Karen D. Fili Southern Nuclear Site Vice President ...Sep 18, 2015 · Karen D. Fili Southern Nuclear Site Vice President Operating Company, Inc. Plant Vogtle Units 3&4 7825 River

Karen D. Fili Southern Nuclear Site Vice President Operating Company, Inc. Plant Vogtle Units 3&4 7825 River Road Waynesboro, GA 30830 Tel 706.848.7717 Fax 706.826.5796 [email protected]

September 18, 2015 Docket Nos.: 52-025 ND-15-1333

52-026 10 CFR 55.46(b) U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Mr. Glenn M. Tracy Director, Office of New Reactors U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

Southern Nuclear Operating Company

Vogtle Electric Generating Plant Units 3 and 4 Request for a Commission-Approved Simulation Facility

Ladies and Gentlemen:

Pursuant to 10 CFR 55.46(b), Southern Nuclear Operating Company (SNC) requests a

Commission-Approved Simulation Facility for Vogtle Electric Generating Plants (VEGP) Units 3

& 4.

The enclosures provide information required by 10 CFR 55.46(b) for facility licensees that

propose use of a simulation facility other than a plant-referenced simulator in the administration

of operating tests under 10 CFR 55.45(b)(1). Enclosure 1 provides summaries of SNC

evaluations of open simulation facility discrepancies with respect to; AP1000 simulation facility

Unresolved Items (UIs) issued by the Nuclear Regulatory Commission (NRC), their cumulative

effect on operator performance, simulator conformance with the AP1000 plant design, and

variances between AP1000 simulation facilities. Subsequent enclosures provide supporting

details.

Pursuant to 10 CFR 2.390, SNC requests that the specified information be withheld from public

disclosure. In support of this request for withholding, SNC has attached to this letter the

following documents:

Enclosure 5P contains an evaluation of AP1000 simulation facility Unresolved

Items (UIs) that were issued by the NRC.

U. S. Nuclear Regulatory Commission ND-15-1333 Page 2 of 5

Enclosure 6P contains an evaluation of open simulator deficiencies and their

aggregate impact on 10 CRF 55.45 criteria.

Enclosure 8P contains an evaluation of Priority One (1) Potential Human

Engineering Discrepancies (PHEDs) from Integrated Systems Validation (ISV)

Daily Assessments.

Enclosure 9P contains a list of simulator discrepancies that were open as of May

15, 2015.

Enclosure 12 contains an affidavit for withholding proprietary information from

public disclosure, executed by Westinghouse. The Affidavit sets forth the basis

on which the information may be withheld from public disclosure by the

Commission and addresses with specificity the considerations listed in paragraph

(b)(4) of Section 2.390 of the Commission’s regulations.

Accordingly, it is respectfully requested that the information which is proprietary to WEC be

withheld from public disclosure in accordance with 10 CFR 2.390. Correspondence with respect

to the copyright or proprietary aspects of the items listed above or the supporting WEC Affidavit

should reference CAW-15-4260 and should be addressed to James A. Gresham, Manager,

Regulatory Compliance, Westinghouse Electric Company, 1000 Westinghouse Drive, Building 3

Suite 310, Cranberry Township, Pennsylvania 16066.

To support the operator licensing schedule, SNC respectfully requests NRC approval of this

request by December 18, 2015.

This letter contains no regulatory commitments. If you have any questions, please contact

Michael Yox at (706) 848-6459.

Respectfully submitted, SOUTHERN NUCLEAR OPERATING COMPANY

Karen D. Fili

Vice President,

VEGP 3&4 Operational Readiness

Nuclear Development

KDF/MC/sdc

sdcheste

Stamp

U. S. Nuclear Regulatory Commission ND-15-1333 Page 3 of 5 Enclosure 1: Information Provided Pursuant to a 10 CFR 55.46(b) - Request for a

Commission-Approved Simulation Facility

Enclosure 2: Description of the Components of the Simulation Facility Intended to be Used for Each Part of the Operating Test - 10 CFR 55.46(b)(1)(i)

Enclosure 3: Description of the Performance Tests for the Simulation Facility and Results of the Tests - 10 CFR 55.46(b)(1)(ii)

Enclosure 4: Description of the Procedures for Maintaining Examination and Test Integrity Consistent with the Requirements of 10 CFR 55.49 - 10 CFR 55.46(b)(1)(iii)

Enclosure 5: Evaluation of AP1000 Simulation Facility Summary of Unresolved Items (UIs) Issued By the NRC

Enclosure 5P: Evaluation of AP1000 Simulation Facility Summary of Unresolved Items (UIs) Issued By the NRC (Withhold from Public Disclosure)

Enclosure 6: Commission Approved Simulator Aggregate Study - Simulator Training System Deficiency Impact on 10 CRF 55.45

Enclosure 6P: Commission Approved Simulator Aggregate Study - Simulator Training System Deficiency Impact on 10 CRF 55.45 (Withhold from Public Disclosure)

Enclosure 7: List of Westinghouse Simulator Corrective Actions

Enclosure 8: Evaluation of Priority One (1) Potential Human Engineering Discrepancies (PHEDs) from Integrated Systems Validation (ISV) Daily Assessments

Enclosure 8P: Evaluation of Priority One (1) Potential Human Engineering Discrepancies (PHEDs) from Integrated Systems Validation (ISV) Daily Assessments (Withhold from Public Disclosure)

Enclosure 9: List of Open Simulator Discrepancies

Enclosure 9P: List of Open Simulator Discrepancies (Withhold from Public Disclosure)

Enclosure 10: BEACON

Enclosure 11: Acronyms & Definitions

Enclosure 12: Westinghouse Authorization Letter CAW-15-4260, Application for Withholding Proprietary Information From Public Disclosure, Accompanying Affidavit, Proprietary Information Notice and Copyright Notice

U. S. Nuclear Regulatory Commission ND-15-1333 Page 4 of 5 cc:

Southern Nuclear Operating Company / Georgia Power Company

Mr. S. E. Kuczynski (w/o enclosures) Mr. J. A. Miller Mr. D. A. Bost (w/o enclosures) Mr. M. D. Meier Mr. M. D. Rauckhorst (w/o enclosures) Mr. J. T. Gasser (w/o enclosures) Mr. D. H. Jones (w/o enclosures) Ms. K. D. Fili Mr. D. R. Madison Mr. T. W. Yelverton Mr. B. H. Whitley Mr. C. R. Pierce Mr. D. L. Fulton Mr. M. J. Yox Mr. T. R. Takats Mr. W. A. Sparkman Mr. J. P. Redd Document Services RTYPE: VND.LI.L00 File AR.01.02.06

Nuclear Regulatory Commission

Mr. V. M. McCree (w/o enclosures) Mr. M. Delligatti (w/o enclosures) Mr. L. J. Burkhart (w/o enclosures) Mr. P. Kallan (w/o enclosures) Mr. C. P. Patel Ms. D. L. McGovern Mr. B. M. Bavol Ms. R. C. Reyes Ms. M. A. Sutton Mr. M. E. Ernstes Mr. G. J. Khouri Mr. L. M. Cain Mr. J. D. Fuller Mr. C. B. Abbott Ms. S. E. Temple Mr. I. A. Anchondo

Oglethorpe Power Corporation

Mr. M. W. Price Ms. K. T. Haynes Ms. A. Whaley

U. S. Nuclear Regulatory Commission ND-15-1333 Page 5 of 5 Municipal Electric Authority of Georgia

Mr. J. E. Fuller Mr. S. M. Jackson

Dalton Utilities

Mr. D. Cope Mr. T. Bundros

Westinghouse Electric Company, LLC

Mr. R. Easterling (w/o enclosures) Mr. J. W. Crenshaw (w/o enclosures) Mr. C. D. Churchman (w/o enclosures) Mr. L. Woodcock Mr. P. A. Russ Mr. T. G. Rubenstein Mr. G. F. Couture Mr. M. Y. Shaqqo

Other

Mr. J.E. Hesler, Bechtel Power Corporation Ms. L. Matis, Tetra Tech NUS, Inc. Dr. W. R. Jacobs, Jr., Ph.D., GDS Associates, Inc. Mr. S. Roetger, Georgia Public Service Commission Ms. S. W. Kernizan, Georgia Public Service Commission Mr. K. C. Greene, Troutman Sanders Mr. S. Blanton, Balch BinghamMr. R. Grumbir, APOG


Vogtle Electric Generating Plant (VEGP) Units 3 and 4

ND-15-1333

Enclosure 1

Information Provided Pursuant to a 10 CFR 55.46(b) Request for a Commission-Approved Simulation Facility

(This Enclosure consists of 6 pages, including this cover page)

ND-15-1333 Enclosure 1, Page 2 of 6 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility TABLE OF CONTENTS

1.0 Summary

2.0 Description of Simulator Discrepancies

2.1 Cumulative Effect of Simulator Discrepancies on Operator Performance

2.2 Evaluation of AP1000 Simulation Facility Unresolved Items (UIs) Issued by the NRC

2.3 Simulator Conformance with the AP1000 Plant Design

3.0 Variances between AP1000 Simulator Facilities

4.0 Conclusion

5.0 References

Special Note:

When referring to the “VEGP Units 3&4 Simulator Training System (STS)”, the word “simulator”

will be used throughout this and subsequent enclosures.

ND-15-1333 Enclosure 1, Page 3 of 6 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility 1.0 Summary

Pursuant to 10 CFR 55.46(b), Southern Nuclear Operating Company (SNC) requests a

Commission-Approved Simulation (CAS) Facility for Vogtle Electric Generating Plants

(VEGP) Units 3 & 4 for the administration of operating tests under 10 CFR 55.45(b)(1).

This document and the related enclosures provide the information required by 10 CFR

55.46(b) for facility licensees that propose use of a simulation facility other than a plant-

referenced simulator in the administration of operating tests under 10 CFR 55.45(b)(1).

10 CFR 55.46(b)(1) states:

Facility licensees that propose to use a simulation facility, other than a plant-referenced

simulator, or the plant in the administration of the operating test under §§ 55.45(b)(1) or

55.45(b)(3), shall request approval from the Commission. This request must include:

(i) A description of the components of the simulation facility intended to be used, or the

way the plant would be used for each part of the operating test, unless previously

approved; and

(ii) A description of the performance tests for the simulation facility as part of the

request, and the results of these tests; and

(iii) A description of the procedures for maintaining examination and test integrity

consistent with the requirements of § 55.49.

Enclosure 2 contains a description of the components of the simulation facility per

paragraph (i) above.

Enclosure 3 contains a description of the performance tests for the simulation facility and

the results of those tests per paragraph (ii).

Enclosure 4 contains a description of the procedures for maintaining examination and test

integrity per paragraph (iii).

10 CFR 55.46(b)(2) states:

The Commission will approve a simulation facility or use of the plant for administration of

operating tests if it finds that the simulation facility and its proposed use, or the proposed

use of the plant, are suitable for the conduct of operating tests for the facility licensee's

reference plant under § 55.45(a).

Southern Nuclear Operating Company commissioned a team to evaluate the known

discrepancies in the simulator to determine if the 13 criteria established in 10 CFR

55.45(a), “Operating Tests,” would be challenged. The team was comprised of

representatives from SNC (Operations, Training and Engineering), SCANA (Training) and

Westinghouse (Human Factors Engineering).

ND-15-1333 Enclosure 1, Page 4 of 6 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility

The team examined all Discrepancy Reports (DRs) that were open as of May 15, 2015,

and determined that 101 DRs were relevant to acceptability of one or more of the first nine

(9) criteria of 10 CFR 55.45(a). No DRs were found to be relevant to the last four criteria;

55.45(a)(10) through 55.45(a)(13). The team also determined that no singular DR posed a

challenge to the suitability of the simulation facility for the conduct of operating tests;

however, when considered in the aggregate, 42 of the DRs challenged criterion (3) and (5)

of 10 CFR 55.45(a) (See section 2.1 below for additional details).

In order to ensure the simulator is suitable for the conduct of operating tests, corrective

actions were initiated to resolve the subject 42 DRs. This assessment was communicated

to Westinghouse Electric Company (WEC) and WEC committed to implement

improvements aimed at resolving these issues in a patch deliverable to SNC by August

14, 2015. Based on this commitment, the CAS Aggregate Study Team reconvened on

July 7, 2015 and determined that the proposed changes would be adequate so that the

aggregate impact of the remaining discrepancies would not pose a challenge to any of the

10 CFR 55.45(a) criteria.

On August 14, 2015, WEC delivered a patch to SNC which contained corrections for the

42 items previously identified along with some additional corrections. After performing

Verification and Validation (V&V), 11 were determined to require further

investigation. After confirming the corrections that successfully passed the V&V process,

the CAS Aggregate Study Team reconvened on September 1, 2015, to review the impact

of the 11 outstanding items. The Aggregate Study Team determined that, in aggregate,

the impact of the 11 outstanding items, combined with the improvements in the area of

Alarm Response and the other remaining open items, would not impact the suitability of

the simulator for the conduct of operating tests.

Enclosure 6 contains the Aggregate Study mentioned above. Enclosure 7 contains a list

of the items WEC corrected. Enclosure 9 contains a list of the open DRs as of May 15,

2015.

2.0 Description of Simulator Discrepancies

Simulator discrepancies identified by SNC, other domestic AP1000 simulator owners and

those discrepancies that were issued as Unresolved Issues (UIs) by the NRC were

evaluated for applicability to SNC’s simulator. Discrepancies that were determined to be

applicable were entered into SNC’s Configuration Management System (CMS) Mantis

database as Simulator Change Requests (SCRs). If the SCR could be corrected by SNC,

it was corrected. Those SCRs that could not be corrected were evaluated by performing a

Training Needs Assessment to determine the impact on training. If the Training Needs

Assessment determined that there was an impact on training, a Training Needs Analysis

was performed to determine the extent of effect and to develop mitigation under the

Systematic Approach to Training (SAT) process. If the Training Needs Assessment

determined that there was no impact on training, then the discrepancy was entered into

the global tracking book as a historical record.


2.1 Cumulative Effect of Simulator Discrepancies on Operator Performance

As stated in Section 1.0 above, SNC commissioned a team to determine if, in

aggregate, open discrepancies would present a challenge to the simulation facility’s

suitability for the conduct of operating tests. The results of this study are

documented in the report “Commission Approved Simulator Aggregate Study -

Simulator Training System Deficiency Impact on 10 CFR 55.45” (Enclosure 6).

Initially, the team determined that the aggregate of the open discrepancies would

challenge the ability of licensed operators to respond to simulator scenarios in

normal, off-normal, and emergency conditions based on criterion (3) and (5) of 10

CFR 55.45(a). The team analyzed the items that challenged these two criteria and

determined that the suitability of the simulator for the conduct of operating tests

would not be challenged if 42 of the discrepancies were corrected. SNC requested

Westinghouse Electric Company (WEC) to correct these discrepancies.

Based on WEC’s commitment to correct these items, the CAS Aggregate Study

Team reconvened on July 7, 2015 to determine if the remaining DRs would still

present a challenge to SNC’s ability to conduct an operating examination in

accordance with 10 CFR 55.45. The team concluded that the aggregate impact of

the remaining items would not pose a challenge to any of the 10 CFR 55.45(a)

criteria.

2.2 Evaluation of AP1000 Simulation Facility UIs Issued By the NRC

SNC performed a review of AP1000 simulation facility UIs issued by the NRC

(References 1 and 2). UIs were screened for applicability to the VEGP 3&4

simulation facility. Applicable UIs were entered into SNC’s CMS using NMP-TR-422,

“Simulator Configuration Control Procedure.”

Enclosure 5 contains the results of SNC’s evaluation of AP1000 Simulation Facility

UIs issued by the NRC.

2.3 Simulator Conformance with the AP1000 Plant Design

SNC accepted turnover of the VEGP Units 3&4 Simulator Training System (STS)

from Westinghouse, as described by letter (Reference 3) on December 30, 2014.

Subsequent to the STS turnover, SNC identified simulator discrepancies that were

determined to be plant design issues. Westinghouse retains design authority of the

plant configuration until Unit 3 turnover. Simulator discrepancies that are determined

to be plant design issues prior to Unit 3 turnover will be tracked until the AP1000

plant design changes have been approved. Approved design changes will be

incorporated into the simulation facility in accordance with SNC simulator fidelity and

configuration management programs per 10 CFR 55.46(d).

ND-15-1333 Enclosure 1, Page 6 of 6 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility 3.0 Variances between AP1000 Simulator Facilities

It has been noted that variances exist between the various AP1000 simulators.

As described in this letter and Enclosures, SNC reviews simulator discrepancy reports

from other licensees and the vendor. The item is entered into the Configuration

Management System (CMS) Mantis database and tracked to resolution.

For example, if SNC identifies an issue or is able to duplicate an issue identified on one of

the other AP1000 simulators, SNC will generate an SCR. If it is within SNC’s capability to

do so, SNC will develop and implement a resolution to address the issue. SNC shares

these solutions with WEC. WEC, at its discretion, may immediately distribute the solution

with other AP1000 simulator owners or wait to incorporate the correction as part of a future

software update.

4.0 Conclusion

SNC evaluated all open simulator discrepancies that existed through May 15, 2015 and

corrected discrepancies that, in aggregate, could impact the suitability of the simulators for

the conduct of operating tests. SNC has determined that there is no open simulator

discrepancy, individually or in aggregate, that would challenge the ability of licensed

operators to respond to simulator scenarios in normal, off-normal, or emergency

conditions.

The material SNC is presenting is required for a Commission-Approved Simulation Facility

as defined in 10 CFR 55.46(b). SNC is requesting the Commission’s approval of the

VEGP Units 3&4 simulation facility for use in the administration and conduct of operating

tests in accordance with 10 CFR 55.46(b)(2).

5.0 References

1. NRC Email dated 2015-05-13, Meeting Materials for May 14, 2015- VCSNS 2 and 3

Commission-Approved Simulator - CAS-Summer-RAI 5-7-15_b Redacted,

ML#15133A497

2. NRC Letter dated 2015-07-02, Virgil C. Summer Nuclear Station Units 2 and 3 - Request

For A Commission Approved Simulation Facility - ML15182A097

3. SNC Letter dated 2014-12-30, Vogtle Electric Generating Plant, Units 3 & 4 - Response to

SVP_SVO_002964 and Simulator Training System (STS) Acceptance



ND-15-1333

Enclosure 2

Description of the Components of the Simulation Facility Intended to be Used for Each Part of the Operating Test - 10 CFR 55.46(b)(1)(i)



1.0 Summary Description

The Vogtle simulation facility is comprised of two AP1000 full scope simulators,

designated “3A” and “3B.” Both simulators are referenced to Vogtle Unit 3 and are

intended to be maintained functionally identical. The simulators are licensed to conform to

the requirements of ANSI/ANS-3.5-1998, “Nuclear Power Plant Simulation Facilities for

Use in Operator Training and License Examination,” as endorsed by Revision 3 of NRC

Regulatory Guide 1.149, “Nuclear Power Plant Simulation Facilities for Use in Operator

Training and License Examinations.”

2.0 Functional Description

Instructor-controlled normal plant evolutions, system malfunctions, Component Level

Failures (CLFs) and Local Operator Actions (LOAs) are used to provide simulated plant

performance and failure or degradation of simulated plant systems or equipment. To

achieve this level of functionality, plant systems listed in Table E2-1 are simulated.

3.0 Detailed Description of the VEGP Simulators

The simulation facility design, models, and software are based upon the Westinghouse

“Baseline 7” milestone for Instrumentation and Controls (I&C) design. The Baseline 7

milestone document established a set of requirements to ensure the integrated I&C

system design is consistently implemented within various core I&C platforms and systems.

The Vogtle simulation facility has also been updated with various modifications, in

coordination with Westinghouse as new I&C issues or design changes have been

identified.

The Vogtle Unit 3A and 3B simulators are referenced to Unit 3. Unit 3 is approximately

one year ahead of construction for Unit 4. The only meaningful difference noted between

Unit 3 and Unit 4 design documentation at this time is the switchyard. Unit 3 is tied to the

230kV switchyards. This is shown in the ZBS Ovation screens in both simulators. Unit 4

will be tied to the 500kV switchyard. Presently, this is the only identified difference in Unit

3 and Unit 4.

The VEPG Unit 3 Simulator Training System is tested to the AP1000 design. Simulator

fidelity is maintained in accordance with 10 CFR 55.46(d) as documented by the NRC

(Reference 1). Any significant outstanding discrepancies will be resolved during the

finalization of the AP1000 design and/or initial test program. SNC continues to update the

simulator with corrections to minor programming discrepancies as they are identified and

information associated with these updates is forwarded to Westinghouse for inclusion in

the next baseline update or patch as appropriate.

ND-15-1333 Enclosure 2, Page 3 of 5 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility 4.0 Detailed Description of Simulated Systems

The systems listed in Table E2-1 are simulated. A detailed description of each of these

systems can be found in the VEGP 3&4 Updated Final Safety Analysis (UFSAR), Rev. 4.0.

The systems listed in Table E2-2 are systems that are listed in the UFSAR, but are not

modeled for the reasons stated in the table.

Table E2-1 List of Plant Systems Simulated

System Code

System Title

ASS Auxiliary Steam Supply System

BDS Steam Generator Blowdown System

CAS Compressed and Instrument Air Systems

CCS Component Cooling Water System

CDS Condensate System

CES Condenser Tube Cleaning System

CFS Turbine Island Chemical Feed System

CMS Condenser Air Removal System

CNS Containment System

CPS Condensate Polishing System

CVS Chemical and Volume Control System

CWS Circulating Water System

DAS Diverse Actuation System

DDS Data Display and Processing System

DOS Standby Diesel Fuel Oil System

DTS Demineralized Water Treatment System

DWS Demineralized Water Transfer and Storage System

ECS Main AC Power System

EDS Non Class 1E DC and UPS System

EHS Special Process Heat Tracing System

ELS Plant Lighting System

FPS Fire Protection System

FWS Main and Startup Feedwater System

GSS Gland Seal System

HCS Generator Hydrogen and CO2 Systems

HDS Heater Drain System

HSS Hydrogen Seal Oil System

IDS Class 1E DC and UPS System

IIS Incore Instrumentation System

LOS Main Turbine and Generator Lube Oil System

MES Meteorological and Environmental Monitoring System

MHS Mechanical Handling System

MSS Main Steam System

MTS Main Turbine System

OCS Operation and Control Centers System

PCS Passive Containment Cooling System

PGS Plant Gas Systems

PLS Plant Control System

PMS Protection and Safety Monitoring System

PSS Primary Sampling System


Table E2-1 (continued)

System Code

System Title

PWS Potable Water System

PXS Passive Core Cooling System

RCS Reactor Coolant System

RMS Radiation Monitoring System

RNS Normal Residual Heat Removal System

RWS Raw Water System

RXS Reactor System

SDS Sanitary Drainage System

SFS Spent Fuel Pool Cooling System

SGS Steam Generator System

SJS Seismic Monitoring System

SMS Special Monitoring System

SSS Secondary Sampling System

SWS Service Water System

TCS Turbine Building Closed Cooling Water System

TDS Turbine Island Vents, Drains and Relief System

TOS Main Turbine Control and Diagnostics System

VAS Radiologically Controlled Area Ventilation System

VBS Nuclear Island Nonradioactive Ventilation System

VCS Containment Recirculation Cooling System

VES Main Control Room Emergency Habitability System

VFS Containment Air Filtration System

VHS Health Physics and Hot Machine Shop HVAC System

VLS Containment Hydrogen Control System

VRS Radwaste Building HVAC System

VTS Turbine Building Ventilation System

VUS Containment Leak Rate Test System

VWS Central Chilled Water System

VXS Annex/Aux Building Nonradioactive Ventilation System

VYS Hot Water Heating System

VZS Diesel Generator Building Heating and Ventilation System

WGS Gaseous Radwaste System

WLS Liquid Radwaste System

WRS Radioactive Waste Drain System

WSS Solid Radwaste System

WWS Waste Water System

ZAS Main Generation System

ZBS Transmission Switchyard and Offsite Power System

ZOS Onsite Standby Power System

ZRS Offsite Retail Power System

ZVS Excitation and Voltage Regulation System


Table E2-2 List of Plant Systems NOT Simulated

System Code

System Title Not Simulated Because . . .

DFS Diesel Fuel Offloading System No control room interface

DRS Storm Drain System No control room interface

EFS Communication Systems

The simulator does not model the EFS networking scheme but does provide similar functions for communication systems. The simulator mimics the plant communication systems with a Private Branch Exchange (PBX) phone system for the Training Center.

EGS Grounding and Lightning Protection System No control room interface

EQS Cathodic Protection System No control room interface

FHS Fuel Handling and Refueling System No control room interface

NCS Network Connection System No control room interface

OWS Offsite Water Treatment System No control room interface

RDS Gravity and Roof Drain Collection System No control room interface

RLS Radiochemistry Laboratory System No control room interface

SES Plant Security System No control room interface

TVS Closed Circuit TV System No control room interface

VDS Demineralized Water Treatment Building HVAC System

No control room interface

VGS Auxiliary Boiler Building Ventilation System No control room interface

VIS Transmission Switchyard Ventilation System No control room interface

VNS Switchyard Control Building HVAC System No control room interface

VPS Pump House Building Ventilation System No control room interface

VQS Chlorination Workshop HVAC System No control room interface

VVS Waste Water Treatment Plant Ventilation System

No control room interface

YFS Yard Fire Water System No control room interface

ZFS Offsite Communications System

The simulator does not model the ZFS but does provide similar functions for communication systems. The simulator mimics the plant and offsite communication systems with a PBX for the Training Center.

5.0 References

1. Vogtle Electric Generating Plant Units 3 and 4 - NRC Simulator Inspection Reports

05200025/2015301 and 05200026/2015301, dated April 21, 2015 - ML15113A028



ND-15-1333

Enclosure 3

Description of the Performance Tests for the Simulation Facility and Results of the Tests - 10 CFR 55.46(b)(1)(ii)


ND-15-1333 Enclosure 3, Page 2 of 6 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility 1.0 Summary Description of the Performance Tests for the Simulation Facility and

Results of the Tests

Performance tests were conducted on site, in addition to the earlier factory acceptance

tests performed by the vendor at the vendor’s facility, in order to demonstrate simulator

fidelity. The performance testing concluded that simulator performance met the

requirements of ANS/ANSI-3.5 testing for the current simulator design. On April 8, 2015,

the NRC completed an inspection of the VEGP Units 3&4 simulation facilities to “ensure

that the Vogtle 3A and 3B simulation facilities were being tested in accordance with

ANSI/ANS-3.5-1998, ‘Nuclear Power Plant Simulators for Use in Operator Training

Examination,’” (Reference 1) with no findings of significance. The following is a summary

of the simulator performance test licensing basis, a description of the performance tests

conducted and the test results.

2.0 Detailed Description of the Performance Tests for the Simulation Facility and

Results of the Tests

2.1 Vendor and Other Testing

The VEGP Units 3&4 Simulator Training System was developed by WEC and turned

over to SNC on December 30, 2014. In parallel, WEC continued design finalization

activities which included Human Factors Engineering Validation tasks such as

Integrated System Validation (ISV). The ISV used and exercised the simulator

extensively. The ISV shakedown, pilot and final ISV testing identified a number of

integration issues. The NRC had requested licensees to report and assess these

issues for NRC consideration. Some of the issues reported by other licensees were

not able to be duplicated in the performance tests on the VEGP Units 3&4 STS.

Refer to Enclosure 1 Section 3.0 for a discussion on simulator variations.

A preliminary review of ISV testing identified twenty one potential Priority-1 Human

Engineering Discrepancies (HEDs). The NRC had requested licensees to report and

assess those potential HEDs. The assessment results, including resolutions, were

provided to the NRC. The NRC did not fully accept some of the resolutions and

indicated that they would need additional information (Reference 2). SNC reviewed

the information requested by the NRC and developed resolutions which are provided

in Enclosure 8.

2.2 Simulator Performance Testing Licensing Bases

Simulator licensing bases are described in Vogtle 3&4 UFSAR Chapter 1, Appendix

1A. Per the UFSAR, Vogtle conforms to section C.1 of Regulatory Guide 1.149,

Revision 3 “Nuclear Power Plant Simulation Facilities for Use in Operator License

Examinations.” Operator Licensing examinations are conducted on a simulator

meeting the applicable requirements of ANSI/ANS-3.5-1998.” This Regulatory Guide

endorses ANSI/ANS 3.5-1998 “Nuclear Power Plant Simulation Facilities for Use in


Operator License Examinations.” Section 4.4.3 of the ANSI/ANS 3.5 describes the

simulator performance testing.

2.3 VEGP Units 3&4 Simulator Performance Testing Description

Simulator performance testing is made up of operability testing and scenario-based

testing. Simulator performance testing is performed in a fully integrated mode of

operation. The test procedures are documented in Vogtle procedures NMP-TR-422-

006, “Plant Vogtle 3-4 Simulator Testing Instruction,” and NMP-TR-422-006-001,

“Simulator Configuration and Performance Criteria Instruction.” These tests are

based on the ANS-3.5-1998 standard. The test cases included:

1. Simulator Operability Testing - Simulator Operability Testing is conducted to confirm

overall simulator model completeness and integration. Operability testing:

Is intended to demonstrate overall simulator model completeness and

integration

Includes simulator transient performance for a benchmark set of transients as

shown below

Item Title Test Type

1 Manual Reactor Trip Transient

2 Simultaneous trip of Main Feedwater Pumps Transient

3 Simultaneous Trip of all Feedwater Pumps Transient

4 Simultaneous closure of All Main Steam Isolation Valves Transient

5 Simultaneous trip of All Reactor Coolant Pumps Transient

6 Single Reactor Coolant Pump Trip Transient

7 Main Turbine Trip Without a Reactor Trip Transient

8 Maximum Rate Power Ramp Transient

9 Maximum Size Reactor Coolant System Rupture with Loss of Offsite Power

Transient

10 Maximum Size Unisolable Main Steam Line Rupture Transient

11 Slow Primary System Depressurization to Saturated Condition (Pzr Safety)

Transient

12 Slow Primary System Depressurization to Saturated Condition (ADS) Transient

13 Maximum Design Load Rejection Transient

Includes real time and repeatability tests


14 Computer Real Time Test Real Time

15 Simulated Limits Exceeded Test Real Time

16 Repeatability Test Repeatability

Includes simulator steady-state performance tests


17 Steady State Performance at 50% Power Steady State




Includes the following normal evolutions


20 Plant Startup from Cold to Hot Standby Normal Evolution

21 Nuclear Startup from Hot Standby to Rated Power Normal Evolution

22 Reactor Trip with Recovery to Rated Power Normal Evolution

23 Plant Shutdown from Rated Power to Cold Shutdown Normal Evolution

24 Surveillance Testing Normal Evolution

Includes the following core tests


25 Shutdown Margin Determination Core

26 Core Reactivity coefficients Test Core

27 Isothermal Temperature Coefficient Core

28 Control Rod Worth Test Core

2. Malfunction testing is conducted on an as needed basis. Malfunction testing was

conducted to gather base line data.


29 Steam Generator Tube Rupture Malfunction

30 Loss of Coolant Outside Containment Malfunction

31 Large Break Loss of Coolant Accident Malfunction

32 Small Break Loss of Coolant Accident Malfunction

33 Loss of Instrument Air Malfunction

34 Loss of IDS Division A Instrument Busses Malfunction

35 Loss of IDS Division B Instrument Busses Malfunction

36 Loss of IDS Division C Instrument Busses Malfunction

37 Loss of IDS Division D Instrument Busses Malfunction

38 Loss of Offsite Power with Loss of Diesel Generators Malfunction

39 Loss of Electrical Distribution Bus ES-1 Malfunction






45 Loss of Electrical Distribution Bus EK-11 Malfunction










55 Loss of EDS Instrument Buses Malfunction

56 Loss of IDS Division B and C 72 Hour Instrument Busses Malfunction


58 Loss of Condenser Vacuum Malfunction

59 Loss of Condenser Level Control Malfunction

60 Loss of Service Water Malfunction

61 Loss of Shutdown Cooling Malfunction



62 Loss of Component Cooling Water Malfunction

63 Loss of Normal Feedwater Malfunction

64 Loss of All Heat Sinks Malfunction

65 Loss of Division A PMS Malfunction

66 Loss of Division B PMS Malfunction

67 Loss of Division C PMS Malfunction

68 Loss of Division D PMS Malfunction

69 Rod F06 Stuck Malfunction

70 Rod B06 Uncouples Malfunction

71 Rod G07 Drops Malfunction

72 Misaligned Rod Malfunction

73 Inability to Drive Rods Malfunction

74 Fuel Clad Failure Malfunction

75 Main Generator Trip Malfunction

76 Inadvertent Operation of Core Makeup Tanks at Power Malfunction

77 Inadvertent Actuation of Passive Residual Heat Exchanger at Power Malfunction

78 Increase in RCS Inventory Malfunction

79 Failure of Pressurizer Pressure Control Malfunction

80 Main Steam Line Break Inside Containment Malfunction

81 Main Steam Line Break Outside Containment Malfunction

82 Main Feed Line Break Inside Containment Malfunction

83 Main Feed Line Break Outside Containment Malfunction

84 Failure of Power Range Nuclear Instrument Malfunction

85 Failure of Intermediate Range Nuclear Instrument Malfunction

86 Failure of Source Range Nuclear Instrument Malfunction

87 Failure of the Alarm Presentation System Malfunction

88 Anticipated Transient without SCRAM without DAS Malfunction

89 Anticipated Transient without SCRAM with DAS Malfunction

3. Simulator Scenario-Based Testing (SBT) – The VEGP Units 3&4 simulator facility is

committed to the SBT methodology described in the 1998 ANS-3.5 standard as

endorsed by Reg. Guide 1.149 Rev. 3 and in NEI 09-09. SBT is the parallel testing

and evaluation of simulator performance while instructors validate NRC Initial license

examination scenarios, licensed operator requalification annual examination

scenarios, and scenarios used to satisfy the reactivity control manipulation

requirements for license candidates in 10 CFR 55.31 (a)(5). As instructors validate

satisfactory completion of training or evaluation objectives, procedure steps and

scenario content, they are also ensuring satisfactory simulator performance in

parallel, not series, making the process an “online” method of evaluating simulator

performance. SBT is conducted to ensure the simulator was capable of producing

the expected “reference unit” response to satisfy predetermined learning or

examination objectives by utilizing the existing training and examination scenario

validation process. The term, “reference unit,” as used above, refers to the AP1000

plant design since both of the VEGP units are still under construction.


2.4 Simulator Operability Test Results

Operability tests were completed satisfactorily in 2014 with one test (Reactor Trip

Recovery) deviation. A Condition Report was initiated to document the deviation.

The performance tests resulted in the documentation of SCRs. These discrepancies

were captured in the final test report and the Simulator CMS SCR database.

On April 8, 2015, the NRC completed an inspection that included a selection of

simulator test procedures and test records. Based on the results of that inspection,

no findings of significance were identified (Reference 1).

3.0 Maintenance of Simulator Fidelity

As mentioned in Enclosure 2, the simulators are maintained in conformance with the

requirements of ANSI/ANS-3.5-1998, “Nuclear power Plant Simulation Facilities for Use in

Operator Training and License Examination,” as endorsed by Revision 3 of NRC

Regulatory Guide 1.149, “Nuclear Power Plant Simulation Facilities for Use in Operator

Training and License Examinations.”

The NRC performed an inspection of the VEGP Units 3&4 simulation facility on April 8,

2015. The inspection included a review of SNC’s programs and processes related to

continued assurance of simulator fidelity in accordance with 10 CFR 55.46(d). The

inspection yielded no findings of significance and determined that SNC’s programs to

assure continued simulator fidelity were adequate (Reference 1).

4.0 Summary Conclusion

Simulator operability tests and simulator scenario-based tests were conducted and

completed with no major differences identified between the AP1000 plant design and the

simulator. As a result, SNC believes NRC examiners should be able to make pass-fail

judgments with confidence as required by Reg. Guide 1.149.

5.0 References

1. Vogtle Electric Generating Plant Units 3 and 4 - NRC Simulator Inspection Reports

05200025/2015301 and 05200026/2015301, dated April 21, 2015 - ML15113A028

2. Virgil C. Summer Nuclear Station Units 2 and 3 - Request for a Commission-Approved

Simulation Facility, Dated July 2, 2015 - ML15182A097



ND-15-1333

Enclosure 4

Description of the Procedures for Maintaining Examination and Test Integrity Consistent with the Requirements of 10 CFR 55.49 - 10 CFR 55.46(b)(1)(iii)



1.0 Summary Description of the Procedures for Maintaining Examination and Test

Integrity Consistent with the Requirements of 10 CFR 55.49

Security for examination development and implementation is accomplished with NMP-TR-423, “Regulatory Exam Development.” This procedure conforms to the requirements of NUREG-1021, “Operator Licensing Examination Standards for Power Reactors,” which is founded in the requirements of 10 CFR 55.49. The procedure includes:

Door Security Access Control;

Encryption of Initial Condition (IC) sets and Application (APP) and Trigger files;

Disabling of Video Recording Equipment; and,

Physical Security of Examination Material.



