research, demonstration and deployment of emerging small modular reactor (SMR) designs

Impact Potential benefits from this project include the following: Accelerate and direct the development, demonstration, and deployment of advanced reactor technology

for future energy markets. The commercial deployment of new technologies can help address critical issues of energy security and global climate change.

Investigate alternate processes to enable hydrogen production at lower costs and with less energy input. Identify member-driven, common technical requirements for design, development, and operation of small

modular reactors, and define potential research gaps.

How to Apply Results Members will review research and development results to guide awareness and development of advanced nuclear power plants, including small modular reactor designs and high-temperature gas reactors.

Chemistry, Low-Level Waste and Radiation Management

Develop guidance and technologies to improve water chemistry practices, enhance low-level waste management, reduce radiation exposure,

and inform plant decommissioning efforts.

Programs Low-Level Waste and Radiation Management Decommissioning and Technology Development Water Chemistry

Low-Level Waste and Radiation Management

Program Overview

Program Description Nuclear power plants face significant regulatory, economic, environmental, and public perception pressures with respect to low- and intermediate-level waste (LLW) management and personnel exposure to radiation. The safe processing, handling, and disposing of low-level waste requires a detailed familiarity with both technical and regulatory issues. Similarly, as regulatory limits on personnel exposure to radiation decrease, greater effort is needed to develop and demonstrate effective radiation protection and source-term reduction technologies. The Low-Level Waste and Radiation Management program investigates improvements to nuclear plant operational practices that can reduce risks associated with waste management and radiation exposure. The program develops guidelines and technologies for waste disposal volume reduction, dose and radiation field reduction, and nuclear plant decommissioning, resulting in lower electricity production costs, better informed regulatory oversight, and improved public perception. The program also develops technical guidance for early detection, mitigation, and remediation of groundwater contamination, an issue of increasing public concern and regulatory oversight.

Research Value Effective management of low- and intermediate-level waste and radiation exposure enables nuclear plants to operate safely, cost-effectively, and with minimal risk to plant personnel, the public, and the environment. Research results are used by radiation protection managers to develop strategies for minimizing waste generation and reducing handling and storage costs. Research results are used by radiation protection managers to minimize radiation fields and reduce activity generation. Low-Level Waste and Radiation Management Program participants gain access to the following: Technologies, assessments, and guidelines that can reduce solid and liquid waste volumes. LLW

assessments, for example, have identified optimization recommendations valued at more than $75 million per year.

New source-term reduction and radiation protection techniques that can reduce radiation dose. Source-term reduction studies have identified methods for reducing radiation fields by as much as 50% over 5 years.

Technical guidance for risk-informed regulations in LLW, radiation protection, and groundwater protection that can address public safety and environmental stewardship concerns. Operational strategies for reducing the volume of Class B/C LLW could save the industry more than $27 million per year when fully implemented.

Approach The Low-Level Waste and Radiation Management Program develops knowledge, guidance, and tools to reduce the risks and costs associated with waste management and radiation protection. The program also conducts plant assessments to provide expert support and to capture lessons learned that can be shared across the industry. Provide cost-effective, risk-based alternatives for waste disposal due to limited LLW site access (Barnwell

closure). Develop guidelines and technologies for reducing waste volumes and worker radiation dose. Establish technical foundations for reduced regulatory burden in the areas of radiation protection and low-

level waste. Provide tools for improved public perception regarding groundwater protection programs.

Low-Level Waste and Radiation Management - Program 41.09.01 p. 1


Accomplishments The Electric Power Research Institute (EPRI’s) Low-Level Waste and Radiation Management Program supports industry efforts to reduce the costs and regulatory burdens associated with low-level waste and to drive reductions in public, environmental, and personnel exposure to radiation. The Program develops and demonstrates innovative technologies, converts industry operating experience into practical guidelines, and explores alternative approaches for more effective LLW and radiation management. Received regulatory endorsement for EPRI's On-Site LLW Storage Guidelines, which provide consistent,

industry-driven guidance for operation of on-site LLW storage facilities. Submitted to U.S. Nuclear Regulatory Commission the technical basis for LLW concentration averaging,

which would provide $36 million per year in industry cost savings upon implementation. Source-term reduction recommendations implemented at Brown's Ferry Unit 1 helped the unit achieve the

lowest dose rates in the boiling water reactor (BWR) fleet following a restart. Based on analysis of industry pressurized water reactor (PWR) dose rate data, identified technical

solutions that can provide measurable reductions in plant radiation fields. For example, zinc injection and electropolishing show strong benefit in reducing dose rates.

Completed multi-year review of updated research on the health effects associated with low-dose radiation. Analysis concluded that the radiation damage/response paradigm should be expanded to account for increased complexity in biological response mechanisms. Results shared with regulatory community to inform revisions to radiation protection standards.

Developed guidance to establish industry-wide practices for monitoring the release of materials and personnel from radiologically controlled and protected areas. Recommendations encompass tools, equipment, volumetric and non-volumetric materials, small quantity materials, non-radioactive shipment vehicles, and personal items.

Developed tools for implementing an effective, optimized groundwater monitoring and remediation program, including technical guidelines and supplemental site evaluations. Nuclear Regulatory Commission endorsement of the EPRI Guidelines would provide significant site relief.

Current Year Activities Low-Level Waste and Radiation Management program research and development for 2011 will sustain progress toward lower-cost waste handling and disposal, reduced worker dose, and improved detection and monitoring of groundwater. Specific efforts will include the following: Develop the technical basis for regulatory changes to low-level waste classification criteria. Develop EPRI Cobalt Reduction Guidelines, Rev.1. Develop EPRI groundwater remediation guidelines and technologies. Provide site-specific implementation support for source-term reduction technical guidance being

developed by the program.


Program Manager Lisa Edwards, 469-586-7468, [email protected]



Summary of Projects


P41.09.01.01 Low-Level Waste R&D Program (base)

The Low-Level Waste R&D Program conducts research and analysis that can reduce costs and improve efficiency and performance of LLW management strategies, reduce the generation of Class B/C LLW requiring on-site storage, provide clear and concise industry guidance for the safe and regulatory compliant on-site storage of LLW, and where appropriate, provide technical justifications for changes to existing disposal regulations to expand the disposal options available to LLW generators.

P41.09.01.02a Low-Level Waste Assessment (supplemental)

The low-level waste assessment activities available through this project are designed to help plant personnel fully benefit from base research program results. Participating members will select the assessment area of focus. Common areas of focus include: On-Site Storage, BC Reduction, Solid LLW, Liquid LLW, and Liquid System Manager (LSM) installation.

P41.09.01.02b LLW Technical Strategy Group (supplemental) - Annual Membership OR 3-Year Membership + Assessment

The LLW Technical Strategy Group provides a forum for discussing technical issues and sharing lessons learned regarding strategic LLW management. Members also receive expert technical consulting as part of their membership. The LLW Technical Strategy Group is available in 3-year and 1-year membership options. The 3-year membership includes one full LLW assessment once during the three-year period.

P41.09.01.03 Radiation Management R&D Program (base)

The Dose reduction for cumulative and individual exposure tasks will be accomplished by developing a systematic method for evaluating the tasks and determining the optimum strategy for technology evaluation and implementation.

P41.09.01.04a Radiation Management Assessment (supplemental)

A series of assessment activities have been designed to help plant personnel gain the full benefit of the base program results by providing site-specific or unit-specific guidance in implementing the technologies and strategies produced by the base Radiation Management Program. Assessments include as low as reasonably achievable (ALARA) technologies, radiation source term, remote monitoring technology, and scaffolding.

P41.09.01.04b Radiation Management/Source Term Strategy Group (supplemental) - Annual Membership OR 3-Year Membership + Assessment

The Radiation Management/Source Term Strategy Group provides an interactive forum for members to share and get expert advice in applying ALARA technologies and to gain insights on how to effectively reduce source term. Industry lessons learned and discussion of emergent issues will provide members with the most up-to-date information for making informed decisions on job planning and preparation.

P41.09.01.05 Groundwater Protection R&D Program (base) (QA)

EPRI's Groundwater Protection Project provides members with advanced strategies and technologies for improved management of situations involving radiologically contaminated groundwater. This project develops technical guidance for implementing site-specific groundwater monitoring programs geared toward mitigation, early detection, and remediation of groundwater contamination. Implementing these programs will enhance site knowledge and increase confidence and accuracy in communications with stakeholders.This project will evaluate and develop advanced technologies and methodologies for environmental and groundwater protection, monitoring, and remediation at nuclear power plants.





P41.09.01.06a Groundwater Assessment (supplemental)

Three types of assessments are available to members through this supplemental assistance program: Groundwater Characterization and Protection, Groundwater Technology Demonstration, and Tritium and Water Management Model. Participants will work with EPRI to select the assessment type that is of most interest to the plant. The generic lessons learned and experiences from each assessment project will be incorporated into the EPRI base program to support further development of technical guidance and technologies.

P41.09.01.06b Groundwater Strategy Group (supplemental)

The Groundwater Strategy Group provides members a forum for discussing groundwater protection experiences, lessons learned, and advanced technologies. Members will have access to the Groundwater Strategy Group collaboration website and quarterly conference calls. Those utilities participating in the 3-year membership are eligible for one site-specific Groundwater assessment once during the 3-year period.

Low-Level Waste R&D Program (base) (061432)

Key Research Question Nuclear power plants continually evaluate opportunities to reduce the costs and improve the efficiency and performance of low-level waste (LLW) management, handling, storage, and disposal programs. EPRI’s Low-Level Waste Program helps nuclear plants optimize LLW management programs through advanced media testing, improved technologies and tools, safe and efficient on-site storage of LLW, and the development of technical bases for improved flexibility and risk-informed regulations for LLW disposal.

Approach Currently, 85% of the U.S. industry does not have access to Class B/C LLW disposal. EPRI has developed a three-part strategy to address this issue: 1) minimize the generation of Class B and C waste, 2) provide industry guidance for on-site storage of waste, and 3) examine alternatives to existing disposal regulations and guidance. Multiple products have been developed and are under development in support of this strategy. Waste Class B/C Reduction Guide (1015115): Revision 1 is a planned product for 2011 Guidelines for Operating an Interim On-Site Low Level Radioactive Waste Storage Facility, Revision 1

(1018644) and the Low Level Waste On-Site Storage Operating Guideline — Supplemental Information Manual (1018651)

Proposed Modification to the NRC Branch Technical Position on Concentration Averaging and Encapsulation (BTP), Technical Basis and Consequence Analysis (1016761)

Technical Progress Report on Options for Improved LLW Disposal Flexibility Using 10CFR61.58 (1019222): Final report planned for 2010

Specialized media and other operations continue to be a developing field. The need to reduce Class B and C waste has led to the cost-effective development of nuclide-specific separation methods and innovative processing technologies both within EPRI and by industry service providers. EPRI's LLW R&D Program identifies and conducts performance testing of media and processing strategies that may be considered for reducing Class B and C waste generation and for optimizing liquid radwaste processing system performance. In 2010, this project will evaluate the use of nuclide-specific media by treating plant-generated liquid radioactive waste in a series of test columns. The tests will evaluate the relative removal efficiency as compared to other media types and consolidate existing guidance on multiple media evaluations.


In 2010, EPRI's LLW R&D Program will complete the work on developing a technical basis for options for improved LLW disposal flexibility using 10CFR61.58. Performance testing of multiple media for the selective removal of Cs-137 and colloidal cobalt in liquid radwaste processing systems will be conducted. Work will also begin on the revision of the 2007 Waste Class B/C Reduction Guide, the development of global radwaste profiles for member countries, and the development of a technical basis for Very Low Level Waste as a separate waste category.

Impact Reduced generation of Class B/C LLW requiring on-site storage. Avoided potential new rule making regarding on-site storage of LLW. The On-Site Storage Guidelines have been reviewed and endorsed by the Nuclear Regulatory

Commission as "providing an acceptable method for recordkeeping, determining waste forms and waste containers and monitoring and inspecting the interim long-term storage of LLRW." This effort provides concise guidance to plant operators on how to operate their interim storage facilities in compliance with regulatory expectations.

Provided technical basis for modification of the regulatory branch technical position to allow broader blending of compatible waste. If adopted, these changes will favorably increase the volume of waste that is classified as Class A and reduce the amount that is Class B or C.

Expanded disposal options for the nuclear industry through more clearly established protection requirements.

(Future) Provide technical justification for regulatory relief to the concentration limits defined in 10CFR61 based on existing knowledge of performance assessment technology, realistic dose-pathways of existing disposal facilities, and updated dose conversion factors.

How to Apply Results Program results are used by radwaste managers in developing strategies for minimizing the generation of Class B/C waste and optimizing solid and liquid LLW management programs. Members also can apply technical guidance to ensure compliance with regulatory concerns regarding interim waste storage. Long-term research results will provide the technical basis for future risk-informed regulations and improved flexibility of disposal options.

Low-Level Waste Assessment (supplemental)

Key Research Question EPRI understands that the implementation of guidance, technologies, and strategies produced by the LLW base research program can be done more quickly and effectively with site-specific assistance. A number of LLW assessment activities can be deployed to help plant personnel gain the full benefit of base EPRI research results.

Approach Most assessment activities provide an on-site evaluation of how specific research results, technologies, industry experience, and industry best practices could be applied at a given plant. The assessment delineates actions with the largest potential benefit to the site and identifies potential gaps that, if closed, could provide economic, performance, and/or regulatory margin benefits. Participants can select from several assessment focus areas: on-site storage, BC reduction, solid LLW, liquid LLW, and Liquid System Manager software installation. The utility also may specify a focus area currently challenging the plant. The utility and EPRI Project Manager will then work together to define the scope.



Impact Expected impacts vary depending on the area of focus, but impacts can include the following: Optimized LLW management program performance Improved regulatory margin Cost savings due to reduced generation of Class B and C waste Improved effluent program performance Cost savings due to reduced wet and dry LLW generation Improved program performance through advanced technology applications Evaluation of site implementation of guidance documents

How to Apply Results Members participate with an industry expert during the on-site assessment. The area of focus is determined by the assessment activity selected. The process helps plant personnel gain insights about their plant-specific performance and how to apply EPRI guidance, technologies, and tools to the plant's advantage. After the assessment, a confidential site-specific report details the strengths and gaps associated with program implementation and prioritizes recommendations and potential benefits. Later, generic results and lessons learned may be compiled in program reports for industry use.

LLW Technical Strategy Group (supplemental) - Annual Membership OR 3-Year Membership + Assessment (004514)

Key Research Question Nuclear plants frequently benefit from broader awareness of the LLW management activities practiced at other plants. The LLW Technical Strategy Group provides a forum for discussing technical issues and sharing lessons learned regarding strategic LLW management. Emerging technical issues include the Nuclear Regulatory Commission (NRC) Branch Technical Position on LLW concentration averaging; potential changes to 10CFR61; and LLW disposal site development, blending, encapsulation, solidification, and economics. Members also receive expert technical consulting as part of their membership. The LLW Technical Strategy Group is available in 3-year and 1-year membership options. The 3-year membership includes one full LLW assessment once during the 3-year period

Approach The Technical Strategy Group conducts periodic conference calls to keep membership appraised of emerging issues and to solicit input on industry responses to these issues. Webcasts are used to provide members with up-to-date status of LLW disposal options, presentations on new processing strategies, information on new regulatory notices, technical exchanges of lessons learned, and new ideas on cost control. Members of the LLW Technical Strategy Group receive annual on-site expert technical consulting as part of their membership. This consulting time is typically used for continuous improvement of LLW program management strategies and for analysis of special projects. On-site consultation topics are scheduled with individual members. International members participating in the LLW Technical Strategy Group will receive their site-specific support remotely. As noted above, those utilities participating in the 3-year membership are eligible for one site-specific LLW assessment once during the 3-year period. This assessment is conducted on-site for both U.S. and international members.



Impact Participation in the Technical Strategy Group keeps members abreast of emerging issues surrounding LLW management and provides members with a forum for technical exchange. Site-specific consulting time provides expert support for specific plant or corporate project requests. Individual plant and fleet strategies for LLW management are frequently evaluated with this support. Cost evaluations conducted during these consultations often identify significant cost-saving measures. The on-site assessment provided for utilities participating in the 3-year membership evaluates how specific research results, technologies, industry experience, and industry best practices could be applied at a given plant. The assessment delineates actions with the largest potential benefit to the site and identifies potential gaps that, if closed, could provide economic, performance, and/or regulatory margin benefits.

How to Apply Results On-site consultation time is used to ensure EPRI guidance is applied to emerging and critical plant-specific LLW management issues. Participation in periodic webcasts keeps members abreast of emerging issues in the rapidly changing climate surrounding LLW management and provides members with a forum for technical exchange. The LLW assessment helps plant personnel gain insights about their plant-specific performance and how to apply EPRI guidance, technologies, and tools to the plant's advantage.

Radiation Management R&D Program (base) (052350)

Key Research Question The nuclear industry's radiation protection organizations are continually challenged to better manage and reduce radiation exposure to workers while balancing increased maintenance work resulting from plant aging issues. There is a worldwide trend to adopt more restrictive worker dose limits while minimizing total collective exposure. In the United States, the Nuclear Regulatory Commission announced its intentions to revise radiation protection regulations in the next several years. Perhaps the most challenging change will be the adoption of lower occupational dose limits, potentially reducing the current limit of 5 rem/yr to as low as 2 rem/yr. EPRI's Radiation Management Program, the Nuclear Energy Institute (NEI), the Institute of Nuclear Power Operations (INPO), and the nuclear industry worked together to develop the RP2020 Initiative, which aims to make radiation exposure a non-issue by the year 2020. Use of a wide spectrum of dose reduction and source term reduction technologies will likely be required to meet the regulatory challenges and industry goals of improved radiation protection practices. In addition, better incorporation of the as low as reasonably achievable (ALARA) principles by vendors and other stakeholders in the development of new tools and techniques is needed for plant inspection and maintenance. A new framework for more efficient human resource allocation and coordination of key workers also is needed. To meet these challenges, EPRI's Radiation Management Program will develop guidance, technologies, and operational practices to more aggressively reduce radiation fields (source term) and minimize worker dose to ALARA standards.

Approach The Radiation Management Program is divided into two major areas: 1) Source Term Reduction technical area, which is focused on minimizing radiation fields, and 2) Radiation Protection technical area, which is focused on improving the use of dose reduction technologies and improving worker efficiency. The immediate priority for the program is to reduce the number of high-dose workers receiving greater than 1.0 rem/yr through the optimization of tasks and the radiological work environment.



The Radiation Protection technical area enhances worker efficiency and minimizes dose through the improved use of engineering controls, administrative controls, and personal protective equipment. To meet the potential reduced occupational dose limit, a five-step approach is planned to prioritize research and development efforts: 1) survey the industry to determine which jobs have the highest cumulative and individual exposures; 2) reach out to other EPRI programs and vendors/utilities to identify the detailed tasks associated with the high-dose jobs; 3) organize the jobs and rank them for feasibility to reduce dose through technology implementation; 4) identify technologies that can be useful for reducing the time, distance, or shielding requirements for each of the high-dose tasks; and 5) document the implementation of these technologies at host utilities. During 2011-2015, this technical area will focus on demonstrations of key dose reduction technologies. Concurrently, the Radiation Source Term Reduction technical area aims to minimize radiological fields (source term) in key work environments through a more thorough understanding of the relationship between the operating environment (for example, chemistry and core duty) and metal surfaces (material). The Radiation Source Term Reduction Technical Area develops operating strategies and technologies that can intervene at each of the five major processes that occur in the creation of radiation fields: 1) corrosion and release, 2) corrosion product transport, 3) deposition and activation, 4) activated product transport, and 5) deposition and incorporation of activated products on surfaces. Efforts in 2011-2012 will focus on the impact of high core duty on ex-core radiation fields. Industry radiation field measurements will be based on industry standard measurement points (BWR radiation assessment and control [BRAC] and the Standard Radiation Monitoring Program [SRMP]).

Impact Aggressive and coordinated industry implementation of key dose reduction strategies will help ensure regulatory compliance for all workers prior to any actual regulatory change. This systematic method of applying technology will provide the industry with resources such as technology identification, sample engineering qualifications, and lessons learned. Additionally, the implementation of strategic source-term reduction techniques may result in reduced/managed crud bursts, lower-radiation fields, improved fuel performance (less curd on fuel leads to few failures), and reduced low-level waste generation.

How to Apply Results The results of the proposed method to prioritize tasks and implement individual plant demonstrations will provide members with resources such as technology identification, engineering qualifications of the technologies, and implementation lessons learned. Members also will be given recommendations on how best to reduce radiation fields and activity generation based on plant design and operating factors.

Radiation Management Assessment (supplemental)

Key Research Question Nuclear power plants face significant regulatory, economic, environmental, and public perception pressures with respect to personnel exposure to radiation. As regulatory limits and industry goals associated with personnel exposure to radiation become more challenging, greater effort is needed to develop and demonstrate effective radiation protection and radiation source-term reduction. The advances developed through EPRI's Radiation Management Program can improve plant performance associated with management of radiation exposure, but are often under-utilized. To help plant personnel gain the full benefit of EPRI research results, a series of assessment activities have been designed to help implement the guidance, technologies, and strategies produced by the base program.

Approach EPRI understands that the implementation of guidance, technologies, and strategies produced by the LLW base research program can be done more quickly and effectively with site-specific assistance. A number of Radiation Management assessment activities are available to assist nuclear power plants in implementing research results from the EPRI Radiation Management base research program:



1. ALARA Assessment (by site): This project entails a 5-day on-site assessment of the application of ALARA

in the daily activities of the plant. The information is used to develop a comprehensive report that summarizes the program's strengths and status using EPRI and industry exposure management databases. The approach uses a series of established "filters" to identify relevant ALARA criteria from various sources, including EPRI technology, the information system of occupational exposure databases, and industry "good practices." To date, the EPRI methodology has been used to evaluate the ALARA programs at 16 sites.

2. Radiation Source Term Assessment (by unit): This project entails a 3-day site visit to evaluate a plant's source-term reduction potential by evaluating material and operating status, normal chemistry program, radiation monitoring program, and plant chemistry controls for shutdown and startup. When possible, performance and parameters will be benchmarked against similarly designed units.

3. Remote Monitoring Technology Assessment (by site): This project entails a 3-day assessment of a station's remote monitoring program and existing capabilities to identify opportunities for near-term and long-range expansion of this technology. A review matrix based on the EPRI Remote Monitoring Guidelines will be used to evaluate deployment scope, interface with other plant processes, determine use of procedures and standards, and examine training/qualifications.

4. Scaffold Assessment (by site): This project entails a 4-day, on-site review of the plant's current vertical access program. The site-specific data are analyzed to define current program performance, identify opportunities for improvement, and to benchmark the station's performance relative to similar stations. The methodology focuses on practical elements and includes an assessment of the integration and effectiveness of the overall organization.

Impact EPRI assessments provides the station with a success path for implementing exposure management technologies and techniques, resulting in the following benefits: Reduced personnel exposure Reduced radiation fields In-depth assessment of source term with benchmark relative to industry Reduced source term Reduced outage dose, duration, and costs Enhanced technologies for remote monitoring of personnel Improved integration of station goals and focus

How to Apply Results Members participate with an industry expert during an on-site assessment. The area of focus is determined by the assessment activity selected. The process helps plant personnel gain insights about their plant-specific performance and how to apply the EPRI Radiation Management Program guidance, technologies, and tools to the plant's advantage. After the assessment, a confidential site-specific report that details the strengths and gaps associated with the program implementation is developed. Prioritized recommendations and potential benefits are highlighted. Later, generic results and lessons learned may be compiled in program reports for industry use.



Radiation Management/Source Term Strategy Group (supplemental) - Annual Membership OR 3-Year Membership + Assessment

Key Research Question While the industry’s annual collective radiation exposure continues to trend down, aggressive industry goals to further minimize station dose are challenging to meet in the life extension environment. The health of radiation protection programs is regularly assessed using cumulative exposure and exposure estimating, and utilities remain obligated to minimizing the impact of ionizing radiation on plant personnel. Technology transfer and sharing of lessons learned can assist plants in driving greater implementation of EPRI guidance, technologies, and strategies related to ALARA and radiation management programs. The Radiation Management/Source Term Strategy Group is designed to enhance the technology transfer between EPRI and member utilities through interactive forums, workshops, and expert assistance. Such engagement helps plants "take the research off the shelf and put it into the plant." Generic results and lessons learned will be used in base research and development (R&D) program products and reports. The Radiation Management/Source Term Strategy Group is available in 3-year and 1-year membership options. The 3-year membership includes a full Radiation Management assessment once during the 3-year period.

Approach The Radiation Management/Source Term Strategy Group provides an interactive forum for members to share and get expert advice in applying ALARA technologies and to gain insights on how to effectively reduce source term. Industry lessons learned and discussion of emergent issues will provide members with the most up-to-date information for making informed decisions on job planning and preparation. This comprehensive approach to managing radiation exposure will help educate and cross-train personnel with differing backgrounds on topics highly relevant to dose minimization and management. The Strategy Group focuses on best practices, advanced technologies, the most efficient implementation options of ALARA source-term reduction technologies, and cost-effective sustainable ALARA program success. The group typically sponsors workshops on topics of interest to dose management such as source term reduction and advanced shielding applications. These workshops bring together plant personnel and service providers to ensure members are kept abreast of emerging technologies and have the advantage of a peer-to-peer forum for exchanging ideas and information. Results obtained from these workshops and information exchanges will be integrated into a summary report that will provide Strategy Group members with the benchmarking information they would need to develop a site-specific dose reduction strategy. The information also will supplement research performed within the EPRI base-funded Radiation Management Program. As noted above, those utilities participating in the 3-year membership are eligible for one site-specific Radiation Management assessment once during the 3-year period. This assessment is conducted on-site for both U.S. and international members and provides a detailed evaluation of how specific research results, technologies, industry experience, and industry best practices could be applied at a given plant.

Impact Plants that have implemented relevant EPRI technology in the ALARA area have realized significant benefits in personnel exposure control. EPRI’s radiation field control technologies provide a menu of techniques to reduce out-of-core shutdown radiation fields and for continuing development of worker risk-minimization techniques that target increased worker productivity. The Radiation Management/Source Term Strategy Group provides an opportunity to drive exposure performance success that, in turn, can impact regulatory requirements, insurance premiums, and benchmarking metrics.



Utilities participating in the 3-year membership are eligible for one site-specific Radiation Management assessment once during the 3-year period. The assessment delineates actions with the largest potential benefit to the site and identifies potential gaps that, if closed, could provide economic, performance, and/or regulatory margin benefits.

How to Apply Results Members sponsor annual meetings and workshops that address key ALARA and radiation source term issues. These workshops provide an interface for peer-to-peer and utility-to-service-provider interactions on topics of specific interest to dose management. Lessons learned and insights are brought back to the plant for implementation. Members also have access to Radiation Management/Source Term Strategy Group industry experience databases and reports. These mechanisms enable radiation protection personnel to more effectively address technical ALARA issues and implement worker risk-minimization techniques. For utilities participating in the 3-year membership option, the full Radiation Management Assessment can be used to gain insights about the plant’s performance and how to apply the EPRI guidance, technologies, and tools to the plant's advantage. The assessment team will develop a confidential site-specific report that details the strengths and gaps associated with the program and highlights prioritized recommendations and potential benefits. Later, generic results and lessons learned may be compiled in program reports for industry use.

Groundwater Protection R&D Program (base) (QA) (068229)

Key Research Question Leaks and spills from nuclear power plant operations can potentially impact site groundwater throughout the life of the plant. Experience from decommissioning and operating nuclear power plants shows that, although these leaks pose little to no public risk, they damage the industry’s public credibility. Understanding the extent of groundwater contamination during operation and in the early phase of decommissioning is important for the eventual decommissioning and site release of the nuclear power plant. In 2007, the U.S. nuclear industry committed to the Groundwater Protection Initiative [NEI 07-07] to implement groundwater monitoring programs at all nuclear power plant sites. International operating and decommissioning nuclear power plant sites also are implementing groundwater protection and characterization programs. As utilities implement groundwater protection programs at their sites, many encounter new challenges and lessons learned related to groundwater contamination. Further, new environmental issues can arise via regulatory changes, new standards, and stakeholder and public concern that require attention. Examples include carbon-14 in gaseous effluents and the impact of plant-related radioactivity on non-human biota. Greater fundamental understanding of such issues will be needed to develop effective environmental monitoring and mitigation approaches to meet public expectations and regulatory requirements.

Approach EPRI's Groundwater Protection Project provides members with advanced strategies and technologies for improved management of situations involving radiologically contaminated groundwater. This project develops technical guidance for implementing site-specific groundwater monitoring programs geared toward mitigation, early detection, and remediation of groundwater contamination. Implementing these programs will enhance site knowledge and increase confidence and accuracy in communications with stakeholders. Advanced and cost-effective technologies for the early detection, monitoring, and remediation of groundwater contamination will be developed and demonstrated for the nuclear industry. Methods and technologies for evaluating and preventing the failure of systems, structures, and components containing radioactive liquids will be explored for proactive action against groundwater contamination.



The EPRI Groundwater Protection group will provide members with technically sound methods, strategies, and technologies to address new environmental protection topics, such as carbon-14 and radioactivity to non-human biota. EPRI also collaborates with industry and regulatory entities to provide technical data that can inform policies related to groundwater and environmental protection.

Impact Utilities will be able to allay stakeholder concerns by taking action on groundwater and environmental protection issues. By implementing site-specific groundwater and environmental protection programs and the best technologies available, utilities will be able to optimize costs and reduce waste. Improve relationship with communities, government, and regulatory agencies about the industry’s

commitment to public radiation safety and environment protection Achieve cost savings at the decommissioning stage due to preemptive action during the operating stage Achieve cost savings due to advanced and efficient monitoring and remediation technologies Achieve cost savings due to prevention of radioactive liquid leakage to the environment

How to Apply Results Members can implement site-specific groundwater protection programs using the guidance provided in the EPRI Groundwater Protection Guidelines (1015118.) Members also can apply the technologies that are developed and demonstrated by EPRI at their plant sites. Members can interact with fellow groundwater protection colleagues and share lessons learned, experiences, good practices, and technologies at the annual EPRI Groundwater Protection Workshop.

2011 Products

Product Title & Description Planned

Completion Date Product Type

Technology Demonstration Experience Reports: EPRI will support and facilitate the demonstration of groundwater monitoring and remediation technologies at nuclear power plant sites. The results of these demonstration projects will be documented in EPRI Technical Updates.

12/31/11 Technical Update

Groundwater Protection Lessons Learned Report: The nuclear power plant industry began to implement groundwater protection programs at their sites in 2007 per the Groundwater Protection Initiative. This technical update will summarize the lessons learned and experiences from the implementation of groundwater monitoring and remediation projects at nuclear power plant sites. These lessons learned and experiences will be based on EPRI Groundwater Protection Assessment and NEI Groundwater Protection Initiative Peer Assessment results.


Future Year Products



Remedial Option Analysis Module (ROAM) for Nuclear Power Plants: The Remedial Option Analysis Module (ROAM) is an EPRI Environment Group software tool for modeling groundwater remediation options. The Groundwater Protection Program will collaborate with the Environment Group to develop radionuclide-modeling capabilities for ROAM.

12/31/12 Software






Software Tool for the Estimation of Carbon-14 in Effluents: In 2010, EPRI explored the methods and tools available to the nuclear power industry for the estimation of carbon-14 in plant gaseous effluents. The objective of this project was to develop a methodology for a more accurate estimation of the generation and release of carbon-14 from nuclear power plants. EPRI will develop an Excel based software tool to help members implement the EPRI-developed methodology for carbon-14 estimation.

12/31/12 Soft ware

EPRI Groundwater Protection Guidelines for Nuclear Power Plants, Rev. 1: In 2009-2010, the EPRI Groundwater Protection Committee reviewed the EPRI Groundwater Protection Guidelines. The review resulted in the identification of several items to include in future revisions. EPRI will continue the review and revision cycle of the Guidelines by assessing the need for revision in 2012.

12/31/12 T echnical Report

Groundwater Assessment (supplemental)

Key Research Question Leaks and spills from nuclear power plant operations can potentially impact site groundwater throughout the life of the plant. EPRI's Groundwater Protection Project has developed advanced strategies and technologies for improved management of situations involving radiologically contaminated groundwater, but nuclear power plants are often challenged in fully implementing these solutions. The EPRI Groundwater Supplemental Assessment Program provides members with site-specific assistance for implementing groundwater protection programs and technologies at their sites. This assistance is based on the implementation of NEI 07-07, the EPRI Groundwater Protection Guidelines (1015118), and promising groundwater protection technologies identified or developed by EPRI through the base program (1016765, 1016763, 1016764.)

Approach Three types of assessments are available to members through this supplemental assistance program. Participants will work with EPRI to select the assessment type that is of most interest to the plant. The generic lessons learned and experiences from each assessment project will be incorporated into the EPRI base program to support further development of technical guidance and technologies. 1) Groundwater Characterization and Protection Assessment Program: A team of groundwater protection experts will review and evaluate a plant's groundwater protection program based on the technical guidance provided in the EPRI Groundwater Protection Guidelines, NEI 07-07, and industry lessons learned. The team will provide recommendations for identifying and assessing system, structures, and components (SSCs) of concern; developing an adequate site conceptual model; characterizing groundwater; implementing a groundwater data management program; and implementing advanced technologies to augment a site's groundwater protection program. A site-specific report documenting the results of the Assessment will be developed for the member. 2) Groundwater Technology Demonstration: EPRI has evaluated and developed technologies for monitoring and remediation of groundwater at nuclear power plant sites. Through this Technology Demonstration Project, a team of groundwater protection experts will provide members with site-specific support for demonstrating a technology at their site. This support will include consultation on promising technologies, development of demonstration procedures, development of evaluation criteria, and on-site support for technology demonstration. The observations, results, and recommendations from the technology development will be documented in a site-specific report for the member.



3) Tritium and Water Management Model: Tritium at nuclear power plants accumulates in various tanks and systems before being released through monitored release pathways. The EPRI Tritium and Water Management Model tracks the generation and accumulation of tritium in systems and components and calculates the concentration of tritium in monitored releases. This Excel-based tool is site-specific and allows each plant/unit to understand and manage tritium impact on release pathways such as rivers, lakes, and oceans. This assessment includes the development of a site-specific tritium model based on plant design and operation.

Impact By taking action against groundwater contamination, utilities will be able to allay stakeholder concerns about environmental contamination. By implementing site-specific groundwater protection programs and the best technologies available for groundwater protection at nuclear power plants, utilities will be able to optimize costs and reduce waste due to groundwater monitoring and remediation: Improve relationship with communities, government, and regulatory agencies about the industry’s


How to Apply Results Members can use the site-specific results from the Groundwater Characterization assessment to optimize their tritium management and groundwater protection program, and to improve a plant's understanding of the site hydrogeology and risk for groundwater contamination. The results of the Technology Demonstration Program can be used to implement new groundwater protection technologies at the site. The results of the Tritium and Water Management Model can be used to optimize tritium inventory in various systems and monitored releases.

Groundwater Strategy Group (supplemental)

Key Research Question Leaks and spills from nuclear power plant operations can potentially impact site groundwater throughout the life of the plant. Experiences from other nuclear power plants can provide valuable insight into effective practices for addressing both technical and non-technical elements associated with groundwater protection. The Groundwater Strategy Group provides a forum for sharing such experiences with industry colleagues and defining best practices applicable across the industry. The Groundwater Strategy Group is available in 3-year and 1-year membership options. The 3-year membership includes one full Groundwater assessment once during the 3-year period.

Approach The Groundwater Strategy Group will be composed of members interested in sharing and discussing groundwater experiences and lessons learned, new technologies, and EPRI projects. Members will have access to the Groundwater Strategy Group collaboration website and quarterly conference calls. The collaboration website is a digital portal where documents can be shared and forum discussions can be held. Relevant groundwater experiences, lessons learned, and technology information will be uploaded to the collaboration website for member access. Members also will be able to post their own experiences, lessons learned, and technology ideas to spark discussion with other members. Questions on groundwater topics also can be posted so that EPRI groundwater experts and other members can provide answers and associated information.



Conference calls on groundwater protection experiences, lessons learned, and technologies will be held each quarter (February, May, August, November.) These conference calls will be used to discuss promising technologies, key experiences and lessons learned, and EPRI projects. As previously noted, those utilities participating in the 3-year membership are eligible for one site-specific Groundwater assessment once during the 3-year period. This assessment is conducted on-site for both U.S. and international members, and provides a detailed evaluation of how specific research results, technologies, industry experience, and industry best practices could be applied at a given plant.

Impact By taking action against groundwater contamination, utilities will be able to allay stakeholder concerns about environmental protection. By implementing site-specific groundwater protection programs and the best technologies available for groundwater protection at nuclear power plants, utilities will be able to optimize costs and reduce waste due to groundwater monitoring and remediation: Improve relationship with communities, government, and regulatory agencies about the industry’s


Utilities participating in the 3-year membership are eligible for one site-specific Groundwater assessment once during the 3-year period. The assessment delineates actions with the largest potential benefit to the site and identifies potential gaps that, if closed, could provide economic, performance, and/or regulatory margin benefits.

How to Apply Results Members can use the real-time information from the Groundwater Strategy Group to implement improvements to their groundwater protection programs and to evaluate new technologies. For utilities participating in the 3-year membership option, the Groundwater assessment can be used to gain insights into the plant’s performance and into applying the EPRI guidance, technologies, and tools to the plant's advantage. The assessment team will develop a confidential site-specific report that details the strengths and gaps associated with the program, and highlights prioritized recommendations and potential benefits. Later, generic results and lessons learned may be compiled in program reports for industry use.

Decommissioning and Technology Development

Program Overview

Program Description Decommissioning a nuclear power plant requires expertise in safe industrial dismantling and demolition, nuclear power plant operations, radiation protection, radiological characterization, environmental protection, radwaste management, and other specialized disciplines. Because of the complex, multi-disciplinary activities involved in decommissioning nuclear power plants, experience must be captured to serve as guidance for ongoing and future decommissioning projects around the world. The Decommissioning and Technology Development program provides a structured approach for capturing lessons learned from decommissioning efforts and incorporating them into guidance for the entire nuclear power industry. Several nuclear power plants, for example, have gained experience in addressing both technical and regulatory challenges. These challenges include final site surveys and site release criteria, license termination planning, transition of regulations from operation to decommissioning plants, plant structure demolition, reactor vessel segmentation, and waste disposal.

Research Value The Decommissioning and Technology Development program develops guidance and assesses technologies that can assist in the safe, cost-effective decommissioning of a nuclear power plant. Program participants gain access to the following: Data and information leading to lower decommissioning costs and risks Enhanced planning tools to guide decommissioning Lessons learned from decommissioning activities at other plants Application results from the use of advanced technology Guidance on unresolved issues in low-level waste management, site characterization, radiation dose

modeling for site release, and license termination plans

Approach The Decommissioning and Technology Development program evaluates industry practices to distill generic guidance that nuclear plant owners can incorporate into decommissioning plans. Members use lessons-learned reports and advanced technologies to establish and implement efficient decommissioning programs at plant sites. Members also enhance technology transfer through participation in Electric Power Research Institute (EPRI) decommissioning workshops and plant-specific decommissioning support meetings.

Archive experience and lessons learned related to decommissioning regulations and technology Identify critical elements associated with maintaining an effective decommissioning plan Evaluate options for disposing wastes from decommissioning plants Develop and demonstrate advanced technologies and improved methodologies for decommissioning Anticipate and address needs arising from premature (unplanned) shutdown of nuclear units

Accomplishments The Decommissioning and Technology Development program supports nuclear power industry activities to safely and cost-effectively decommission nuclear power plants. EPRI has archived best practices, lessons learned, and technology experiences ranging from decommissioning planning and execution to final site release and license termination. This information is available to members through technical reports and through direct interaction with decommissioning experts.

Decommissioning and Technology Development - Program 41.09.02 p. 1


Developed guidance on program change management during decommissioning. The guidance defines decommissioning in terms of a sequence of major milestones, and then identifies the plant programs, associated plans and actions, and staff for each milestone.

Documented full system chemical decontamination experience at Spain's José Cabrera Nuclear Power Plant, capturing new practices and useful lessons learned.

Captured lessons learned and good practices involved in managing radiologically impacted soils, sediments, and bedrock at decommissioning nuclear power plants.

