Reactors Annual Conference September 17th 21st, …muconf.missouri.edu/trtr2017/TRTR2017Program.pdf · Reactors Annual Conference September 17th ... USS Midway (CV-41) ... Amgad Shokr,

The University of Missouri Research Reactor Center Presents:

Test, Research, and Training

Reactors Annual Conference

September 17th—21st, 2017

Manchester Grand Hyatt

San Diego, CA

Background & Information

TRTRThe National Organization of Test, Research, and Training

Reactors (TRTR) represents research reactor facilities across the

nation from government, major universities, national laboratories,

and industry. TRTR's primary mission is education, fundamental

and applied research, application of technology in areas of national

concern, and improving U.S. technological competitive-ness around

the world. TRTR membership includes managers and directors of

research reactors, educators, administrators, regulators, and

research scientists and engineers.

Begun as a small technical group in the sixties, TRTR

quickly grew into a national organization and adopted its current

name in 1976. The organization regularly holds an annual

conference, hosted by a member institution, to discuss current

technical and regulatory issues, advances in research and

education, operating experience, and development of new

applications in medicine, materials, health and safety, information

technology, and environmental sciences, among others. TRTR

provides expert technical assistance to member institutions and

others through peer reviews, audits, and assessments.

It also publishes a quarterly newsletter, which provides the

latest information in all areas of interest to the membership. The

newsletter is widely distributed within and outside TRTR, in the U.S.

and abroad.

2017 TRTR MeetingUniversity of Missouri Research Reactor Center (MURR)

1513 Research Park DR

Columbia, MO 65211

(573) 882-4211

Manchester Grand Hyatt San Diego

1 Market PL

San Diego, CA 92101

(619) 232-1234

**Program provided by sponsorship from STS Nuclear**

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Background & Information

USS Midway (CV-41)The USS Midway was the longest-serving aircraft carrier in

the 20th century. Named after the climatic Battle of Midway of June

1942, Midway was built in only 17 months, but missed World War II

by one week when commissioned on September 10, 1945.

In 1946 it became the first American carrier to operate in

the midwinter sub-Arctic, developing new flight deck procedures.

The following year Midway became the only ship to launch a

captured German V-2 rocket. The trial’s success became the dawn

of naval missile warfare. Just two years after that, Midway sent a

large patrol plane aloft to demonstrate that atomic bombs could be

delivered by a carrier.

Midway served with the Atlantic Fleet for ten years, making

seven deployments to European waters, patrolling “the soft

underbelly” of NATO. A round-the-world cruise took Midway to the

west coast in 1955, where it was rebuilt with an angled deck to

improve jet operations.

Midway’s first combat deployment came in 1965 flying

strikes against North Vietnam. Midway aircraft shot down three

MiGs, including the first air kill of the war. However, 17 Midway

aircraft were lost to enemy fire during this cruise. Over a chaotic two

day period during the fall of Saigon in April 1975, Midway was a

floating base for large Air Force helicopters which evacuated more

than 3,000 desperate refugees during Operation Frequent Wind.

In 1990 Midway deployed to the Arabian Gulf in response to

the Iraqi seizure of Kuwait. In the ensuing Operation Desert Storm,

Midway served as the flagship for naval air forces in the Gulf and

launched more than 3,000 combat missions with no losses.

Its final mission was the evacuation of civilian personnel

from Clark Air Force Base in the Philippines after the 20th century’s

largest eruption of nearby Mount Pinatubo. On April 11, 1992 the

Midway was decommissioned in San Diego and remained in

storage in Bremerton, Washington until 2003 when it was donated

to the San Diego Aircraft Carrier Museum. It opened as the USS

Midway Museum in June 2004. and abroad.

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Schedule

Sunday, September 17th, 2017

8:00am-12:00pm ANS Stds. Mtg Nautical

2:00pm-5:00pm Exec. Comm. Mtg. Nautical

3:00pm-8:00pm Registration Coronado Foyer

6:00pm-8:00pm Reception Coronado Foyer

Monday, September 18th, 2017

7:00am-5:00pm Registration Coronado Foyer

7:00am-5:00pm Exhibits Coronado Foyer

7:00am-8:30am Breakfast Buffet Coronado Foyer/Terrace

8:30am-9:00am Welcome Coronado AB

9:00am-10:00am General Session Coronado AB

10:00am-10:30am AM Break Coronado Foyer/Terrace

10:30am-12:00pm General Session Coronado AB

12:00pm-1:30pm LUNCH ON OWN ON YOUR OWN

1:30pm-2:50pm General Session Coronado AB

2:50pm-3:20pm PM Break Coronado Foyer/Terrace

3:20pm–4:40pm General Session Coronado AB

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Schedule

Tuesday, September 19th, 2017

7:00am-5:00pm Registration Coronado Foyer

7:00am-5:00pm Exhibits Coronado Foyer


8:30am-10:00am NRC Session Coronado AB


10:30am-12:00pm NRC Session Coronado AB


1:30pm-6:00pm Sub Base Tour Submarine Base

Wednesday, September 20th, 2017

7:00am-11:00am Exhibits Coronado Foyer


8:30am-9:00am NRC Session Coronado AB

9:00am-10:00am General Session Coronado AB


10:30am-12:00pm General Session Coronado AB


1:30pm-2:50pm General Session Coronado AB

2:50pm-3:20pm PM Break Coronado Foyer/Terrace

3:20pm–4:20pm Business Meeting Coronado AB

7:00pm-10:00pm Midway Reception U.S.S. Midway

Thursday, September 21st, 2017

7:00am-8:00am Continental Breakfast Coronado Foyer/Terrace

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Monday, September 18th

8:30am-12:00pm – Session 1

INFRASTRUCTURE, SUPPORT, AND STATUS REPORTS

8:30am-9:00am WELCOME

9:00am-9:20am 2017 STATUS REPORT - RESEARCH REACTOR

INFRASTRUCTURE PROGRAM – Doug Morrell

9:20am-9:40am IAEA ACTIVITIES IN SUPPORT OF RESEARCH

REACTORS – Yeonggarp Cho

9:40am-10:00am KSU REACTOR CONSOLE REPLACEMENT STATUS

REPORT – Jeffrey Geuther

10:00am-10:30am MORNING BREAK

10:30am-10:50am STATUS OF THE ZERO ENERGY DEUTERIUM (ZED-

2) REACTOR – Jake Horner

LEU FUEL CONVERSION AND FUEL DEVELOPMENT

10:50am-11:10am FIRST REUSE OF TRIGA FUEL – Timothy Koeth

11:10am-11:30am ADVANCED TEST REACTOR (ATR) CONVERSION TO

LOW ENRICHED URANIUM (LEU) FUEL – Jeffrey

Bower

11:30am-11:50am SAFETY MARGIN EVALUATIONS FOR ATR IN-CORE

EXPERIMENTS SUPPORTING LEU U-MO FUEL

DEVELOPMENT – Kaichao Sun

12:00pm-1:30pm LUNCH ON YOUR OWN

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Monday, September 18th

1:30pm-5:00pm – Session 2

OPERATIONS AND MAINTENANCE OF REACTOR FACILITIES

1:30pm-1:50pm INSTALLING A PHOTONIS CFUF43 INCORE FISSION

CHAMBER TO CHARACTERIZE STEADY STATE

NEUTRON FLUX VARIATIONS IN TRIGA REACTOR –

Matthew Stokley

1:50pm-2:10pm INSTALLING A LOG CHANNEL AT THE REED

RESEARCH REACTOR – Melinda Krahenbuhl

2:10pm-2:30pm NBSR DEUTERIUM COLD SOURCE: COMPRESSOR

MOTOR STARTUP – Andrew Main

2:30pm-2:50pm IMPLEMENTING AGING REACTOR MANAGEMENT:

IF IT WERE EASY, EVERYONE WOULD DO IT –

Marcus Schwaderer

2:50pm-3:20pm AFTERNOON BREAK

3:20pm-3:40pm NUCLEAR CAMERAS FOR VISUAL INSPECTIONS

INCREASE OPERATIONAL SAFETY – Aaron Huber

REACTOR SAFETY

3:40pm-4:00pm EVALUATION OF GASEOUS EFFLUENT MONITOR

MAJOR SCRAM SETPOINT – Timothy Barvitskie

4:00pm-4:20pm ANALYSIS OF LOSS OF ELECTRICAL FEED TO THE

VITAL NETWORK OF THE BR2 REACTOR – Steven

Van Dyck

4:20pm-4:40pm TECHNICAL ANALYSIS AND ADMINISTRITIVE

ISSUES OF CRITICALITY STUDY FOR DIFFERENT

MITR FACILITIES – Kaichao Sun

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Tuesday, September 19th

8:30am-12:00pm – NRC Session

8:30am-10:00am NRC Presentations


10:30am-12:00pm NRC Presentations


1:30pm-6:00pm – Submarine Base Tour

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Wednesday, September 20th

8:30am-9:00am – NRC Session

8:30am-9:00am NRC Presentations

9:00am-12:00pm – Session 3

INSTRUMENTATION AND CONTROL AT RTR FACILITIES

9:00am-9:20am INVESTIGATION AND CORRECTION OF THE HIGH

THERMAL TO N16 RATIO IN THE SOUTHWEST LOBE

IN THE ATR – Mayra Morrison

9:20am-9:40am NBSR SECONDARY COOLANT FLOW CONTROL

SYSTEM – Marcus Schwaderer

9:40am-10:00am DIGITAL CONTROL AND SAFETY SYSTEM

MODERNIZATION FOR THE PENN STATE TRIGA

REACTOR – James Turso


10:30am-10:50am PENN STATE TRIGA REACTOR REACTIVITY

COMPUTER OBSOLESCENCE UPGRADE – James

Turso

10:50am-11:10am PUR-1 DIGITAL I&C PROJECT REVIEW AND

LESSONS LEARNED – Clive Townsend

11:10am-11:30am MANAGING SYSTEMATIC ERRORS IN THE NBSR

THERMAL POWER DETERMINATION – Marcus

Schwaderer

11:30am-11:50am WIDE RANGE CHANNEL AND MICROCONTROLLER

BASED SIGNAL CONVERTER FOR AN EXISTING

RESEARCH, TRAINING AND ISOTOPE PRODUCTION

NUCLEAR FACILITY – Benjamin Schlottke


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Wednesday, September 20th

1:30pm-5:00pm – Session 4

RESEARCH AND TEST REACTOR (RTR) UTILIZATION

1:30pm-1:50pm IRRADIATION TOOLS FOR NSUF MATERIALS

RESEARCH – Brenden Heidrich

1:50pm-2:10pm DELAYED NEUTRON SPECTROSCOPY FOR

CHARACTERIZATION OF SPECIAL NUCLEAR

MATERIAL – A. J. Shaka

2:10pm-2:30pm ACCESS TO IRRADIATION CAPACITY IN THE BR2

REACTOR – Steven Van Dyck

INFRASTRUCTURE, SUPPORT, AND STATUS REPORTS

2:30pm-2:50pm STATUS UPDATE ON MITR NUCLEAR SAFETY

SYSTEM UPGRADE – Edward Lau

LEU FUEL CONVERSION AND FUEL DEVELOPMENT

2:50pm-3:10pm TRIGA FUEL CLADDING CHEMICAL INTERACTIONS

– Eric Woolstenhulme

3:10pm-3:40pm AFTERNOON BREAK

3:40pm-4:40pm – TRTR Business Meeting

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Session 1 Abstracts:

9:00am-9:20am

2017 STATUS REPORT

RESEARCH REACTOR INFRASTRUCTURE PROGRAM

Douglas K. Morrell

Idaho National Laboratory

P.O. Box 1625

Idaho Falls, ID 83415-3890

This presentation will discuss the purpose and scope of the Department of

Energy - Research Reactor Infrastructure (RRI) Program. Personnel involved

in the program will be introduced and contact information will be provided for

team member. Information will be provided to conference attendees as to the

status of the core activities of the program. These activities include fresh fuel

element fabrication and spent nuclear fuel shipment returns to the DOE.

Current and future issues pertinent to the RRI program will also be presented.

The RRI program maintains fuels support contracts and provides nuclear

reactor fuel at no or low cost to 24 U.S. universities operating a total of 25

reactor facilities. These facilities include:

• Twelve TRIGA facilities

• Eight plate fueled facilities

• Three AGN facilities

• One Pulstar fueled facility

• One Critical facility

The title for the fuel remains with the United States government and when the

universities are finished with the fuel, the fuel is returned to the United States

government for long-term storage.

Mission of the Research Reactor Infrastructure Program:

The Research Reactor Infrastructure Program is funded by the U.S.

Department of Energy, Office of Nuclear Energy and is managed by the Idaho

National Laboratory (INL) in Idaho Falls, Idaho. The program goals are:

• Keep all U.S. operating university reactor programs supplied with

nuclear fuel.

• Provide assistance for movement of irradiated nuclear fuel from U.S.

universities, after the DOE receipt facility authorizes the fuel receipt.

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9:20am-9:40am

IAEA ACTIVITIES IN SUPPORT OF RESEARCH REACTORS

Yeonggarp Cho, Ram Charan Sharma, Andrea Borio di Tigliole,

Amgad Shokr, and Nuno Pessoa Barradas

International Atomic Energy Agency (IAEA)

Vienna International Centre

PO Box 100, 1400 Vienna, Austria

The IAEA, in its programme on research reactors, supports Member States in

enhancing safe and sustainable operation and effective utilization of research

reactors. This includes: (1) support for the development and implementation of

plans for operation and maintenance (O&M), ageing management, human

resource development, refurbishment and modernization and decommissioning

of research reactors as well as for the establishment of Integrated Management

System; (2) support to address various safety issues of concern and

strengthening regulatory supervision; (3) establishing and implementing

leadership and management for safety including enhancing safety culture; (4)

support for the development of the national infrastructure for new research

reactor projects following the IAEA Milestones approach; (5) support to

address research reactors fuel cycle issues related to fuel supply, best

practices for research reactor core management, development of strategies for

spent fuel management, HEU to LEU conversion and HEU minimization

activities and (6) support to preparation of Strategic and Business Plans for

research reactors, surveying and involving Stakeholders and enhancement of

research reactors utilization and promotion of research reactors products &

services for socio-economic development.

The programme on research reactors is implemented through various activities

that include: (a) development of the safety standards and supporting Member

States in their application; (b) peer review missions upon request by Member

States such as Integrated Safety Assessment for Research Reactors

(INSARR), Operation and Maintenance Assessment for Research Reactors

(OMARR) and the Integrated Nuclear Infrastructure Review for Research

Reactors (INIR-RR), and specific missions to address an area of concern; (c)

capacity building through organization of training workshops and courses at

national, regional, and international levels to address topical areas of concern

to research reactors; (d) establishment of research reactors networks and

coalitions (regional and/or thematic) including nuclear safety networks, regional

advisory safety committees, Internet Reactor Laboratory project (IRL), and

International Centre based on Research Reactor (ICERR); (d) Exchange of

experiences among Member States through organization of technical meetings

and conferences; (e) Co-ordinated research projects of interest to research

reactor community to address gaps in the existing knowledge.

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9:20am-9:40am

Additionally, IAEA manages information sources such as Research Reactor

Data Base (RRDB), the Research Reactor Ageing Data Base (RRADB) and

Research Reactor Material Properties Data Base (RRMPDB-under

implementation) and the Incident Reporting System for Research Reactors for

an effective operating experience feedback.