ND-15-1333

Enclosure 5

Evaluation of AP1000 Simulation Facility Summary of Unresolved Items (UIs) Issued By the NRC

Redacted (Non-Proprietary)


ND-15-1333 Enclosure 5, Page 2 of 19 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility 1.0 Summary Evaluation of AP1000 Simulation Facility Unresolved Items (UIs) Issued

by the NRC

SNC performed a review of AP1000 simulation facility UIs issued by the NRC (References

1 and 2). UIs were screened for applicability to the VEGP 3&4 simulation facility.

Applicable UIs were entered into SNC’s CMS using NMP-TR-422, “Simulator

Configuration Control Procedure.”

2.0 Detailed Description of SNC’s Process

Simulator discrepancies are captured by the SNC Simulator Change Request

process. The main method SCRs are identified is through direct simulator response

during training, validation, or performance testing. Noticeable differences and

discrepancies in expected simulator response are entered into the SCR CMS database

(Mantis). The Mantis system is used for issue reporting, change management,

tracking/querying issues, software change documentation, hardware change

documentation, and other simulator related administrative issues. Alterations to the

simulator models, simulated I&C, and Design Change Package (DCP) implementation are

all documented via Mantis. SNC also receives simulator discrepancies from VC Summer

and Westinghouse. These issues are examined for applicability to the Vogtle 3 STS and

processed through the SNC simulator configuration management process where

applicable.

During analysis of reported simulator discrepancies, design documents are reviewed in

order to determine the response that should be expected from SNC’s simulator. Design

documentation is the main method of analyzing appropriate response due to the lack of an

operating reference unit.

Westinghouse is informed of any I&C or design issue via the SNC corrective action

process. Model issues are corrected by SNC where feasible and appropriate via the SNC

Fleet SCR process. Vendor issued fixes are also processed through SNC’s corrective

action process. All simulator changes are tested in accordance to the ANSI/ANS-3.5.1998

standard. Verification and Validation are performed to examine the effectiveness of the

simulator repair.

The Simulator Review Committee (SRC) reviews SCR disposition at least quarterly. The

SRC also determines if uncorrected simulator discrepancies introduce negative training to

the licensed operator curriculum. Issues which do not introduce negative training and

remain uncorrected for any given length of time are presented to the students at the

beginning of each segment/class. Issues having the potential to introduce negative

training undergo a Training Needs Analysis to exercise the SAT process to prevent

negative training.

ND-15-1333 Enclosure 5, Page 3 of 19 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility 3.0 Nuances Related to the Resolutions in Table E5 and the Aggregate Study

There are numerous times within the Aggregate Study (Enclosure 6), where it states, “This

issue was dispositioned as unacceptable by the Simulator Review Committee (SRC),” yet

the issue was ultimately dispositioned as acceptable in Table E5-1. The review of these

items by the aggregate study team focused on the impact the lack of designed protective

functions (UI #26) and the invalid indications (UI #15) had on an operator when combined

with all other issues under review. The SRC was focused solely on the impact of the

individual items. However, an understanding of the SRC’s dispositioning process and its

use of the terms “acceptable deviation” and “unacceptable deviation” is beneficial.

Acceptable Deviation - After an issue is entered as an SCR in Mantis and proved valid via

investigation, the issue is evaluated by the SRC, supplemented by Subject Matter

Experts. If the issue screens as an acceptable deviation by the SRC, using the guidance

provided in section 4.2.1.4 of ANSI/ANS-3.5-1998, the item remains in Mantis as a

historical record for reference. A running log of these acceptable deviations is available to

the Operations Instructional staff for reference and is provided to licensed operator

candidates at the beginning of simulator training as a reference.

Unacceptable Deviation - If the issue screens as an unacceptable deviation by the SRC,

then a detailed evaluation of the deviation is conducted via a Training Needs Analysis

(TNA). The TNA will determine if any compensatory actions can be taken to mitigate the

deviation’s impact to students until a software or hardware solution is implemented. If

compensatory actions can be taken or the issue is corrected, the deviation status

becomes acceptable.

4.0 Simulator Review Committee (SRC)

The Simulator Review Committee (SRC) is composed of one member of Operations,

selected by the Operations Director; the Operations Training Manager or designee; and

the Simulator Coordinator or designee.

The SRC is supplemented by additional personnel, as necessary, to serve as subject

matter experts to conduct a Training Needs Assessment (an appraisal by a subject matter

expert of a simulator deviation, deficiency, or modification, and its relative importance to

the operator as required tasks are performed). Additional members of the SRC are

Operations individuals who have completed AP1000 certification training.

Representatives from Engineering attend the SRC when required to discuss plant design

changes and their impact on the simulator.


Table E5-1

SNC Evaluation of NRC UIs and Cross-Reference List

NRC

# {*} NRC UI Description

{*}

Ref

# {*}

SNC SCR

#

SNC Evaluation

1

Subcriticality Critical Safety Function (CSF) alarm block is turning magenta (bad input) intermittently

TO-40 5627

Closed. Fixed with patch WEC provided to SNC on August 14, 2015.

V&V testing was performed successfully.

For V&V, two different tests were conducted under simulator conditions replicating those under which the issue was first identified. The first test was a steady state 100% power run for one hour with no operator action. The second test was conducted for an RCS leak. During both tests, the Mode 1/2 Critical Safety Function status wall panel display was observed and [ ]a,c.

2 Control rods rejecting to manual {Rod Control Urgent Alarm}

TO-45 5808

Closed. Fixed with Patch Version 1.0.1.

For Validation and Verification (V&V), a loss of ES-1 for six different plant conditions was conducted. During each tested condition, rod control remained in automatic.

SNC has not seen this since the patch was installed.

3

Wall Panel Information System (WPIS) is cycling between different displays

{Mode 2 is procedurally called when all AO bank rods are off the bottom. Currently, PLS Auto Plant Mode selector changes from Mode 3 to Mode 2 when the RTBs are closed (P-3 is cleared).}

TO-52 6144



For V&V, a General Operating Procedure startup was conducted. Simulator conditions under which the issue was first identified were replicated as nearly as possible. When all banks of AO rods indicated 1 step, it was verified that the auto plant mode selector NAP changed to Mode 2. This test was repeated 3 times with the same results.

4 Feedflow oscillations TO-54 and 58

6151

SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.

The Aggregate Study and the SRC dispositioned this issue as acceptable.

These feedwater oscillations are associated with Startup Feedwater during shutdown conditions with little steam and feed demand.

This is in accordance with the current AP1000 plant design. Engineering calculations of expected flow characteristics were used to establish the initial controller tuning values. The Simulator controller tuning values have been established at these calculated values.



NRC


{*}

Ref

# {*}

SNC SCR

#

SNC Evaluation

During shutdown conditions with low steam and feed demand, Start-up Feedwater Control Valves (SFCVs) are in automatic. Under these conditions, the SFCVs cycle over the entire operating range in short periods of time resulting in start-up feedwater oscillations even though Steam Generator Water Level (SGWL) remains steady in the program band with no noticeable perturbations. If an operator deems it necessary, current operational procedures do allow placing the SFCVs in MANUAL and maintaining SGWL in that mode of operation. At some point, operators would contact maintenance to troubleshoot and repair the cause of the oscillations.

Since the system is performing its design function of maintaining SGWL, SNC has determined that this issue does not impact the suitability of the simulator for the conduct of operating tests.

SNC does consider cycling of the SFCVs and the resulting feedflow oscillations as an undesirable condition for an operating plant. WEC is aware of this issue and plans are in place to obtain more precise tuning data during hot functional testing. Hot functional testing will provide as-built flow characteristics and more accurate controller tuning values. Once hot functional testing has been completed, the expectation is that SGWL will continue to be maintained in the required operating band with the SFCVs in AUTO and with no observable feedflow oscillations.

5 Unexpected hotwell low level during trip recovery

TO-89 5987

Closed. Fixed with patch, Version 1.0.9.


For V&V, hotwell makeup valves were fully opened at 100% power. Simulator conditions under which the issue was first identified were replicated as nearly as possible. Makeup flow was verified to be greater than the required value per AP1000 design documents.

6 Modeling of baseline vs design certification configuration

TO-96 None


This issue relates to how a secondary trip function is administered during training scenarios. During the evaluation of operators, any scenario affecting the operation of the turbine control valves, the instructor will either have all automatic trips fail, requiring operators to take manual action to trip the turbine, OR insert a spurious trip of the turbine, requiring operators to take action based on the sudden loss of turbine load. In either case, the issue is transparent to the operators.

The Cause and Effect document for TOS02 on the Instructor Station references future design information. This is an administrative issue on the Instructor Station. The functionality can be updated by WEC when the appropriate design resolution is available. The current modeling for TOS02 does prevent an automatic turbine trip. When evaluating operators, SNC will have all automatic turbine trips fail so as to require operators to take manual actions in accordance with plant procedures.

7 Control rods reject to manual

TO-101 and 104

5659

Closed. Fixed With patch, Version 1.0.1.




NRC


{*}

Ref

# {*}

SNC SCR

#

SNC Evaluation

SNC has not seen this since the patch was installed. For Validation and Verification (V&V), the simulator was placed in run for 14 hours during a steady state test created for ANSI testing and in that 14 hour period, the rods did not reject to manual. A manual turbine trip was inserted [

]a,c.

8

Moisture Separator Reheater (MSR) valve response is incorrect and causes a reactor coolant system (RCS) temperature transient

TO-128 5618


This is not in accordance with the current AP1000 plant design. The implementation of the AP1000 design requirements into the I&C control scheme was incorrect. The simulator correctly models the installed I&C control scheme. This issue will be corrected when the implementation of the I&C control scheme is updated by WEC in accordance with SNC’s configuration management program.

[

]a,c.

[ ]a,c. This does not affect the performance of simulator

operations as the alignment of the steam dump control system, [ ]a,c, is procedurally controlled during the load reduction and the lineup is established prior to reaching these lower turbine load conditions.

The time period that the incorrect control signal is in effect will vary dependent upon the down-power rate. Throughout this control sequence, operators are performing other actions that are directed by the controlling procedure. The controlling procedure will direct the use of the SOP to verify the proper position of the valves. However, when this direction is provided the valves have already cycled to the closed position and are in the expected position when checked. SNC has developed an APP file to override the controller and close the valves at the required turbine load. This results in the controller error being transparent to the operator.

SNC will continue to use this APP file until a permanent correction is provided by WEC.

9

Aux steam pressure not meeting design requirements

{GSS Header pressure will not maintain pressure as required}

TO-131 5609



For V&V, testing was performed at 100% power. Initial simulator conditions under which the issue was first identified were replicated as nearly as possible during V&V testing. Gland Sealing Steam header pressure maintained [ ]a,c psig under full power conditions.

This issue has not been noted on the simulator since the correction was implemented.

10 Unexpected Main Turbine System alarm

1411-03 5722 Closed. Fixed with patch WEC provided to SNC on August 14, 2015.




NRC


{*}

Ref

# {*}

SNC SCR

#

SNC Evaluation

at power For V&V, turbine load was lowered [ ]a,c. Simulator conditions under which the issue was first identified were replicated as nearly as possible. A second test was then conducted where turbine load was lowered [ ]a,c. During both turbine load reductions, APS was monitored. The unexpected MTS alarm was not received during either test.


11

Rod control urgent failure on loss of EK-12 appears inconsistently without loss of power

1501-08 6726


The plant response identified by this DR is per current design.

SNC has confirmed that the indications and automatic actions are consistent with expected plant response per current rod control software design.

Results of investigation/observations:

[ ]a,c. This cabinet provides the processing and amplification of half the self-powered flux

detector signals. These signals are provided to [ ]a,c When power

is lost to this cabinet it [ ]a,c. The inconsistent results are due to different initial conditions when the event occurs.

[

]a,c. The small variations in plant parameters that are part of normal fluctuations during steady state operations will ultimately change the response to this power loss.

The cause and effect in this situation is per current plant software design and therefore replicates expected actual plant response. Operators will take action accordingly if this were to happen in the plant and therefore are taking the same actions if this event occurs in the simulator. The expected operator action for automatic rod motion that is not expected or is occurring due to a detector failure is to take rod control to manual. This expectation is consistent with fleet expectations.

If/When WEC develops a plant design change, it will be applied per current configuration control management procedures.

12 Axial Offset (AO) rods move inconsistently between tests

1502-10 5585

5659

Closed. Fixed in patch, Version 7F8.1.0.1.


These issues were corrected by RITS 41468 which was delivered in a patch from WEC. The patch was tested on the SDS under conditions that attempted to duplicate the conditions that existed at the time the issue was



NRC


{*}

Ref

# {*}

SNC SCR

#

SNC Evaluation

first identified. The patch was implemented in November of 2014. The patch resolved both issues and the SDRs were closed.

AO rods randomly rejecting to manual has not been observed since the implementation of this patch.

13 RCS wide range pressure dropped from 1400 to 700 psig

1503-03 and 04

6741

This issue was closed as invalid.

An investigation determined that this is the expected plant response to proper operation of the plant passive cooling capabilities.

This observation was made during a simulator scenario which was evaluating operator response to a Loss of Coolant Accident (LOCA). The dynamic conditions of the plant at the time were that plant pressure was lowering and the Pressurizer (PZR) had completely emptied. This resulted in the reactor vessel coolant conditions [

]a,c during their re-creation of this event.

The continued cooling, [

]a,c.

14 Alarm avalanche 1503-16

HED #14

5612

5813



SNC has determined that this issue no longer impacts the simulator’s suitability for the conduct of operating tests.

The large volume of alarms, often referred to as the “Alarm Avalanche”, was significantly lowered due to the combination of WEC’s alarm prioritization project and SNC’s use of the APS “Consequence” feature.

The alarm prioritization project by WEC led to a reevaluation of the alarm points and the priority assigned to each. [

]a,c and response but do not require immediate attention as they did previously.

[ ]a,c. To date; SNC has developed 8 specific consequence files based on

identified transients where the use of the consequence logic has been proven beneficial. An example of this is that [ ]a,c. Those alarms can be set to populate the Consequence tab on APS vice appearing on the current tab. The consequence alarms can be reviewed at any time by selecting the correct tab or by turning off individual consequence functions.



NRC


{*}

Ref

# {*}

SNC SCR

#

SNC Evaluation

Through the combination of these two efforts the overall number of audible alarms received during transients is reduced to those [ ]a,c. This removes a major distraction from operators and allows efforts and attention to be focused upon monitoring and controlling the plant.

Examples of alarm reduction: Reactor Trip:

[ ]a,c

[ ]a,c

Loss of offsite power concurrent with main generator trip:

[ ]a,c

[ ]a,c

This item was also identified during the ISV. See Enclosure 8, HED #14.

15

Inconsistent VRS and VHS radiation monitor indications on a loss of process flow

TO-75 and 76

5828

5914


Two separate V&V tests were performed successfully.

Two tests were conducted to verify response of VRS and VHS. The first test was conducted at 100% power. ECS-ES-1 was de-energized. It was verified that [ ]a,c and no alarm was received upon loss of power. This test was completed successfully.

The second test was conducted at 100% power. Test personnel ensured [ ]a,c. This test

was completed successfully.

16 BEACON operability TO-102 5583


This issue has been determined to be a simulator I&C implementation issue. The issue is not an AP1000 I&C design issue, nor is it a modeling issue.

The ability of BEACON to perform its intended function is directly related to the functionality of [

]a,c. In the event that BEACON is not functional, operators are required to carry out actions in specific Technical Specifications.

For training scenarios where it is desired to fail BEACON (or one of the inputs to BEACON), OPDMS, as it is currently implemented on AP1000 simulators, fails to alert operators that it is no longer operable.

Currently, the status of BEACON is passed to students by instructors when OPDMS displays an incorrect operational status. Therefore, the lack of the ability of BEACON to determine its operability status when the BDP NAP provides a BAD quality signal to BEACON does not impact the suitability of the simulator for the



NRC


{*}

Ref

# {*}

SNC SCR

#

SNC Evaluation

conduct of operating tests.

17

Inconsistent navigation to Protection and Safety Monitoring System (PMS) mimics in Ovation

1504-02 6670


PMS mimics in Ovation have several graphics [ ]a,c, you will be

changed to PMSA.

The PMS mimic in Ovation is an operator aid and not needed for plant operation or PMS actuations. Therefore, this issue does not impact the conduct of operating tests.

Operators are trained to apply Human Performance (HU) tools when operating the plant, including changing from one Ovation screen to another.

CR 10070361 for WEC resolution.

18

Confusing PMS status display

{Stage 3 ADS box unused on PMS Divisions C and D Display}

TO-122 5619

This issue was closed as invalid.

An investigation determined that this is in accordance with the AP1000 design and that this is the expected plant indication.

The ADS Summary graphic provides the following possible indications in regards to ADS Actuation:

[



NRC


{*}

Ref

# {*}

SNC SCR

#

SNC Evaluation

]a,c, just before the sheet navigation flag for the respective valve control sheet.

For this reason SNC has determined that this SCR is not valid. These indications are the designed indications and therefore will be the same indications available to and require the same actions from an operator.

19

Unexplained Steam Generator (SG) level rise following trip of all Reactor Cooling Pumps (RCPs)

1502-03 6471

SNC evaluated this item and determined that the simulator is modeling the AP1000 plant design.

SNC has closed this issue.

The rise in level is due to the downcomer dropping the last bit of water it has as it goes to near dryout conditions.

During AP-OPS-T-004, “Trip of ALL RCPS,” a level rise of [ ]a,c (with a slight delta between SGs). At this time all inputs and outputs from the SGs have been isolated for over [

]a,c

A detailed evaluation of the conditions internal to the SGs determined the level rise is due to the last bit of water dropping out of the downcomer as it enters near dryout conditions. WEC agreed with this. Therefore, the simulator modeling is correct and this is a correct plant response for the transient.

20

Pressurizer (PZR) Level went down in 2 of 3 training scenarios with the leak through the PZR safety

1502-08 6484

Closed. Could Not Duplicate.

This issue was reported at a non-SNC AP1000 simulator.

The same initial conditions were established on SNC’s simulator. This was facilitated by using the same Simulator APP file that was used by the discovering simulator group. The APP file provides the ability to save a set of malfunctions such that the same scenario can be reset and the same exact malfunctions can be re-inserted. SNC used this file during three trial runs in an attempt to duplicate this issue. This ensured the same exact faults were used for the diagnosis. SNC monitored Wide Range and Narrow Range pressurizer level response and observed no significant difference in the indications. The maximum difference (delta %) between the three runs for wide range was [ . ]a,c Since these values represent no significant difference given the indication response and since this does not indicate a leak through the PZR safety, this issue was closed.

21

Over power control permissives did not respond to steam leak as designed

1502-09 6122



V&V testing was conducted from 100% power under initial simulator conditions similar to those that existed at the time the issue was identified. Turbine control was placed in MWe IN with rods in automatic. Both Power Operated Relief Valves (PORVs) were manually opened to 100% for an excess steam demand. [

]a,c. V&V testing was completed satisfactory.



NRC


{*}

Ref

# {*}

SNC SCR

#

SNC Evaluation

22

PZR Water Level response during Safety valve malfunctions has variations in tests

1502-12 6484

Closed. Could Not Duplicate.

The APP file provides the ability to save a set of malfunctions such that the same scenario can be reset and the same exact malfunctions can be re-inserted. SNC used this file during three trial runs in an attempt to duplicate this issue. This ensured the same exact faults were used for the diagnosis. SNC monitored Wide Range and Narrow Range pressurizer level response and observed no significant difference in the indications. The maximum difference (delta %) between the three runs for wide range was [

]a,c Since these values represent no significant difference given the indication response and since this does not indicate a leak through the PZR safety, this issue was closed.

23

During Load Rejection events, Load Unbalance response is inconsistent causing noticeable deltas in several key parameters

1502-13 6483



V&V testing was conducted with three identical tests under initial simulator conditions similar to those that existed at the time the issue was identified. Each test initiated a 100% load rejection and graphed the response of the turbine intercept valves. All valves responded identically throughout all three tests. The problems that were initially reported under this issue were not observed during these tests. V&V testing was completed satisfactory.

24

TCS heat transfer characteristics through the H2 coolers are unrealistic

1503-33 6181

SNC has determined that this issue is acceptable and that it does not impact the simulator’s suitability for the conduct of operating tests.

The simulator is correctly modeling the present plant design.

This issue is based on the inability of the TCS temperature control valve, controlling H2 cooler temperature, to establish a steady state position. Corrective Action Program And Learning (CAPAL) 100221278 was sent to CB&I for a design or I&C change. CB&I responded by informing SNC that the heat exchanger is too large. This corresponds to the Simulator response. Because the heat exchanger is too large, the temperature control valve is forced closed to prevent over-cooling. Because the valve is fully closed, when the temperature reaches a point where the valve needs to open, the response time is too slow and temperature doesn’t begin to lower before a high temperature alarm is received. As the valve continues to open, temperature turns and begins to lower, but the temperature drop occurs faster than the valve can respond and the valve is once again forced closed. However, even in steady-state conditions a small modulation of the temperature control valve will result in temperature lowering. CB&I will need a better control scheme or an actual heat exchanger design change.

This cycling of temperature from a low to a high value occurs over approximately 2 hours (from points where the over-cooling has occurred and high temperature alarm is received). A mitigating strategy has been put in place to establish the initial conditions and then save those initial conditions immediately after the temperature has been lowered. This provides the maximum amount of time before a high temperature alarm is received. Most training scenarios are either less than 2 hours OR result in placing the plant in a condition where H2 cooling is no longer required prior to this 2 hour window expiring. [



NRC


{*}

Ref

# {*}

SNC SCR

#

SNC Evaluation

]a,c.

25 "Instrument Air" alarm tile has no points assigned to it

1504-01 6669

SNC has determined that this issue is invalid. The issue is closed.

The Alarm Presentation System (APS) has multiple tiles where alarm points are associated with each tile. This issue was initially discovered at an alternate domestic AP1000 simulator. SNC confirmed that APS has associated alarm points with Instrument Air and therefore, this is not an issue at the SNC simulation facility.

26

Control logic functions associated with solid plant operations do not function as described in the design documentation

1504-09 5968



For V&V testing, two testing scenarios were used to verify the proper function of the protective features under simulator conditions replicating those that existed when the issue was first identified. One test was performed to verify the high pressure related functions and another for the low pressure related functions. The observed responses were verified correct per AP1000 design documentation.

The first test was conducted at Mode 5 with the RCS in solid pressure mode. [

]a,c. The test was reset to initial conditions and the “B” CVS makeup pump trip setpoint tested. Both tests resulted in satisfactory V&V.

The second test conditions for V&V were to verify RCP trip setpoints on low pressure while in solid plant operations. [

]a,c. The test was reset to initial conditions and repeated with the same result. Both tests resulted in satisfactory V&V.

27

Control rods rejecting to manual during Anticipated Transient Without Scram (ATWS)

TO-47 None

SNC was unable to duplicate this issue.

A number of Rod Control issues have been observed at SNC and evaluated as a whole. Multiple ATWS scenarios have been reviewed and rods did not reject to manual. All rod control issues at SNC have been associated with one (1) SCR. Refer to SCR 5659 for SNC testing of rod control issues. Over 10 test runs were completed under various plant conditions to verify rods reject to manual. Following patch V3.R1.7F8.1.0.1, five different tests were run. A steady state test at [ ]a,c was allowed to run for [ ]a,c to verify rods did not reject to manual. All tests were completed satisfactory.

Since this particular issue was never observed or duplicated, SNC did not create an SCR.

28 Steam dump capacity appears to be larger than expected

1410-07 6830

SNC evaluated this issue and has developed and implemented a solution that results in turbine bypass valve flow being simulated per design.

V&V testing was conducted under initial simulator conditions similar to those that existed at the time the issue was identified. Main Steam Header pressure and Turbine Bypass valve response was monitored during the spurious trip of the main generator circuit breaker from 100% power. Steam flow through turbine bypass valve,



NRC


{*}

Ref

# {*}

SNC SCR

#

SNC Evaluation

MSS-V001, was verified correct when the valve was full open. [ ]a,c. V&V testing was

successful.

29

Determine if ventilation system response is correct (VAS, VRS, (VFS) systems)

1501-02 6410

SNC evaluated this issue and determined that it constitutes a plant design issue that does not impact the suitability of the simulator for the conduct of operating tests.

After discussions with WEC and other licensees, the simulator was verified to be functioning per plant design. WEC has determined that a plant design change will be necessary to alleviate the condition. The WEC update is the result of a specialized test that was developed by VC Summer. The test was designed to create a LOCA outside the reactor containment through the letdown flow path. [

]a,c.

This condition would only exist for [ ]a,c before the VFS fans trip on low flow. Operators will respond to alarming conditions per the ARPs associated with Containment (CTMT) pressure which require a flow path to be aligned to the VFS exhaust fans.

30

Following SG dryout, SG Wide {Narrow} Range level does not stay at zero. The Level will oscillate

1502-14 6434

SNC evaluated this item and determined that the instruments are responding as per the AP1000 plant design. This issue has no impact on the suitability of the simulator for the conduct of operating tests.

SNC closed this item as invalid.

SNC discovered that as the compartment pressure rises, the narrow range differential pressure slowly falls. This is what is causing the narrow range level to exhibit a slight rise.

This response is expected per the AP1000 design.

31

SG parameters have unexplained damped oscillation following “Main Steamline Break Outside Containment”

1502-15 6482

Closed. Could not duplicate.

The issue was originally noted on one of two simulators at another licensee’s site. The oscillations were noted around 1800 seconds in a malfunction test. While attempting to mimic the conditions that existed at the time the issue was first discovered, SNC identified that the issue was due to the data collection interval being shortened to 0.5 seconds. SNC shortened the data collection interval to 0.5 seconds, but was not able to recreate the issue using this interval with the same malfunction inserted in either of two test runs. This issue was closed.

32 Difficulty determining CMT actuation

1503-08 and 09

HED #2

5998

SNC evaluated this issue and determined that it does not impact the suitability of the simulator for the conduct of operating tests.

This issue is not an AP1000 design issue nor is it a simulator modeling issue.

HED #2 was driven by inconsistencies in determining whether CMTs were in service by crews during ISV. To

correct this, WEC is issuing procedural changes to the Emergency Operating Procedure network. SNC has received the updated procedures and is processing them in accordance with the normal procedural change



NRC


{*}

Ref

# {*}

SNC SCR

#

SNC Evaluation

process.

Related to Enclosure 8, HED #2.

33 Problems during transfer to remote shutdown room

1503-21 6075


This is a simulator I&C network related issue. It is not an AP1000 plant design issue nor is it a simulator modeling issue.

This issue occurs when simulator servers 216/217 are logged out and the displays are disconnected. The software will crash when an operator attempts to log in. If the drops are logged in and the displays are reconnected, the software will lock the interface.

As a safeguard to prevent this from happening during training, the simulation guide directs the instructor to test the functionality of the transfer switch as part of the scenario set-up.

The strategy to prevent future impact also involves the soft control functions at the instructor station. The booth operator has the ability to use these soft controls to ensure the transfer to the RSR occurs in a manner that is transparent to the operating crew.

SNC also determined that the issue only affects RSR operation. Neither the availability of the RSR or the ability of operators to use or shift operations to the RSR is a requirement for licensed operators.

34 (RNS) system over-pressurization

1503-13

HED #11 None

SNC evaluated this issue and determined it to be invalid at the SNC site simulator.

SNC evaluated this issue against its procedures and determined that its procedures were adequate, providing sufficient detail and guidance to prevent an operator or crew from performing this action.

Specifically:

In accordance with 3-RNS-SOP-001 version D 0.3, [ ]a,c (Attachment 4 section 3.0). [

]a,c which is inside containment.

By following regulatory requirements, management expectations to follow procedures and SNC’s “Conduct of Operations” procedure, an event where the need for an interlock on RNS-V061 should not occur. For this event to occur, an SNC operator or an operating crew would have to intentionally violate one or more procedures.

Therefore, SNC has determined that this issue does not affect operator training or the development of exams.


35 CCS low surge tank 1503-15 None Closed. Fixed with patch WEC provided to SNC on August 14, 2015.



NRC


{*}

Ref

# {*}

SNC SCR

#

SNC Evaluation

level alarm priority is incorrect

HED #13 V&V testing was performed successfully.

The priority of the CCS surge tank low level alarm was raised to [ ]a,c which is commensurate with the effects of a low tank level upon the plant.


36

RNS pump does not restart on Diesel Generator (DG) Sequencer

1410-09 6000



For V&V, the simulator was initialized in a Mode 4 initial condition with both RNS pumps running under conditions that attempted to duplicate the conditions that existed at the time the issue was first identified. The power supplies for each RNS pump were de-energized individually. The RNS pumps loaded onto the diesel in the proper load sequence.


37 EDS battery performance

TO-04 5679



SNC replicated the initial conditions that existed at the time the issue was first identified and re-tested the EDS battery performance. The V&V test consisted of a manual turbine trip, loss of offsite power, and failure of the diesel generators to load their respective busses. [

]a,c.

38

Main Steam System (MS) radiation] monitors do not respond during Steam Generator Tube Rupture (SGTR)

TO 89

{TO-10} 5682

SNC evaluated this item and determined that the detectors are responding per the AP1000 plant design.

SNC closed this item as invalid based on WEC input.

The RMS is functioning correctly as designed. The simulator results are correct.

There two sets of radiation detectors on each Main Steam Line (MSL), SGS-RE026B/RE027B and SGS-RE026A/RE027A. SGS-RE026B/RE027B are for detecting low level radioactivity such as a primary-to-secondary leak. SGS-RE026A/RE027A are for detecting high levels of radioactivity during post-accident conditions. Low level primary-to-secondary leakage would only be detected by SGS-RE026B/RE027B not SGS-RE026A/RE027A. If a SG tube rupture does not emit large concentrations of activity it will only be detected by RE026B/RE027B.

The measuring range for SGS-RE026A/RE027A rad monitors is consistent with operating plants and the DCD as dictated by Reg. Guide 1.97. It is a Reg. Guide 1.97 Variable Type E with a dictated range of 1E-1 to 1E+3 µCi/cc for the purpose of measuring noble gas effluent releases. NUREG-0737 and Reg. Guide 1.97 mandated plants add high range detectors on the MSL for the purpose of measuring accident level (magnitude)



NRC


{*}

Ref

# {*}

SNC SCR

#

SNC Evaluation

effluent releases from the MSL relief or atmospheric dump valves for the purpose of evaluation off-site dose releases. SGS-RE026A/RE027A are intended for accident (high range) measurements (Range: [ ]a,c µCi/cc). SGS-RE026B/RE027B are intended for low range measurements (Approximate Range: [

]a,c µCi/cc). SGS-RE026B/RE027B were added to the design to support Tech Spec 3.4.7, RCS Operation Leakage, specifically to address the tech spec limit for detecting 150 gpm per day per SG resulting from primary-to-secondary tube leakage.

Note, not all operating plants have N-16 detectors on their MSLs and cannot detect primary-to-secondary leakage via on-line radiation measurements. They use alternate indications.

In summary, SGS-RE026B/RE027B are for activity or low range measurements. SGS-RE026A/RE027A are for post-accident measurements. The measurement range is dictated by Reg. Guide 1.97. The AP1000 RMS present design for the MSL is in compliance with the licensing basis.

39

When Containment Air Filtration System had no flow, VFS-RY102 alarmed for high iodine

1503-25 6192

5914


V&V testing was performed successfully and the correction was subsequently incorporated into simulator load V3.R1.7F8.1.1.0, which was deployed on August 29, 2015.

40

As the licensee notes in their RAI response, the computer support applications provided by (NAPs) would not be used for Job Performance Measures because they do not assess the applicant’s knowledge. Calculations would be performed manually. This is why many of the discrepancies were considered to be not significant. However, NAPs provides data to the operator during event diagnosis and response. Given the number of NAPs


WEC provided a patch to SNC on August 14, 2015, that included corrections to Nuclear Applications (NAPs). SNC conducted V&V testing for each of the corrections under simulator conditions replicating those under which each issue was first identified.

The following four NAPs corrections successfully passed V&V testing:

1. The Plant Mode Application automatically updates plant mode from [ ]a,c. V&V testing was performed with

similar conditions to those when the issue first identified. V&V testing was satisfactory.

2. The Redundant Sensor Algorithm for Power Range Nuclear Power was updated to [

]a,c. SNC determined that V&V testing was satisfactory.

3. The Inverse Count Ratio application was corrected for proper response of the Intermediate Range 1/M plot. [



NRC


{*}

Ref

# {*}

SNC SCR

#

SNC Evaluation

discrepancies the staff concludes that they could impact operator workload in an inconsistent manner. The staff concludes that there needs to be a reduction in the number of NAPs related discrepancies including those already identified as significant.

]a,c. V&V testing was satisfactory.

4. The Critical Shutdown Safety Function application was corrected to properly display whether a cooldown or heatup was uncontrolled on the Mode 5/6 Critical Safety Function WPIS display. V&V testing was conducted under similar initial conditions to when the issue was first identified. [

]a,c. After waiting for RCS temperature to

stabilize, the “Control HU or CD” was no longer illuminated. V&V testing was satisfactory.

The following two NAPS corrections failed V&V testing: SNC evaluated this issue and determined that it does not impact the suitability of the simulator for the conduct of operating tests.

These three issues were evaluated individually and in aggregate by members of the team that performed the initial Aggregate Study using the same evaluation criteria as before. The team determined that these items do not substantially impact the simulator’s suitability for the conduct of operating tests for the reasons given at the end of each issue.

1. The Leak Rate Monitoring Application was updated as part of this patch. After performing V&V testing, SNC determined that the update was not successful. The operators were still unable to perform a leak rate determination with the NAP. The V&V test consisted of operator performance of an RCS and Main Steam Leak Determination Surveillance in accordance with the surveillance procedure. [

]a,c. The surveillance was unable to be performed. This V&V test failed.

The Leak Rate Monitoring Application is informational only and does not drive any alarms based upon the calculated leakage. For this reason, any leak rate calculations would have to be performed manually per plant procedures vice using the NAP calculated values.

2. Updates to the Time to Boil indications were included as part of this patch. After performing V&V testing, SNC personnel determined that the update was not successful. Time to Boil indication on the Mode 5/6 Primary trend WPIS was observed to be displayed in exponential minutes for the RCS time to boil. The same was true for the Spent fuel pool time to boil. V&V testing was conducted under similar initial conditions as when the issue was first identified. The values indicated by each of these displays should be in hours and minutes. This V&V test failed.

The Time to Boil NAP is a tool that is used for information only. The NAP is active when in Mode 5/6 conditions for RCS time to boil and when fuel is present in the spent fuel pool for spent fuel pool time to boil.



NRC


{*}

Ref

# {*}

SNC SCR

#

SNC Evaluation

The value is displayed in exponential units versus hh:mm. This means that operators will need to convert the scientific notation values into hours and minutes. Although this does take a short amount of time, the net effect is that it does not remove the ability to monitor the time to boil and it has no impact on actions the operator may, or may not, take in response to plant conditions.

41

Provide documentation that the Westinghouse Electric Company’s resolution of HED-1 discrepancies is consistent with the VC Summer (VCS) conclusions provided in the Commission-approved simulator request and its supplements.

See Enclosure 8.

42

Include all open discrepancy reports when the docketed list of simulator discrepancies is submitted.

See Enclosure 9

Note: {*} Numbers and descriptions correspond to the table “Summary of Unresolved Items as of 06-30-2015” as it appeared in an NRC letter dated July 2, 2015

(Reference 2) with the following exceptions. If the information in the “NRC UI Description” or “Ref #” columns was found to be incorrect, that information

was retained, but indicated by using strikethrough. The correct information was added immediately following and was contained in brackets “{ }.”

2.0 References

1. NRC Email dated 2015-05-13, Meeting Materials for May 14, 2015- VCSNS 2 and 3 Commission-Approved Simulator - CAS-

Summer-RAI 5-7-15_b Redacted, ML#15133A497

2. Virgil C. Summer Nuclear Station Units 2 and 3 - Request For A Commission-Approved Simulation Facility dated July 2, 2015,

ML#15182A097



ND-15-1333

Enclosure 6

Commission Approved Simulator Aggregate Study - Simulator Training System Deficiency Impact on 10 CRF 55.45

(Non-Proprietary)

(The Aggregate Study is a standalone document consisting of 105 pages.)

