ML21145A252

Advanced Reactors GEIS Docs - INL - a Proposed Path Forward for Transportation of HALEU

ML21145A252
Person / Time
Issue date:	05/25/2021
From:	NRC
To:	NRC/NMSS/DREFS
References
Download: ML21145A252 (176)

Text

From:

Giacinto, Joseph Sent:

Tuesday, May 25, 2021 2:05 PM To:

AdvancedReactors-GEISDocsPEm Resource

Subject:

INL - A Proposed Path Forward for Transportation of HALEU Attachments:

Jarrell 2018.pdf

 

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Meeting Summary

SUBJECT:

INL-NEI Invitation-Only Technical Workshop on Transportation of High Assay Low-Enriched Uranium ORGANIZER: INL and NEI AUTHOR: Gordon Petersen (INL)

DATE: August 30th and August 31st PURPOSE: The primary objective of this workshop will be to advise DOE on the gaps related to transportation of HALEU and licensing support activities. The goal is to ensure that transportation and handling of HALEU at associated fuel cycle facilities does not delay the ability of advanced reactors to be deployed.

OVERVIEW: The meeting started with lunch provided by NEI. Everett Redmond from NEI then began the meeting by announcing safety procedures and letting all the attendees introduce themselves. He then went over the mission statement of the NEI Fuels Task Force and the letter sent to Secretary Perry by NEI specifiying the amount of HALEU needed over the next ten years. Josh Jarrell from INL took over and introduced the goals of the meeting and reiterated some of the questions Everett proposed. Over the next day, presentations were given by industry, national laboratories, and the NRC. Each presentation concluded with time to ask questions and have discussions. The first day concluded with a discussion in preparation for the NRC visit led by Nima Ashkeboussi. The second day was led off with a presentation from the NRC followed by discussion. Next the labs and industry continued presenting topics related to the capabilities and needs related to HALEU management. The second day concluded with a DOE perspective given by John Herczeg, industry/NEI recommendations for DOE led by Nima, and a wrap up of action items led by Josh. The following notes provide a short overview of the presentations given.

Industry provided information from an enrichment, licensing, and transportation perspective:

1.

Capabilities exist for enrichment up to 20% (Melissa Mann/URENCO) a.

Imperative to develop fuel cycle with consortium (fabricators, convertors, enrichers, reactor operators, transporters, etc.) approach for licensing framework b.

Questions remain concerning transforming Cat III facility into Cat II facility and transportation off site c.

Suggests engaging NRC and ANSI/ASTM standards now 2.

Experience in licensing facilities with enrichments greater than 5.0 wt.% U235 and have transportation packages that can be amended for HALEU (Lon Paulson/GNF) a.

GNFA Wilmington fuel fabrication facility b.

Model RAJ-II Type B fissile package will require SAR update to transport HALEU c.

Model NPC Type A fissile package will require SAR update to transport HALEU d.

Licensing a new package takes 42 weeks minimum for NRC review, but start to finish takes ~5 years 3.

Packages for shipping 20% enriched materials (Andy Langston/DAHER-TLI) a.

Majority of DOE 20% enriched fuel shipped in drum type packages (Versa-Pac) b.

Currently Versa-Pac is under NRC amendment application for 1S/2S cylinder c.

30B cylinder design up to 20% UF6 enrichment currently under development i.

1600 kg ii.

30B-20 can be operated and handled in same way as 30B cylinder

iii.

Licensing overpack and cylinder with French, German, and NRC.

d.

Package for 5B/A cylinders under development i.

VP-55XL is an enhanced version of the TLIs NRC approved VP-55 4.

Licensing transport overpacks and packages with NRC (Rick Migliore/TN Americas) a.

Little concern in ability to license/certify package b.

Industry is not in position to create criticalitybenchmarks c.

More concerned with licensing and packaging on the SNF side after the fuel is removed from the reactor The labs presented on the following capabilities:

1.

Nuclear Data and Benchmarking Program (Brad Rearden/ORNL) a.

High uncertainties in cross sections with-in intermediate and high energy ranges b.

Cross cutting program can support the needs of advance reactors i.

Use correlation coefficients in trending analyses to determine cross section sensitivities ii.

Perform gap analyses for non LWRs c.

Mine existing experiments to determine similarities 2.

INL could bridge material gap for 10 years (Monica Regalbuto/INL) a.

Naval reactor fuel, EBR-II, and ZPPR plates can be available for downblending b.

Issues may exist with uncertainties and dose of U-234 3.

Nuclear Criticality Safety Program (Doug Bowen/ORNL) a.

National Criticality Experiments Research Center (NCERC) best for 20% enrichment experiments b.

Experiments are expensive and time consuming to setup and perform i.

Cost $425k-$2.1M ii.

Time frame24-54 months 4.

Validation discussion (John Scaglione/ORNL) a.

Some techniques do not need experiments but can instead use physics-based solution b.

Criticality validation process for ES-4100 package i.

Requires detailed knowledge of the application system ii.

Used similarity assessment to find how similar experiments were to target (Ck value) iii.