Developed decommissioning pre-planning and planning guidance reports Compiled decommissioning experience reports on U.S. decommissioning projects: Maine Yankee,

Connecticut Yankee, Rancho Seco, and San Onofre. Compiled decommissioning experience reports on reactor vessel and internal segmentation, concrete

radiological characterization and remediation, final status survey and license termination, and groundwater protection

Developed decommissioning waste management software tools Built international participation spanning France, Spain, Sweden, Japan, the United Kingdom, Italy, and

the United States

Current Year Activities Decommissioning and Technology Development program research and development for 2011 will focus on continued collection and evaluation of industry decommissioning experience to derive effective guidance for future plant decommissioning efforts. Project topics may include the following: Software for the automatic estimation of the radiological inventory for the dismantling of nuclear facilities Technical justification for the development and application of derived concentration guidance levels Decommissioning lessons learned, experiences, and their impacts on decommissioning costs and other

resources International experience in segmentation of reactor internals and vessels Software for the collection and analysis of site characterization/final status survey data to show

compliance with site release Updated base material specifications for activated metals Decommissioning waste source term Behavior of chlorine-36 and tritium in irradiated graphite wastes Lessons learned from decommissioning—factoring non-radiological issues into decommissioning

planning

Estimated 2011 Program Funding 1.0M

Program Manager Sean Bushart, 650-855-8752, [email protected]



Summary of Projects


P41.09.02.01 Decommissioning Technology Development (supplemental)

The EPRI Decommissioning Technology Development project provides technical support and technology development for cost- effective, safe, and environmentally sound decommissioning of nuclear power plants.

Decommissioning Technology Development (supplemental) (052386)

Key Research Question Nuclear power plant decommissioning requires expertise in safe industrial dismantling and demolition, nuclear power plant operations, radiation protection, radiological characterization, environmental protection, radwaste management, and other specialized disciplines. U.S. and international experience in decommissioning nuclear power plants can guide decommissioning efforts around the world. Several nuclear power plants have gained experience in both technical and regulatory challenges, such as final site surveys and site release criteria, license termination planning, transition of regulations from operation to decommissioning plants, plant structure demolition, reactor vessel segmentation, and waste disposal. Applying lessons learned and experiences from previous projects to the planning and execution of current and future projects will provide opportunities to optimize costs, increase safety, and reduce waste and impact on the environment.

Approach This project assists members in minimizing the cost and risks of decommissioning through enhanced planning, applying lessons learned from other retired plants, and using advanced technology. Guidance is developed on unresolved issues in low-level waste management, site characterization, radiation dose modeling for site release, and license termination plans. Key project objectives include anticipating and addressing the needs arising from premature (unplanned) shutdown of nuclear units and capturing the lessons learned from current decommissioning work. Best practices, lessons learned, experiences, and recommendations are documented in EPRI technical reports. This information also is available to members through direct interactions with decommissioning experts at technical workshops and through site-specific member support. As new technologies are developed to address challenges from past decommissioning projects and as new technical challenges create the need for new technologies, EPRI works with technology vendors and utilities to evaluate and demonstrate technologies for application in nuclear power plant decommissioning.

Impact The successful decommissioning of nuclear power plants demonstrates responsible management of a nuclear power plant's complete life cycle. Applying the lessons learned and experiences of previous decommissioning projects will allow current and future nuclear power plants to plan and execute successful decommissioning projects that are cost-effective, safe, minimize waste, and minimize impact on the environment, while increasing public acceptance and support for nuclear power. Potential benefits include the following: Access to experience and lessons learned related to decommissioning regulations and technology Reduced costs in developing and maintaining an effective decommissioning plan Reduced costs associated with disposing wastes from decommissioning plants Reduced implementation costs for advanced decommissioning technologies




How to Apply Results Members use lessons-learned reports and advanced technologies to establish and implement efficient decommissioning programs at plant sites. Members also enhance technology transfer through participation in EPRI decommissioning workshops and plant-specific decommissioning support meetings. Plant-specific decommissioning support meetings allow members to tailor technical support to site-specific concerns.

2011 Products



Technical Justification for the Development and Application of Derived Concentration Guidance Levels (DCGL): Derived concentration guidance levels (DCGL) are site-release criteria used by decommissioning nuclear power plants to ensure that the decommissioned site will meet the U.S. Nuclear Regulatory Commission 25 mrem/yr limit to the future user of the site. Over the years decommissioning nuclear power plants and the U.S. NRC have worked together to develop a range of future use scenarios and corresponding DCGLs for site release. This report describes the technical justification used to develop and apply these DCGLs to nuclear power plant decommissioning, final status survey, and site release.

12/23/11 Technical Report

Decommissioning Projects, Lessons Learned, Experiences and Their Impacts on Decommissioning Costs: This report will discuss the experiences with decommissioning cost estimates and the factors that impacted the actual cost of the decommissioning projects. The measures used to manage costs and reduce the impact of unexpected regulatory developments and site conditions also will be discussed. Detailed cost breakdowns by major projects and other cost categories from actual power plant decommissioning experiences will be presented. Required staffing levels during decommissioning projects also will be presented. Additionally, the measures taken by the nuclear power industry to incorporate the lessons learned from the prior decommissioning experiences into planning future decommissioning and design of new plants will be presented.





Automatic Estimation of Radiological Inventory: The Automatic Estimation of Radiological Inventory (AERI) software tool was developed by Enresa to characterize and estimate waste volumes from the decommissioning of the Jose Cabrera Nuclear Power Plant. The Electric Power Research Institute (EPRI), Enresa, and Iberinco will collaborate to upgrade and enhance this software for use by the EPRI decommissioning membership.

12/31/12 Software

International Decommissioning Experience Reports: The Electric Power Research Institute (EPRI) will continue to document the experiences and lessons learned from international decommissioning projects. These reports will add to the existing EPRI archive of experiences reports and will cover specific decommissioning tasks and full decommissioning projects.


Water Chemistry

Program Overview

Program Description Water chemistry conditions at nuclear power plants can impact corrosion rates, fuel performance, and radiation management. In light of increasing demands on chemistry staff and reductions in the number of staff, nuclear power plants are challenged to maintain effective water chemistry control. Improved water chemistry can reduce the frequency of transient fault conditions and overall impurity concentrations. However, continued improvements are needed to optimize water chemistry and balance the resulting impacts and improvements on system materials corrosion, fuel performance, and radiation fields. The Water Chemistry Program develops and updates water chemistry guidelines for nuclear reactors based on industry research and plant experience. The program also develops water chemistry optimization tools to mitigate corrosion, achieve and maintain design fuel performance standards, and minimize plant radiation fields.

Research Value The Water Chemistry Program develops technical guidance that can be incorporated into the day-to-day duties of nuclear plant chemists. Research results help nuclear plants create strategic water chemistry plans for maximizing plant availability and cost efficiency in a manner consistent with safety and regulatory requirements. Water Chemistry Program members gain access to the following: Cost-effective chemistry optimization tools and techniques to improve plant availability and safety. New chemistry applications through first-of-a-kind technology demonstrations. For example, field

demonstration of a polyacrylic acid dispersant showed a 50% reduction in corrosion product fouling, increasing steam generator availability.

Software-enabled improvements in chemistry control, diagnostic capabilities, and staff productivity. Enhanced technology transfer through plant-specific collaborations. On-site assessment support to benchmark plant chemistry controls and identify opportunities to optimize

chemistry protocols.

Approach The Water Chemistry Program combines basic and applied research with industry operating experience to develop guidance and technologies tailored to the needs of the nuclear power industry. The Program provides a comprehensive suite of water chemistry tools, including guidelines, new operating and monitoring technologies, chemistry control and assessment software, user groups, and on-site assessments. Conduct research to improve the understanding of water chemistry impacts on nuclear plant performance

and related impacts on equipment and systems. Develop scientifically and technically based water chemistry guidelines to minimize operational risks to

plant materials, maintain design fuel performance standards, and mitigate plant radiation fields. Conduct first-of-a-kind technology demonstrations targeting improved water chemistry control. Design and deploy software tools that can accurately and cost-effectively monitor and assess water

chemistry.


Accomplishments Electric Power Research Institute's (EPRI’s) Water Chemistry Program supports nuclear power industry efforts to improve water chemistry control and minimize water chemistry impacts on other plant systems and components. Water chemistry research provides members with the guidance and technologies to improve operational flexibility, reduce operations and maintenance costs, reduce dose, and reduce material degradation risks. Supported first continuous operational application of a chemical dispersant to reduce sludge

accumulation in steam generators. Dispersant use maximizes tube life and mitigates future power reduction from loss of heat transfer capability.

Conducted laboratory testing to confirm the technical feasibility of using chemical dispersants to manage steam generator deposits during the long-path recirculation cleanup process. Documented a generic qualification of a lead plant for an initial industry application and prepared a set of recommendations to guide development of an initial application plan.

Evaluated the use of septa and precoat media in controlling feedwater iron in boiling water reactors. Results will provide input to the next revision of the EPRI Condensate Polishing Guidelines.

Continued collecting water chemistry data through EPRI's Chemistry Monitoring and Assessment project to benchmark specific water chemistry regimes and optimize plant operation. As of December 2009, the pressurized water reactor (PWR) database contained 639 cycles of primary and secondary chemistry data from 67 U.S. and 80 non-U.S. PWR plants. The boiling water reactor (BWR) database includes data from all 35 U.S. and 11 non-U.S. BWR plants.

Published first revision of the BWR Shutdown and Startup Chemistry Operating Experience and Sourcebook. This sourcebook summarizes BWR good practices for controlling corrosion product transport during shutdowns, particularly refueling outages, and for startup chemistry control to minimize intergranular stress corrosion cracking.

Optimized procedures for detecting trace inorganic cations in ultrapure water using capillary electrophoresis, a method capable of rapid detection of very low concentrations of simple metallic and non-metal ions.

Improved chemistry software by incorporating the PWR and BWR Shutdown Chemistry Calculator and Analyzer into ChemWorks™ Tools; and updated the MULTEQ Database with new and revised species.

Current Year Activities Water Chemistry Program research and development for 2011 will focus on technology developments and assessments to continually improve guidance and tools for optimized chemistry control. Specific efforts will include the following: Review the boiling water reactor water chemistry guidelines and the pressurized water reactor secondary

water chemistry guidelines to evaluate gaps related to industry initiatives. Based on review, next revision of the pressurized water reactor primary water chemistry guidelines may begin.

Complete a generic safety assessment in support of elevated hydrogen application on the primary side of a four-loop Westinghouse pressurized water reactor. Coordinate field implementation of an elevated hydrogen program at a lead plant, with implementation to begin in 2012.

Revise the pressurized water reactor zinc application guidelines by updating current industry experience, identifying best practices, developing a long-term zinc injection strategy, and providing guidance for operational decision making.

Assess new plant designs and planned operation against the current Water Chemistry Guidelines. This analysis will form the basis of future activities to define chemistry guidance for new plants.

Continue development of boiling water and pressurized water reactor monitoring and assessment tools to benchmark the industry and evaluate chemistry improvement opportunities.

Begin revision of the condensate polishing guidelines for boiling water and pressurized water reactors. Develop recommendations for methods to adjust reactor water chemistry such that reduction of

radioisotope levels can be accomplished during periods of high moisture carryover.

Water Chemistry - Program 41.09.03 p. 2



Program Manager Keith Fruzzetti, 650-855-2211, [email protected]

Summary of Projects


P41.09.03.01 Chemistry Guidelines (base)

This program element provides up-to-date guidelines, develops leading technologies, and performs critical assessments in support of safe, reliable, and optimized water chemistry operation. The program’s key products are the Water Chemistry Guidelines, which are produced in collaboration with the Electric Power Research Institute (EPRI) issue programs, based on ongoing research and plant and industry experience.

P41.09.03.02 First-of-a-kind Technology Demonstrations (base)

This project implements new chemistry technologies at a plant site under controlled and monitored conditions and develops detailed application guidance for optimized application.

P41.09.03.03 Software Development (base)

In support of plant chemists, ChemWorks codes provide a consistent and reliable means for assessing chemistries that could impact component/asset management strategies.

P41.09.03.15a SMART ChemWorksTM User Group - Maintenance and Support (supplemental)

This project provides support to the 24 current users of SMART ChemWorks, including two from the BWR fleet. Plant and corporate personnel have access to the SMART ChemWorks technology through a web interface, and can monitor plant chemistry on a continuous basis. An alert system is customized for each plant.

P41.09.03.15b SMART ChemWorksTM User Group - New Installation (supplemental)

Installation of SMART ChemWorks requires a coordinated effort between EPRI engineers and plant team members. A one-time on-site meeting is required between team members to establish project scope and timelines to ensure a successful implementation plan. SMART ChemWorks requires that a data transfer tool be installed at the plant and that access be provided to the EPRI servers, after which plant personnel will have access to SMART ChemWorks through a web interface. Additional site-specific customizations are supported over the first 3 months as the model is developed and adjusted based on plant information.

P41.09.03.15d ChemWorksTM User Group (supplemental)

The ChemWorks User Group provides several mechanisms for enhancing the ChemWorks software codes and their application at nuclear plants. Through industry forums, newsletters, annual meetings (U.S. and international), and webcast sessions, EPRI technical staff support utility application of the codes and gain insight into user experience that can lead to needed software modifications and improvements.





P41.09.03.15e Chemistry Technical Strategy Group (supplemental)

The Chemistry Technical Strategy Group provides a forum for members to exchange ideas and lessons learned related to strategic management of BWR and PWR chemistry programs. The 3-year membership provides participants with all the benefits of an annual membership plus a full chemistry assessment once during the 3-year period. Participation on this basis is at a lower cost than having one assessment performed and subscribing to the Technical Strategy Group on an annual basis for 3 years.

P41.09.03.16a PWR Secondary Cycle pH Optimization (supplemental)

To help utilities minimize corrosion in the secondary cycle, EPRI has developed a process to evaluate plant pH optimization programs. EPRI experts help plant managers evaluate the status of their corrosion-product transport and local pHs throughout the secondary cycle. Evaluations include the following: A mass balance of corrosion products around the secondary

cycle to determine the corrosion source Calculation of amine concentrations and local pH values at major

locations in the secondary cycle using the EPRI Plant Chemistry Simulator

Calculation of feedwater iron concentrations Estimation of polisher run length as a function of feedwater

amine concentrations Economic analysis comparing the use of different potential

amines

P41.09.03.16d PWR Dispersant Application Support (supplemental)

This project provides plant-specific support for successful dispersant application, which can reduce steam generator fouling by as much as 50% based on plant trials and early results from application at Exelon.

P41.09.03.16e PWR Primary and Secondary Resins and Filters User Group (supplemental)

The PWR Primary and Secondary Resins and Filters User Group provides a forum for collecting industry best practices that can be used by member utilities to optimize operations with regard to demineralizer and filter performance.

P41.09.03.16f PWR Primary Zinc Application User Group (supplemental)

The PWR Primary Zinc Application User Group provides members access to an annual meeting to update members on technology developments and to share experiences and best practices. As part of the Chemistry Monitoring Assessment Program, key parameters from utilities will be tracked and trended based on cycle performance in radiation exposure and zinc injection.




P41.09.03.17a BWR Cycle Chemistry Evaluation (supplemental)

This project will assist plant operators in evaluating their cycle chemistry and in developing a roadmap for chemistry improvements by performing the following: Reviewing the technical basis for the site chemistry trending

program using EPRI guidelines and other industry standards Reviewing the effectiveness and completeness of chemistry-

related corrective actions taken during previous cycles Evaluating and trending chemistry results using EPRI-developed

tools to document behaviors Recommending site chemistry program improvements based on

trends and observations Conducting on-site walkdowns (as needed) of sample systems to

review effectiveness Assessing laboratory operations with respect to communication,

equipment operation and maintenance activities, training, and implementation of industry lessons learned

P41.09.03.17b BWR Startup Hydrogen Injection Evaluation (supplemental)

This project will assist plant operators in identifying preferred locations for early hydrogen injection during BWR startup as well as support initial planning efforts for possible plant modifications to accommodate the injection process. The evaluation team will perform the following: Review plant drawings and documents for injection locations Perform a plant walkdown to confirm preferred locations Identify plant modifications required for injection tie-ins and

services Determine space availability should hydrogen gas cylinders be

used Document the results of the evaluation and provide plant-specific

recommendations

P41.09.03.17c BWR Feedwater Iron Optimization Support (supplemental)

This project supports efforts to improve the performance assessment accuracy of individual condensate filtration/demineralizer vessels for iron optimization. Support at a given plant includes the following: Temporary corrosion product sampler setup Baseline evaluation of current sampling/analysis program Consultation with station personnel for execution of site-specific

sampling program Site-specific sampling program and sample analysis (lab analysis

generally performed on-site by site personnel) Compilation and evaluation of results Site-spec ific recommendations regarding septa, precoated

material and operating conditions to achieve feedwater iron control goals (electronic report included)




P41.09.03.17d BWR Condensate Filter User Group (supplemental)

Participation in the BWR Condensate Filter User Group provides access to an annual conference, electronic reports, newsletters, and industry alerts. Current issues include the following: Pleate d septa experience and septa quality control Septa and precoated materials optimization Use of specialty resins for cobalt removal Equipment upgrades and related issues and resolutions Optimization of iron and soluble species removal Control of sulfate releases due to pleated septa age and

attachment hardware failure Minimization of radwaste generation

P41.09.03.17f BWR Zinc User Group (supplemental)

Zinc injection has become a key technology for minimizing drywell dose rates under the highly reducing chemistry conditions established by hydrogen and noble metal water chemistry programs. While more zinc can be better for dose control, chemistry guidelines limit feedwater zinc concentrations to control the buildup of tenacious crud on the fuel cladding. The BWR Zinc User Group enables plants to accomplish the following: Share plant experiences and lessons learned with zinc

application Review the BWR chemistry monitoring database for trends and

updates Monitor zinc performance results based on available chemistry,

radiation field, and fuel surveillance data Communicate issues and improvement plans for active and

passive zinc injection systems Identify needed research involving zinc addition

Chemistry Guidelines (base) (052415)

Key Research Question Improving water chemistry has contributed to a reduction in the frequency of transient fault conditions and an overall reduction in impurity concentrations. However, continued improvements are sought to optimize water chemistry and balance the resulting impact on system materials corrosion, fuel performance, and radiation fields. In addition, advanced technology is needed to improve water chemistry control.

Approach The program’s key products are the Water Chemistry Guidelines, which are produced in collaboration with the EPRI issue programs, based on ongoing research and plant and industry experience. Each guideline is a technical consensus document developed by industry experts to optimize water chemistry programs and control methods, thereby maximizing the long-term availability and safe operation of nuclear power plants. Guidelines are formally reviewed on an annual cycle and generally revised on a 4-year cycle. Much of the work involves technology developments needed to improve guidelines, including water chemistry control methods, improved monitoring techniques, and chemical additives to control corrosion, reduce radiation fields, and maintain fuel performance.


Impact Improved water chemistry technology to support corrosion mitigation, fuel performance, radiation

management, water chemistry control methods, and improved monitoring techniques Optimization methods that nuclear plants can use to create strategic water chemistry plans customized

for maximizing plant availability and cost-efficiency in a manner consistent with safety and regulatory requirements

How to Apply Results Technology development and Water Chemistry Guidelines can be applied by members to implement improved monitoring techniques, as well as optimize strategic water chemistry plans. These products must be read, understood, and implemented by the chemistry community.

2011 Products



BWR Chemistry Monitoring and Assessment - Sampling: Summarizes sampling/analysis practices (frequencies, analysis methods, feedwater sample line velocity) by the 49 participating boiling water reactors (BWRs) for key reactor water parameters (for example, chloride, sulfate, metals, Co-60, and Zn-65) and feedwater metals.


BWR Chemistry Monitoring and Assessment - Condensate Filtration: Two areas are covered. The first area covers deep bed condensate demineralizer ion exchange resins, providing details on the resin types applied, resin properties, quantities used, and practices such as the use of anion resin underlays at plants with both Deep Bed Only and Filter + Deep Bed condensate polishing systems. The second area covers condensate filter demineralizer precoat materials, providing details on the composition of the precoat materials applied (for example, all powdered resin, resin/fiber mixtures, and mixtures containing weak acid resins), properties, and quantities used.


BWR Chemistry Monitoring and Assessment - Mitigation Performance Indicator: Provides annual updates of industry status and progress in meeting goals for chemistry programs designed to achieve intergranular stress corrosion cracking (IGSCC) mitigation of reactor external piping and internal components and to meet criteria for piping inspection relief.


PWR Chemistry Monitoring and Assessment - Primary Chemistry: Provides information on operating and shutdown chemistry, chemistry control methods, and plant changes from participating utilities. Includes evaluations and specific assessments on various chemistry control programs.


PWR Chemistry Monitoring and Assessment - Secondary Chemistry: Provides information on operating chemistry, chemistry control methods, and plant changes from participating utilities. Includes evaluations and specific assessments of various chemistry control programs.


PWR Zinc Application Sourcebook, Revision 1: Revision 1 to the PWR Zinc Application Guidelines based on research developments, new guidance, and plant experience since the initial publication in 2006. This revision also will focus on additional information and guidance from plants already implementing primary zinc injection.







BWR Chemistry Monitoring and Assessment - Shutdown Chemistry and Dose Rates: A summary of shutdown chemistry and drywell radiation dose rates. This provides the only known detailed industry compilation of such data. This is updated annually. A summary of shutdown drywell radiation dose rates. This provides the only known detailed industry compilation of such data. This is updated after spring and fall refueling outages, when new end-of-cycle drywell radiation survey data are normally available. (February and June)


Generic Safety Assessment for Application of Elevated Hydrogen on the Primary Side of a Four-loop Westinghouse Pressurized Water Reactor: Completion of all necessary safety-related analyses as well as identification of any necessary updates to plant Technical Specifications based on recommendations from EPRI Report 1015017, Evaluation of Altering the Hydrogen Concentration for Mitigation of Primary Water Stress Corrosion Cracking.


BWR Chemistry Monitoring and Assessment - Chemistry Summary: Captures the status of important reactor water, feedwater, and condensate chemistry parameters and relates them to plant design and operating factors.


Waste Class B/C Reduction Guide for Chemistry Managers: Provides technical guidance on Waste Class B/C Reduction Strategies and addresses impacts on chemistry programs.


2011 Interim Review of the Pressurized Water Reactor Secondary Water Chemistry Guidelines, Revision 7: Complete and document the required annual review of the PWR Secondary Water Chemistry Guidelines in light of industry research and plant experience.


First-of-a-kind Technology Demonstrations (base) (052418)

Key Research Question Implementing new chemistry technologies typically requires demonstration at a plant site under controlled and monitored conditions, informed by EPRI expertise and management. Data and information from these demonstrations are instrumental in facilitating technology commercialization.

Approach The project develops and tests new chemistry additives, new analysis methods, new instrumentation, and application guidelines on how to efficiently use new chemistry technology. Field testing is an essential part of the development process, as many new developments would not be implemented without an initial plant demonstration.

Impact Provides experience and plant data from the first demonstration of new technologies Provides industry application guidelines based on lessons learned

How to Apply Results Members gain access to data and guidance that can be used to inform technology application at other facilities.


2011 Products



Capillary Electrophoresis Sourcebook for BWRs: A comprehensive sourcebook for optimized application of capillary electrophoresis for boiling water reactor (BWR) feedwater anion and cation measurements.


Water Chemistry Assessment for New Plants - continuation: Assessment will continue, per the work in 2010, for other new designs (the ABWR and AP1000 designs were completed in 2010). The GE-H economic simplified boiling water reactor (ESBWR), Areva U.S. EPR, and MHI U.S. advanced pressurized water reactor (APWR) will be investigated. The work provides an assessment of plant design and planned operation against the current Water Chemistry Guidelines to identify gaps and the path forward for any needed revised guidance. This project is will be conducted In collaboration with the Advanced Nuclear Technology Program.


Software Development (base) (052419)

Key Research Question Nuclear power plants must meet strict system performance guidelines as specified by EPRI chemistry guidelines, the Institute for Nuclear Power Operations (INPO), and the Nuclear Regulatory Commission (NRC) to ensure pressure boundary integrity, fuel performance, and minimized radiation fields. Members require calculation tools that are robust and consistent with industry practice to manage their programs within the specified guidance.

Approach Chemistry managers, engineers, and technicians rely on EPRI ChemWorksTM tools for developing and optimizing their chemistry programs. The results from EPRI ChemWorks software programs are used in a variety of chemistry system evaluations, including high-temperature pH calculations for reactivity control in pressurized water reactors, hideout return evaluations for the secondary side of the steam generators, and estimating corrosion product inventory during shutdown. From these predictions, chemistry personnel can assess corrosion control, guide life-cycle strategies, and optimize the overall cost of the chemistry program. This research area ensures the ChemWorks tools reflect the latest industry operating experience and have functionality commensurate with industry needs.

Impact ChemWorks codes provide members with the opportunity to evaluate, optimize, and train plant personnel

on various chemistry programs and strategies. Continued optimization of ChemWorks can provide members with direct cost reductions. ChemWorks codes allow assessments of chemistries that impact component/asset management

strategies.

How to Apply Results Chemistry personnel can implement the ChemWorks codes to evaluate the effects of chemistry on a variety of situations, including corrosion mitigation, amine optimization, and resin life management.



2011 Products



ChemWorks Tools version 4.0: Additional functionality and improvements will be implemented based on the 5-year plan. 12/23/11 Software

MULTEQ Database version 7.0: Incorporation of significant updates based on the recommendations and work of the MULTEQ Database Committee. This activity will be conducted in collaboration with the Fuel Reliability and Steam Generator Management Programs.


SMART ChemWorksTM User Group - Maintenance and Support (supplemental) (007452)

Key Research Question Chemistry staffs in the nuclear power industry continue to shrink as a result of economic pressures and the aging work force. Highly skilled personnel spend too much time completing mundane but necessary tasks, while the backlog of important, but less time-sensitive obligations, continues to grow.

Approach EPRI’S SMART ChemWorks™ is a real-time plant water chemistry monitoring and advisory system that aids chemistry staff by completing routine chemistry analysis, identifying early indications of adverse plant chemistry, and alerting personnel to emerging issues. Continued maintenance and support is required to ensure SMART ChemWorks incorporates the latest operating experience and is broadly applicable across the global nuclear industry.

Impact SMART ChemWorks uses sophisticated mathematical models and pattern-recognition techniques to identify abnormal conditions. Once installed, plant managers can look forward to increased efficiency and better use of plant personnel, reduced risk of plant components damage, increased chemistry monitoring coverage and early detection of chemistry problems, improved control of chemical injection systems, reduced cost due to optimization of grab sample frequencies, and reduced “out-of-specification” time for instrumentation.

How to Apply Results The SMART ChemWorks system relies on continuous chemistry monitoring (24 hours a day, 7 days a week) and real-time evaluation of plant chemistry conditions, accurate and prompt diagnosis of abnormal chemistry, and instantaneous alerting to adverse chemistry trends through an email/paging system and web page alerts. The output from SMART ChemWorks represents technical intelligence on which plant chemists can take action as appropriate.



SMART ChemWorksTM User Group - New Installation (supplemental)

Key Research Question Chemistry staffs in the nuclear power industry continue to shrink as a result of economic pressures and the aging work force. Highly skilled personnel spend too much time completing mundane but necessary tasks, while the backlog of important, but less time-sensitive obligations, continues to grow.

Approach EPRI’S SMART ChemWorks™ is a real-time plant water chemistry monitoring and advisory system that aids chemistry staff by completing routine chemistry analysis, identifying early indications of adverse plant chemistry, and alerting personnel to emerging issues. Installation of SMART ChemWorks requires a coordinated effort between EPRI engineers and plant team members. A one time on-site meeting is required between team members to establish project scope and timelines to ensure a successful implementation plan.

Impact SMART ChemWorks uses sophisticated mathematical models and pattern-recognition techniques to identify abnormal conditions. Once installed, plant managers can look forward to increased efficiency and better use of plant personnel, reduced risk of plant components damage, increased chemistry monitoring coverage and early detection of chemistry problems, improved control of chemical injection systems, reduced cost due to optimization of grab sample frequencies, and reduced “out-of-specification” time or instrumentation

How to Apply Results The SMART ChemWorks system relies on continuous chemistry monitoring (24 hours a day, 7 days a week) and real-time evaluation of plant chemistry conditions, accurate and prompt diagnosis of abnormal chemistry, and instantaneous alerting to adverse chemistry trends through an email/paging system and web page alerts. The output from SMART ChemWorks represents technical intelligence on which plant chemists can take action as appropriate.

ChemWorksTM User Group (supplemental) (006521)

Key Research Question Economic pressures, work demand and reduced staffing require chemists to spend more time in the field and less time evaluating changes in the chemistry programs. ChemWorks™ provides users with a simple set of tools to quickly evaluate plant chemistry controls.

Approach The ChemWorks User Group provides several mechanisms for enhancing the ChemWorks software codes and their application at nuclear plants. Through industry forums, newsletters, annual meetings (U.S. and international), and webcast sessions, EPRI technical staff gain insight into user experience that can lead to needed software modifications and improvements.

Impact ChemWorks uses sophisticated mathematical models to aid plant chemists in developing optimal chemistry programs and applying appropriate chemistry controls that support long-term equipment reliability. Users Group members provide input into continued software improvements that reflect industry needs and experience.

How to Apply Results User Group members receive direct support via meetings and product development as well as individual training on ChemWorks™ codes.



Chemistry Technical Strategy Group (supplemental)

Key Research Question The Chemistry Technical Strategy Group provides a forum for discussing technical issues regarding the strategic management of BWR and PWR chemistry programs. Emerging chemical issues challenge plant staff in both the day-to-day and long-term management of chemistry programs. This group enables members to exchange ideas and information related to emerging issues as well as lessons learned. The Chemistry Technical Strategy Group is available in 3-year and 1-year membership options. The 3-year membership inlcudes 1 full cycle chemistry assessment.

Approach The Chemistry Technical Strategy Group, available as a 3-year membership, enables plant chemists from BWRs and PWRs to share best practices and discuss chemistry challenges. Meetings will include general sessions to address common chemistry issues and break-out sessions to address issues specific to BWRs and PWRs. In addition to the member forum for sharing lessons learned, participants are eligible for annual technical consulting and a focused assessment during the 3-year membership period. The chemistry assessment entails an expert review of a plant’s chemistry program and recommendations on how EPRI technology can assist in addressing plant-specific issues and improving chemistry performance.

Impact Participation in the Chemistry Technical Strategy Group keeps members abreast of emerging issues in the rapidly changing climate surrounding chemistry issues and provides members with a forum for technical exchange. On-site consulting time provides expert support for specific plant or corporate project requests focused on long-term strategic planning.

How to Apply Results Participation in periodic webcasts and meetings keeps members abreast of emerging issues. Annual on-site consultation time and the focused chemistry assessment are used to ensure EPRI guidance is applied to emerging and critical plant-specific issues. The annual consultation time is used to address more narrowly focused issues, while the full assessment provides a broader programmatic perspective. Both venues provide plant-specific recommendations to maximize plant benefits.



PWR Secondary Cycle pH Optimization (supplemental) (064149)

Key Research Question Materials corrosion in the condensate, feedwater, and drain systems of pressurized water reactors (PWRs) generates a significant amount of corrosion products in the secondary cycle. Generally, these corrosion products are transported into steam generators (SGs) and deposited on tubing surfaces, tubesheets, and tube support plates. These corrosion products can act as sites for ionic impurity concentration in restricted areas, leading to corrosion of steam-generator tubing.

Approach This project will develop a plant-specific corrosion-products mass balance file based on current plant equipment. The mass balance results are evaluated to assess how changes in the type and concentration of amines could optimize secondary cycle pH.

Impact Local pH strongly affects corrosion of several different types, including intergranular attack and stress corrosion cracking. The right choice of amines and proper operation of a pH control additive will reduce SG fouling and deposit consideration. Secondary-cycle pH optimization aids in the selection of optimum amines for the plant, thus minimizing corrosion product transfer.

How to Apply Results Members receive a detailed report discussing the status of corrosion-product transport and local pHs in the secondary cycle, along with recommendations and supporting documentation for optimum amine selection. Application of these results will support an optimized feedwater iron control.

PWR Dispersant Application Support (supplemental) (061414)

Key Research Question Steam generator deposits (fouling) can inhibit heat transfer, lead to thermal-hydraulic instabilities through blockage of tube supports, and create occluded regions where corrosive species can concentrate along tubes and in tube-to-tube support plate crevices. Steam generator performance is compromised not only by formation of an insulating scale, but by the removal of tubes from service due to corrosion. Dispersant application is a very promising technology for significantly reducing steam generator fouling. A number of utilities are pursuing dispersant applications using the PWR Dispersant Application Sourcebook (1015020) and additional plant-specific support as provided through this project. EPRI also can assist in evaluating newer dispersant application technologies. For example, dispersants can be used for increasing cleanup of corrosion products during steam generator wet layup or during long-path recirculation cleanup of the condensate/feedwater piping prior to power operation.

Approach This EPRI project will support assessment and application of dispersant for steam generator fouling mitigation. Specific work will be determined in collaboration with the funding utilities, depending on need and plant-specific concerns. Assessments could involve the following: Plant-specific materials qualification assessment Chemistry operation and monitoring with dispersant injection (application plan)



Evaluation of steam generator thermal performance, including a baseline evaluation prior to dispersant injection and customization of a thermal performance tracking spreadsheet

Preparation of materials needed to support a 10CFR50.59 evaluation

Impact Successful on-line application of dispersant could reduce steam generator fouling by as much as 50% based on results from previous plant trials. Other applications during steam generator wet layup and startup could provide additional benefits.

How to Apply Results The products from this project are assessments, site-specific reports, and/or recommendations in support of application at a lead pressurized water reactor (PWR) unit. EPRI works with each funding utility to identify and deliver what is needed for successful application, within funding constraints.

2011 Products



Dispersant application assessments: Assessments completed based on member input and needs. 12/22/11 Technical

Resource

PWR Primary and Secondary Resins and Filters User Group (supplemental) (063963)

Key Research Question To minimize ionic and particulate impurity transport in the reactor coolant and the steam generators, PWR plants employ a wide variety of chemical and volume control systems, condensate polishers, and steam generator blowdown demineralizer systems and use different strategies to operate theses systems. The PWR Primary and Secondary Resins and Filters User Group provides a forum for sharing industry experience that can improve and accelerate plant activities to reduce impurity transport.

Approach The User Group will track operating experience and perform focused studies on specific industry issues. Technology reviews and comparisons of plant practices, for example, will help plants improve reactor coolant system cleanup, reduce iron transport to the steam generators, and reduce low-level waste generation. Review and benchmarking activities will be performed through annual meetings, with products defined by members. The PWR Primary and Secondary Resins and Filters User Group will function in a manner similar to the BWR Condensate Filter User Group, which was formed in the late 1990s to address problems and optimization issues with condensate filters.

Impact This group will identify and improve the application of demineralizers, filters, and membrane technologies applied in PWRs by assisting in the evaluation of filter and ion exchange performance and by sharing of information concerning filter design, novel resin use, and other general operating experience.

How to Apply Results Application of the practices identified as optimal will be performed at member utilities on an as-needed basis. A key product target will be development of a demineralizer and filter sourcebook, from which users can reference industry best practices.



PWR Primary Zinc Application User Group (supplemental) (061425)

Key Research Question Many EPRI reports have been published over the years regarding the effectiveness of zinc addition for primary water stress corrosion cracking mitigation (for both initiation and crack growth rate), including the development of PWR Zinc Application Guidelines in 2006. However, there remains a strong need for plant personnel to exchange information first-hand and provide focused input on future research activities.

Approach At each user group meeting, members provide an update on their plant or utility's zinc injection program status. These updates address all aspects of a zinc injection program, including planning, implementation, scheduling, and long-term strategy. The user group members also receive updates related to pressurized water reactor (PWR) reactor coolant system zinc addition.

Impact The Zinc User Group provides a forum for members to benchmark new and existing zinc injection programs throughout the industry. Sharing of lessons learned and annual updates related to EPRI research programs associated with zinc addition will provide utilities with valuable information to guide planning and implementation.

How to Apply Results The plant experiences and challenges shared through the Zinc User Group provide learning opportunities for the entire industry. Plant personnel are able to bring these lessons learned back to their plants for implementation for further analysis.

BWR Cycle Chemistry Evaluation (supplemental)

Key Research Question Plant chemistry programs are instrumental in maintaining nuclear plant reliability and availability. In-depth technical reviews of a boiling water reactor's (BWR's) cycle chemistry can provide specific recommendations aimed at optimizing chemistry control, enhancing plant operations, mitigating stress corrosion cracking, ensuring fuel reliability, and reducing radiation exposure and radioactive waste generation.

Approach This project will assist plant operators in optimizing BWR chemistry. Plant-specific evaluations of the most recent operating cycle will be evaluated and recommendations will be made to support enhanced plant operation.

Impact Improved plant operation and cycle chemistry control Reduced costs for chemicals, additives, and other consumables Reduced stress corrosion cracking of susceptible components Lower radiation dose and radioactive waste Improved fuel performance

How to Apply Results By applying the detailed report recommendations, plant operators can maximize plant performance and aid in extending plant life. Recommendations may include optimization of condensate treatment and reactor water cleanup systems, additive chemistries (such as zinc addition, hydrogen addition, and/or noble metal addition), and improvements in chemistry sampling and analysis programs.



BWR Startup Hydrogen Injection Evaluation (supplemental)

Key Research Question All U.S. boiling water reactors (BWRs) are currently injecting hydrogen to mitigate stress corrosion cracking of susceptible components. However, hydrogen injection is typically not injected until the plant is at 5% power or more. Earlier injection of hydrogen, specifically at plant startup until typical feedwater hydrogen injection can be initiated, could provide additional mitigation value. Utilities interested in this technology may need support in implementing this technology.

Approach This project will assist plant operators in identifying preferred locations for early hydrogen injection during BWR startup as well as support initial planning efforts for possible plant modifications to accommodate the injection process.

Impact Early identification of physical plant constraints Shorter lead times for system implementation Improved project scoping and accuracy Tighter budget and schedule control

How to Apply Results Plant-specific recommendations as documented in the final report will detail how early hydrogen injection can be optimized at the plant.

BWR Feedwater Iron Optimization Support (supplemental) (062736)

Key Research Question The BWR Water Chemistry Guidelines recommend feedwater iron concentration in the range of 0.1-1.0 ppb for plants operating with zinc addition and reducing chemistry conditions (hydrogen water chemistry as well as noble metal application technologies). Operation in the range of 0.1-0.5 ppb is further encouraged to reduce the amount of zinc needed for dose rate control purposes. While many plants have been able to achieve these conditions, plants with non-optimized condensate filtration/demineralizer systems may need additional support to reduce feedwater iron inputs to these recommended levels.

Approach This project supports the optimization of feedwater iron by analyzing current operation and performance and providing recommendations to optimize future performance of the condensate filtration demineralizer system.

Impact The optimization of feedwater iron input results in enhanced plant operation: Reduced radwaste generation and exposure Improved fuel performance and reliability Optimized stress corrosion cracking mitigation technologies (such as hydrogen addition and noble metal

injection)



How to Apply Results This project provides direct consultations with plant staff and a final electronic report detailing the project work and recommendations. Implementation of the report recommendations regarding precoat usage and dosage on septa will be detailed. Application of these results will support optimized feedwater iron control.

BWR Condensate Filter User Group (supplemental) (006388)

Key Research Question While nuclear industry experience with condensate filtration is extensive, the variety of septa types, precoat materials, vessel designs, backwash, and precoating methods present application challenges. Open dialogue among users to share experiences can drive excellence in performance, especially with the ever-increasing emphasis on water chemistry limits and impacts on fuel performance, stress corrosion cracking mitigation, and radwaste and exposure reduction.

Approach The BWR Condensate Filter User Group supports optimization of water chemistry filtration technologies to control important chemistry parameters. Through annual User Group conferences, electronic reports, newsletters, and industry alerts, members gain access to industry experience that can be tailored to individual plant applications. A database of operational and technical information is maintained to address condensate filter system challenges.

Impact This User Group has been successfully supporting the nuclear industry for more than 10 years. The database of operational and technical information provides extensive insight into the understanding of condensate filter system challenges. Lessons learned and implemented from the User Group results in reduced radwaste and worker exposure, optimized fuel performance, and enhanced overall plant operation.

How to Apply Results Members of the BWR Condensate Filter User Group apply the experience of other member's situations to their own utility. Special studies also are made available on a plant-specific basis and shared with the membership group.

BWR Zinc User Group (supplemental)

Key Research Question Many Electric Power Research Institute (EPRI) reports have been published over the years regarding plant experience with zinc injection on drywell shutdown radiation dose rates and laboratory investigations on zinc effectiveness for intergranular stress corrosion cracking (IGSCC) mitigation for sensitized Alloy 600, 304 stainless steel, and Alloy 182 weld metal. However, given the need to balance the chemistry and radiation benefits of zinc with fuel concerns regarding tenacious crud deposits and the potential effects of crud/oxide spallation, there remains a strong need for plant personnel to exchange information first-hand and provide focused input on future research activities. These activities can be supported through a BWR Zinc User Group.

Approach The BWR Zinc User Group will enable participants to update peers on plant/utility zinc injection programs. Updates will address all aspects of the zinc program, including planning, implementation, scheduling, and long-term strategy. User Group members will receive updates related to radiation field control, reactor water and feedwater zinc data trends, reactor water Co-60(s)/Zn(s) ratio control, Co-60 trends, and effectiveness of Zn injection following noble metal applications and reactor recirculation piping chemical decontaminations.