SOUTHERN NUCLEAR COMPANY

Commission Approved Simulator Aggregate Study

Simulator Training System Deficiency Impact On 10CFR55.45(a) Compliance

7/17/2015

ND-15-1333

Enclosure 6

VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility

2 | P a g e

Contents

Introduction ....................................................................................................................................................... 9

Executive Summary ........................................................................................................................................... 9

Method of Study ................................................................................................................................................ 9

Participants ...................................................................................................................................................... 10

Simulator Review Committee (SRC) ................................................................................................................ 10

Aggregate Study Evaluation Results ................................................................................................................ 11

Index of Proposed Corrections......................................................................................................................... 12

10 CFR 55.45(a)(1) ............................................................................................................................................... 14

Executive Summary ......................................................................................................................................... 14

Reactor Coolant Pump (RCP) Net Positive Suction Head (NPSH) Curve has Inadequate Range for Operation 14

Rod Withdrawal button deselects During Continuous Operation ................................................................... 15

Issue with Automatic Control of Deaerator Storage Tank (DST) level and Auto Start of Standby Condensate Pump ................................................................................................................................................................ 15

Model Instability during pressurizer (PZR) Fill to Solid (no vapor bubble remaining) ..................................... 15

OPDMS Rod Insertion Limit (RIL) Indication Does Not Align to Combined Operating Limits Report (COLR) Rev. 0 ....................................................................................................................................................................... 16

Decay Heat Calculation Summary - Assembly Move NAP Function Not Functional ........................................ 16

M Control Rod Banks B & C Reversed on DRPI Health Screen ......................................................................... 17

Liquid Radwaste System WLS-MP-08C improperly Pumps Monitor Tank C .................................................... 17

Excessive Startup Feedwater (SFW) Control Valve Cycling .............................................................................. 18

Redundant Sensors Algorithm Application NAP Does Not Process Failed Channels Correctly ........................ 18

Manual Reactor Trip Alarm Occurred without a Reactor Trip Request ........................................................... 19

10 CFR 55.45(a)(2) ............................................................................................................................................... 20



Issue with Automatic Control of DST level and Auto Start of Standby Condensate Pump .............................. 21

OPDMS RIL Indication Does Not Align to COLR Rev. 0 ..................................................................................... 21

NAP for 1/M Intermediate Range Not Functional ........................................................................................... 22

MA Bank Rods Sometimes Stop at 263 steps during a CRE ............................................................................. 22

Excessive SFW Control Valve Cycling ............................................................................................................... 23

Audible Rod Step Skips .................................................................................................................................... 23

Improper Bank Overlap Occurs when Data Point OCB07CE00C_OUTAV Increments Improperly During

ND-15-1333

Enclosure 6


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Startup ............................................................................................................................................................. 24


10 CFR 55.45(a)(3) ............................................................................................................................................... 25


Containment Cooling System (VCS) fan response due to loss of power .......................................................... 25

EDS Power Supply Assignments to PLS/DDS Cabinets Incomplete .................................................................. 25

Modeled BEACON Data Cannot Determine Quality ......................................................................................... 26


Containment Radiation Alarm Reset Points Incorrect ..................................................................................... 27

Pressurizer Heater Current Indicates BAD Quality at Limits ............................................................................ 27

Unidentified and Identified Leak Rate Always Indicates BAD Data .................................................................. 28

Low Flow Alarm on TCS-FT007 Occurs Earlier than Expected .......................................................................... 28

MFP 'B' Alarm Response Differs For Identical Fault ......................................................................................... 28

Unexpected Response of Alarm Cutout of RWS Pressure Alarms ................................................................... 29

Pressurizer Pressure Out of Range Indication Not Properly Displayed ............................................................ 30

Subcriticality Indication on Critical Safety Function Screen Drops to Bad Quality ........................................... 30


Degasifier Level Alarm Limits ........................................................................................................................... 31

PMS Mimic Screens .......................................................................................................................................... 31


Nuisance Valve Modulating Status Alarms ...................................................................................................... 32

Unexpected VRS High Rad Alarm ..................................................................................................................... 32

VFD Transformer Temperature........................................................................................................................ 33

Print Feature from NAP non-functional ........................................................................................................... 33

VHS Rad Monitor Response to Loss of Process Flow ....................................................................................... 34

Pressurizer Narrow Range Pressure Does Not Indicate Bottom of Scale ......................................................... 34

CDS-TE040A/B Range is Inadequate ................................................................................................................ 34

DRPI Health Screen Alarms for Data Cabinet A and B Crossed ........................................................................ 35

Digital Rod Position Indication (DRPI) Health Screen Incorrect Logic Cabinet Alarms ..................................... 35

Containment Recirculation Actuation Indication Issue .................................................................................... 35

Uncontrolled Heat-up (H/U) Indication Incorrect ............................................................................................ 36

DHC Summary - Assembly Move NAP Function Not Functional ...................................................................... 36

RCP Vibration Alarm Naming ........................................................................................................................... 37

HSS Display does not Include ESOP Discharge Pressure .................................................................................. 37

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Enclosure 6


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NAP for 1/M Intermediate Range Not Functional ........................................................................................... 38

ECS Penetration Temperature off Scale Low ................................................................................................... 38

Plant Mode Selector NAP Inconsistent with Procedure ................................................................................... 38

NAPS display issues .......................................................................................................................................... 39

WPIS Downscale Arrow Absent ....................................................................................................................... 39

RSA NAP Does Not Process Failed Channels Correctly .................................................................................... 39

ZVS and ZBS Alarm Scaling Incorrect ............................................................................................................... 40

Flux doubling difference between divisions .................................................................................................... 40

Time to Boil Calculation ................................................................................................................................... 41


WPIS Display VARs ........................................................................................................................................... 41

CMT WR Level Indications go Bad Quality ....................................................................................................... 42

Unexpected Bank Sequence Out of Sequence Alarm ...................................................................................... 42

Urgent Alarm Occurs During Case 2 CRE.......................................................................................................... 43

Improper Bank Overlap Occurs when Data Point OCB07CE00C_OUTAV Increments Improperly During Startup ............................................................................................................................................................. 43


Controller Fault Alarms Received on Turbine Trip ........................................................................................... 44

Diesel Fuel Oil Day Tank Level Transmitter Operation..................................................................................... 45

Inconsistent UAT Line Voltage Alarm Priorities ............................................................................................... 45

Any Rods at Bottom Alarm .............................................................................................................................. 46

WGS Sample Package Ovation Interface ......................................................................................................... 46

WGS Sample Package Digital Indication .......................................................................................................... 47

RSA NAP for Power Range Power does not Eliminate Erroneous Input .......................................................... 47

Inconsistent DPU Alarm Priority Levels ............................................................................................................ 47

Safety Mimic Display for SGS-V255A& B Indicates Bad Quality Following a SFW Isolation ............................. 48

10 CFR 55.45(a)(4) ............................................................................................................................................... 49


Stage 3 ADS Box Unused on Divisions C and D ................................................................................................ 49

Subcriticality Indication on Critical Safety Function Screen Drops to Bad Quality ........................................... 49

VWS-TE079 Point Named Incorrectly .............................................................................................................. 50

Calorimetric Data Precision ............................................................................................................................. 50

Inconsistent OPDMS QPT Indications .............................................................................................................. 50

VFD Transformer Temperature ........................................................................................................................ 51


ND-15-1333

Enclosure 6


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CVS-PT040 does not Provide Proper Protective Functions .............................................................................. 52

M Banks B & C Reversed on DRPI Health Screen ............................................................................................. 52

Quality of RWS-V503 BAD at Limits ................................................................................................................. 53

Reactor Coolant Pump (RCP) Stator Temperature Indication off Scale Low at Lower Speeds ........................ 53

HSS Display does not Include Emergency Seal Oil Pressure (ESOP) Discharge Pressure ................................. 54

DWS-LT006 has Insufficient Range .................................................................................................................. 54


ECS Penetration Temperature off Scale Low ................................................................................................... 55

Improper function of C-2 reactor power control interlock .............................................................................. 55

WPIS RCS Inventory Issues ............................................................................................................................... 55

WPIS Downscale Arrow Absent ....................................................................................................................... 56

Tuning of VBS Required for Stability ................................................................................................................ 56

Condensate Polisher Bypass Valve Control ...................................................................................................... 57




Main Generator Output breaker logic ............................................................................................................. 58

Excitation Transformer Graphic Issue .............................................................................................................. 59

IDS Charger Capacity and Design Float Voltage Requirement are Incompatible ............................................. 59

Graphic 1805 has reversed rods ...................................................................................................................... 60

Residual Bus Transfer Issues ............................................................................................................................ 60


WGS Sample Package Digital Indication .......................................................................................................... 61


Safety Mimic Display for SGS-V255A& B Indicates Bad Quality Following a SFW Isolation ............................. 62

Safety Mimic Display Navigation Issue............................................................................................................. 63

10 CFR 55.45(a)(5) ............................................................................................................................................... 64


EDS Power Supply Assignments to PLS/DDS Cabinets Incomplete .................................................................. 64

Modeled BEACON Data Cannot Determine Quality ......................................................................................... 64

Repeatability issues involving CL 1B ................................................................................................................ 65

Unidentified and Identified Leak Rate Always Indicates BAD Data .................................................................. 65

Primary Dedicated Safety Panel Screens Do Not Update during MCR/RSR Transfer ...................................... 66

PMS Mimic Screens .......................................................................................................................................... 66

ND-15-1333

Enclosure 6


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Hot Leg Fluctuations at Mid-loop .................................................................................................................... 67


Inconsistent OPDMS QPT Indications .............................................................................................................. 68

Print Feature from NAP non-functional ........................................................................................................... 68



Containment Recirculation Actuation Indication Issue .................................................................................... 70

CVS-V094 Power Failure Response .................................................................................................................. 70


HSS Display does not Include Emergency Seal Oil Pump (ESOP) Discharge Pressure ...................................... 71

DWS-LT006 has Insufficient Range .................................................................................................................. 71


Improper function of C-2 ................................................................................................................................. 72


SWS temperature control ................................................................................................................................ 73

FWS-V037 Control Issue ................................................................................................................................... 74

SGS MSL drain pot erratic indication ............................................................................................................... 74

Stuck Rod Recovery Malfunction ..................................................................................................................... 74

Tuning of VBS Required for Stability ................................................................................................................ 75

RSA NAP Does Not Process Failed Channels Correctly .................................................................................... 75

Flux doubling difference between divisions .................................................................................................... 76



VFS Radiation Monitoring Issue ....................................................................................................................... 77



Improper Bank Overlap Occurs when Data Point OCB07CE00C_OUTAV Increments Improperly During Startup ............................................................................................................................................................. 78


Main Generator Output breaker logic ............................................................................................................. 79



VES Supply Header Pressure Response to Temperature Changes ................................................................... 81

ECS-EC-313 Loads not modeled ....................................................................................................................... 81

D/G Sequencer Operation ............................................................................................................................... 82

ND-15-1333

Enclosure 6


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VZS Dampers do not Fail As-Is after Loss of Power .......................................................................................... 83

Turbine Bypass Control Valve Control Logic cannot Support Design Power Supplies ..................................... 83

Battery Temperature does not change ............................................................................................................ 84

Fire Protection System is not modeled in Containment .................................................................................. 84

IRWST Temperature Response ........................................................................................................................ 84

10 CFR 55.45(a)(6) ............................................................................................................................................... 86



Unstable VFS Containment Exhaust Flow ........................................................................................................ 87

GSS Header Pressure Response ....................................................................................................................... 87


Model Instability during PZR Fill to Solid ......................................................................................................... 88

Steam Generator Level Instability with Control Valves Shut ........................................................................... 89

CVS-V094 Power Failure Response .................................................................................................................. 89


WLS-MP-08C improperly Pumps Monitor Tank C ............................................................................................ 90


WRS Sump Pump B Discharge Pressure Inadequate ....................................................................................... 91


Stuck Rod Recovery Malfunction ..................................................................................................................... 92

Polisher Bypass Valve Control .......................................................................................................................... 92


IDS Charger Capacity and Design Float Voltage Requirement are Incompatible ............................................. 93


ECS-EC-313 Loads not modeled ....................................................................................................................... 94

D/G Sequencer Operation ............................................................................................................................... 95

Turbine Bypass Control Valve Control Logic cannot Support Design Power Supplies ..................................... 95

Fire Protection System is not modeled in Containment .................................................................................. 96

10 CFR 55.45(a)(7) ............................................................................................................................................... 97


Repeatability issues involving CL 1B ................................................................................................................ 97



ND-15-1333

Enclosure 6


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CMT WR Level Indications go Bad Quality ..................................................................................................... 100

Turbine Bypass Control Valve Control Logic cannot Support Design Power Supplies ................................... 100

IRWST Temperature Response ...................................................................................................................... 101

10 CFR 55.45(a)(8) ............................................................................................................................................. 102

Executive Summary ....................................................................................................................................... 102

EDS Power Supply Assignments to PLS/DDS Cabinets Incomplete ................................................................ 102

CVS-V094 Power Failure Response ................................................................................................................ 102

CMT WR Level Indications go Bad Quality ..................................................................................................... 103

Fire Protection System is not modeled in Containment ................................................................................ 103

IRWST Temperature Response ...................................................................................................................... 104

10 CFR 55.45(a)(9) ............................................................................................................................................. 105

Executive Summary ....................................................................................................................................... 105

Simulator MCR missing Rad Monitoring Panel .............................................................................................. 105

ND-15-1333

Enclosure 6


9 | P a g e

Introduction As part of the Commission Approved Simulator (CAS) request, Southern Nuclear Operating Company (SNC) commissioned a team to evaluate the known deficiencies in the simulator to determine if the 13 criteria established in 10 CFR 55.45(a), “Operating Tests,” would be challenged. The team was comprised of representatives from SNC (Operations, Training and Engineering), SCANA (Training) and Westinghouse (Human Factors Engineering).

Executive Summary The team examined all Discrepancy Reports (DRs) that were open as of May 15, 2015, and determined that 101 DRs were relevant to acceptability of one or more of the first nine criteria of 10 CFR 55.45(a). No DRs were found to be relevant to the last four criteria; 55.45(a)(10) through 55.45(a)(13). The team also determined that no singular DR posed a challenge to the suitability of the simulation facility for the conduct of operating tests; however, when considered in the aggregate, 42 of the DRs challenged criterion (3) and (5) of 10 CFR 55.45(a) (See section 2.1 below for additional details).

In order to declare the simulator suitable for the conduct of operating tests, corrective actions were initiated to resolve the subject 42 DRs. This assessment was communicated with Westinghouse Electric Company (WEC) and WEC committed to implement improvements aimed at resolving these issues in a patch deliverable to SNC by August 14, 2015. Based on this commitment, the CAS Aggregate Study Team reconvened on July 7, 2015 and determined that the proposed corrective actions would be adequate so that the aggregate impact of the remaining discrepancies would not pose a challenge to any of the 10 CFR 55.45(a) criteria.

On August 14, 2015, WEC delivered a patch to SNC which contained corrections for the 42 items previously identified along with some additional corrections. After performing Verification and Validation (V&V), 11 were determined to require further resolution. After confirming the corrections that successfully passed the V&V process, the CAS Aggregate Study Team reconvened on September 1, 2015, to review the impact of the remaining 11 items. Based on the combination of these successful corrections and additional improvement in the area of Alarm Response, through the use of the Consequence Alarm feature of the Alarm Presentation System (APS), the team concluded that the aggregate impact of the remaining items would not impact the suitability of the simulator for the performance of operating tests.

Method of Study A multi-disciplined team was formed (see “Participants” below) consisting of individuals from both internal and external of SNC. The participants were requested to answer the following questions for the individual and aggregate impact of the deficiencies.

General Questions for each deficiency of the study

• Does an individual item fit in the assigned category?

• Could an individual item affect another part of 55.45 that it is not currently assigned to?

• Could an individual item come off the list entirely? (I.e. does not affect 55.45)

• Does an individual item make sense? Do you understand the problem it introduces?

• Are we working around the issue with procedure changes or special training? If so, how is that documented?

Each of the identified deficiencies were evaluated using the above criteria and categorized into the appropriate criteria of 10 CFR 55.45(a). Individual deficiencies were not limited to one category, but

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were added to all applicable category bins associated with the nature of the deficiency. Deficiencies associated with design of the plant or deficiencies that did not apply to the criteria in 10 CFR 55.45(a), were removed from the list. Upon completion of binning of the individual deficiencies, the team evaluated each of the criterion of 10 CFR 55.45(a)(1) – (13) to determine if the simulators are suitable for conducting operating tests with the existing deficiencies.

Participants The team was composed of participants with a diverse mix of backgrounds including Operations, Instrument & Controls (I&C) Engineering, Training, and Human Factors Engineering from both Vogtle 3&4, V.C. Summer Units 2&3 and Westinghouse. The team was composed of the following members:

Tom Arnette – Shift Manager, Vogtle Units 3&4 (12.5 years Nuclear Navy, Reactor Operator; SRO license holder and Shift Manager at Kewaunee)

Chris Parkes – Shift Supervisor, Vogtle Units 3&4 (BS- Computer Information Systems; 23 years Nuclear navy, Qualified Engineering Officer of the Watch and Engineering Watch Supervisor for 18 years)

Shawn Wolfgong – Shift Support Supervisor, Vogtle Units 3&4 (BS – Applied Nuclear Technology; 20 years Nuclear Navy, EWS & EOOW)

Chris Cannon – Nuclear Plant Operator, Vogtle Units 3&4 (BS – Nuclear Engineering Technology; 6 years Nuclear Navy, Reactor Operator; 3 years Lockheed Martin Electronics testing and repair)

Matt Schmader – Training Lead, Vogtle Units 3&4 (BS – Physics; 9 years Nuclear Navy Officer; 5 years Operations training at Watts Barr, SRO-certified.

Allahondra Manning – Engineer, Vogtle Units 3&4 (BS- Electrical Engineering; 6.5 years Electrical and I&C systems engineer, design authority SRS)

Kim Yennerell – Engineer, Vogtle Units 3&4 (BS – Electrical Engineering; 10.5 years Nuclear design and program engineering)

Korrie Hoffman – Simulator Engineer, Vogtle Units 3&4 (BS – Nuclear Engineering; 3.5 years core design engineer with WEC)

Kevin Balch – Simulator Engineer, V.C. Summer Units 2&3 (BS – Nuclear Engineering; 20 years simulator software engineer, 10 years nuclear fuel engineer)

H. Adrian Fletcher – Human Factors Engineering Operations Specialist, Westinghouse (AS – Nuclear Engineering Technology; NRC license holder for 18 years, 6 years Reactor Operator, 8 years Unit Supervisor, 4 years Shift Manager)

Simulator Review Committee (SRC) The SRC is composed of one member of Operations, selected by the Operations Director; the Operations Training Manager or designee; and the Simulator Coordinator or designee. The SRC is supplemented by incumbents, as necessary, to serve as subject matter experts to conduct a Training Needs Assessment (an appraisal by a subject matter expert of a simulator deviation, deficiency, or modification, and its relative importance to the operator as required tasks are performed). Incumbents are Operations individuals who have completed AP1000 certification training.

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Aggregate Study Evaluation Results The team evaluated each of the individual deficiencies and determined that none of the issues, by themselves, constituted a challenge to any of the 13 criteria of 10 CFR 55.45(a). The team did determine that, in the aggregate, some of the deficiencies could challenge 10 CFR 55.45(a) criterion (3) and (5).

10 CFR 55.45(a)(3):

“Identify annunciators and condition-indicating signals and perform appropriate remedial actions where appropriate.”

10 CFR 55.45(a)(5):

“Observe and safely control the operating behavior characteristics of the plant.”

The reasoning for the determination involved four (4) main areas: 1. Indication deficiencies 2. Alarms management deficiencies and challenges 3. Rod Control System deficiencies 4. Secondary control challenges

Indication deficiencies: The key drivers for this area are associated with challenges to the simulator in providing the operator with the necessary and correct information in order to monitor and control the plant. The effect of the identified deficiencies is that operators do not always have the correct information presented to them to make appropriate decisions required for safe operation. Additionally, with many identified deficiencies, operators will tend to question the validity of all indications, including the ones that are working correctly.

Specific DRs associated: 6169, 6621, 5689, 5599, 6175, 6315, 6159, 6089, and 5623.

Alarm Management deficiencies and challenges: The key drivers for this area are associated with the excessive number of alarms, the absence of some required alarms, and the distraction presented to the operators in managing the Alarm Presentation System (APS). Alarm management is a significant operator burden placed on the crew throughout all scenarios and plant conditions. Alarm management currently makes simple and routine evolutions difficult.

Specific DRs associated: 5813, 5613, and 6651.

Rod Control System deficiencies: The key drivers for this area are associated with the inconsistent and unpredictable nature of the Rod Control System during performance of reactivity related tasks. In effect, whenever the operator operates the controls to move rods, the operator could encounter no issues or he/she may encounter many issues:

- The control button can deselect during manual rod motion - The audible clicking sound can skip during rod motion or continue after rod motion has stopped - The Bank Out-of-Sequence alarm can come in when conditions do not warrant - The rod group indication may indicate values below the fully inserted position or above the fully

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withdrawn position - The bank overlap could be incorrect - During a Rod Exchange, an Urgent Failure alarm may or may not occur when AO bank temperature

control is demanded The overall affect is that these issues will inhibit the operator’s ability to timely perform reactivity manipulations in a precise and controlled manner.

Specific DRs associated: 5584, 6259, 6302, 6186, and 6267.

Secondary control challenges: The key drivers for this area are associated with the inability of the automatic control of the secondary plant and subsequent operator required action to manually control systems in order to respond to simulator scenarios to prevent automatic actuation of standby components.

Specific DRs associated: 6151, 5655, and 6156.

Index of Proposed Corrections The following is a summary of the proposed corrections that the team determined would result in a valid combination of corrections that the aggregate impact of the remaining discrepancies would not pose a challenge to any of the 10 CFR 55.45(a) criteria. [See Enclosure 7 of the CAS Submittal Letter for an updated status of the following items.]

1. SCR-DR-5584 (Parts 1, 2, 3, 6)1 - Rod Withdrawal button deselects During Continuous Operation 2. SCR-DR-5597 (Part 3) - Containment Radiation Alarm Reset Points Incorrect

3. SCR-DR-5599 (Parts 3, 5)1 - Unidentified and Identified Leak Rate Always Indicates BAD Data 4. SCR-DR-5627 (Parts 3, 4) - Sub-criticality Indication on Critical Safety Function Screen Drops to Bad

Quality 5. SCR-DR-5643 (Part 4) - VWS-TE079 Point Named Incorrectly (TE079 is a temperature indication) 6. SCR-DR-5644 (Not part of aggregate study) – Display 17600 indicating wrong flowpath 7. SCR-DR-5688 (Not part of aggregate study) – RCS graphic 50308 incorrect 8. SCR-DR-5689 (Parts 3,5) – PMS mimic screens

9. SCR-DR-5702 (Not part of aggregate study) – IDS screens show inaccurate power supplies 10. SCR-DR-5712 (Part 4) - Calorimetric Data Precision

11. SCR-DR-5813 (Part 3)1 - Nuisance Valve Modulating Status Alarms 12. SCR-DR-5909 (Not part of aggregate study) – Graphic 11181 has orphaned “n” 13. SCR-DR-5920 (Part 3) - Pressurizer Narrow Range Pressure Does Not Indicate Bottom of Scale 14. SCR-DR-5924 (Part 3) – Digital Rod Position Indication (DRPI) Health Screen Alarms for Data Cabinet A

and B Crossed 15. SCR-DR-5925 (Part 3) - DRPI Health Screen Incorrect Logic Cabinet Alarms 16. SCR-DR-5968 (Parts 4, 5, 7) - CVS-PT040 does not Provide Proper Protective Functions (PT040 is a

pressure transmitter) 17. SCR-DR-6009 (Part 3) - Uncontrolled heatup or cooldown indication incorrect 18. SCR-DR-6030 (Parts 1, 4) - M Banks (control rods) B & C Reversed on DRPI Health Screen 19. SCR-DR-6078 (Parts 3, 4, 5) – Hydrogen Seal Oil System (HSS) Display does not Include Emergency Seal

Oil Pump (ESOP) Discharge Pressure

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20. SCR-DR-6089 (Parts 2, 3)1 – NAP(Nuclear Application) for 1/M(inverse count rate ratio plot) Intermediate Range Not Functional

21. SCR-DR-6102 (Parts 2, 4, 5, 6) - MA Bank Rods Sometimes Stop at 263 steps during a CRE (Control Rod Exchange)

22. SCR-DR-6129 (Not part of aggregate study) – Display 40023 units issue

23. SCR-DR-6144 (Part 3) - Plant Mode Selector NAP Inconsistent with Procedure

24. SCR-DR-6159 (Part 3)1 - NAPS display issues 25. SCR-DR-6160 (Not part of aggregate study) – Component Cooling System (CCS) screen issue 26. SCR-DR-6164 (Parts 3, 4) – Wall Panel Information System (WPIS) Downscale Arrow Absent 27. SCR-DR-6165 (Not part of aggregate study) – WPIS Tavg scale 28. SCR-DR-6169 (Parts 1, 3, 5) – Redundant Sensors Algorithm (RSA) NAP Does Not Process Failed

Channels Correctly 29. SCR-DR-6170 (Not part of aggregate study) – Radioactive Waste Drain (WRS) graphic issue 30. SCR-DR-6180 (Not part of aggregate study) – Trend for Time to Boil unit indication 31. SCR-DR-6187 (Not part of aggregate study) - Rod sequence skips steps

32. SCR-DR-6259 (Part 3)1 - Unexpected Bank Sequence Out of Sequence Alarm

33. SCR-DR-6267 (Parts 3, 5, 6)1 - Urgent Alarm (Causes control rods to swap to manual and stop) Occurs During Case 2 CRE

34. SCR-DR-6278 (Not part of aggregate study) – Battery bank indications mislabeled for EDS1, EDS2, and EDS4

35. SCR-DR-6302 (Parts 2, 3, 5)1 - Improper Bank Overlap Occurs when Data Point OCB07CE00C_OUTAV Increments Improperly During Startup

36. SCR-DR-6315 (Parts 1, 2, 3, 4, 5)1 - Manual Reactor Trip Alarm Occurred without a Reactor Trip Request 37. SCR-DR-6398 (Part 4) - Excitation Transformer Graphic Issue

38. SCR-DR-6409 (Part 4) - Graphic 1805 has reversed rods

39. SCR-DR-6621 (Parts 3, 4, 5)1 - RSA NAP for Power Range Power does not Eliminate Erroneous Input

40. SCR-DR-6651 (Part 3)1 - Inconsistent Digital Processing Unit (DPU) Alarm Priority Levels 41. SCR-DR-6698 (Parts 3, 4) - Safety Mimic Display for SGS-V255A& B Indicates Bad Quality Following a

Startup Feedwater Isolation

42. Alarm Server update with alarm prioritizations. This update resolves HED issues such as the CCS surge

tank leak going unidentified due to excessive alarms1

1These SCR-DRs were identified as impacting 10 CFR 55.45 in the aggregate.

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10 CFR 55.45(a)(1) Perform pre-startup procedures for the facility, including operating of those controls associated with plant equipment that could affect reactivity.

Executive Summary

The team determined that there was not anything in this section that could not be handled by a standard three (3) person control room crew. The individual SCRs listed here do not influence each other with respect to responding to simulator scenarios. From a procedural standpoint, they could be evaluated effectively and do not impact responding to simulator scenarios.

The team determined there are no negative aggregate impacts from these issues that will affect operator training or operations in the simulator.

In terms of examination there is still enough of a representative sample that could test operator effectiveness. There remains a large population of JPM tasks which could be combined to evaluate this otherwise.

Reactor Coolant Pump (RCP) Net Positive Suction Head (NPSH) Curve has Inadequate Range for Operation

SCR-DR-5577 This issue impacts the following RO/SRO task: RO-PRI-RCS-005-00 Operate the RCS during shutdown/cooldown conditions

Disposition

This issue was dispositioned as acceptable by a Subject Matter Expert (SME). The current display requires

finer pressure control by the operators, but the procedures can still be used to successfully accomplish the task.

Description

The Reactor Coolant Pump (RCP) minimum Net Positive Suction Head (NPSH) display (60029) shows the required Reactor Coolant System (RCS) pressures for given RCS temperatures for starting RCPs. The display shows the limits based on the instruments used for indication, RCS Wide Range (WR) pressure (RCS-PT140A/B/C/D) or Normal Residual Heat Removal System (RNS) pump suction pressure (RNS- PT011A/B).

The current display only indicates the RNS limits below 275oF. Per RCS Component Control Requirements (APP-RCS-M3C-100 Rev. 9) logic sheets RCS-13 and RCS-18, the pressure indication used for determining NPSH should be [ ]a,c. Per procedure, RNS is placed on

service when [ ]a,c.

During the subsequent cooldown and depressurization using RNS, this display does not indicate the larger

margin allowed for NPSH using the RNS suction pressure above 275oF as the margin to NPSH limits using the RCS WR pressure instruments is very small at these lower temperatures and pressures.

Area of Impact

Decay heat removal (forced circulation) during startup/shutdown

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Rod Withdrawal button deselects During Continuous Operation

SCR-DR-5584 This issue impacts the following RO/SRO task: RO-INC-PLS-003-02 Monitor the Control Rod Drive System

Disposition

Subsequent to the performance of the Aggregate Study this item has been corrected and closed.

Description

While performing extended rod withdrawals during startups, depressing the rod withdrawal button (UP ARROW) may cause the UP ARROW button to un-highlight and momentarily flash gray even though still depressed. Rod motion will still occur.

Area of Impact

Reactivity Management

Issue with Automatic Control of Deaerator Storage Tank (DST) level and Auto Start of Standby Condensate Pump

SCR-DR-5655 This issue impacts the following RO/SRO task: RO-SEC-FWS-002-03 Establish level in the DST

Disposition

This issue was dispositioned as acceptable by the Simulator Review Committee (SRC). The reasoning was that the current controller tuning is at the best estimated values. Hot functional testing would provide as- built tuning. However, this issue did contribute to the aggregate impact of sections 3 and 5 with regards to secondary plant control. Description

During plant startup from Mode 5 to 100% power, the Condensate system pressure would lower to the auto start setpoint of the standby Condensate pump. This pressure drop is due to Condensate System (CDS) valves, CDS-V022 and CDS-V025 modulating to maintain level in the Deaerated Storage Tank (DST). In accordance with reference plant procedures for normal operation, the second condensate pump is started at 40-45% power. However, the second condensate pump will have already auto started in the heatup and startup procedures, due to the slow response of CDS-V022 and CDS-V025.

Area of Impact

Plant design deficiency impacts operations during startup

Model Instability during pressurizer (PZR) Fill to Solid (no vapor bubble remaining)

SCR-DR-5698 This issue impacts the following RO/SRO task: RO-PRI-CVS-003-04 Operate the Chemical and Volume Control System to control the primary system pressure in water solid mode

Disposition

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This issue was dispositioned as unacceptable by the Simulator Review Committee (SRC). Training Needs

Analysis was performed under Condition Report (CR) 10000465. Training Needs Analysis determined training involving establishment of solid plant should not be performed until issue is corrected. A review of current training material did not reveal any scenarios where this was required.

Added to the DR Global Issues list, this will be briefed to the students at the beginning of the Simulator portion of training. Scenario AP-LT-I-SIM-GOPSDCD (Covering GOP-205, Plant Cooldown MODE 3 to

MODE 5) does not train on Solid Plant Operations.

Description

The Liquid Radwaste System (WLS) model is prone to failure during evolutions involving near solid pressurizer operations if the Effluent Holdup Tank is filled too rapidly.

Area of Impact

Difficulty in achieving solid plant operations continuously

OPDMS Rod Insertion Limit (RIL) Indication Does Not Align to Combined Operating Limits Report (COLR) Rev. 0

SCR-DR-5736 This issue impacts the following RO/SRO task: RO-INC-PLS-002-03 Control Rods: Verify Sequence and Overlap within COLR specifications

Disposition

This issue was dispositioned as unacceptable by the Simulator Review Committee (SRC). Training Needs Analysis was performed under CR 10000474. The following is taken from the analysis performed. The Training Needs Analysis determined that the M1 Bank Insertion Low-2 alarm would be received prior to the M2 Bank Insertion Low-2 alarm due to rod sequencing and bank overlap. The M1 Bank alarm is set at the correct value.

Description

It is noted that the Core Operating Limits Report (COLR) Rev. 0 Rod Insertion Limit (RIL) for M2 bank at all power levels is [ ]a,c Steps (fully withdrawn). The Online Power Distribution Monitoring System (OPDMS) Rod Insertion Limit (RIL) display ([ ]a,c) indicates the RIL for M2 bank is [ ]a,c steps. Further investigation indicates the M2 RIL indication high limit is [ ]a,c steps and therefore cannot indicate above this level (determined using point information page instrumentation tab for RB-INSERT- M2LIM.SV3@NET0). All Shutdown (SD) bank indications are capable of indicating a maximum of [ ]a,c steps.

Area of Impact

Reactivity Management with regards to indication

Decay Heat Calculation Summary - Assembly Move NAP Function Not Functional

SCR-DR-6022

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This issue impacts the following RO/SRO task: RO-INC-PLS-004-03 Perform decay time surveillance

Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to the minimal training impact this particular Nuclear Application (NAP) has.

Description When attempting to simulate fuel assemblies being moved from the core to the Spent Fuel Pool (SFP), it was noted that the Decay Heat Calculation (DHC) NAP to maintain the administrative location of fuel does not work correctly. On display 40203 the assembly move buttons on the lower right portion indicate they are only available when the light DDS-AP-DHC Status indicates it is ACTIVE. This light is driven by the automatic mode selector and is INACTIVE when in MODES 1&2 and ACTIVE in MODES 3- 6. However, when the light indicates INACTIVE the buttons for moving are raised and available. When the light changes status to ACTIVE the buttons for moving are grayed out and no longer available. The light being active or inactive is currently driven by the auto mode selector and becomes active in MODES 3-6. However, fuel cannot be moved from the core into the SFP in any MODE other than MODE 6. The light should be driven by the manual input of the Rx vessel head being removed or installed or upper internals position on display 40004.

Area of Impact Reactivity Management with regards to indication and administration

M Control Rod Banks B & C Reversed on DRPI Health Screen

SCR-DR-6030 This issue impacts the following RO/SRO task: RO-INC-DDS-001-00 Access plant information using the DDS

Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed.

Description The DRPI Health Screen (1805) control rod banks M-B and M-C have the wrong rods listed as being in each bank. The rods listed as being M-B are actually M-C and the rods listed for M-C are the M-B rods. The correct arrangement of rods is shown in APP-RXS-M3-001 Rev 4 Figure 4-1 as well as on the DRPI M Bank screen (11172).

Area of Impact Plant indications

Liquid Radwaste System WLS-MP-08C improperly Pumps Monitor Tank C

SCR-DR-6068 This issue impacts the following RO/SRO task: RO-SUP-WLS-002-00 Operate the WLS

Disposition


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Description While performing a startup from Mode 6 it was discovered that the Liquid Waste System (WLS) WLS- MP-08C will not pump Monitor tank C around 37 inches. The pump will turn on and occasionally the downstream check valve will throttle open and shut but there is little or no evidence of flow. Also, discharge pressure never goes above 12-13psig. Normal discharge pressure for the other monitor tank pumps is around [ ]a,c.

Note that it does pump when level is above 37 inches as the tank has been pumped down to 37 inches successfully. It appears to exhibit strange behavior at 37 inches and below.

Area of Impact Correct operation of plant systems

Excessive Startup Feedwater (SFW) Control Valve Cycling

SCR-DR-6151 This issue impacts the following RO/SRO task: RO-SEC-FWS-002-01 Monitor SFWS and MFWS system and component parameters

Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC). The reasoning was that the current controller tuning is at the best estimated values. Hot functional testing would provide as- built tuning. However, this issue did contribute to the aggregate impact of sections 3 and 5 with regards to secondary plant control.

Description At low pressure conditions less than 350 psig, the operator often has to take manual control of Startup Feed Water (SFW) control valves due to excessive cycling of the valves. Indicated flow rates oscillate erratically between no flow and max flow every 10 to 15 seconds. This requires 100% of the operator’s attention until RNS can be placed into service removing cooldown function from the steam dumps.

Area of Impact Plant control during startup and shutdown

Redundant Sensors Algorithm Application NAP Does Not Process Failed Channels Correctly

SCR-DR-6169 This issue impacts the following RO/SRO task: RO-INC-PMS-003-00 Monitor System Indications

Disposition


Description The Redundant Sensors Algorithm Application (RSA) driven source range counts on the WPIS displays (main, trends, and safety functions) will still reflect an abnormally high value for source range power after a source range channel failure. The RSA NAP should account for the failure and remove it from the calculation.