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9:40am-10:00am

KSU REACTOR CONSOLE REPLACEMENT PROGRESS REPORT

Jeffrey A. Geuther

Pennsylvania State University

101 Breazeale Nuclear Reactor

University Park, PA 16801

Amir A. Bahadori and Max E. Nager

Kansas State University

3002 Rathbone Hall

Manhattan, KS 66506

The Kansas State University TRIGA Mk II nuclear reactor facility was the

recipient of a $1.5 M DOE NEUP reactor infrastructure grant in 2015 for the

replacement of the nuclear reactor control console and flux monitoring

instrumentation. The period of performance for the grant ends in September,

2018, but it is expected that the new control console will be operational by the

end of January, 2018. The existing KSU console is a General Atomics design

from 1969, procured second-hand from the USGS reactor in 1991. While some

of the console instrumentation has been updated, much of the electronics are

original, and reliability issues have become increasingly common. The new

console and flux monitors are being designed by Thermo Fisher Scientific with

some auxiliary digital instrumentation provided by Schneider Electric. The new

console is intended primarily to address reliability concerns, however, it will also

include additional capabilities, an improved operator interface, and increased

redundancy of power channels. The following topics will be presented: console

design, including the use of analog controls and safety circuits with digital

auxiliary displays; 50.59 screening process; project status; anticipated

challenges during installation; and plans for installation, including startup

testing.

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10:30am-10:50am

STATUS OF THE ZERO ENERGY DEUTERIUM (ZED-2) REACTOR

Jake T. Horner

Canadian Nuclear Laboratories (CNL), Chalk River, Ontario, Canada

286 Plant Road, Chalk River

Ontario, Canada, K0J 1P0

The Zero Energy Deuterium Research Reactor (ZED-2) is a low power (200

W), heavy water moderated reactor located in Chalk River, Ontario, Canada at

the Canadian Nuclear Laboratories (CNL). ZED-2 first went critical in

September 1960. The vessel is 3.3 m tall and 3.3 m in diameter, and is open at

the top. Fuel assemblies are hung vertically in the reactor in a versatile manner

from movable beams that allow the fuel to be arranged in virtually any desired

configuration or lattice geometry. The ZED-2 credited safety system and reactor

protection is the three dump valves located at the bottom of the vessel. ZED-2

is controlled by moderator level, and upon any trip signal, these three dump

valves open and drain moderator from the vessel into three dump tanks.

Historically, the ZED-2 Reactor was used for activities such as CANDU reactor

development, advanced fuel cycles, and detector calibrations. The ZED-2

Reactor is currently undertaking a multi-year experiment to study kinetics with

mixed oxide fuels, including Pu-U, Pu-Th, and 233U-Th. Furthermore, a flux

perturber has been used in these kinetics experiments to probe the kinetics

parameters of the core during reactor operation. Moderator drains and online

coolant changes have also been used to study transients using the different

fuel types.

The ZED-2 Reactor has a counting laboratory to calibrate neutronic instruments

for the nuclear industry, internal use, and academia. Some examples of

calibrations include: self-powered flux detectors for power reactors; National

Research Universal (NRU) Reactor ion chambers; and fission chamber

calibrations.

Future proposals for work with the ZED-2 Reactor may include, but are not

limited to: potential usage of a light water moderator; continued support for the

annual ZED-2 Reactor Physics School; Small Modular Reactor support; and

possible design and construction of Supercritical Water Reactor (SCWR)

channels to support the Canadian Gen IV reactor design, the SCWR.

The ZED-2 Reactor has undergone several component replacements and two

large data acquisition system installations over the past few years. The

replacement of the pump timers, overhaul of the safety system comparator, and

installation of a safety system data acquisition system were recently completed

in the last two years as part of modernization efforts. A dump valve data

acquisition system was installed and is currently being tied into the safety

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10:30am-10:50am

system data acquisition system. Underway, is the installation of new

pushbuttons and indicators for the entire control room console.

A System Health Program (SHP) has been developed for ZED-2 which includes

the aging management program, obsolescence management program,

maintenance activities, and spare parts inventory. The SHP provides a

framework to increase the safety and reliability of the systems. Each system

undergoes boundary definition, aging degradation mechanisms to monitor,

source and location of relevant system information, and a spare parts inventory

verification. All safety related systems in the ZED-2 Reactor are included in the

SHP under one of eight ‘mega’ systems; the program has been implemented

for five of the eight systems. The SHP results are presented to management to

direct resources appropriately to ensure reliability and plan for the future.

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10:50am-11:10am

FIRST REUSE OF TRIGA FUEL

Timothy W. Koeth and Amber S. Johnson

Maryland University Training Reactor

Materials Science and Engineering, University of Maryland

4418 Stadium Dr.

College Park, MD, 20742

The Maryland University Training Reactor (MUTR) is pleased to present on the

successful collaboration with the U.S. Department of Energy’s Idaho National

Laboratory for the first ever transfer of lightly irradiated TRIGA fuel to a

university campus. We will review the necessary agency requirements that had

to met before this delivery could become a reality. Also to be discussed is the

timeline for the installation of this repurposed fuel in our core. Finally, we would

like to recognize the efforts of the many people who made this a successful

shipment.

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11:10am-11:30am

ADVANCED TEST REACTOR (ATR) CONVERSION TO LOW ENRICHED

URANIUM (LEU) FUEL

Jeffrey O. Brower

ATR Fuel Management Support

Design Authority Lead for LEU Conversion

Idaho National Laboratory (INL) Battelle Energy Alliance (BEA)

PO Box 1625, 2525 Fremont Ave., Mail Stop 3890,

Idaho Falls, ID 83415-3890

The United States Department of Energy (DOE), National Nuclear Security

Administration (NNSA), Office of Material Management and Minimization

(MMM) is working to convert research reactors globally from highly enriched

uranium (HEU) fuel to low-enriched uranium (LEU) fuel. MMM and its

predecessor programs have converted or verified the shutdown of 88 HEU

research reactors and isotope production facilities, using LEU fuel developed in

the 1980s. However, there is a small set of high performance research reactors

which require a new high density LEU fuel for conversion, including six U.S.

high performance research reactors (USHPRRs), which include the Advanced

Test Reactor (ATR) and the ATR Critical (ATRC) Facility at the Idaho National

Laboratory (INL) near Idaho Falls, ID; University of Missouri Research Reactor

(MURR) at Columbia, MO; Massachusetts Institute of Technology Reactor

(MITR) at MIT in Cambridge, MA; High Flux Isotope Reactor (HFIR) at Oak

Ridge National Laboratory (ORNL) near Oak Ridge, TN; and National Bureau

of Standards Reactor (NBSR) at the National Institute of Standards and

Technology (NIST) Center for Neutron Research (NCNR) in Gaithersburg, MA.

These reactors will share the same uranium-molybdenum (U-Mo) fuel plate

design. The U-Mo fuel meat foil will have a zirconium (Zr) diffusion barrier with

aluminum cladding. Three different application methods for the Zr diffusion

barrier layer are being pursued – co-rolled, electroplating and plasma spraying.

Fabrication of the higher density LEU U-Mo fuel has presented several

challenges. Because the U-Mo fuel fabrication process uses the existing

stockpile of 93% highly enriched uranium (HEU) to be blended with

molybdenum and depleted uranium (DU) to form 19.75% LEU, cost benefits of

enriching to 19.75% will not be realized. Homogeneity of the U-235 and

molybdenum in the metal foil is a challenge due to U-Mo casting process

results in variations between adjacent cast U-Mo plates. LEU will not require

less U-235 as previously believed due to self-shielding properties of U-238

requiring additional U-235 to be loaded in each fuel element. Although the

overall dimensions of the ATR fuel element remain unchanged, the weight of

the LEU ATR fuel element is 72% greater than a HEU ATR fuel element.

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11:10am-11:30am

During waste recovery and recycling efforts, separation of aluminum cladding

from the HEU fuel meat is a relatively simple chemical dissolution, however,

separation of the Zr diffusion barrier from the U-Mo fuel meat is significantly

more challenging.