Impact The BWR Zinc Users Group will provide a forum for utilities to benchmark zinc injection programs throughout the industry. Sharing of lessons learned and annual updates related to EPRI research programs associated with zinc addition will provide utilities with valuable information to guide planning and implementation.

How to Apply Results The EPRI BWR Zinc User Group will provide members with access to industry data relevant to zinc injection, data correlations, an annual User Group meeting to share and benchmark plant experiences related to zinc injection, and discussion of EPRI zinc-related projects and work prioritization.

Equipment Reliability

Develop the technical basis for preventing fuel failures through research into failure mechanisms, analysis of failed fuel rods, and development of fuel reliability guidelines.

Programs Nuclear Maintenance Application Center (NMAC) Plant Support Engineering (PSE)

Nuclear Maintenance Application Center (NMAC)

Program Overview

Program Description Maintenance practices at nuclear power plants play a critical role in a unit’s ability to achieve or maintain high reliability and capacity factor levels. To this end, maintenance practices must be continuously reviewed and updated based on industry operating experience and emerging issues. The Nuclear Maintenance Application Center (NMAC) develops maintenance guides and coordinates worldwide technology transfer to drive improvements in nuclear maintenance activities. The program’s technical guides, user groups, and workshops reflect best practices and engineering judgment gathered from nuclear plant experience, providing actionable maintenance activities that lead to lower costs and higher reliability. NMAC also conducts research to identify maintenance advances with the potential to produce substantial plant performance improvements. These activities require accurate assessment of plant needs and tend to be strategic, complex, and longer in duration.

Research Value Research results from the Nuclear Maintenance Application Center provide knowledge and guidance that enable nuclear plants to reduce operations and maintenance costs and improve equipment reliability. NMAC participants gain access to the following: Industry data and best practices from more than 35 plant visits each year, helping members effectively

implement maintenance program improvements More than 250 maintenance guides for nuclear equipment and systems, which provide source documents

for improved procedures and training packages A worldwide network of maintenance professionals to help participants resolve nuclear plant maintenance

issues Quicker identification of failure-related root causes through the use of the NMAC Hotline, NMAC staff, and

other members A broader range of maintenance solutions with reduced implementation risks due to collaboration with

subject matter experts from all over the world Templates for establishing defensible preventive maintenance practices and intervals for key components

and systems

Approach The Nuclear Maintenance Application Center conducts near-term and long-term research to drive maintenance improvements at nuclear plants. Near-term research focuses on maintenance methods and guidance that can help reduce operations and maintenance costs and improve equipment reliability. Long-term research focuses on new methods and approaches that drive sustained improvements to plant equipment, processes, and practices. Identify maintenance program improvements that can increase equipment reliability and plant

performance Provide technical assistance in defining, implementing, and sustaining high-quality maintenance

programs Develop practical and actionable maintenance guidelines for various equipment and systems Ensure maintenance best practices are informed by industry experience and sound engineering

judgment, leading to more consistent application Leverage worldwide maintenance expertise and experience through user groups, workshops, and

databases

Nuclear Maintenance Application Center (NMAC) - Program 41.05.01 p. 1


Accomplishments Electric Power Research Institute's (EPRI’s) Nuclear Maintenance Application Center distills global operations and maintenance experience into actionable guidance for nuclear plant systems and components. Lessons learned from nuclear plants around the world are incorporated into industry- and vendor-specific technical guidance. Developed new or revised maintenance and process guides for key plant systems and equipment,

including feedwater system, the generator portion of emergency diesel generators, oil lubrication systems, and heat exchangers.

Evaluated alternatives to wire brushes for routine nuclear plant maintenance. Report recommends alternatives, with emphasis on tooling requirements, product speeds, temperature limitations, and brush applications.

Updated EPRI guidance on nuclear plant fluid sealing programs to incorporate lessons learned from field experience. The update recommends a philosophical shift away from "fluid leakage" and toward "fluid sealing" that would integrate procedures, materials, training, and management support.

Reviewed diagnostic techniques to monitor printed circuit board performance. Identified proven techniques that can enhance a plant’s preventive maintenance program for printed circuit boards.

Updated the Preventive Maintenance Basis Database (Version 2.1). This product enables utility engineers to readily access technically applicable and cost-effective preventive tasks.

Coordinated industry technical response to emerging issues on gas accumulation, lessons learned from power uprates, and foreign material methods and procedures.

Sustained industry engagement and critical issue awareness through equipment workshops and user group meetings.

Developed web-based technical training on emerging and high-priority nuclear maintenance issues (systematic troubleshooting, isolated phase bus maintenance, and expansion joint maintenance).

Current Year Activities NMAC Program research and development for 2011 will focus on updates to key equipment maintenance guides, greater outreach to domestic and international participants, and focused attention to emerging industry maintenance issues. Specific efforts will include the following: Develop guidelines on maintenance performance indicators to complement guidelines for assessing

nuclear plant maintenance programs Provide guidance for cooling tower inspection and maintenance in response to industry failures Develop comprehensive guidance regarding seal selection and maintenance of reactor coolant pumps

and reactor recirculating pumps Develop maintenance guide for control rod drive hydraulic systems Develop diagnostics for rotating equipment health monitoring using torsional vibration indicators Conduct workshops and manage user group meetings to facilitate knowledge capture and sharing of

lessons learned


Program Manager Martin Bridges Jr., 704-595-2672, [email protected]



Summary of Projects


P41.05.01.01 Equipment Issues and Maintenance Guides (base)

NMAC maintenance guides provide specific technical information and human performance information, contained as component descriptions, failure mode identification, troubleshooting information, preventive and predictive maintenance advice, and detailed specific maintenance tasks. The guidelines collect the best available information from industry subject matter experts and from experienced equipment service personnel.

P41.05.01.02 Operations and Maintenance Procedures (base)

This project provides access to NMAC expertise that can be applied to plant operation and maintenance concerns through plant visits, on-site assistance, phone/email interaction, and other mechanisms.

P41.05.01.03 Equipment Issues Maintenance Guidelines (supplemental)

NMAC maintenance guides provide specific technical information and human performance information, contained as component descriptions, failure mode identification, troubleshooting information, preventive and predictive maintenance advice, and detailed specific maintenance tasks. The guidelines collect the best available information from industry subject matter experts and from experienced equipment service personnel.

P41.05.01.04 Maintenance Process Guides (supplemental)

NMAC process guides provide technical and human performance information—in the form of process descriptions, process flow, and specific task guidance—to enable power plant staff personnel to develop training packages, work procedures, and work planning packages.

P41.05.01.05 User Groups (supplemental)

This project supports equipment-specific user groups and forums where experienced plant and NMAC staff can help newly assigned individuals more quickly and competently satisfy their new responsibilities. The effectiveness and usefulness of each NMAC user group is routinely reviewed by the NMAC Steering Committee to ensure ongoing value.

P41.05.01.05b Member Requested Support (supplemental)

NMAC offers assistance to members in evaluating the extent to which various NMAC products can provide value to their organization. Such assistance may include training, software implementation support, work package planning, and on-line maintenance assessment.

P41.05.01.05c EPRI MOV Performance Prediction Methodology User Group (EMPUG) (QA) (supplemental)

The EPRI MOV PPM User Group (EMPUG) provides ongoing code maintenance, user technical support, and training. EMPUG supports users of the EPRI MOV Performance Prediction Methodology and the EPRI MOV Performance Prediction Program.

P41.05.01.05d Maintenance Rule User Group (MRUG) (supplemental)

The Maintenance Rule Users Group (MRUG) provides a forum for information exchange between participants to resolve technical issues from baseline inspections and ongoing revisions to the maintenance rule. MRUG develops technical guides and compiles good practices that can reduce implementation costs, increase consistency among participants, leverage rule activities to improve plant performance, and reduce vulnerability to regulatory compliance issues.

P41.05.01.05e Preventive Maintenance Basis Database (PM Basis) (supplemental)

The Preventive Maintenance Basis Database (PMBD) User Group serves as the primary source for input to guide new and revised functionality for the database. Members share experience with the database and suggest new component types that may need to be developed.



Equipment Issues and Maintenance Guides (base) (052441)

Key Research Question The performance and reliability of nuclear plant systems and components depend on component design and the effective planning and application of maintenance. Maintenance strategies, fully informed by operating experience and technology advances, can result in improved equipment reliability, lower operating costs, and higher overall plant reliability.

Approach This program area identifies and addresses important maintenance and equipment issues by conducting more than 30 plant visits each year and compiling data from vendors, the Institute of Nuclear Power Operations (INPO), and other industry sources. The project develops guides to aggregate relevant diagnostic and mitigating technical advice for addressing key maintenance issues. These guides include problem identification, troubleshooting information, preventive and predictive maintenance advice, and detailed specific maintenance tasks, contained as applicable component descriptions and tutorials, application advice, and failure modes. NMAC equipment issues guides typically represent industry consensus positions on important items and occasionally provide technical and tactical support for accomplishing strategic industry initiatives.

Impact Potential benefits form this project include the following: Reduce operations and maintenance costs Improve equipment reliability Improve access to technical and human performance information Develop maintenance guidelines that represent industry consensus Provide technical support for strategic industry initiatives

How to Apply Results NMAC produces four to six equipment issues guides each year. These are sent directly to the maintenance manager, the NMAC site coordinator, and the technical library at each NMAC member plant. Members also can access NMAC staff through a phone and email hotline to respond to emergent plants questions and needs. The NMAC portion of www.epri.com contains all NMAC guides in full text, downloadable, and word searchable.

Operations and Maintenance Procedures (base) (065807)

Key Research Question Operations and maintenance practices constitute one of the three principal areas that drive nuclear plant performance, along with equipment and structures and processes. Bringing essential information to the point of decision-making can drive successful plant and fleet performance. Access to equipment information and personnel knowledgeable in operations and maintenance practices from outside one's immediate plant or company can provide useful perspective and insight.

Approach NMAC expertise is applied to plant operation and maintenance concerns through direct phone and email interaction, routine plant visits to talk with maintenance and engineering personnel, assistance with selected plant assessments/evaluations, and specific field response when plants face emergent issues. EPRI staff also use this expertise and their many contacts to develop reports and guidelines that address new or improved methods and processes to accomplish tasks necessary to successful operation of nuclear power facilities.



NMAC staff members travel to more than 30 plants each year to discuss recent NMAC guides and plant maintenance problems. NMAC maintains a (1-800) hotline to respond to plant questions and needs. NMAC staff members pursue plant needs through interaction with the broader NMAC network of member plant personnel and industry vendors. NMAC has primary contacts in more than 100 domestic and international nuclear facilities. The NMAC portion of www.epri.com contains a significant amount of maintenance-related information, including all NMAC guides in full text, downloadable, and word searchable. Additionally, NMAC members share the collective knowledge and the information obtained through their interactions in newsletters, webcasts, and user group meetings.

Impact Potential benefits from this project include the following: Reduce maintenance costs Improve equipment reliability by providing specific information and advice on specific plant problems Access expertise and resources that serve as a complement to member companies’ in-house staffs Move more quickly to root-cause identification, identify a broader range of solutions, and target

implementation risks associated with selected corrective action paths

How to Apply Results Members apply expertise and guidance recommended by NMAC experts to specific operations and maintenance issues. Guidance may include problem identification, troubleshooting information, preventive and predictive maintenance advice, and detailed specific maintenance tasks, contained as applicable component descriptions and tutorials, applications concerns and advice, and failure modes.

Equipment Issues Maintenance Guidelines (supplemental) (052443)

Key Research Question The performance and reliability of nuclear plant systems and components depend on component design and the effective planning and application of maintenance. Maintenance strategies, fully informed by operating experience and technology advances, can result in improved equipment reliability, lower operating costs, and higher overall plant reliability.

Approach The Nuclear Maintenance Application Center (NMAC) concentrates industry efforts aimed at improving nuclear plant maintenance. NMAC produces materials and services that support component engineers, maintenance supervisors, technicians, system engineers, and design engineers. NMAC maintenance guides distill the experience of industry maintenance professionals into practical technical products, providing proven maintenance techniques and methods. NMAC produces numerous guides on priority issues such as circuit breakers, pumps, motors, valves, and other equipment.

Impact Directly addresses day-to-day plant maintenance activities and a wide array of priority issues Reflects an industry consensus approach to in-plant problem-solving Distills nuclear plant maintenance expertise and experience into practical technical products Provides multiple dissemination channels to members, including technical reports, newsletters,

conferences and workshops, user groups, webcasts, hotline support, and the NMAC website Focuses input from broad domestic and international participation



How to Apply Results NMAC guides and services are used by members as the basis for craft training, the detailed instructions in maintenance procedures, and the foundation of troubleshooting plans and general references. This information is disseminated to member companies through technical reports, newsletters, conferences and workshops, user groups, webcasts, telephone/fax/e-mail hotline support, and the NMAC web site. To support international technology transfer, NMAC has translated the Table of Contents, List of Tables, and List of Figures of each guide into Japanese, French, and Spanish since 2004. Beginning in 2006, Chinese also was added.

Maintenance Process Guides (supplemental) (052444)

Key Research Question Nuclear power plant performance can fluctuate due to personnel turnover and inconsistent industry guidance for maintenance. Early identification of important maintenance process issues and corresponding best practices can help achieve and maintain plant performance.

Approach This project documents strategic and tactical maintenance standards by capturing industry best practices and operating experience. Maintenance process guidelines define improved maintenance processes at existing and new generation plants through process descriptions and tutorials, implementation concerns and advice, troubleshooting information, industry perspective, and detailed insights. Information is collected by conducting plant visits, using industry capacity loss data, compiling surveys, and accumulating Maintenance Rule (a)1 populations. All data available from vendors, the Institute of Nuclear Power Operations, and other industry sources are assembled into NMAC guides that address industry best practices for improving maintenance processes.

Impact Potential benefits form this project include the following: Continuously improve industry performance through proven maintenance methods and processes Reduce operation and maintenance (O&M) costs Improve access to technical and human performance information

How to Apply Results NMAC maintenance guides, workshops, and user groups are used directly by member maintenance and engineering staffs. Guidelines are sent directly to the maintenance manager, the NMAC site coordinator, and the technical library at each NMAC member plant. NMAC staff members travel to more than 30 plants each year to discuss recent guides and enhance understanding of plant maintenance problems. The NMAC portion of www.epri.com contains all NMAC guides in full text, downloadable, and word searchable.



User Groups (supplemental) (061649)

Key Research Question Data taken at several industry meetings indicate that plant maintenance personnel rotate frequently, making it more difficult for newly assigned personnel to grasp the essentials of their new responsibilities. Frequent and regular interaction with industry colleagues can facilitate education, training, and staff productivity.

Approach This project supports equipment-specific user groups and forums where experienced plant and NMAC staff can help newly assigned individuals more quickly and competently satisfy their new responsibilities. The user group membership becomes a ready resource for information and advice as plant problems are encountered. The effectiveness and usefulness of each NMAC user group is routinely reviewed by the NMAC Steering Committee to ensure ongoing value. NMAC users groups include the following: NMAC Circuit Breaker User Groups (ABB, GE, Westinghouse, Siemens Vacuum) NMAC Hoisting, Rigging, and Cranes User Group NMAC Large Electric Motor User Group NMAC Pressure Relief Devices User Group NMAC Pump User Group NMAC Terry Turbines User Group NMAC Transformer and Switchyard User Group NMAC Work Planners User Group Japanese RCM/CBM User Group Japanese Valve Maintenance User Group Access to the EPRI Maintenance Rule User Group (MRUG) Access to the EPRI MOV Performance Prediction Program User Group (EMPUG) Access to the EPRI Preventive Maintenance Basis Database User Group

Impact Potential benefits from this project include the following: Accelerate ability of newly assigned individuals to contribute to plant maintenance issues and fulfill new

responsibilities Provide a ready resource for information and advice as plant problems are encountered Provide a forum for member discussions regarding specific plant issues

How to Apply Results Members participate in NMAC user groups through annual issues meetings. Action items are addressed through working groups via phone, e-mail, and small meetings. Products are distributed and posted to the NMAC website.



Member Requested Support (supplemental) (069088)

Key Research Question Members are often challenged in applying NMAC products. Therefore, NMAC offers assistance to members in evaluating the extent to which these products can provide value to their organization.

Approach Support services can include the following: Traini ng Implementation of the Preventive Maintenance Basis Database Work package planning and preparation Foreign material exclusion (FME) program support Sealing technology and bolting techniques Implementation support in the area of on-line maintenance Specific component maintenance strategy support Expert assistance via evaluations and audits, or as consultants for resolving problems

NMAC provides on-site member-requested support to participating utilities on a cost recovery basis. The scope of member-requested support activities should be consistent with the overall objectives of the NMAC program. Examples of typical activities include programmatic and technical review of specific engineering programs, implementation of NMAC products, and response to a plant’s technical issues. The costs of these services vary, depending on the level of support requested.

Impact Reduced engineering staffs, aging plants, and dwindling vendor and architect/engineer (A/E) support make solving engineering system and component problems more difficult for operating nuclear plants. In this environment, utility engineers need a variety of tools available to assist them with problem resolution. This program provides utility personnel with information and technology solutions that decrease the time and cost needed to resolve specific technical issues or implement specific programs or products. Improved decision-making is the greatest utility benefit delivered by this program.

How to Apply Results This program will be delivered through on-site technical assistance. This program is offered in one-week increments of a full-time equivalent NMAC representative.



EPRI MOV Performance Prediction Methodology User Group (EMPUG) (QA) (supplemental) (004433)

Key Research Question EPRI’s Motor-Operated Valve Performance Prediction Methodology (MOVPPM) provides a low-cost alternative to prototype (or in situ) design basis differential pressure testing of motor or air-operated valves. MOVPPM software (QA) validates the thrust/torque requirements under design basis flow and differential conditions of gate, globe, and butterfly valve designs commonly found in both motor and air-operated valve service. In addition to the code, several hand calculation methods have been developed to address specific designs not covered by the code. The methodology has been approved by the U.S. Nuclear Regulatory Commission. Feedback and dialogue among software users can lead to functional improvements and more effective application.

Approach The EPRI MOV PPM User Group (EMPUG) provides ongoing code maintenance, user technical support, and training for the MOV Performance Prediction Methodology (MOVPPM) and the MOV Performance Prediction Program (MOVPPP). The users forum enables exchange of information pertaining to utilization of the methodology and a vehicle for maintaining and modifying the MOVPPM code.

Impact Use of this methodology obviates the need for differential pressure testing either as an initial demonstration of or periodic verification of design basis capability. Not only does MOVPPM greatly decrease downtime for valve testing in general, it can significantly reduce valve failures due to insufficient torque or thrust.

How to Apply Results MOVPPM is Windows™-based and runs on personal computers and comes with support documentation. Use of the method requires the utility to obtain valve internal design information from valve vendors, which requires about one man-week per valve.

Maintenance Rule User Group (MRUG) (supplemental) (006893)

Key Research Question The Maintenance Rule User Group (MRUG) develops solutions to generic technical issues associated with implementation of the maintenance rule. High-priority issues identified by MRUG members have included guidance on (a)(3) maintenance effectiveness assessments, guidance on component “run-to-failure” justification, and clarification of unavailability times for standby equipment. Feedback and dialogue among members can identify project opportunities to improve maintenance rule application.

Approach The Maintenance Rule User Group provides an information exchange among participants through periodic meetings, newsletters, and website and email communications. Such information exchange between participants and EPRI helps resolve technical issues from baseline inspections and ongoing revisions to the rule. MRUG also develops technical guides and documentation of good practices that can reduce costs of implementation, increase consistency among participants, leverage rule activities to improve plant performance, or reduce vulnerability to regulatory compliance issues.




Impact MRUG identifies best practices for issues such as balancing availability and reliability, monitoring of structures, improving timeliness of (a)(1) actions, coordination with the Equipment Performance Information Exchange (EPIX), and uses of condition monitoring for performance criteria. These efforts do not aim to commit plants to a single approach, but rather to identify cost-effective options, foster discussions on the strengths and weaknesses of the various options, and provide guidelines on their effective utilization.

How to Apply Results Membership in MRUG allows plants to participate in the development of implementation guidance and to provide comments and feedback to the process.

Preventive Maintenance Basis Database (PM Basis) (supplemental) (068039)

Key Research Question Effective industry use of preventive maintenance strategies relies on widespread availability of component-specific maintenance data and information. A comprehensive repository of preventive maintenance basis information for power plant equipment can support effective maintenance. Feedback and dialogue among database users leads to functional improvements and more effective application.

Approach The Preventive Maintenance Basis Database (PMBD) collects data from worldwide industry sources to develop a comprehensive repository of PM basis information for power plant equipment. The PMBD contains the data related to PM tasks, task intervals, and the technical bases of these tasks for all defined failure and degradation mechanisms. The foundation of this repository was the EPRI 38-volume PM Basis Reports and Handbook (TR-112500) and, subsequently, PM Basis Database Client /Server Version 2.0 and subsequent versions.

Impact The PMBD User Group serves as the primary source for new or revised functionality of the database. User group members serve as the beta testers for new versions of the database. The group also will suggest new component types that may need to be developed, and the group will provide input on what interfaces should be developed for the database.

How to Apply Results Members in the PMBD User Group receive copies of the current version of EPRI PM Basis Database containing information on the preventive maintenance programs recommended for 130+ component types. Updates to the existing component data tables and the addition of data tables for new component types will be communicated to members so they can download the new data as desired.

Plant Engineering

Program Overview

Program Description Safe, reliable, cost-effective nuclear plant operation is supported by detailed, technically sound engineering practices. Engineering analysis, for example, is important in assessing the condition of plant components and whether they should be replaced or repaired. Engineering also is critical when investigating life-limiting conditions, evaluating plant performance improvements, and assessing component and vendor quality. The Plant Engineering Program performs research to support the long-term, cost-effective operation of the nuclear fleet, addressing key equipment issues and enhancing the effectiveness of plant engineering programs. Issues addressed include product and vendor quality, cable aging, buried piping, workforce and skill development, life-cycle management, and obsolescence. The program also supports technology transfer through technical assistance programs, training, and user group workshops.

Research Value Research results from the Plant Engineering Program provide engineering-based guidance that enables nuclear plants to reduce capital and operations and maintenance costs and improve equipment reliability. Program participants gain access to the following: Definitive cable condition assessment methods that enhance the ability to identify, assess, and manage

aging. Enhanced validation of product quality and improved procurement specifications to reduce procurement

costs, solve obsolescence issues, and define needed engineering process changes. Improved long-term planning on key components to avoid in-service failures and potential plant outages. Enhanced workforce skills development tools to address gaps in utility training programs, validate worker

skills prior to use, and facilitate worker movement between sites. Technical results enabling the use of high-density polyethylene pipe as a replacement option for

degraded metal pipe. Risk-ranking software and inspection and mitigation technology to characterize and address buried pipe

degradation. Inspection and mitigation technologies for components susceptible to flow-accelerate corrosion.

Approach The Plant Engineering Program investigates engineering process improvements to more effectively inform and respond to plant, system, and component issues. The program targets issues such as unanticipated material degradation deficiencies that can reduce the inherent design margins in plant equipment and impact equipment reliability. Identify management and mitigation options to address cable aging, buried pipe degradation, flow-

accelerated corrosion, coating aging, service water system degradation, heat exchanger performance, and other life-limiting issues.

Define procurement and product quality standards to maintain high equipment reliability. Develop long-term planning tools to guide life-cycle decisions for key components. Conduct plant thermal performance assessments to identify engineering-based opportunities for

performance improvements. Develop tools to address component aging and obsolescence management. Develop training and workforce skills assessment tools to safely and reliably operate and maintain

nuclear power plants. Support license renewal and life extension through an array of engineering products.

Plant Engineering - Program 41.05.02 p. 1


Accomplishments The Electric Power Research Institute's (EPRI’s) Plant Engineering Program produces an array of guidance documents, training tools, and assessment methodologies that support safe, reliable nuclear plant operation and reduce risks associated with extended plant operation. Issued guidance on the development of an effective obsolescence management program, with emphasis

on the ability to effectively identify and prioritize known obsolescence issues. Revised and updated industry guidance on nuclear safety-related coatings. Guidance covers qualification

and selection of coating systems, procurement and materials management, surface preparation and coating application, inspection, condition assessment, management of non-conforming coatings, and personnel training and qualification.

Developed guidance on mitigating the risks associated with counterfeit, fraudulent, and substandard items. Report identifies measures that utilities can immediately take to reduce exposure to counterfeit items.

Formulated a 15-step process for implementing a critical spares program at nuclear power plants based on industry responses to a benchmarking questionnaire and in-depth reviews of successful critical spares programs at two nuclear plants.

Developed engineering training modules on instrument uncertainty determination, seismic analysis, relief and safety valves, water hammer, valve actuators, and finite element analysis. Developed engineering fundamental courses on basic atomic/nuclear physics and core protection.

Developed guidance for establishing an effective program for managing buried piping. Released BPWORKS™ Version 1.0a, which performs risk ranking to help plant owners prioritize the

inspections of buried piping. Investigated the possibility of applying higher allowable stresses to evaluate the fitness-for-service of

existing piping systems subject to wall thinning from flow-accelerated corrosion. Conducted a five-plant assessment to evaluate the effectiveness of flow-accelerated corrosion programs

at nuclear plants. Recommendations will enable plants to minimize or eliminate situations that are more expensive to correct than to prevent.

Compiled interim results of slow crack growth rate testing of high-density polyethylene piping to support technical justification for use of high-density polyethylene for safety and non-safety piping systems (American Society of Mechanical Engineers [ASME] Code Case N-755).

Current Year Activities Plant Engineering Program R&D for 2011 will focus on cabling, obsolescence, life-cycle planning guidance, secondary plant and buried piping corrosion phenomena, and training/qualification. Specific efforts will include the following: Develop and assess electrical cable aging management strategies Continue development of long-term planning products for identifying replacement needs associated with

major plant components Initiate project to develop advanced heat exchanger performance analysis techniques Complete 480V Motor Control Center Replacement Guide Develop resource materials to help new plant engineers learn and adapt to their new roles and

responsibilities Conduct high-density polyethylene materials (HDPE) research supporting efforts to develop a regulator-

accepted code case permitting the use of HDPE in ‘Code’ applications Support research to identify and develop technologies for interrogating buried pipe to discern its condition Develop reference materials and calculation tools pertaining to pipe and component erosion phenomena

involving liquid droplet impingement, flashing, and cavitation




Program Manager Brozia Clark, 704-595-2684, [email protected]

Summary of Projects


P41.05.02.01 PSE Equipment Reliability & Engineering Processes (base) (QA)

This project develops guidance on resolution of generic and specific aging issues to support plant engineering staffs with emergent and end-of-life component decisions. Both theoretical and practical guidance is developed, including aging models, data, and acceptance criteria for components and cables; field guides for walk-downs and inspections; and development of condition monitoring techniques. Longer-term products that support contingency planning include sourcebooks for key components and reports identifying end-of-expected life, related monitoring, and logistics issues.

P41.05.02.02 PSE Procurement and Quality Issues (base)

Provides a utility forum for sharing procurement-related concerns and experience. Conducts research on actions to enhance vendor quality, develop common specifications, establish vendor surveillance, and detect counterfeit items. Supports an industry-wide approach to prioritization and management of obsolete items.

P41.05.02.03 PSE Workforce Skills, Knowledge & Assessment (base)

Develop training tools for engineers that can be delivered via computer-based training methodologies. Develop a methodology to validate knowledge and skills competencies and record successful completion in an industry database.

P41.05.02.04 BOP Corrosion (base) (QA)

Through guidance for effective buried pipe programs, risk-ranking software, and industry dialogue including interactive web, industry conferences, and other communication, this project develops tools for organizing and prioritizing nuclear power plant approaches to code acceptance of degraded piping. Through guidance for effective flow-accelerated corrosion programs, comprehensive software, and member dialogue including interactive web, industry conferences, and other communication, this project helps nuclear power plants to maintain strong stewardship over generation assets.

P41.05.02.15a PSE - Cable Program (supplemental) (QA)

The Cable Program provides the nuclear industry with up-to-date information on cable aging and cable aging management practices from both a technical and regulatory perspective.

P41.05.02.15b PSE - Engineering Technical Training Modules (supplemental)

Engineering computer-based training modules help meet increasing industry needs for position-specific and continuing training as new personnel are brought on board and as seasoned personnel take on new assignments. Existing training modules are being converted to computer-based training format and made available to supplemental program participants.





P41.05.02.15c PSE - Heat Exchanger Performance Users Group (HXPUG) (supplemental)

The Heat Exchanger Performance Users Group offers a forum for industry personnel to improve the reliability, availability, and operational capability of heat exchangers through user group meetings and reports.

P41.05.02.15f PSE - EQMS Users Group (supplemental) (QA)

The EQMS (Environmental Qualification Management System) software has become a key tool for documenting environmental qualification. The EQMS Users Group maintains the EQMS software and provides continuing support for software users.

P41.05.02.15g-h

PSE - Seismic Qualification Reporting and Testing Standardization (SQURTS) (supplemental) (QA)

The Seismic Qualification and Reporting and Testing Standardization (SQURTS) program addresses nuclear plant replacement part obsolescence and attendant seismic qualification issues. Nuclear power plant members share equipment seismic testing costs and test results. A “library-only” membership option provides access to past completed component test reports, but without participation in the active testing program.

P41.05.02.15i Standardized Task Evaluations for Portable Qualifications (supplemental)

Standardized task evaluations can help reduce or eliminate industry's duplication of effort in assessing an individual's competency and subsequent tracking of their status, which is an important element in the industry's portable qualification efforts. This program also provides guidelines for administering practical qualifications.

P41.05.02.16a CHECWORKS User Group (CHUG) (QA) (supplemental)

The CHECWORKS User Group (CHUG) applies experience from about 260 nuclear plants worldwide to address existing and emerging issues related to flow-accelerated corrosion. CHUG provides training to new and reassigned personnel, maintains and provides updates to the CHECWORKS software, operates a dedicated website, and sponsors related research as requested by members.

P41.05.02.16b Buried Pipe Integrity Group (BPIG) (supplemental)

The Buried Pipe Integrity Group (BPIG) provides a forum for exchanging plant experience and supporting the implementation of advanced buried pipe assessment and mitigation technologies.

P41.05.02.16c Service Water Assistance Program (SWAP) (supplemental)

This project provides several forums for acquiring or sharing information pertinent to the successful operation and maintenance of nuclear plant service water systems.

PSE Equipment Reliability & Engineering Processes (base) (QA)

Key Research Question Sustained equipment reliability at nuclear plants depends on attention to both near-term and long-term degradation mechanisms that can lead to failure, unacceptable performance, or premature replacement. In many cases, aging models and condition monitoring techniques do not exist for plant equipment, and where they do they exist, readily understandable acceptance criteria may not exist. When replacements are necessary, superior materials or components may not have been identified or recognized as acceptable for nuclear service. For long-term equipment reliability concerns, methods are needed for identifying end-of-life conditions and selecting the best alternative for continued use, refurbishment, or replacement. Improved engineering processes also are needed to ensure efforts are commensurate with and properly focused on plant needs. Developing guidance, benchmarking, and good practice recommendations is important in refining these processes and helping plants improve performance.


Approach This project develops guidance on resolution of generic and specific aging issues to support plant engineering staffs with emergent and end-of-life component decisions. Both theoretical and practical guidance is developed, including aging models, data, and acceptance criteria for components and cables; field guides for walk-downs and inspections; and development of condition monitoring techniques. Longer-term products that support contingency planning include sourcebooks for key components and reports identifying end-of-expected life, related monitoring, and logistics issues. To identify best practices, lessons learned, and the need for additional industry guidance, EPRI benchmarks engineering processes at selected plants. Recent examples include optimization of the engineering change process and equivalency versus design change guidance. EPRI also convenes an annual Equipment Reliability Forum where member utilities can discuss successes and challenges in addressing equipment reliability issues. Finally, Plant Engineering operates technical assistance programs on equipment qualification, plant performance enhancement, and nuclear coatings to transfer research results, address member questions, and provide inputs for additional research.

Impact Avoid in-service failures and potential plant outages through improved detection of component

degradation Predict remaining life and evaluate the seriousness of equipment degradation through interpretation of

aging data and development of acceptance criteria Enhance ability to identify, assess, and manage aging through field guides and aging management

guidance Assure broad distribution of degradation research results and information through meetings such as the

Equipment Reliability Forum Improve assessment techniques to identify components and materials prone to early aging Develop timely, cost-effective strategies for managing the life cycles of capital-intensive components Optimize engineering costs by identifying and disseminating best processes that can more effectively

implement utility-limited engineering resources.

How to Apply Results Because multiple tasks are performed under this project, member applications vary. Field guides, aging management guides, and end-of-life guides are applied directly. Engineering best practices and lessons learned will be incorporated into utility programs according to their compatibility with existing practices or planned changes. Information provided through the Equipment Reliability Forum can be used to promote broader understanding and implementation of research results and can be incorporated into training courses.

PSE Procurement and Quality Issues (base) (065801)

Key Research Question Product quality issues have negatively impacted plant reliability and costs for replacement items. Causes of poor product quality include loss of vendor expertise, lack of vendor understanding, and poor specification development. Additional focus and guidance are needed to better understand the root causes of poor product quality and actions needed to improve quality, particularly for hardware and hardware refurbishments. Guidance and sharing of experiences also is needed to more effectively use the supply chain and procurement engineering functions at nuclear power plants. Finally, as plants age, additional emphasis is needed on developing collaborative solutions to obsolescence.



Approach This project consists of three principal elements: 1) continuing support of utility forums for sharing procurement-related concerns and experience through the Joint Utility Task Group (JUTG) and the Nuclear Supply Chain Strategic Leadership Council (NSCSL); 2) research on the root cause and corrective actions to enhance vendor quality, with possible spin-off projects addressing common procurement specifications, source surveillance templates, and guidance on detecting fraudulent and counterfeit items; and 3) support of an industry-wide approach to prioritization and management of obsolete items, including a pilot project to demonstrate methods being developed.

Impact Reduce procurement-related costs for components Improve equipment reliability and performance through better understanding and improvement of product

quality Reduce procurement costs through application of commercial-grade dedication processes Access to cost-effective, collaborative solutions to parts obsolescence

How to Apply Results Members use project information to enhance procurement programs, develop improved supplier partnerships, improve specification development, and solve obsolescence issues. Members also gain insights into needed engineering process changes (such as equivalency versus design change).

PSE Workforce Skills, Knowledge & Assessment (base) (065802)

Key Research Question Turnover of nuclear plant technical staff will be significant in the next 5 to 10 years. During the same timeframe, nuclear utilities will be training personnel to support construction and operation of new nuclear plants. Pressures to reduce operations and maintenance costs often result in impacts to training budgets. Also, fewer and fewer skilled supplemental workers are available for working power plant outages. Often, as the supplemental outage workers travel from plant to plant, they receive the same training and examinations at each plant. In this environment, nuclear plants need cost-effective methods to develop and deliver high-quality, effective training and be able to quickly validate the skills competencies of the supplemental workers.

Approach Computer-based training technology can improve the effectiveness of engineering training and reduce the costs associated with providing this training. Plant Engineering (PSE) is engaged in developing computer-based training for nine engineering fundamentals topics included in the Institute of Nuclear Power Operations (INPO) guidelines for orientation of new engineers. PSE also is transferring materials related to its Standardized Task Evaluations onto NANTeL; these knowledge examinations and skills proficiency demonstration examinations can be administered to supplemental workers to verify skills competencies. Any power plant with access to NANTeL can download the examination materials, administer the examinations, and record an individual's successful completion within the NANTeL system. This will serve as a basis for accepting prior qualification testing in lieu of re-administering knowledge and skills training and examination.

Impact Stations are using the engineering computer-based training (CBT) modules in lieu of classroom-conducted training sessions to provide orientation training to new engineers. This results in fewer disruptions for engineering organizations and also frees up instructor time associated with class lecture and examination preparations.



By verifying prior completion of knowledge and skills qualification-related examinations using the NANTeL or EPRI databases, utilities are experiencing cost savings associated with streamlining the in-processing, training, and qualification of supplemental personnel.

How to Apply Results The content of the engineering training modules will be available both from EPRI as well as an industry web-based training delivery system (INPO’s NANTEL system) where students can complete the CBT and take the course examination. For qualification of outage supplemental personnel, knowledge examinations and skills proficiency demonstration examinations can be accessed via NANTeL, administered, and results recorded within the NANTeL database. Once the record of successful completion is recorded in the database, the record of successful completion can be used by other utilities as a basis for exempting their examination requirements when the supplemental worker arrives at their station for outage work.

2011 Products



Reference Handbook for New Systems Engineers 02/28/11 Technical Report

Web-Based Engineering Training: Engineering Fundamentals - Heat Transfer & Fluid Flow, Version 4.0 02/15/11 Software

BOP Corrosion (base) (QA) (052459)

Key Research Question Corrosion in the secondary systems of nuclear plants can result in annual costs of up to $25 million per plant. These costs are primarily associated with corrosion product transport in boiling water reactors, flow-accelerated corrosion in steam and feedwater systems of all types of nuclear plants, degradation in service water systems, and degradation in raw water heat exchangers, including the main condenser. Without intervention, these costs will increase as plants age. A specific issue impacting plants considering life extension is the health of buried piping. Inspection, repair, and replacement of these lines can be extremely expensive, particularly in buried lines that pass beneath buildings and equipment.

Approach The Balance of Plant Corrosion (BOP) Corrosion program develops the technology, tools, and software to cost-effectively address corrosion issues in the BOP portions of nuclear power plants. BOP corrosion has spearheaded the development of improved inspection technology to assess the health of secondary systems and the use of alternate materials to reduce cost and improve the service life of BOP piping and components.

Impact Data and methodology to allow use of high-density polyethylene as an option to repair or replace

corroding steel pipe in Class 3 service water systems. Computer-based modules to train new and reassigned plant personnel on the most common forms of

corrosion in the secondary systems of nuclear plants. Mitigation technologies to address buried pipe degradation. Risk-ranking software tool to prioritize buried pipe inspections. This tool calculates the probability of a

leak occurring in each segment of buried piping, considers the consequences of a leak at each specific location, and derives a calculated “risk.”



Robust inspection technology and guidance to assess the health of large-diameter and intermediate-diameter buried piping. The “proof-of-concept” of a large-bore pipe inspection tool (3-foot to 12-foot diameter) was completed in 2008. An intermediate pipe diameter tool is currently in development and scheduled for “proof-of-concept” testing at a nuclear power plant during the spring 2011 outage season.

Research to support the American Society of Mechanical Engineers (ASME) in developing “design rules” and “fitness-for-service” rules for buried pipe.

How to Apply Results Data supporting the use of high-density polyethylene as a repair and replacement option for corroded steel service water systems has been provided to ASME and is available for members to incorporate into code cases. Computer-based training modules are available to members and can be modified for plant-specific information.

PSE - Cable Program (supplemental) (QA) (005614)

Key Research Question The aging of medium- (4160 V+) and low-voltage (<1000 V) cable systems has raised regulator interest in the ability of these systems to perform their safety and support functions. This program supports the industry by disseminating information on how cable systems age and the best means for detecting and mitigating aging effects.

Approach This project provides an industry forum for discussing issues related to cable system aging management and for transferring cable research results to members in practical terms. As funding permits, technical reports are produced through the project on cable-aging-related topics of interest to program members. Feedback from Cable Users Group meetings serves to help guide cable research on aging-model and condition-monitoring development. The Program also provides access to EPRI personnel conversant in cable aging management issues, allowing utility personnel to discuss plant problems and their resolution.

Impact Benefits accrue through direct access to experts in cable aging management and cable condition monitoring. Participants also advise on cable research to ensure its pertinence to nuclear plant applications.

How to Apply Results Cable User Group attendees have direct access to EPRI and industry experts in condition monitoring, cable manufacture and installation, and the discussion of the latest industry issues and practices. Research results are provided in EPRI research reports and meeting minutes from the Cable User Group Meetings.



PSE - Engineering Technical Training Modules (supplemental) (005556)

Key Research Question As new engineering personnel are brought into the workforce and as individuals are moved into different assignments, position-specific training modules can accelerate their acclimation and value to the nuclear industry. If training modules are not readily available to meet the needs in these cases, organizations typically have to develop specialized courses or find course offerings available elsewhere in the industry. Unless the topic is one that is routinely offered, the availability of a course will not likely meet scheduling needs.

Approach EPRI’s computer-based training (CBT) modules can be used for position-specific and continuing training needs for selected topics. Forty-five modules were developed some years ago using PowerPoint™ slides and companion Word™ documents. This information is being used as a basis for the new modules; however, the content is being updated and photographs and graphics are being used along with interactive features to enhance the training. The CBT modules are much more in line with expectations of new engineers entering the workforce. Seventeen modules were previously converted, and more will be converted in 2010. Modules are being selected on a priority basis to meet industry needs.

Impact Based on today’s demographics, personnel turnover in the nuclear power industry will be considerable in the coming years. The need for this training is increasing as new personnel are hired and seasoned employees are reassigned as a result of personnel turnover. These CBTs have the following attributes: Can be downloaded for use when needed from http://www.epri.com Provide basic position-specific training for new hires and individuals reassigned to new jobs Can be used for continuing training

How to Apply Results Engineering supervisors and training personnel should be aware of these modules and use them for position-specific and continuing training as appropriate.