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Area of Impact Reactivity management

Manual Reactor Trip Alarm Occurred without a Reactor Trip Request

SCR-DR-6315 This issue impacts the following RO/SRO task: RO-INC-PMS-005-02 Respond to a loss of a

Protection and Safety Monitoring System (PMS) division or failure of PMS components


Description The below failure sequence was performed and the reactor did not trip, but a Reactor Trip alarm was received. The alarm was for a Manual Reactor Trip, but a manual Rx Trip was not inserted. A P-4 was not received.

• RCS TE122C -> Open Circuit: PMS indicates Low-2 Tcold Cold Leg 2 Bistable Trip with Bad Quality

& Maintenance Bypass for Division C & PMS Cabinet Fault Alarm for Division C • RCS TE122D -> Open Circuit: PMS indicates Low-2 Tcold Cold Leg 2 Bistable Trip with Bad Quality

& Maintenance Bypass for Division D with other alarms

• Cold Leg 2 Temperature Low-2 Bypass inserted for Division A.

• Open the circuit for ECS-TE121B

Manual Reactor Trip Alarm (PMS-RXTR-MA-X0) actuate though none of the PMS divisions indicated that a Manual Reactor Trip had been inserted. PMS-J3-308 shows that the PMS-RXTR-MA alarms should only be activated by the Reactor Trip Switches at the PDSP or the RSR.

Area of Impact Reactivity Management

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10 CFR 55.45(a)(2) Manipulate the console controls as required to operate the facility between shutdown and designated power levels.

Executive Summary

Rod control issues during startup were grouped together for analysis (5736, 6089, 6102 and 6302). The letdown heat exchanger issue will impact startup due to its need to be manipulated by the Balance of Plant (BOP) operator at the same time (6158). Crews that plan ahead will be prepared for this combination of issues but the study also looked at a crew that isn’t planning ahead appropriately. There are some procedural controls in place to prevent these issues from manifesting, but the need to continue to strengthen our procedures remains.

All rod control issues in this section will apply during plant startup. Analysis took into account if the controls manipulated during startup with these rod control issues encroach on the ability to successfully manage simulator scenarios. The inverse count-rate ratio (1/M) plot may be performed manually by procedure so that wasn’t determined an issue. Issues with the Rod Insertion Limit screen are mitigated because the COLR has precedence over a graphic. Operators understand that the COLR is the definitive document on Rod Insertion Limits.

A concern with this particular combination of issues potentially disrupting operational analysis, decision making, and action was mitigated during the analysis by a belief that the crews will be able to handle these issues effectively (rod control group with letdown heat exchanger issue).

The issues with rods out of sequence alarms occurring were analyzed for operation impact during a startup. If a startup is occurring and a rod out of sequence alarm actuates, the operator may just stop and say that there won’t be a rods out of sequence during this start up and subsequently commence a shutdown and retry. Since this is an identified issue, the instructor would have to intervene to continue the startup.

For the aggregate study, an assumption was made that these issues will not manifest simultaneously or in a combination such that the students cannot dissipate the alarms in the proper order without putting rods to manual. This is backed up by nearly a year of simulator operation. These issues have been identified one at a time over a period of time. There is a need for additional reinforcement of the skills required for the task due to these issues. The instructors will need to provide additional information to the students in order to effectively deal with rods out of sequence alarms in the course of the licensed operator training program. The study recognized the risk, but determined additional skill reinforcement will allow crews to successfully manage simulator scenarios.



Disposition


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Description

While performing extended rod withdrawals during startups, depressing the rod withdrawal button (UP ARROW) may cause the UP ARROW button to un-highlight and momentarily flash gray even though still depressed. Rod motion will still occur.


Issue with Automatic Control of DST level and Auto Start of Standby Condensate Pump



Description During plant startup from Mode 5 to 100% power, the Condensate system pressure would lower to the auto start setpoint of the standby Condensate pump. This pressure drop is due to Condensate System (CDS) valves CDS-V022 and CDS-V025 modulating to maintain level in the Deaerated Storage Tank (DST). In accordance with reference plant procedures for normal operation, the second condensate pump is started at 40-45% power. However, the second condensate pump will have already auto started in the heatup and startup procedures, due to the slow response of CDS-V022 and CDS-V025.

Area of Impact Plant design deficiency impacts operations during startup.

OPDMS RIL Indication Does Not Align to COLR Rev. 0


Disposition This issue was dispositioned as unacceptable by the Simulator Review Committee (SRC). Training Needs Analysis was performed under CR 10000474. The following is taken from the analysis performed. The Training Needs Analysis determined that the M1 Bank Insertion Low-2 alarm would be received prior to the M2 Bank Insertion Low-2 alarm due to rod sequencing and bank overlap. The M1 Bank alarm is set at the correct value.

Description It is noted that the Core Operating Limits Report (COLR) Rev. 0 Rod Insertion Limit (RIL) for M2 bank at all power levels is [ ]a,c Steps (fully withdrawn). The Online Power Distribution Monitoring System

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(OPDMS) Rod Insertion Limit (RIL) display ([ ]a,c) indicates the RIL for M2 bank is [ ]a,c steps. Further investigation indicates the M2 RIL indication high limit is [ ]a,c steps and therefore cannot indicate above this level (determined using point information page instrumentation tab for RB-INSERT- M2LIM.SV3@NET0). All SD bank indications are capable of indicating a maximum of [ ]a,c steps.

Area of Impact Reactivity Management with regards to indication

NAP for 1/M Intermediate Range Not Functional

SCR-DR-6089 This issue impacts the following RO/SRO task: RO-PRO-GEN-014-00 Perform an Inverse Count Rate Plot using GOP-307

Disposition This issue was dispositioned as unacceptable by the Simulator Review Committee (SRC). Training Needs Analysis was performed. It also was determined to impact the third section of 55.45 section 3 in the aggregate. The functionality of the Intermediate Range (IR) Inverse Count Rate Ratio Nuclear Application has been fixed such that it is usable by operators but additional work will be performed by WEC to restore this to full use. WEC RITS 38306 for tracking.

Description The intermediate range 1/M plot Nuclear Application (NAP) does not work. Once P-6 (Permissive 6) was blocked and source range de-energized, the operator no longer had a 1/M plot generated.


MA Bank Rods Sometimes Stop at 263 steps during a CRE

SCR-DR-6102 This issue impacts the following RO/SRO task: RO-INC-PLS-003-06 Perform grey rod exchange


Description The plant control system operating procedure allows for a case 1 Control Rod Exchange in the event that Case 2 is not functioning. With MA rods at 234 steps and AO rods in manual at 218 steps, the AO rods were stepped into [ ]a,c steps, which will drive MA out. The MA rods stopped at 263 steps, with an outward demand still in. The audible rod clicking stopped as well.

If the CRE is still continued in accordance with the procedure, MD will step into the core and MA will remain at [ ]a,c. This generates a “Rods out of sequence alarm”.

Area of Impact

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Excessive SFW Control Valve Cycling



Description At low pressure conditions less than 350 psig, the operator often has to take manual control of Startup Feed Water (SFW) control valves due to excessive cycling of the valves. Indicated flow rates range from 0 to greater than 600 gpm within 10 to 15 second cycles. This requires 100% of the operator’s attention until RNS can be placed into service removing cooldown function from the steam dumps.


Audible Rod Step Skips

SCR-DR-6186 This issue impacts the following RO/SRO task: RO-INC-PLS-003-01 Monitor the reactor power control system

Disposition This issue was dispositioned as unacceptable by the Simulator Review Committee (SRC). A training needs analysis was performed under CR 10025690.

Explain/Brief students prior to beginning a simulator training phase or segment. Update a “SIMULATOR TRAINING STUDENT HANDOUT” (example is attached) and file in the “Operator Aids” notebook. Reference the SIMULATOR TRAINING STUDENT HANDOUT in each sum guide.

Description

During outward rod motion, the audible step counter randomly can have an extra second pause in it with rod motion continuing. The step counter indication does not update at the same rate as the audible cue occurs.


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Improper Bank Overlap Occurs when Data Point OCB07CE00C_OUTAV Increments Improperly During Startup


Disposition Subsequent to the performance of the Aggregate Study this item was determined to be invalid. This behavior is per current Westinghouse design.

Description An Ovation data point in rod control is initially set to 0 when the Digital Rod Control System (DRCS) is reset. However, during SD1 withdrawal, OCB07CE00C_OUTAV (the Ovation data point) will increment to a value of 2 and then stay at this value. This data point is only supposed to increment for inward rod motion during M bank rod movement. The end result is that bank overlap will be incorrect if not manually corrected in Ovation.



SCR-DR-6315 This issue impacts the following RO/SRO task: RO-INC-PMS-005-02 Respond to a loss of a PMS

division or failure of PMS components

Disposition

Subsequent to the performance of the Aggregate Study this item has been corrected and closed. Description The below failure sequence was performed and the reactor did not trip, but a Reactor Trip alarm was received. The alarm was for a Manual Reactor Trip, but a manual Rx Trip was not inserted. A P-4 was not received.

• RCS TE122C -> Open Circuit: PMS indicates Low-2 Tcold Cold Leg 2 Bistable Trip with Bad Quality &

Maintenance Bypass for Division C & PMS Cabinet Fault Alarm for Division C • RCS TE122D -> Open Circuit: PMS indicates Low-2 Tcold Cold Leg 2 Bistable Trip with Bad Quality &

Maintenance Bypass for Division D with other alarms • Cold Leg 2 Temperature Low-2 Bypass inserted for Division A. • Open the circuit for ECS-TE121B

Manual Reactor Trip Alarm (PMS-RXTR-MA-X0) actuate though none of the PMS divisions indicated that a Manual Reactor Trip had been inserted. PMS-J3-308 shows that the PMS-RXTR-MA alarms should only be activated by the Reactor Trip Switches at the Primary Dedicated Safety Panel (PDSP) or the Remote Shutdown Workstation (RSR).

Area of Impact

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10 CFR 55.45(a)(3) Identify annunciators and condition-indicating signals and perform appropriate remedial actions where appropriate.

Executive Summary

The 1/M NAP issues (DR-6089) was determined to be mitigated by the manual performance option. The issues with NAP Surveillance screen (DR-6159) is not a screen any of the operators use or are trained to use.

Excessive valve modulating status alarms (DR-5813) challenges responding to simulator scenarios by the operators. There are overabundances of these alarms which come in that require operator attention. Additionally, there is a risk of desensitizing operators to alarms which may ultimately be important in the plant during testing, but are not important in the simulator during training. The Initial Test Program will occur in the plant and not the simulator.

The indication issues and the alarm issues combined are significant enough to conclude that they will impact operation in the aggregate. The magnitude and influx of alarms and indications (especially when the indications and the alarms are not aligned) is too great to mitigate. Training may be able steer the operators to properly prioritize the information, but ultimately this is not a preferred way to train. Operators can’t be permitted to operate differently in the plant than the simulator. This is unacceptable.

Containment Cooling System (VCS) fan response due to loss of power

SCR-DR-216 This issue impacts the following RO/SRO task: RO-VNT-VCS-002-01 Monitor the VCS parameters

Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to the existence of a design change CAPAL 100044029 and current documentation states that the behavior is correct.

Description During normal operations, the “A” and “B” Containment Cooling System (VCS) fans are running in fast speed. Under a loss of power condition, the “A” VCS fan will trip and the “C” VCS fan should automatically start in fast speed at a low flow setpoint on VCS-FT010C. This was not observed on the STS. The “A” VCS fan does trip, but the “C” VCS fan does not auto start at a low flow. Flow indication, VCS-FT010C reads 0 cfm.

Area of Impact Containment cooling operations

EDS Power Supply Assignments to PLS/DDS Cabinets Incomplete

SCR-DR-5546 This issue impacts the following RO/SRO tasks: RO-LT-R-EDS.001 Monitor the Non Class 1E DC and UPS system (EDS) for proper operation

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RO-LT-R-EDS.004 Respond to a loss of EDS DC power Disposition

This issue was dispositioned as acceptable by the Simulator Review Committee (SRC). SMEs determined that the current power supply arrangement was adequate to teach since it is per design documentation. Power supplies are an item that will be continuously taught as they are updated and changed.

Description A loss of individual Non Class 1E DC and UPS System (EDS) busses will result in incomplete system response. Some Ovation drops (computers and other equipment that are part of the plant computer system) are not dynamically powered by the EDS model but are powered by a permanently energized model constant (specifically DPU047, DPU048, and DPU044). The load lists for the STS do not assign a power supply to all the Ovation drops so there is no plant design data to insert into the simulator.

Area of Impact Effective plant response to loss of power

Modeled BEACON Data Cannot Determine Quality

SCR-DR-5583 This issue impacts the following RO/SRO task: RO-INC-IIS-004-00 Determine functionality of the On-line Power Distribution Monitoring System

Disposition SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests. Operator training simulator guides inform the training instructors on whether BEACON is operable or not operable. (Note: This evaluation is actually based on whether or not BEACON if functional or not functional. BEACON is not safety-related and does not have applicable Technical Specifications.)

Description Failure of the BEACON Data Processing (BDP) NAP causes the manual override signal originated by BDP to have BAD quality as expected. The BAD indication is passed through the BEACON operability calculations in the Plant State Monitoring (PST) NAP and appears on the OPDMS displays as 'operable' with BAD quality. Failure of the BDP application will cause BEACON to indicate ‘inoperable’ in the reference unit. However, the current STS scope of limitation has the BEACON outputs driven by the core model. There is currently no ability to pass quality over the interface for outputs. Inputs are not taken from the BDP NAP but from other plant process data. The core model does not know the status of the BDP NAP. The core model does not pass operability information.

Area of Impact Plant control



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Description While performing extended rod withdrawals during startups, depressing the rod withdrawal button (UP ARROW) may cause the UP ARROW button to un-highlight and momentarily flash gray even though still depressed. Rod motion will still occur.

Area of Impact


Containment Radiation Alarm Reset Points Incorrect

SCR-DR-5597 This issue impacts the following RO/SRO task: RO-INC-RMS-003-07 Startup and operate a containment high-range area radiation monitor (safety related)

Disposition Acceptance criteria not fully met. SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests. This is an alarm deadband issue that WEC must resolve. If operators encounter this condition, they will follow their procedures. The procedure provides the steps necessary for operators to respond to the condition.

Description The current high setpoint and deadband combinations for Passive Core Cooling System (PXS) PXS-RY160, RY161, RY162, and RY163 do not allow for the High-1 and High-2 alarms to clear.

Area of Impact Radiation control

Pressurizer Heater Current Indicates BAD Quality at Limits

SCR-DR-5598 This issue impacts the following RO/SRO task: RO-PRI-RCS-003-03 Operate the pressurizer level control system in manual and automatic

Disposition This issue was dispositioned as acceptable by the Subject Matter Expert (SME). This issue was dispositioned as acceptable because the SME determined that the issue did not impact any operator actions during training or examination.

Description RCS-EH-04A-1-AMPB indicates 0 and BAD quality when the PZR backup heaters are off. It indicates 100 and BAD quality when the backup heaters are on. This applies for 04B, 04C, 04D points as well. All points can be viewed on Ovation display 33001.

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The control heaters (RCS-EH-03-1-AMPB) will display BAD quality at a value of 0 when they are off. These heaters are typically at some value other than 0 or 100 and the quality is good, however, it seems to have the same issue as the backup heaters at the limits.


Unidentified and Identified Leak Rate Always Indicates BAD Data

SCR-DR-5599 This issue impacts the following RO/SRO task: RO-PRO-AOP-053-00 Respond to a Reactor Coolant Leak using AOP-112 Disposition

After performing V&V testing, SNC determined that the update was not successful. The Leak Rate Monitoring Application is informational only and does not drive any alarms based upon the calculated leakage. For this reason, any leak rate calculations would have to be performed manually per plant procedures vice using the NAP calculated values.

Description The Leak Rate Monitor (LRM) has BAD quality point indication for the Identified and Unidentified leak rates. They never change to good quality and indicate BAD when using the on demand leak rate calculation.


Low Flow Alarm on TCS-FT007 Occurs Earlier than Expected

SCR-DR-5603 This issue impacts the following RO/SRO task: AP-LT-R-TCS.003 Monitor the TCS

Disposition This issue was dispositioned as unacceptable by the Simulator Review Committee (SRC). Training Needs Analysis was performed under CR 10000466. Information sharing of the discrepancy was selected as the solution to this issue.

Description A low flow alarm occurs on Turbine Building Closed Cooling Water System (TCS) TCS-FT007 at approximately 80% power during a down power evolution which is earlier than SMEs expected.


MFP 'B' Alarm Response Differs For Identical Fault

SCR-DR-5613 This issue impacts the following RO/SRO task: RO-SEC-FWS-002-01 Monitor SFWS and MFWS system

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and component parameters Disposition

Subsequent to the performance of the Aggregate Study, this item was reinvestigated at Westinghouse’ request and could not be replicated after multiple attempts. SNC performed additional investigation and determined that the original SCR entry was invalid (SNC was the entity from whom this issue originated). This item has been dispositioned as invalid and closed.

Description

A spurious trip of Main Feed Pump (MFP) 'B' responds differently than the on MFPs 'A' or 'C'. A spurious trip of MFP 'A' or 'C'’s respective supply breaker will cause three alarms: FWS-FT011A/C, FWS-FT012A/C and Feedwater Pump Control Status 1. The flow transmitter alarms are automatically taken to Cutout (CO) by the Alarm Presentation System (APS) and removed from the typical screens available to the operator but the control status alarm remains. When MFP 'B' has a spurious trip inserted on the supply breaker, 2 alarms are received: FWS-FT011B and FWS-FT012B. Since both of these alarms are automatically taken to CO the only indication is an audible noise from APS with no visual cue as to what caused the alarm. This results in a spurious trip of MFP 'B' supply breaker indicating exactly the same as if an operator took the controller to STOP.

Area of Impact

Plant control

Unexpected Response of Alarm Cutout of RWS Pressure Alarms

SCR-DR-5621 This issue impacts the following RO/SRO task: AP-RO-ADM.015 Respond to alarms using the Alarm Presentation System (APS)

Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC). The SMEs determined that adding this issue to the known issues list was adequate. The student would be aware of the issue’s existence if it ever manifested and there are a very few number of pumps which cause this to occur.

Description The Raw Water System (RWS) pump discharge pressures P003A/B/C (associated with pumps RWS-MP- 02A/B/C) each have a Low-1 alarm setpoint at 66 psig that is only active when the pump is running (i.e., alarm is cutout when the associated pump is off). However an audible nuisance alarm occurs whenever the operators take normal action on these pumps.

This “ghost alarm” occurs also during normal operation of CCS-MP-01A/B.


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Pressurizer Pressure Out of Range Indication Not Properly Displayed

SCR-DR-5623 This issue impacts the following RO/SRO task: RO-PRI-RCS-003-02 Operate the RCS/pressurizer pressure master controller in manual and automatic

Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to the isolated nature of the issue. However, this issue was found to impact operator indications in the aggregate.

Description The Primary Dedicated Safety Panel (PDSP) Reactor Coolant System (RCS) Parameters of Pressurizer Pressure (PT191A-D) stops lowering at [ ]a,c, but a low out of range arrow does not display indicating bottom of range.


Subcriticality Indication on Critical Safety Function Screen Drops to Bad Quality

SCR-DR-5627 This issue impacts the following RO/SRO task: RO-PRO-EOP-031-00 Implement and evaluate Critical Safety Function Status trees using CSF-F-0


Description The Nuclear Application (NAP) point DDS-SPD31-X0 randomly cycles to magenta bad quality due to DDS- RSA3J-J1 driving into bad quality. The result of the former point going bad quality is subcriticality and reactivity control display on the Critical Safety Function Wall Panel Information System (WPIS) screen goes magenta.





Description During plant startup from Mode 5 to 100% power, the Condensate system pressure would lower to the

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auto start setpoint of the standby Condensate pump. This pressure drop is due to Condensate System (CDS) valves CDS-V022 and CDS-V025 modulating to maintain level in the Deaerated Storage Tank (DST). In accordance with reference plant procedures for normal operation, the second condensate pump is started at 40-45% power. However, the second condensate pump will have already auto started in the heatup and startup procedures, due to the slow response of CDS-V022 and CDS-V025.

Area of Impact Plant design deficiency impacts operations during startup

Degasifier Level Alarm Limits


Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to minimal impact on training. Description CVS-M3C-100 Rev 8 states that Liquid Radwaste System (WLS) WLS-LICA-016 High-3 is the degasifier level setpoint that controls the operation of letdown. The control circuit is working as described, however the Point Information Limits show only High-1 has a value of 85”.


PMS Mimic Screens

SCR-DR-5689 This issue impacts the following RO/SRO task: RO-INC-DDS-001-00 Access plant information using the DDS (DDS is Data Display and Processing System)

Disposition

SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests. This is a simulator I&C issue. PMS mimic screens are used for verification of indications only. If a question arises regarding an indication on the PMS mimic screen, operators will use primary indications from the PMS displays on the Primary Dedicated Safety Panel (PDSP). No operator action is available through the PMS mimic screens. All actions must be taken from the division’s PMS PDSP. Description Protection and Safety Monitoring System (PMS) Mimic screens on Ovation do not reflect what is shown on the associated Primary Dedicated Safety Panel (PDSP). This is especially true when there is any FAULT condition shown on the Primary Dedicated Safety Panel (PDSP).


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Description It is noted that the Core Operating Limits Report (COLR) Rev. 0 Rod Insertion Limit (RIL) for M2 bank at all power levels is [ ]a,c Steps (fully withdrawn). The Online Power Distribution Monitoring System (OPDMS) Rod Insertion Limit (RIL) display ([ ]a,c) indicates the RIL for M2 bank is [ ]a,c steps. Further investigation indicates the M2 RIL indication high limit is [ ]a,c steps and therefore cannot indicate above this level (determined using point information page instrumentation tab for RB-INSERT- M2LIM.SV3@NET0). All SD bank indications are capable of indicating a maximum of [ ]a,c steps.


Nuisance Valve Modulating Status Alarms



Description The following valves modulations alarms cause an excessive nuisance.

• Pressurizer Spray Valves (RCS-V110A & B)

• Main Feedwater Control Valves (SGS-V250A & B)

• Startup Feedwater Control Valves (SGS-V255A & B)

Area of Impact Plant Control

Unexpected VRS High Rad Alarm

SCR-DR-5828

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This issue impacts the following RO/SRO task: RO-VNT-VRS-005-01 Respond to Radwaste building HVAC system alarms

Disposition

Subsequent to the performance of the Aggregate Study this item has been corrected and closed. Description When ES-1 is re-energized by the diesel sequencer after a loss of power incident, the Radwaste Building HVAC System (VRS) VRS-RY023 high radiation alarm will actuate. During this actuation, process flow through VRS is unavailable per design. Only the detector receives power.

Area of Impact Radiation Control

VFD Transformer Temperature

SCR-DR-5910 This issue impacts the following RO/SRO task: RO-PRI-RCS-005-01 Operate RCP VFDs

Disposition


Description ECS-EV-X1-TMPC points associated with the Variable Frequency Drive (VFD) hottest cell parameters are not being driven by the models. They are a constant value.

Area of Impact Plant Operations

Print Feature from NAP non-functional

SCR-DR-5913 This issue impacts the following RO/SRO task: AP-RO-ADM.020.07 Document surveillance test in log book

Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to minimal training impact as this simply means the automated system is not capable of being used to complete surveillance requirement testing, operators are still capable of using the paper copies.

Description While performing surveillance procedure "Incore Detector Comparison to Nuclear Instrument Channel Axial Flux Difference", the select PRINT SURVEILLANCE REPORT does not result in a printout.

Area of Impact Plant response – this prevents or impacts the performance of most Surveillance Tests

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VHS Rad Monitor Response to Loss of Process Flow

SCR-DR-5914 This issue impacts the following RO/SRO task: RO-VNT-VHS-005-01 Respond to health physics and hot machine shop HVAC alarms

Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed. Description Health Physics and Hot Machine Shop HVAC System (VHS) VHS-RE001 goes up by 4 decades in 10 minutes on a loss of process flow. This gives a Priority 1 alarm on VHS and RADIATION MONITORING.

Area of Impact Radiation Control

Pressurizer Narrow Range Pressure Does Not Indicate Bottom of Scale

SCR-DR-5920 This issue impacts the following RO/SRO task: RO-PRI-RCS-003-02 Operate the RCS/pressurizer pressure master controller in manual and automatic


Description With RCS pressure less than [ ]a,c, the Wall Panel Information System (WPIS) for Mode 1-4 does not indicate the instrument is bottom of scale via graphical down arrow.


CDS-TE040A/B Range is Inadequate

SCR-DR-5921 This issue impacts the following RO/SRO task: RO-SEC-BDS-005-07 Verify automatic blowdown isolation upon high-2 temperature in heat exchanger shell outlet (CDS fluid) or high-2 DST level

Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) since an open item exists in the design documentation.

Description When the condensate outlet of the blowdown heat exchanger temperature element is failed high the blowdown flow remains un-isolated. The high-2 temperature ([ ]a,c) should isolate blowdown flow in accordance with APP-CDS-M3C-101 Rev 3. However, the range of the instrument (CDS-T-040A/B) listed in APP-CDS-M3C-101 Rev 3 is [ ]a,c which would never allow blowdown isolation on high temperature.

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DRPI Health Screen Alarms for Data Cabinet A and B Crossed

SCR-DR-5924 This issue impacts the following RO/SRO task: RO-INC-DDS-001-00 Access plant information using the DDS Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to not being a frequently used indication and the associated alarm points functioning properly.

Description The Digital Rod Position Indication (DRPI) Health screen (1805) has the following Data Cabinet alarms addressing the wrong point:

1) "A (-15V)" is addressing RM-DATAB4-ALM.SV3 and it should be addressing DATAA4 2) "A (+15V)" is addressing RM-DATAB3-ALM.SV3 and it should be addressing DATAA3

3) "B (-15V)" is addressing RM-DATAA4-ALM.SV3 and it should be addressing DATAB4 4) "B (+15V)" is addressing RM-DATAA3-ALM.SV3 and it should be addressing DATAB3


Digital Rod Position Indication (DRPI) Health Screen Incorrect Logic Cabinet Alarms

SCR-DR-5925 This issue impacts the following RO/SRO task: RO-INC-DDS-001-00 Access plant information using the DDS. Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed.

Description

The DRPI Health Screen (1805) should have the following Logic Cabinet Alarms with associated points:

1) "A (-15V)" point number RM-DCLON15VA-ALM.SV3 2) "A (+15V)" point number RM-DCLOP15VA-ALM.SV3 3) "B (-15V)" point number RM-DCLON15VB-ALM.SV3 4) "B (+15V)" point number RM-DCLOP15VB-ALM.SV3


Containment Recirculation Actuation Indication Issue

SCR-DR-5972

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This issue impacts the following RO/SRO task: RO-INC-PMS-003-00 Monitor System Indications

Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to minimal impact on training.

Description Once containment recirculation is actuated, the actuation indication for Divisions C and D did not have the white box with an X on the ESF Act Status Screen for the divisional PDSPs or the Non-Safety Operational Overview screen (33020). The individual PMS division screen for CNMT Recirc actuation (IRWST/INJT Recirc) did show that it had been actuated on all 4 divisions.

Area of Impact Verifying plant response

Uncontrolled Heat-up (H/U) Indication Incorrect

SCR-DR-6009 This issue impacts the following RO/SRO task: RO-PRO-GEN-008-00 Perform Plant Cooldown from Mode 5 to Refueling Mode using GOP-206 Disposition


Description The "Uncontrolled HU or CD" (point DDS-SSF28-X0) indication on screen 60032 does not appear to change whether RCS temps are stable or changing. The only time the point driving the uncontrolled HU/CD indication would change state was when the Plant Mode Control (Screen 40003) was cycled to Manual Mode 4 during a RCS heatup IC in Mode 4. The controller was then cycled back to Auto and that uncontrolled HU/CD light energized. These same actions were repeated with a 100% steady-state IC, mode 5 IC and mode 3 IC without any changes in the indication.


DHC Summary - Assembly Move NAP Function Not Functional

SCR-DR-6022 This issue impacts the following RO/SRO task: RO-INC-PLS-004-03 Perform decay time surveillance

Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to the minimal training impact this particular Nuclear Application (NAP) has.

Description When attempting to simulate fuel assemblies being moved from the core to the Spent Fuel Pool (SFP), it was noted that the Decay Heat Calculation (DHC) NAP to maintain the administrative location of fuel does not work correctly. On display 40203 the assembly move buttons on the lower right portion

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indicate they are only available when the light DDS-AP-DHC Status indicates it is ACTIVE. This light is driven off of the automatic mode selector and is INACTIVE when in MODES 1&2 and ACTIVE in MODES 3- 6. However, when the light indicates INACTIVE the buttons for moving are raised and available. When the light changes status to ACTIVE the buttons for moving are grayed out and no longer available. The light being active or inactive is currently driven by the auto mode selector and becomes active in MODES 3-6. However, fuel cannot be moved from the core into the SFP in any MODE other than MODE 6. The light should be driven by the manual input of the Rx vessel head being removed or installed or upper internals position on display 40004.


RCP Vibration Alarm Naming

SCR-DR-6025 This issue impacts the following RO/SRO task: RO-PRI-RCS-008-00 ****Respond to RCS alarms


Description Reactor Coolant Pump (RCP) Vibration alarms are received at [ ]a,c. When looking at the individual RCP display (12105 for example) the bottom right corner has a light indication that displays "RCP 1A Vibration" and is grayed out when no alarming condition is met. When any vibration monitor goes above [ ]a,c mils the light will illuminate and change to "RCP 1A High-1". When the H2 setpoint is reached the light [ ]a,c alarm even though the H2 alarm is a Pri-2 alarm. The small button poke next to the light will also be available and will provide indication that both the H1 and H2 alarms are in. The point identifier is a good indication that these alarms are HIGH alarm (have H1 or H2 in the identifier). However, the Point information has L1 and L2 in the descriptions for the H1 and H2 alarms. This is an error likely naming convention as the L1 and L2 could mislead assumptions to LOW1 and LOW2.


HSS Display does not Include ESOP Discharge Pressure

SCR-DR-6078 This issue impacts the following RO/SRO task: RO-SUP-HSS-002-00 Operate the HSS


Description Hydrogen Seal Oil System (HSS) HSS-MP02 discharge pressure HSS-PT017 is not on the HSS display (15100). APP-HSS-M6-001 Rev 2 indicates that the transmitter should have an available point reference in ovation (PT-017 has a box with PIA - Pressure Indication/Alarm).

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NAP for 1/M Intermediate Range Not Functional

SCR-DR-6089 This issue impacts the following RO/SRO task: RO-PRO-GEN-014-00 Perform an Inverse Count Rate Plot using GOP-307

Disposition This issue was dispositioned as unacceptable by the Simulator Review Committee (SRC). Training Needs Analysis was performed. It also was determined to impact criterion (3) of 55.45(a) in the aggregate. The functionality of the Intermediate Range (IR) Inverse Count Rate Ratio Nuclear Application has been fixed such that it is usable by operators but additional work will be performed by WEC to restore this to full use. WEC RITS 38306 for tracking.

Description The intermediate range 1/M plot NAP does not work. Once P-6 (Permissive 6) was blocked and source range de-energized, the operator no longer had a 1/M plot generated.


ECS Penetration Temperature off Scale Low

SCR-DR-6103 This issue impacts the following RO/SRO task: AP-LT-R-ECS.005 Monitor the ECS

Disposition

Subsequent to the performance of the Aggregate Study this item has been corrected and closed. Description ECS penetration temperature reading is off scale low on display 22503. This is for the penetration to containment for the power cables for the RCPs as indicated on ECS-TE001A/B and TE002A/B which currently show the electrical penetration temperature as 0 degree F. The temperature should be reading something slightly higher than the ambient conditions.


Plant Mode Selector NAP Inconsistent with Procedure

SCR-DR-6144 This issue impacts the following RO/SRO task: RO-PRO-GEN-016-00 Perform Plant Power Escalation From 2% Power to 100% Power using GOP-306

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Description In accordance with reference procedures, Mode 2 is entered when all Axial Offset (AO) bank rods are off the bottom. Currently Plant Control System (PLS) Auto Plant Mode selector changes from Mode 3 to Mode 2 when the RTBs are closed (P-3 is cleared).

Area of Impact Plant Startup

NAPS display issues



Description Several human factor related issues exist on the NAP surveillance screens. They involve unexplained acronyms, grammar errors, and inconsistent color coding.

Area of Impact Accessing information from DDS

WPIS Downscale Arrow Absent

SCR-DR-6164 This issue impacts the following RO/SRO task: RO-PRI-RCS-005-03 ****Cool down the pressurizer

Disposition This is a backup indication to alert the operator that the instrument is at its lower limit; the numeric indication is still available. This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to the isolated issue and minimal impact on training once students are briefed on issue. Description No downscale arrow on Wall Panel Information System (WPIS) trend display (mode 3 / 4) exists for Tavg when bottom of scale.


RSA NAP Does Not Process Failed Channels Correctly


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ZVS and ZBS Alarm Scaling Incorrect

SCR-DR-6171 This issue impacts the following RO/SRO task: RO-ELE-ZVS-001-00 Respond to Excitation and Voltage Regulation System (ZVS) abnormalities

Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due the fact that there are other faults available that will result in the alarm notification to the operator.

Description The large APS tile EXCITATION VOLTAGE REG or any ZVS alarm will not come in on a regulator failure which causes voltage to peg high. The alarm setpoints in the database are maxed out so alarm will never come in on these parameters at this point.


Flux doubling difference between divisions

SCR-DR-6175 This issue impacts the following RO/SRO task: RO-INC-PMS-003-00 Monitor System Indications Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) as the protective functions associated with these signals will still occur.

Description The alarms/alarm response for A/D Divisions differs significantly from B/C divisions. A/D divisions activates at 1.6 in 50 seconds whereas B/C at 2.2 in 10 seconds. J3 documents specify 2.2 in 10 sec.

Area of Impact Safety System Operation

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Time to Boil Calculation

SCR-DR-6179 Disposition

SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests. The Time to Boil NAP is a tool that is used for information only. The NAP is active when in Mode 5/6 conditions for RCS time to boil and when fuel is present in the spent fuel pool for spent fuel pool time to boil. The value is displayed in exponential units versus hh:mm. This means that operators will need to convert the scientific notation values into hours and minutes. Although this does take a short amount of time, the net effect is that it does not remove the ability to monitor the time to boil and it has no impact on actions the operator may, or may not, take in response to plant conditions. Description When core exit temperature was 300oF, Thot was > 212oF and TTB was >0 (15-25 min range). Either the NAP is calculating Time to Boil (TTB) incorrectly or the inputs to the NAP are wrong. TTB should reflect actual plant conditions.

Area of Impact NAP




Explain Brief students prior to beginning a simulator training phase or segment. Update a “SIMULATOR TRAINING STUDENT HANDOUT” (example is attached) and file in the “Operator Aids” notebook. Reference the SIMULATOR TRAINING STUDENT HANDOUT in each sim guide.

Description During outward rod motion, the audible step counter randomly can have an extra second pause in it with rod motion continuing. The step counter indication does not update at the same rate as the audible cue occurs.


WPIS Display VARs

SCR-DR-6190

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This issue impacts the following RO/SRO task: RO-ELE-ZAS-002-05 Maintain generator power factor and reactive load within acceptable ranges

Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to the fact that the numeric value displayed in all locations is the same while only the units of measurement change.

Description The Wall Panel Information System (WPIS) display has VARS indicated rather than Mega Volt-Amps Reactive (MVARS) for Generator Output.


CMT WR Level Indications go Bad Quality

SCR-DR-6217 This issue impacts the following RO/SRO task: AP-LT-S-EOP.007 Direct implementation of E-1, AP1000 Loss of Reactor or Secondary Coolant

Disposition The Aggregate Study team determined this issue as not impacting simulator training due to only providing indication function. All protective functions are still available via the CMT Narrow Range level instruments.

Description The Wide Range (WR) Core Makeup Tank (CMT) level indications shift to Bad Quality once Automatic Depressurization System 1-3 (ADS 1-3) Actuate. Prior to this event, they would toggle to Bad Quality intermittently. The Bad Quality status is on indications PXS-LT009A/B & -LT010A/B (on Passive Core Cooling System (PXS) Supplemental Ind. Screen) and DDS-RSA11-L1 & DDS-RSA13-L1 (on WPIS screen 60017). The NAP driving the calculation of this indication drives them to bad quality whenever it determines voiding is occurring in the CMT (which is expected per design transients).