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11:30am-11:50am

SAFETY MARGIN EVALUATIONS FOR ATR IN-CORE EXPERIMENTS

SUPPORTING LEU U-MO FUEL DEVELOPMENT

Akshay Dave, Jonathan Morrell, Kaichao Sun, Lin-wen Hu

Massachusetts Institute of Technology

Cambridge, MA 02139

Joseph Nielsen, Paul Murray, Ryan Marlow


Idaho Falls, ID 83415

The MIT Research Reactor (MITR) and Advanced Test Reactor (ATR) are two

of the remaining five High Performance Research Reactors in the U.S. that are

still using High Enriched Uranium (HEU) fuel. In the framework of non-

proliferation policy, the international community aims to minimize the amount of

HEU used in civilian facilities. A new type of Low Enriched Uranium (LEU) fuel

based on an alloy of uranium and molybdenum (U-Mo) is expected to allow the

conversion. Developing and qualifying the LEU fuel is critical to the conversion

program. Multiple series of irradiation tests are being performed at the ATR.

The Mini-Plate-1 (MP-1) is one of the experiments that are designed to irradiate

aluminum-clad, monolithic U-Mo fuel plates. The projected "high power" fuel

samples face very limited safety margin according to the ATR safety basis. The

current criteria are maintained for all Condition 2 events (coastdown transient

and reactivity insertion accident) by verifying the critical heat flux ratio (CHFR)

is greater than two. However, the basis used to establish this limit is not well

defined and may be traced to research reactor licensing based on overly

conservative thermal hydraulic criteria.

This ongoing research starts from a thorough literature review of empirical

correlations for critical heat flux (CHF), onset of flow instability (OFI) and onset

of nucleate boiling (ONB). Monte-Carlo method based sampling technique has

been employed to understand the sensitivity of different thermal-hydraulic

parameters to applicable correlations. Thereafter, RELAP5 modeling and

uncertainty quantification will be involved for the considered Condition 2 events.

Utilizing the DAKOTA and RAVEN codes, the uncertainty and safety margin

with respect to CHF, OFI and ONB will be evaluated. Once the safety margin is

quantified, recommendations can be made to support potential safety basis

modifications that can expand the experimental operating envelope of the ATR

without compromising safety.

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1:30pm-1:50pm

INSTALLING A PHOTONIS CFUF43 INCORE FISSION CHAMBER TO

CHARACTERIZE STEADY STATE NEUTRON FLUX VARIATIONS IN TRIGA

REACTOR FOR A MORE ACURATE SHORT DURATION NEUTRON

ACTIVATION ANLYAISIS

Matthew B. Stokley and Larry E. Hall

The University of Texas at Austin

Nuclear Engineering Teaching Lab

10100 Burnet Rd., Bldg. 159

MC R9000

Austin, TX 78758-4445

The Neutron Activation Analysis (NAA) Pneumatic Sample System is a highly

utilized experimental facility within The University of Texas at Austin (UT at

Austin) Nuclear Engineering Teaching Laboratory (NETL). The NAA Pneumatic

Sample System is located within the outer grid plate ring of the NETL TRIGA II

The reactor power and neutron flux monitoring instruments, The NM-1000 and

two NP(P)-1000s, are not sensitive enough to detect minor variations in power

level that can introduce error in some NAA Pneumatic Sample System

applications. This becomes an even larger discrepancy when performing short

duration irradiations where ICS controlled steady state power can have neutron

flux variation up to 10% of desired power.

We proposed the installation of a Photonis CFUF43 inner core fission chamber

adjacent to the NAA Pneumatic Sample System terminus. The CFUF43 is a

hardened stainless steel argon filled ion chamber with the inner cylinder doped

with >90% 235U enriched Uranium. The casing is approximately 4.7x83mm with

an integral mineral coaxial cable. The designed measurement range is 1010–

1014 n/cm2-s and it has a maximum temperature rating of 350°C.

The CFUF43 will be installed utilizing a 5/8’’ diameter access hole within the

upper grid plate of the NETL TRIGA. This access hole is intended to be utilized

for activation foil and wire irradiations in order to determine neutron flux within

the core. In this location, the CFUF43 will be within approximately 3 inches of

the NAA Pneumatic Sample System horizontally and an aluminum casing has

been designed to position the CFUF43’s sensitive region, semi-permanently, at

the same vertical position of the NAA Pneumatic Sample System terminus

within the core.

The CFUF43 is powered by a Keithley 6487 Picoammeter/Voltage Source and

the signal is captured by a National Instruments LabVIEW designed data

logging program. This setup enables this system to be used for future sample

delivery and retrieval automation, if deemed necessary. By installing the

CFUF43, the goal is to be able to better monitor sample irradiance in real-time

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1:30pm-1:50pm

and lower the NAA sample variance for short duration samples. By

accomplishing this, the accuracy of experimentation can be drastically

increased allowing more credibility to data received.

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1:50pm-2:10pm

INSTALLING A LOG CHANNEL AT THE REED RESEARCH REACTOR

Melinda Krahenbuhl

Reed College

3203 SE Woodstock Blvd

Portland Oregon 97202

Robert Barnes

Thermo Fisher Scientific

10010 Mesa Rim Road

San Diego CA 92121

In 2015, Reed College obtained Department of Energy funding and ordered a

new log channel for extending research capabilities. Specifically, of interest

was an upgrade to display source range level in counts per second to enhance

the training of subcritical multiplication. This new channel was not originally

intended for regulatory compliance, nor to replace the existing linear and log

channels in use at Reed College. In 2016, Reed College experienced a

malfunction of an auto water fill system for the reactor tank. The result of the

malfunction was the failure of both the linear channel and the log channel

installed at Reed College. The presentation will address the challenges,

lessons learned while replacing the linear and log channels and the successful

return to normal operations.

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2:10pm-2:30pm

NBSR DEUTERIUM COLD SOURCE: COMPRESSOR MOTOR STARTUP

Andrew W. Main

NIST Center for Neutron Research (NCNR)

100 Bureau Drive, Stop 6101

Gaithersburg, MD 20899

The National Bureau of Standards Reactor (NBSR), located at the NIST Center

for Neutron Research (NCNR) in Gaithersburg, MD, is currently equipped with

a liquid hydrogen cold neutron source. NCNR is in the process of replacing the

existing cold neutron source with a new liquid deuterium cold neutron source.

One of the first phases of this project includes the installation of a new 7kW

refrigerator system complete with two 100% redundant, 800 HP, 4160V Screw

Compressors. Installation of the new refrigerator system was completed in

early 2016. However, initial attempts to start the large refrigerator compressor

motors were unsuccessful due to significant voltage drops generated on the

electrical distribution system during motor startup. These large voltage drops

were largely due to high motor inrush currents on an atypical electrical

distribution system design. Several electrical contractors were consulted for

suggestions to resolve the issue. Their suggestions ranged from adding

additional sets of cable to a complete redesign of the electrical distribution

system. Some of the suggestions fell short of resolving the issue while others

were too expensive and/or would cause significant delays to the overall project

schedule. Therefore, NCNR decided to perform an in-house electrical analysis

of the overall distribution system as well as on each individual component (i.e.,

buses, transformers, soft starters, cables) within the distribution system to

determine the best way to move forward. The results of the analysis identified

several options that were less expensive and allowed the overall project to

remain on schedule.

I would like to present to the TRTR community an overview of the compressor

motor startup issues encountered, the in-house electrical analysis performed on

the system, the modification options considered and implemented, the motor

startup tests performed, and the lessons learned from the project.