PSE - Heat Exchanger Performance Users Group (HXPUG) (supplemental) (45060)

Key Research Question This project offers a forum for industry personnel to improve the reliability, availability, and operational capability of heat exchangers (with the exception of steam generators and boilers). The project allows participants to share experiences and to resolve technical issues associated with heat exchangers.

Approach Information sharing with Heat Exchanger Performance User Group (HXPUG) members Technical investigations into high-priority technical issues as directed by the membership Plant support and services as related to heat exchanger testing Report creation on topics related to heat exchangers that are of value to the group Collaboration on common industry problems and solutions as they relate to heat exchanger testing and

performance



Impact The project offers an estimated cost savings of $25,000 to $100,000 annually per plant using the information available through the HXPUG group facilitating the following: Improved testing methods through collaboration with industry personnel and use of EPRI guidelines Avoided costs through the reduction of unnecessary heat exchanger testing Improved plant performance through improved thermal performance of the feedwater heater, moisture

separator reheater, and condenser Collaboration on common industry problems and solutions as they relate to heat exchanger testing and

performance

How to Apply Results Participating members in the Heat Exchanger Performance User Group can implement the lessons learned and information generated in this group. Examples include improved heat exchanger performance testing methods, new approaches to calculating end-of-life for plant heat exchangers, and avoiding issues experienced at other plants.

2011 Products



Heat Exchanger Program Owner and System Engineer Handbook 03/31/11 Technical Report

PSE - EQMS Users Group (supplemental) (QA) (007529)

Key Research Question The Environmental Qualification Management System (EQMS) is a state-of-the-art, integrated software management tool designed to assist utilities in managing their environmental qualification programs more efficiently and cost-effectively. EQMS also integrates the calculations required to support the documentation of environmental qualification directly in the software, which simplifies the evaluation process and further reduces the effort required to maintain environmental qualification files.

Approach The EQMS User Group provides for continued maintenance and upgrade of the EQMS software and sharing of lessons learned and improvement opportunities through periodic users group meetings. Maintenance and upgrade activities include documenting identified problems, considering desired modifications, developing and testing approved changes, and guiding the software through the EPRI software quality assurance process. Quality Assurance Statement: EPRI develops and maintains the EQMS, which is considered a nuclear safety-related product, in accordance with Title 10, Code of Federal Regulations; Part 50, Domestic Licensing of Production and Utilization Facilities (10CFR50); Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants; and Part 21, Reporting of Defects and Noncompliance (10CFR21).

Impact The EQMS software allows users to accomplish the following: Document their environmental qualification records in a consistent fashion Quickly evaluate changes to environmental changes Manage generic and plant-specific information from corporate offices or from individual sites Share environmental qualification files between plant sites and utilities



Significantly reduce the effort required to maintain EQ files.

How to Apply Results Nuclear plant owners use the EQMS database to document environmental qualification. The EQMS database includes four basic modules for storing EQ-related data: Generic Qualification Evaluation (GQE) module Plant Qualification Evaluation (PQE) module Environments module Equipment module

EQMS also includes procurement requirements sheets, maintenance requirements sheets, and an update center that is used for importing and exporting selected database records. EQMS Users Group funders have access to nonproprietary generic qualification evaluations (GQEs) through the Scientech EQDB website (http://eqdb.scientech.com).

PSE - Seismic Qualification Reporting and Testing Standardization (SQURTS) (supplemental) (QA) (004414)

Key Research Question Component obsolescence remains an industry challenge in the maintenance of nuclear plants, and with the prospect of even longer-term operation, replacement parts for existing facilities will continue to drive higher costs. Component qualification to individual utility design specifications constitutes a significant cost in the dedication of replacement parts. The Seismic Qualification and Reporting and Testing Standardization (SQURTS) program, conceived in the early 1990s to address nuclear plant component obsolescence issues, applies the economies of scale of member utility owners and operators to share component seismic testing specifications, costs, and test results.

Approach Seismic testing conducted through the SQURTS program involves component testing at a service vendor facility nominally 6 to 8 weeks per year. Utility participation is critical in developing generic test specifications and component test procedures, witnessing test performance and approving test reports, and participating in user meetings. Members also have access to a seismic test report database comprised of SQURTS-performed test results and individual member test reports (should they choose to enter them). EPRI provides project management for the program, including contracting test services, budget forecast, tracking and reporting, database management, test report distribution, user communication, initiative coordination, and member meetings. EPRI provides an option for membership in the Library only; this provides access to past completed component test reports. However, participation in the active testing program is not permitted if only this option is selected.

Impact The program enables members to reduce component seismic testing costs through utility economies of scale and the shared database that members can use for component evaluations.

How to Apply Results Results are generally implemented immediately by participants. Testing is driven by the needs of the members, and the database is accessed on member demand. Design engineers, seismic subject matter experts, and procurement engineers are the usual customers of the SQURTS program.


http://eqdb.scientech.com/


2011 Products



SQURTS User Group Meeting 12/31/11 Technical Resource

Standardized Task Evaluations for Portable Qualifications (supplemental) (005354)

Key Research Question Utility and supplemental personnel are critical to a plant's ability to conduct quick-turnaround refueling outages. Recent trends show a disproportionate occurrence of events associated with supplemental personnel. EPRI's Standardized Task Evaluation program (formerly called the Task Proficiency Evaluation program) provides a proven knowledge and skills evaluation process to efficiently evaluate the capabilities of entry-level, incumbent, and contractor personnel. The program is working within the framework of the Nuclear Energy Institute's Workforce Issues initiative and with the Institute of Nuclear Power Operations' National Academy for Nuclear Training e-Learning (NANTeL) portable qualification project to establish an infrastructure that ensures the competency of the industry's craft and technician workforce. The standardized task evaluations are not restricted to U.S. applications; utilities in France, South Africa, and Canada have expressed interest in adapting such evaluations to their own countries. Additionally, the need to implement an industry consensus for standards for administering practical qualifications has been identified.

Approach Standardized task evaluations are used to ensure that the workforce is competent to reliably perform the many tasks associated with operating and maintaining industry facilities. Program participants continue to collaboratively develop evaluation tests that support high-priority industry needs. More than 60 evaluations have been developed within the STE Program and are available on www.epri.com. These evaluations, which cover tasks performed by utility and supplemental workforce during outage work, include a task analysis and objectives, written test items, and performance (practical) evaluations. Additionally, the results from these evaluations are documented into a national registry of personnel who have demonstrated competency in specific task areas.

Impact Participating organizations can use the STE evaluations to assess the competency of their workforce, thus eliminating unnecessary training or retraining. Further, because the modules were developed according to EPRI's Administration Protocol for Portable Practicals (AP3), they reflect industry consensus standards for administering practical evaluations.

How to Apply Results Training and maintenance managers can directly access the STE modules through multiple channels: Identifying and downloading evaluations through www.epri.com for use by participating organizations with

specific task needs. Accessing evaluations available on INPO's NANTeL System (http://www.nantel.org) for use by

participating organizations for on-line testing and for reporting results. Accessing the registry of qualified personnel on www.epri.com. EPRI Report 1015074, Administration Protocol for Portable Practicals (AP3) in Task Proficiency

Evaluations



CHECWORKS User Group (CHUG) (QA) (supplemental) (052460)

Key Research Question Although industry efforts have been effective in reducing the number of piping and equipment failures caused by flow-accelerated corrosion (FAC), piping and components remain susceptible to degradation as plants age. Refined guidance on where to inspect, chemistry improvements to reduce damage rates, and material upgrades for replaced components is needed to offset challenges posed by economic considerations, short outages, and personnel turnover. Mechanical pipe degradation caused by cavitation, liquid droplet impingement, flashing, and solid particle erosion can affect personnel safety and cause power losses. Damage caused by these mechanisms is nonlinear with time and often results from off-normal operations.

Approach The CHECWORKS User Group (CHUG) applies experience from about 260 nuclear plants worldwide to address existing and emerging issues related to flow-accelerated corrosion. CHUG provides training to new and reassigned personnel, maintains and provides updates to the CHECWORKS software, operates a dedicated website, and sponsors related research as requested by members. This includes research and guidance to address detection of erosion damage in high-energy piping systems.

Impact Minimize risk to personnel by reducing the probability of large-bore pipe ruptures Reduce forced power reductions through FAC mitigation Reduce the number of piping inspections through improved guidance, predictive software, and piping

replacements (some plants have observed a 25% reduction over a 6-year period with average estimated savings of $2,150,000 per outage per plant)

Develop practical tools to reduce FAC risks, such as material alloy analyzers that can be used to reduce the number of inspections

Identify new FAC vulnerabilities before leaks occur Train new and reassigned plant engineers on FAC identification, monitoring, and mitigation Facilitate interaction with industry peers and ease access to reports and other information

How to Apply Results Members use CHECWORKS to predict plant degradation and reduce unneeded piping inspections. Technical guidance related to pipe alloy analyzers, erosion, and low-temperature FAC provide members with information to optimize inspection locations. Members can access training for new and reassigned personnel and can use the CHUG website to facilitate communications between FAC personnel at member plants.



Buried Pipe Integrity Group (BPIG) (supplemental) (068213)

Key Research Question Buried piping has become a more visible issue with regulatory emphasis on material aging issues and plant life extension requirements. Due to aging of external protective coatings as well as a multitude of internal and external piping corrosion mechanisms, each nuclear plant’s buried pipe infrastructure is susceptible to leaks and failures. Leaks can be difficult to locate. Also, some may contribute to contamination of groundwater. All buried pipe leaks can be expensive to repair due to accessibility issues. Some leaks may require a plant shutdown for repair. A broad-based and comprehensive program is needed to support plant efforts to reduce the probability and consequences of failure to an acceptable level.

Approach The Buried Pipe Integrity Group (BPIG) provides a forum for exchanging plant experience and provides counsel and recommendations as to the implementation of advanced buried pipe assessment and corrosion mitigation technology.

Impact Assess the health of existing piping and determine remaining service life Develop methods to repair buried piping in situ Select and qualify alternate materials and service environments (for example, high-density polyethylene,

water treatment, and cathodic protection) Provides a forum for the buried pipe services industry to interface with buried pipe engineers from

participating utilities All United States Nuclear utilities and several international participants were members of the BPIG in

2010.

How to Apply Results Members will apply the results of this project in developing effective buried pipe integrity programs and in assessing and maintaining existing buried piping systems.

Service Water Assistance Program (SWAP) (supplemental)

Key Research Question Nuclear plant service water systems are complex systems that can occasionally provide engineers with day-to-day challenges. Service water system performance can be improved by providing engineers with access to a collaborative environment where thoughts, ideas, and solutions can be readily available and shared.

Approach Project participants gain access to the SWAP web page, which includes the SWAP technical library, SWAP surveys, a listing of SWAP coordinators, and easy access to SWAP products. Members also can query the nuclear industry on service water problems through industry SWAP surveys. Members also can obtain personal assistance from EPRI personnel via phone or email. EPRI also sponsors an annual meeting of SWAP coordinators for sharing operating experience and discussing solutions to field problems. Training courses for service water engineers are available on piping and corrosion mechanisms.

Impact The SWAP technical library, which contains more than 2350 documents on 300 subjects, can be searched by subject, author, and/or date. Many of the titles are available for download as PDF files. The annual meeting and access to EPRI experts provide opportunities to discuss plant issues and identify potential solutions.




How to Apply Results The SWAP coordinators serve as the point of contact between the EPRI SWAP program and the plant. Active participation facilitates technology transfer and maximizes benefits received. A SWAP Coordinator's Manual helps guide access to SWAP resources and services.

2011 Products



Annual Service Water Assistance Program Meeting: Annual meeting for sharing information from the U.S. Department of Energy's Office of Electricity Delivery and Energy Reliability and technologies related to nuclear plant service water systems.

12/23/11 Workshop, Training, or Conference




Annual Service Water Assistance Program Meeting 12/21/12 Workshop, Training, or Conference

Fuel Reliability

Provide critical information needed to reduce uncertainties and support planning decisions about today’s operations and tomorrow’s assets.

Program Fuel Reliability

Fuel Reliability

Program Overview

Program Description Fuel failures and other fuel-related issues can have significant operational impacts on nuclear power plants. Fuel failures, for example, have cost some nuclear power plants $40 million or more per event to cover replacement power costs and the costs of a fuel reload. While the industry has made substantial progress in reducing the frequency of fuel failures, continued attention to technical gaps impacting fuel reliability is needed. The Fuel Reliability Program drives improvements in nuclear fuel performance and reliability based on issues encountered at operating plants around the world. Research addresses multiple aspects of fuel performance and reliability, including fuel failure root-cause resolution, fuel/water chemistry interactions, and operational margins through end of life. The new knowledge is then applied to update fuel reliability guidance and to provide technical feedback for new fuel designs. The program also engages nuclear regulatory agencies and provides technical input regarding fuel-related rulings and regulations.

Research Value About $45-$50 billion of nuclear fuel is operating in nuclear reactors worldwide. Optimizing the use of this fuel, while ensuring its safe operation, is paramount to reliable, cost-effective nuclear plant operation. Fuel Reliability Program participants gain access to the following: Technical guidance to improve fuel reliability and reduce economic risks associated with fuel failures,

which have cost the U.S. nuclear industry more than $300 million over the past decade, and the global nuclear industry 2-3 times that amount.

Global operating experience with all types of nuclear fuel assemblies and reactor types to inform decision-making.

State-of-the-art software tools to analyze fuel rod performance, helping utilities manage power margins by ensuring that stresses don’t exceed a threshold that could lead to fuel failures.

Data and technical insights pertaining to the use of high-burnup nuclear fuels. Electric Power Research Institute (EPRI) studies show the industry can save about $200 million/year by moving closer to the current burnup limit and another $200 million/year by increasing the licensed limit.

Technical studies to inform regulators and ensure regulations impacting nuclear fuel are technically based and not unnecessarily conservative.

Approach The Fuel Reliability Program develops knowledge, guidance, and tools to maximize the reliability of nuclear fuel and core components. The program also participates in international research consortiums to improve fundamental understanding of in-reactor behavior of fuel, cladding, control materials, and other core components. The overall program is organized into four Technical Advisory Committees (TACs). Fuel Performance and Reliability: Quantify fuel operational margins and identify fuel failure mechanisms

through poolside and hot cell examinations. Projects are cost-shared with fuel vendors to ensure a direct connection exists and information is factored into subsequent fuel designs.

Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) Corrosion and Crud Control: Improve industry understanding of the links between water chemistry, crud, and fuel reliability in pressurized and boiling water reactors. The research combines fuel surveillance programs and mechanistic studies to develop various guidelines and improve predictive capabilities.

Fuel Regulatory Issues: Serve as the industry focal point on technical aspects of regulatory issues by participating in experimental programs and performing independent analyses to ensure the adequacy of proposed modifications to various accident criteria.

Fuel Reliability - Program 41.02.01 p. 1


Accomplishments EPRI’s Fuel Reliability Program distills global experience with nuclear fuel into actionable guidance and insights that drive measurable improvements in fuel reliability. Developed a comprehensive industry plan to identify operational impacts related to channel distortion and

formulate near-term mitigation guidance. Performed healthy fuel examinations in 2009 to define industry standard and kick-start measurements in

support of industry efforts toward zero fuel failures; initiated vendor assessments in 2010 of all healthy fuel inspections as part of "check and adjust."

Released Version 3.1 of the Fuel Reliability Database, which includes new graphs on fuel integrity and more detailed information on debris filter types and spacer grid materials. The database includes data from about 600 global fuel cycles and can be used to benchmark nuclear fuel performance, evaluate alternate fuel designs, and gauge industry fuel-related practices.

Performed hot cell examination of fuel from Browns Ferry and identified a new failure mechanism where elevated levels of corrosion generated enough hydrogen in the cladding for the cladding to fail under otherwise insignificant levels of stress. Both cladding corrosion and water chemistry conditions have been improved to avoid future failures by this mechanism.

Teamed with the Nuclear Maintenance Application Center (NMAC) and the Institute for Nuclear Power Operations (INPO) to define an improved organization for foreign material exclusion.

Completed feasibility analysis of chemically enhanced ultrasonic fuel cleaning for reducing the mass of corrosion products carried over from one cycle to another.

Issued five fuel reliability guidelines aimed at enabling industry to achieve zero fuel failures and continued research efforts to resolve gaps identified in guidelines.

Coordinated failure root cause identification in several fuel failure cases with industry-wide implications, including fuel pellet quality issues in AREVA and Westinghouse fuel.

Evaluated zinc addition in high-duty pressurized water reactors and developed a framework for demonstrating the impact of elevated coolant hydrogen.

Evaluated noble chemical addition, including Online NobleChem, and zinc in boiling water reactors to optimize operation.

Current Year Activities Fuel Reliability Program research and development for 2011 will focus on a number of remaining fuel reliability gaps, including the continuing threat of foreign materials, the effects of new water chemistry regimes on fuel reliability, and fuel reliability training. Specific efforts will include the following: Perform fuel failure investigations for events with industry-wide implications. Improve risk assessment tool to avoid axial offset anomaly. Coordinate input for revisions to water chemistry guidelines for pressurized water reactors, including a

better understanding of zinc. Refine risk assessment tool for boiling water reactors to better understand interactions between water

chemistry, cladding, and fuel duty. Conduct hot cell examinations of an ultra-high fluence control rod to support accurate lifetime predictions.


Program Manager Kurt Edsinger, 650-855-2271, [email protected]



Summary of Projects


P41.02.01.01 Fuel Performance & Reliability (base)

The project conducts research and develops guidance to address generic fuel performance and reliability issues. The project relies on detailed poolside and hot cell exams to establish current margins, identify failure mechanisms, and provide technical insight for other performance issues.

P41.02.01.02 Channel Distortion Industry Action Plan (supplemental) (QA)

The BWR channel distortion action plan represents a coordinated industry effort to develop short-term guidance and drive research needed to improve the industry’s ability to manage and eliminate channel distortion as an issue. Selected reports and products may be prepared in whole or in part in accordance with the EPRI Quality Program Manual that fulfills the requirements of 10CFR50 Appendix B, 10CFR21 and ANSI N45.2-1977.

P41.02.01.03 PWR Corrosion & Crud Control (supplemental) (QA)

The project conducts research and develops guidance to minimize fuel performance issues related to fuel corrosion and crud in pressurized water reactors.

P41.02.01.04 BWR Corrosion & Crud Control (supplemental) (QA)

This project conducts research and develops guidance to minimize fuel performance issues related to fuel corrosion and crud in BWRs.

P41.02.01.05 Fuel Regulatory Issues (supplemental) (QA)

The project conducts research in support of fuel regulatory issues and acts as the industry focal point with regulatory agencies on technical issues.

P41.02.01.06b FALCON User Group (supplemental)

The FALCON User Group provides a forum for nuclear utilities to further develop the fuel performance and analysis capabilities of the FALCON software code.

P41.02.01.07 NFIR-VI (supplemental) Through the Nuclear Fuel Industry Research (NFIR) Program, EPRI coordinates research on behalf of an international consortium of utilities, fuel vendors, and research laboratories. Research activities focus on generic, long-term issues and opportunities to ensure safe and reliable use of light water reactor core materials and components. The NFIR program seeks to understand fundamental in-reactor behavior of fuel, cladding, control materials, and other core components and to share this valuable knowledge with the industry.



Fuel Performance & Reliability (base) (052388)

Key Research Question A wide range of research is needed to quantify fuel reliability margins for current and new fuel designs. This information addresses the behavior of fuel, cladding, and assembly components (for example, spacer grids, guide tubes, water rods, and boiling water reactor channels). There also is a need to understand and resolve fuel failure root causes for mechanisms with generic industry implications. Other industry needs include quantifying the performance of control rods and control blades.

Approach This research area focuses on root-cause evaluations of failure mechanisms with industry-wide implications and detailed assessments of fuel operating margins under limiting conditions. Ongoing root-cause investigations include corrosion-related failures at two BWRs, duty-related failures in barrier fuel, and an unknown failure mechanism in PWR fuel. Performance assessments have been initiated for key PWR and BWR fuel designs under bounding conditions to ensure fuel will operate safely and economically up to current design limits. These assessments include substantial poolside and hot cell programs on fuel (past examples included AREVA M5, several variations from GNF, and Westinghouse ZIRLO). The hot cell examinations are closely coordinated with the particular fuel supplier to leverage funding and ensure the vendor can fully apply the investigation results. This program also has responsibility for the reliability of other fuel-related components, including balance of assembly components (spacers grids, guide tubes, water rods, and BWR channels) and reactivity control components (PWR control rods and BWR control blades). A small but important program area focuses on advancing nondestructive poolside techniques to quantify fuel performance margins, locate failed rods, and identify the cause of failure. The industry’s fuel reliability database (FRED), which compiles data on failure root causes, fuel reliability statistics, and good operating practices, is maintained through this program area.

Impact Resolution of failure root causes to avoid recurrence and identify technical remedies, both on the vendor

and utility side Improvements in fuel cladding properties (primarily corrosion and hydrogen pickup) Guidelines to assist the industry in applying research results

How to Apply Results Many of the key results flow into fuel reliability guidelines; some of the results inform decisions on fuel designs or strategies; other results can be used by an individual utility in working with its fuel supplier.

2011 Products



Westinghouse Assessment of Fuel Surveillance and Inspection Guideline (FSIG) Baseline Inspections 11/01/11 Technical Report

AREVA Assessment of Fuel Surveillance and Inspection Guideline (FSIG) Baseline Inspections 10/01/11 Technical Report

Ramp Testing of Fuel Rods with Pellet-Pellet Gaps 04/30/11 Technical Report

Swelling Model for AgInCd Absorber Material used in RCCA 11/30/11 Technical Report



Channel Distortion Industry Action Plan (supplemental) (QA)

Key Research Question Seventeen of the 35 BWRs in the United States have reported control blade interference due to channel distortion in the last 8 years. Affected fuel designs include Zircaloy-2 channels manufactured by all three U.S. fuel vendors. Multiple mechanisms have been identified that can give rise to channel distortion, which manifest as bow, bulge, and twist. Recent experience indicates that fuel vendor models can not sufficiently predict the channel distortion.

Approach A Channel Distortion Industry Action Plan (CDIAP) has been developed to coordinate efforts among utilities, EPRI, the BWR Owners Group, fuel vendors, and INPO. The purpose of this plan is to provide the industry with an effective short-term and long-term strategy to address channel distortion as an operational issue. Integral to this plan is to understand each mechanism of channel distortion at the scientific (or mechanistic) level. Therefore, the goals of the plan include the following: Update guidance to effectively manage channel distortion until more effective solutions are available Collect and analyze channel performance data, including operational performance data, poolside

dimensional measurements, and hot cell examination data Develop a mechanistic understanding of channel distortion and identify gaps in understanding Conduct more detailed examinations to understand and quantify each distortion mechanism Develop improved distortion models and provide to fuel vendors for incorporation into their channel

management tools; validate the performance of distortion models via examination and surveillance Ensure proposed materials solutions are appropriately validated

Impact Despite substantial past efforts to understand the issue, recent experience indicates channel distortion is not predicted well enough. Results from research in this area will lead to model improvements that ensure core designs are less susceptible to channel distortion.

How to Apply Results The initial effort to provide channel distortion guidance will be directly applicable by the utility. The information from poolside measurements and laboratory analysis will be used to improve the models that utilities use to guide core design and manage operational aspects of channel distortion.

PWR Corrosion & Crud Control (supplemental) (QA) (052390)

Key Research Question Changes in pressurized water reactor fuel duty and reactor coolant chemistry have had adverse effects on nuclear fuel performance and reliability. The consequences have led to axial offset anomaly (AOA), reduced shutdown margins (safety), fuel failures, and elevated radiation fields at many plants. The industry requires a focused and integrated research and development (R&D) approach to develop means to prevent these problems.

Approach This project combines plant-generated data with laboratory research and theoretical modeling to provide members with tools and skill sets to avoid the costly consequences of significant corrosion product deposition on fuel rods that can lead to deteriorating fuel performance. Plant demonstrations verify that new technologies and changes in new chemistry regimes or core operating strategies do not adversely affect fuel performance.



Impact Improves fuel reliability by minimizing corrosion product (crud) formation on fuel rods. Excessive crudding

can lead to fuel failures, which can cost members several hundred thousand to millions of dollars. Reduces economic and safety risks associated with issues arising from heavy crudding, such as axial

offset anomaly, which can reduce shutdown margins and elevate radiation fields during plant outages. Integrates plant data collection, laboratory testing, and innovative loop research with theoretical modeling

to understand the in-core crud deposition process and to deliver integrated strategies for avoiding crud deposition.

Supports the industry’s zero fuel failure goal.

How to Apply Results Members use the reports generated by the project to support on-site changes in programs and procedures. Research products emerging from this project include PWR Axial Offset Anomaly Guidelines, ultrasonic fuel cleaning technology, and the Boron-Induced Offset Anomaly Risk Assessment Tool (BOA software). Project results are appropriately integrated with other EPRI research programs, including the Materials Reliability Program (MRP) and the Chemistry and Radiation Management Programs. For example, applying zinc to pressurized water reactors to mitigate stress corrosion cracking or abate radiation field requires a carefully orchestrated effort to ensure the application of zinc does not adversely impact the performance and integrity of nuclear fuel. Members use the reports generated by the project to support on-site changes in programs and procedures, including the plant final safety analysis report (FSAR). Particularly valuable have been the following products, which continue to evolve and improve as a result of ongoing research: PWR Axial Offset Anomaly Guidelines, ultrasonic fuel cleaning technology, and the Boron-Induced Offset Anomaly Risk Assessment Tool (BOA software). Project results are appropriately integrated with other EPRI research programs, including the Materials Reliability Program (MRP) and the Chemistry and Radiation Management Programs. For example, applying zinc to pressurized water reactors to mitigate stress corrosion cracking or abate radiation field requires a carefully orchestrated effort to ensure the application of zinc does not adversely impact the performance and integrity of nuclear fuel.

BWR Corrosion & Crud Control (supplemental) (QA) (061376)

Key Research Question Since 2001, four boiling water reactors have suffered crud- and corrosion-induced fuel failures with a significant impact on plant performance and fuel economics. While the precise conditions leading to failure are still not fully understood, the failures are generally attributed to a combination of water chemistry, fuel duty, and cladding materials. Tenacious crud and surface spallation also have been increasingly observed as water chemistry conditions evolve. For example, to mitigate stress corrosion cracking of core internals and other components, most boiling water reactors in the United States have moved from normal water chemistry to hydrogen water chemistry to noble metal chemical application and, more recently, to online NobleChem. Zinc injection has been widely adopted for shutdown dose rate reduction, but also has implications for crud on the fuel.

Approach The objective of this project is to mitigate the impact of fuel crud deposition and cladding corrosion on fuel reliability by studying the roles of various factors on crud deposition and cladding corrosion and associated hydriding characteristics. In particular, this program will 1) continue to refine guidance in the Boiling Water Reactor Fuel Cladding Corrosion & Crud Guideline; 2) work with the Water Chemistry Guidelines Committee to establish limits on known and potentially harmful chemical species in the feedwater and reactor water; 3) perform fuel surveillances at plants with new or bounding water chemistry conditions; 4) study the important characteristics of tenacious crud and the conditions that form types of crud carrying the most risk, including the role of specific chemistry impurities (for example, Fe, Zn, Cu, and SiO2); and 5) develop a predictive capability



for crud deposition. Research activities will be coordinated with the Boiling Water Reactor Vessels and Internals Project (BWRVIP) Mitigation Committee and Water Chemistry program to ensure maximum benefits to members.

Impact Evaluate interactions among key water chemistry parameters, fuel duty, and fuel assembly materials in

relation to fuel corrosion and related aspects of performance through a combination of fuel surveillance programs, fundamental research, and advanced methodologies.

Develop predictive capabilities for crud deposition, optimize fuel operation, and provide input for boiling water reactor fuel and water chemistry guidelines.

Support the industry’s zero fuel failure goal.

How to Apply Results Much of the technology developed will be directly applied by member nuclear power plant personnel. Since a substantial fraction of the information also will have implications for core design and fuel assembly properties, members also can use the information while working with their fuel supplier.

Fuel Regulatory Issues (supplemental) (QA) (052391)

Key Research Question Recent experimental evidence has raised questions about the applicability of current reactivity-initiated accident (RIA) and loss-of-coolant-accident (LOCA) criteria to high- or intermediate-burnup fuel. The Fuel Reliability Program is the industry’s focal point for coordinating technical interactions with the Nuclear Regulatory Commission on fuel-related issues. The aim of this project is to improve fuel utilization, operational flexibility, and cycle economics without undue restrictions resulting from overly conservative regulatory assumptions about potential fuel behavior under postulated transients. This requires ensuring that proposed modifications to current licensing criteria are realistic and appropriate for such fuel when it is operated under high-duty conditions out to the currently licensed burnup limits. For additional flexibility, it also is desirable that a clear path be established for licensing even higher burnups. Recent experimental evidence has raised questions about the applicability of current reactivity-initiated accident (RIA) and loss-of-coolant-accident (LOCA) criteria to high- or intermediate-burnup fuel. The Fuel Reliability Program is the industry’s focal point for addressing such concerns on a coordinated, industry-wide basis and for technical interactions with the Nuclear Regulatory Commission on fuel-related issues. The aim of this project is to improve fuel utilization, operational flexibility, and cycle economics without undue restrictions resulting from unnecessarily overly conservative regulatory assumptions about potential fuel behavior under postulated transients. From a regulatory perspective, this requires ensuring that proposed modifications to current licensing criteria are realistic and appropriate for such fuel when it is operated under high-duty conditions out to the currently licensed burnup limits. For additional flexibility, it also is desirable that a clear path be established for licensing even higher burnups.

Approach EPRI engages U.S. and overseas regulators on fuel issues by sponsoring focused separate-effects experiments as needed and performing independent analyses and evaluations. These efforts ensure the adequacy of any proposed modifications to current reactivity-initiated accident and loss-of-coolant-accident criteria for currently licensed or anticipated new fuel designs. The project also aims at developing a framework for licensing burnup extensions beyond the current limits.



The issue of appropriate criteria for RIA and LOCA accidents continues to be actively debated with the Nuclear Regulatory Commission (NRC). With an interim RIA criteria issued by NRC’s office of Nuclear Reactor Regulation (NRR) in 2007 focused on new plants, the working group will continue to provide input in support of the final criteria for the existing fleet. This project provides the evidence needed to demonstrate the extent of unnecessary conservatisms in both areas.

Impact Independent analysis of international tests being assessed by NRC and others Application of the most advanced tools for those assessments Access to key supporting information from other Fuel Reliability Program activities Reduced regulatory impact

How to Apply Results The objective of this work is to influence NRC regulations in a manner that will allow nuclear power plants to operate safely without sacrificing efficiency and with minimal impact on current operations.

FALCON User Group (supplemental) (006165)

Key Research Question This project provides a forum for FALCON users and developers to share lessons learned and discuss code improvements necessary to enhance nuclear fuel reliability and performance. FALCON is the Electric Power Research Institute's (EPRI's) thermal-mechanical fuel performance code for steady-state and transient applications.

Approach The user group meets roughly once per year. The user group provides guidance in upgrading the FALCON code in response to industry needs and operating experience. Support also is available from EPRI throughout the year.

Impact The user group allows members to share experience, suggest improvements, and maximize the value they receive from the code.

How to Apply Results Through user group meetings and other communication mechanisms, participants share operating experience with the FALCON code and identify software improvements to address emerging needs.




NFIR-VI (supplemental) (058707)

Key Research Question The nuclear fuel industry has long recognized the need for a generic, long-term R&D program to ensure safe and reliable use of light water reactor core materials and components. Since 1982, EPRI has led the Nuclear Fuel Industry Research (NFIR) group, an international consortium of utilities, fuel vendors, and research laboratories. The NFIR program seeks to understand fundamental in-reactor behavior of fuel, cladding, control materials, and other core components and to share this valuable knowledge with the industry.

Approach In the current phase (NFIR-VI, 2010-2015), planned projects to be conducted through NFIR will address topics such as the following: Channel bow (irradiation test) Hydrogen pickup in high-burnup BWR fuel Pellet cladding interactions (test reactor irradiation) Thermal stability of irradiation and cold work defects Fuel pellet properties at high burn-up (e.g., melting point, helium release, and additives behavior)

Impact NFIR enables cost-effective, collaborative work on generic issues important to the industry, but not

necessarily tied to a specific fuel design or plant operation. All major vendors, international utilities, and research labs are members of the program wherein utility

members can advise on fuel R&D issues affecting all vendors, not just their own fuel supplier. Through NFIR participation, members have the opportunity to network with industry experts from around

the world and learn about current and anticipated issues.

How to Apply Results NFIR-VI projects will continue to provide fundamental materials properties and behavior data that lead to improved fuel products by improving the knowledge about the behavior of fuel and core components materials. The knowledge and data obtained through these projects will be factored into fuel design modifications, new fuel designs, and operational strategies targeting higher fuel reliability.


Provide the technical bases to apply advanced I&C technologies so nuclear plants can tap into functionality and capabilities underutilized to date in the nuclear sector.

Programs Instrumentation and Control


Program Overview

Program Description Instrumentation and control (I&C) systems affect all areas of plant operation and can profoundly impact plant reliability, efficiency, and operations and maintenance costs. Plants are facing changes that involve serious I&C-related challenges — equipment is getting older and cost-effective operation is more critical. The Electric Power Research Institute's (EPRI's) Instrumentation and Control program provides the technical bases to apply advanced I&C and information technologies so that existing and new nuclear plants can tap into functionality and capabilities underutilized to date in the nuclear sector. These capabilities will enable nuclear plants to maintain safe operation while managing I&C obsolescence with higher equipment reliability and personnel productivity. Three initiatives support the I&C Program mission:

1. Improve reliability of existing I&C systems & components 2. Enable the implementation of replacement I&C systems 3. Use advanced I&C to enhance plant health and productivity

Research Value Research results from the Instrumentation and Control Program enable nuclear plants to realize direct and indirect cost savings, to make technically sound system- and component-level decisions, and to comply with regulatory requirements. Instrumentation and Control Program participants gain access to the following: Life-cycle management and maintenance guidelines for generic existing I&C systems and components Regulatory compliance support such as the generic resolution of regulatory issues for new and operating

plants, risk-informed defense-in-depth and diversity assessment guidance, cyber security guidance, and guidelines for electromagnetic interference testing and digital upgrades

Technical evaluations for new technologies in nuclear applications, such as programmable controllers, “smart” sensors, and wireless communications

Guidance in setting up automated asset- and equipment-monitoring systems that will improve overall plant reliability

Improved decision making tools such as control room human factors guidelines, improved information access and visualization, and visualization-enhanced approaches for tacit knowledge capture and training

Training, operating experience, and lessons learned on I&C replacement projects that will enable plants to avoid costly mishaps and electromagnetic interference events and to implement plant strategies to cost-effectively manage I&C obsolescence

Approach The I&C Program is designed around three main initiatives: Improve Existing I&C System and Component Reliability — Develop generic technical bases for effective

maintenance and life-cycle management of I&C systems and components already installed in the plant, which will always be required to maintain and improve the reliability of the existing I&C systems and equipment.

Enable Replacement I&C System Implementation — Develop the technical bases to support the deployment and licensing of I&C and human system interface (HSI) replacement systems; develop guidelines for implementing new I&C, information, and HSI technologies in nuclear applications; and document operating experience and lessons learned.

Instrumentation and Control - Program 41.06.01 p. 1


Use Advanced I&C to Enhance Plant Health and Productivity — Reduce costly downtime and repairs by integrating new technology and techniques such as remote monitoring, wireless communication, early prognosis, and data visualization into operating and new nuclear plants. Adopt simulation and visualization technologies to streamline tasks such as training, maintenance planning, and testing, while reducing the likelihood of information overload and human error.

Accomplishments EPRI’s Instrumentation and Control Program has provided much of the fundamental basis supporting digital implementation in the nuclear industry and in identifying and overcoming many of the barriers to implementing newer technology. These include the following: Assessed the benefits of I&C defense-in-depth and diversity from a risk perspective. Higher-frequency

events such as turbine trip and loss-of-feedwater showed greater safety benefits than rarer accident sequences such as loss-of-coolant.

Developed technical guidelines for using field programmable gate arrays (FPGAs) in nuclear safety-related applications.

Captured the implementation challenges and benefits associated with on-line monitoring to support transmitter calibration interval extension at the Sizewell B plant in the United Kingdom. The use of on-line maintenance allowed a significant amount of safety-related calibrations to be removed from the outage schedule, resulting in savings of more than $1 million per outage day saved, or $5 million per operating cycle.

Issued guidance for the implementation of wireless networks in nuclear power plants, with a secondary emphasis on the use of wireless sensors for asset condition monitoring. Guidance includes technical details and real-life experiences from industry and addresses concerns such as cyber security and electromagnetic and radio frequency interference.

Assessed the benefits of I&C defense-in-depth and diversity from a risk perspective. Higher-frequency events such as turbine trip and loss-of-feedwater showed greater safety benefits than rarer accident sequences such as loss-of-coolant.

Obtained U.S. Nuclear Regulatory Commission approvals in safety evaluation reports on various guidelines/requirements (digital platforms, commercial off-the-shelf components, electromagnetic interference testing).

Current Year Activities I&C Program research and development for 2011 will focus on life-cycle management, new I&C system implementation, equipment reliability, and plant productivity. Specific efforts will include the following: Develop generic life-cycle management guidance for printed circuit card systems Develop improved failure analysis techniques for replacement I&C systems Develop training modules for implementing cyber-security technical solutions in new I&C system designs Update computer-based training modules on implementing digital I&C to facilitate utility application with

reduced cost and greater convenience Develop algorithms and techniques to integrate modeling and monitoring results to provide better

indication of equipment health


Program Manager Robert Austin, 704-595-2529, [email protected]



Summary of Projects


P41.06.01.01 I&C - Improve Reliability of Existing Systems and Components (base)

This project provides generic maintenance and life-cycle guidance for existing nuclear plant I&C systems and components, such as printed circuit card systems.

P41.06.01.01b I&C Reliability (formerly Maintenance & Life Cycle Management) (supplemental)

Instrumentation and control (I&C) maintenance and life-cycle management has emerged as a critical reliability issue for operating nuclear plants. The I&C Reliability project provides participants with implementation support for I&C maintenance and life-cycle management programs and a forum for sharing experiences and identifying research needs to address emerging problems.

P41.06.01.02 I&C - Enable Replacement System Implementation (base)

Research activities in this project aim to develop the technical bases to support the licensing of I&C and human-system interface (HIS) replacement technologies with nuclear regulators; develop guidelines for implementing new I&C and HSI technologies in nuclear applications; and document operating experience and lessons learned.

P41.06.01.02a Digital I&C Implementation (supplemental)

The Digital I&C Implementation project group coordinates two workshops per year to promote discussion and resolution of problematic digital I&C implementation issues and development of new solutions, guidance, and training materials when needed.

P41.06.01.03 I&C - Use Advanced I&C to Improve Overall Plant Health and Productivity (base)

Research activities in this project aim to reduce costly downtime and repairs by integrating new technology and techniques such as remote monitoring, wireless communication, early prognosis, and data visualization into operating and new nuclear plants. The project also develops simulation and visualization technologies to streamline tasks such as training, maintenance planning, and testing while reducing the likelihood of information overload and human error.

P41.06.01.03c I&C Monitoring (supplemental)

The I&C Monitoring project provides a forum for participants to discuss the latest progress in centralized on-line monitoring. The meeting includes a member roundtable to discuss open issues in the application of fleet-wide monitoring tools and identify gaps that can be addressed through separate research projects. A website bulletin board provides a forum for correspondence among members as well as an archive of meeting materials and links to related projects.

P41.06.01.04d I&C Productivity Improvements (supplemental)

The ability to improve plant performance and reduce operations and maintenance costs over the extended life of plants is becoming increasingly difficult with current technology and workloads. The I&C Productivity Improvements project will identify implementation opportunities through new technologies and new work task definitions that can cost-effectively improve performance, reduce costs, and lead to new plant capabilities.



I&C - Improve Reliability of Existing Systems and Components (base) (060634)

Key Research Question Nuclear plants pursuing extended operation will continue operating aging and obsolete instrument and control (I&C) systems and components such as printed circuit cards past the period of effective vendor support. Many of the maintenance and reliability concerns associated with these components are generic to component classes, independent of a particular manufacturer. Further, the transition to digital technology presents many new challenges for nuclear utilities and traditional suppliers for operating long-term, and industry guidance is needed.

Approach This project develops generic technical bases for effective maintenance and life-cycle management of I&C systems and components already installed in the plant. Near-term research includes the following: Life-cycle management guidance for analog and digital circuit card systems: EPRI will work with industry

to provide useful tools to plant engineers to improve circuit card reliability. EPRI will develop a “Gold Card” life-cycle management program for circuit cards and perform research to develop new or improved capabilities such as monitoring, testing, diagnosing, and estimating remaining life.