Area of Impact Plant control during decay heat removal

Unexpected Bank Sequence Out of Sequence Alarm


Disposition

The Bank Out of Sequence alarm and corresponding Alarm Response Procedure drives operators to perform AOP-104, “Rod Control Malfunction.” This will lead the crew to ensure all equipment is operating properly and to confirm that the rod alignment requirements of the Core Operating Limits Report are

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met. This issue was dispositioned as acceptable by the Simulator Review Committee (SRC). Description A Bank Sequence Out of Sequence (DDS-RSU01-X0) alarm actuates anytime banks M1 and MD (MA) are in overlap. When M1 and MD (MA) are in overlap the NAP generates an alarm showing bank M2 as being Out of Sequence (OOS). The NAP may be incorrectly calculating the OOS condition.

Area of Impact Reactivity control

Urgent Alarm Occurs During Case 2 CRE

SCR-DR-6267

This issue impacts the following RO/SRO task: RO-INC-PLS-003-06 Perform grey rod exchange

Disposition SNC evaluated this issue and determined that it does not impact the suitability of the simulator for the conduct of operating tests.

This is an AP1000 plant design issue. The simulator models the plant design. If operators encounter this condition, they will follow procedural guidance. The procedures provide the steps necessary for operators to respond to the event.

Description The Urgent Failure Alarm (UA) occurs when MA and MD banks are in motion and the Tavg-Tref deviation requires the AO rods to move to restore Tavg-Tref back into band. This only occurs if MA and MD rods are in motion. For plant conditions where only the MA or MD rods are in motion and the Tavg-Tref deviation requires AO rods to move, then an UA does not occur.

The UA appears to be a timing issue that occurs only when MA and MD banks are both in motion when the Tavg-Tref deviation occurs. Basically, the Ovation controllers briefly generate a RODS IN and a RODS OUT signal to the MA bank and a RODS IN and a RODS OUT signal to the MD bank which results in a UA from the Power Cabinets.





Description

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An Ovation data point in rod control is initially set to 0 when the Digital Rod Control System (DRCS) is reset. However, during SD1 withdrawal, OCB07CE00C_OUTAV (the Ovation data point) will increment to a value of 2 and then stay at this value. This data point is only supposed to increment for inward rod motion during M bank rod movement. The end result is that bank overlap will be incorrect if not manually corrected in Ovation.








& Maintenance Bypass for Division C & PMS Cabinet Fault Alarm for Division C • RCS TE122D -> Open Circuit: PMS indicates Low-2 Tcold Cold Leg 2 Bistable Trip with Bad Quality

& Maintenance Bypass for Division D with other alarms





Controller Fault Alarms Received on Turbine Trip



Description

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The Priority 4 alarms for Controller 34 (Drop 34) occur for Turbine Trips from 100%, 75% and 50% power. A Controller 21 (Drop 21) alarm occurs after the Turbine trip from 100% power.

Area of Impact Alarm Management

Diesel Fuel Oil Day Tank Level Transmitter Operation

SCR-DR-6491 This issue impacts the following RO/SRO task: RO-SUP-DOS-001-00 Operate Standby Diesel Fuel Oil System (DOS)

Disposition SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests. The initial tank levels established in the initial conditions provides suitable inventory for at least 2 hours of operation before refill of the tank would initiate at the incorrect level. Most scenarios are established such that the scenario would be complete prior to this refill level being achieved and the issue does not result in a loss of the DG. In addition, there is no procedural guidance that would direct an operator to verify the day tank level or proper operation of the day tank level control system. SNC Simulator Group continues to investigate the issue.

Description Diesel Fuel Oil System (DOS) level transmitters DOS-LT016A/017A and 016B/017B on the day tank control the refilling of the day tank based on level. The refilling should start when day tank level reaches low level ([ ]a,c) and stop at high level ([ ]a,c). The refilling of the day tank actually begins at 44.67% and stops at 100%. Additionally as level rises at ~85% the level indication jumps to 100%.


Inconsistent UAT Line Voltage Alarm Priorities


Disposition Subsequent to the performance of the Aggregate Study this item was retested after the new APS was loaded. All UAT Undervoltage alarms except ES-7 had the proper priority assigned to them, ES-7 is still defined as a priority 1 alarm. The improper alarm priority being assigned to ES-7 does not impact the simulator’s suitability for the conduct of operating tests. The operator response to this alarm is consistent with ES-1 through ES-6 UAT Line Undervoltage alarm response procedures.

Description While performing a LOOP with Fast Bus Transfer to Reserve Auxiliary Transformer (RAT), it was noted that the Unit Auxiliary Transformer (UAT) Breaker Line Undervoltage Alarms for ES-2, 3 and 6 come in as Priority 2 alarms. Alarms for ES-4, 5, and 7 come in as Priority 1 alarms. These should be consistent.

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Area of Impact Plant Indication

Any Rods at Bottom Alarm

SCR-DR-6532 This issue impacts the following RO/SRO task: RO-INC-PLS-005-01 Respond to control rod position alarms

Disposition Subsequent to the performance of the Aggregate Study the simulator group developed a method for ensuring this issue is transparent to the to operators. Specific data points in the I&C architecture are configured in the Initial Conditions File prior to the start of a scenario such that the alarm does not appear. The deficiency remains open pending final correction of the underlying issue from WEC. This does not impact the simulator’s suitability for the conduct of operating tests. Description The "Any Rods at Bottom" alarm is actuating anytime rods are being driven through [ ]a,c steps. Based on APP-PLS-J1-023 Rev 2 3.1.2 Rev. 16 [

]a,c.


WGS Sample Package Ovation Interface

SCR-DR-6612 This issue impacts the following RO/SRO task RO-SUP-WGS-003-00 Monitor WGS operation

Disposition

This issue does not impact the suitability of the simulator for the conduct of operating tests.

The functions associated with PS-001 are covered by APP-MS27-M6-001, APP-MS27-E5-001 and APP-WGS-M3C-101. These different design documents provide conflicting guidance as to what functions should and should not be present. The simulator modeling appears to be per APP-WGS-M3C-101 which does not include any functions based on PS-001. There is not any procedural guidance in place to take any actions based upon the indications that would be provided if PS-001 were modeled.

Description

Per drawing APP-MS27-E5-001 & APP-MS27-M6-001, PS-001 for monitoring N2 pressure to the Gaseous Radwaste System (WGS) Sample Package should be modeled to give an Ovation alarm when pressure decreases below 60 psig. PS-001 does not seem to be monitored in Ovation or the WGS model. Per the listed references, PS-001 will generate an Ovation alarm on low pressure or loss of power by de- energizing relay CR-3.

Area of Impact

WGS Indications

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WGS Sample Package Digital Indication


Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed. Description Per drawing APP-MS27-E5-001, APP-MS27-M6-001, APP-MS27-VMM-004 pages 20 & 354, APP-WGS- MC3-101 page 16, H2 monitor AT032 (AE032) provides only a digital output.

Ovation drawing 16100 shows WGS-AT032 as having continuous indication. This continuous indication is inferred when looking at APP-MS27-M6-001, APP-WGS-M3C-101 page 23, and APP-WGS-M6-001. However per APP-MS27-VMM-004 page 354, AE032 provides only a digital output via a normally closed contact.

Area of Impact WGS Indications

RSA NAP for Power Range Power does not Eliminate Erroneous Input



Description When the power range (PR) B lower detector fails high, the Redundant Sensors Algorithm Application (RSA) NAP for Power Range Power does not eliminate this input. This causes an erroneous PR power reading on the WPIS. Area of Impact Off Normal Event Response

Inconsistent DPU Alarm Priority Levels



Description Controller 39 Alarm (DROP39_) is ranked as Priority 1 while the other Controller alarms are all Priority 4. The data process unit (DPU) alarms are all Priority 4 (including DPU 39).

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Safety Mimic Display for SGS-V255A& B Indicates Bad Quality Following a SFW Isolation SCR-DR-6698 This issue impacts the following RO/SRO task: RO-INC-DDS-001-00 Access plant information using the DDS


Description The PLS Safety Mimic display for all 4 divisions indicate bad quality for SGS-V255A&B following a Startup Feedwater System (SFW) Isolation Signal. The valve is closed as verified by the FW Components Status tab on the PDSP.


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10 CFR 55.45(a)(4) Identify the instrumentation systems and the significance of facility instrument readings.

Executive Summary

The inability of operators to understand the operational significance of their indications due to the impact from these issues could cause a delay in tripping the unit or prevent them from tripping the unit. By the definition of safety, this would cause a failure and is not acceptable. Additional reinforcement during training of the operator skill set is required to mitigate this effect of particular issues. The issues affecting the operators are the Wall Panel Information System, Redundant Sensors Algorithm Application, Nuclear Application (WPIS RSA NAP) indications (especially Power Range (PR) and Intermediate Range (IR)).

The study team believes that the current issues impacting instrumentation will not preclude successful and safe operation. The team feels that the issues in this section have no aggregate impact. With proper training and reinforcement of skills, there is no overall impact on safety.

Stage 3 ADS Box Unused on Divisions C and D


Disposition SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests. The simulator has N/A instead of a box for Stage 3 ADS. A review of current procedures confirms that the Division C and D Stage 3 ADS signal is not required. Therefore, there is no impact to the operator. Description The PMS divisions C and D indication for Stage 3 ADS status do not ever indicate actuation status.


Subcriticality Indication on Critical Safety Function Screen Drops to Bad Quality

SCR-DR-5627 This issue impacts the following RO/SRO task: RO-PRO-EOP-031-00 Implement and evaluate Critical Safety Function Status trees using CSF-F-0


Description The Nuclear Application (NAP) point DDS-SPD31-X0 randomly cycles to magenta bad quality due to DDS- RSA3J-J1 driving into bad quality. The result of the former point going bad quality is subcriticality and reactivity control display on the Critical Safety Function Wall Panel Information System (WPIS) screen goes magenta.

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VWS-TE079 Point Named Incorrectly

SCR-DR-5643

This issue impacts the following RO/SRO task: AP-LT-R-VWS.007 Monitor VWS parameters


Description Central Chilled Water System (VWS) VWS-TE079 is the inlet temperature for VWS Low Capacity Chiller #2. However, the point name is currently Low Cap Chiller 3 Inlet Temp.


Calorimetric Data Precision

SCR-DR-5712 This issue impacts the following RO/SRO task: RO-INC-PMS-013-04 Perform MCR actions associated with calorimetric calibration of the excore NIS power range instrumentation


Description The following points do not have the sufficient precision to perform the startup to 100% procedure steps regarding verification of power.

DDS-PPP08-J0 (SG FW RTO) DDS-PPP08-J0-AVP (SG RTO 1 Hr)


Inconsistent OPDMS QPT Indications

SCR-DR-5903 This issue impacts the following RO/SRO task: RO-INC-IIS-005 Respond to On-line Power Distribution Monitoring System (OPDMS) malfunctions.

Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to the specific nature of the tasks associated with this screen.

Description During dropped control rods at core locations C7 and H8, it was noted that the indications provided by

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the OPDMS Excore and Incore Quadrant Power Tilt (QPT) monitors were not consistent. The following observations were made:

• The Excore display lists the detectors as N41-N44. A detailed search of AP1000 documentation

found no reference material in which the PR excore detectors are referred to by N41-N44 except on page M-4 of APP-OCS-J4V-207, “Operation and Control Centers Display Design Document for Online Power Distribution Monitoring System.” Page M-4 has a table showing that OPDMS has N41, N42, N43, and N44 mapped to PR C, D, B, and A respectively, which appears to be consistent with what was seen in the PRS.

• The values for all of the excore detectors read 1.0 which appears to be just some sort of default. There are two problems with this. 1) One decimal point worth of data is not enough to adequately assess QPTR. 2) In this scenario, they should definitely not be reading 1.0.

• On the values displayed on this graphic, there are 8 labels titled “PR Power Upper Detector”, with no additional label as to what division.

• For the Incore QPT display, the letters in the corners do not match up with the data. For example, in this scenario the flux shifted towards PR A and C but the display shows it greatest near A and D and suppressed at C.


VFD Transformer Temperature

SCR-DR-5910 This issue impacts the following RO/SRO task: RO-PRI-RCS-005-01 Operate RCP VFDs

Disposition


Description ECS-EV-X1-TMPC points associated with the Variable Frequency Drive (VFD) hottest cell parameters are not being driven by the models. They are a constant value.

Area of Impact Plant Operations




Description

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When the condensate outlet of the blowdown heat exchanger temperature element is failed high the blowdown flow remains un-isolated. The high-2 temperature ([ ]a,c) should isolate blowdown flow in accordance with APP-CDS-M3C-101 Rev 3. However, the range of the instrument (CDS-T-040A/B) listed in APP-CDS-M3C-101 Rev 3 is [ ]a,c which would never allow blowdown isolation on high temperature.


CVS-PT040 does not Provide Proper Protective Functions

SCR-DR-5968

This issue impacts the following RO/SRO task: RO-PRI-CVS-004-00 Monitor CVS operations


Description CVS-PT040 (pressure transmitter upstream of the letdown control valve) does not provide the proper protective functions for low pressure and high pressure protection in accordance with design documentation.

Per APP-CVS-M3C-101 Rev 6 Appendix C.3.2, at [ ]a,c when CVS-V047 is in automatic pressure control mode there is supposed to be a signal sent to “trip the CVS Makeup Pumps.” The high pressure signal is generated but does not trip the pumps; it presently feeds a Pump Auto Stop Demand signal. This signal will stop any pumps that are running in automatic only. When the plant is in water-solid mode, as determined in logic diagrams as having CVS-V047 in automatic pressure control mode, Chemical and Volume Control System (CVS) makeup pumps must be operated in manual; see APP-CVS-M3C-100 Rev 11 Logic Sheet CVS-4 Note10. Since the CVS makeup pumps are in manual the Auto Stop Demand signal will not shut the pump off to provide overpressure protection.

Per APP-CVS-M3C-101 Rev 6 Appendix C.3.2, at [ ]a,c when CVS-V047 is in automatic pressure control mode there is supposed to be a signal sent to “trip the Reactor Coolant Pumps…in order to protect them from reduced suction pressure.” This signal is also discussed in APP-RCS-M3C-100 Rev 9 Logic Sheet RCS-13 table and note 11. As presently designed there is no logic tie between CVS-PT040 and the Reactor Coolant Pumps (RCPs) to prevent damaging the RCPs upon a loss of Reactor Coolant System (RCS) pressure.


M Banks B & C Reversed on DRPI Health Screen


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Description The DRPI Health Screen ([ ]a,c) control rod banks M-B and M-C have the wrong rods listed as being in each bank. The rods listed as being M-B are actually M-C and the rods listed for M-C are the M-B rods. The correct arrangement of rods is shown in APP-RXS-M3-001 Rev 4 Figure 4-1 as well as on the DRPI M Bank screen ([ ]a,c).


Quality of RWS-V503 BAD at Limits

SCR-DR-6038 This issue impacts the following RO/SRO task: RO-SUP-RWS-005-00 Monitor the Raw Water System

Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due minimal training impact based on still having other indications of the valve position available.

Description RWS-V503 and Circulating Water System (CWS) CWS-V514 are BAD quality at valve operating limits (open and closed).


Reactor Coolant Pump (RCP) Stator Temperature Indication off Scale Low at Lower Speeds

SCR-DR-6071 This issue impacts the following RO/SRO task: RO-PRI-RCS-003-01 Monitor the RCS during steady-state- power operation of the plant

Disposition

Subsequent to the performance of the Aggregate Study this item has been corrected and closed. Description RCP Stator Temperature indication is off scale low at lower speeds. With RCPs at 50% speed, the stator temperature (RCS-TE271, 272, 273, 274) indicates off scale low of 50F 'V'. This does not appear to be a valid temperature reading as the CCS temperature is 72F, SG cubicle temperature is 72F, and bearing temperature is 82F. With the ambient temperature above 50 degrees, it would be expected to have the motor at the same or slightly higher temperature.


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HSS Display does not Include Emergency Seal Oil Pressure (ESOP) Discharge Pressure



Description Hydrogen Seal Oil System (HSS) HSS-MP02 discharge pressure HSS-PT017 is not on the HSS display (15100). APP-HSS-M6-001 Rev 2 indicates that the transmitter should have an available point reference in ovation (PT-017 has a box with PIA - Pressure Indication/Alarm). Area of Impact Plant Control

DWS-LT006 has Insufficient Range

SCR-DR-6099 This issue impacts the following RO/SRO task: RO-SUP-DWS-002-00 Monitor DWS operation

Disposition The simulator is modeling the plant as currently designed; once the design has been updated the simulator model will be updated as well. This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to minimal impact on training.

Description Demineralized Water Transfer and Storage System (DWS) DWS-LT006 is the level indication for the Condensate Storage Tank (CST). This level spans [ ]a,c per APP-DWS-M3C-101. However, the calculation note for DWS (APP-DWS-M3C-002) has the high 2 alarm at [ ]a,c. This is an important alarm because it is designed to give operators time to determine why the CST is full before it overflows to a drain. Overflow will occur at [ ]a,c.





Description The plant control system operating procedure allows for a case 1 Control Rod Exchange in the event that Case 2 is not functioning. With MA rods at 234 steps and AO rods in manual at 218 steps, the AO rods were stepped into [ ]a,c steps, which will drive MA out. The MA rods stopped at 263 steps, with an outward demand still in. The audible rod clicking stopped as well. If the CRE is still continued in

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accordance with the procedure, MD will step into the core and MA will remain at [ ]a,c. This generates a “Rods out of sequence alarm”.


ECS Penetration Temperature off Scale Low


Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed. Description

ECS penetration temperature reading is off scale low on display 22503. This is for the penetration to containment for the power cables for the RCPs as indicated on ECS-TE001A/B and TE002A/B which currently show the electrical penetration temperature as 0 deg F. The temperature should be reading something slightly higher than the ambient conditions.


Improper function of C-2 reactor power control interlock

SCR-DR-6122 This issue impacts the following RO/SRO task: RO-INC-PLS-008-00 Respond to PLS - related abnormalities

Disposition

Subsequent to the performance of the Aggregate Study this item has been corrected and closed. Description C-2 control interlock utilizes a non-conservative power input for operation. It currently uses BDP corrected power which is lower than NI power during over power scenarios.


WPIS RCS Inventory Issues


Disposition This is a graphic issue that does not effect operational decisions and is representative of the information in GOP-114. This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to

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the fact the procedure will always be the preferred reference.

Description The WPIS RCS Inventory screen has reference level lines for hot leg top and bottom which appear incorrect. They are only 18 inches apart. It is, however a faithful reproduction of the chart in GOP-114. The procedure and display both need to be looked at. Issues found in calculation notes, procedure and display.


WPIS Downscale Arrow Absent

SCR-DR-6164 This issue impacts the following RO/SRO task: RO-PRI-RCS-005-03 ****Cool down the pressurizer

Disposition This is a backup indication to alert the operator that the instrument is at its lower limit; the numeric indication is still available. This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to the isolated issue and minimal impact on training once students are briefed on issue. Description No downscale arrow on WPIS trend display (mode 3 / 4) exists for Tavg when bottom of scale.


Tuning of VBS Required for Stability

SCR-DR-6168 This issue impacts the following RO/SRO task: RO-VNT-VBS-001-15 Purge smoke from MCR/CSA

Disposition This issue was dispositioned as unacceptable by the Simulator Review Committee (SRC). A training needs analysis was performed under CR 10025689. SNC evaluated this issue and determined that even though it is unacceptable as a design issue, it does not impact actions taken by the operator.

Therefore, it does not impact the suitability of the simulator for the conduct of operating tests.

Description During Main Control Room (MCR) purge operations, Nuclear Island Nonradioactive Ventilation System (VBS) air handling unit trains cannot maintain stable flow and as a result, enter an indefinite cycling between two trains. The current tuning does not allow enough time to establish stable flow.


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Condensate Polisher Bypass Valve Control

SCR-DR-6172 This issue impacts the following RO/SRO task: RO-SEC-CPS-002-02 Place CPS in service

Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to minimal impact on training and examination as the chemistry levels in the secondary plant are transparent to operators.

Description CPS-V001 (CDS Polisher Bypass Valve) Setpoint controls are confusing. The procedure directs placing the controller in auto and never has a setpoint to control to. The current setpoint is set at the high end of the scale, so the bypass valve will never modulate closed. The calc note states that signals will be set based on CDS header and polisher flow. No setpoint is yet determined.



SCR-DR-6179 Disposition

SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests. The Time to Boil NAP is a tool that is used for information only. The NAP is active when in Mode 5/6 conditions for RCS time to boil and when fuel is present in the spent fuel pool for spent fuel pool time to boil. The value is displayed in exponential units versus hh:mm. This means that operators will need to convert the scientific notation values into hours and minutes. Although this does take a short amount of time, the net effect is that it does not remove the ability to monitor the time to boil and it has no impact on actions the operator may, or may not, take in response to plant conditions.

Description

When core exit temperature was 300oF, Thot was > 212oF and TTB was >0 (15-25 min range). Either the NAP is calculating time to boil (TTB) incorrectly or the inputs to the NAP are wrong. TTB should reflect actual plant conditions.

Area of Impact NAP



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Description The WR Core Makeup Tank (CMT) level indications shift to Bad Quality once Automatic Depressurization System 1-3 (ADS 1-3) Actuate. Prior to this event, they would toggle to Bad Quality intermittently. The Bad Quality status is on level transmitter indications PXS-LT009A/B & -LT010A/B (on PXS Supplemental Ind. Screen) and DDS-RSA11-L1 & DDS-RSA13-L1 (on WPIS screen 60017). The NAP driving the calculation of this indication drives them to bad quality whenever it determines voiding is occurring in the CMT (which is expected per design transients).



SCR-DR-6315 This issue impacts the following RO/SRO task: RO-INC-PMS-005-02 Respond to a loss of a PMS division or failure of PMS components

Disposition

Subsequent to the performance of the Aggregate Study this item has been corrected and closed. Description The below failure sequence was performed and the reactor did not trip, but a Reactor Trip alarm was received. The alarm was for a Manual Reactor Trip, but a manual Rx Trip was not inserted. A P-4 was not received.


& Maintenance Bypass for Division C & PMS Cabinet Fault Alarm for Division C

• RCS TE122D -> Open Circuit: PMS indicates Low-2 Tcold Cold Leg 2 Bistable Trip with Bad Quality & Maintenance Bypass for Division D with other alarms





Main Generator Output breaker logic

SCR-DR-6392

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This issue impacts the following RO/SRO task: RO-ELE-ZAS-002-04 Synchronize the generator in manual/auto modes

Disposition

SNC evaluated this issue and determined that since the MCR operator will not see this anomaly when the procedure is followed; it does not impact actions taken by the operator. Therefore, it does not impact the suitability of the simulator for the conduct of operating tests. Description A discrepancy in the turbine generator synchronization logic results in the following:

• IF you have the ‘acknowledge ready for auto-sync’ poke selected or not, the Generator breaker will close when ‘GEN’ is selected and operator action ceases for ~90 seconds. If you continue in a timely manner with the procedure before or after the edits one will think the plant response is correct due to the time it takes to auto sync. Apparently, selecting ‘GEN’ is the trigger for auto sync actuation regardless the state of the ‘sync check’ poke on 50212.

• IF you select Manual on the ZAS-EP-05 controller prior to depressing GEN (initially the controller comes up with neither selected), depressing GEN will NOT cause the ZAS-ES-01 breaker to close until ZAS-EP-05 controller is selected to AUTO. Again, the status of Acknowledge Ready for Auto Sync poke is irrelevant. The generator syncs.

Area of Impact Secondary plant management

Excitation Transformer Graphic Issue

SCR-DR-6398 This issue impacts the following RO/SRO task: RO-ELE-ZVS-001-00 Respond to Excitation and Voltage Regulation System (ZVS) abnormalities

Disposition

SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests. This is a simulator I&C issue related to a non-consequential graphic indication. Description On screen 22101 there is indication of downstream voltage of the excitation transformer. The indication displays ~26Kv when in fact this voltage should be around 900Vac. The drawing on the screen could be changed to connect to the upstream side of the transformer and then the operator would be able to see voltage on both sides of the main generator breaker when synchronizing to the grid.


IDS Charger Capacity and Design Float Voltage Requirement are Incompatible

SCR-DR-6400 This issue impacts the following RO/SRO task: RO-LT-R-IDS.002 Monitor the IDS

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Disposition

SNC has determined that this issue does not impact the simulator’s suitability of the simulator for the conduct of operating tests. This discrepancy is due to partially implementing a forthcoming design change in different documents. The output voltage of the IDS chargers will be increased at a later date. The simulator is properly modeling the design documentation that it was built to. The lower output voltage indication will not drive operators to perform or not perform any actions.

Description Reference documentation for IDS currently states the rated voltage for the battery charger is limited to 250 VDC. Other documentation also states that the battery (IDS) should be normally on a float charge of 264 VDC. This cannot happen without a larger battery charger.


Graphic 1805 has reversed rods



Description MB and MC rods are reversed on Graphic 1805.

Area of Impact Plant Indication

Residual Bus Transfer Issues

SCR-DR-6481 This issue impacts the following RO/SRO task: AP-LT-R-ECS.012 Block fast bus transfer

Disposition

SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests. The current simulator implementation does not provide the capability of the instructor to insert a malfunction that will result in the actuation of a Residual Bus Transfer. However, the Fast Bus Transfer and the Diesel Generator starting sequence function properly and provide the capability to examine the operators on electrical failures that would result in similar indications and Abnormal Operating Procedure entries. Description

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For undervoltage conditions (loss of power) sensed by 27B2 (two-out-of-two or two-out-of-three logic) in conjunction with a source undervoltage condition sensed by 27S, the Unit Auxiliary Transformer (UAT) breaker will be tripped, leading to automatic closing of the RAT breaker completing residual bus transfer after establishing that the Reserve Auxiliary Transformer (RAT) source is live (59S1), the bus is dead (27B2), and all motor feeders are tripped.

The sequence of events for a residual bus transfer to occur is as follows:

• At 75% rated voltage (~ 3 sec time delay) the load shed occurs. The associated bus output breakers are tripped open. FOR ES-1 and 2 ONLY the associated DG starts.

• At 30% rated voltage the residual bus transfer occurs, the UAT source supply breaker opens and the RAT source supply breaker shuts for the associated bus.

This is not modeled currently.



SCR-DR-6491 This issue impacts the following RO/SRO task: RO-SUP-DOS-001-00 Operate Standby Diesel Fuel Oil System (DOS)




WGS Sample Package Digital Indication


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Description Per drawing APP-MS27-E5-001, APP-MS27-M6-001, APP-MS27-VMM-004 pages 20 & 354, APP-WGS- MC3-101 page 16, H2 monitor AT032 (AE032) provides only a digital output.

Ovation drawing 16100 shows WGS-AT032 as having continuous indication. This continuous indication is inferred when looking at APP-MS27-M6-001, APP-WGS-M3C-101 page 23, and APP-WGS-M6-001. However per APP-MS27-VMM-004 page 354, AE032 provides only a digital output via a normally closed contact.

Area of Impact WGS Indications




Description When the power range (PR) B lower detector fails high, the RSA NAP for Power Range Power does not eliminate this input. This causes an erroneous PR power reading on the WPIS.

Area of Impact Off Normal Event Response

Safety Mimic Display for SGS-V255A& B Indicates Bad Quality Following a SFW Isolation SCR-DR-6698 This issue impacts the following RO/SRO task: RO-INC-DDS-001-00 Access plant information using the DDS


Description The PLS Safety Mimic display for all 4 divisions indicate bad quality for SGS-V255A&B following a SFW Isolation Signal. The valve is closed as verified by the FW Components Status tab on the PDSP.


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Safety Mimic Display Navigation Issue

SCR-DR-6670 This issue impacts the following RO/SRO task: RO-INC-DDS-001-00 Access plant information using the DDS Disposition


PMS mimics in Ovation have several graphics that change division when using bottom links from a high level page. CVS for example, when in PMSC or PMSD and select CVS, then select Status, you will be changed to PMSA.

The PMS mimic in Ovation is an operator aid and not needed for plant operation or PMS actuations. Therefore, this issue does not impact the conduct of operating tests.

Operators are trained to always apply Human Performance (HU) tools when operating the plant, including changing from one Ovation screen to another.

CR 10070361 for WEC resolution.

Description PMS mimics in Ovation have several graphics than change division when using bottom links from a high level page. CVS for example, When in PMSC or PMSD and select CVS, then Status, you will changed to PMSA.

Area of Impact Plant Navigation

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10 CFR 55.45(a)(5) Observe and safely control the operating behavior characteristics of the facility.

Executive Summary

Nine issues associated with rod control were assessed on an individual basis and deemed acceptable individually. The team believes operators can respond to simulator scenarios with these rod control issues and the NAPs issues. However, some of these issues require the operators to ignore the NAP indication or rod control behavior. There are some NAP issues having the potential to be resolved via a procedure update.

Issues with the RSA NAP concerned the team. Unnecessary reactor trips could happen due to misleading indications on the WPIS. This was deemed as a threat to quality examination.

The team discussed the numerous alarms (modulating status alarms in this case) and the potential to desensitize the operator to not pay attention to these alarms. The excessive alarm count coupled with the rod control and NAPs deficiencies led the team to determine that the issues in the section as an aggregate were unacceptable. Operations and Training could not deal with these issues and effectively train.


SCR-DR-5546 This issue impacts the following RO/SRO tasks: RO-LT-R-EDS.001 Monitor the Non Class 1E DC and UPS system (EDS) for proper operation RO-LT-R-EDS.004 Respond to a loss of EDS DC power

Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC). SMEs determined that the current power supply arrangement was adequate to teach since it is per design documentation. Power supplies are an item that will be continuously taught as they are updated and changed.

Description A loss of individual Non Class 1E DC and UPS System (EDS) busses will result in incomplete system response. Some Ovation drops are not dynamically powered by the EDS model but are powered by a permanently energized model constant (specifically DPU047, DPU048, and DPU044). The load lists for the STS do not assign a power supply to all the Ovation drops so there is no plant design data to insert into the simulator.


Modeled BEACON Data Cannot Determine Quality

SCR-DR-5583 This issue impacts the following RO/SRO task: RO-INC-IIS-004-00 Determine functionality of the On-line Power Distribution Monitoring System

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Disposition SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests. Operator training simulator guides inform the training instructors on whether BEACON is operable or not operable. (Note: This evaluation is actually based on whether or not BEACON if functional or not functional. BEACON is not safety-related and does not have applicable Technical Specifications.)

Description Failure of the BEACON Data Processing (BDP) NAP causes the manual override signal originated by BDP to have BAD quality as expected. The BAD indication is passed through the BEACON operability calculations in the Plant State Monitoring (PST) NAP and appears on the OPDMS displays as 'operable' with BAD quality. Failure of the BDP application will cause BEACON to indicate ‘inoperable’ in the reference unit. However, the current STS scope of limitation has the BEACON outputs driven by the core model. There is currently no ability to pass quality over the interface for outputs. Inputs are not taken from the BDP NAP but from other plant process data. The core model does not know the status of the BDP NAP. The core model does not pass operability information.


Repeatability issues involving CL 1B

SCR-DR-5594 This issue impacts the following RO/SRO task: RO-PRO-AOP-054-00 Respond to Reactor Coolant Pump Malfunctions using AOP-114

Disposition This issue was dispositioned as unacceptable by the Simulator Review Committee (SRC). This issue was further evaluated and determined to relate with a current design issue involving a lead/lag circuit associated with the main steam line pressure detection and input to the Safeguards ESF actuation. The pressure drop in the main steam line turns very close to the actuation set-point and the lead/lag circuit amplification of the rate of change may or may not cause the actuation. This is the expected plant response with the current actuation logic/software. Operators have sufficient procedure guidance directing them to respond to this event. For this reason, it was determined that this issue does not impact the suitability of the simulator for the conduct of operating tests.

Description The 1B RCP shaft shear malfunction may cause an unexpected safeguards depending on the initial conditions when the malfunction is inserted. Due to response caused by lead/lag filters in the steam line pressure logic, the signal is driven into the dead-band range of actuation at certain initial steam pressures. Elevating the initial steam pressure completely mitigates this occurrence.

Area of Impact Abnormal operation response

Unidentified and Identified Leak Rate Always Indicates BAD Data

SCR-DR-5599

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This issue impacts the following RO/SRO task: RO-PRO-AOP-053-00 Respond to a Reactor Coolant Leak using AOP-112

Disposition

After performing V&V testing, SNC determined that the update was not successful. The Leak Rate Monitoring Application is informational only and does not drive any alarms based upon the calculated leakage. For this reason, any leak rate calculations would have to be performed manually per plant procedures vice using the NAP calculated values.

Description The Leak Rate Monitor (LRM) has BAD quality point indication for the Identified and Unidentified leak rates. They never change to good quality and indicate BAD when using the on demand leak rate calculation.


Primary Dedicated Safety Panel Screens Do Not Update during MCR/RSR Transfer SCR-DR-5680 This issue impacts the following RO/SRO task: RO-INC-DDS-008-00 Transfer function from the Main Control Room to the Remote Shutdown Workstation

Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) since the operators would have already evacuated the MCR and the indications would not been seen.

Description During the transfer between Main Control Room (MCR) and Remote Shutdown Room (RSR), it was noted that Protection and Safety Monitoring System (PMS) screens were not reflecting the position of the transfer switch correctly. Once the PDSP screen was refreshed, it displayed “correcting”.

Areas of Impact Impacts operation during unavailability of MCR for emergency safe shutdown

PMS Mimic Screens


Disposition

SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests. This is a simulator I&C issue. PMS mimic screens are used for verification of indications only. If a question arises regarding an indication on the PMS mimic screen, operators will use primary indications from the PMS displays on the Primary Dedicated Safety Panel (PDSP). No operator action is available through the PMS mimic screens. All actions must be taken from the division’s PMS PDSP.

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Description PMS Mimic screens on Ovation do not reflect what is shown on the associated PDSP. This is especially true when there is any FAULT condition shown on the PDSP.


Hot Leg Fluctuations at Mid-loop

SCR-DR-5707 This issue impacts the following RO/SRO task: RO-PRI-CVS-003-12 ****Fill the IRWST

Disposition This issue was dispositioned as unacceptable by the Simulator Review Committee (SRC). A training needs analysis was performed. The training needs analysis identified that this only occurs if CVS is used to refill the IRWST while isolated from the RCS. Since this is an abnormal lineup, the operator is unlikely to see this. For this reason, SNC has determined that the issue does not impact the suitability of the simulator for the conduct of operating tests.

Description While at mid-loop, the hot leg level and pressurizer wide range level began to fluctuate erratically. The hot leg was ~80% full.

Area of Impact

Plant control




Description It is noted that the Core Operating Limits Report (COLR) Rev. 0 Rod Insertion Limit (RIL) for M2 bank at all power levels is [ ]a,c Steps (fully withdrawn). The Online Power Distribution Monitoring System (OPDMS) Rod Insertion Limit (RIL) display ([ ]a,c) indicates the RIL for M2 bank is [ ]a,c steps. Further investigation indicates the M2 RIL indication high limit is [ ]a,c steps and therefore cannot indicate above this level (determined using point information page instrumentation tab for RB-INSERT- M2LIM.SV3@NET0). All SD bank indications are capable of indicating a maximum of [ ]a,c steps.

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Inconsistent OPDMS QPT Indications

SCR-DR-5903 This issue impacts the following RO/SRO task: RO-INC-IIS-005 Respond to On-line Power Distribution Monitoring System (OPDMS) malfunctions.

Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to the specific nature of the tasks associated with this screen.

Description During dropped control rods at core locations C7 and H8, it was noted that the indications provided by the OPDMS Excore and Incore Quadrant Power Tilt (QPT) monitors were not consistent. The following observations were made:

• The Excore display lists the detectors as N41-N44. A detailed search of AP1000 documentation

found no reference material in which the PR excore detectors are referred to by N41-N44 except on page M-4 of APP-OCS-J4V-207, “Operation and Control Centers Display Design Document for Online Power Distribution Monitoring System.” Page M-4 has a table showing that OPDMS has N41, N42, N43, and N44 mapped to PR C, D, B, and A respectively, which appears to be consistent with what was seen in the PRS.

• The values for all of the excore detectors read 1.0 which appears to be just some sort of default. There are two problems with this. 1) One decimal point worth of data is not enough to adequately assess QPTR. 2) In this scenario, they should definitely not be reading 1.0.

• On the values displayed, there are 8 titled “PR Power Upper Detector”, with no additional label as to what division.

• For the Incore QPT display, the letters in the corners do not match up with the data. For example, in this scenario the flux shifted towards PR A and C but the display shows it greatest near A and D and suppressed at C.


Print Feature from NAP non-functional

SCR-DR-5913 This issue impacts the following RO/SRO task: AP-RO-ADM.020.07 Document surveillance test in log book

Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to minimal training impact as this simply means the automated system is not capable of being used to complete surveillance requirement testing, operators are still capable of using the paper copies.

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Description While performing surveillance procedure "Incore Detector Comparison to Nuclear Instrument Channel Axial Flux Difference", the select PRINT SURVEILLANCE REPORT does not result in a printout.