24


2:30pm-2:50pm

IMPLEMENTING AGING REACTOR MANAGEMENT:

IF IT WERE EASY EVERYONE WOULD DO IT

Marcus D. Schwaderer, PPM, MBA


100 Bureau Drive, MS 6100

Gaithersburg, MD 20899-6100

Aging-reactor management (ARM) is the application of engineering, operation,

and maintenance strategies to control, within acceptable safety, reliability limits,

the age degradation of structures, systems, and components (SSC) of nuclear

reactors. Aging management is a proactive program that will detect and

evaluate the degradation of components and systems due to aging effects,

further utilized to repair, refurbish, and replace SSCs to improve the longevity of

the facility. The ARM program would benefit the NCNR across safety, health,

environmental, security, operational and economic objectives of the

organization in the operation of the NCNR test reactor. The ARM program is

not about how old the equipment is; it is about its current condition, how it has

changed over time and attempting to trend how it will continue to change over

time. Condition monitoring will allow NCNR management to make fully

informed decisions on capital improvements and refurbishments that will be

required as time goes on. This will ensure an effective and efficient use of

funds in the maintenance of the reactor and its subsystems, thereby allowing

the NCNR to continue to operate safely and reliably.

The NCNR is a 50-year-old facility that needs a dedicated and well-funded

program to sustain and elongate its useful lifespan. The development of the

ARM program will result in greater reliability and lifespan with regards to NCNR

resources and processes. The ARM program is also aligned with NCNR

strategy and objectives since it is focused on safety and reliability and uses

technology to improve the way the facility operates.

As with any well-meaning program, simply stating the need for such a program

is barely scratching the surface on the work that needs to be done. This

presentation will focus on the actual work that needs to be carried out by

management and technical persons alike. Challenges and obstacles abound

when managers attempt to launch sweeping programs that affect all facets of

the operation of a nuclear facility. Challenges include the procurement of

funding, the creation of an organizational vision for both the long- term and the

near-term futures, and the development of buy-in from all persons affected.

Finally, the execution of the ARM program actions that constitute aging

management is considered. An aging management program manager must

25


2:30pm-2:50pm

demonstrate proficiency in each of these following areas: Leading change,

leading people, results driven, business acumen, and coalition building.

The presentation will touch upon each of these areas and will give examples of

personal experiences in either success or more importantly, failures. It is the

goal of the talk to provide “real-life” experiences for others to use and learn

from. A roadmap of how NCNR intends to learn, construct and move forward

on initiatives in the future will be presented to facilitate discussion within the

Test Research and Training Reactors further.

26


3:20pm-3:40pm

NUCLEAR CAMERAS FOR VISUAL INSPECTIONS INCREASE

OPERATIONAL SAFETY

Aaron Huber

Diakont

3853 Calle Fortunada

San Diego, CA 92123

Nuclear cameras are essential for maintaining reactor safety through early

defect detection. The safe operation of test and research reactors requires

reliable detection of defects on internal components. Defects or component

failures can lead to serious issues such as; reactor downtime, impacted safety,

increased costs, increased regulatory scrutiny, and increased stakeholder

concern.

Fuel inspection:

Fuel serial numbers must be inspected to verify proper positioning. Fuel

cladding must also be inspected for defects. Overlooked defects could cause

containment leakage into primary reactor coolant systems which could lead to

increased radiation dose exposure to operators and maintenance workers.

Periodic Inspection Programs:

Periodic inspection programs are imperative to maintain safe operating

conditions of test and research reactors. Ultrasonic and Eddy Current

inspection methods do not reveal surface defects on reactor internals. Visual

inspection technologies are considered the best method for this inspection.

This presentation will provide details on visual inspection programs along with

real-world examples of defects uncovered with VT-1 inspections.

27


3:40pm-4:00pm

EVALUATION OF GASEOUS EFFLUENT MONITOR MAJOR SCRAM

SETPOINT

Timothy J, Barvitskie, Health Physicist

Daniel Mattes, Mechanical Engineer


100 Bureau Drive, Bldg. 235 Rm. B106


The National Bureau of Standards Reactor (NBSR) normally operates with

three installed gaseous effluent monitor channels: Normal Air, Irradiated Air,

and Stack. If the count rate on any one of these monitors exceeds a preset

alarm level, a “major scram” is initiated, closing the normal exhaust pathway

and triggering the start of the emergency ventilation system. The reactor

Technical Specifications require that two out of the three gaseous effluent

channels be operational in order to operate the reactor. The Normal Air monitor

has recently failed and due to obsolescence and lack of replacement parts is

not able to be repaired and returned to service. The NCNR is focusing efforts

on installing a new monitoring system having modern electronics and detector

systems to replace the older, failed Normal Air channel. In order to

successfully integrate this new monitor into the current reactor effluent

monitoring system an alarm set point will need be established for the new

system. Therefore we have evaluated potential alarm set points for agreement

with reactor Technical Specifications and 10CFR 20. The response of the

monitor was evaluated based on detector functional parameters, plant

operating conditions, estimated atmospheric dilution factors, and dose limits.

This presentation will outline the process used in this evaluation.

28


4:00pm-4:20pm

ANALYSIS OF LOSS OF ELECTRICAL FEED TO THE VITAL NETWORK OF

THE BR2 REACTOR

S. Van Dyck, G. Van den Branden, S. Declercq

Belgian Nuclear Research Centre SCK●CEN

200, Boeretang, B2400 Mol, Belgium

During the third operation cycle of the BR2 reactor at SCK●CEN in Mol,

Belgium, the electrical feed of the vital network was lost due to a combination of

disturbance on the external electrical feed due to lightning and the failure to

start up the diesel generators to provide alternative power to the vital network.

The presentation gives the analysis of the impact of the event on the safety and

operation of the installation, as well as the identification of the root cause of the

failure to start the diesel generators. As a consequence of the analysis, an

action plan in order to improve both the hardware of the facility as well as the

testing and maintenance procedures for the electrical feed system. Some

generic lessons can be taken from this event for other installations, operating

diesel generators or other UPS systems.

29


4:20pm-4:40pm

TECHNICAL ANALYSIS AND ADMINISTRITIVE ISSUES OF CRITICALITY

STUDY FOR DIFFERENT MITR FACILITIES

Kaichao Sun

Massachusetts Institute of Technology

Nuclear Reactor Laboratory

138 Albany Street

Cambridge, MA 02139

During recent years, U.S. Nuclear Regulation Commission (NRC) enhances the

criticality safety regulations, emphasis being placed on the validation

requirements for the corresponding neutronic calculations. In the past year

only, there are three criticality studies being required to the Criticality Officer of

the MIT Research Reactor (MITR) for analyzing facilities with fissionable

material involved: 1) Wet Storage Systems, 2) Special Nuclear Material Vault,

and 3) Exponential Graphite Pile. Most existing criticality reports (if there is any)

for the above mentioned facilities are out dated and lack of sufficient technical

details. There are needs to perform up-to-date calculations for the license

renewal (and/or accommodate the new regulation requirements). State-of-the-

art computational tools and the newly available cross-section libraries will be

used. In addition, there is a clear trend that NRC pushed to implement

neutronic validations for the calculation results, where newer versions of

ANSI/ANS Standards (Series 8) is particularly requested to be followed.

In this presentation, the technical results of the criticality analysis for the three

MITR facilities will be briefly presented. By considering double contingency, all

the cases satisfy the MITR technical specifications (i.e., keff shall be less than

0.90) with significant safety margins. More importantly, additional discussions

will focus on the required validation efforts, since these typically even require

more work than the criticality analysis itself. Last but not least, certain

corresponding administrative issues may worth discussion within the TRTR

community, such as communications for the criticality analysis, generic

validation efforts, financial budget management, and etc.