"Digital" I&C systems preventive maintenance: The lack of industry accepted preventive maintenance guidelines for digital systems may be leading to failures and inappropriate maintenance activities. EPRI will evaluate the data for components of interest and develop preventive maintenance guidelines.

Electronics Laboratory: EPRI will study the feasibility of starting an "EPRI Electronics Laboratory" where I&C research, operating experience consolidation, test techniques, and other items can be further developed.

Impact I&C systems and components such as printed circuit cards must be reliable to avoid unplanned plant trips and down-powers. However, too much maintenance or surveillances can lead to maintenance- or operator-induced errors. The results of this project will enable plant owners and operators to specify system tests and maintenance that will minimize the risk of plant transients while maximizing the use of scarce plant resources.

How to Apply Results Members apply project results to internal procedures for plant monitoring and maintenance to ensure that potentially problematic issues are specifically addressed and tracked through implementation. Members also should train systems engineers and maintenance personnel on the overall findings of this research for general awareness and appreciation.



I&C Reliability (formerly Maintenance & Life Cycle Management) (supplemental)

Key Research Question Aging or poorly maintained instrumentation and control (I&C) systems have resulted in numerous plant trips and power derates and have compelled nuclear plants to take repair or replacement actions to maintain plant availability and reliability. The Institute of Nuclear Power Operations has identified I&C components, in particular circuit cards, as an "Area for Improvement" at many plants. Because of the time and resources required to replace older I&C systems with modern systems, it may never make business sense in some instances to upgrade these systems. Plants will have to maintain existing systems long past the period where effective vendor support is available. In addition, replacement systems will require maintenance and life-cycle planning for eventual replacement.

Approach Many nuclear plant operators have requested EPRI assistance in developing programs for managing I&C obsolescence focused on maintaining existing components. The I&C Reliability project provides a forum to exchange plant experience, best practices, and lessons learned. Such interactions support the implementation of effective I&C maintenance and life-cycle management technology and approaches for generic I&C maintenance issues that cut across multiple systems and/or suppliers. The group ensures that EPRI research and guidance documents respond to, and evolve with, the expanding knowledge base regarding I&C maintenance and life-cycle management.

Impact I&C system and component failures are expected to increase as plants age, unless aging is carefully managed. This project provides many benefits: Broad cross-section of operating experience from which to capture lessons learned Identification of high-priority research activities to resolve I&C maintenance and life-cycle planning issues Opportunity to advise EPRI on I&C research to ensure activities address industry needs

How to Apply Results Members apply the results of this project by adapting industry lessons learned into their plant I&C programs to more effectively maintain existing I&C systems and components.

I&C - Enable Replacement System Implementation (base) (052365)

Key Research Question The nuclear industry is transitioning from analog to digital technology. For both operating and new plants, there are several I&C-related issues for which the available technical and regulatory guidance is unclear, incomplete, or evolving. Examples include failure analysis, cyber security, defense-in-depth and diversity (software common-cause failure), various design considerations for digital control rooms, and the impact of new technologies such as Field Programmable Gate Arrays (FPGAs). Having to develop technical guidance and resolve unsettled regulatory issues as part of an I&C upgrade or a licensing submittal significantly increases project costs and risks and can introduce substantial delays.

Approach Recognizing that the plant and workforce of tomorrow will demand advanced I&C and human-system interface (HSI) technology, the transformation to digital systems is inevitable. Such a transformation will equip the workforce with tools to drive improvements in plant design, operation, and maintenance. EPRI research



identifies new I&C and HSI technology for the nuclear industry and develops the technical bases necessary for use in nuclear plants. Specific research activities include the following: Provide generic methodology and guidance in support of licensing efforts for nuclear safety-related I&C

systems Develop improved safety and non-safety capabilities and establish the technical bases for their

implementation Develop guidelines to address new issues that arise due to technology advances (such as cyber security

and electromagnetic interference) Develop guidelines reflecting operating experience and lessons learned from implementing digital I&C

systems

Impact Facilitate shift toward more predictable licensing process for operating plant modernization through

generic approaches to digital I&C and human factors technical and regulatory issues. Reduce the risk, cost, and time for regulatory approval of license amendment requests and new plant

operating licenses through generic resolution of digital I&C and human factors technical and regulatory issues. For operating plants, avoiding uncertainty and delays in regulatory reviews can save years on

schedules and millions of dollars per upgrade project in extra work. For new plants, unresolved regulatory issues can add months to years to the schedule and result in

significant lost revenue opportunities, on the order of $1 million per day. Enable acceptable application of plant simulators and alternate simulation devices for engineering design

and evaluation through early user input on new designs and earlier evaluations of new designs for operating plants and new plants.

Reduce the risks and costs of implementing new technologies such as field programmable gate-array-based safety systems.

Reduce the risks and costs of responding to new issues such as cyber security Feed operating experience and lessons learned into design and maintenance processes to avoid past

mistakes and leverage past successes.

How to Apply Results Research results and interactions with industry groups enable nuclear plant owners and operators to identify, evaluate, and resolve outstanding technical issues with industry-developed guidelines. Utilities, suppliers, and third parties will use these generic solutions for design and license submittals to reduce licensing risk, cost, and time. Regulators will use the technical guidance to develop review and acceptance criteria. Guidance on new technologies such as field programmable gate-array-based safety systems will allow nuclear plant staff to carry out activities acceptable to the regulator for reducing implementation costs and risks. Guidance on new or evolving issues such as cyber security and electromagnetic interference (EMI) will reduce the risks of these external events impacting plant operations. Early use of plant simulators and alternative simulation devices will improve designs and reduce potential costly redesigns.



Digital I&C Implementation (supplemental)

Key Research Question Digital upgrades at several plants have involved significant unanticipated costs due to problems coping with various implementation issues. Examples of problematic issues with digital upgrades include unanticipated behaviors of digital equipment, software verification and validation, configuration management, evaluation of failure modes and effects, commercial grade dedication, and inadequate vendor oversight. Adverse impacts have included the following: Large increases in vendor and utility staff costs Significant project delays, as much as one or two refueling cycles Plant trips Extended outages to correct problems Additional engineering to correct problems Increased regulatory scrutiny

The problems are typically caused by inadequate knowledge and processes at the utility and its suppliers that prevent utility staff from managing the issues cost-effectively. In some cases, emerging instrumentation and control (I&C) and human-system interface (HSI) technologies include standard features that can eliminate or mitigate problems.

Approach Many nuclear plant operators have requested EPRI assistance in improving plant programs for managing the problematic issues associated with digital upgrades. In some cases, industry guidance and good practices already exist, but have not been broadly communicated or widely practiced. In other cases, practical guidance for utility engineers is simply not available. This project coordinates two meetings per year to address one or two specific application issues that are proving problematic for current digital upgrade projects. Topics include ensuring high reliability in non-safety systems, performing failure modes and effects analyses (FMEA) for digital systems, vendor interaction and oversight, and factory acceptance testing. Participants propose meeting topics, share plant experiences, discuss lessons learned, and identify areas that need additional research or guidance for utility engineers. Where appropriate, participants may develop or request new guidance and/or technical transfer mechanisms to provide practical, useful tools to plant engineers.

Impact This interest group develops and promulgates practical guidance that will help utilities anticipate, detect, and mitigate potential problems before they result in expensive learning-curve events that can cost millions of dollars. The group promotes technology transfer of the latest industry and EPRI guidance on key issues and opportunities to identify current and future research needs for solutions that will smooth the transition to digital instrumentation and control (I&C) and ensure its long-term viability. Specific technical benefits include the following: Practices that will improve utility handling of problematic digital system issues Practices that will increase utility engineers’ ability to detect and manage weaknesses in suppliers’

designs and processes for key issues, such as failure analysis, software verification and validation, and software configuration management

Technologies, strategies, and guidance that enable plant engineers to ensure long-term obsolescence management of digital systems using "design for replacement" approaches

Practical guidance and training materials for utility engineers



How to Apply Results Members will incorporate the lessons learned, guidelines, and training materials generated in this project into their processes, procedures, and training for digital upgrades.

I&C - Use Advanced I&C to Improve Overall Plant Health and Productivity (base) (052363)

Key Research Question Existing plant instrumentation and control (I&C) equipment and functionality do not accommodate up-to-date features and techniques that can reduce costs and enhance reliability and productivity. Expanded capabilities of I&C equipment and emerging technologies can streamline many plant tasks and procedures to reduce operations and maintenance costs while improving reliability and extending component lifetimes. Examples of such benefits include calibration interval extension, on-line equipment condition assessment, self-testing and diagnostics, and greatly improved access and presentation of plant data via simulation and visualization.

Approach Tapping into the capabilities of advanced I&C, human-system interface (HSI), and information technologies can increase reliability and productivity while supporting safe, cost-effective, long-term nuclear plant operation. By integrating new technology and techniques such as remote monitoring, wireless communication, early prognosis and data visualization into design, operation, and maintenance practices, nuclear plants can reduce costly downtime and repairs. Digital I&C, HSI, and information technologies also allow greater personnel productivity with expected smaller future workforces. Adoption of simulation and visualization technologies, for example, can streamline tasks such as training, maintenance planning, and testing while reducing the likelihood of information overload and human error. This research area will identify, develop, and demonstrate advanced I&C, human-system interface (HSI), and information technologies that offer enhanced equipment reliability and plant productivity.

Impact Improve work efficiency of plant engineers through the use of automated on-line monitoring tools Improve plant equipment reliability by providing more information on equipment condition Improve instrument monitoring, allowing extension of instrument calibration intervals Reduce cost of adding sensors by using wireless technology rather than cabling to support equipment

condition assessment and other applications Improve efficiency and reduce cost associated with capturing tacit knowledge from experts and

presenting it to others along with simulation and visualization of design and work planning

How to Apply Results On-line monitoring products will be applied by using the guidance and lessons learned to improve current and future on-line monitoring implementations. The wireless technology products serve as guidance when planning and implementing wireless sensors and networking technology. Results from the tacit knowledge capture task will be used to determine where and how visualization can be used to benefit the capture and dissemination of expert tacit knowledge.



I&C Monitoring (supplemental)

Key Research Question The advancing state of centralized on-line monitoring (COLM) provides a broad range of technical solutions to asset management. Several early adopters have progressed to establishing and operating fleet-wide monitoring control centers, but the industry has much to learn with respect to best practices, implementation challenges, and problem resolution. Sharing of such information among users could accelerate COLM application and value.

Approach The I&C Monitoring project provides a forum for high-level technical information exchange, distribution of current information, identification of common problems and barriers, identification of needed actions by members and EPRI, and a platform for introducing centralized on-line monitoring technology to new adopters. This forum will serve as the gateway for supporting the development and distribution of EPRI technologies. Meetings are held twice per year to discuss the latest progress in centralized on-line monitoring. The meeting includes a member roundtable to discuss open issues in the application of fleet-wide monitoring tools and to identify gaps that can be addressed through separate research projects.

Impact Project participation will further the use of COLM in the nuclear power industry. Research and development advances for COLM in the nuclear industry lag behind those in the fossil generation sector. Greater interaction and sharing of lessons learned with participants in the fossil generation sector's Fleet-wide Monitoring Interest Group will support faster distribution and application of the technology in the nuclear sector.

How to Apply Results The meeting activities will highlight application results. Attendance at the meeting will provide technology implementation opportunities through sharing of information, experience sharing, and technology transfer.

I&C Productivity Improvements (supplemental)

Key Research Question The nuclear power industry is concerned about its ability to maintain current high plant performance levels due to aging and obsolescence, knowledge drain, and fewer plant staff. Current plant operations are labor-intensive due to the vast number of operational and support activities required by the current technology. These concerns increase as plants extend their operating life. To further improve performance while reducing human errors, nuclear plants increasingly focus on operations and maintenance costs, of which labor is typically the largest contributor. New productivity improvement capabilities with measurable economic benefits are needed so that a successful business case can be made for their use.

Approach Improved and new instrumentation and control (I&C), human-system interface (HSI), information, and communications technologies can address concerns about cost-effectively maintaining current performance levels and enable shifts to even higher performance levels. This project will facilitate new technology implementation to improve productivity. Efforts will include demonstration of new technologies and how they can be used for plant and personnel productivity improvements, as well as providing pros and cons of their uses. Based on member input, required guidance for the application of technologies and pilot demonstration applications will be developed or requested to be developed.




Impact Implementation of modern technologies can provide multiple benefits: Automation of appropriate tasks will reduce workload and human stress levels, remove human error-

prone activities, and perform repetitive and time-consuming activities more effectively, allowing humans to better focus on essential activities requiring human capabilities.

Simulation and visualization will support planning and decision-making, improve designs and facilitate early input from users, support development and testing, facilitate knowledge capture and training, improve job performance, and reduce the likelihood of human errors.

HSIs and information technology will provide better user-friendliness, reduce the likelihood of human error, improve situation awareness, enable rapid access to data, and support decision-making.

Communications technologies will enable collaborative activities, including rapid access to remote expertise, which will be even more effective with the use of visualization and simulation.

How to Apply Results Members will apply the results of this project by learning how to implement advanced technologies for productivity improvements into plant modernization and workload definition plans. Potentially, pilot projects will be developed from which members can implement plant-specific applications.

Long-Term Operations

Develop the technical basis enabling nuclear plant owners to make informed decisions regarding plant operation to 80 years and beyond.

Program Long-Term Operations

Long-Term Operations (QA)

Program Overview

Program Description High capacity factors and low operating costs make nuclear power plants some of the most economical power generators available. Even when major plant components must be upgraded to extend operating life, these plants often represent a cost-effective, low-carbon asset. The decision to extend nuclear plant life involves a host of inter-related technical, economic, regulatory, and public policy issues. Unknown or uncertain technical inputs impact the decision-making process both directly and indirectly: directly, through design and operational contingencies; and indirectly, through impacts on regulatory actions and public policy. Recognizing the many technical challenges confronting nuclear plant operation in the next 60 or 80 years and beyond, the Electric Power Research Institute (EPRI) has launched the Long-Term Operations project. Participants gain access to the technical solutions needed to address long-term operational challenges, and also demonstrate a tangible commitment to addressing key societal issues related to climate change and energy security.

Research Value The Long-Term Operations Project will provide a sound, defensible technical basis for extended nuclear plant operations to 60 or 80 years and beyond through the following: Identifying and overcoming key technical barriers Investigating cost-effective modernization opportunities Positioning the industry in front of potential regulatory issues Capitalizing on substantial government, participant, and global research and development investments

Approach The LTO Project provides value by identifying and developing R&D activities that will provide the technical basis for license renewal beyond 60 years and for extended high-performance operation. The project accomplishes its objectives through an integrated strategy that encompasses research activities defined and funded through EPRI and its membership, collaboration on complementary research activities through the Department of Energy's Light Water Reactor Sustainability Program, and engagement with other key stakeholders such as the Materials Aging Institute. This strategic integration and coordination ensures a technical basis will be in place to support license renewal and life extension decisions by 2014. The activities conducted through the Long Term Operations Project will be identified and prioritized in association with nuclear plant owners, regulators, and other key stakeholders. The project will also build on the technical experience and expertise accumulated through EPRI leadership in the U.S. license renewal effort in the 1990s and early 2000s. Research products will focus on the following: Modernization and enhancement opportunities for existing plants that offer significant cost and/or

performance benefits Technical bases for evaluating continued operation of systems or components likely to be subject to

aging and considerable public and/or regulatory scrutiny Enhanced analytical capabilities that enable defensible technical assessments without long-term testing

p. 1


Accomplishments EPRI's Long-Term Operations Project has grown into a fairly large research effort after just one year. Total funding is about $5 million, including funding of $2 million from 15 companies. Significant research results will be emerging over the next few years. Key results from initial research activities include the following: Identified and prioritized long-term operations issues in an LTO Issue Tracking Table. This table,

available on the EPRI LTO website (www.epri.com/nuclearLTO), is the basis for collaborative R&D programs at the Department of Energy (DOE) and EPRI.

Compiled a report documenting concrete structures at U.S. nuclear power plants that can be used to identify and coordinate research targeting concrete aging.

Issued a report documenting good practices, barriers, and gaps related to the use of information technology for driving improvements in equipment reliability.

Initiated research to develop and validate an integrated framework and advanced tools that will enable accurate characterization and visualization of nuclear power plant safety margins. Such tools are needed to account for plant operational changes that can affect original design margins over time, such as power uprates.

Developed a project plan defining flexible functional requirements for control room and underlying instrumentation and information technology infrastructure, architecture, and associated capabilities that will support plants throughout their extended operating life.

Developed functional requirements for the Phoenix software, an advanced risk code that would enable analysis of all modes and hazards, and an integrated risk profile of the entire plant.

Current Year Activities Long-Term Operations research for 2011 will focus on the following: Applicatio n of Concrete Structures Reference Manual and Degradation Database Characterization, modeling, and mitigation of intergranular stress corrosion cracking in nickel alloys and

irradiation-assisted stress corrosion cracking in stainless steel Enhanced safety analysis and tools development for safety margin characterization Enhanced centralized online monitoring methods and pilot studies for critical systems, structures, and

components Methods and database enhancements for life-cycle management of key components, refurbishments,

and uprates Selected reports and products may be prepared in whole or in part in accordance with the EPRI Quality Program Manual that fulfills the requirements of 10CFR50 Appendix B, 10CFR21 and ANSI N45.2-1977.

Estimated 2011 Program Funding $6.0M

Program Manager John Gaertner, 704-595-2666, [email protected]

Long Term Operations (QA) - Program 41.10.01 p. 2


Summary of Projects


P41.10.01.01 Long-Term Operations (supplemental) (QA)

The Long-Term Operations Project is intended to provide the technical basis and support the decision-making process for nuclear plant operation up to and beyond 60 years. Results will also clarify significant aging concerns and modernization opportunities.

Long-Term Operations (supplemental) (QA) (069524)

Key Research Question Operators of existing nuclear plants need a technical basis for considering license renewal beyond 60 years of life. Data, analysis, and tools are needed to support decisions related to license renewal or life extension beyond 60 years, refurbishment or uprate projects, and effective management of aging systems and components while maintaining high performance and safety levels.

Approach The Long-Term Operations (LTO) Project accomplishes its objective through the following: Maintaining an integrated strategic plan to identify, prioritize, facilitate, and monitor R&D necessary to

support effective decision-making pertaining to life extension in the 2014-2019 timeframe. An LTO Issue Tracking Table will be maintained.

Focusing near-term R&D activities in two primary areas—Integrated Life Cycle Management and On-line Monitoring of Critical Assets—that will have products for implementation in 2011.

Interfacing with other EPRI Nuclear Sector technical programs to identify, prioritize, and integrate research projects that support strategic LTO objectives. Technical areas include primary metals aging, containment and concrete degradation, advanced instrumentation and controls (I&C) systems and technology, enhanced safety and risk analysis to ensure safety margins, fuel performance simulation and modeling, welding technology for reactor vessel internals, and cable testing tools and methods

Collaborating with other stakeholders to coordinate complementary long-term operations and license renewal activities. In particular, collaboration with the U.S. government and its expected $25 million of annual funding of the Light Water Reactor Sustainability Program is a primary objective. Other collaborators will include vendors, engineering service providers, Nuclear Regulatory Commission (NRC) research, and the Materials Aging Institute.

Currently, 40 percent of all funding that supports LTO objectives comes from this supplemental project and directly funds the activities above. The remaining 60 percent of funds comes from Nuclear Sector membership or EPRI Corporate and is directed to R&D activities identified in the third bullet above.

Impact Results of the project are intended to provide the technical basis and support the process for nuclear plant operation up to and beyond 60 years. Results will also clarify significant aging concerns and modernization opportunities. If this basis is largely established by 2014, nuclear plant owners can potentially secure license renewal, as well as make informed decisions regarding refurbishment, uprate, and life-cycle management to sustain high performance. Because such activities are capital-intensive, this R&D project can favorably impact costs by providing strategic planning and collaboration to augment industry resources.




How to Apply Results Demonstration of results and pilot application of technical products is an objective of this project. Key technical areas, which currently include Integrated Life Cycle Management and On-line Monitoring of Critical Assets, will have intermediate products that can be applied in the near-term. In addition, as industry identifies lead plants for license renewal, this project can support those efforts.

Materials Degradation/Aging

Increase the understanding of materials aging mechanisms in nuclear reactors and develop technologies to mitigate and monitor degradation.

Programs Primary System Corrosion Research PWR Steam Generator Management Boiling Water Reactor Vessel and Internals Pressurized Water Reactor Materials Reliability Welding & Repair Technology Center (WRTC)

Primary Systems Corrosion Research

Program Overview

Program Description Materials degradation problems due to environmentally assisted cracking in nuclear power plants have cost the nuclear industry at least $10 billion in the last 30 years because of forced and extended outages, increased inspection requirements, and component repairs and replacements. Inadequate understanding of the processes that lead to stress corrosion cracking has hampered the development of reliable predictive models and cost-effective mitigation technologies. As the majority of existing plants move toward the extended operation period, understanding of materials-aging phenomenon becomes critically important in developing sound management strategy to ensure the long-term reliability of primary system components. The Primary Systems Corrosion Research Program improves the useful life of primary system components in boiling water and pressurized water reactors through a deeper understanding of the crack initiation and early propagation processes involved in stress corrosion cracking and irradiation-assisted stress corrosion cracking. Extensive international collaboration ensures that research findings reflect a wide range of nuclear technologies, operating conditions, and service environments.

Research Value The Primary Systems Corrosion Research Program enhances nuclear industry understanding of the early stages of damage to irradiated materials used in boiling water and pressurized water reactor internals. Research results lead to improved predictive models and potential countermeasures that can significantly extend the useful life of plant components. International collaboration with utilities, vendors, regulators, and research organizations capitalizes on materials-related corrosion research around the world. Primary Systems Corrosion Research Program participants gain access to the following: Predictive models and mitigation techniques to manage material corrosion issues in reactor internals Better understanding of early stages of stress corrosion cracking in nickel-base alloys and stainless steels Methods to mitigate damage and significantly extend component life Global knowledge sharing through international collaboration with utilities, vendors, regulators, and

research organizations around the world

Approach The Primary Systems Corrosion Research Program conducts and coordinates experimental and theoretical studies to advance the mechanistic understanding and predictive modeling of crack initiation and early crack propagation. Program results are transferred to the appropriate Electric Power Research Institute (EPRI) Issues Programs (Boiling Water Reactor Vessels and Internals Project, Materials Reliability Program, and Steam Generator Management Program) for application via inspection and evaluation guidelines and other mechanisms. Develop improved detection and mitigation technologies to manage materials degradation and corrosion-

related issues Participate in international research consortia to better understand the role of key parameters on

irradiation-assisted stress corrosion cracking of reactor materials Conduct crack growth rate testing of irradiated alloy materials to evaluate the effect of material

composition, fluence, flux, stress intensity, and temperature Develop test methods to study stress corrosion cracking initiation, coalescence, and growth in nickel-base

alloys in pressurized water reactors Design management strategies to address corrosion-related plant impacts, including forced and extended

outages, increased inspection requirements, component repairs and replacements, and increased regulatory scrutiny

p. 1


Accomplishments EPRI’s Primary Systems Corrosion Research Program supports nuclear power industry efforts to identify and mitigate early stages of materials degradation to extend component life. Created an on-line Materials Information Portal to provide easy and integrated access to the Issue

Management Tables, Materials Degradation Matrices, and supporting technical reports that underlie EPRI's materials research activities

Quantified the effect of key parameters on crack growth rates of fast reactor irradiated stainless steels in boiling water reactor (BWR) and pressurized water reactor (PWR) environments

Demonstrated the strong correlation between the degree of localized deformation and irradiation-assisted stress corrosion cracking in austenitic alloys

Issued a comprehensive review of interaction between deformation and oxidation/corrosion in environmentally assisted cracking of light water reactor (LWR) materials

Current Year Activities Primary Systems Corrosion Research Program research and development for 2011 will focus on improving the understanding of corrosion mechanisms in nuclear materials, leading to more effective management and mitigation strategies for extended operation life of existing plants and better materials selection for new plants. Specific efforts will include the following:

Address emerging issues associated with extended operation life beyond 60 years Investigate susceptibility of neutron-irradiated stainless steels to irradiation-assisted stress corrosion

cracking Study the mechanisms of stress corrosion cracking initiation and short crack growth in nickel alloys Study the effect of localized deformation on stress corrosion cracking in stainless steels and nickel

alloys Investigate the effect of light water reactor environments on fracture resistance of irradiated stainless

steels and nickel-base weld metals Investigate the effect of inhomogeneous microstructure and deformation in the heat-affected zone of

Alloy 690 welds on stress corrosion cracking Explore new experiment methodology in studying stress corrosion cracking initiation in Alloys

690/52/152


Program Manager Rajeshwar Pathania, 650-855-8762, [email protected]

Primary Systems Corrosion Research - Program 41.01.01 p. 2


Summary of Projects


P41.01.01.01 Primary System Corrosion Research (QA)

The Primary System Corrosion Research program aims to develop a deeper understanding of degradation mechanisms associated with stress corrosion cracking (SCC) and irradiation-assisted stress corrosion cracking (IASCC) in light water reactors. Research projects address: IASCC susceptibility of neutron-irradiated stainless steels Mechanisms of SCC initiation and short crack growth in nickel

alloys Effect of localized deformation on SCC in stainless steels and

nickel alloys Effect of light water reactor environments on fracture resistance

of irradiated stainless steels and nickel-base weld metals Effect of inhomogeneous microstructure and deformation in the

heat-affected zone of Alloy 690 welds on SCC Experimental methodologies for studying SCC initiation in Alloys

690/52/152 Ongoing maintenance of the Materials Degradation Matrix, Materials Information Portal and materials

Primary System Corrosion Research (QA)

Key Research Question Materials degradation problems due to environmentally assisted cracking have cost the nuclear industry at least $10 billion in the last 30 years because of forced and extended outages, increased inspection requirements, component repairs and replacements, and increased scrutiny by regulators. Fundamental understanding of the early stages of stress corrosion cracking (SCC) and irradiation-assisted stress corrosion cracking (IASCC) in light water reactor materials can lead to reliable models for predicting damage progression and more effective mitigation technologies. Such understanding also will enable nuclear plants to more effectively manage materials aging issues over the life of the plant.

Approach The major elements of the Primary Systems Corrosion Research Program include the following: Address the knowledge gaps identified in the Materials Degradation Matrix and Issue Management

Tables associated with near-term and long-term operation reliability Maintain and update the Materials Information Portal, which integrates the Materials Degradation Matrix,

Issue Management Tables, issue resolution roadmaps, and materials research project information Develop experimental data, mechanistically based predictive models, and countermeasures for

irradiation-assisted stress corrosion cracking and intergranular stress corrosion cracking Investigate the effects of long-term exposure in light water reactor environments on fracture resistance

and performance of the component materials Investigate damage initiation processes using state-of-the art material characterization techniques Work with EPRI’s Technology Innovation Office to conduct strategic materials research that complements

Primary Systems Corrosion Research work Collaborate with international utilities, vendors, regulators, and research organizations around the world




Primary Systems Corrosion Research work is closely coordinated with other EPRI materials issue programs. The PSCR program will continue to develop alliances with the U.S. Department of Energy and international partners to leverage EPRI resources to address material degradation and aging issues associated with long-term operation of light water reactors.

Impact Identify key knowledge gaps in material degradation that could pose a threat to long-term reliable

operation of light water reactors, in particular to extended service life up to and beyond 60 years Develop improved predictive models and potential countermeasures for irradiation-assisted degradation

of reactor internal materials in both boiling water and pressurized water reactors Develop reliable methods to predict and mitigate the early stages of damage and to significantly extend

useful life of components in existing plants

How to Apply Results The results of the Primary Systems Corrosion Research Program are transferred to the appropriate EPRI Issues Programs (Boiling Water Reactor Vessels and Internals Program, Materials Reliability Project, and Steam Generator Management Program) for further development and application by members.

2011 Products



Crack Growth Disposition Curves for Irradiated Stainless Steels in BWR & PWR Environments -- Technical Basis Document 12/20/11 Technical Report

Compact Tension Specimen Size Validity Study for Irradiated Stainless Steels 12/20/11 Technical Report

Damage Processes Prior to Crack Initiation in Ni-Based Alloys 12/20/11 Technical Report

Effects of Environments on Fracture Resistance of Irradiated Stainless Steels and Non-irradiated Alloy 182 11/30/11 Technical

Update

PWR Steam Generator Management (QA)

Program Overview

Program Description Many factors affect materials degradation in steam generators, including water chemistry, inspection limitations, material performance issues, and the presence of foreign objects. Greater understanding of these factors and their integrated impacts will lead to more effective tools to predict the potential for degradation and more effective inspection and mitigation techniques to identify and address degradation. The Steam Generator Management Program (SGMP) conducts research to ensure the safe, reliable, and economic operation of steam generators in pressurized water reactor plants. Research activities target identification and mitigation of various forms of steam generator degradation, foreign object assessments, optimized operation of replacement steam generators, water chemistry, in-service inspections, and tube integrity.

Research Value The Steam Generator Management Program drives greater consistency in managing steam generator issues across the nuclear fleet. The program develops guidance for existing issues such as degradation in steam generators with original Alloy 600 MA tubes, as well as emerging issues such as the early and reliable detection of degradation in steam generators with the more corrosion-resistant Alloys 600 TT and 690 TT tubes. SGMP participants gain access to the following: Guideline documents that reduce the potential for steam generator tube ruptures and forced leakage

outages, which can cost an estimated $5 to $20 million per event Better tools for integrity assessments, reducing unnecessary examinations that can cost an estimated $1

to $2 million per plant Chemistry controls that can delay the onset of corrosion and mitigate steam generator fouling A database of worldwide steam generator information related to degradation, used to assist utilities with

decisions on steam generator operation and maintenance

Approach The Steam Generator Management Program applies an integrated approach for managing steam generator materials degradation in pressurized water reactors. The program develops guidance through improved understanding of how multiple variables impact steam generator operation and maintenance, including thermal hydraulics, water chemistry, tubing materials, inspection techniques, and tube-plugging/repair criteria. Coordinate industry response to unanticipated technical and regulatory issues that affect the operation of

steam generators Develop water chemistry control techniques to minimize corrosion product transport, fouling, and

corrosion damage to steam generator tubes Maintain steam generator degradation database to catalog industry experience with degradation

mechanisms and mitigation options Develop faster, more accurate methods for examining steam generator tubes, plugs, tube support

structures, secondary side deposits, and foreign objects Conduct thermal hydraulic studies to evaluate conditions in operating steam generators that could lead to

tube wear, foreign object damage, flow-induced vibration damage, and reduced performance due to deposit buildup

PWR Steam Generator Management (QA) - Program 41.01.02 p. 1


SGMP closely collaborates with other EPRI programs, including Materials Reliability, Nondestructive Evaluation, and Chemistry, to ensure appropriate technologies and technical guidance are effectively integrated into research activities.

Accomplishments The Electric Power Research Institute's (EPRI’s) Steam Generator Management Program supports nuclear power industry efforts to minimize the potential for steam generator tube ruptures, forced leakage outages, and other steam generator integrity issues. Accomplishments include both technology development and technical support, spanning more cost-effective nondestructive evaluation techniques for steam generators to technical justification for regulatory issues.

Developed and conducted a Steam Generator Engineer Training course and compiled companion training materials that can be incorporated into plant training programs

Updated guideline and supporting technical documents implementing the requirements of NEI 97-06, which imposes industry requirements for a nuclear plant’s steam generator program

Conducted research in response to the fatigue issue at Cruas Nuclear Plant (France) to assess the significance of chimney region and tube support plate buildup on steam generator tube degradation

Developed detection guidance for various types and sizes of steam generator foreign objects Upd ated the PWR Water Chemistry Guidelines, which define needed requirements and provide

guidance in optimizing a plant’s chemistry program Developed new applications for dispersant use beyond online addition to significantly reduce steam

generator fouling: dispersant addition during steam generator wet layup as well as during the long-path recirculation cleanup of the condensate and feedwater systems just prior to plant startup

Analyzed the impact of advanced amine use on iron transport, flow accelerated corrosion, and steam generator fouling in pressurized water reactor (PWR) secondary systems

Developed a technical basis for modifications to water chemistry guidelines to reflect alloy-specific improvements in degradation resistance

Conducted research on electrochemical kinetics of Ni and Pb to understand the mechanics of Pb stress corrosion cracking

Current Year Activities Steam Generator Management Program research and development for 2011 will focus on continued development of dispersant applications; advanced inspection and inspection analysis methods; root causes of steam generator degradation; and guideline revisions related to water chemistry. Specific efforts include the following: Publish PWR Primary-to-Secondary Leak Guidelines, Revision 4 Develop a model for assessing tube wear due to foreign objects Develop a technique to more rapidly perform required eddy current technique equivalencies Develop PC-based software to perform site-specific data analyst performance demonstrations Develop experience-based models that can be used to predict the rate of deposit buildup and the rate of

tube corrosion under the deposits as a function of time Conduct performance demonstration of a circumferential primary water stress corrosion cracking data set

to establish total system uncertainties for eddy current inspection techniques Conduct field trial of dispersant application during the long-path recirculation cleanup process Improve understanding of Pb stress corrosion cracking fundamentals in order to identify possible field

mitigating actions Selected reports and products may be prepared in whole or in part in accordance with the EPRI Quality Program Manual that fulfills the requirements of 10CFR50 Appendix B, 10CFR21 and ANSI N45.2-1977. Reports and products developed under the EPRI QA program will be marked and identified as such.




Program Manager James Benson, 704-595-2550, [email protected]

Summary of Projects


P41.01.02.01 Materials Performance and Thermal Hydraulics (base)

This project conducts experiments and develops computational simulations that more accurately estimate foreign object movements and tube wear rates from steam generator foreign objects. This project also performs stress corrosion cracking tests that can be used to estimate steam generator tube degradation.

P41.01.02.02 Nondestructive Evaluation R&D (base)

This project develops tools such as software algorithms, improved inspection techniques, and database libraries to enhance the accuracy and efficiency of steam generator inspections. This project also creates realistic stress corrosion flaw samples in steam generator tubing for use in qualification of inspection techniques.

P41.01.02.03 Advanced Water Chemistry (base)

This project develops guideline documents, chemistry technologies, and predictive models to provide utility chemists with resources to optimize chemistry for safe, reliable, and long-term steam generator operation.

P41.01.02.04 Steam Generator Degradation Database (base)

EPRI has developed a web-accessible database of worldwide steam generator information, the Steam Generator Degradation Database, that is available to all EPRI members. This project maintains this database, which is essential for meeting utility information needs, with complete and accurate data, reported in a consistent manner.

P41.01.02.07 Supplemental Research and Emerging Issues (supplemental)

The products resulting from base-funded projects often lead to further research that addresses a specific issue for a subset of the original funding group. Such research may include additional capabilities for available software products, database maintenance for alternate tube repair criteria, and reviews of newly developed chemical additives. The industry also occasionally encounters operational and regulatory issues not anticipated and included in annual plans. Such emerging issues can be addressed and funded through this project.

P41.01.02.08 Structural Integrity Assessment and Nondestructive Evaluation Field Support (QA)(supplemental)

This project develops products to ensure steam generator tube integrity through thorough inspections, condition monitoring, and operational assessments. Cornerstone products include the SG Examination Guidelines and its qualification program, the SG Integrity Assessment Guidelines, the SG Primary-to-Secondary Leakage Guidelines, and the SG In Situ Pressure Test Guidelines.



Materials Performance and Thermal Hydraulics (base) (061428)

Key Research Question Foreign objects and tube wear can threaten safe and reliable steam generator operation. An understanding of these phenomena is needed to develop predictive tools and take actions to minimize the potential for degradation that exceeds tube structural limits. Prediction tools and on-line measurement techniques also are needed to enable utilities to manage issues resulting from buildup on tube support plates. Finally, corrosion studies on steam generator tubing materials are needed to determine the effect of various steam generator environments on the rate of tube degradation and allow accurate long-term predictions on the initiation and growth of tube degradation.

Approach This project conducts experiments and develops computational simulations that more accurately estimate foreign object movements and tube wear rates from steam generator foreign objects. Models are developed to predict tube support plate build up as a function of time. This project also performs stress corrosion cracking tests that can be used to estimate steam generator tube degradation.

Impact Determine maximum inspection intervals based on predictions of tube wear depths Provide guidance on long-term tube repair decisions and strategies Improve understanding of stress corrosion cracking

How to Apply Results Plant engineers use EPRI reports to evaluate potential steam generator tube wear damage from foreign objects and wear at steam generator support structures. Plant engineers also use research results to better evaluate the potential for stress corrosion cracking in 600TT and 690TT tubes.

2011 Products



Steam Generator Management Program: Prediction of Tube Support Plate Build Up 12/23/11 Technical Report

Steam Generator Management Program: Conditions Causing Lead Stress Corrosion Cracking of Steam Generator Tubing 12/23/11 Technical

Update

Nondestructive Evaluation R&D (base) (061427)

Key Research Question Current inspection methods for steam generator tubes are limited in their ability to detect and size tube degradation. For utilities with replacement steam generators, inspection limitations may limit the length of the inspection interval, eliminating possible cost savings. In addition, inherent errors and inconsistencies associated with manual data analysis can potentially be eliminated through the development and qualification of automatic data analysis algorithms. Documenting the performance of NDE techniques at different locations and for different forms of degradation requires a large variety and number of realistic flaw samples. Fabrication of laboratory-induced flaws is often needed when tubes removed from service do not provide an adequate number of realistic flaws.



Approach Tools will be developed to improve the accuracy and efficiency of steam generator inspections. Tools may include software algorithms for automatic analysis of inspection data, a database library for documenting the performance of automatic data analysis software, procedures for determining examination technique equivalency, guidance for improved data analysis, and improved inspection techniques. This project will fabricate tube flaw samples for current and new generation steam generators. Samples of the most generic flaw types and configurations will be fabricated that replicate service-induced stress corrosion cracking.

Impact Potentially reduce steam generator inspection costs and duration Improve steam generator flaw detection and sizing accuracy to justify longer inspection intervals Provide realistic flaw samples to validate the performance of NDE techniques for field use

How to Apply Results With improved flaw detection and flaw-sizing accuracy capabilities, nuclear plants could potentially justify longer operating intervals. Algorithms for automatic analysis of inspection data could be used by nondestructive evaluation vendors to improve inspection speed and accuracy during steam generator inspections. Utilities will implement inspection techniques qualified through use of flaw samples developed under this project.

2011 Products



Accurately Size Foreign Object Wear from the Steam Generator Tube Secondary Side: As a result of a feasibility study, develop a technique to accurately size foreign object wear from the secondary side of the steam generator. A prototype will be developed and tested during this phase of the project.


Evaluation of NRC/ANL Data Analysis Algorithms: Algorithms for analysis of steam generator tube eddy current data developed by Argonne Laboratories will be evaluated for potential enhancement to AutoAnalysis algorithms.


Development of AutoAnalysis Algorithms for SG Tubing: AutoAnalysis algorithms for steam generator tubing eddy current testing will be developed for enhanced detection of degradation. Algorithms will be available for implementation in commercial automated analysis software products.

04/30/11 Software

Steam Generator Eddy Current Simulation Tool Ver. 3: Develop an accurate computational model for simulating signals representing steam generator eddy current degradation and incorporate the model into a user friendly software tool.

03/30/11 Software

Fabrication of Steam Generator Tubing Samples 12/31/11 Technical Report

Advanced Water Chemistry (base) (052334)

Key Research Question Corrosion product transport into steam generators can foul tube surfaces and create crevice environments for the concentration of corrosive impurities. Improved water chemistry control can minimize this transport mechanism, leading to reduced fouling and corrosion damage within the steam generators.



Approach Secondary and primary water chemistry guidelines are developed and periodically reviewed and revised as needed to reflect technology developments and industry experience. Advanced technology developments are incorporated in application sourcebooks, which provide assistance to plant chemists on water chemistry control, including improved amines, dispersant, molar ratio control, and intergranular stress corrosion cracking inhibition (boric acid and TiO2 addition). The influence of amines and dispersants on corrosion product deposition and removal from steam generators will be investigated, as well as the role of lead in stress corrosion cracking.

Impact Industry guidance in primary chemistry, secondary chemistry, steam generator lay-up, and hideout return Continued assessment of dispersant application to mitigate steam generator fouling Improved understanding of high-temperature lead chemistry as a means to develop remedial strategies

against lead-induced stress corrosion cracking

How to Apply Results Plant chemists incorporate EPRI chemistry guidance into plant operating procedures to minimize steam generator tube corrosion and steam generator tube fouling. In addition, plant chemists will review the results of EPRI studies on various additives for controlling steam generator tube corrosion and fouling to assess application at their plants.

2011 Products



Development of Predictive Models for Deposit Accumulation and Corrosion on the Secondary Side of Steam Generators: Experience-based empirical models can predict the probable rates of deposit accumulation and tube corrosion at support intersections and at top of tubesheet areas of steam generators. This project will build on completed work on a quantitative model characterizing line contact support plate fouling, which was funded by the Electric Power Research Institute (EPRI) Technology Innovation program.