Area of Impact Plant response – this prevents or impacts the performance of most Surveillance Tests




Description When the condensate outlet of the blowdown heat exchanger temperature element is failed high the blowdown flow remains un-isolated. The high-2 temperature ([ ]a,c) should isolate blowdown flow in accordance with APP-CDS-M3C-101 Rev 3. However, the range of the instrument (CDS-T-040A/B) listed in APP-CDS-M3C-101 Rev 3 is [ ]a,c which would never allow blowdown isolation on high temperature.



SCR-DR-5968 This issue impacts the following RO/SRO task: RO-PRI-CVS-004-00 Monitor CVS operations



Per APP-CVS-M3C-101 Rev 6 Appendix C.3.2, at [ ]a,c when CVS-V047 is in automatic pressure control mode there is supposed to be a signal sent to “trip the CVS Makeup Pumps.” The high pressure signal is generated but does not trip the pumps; it presently feeds a Pump Auto Stop Demand signal. This signal will stop any pumps that are running in automatic only. When the plant is in water-solid mode, as determined in logic diagrams as having CVS-V047 in automatic pressure control mode, Chemical and Volume Control System (CVS) makeup pumps must be operated in manual; see APP-CVS-M3C-100 Rev 11

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Logic Sheet CVS-4 Note10. Since the CVS makeup pumps are in manual the Auto Stop Demand signal will not shut the pump off to provide overpressure protection.

Per APP-CVS-M3C-101 Rev 6 Appendix C.3.2, at [ ]a,c when CVS-V047 is in automatic pressure control mode there is supposed to be a signal sent to “trip the Reactor Coolant Pumps…in order to protect them from reduced suction pressure.” This signal is also discussed in APP-RCS-M3C-100 Rev 9 Logic Sheet RCS-13 table and note 11. As presently designed there is no logic tie between CVS-PT040 and the Reactor Coolant Pumps (RCPs) to prevent damaging the RCPs upon a loss of Reactor Coolant System (RCS) pressure. Area of Impact Plant Control

Containment Recirculation Actuation Indication Issue



Description Once containment recirculation is actuated, the actuation indication for Divisions C and D did not have the white box with an X on the ESF Act Status Screen for the divisional PDSPs or the Non-Safety Operational Overview screen (33020). The individual PMS division screen for CNMT Recirc actuation (IRWST/INJT Recirc) did show that it had been actuated on all 4 divisions.

Area of Impact Verifying plant response

CVS-V094 Power Failure Response


Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC). When this issue manifests itself on the simulator, the operators will still comply with Technical Specifications requirements. An additional Tech Spec call would have to be made since there is an issue with Zinc addition in the CVS when power is lost. For this reason, this issue does not impact the suitability of the simulator for the conduct of operating tests.

Description CVS-V094 does not close upon a loss of power to ILCA02 as expected. It did close on loss of power to ILCA03, which is not in accordance with design documentation.

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Area of Impact Response to power loss



Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to the minimal training impact this particular NAP has.

Description When attempting to simulate fuel assemblies being moved from the core to the Spent Fuel Pool (SFP), it was noted that the Decay Heat Calculation (DHC) NAP to maintain the administrative location of fuel does not work correctly. On display 40203 the assembly move buttons on the lower right portion indicate they are only available when the light DDS-AP-DHC Status indicates it is ACTIVE. This light is driven off of the automatic mode selector and is INACTIVE when in MODES 1&2 and ACTIVE in MODES 3-6. However, when the light indicates INACTIVE the buttons for moving are raised and available. When the light changes status to ACTIVE the buttons for moving are grayed out and no longer available. The light being active or inactive is currently driven by the auto mode selector and becomes active in MODES 3-6. However, fuel cannot be moved from the core into the SFP in any MODE other than MODE 6. The light should be driven by the manual input of the Rx vessel head being removed or installed or upper internals position on display 40004.


HSS Display does not Include Emergency Seal Oil Pump (ESOP) Discharge Pressure



Description Hydrogen Seal Oil System (HSS) HSS-MP02 discharge pressure HSS-PT017 is not on the HSS display (15100). APP-HSS-M6-001 Rev 2 indicates that the transmitter should have an available point reference in ovation (PT-017 has a box with PIA - Pressure Indication/Alarm).


DWS-LT006 has Insufficient Range

SCR-DR-6099 This issue impacts the following RO/SRO task: RO-SUP-DWS-002-00 Monitor DWS operation

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Disposition The simulator is modeling the plant as currently designed; once the design has been updated the simulator model will be updated as well. This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to minimal impact on training.

Description Demineralized Water Transfer and Storage System (DWS) DWS-LT006 is the level indication for the Condensate Storage Tank (CST). This level spans [ ]a,c per APP-DWS-M3C-101. However, the calculation note for DWS (APP-DWS-M3C-002) has the high 2 alarm at [ ]a,c. This is an important alarm because it is designed to give operators time to determine why the CST is full before it overflows to a drain. Overflow will occur at [ ]a,c.



SCR-DR-6102 This issue impacts the following RO/SRO task: RO-INC-PLS-003-06 Perform grey rod exchange Disposition Subsequent to the performance of the Aggregate Study this item has been corrected and closed.

Description The plant control system operating procedure allows for a case 1 Control Rod Exchange in the event that Case 2 is not functioning. With MA rods at 234 steps and AO rods in manual at 218 steps, the AO rods were stepped into [ ]a,c steps, which will drive MA out. The MA rods stopped at 263 steps, with an outward demand still in. The audible rod clicking stopped as well.

If the CRE is still continued in accordance with the procedure, MD will step into the core and MA will remain at [ ]a,c . This generates a “Rods out of sequence alarm”.


Improper function of C-2

SCR-DR-6122 This issue impacts the following RO/SRO task: RO-INC-PLS-008-00 Respond to PLS - related abnormalities

Disposition

Subsequent to the performance of the Aggregate Study this item has been corrected and closed. Description C-2 control interlock utilizes a non-conservative power input for operation. It currently uses BDP corrected power which is lower than NI power during over power scenarios.

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Description At low pressure conditions less than 350 psig, the operator often has to take manual control of Startup Feed Water (SFW) control valves due to excessive cycling of the valves. Indicated flow rates range from 0 to greater than 600 gpm within 10 to 15 second cycles. This requires 100% of the operator’s attention until Normal Residual Heat Removal System (RNS) can be placed into service removing the cooldown function from the steam dumps.

Area of Impact

Plant control during startup and shutdown

SWS temperature control

SCR-DR-6152 This issue impacts the following RO/SRO task: RO-LT-R-SWS.019 Monitor the Service Water System

Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to minimal operator training impact. Operators might question the low SWS temperature reading, but at this reading, procedures require no operator actions be taken. Since this issue does not impact actions

taken by the operator, SNC determined it does not impact the suitability of the simulator for the conduct of operating tests.

Description SWS temperature got as low as about 51 deg F with an ambient air temp of 70 deg F. When the cooling fan kicked off, temperature returned to about 60 deg F. Cooling tower fan seems modeled more like an air conditioner than an evaporative cooler.

Area of Impact Plant performance

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FWS-V037 Control Issue

SCR-DR-6156 This issue impacts the following RO/SRO task: RO-SEC-CDS-004-06 Respond to abnormal DST water level

Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC). The reasoning was that the current controller tuning is the current design. Hot functional testing would provide as-built tuning. However, this issue did contribute to the aggregate impact of section 5 with regards to secondary plant control. The team determined that there is procedural guidance already in place which mitigates the impact on operations. Therefore, it does not impact the suitability of the simulator for the conduct of operating tests.

Description The Deaerated Water Storage Tank (DST) level control presents an undue operator workload. DST level control often cannot maintain the DST above 95% to prevent Main Feedwater System (FWS) FWS-V037 rejecting to manual. Manual control of the input and output flow streams to the DST does not result in stable level control either due to the lag in valve response.


SGS MSL drain pot erratic indication

SCR-6157 This issue impacts the following RO/SRO task: RO-SEC-SGS-006-00 Monitor SG system and component parameters

Disposition This issue does not impact the suitability of the simulator for the conduct of operating tests.

The drain pot level going high will cause operators to take the procedurally directed actions in the Alarm Response Procedure. The erratic indications will not solely drive the operators to perform plant maneuvers.

Description

SGS MSL drain pots became erratic and were flashing on both main steam lines at 53% Rx Power. By 90%, both pots filled with water, even with the drains open.

Area of Impact

Plant Indications

Stuck Rod Recovery Malfunction

SCR-DR-6162 This issue impacts the following RO/SRO task: RO-INC-PLS-005-00 Respond to DRCS related abnormalities

Disposition

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This SCR is specifically related to rod K10 and an individual fault. This fault is functional for all other rods and the reset capability described in the SCR is functional for all other rods. Training and exam scenarios are written to avoid this specific rod with no impact on the ability to train or examine on this fault.

WEC has created a tracking RITS (42516).

Description Rod K10’s keyboard reset does not go to the K10 algorithm in rod control, but K6. Rod K10 can never get an individual reset via the operator keyboard.

Area of Impact Rod control abnormality recovery

Tuning of VBS Required for Stability

SCR-DR-6168 This issue impacts the following RO/SRO task: RO-VNT-VBS-001-15 Purge smoke from MCR/CSA

Disposition This issue was dispositioned as unacceptable by the Simulator Review Committee (SRC). A training needs analysis was performed under CR 10025689. SNC evaluated this issue and determined that even though it is unacceptable as a design issue, it does not impact actions taken by the operator.

Therefore, it does not impact the suitability of the simulator for the conduct of operating tests.

Description During Main Control Room (MCR) purge operations, Nuclear Island Nonradioactive Ventilation System (VBS) air handling unit trains cannot maintain stable flow and as a result, enter an indefinite cycling between two trains. The current tuning does not allow enough time to establish stable flow.


RSA NAP Does Not Process Failed Channels Correctly




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Flux doubling difference between divisions

SCR-DR-6175 This issue impacts the following RO/SRO task: RO-INC-PMS-003-00 Monitor System Indications Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) as the protective functions associated with these signals will still occur.

Description The alarms/alarm response for A/D Divisions differs significantly from B/C divisions. A/D divisions activates at 1.6 in 50 seconds whereas B/C at 2.2 in 10 seconds. J3 documents specify 2.2 in 10 sec.

Area of Impact Safety System Operation


SCR-DR-6179

Disposition

SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests. The Time to Boil NAP is a tool that is used for information only. The NAP is active when in Mode 5/6 conditions for RCS time to boil and when fuel is present in the spent fuel pool for spent fuel pool time to boil. The value is displayed in exponential units versus hh:mm. This means that operators will need to convert the scientific notation values into hours and minutes. Although this does take a short amount of time, the net effect is that it does not remove the ability to monitor the time to boil and it has no impact on actions the operator may, or may not, take in response to plant conditions. Description When core exit temperature was 300oF, Thot was > 212oF and TTB was >0 (15-25 min range). Either the NAP is calculating TTB incorrectly or the inputs to the NAP are wrong. TTB should reflect actual plant conditions.

Area of Impact NAP



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Explain Brief students prior to beginning a simulator training phase or segment. Update a “SIMULATOR TRAINING STUDENT HANDOUT” (example is attached) and file in the “Operator Aids” notebook. Reference the SIMULATOR TRAINING STUDENT HANDOUT in each sim guide.

Description During outward rod motion, the audible step counter randomly can have an extra second pause in it with rod motion continuing. The step counter indication does not update at the same rate as the audible cue occurs.


VFS Radiation Monitoring Issue

SCR-DR-6192 This issue impacts the following RO/SRO task: RO-VNT-VFS-002-01 Monitor VFS parameters


Description VFS-RY102 indicates off scale high on a loss of Shutdown Cooling. The alarm appears to be related to the cycling flows in the Containment Air Filtration System (VFS) system. Alarms will also occur during containment purge activities with no RCS leakage or increased activity providing a false radiation alarm.

Area of Impact Radiation management




Description The WR Core Makeup Tank (CMT) level indications shift to Bad Quality once Automatic Depressurization System 1-3 (ADS 1-3) Actuate. Prior to this event, they would toggle to Bad Quality intermittently. The Bad Quality status is on indications PXS-LT009A/B & -LT010A/B (on PXS Supplemental Ind. Screen) and

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DDS-RSA11-L1 & DDS-RSA13-L1 (on WPIS screen 60017). The NAP driving the calculation of this indication drives them to bad quality whenever it determines voiding is occurring in the CMT (which is expected per design transients).












Description An Ovation data point in rod control is initially set to 0 when the Digital Rod Control System (DRCS) is reset. However, during SD1 withdrawal, OCB07CE00C_OUTAV (the Ovation data point) will increment to a value of 2 and then stay at this value. This data point is only supposed to increment for inward rod motion during

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M bank rod movement. The end result is that bank overlap will be incorrect if not manually corrected in Ovation.








& Maintenance Bypass for Division C & PMS Cabinet Fault Alarm for Division C

• RCS TE122D -> Open Circuit: PMS indicates Low-2 Tcold Cold Leg 2 Bistable Trip with Bad Quality & Maintenance Bypass for Division D with other alarms





Main Generator Output breaker logic

SCR-DR-6392 This issue impacts the following RO/SRO task: RO-ELE-ZAS-002-04 Synchronize the generator in manual/auto modes

Disposition

SNC evaluated this issue and determined that since the MCR operator will not see this anomaly when the procedure is followed; it does not impact actions taken by the operator. Therefore, it does not impact the suitability of the simulator for the conduct of operating tests. Description A discrepancy in the turbine generator synchronization logic results in the following:

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• IF you have the ‘acknowledge ready for auto-sync’ poke selected or not, the Generator breaker will close when ‘GEN’ is selected and operator action ceases for ~90 seconds. If you continue in a timely manner with the procedure before or after the edits one will think the plant response is correct due to the time it takes to auto sync. Apparently, selecting ‘GEN’ is the trigger for auto sync actuation regardless the state of the ‘sync check’ poke on 50212.

• IF you select Manual on the ZAS-EP-05 controller prior to depressing GEN (initially the controller comes up with neither selected), depressing GEN will NOT cause the ZAS-ES-01 breaker to close until ZAS-EP-05 controller is selected to AUTO. Again, the status of Acknowledge Ready for Auto Sync poke is irrelevant. The generator syncs.

Area of Impact Secondary plant management



Disposition

SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests. The current simulator implementation does not provide the capability of the instructor to insert a malfunction that will result in the actuation of a Residual Bus Transfer. However, the Fast Bus Transfer and the Diesel Generator starting sequence function properly and provide the capability to examine the operators on electrical failures that would result in similar indications and Abnormal Operating Procedure entries. Description For undervoltage conditions (loss of power) sensed by 27B2 (two-out-of-two or two-out-of-three logic) in conjunction with a source undervoltage condition sensed by 27S, the Unit Auxiliary Transformer (UAT) breaker will be tripped, leading to automatic closing of the Reserve Auxiliary Transformer (RAT) breaker completing residual bus transfer after establishing that the RAT source is live (59S1), the bus is dead (27B2), and all motor feeders are tripped.


• [ ]a,c

• [ ]a,c




SCR-DR-6491

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This issue impacts the following RO/SRO task: RO-SUP-DOS-001-00 Operate Standby Diesel Fuel Oil System (DOS)




VES Supply Header Pressure Response to Temperature Changes

SCR-DR-6547 This issue impacts the following RO/SRO task: RO-VNT-VBS-001-11 Monitor the main control room/control support area HVAC subsystem parameters

Disposition

Subsequent to the performance of the Aggregate Study this item has been corrected and closed. Description When the Radiological Controlled Area Ventilation System (VAS) ventilation is secured to the Main Control Room Emergency Habitability System (VES) Air Storage Area (Rm 12555), room temperature will rise as expected, as indicated on VAS-TE080A/B. This rise in temperature should cause VES Supply Header Pressure (VES-PT001A and B) to rise, since the volume of air in the VES tanks is not changing. As depicted in the attached trend, this does not happen in the current model. Since VES air storage tank pressure does not change in the model, this causes the calculated air quantity (VES-QIY008A and B) to lower, which should not be the case. The VES air quantity should only change if you are filling VES or depressurizing it.

Area of Impact Observation of fundamental parameters

ECS-EC-313 Loads not modeled


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Disposition The Aggregate Study team determined this issue does not affect the operators because they will be fairly transparent. The students will most likely not notice these issues during training. The team determined this issue in the aggregate is acceptable. While there will be manual actions that occur, it is still acceptable and should be addressed in a training moment.

Description APP-ECS-E3-EC31302 Rev 1 depicts ECS-ET-3131. Model requirements document SV3-STS-J4-119 Rev 1 required WEC/GSE to model the loads of EC-313. Work in SCR-DR-6568 determined they neglected to meet this requirement. SV3-STS-J1-119 Rev 0 shows loads off of EC-312 but nothing on EC-313.

Area of Impact Plant response

D/G Sequencer Operation

SCR-DR-6610 This issue impacts the following RO/SRO task: RO-ELE-ZOS-002-06 Perform priority load AC load sequencing and verify proper 6.9 kV priority load system voltage


Description According to APP-ZOS-E0C-001 (Onsite Standby Power System Diesel Generator Sizing Calculation) Rev 1, the following busses should get re-energized by the sequencer at Load Step [

]a,c

[ ]a,c [ ]a,c [ ]a,c [ ]a,c [ ]a,c [ ]a,c

These busses are not currently sequenced once ECS-ES-1/2 gets re-energized from its respective DG. Area of Impact Effective plant response to loss of power


SCR-DR-6621

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This issue impacts the following RO/SRO task: RO-INC-PMS-003-00 Monitor System Indications


Description When the power range (PR) B lower detector fails high, the RSA NAP for Power Range Power does not eliminate this input. This causes an erroneous PR power reading on the WPIS.

Area of Impact Off Normal Event Response

VZS Dampers do not Fail As-Is after Loss of Power

SCR-DR-6623 This issue impacts the following RO/SRO task: RO-VNT-VZS-005-00 Respond to Abnormal VZS Conditions

Disposition This issue does not impact the suitability of the simulator for the conduct of operating tests. This incorrect modeling issue is transparent to the operators and therefore has no impact on operator actions. The specific dampers in question are verified to be in the proper position by AOP-302 Attachment 1. The only time the positions of the dampers in question are checked is when the associated diesel is operating. During this scenario the dampers will be operating properly as they are energized. If the diesel generators are not powering their respective buses then the steps directing the operators to verify the position of these dampers will not be performed. This procedural guidance will cause the incorrect modeling of these dampers to not be observed.

Description Diesel Generator Building Heating and Ventilation dampers VZS-D014A/B and VZS-D015A/B do not fail as is on loss of power.

Area of Impact Plant Control during loss of power

Turbine Bypass Control Valve Control Logic cannot Support Design Power Supplies

SCR-DR-6634 This issue impacts the following RO/SRO task: RO-SEC-MSS-003-01 Monitor MSS system and component parameters


Description The turbine bypass control valves (MSS-PL-V001, MSS-PL-V002, MSS-PL-V003, MSS-PL-V004, MSS-PL- V005, MSS-PL-V006) have four solenoid valves that are designed to share two different power supplies. However, the current design control logic) uses one power supply for all 4 solenoids.

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Area of Impact Plant operations during loss of power

Battery Temperature does not change

SCR-DR-6645 This issue impacts the following RO/SRO task: RO-LT-R-EDS.001 Monitor the Non Class 1E DC and UPS system (EDS) for proper operation


Description During a loss of all AC sources discharge test it was identified that the battery temperature did not change during the 15 hour run. Further research showed the battery room temperature does not change during loss of ventilation

Area of Impact

Effective battery temperature indication

Fire Protection System is not modeled in Containment

SCR-DR-6657 This issue impacts the following RO/SRO task: RO-SUP-FPS-002-00 Operate the FPS

Disposition This issue does not impact the suitability of the simulator for the conduct of operating tests. Due to the low importance values in NUREG-2103 of the items associated with this failure no simulator scenarios have been developed that would make use of this malfunction. The modeling of this malfunction will be investigated as time permits and corrected if possible.

Description While attempting to create a FR-Z.2 (Response to High Containment Level) scenario, it was discovered that a leak from the Fire Protection System (FPS) header in containment had no effect on any containment parameters (ex. Containment Sump Level, Containment Humidity). FPS Containment Spray also had no effect on any containment parameters.


IRWST Temperature Response

SCR-DR-6701 This issue impacts the following RO/SRO task: RO-PRO-EOP-003-00 Respond to a loss of Reactor or Secondary Coolant using EEP-E-1

Disposition

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The Aggregate Study team determined this issue does not have an aggregate impact on operations. Additionally, this modeling of the differential temperatures has no impact on operator actions because, by procedure, there are no operator actions related to IRWST temperature. For this reason, it was determined that this issue does not impact the suitability of the simulator for the conduct of operating tests.

Description [

]a,c


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10 CFR 55.45(a)(6) Perform control manipulations required to obtain desired operating results during normal, abnormal, and emergency situations.

Executive Summary

Issues 5655, 5926, 6151, 6156, and 6172 were grouped together for analysis by the team. The team acknowledged that there is procedural guidance already in place which mitigates some of these items impact on operation. The team discussed the potential burden they do place but also discussed examples of successful navigation through these items during ISV.

The issues involving having to take manual control to enhance stability were discussed by a team. For example, Nuclear Island Nonradioactive Ventilation System (VBS) issue will require manual control (6168) during VBS operations. In discussing if these issues presented too much of an aggregate challenge, the team determined a manual reactor trip would likely not be initiated. The issues described in this section are always in, thus making them a part of the standard training session. The Initial Conditions (IC) or scenario guide would be setup to already have these issues placed in a status where the scenario wouldn’t be adversely affected. The team justified the first group of issues being acceptable. Issues like having to manually control FWS-V037 are part of the training process. Each operator candidate is already familiar with how to handle them. Issues with VBS and VFS can be handled because they are out of the way items during training and operations. These systems are rarely operated as a normal process.

Issues 6610 and 6593 were grouped together and determined not to affect the operators because they will be fairly transparent. The students will most likely not notice these issues during training. Issue 5609 has been addressed as a procedure change.

The team determined that the combination of these issues in the aggregate were acceptable. While there will be manual actions that occur, it is still acceptable and should be addressed in a training moment.

These controllers are expected to work in automatic, and when manual control is required reinforcement to the operator’s skills will be required to establish proficiency. The absence of this ability will not preclude safety, but additional training will provide additional proficiency. Simulator JIT will be a valuable tool in preparing the operators for encountering these issues in the plant.




Description While performing extended rod withdrawals during startups, depressing the rod withdrawal button (UP ARROW) may cause the UP ARROW button to un-highlight and momentarily flash gray

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even though still depressed. Rod motion will still occur. Area of Impact Reactivity Management

Unstable VFS Containment Exhaust Flow

SCR-DR-5593 This issue impacts the following RO/SRO task: RO-VNT-VFS-002-05 Regulate Containment Air Pressure

Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to low impact on training and the ability to take some controllers to manual to control flow.

Description When Containment Air Filtration System (VFS)-MS-02A/B are placed in service to establish containment exhaust flow in AUTO with a setpoint of 4000 scfm, a stable flow of 4000 scfm cannot be obtained on VFS-FT011A/B. Flow gradually increases above 4000 scfm and oscillates. 4000 scfm can be achieved in MANUAL control, but when placed back into AUTO, the oscillations continue. This point is on screen 20102.


GSS Header Pressure Response

SCR-DR-5609 This issue impacts the following RO/SRO task: RO-SEC-GSS-002-00 Monitor the GSS


Description In both at-power ICs with Main Steam System (MSS) supplying, GSS-V002 was able to maintain 4 psig. Once the shutdown IC, 32, was loaded with Auxiliary Steam Supply System (ASS) supplying the Gland Seal System (GSS) header GSS-V002 was unable to maintain downstream pressure. If GSS-V004 (bypass for GSS-V002) was opened to > 50% there was sufficient pressure/flow to allow GSS-V002 to control pressure at 4 psig.

To determine if flow was the root of the problem aligned MSS to supply GSS while still using IC 32. Once steam load was on MSS, the GSS header pressure lowered until GSS-V002 was fully open (GSS pressure was ~ 1.8psig); MSS-V022A was ~ 28% open. Adjustment was required to the automatic pressure setpoint of MSS-V022A to 240 psig to increase the steam pressure upstream of GSS-V002. Once MSS-PT015A was ~ 240 psig GSS-V002 closed to 98% to control at 4 psig.


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SCR-DR-5655

This issue impacts the following RO/SRO task: RO-SEC-FWS-002-03 Establish level in the DST


Description During plant startup from Mode 5 to 100% power, the Condensate system pressure would lower to the auto start setpoint of the standby Condensate pump. This pressure drop is due to Condensate System (CDS) CDS-V022 and CDS-V025 modulating to maintain level in the Deaerated Storage Tank (DST). In accordance with reference plant procedures for normal operation, the second condensate pump is started at 40-45% power. However, the second condensate pump will have already auto started in the heatup and startup procedures, due to the slow response of CDS-V022 and CDS-V025.

Area of Impact Plant design deficiency impacts operations during startup

Model Instability during PZR Fill to Solid

SCR-DR-5698 This issue impacts the following RO/SRO task: RO-PRI-CVS-003-04 Operate the Chemical and Volume Control System to control the primary system pressure in water solid mode

Disposition This issue was dispositioned as unacceptable by the Simulator Review Committee (SRC). Training Needs Analysis was performed under CR 10000465. Training Needs Analysis determined training involving establishment of solid plant should not be performed until issue is corrected. A review of current training material did not reveal any scenarios where this was required. Wording from analysis follows.

Added to the DR Global Issues list, this will be briefed to the students at the beginning of the Simulator portion of training. Scenario AP-LT-I-SIM-GOPSDCD (Covering GOP-205, Plant Cooldown MODE 3 to MODE 5) does not train on Solid Plant Operations.

Description The Liquid Radwaste System (WLS) model is prone to failure during evolutions involving near solid pressurizer operations if the Effluent Holdup Tank is filled too rapidly.

Area of Impact Difficulty in achieving solid plant operations continuously

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Steam Generator Level Instability with Control Valves Shut

SCR-DR-5926 This issue impacts the following RO/SRO task: RO-SEC-SGS-006-00 Monitor SG system and component parameters

Disposition This issue was dispositioned as unacceptable by the Simulator Review Committee (SRC). Training Needs Analysis was performed under CR 10000465. Further analysis determined that this condition has no impact on operator actions. Under this condition, the Steam Generator has been isolated and there are no further actions required by operators. For this reason and because the level stabilized after a relatively short period of time, it was determined that this issue does not impact the suitability of the simulator for the conduct of operating tests. Description Rapid steam generator level oscillations can be observed with the steam generator isolated and all control valves shut for a period of three to five minutes before stabilizing. Area of Impact Plant Response


SCR-DR-6019 This issue impacts the following RO/SRO task: RO-INC-PMS-005-02 Respond to a loss of a Protection and Safety Monitoring System (PMS) division or failure of PMS components

Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC). When this issue manifests itself on the simulator, the operators will still comply with Technical Specifications requirements. An additional Tech Spec call would have to be made since there is an issue with Zinc addition in the CVS when power is lost. For this reason, this issue does not impact the suitability of the simulator for the conduct of operating tests.





Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to the minimal

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training impact this particular NAP has.

Description When attempting to simulate fuel assemblies being moved from the core to the Spent Fuel Pool (SFP), it was noted that the Decay Heat Calculation (DHC) NAP to maintain the administrative location of fuel does not work correctly. On display 40203 the assembly move buttons on the lower right portion indicate they are only available when the light DDS-AP-DHC Status indicates it is ACTIVE. This light is driven off of the automatic mode selector and is INACTIVE when in MODES 1&2 and ACTIVE in MODES 3-6. However, when the light indicates INACTIVE the buttons for moving are raised and available. When the light changes status to ACTIVE the buttons for moving are grayed out and no longer available. The light being active or inactive is currently driven by the auto mode selector and becomes active in MODES 3-6. However, fuel cannot be moved from the core into the SFP in any MODE other than MODE 6. The light should be driven by the manual input of the Rx vessel head being removed or installed or upper internals position on display 40004.


WLS-MP-08C improperly Pumps Monitor Tank C


Disposition The cause of this issue was determined to be a modeling issue with the variables associated with piping arrangements. SNC Simulator Group has corrected this deficiency and resolved the issue.

Description

While performing a startup from Mode 6 it was discovered that WLS-MP-08C will not pump Monitor tank C around 37 inches. The pump will turn on and occasionally the downstream check valve will throttle open and shut but there is little or no evidence of flow. Also, discharge pressure never goes above 12-13psig. Normal discharge pressure for the other monitor tank pumps is around [ ]a,c.

Note that it does pump when level is above 37 inches as the tank has been pumped down to 37 inches successfully. It appears to exhibit strange behavior at 37 inches and below.

Area of Impact

Correct operation of plant systems



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Description The plant control system operating procedure allows for a case 1 Control Rod Exchange (CRE) in the event that Case 2 is not functioning. With MA rods at 234 steps and AO rods in manual at 218 steps, the AO rods were stepped into [ ]a,c steps, which will drive MA out. The MA rods stopped at 263 steps, with an outward demand still in. The audible rod clicking stopped as well.

If the CRE is still continued in accordance with the procedure, MD will step into the core and MA will remain at [ ]a,c. This generates a “Rods out of sequence alarm.”


WRS Sump Pump B Discharge Pressure Inadequate

SCR-DR-6126 This issue impacts the following RO/SRO task: RO-SUP-WRS-002-00 Operate WRS


Description Radioactive Waste Drain System (WRS) sump pump WRS-MP-01B indicates a low pressure when pumping with an automatic start signal. This is evident when a leak is inserted that fills the WRS sump (such as a RNS leak). The "A" pump has proper discharge pressure, but "B" does not indicating low pressure in alarm. If the "A" pump is taken to manual and secured, the "B" pump still does not develop proper pressure.





Description At low pressure conditions less than 350 psig, the operator often has to take manual control of Startup Feed Water (SFW) control valves due to excessive cycling of the valves. Indicated flow rates range from 0 to

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greater than 600 gpm within 10 to 15 second cycles. This requires 100% of the operator’s attention until RNS can be placed into service removing cooldown function from the steam dumps.


Stuck Rod Recovery Malfunction

SCR-DR-6162 This issue impacts the following RO/SRO task: RO-INC-PLS-005-00 Respond to DRCS related abnormalities

Disposition SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.

This SCR is specifically related to rod K10 and an individual fault. This fault is functional for all other rods and the reset capability described in the SCR is functional for all other rods. Training and exam scenarios are written to avoid this specific rod with no impact on the ability to train or examine on this fault.

WEC has created a tracking RITS (42516).

Description Rod K10’s keyboard reset does not go to the K10 algorithm in rod control, but K6. Rod K10 can never get an individual reset via the operator keyboard.

Area of Impact Rod control abnormality recovery

Polisher Bypass Valve Control

SCR-DR-6172 This issue impacts the following RO/SRO task: RO-SEC-CPS-002-02 Place CPS in service

Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to minimal impact on training and examination as the chemistry levels in the secondary plant are transparent to operators.

Description CPS-V001 (CDS Polisher Bypass Valve) Setpoint controls are confusing. The procedure directs placing the controller in auto and never has a setpoint to control to. The current setpoint is set at the high end of the scale, so the bypass valve will never modulate closed. The calc note states that signals will be set based on CDS header and polisher flow. No setpoint is yet determined.


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IDS Charger Capacity and Design Float Voltage Requirement are Incompatible

SCR-DR-6400 This issue impacts the following RO/SRO task: RO-LT-R-IDS.002 Monitor the IDS

Disposition

SNC has determined that this issue does not impact the simulator’s suitability of the simulator for the conduct of operating tests. This discrepancy is due to partially implementing a forthcoming design change in different documents. The output voltage of the IDS chargers will be increased at a later date. The simulator is properly modeling the design documentation that it was built to. The lower output voltage indication will not drive operators to perform or not perform any actions.

Description Reference documentation for IDS currently states the rated voltage for the battery charger is limited to 250 VDC. Other documentation also states that the battery (IDS) should be normally on a float charge of 264 VDC. This cannot happen without a larger battery charger.


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Disposition

SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests. The current simulator implementation does not provide the capability of the instructor to insert a malfunction that will result in the actuation of a Residual Bus Transfer. However, the Fast Bus Transfer and the Diesel Generator starting sequence function properly and provide the capability to examine the operators on electrical failures that would result in similar indications and Abnormal Operating Procedure entries. Description For undervoltage conditions (loss of power) sensed by 27B2 (two-out-of-two or two-out-of-three logic) in conjunction with a source undervoltage condition sensed by 27S, the UAT breaker will be tripped, leading to automatic closing of the RAT breaker completing residual bus transfer after establishing that the RAT source is live (59S1), the bus is dead (27B2), and all motor feeders are tripped.


• At 75% rated voltage (~ 3 sec time delay) the load shed occurs. The associated bus output breakers are tripped open. FOR ES-1 and 2 ONLY the associated DG starts.

• At 30% rated voltage the residual bus transfer occurs, the UAT source supply breaker opens and the RAT source supply breaker shuts for the associated bus.



ECS-EC-313 Loads not modeled



Description APP-ECS-E3-EC31302 Rev 1 depicts ECS-ET-3131. Model requirements document SV3-STS-J4-119 Rev 1 required WEC/GSE to model the loads of EC-313. Work in SCR-DR-6568 determined they neglected to meet this requirement. SV3-STS-J1-119 Rev 0 shows loads off of EC-312 but nothing on EC-313.


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D/G Sequencer Operation

SCR-DR-6610 This issue impacts the following RO/SRO task: RO-ELE-ZOS-002-06 Perform priority load AC load sequencing and verify proper 6.9 kV priority load system voltage


Description According to APP-ZOS-E0C-001 (Onsite Standby Power System Diesel Generator Sizing Calculation) Rev 1, the following busses should get re-energized by the sequencer at Load Step [

]a,c

[ ]a,c [ ]a,c [ ]a,c [ ]a,c [ ]a,c [ ]a,c

These busses are not currently sequenced once ECS-ES-1/2 gets re-energized from its respective DG.





Description The turbine bypass control valves (MSS-PL-V001, MSS-PL-V002, MSS-PL-V003, MSS-PL-V004, MSS-PL- V005, MSS-PL-V006) have four solenoid valves that are designed to share two different power supplies. However, the current design control logic) uses one power supply for all 4 solenoids.


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Description While attempting to create a FR-Z.2 (Response to High Containment Level) scenario, it was discovered that a leak from the Fire Protection System (FPS) header in containment had no effect on any containment parameters (ex. Containment Sump Level, Containment Humidity). FPS Containment Spray also had no effect on any containment parameters.


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10 CFR 55.45(a)(7) Safely operate the facility’s heat removal systems, including primary coolant, emergency coolant, and decay heat removal systems, and identify the relations of the proper operation of these systems to the operation of the facility.

Executive Summary

Issues 6701, 6217, and 6179 are not issues impacting simulator training. For example, Time to boil is not a decision-making factor.

Issue 6022 currently has a work-around in place. Numerous work-around solutions become problematic, but there are currently few work-around solutions in place. The instructors make the present work-around solutions transparent. Continued work-around solutions are also tedious for the operators and it causes the NAPs to lose credibility.

Issue 5594 was determined not to impact reactor safety because it causes safeguards when it is required.

Issue 5968 only occurs during solid plant conditions so the impact to training is minimal. With operator action, simulator scenarios can still be successful. Also, if pressure control wasn’t a focus for one operator when the plant is solid, then this could be an issue, but that isn’t the concern.

Issue 6151 is mitigated by the manual operation of these valves when this issue manifests.

Issue 6634 was evaluated as not having an aggregate impact on operations.

The team determined that no additional reinforcements were required for completion of these tasks. Therefore, the team evaluated the collection of these issues as acceptable in the aggregate.

Repeatability issues involving CL 1B

SCR-DR-5594 This issue impacts the following RO/SRO task: RO-PRO-AOP-054-00 Respond to Reactor Coolant Pump Malfunctions using AOP-114 Disposition This issue was dispositioned as unacceptable by the Simulator Review Committee (SRC). This issue was further evaluated and determined to relate with a current design issue involving a lead/lag circuit associated with the main steam line pressure detection and input to the Safeguards ESF actuation. The pressure drop in the main steam line turns very close to the actuation set-point and the lead/lag circuit amplification of the rate of change may or may not cause the actuation. This is the expected plant response with the current actuation logic/software. Operators have sufficient procedure guidance directing them to respond to this event. For this reason, it was determined that this issue does not impact the suitability of the simulator for the conduct of operating tests.