Over 175 relevant experiments with Ck over 0.9 and just under 700 with Ck over 0.8, when considering HALEU UF6 in the ES-4100 package. Therefore, optimism that experiments exist to defend future package designs for HALEU transport.

The NRCs also gave a short presentation followed by a discussion (Drew Barto/NRC) 1.

Stressed the lack of information from >5% x <19.75% enrichment 2.

Explained difficulty in changing existing regulation, especially regarding moderator exclusion for >5% enriched UF6.

3.

Gave timeline for expected review a.

Complete entire process from day of acceptace of application to certifying in 7.4 months for 80% of transportation reviews and 2 years for all transportation reviews ACTION ITEMS/IMPORTANT TAKE-AWAYS 1.

DOE is committed to transportation of material regardless of form, and NEI will be be the focal point for prioritization of different strategies.

2.

Although the labs can provide additional criticality experiments, industry has enough data to license facilities, overpacks, and cylinders. Validation to find more critical experiments to establish less uncertainty in the benchmarks will be helpful.

3.

A collective effort from industry is needed to express consistency on how much information exists or is needed in regards to criticality.

4.

NEI will change HALEU white paper concerning criticality.

5.

NRC needs to validate methodology is applicable at >5% enriched.

6.

NRC already has group that meets bi-weekly concerning HALEU.

a.

It will be very difficult and time-consuming to change NRC regulations INDUSTRIES REQUEST FOR NEI, DOE, LAB COMPLEXES 1.

DOE and the lab complex should communicate and educate the NRC on criticality issues related to HALEU.

2.

INL should support work needed to certify package design for the transportation of HALEU.

a.

Suggest amending the COC of an existing package used for the shipment of commercial quantities.

b.

Suggest DOE provide funding to package designer(s) for analysis and engineering work for a package to be submitted to NRC for approval.

3.

INL should provide specific, or a range, on the expected impurities that will be present in recycled naval fuel.

4.

In the longer term, DOE and the lab complex should increase the availabilitiy of criticality benchmark data (i.e., by performing, sponsoring, or data mining additional criticality benchmarks) to further reduce conservatism in package design.

ATTACHMENTS x

Part I: Agenda x

Part II: Attendee List x

Part III: Presentations









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Part II

Part III

Everett Redmond, Ph.D.

Nuclear Energy Institute HALEU WORKSHOP

NEI FUELS TASK FORCE Mission: Lead industry efforts in identifying and resolving regulatory and policy issues for the development of the nuclear fuel supply chain for advanced reactors with an emphasis on challenges related to the utilization of high assay low enriched uranium.

Year Total

Cumulative

2018

0.026

0.026

2019

1.506

1.532

2020

2.21

3.7

2021

4.2

7.9

2022

3.7

11.6

2023

18.8

30.4

2024

10.3

40.7

2025

12.4

53.1

2026

57.4

110.5

2027

73.6

184.1

2028

108.1

292.2

2029

111.8

404.0

2030

185.5

589.5

Values in MTU Current fleet uses about 2000 MTU/year Letter to Secretary Perry July 5, 2018 Data from eight companies Not all ARs or advanced fuels need HALEU INDUSTRY NEEDS

QUESTIONS TO CONSIDER Will the fuel cycle process be similar to current fleet?

Mining Conversion Enrich Fab Reactor What differences might exist - material form, etc.?

Should the task force engage publicly with NRC on the issues from this workshop?

What other topics should the task force tackle?

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Presented by:

Bradley T. Rearden, Ph.D.

National Technical Director Nuclear Data and Benchmarking Program Presented to:

Technical Workshop on Transportation of High Assay Low-Enriched Uranium August 30-31, 2018 Nuclear Energy Institute August 30, 2018 Validation and Role of Critical Experiments and Nuclear Data

energy.gov/ne 2

en en en en en en en en ener er er er er er er er ergy gy gy gy gy gy gy gy gy.g

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/ne 2

• New Nuclear Energy Enabling Technology (NEET)

Crosscutting Program

• Partner with industry, NRC, and other programs to:

- Identify priority needs for nuclear data and benchmarking

- Perform new data measurements and evaluations

- Support integral experiments and handbooks

- Participate in application benchmark studies Nuclear Data and Benchmarking Program Office of Nuclear Energy ENDF/B-VII.1 ENDF/B-VII.0 JEFF-3.2

energy.gov/ne 3

en en en en en ener er er er er ergy gy gy gy gy gy.g

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/ne 3

Reactor Type Companies Red = NRC Priority Licensing action expected Fuel /

Enrichment Thermal spectrum Fast Spectrum Coolant Radial core expansion Flowing Fuel Fuel Form Control elements HPR Oklo 2019

~20%

Sodium heat pipes

Metallic Castings External drums Westinghouse (eVinci) 2019 19.75%

Thermal/

Epithermal Sodium heat pipes (dual condenser)

Oxide External drums SFR TerraPower (TWR)

~20%

Sodium

Metallic Rods Internal rods GE PRISM

~20%

Sodium

Metallic Rods Internal rods LFR Westinghouse 15-20%

Lead

Oxide/

Nitride Internal rods HTGR X-energy (Xe-100) 2020s 15.5%

Helium Pebbles TRISO External rods Areva (SC-HTGR)