30


9:00am-9:20am

INVESTIGATION AND CORRECTION OF THE HIGH THERMAL TO N-16

RATIO IN THE SOUTHWEST LOBE IN THE ATR

Marya K. Morrison, Reactor Engineering

Robert L. Fulks, System Engineering Manager

Gerald V. Mullen, System Engineering

Daren R. Norman, Reactor Engineering

Darrin G. Robinson, System Engineering

Idaho National Laboratories

Advanced Test Reactor

1955 North Fremont Avenue


The Advanced Test Reactor (ATR) located at the U.S. Department of Energy’s

(DOE) Idaho National Laboratory (INL) is designated by DOE as a National

Scientific User Facility (NSUF) and has been in operation continuously since

1967.

The core power of the ATR is analyzed by five lobes or four quadrants. The

absolute and relative power in each of the five lobes must be known both for

safety considerations and to provide data for evaluation of experiment

performance. During reactor power operations, a portion of the O-16 that

makes up water is activated to N-16. The decay of N-16 results in emission of

beta particles that is proportional to the fast neutron flux to which the water was

exposed. Measurement of the N-16 beta decay from 10 different locations in

the core supports calculations of the lobe and quadrant power distributions

within the core.

The signals from the N-16 system are used in conjunction with Reactor Data

Acquisition System (RDAS) computer programs to calculate lobe and quadrant

power distributions. The Lobe Power Calculation and Indication System

(LPCIS) provides the only method by which lobe powers can be calculated and

it also is calibrated to provide quadrant power distributions. A quadrant power

distribution available from the Water Power Calculator permits comparison of

the two independent distributions to provide a check on both the N-16 system

operation and the Water Power Calculator operation.

Over the course of operation cycles 154A-1 to 157D-1 the thermal power to N-

16 quadrant power ratio in the Southwest was observed to be trending high in

comparison to the other quadrant power ratios. The problem was investigated

from the operability of the system hardware to the condition of hardware

materials. During the outage prior to operation cycle 158A-1 the SW N-16

sample tubing was replaced to resolve the issue.

31


9:20am-9:40am

NBSR SECONDARY COOLANT FLOW CONTROL SYSTEM

Dağıstan Şahin, Marcus Schwaderer


100 Bureau Drive, MS 6100


Secondary coolant system removes heat from the main coolant system, and

other auxiliary reactor systems, such as the Thermal Shield Cooling System,

Purification System and Thermal Column Tank Cooling System. Secondary

coolant system heat load is dissipated to the atmosphere by cooling towers.

Secondary coolant system flow is supplied by four identical pumps arranged in

parallel. In normal operation, only three of the pumps are used keeping one

pump for backup. These pumps draw cold water from the basin of the cooling

towers and heated water from the bypass piping around the cooling water.

Water from the pumps passes through discharge strainers. The pumps

start/stop conditions are controlled by manual switches located in the reactor

control room. The main secondary pump VFD’s are given a start signal and the

pumps ramp up to their preset minimum operational frequency. That frequency

was operationally set by observing total system flow and increasing the

minimum set point until the desired total system flow was achieved (~11,000

GPM). An issue arises when the pump speeds are fixed (previously 47 Hz) and

any one of the main secondary strainers begins backwashing. When a strainer

backwashes, ~600 GPM is taken from the main flow and brought back to the

inlet of the pumps. This 600 GPM is required to force debris off the strainer

basket, but has the negative impact of affecting total system flow. The reduction

in the flow is not permanent, and lasts about a few minutes. A postponed

thermal transient occurs after each flow cascade, causing fluctuations in the

reactor primary coolant temperatures. To stabilize the secondary system, we

implemented a control system using pump VFD’s with the total system flow as

the controlling parameter. A PID controller was installed in the control room for

the operators to adjust the set-point for the secondary flow, enable/disable auto

control by pressing A/M (Auto/Manual) button, and select a pump to auto-

control. The difference allows the pump to make-up for the flow reductions

when strainers actuate. Continuous PID control was not practical due to cabling

issues and unnecessarily adjustments to the VFD’s. Furthermore, a single PID

controller was not capable of allowing to change/select an auto-control pump.

Therefore, a PLC system reads settings on the PID controller and adjusts pump

VFD frequency to keep the flow around the set-point. The VFD settings are

used ensure automatic control will stay within safe operating frequencies. The

minimum motor speed is set to 47 Hz (previous normal operation value). The

maximum motor speed is set to 56 Hz. We will present our lessons learned and

application experiences for flow control systems using VFDs.

32


9:40am-10:00am

DIGITAL CONTROL AND SAFETY SYSTEM MODERNIZATION FOR THE

PENN STATE TRIGA REACTOR

James A. Turso

Assistant Director

Radiation Science and Engineering Center

Pennsylvania State University


Penn State has historically been a leader in reactor control for the research

reactor community. The current console, installed in 1991, continues in

operation to this day with only minor updates of computer components.

However, as is the case with all operating nuclear plants, obsolescence issues

must eventually be addressed which require replacement of hardware and

software. The present system is a hybrid system - digital control and monitoring

and analog safety system. The anticipated instrumentation and control (I&C)

replacement will replace the existing digital equipment (26+ year-old

technology) with more versatile and supportable state-of-the-art technology.

Additionally, the existing analog safety system will be incrementally phased-out,

replaced with a 1E certified digital safety system used in the commercial

nuclear power industry. PSU proposes to partner with Schneider-Electric

subsidiaries Foxboro Controls and Invensys Automation to implement this first-

of-a-kind safety and control system upgrade to a university research reactor. As

a first step in assessing the capabilities of the new system, Schneider-Electric

donated Foxboro Process Automation I/A System and a TRICON 1E digital

safety system, valued at $276,194.00, to the Penn State RSEC to develop a

laboratory for the staff to become familiarized with the equipment and conduct

preliminary control experiments. Many of the aspects of a full-scale control

system are demonstrated on a small-scale with the laboratory equipment

provided - from specification development and factory acceptance testing, to

preliminary testing on the TRIGA using an Auxiliary Control Rod. This

equipment would ultimately be used as a staging area for assessing future

software and hardware changes on an upgraded console.

The PSU TRIGA control system replacement would avert the possibility and

negative impact of the existing, obsolete system failing prior to replacement,

and provide continued support for all of the operational and research activities

currently performed at the RSEC. The proposed replacement will maintain the

full-functionality of the present digital control system using modern digital

technology. The Foxboro I/A control equipment will be the interface between

the reactor operators, the TRICON and the TRIGA, and will be housed in the

existing control console enclosure to retain the same “feel” as the existing

system (and to minimize costs). Similarly, the graphical user interface will retain

the same form, fit, and function as the existing system to minimize impact to

33


9:40am-10:00am

reactor operations. The flexibility to easily update control code and incorporate

new features, such automated nuclear instrument and control rod worth

calibration, would be cost-effectively enhanced by use of the interactive

software development environment and modular hardware architecture of the

Foxboro I/A and TRICON systems. The code and equipment architecture

developed for the Penn State TRIGA control system upgrade would be “open

source”, in that all technical and regulatory content would be shared among the

TRIGA Reactor User’s Group, and potentially serve as an open-source model

for power reactor control system upgrades. Additionally, the console will be

incorporated into the nuclear engineering curriculum at Penn State,

demonstrating to senior-level and graduate engineering students state-of-the-

art control engineering concepts and implementation. It is anticipated that the

PSU RSEC will provide training opportunities in digital instrumentation and

control to the Nuclear Regulatory Commission, Department of Energy, and

nuclear utilities and service providers using the new control system as the focal

point.