MULTEQ Database Update: MULTEQ is the leading tool for understanding the primary and secondary chemistry environment. The MULTEQ database is continually reviewed and revised by a select committee of experts. This project will provide the latest update and improvements to the MULTEQ database.


Evaluation of Additives for Inhibiting PbSCC: The objective of this work is to identify possible field remedial actions that can be implemented to mitigate lead stress corrosion cracking (PbSCC). Electrochemistry will be used to evaluate the effectiveness of additives that slow plating kinetics and precipitate Pb in removing Pb from solution.


PWR Monitoring and Assessment: This product will investigate and identify correlations between chemistry parameters and operating experiences, including benchmarking. In addition, it will provide supporting information to strengthen and improve the technical guidance documented in the Electric Power Research Institute (EPRI) pressurized water reactor (PWR) water chemistry guidelines.




Steam Generator Degradation Database (base) (052592)

Key Research Question Utility management and steam generator engineers can make more effective operational and maintenance decisions if fully informed by industry steam generator operating experience. A common source of up-to-date and easily retrieved steam generator information, including degradation mechanisms, can provide this capability.

Approach EPRI has developed a web-accessible database of worldwide steam generator information, the Steam Generator Degradation Database (SGDD), that is available to all EPRI members. This project maintains this database, which is essential for meeting utility information needs, with complete and accurate data, reported in a consistent manner.

Impact SGDD helps to ensure the safe and reliable operation of steam generators by providing data to plant engineers to help in determining inspection scope, planning for tube repair activities, and determining the effectiveness of various steam generator corrective action programs (for example, chemical cleaning).

How to Apply Results Steam Generator Database information assists plant engineers in preparing various steam generator assessment documents. Steam generator experience is essential in determining if negative industry events could occur at a plant and in planning for potential events that affect steam generator safety or operation.

2011 Products



Steam Generator Degradation Database, Version 7.1 12/23/11 Software

Supplemental Research and Emerging Issues (supplemental) (058768)

Key Research Question Research is often needed to address issues of interest to a smaller subset of utility steam generator operators or to address emerging issues that require urgent attention. The products resulting from base-funded projects sometimes lead to further research that addresses a specific issue applicable to only certain nuclear plants; concurrently, not all steam generator operational and regulatory issues can be anticipated and included in annual plans. Flexibility is needed within the steam generator research scope to respond to these issues, which could include development of additional capabilities for software products, databases to support alternate tube repair criteria, and reviews of newly developed chemical additives.

Approach Projects without generic application to all steam generator owners can be pursued through this project, which enables individual utilities or multiple utilities to define and fund targeted research efforts. For example, EPRI software codes have been developed to assess the thermal hydraulic conditions present in steam generators and to determine the rate of tube wear from both steam generator support structures and foreign objects. These codes are maintained and updated based on input from a user group that captures lessons learned and improvement ideas.



This project also manages an Emerging Issues fund to address unanticipated and emerging issues of high priority throughout the year. The Steam Generator Management Program (SGMP) Integration Committee evaluates and determines candidate emerging issues to be addressed by this fund.

Impact Address plant-specific research needs and generic emerging issues in a timely manner Address research needs of interest to a subset of utilities Eliminate or reduce impact of emerging issues on ongoing or planned projects

How to Apply Results Information will be distributed to SGMP members via information letters, interim guidance, technical reports, and workshops. For targeted research projects, results will be provided to those companies funding the work. The method of applying the research would depend on the specific product.

Structural Integrity Assessment and Nondestructive Evaluation Field Support (QA) (supplemental) (061426)

Key Research Question Nuclear plant owners and operators maintain steam generator programs to resolve current and near-term engineering, regulatory, and inspection issues. A principal concern is steam generator tube integrity. Other issues of concern include divider plate cracking, the detection and sizing of tube wear caused by foreign objects, improved leak rate modeling, and inspection system uncertainties for all degradation mechanisms.

Approach This project develops products to ensure steam generator tube integrity through thorough inspections, condition monitoring, and operational assessments. Cornerstone products include the SG Examination Guidelines and its qualification program, the SG Integrity Assessment Guidelines, the SG Primary-to-Secondary Leakage Guidelines, and the SG In Situ Pressure Test Guidelines.

Impact Minimize the potential for steam generator tube ruptures and forced leakage outages, which can cost $5

to $20 million per event Develop better tools for integrity assessments, which can reduce unnecessary examinations costing $1 to

$2 million per plant

How to Apply Results Results are detailed in the SGMP guidelines, whose implementation is mandated by industry initiatives such as NEI 97-06 and NEI 03-08. Members use this information to develop in-house procedures in compliance with Technical Specifications and NEI 97-06 requirements.




2011 Products



Rev. 4 of In Situ Pressure Test Guidelines: Guideline document that provides requirements for performing in situ pressure testing of steam generator tubes. This document is a required guideline for implementation of NEI 97-06.


Rev. 8 of SG Examination Guidelines: Guideline document that supports requirements of NEI 97-06 and NEI 03-08 for nondestructive evaluation of steam generator tubing.


Steam Generator Management Program: Steam Generator Secondary Side Corrosion 09/30/11 Technical Report

Steam Generator Management Program: Onset of Fatigue Cracking 12/23/11 Technical Report

Steam Generator Management Program: PWR Primary-to-Secondary Leak Guidelines, Revision 4 12/23/11 Technical Report

Boiling Water Reactor Vessel and Internals (QA)

Program Overview

Program Description As boiling water reactors have aged, various forms of operation-limiting stress corrosion cracking have appeared, first in the recirculation piping, then in the reactor pressure vessel internals. Typically, cases of poor materials performance have been addressed by analyses focused on the specific component or system. This near-term, reactive approach has resulted in costly unplanned outages and expensive weld-by-weld mitigation and repair methods. A longer-term, strategic approach can address a broader range of factors impacting pressure vessel internals. The Boiling Water Reactor Vessel and Internals Project (BWRVIP) provides an integrated approach for managing materials-related degradation issues in reactor coolant system components in boiling water reactors. The program assesses all facets of operation, maintenance, and repair to develop reliable and cost-effective detection, inspection, and mitigation techniques.

Research Value The BWRVIP Program maintains alignment with current industry internals integrity concerns affecting boiling water reactors. Research results lead to cost-effective solutions to reduce damage related to stress corrosion cracking; cost savings due to reduced inspection scope, extended intervals between inspections, and improved operating characteristics; reduced personnel radiation exposure; and improved models to better characterize the mitigation of internals components. BWRVIP participants gain access to the following: Technologies and technical guidance that drive increased capacity factors (less unplanned or extended

outages). Cost-effective techniques to mitigate stress corrosion cracking of reactor internal components. Economic

evaluations indicate that cost savings for implementing hydrogen water chemistry or noble metal chemical application exceed $40 million per plant.

Cost-effective options for replacing or repairing reactor components. Technical solutions to internals inspection needs. Industry operating experience and technical insights driving reduced inspection requirements, outage

critical path times, and regulatory scrutiny.

Approach The BWRVIP Program takes an integrated approach to degradation management, encompassing assessment, mitigation, and inspection. Through improved inspection techniques, new results from materials research and development, and plant operating experiences, best practices can be deployed to make cost-effective decisions. Develop guidelines to ensure prompt detection of material degradation and a variety of solutions for

addressing observed problems Develop and demonstrate cost-effective means to implement techniques to mitigate stress corrosion

cracking of reactor internal components Devise unique solutions to internals inspection needs such that boiling water reactor (BWR) plants have a

wider selection of nondestructive evaluation offerings Formulate design criteria and develop unique solutions to repair or replace reactor internals and piping Improve the understanding of materials performance in areas such as fracture toughness of stainless

steel exposed to high fluence levels, weldability of irradiated materials, and crack growth rates BWRVIP closely collaborates with other EPRI programs, including Nondestructive Evaluation and Chemistry, to ensure appropriate technologies and technical guidance are effectively integrated into research activities.

Boiling Water Reactor Vessel and Internals (QA) - Program 41.01.03 p. 1


Accomplishments The BWRVIP Program supports nuclear power industry efforts to assess and implement effective countermeasures for stress corrosion cracking of reactor internal components. BWRVIP research provides utilities with the information necessary to make cost-effective decisions for managing degradation of boiling water reactor vessel and internal components. Issued NRC-approved guidelines on steam dryer inspection and flaw evaluation that define visual

inspection requirements for BWR steam dryer assemblies. Baseline inspection results can be compared to subsequent results to assess potential effects of time and power uprates.

Completed construction, shakedown testing, and initial operation of full-scale jet pump facility to advance understanding of observed degradation mechanisms.

Issued guidelines for performing weld repairs on irradiated BWR internals, providing a mechanism for determining the weldability of reactor components.

Compiled data to advance understanding of the relationship between fracture toughness and neutron fluence in highly irradiated stainless steel materials. Also conducted tests to collect crack growth rate data on irradiated stainless steels that can be used to extend EPRI's flaw evaluation methodology to higher neutron doses.

Designed, developed, and demonstrated a radiographic testing system to facilitate corrosion detection in difficult-to-access BWR drain lines.

Demonstrated online noble metal chemical addition as a mitigation technique for stress corrosion cracking. Field tests indicate critical path savings of up to 60 hours.

Current Year Activities BWRVIP research and development for 2011 will continue to focus on the technical gaps defined in the BWR Issue Management Tables. Highest priority gaps include the impacts of fluence on the material properties of BWR materials, high-cycle fatigue in jet pump assemblies, and flow-assisted corrosion of the BWR bottom head drain line. Specific efforts will include the following: Continue testing of the full-scale jet pump facility to better understand the operating degradation

mechanism and to evaluate mitigation technologies Support the steam dryer loads methodology as it works its way through Nuclear Regulatory Commission

review Continue crack growth and fracture toughness evaluations of highly irradiated materials Develop advanced mitigation techniques for stress corrosion cracking Complete inspection and evaluation guidelines for all BWR internals’ components

Selected reports will be developed in whole or in part under Title 10 of the Code of Federal Regulations Part 50 (10 CFR50) Appendix B, Quality Assurance, 10 CFR 21, and the EPRI Quality Assurance Program. Additional products may be developed under 10 CFR 50 Appendix B, and 10 CFR 21 at the discretion of the BWRVIP member utilities or EPRI, when such action is deemed appropriate.


Program Manager Randal Stark, 650-855-2122, [email protected]



Summary of Projects


P41.01.03.01 BWRVIP Improved Materials Performance (base)

Utilities must manage current and potential future degradation of boiling water reactor internal components. This program provides members with improved understanding of materials performance in the areas of fracture toughness of stainless steel exposed to high fluence levels, weldability of irradiated materials, and crack growth rates of internal components subjected to irradiation-assisted stress corrosion cracking and intergranular stress corrosion cracking.

P41.01.03.02 BWRVIP Integration (supplemental)

This task provides overall BWRVIP technical and administrative program management. Activities include oversight and attendance at BWRVIP meetings; interface with the Nuclear Regulatory Commission on all BWRVIP matters; coordination with international BWRVIP members and potential members; BWRVIP report licensing and distribution; contract management with BWRVIP contractors; and coordination regarding EPRI contracts with U. S. and international BWRVIP members. This task includes preparing materials for BWRVIP training sessions and conducting domestic and international training sessions.

P41.01.03.03 BWRVIP Assessment, Inspection and Repair (supplemental)

BWRVIP provides utilities with the tools and information needed to manage degradation of boiling water reactor vessel and internal components. The products that come from this task target three areas: assessment, inspection, and repair. Assessment: Inspection and evaluation guidelines provide the

scope for what needs to be inspected and a methodology for evaluating or repairing any indications that are found.

Inspection: Advanced nondestructive evaluation techniques improve detection of indications in internals’ components so they can be assessed and repaired/mitigated to maintain safety margins.

Repair: Technically based repair criteria for degraded components equip nuclear plants with the information needed to confidently plan and implement repairs.

P41.01.03.04 BWRVIP Mitigation (supplemental)

This program provides guidance for implementing effective chemistry-based countermeasures for stress corrosion cracking of reactor internal components. Research results and guidelines will enable members to effectively implement techniques for mitigating stress corrosion cracking, such as hydrogen water chemistry and noble metal chemical application. Work also will be conducted to demonstrate the effectiveness of mitigation techniques and the effect of noble metal chemical application on fuel.

P41.01.03.06 BWRVIP Integrated Surveillance (supplemental)

Each BWR has a surveillance program for monitoring changes in reactor pressure vessel material properties due to neutron irradiation. Substantial cost savings and improvements in data quality are possible by integrating these individual surveillance programs. This program helps utilities optimize the quality of data and number of materials used to monitor embrittlement of BWR reactor vessel materials.





P41.01.03.07B BWRVIA - BWR Vessel and Internals Application User Group (ISO 9001) (supplemental)

The BWR Vessel & Internals Application (BWRVIA) computer code, which performs radiolysis analysis and electrochemical corrosion potential calculations for BWRs, has progressed through several upgrades to enhance its value to the nuclear power industry. Continued maintenance through this project ensures the code reflects the latest industry operating experience and is equipped to address analytical needs. The BWRVIA User Group provides training and ongoing support to all BWR utilities using the model and participating in this program.

BWRVIP Improved Materials Performance (base) (052368)

Key Research Question Utilities must manage current and potential future degradation of boiling water reactor internal components. Because changes in materials, welding practices, and fluence levels can all impact materials aging issues, the industry must continually evaluate the interactions of such changes with materials performance.

Approach This program provides members with improved understanding of materials performance in the areas of fracture toughness of stainless steel exposed to high fluence levels, weldability of irradiated materials, and crack growth rates of internal components subjected to irradiation-assisted stress corrosion cracking and intergranular stress corrosion cracking.

Impact Cost-effective tools to enable members to identify and manage degradation for current license terms and

for the license renewal period Regulatory approval of many of the products allows members to effectively address regulatory issues Reduced inspection scope, extended intervals between inspections, and improved operating

characteristics

How to Apply Results This program will be delivered through a combination of guidance documents and technical reports throughout its duration. Many of the guidance documents will be submitted to the NRC for approval, which will result in effective member implementation of regulatory-approved guidance.

BWRVIP Integration (supplemental) (062248)

Key Research Question The broad technical scope of materials issues impacting BWRs, coupled with the extensive regulatory interactions required to address these issues, call for an integrated coordination effort. This project ensures that the Boiling Water Reactor Vessel and Internals Project (BWRVIP) Integration Committee remains aligned with current industry internals integrity concerns. The overall BWRVIP issue management strategy is reviewed and modified as necessary, and assistance is provided to the Integration Committee to ensure that BWRVIP work priorities reflect the best industry information. Maintain alignment of the Boiling Water Reactor Vessel and Internals Project (BWRVIP) Integration

Committee with current industry internals integrity concerns


Manage, develop, and modify, as appropriate, the overall BWRVIP issue management strategy Assist the Integration Committee in ensuring that BWRVIP task activities are coordinated and that work

priorities reflect the best industry information.

Approach EPRI activities in this task include overall BWRVIP technical and administrative program management to support the Integration Committee; costs for conducting and attending BWRVIP meetings; interface with the Nuclear Regulatory Commission (NRC) on all associated BWRVIP matters; coordination and interface with international BWRVIP members and potential members; BWRVIP report licensing and distribution; contract management with BWRVIP contractors; and coordination and interface regarding EPRI contracts with all U.S. and international BWRVIP members. This task includes preparing materials for BWRVIP training sessions and conducting domestic and international training sessions. The task also includes participating in Institute of Nuclear Power Operations’ review visits and developing and documenting interpretations and implementation issues associated with BWRVIP products.

Impact Successful overall BWRVIP program management and regulatory interface Close coordination and prioritization of the various task activities within BWRVIP

How to Apply Results The work performed under this task provides the information necessary for members to better manage BWR power plants. Program results enable nuclear plants to apply the operating experience and lessons learned from other plants, respond to emerging industry issues, and better understand the regulatory aspects of the program.

BWRVIP Assessment, Inspection and Repair (supplemental) (052371)

Key Research Question As boiling water reactors have aged, various forms of operation-limiting stress corrosion cracking have appeared, first in the recirculation piping and then in the reactor pressure vessel internals. To systematically address such degradation issues—from assessment to inspection to repair—the industry developed the Boiling Water Reactor Vessels and Internals Project. Through standard guidelines for inspection and evaluation, improved inspection techniques, and detailed repair criteria, best practices can be deployed to make cost-effective decisions. To best serve the industry, these best practices should be communicated fleet-wide such that they are consistently applied and continually updated as new information is available.

Approach To assist utilities in properly characterizing and planning for potential materials degradation, BWRVIP develops assessment tools such as inspection and evaluation guidelines for internals components and standardized methodologies for fluence evaluation. These tools provide utilities with the necessary information to make cost-effective decisions for managing degradation. To improve detection of indications in internals’ components so they can be assessed and repaired/mitigated to maintain safety margins, BWRVIP develops and refines nondestructive evaluation techniques. Past examples of inspection programs developed under this activity include creeping wave ultrasonic testing; phased-array ultrasonic and eddy current testing for the shroud; core plate bolt nondestructive evaluation from the annulus; ultrasonic testing for shroud support legs; and alternative ultrasonic testing for jet pump beams. To provide the industry with the technical bases for effective repairs, BWRVIP develops detailed repair criteria that define important considerations when evaluating and planning structural or mechanical repairs to BWR vessel and internals components. Repair approaches are provided to the regulator for review and approval, increasing confidence in their application.



Impact Standardized and regulator-approved methodology for fluence evaluation Inspection and Evaluation Guidelines to ensure prompt detection of degradation, reduce outage time due

to unanticipated degradation, and provide cost-effective solutions for reducing inspections and damage related to stress corrosion cracking

Eliminated testing of some surveillance program capsules of low value to the industry Improved data and information on changes in reactor pressure vessel material properties due to neutron

irradiation Nondestructive evaluation solutions to internals inspection needs that reduce inspection and outage

critical path times, reduce personal radiation exposure, and result in an estimated $25,000 to $100,000 savings per member per year

Availability of unique nondestructive evaluation solutions such that BWRVIP participants are not limited to vendors’ nondestructive evaluation offerings

Generic design criteria approved by the Nuclear Regulatory Commission (NRC) Generic repair and replacement options

How to Apply Results The products that come from this task are directly applied at nuclear power plants through inspection, maintenance, and repair programs. The inspection and evaluation guidelines provide the scope for what needs to be inspected and a methodology for evaluating or repairing any indications that are found. Nondestructive evaluation results provide information that members can use to determine which vendor and which inspection technique should be used for specific inspection needs. Mitigation research results enable members to optimize water chemistry programs, allowing plants to mitigate stress corrosion cracking of reactor internals and recirculation piping without affecting other plant parameters (for example, dose and fuel).

BWRVIP Mitigation (supplemental) (052372)

Key Research Question As boiling water reactors have aged, various forms of operation-limiting stress corrosion cracking have appeared, first in the recirculation piping, and then in the reactor pressure vessel internals. Chemistry-based technologies can be applied to help mitigate such corrosion, but require laboratory and field demonstration to confirm their capabilities.

Approach This program will provide guidance for implementing effective countermeasures for stress corrosion cracking of reactor internal components. Research results and guidelines will enable members to effectively implement techniques for mitigating stress corrosion cracking, such as hydrogen water chemistry and noble metal chemical application. Work also will be conducted to demonstrate the effectiveness of mitigation techniques and the effect of noble metal chemical application on fuel.

Impact Cost-effective techniques to mitigate stress corrosion cracking of reactor internal components; economic

evaluations conducted for five plants indicate that cost savings for implementing hydrogen water chemistry or noble metal chemical application exceed $40 million per plant

Cost savings due to reduced inspection scope, extended intervals between inspections, and improved operating characteristics

Savings also expected in optimized use of costly chemicals



How to Apply Results Members would use the results from this project to optimize water chemistry programs, allowing plants to mitigate stress corrosion cracking of reactor internals and recirculation piping without affecting other plant parameters (for example, dose and fuel).

BWRVIP Integrated Surveillance (supplemental) (052373)

Key Research Question Each boiling water reactor has a surveillance program for monitoring changes in reactor pressure vessel material properties due to neutron irradiation. Substantial cost savings and improvements in data quality are possible by integrating these individual surveillance programs.

Approach The Integrated Surveillance Program provides an integrated plan for monitoring BWR reactor pressure vessel embrittlement, resulting in significant savings compared to individual programs. Materials chosen for the Integrated Surveillance Program best represent the limiting plate and weld materials for each plant using specimens from the entire BWR fleet. The Integrated Surveillance Program is a regulatory commitment for all U.S. BWRs through the end of life.

Impact Neutron irradiation exposure reduces the toughness of reactor vessel steel plates, welds, and forgings. Accurate methods for monitoring radiation embrittlement are important for evaluating the remaining life of reactor pressure vessel materials. The Integrated Surveillance Program will result in significant cost savings to the BWR fleet and provide more accurate monitoring of embrittlement in BWRs.

How to Apply Results This program and sourcebook data will help member utilities optimize the quality of data and number of materials that will be used to monitor embrittlement of BWR reactor vessel materials and ensure that the Integrated Surveillance Program will comply with the requirements of 10CFR50, Appendix H.

BWRVIA - BWR Vessel and Internals Application User Group (ISO 9001) (supplemental) (047065)

Key Research Question Two technologies—moderate hydrogen injection, known as hydrogen water chemistry, and noble metal chemical addition—have been applied in boiling water reactors to mitigate intergranular stress corrosion cracking (IGSCC) by lowering primary water electrochemical corrosion potential. Analytical capabilities are needed to determine appropriate injection concentrations that can maintain electrochemical corrosion potential values at levels that mitigate corrosion. This user group provides information and training on the use of the radiolysis and electrochemical corrosion potential models used in EPRI’s BWR Vessel and Internals Application (BWRVIA) software program. Ongoing development of the codes also is evaluated and reported out at the annual member's meeting.

Approach The BWRVIA User Group provides technology that operating BWRs can use to help mitigate IGSCC of reactor piping and internals. The technical project team performs comprehensive reviews of research and development in the areas of radiation chemistry and electrochemical corrosion potential modeling. Sensitivity analyses are performed to evaluate the model's response due to changes in input parameters such as chemical reaction rate constants and dose rate profiles. Adjustments are then made to these sensitive parameters to provide the best




possible correlations. Finally, the results of the sensitivity analyses are compared to actual plant data to provide a technical basis for plant application of the calculated results.

Impact By incorporating the current state of the art in radiation chemistry and electrochemical corrosion potential formulation into the BWRVIA code, and benchmarking the revised code against all plant and laboratory data available, this user group ensures the availability of an accurate model for BWR plant owners. The model can then be used, for example, to predict the amount of hydrogen injection needed for IGSCC mitigation of susceptible reactor internals and piping in BWRs.

How to Apply Results The BWRVIA User Group provides annual training workshops and ongoing support to run the software.

Pressurized Water Reactor Materials Reliability Program (QA)

Program Overview

Program Description Stress corrosion cracking and general environmental corrosion of reactor coolant system components have cost the nuclear industry billions of dollars due to forced and extended outages, increased inspection requirements, component repairs and replacements, and increased regulatory scrutiny. Materials aging effects must be effectively managed to ensure safe and reliable functionality is maintained throughout the life of the plant. Further, a better mechanistic understanding of crack initiation and propagation processes and environmental corrosion in the reactor coolant system components is needed to develop reliable predictive models and cost-effective mitigation technologies. The Materials Reliability Program (MRP) conducts research to identify and resolve existing and potential issues impacting pressure boundary materials in pressurized water reactors. Research activities inform operational and maintenance decisions for existing plants, design choices for new reactors, and regulatory actions pertaining to material aging and degradation mechanisms. These activities are coordinated among pressurized water reactor owners and operators to ensure the plants are aggressively addressing materials degradation and aging and meeting the intent of industry materials initiatives.

Research Value MRP aligns with industry and regulatory concerns regarding materials degradation in pressurized water reactors and pursues cost-effective inspection, evaluation, and mitigation approaches for addressing degradation. Coordinated activities ensure plants can maintain safe operation and avoid unnecessary outages. MRP participants gain access to the following: Detailed inspection and evaluation guidelines for susceptible areas of the reactor coolant system in

pressurized water reactors Safety and operational assurance promoting long-term reliable operation of pressurized water reactors Chemical and mechanical mitigation technologies for aging degradation mechanisms Increased credibility with regulators by effectively managing in-service degradation without the need for

extra regulatory mandates Guidance and tools for fatigue-specific materials management in existing plants and design guidance for

new plants to address environmentally assisted fatigue

Approach The MRP takes an integrated approach to degradation management in pressurized water reactors, encompassing assessment, mitigation, and inspection. Through improved inspection techniques, new results from materials research and development, and plant operating experiences, best practices can be deployed to make cost-effective decisions. Specific activities include the following: Improved understanding of issues affecting pressure boundary materials in pressurized water reactors:

vessels, piping and piping components, and reactor pressure vessel internals. Better mechanistic understanding of crack initiation and propagation processes observed in pressurized

water reactors. Technical and analytical options for resolving existing and emerging materials performance, safety, and

reliability issues. Standardized guidelines for monitoring and managing degradation of plant components. Dissemination of research results to inform the regulatory process.

p. 1


MRP closely collaborates with other EPRI programs, including Steam Generator Management, Nondestructive Evaluation, and Chemistry, to ensure appropriate technologies and technical guidance are effectively integrated into research activities.

Accomplishments The Electric Power Research Institute's (EPRI’s) Materials Reliability Program supports nuclear power industry efforts to assess and implement countermeasures for degradation mechanisms impacting pressure boundary materials in pressurized water reactors. MRP research provides utilities and regulatory agencies with the information necessary to make technically sound and cost-effective decisions for managing degradation. Developed probability of detection curves to support continued use of leak-before-break assessments for

components containing dissimilar metal welds. The curves show with high confidence that inspection procedures are reliable and support leak-before-break principles.

Quantified the benefits of zinc addition and hydrogen optimization to mitigate primary water stress corrosion cracking initiation and growth. Such quantification is intended to provide the technical basis to support modifications to inspection intervals.

Developed pragmatic, technically defensible inspection and evaluation guidance for reactor internals. Assessed the usefulness of the guidance through pilot tests at three plants; test results were subsequently used to refine the guidelines.

Developed a predictive model (DISFRAC) for fracture toughness of ferritic steels in the transition temperature region. Recent work added a crack propagation model to more accurately model temperature effects.

Developed generic safety- and reliability-driven strategies for degraded materials management for Alloy 600 components, reactor vessel internals, reactor pressure vessels, and piping degradation due to thermal and environmental fatigue.

Current Year Activities Materials Reliability Program research and development for 2011 will focus on reactor internals degradation, fatigue susceptibility and Alloy 600 management to inform regulations. MRP also will develop data needed to revise materials management guidelines by conducting testing programs related to boric acid corrosion, Alloy 690, and nondestructive evaluation (NDE) techniques. Specific efforts will include the following: Investigate reactor internals degradation management through materials modeling, inspection method

development and demonstration, and continued testing of irradiated materials Evaluate crack growth rates in pressurized water reactor environments optimized for primary water stress

corrosion cracking mitigation Assess Alloy 690/52/152 resistance to primary water stress corrosion cracking (includes crack growth

rates) Revise Alloy 600/82/182 guidelines as needed based on industry experience Analyze welding residual stresses in Alloy 600 materials Support plant demonstrations of mechanical mitigation through peening Define and conduct necessary research to modify Generic Design Criteria No. 4 associated with “leak

before break" calculation for Alloy 600 materials Selected reports have been developed in whole or in part under Title 10 of the Code of Federal Regulations Part 50 (10CFR50) Appendix B Quality Assurance and 10CFR21 and the EPRI Quality Assurance Program. Additional products may be developed under 10CFR50 Appendix B and 10CFR21 at the discretion of the Pressurized Water Reactor Materials Reliability Program (PWRMRP) member utilities or EPRI MRP, when such action is deemed appropriate.

Pressurized Water Reactor Materials Reliability Program (QA) - Program 41.01.04 p. 2



Program Manager Christine King, 650-855-2164, [email protected]

Summary of Projects


P41.01.04.01 PWR Reactor Vessel Integrity (base) (QA)

The project works with industry stakeholders to develop improved understanding of reactor pressure vessel performance as a result of aging due to neutron irradiation. The improved understanding provides operating flexibility in plant heat-up and cool-down evolutions.

P41.01.04.02 Fatigue Management in PWR Reactor Coolant System Components (QA) (base)

This project develops the tools and guidelines required for PWR utilities to effectively manage environmental and thermal fatigue issues in PWR systems and components.

P41.01.04.03 PWR Alloy 600/82/182 Materials Degradation Management (supplemental)

This project focuses on improved understanding of primary water stress corrosion cracking in PWRs in order to develop standard protocols for effectively managing Alloy 600/82/182 degradation. These protocols transform managing Alloy 600 degradation from a crisis situation to a well-planned evolution.

P41.01.04.08 PWR Pipe Rupture Probability Reassessment (xLPR) (supplemental)

Because the analytical basis for leak-before-break is now considered too limiting, this project develops new analytical tools and methods to define a more robust yet flexible technical basis for leak-before-break. Collaborative activities between NRC Research and MRP support this project.

P41.01.04.10 PWR Reactor Internals Aging Management (supplemental) (QA)

This project integrates information and insights from various sources to develop inspection and evaluation guidelines for PWR internals. Information sources include the irradiated materials behavior database, functionality/safety analysis results, inspection strategies, flaw evaluation methods and criteria, plant design information and design bases, and plant operation data and experience.



PWR Reactor Vessel Integrity (base) (QA) (052342)

Key Research Question The reactor pressure vessel (RPV) is arguably the most critical safety-related component in the primary pressure boundary of a nuclear power plant. Repair/replacement of the RPV is not practical, yet its mechanical integrity must be conservatively demonstrated for up to 80 years of operation if the electric power industry is to realize continued operation of the nuclear fleet. During operation, the RPV is exposed to neutron radiation that ultimately reduces its mechanical properties. To maintain operation within established limits, the structural integrity of the RPV must be conservatively demonstrated under a series of normal operational conditions such as heat-up and cool down and against specific, more severe postulated scenarios such as pressurized thermal shock. Early analysis methods established limits that have been shown to be overly conservative. More accurate methodologies are needed to analyze vessel integrity for neutron attenuation through the vessel wall, evaluate the effect of irradiation on forged nozzles, and develop models for predicting fracture toughness shifts as vessels operate beyond their original design lives.

Approach MRP will work with NRC Research, national laboratories (for example, Oak Ridge National Laboratory), the appropriate ASME Code Committees, and other stakeholders to develop strategies for ensuring RPV integrity. This includes the following: Complete Code revision for risk informing ASME Section XI, Appendix G Develop and validate a fracture toughness prediction model that uses research based on studies of

neutron attenuation through vessel walls Evaluate impact of embrittlement correlations proposed by NRC and ASTM Coordinate a vessel capsule pull program that will obtain high fluence data from power reactors

applicable to 80 years of operation Evaluate the long-term effects of irradiation on vessel components outside of the core region such as

nozzles to ensure continued safe operation

Impact Potential benefits from this project include the following: Improved asset management. The research would eliminate or minimize economic impact of vessel

annealing or neutron flux reduction schemes. Reduction of regulatory uncertainty (as related to 10CFR50.61 and Appendix G requirements). Further

research can provide technical basis for improving the accuracy of the embrittlement correlation and fracture toughness shifts.

How to Apply Results The products from the vessel integrity research provide members with supportable (and NRC-accepted) data that allow plant heat-up and cool-down operating flexibility. This will include the following: ASME Section XI Appendix G revision to incorporate findings from studies that offer justification for risk

informing heat up and cool down New and updated industry guidelines that reliably address concerns for toughness shifts, neutron

attenuation, and coordinated surveillance capsule pulls with research-backed technical bases



Fatigue Management in PWR Reactor Coolant System Components (QA) (base) (065370)

Key Research Question PWR reactor coolant system components are susceptible to both environmental and thermal fatigue. Because the effects of the reactor coolant environment are not fully understood, many plants have had to implement fatigue-monitoring programs for their initial license renewal. The prospect of additional license renewal terms (80-year service lives) presents a need for more accurate characterization of the effects of high-temperature coolant on component fatigue life, including design rules and inspection requirements. In addition, although industry guidance is in place to address high-cycle thermal fatigue degradation in reactor coolant system components due to cyclic stratification induced by swirl penetration, the guidance needs to be evaluated to assure it adequately addresses this concern for all reactor coolant system components. This includes reactor pressure vessel safety injection and core flood line locations, as well as for socket welds and cast austenitic stainless steel.

Approach This project includes several tasks related to environmental and thermal fatigue management: Environmental Fatigue Issues Develop code case(s) to incorporate findings from studies that offer justification for reduced stress indices

into ASME code Develop a program plan to address environmental fatigue in new plants and existing plants (license

renewal, configuration changes) by use of an “expert panel” of NSSS personnel, laboratory researchers, and ASME representatives

Obtain regulatory concurrence/approval by incorporation of research-backed justifications into the ASME code for existing and new plants

Thermal Fatigue Issues Provide/continue training on existing MRP guidance, and document implementation survey results to

determine future research needs Revise MRP inspection guidance to account for new ASME inspection procedures for smaller lines and

evaluate current NDE technologies for their effectiveness in detecting thermal fatigue damage in piping Develop flaw tolerance evaluation tools and evaluate the effect of weld overlays in support of MRP

specified analyses Develop inspection technologies for CASS piping and guidance for inspection of socket welded

configurations where thermal cycling is predicted Complete a scoping assessment of the potential for high-cycle thermal fatigue in RPV-connected safety

injection and core flood lines

Impact Potential benefits from this project include the following: Improved asset management. The research would eliminate or minimize economic impact of fatigue-

degradation-related operating events. Reduced need for unnecessary inspections related to thermal fatigue. Further research can provide

technical basis for eliminating some of the currently mandated inspections or decreasing the inspection frequency.

Optimum environmental fatigue management. Development of an optimum approach to address effect of environmental fatigue would be of great economic value to a second license renewal and to new plants.



How to Apply Results Members will use MRP-developed tools and guidelines to manage and address environmental and thermal fatigue issues. MRP staff will provide training and support as needed to assist members in this effort.

PWR Alloy 600/82/182 Materials Degradation Management (supplemental) (061346)

Key Research Question Alloy 600 and its weld metal formulations (A82, A132, and A182) have been used extensively in PWR reactor coolant system applications. Since about 2000, Alloy 600 degradation has demanded the attention of the PWR fleet primarily due to the leaks in reactor pressure vessel heads and leak and flaw indications in Alloy 82/182 butt weld locations. Within the standard operating water chemistry of the PWR fleet, Alloy 600 and its weld metals are susceptible to primary water stress corrosion cracking (PWSCC). Comprehensive Alloy 600 management to address PWSCC involves a complex variety of actions and activities depending on specific attributes of the location in question, ranging from inspection and repair to pre-emptive mitigation strategies and outright replacement. Guidelines, inspection techniques, and mitigation measures are needed for the fleet to manage materials degradation.

Approach This project develops pragmatic, technically defensible guidance for inspecting, mitigating, and managing Alloy 600 and dissimilar metal butt welds across the pressurized water reactor fleet to ensure safe operation and a low probability of safety-significant leakage. By reviewing the latest field results and comparing them to original assumptions in technical basis for the guidelines, the MRP identifies the best practices for the fleet necessary to manage the Alloy 600 issues. When needed, more detailed analyses can be performed to justify alternate inspection/mitigation strategies. In addition, the MRP monitors regulatory inquiries to ensure consistent fleet implementation of MRP guidelines, which is a necessary component of successful self-regulation.

Impact Potential benefits from this project include the following: Allows for self-regulation through MRP guidelines Prevents degradation of Alloy 82/182 butt welds through mitigation rather than inspection Clarifies inspection requirements for Alloy 82/182 butt welds that have not had mitigation measures

applied

How to Apply Results Member nuclear power plants will adapt generic degradation management guidelines in creating plant-specific programs addressing flaw evaluation procedures, inspection standards, acceptance criteria, and mitigation measures.



PWR Pipe Rupture Probability Reassessment (xLPR) (supplemental) (064687)

Key Research Question Many pressurized water reactor (PWR) plants are licensed for “leak-before-break” (LBB) in various locations within the reactor coolant system piping. In the original LBB regulatory construct, certain limits were placed on its applicability, including “no active degradation mechanisms or repairs.” However, the set of approved LBB lines includes weld locations now known to be susceptible to primary water stress corrosion cracking and subject to weld overlay as a mitigative and repair activity. The analytical base upon which LBB was developed is now considered too limiting, and methods are needed to define a more robust yet flexible technical basis for LBB.

Approach A collaborative effort between NRC Research and the Materials Reliability Program (MRP) is addressing the LBB issue. MRP will be responsible for several tasks related to the new calculation, including dissimilar metal weld residual stress measurements, mapping the extremely low probability of rupture (xLPR) calculation, and implementing the xLPR calculation plans. Detailed tasks and overall schedule are developed through joint meetings and conference calls.

Impact This research is modeled after the industry's approach to revising the Pressurized Thermal Shock rules. As such, the calculation cannot be redesigned by one organization and will take a large collaborative effort to develop the final tool. In the end, the industry will have a tool for handling the active degradation of a LBB location.

How to Apply Results Member utilities will use the tool to re-analyze their LBB locations and address active degradation mechanisms if necessary.

PWR Reactor Internals Aging Management (supplemental) (QA) (065838)

Key Research Question Pressurized water reactor internals structurally support the core, the control rod assemblies, the thermal and neutron instrumentation, and the reactor pressure vessel surveillance capsules. The reactor internals also maintain a distributed flow of water through the core and to certain bypass flow paths for cooling purposes. As reactor owners pursue license renewal or life extension, they must pay close attention to aging and degradation of reactor internals and must demonstrate an effective aging management inspection and evaluation program.

Approach To support the implementation and execution of effective aging management programs, MRP develops inspection and evaluation guidelines for PWRs. These guidelines are developed and updated by integrating information and insights from the irradiated materials behavior database, functionality/safety analysis results, inspection strategies, flaw evaluation methods and criteria, plant design information and design basis, and plant operation data and experience. To promote awareness and consistent implementation of the guidelines, MRP also develops program templates for plant use and convenes workshops to discuss industry best practices and share lessons learned.




Impact Potential benefits from this project include the following: Self-regulation through MRP guidelines Management of aging/degradation effects in reactor internals for nuclear plants considering life extension

or license renewal

How to Apply Results Each member utility will take generic degradation management guidelines and create a plant-specific program that addresses flaw evaluation, inspection standards, and acceptance criteria.

Welding & Repair Technology Center

Program Overview

Program Description

High-quality, reliable welds are critical to safe nuclear plant operation. Because of the safety significance of welds in many metal components and plant systems, the nuclear power industry must be confident in the quality and integrity of welds when joining various metals and using different welding systems. While maintaining this commitment to quality and safety, nuclear power plants also are interested in productivity improvements and new technology applications that can provide cost savings in terms of maintenance and operations. The Welding & Repair Technology Center (WRTC) develops advanced materials, joining, and repair technologies for nuclear power plants, contributing to reduced operation and maintenance costs and improved plant availability. The program also supports technical interactions with code and regulatory entities to reduce the time and cost associated with implementing new technologies and repair rules.

Research Value

Research results from the Welding & Repair Technology Center help nuclear power plants find faster, less costly ways of making repairs using novel welding techniques or by applying existing techniques in new situations. Research results also are commonly used to support technical justifications that enable utilities to pursue regulatory relief from code requirements, typically saving time and money. WRTC participants gain access to the following: Repair options for key components, supplemented by application guidelines, procedures, and training Materials, welding, and repair experts across the Electric Power Research Institute (EPRI) and the

nuclear industry Direct support during implementation of critical plant repair applications involving material interactions,

weld process control, and code requirements Techniques for reducing repair costs, reducing component downtime, and increasing plant availability Demonstrated repair techniques and technologies that improve material performance and enable

component life extension

Approach

The Welding & Repair Technology Center combines extensive laboratory capabilities with detailed familiarity with industry and regulatory needs to investigate and evaluate welding and repair techniques. WRTC staff can replicate welding setups in the field—power supplies, weld heads, and other equipment—to create realistic welding environments in the laboratory. Through participation in many American Society of Mechanical Engineers (ASME) and industry technical committees, WRTC staff can then ensure that the program’s research results can support code requirements. The WRTC has the ability to accomplish the following: Develop, test, and evaluate safe, effective, and reliable repair techniques that contribute to shorter

outages and meet or exceed all code, regulatory, and design requirements Ensure broad dissemination and application of industry lessons learned and benchmarking practices Provide information exchange and peer review of current and past repair applications Develop and assess technologies that enable repairs to be made more quickly while the plant is off-line Develop and assess technologies that enable repairs to be made while the plant is on-line

Accomplishments

EPRI’s Welding & Repair Technology Center supports nuclear power industry efforts to develop and apply welding and repair techniques that ease regulatory concerns, reduce maintenance costs, and improve productivity. WRTC activities have specifically accomplished the following:

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Provided technical guidance for conducting detailed root-cause analysis to support repair option selection

for critical pressure components. Evaluated nickel high chromium weld filler materials for resistance to typical welding related defects, such

as direct digital control (DDC), hot cracking, and general weldability. Alloy 52 type materials and derivatives have been evaluated to support alloy selection, and grading to distinguish heat-to-heat variations.