Description

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The 1B RCP shaft shear malfunction may cause an unexpected safeguards depending on the initial conditions the malfunction is inserted. Due to response caused by the lead/lag filters in the steam line pressure logic, the signal is driven into the dead band range of actuation at certain initial steam pressures. Elevating the initial steam pressure completely mitigates this occurrence.

Area of Impact Abnormal operation response


SCR-DR-5968

This issue impacts the following RO/SRO task: RO-PRI-CVS-004-00 Monitor CVS operations



Per APP-CVS-M3C-101 Rev 6 Appendix C.3.2, at [ ]a,c when CVS-V047 is in automatic pressure control mode there is supposed to be a signal sent to “trip the CVS Makeup Pumps.” The high pressure signal is generated but does not trip the pumps; it presently feeds a Pump Auto Stop Demand signal. This signal will stop any pumps that are running in automatic only. When the plant is in water-solid mode, as determined in logic diagrams as having CVS-V047 in automatic pressure control mode, Chemical and Volume Control System (CVS) makeup pumps must be operated in manual; see APP-CVS-M3C-100 Rev 11 Logic Sheet CVS-4 Note10. Since the CVS makeup pumps are in manual the Auto Stop Demand signal will not shut the pump off to provide overpressure protection.

Per APP-CVS-M3C-101 Rev 6 Appendix C.3.2, at [ ]a,c when CVS-V047 is in automatic pressure control mode there is supposed to be a signal sent to “trip the Reactor Coolant Pumps…in order to protect them from reduced suction pressure.” This signal is also discussed in APP-RCS-M3C-100 Rev 9 Logic Sheet RCS-13 table and note 11. As presently designed there is no logic tie between CVS-PT040 and the Reactor Coolant Pumps (RCPs) to prevent damaging the RCPs upon a loss of Reactor Coolant System (RCS) pressure.




Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC) due to the minimal training impact this particular NAP has.

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Description When attempting to simulate fuel assemblies being moved from the core to the Spent Fuel Pool (SFP), it was noted that the Decay Heat Calculation (DHC) NAP to maintain the administrative location of fuel does not work correctly. On display 40203 the assembly move buttons on the lower right portion indicate they are only available when the light DDS-AP-DHC Status indicates it is ACTIVE. This light is driven off of the automatic mode selector and is INACTIVE when in MODES 1&2 and ACTIVE in MODES 3- 6. However, when the light indicates INACTIVE the buttons for moving are raised and available. When the light changes status to ACTIVE the buttons for moving are grayed out and no longer available. The light being active or inactive is currently driven by the auto mode selector and becomes active in MODES 3-6. However, fuel cannot be moved from the core into the SFP in any MODE other than MODE 6. The light should be driven by the manual input of the Rx vessel head being removed or installed or upper internals position on display 40004.





Description At low pressure conditions less than 350 psig, the operator often has to take manual control of Startup Feed Water (SFW) control valves due to excessive cycling of the valves. Indicated flow rates range from 0 to greater than 600 gpm within 10 to 15 second cycles. This requires 100% of the operator’s attention until RNS can be placed into service removing cooldown function from the steam dumps.



SCR-DR-6179

Disposition

SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests. The Time to Boil NAP is a tool that is used for information only. The NAP is active when in Mode 5/6

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conditions for RCS time to boil and when fuel is present in the spent fuel pool for spent fuel pool time to boil. The value is displayed in exponential units versus hh:mm. This means that operators will need to convert the scientific notation values into hours and minutes. Although this does take a short amount of time, the net effect is that it does not remove the ability to monitor the time to boil and it has no impact on actions the operator may, or may not, take in response to plant conditions. Description When core exit temperature was 300oF, Thot was > 212oF and TTB was >0 (15-25 min range). Either the NAP is calculating TTB incorrectly or the inputs to the NAP are wrong. TTB should reflect actual plant conditions.

Area of Impact NAP




Description The WR Core Makeup Tank (CMT) level indications shift to Bad Quality once Automatic Depressurization System 1-3 (ADS 1-3) Actuate. Prior to this event, they would toggle to Bad Quality intermittently. The Bad Quality status is on indications PXS-LT009A/B & -LT010A/B (on PXS Supplemental Ind. Screen) and DDS-RSA11-L1 & DDS-RSA13-L1 (on WPIS screen 60017). The NAP driving the calculation of this indication drives them to bad quality whenever it determines voiding is occurring in the CMT (which is expected per design transients).





Description The turbine bypass control valves (MSS-PL-V001, MSS-PL-V002, MSS-PL-V003, MSS-PL-V004, MSS-PL- V005, MSS-PL-V006) have four solenoid valves that are designed to share two different power supplies.

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However, the current design control logic) uses one power supply for all 4 solenoids.




Disposition The Aggregate Study team determined this issue does not have an aggregate impact on operations. Additionally, this modeling of the differential temperatures has no impact on operator actions because, by procedure, there are no operator actions related to IRWST temperature. For this reason, it was determined that this issue does not impact the suitability of the simulator for the conduct of operating tests.

Description [

]a,c


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10 CFR 55.45(a)(8) Safely operate the facility’s auxiliary and emergency systems, including operation of those controls associated with plant equipment that could affect reactivity or the release of radioactive materials to the environment.

Executive Summary

It would be desirable to repair the Core Makeup Tank Nuclear Application (CMT NAP) issue in this section (6217), but it doesn’t affect the section in aggregate. When issues 5546 and 6019 manifest on the simulator, the operators will still comply with Technical Specifications requirements. An additional Tech Spec call would have to be made with 6019 since there is an issue with Zinc addition in the CVS when power is lost. Ultimately, the containment isolation is the issue, but the team believes that this can be dealt with safely.

Issue 5546 will be transparent with most operators without a reference diagram and detailed exploration on the Ovation status screens. Ultimately, this is not an issue for the aggregate study.


SCR-DR-5546 This issue impacts the following RO/SRO tasks: RO-LT-R-EDS.001 Monitor the Non Class 1E DC and UPS system (EDS) for proper operation RO-LT-R-EDS.004 Respond to a loss of EDS DC power

Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC). SMEs determined that the current power supply arrangement was adequate to teach since it is per design documentation. Power supplies are an item that will be continuously taught as they are updated and changed.

Description A loss of individual Non Class 1E DC and UPS System (EDS) busses will result in incomplete system response. Some Ovation drops are not dynamically powered by the EDS model but are powered by a permanently energized model constant (specifically DPU047, DPU048, and DPU044). The load lists for the STS do not assign a power supply to all the Ovation drops so there is no plant design data to insert into the simulator.




Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC). When this issue manifests itself on the simulator, the operators will still comply with Technical

ND-15-1333

Enclosure 6


103 | P a g e

Specifications requirements. An additional Tech Spec call would have to be made since there is an issue with Zinc addition in the CVS when power is lost. For this reason, this issue does not impact the suitability of the simulator for the conduct of operating tests.






Description The WR Core Makeup Tank (CMT) level indications shift to Bad Quality once Automatic Depressurization System 1-3 (ADS 1-3) Actuate. Prior to this event, they would toggle to Bad Quality intermittently. The Bad Quality status is on indications PXS-LT009A/B & -LT010A/B (on PXS Supplemental Ind. Screen) and DDS-RSA11-L1 & DDS-RSA13-L1 (on WPIS screen 60017). The NAP driving the calculation of this indication drives them to bad quality whenever it determines voiding is occurring in the CMT (which is expected per design transients).





Description While attempting to create a FR-Z.2 (Response to High Containment Level) scenario, it was discovered that a leak from the Fire Protection System (FPS) header in containment had no effect on any containment

ND-15-1333

Enclosure 6


104 | P a g e

parameters (ex. Containment Sump Level, Containment Humidity). FPS Containment Spray also had no effect on any containment parameters.




Disposition The Aggregate Study team determined this issue does not have an aggregate impact on operations. Additionally, this modeling of the differential temperatures has no impact on operator actions because, by procedure, there are no operator actions related to IRWST temperature. For this reason, it was determined that this issue does not impact the suitability of the simulator for the conduct of operating tests.

Description [

]a,c


ND-15-1333

Enclosure 6


105 | P a g e

10 CFR 55.45(a)(9) Demonstrate or describe the use and function of the facility’s radiation monitoring systems, including fixed radiation monitors and alarms, portable survey instruments, and personnel monitoring equipment.

Executive Summary

There is only a single issue categorized here and therefore the team did not deem this an aggregate issue.

Simulator MCR missing Rad Monitoring Panel

SCR-DR-237 This issue impacts the following RO/SRO task: RO-INC-RMS-003-00 Startup and operate radiation monitors

Disposition This issue was dispositioned as acceptable by the Simulator Review Committee (SRC). There is little training value with this panel at this time. The indications that would otherwise be provided by this panel are available in Ovation. Therefore, operator actions are not impacted and it does not impact the suitability of the simulator for the conduct of operating tests.

Description Simulator MCR does not have the radiation monitoring panel on the back wall as depicted in the design reference.

Area of Impact Physical fidelity of simulator MCR and radiation control

ND-15-1333

Enclosure 6




ND-15-1333

Enclosure 7

List of Westinghouse Simulator Corrective Actions


ND-15-1333 Enclosure 7, Page 2 of 8 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility 1.0 Summary Description of Westinghouse’ Simulator Corrective Actions

In July, SNC commissioned a team to perform an aggregate study of all open

simulation discrepancies.

As of May 15, 2015, SNC had 166 open discrepancies associated with its simulator.

SNC evaluated these discrepancies and their impact on the suitability of the

simulation facility for the conduct of operating tests. Out of the 166 discrepancies,

101 were determined to be relevant to 9 of the 13 criteria listed under 10 CFR

55.45(a). SNC evaluated each of these 101 discrepancies and determined that no

singular issue challenged the simulation facility’s suitability for the conduct of

operating tests. SNC then commissioned a team with representatives from

Operations, Training, and Human Factors Engineering to perform a study to

determine if, in aggregate, these 101 discrepancies presented a challenge to the

simulation facility’s suitability for the conduct of operating tests. The team

determined that, in the aggregate, the discrepancies could challenge the suitability

of the simulator for the conduct of operating tests. Specifically, 10 CFR 55.45(a)

criterions (3) and (5) were impacted. It was further determined that these criteria

would no longer be challenged if a specified subset of discrepancies could be

corrected by WEC.

On July 7 h, SNC confirmed that the scope of this subset satisfied the near term

request to support the VEGP Units 3&4 STS CAS.

2.0 Itemized List of Westinghouse’ Simulator Corrective Actions

Table E7-1 lists the items that Westinghouse included in a patch delivered to SNC

on August 14, 2015. Verification and Validation (V&V) testing was performed and

the patch was deployed in simulator load V3.R1.7F8.1.1.0.


Table E7-1

List of Westinghouse’ Simulator Corrective Actions

A.S.

# {1}

NRC

UI # {*}

WEC RITS/CAPA

I:

SNC SCR

#

Other Ref #:

Summary V&V Test Results

1 RITS 37569 5584 Rod Withdrawal button un-highlights during continuous operation

Pass

2

RITS 39523

RITS 39409

CAPAL-100025685

5597 TO-117

Containment Radiation Alarm Reset Points

Acceptance criteria not fully met.


This is an alarm deadband issue that WEC must resolve.

If operators encounter this condition, they will follow their procedures. The procedure provides the steps necessary for operators to respond to the condition.

3 RITS 39633 5599 TO-09

Unidentified and Identified Leak Rate always

indicates BAD quality


See Table E5-1 Item 40.

4 1 5627 TO-40

Sub criticality on Critical Safety Function Screen bad quality

Pass

5 RITS 37623 5643 1503-

02 VWS-TE079 Labeled incorrectly Pass

6 5644 SSS Display # 17600 incorrect



This is a simulator I&C issue related to a non-consequential graphic indication.

7 RITS 39605 RITS 42463

5688 TO-25

Graphic 50308 Issue Pass

ND-15-1333 Enclosure 7. Page 4 of 8 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility


A.S.

# {1}

NRC

UI # {*}

WEC RITS/CAPA

I:

SNC SCR

#

Other Ref #:


8 RITS 39466 5689 TO-28

PMS Mimic Screens


(This item was not included in the patch received from WEC.)


This is a simulator I&C issue.

PMS mimic screens are used for verification of indications only. If a question arises regarding an indication on the PMS mimic screen, operators will use primary indications from the PMS displays on the Primary Dedicated Safety Panel (PDSP). No operator action is available through the PMS mimic screens. All actions must be taken from the division’s PMS PDSP.

9 RITS 42477 5702 1501-

07 IDS screens in simulator show inaccurate Power supplies

Pass

10 5712 1411-

06 Calorimetric Power Data Points do not have required precision

Pass

11

14

5813

1503-16

ISV Pri-1 HED-14 Alarm Overload Pass

12 RITS 42461 5909 TO-59

Digital Rod Control System (DRCS) M bank rod control graphic

Pass

13 RITS 38522 5920 1410-

6

Pressurizer narrow range pressure does not indicate bottom of scale on WPIS 2 for Mode 1-4 screen

Pass



A.S.

# {1}

NRC

UI # {*}

WEC RITS/CAPA

I:

SNC SCR

#

Other Ref #:


14 5924 1410-

2 DRPI Health Screen has alarms for Data Cabinet A and B crossed




15 5925 1410-

3 DRPI Health Screen (1805) Incorrect Logic Cabinet Alarms

Pass

16 26 RITS 38825 5968 1504-

09 CVS-PT040 do not function as described Pass

17 RITS 40030 6009 1504-

3

Uncontrolled H/U or C/D light on Mode 5/6 CSFST WPIS Display does not indicate correctly

Pass

18 6030 M Banks B & C Reversed on DRPI Health Screen

Pass

19 6078 HSS Display does not include ESOP Disch Pressure

Pass

20 RITS 38306 6089 NAP for 1/M Intermediate Range does not work


See Table E9-2.

21

RITS 43153 6102

MA Bank Rods Sometimes Stop at 263 steps during a CRE (Control Rod Exchange)

Pass

(This item was corrected locally by SNC. It was not included in the patch received from WEC.)

22 RITS 41846 6129 Display 40023 units issue Pass

23 3 6144 TO-52

PLS Auto Plant Mode Selector not consistent with Mode 3 to Mode 2 entry

Pass

24 40 RITS 39434 6159 NAPS display issues Pass



A.S.

# {1}

NRC

UI # {*}

WEC RITS/CAPA

I:

SNC SCR

#

Other Ref #:


25 RITS 39470 6160 CCS Screen issue Pass

26 RITS 39588 6164 WPIS downscale arrow issue



This is a simulator I&C issue.

This serves as backup indication only and has no impact on operator decision making.

27 RITS 39403 6165 WPIS Tavg Scale Pass

28 RITS 40059 6169 Erroneous NAP RSA behavior Pass

29 RITS 39783 6170 WRS graphic issue Pass

30 RITS 38601 6180 TTB unit indication Acceptance criteria not fully met.

See Table E5-1 Item 40.

31 RITS 40314 RITS 41693

6187 Rod stop logic Pass

32 RITS 41193 6259 Received Bank Sequence Out of Sequence Alarm

Pass

33

6267

Urgent Alarm (Causes control rods to swap to manual and stop) Occurs During Case 2 CRE


(This item was not included in the patch received from WEC.)


This is an AP1000 plant design issue. The simulator models plant design.

If operators encounter this condition, they will follow their procedures. The procedure provides the



A.S.

# {1}

NRC

UI # {*}

WEC RITS/CAPA

I:

SNC SCR

#

Other Ref #:


steps necessary for operators to respond to the condition.

34 6278 Battery bank indications are mislabeled for EDS1, EDS2, and EDS4

Pass

35

6302


Closed. Invalid.

Subsequent to the Aggregate Study team forwarding this issue to WEC for correction, WEC evaluated the issue and determined that this is per the AP1000 design and that the overlap is not improper. Therefore, no correction was required.

36 6315 Manual reactor trip alarm when one is not requested

Pass

37 6398 Screen 22101 has incorrect indication




38 6409 1501-

05 Graphic 1805 has reversed rods Pass

6483 1502-

13 During Load Rejection events, Load Unbalance response is inconsistent

Pass

39 6621 PR B lower detector failure is not compensated for by the RSA NAP

Pass

40 14 6651 Inconsistent priority levels of Data Processing Unit alarms.

Pass

41 RITS 45989 6698 Safety Mimic Display for SGS-V255A and B indicates BQ following a SFW Isolation

Pass



A.S.

# {1}

NRC

UI # {*}

WEC RITS/CAPA

I:

SNC SCR

#

Other Ref #:


35 None 1503-

15 Failure to identify CCS leak Pass

28 6830 1410-

07 Steam Dump Capacity Pass

37 RITS 40480 5679 TO-04

EDS performance on Battery Pass

5 5987 TO-89

Condensate Makeup flow rate Pass

10 RITS 42360 5722 1411-

03 MTS Alarm at 75% Power Pass

9 RITS 39748 5609 TO-131

Gland Seal Steam system pressure discrepancy between procedure and actuals

Pass

Notes: {*}

Numbers and descriptions correspond to the table “Summary of Unresolved Items as of 06-30-2015” as it appeared in an NRC letter dated July 2, 2015,

(Reference 1). {1}

Numbers correspond to the items as they appear in the “List of Proposed Actions” on page 13 of the Aggregate Study, Enclosure 6.

3.0 References

1. Virgil C. Summer Nuclear Station Units 2 and 3 - Request For A Commission-Approved Simulation Facility dated July 2, 2015 -

ML15182A097



ND-15-1333

Enclosure 8

Evaluation of Priority One (1) Potential Human Engineering Discrepancies (PHEDs) from Integrated Systems Validation (ISV) Daily Assessments

(Non-Proprietary)


ND-15-1333 Enclosure 8, Page 2 of 6 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility 1.0 Summary of Priority One (1) Potential Human Engineering Discrepancies (PHEDs)

from Integrated Systems Validation (ISV) Daily Assessment

Integrated System Validation (ISV) was conducted as part of the Human Factors Engineering

(HFE) Verification and Validation process. Conduct of the AP1000 ISV resulted in a number

of Potential HEDs (PHEDs). These PHEDs have the potential to impact the simulator, plant

design, operator training, and/or procedures.

In order to elevate simulator fidelity to a level suitable for the conduct of operating tests

commensurate with the requirements of 10 CFR 55.45(b), Priority 1 PHEDs specific to the

operation of the simulator have been evaluated and corrected. A list of all ISV Priority 1

PHEDs and how they were resolved appears in Table E8-1. Table E8-1 also includes

PHEDs related to other elements of the Integrated System Validation, specifically procedures

and training.

2.0 PHED Assessment

Westinghouse performed a preliminary assessment of ISV issues during the conduct of the

ISV to identify those scenario failures that could have benefited from a fourth trial run while

the Utility Crews were present. These assessments identified 15 Priority 1 PHEDs.

Westinghouse’ resolution of these 15 PHEDs identified four items requiring simulator model

or HSI changes and 11 PHEDs specific to procedures and training. The changes resolved

the Alarm Workload PHED.


Table E8-1

Priority 1 PHEDs

# PHED Description - (# trial failures of # trials

performed) Resolution

1

[

]a,c

([ ]a,c)

[ ]a,c

[ ]a,c

[ ]a,c

[ ]a,c

2

[

]a,c

([ ]a,c)

[ ]a,c

[ ]a,c

[ ]a,c

[ ]a,c

[ ]a,c

[ ]a,c

[ ]a,c

[ ]a,c

3

[

]a,c

([ ]a,c)

[ ]a,c

[ ]a,c

[ ]a,c

4

[

]a,c

([ ]a,c)

[ ]a,c

[ ]a,c

[

]a,c





[ ]a,c

[ ]a,c

5

[

]a,c.

([ ]a,c)

[ ]a,c

[

]a,c

6

[

]a,c

([ ]a,c)

[ ]a,c

7

[

]a,c

([ ]a,c)

[ ]a,c

8

[

]a,c

([ ]a,c)

[ ]a,c

9

[

]a,c

([ ]a,c)

[ ]a,c

[ ]a,c

[





]a,c

[

]a,c

10

[ ]a,c

([ ]a,c)

[ ]a,c

[ ]a,c

[

]a,c

11

[

]a,c

([ ]a,c)

[ ]a,c

[ ]a,c

[ ]a,c

[ ]a,c

[

]a,c

12

[

]a,c

([ ]a,c)

[ ]a,c

[ ]a,c

[ ]a,c

[ ]a,c

[

]a,c

13

[

]a,c

([ ]a,c)

[ ]a,c

[ ]a,c

[ ]a,c

[ ]a,c





14

[

]a,c

[ ]a,c

[ ]a,c

[ ]a,c

[ ]a,c

[

]a,c

[

]a,c

Using the above criteria, WEC provided an update that enhances alarm prioritization, greatly lowering the number of alarms that have to be addressed by the operator post-event. In addition, SNC enabled the [

]a,c that is part of the Alarm Presentation System design. As can be seen from the results below, the number of alarms are now in an acceptable range and the operator workload has been greatly reduced.

Event Alarms

Turbine Trip [ ]a c

Reactor Trip [ ]a c

ES1 [ ]a,c

ES2 [ ]a c

ES3 [ ]a,c

ES4 [ ]a,c

ES5 [ ]a c

ES6 [ ]a,c

LOOP with TT [ ]a,c

LOOP= loss of offsite power TT= turbine trip

15

[

]a,c

([ ]a,c)

[ ]a,c

[

]a,c

Southern Nuclear Operating Company Vogtle Electric Generating Plant (VEGP) Units 3 and 4

ND-15-1333

Enclosure 9

List of Open Simulator Discrepancies (Non-Proprietary)


ND-15-1333 Enclosure 9, Page 2 of 23 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility Summary Description of Open Simulator Discrepancies

There were 166 open simulator discrepancies as of May 15, 2015. Each discrepancy was

evaluated to determine if it impacted any of the criteria listed under 10 CFR 55.45(a). Out of the

166 discrepancies, 101 were determined to be relevant to the 13 criteria listed under 10 CFR

55.45(a). Since SNC’s evaluations and conclusions for these items are reflected and/or

imbedded in the Aggregate Study contained in Enclosure 6, they are not repeated here. Table

E9-1 lists these 101 items.

The remaining 65 discrepancies that were determined not to pose a challenge to the criteria

listed under 10 CFR 55.45(a) are listed in Table E9-2. A Training Needs Assessment as

defined in ANSI/ANS-3.5-1998, Section 4.2.1.4, was performed for these 65 discrepancies It

was determined that none of these issues impacted any of the six criteria listed under

ANSI/ANS-3.5-1998, Section 4.2.1.4. or any of the 13 criteria listed under 10 CFR 55.45(a).


Table E9-1

SCRs Impacting 55.45(a) Criteria Evaluated In Aggregate

SCR# Description Current Status

216 VCS fan response due to loss of power Open

237 Simulator MCR missing Rad Monitoring Panel Open

5546 EDS Power Supply to PLS/DDS cabinets not IAW EDS Load List Open

5577 RCP Net Positive Suction Head Curve - Display 60029 needs extension Open

5583 BEACON inoperability calculation in reactor core model Open

5584 Rod Withdrawal button un-highlights during continuous operation Closed

5593 Unstable VFS Containment Exhaust Flow Open

5594 Repeatability issues involving CL 1B Open

5597 Containment radiation alarm reset points Open

5598 PZR heater current indicates BAD quality at limits Open

5599 Unidentified and Identified Leak Rate always indicates BAD data Open

5603 Investigate validity of low flow alarm on TCS-FT007 Open

5609 GSS Header pressure will not maintain pressure as required Closed

5613 MFP 'B' Alarm Response Differs For Identical Fault Closed

5619 Stage 3 ADS box unused on Divisions C and D Open

5621 Problems with Alarm Cutout and RWS pressure alarms Open

5623 PZR pressure out of range indication not properly displayed Open

5627 Subcriticality on Critical Safety Function Screen bad quality Closed

5643 VWS-TE079 Labeled incorrectly Closed

5655 Plant issue with automatic control of DST level and autostart of standby Condensate pump

Open

5680 PMS Screens not updating during MCR/RSR transfer Open

5686 Degasifier Level Alarm Limits Open

5689 PMS Mimic Screens Open

5698 ISS freeze during PZR fill to solid Open

5707 HL Level and PZR WR Level Fluctuations during Midloop when filling In-Containment Refueling Water Storage Tank (IRWST) with CVS

Open

5712 Calorimetric Power Data Points do not have required precision Closed

5736 OPDMS RIL for M2 does not match COLR Rev. 0 Open

5813 Valve Modulating Status Alarms are a nuisance Closed

5828 VRS High Rad alarm when ES1 de-energizes and is re-energized from DG Closed

5903 Inconsistent OPDMS QPTR Indications Open

5910 VFD Transformer temperature Closed

5913 Print from NAP SRM not working Open

5914 VHS rad monitor response on loss of flow Closed

5920 Pressurizer narrow range pressure does not indicate bottom of scale on WPIS 2 for Mode 1-4 screen

Closed

5921 During simulator scenario validation CDS-TE040A/B range found to be inadequate. Open

5924 DRPI Health Screen has alarms for Data Cabinet A and B crossed Open

5925 DRPI Health Screen (1805) Incorrect Logic Cabinet Alarms Closed

5926 Startup Feedwater Control Valve is opening and closing with level high is associate Open




steam generator

5968 CVS-PT040 Does not provide the proper protective functions Closed

5972 CNMT Recirc Actuation for Div C and D not visible on PMS ESF Screen Open

6009 Uncontrolled H/U or C/D light on Mode 5/6 CSFST WPIS Display does not indicate correctly

Closed

6019 CVS-V094 Power Failure Open

6022 DHC Summary - Assembly Move NAP function doesn't work Open

6025 RCP Vibration Alarms Open

6030 M Banks B & C Reversed on DRPI Health Screen Closed

6038 Quality of RWS-V503 BAD at limits Open

6068 WLS-MP-08C will not pump monitor tank C Closed

6071 RCP stator temperature indication off scale low at lower speeds Closed

6078 HSS Display does not include ESOP Disch Pressure Closed

6089 NAP for 1/M Intermediate Range does not work Open

6099 DWS-LT006 has insufficient range Open

6102 M Bank A Rods would not step out beyond 263 steps during a CRE Closed

6103 ECS penetration temperature reading off scale low Closed

6122 Improper function of C-2 Closed

6126 WRS sump pump B does not indicate proper discharge pressure Closed

6144 PLS Auto Plant Mode Selector not consistent with Mode 3 to Mode 2 entry Closed

6151 SFW control valve cycling Open

6152 SWS temperature control Open

6154 WPIS RCS inventory screen issues Open

6156 FWS-V037 Control issue Open

6157 SGS MSL drain pot erratic indication Open

6159 NAPS display issues Closed

6162 Stuck Rod Recovery Malfunction Open

6164 WPIS downscale arrow issue Open

6168 Tuning of VBS required Open

6169 NAP RSA behavior Closed

6171 APS ZVS and ZBS alarm scaling Open

6172 Polisher bypass valve control Open

6175 Flux doubling difference between divisions Closed

6179 TTB calculation Open

6186 Tracking issue for rod step sound problems Open

6190 WPIS display for VARs Open

6192 VFS radiation monitoring Closed

6217 CMT WR Level Indications go Bad Quality Open

6259 Received Bank Sequence Out of Sequence Alarm Closed

6267 Urgent Alarm during Case 2 CRE at 90% Power Open

6302 Improper bank overlap occurs when data point OCB07CE00C_OUTAV is incremented during Rx startup

Closed




6315 Manual reactor trip alarm when one is not requested Closed

6366 During Turbine Trips, Pri 4 controller faults for Drop 21 and 34 are received Closed

6392 Main Generator output breaker closes before it is supposed to when syncing the main generator to the grid.

Open

6398 Screen 22101 has incorrect indication Open

6400 IDS charger and documented float voltage are incompatible Open

6409 Graphic 1805 has reversed rods Closed

6481 Currently there is no way to cause a residual bus transfer of ECS-ES-1 thru 6 Open

6491 Diesel Fuel oil day tank level transmitters not operating correctly Open

6492 UAT Bkr Line Undervoltage Priorities are Wrong Open

6532 Any Rods at Bottom Alarm Open

6547 VES Supply Header Pressure not modeled correctly for a change in temperature w/o a change in mass

Closed

6593 Model ECS-EC-313 loads Open

6610 Potential issue with DG Sequencer Open

6612 Possible modeling and/or Ovation issues with WGS Sample Package MS-01 PS-001 Open

6613 Possible modeling and/or Ovation issues with WGS Sample Package MS-01 AT-032 (AE032)

Closed

6621 PR B lower detector was failed high, the RSA for Power Range Power did not eliminate this input causing an erroneous PR PWR read

Closed

6623 VZS-D014A/B and VZS-D015A/B do not fail as-is after loss of power. Open

6634 The Turbine Bypass Control Valve control scheme does not support multiple power supplies

Closed

6645 Battery Temperature does not trend during battery operations Closed

6651 Inconsistent priority levels of Data Processing Unit alarms. Closed

6657 Fire Protection System is not modeled in Containment Open

6670 PMS mimic screen navigation issues Open

6698 Safety Mimic Display for SGS-V255A and B indicates BQ following a SFW Isolation Closed

6701 Examine IRWST temperature Change Open

Total Remaining Open 62

Total Closed 39


Table E9-2

SCRs Not Impacting 55.45(a) Criteria

SCR# Summary Description Evaluation Basis

233

Simulator MCR missing cooling fins:

Simulator MCR is missing cooling fins as designated by the Unit 3 design documentation


Missing [ ]a,c was determined to be a physical fidelity issue that imposes no operational restrictions on the indications available to or actions taken by the operator. A Training Needs Assessment was performed and it was determined that this issue does not impact any of the six criteria listed under ANSI/ANS-3.5-1998, Section 4.2.1.4 or any of the 13 criteria listed under 10 CFR 55.45(a).

235

Simulator MCR lights not hanging from chains:

Simulator Main Control Room (MCR) lights are not located as designated by the Unit 3 Design Document APP-1242-EL-001.


The MCR lights not hanging from chains in the MCR was determined to be a minor physical difference between MCR and simulation facility that imposes no operational restrictions on the indications available to or actions taken by the operator. The current MCR lighting responds properly to loss of power scenarios. A Training Needs Assessment was performed and it was determined that this issue does not impact any of the six criteria listed under ANSI/ANS-3.5-1998, Section 4.2.1.4 or any of the 13 criteria listed under 10 CFR 55.45(a).

5540

ECS (1) Components Tagout Graphic display wrong:

When the Main AC Power System (ECS) ‘ECS (1) Components’ poke is selected from the Tagout Navigation Display, the graphic that appears is the CWS Components Tagout interface.

This issue has been corrected.

The display graphic was corrected with the recent simulator software update. The change has been tested and the graphics found to be functioning correctly.

5588

Tagged Components are not being captured in Snap:

When tagging out components, the current Initial Condition (IC) is snapped to incorporate this change in component status. When the system is reset to the same IC (newly snapped) the tagged out component does not reflect being tagged out. It seems the graphics are overriding the bits that allow the tagged-out component to remain in that state. After a second reset to the same IC, the graphics allow the component to be reflected as tagged-out (i.e., the correct status).


This issue is of the same nature as that described in SCR 5590 and 5905. Many simulators allow the booth operator to save a scenario’s initial conditions. This gives the booth operator the ability to efficiently reset the simulator to the same conditions when a scenario is to be repeated. [

]a,c and the booth operator must perform this task manually. The

scenario guide used by the booth operator contains guidance to this effect.




In either case, this activity is performed prior to trainees being admitted to the simulator and is completely transparent to the trainee.

A Training Needs Assessment was performed and it was determined that this issue does not impact any of the six criteria listed under ANSI/ANS-3.5-1998, Section 4.2.1.4 or any of the 13 criteria listed under 10 CFR 55.45(a).

5590

Clearance Tags Remaining after Simulator Reset:

Several components were tagged-out while running a scenario; the current Initial Condition (IC) was not snapped. The system was reset to a different IC and the components tagged-out in the previous IC were still reflecting a tagged-out status. The half of the components were manually cleared out and the system then was reset to a different IC, the components that were manually cleared remained clear and the other half remained tagged-out.


This issue is of the same nature as that described in SCR 5588 and 5905. The scenario guide used by the booth operator directs him/her to perform a check for clearance tags when resetting the simulator to a different initial condition.


5608

Simulator Operations (SIMOPS) datalink alarms incorrect:

When the Map and Mitigation Tools are setup to assign each individual datalink a unique entry in the configuration, the digital alarm points respond incorrectly when datalink failures are inserted.


[ ]a,c These alarms

are inconsequential to the MCR operator as the only action that would be required would be to notify the associated building operator and I&C to investigate the local equipment. This issue does not impact indications available to or actions taken by the operator.

Westinghouse stated that this is a software conversion tool issue.

5614

Ovation indicating incorrect time after backtrack:

After placing the simulator in FREEZE a BACKTRACK was performed. Once the simulator was reset into the backtrack IC and placed in Run, it was observed: the real time on Ovation was incorrect.


The BACKTRACK feature is a tool used during training scenarios to return to a given point in a scenario or evolution. The instructor uses this feature as a training tool to reinforce a particular point of interest relative to operator actions and plant transient responses. The existence of this fault is an inconvenience to the instructor and, worse case, deprives the instructor of an additional training tool. This issue is transparent to the trainees and does not impact the suitability of the simulator for the conduct of operating tests.




It is the expectation during exam scenarios that the entire scenario would be reset and re-performed or a different scenario would be used for evaluation if a simulator fault occurred. This fault does not affect the indications available to or actions taken by operators.


WEC is aware of the issue and a RITS number was provided. Since this issue does not impact the suitability of the simulator for conducting operating tests, no further action is planned to be taken by SNC simulator staff and the issue will be closed only after a WEC solution is made available.

5618

MSS-V016A/B Response during reactor shutdown:

During a reactor shutdown, it is instructed to shut down the MSR. Based on Document APP-MSS-GJP-101 Attachment 4, step 4.3.3 when the turbine load is 10%, control valves should remain closed but they open 5% causing an undesired RCS temperature transient.


This issue is related to a controller design implementation error. [

]a,c which is not consistent with plant design.

SNC has developed an APP file to override the controller and close the valves at the required turbine load. This results in the controller error being transparent to the operator.

SNC will continue to use this APP file until a permanent correction is provided by WEC.

5644

SSS Display # 17600 incorrect:

The Secondary Sampling System (SSS) display in Ovation indicates incorrect flow passing these valves SSS-V920 or SSS-V921.


This issue is related to the system mimic presented on the graphical display. [

]a,c The error does not change the operation of the valves or the system, just improperly represents the process fluid flow paths that will be diverted when the valves reposition.

These displays are infrequently used during normal routine operations.




5649

Reference plant procedure and limitations of CVS design to support plant heat-up are incompatible:

During a scenario it was discovered that the design of the CVS purification loop and letdown system was unable to handle a plant heat-up in accordance with GOP-107.

This is an AP1000 plant design issue. The simulator models plant design.

This issue does not impact the suitability of the simulator for the conduct of operating tests.

The AP1000 plant design allows for a heat-up rate that the letdown system cannot keep up with. [

]a,c

SNC operators are trained to follow their procedures. [

]a,c


SNC has determined that the guidance contained in the procedures is adequate to address the design issue and maintain safe operation of the plant.

This issue is similar to SCR-6158.

5656

Failed test-AP-OPS-EVO-003:

During ANSI ANS-3.5 test AP-OPS-EVO-003, Reactor Trip Recovery, an issue was identified resulting in the inability of the operator to recover the plant using reference plant procedures. Reference plant procedures do not provide guidance for proper plant alignment to perform a plant startup following an exit from the E-network. The test began with the plant at 100% full power and steady state. A manual reactor trip was directed and entry into E-0 commenced. From E-0, the operator [

]a,c When performing the Rx start-up using GOP-108, shutdown [

]a,c The Rapid


The Nuclear Application Program was designed [ ]a,c The

simulator modeling reflects the design. The issue here is that when the procedure was developed, it failed to include a step directing the operator to perform this action.

CR 898848 was written to identify the issue and GOP-108 was revised [ ]a,c A retest

of this evolution was performed and this issue was verified to be resolved by this procedural change.




power reduction latched signal was found still locked in following the manual reactor trip blocking the withdrawal of [ ]a,c The reference plant procedure (GOP-108) does not direct resetting the rapid power reduction latched signal prior to rod withdrawal. The plant is unable to recover to 100% full power from a reactor trip using reference plant procedures.

5688

Graphic 50308 Issue:

Reactor Cooling System (RCS) – TASK RCS Heatup/Cooldown Primary Side, Ovation screen shows the outlet of RCS-V007C going to IRWST. APP-RCS-M6-002 Rev 13, shows this line going to RCDT.



5702

IDS screens in simulator show inaccurate Power supplies:

IDS Screens (22701, 22702 and 22703) in Simulator show inaccurate Power Supplies for 24 Hr and 72 Hr Battery Chargers and Voltage regulating Transformers.