34


10:30am-10:50am

PENN STATE TRIGA REACTOR REACTIVITY COMPUTER

OBSOLESCENCE UPGRADE

James A. Turso

Assistant Director

Penn State University Radiation Science and Engineering Center

122 Breazeale Nuclear Reactor


The existing reactivity computer used to determine control rod worth curves for

the Penn State TRIGA reactor is currently 15 years old and executes on a 15

year-old PC using a version of Matlab that is no longer supported by the

Mathworks. The source code is not available and the data acquisition card is of

the same vintage as the computer. A new reactivity computer is being

developed by recoding in a commercially-popular, user-friendly software

package - LabView software (National Instruments) - and is being deployed in

National Instruments hardware. The motivation for using National Instruments

hardware and software are 1) Large domestic customer base – solutions

shared among customers 2) In-house knowledge – Penn State Reactor staff

can upgrade/alter as necessary 3) Software and hardware relatively

inexpensive for academic end-users 4) LabView provides for Data

Acquisition/reactivity calculation/curve fitting/data archiving all in one program

5) LabView enables stand-alone deployment, where an executable may be

produced and run on PC’s without LabView. The LabView-based reactivity

computer calculates step changes in control rod reactivity based on the prompt-

jump approximation, and integration of 6 delayed neutron group precursor

differential equations. For the control rods that are not being calibrated during

the test (which necessarily need to offset the positive reactivity introduced by

changes in calibrated rod position), the LabView software provides suggested

positions based on previous control rod worth curves. A description of the

software and hardware will be provided along with example results from tests.

The software is designed so that other university research reactors may easily

adapt and use for control rod calibrations.

35


10:50am-11:10am

PUR-1 DIGITAL I&C PROJECT REVIEW AND LESSONS LEARNED

Clive H Townsend, David L Storz, Robert S Bean

Purdue University

400 Central Drive

West Lafayette, Indiana 47906

The Purdue University Reactor Number One (PUR-1), a Materials Test Reactor

built in 1962, has undergone a complete overhaul of the original

Instrumentation and Control (I&C) System. This paper details the project’s

aggressive timeline, implementation strategies, licensing efforts, and lessons

learned. As the nation’s Test, Research, and Teaching Reactor (TRTR) fleet

continues to mature, it will face an increasing need for phased or integral

replacements of safety significant systems including the I&C. The lack of

availability or access to current manufacturing of like-for-like replacements,

allowing a 10 CFR 50.59 replacement, will necessitate the TRTR community

break bonds with current temporary mitigants and consider approaches which

utilize the suite of modern equipment available. The PUR-1 I&C replacement,

which underwent contract negotiations, design, and site staging in slightly more

than one calendar year, is a positive example of the speed at which this work

can be done. Frequent communication with an open and flexible document

development and design strategy ensured the vendor was knowledgeable of

the system while Purdue staff minimized changes to the design principles

behind the integral system. Licensing was focused around early communication

between PUR-1 staff and regulatory personnel. This open and positive dialogue

ensured all parties were aware of expectations in documentation and will lead

to a significantly reduced licensing timeline. Through the course of the project,

missteps were documented to prepare this paper. The major lesson learned is

to balance the creation of a Functional Requirement Specification which meets

the needs for safe operation of the facility in the framework of a historical

perspective but not being chained by historic design decisions. The PUR-1

Replacement I&C was built such that nearly every functionality of the original

system is preserved in the digital replacement. This confinement created more

difficulty than was needed and should have been avoided earlier in the project.

The new PUR-1 digital I&C replacement will prepare the Purdue School of

Nuclear Engineering to serve as a preeminent teaching facility and completely

remove historic operational issues with noise laden equipment. This major

project can serve as a baseline for all TRTRs to begin work in their own I&C

replacement.

36


11:10am-11:30am

MANAGING SYSTEMATIC ERRORS IN THE NBSR THERMAL POWER

DETERMINATION

Dağıstan Şahin, Samuel J. MacDavid, Marcus Schwaderer


100 Bureau Dr., 20899 Gaithersburg, MD, USA

There has been a calorimetric misbalance between the primary and secondary

loops of the National Bureau of Standards Reactor (NBSR) since the

installation of a set of new heat exchangers in the early 90’s. The NBSR’s

primary and secondary process instrumentation were investigated to resolve

the underlying causes. The main issue was found to be the immersion length of

the thermowells, resulting in non-exemplary process measurements.

Furthermore, thermal insulation on various temperature sensors was found to

be degraded. Hence, the following were found to be inconsistent: the heat

exchanger output temperatures, previously-used differential temperature

sensors, reactor inlet sensor, and the reactor outlet temperature indications.

Therefore, thermodynamic analyses using these sensor measurements were

inconclusive. Secondly, the discrepancy between primary and secondary loops

and the gradual inconsistency between primary side sensors went largely

unnoticed. We implemented sustainable, state-of-the-art upgrades to resolve

systematic errors in the NBSR reactor thermal process instrumentation. Several

digital upgrades were completed, along with detailed 50.59 reviews.

Redundancy, defense-in-depth, reliability, diversity and accuracy of the thermal

monitoring system was established by implementing an inclusive engineering

approach by analyzing the sensors as a whole system instead of individual

assessments. The upgrades produced two important outcomes. First, the long-

term existing calorimetric discrepancy between the primary and secondary

loops was resolved. Secondly, excellent agreement was achieved within the

primary process measurement instrumentation, resulting in reliable and stable

thermal power assessment. We will present lessons learned performing

instrumentation upgrades and our future enhancement plans in process

monitoring for the NBSR reactor.

37


11:30am-11:50am

WIDE RANGE CHANNEL AND MICROCONTROLLER BASED SIGNAL

CONVERTER FOR AN EXISTING RESEARCH, TRAINING AND ISOTOPE

PRODUCTION NUCLEAR FACILITY

Benjamin Schlottke 1), Franz-Josef Terheiden 2)

1) Project Manager, 2) Software Development

Mirion Technology (MGPI H&B) GmbH

Landsberger Str. 328a, 80687 Munich, Germany

+49-(0)89-51513-124

For an existing research reactor, with the purpose to also serve as an isotope

production facility, Mirion was awarded the contract to provide the wide range

channel and microcontroller based signal converter.

The research reactor is a 2 MW open pool, light water moderated and cooled

reactor of the TRIGA conversion type.

The research reactor wants to replace the existing NM-1000 but at the same

time keep the existing Reuter Stokes fission chamber and the cable to the

detector and also the installed wire to the control room from the existing NM-

1000.

Mirion’s solution is to replace the existing NM-1000 with Mirion’s wide range

channel proTK™ DWK250. Due to the customers special I&C system, Mirion

works together with his partner at Plantation Production, Inc. to integrate a

microcontroller based signal converter. This signal converter will compile the

signals from Mirion pro TK™ DWK 250 channel to the customers I&C system.

This solution covers also all necessary Cyber Security aspects for the proTK™

equipment. Mirion will provide information about the implemented Cyber

Security features.

38


1:30pm-1:50pm

IRRADIATION TOOLS FOR NSUF MATERIALS RESEARCH

Brenden J. Heidrich

Nuclear Science User Facilities, Idaho National Laboratory

Center for Advanced Energy Studies,

995 University Blvd., Idaho Falls, ID 83402

and

Jieun Lee, Emma Redfoot, and Kelly Verner,- University of Idaho, and

Nicholas Herring – North Carolina State University

An irradiation toolkit was developed and deployed by the Nuclear Science User

Facilities (NSUF) to assist researchers and technical staff to guide neutron

irradiation experiments. NSUF is the Department of Energy Office of Nuclear

Energy’s first and only user facility. It offers unparalleled research opportunities

for nuclear energy researchers by providing access (at no cost to the

researcher) to world-class nuclear research facilities along with technical

expertise from experienced scientists and engineers for assistance with

experiment design, assembly, safety analysis and examination. NSUF is not a

typical user facility in that it is not a single self-contained facility but represents

a consortium of capabilities distributed across the U.S. at a number of

institutions. NSUF provides access through a peer-reviewed competitive

process to the Idaho National Laboratory (INL) and fourteen partner facilities in

the United States and one international affiliate.