Demonstrated the feasibility of in-vessel underwater laser beam weld repair of critical nickel alloy welds, eliminating the need to drain the reactor vessel. Areas that have been addressed include seal welding capabilities, temper bead welding, hot cracking susceptibility.

Developed a roadmap to assist power plant personnel in conducting failure analyses for various components or for directing other organizations responsible for such work, with a focus on metallurgical and mechanical aspects. The report provides a roadmap of the sequential steps for the investigation and a guide to the laboratory equipment used.

Provided technical support for implementing new technologies, including application guides for advanced welding methods, guidelines for installing and examining dissimilar metal weld overlays, and repair/mitigation of socket weld fatigue failures.

Supported development of realistic code rules, including new code cases to reduce post-weld examination hold times and use dissimilar metal weld overlays for stress corrosion cracking mitigation.

Developed guidance for overlay applications based on lessons learned, best practices, and weld studies to support current technology. Also supported development of new and higher production welding processes such as gas-metal arc welding and dual wire feed gas tungsten arc welding.

Conducted testing to determine the way in which concrete and reinforcing steel are affected by exposure to boric acid in concentrations typical of spent fuel pool chemistry. Improved understanding of the degradation mechanisms and degradation rates will support life extension decisions.

Evaluated the application of new repair techniques for high-density polyethylene piping, which is gaining traction as an alternative to steel in low-energy applications.

Current Year Activities

Welding & Repair Technology Center research and development for 2011 will focus on developing repair and fabrication technologies to reduce outage time and expand the availability of repair options that may be performed during plant operation. WRTC also will provide technical support to address the challenges of new plant construction. Specific efforts will include the following: Conduct failure analysis and stress measurements to assist utilities in repair decisions that are cost

effective, reduce downtime, and improve quality Evaluate advanced filler/welding materials (Alloy 52M) for critical plant repair applications Develop training for new repair and replacement engineers Evaluate welding methods for small bore piping and alternative joining methods for socket welded joints Provide benchmarking support for utility repair and replacement programs Identify repair/mitigation options that address buried piping issues, fuel pool leakage and components

susceptible to stress corrosion cracking

Estimated 2011 Program Funding

$3.3 million

Program Manager

Gregory Frederick, 704-595-2571, [email protected]

Welding & Repair Technology Center - Program 41.01.05 p. 2


Summary of Projects


P41.01.05.01 WRTC Information Exchange and Welding Conferences (base)

This project supports member access to Welding and Repair Technology Center expertise through meetings and information products. Members also may receive case history information for unique repair/replace applications, including welding qualification (PQR) database support, procedure review, repair/welding program assessments, benchmarking activities, research and development results, and specific code support.

P41.01.05.02 WRTC Codes & Standards (base) (QA)

This project supports Welding and Repair Technology Center activities in the development of ASME (American Society of Mechanical Engineers) code cases, revisions, and technical interpretations to address repairs of a wide range of components. The project also engages other code organizations, as needed, to expand the availability of repair options.

P41.01.05.04 WRTC Subscriber Requested Assistance (supplemental)

Subscriber Requested Assistance provides members with technical support in a broad range of repair-related areas, including materials and joining evaluations, benchmarking of welding and repair programs, and process and procedure development. Members determine the workscope of the individual projects.

P41.01.05.05 WRTC Advanced Weld Application (supplemental)

This project provides Weld Repair and Technology Center (WRTC) members with advanced welding and repair technology to reduce the time and cost of repairs or modifications to critical plant components.

P41.01.05.06 WRTC Benchmarking and Guidelines (supplemental)

This project assembles valuable case history and lessons-learned data and distributes it to WRTC members to assist in the implementation of repair and replacement activities.

P41.01.05.08b Repair and Replacement Options for High-Cycle Fatigue Failures in Socket Welds (QA)

This project supports the revision of guidelines for repair and replacement of socket welds susceptible to high-cycle fatigue failures. This includes code-approved technologies to enable on-line repair of defective socket welds in operating nuclear plants and validation of modified weld joint geometries.

P41.01.05.c Weld Mitigation Interest Group (supplemental)

The Weld Mitigation Interest Group provides a forum for exchanging industry experience, solutions, and technology. The group evaluates emerging repair options that could reduce the time, cost, and radiation exposure related to welding operations; provide enhanced inspectability; expand repair options to include inlays, onlays, and underwater laser beam welding; validate filler metal weldability; and document industry experience.



WRTC Information Exchange and Welding Conferences (base) (065821)

Key Research Question

Nuclear plant staffs have been reduced, reducing the time available for personnel to participate in technology transfer activities, stay abreast of changing codes and regulations, and monitor improved repair technologies and processes. Accelerated technology transfer would help plant personnel meet immediate needs and provide peer support among repair personnel.

Approach

This project supports member access to Welding and Repair Technology Center (WRTC) expertise through meetings and information products. Members also may receive case history information for unique repair/replace applications, including welding qualification (PQR) database support, procedure review, repair/welding program assessments, benchmarking activities, research and development (R&D) results, and specific code support.

Impact

Potential benefits include the following: Immediate access to repair experts Peer support from the repair community through focused workshops Access to welding and repair procedures to meet new challenges Review/evaluation of new repair and welding procedures Support on repair issues and implementation of WRTC technology

How to Apply Results

Project results directly support utility member needs for ready access to repair and welding information. Products are aimed at supporting technology transfer of R&D results to WRTC subscribing members.

WRTC Codes & Standards (base) (QA) (065822)


The continued downsizing of plant staffs and increasing travel restrictions make it difficult for nuclear power plants to maintain the required level of contact with code organizations and other technical bodies. Such engagement is necessary to ensure technical revisions permit the implementation of new repair methods, welding procedures, and weld materials within acceptable safety limits. Realistic code rules for welding and repair and the availability of technical support can assist nuclear plants in pursuing regulatory relief and obtaining regulatory approval of new code rules.

Approach

This project supports Welding and Repair Technology Center (WRTC) activities in the development of ASME (American Society of Mechanical Engineers) code cases, revisions, and technical interpretations to address repairs of a wide range of components. The project also engages other code organizations, as needed, to expand the availability of repair options. As part of this work, WRTC provides technical information to appropriate regulatory bodies in support of new and emerging technologies. Examples of this work include boiling water reactor control rod drive leakage repair, pre-emptive dissimilar metal weld overlay to address Alloy 600 repair applications, and use of specialized methods to seal leakage while under power. Activities also support code updates and revisions to American Welding Society (AWS) structural and pipe/tubing codes, National Board Inspection Code (NBIC), and international codes.



Impact

Potential benefits include the following: Provide technical basis supporting code and regulatory rules that permit the use of innovative repair

techniques, materials, and technology for critical reactor components Reduce the cost and complexity of repairs through the availability of practical code rules Reduce reliance on highly skilled staff in the application of more complex requirements Engage code and regulatory bodies during the development of new technologies to reduce the time and

cost for implementing new technologies and repair rules


The results of this project will be implemented through adoption of new code rules and by modifications in regulations related to repair activities.

WRTC Subscriber Requested Assistance (supplemental) (045710)


Nuclear plant staffs have been reduced, reducing the time available for personnel to participate in technology transfer activities, stay abreast of changing codes and regulations, and monitor improved repair technologies and processes. Accelerated technology transfer would help plant personnel meet immediate needs and provide peer support among utility repair personnel.

Approach

Subscriber Requested Assistance facilitates technical support for application of Welding Repair and Technology Center (WRTC)-developed technology products. Through this feature, members also may receive assistance for unique repair/replace application needs, including welding qualification database support, procedure review, repair/welding program assessments, benchmarking activities, and specific code support.

Impact

Immediate access to EPRI repair experts Peer support from the larger nuclear repair community Access to welding and repair procedures to meet new challenges Support in applying new repair and welding procedures and implementing WRTC technology


Subscriber Requested Assistance projects directly support utility member needs. Each product is tailored to meet the request of a participating utility or utility group.

WRTC Advanced Weld Application (supplemental) (065815)


Advanced repair technologies and processes can improve plant availability while improving the quality/reliability of repairs. New materials and increasing challenges to maintain existing plant components have increased the value of welding as a repair option. Repair and replacement of existing pressure systems, for example, relies heavily on welding technology. Development and demonstration of advanced welding processes is needed to evaluate limitations, identify the range of applications, and ensure weld performance meets quality requirements.

Approach

This project develops advanced repair equipment and technologies to meet emerging issues in power plant repairs. Recent examples include cleaning of contaminated surfaces, fuel pool repairs, and use of weld



overlays for repair of Alloy 600 components. This project also develops and evaluates advanced equipment and processes such as laser cleaning systems and wave-form controlled gas metal-arc welding systems for underwater or overlay welding applications. Other studies will include use of these advanced repair systems to support improvement of special methods such as temperbead welding to further reduce critical path time for repairs

Impact

Potential benefits include the following: Reduce the time and cost of performing repairs by improving the deposition rate of welds for

repair/replacement applications Extend the life of repaired components through advanced technology for repairs Reduce the cost of post-repair examinations and future monitoring through improved repair quality Develop specialized processes such as temperbead welding to reduce repair cost and critical path time

and provide alternatives to repair techniques based on original construction code rules


Members implement repair techniques and products on-site using their own staff or by sharing the methods with vendor organizations. A number of the developments from this project will be approved by code and regulatory agencies, further assisting in implementation.

WRTC Benchmarking and Guidelines (supplemental) (065816)


The development of repair and replacement technology for power plants is an evolutionary process that requires continuous sharing of experience from nuclear plants, vendors, and research organizations. These experiences and case histories are often passed along from outage to outage within a utility or a specific vendor. Sharing this information across multiple plants informs decisions related to repair and replacement activities.

Approach

The Weld Repair and Technology Center (WRTC) supports upcoming repair and replacement activities by compiling and sharing best practices, experience information, and benchmarking data. This project continues those efforts by using a number of resources to acquire and distribute this information to members.

Impact

Realize substantial time and cost savings through implementation of key products such as Welding Best Practices and the Repair & Replacement Program Checklist. In many cases, these documents also assist in ensuring compliance with the latest code or regulatory requirements.

Access to the WRTC Sharepoint site, which provides continuously updated information on utility procedures, administration documents, repair and replacement programs, welding programs, experience reports.

Provide peer support and benchmarking activities through information exchange services, SharePoint website, and issues meetings.


This project is implemented by WRTC members through email and web-based communication and document-sharing tools. These tools provide quick access to case history information, technical reports, and procedures/practices used by other members. Members realize value by integrating such information into plant procedures or practices.



Repair and Replacement Options for High-Cycle Fatigue Failures in Socket Welds (QA) (049937)


Failures of small-bore piping connections (2-inch and smaller) continue to occur at nuclear power plants, resulting in degraded plant systems and unscheduled plant downtime. Fatigue-related failures are generally detected as small cracks or leaks, but in many cases the leak locations are not isolable from the primary reactor coolant system and result in extended outages. These failures are typically accelerated by poor weld quality at the root and toe locations (lack of fusion, undercut).

Approach

This project will support the revision of guidelines for repair and replacement of socket welds susceptible to high-cycle fatigue failures. This includes code-approved technologies to enable on-line repair of defective socket welds in operating nuclear plants and validation of modified weld joint geometries. The socket weld configurations will be evaluated through high-cycle fatigue testing, mockup testing, and finite element analyses. Consequently, an understanding of socket weld fatigue failures related to weld quality and design will be improved, and successful resolutions will be implemented to reduce costs associated with forced outages and repeat failures.

Impact

Potential benefits include the following: Improved weld geometry specifications that can enhance the reliability of socket welds through reduced

failures and extended life of susceptible socket weld locations Online repair applications that can extend operating time, allowing permanent replacement of the repaired

connection to be scheduled during a routine outage with little or no impact on power production Standardized test configuration and vibration fatigue test conditions that support American Society of

Mechanical Engineers (ASME) code actions, overlay repair design criteria, and fitting and weld geometry modifications


This project provides test data and guideline information for modification of existing socket welds or use of improved weld geometries for new installations. The guidelines also address online repair methods that can prevent outages or derating through the use of code-approved techniques.

Weld Mitigation Interest Group (supplemental)


New and emerging repair and mitigation technologies can provide safe, cost-effective options for nuclear plant owners. Evaluation of such technologies is necessary to ensure their application is safe, effective, and can meet code requirements. Candidate technologies may include the following: Underwater laser beam welding for inside diameter (ID) mitigation and repair High-deposition welding processes Inlay and only repair technologies for ID mitigation and repair Improved weldability of Alloy 52, a high-chromium filler material with high resistance to primary water

stress corrosion cracking (PWSCC) Improved welding methods to enhance inspectability, weld quality, and productivity




Approach

The Weld Mitigation Interest Group evaluates emerging repair options that could reduce the time, cost, and radiation exposure related to welding operations; provide enhanced inspectability; expand repair options to include inlays, onlays, and underwater laser beam welding; validate filler metal weldability; and document industry experience. The Weld Mitigation Interest Group provides a forum for exchanging industry experience, solutions, and technology. The Interest Group produces a Stress Corrosion Cracking Repair and Mitigation Handbook to support effective management of primary system components susceptible to stress corrosion cracking (SCC). Related products include evaluation of advanced welding technologies and a database of high-chromium filler metal weldability tests.

Impact

Potential benefits include the following: Availability of proven, cost-effective repair and mitigation techniques. Consistent approach for compiling operating experience and addressing regulatory issues. Identification of high-priority research activities to address weld overlay issues.


Participants use technology evaluation results to analyze repair options for future plant application. The Mitigation Handbook provides specific guidance that can be incorporated into management and inspection plans for primary systems components susceptible to SCC. Workshop participation provides access to industry experts on weld repair and mitigation.

Nondestructive Evaluation and Material Characterization

Develop and qualify nondestructive evaluation technologies that can be used to guide decisions on whether and when to replace, repair, or continue operation of components.

Program Nondestructive Evaluation Program

Nondestructive Evaluation Program (QA)

Program Overview

Program Description Nuclear plant components and materials can undergo substantial changes in a nuclear environment. Plant owners must remain attuned to these changes to ensure steps can be taken if necessary to avoid unsafe operations or minimize mitigation costs. Nondestructive evaluation (NDE) is an important tool for assessing component condition, but depends on the availability of qualified technologies, personnel, and procedures. The Nondestructive Evaluation Program develops technologies and procedures to quickly, accurately, and cost-effectively inspect and characterize nuclear component condition and guide strategic decisions on whether and when to replace, repair, or continue operation. Research results are used to inform regulatory actions related to pre-service and in-service inspections. The Program also supports industry efforts to expand and accelerate the supply of qualified NDE workers in the nuclear industry.

Research Value Research results from the Nondestructive Evaluation Program enable the accurate deployment of advanced inspection technologies to the nuclear power industry. Program activities also support the use of performance-based and risk-informed methodologies to improve inspection reliability. Collectively, these activities increase the accuracy of information used to assess material condition, lower operating costs, lower radiation exposure to workers, and assist plant owners in meeting regulatory commitments. NDE Program participants gain access to the following: NDE technologies, training, and regulatory/code support that can shorten plant outages, leading to

savings of $1 million or more per plant per day saved Technical justification for the use of NDE diagnostic capabilities during extended plant operation,

supporting regulatory approval of license renewal or extension Implementation support for inspection and evaluation guidelines promulgated through the Electric Power

Research Institute (EPRI) and industry materials analysis programs Qualification process for NDE personnel, procedures, and equipment in accordance with American

Society of Mechanical Engineers (ASME) Section XI, Appendix VIII Technical guidance supporting code changes that can lead to improved pre-service and in-service

inspections Strengthened NDE workforce through industry-focused training and central qualification resources

Approach The Nondestructive Evaluation Program develops and demonstrates cost-effective and reliable inspection methods and analysis programs that can be integrated with structural and lifetime evaluations of power plant components and systems. NDE results contribute to more accurate characterization of nuclear component condition and help inform decisions on whether to replace, repair, or continue operation. Develop new and improved NDE hardware, software, databases, methods, and delivery of NDE products

to support nuclear plant inspection programs Provide NDE qualification programs addressing personnel, procedures, and equipment that enable

nuclear plants to comply with regulatory and industry requirements Integrate NDE research and development with risk-informed technology and human performance

research Provide technical basis to guide regulatory and code activities related to pre-service and in-service

inspections

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Accomplishments EPRI’s Nondestructive Evaluation Program supports nuclear industry efforts to address emergent material- and inspection-related issues through innovative NDE technologies, technical guidance, and qualification support. Published guidance for planning and executing efficient, effective examinations of dissimilar metal welds.

Guidance was published publicly so that NDE vendors and utilities could collaborate to achieve greater success.

Developed a buried piping NDE test facility. The test facility’s first installation is a 220-foot run of 24-inch piping containing several elbows and controlled defects. It will be used both for technology development and for NDE vendor capability benchmarking.

Completed the first qualification for ultrasonic examinations of pressurized water reactor (PWR) vessel head penetrations. Although the qualification process was difficult, it achieved its main objectives before the aggressive regulatory deadline.

Successfully pilot-tested a risk-informed repair/replacement activity methodology (RI-RRA) and received regulatory approval from its use. The methodology allows plants to replace low safety-significant components with commercial components rather than N-stamp components. Because ASME Section XI requirements would no longer apply, cost reductions of 3-10X are expected.

Evaluated the use of guided wave ultrasonics and alternating current field measurements for inspecting spent fuel pool liner welds. If undetected, leaks from these welds could lead to contamination of groundwater and surrounding soils.

Developed qualified phased array ultrasonic procedures for fast, high-coverage examinations of reactor pressure vessel welds, austenitic and ferritic piping welds, dissimilar metal welds, and weld overlays. Phased array technology can examine a complex weld overlay configuration using only two scans, while a conventional approach would require as many as 20 scans.

Developed 3D laser imaging and mathematical modeling tools for examining reactor vessel nozzles. Published initial guidance for NDE detection and measurement of gas accumulation in piping. Developed guidance for planning and executing effective examinations of dissimilar metal welds.

Current Year Activities Nondestructive Evaluation Program research and development for 2011 will address a number of nuclear component materials issues where effective NDE is essential. Specific efforts will include the following: Identify and characterize best practices for examining buried piping Maintain qualification program for NDE of reactor pressure vessel head penetrations Convert risk-informed programs to applicable engineering code standards

Selected NDE program activities are conducted in whole or in part in accordance with Title 10, Code of Federal Regulations, Part 50, Domestic Licensing of Production and Utilization Facilities (10CFR50), Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, and may invoke Part 21, Reporting of Defects and Noncompliance (10CFR21). Additional NDE program activities may be conducted in accordance with 10CFR50 Appendix B and 10CFR21 at the discretion of the Nondestructive Evaluation Center, member utilities or EPRI, when such action is deemed appropriate.


Program Manager Greg Selby, 704-595-2595, [email protected]

Nondestructive Evaluation Program (QA) - Program 41.04.01 p. 2


Summary of Projects


P41.04.01.01 NDE Research and Development (base) (QA)

The base program conducts collaborative research and development of nondestructive evaluation technology (NDE) to address nuclear plant materials issues and characterize component degradation. Collaborative research and development of nondestructive evaluation technology to address nuclear plant materials issues and characterize component degradation.

P41.04.01.02 NDE Applications and Technology (QA) (supplemental)

The Nondestructive Evaluation Applications and Technology Transfer project develops products that support the demonstration, qualification, and technology transfer of NDE technology to field-usable products and services.

P41.04.01.03 NDE PDI Training and Qualification (QA) (supplemental)

This project provides general operation and maintenance support of the Performance Demonstration Initiative (PDI) program. Through the PDI program, utilities and their service providers, NDE personnel, NDE procedures, and equipment are qualified in accordance with ASME Section XI, Appendix VIII.

NDE Research and Development (base) (QA) (068030)

Key Research Question The existing nuclear fleet has a proven track record of operating safely and effectively. Extending this track record will depend on continued, reliable insight into the condition of materials to assess and address aging and degradation. Nondestructive evaluation (NDE) plays a vital role in managing material aging issues. In recognition of this increasing role, EPRI conducts NDE research and development to develop NDE solutions that assist the nuclear power industry in safely operating and maintaining existing nuclear assets.

Approach The Nondestructive Evaluation (NDE) research and development program focuses on developing more efficient and more accurate NDE devices and techniques. Staying ahead of materials issues requires sustained R&D. The program addresses NDE of all major plant components, including piping, vessels, balance-of-plant, and reactor internals. Emerging materials issues also are being tackled, including buried pipe, cast stainless steel, civil structures, and spent fuel. A wide range of NDE technologies are being evaluated, including ultrasonic guided-wave, low-frequency phased arrays, and laser-based technologies. R&D activities employing the latest NDE techniques will address a variety of challenging materials inspection issues: Buried components Cast stainless steel Civil structures and materials Reactor vessel internal components Spent fuel pool and transfer canal liners

Impact Inspection strategies targeting existing and emerging degradation issues New inspection technologies for buried pipe, concrete, and cast stainless steel NDE approaches that facilitate life extension



How to Apply Results The results of the NDE Program are primarily delivered as technical reports or guideline documents that are used by utilities and their inspection vendors to develop inspection strategies. Additionally, the results of the NDE Research & Development Program are often used as the technical basis for generic inspection strategies and procedures.

NDE Applications and Technology (QA) (supplemental) (068025)

Key Research Question Nondestructive evaluation (NDE) capabilities address a number of nuclear power industry needs, including pre-service and periodic in-service inspection of components to satisfy regulatory requirements, inspection to characterize component condition, and inspection to guide strategic decisions on whether and when to replace, repair, or continue operation of components. The nuclear power industry must remain cognizant and receptive to the introduction of new technology while ensuring confidence in the overall reliability of the NDE processes employed. Often, the industry requires technical assistance to support the implementation and effective application of advanced NDE technologies.

Approach The NDE Applications and Technology program provides the nuclear power industry with a collaborative "ready now" NDE resource focused on today's operations and emerging issues. A team of more than 40 EPRI NDE technical staff target nuclear power applications and enable NDE technology transfer through the following activities: Inspection strategies and application plans using the results from the EPRI NDE research and

development program Regulatory and code support for pre-service and in-service inspections Evaluation of NDE technology to address aging plant and equipment reliability issues Resources to facilitate the supply of qualified NDE workers in the nuclear industry Supporting member utilities with independent assessments of NDE technology and results

Impact NDE Applications and Technology Transfer program participants benefit from the development and application of NDE hardware, software, databases, and methods. These products include technical services support to participating nuclear units, as well as technical support to other EPRI Nuclear programs, including the Performance Demonstration Initiative (PDI), materials programs (Materials Reliability Program [MRP], Boiling Water Reactor Vessel and Internals Project [BWRVIP], Steam Generator Management Program [SGMP]), Low-Level Waste, Balance of Plant Corrosion, Advanced Nuclear Technology, and Fuel Reliability.

How to Apply Results The results of the NDE Applications and Technology Transfer program are provided through EPRI technical reports and guidelines that member utilities directly apply. In some cases, program participants may elect to apply the results via service organizations that provide in-service inspection and other related inspection services. Participants also apply the results of the program through collaborative resources such as NDE mock-ups and generic procedures. Indirect methods of application also are used, including technical changes to inspection codes, NDE training materials for participants and their suppliers, and the release of good practice documents targeting increased public awareness and confidence in the reliability of NDE.




NDE PDI Training and Qualification (QA) (supplemental) (061738)

Key Research Question Revisions to 10CFR50.55(a), initially published September 22, 1999, mandate the implementation of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI, Division 1, Appendix VIII, Performance Demonstration for Ultrasonic Examination Systems. Appendix VIII requires qualification of the procedures, personnel, and equipment used to detect and size flaws in certain piping, bolting, and reactor pressure vessel components.

Approach This program includes the general operation and maintenance of the Performance Demonstration Initiative (PDI) program. This includes quality assurance (QA) activities, document control, and database maintenance.

Impact The PDI program is administered by EPRI to address the requirements of the ASME Boiler and Pressure Vessel Code, Section XI, Division 1, Appendix VIII, in a collaborative, efficient, cost-effective, and technically sound manner.

How to Apply Results Results of the PDI program may be applied by member utilities and organizations that provide in-service inspection services as defined in the PDI use agreement. Additional information on the application of the results is available at www.epriq.com.

2011 Products



PDI Program 2011: For 2011 and the foreseeable future, maintaining the Performance Demonstration Initiative (PDI) Program as a collaborative program is expected to be the most efficient method for utilities to comply with the mandatory Performance Demonstration requirements within American Society of Mechanical Engineers (ASME) Section XI, Appendix VIII. The PDI Program includes the management, oversight, and implementation per the PDI quality program. This includes quality assurance (QA) activities, document control, and database maintenance. Through the PDI Program, utilities and their service providers, NDE personnel, NDE procedures, and equipment are qualified in accordance with ASME Section XI, Appendix VIII. The PDI Technical Advisory Committee provides oversight of PDI program operations and via the NDE Program integration Committee recommends policy and procedures to the NDE Action Plan Committee. The costs associated with conducting performance demonstrations are not included in this project; these are recovered from user fees.

12/31/11 Technical Resource


Develop risk and safety assessment tools and techniques that enable nuclear plant owners to make technically sound design, maintenance and operational decisions.

Program Risk and Safety Management


Program Overview

Program Description Risk and safety assessment techniques provide information that enable nuclear plant owners to make technically sound design, maintenance, and operational decisions, contributing to safer and more cost-effective plant operation. Risk models and techniques are being used in a growing number of applications, including online maintenance, surveillance interval extensions, flexible allowed outage times, and risk-informed performance-based fire protection. Continuous refinement of risk and safety models and application approaches is necessary to ensure decisions based on these models reflect industry operating experience and state-of-the-art computational advances. The Risk and Safety Management Program develops risk assessment tools to enhance the safety and improve the economics of existing and future nuclear power plants. Electric Power Research Institute (EPRI) risk and safety software codes assist utilities in performing detailed analyses to quantify the level of risk, contributing to increased plant safety, more efficient and flexible plant operations and maintenance, and reduced electricity production costs.

Research Value Risk-informed performance-based approaches provide a “win-win” for the regulator, the nuclear plant owner, and the public. The regulator can focus on issues truly important to safety that benefit the public, while the plant owner gains operational flexibility and an opportunity for cost reduction. Since 1992, overall industry core damage frequency has dropped by a factor of five. This improvement has been driven by risk-informed initiatives, continued plant and equipment performance improvements, and probabilistic risk assessment model improvements. Risk and Safety Management Program members gain access to the following: Research results and technical input that foster a risk-informed, performance-based regulatory

environment, including the significance determination process (SDP), the mitigating system performance index (MSPI), and internationally with periodic safety cases and license renewal

Tools and methodologies that increase plant safety and reduce plant and resource requirements Shortened outages, fewer unnecessary shutdowns, reduced inspections and testing, and less-intensive

treatment of low-safety significant safety-related equipment Robust, plant-specific framework for more focused and stable regulatory interaction

Approach The Risk and Safety Management Program conducts research to facilitate the development of a risk-informed framework that can provide both operational flexibility and safety benefits to nuclear power plants. Operational flexibility encompasses on-line maintenance, flexible testing, flexible technical specifications, and enforcement discretion. Safety benefits include tangible and measured risk reductions as well as intangible items such as improved safety focus. Refine probabilistic risk assessment (PRA) models to guide effective design, operation, and asset

management decisions for critical plant issues Provide technical analyses supporting continued regulatory acceptance of risk-informed activities Develop analytical and software tools for safety evaluations, configuration risk management, fault tree

analysis, and security assessments Support the development of a larger pool of trained nuclear risk professionals

p. 1


Accomplishments EPRI’s Risk and Safety Management Program supports industry efforts to ensure risk-informed approaches can be used in making operational, maintenance, and regulatory decisions impacting nuclear power plants. Developed the first in a series of computer-based training modules on the fundamentals of probabilistic

risk assessment and risk-informed regulation. These modules provide a convenient and easy-to-understand mechanism for conveying risk principles.

Developed functional requirements for the Phoenix software, an advanced risk code that would enable analysis of all modes and hazards, and an integrated risk profile of the entire plant.

Continued training program for the next generation of risk professionals. More than 75 people have completed the course, which reduces the qualification time to develop a contributing risk engineer.

Provided improvements to the methods for developing Fire Probabilistic Risk Assessments in support of risk-informed regulation including transition to National Fire Protection Association (NFPA) 805.

Developed guidance for performing an Internal Flood Probabilistic Risk Assessment (IFPRA). This guidance will help users meet the requirements of ASME/ANS RA-Sa-2009 while saving resources in development, maintenance, and review.

Formulated guidelines for developing consistent high-quality probabilistic risk assessments (PRAs) and risk-informed regulatory submittals.

Assisted nuclear plants in pursuing regulatory relief through individual and emergency technical specification changes such as diesel generator “allowed outage time” modifications.

Current Year Activities Risk and Safety Management Program research and development for 2011 will focus on the continued socialization of risk technology with both regulators and industry management and staff. Specific efforts will include the following: Continued development of the next generation of risk professionals through the Education of Risk

Professionals course Computer-based training overview of PRA fundamentals and risk-informed regulation suitable for

management and the end users of risk information Update of the Safety and Operational Benefits of Risk-Informed Initiatives report (1016308) Continued improvements and enhancements of PRA and risk technology, most notably in the fire and

seismic hazard areas

Estimated 2011 Program Funding $10.0 M

Program Manager Ken Canavan, 704-595-2731, [email protected]

Risk and Safety Management - Program 41.07.01 p. 2


Summary of Projects


P41.07.01.01 Internal Event PRA (base) This project addresses methods research, tools development, and data collection for Internal Events Probabilistic Risk Assessments (PRAs). In the area of methods, the project solicits input from nuclear plant operators, regulators, and the public on the current technical issues facing those who prepare PRAs and who apply them in a risk-informed framework. In the area of tools, this project performs fundamental research in the area of numerical and logic modeling techniques that result in improved software tools or approaches. In the area of data, this project collects data, such as loss of offsite power data and experience that supports the development of PRAs.

P41.07.01.02 Internal Fire PRA (base) As with internal events PRA, EPRI develops the tools, data, and methods to support the development of fire PRAs including the accurate assessment of fire growth, propagation, suppression, mitigation, and the effects of the fire on equipment and cables. This project includes activities and products such as fire PRA methods, the Fire Events Database, fire growth and propagation tools, and fire effects testing.

P41.07.01.03 Seismic Risk Analysis (base) (QA)

This project maintains and improves seismic methods to resolve issues and decrease the cost of seismic assessments and seismic equipment qualification. Project activities focus on both simplified and extensive seismic risk methods that are "used and useful" by plant engineers and regulators. The project also monitors and informs regulatory programs to minimize costs for complying with seismic requirements.

P41.07.01.04 Risk-Informed Regulation (base)

This project supports engagement with stakeholders to advance risk-informed methods within the regulatory context through meetings, risk-informed programs, and EPRI research activities. This project also provides technical insight and direction for specific initiatives, such as the treatment of external events and low power and shutdown; conducts the annual Configuration Risk Management Forum meeting; and supports participation on international standards and working committees.

P41.07.01.05 PRA and Safety Training (base)

This project continues to develop and refine the Education of Risk Professionals course materials, including classroom presentations, speaker notes, demonstrations, and test banks. These courses will continue to be delivered as part of the Education of Risk Professionals training class. Five fee-based courses are scheduled for 2011 based on these materials. This project also includes the development of a series of computer-based probabilistic risk assessment (PRA) training modules.

P41.07.01.06 Analysis of Other Risks and Hazards (base) (QA)

Nuclear plants are potentially exposed to a number of external risks and hazards, including the crashing of a commercial aircraft into a nuclear plant, high wind loads, toxic chemical releases, and various business, financial, and regulatory risks. Advanced methods and tools are needed to support the evaluation and management of these diverse hazards.





P41.07.01.09a PHOENIX Technology Development (supplemental) (QA)

This project support development of PHOENIX, an all-modes and all-hazards advanced risk tool that provides an integrated risk profile of the plant. PHOENIX will be an advanced risk tool that enables users to address multiple hazards and modes of operation efficiently, obviating the current need for multiple tools. Additional planned capabilities include the ability to interface with inspection and automated log-books, remote equipment monitoring devices, and materials degradation matrices.

P41.07.01.10 PRA Scope & Quality (supplemental)

The PRA Scope and Quality project provides a vehicle for developing PRA guidance based on state-of-the-art technology. This project also provides a means for transmitting this technology to end users, through seminars, workshops, and webcasts.

P41.07.01.11d PRA Documentation Assistant (supplemental)

This project develops software and compiles best practices to reduce the resource burden associated with PRA documentation and associated risk-informed applications.

P41.07.01.12 NFPA 805 and Fire PRA (supplemental)

This project addresses issues impacting plant and regulatory actions related to fire PRAs, including the change process, change evaluations, fire-induced multiple spurious operations, fire-related human reliability analysis (HRA), operator manual actions, and non-conforming barriers. The Fire PRA User Group responds to technical and procedural questions on the use of fire PRA methods.

P41.07.01.13 SQUG/SEQUAL (supplemental) (QA)

This project supports management and operation of the Seismic Qualification Utility Group (SQUG) and the Seismic Experience-Based Qualification (SEQUAL) program, which provide a cost-effective methodology for evaluating the seismic ruggedness of nuclear plant equipment without costly seismic shake table testing or analysis.

P41.07.01.14f Risk-Informed Option 2 User Group (supplemental)

This project develops guidance to streamline the licensing process and define appropriate treatment practices for low-safety-significant structures, systems, and components. A project technical steering committee consisting of licensee personnel has been established to provide a forum for plant personnel to communicate issues, discuss resolution possibilities, review best practices, and benchmark against other plants’ practices.

P41.07.01.15g GOTHIC Advisory Group (QA) (supplemental)

This project develops software, procedures, and applications guidance for conducting three-dimensional analyses of reactor containment buildings under accident conditions. The user group provides guidance on new software features, as well as training and user support for the software.

P41.07.01.15h HRA/PRA Tool User Group (QA) (supplemental)

This users group supports enhancements to and application of the HRA Calculator software tool. Due to the importance of human actions to nuclear plant safety risk and the impact of these actions on PRA results, the HRA User Group also provides training on HRA methods and the HRA Calculator software at least twice per year.

P41.07.01.15j Risk and Reliability User Group (supplemental)

EPRI has developed a diverse suite of computer software tools to support risk evaluations and plant operations, including fault tree codes (CAFTA), scheduling and risk mitigation tools (EOOS), and software for specialized analyses (for example, FRANX for fire PRA). This project provides software training, user support for these applications, and industry meetings that provide a forum for sharing results and methods among the user community.




P41.07.01.16b MAAP 5 (QA) (supplemental)

This project supports the MAAP5 software code, which was developed to model the core, primary system, and balance of plant of a nuclear reactor following a postulated severe core damage accident. The software follows core melt and relocation, primary system thermal hydraulics, and fission product transport. This software is benchmarked against numerous physical experiments. It provides finer nodalization than MAAP4 and will be particularly useful for designing and licensing new plants. The software is developed under 10CFR50 Appendix B QA requirements for regulatory applications.

P41.07.01.16i MAAP 4 (QA) (supplemental)

This project supports the Modular Accident Analysis Package Version 4 (MAAP4) software code, which models the core and primary system of a nuclear reactor following a postulated severe core damage accident. The software models accident sequence analysis through core heat-up, core melt and relocation, primary system thermal hydraulics, and fission product transport. This software is benchmarked against numerous physical experiments as well as other thermal-hydraulic and severe accident codes. The software is developed under 10CFR50 Appendix B QA requirements for regulatory applications.

P41.07.01.17 On-Line Maintenance (OLM) Assessment (supplemental)

Many countries have observed U.S. on-line maintenance (OLM) practices and inquired as to their applicability and benefits outside the U.S. regulatory arena. Through an on-line maintenance assessment, EPRI supports utility efforts in evaluating and implementing on-line maintenance. The assessment consists of a review of U.S. OLM experience, interviews with utility staff to assess current situation and identify goals, and recommendations for potential OLM implementation strategies

P41.07.01.17l FTREX (supplemental) This project supports continued development and maintenance of the FTREX software, which is widely used in the quantification of large fault trees. FTREX 1.0 was first issued in 2007, and faster, improved updates are released every 6 to 9 months.

Internal Event PRA (base) (068078)

Key Research Question Risk technology and probabilistic risk assessments (PRAs) are currently being used in design, maintenance, and operational decision-making as well as in various regulatory interactions. The foundation for all of these applications, and for PRAs performed for other hazards (such as fires and earthquakes) is the PRA developed to address internal initiating events. Internal events are those events resulting from upsets in plant systems, ranging from turbine trips to design-basis loss-of-coolant accidents (LOCAs). Although methods for these PRAs have evolved significantly over the past 30 years, some areas continue to require further attention. In addition, as applications of PRA technology are increasingly a part of supporting plant operations, the need for improved tools to develop, maintain, and use the PRAs becomes increasingly important as well.

Approach This project addresses methods research, tools development, and data collection for Internal Events PRAs. In the area of methods, the project solicits input from nuclear plant operators, regulators, and the public on the current technical issues facing those who prepare PRAs and who apply them in a risk-informed framework. These issues are addressed through appropriate research efforts aimed at arriving at practical solutions. Among the issues that are addressed are those relating to defining and evaluating accident sequences, evaluating reliability data, assessing human reliability, and quantifying relevant risk measures. The solutions are developed


in the form of technical reports and exchanges with member users through seminars and workshops. In the area of tools, this project performs fundamental research in the area of numerical and logic modeling techniques that result in improved software tools or approaches. In the area of data, this project collects data, such as that data relating to loss of offsite power or other industry experience, that supports the development of PRAs.

Impact Enhance quality of PRAs to support risk-informed applications more effectively Promote consistency and stability in the use of PRA in the regulatory environment Provide tools to make performance and application of PRAs more effective and efficient Foster a safety and risk-informed culture

How to Apply Results Members incorporate the guidance and tools produced through the Internal Events PRA effort into the development and use of plant-specific PRAs. These products are made available via technical reports, topic-specific workshops, and improved software tools. Members also can request individual expert support for more sophisticated guidance.

2011 Products



Resolution or Update of One PRA Scope and Quality Issue: Continue to resolve issues of Probabilistic Risk Assessment (PRA) Scope and Quality associated with Internal Event PRAs.


Internal Fire PRA (base) (068084)

Key Research Question To support risk-informed regulation and continue improving plant safety and performance, key risk contributors must be analyzed in a cost-effective and realistic fashion. Events initiated as a result of fires within nuclear plants have been shown to contribute to the overall risk profile. In addition, risk-informed applications such as the Maintenance Rule and Risk-Managed Technical Specification (RMTS) need to address the potential contribution from fire events to meet regulatory expectations. Other risk-informed applications of fire risk include the National Fire Protection Association (NFPA) 805, which defines deterministic and probabilistic rules for nuclear plant fire protection. Lastly, the American Nuclear Society, in collaboration with the American Society of Mechanical Engineers, has developed a Fire Probabilistic Risk Assessment Standard. Research to advance fire PRA methods to ensure that they are realistic and cost-effective is necessary to ensure that current, new, and proposed risk-informed regulations are cost effective, improve safety, and improve operational effectiveness.

Approach Fire PRAs are useful tools in evaluating the risk of fire-initiated events. Fire PRAs depend on a number of inputs including accurate data as well as a realistic assessment of fire growth, propagation, suppression, mitigation, and the effects of the fire on equipment and cables. The Fire Risk Methods projects develop and/or maintain the following: Fire PRA Methods to ensure consistent and robust methods for fire risk analysis



Fire Event Database (FEDB) to provide fire data for establishing fire PRA inputs and evaluating the effectiveness of a variety of activities and event significance

Fire Growth and Propagation tools and methods to model the growth, propagation, suppression, and mitigations of fire-initiated events in a realistic fashion

Fire Effects Testing to benchmark and improve existing physical models of fire growth, propagation, and impacts on equipment and cables

Risk-informed applications methods that provide the industry with guidance for using risk information Training to encourage and facilitate product use

Impact Potentially reduce compliance costs Contribute to efforts to provide regulatory stability Improve safety through use of risk-informed/performance-based approaches to fire protection Facilitate assessments of plant fire programs and define the fire risk significance of plant changes

How to Apply Results Members participate in training courses to learn how to develop fire scenario models and perform fire PRAs in-house. The results of these efforts will be used to make risk-informed decisions on plant changes and program aspects and to concentrate effort on the most risk-significant areas. Members also participate in the development of the fire PRA and fire modeling research efforts by providing input and feedback to the various technical guides and reports, databases, or tools.