5708

Plant Radiation Detector response to entering Mode 6:

While establishing Midloop and Mode 6, plant radiation detectors steadily rose throughout entire evolution. Containment radiation shows high levels with the Refueling Cavity at its normal refueling level.


Determined that the required RCS cleanup and degasification were not completed prior to establishing the MODE 6 ICs. [

]a,c Therefore no effect on the indications available to or actions taken by an operator.

WEC is tracking RITS (41660) as the radiation systems are in preliminary development and expected to have setpoint and range changes during system testing/startup.

5905 User defined alarm limits not clearing on simulator reset:

Defined Alarm Limits (user) do not clear on RESET of


This issue is of the same nature as that described in SCR 5588 and 5590. The




an IC. They stay at the same value of the previous training session.

scenario guide used by the booth operator directs him/her to perform a check for any User Defined Alarm Limits that did not reset when resetting the simulator to a different initial condition.


This issue will remain open pending WEC correction.

5908

Shadow trails on Inverse Count Rate Ratio (ICRR) graphic 40052:

ICCR vs. Control Bank Withdrawal (Source Range) graphic 40052 is leaving "shadow" trails of previous calculations on WPIS monitors but not on any workstation screen. If the screen is refreshed the shadows disappear.


This affects the WPIS display ONLY. The operating station displays update properly. In addition, the “shadow” trail is similar to performing a manual ICRR plot with pencil and paper. The previous plot is not erased each time the plot is performed. Therefore, the “shadow” is an inconvenience, but all the proper information is still available to and will not lead to incorrect action by the operator.

5909

DRCS M bank rod control graphic:

Graphic 11181 – M and SD Bank Control, there is a “n” showing after an instance where the Rod Control rejects to Man (manual). Once the issue is cleared and Rods are returned to AUTO the “n” shows until the screen is refreshed.

The display graphic was corrected with the August 2015 update and tested to function correctly.

5911

MTS control valve position indication:

GOP-109, Rev. 1, checks the turbine control valves to be partially OPEN, the control valves are open (using signal diagrams) but indicated close on graphic 50211 of ovation.


This issue is observed during a turbine start-up where the control valves were previously closed, to perform the low speed rub checks, and the valves are now being re-opened to continue the turbine start-up. [

]a,c The issue has been tested at multiple different times in the evolution and proper plant response was observed in all instances.




As part of the training process, the operators are trained to monitor the valve positions as the step is performed and monitor turbine RPM. Although valve position is difficult to use, the turbine RPM rising and stabilizing at the required RPM verifies the control valves are performing the required function and the step can be considered to be met.

This does not require an operator to remember a specific detail about the Simulator, but enforces the need to use multiple indications to verify proper plant response to operator actions.

This is current plant design in regards to the graphics indications. The normal process for plant design change recommendations is in progress to consider changing the accuracy provided on the valve position graphic (add additional decimal places to indication to see smaller changes in valve positions).

CR 10072985 was created to provide recommendations for procedure changes to more accurately portray all the indications the operator will need to use.

5915

Xenon graphic response:

OPDMS XENON CONTROL graphic is not responding. Using NRFE-07-100, Revision 4 information, all the data points that feed the "tail" of the DOT never change off of zero. All we have seen is a DOT which never moves or has a tail.

The display graphic was corrected with the Aug 2015 update and tested to function correctly.

5919

Investigate cause of MOV malfunction discrepancy on PMS division B RNS-V002A/B:

During ANSI testing for a plant shutdown when the procedure directed the operator to open RNS-V002A/B from Division B PMS PDSP, a MOV malfunction discrepancy alert was highlighted on the PMS screen.


Per reference documentation V002A/B [

]a,c Therefore this is an expected indication for the operator and should not affect any operator actions while manipulating these components.

5958

Unexpected rise in indicated radiation on loss of process flow:

Ventilation System, including VHS, VRS and VFS radiation monitors are indicating high radiation after loss of process flow.


Two separate V&V tests were performed successfully.




5969

Inconsistent response of alarm cutout operation:

During a scenario an alarm on the annunciator board came in briefly. The alarm was promptly placed in the cutout alarm list for APS and did not appear on the current alarms.


This SCR is the parent (tracking) of two other SCRs (5606 & 5621) which are associated with [

]a,c It was determined that the evolutions performed that lead to these anomalies are infrequent and while it is not expected to get an audible alarm [

]a,c

5973

RCPs indicate 5 RPM when secured:

While performing a shutdown and cooldown, after RCPs secured, RCS-SI263/264 indicated they are at 5 rpm.


This issue is consistent with current plant design.

The operating procedures [

]a,c

5984

JStation will freeze randomly:

Identical to DR 5891, this issue will track final resolution of the overall problem of AOI sims timing out.

This issue is related to the instructor station randomly freezing such that the booth instructor was unable to provide the required simulations. A patch was provided to SNC in October 2014 which corrected the issue.

SNC is maintaining this SCR in an open/pending status to ensure any future baseline updates do not re-create this issue.

5995

JStation remains open when Stop Trainer is performed in the ISS in some occasions.

Trainer does stop but JStation windows do not close.


This is a training instructor inconvenience during shutdown of the simulator during the post training evolution. This process is transparent to the operator and does not affect indications available to or actions taken by an operator.

5996 JStation RUN button enabled when DCIS not-ready.


This is a training instructor inconvenience during ‘RESET’ of the simulator. The operators are not actively being trained or examined during this process and therefore




this issue is transparent to the operator.

5997

Turbine Runback Signal drop assignments:

The Load Correction Runback signal should originates in Drop 5 but the Simulator has it from Drop 6. The DeltaT Runback signal should originates in Drop 6 but the simulator has it from Drop 5.


The issue is related to what Ovation cabinet controller generates the signal. The Simulator performs a turbine runback as required when a demand is generated. All expected responses occur and therefore the origin of the runback signal is transparent to the operator.

6058

Simulation of VES Airflow Noise is not currently modeled.


The noise discussed being present or not present does not change the actions taken by an operator.

6081 CPS failure needs to respond to the simulator requirements for CPS as specified in design documentation.


Although the available fault from the instructor station is not working as expected, the ability still remains for the instructor to prevent proper CPS operation. The method of the fault is transparent to the MCR operator and therefore does not affect the indication available to or actions taken by the operator.

6091 Kirk Key interlock not operable on spare battery LOAs


The Kirk Key indication in question is associated [

]a,c the booth operator can make the appropriate reports to the MCR and this issue is transparent to the MCR operators.

6092 EDSS-DF-1 nomenclature and switch operability issue on simulator diagram for EDS system.


This issue is similar to SCR 6091. The spare battery is still able to be placed in service when required. The nomenclature and operability issues are related to the simulator instructor station and are transparent to the operator.




6101 Control Rod Drive Mechanism (CRDM) Fan Vibrations units in in/s instead of units in mils to be in alignment with plant reference procedures.


This issue was discovered while verifying surveillance procedures and determining which ones could be used for operational exams. The need to check these indications and the expected units is only provided in the System Operating Procedure (SOP) but not the Abnormal Operating Procedures (AOPs) and Emergency Operating Procedures (EOPs) this issue will be transparent to the MCR operators..

6128 Ovation 40023 graphic issue for the calibration offset AFD for quadrant A point name incorrect.


The graphics display error is that the nomenclature dropped off the final number; DDS-AP-BDP-01959 is displayed with the final ‘9’ dropped off as DDS-AP-BDP-0195. The graphic in question is used during the performance of the calorimetric surveillance. The graphic problem does not impact the operation of the simulator model and only requires instructor intervention during the performance of the surveillance to identify the incorrect graphic nomenclature and has been captured in the training material.

6129

Display 40023 units issue for points associated with Nuclear Power Correction Factor and Axial Flux Difference Correction factor:

Request units of %power instead of %.

This issue was corrected subsequent to the Aggregate Study.

V&V testing was completed satisfactorily.

6153 Letdown control setpoints do not respond accurately in automatic control to changing plant parameters during a depressurization.


These indications match current plant design.

The conditions at the time of the observation were that the plant was being depressurized using Aux Spray. [

]a,c This item is transparent to the operator.

6158 Letdown tuning:

CVS letdown heat exchanger alarm comes in when


SNC evaluated this item and determined that the simulator is modeling the AP1000




CVS makeup secured and letdown is in progress. plant design. This issue represents an AP1000 design inadequacy.

Current plant design [

]a,c Since the actions operators are expected to take in response to this condition are identical to those they would otherwise take, the issue was determined to have no impact on the simulator’s suitability for the conduct of operating tests.

This issue is similar to SCR-5649.

6160 CCS Screen issue for CCS header flow in scientific notation.



6165 WPIS Tavg Scale not precise for use by operator. This issue was corrected subsequent to the Aggregate Study.


6166 Turbine Control Protection System (TCPS) rotor stress units are metric. Request change to standard units.


The rotor stress inputs into the maximum RPM rate of change during a turbine startup based on the temperature variants across the rotor. [

]a,c Therefore, this does not change the available indications the operator will evaluate or affect the actions taken by an operator during turbine startup.

6167 ASS display units currently in psia. Request change to psig.


The pressure indication in question resides on the graphic associated with the Aux Boiler and the header pressure just downstream of the Aux Boiler outlet. [




]a,c The

display in question does indicate the correct pressure in ‘psia’. However, this graphic is not used by operators to verify header pressure and does not affect operator actions.

6170

WRS graphic shows the WHT room sump drain lines routed to Auxiliary Building sump. This is incorrect. The WHT room sump drain lines are routed to pump suction.



6173 WPIS nomenclature issue regarding Containment Floodup Level.


This issue stems from the procedure nomenclature differing from the graphic display nomenclature (i.e. procedure directs verification of CTMT FLOODUP LVL vice CTMT WR LEVEL). The two names are interchangeable. The graphic that was viewed was a wall panel display for Critical Safety Functions (CSFs). [

]a,c Therefore, these indications are the same and do

not affect actions taken by an operator.

6180 Time to Boil units on Mode 5/6 WPIS is in exponential units. Request change to standard units


Updates to the Time to Boil indications were included as part of the August 14, 2015, patch. After performing V&V testing, SNC personnel determined that the update was not successful. Time to Boil indication on the Mode 5/6 Primary trend WPIS was observed to be displayed in exponential minutes for the RCS time to boil. [

]a,c V&V testing was conducted under similar initial conditions as when the issue was first identified. The values indicated by each of these displays should be in hours and minutes. This V&V test failed.

The Time to Boil NAP is a tool that is used for information only. [

]a,c The value is displayed in exponential units versus hh:mm. This means that operators will need to convert the scientific notation values

into hours and minutes. Although this does take a short amount of time, the net effect is that it




does not remove the ability to monitor the time to boil and it has no impact on actions the operator may, or may not, take in response to plant conditions.

6181 Tuning of EHC HX is requested to prevent outlet temperature high alarm being received during normal system operation.


This has been identified as a plant design issue.

The applicable HX has been identified as being too large. Therefore, the temperature control valve is not able to throttle cooling flow sufficiently to prevent the process variable flow to reach a temperature which drives the temperature control valve closed. When temperature reaches a point to demand the temperature valve to open, [

]a,c This cycle has been observed to occur over an approximate 2 hour time period. If the alarm were to be received the operator is expected to take the required actions.

6182 Primary trend screen rendering issue for RCS Tcold and RNS flow when called up from two locations at the same time


[

]a,c The only observed anomaly was a fluctuation of the graphic display when selecting “Print.” This fluctuation has no impact on actions the operator may, or may not, take in response to plant conditions.

6185 Default trend screen color for ovation trend screen is yellow and makes reading text difficult


The ovation trend program was opened and the settings associated with trend color were observed. The first two points that are selected to be trended default to red and light blue. If more points than this are required the operator is able to change the trend color using the properties menu. The operator is also able to shift to a tabular view vice a graphical view to see the information. Although, inconvenient, the capability is available to change the trend color if sufficient trends are added to a single trend window that results in one of them being yellow and therefore does not affect the indications available to or actions taken by an operator.

6187 Rod stop logic . When Shutdown Bank 1 fully SNC has determined that this issue does not impact the simulator’s suitability for the




withdrawn, step counter indicates 265 steps. conduct of operating tests.

SNC evaluated this item and determined that the simulator is modeling the AP1000 plant design.

SNC was notified by WEC that the actual rod position will stop at 264 steps based on rod withdrawal limits. [

]a,c

WEC is updating [

]a,c The need for training on the revised procedure will be evaluated through

the Training Needs Analysis (TNA) process.

6188 Plant Mode NAP Temperature Input for the Automatic plant mode calculation uses average cold leg temperature instead of RCS average temperature.


The plant conditions which have [

]a,c and the input that caused the MODE change is transparent to the operator.

6197 OCS Wall Panel Navigation System (WPNS) and Reactor Operator Peer Check System (ROPCS) Rebuild Required


The OCS WPNS and ROPCS have deficiencies in the software resulting in a loss of communication between the various WPIS displays and operator stations. The frequency with which this occurs is low. When it does occur, operators respond as they would to any problem with DDS (perform actions of the Abnormal Operating Procedure) and therefore are taking the actions they would be expected to take if the same event were to occur in the plant.

6207 Can't open signal diagrams from APS


The inability to open signal diagrams specifically from APS does not hinder an operator’s ability to view the desired diagram and has no impact on the actions an operator.

6241 SBT/Replay tool issues SNC has determined that this issue does not impact the simulator’s suitability for the conduct of operating tests.




The replay tool is a capability of the simulator to be reset to a point in the past in order to re-run a scenario. This is a training tool.

6246

SMS Detector ranges not consistent with design statements:

Maximum indication on Ovation screens is [ ]a,c. Design documents list peak values of [

]a,c


Current training material directs the booth operator to insert user defined alarms to generate the alarms to allow operators to respond accordingly prior to the initiation of the training event, thus making this issue transparent to the operators.

6278 Battery bank indications are mislabeled for EDS1, EDS2, and EDS4


V&V completed satisfactorily.

6306 EHS Power Supplies not modeled


EHS is the system associated with heat trace for outside components that require cold weather protection. The impact of this issue would only be associated with scenarios specifically designed for cold weather mitigation. There are no training scenarios that require the use of EHS at this time. Therefore has no effect on the indications provided to or actions taken by an operator.

Although some portions of EHS were modeled, modeling of this portion was considered to be beyond the scope necessary to be simulated.

The SCR will continue to be tracked for possible model changes in the future to allow scenarios that result in temperature swings in plant areas.

6410 VFS/VAS performance during outside containment loss of cooling accident scenarios.


This issue was determined to be a plant design issue. The simulator is functioning consistent with current plant design.

WEC has corrected, but not implemented, a plant design change to address this issue.

SNC is keeping this SCR open for tracking purposes until the change has been implemented.

6418 Same points not available for similar equipment in the radiation monitoring system.


The identified points are associated with power availability to the radiation detectors. These points were identified as being additional information that would be beneficial for




incorporation into CPS logics for entry conditions into the associated Abnormal Operating Procedure. However, these specific indications are not required by any procedure to be viewed and a power loss would also be indicated by the detector values indicating ‘Bad Quality’. Therefore this does not impact the indications available to or actions taken by an operator.

6447 PCS Indications - Inconsistent Naming for containment pressure


The issue is associated with the nomenclature associated with CTMT pressure on the Critical Safety Function (CSF) WPIS display. On this display the CTMT pressures are labeled as ‘ExtR’ and ‘NormR’, meaning ‘Extended Range’ and ‘Normal Range’. Although the naming of the points on the graphics is not consistent, the information being provided by the graphics is readily identifiable. On the PCS graphic, which contains the instruments that provide the input to the CSF display, the instruments are labeled as ‘Wide Range’ and ‘Narrow Range’. For this reason it does not affect the indications provided to or actions taken by an operator.

6470

APS allows priority 1 alarm suppression via consequence logic:

This was discovered during Alarm Avalanche evaluation.

Closed. Invalid.

This issue has been determined to be invalid based upon updated information received from WEC. The AP1000 plant design permits the [

]a,c

6471 RCS flow during loss of all RCP Verification – SCR is to investigate a [ ]a,c level rise during the loss of all RCPs

SNC has determined that this issue is invalid. The issue is closed.

SNC investigated this issue and determined that this is in accordance with the AP1000

design and that this is the expected plant response.

SCR resolution [ ]a,c This issue was closed because this is

the expected simulator response.

6482 Steam generator parameters oscillating during a main steamline break outside containment.

Closed, could not duplicate.

This scenario was re-performed twice by SNC simulator staff under the conditions under which the issue was first identified and was unable to elicit the same response on its simulator. Additionally, the domestic AP1000 simulator that identified this issue was only able to reproduce the issue on one of its two simulators. Because SNC was unable to reproduce the issue and because the discovering organization was unable to reproduce the issue on their second simulator, the issue was closed as invalid.




6498

Demineralized Water Feed Pump A and B (DWS-MP-01A/B) do not have a poke for operator use nor are they modeled in instructor station for the instructor to locally control.


Design change APP-GW-GEE-2907 replaced the motors for the DWS pumps AND moved the control of the pumps to the Local Control Station. Therefore, operation of these components from the MCR is no longer in the plant design and is consistent with the controls currently available to the MCR operators.

6626

TCS-V025, Main Turbine Lube Oil Coolers Flow Control Valve, should switch setpoint values (SP) in AUTO after transition from normal operations to turbine startup


This issue is related to a design document that infers an automatic change to the setpoint based [ ]a,c However, only one design document infers this with no discussion of this feature anywhere else in the document. Also, the instruments stated as providing an input to this automatic feature do not demonstrate having this output to the controller. This may be a design change that has not fully been implemented in all documentation. WEC has been tasked with determining the actual plant response required. Procedures call for operators to “Ensure” the controller setpoint. When this step is performed, the operator manually adjusts the setpoint to the correct value. The system then maintains the correct temperature in AUTO.

Finally, the system training material does not contain any reference to this automatic setpoint change and therefore is currently not an automatic function the operators would identify as not operating correctly.

6635 All Buildings Drain System (WWS) not modeled correctly.


This modeling deficiency was identified and is an ongoing simulator group project to ensure proper building drain indications. Significant items, ones that involve current training material (and the stem of this SCR being created), were prioritized and were corrected first. Therefore, the current training material will not be affected and there will be no effect on the indications available to, or the actions taken by, the operator.

6646 ECS 120V Bus voltage not properly modeled. Instead of 120V, they are modeled to 208V.


The 120V bus indications are not available to the student on any graphic. Therefore this issue does not have an effect on indications available to or actions taken by an operator. When a component powered by these busses is required it will energize and




function properly.

6668

“Show all points” tiles not printing to CSV (Comma-Separated Value). The points can be displayed on the APS screen, but will not print to a CSV file. The points can be printed to paper.


This feature is used during simulator testing and data collection. In the normal case of training or exam scenarios the operators would not use or be required to use this function of APS. This inability to create these files in a .csv file format does not impact the indications available to, or actions taken by, an operator in the MCR.



ND-15-1333

Enclosure 10

BEACON


ND-15-1333 Enclosure 10, Page 2 of 2 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility 1.0 Summary Description of BEACON

Definition

BEACON (Best Estimate Analyzer for Core Operations – Nuclear) is a Westinghouse reactor core monitoring application. BEACON converts plant instrument readings and performs calculations in order to provide measurement and analysis of core performance.

Description

The AP1000 seamlessly integrates BEACON into everyday plant operations; all calculations and programming are transparent to the Control Room operators. BEACON directly interfaces with the plant data network. This allows operators to monitor core performance using the same workstations and software used for daily operations. This data (originating from BEACON) is literally one click from their top-level display. Unlike the AP1000, BEACON was a post-design modification for legacy Westinghouse plants. Those plants were originally designed with other means to monitor in-core reactor performance. Starting in approximately 1990, legacy plants began slowly incorporating BEACON into their systems in varying degrees, requiring additional hardware and software in the plant and simulator.

2.0 BEACON Simulation

Core power distribution data in the AP1000 simulator is generated by complex algorithms and code to accurately represent reactor behavior. Because the AP1000 core monitoring interface is fully integrated with the normal control system (and simulator), there is no need for any additional hardware or software systems in the simulator. The simulator has the ability to model a BEACON failure. With this level of functionality, the core parameters modeled and displayed in the simulator are indistinguishable from the actual plant.



ND-15-1333

Enclosure 11

Acronyms & Definitions


ND-15-1333 Enclosure 11, Page 2 of 6 VEGP Units 3 and 4 Request for a Commission-Approved Simulation Facility 1.0 Acronyms

ADS Slow Primary System Depressurization to Saturated Condition

ANS American National Standards

AO Axial Offset

APP Application

APS Alarm Presentation System

ATWS Anticipated Transient Without SCRAM

BEACON Best Estimate Analyzer for Core Operations – Nuclear

BL7 Baseline 7

BL8 Baseline 8

CAP Corrective Action Program

CAPAL Corrective Action, Prevention and Learning (WEC electronic document

control)

CAS Commission Approved Simulator

CB&I Chicago Bridge and Iron

CET Core Exit Thermocouple

CLFs Component Level Failures

CFR Code of Federal Regulations

CMS Configuration Management System

CMT Core Makeup Tank

CR Condition Report

CSF Critical Safety Function

CSV Comma-Separated Value

CTMT Containment

DCP Design Change Package

DG Diesel Generator


DR Discrepancy Report

DRCS Digital Rod Control System

EHC Electro-Hydraulic Control

FW Feedwater

FRP Functional Restoration Procedure

GOP General Operating Procedure

GSE GSE Systems. This is the name of the simulator vendor contracted by

Westinghouse

HEDs Human Engineering Discrepancies

HFE Human Factors Engineering

HSI Human-System Interface

HX Heat Exchanger

I&C Instrumentation and Controls

IC Initial Condition

ICRR Inverse Count Rate Ratio

IIS Incore Instrumentation System

ISV Integrated Systems Validation

ITAAC Inspection, Test, Analysis and Acceptance Criteria

IR Intermediate Range Power

IRWST In-Containment Refueling Water Storage Tank

IVR In-Vessel Retention

LOAs Local Operator Actions

LOCA Loss of Coolant Accident

MS Main Steam

MSL Main Steam Line

MSR Moisture Separator Reheater


NAPS Nuclear Application Programs

NRC Nuclear Regulatory Commission

OPDMS Online Power Distribution Monitoring System

P1 Priority 1

P2 Priority 2

PBX Private Branch Exchange

PDSP Primary Dedicated Safety Panel

PHED Potential Human Engineering Discrepancy

PORVs Power-Operated Relief Valves

PRA Probabilistic Risk Assessment

PRS Plant Referenced Simulator

PZR Pressurizer

RAIs Requests for Additional Information

RCPs Reactor Cooling Pumps

REN-MAN03 Reference to a specific RIHA

RIHA Risk Informed Human Actions

RITS RRAS Issue Tracking System (WEC CAP)

ROPCS Reactor Operator Peer Check System

Rx Reactor

SAT Systematic Approach to Training

SBT Simulator Scenario-Based Testing

SCANA SCANA is not an acronym, but is taken from the letters in South Carolina

SCR Simulator Change Request

SFCVs Start-up Feedwater Control Valves

SFW Startup Feedwater

SG Steam Generator


SGWL Steam Generator Water Level

SNC Southern Nuclear Operating Company

SOP System Operating Procedure

STGR Steam Generator

STS Simulator Training System

TB Turbine Building

TNA Training Needs Analysis

UIs Unresolved Items

UFSAR Updated Final Safety Analysis

VCS Virgil C. Summer Nuclear Station

V&V Verification and Validation

VEGP Vogtle Electric Generating Plant

VHS Health Physics and Hot Machine Shop HVAC System

WEC Westinghouse Electric Company

WPIS Wall Panel Information System

WPNS Wall Panel Navigation System

WR Wide Range

2.0 Definitions

Mantis An SNC database program used for tracking, software changes,

hardware changes, administrative issues, modeling, I&C and Design

Change Packages (DCPs) that affect the simulator

Priority 1 Westinghouse Alarm Criteria (based on App-DDS-J4-010, Appendix B,

Rev 2) - Less than five minutes to respond, consequence can be a plant

trip or ESF actuation. This also includes radiation release or protection of

personnel.

Priority 2 Westinghouse Alarm Criteria (based on App-DDS-J4-010, Appendix B,

Rev 2) - five to 20 minutes to respond prior to degradation to a P1

condition. This may also include alarms that are important to operability


requirements with time-sensitive actions. Examples include bistable trips

that result in a P1 condition.

Simulator Simulator Training System

Training Needs Assessment

An appraisal by a subject matter expert of a simulator deviation,

deficiency, or modification, and its relative importance to the operator as

required tasks are performed.



ND-15-1333

Enclosure 12

Westinghouse Authorization Letter CAW-15-4260, Application for Withholding Proprietary Information From Public Disclosure, Accompanying Affidavit, Proprietary

Information Notice and Copyright Notice


@ Westinghouse

Document Control Desk U S Nuclear Regulatory Commission Washington, DC 20852-2738

Westinghouse Electric Company New Plants and Major Projects 1000 Westinghouse Drive, Building 1 Cranberry Township, Pennsylvania 16066 USA

Direct tel: (412) 374-3382 Direct fax: (724) 940-8519

e-mail: [email protected] Proj letter: SVP _SV0_003472

CAW-15-4260 9/17/2015

APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE

Subject: Transmittal of Evaluation of APlOOO Simulation Facility Summary of Umesolved Items (Uis) Issued By the NRC, Commission Approved Simulator Aggregate Study - Simulator Training System Deficiency Impact on 10 CRF 55.45(a) Compliance, Evaluation of Priority One (1) Potential Human Engineering Discrepancies (PHEDs) from Integrated Systems Validation (ISV) Daily Assessments, and List of Open Simulator Discrepancies

The proprietary information for which withholding is being requested in the above-referenced reports is further identified in Affidavit CAW -15-4260 signed by the owner of the proprietary information, Westinghouse Electric Company LLC. The Mfidavit, which accompanies this letter, sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)( 4) of 10 CFR Section 2.390 of the Commission's regulations.

Accordingly, this letter authorizes the utilization of the accompanying Mfidavit by Southern Nuclear Company (SNC).

Correspondence with respect to the proprietary aspects of the Application for Withholding or the Westinghouse Affidavit should reference CAW-15-4260, and should be addressed to James A. Gresham, Manager, Regulatory Compliance, Westinghouse Electric Company, 1000 Westinghouse Drive, Building 3 Suite 310, Cranberry Township, Pennsylvania 16066.

Very truly yours,

(]~a_~ Paul A. Russ, Director

U.S. Licensing & Regulatory Support

cc: Richard Paese Sarah DiTommaso Gerry Couture Steven Radomski Mark Chitty Mark Crosby David Midlik Wes Sparkman

Westinghouse Westinghouse Westinghouse Westinghouse SNC SNC SNC SNC

CAW-15-4260 September 17, 2015

Page 2 of2

COMMONWEALTH OF PENNSYL V ANlA:

COUNTY OF BUTLER:

AFFIDAVIT

ss

CAW-15-4260

September 17, 2015

I, Paul A. Russ, am authorized to execute tbis Affidavit on behalf of Westinghouse Electric

Company LLC (Westinghouse), and tbat tbe averments of fact set forth in tbis Affidavit are true and

correct to tbe best of my knowledge, information, and belief.

Paul A. Russ, Director

U.S. Licensing & Regulatory Support

2 CAW-15-4260

September 1 7, 2015

(1) I am Director, U.S. Licensing &Regulatory Support, Westinghouse Electric Company LLC

(Westinghouse), and as such, I have been specifically delegated the function of reviewing the

proprietary information sought to be withheld from public disclosure in connection with nuclear

power plant licensing and rule making proceedings, and am authorized to apply for its

withholding on behalf of Westinghouse.

(2) I am making this Affidavit in conformance with the provisions of 10 CFR Section 2.390 of the

Commission's regulations and in conjunction with the Westinghouse Application for Withholding

Proprietary Information from Public Disclosure accompanying this Affidavit.

(3) I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating

information as a trade secret, privileged or as confidential commercial or financial information.

( 4) Pursuant to the provisions of paragraph (b)( 4) of Section 2.390 of the Commission's regulations,

the following is furnished for consideration by the Commission in determining whether the

information sought to be withheld from public disclosure should be withheld.

(i) The information sought to be withheld from public disclosure is owned and has been held

in confidence by Westinghouse.

(ii) The information is of a type customarily held in confidence by Westinghouse and not

customarily disclosed to the public. Westinghouse has a rational basis for determining

the types of information customarily held in confidence by it and, in that connection,

utilizes a system to determine when and whether to hold certain types of information in

confidence. The application of that system and the substance of that system constitute

Westinghouse policy and provide the rational basis required.

Under that system, information is held in confidence if it falls in one or more of several

types, the release of which might result in the loss of an existing or potential competitive

advantage, as follows:

(a) The information reveals the distinguishing aspects of a process (or component,

structure, tool, method, etc.) where prevention of its use by any of

3 CAW-15-4260

September 17, 2015

Westinghouse's competitors without license from Westinghouse constitutes a

competitive economic advantage over other companies.

(b) It consists of supporting data, including test data, relative to a process (or

component, structure, tool, method, etc.), the application of which data secures a

competitive economic advantage, e.g., by optimization or improved

marketability.

(c) Its use by a competitor would reduce his expenditure of resources or improve his

competitive position in the design, manufacture, shipment, installation, assurance

of quality, or licensing a similar product.

(d) It reveals cost or price information, production capacities, budget levels, or

commercial strategies of Westinghouse, its customers or suppliers.

(e) It reveals aspects of past, present, or future Westinghouse or customer funded

development plans and programs of potential commercial value to Westinghouse.

(f) It contains patentable ideas, for which patent protection may be desirable.

(iii) There are sound policy reasons behind the Westinghouse system which include the

following:

(a) The use of such information by Westinghouse gives Westinghouse a competitive

advantage over its competitors. It is, therefore, withheld from disclosure to

protect the Westinghouse competitive position.

(b) It is information that is marketable in many ways. The extent to which such

information is available to competitors diminishes the Westinghouse ability to

sell products and services involving the use of the information.

(c) Use by our competitor would put Westinghouse at a competitive disadvantage by

reducing his expenditure of resources at our expense.

4 CAW-15-4260

September 17, 2015

(d) Each component of proprietary information pertinent to a particular competitive

advantage is potentially as valuable as the total competitive advantage. 1f

competitors acquire components of proprietary information, any one component

may be the key to the entire puzzle, thereby depriving Westinghouse of a

competitive advantage.

(e) Unrestricted disclosure would jeopardize the position of prominence of

Westinghouse in the world market, and thereby give a market advantage to the

competition of those countries.

(f) The Westinghouse capacity to invest corporate assets in research and

development depends upon the success in obtaining and maintaining a

competitive advantage.

(iv) The information is being transmitted to the Commission in confidence and, under the

provisions of 10 CFR Section 2.390, it is to be received in confidence by the

Commission.

(v) The information sought to be protected is not available in public sources or available

information has not been previously employed in the same original manner or method to

the best of our knowledge and belief.

(vi) The proprietary information sought to be withheld in this submittal is that which is

appropriately marked in ND-15-1333, Enclosure 5P "Evaluation of AP1000 Simulation

Facility Summary of Unresolved Items (U1s) Issued By the NRC" (Proprietary),

"Commission Approved Simulator Aggregate Study - Simulator Training System

Deficiency Impact on 10 CRF 55.45" (Proprietary), ND-15-1333, Enclosure 8P

"Evaluation of Priority One (1) Potential Human Engineering Discrepancies (PHEDs)

from Integrated Systems Validation (ISV) Daily Assessments" (Proprietary), and

ND-15-1333, Enclosure 9P "List of Open Simulator Discrepancies" (Proprietary), for

submittal to the Commission, being transmitted by Southern Nuclear Company (SNC)

letter and Application for Withholding Proprietary Information from Public Disclosure,

5 CAW-15-4260

September 17, 2015

to the Document Control Desk. The proprietary infonnation is submitted to support the

review of the Southern Nuclear Company Vogtle commission approved simulator.

(a) This information is part of that which will enable Westinghouse to:

(i) Manufacture and deliver products to utilities based on proprietary

designs

(b) Further this infonnation has substantial commercial value as follows:

(i) Westinghouse plans to sell the use of similar information to its customers

for the purpose of licensing new nuclear power stations.

(ii) Westinghouse can sell support and defense of industry guidelines and

acceptance criteria for plant-specific applications.

(iii) The information requested to be withheld reveals the distinguishing

aspects of a methodology which was developed by Westinghouse.

Public disclosure of this proprietary information is likely to cause substantial harm to the

competitive position of Westinghouse because it would enhance the ability of

competitors to provide similar technical evaluation justifications and licensing defense

services for commercial power reactors without commensurate expenses. Also, public

disclosure of the information would enable others to use the information to meet NRC

requirements for licensing documentation without purchasing the right to use the

information.

The development of the technology described in part by the information is the result of

applying the results of many years of experience in an intensive Westinghouse effort and

the expenditure of a considerable sum of money.

6 CAW-15-4260

September 17, 2015

In order for competitors of Westinghouse to duplicate this information, similar teclmical

programs would have to be performed and a significant manpower effort, having the

requisite talent and experience, would have to be expended.

Further the deponent sayeth not.

PROPRIETARY INFORMATION NOTICE


Transmitted herewith are proprietary and/or non-proprietary versions of documents furnished to the NRC in connection with requests for generic and/or plant -specific review and approval.

In orderto conform to the requirements of 10 CFR 2.390 of the Commission's regulations concerning the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary versions is contained within brackets, and where the proprietary information has been deleted in the non-proprietary versions, only the brackets remain (the information that was contained within the brackets in the proprietary versions having been deleted). The justification for claiming the information so designated as proprietary is indicated in both versions by means of lower case letters (a) through (t) located as a subscript innnediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (4)(ii)(a) through (4)(ii)(t) of the Affidavit accompanying this transmittal pursuant to 10 CFR 2.390(b)(l).

COPYRIGHT NOTICE

The reports transmitted herewith each bear a Westinghouse copyright notice. The NRC is permitted to make the number of copies of the information contained in these reports which are necessary for its internal use in connection with generic and plant-specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding. With respect to the non-proprietary versions of these reports, the NRC is permitted to make the number of copies beyond those necessary for its internal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary.

Southern Nuclear Company (SNC) Letter for Transmittal to the NRC


The following paragraphs should be included in your letter to the NRC Document Control Desk:

Enclosed are:

l. One copy of ND-15-1333, Enclosure 5P "Evaluation of APlOOO Simulation Facility Summary of Unresolved Items (Uis) Issued By the NRC" (Proprietary)

2. One copy of ND-15-1333, Enclosure 5NP "Evaluation of APlOOO Simulation Facility Summary of Unresolved Items (Uis) Issued By the NRC" (Non-Proprietary)

3. One copy of "Commission Approved Simulator Aggregate Study -Simulator Training System Deficiency Impact on 10 CRF 55.45(a) Compliance" (Proprietary)

4. One copy of "Commission Approved Simulator Aggregate Study - Simulator Training System Deficiency Impact on 10 CRF 55.45(a) Compliance" (Non-Proprietary)

5. One copy of ND-15-1333, Enclosure 8P "Evaluation of Priority One (1) Potential Human Engineering Discrepancies (PHEDs) from Integrated Systems Validation (ISV) Daily Assessments" (Proprietary)

6. One copy ofND-15-1333, Enclosure 8NP "Evaluation of Priority One (1) Potential Human Engineering Discrepancies (PHEDs) from Integrated Systems Validation (ISV) Daily Assessments" (Non-Proprietary)

7. One copy of ND-15-1333, Enclosure 9P "List of Open Simulator Discrepancies" (Proprietary)

8. One copy ofND-15-1333, Enclosure 9NP "List of Open Simulator Discrepancies" (Non-Proprietary)

Also enclosed is the Westinghouse Application for Withholding Proprietary Information from Public Disclosure CAW-15-4260, accompanying Affidavit, Proprietary Information Notice, and Copyright Notice.

As Item 1 contains information proprietary to Westinghouse Electric Company LLC, it is supported by an Affidavit signed by Westinghouse, the owner of the information. The Affidavit sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)( 4) of Section 2.390 of the Commission's regulations.

Accordingly, it is respectfully requested that the information which is proprietary to Westinghouse be withheld from public disclosure in accordance with 10 CFR Section 2.390 of the Commission's regulations.

Correspondence with respect to the copyright or proprietary aspects of the items listed above or the supporting Westinghouse Affidavit should reference CAW-15-4260 and should be addressed to James A. Gresham, Manager, Regulatory Compliance, Westinghouse Electric Company, 1000 Westinghouse Drive, Building 3 Suite 310, Cranberry Township, Pennsylvania 16066.

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