Founded in 2007 as the Advanced Test Reactor National Scientific User

Facility, the NSUF has ten reactors available to provide neutron irradiations: 1)

Advanced Test Reactor at INL, 2) High Flux Isotope Reactor at the Oak Ridge

National Laboratory, 3) the MITR-II reactor at the Massachusetts Institute of

Technology, 4) PULSTAR reactor at the North Carolina State University, and 5)

Belgian Reactor-2 at SCK-CEN in Belgium. In addition, there is limited access

to the Neutron Radiography Reactor and ATR-C reactors at INL and the

Annular Core Research Reactor and the SPR-CX critical facility at Sandia

National Laboratory. The TREAT transient testing reactor will be part of NSUF

when it is recommissioned in 2018.

With all of these options available to nuclear materials researchers, there can

be considerable confusion during the proposal process. Researchers as well

as many of the NSUF technical leads (assigned to guide researchers through

the proposal, experiment design and execution processes) are not experts in

irradiation engineering. A toolkit was needed to make the selection of a

suitable reactor and irradiation position as easy as possible. The author, along

with a team of interns and graduate students has designed and built a web-

based set of tools to guide researchers and technical leads. The toolkit will

39


1:30pm-1:50pm

allow the researchers and NSUF to appropriately allocate scarce resources to

their best use. The toolkit has three parts:

1. Convert from units of displacements per atom (dpa) to fluence (based on

MCNP models of the various reactors).

2. Select from approximately 100 possible irradiation positions (by providing

estimated irradiation times in each position, based on the required fluence)

3. Estimate the radioactivity of the specimens following irradiation.

40


1:50pm-2:10pm

DELAYED NEUTRON SPECTROSCOPY FOR CHARACTERIZATION OF

SPECIAL NUCLEAR MATERIAL

A. J. Shaka

Chemistry Department

University of California

Irvine CA 92692-2025

Jose Ocampo, Jonathan Wallick, George Miller

Chemistry Department

University of California

Irvine CA 92692-2025

Fission products are neutron rich and typically undergo β--decay. In a few

cases, the daughter is produced in an excited nuclear state and itself emits a

neutron. These β--delayed neutrons provide a highly selective and quantitative

means to interrogate fissile material. Several groups have proposed numerical

methods to try to ascertain the composition of an unknown that may contain,

e.g. 239Pu and/or 235U by exploiting the small differences in the populations of

the delayed neutron precursors and then analyzing the tally of delayed

neutrons versus the time after irradiation. This time series approach can work

well, but requires a reactor with high neutron flux. For example, some work

was carried out at the HFIR facility at Oak Ridge National Laboratory.

Simulations of the expected performance of the time series approach suggest

that millions of delayed neutrons need to be detected to achieve good

performance.

Using our small 250 kW TRIGA, with thermal flux of the order of 1012 cm–2s–1,

we have taken a slightly different approach, optimizing the figure of merit for the

technique by maximizing detectable neutrons per unit experiment time. By

altering the irradiation conditions by means of various neutron absorbing

materials, it is possible to achieve decent discrimination between 235U and239Pu or to determine the approximate percent composition of a mixed sample.

We will show that even with relatively low-flux irradiation conditions it is

possible to get quantitative information on samples containing microgram

quantities of fissile material. Some interesting side effects of control rod

positions (with the same average power) on the performance of the experiment

highlight the importance of knowing all the core parameters. It appears to be

more important to lower the reactor power to improve temporal stability and the

exact neutron flux as a function of energy at the terminus, rather than

increasing the power to maximize the flux. As a quick and reliable analytical

method, delayed neutron spectroscopy has a lot of advantages compared with

alternative radiochemical techniques.

41


2:10pm-2:30pm

ACCESS TO IRRADIATION CAPACITY IN THE BR2 REACTOR

Steven Van Dyck, Sven Van den Berghe

Belgian Nuclear Research Centre SCK●CEN

200, Boeretang, B2400 Mol, Belgium

The BR2 reactor, operated by the Belgian nuclear research centre SCK●CEN

in Mol, Belgium, is a high flux material test reactor. The BR2 reactor offers a

unique combination of high performance (access to thermal and fast neutron

flux up to 1015n/cm²s and 6 1014 (E>0.1MeV)n/cm²s respectively) and flexibility

(the core configuration is adaptable on a cycle by cycle base to accommodate

different irradiations). For rapid turnaround experiments, a number of

standardised irradiation devices are available for material and fuel irradiation.

For material irradiations in support of LWR component ageing studies, a new

device with precise temperature control on the samples before, during and after

irradiation is installed. The device allows to irradiate sufficient volume material

in order to complement existing reactor vessel surveillance databases. For

qualification of materials for advanced nuclear systems at high temperature (up

to 1000°C) and high fast flux is available with active temperature control of the

sample irradiation temperature. For fundamental irradiation damage studies,

low temperature irradiation capsules with flexible irradiation time and passive or

active temperature control can be used.

For fuel irradiation, capsules with pressurised water for LWR applications as

well as baskets for MTR fuel plate irradiation are routinely available. The BR2

reactor also offers to possibility to install integrated test loops and large

experiments (up to 200mm diameter). This presentation reviews the

characteristics of the different standard irradiation tools as well as the practical

aspects for preparation of irradiation experiments, which are part of a short

course available at SCK●CEN.

42


2:30pm-2:50pm

STATUS UPDATE ON MITR NUCLEAR SAFETY SYSTEM UPGRADE

Edward S. Lau, Shawn W. Hanvy, Dane E. Kouttron, and Paul T. Menadier

MIT Nuclear Reactor Laboratory

138 Albany St., Room NW12-122

Cambridge, MA 02139

The MIT Research Reactor is replacing its 1975 nuclear safety system (NSS)

with a new one that utilizes digital neutron flux monitoring along with analog

scram logic and signal processing systems. MIT will share its experience in the

technical design, development, and testing of the new NSS, along with

interaction with U.S. NRC on its License Amendment Request. The new

system has been built, and is under operational testing in parallel with the

existing system, except that the new system does not transmit any scram

signals to the existing scram circuit.

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2:50pm-3:10pm

TRIGA FUEL CLADDING CHEMICAL INTERACTIONS

Eric Woolstenhulme, Dennis Keiser



In 1986, the Department of Energy's Reduced Enrichment for Research and

Test Reactors (RERTR) Program, in conjunction with General Atomics,

submitted test and evaluation reports to the U.S. Nuclear Regulatory

Commission (NRC) supporting the qualification of new low enriched uranium

(LEU) 30/20 (30% LEU uranium in the fuel matrix) TRIGA elements for use in

converting TRIGA reactors from high enriched uranium (HEU) to LEU fuel.

A subsequent report in the January 2006 Nuclear Technology Journal, “Fission

Gas Release From Uranium-Zirconium Hydride Fuel During Heating to Melting”

presented results from an experiment that was preformed to evaluate the

fractional gaseous fission product release from uranium-zirconium hydride fuel

as a function of temperature. The article suggests that during the experiment it

was observed that the manner in which the fuel section melted could have

indicated a fuel/cladding chemical interaction (FCCI), or eutectic melting.

An attempt was made to resolve the issue by researching previous FCCI

experiments. The results of the investigation lead researchers to believe that

long standing safety limits for TRIGA fuel provided a reasonable assurance that

the reactors were safe. However, no prototypical studies on the TRIGA FCCI

behavior had been done. Therefore, to determine how and where melting might

occur for solid solutions or compounds at the fuel/cladding interface in TRIGA

fuel elements, annealing experiments have been performed using as-fabricated

and irradiated fuel. This is a presentation on the results of those studies.

44

Notes

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Notes

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Notes

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Notes

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Sponsors / Exhibitors

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PLATINUM SPONSORS:

The National Organization of Test, Research, and Training Reactors

GOLD SPONSORS:

TRIGAINTERNATIONAL

Sponsors / Exhibitors

50

SILVER SPONSORS:

EXHIBITORS:

Documents

Reactors Annual Conference September 17th 21st, …muconf.missouri.edu/trtr2017/TRTR2017Program.pdf · Reactors Annual Conference September 17th ... USS Midway (CV-41) ... Amgad Shokr,