Seismic Risk Analysis (base) (QA) (068085)

Key Research Question Seismic issues arise periodically due to regulatory concerns or actual seismic events. Because seismic issues can expose plants to significant economic risks, research is needed to develop consistent and realistic methods to analyze, address, and mitigate seismic risk. Although the use of risk-informed methods associated with internal events is fairly well established, those for external events are less so. The two established methods for assessing seismic risk are seismic probabilistic risk assessment (SPRA) and seismic margins assessments (SMA). Both methods have been used in the past, but not routinely or to support risk-informed applications for reducing unnecessary costs. Neither has been integrated with the internal events probabilistic risk assessments (PRAs) on which they depend.

Approach This project maintains and improves seismic methods to resolve issues and decrease the cost of seismic assessments and seismic equipment qualification. Project activities focus on simplified and extensive seismic risk methods that are "used and useful" by plant engineers and regulators; SMA methods useful for developing plant seismic risk insights without SPRA; and approved risk-informed equipment seismic qualification methods. The project also monitors and informs regulatory programs to minimize costs for complying with seismic requirements. Specific areas of research and related activities include the following: Seismic Engineering (deterministic research)

Seismic Hazard Revision Earthquake Experience Collection and Analysis New Technologies (base isolation)



Seismic PRA (probabilistic research) Seismic PRA Methodology Seismic Risk Application Risk-Info rmed Seismic Margins

Impact Member ability to perform or manage seismic PRAs for their plants Risk-informed decision-making that incorporates seismic contributions Address significance determination of seismic related events or findings Improvements and maintenance of other seismic efforts such as margins analysis

How to Apply Results Member engineers will use the products to perform or manage seismic PRAs and seismic margins assessments.

Risk-Informed Regulation (base) (068094)

Key Research Question Risk-informed activities have become ingrained in nuclear plant operation in many countries, becoming a “win-win” for both the regulator and the plant owners. Risk-informed approaches allow the regulator to focus on issues truly important to safety while encouraging safety improvements, enabling operational flexibility, and providing an opportunity for cost reductions. Continued development and support is necessary to enable risk-informed activities to drive improvements in plant operational performance while continuing to improve safety.

Approach This project supports activities enabling the broad use of risk-informed approaches in multiple applications: the maintenance rule, risk informed in-service inspection, Risk-Managed Technical Specifications, Regulatory Guide 1.200, the design basis for large loss of coolant accident (LOCA), and applications to digital instrumentation and control systems. As a risk-informed specialty, continued updates and improvements to configuration risk management processes can improve the safety and efficiency of plant maintenance activities. This project supports global dialogue on configuration risk management. The Configuration Risk Management Forum (CRMF) provides a forum for configuration risk management experts to coordinate the needs of the configuration risk management community and identify ways to enhance capabilities and benefits at nuclear power plants. This project also supports low-power and shutdown qualitative risk assessment methods, including participation in international PRA standards development that impacts risk assessment methods.

Impact Demonstrate robust application of PRA methods, tools, and results to support risk-informed regulatory

initiatives Permit plant management to effectively prioritize and allocate resources to improve plant safety and

economic performance Improve effectiveness of plant configuration risk management programs Support regulatory compliance



How to Apply Results Research in risk-informed industry initiatives often results in documentation that serves as implementation guidance for members. Methods and results also are communicated in various forums, including conferences, industry meetings, and specific application workshops and training sessions. The Configuration Risk Management Forum includes an annual meeting for configuration risk management stakeholders to discuss technical research and industry issues. The CRMF Steering Committee identifies and prioritizes areas where research and development is needed to enhance nuclear plant capabilities/competitiveness or to address emerging regulatory requirements.

PRA and Safety Training (base) (068104)

Key Research Question Nuclear plants face challenges in maintaining and expanding probabilistic risk assessment (PRA) staffs to support risk analyses and programs and in maintaining expertise with safety management and thermal-hydraulic codes. Some of the nuclear industry's trained personnel have left for other industries, been hired by regulatory agencies, or will retire in the near future. Because university-based PRA training programs are rare, nuclear plants must devote resources to on-site development of engineering staffs. Improved training and qualification programs are needed to more effectively develop trained and certified PRA personnel to perform risk management tasks, support operations through equipment configuration risk management, and support risk-informed applications and Nuclear Regulatory Commission (NRC) submittals. Easily accessible and digestible management and staff training on risk, risk-informed regulation, and PRA applications is needed to socialize risk into the fabric of nuclear power plant operation. To address staff turnover and expanded use of risk assessment tools, training is also needed on safety management codes such as GOTHIC, MAAP, RETRAN, and VIPRE.

Approach PRA training needs are addressed by providing comprehensive training to educate the next generation of risk professionals and by providing overview training to decision-makers and nuclear plant specialists. The detailed training for new risk professionals is accomplished through a series of six one-week courses designed to provide the formal instruction necessary to qualify an engineer to work on a level 1 probabilistic risk assessment. The overview training is provided through one-hour computer-based training (CBT) modules designed for management and end users of risk technology. The courses cover PRA fundamentals, risk-informed and performance-based regulation, and PRA applications. To ensure that expertise in safety management codes is maintained, EPRI routinely offers training on GOTHIC, MAAP, RETRAN, and VIPRE.

Impact The Education of Risk Professionals program has exceeded expectations as an integral part of developing new risk professionals. The qualification time for new risk professionals can potentially be reduced from 3 years to less than 18 months by combining the Education of Risk Professionals training with mentoring and signoff by supervision at the workplace. Training on safety management codes provides the background necessary to apply these codes and keep current with new code developments.



How to Apply Results The EPRI Education of Risk Professionals PRA training provides the formal training portion, including “hands-on” exercises in support of utility mentoring and qualification programs. Computer-Based Training is downloadable from the EPRI website and available for training of management and engineering support personnel. Members needing safety management code training attend the courses of interest, which are arranged based on user needs.

Analysis of Other Risks and Hazards (base) (QA) (068106)

Key Research Question Nuclear plants are potentially exposed to a number of external risks and hazards, including the crashing of a commercial aircraft into a nuclear plant, an armed attack, high wind loads, toxic chemical releases, and various business, financial, and regulatory risks. Advanced methods and tools are needed to support the evaluation and management of these diverse hazards. For example, the industry has recognized the need to update the technical basis for emergency planning, which is currently rooted in analysis technology 30 years old. An updated emergency planning technical basis would be risk-informed and would account for the great increase in knowledge over the past 30 years.

Approach This project develops a wide range of methods and tools to address the many external issues that can impact nuclear plant safety risk. This research area develops methods (both qualitative and quantitative) and implements tools to permit evaluation and effective management of specific specialized hazards as they are identified. With respect to security, project activities establish the relative risks of security vulnerabilities at nuclear plants compared to other energy options (for generation decisions) and compared to other infrastructure threats (for security decisions). Past projects have included structural integrity studies in response to various external attack scenarios and an analysis of risk-informed defensive strategies to lessen or thwart the effectiveness of an imminent attack. With respect to emergency planning, the project develops improved modeling of plume radionuclide concentrations, dose rate tracking, evacuation modeling, and health effects assessment. With respect to other hazards and risks, the project develops methods to evaluate risks to power production, methods to effectively prioritize a portfolio of capital investment projects, and methods to characterize plant safety margins. This activity has significant interface with Life-Cycle Management (LCM) research in the EPRI Equipment Reliability Program and EPRI Long-Term Operations (LTO) research initiative.

Impact Evaluate the effects of aircraft impact on new and existing nuclear plant designs and assess inherent

defensive capabilities of existing plant structures Simplify vulnerability assessments compared to past efforts Provide the basis for a risk-informed methodology quantifying the relative effectiveness of various offsite

protective action strategies Provide an updated technical basis for emergency planning, including consideration of a risk-informed

approach and quantification of the margin in the required 10-mile emergency planning zone Demonstrate robust application of risk and hazard analyses to support effective, efficient risk

management Permit plant management to effectively prioritize and allocate resources to improve plant safety and

economic performance



How to Apply Results EPRI research addressing other plant risks and hazards results in methods and application software that supports effective and efficient analysis of these risks—in terms of their impact on plant safety, plant operations, plant economics, and other parameters. Results obtained are used to support effective decision-making that accounts for these risks and their potential impacts. Methods and results also are communicated in various forums, including conferences, industry meetings, and specific application workshops and training sessions.

PHOENIX Technology Development (supplemental) (QA)

Key Research Question As the use of risk technology permeates nuclear plant design, maintenance, and operation, risk practitioners need advanced tools capable of reflecting the current plant configuration and condition (including the operation state, ambient conditions, any degraded equipment or states). Greater regulatory attention to additional risks and hazards reinforces the need for an all-modes risk tool.

Approach This project supports development of PHOENIX, an all-modes and all-hazards advanced risk tool that provides an integrated risk profile of the plant. PHOENIX will be capable of interfacing with inspection and automated log books, remote equipment monitoring devices, and materials degradation matrices to provide the most current information to operators and other decision-makers when these tools and interfaces become available. PHOENIX highly leverages existing technology and its development is planned in several major phases over a 5-year period. The first phase is the integration of various probabilistic risk assessment (PRA) software for model building into a more powerful and consistent platform. The second phase is the addition of risk-monitoring phases and the addition of existing advanced modeling and quantification technology. Follow-on phases include the addition of new technology capable of integrating additional databases and information archives, and external interfaces will be added.

Impact PHOENIX represents the next generation of risk tools that are consistent, integrated, and capable of expansion beyond traditional risk analysis.

How to Apply Results PHOENIX will be used by existing staff in a similar manner as existing risk tools and monitors.

2011 Products



PHOENIX - Integrated Platform - Release 1: PHOENIX highly leverages existing technology, and the first step is the integration of various probabilistic risk assessment (PRA) software, both model building and risk monitoring, into a more powerful and consistent platform. Release 1.0 provides the first version of the fully integrated risk tool platform, including risk-model development and risk-monitoring software.

12/31/11 Software



PRA Scope & Quality (supplemental) (059410)

Key Research Question Risk technology and probabilistic risk assessments (PRAs) are currently being used in design, maintenance, and operational decision-making as well as in various regulatory interactions. A significant issue in the use of risk technology is the scope and quality, or technical adequacy, of the PRAs supporting these uses. For PRAs to have maximum utility, technology improvements must be captured and transmitted to PRA practitioners.

Approach The PRA Scope and Quality project provides a vehicle for developing PRA guidance based on state-of-the-art technology. This project also provides a means for transmitting this technology to end users through seminars, workshops, and webcasts. The PRA Scope and Quality project solicits input from nuclear plant operators, regulators, and the public on the current technical issues facing risk-informed regulation. These issues are prioritized and resolved through a combination of research and development and consensus building among the stakeholders. The result is a consensus guide that provides a clear, cost-effective, and consistent approach for addressing specific technical issues.

Impact Enable risk-informed applications Create consistent high-quality PRA and submittals Encourage sStability in the regulatory environment Foster a safety and risk-informed culture

How to Apply Results The PRA Scope and Quality committee meets several times a year to review, prioritize, and resolve PRA scope and quality issues. The research and development activities are accomplished using a collaborative process to develop issue-specific guidance, subject to peer review and pilot evaluation. Members apply the resulting risk guidance in plant-specific PRAs and risk-informed applications.

PRA Documentation Assistant (supplemental) (063062)

Key Research Question Maintaining and upgrading a probabilistic risk assessment (PRA) is resource intensive. Documenting the PRA model changes is daunting and time consuming. Tools and technologies that minimize the manual handling of PRA documentation and data can reduce the overall resource burden associated with PRAs.

Approach The PRA Documentation Assistant (PRA DocAssist) Project develops software tools and features for automating and managing the documentation process. This includes capturing key information about the model, developing the tools and processes that preserve cross-links to related information, and tracing the evolution of PRA assumptions and models. In addition, sharing of industry best practices on the configuration control of the PRA model and its associated documentation will lead to a more consistent industry approach and further resource reduction. The project will build on existing EPRI tools with input from a utility committee. Training, support, and a maintenance version of the code will be available to participants.

Impact Reductions in the required resources through automation and process improvements can significantly increase PRA staff productivity. Furthermore, ready access to the documentation can reduce the resource burden associated with regulatory interaction on risk-informed applications (for example, Significance Determination Process [SDP], Mitigating Systems Performance Index [MSPI], and others) and peer reviewers, as well as



enhance the ability to demonstrate compliance with PRA standards. An additional benefit of the project will be the ability to more quickly familiarize new staff, the Nuclear Regulatory Commission (NRC), and PRA peer reviewers with the PRA model and documentation.

How to Apply Results The software and approaches developed in this project can be immediately implemented by the plant PRA staff by importing any existing documentation into the system. Further benefits are obtained by making the software part of daily use by the PRA engineers. This project provides the tools, techniques, and plant experience support to help members best customize the software configurations and work flows to meet the needs of their PRA models and staff capabilities.

NFPA 805 and Fire PRA (supplemental) (068087)

Key Research Question As the nuclear industry addresses current and emerging fire protection issues, and as some plants begin the transition to National Fire Protection Association (NFPA)-805, technical assistance is often needed to support the development of Fire Probabilistic Risk Assessments (PRAs) and risk-informed, performance-based fire protection. There is an ancillary need to provide a forum for communication among utilities going through the transition.

Approach EPRI and the Nuclear Regulatory Commission (NRC) Research published a Fire PRA Methodology Guide (EPRI 1011989 / NUREG/CR 6850) in 2006. The methods in this report have been piloted and a joint EPRI/NRC effort is in progress to revise the Guide. Among the issues being addressed to date are change process, change evaluations, fire-induced multiple spurious operations, fire-related human reliability analysis (HRA), operator manual actions, and non-conforming barriers. A Fire PRA User Group has been established to respond to technical and procedural questions on the use of fire probabilistic risk assessment (PRA) methods.

Impact The cooperative effort between EPRI and NRC has resulted in the availability of the methodology guide and the resolution of several previously contentious issues. The project communication forum will allow for timely resolution of plant-specific issues as well as generic issues as they emerge in the transition process.

How to Apply Results Results will be reflected in revisions to of the Guide and in individual EPRI-issue guidance reports. The first of these reports is EPRI 1013489, Use of Fire PRA Methodology in Estimating Risk Impact of Plant Changes. Other guidance documents will be used by utility personnel in their development of fire probabilistic risk assessments (1019259) and their implementation of risk-informed and performance-based fire protection.



SQUG/SEQUAL (supplemental) (QA) (057060)

Key Research Question Seismic qualification of nuclear plant equipment and components remains an ongoing concern as additional data from seismic events are collected and analyzed. In the United States, Nuclear Regulatory Commission (NRC) Unresolved Safety Issue (USI) A-46 required seismic re-evaluation of the equipment in older operating plants to render plant seismic ruggedness comparable to newer plants in which the equipment was qualified to newer standards.

Approach EPRI instituted the Seismic Qualification Utility Group (SQUG) to resolve NRC Unresolved Safety Issue A-46, “Seismic Qualification of Equipment in Older Nuclear Power Plants.” EPRI and the nuclear plant community successfully resolved the issue through development and implementation of an experience-based method that uses equipment performance data from power and industrial facilities that have undergone actual earthquakes. The SQUG/ Seismic Experience-Based Qualification (SEQUAL) program continues to investigate earthquakes to add to the database, add new equipment classes, and develop and implement methods for experience-based seismic qualification of replacement equipment and parts. SQUG/SEQUAL also promotes the use of the experience-based methodology to its international members and to organizations beyond the nuclear community.

Impact Provide a cost-effective methodology for evaluating the seismic ruggedness of nuclear plant equipment without costly seismic shake table testing or analysis.

How to Apply Results Member engineers apply the SQUG methods to assess the seismic ruggedness of plant equipment; perform seismic evaluations of plant changes for heating, ventilating, and air conditioning (HVAC), overhead cranes, and piping; and qualify new and replacement equipment and parts.

Risk-Informed Option 2 User Group (supplemental) (061435)

Key Research Question 10CFR50.69 allows the licensee to reduce the “nuclear special treatment” requirements currently imposed upon structures, systems, and components (SSCs) for those SSCs determined to be low-safety significant. Resources have been mainly dedicated to defining and conducting trial applications of risk-informed classification criteria, responding to Nuclear Regulatory Commission (NRC) comments, and supporting issuance of the final rule language and supporting guidance (for example, Regulatory Guide 1.201, Rev1). Now that the rule and accompanying guidance have been approved, there is a need to investigate, develop, and document appropriate “good practice” processes for plant-specific implementation of 10CFR50.69.

Approach This project develops guidance to streamline the licensing process and define appropriate treatment practices for low safety significant SSCs. A project technical steering committee consisting of licensee personnel has been established to provide a forum for plant personnel to communicate issues, discuss resolution possibilities, review best practices, and benchmark against other plants’ practices. The technical steering committee will work with EPRI to prioritize the list of activities and provide input/direction for the program. EPRI will coordinate with other user groups and industry efforts, including licensees and owners groups, and will interface with other EPRI projects addressing equipment qualification, seismic issues, pressure boundary components, procurement to develop good practice documents, and technical basis.



Impact Defining effective practices for classifying applicable SSCs and treatment practices for low-safety significant SSCs is an important part of achieving a successful 10CFR50.69 program. This step can result in significant cost savings while maintaining and/or improving the reliability and performance of SSCs.

How to Apply Results Project participants will use the results to streamline classification of applicable SSCs and develop treatment processes for low-safety significant SSCs.

GOTHIC Advisory Group (QA) (supplemental) (004444)

Key Research Question Software codes are used to perform thermo-hydraulic safety analyses of reactor containment buildings for regulatory applications. Feedback and input from users can help define needed improvements, enhance training effectiveness, and increase industry application.

Approach This project develops software, procedures, and applications guidance for conducting three-dimensional analyses of reactor containment buildings under accident conditions. The user group provides guidance on new software features, as well as training and user support for the software. GOTHIC Version 7.2b (QA) and prior versions of software (QA) were developed and are being maintained by EPRI in a manner compliant with Title 10 of the Code of Federal Regulations Part 50 (10 CFR 50) Appendix B Quality Assurance and 10 CFR 21.

Impact The GOTHIC code is widely used by utilities for performing thermo-hydraulic safety analyses of reactor containment buildings. The large user group results in ongoing refinement and modifications that enhance its value in addressing regulatory applications and responding to regulatory requests for additional information.

How to Apply Results The Gothic User Group provides the software, upgrades, technical support, and training required to use GOTHIC.

2011 Products



GOTHIC 8.1.beta 11/30/11 Software



HRA/PRA Tool User Group (QA) (supplemental) (049250)

Key Research Question EPRI formed the Human Reliability Analysis (HRA) User Group and developed the HRA Calculator in response to member needs to promote quality and consistency in the use of HRA calculation methods and to consolidate these methods into a comprehensive software tool. The mission has expanded to include development of fire HRA methods, involvement in HRA benchmarking projects, and tracking of other research efforts in HRA. Continued development and maintenance of the software are necessary to sustain and expand its usefulness to the nuclear industry.

Approach The HRA User Group has developed a software tool, the HRA Calculator, to support increased consistency in the use of HRA methods. The User Group seeks to maintain the software, improve its usefulness, and add capabilities where needed. The HRA User Group also provides training on the software at least twice each year. In addition, the User Group participates in the benchmarking of HRA methods and in other research areas aimed at improving HRA methods.

Impact The HRA Calculator provides users with an effective software tool capable of applying a variety of HRA methods. The major methods implemented in the HRA Calculator are those developed by EPRI, including the Cause-Based Decision Tree Method (CBDTM) and the Human Cognitive Reliability with Operator Reliability Experiment (HCR/ORE) approach. The HRA Calculator also incorporates such methods as the Technique for Human Error Rate Prediction (THERP) and the Standardized Plant Analysis Risk-HRA (SPAR-H) method. The user group continues to grow in its industry leadership role as recognized by the NRC and other entities seeking EPRI participation in the area of human reliability analysis.

How to Apply Results The methods and software developed in this project can be immediately implemented by plant PRA staff responsible for performing human reliability analysis. Additionally, training on the HRA methods and the HRA Calculator software provides an efficient mechanism for members to expand capabilities to perform these evaluations.

Risk and Reliability User Group (supplemental) (003888)

Key Research Question Risk practitioners rely on tools such as the Risk and Reliability Workstation to achieve substantial efficiencies when performing risk analyses for probabilistic risk assessment (PRA) applications. Feedback and input from users can help define needed improvements, enhance training effectiveness, and increase industry application.

Approach A suite of computer software tools has been developed to support risk evaluation and plant operations. This includes fault tree codes (CAFTA), scheduling and risk mitigation tools (EOOS), and software for specialized analyses (for example, FRANX for fire PRA). Guidelines are produced to aid the engineer, and regular training sessions are conducted. Meetings are held several times per year to promote training on the use of the software and provide a forum for sharing results and methods among peers.

Impact The project has resulted in near 100% use of the Risk and Reliability Workstation software by U.S. plants and is widely used in international nuclear units and in other industries. The project provides a cost-effective method



for members to provide input into and prioritize needed enhancements to the risk tools. It also provides a useful forum from which experiences from the user community can be shared.

How to Apply Results The software, training, and user support available to members of the Risk and Reliability User Group is directly applicable by risk practitioners. Lessons learned from user group meetings can be incorporated into utility risk approaches as necessary. Software, training, and user support are available to members of the Risk and Reliability (R&R) User Group. Meetings held several times a year promote training on the use of the software and provide a forum for sharing results and methods among peers.

2011 Products



FRANX 5.0: FRANX version to do flooding evaluation analysis 11/15/11 Software




FRANX 6.0: Advanced version of FRANX to do seismic PRA analyses 11/19/12 Software

MAAP 5 (QA) (supplemental) (050159)

Key Research Question New regulatory applications and new nuclear plant designs introduce accident conditions and scenarios that the existing Modular Accident Analysis Package (MAAP4) can’t address. An improved version, therefore, is needed for robust safety analyses related to the post-accident behavior of the core and thermal hydraulics of the primary and secondary systems, including fission product release and transport.

Approach This project supports continued development and maintenance of the MAAP5 software code, which was developed to model the core, primary system, and balance of plant of a nuclear reactor following a postulated severe core damage accident. The software follows core melt and relocation, primary system thermal hydraulics, and fission product transport. This software is benchmarked against numerous physical experiments. It provides finer nodalization than MAAP4 and will be particularly useful for designing and licensing new plants. The software is developed under 10CFR50 Appendix B QA requirements for regulatory applications. Both Windows and Linux versions of the software are available.

Impact MAAP5 will be extensively used for the design of current and new generations of nuclear reactors. The analytical results generated by MAAP5 will be important in the final licensing of the new designs.

How to Apply Results The software comes with a complete users manual, including solved examples. An applications manual has been developed to provide users guidance for specific, frequently encountered analyses. User Group meetings and training (both in person and web-based sessions) are provided.



2011 Products



MAAP 5.2 Beta: Advanced version of MAAP5.1 focusing on additional experimental benchmarks

MAAP 4 (QA) (supplemental) (003068)

Key Research Question Nuclear plant owners must perform severe accident analysis in support of regulatory and safety applications. Such analyses depend on the ability to model and assess the behavior of the reactor core and fission products (for example, probability risk assessment [PRA] success criteria and accident sequence time and human reliability analysis). MAAP4 software and applications guidance were developed to fill these needs.

Approach This project supports continued development and maintenance of the Modular Accident Analysis Package Version 4 (MAAP4) software code, which models the core and primary system of a nuclear reactor following a postulated severe core damage accident. The software models accident sequence analysis through core heat-up, core melt and relocation, primary system thermal hydraulics, and fission product transport. This software is benchmarked against numerous physical experiments as well as other thermal-hydraulic and severe accident codes. MAAP4 remains one of the few integrated thermal-hydraulic severe accident codes available. The software is developed under 10CFR50 Appendix B QA requirements for regulatory applications. Both Windows and Linux versions of the software are available.

Impact Utilities use MAAP software to assess the Final Safety Analysis Report (FSAR) Chapter 15 analyses and to identify appropriate success criteria, perform accident sequence analysis, and identify human reliability timing in probabilistic risk assessments (PRAs). It has a large user group and is the code of choice for members to address regulatory applications and to respond to Nuclear Regulatory Commission (NRC) Requests for Additional Information (RAIs).

How to Apply Results The software comes with a complete users manual, including solved examples. An applications manual has been developed to provide user guidance for specific, frequently encountered analyses. User group meetings and training (both in person and web-based sessions) are provided.

2011 Products



MAAP4 version 4.0.8: Updated version of MAAP4 09/15/11 Software



On-Line Maintenance (OLM) Assessment (supplemental)

Key Research Question On-line maintenance (OLM) can offer significant operational, economic, and safety benefits. Many countries have observed U.S. on-line maintenance practices and inquired as to their applicability and recognized benefits outside the U.S. regulatory arena. Realizing these benefits and ensuring effective implementation, however, requires a careful consideration of plant capabilities, maintenance culture, and regulatory limitations.

Approach Through an OLM assessment, EPRI supports utility efforts in evaluating and implementing on-line maintenance. The assessment consists of working sessions with the following elements: Review of U.S. OLM experience—practices, support systems and evolution to current status Interviews with key utility staff to determine the utility’s current situation and goals related to OLM Recommendations identifying of key considerations, potential benefits, and potential OLM implementation

strategies specific to the utility’s situation and goals The assessment will be conducted over a 4-day period at the utility with two EPRI staff to support the workshop. The first two elements (review of U.S. OLM experience and discussions and interviews with key utility staff) are anticipated to take place in the first two days and the third element (EPRI recommendations) on the fourth day.

Impact The assessment will provide member utilities with an understanding of the several areas in which OLM differs from outage maintenance, but that are important in planning for the introduction of OLM. These include risk and safety, regulatory, maintenance practice, and work management considerations. Additionally, the EPRI recommendations will help the member utility determine where it has strengths and gaps related to moving forward with OLM and, thus, the most appropriate implementation strategies.

How to Apply Results The results of the assessment can immediately support decisions related to use of OLM and assist in identifying OLM implementation strategies. The primary products will be a review of U.S. OLM experience, a summary of utility interviews, and EPRI recommendations.

FTREX (supplemental) (063727)

Key Research Question As risk is increasingly employed in routine decision-making, the ability to rapidly produce results is needed. Advanced quantification engines such as FTREX are needed to support state-of-the-art analysis of large fault trees.

Approach This project supports continued development and maintenance of the FTREX software, which is widely used in the quantification of large fault trees.

Impact FTREX is currently the fastest solution engine for solving PRA fault trees currently available from any source. It is especially useful for large models of systems and plants. It typically reduces the solution times from many hours to a few minutes of computer time.




How to Apply Results FTREX is a simple add-in module for PRA type software.

2011 Products



FTREX 1.6 beta 09/15/11 Software

Used Fuel and High-Level Waste Management

Inform industry and regulatory decisions affecting waste management, including advanced fuel cycles, used fuel transportation, geologic repository

capacity, and interim storage requirements.

Program Used Fuel and High-Level Waste Management

Used Fuel and High-Level Waste Management (QA)

Program Overview

Program Description Nuclear power's long-term viability depends on safe, cost-effective management of used fuel and high-level waste. An integrated solution encompassing fuel cycle closure, centralized interim storage, transportation, and permanent geologic disposal is achievable, but requires significant research and technology development. The Used Fuel and High-Level Waste (HLW) Management program examines a range of complex scientific and technical issues affecting waste management, including advanced fuel cycles, used fuel transportation, geologic disposal, and on-site and interim storage requirements. Research related to these issues informs regulatory decisions on waste management and provides insight into technical advances required to develop an integrated solution.

Research Value Better understanding of the options available for managing the technical, economic, environmental, and institutional aspects of high-level waste and advanced fuel cycles is needed to inform decision-making. Electric Power Research Institute's (EPRI’s) Used Fuel and HLW Management program conducts research to characterize the various technical risks and risk mitigation options related to used fuel storage, transportation, and disposal. The program also pursues the identification of promising technologies for closing the fuel cycle, which could lead to significant increases in the amount of energy extracted from uranium and to reductions in the volume of waste requiring long-term geologic disposal. The Used Fuel and High-Level Waste Management program members gain access to the following: Technical guidance to support operation of long-term on-site and geologic repositories Technical basis to resolve generic used fuel storage and transportation issues impacting plant operability,

license renewal, decommissioning, and the licensees’ ability to move fuel off-site More efficient dry storage and transportation canister designs that offer economic savings and enhance

the licensees’ ability to move high-burnup used fuel off-site Research gaps in bringing new fuel cycle technologies to full commercial scale

Approach The Used Fuel and HLW Management program develops knowledge, guidance, and tools to reduce the risks and maximize the technical options for safe, cost-effective handling, storage, and disposal. The program incorporates international experience with fuel handling and fuel reprocessing, and collaborates with international entities on cross-cutting research addressing the nuclear fuel cycle and supporting elements. Identify efficient used fuel management strategies to maintain plant operability Conduct technical analyses to assess used fuel transportation risks and identify infrastructure issues

related to developing an efficient used fuel transportation system Evaluate technical risks associated with long-term storage of high-level waste at nuclear plant sites Develop tools to optimize on-site wet and dry storage of used nuclear fuel Conduct technical and economic analyses of advanced fuel cycles (design, operations, scale-up,

research needs)

Accomplishments EPRI’s Used Fuel and HLW Management program provides specialized technical analysis to ensure the safe, economic handling and disposal of used fuel and HLW and to inform decisions regarding the viability of various nuclear fuel cycles.


Completed a report outlining the technical considerations that utilities should be aware of when evaluating their readiness to use mixed oxide fuel (MOX) in existing and new reactors.

Developed cost estimates evaluating trade-offs between front-end issues (for example, additional uranium ore versus cost of conversion and enrichment), back-end issues (for example, direct disposal of all used fuel versus cost of reprocessing with direct disposal), and centralized storage issues (for example, storage facility size and operating lifetime).

Issu ed the Handbook of Neutron Absorber Materials for SNF Transportation and Storage, which assists utilities in selecting and managing such materials in the spent fuel pool and in dry storage and transportation systems.

Launched a new project to develop the technical bases for extended (>60 years) used fuel/high-level waste storage. To encourage industry technical consensus, EPRI hosted a workshop to coordinate parallel efforts within the Nuclear Regulatory Commission (NRC) and the Department of Energy (DOE). The generic nature of this work also provides a valuable collaboration opportunity with EPRI’s international members.

Released Cask Loader Version 2.2a, which assists utilities in planning and executing the loading of spent fuel assemblies and their components into dry cask storage. The new version optimizes cask loading based on cost and schedule constraints and maintains data for DOE and NRC reporting.

Determined that the probability of a criticality occurring during used fuel transport is extremely low. This supports industry efforts to obtain transportation licenses for used fuel with high burnups.

Performed first dry cask storage probabilistic risk assessment, which showed the storage risk to be extremely low.

Analyzed high-burnup advanced cladding properties to support their continued use while allowing for used fuel storage and transportation.

Current Year Activities Used Fuel and HLW Management program research and development for 2011 will focus on generic used fuel storage and transportation issues and on evaluating strategies for the future of a sustainable nuclear energy fuel cycle. Specific efforts will include the following: Develop technical bases supporting licensing of high-burnup used fuel transportation systems Identify existing and near-future light water reactors that could burn mixed oxide fuel Obtain additional data on advanced nuclear fuel claddings to support used fuel storage, transportation,

and disposal Develop a program to provide the technical bases for very long-term (>60 years) used fuel storage

Selected reports and products may be prepared in whole or in part in accordance with the EPRI Quality Program Manual that fulfills the requirements of 10CFR50 Appendix B, 10CFR21 and ANSI N45.2-1977.

Estimated 2011 Program Funding $3.1M

Program Manager John Kessler, 704-595-2737, [email protected]

Used Fuel and High-Level Waste Management (QA) - Program 41.03.01 p. 2


Summary of Projects


P41.03.01.01 Used Fuel Transportation and Storage (base)

This project develops technical bases, white papers, and comments to assist in resolving generic issues associated with used fuel transportation and storage. Such issues include licensing, risk-informed treatment, and high burnup fuel. Activities conducted under this project support efforts to articulate industry responses to potential regulatory actions.

P41.03.01.02 Neutron Absorber Materials Studies (base)

This project organizes annual meetings of neutron absorber material users and manufacturers to share experience and guide future needs. It also maintains an up-to-date handbook on available neutron absorber material information related to materials properties, manufacturing processes, and field experience.

P41.03.01.03 Advanced Fuel Cycle Modeling and Waste Management (base)

This project will develop appropriate nuclear fuel cycle models that track materials and radionuclides through the system; determine the appropriate mixture of light water reactor and fast reactor technology for both start-up and long-term, steady-state operation; and evaluate the health and economic impacts. Detailed sensitivity and uncertainty analyses will be conducted to shed light on R&D, health, and economic risks. This project also will evaluate the readiness of the existing and near-future light water reactor fleet to utilize mixed oxide fuel.

P41.03.01.04 Very Long-Term Used Fuel Storage (base)

This project will identify issues that need to be addressed to achieve very long-term dry storage in existing canisters, an appropriate aging management plan (including surveillance, testing, and eventual repackaging), and identification of regulatory issues to be addressed.

P41.03.01.08 Fuel Works/Cask Loader User's Group (supplemental)

This project provides a mechanism for continued development, maintenance, and sharing of lessons learned regarding the Cask Loader software package. Cask Loader provides guidance for loading dry storage casks that enables plant personnel to minimize the need for fuel movement and significantly reduce documentation errors.

Used Fuel Transportation and Storage (base) (052406)

Key Research Question Management of used fuel and high-level waste encompasses an array of activities related to storage, transportation, and disposal. Technical support is needed in a number of generic and specific areas to improve used fuel management: Licensing issues related to transportation and storage Risk-informed treatment of wet and dry storage and transportation Storage and transportation of high-burnup used fuel Generic disposal issues common to countries with nuclear power

Approach This project develops technical bases, white papers, and comments to assist in resolving generic issues associated with used fuel transportation and storage. Activities conducted under this project support efforts to articulate industry responses to potential regulatory actions. Also, because nuclear countries face many



common technical issues regarding used fuel management, this project will identify generic disposal needs and potential research gaps. Risk-based approaches for the management of used fuel are only emerging. Probabilistic risk assessments enable risk-informed regulations and regulatory practices to be developed for wet and dry storage and transportation, providing a framework for risk-informed decision-making. EPRI will use two probabilistic risk assessments of bolted storage casks to evaluate the effectiveness of regulations and regulatory practices governing wet and dry storage activities. This project also develops technical bases and collects confirmatory experimental data to support regulatory acceptance of practical approaches for implementing dry storage and transport of used nuclear fuels with initial enrichment up to 5% and discharge burnup greater than 45 GWd/MTU. Research elements being pursued include probabilistic assessment of a criticality event during transportation; options for excluding intrusion of water in the waste package, or moderator exclusion; and data collection for full burnup credit methodology benchmarking.

Impact Potential benefits from this project include the following: Reduce unnecessary regulatory burden that could result from regulatory guidance on transportation and

storage Reduce unnecessary used fuel storage expenses, analyses, operations, and worker dose through risk-

informed regulations Provide predictable approaches for licensing transportation of used fuel without any discharge burnup

limitations, enabling nuclear plants to move used fuel off-site as soon as permitted Identify and address generic research needs related to long-term disposal of used fuel and high-level

waste

How to Apply Results Research results provide nuclear plant owners and operators with perspective on the viability of long-term disposal options. Such perspective can inform decisions regarding long-term operations and future investments in nuclear power. Technical information also can be used in many cases to support regulatory interactions and potentially minimize regulatory burden.

Neutron Absorber Materials Studies (base) (052409)

Key Research Question Neutron absorber materials providing criticality control in fuel storage pools and in used spent-fuel dry storage and transportation canisters have experienced some degradation and performance issues. Costs incurred to manage degrading neutron absorber materials can reach into the millions of dollars. Data regarding existing and proposed neutron absorber materials is needed to assist in managing existing materials and developing appropriate manufacturing controls for new neutron absorber materials.

Approach This project organizes annual meetings of neutron absorber material users and manufacturers to share experience and guide future needs. It also maintains an up-to-date handbook on available neutron absorber material information related to materials properties, manufacturing processes, and field experience.

Impact Project activities can result in more efficient management of existing neutron absorber materials, help maintain choices between available materials, and guide future development of new materials.



How to Apply Results Material development information can be used by potential new neutron absorber material vendors to establish the necessary quality assurance program for material property control and by utilities to remain aware of potential neutron absorber materials performance issues.

Advanced Fuel Cycle Modeling and Waste Management (base) (066959)

Key Research Question Nuclear fuel cycles beyond a "once-through" cycle involve a complex arrangement of many components, from uranium mining and milling through reprocessing, recycling, and disposal. To determine how a more "closed" nuclear fuel cycle should be developed and implemented, research is needed to understand material and radionuclide tracking, the appropriate mixture of light water and fast reactors, economics, and start-up issues. Nuclear fuel cycle models can be instrumental in analyzing these issues.

Approach This project will develop appropriate nuclear fuel-cycle models that track materials and radionuclides through the system; determine the appropriate mixture of LWR and fast reactor technology for both start-up and long-term, steady-state operation; and advance understanding of the health and economic impacts. Almost all of the components for each model are uncertain such that sensitivity and uncertainty analyses will be conducted. This project also will evaluate the readiness of the existing and near-future light water reactor fleet to utilize mixed oxide fuel.

Impact The models can help decision-makers evaluate research and development funding needs, investigate how to begin introducing closed fuel cycle(s) at a commercial scale, and improve understanding of the health and economic risks.

How to Apply Results Reports will be available to industry and federal decision-makers to inform funding decisions and fuel-cycle planning.

2011 Products



Advanced Fuel Cycle Modeling Case Studies 09/30/11 Technical Update

Very Long-Term Used Fuel Storage (base) (068536)

Key Research Question Used fuel will likely need to be stored for many decades until either reprocessing or final disposal becomes available. The use of dry used fuel storage technology is now in widespread use. Given the economic and worker dose costs involved in moving used fuel from fuel pools into dry storage, it is in the interest of the utilities and rate payers to be able to continue to store used fuel in the original dry storage systems rather than incurring the additional economic and worker dose related to repackaging the used fuel into a new dry storage system.



Approach Currently available information on the long-term aging of dry storage systems is enabling some utilities to obtain license extensions for their existing dry storage systems up to approximately 60 years after initial used fuel loading. Given the uncertainty of the availability of options for the ultimate disposition of used fuel, it may be necessary to store used fuel for longer than 60 years. This project will identify the factors that need to be considered for very long-term storage, such as economics and used fuel and dry storage system aging mechanisms.

Impact This project will identify technical barriers and the technical bases necessary for continuing to store used fuel in its original dry storage systems beyond 60 years.

How to Apply Results Utilities and dry storage system vendors can use the information with regulator agencies to pursue approval for extended dry storage beyond 60 years.

2011 Products



Used Fuel and HLW Storage Time-Limited Aging Analyses—Data Needs for Extended Storage 04/01/11 Technical Report

Technical and Regulatory Considerations for a New U.S. Geologic Disposal Program for Used Nuclear Fuel and High-Level Radioactive Waste


High Burnup Used Fuel Dry Storage Demonstration Program—Description of Experimental Approach 09/30/11 Technical

Update

Fuel Works/Cask Loader User's Group (supplemental) (052410)

Key Research Question Fuel pools at many nuclear power plants are near capacity. While some used fuel can be stored outside the reactor building on concrete pads, the data needed to load used fuel into dry casks are often in disparate locations, and some are not in electronic form. A software program containing the necessary data to meet plant and cask vendor technical specifications is needed to choose fuel that would meet the specifications in the cask’s Certificate of Compliance.

Approach This project provides a mechanism for continued development, maintenance, and sharing of lessons learned regarding the Cask Loader software package. Cask Loader provides a tool for loading dry storage casks that enables plant personnel to minimize the need for fuel movement and significantly reduce documentation errors. Cask Loader helps select the appropriate used-fuel assembly for each cask and then prints the utility move sheets and other required forms.

Impact Potential benefits from this project include the following: Optimize cask loading for cost and schedule Maintain data for Department of Energy (DOE) and Nuclear Regulatory Commission (NRC) reporting Interact with cask vendor models and technical specifications




Ease crowding in used fuel pools Mitigate personnel concerns on efficiency and dose Import current utility data and obtain reports in utility-specific format Facilitate easy updates for new utility or regulatory requirements Allow customization of new cask vendor and reports Perform gamma and neutron calculations Perform decay heat calculations based on Reg. Guide 3.54 (default) or NRC Branch Technical Position

ASB 9-2

How to Apply Results Members input data into Cask Loader from spreadsheets or ShuffleWorks files. The data include bundle/assembly as-built data, exposure, core location, and failure status; core data, including cycle dates and exposures; and cask data including as-built data. Cask Loader populates the chosen casks based on cask technical specification requirements and fuel available that meet the requirements of the cask. Members may choose to replace automatically chosen bundles with other candidates from the candidate pool. Reports can be printed and data output to spreadsheets or other proprietary report formats.

Documents

2011 Research Portfolio / All Nuclear - EPRImydocs.epri.com/docs/Portfolio/PDF/2011_Portfolio...Electric Power Research Institute 2011 Research Portfolio being implemented into new