~20%

Helium TRISO Internal rods FHR Kairos 2020s

~17%

FLiBe Pebbles TRISO External rods Abbreviated advanced reactor technology matrix (1/2)

energy.gov/ne 4

en en en en en en en en ener er er er er er er er ergy gy gy gy gy gy gy gy gy.g

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/ne 4

Reactor Type Companies Red = NRC Priority Licensing action expected Fuel /

Enrichment Thermal spectrum Fast Spectrum Coolant Radial core expansion Flowing Fuel Fuel Form Control elements MSR Terrestrial Energy (IMSR) 2019

~5%

Proprietary Salt Molten Salt Internal rod Transatomic 2020s

~5%

Thermal/

Epithermal FLiBe Salt Molten Salt Internal ZrH moderating rods TerraPower (MCFR) 2020s

~20%

Chloride salt Salt Molten Salt External rods?

Elysium

~20%

Chloride salt Salt Molten Salt FLiBe Energy Thorium

FLiBe Salt Molten Salt Internal rods Abbreviated advanced reactor technology matrix (2/2)

Send updates to Brad - reardenb@ornl.gov

energy.gov/ne 5

en en en en en en en en ener er er er er er er er ergy gy gy gy gy gy gy gy gy.g

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/ne 5

Neutronics calculations rely on nuclear data for criticality, reactivity, power distributions, depletion, decay heat, and more.

Nuclear data is of fundamental importance in nuclear science and engineering

energy.gov/ne 6

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/ne 6

Different reactor designs have different nuclear data needs

energy.gov/ne 7

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/ne 7

0.985 0.990 0.995 1.000 1.005 1.010 1.015 1.020 1.025 1.030 C/E v7.1-56 v7.1-252 v7.1-200 ce_v7.1 Exp. Unc.

MG XS Unc.

HEU-SOL-THERM-014 and -016 Nuclear data lifecycle v7 1 56 v7 1 56 1 5 v7 1-56 v7 1-56 7 1-v7.1-56 v7.1 56 v7 1 25 v7 1 25 v7 1-25 v7 1-25 25 v7.1 25 v7.1 252 v7 1 20 v7 1 20 v7 1-20 v7 1-20 v7.1 20 v7.1 200 Differential Data Measurements Data Evaluation (SAMMY)

Evaluated Nuclear Data Files (ENDF)

Nuclear Data Processing (AMPX)

Nuc Validation and Applications (SCALE)

. Unc.

MG 7.1 ce_v7 ce ce ce c

ation and Nuclear Data Needs D

energy.gov/ne 8

en en en en en en en en ener er er er er er er er ergy gy gy gy gy gy gy gy gy.g

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/ne 8

Cross section components:

Typically generated separately, then combined for distribution Thermal Scattering Resolved Resonances Unresolved Resonances Fast Energy Range

energy.gov/ne 9

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/ne 9

• A specific program (DOE-SC, NNSA/NCSP, NNSA/NA-22, DOD, international participant) funds an update in a nuclear data evaluation

- New differential physics experiments

- Data processing

- Comparison to and optimization with applications in their interest

• National Nuclear Data Center - Cross Section Evaluation Working Group (CSEWG)

- Updates are exchanged through a beta repository for ENDF and reviewed by a global team

- Meets twice annually, with participation from IAEA, OECD/NEA, and others to review proposed updates

- If changes benefit, or do not disrupt, applications of interest to these teams, the new evaluation is approved

• Until now, no official representation for Nuclear Energy applications How are these general purpose libraries generated?

energy.gov/ne 10 gy gy gy gy gyy gov ov ov ov ov ov/n

/n

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/n

/ne Compensating Errors in the Jezebel keff Eric Bauge* reported on an analysis where components of the Bruyres-le-Châtel (BRC) 239Pu evaluation were replaced with those from ENDF/B-VII.1. At each step in the replacement process, keff of the Jezebel critical assembly was computed. While both the BRC and ENDF/B-VII.1 give the same keff for Jezebel, they do so for very different reasons. This replacement study shows how different parts of the evaluation substantially shift the reactivity of Jezebel. We do not know if either evaluation is correct but both get the correct answer.

en en en en en ener er er er er ergy gy gy gy gy gy.g

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.go 10 We do not know if either evaluation is correct but both get the correct answer.

• E. Bauge et al., Eur. Phys. J. A (2012) 48: 113

-16 p.c.m.

+275 p.c.m.

-638 p.c.m.

+522 p.c.m.

-14 p.c.m.

-122 p.c.m.

energy.gov/ne 11 en en en en en en en en ener er er er er er er er ergy gy gy gy gy gy gy gy gy.g

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/ne 11 Generation of Cu evaluation for ENDF/B-VIII.0 ENDF/B-VII.1 Proposed ENDF/B-VIII.0 Final ENDF/B-VIII.0 F/DOF

Measured Proposed Measured Proposed Measured Final Measured Final Angular Distribution Proposed Final V. Sobes - ORNL

energy.gov/ne 12 en en en en en en en en ener er er er er er er er ergy gy gy gy gy gy gy gy gy.g

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/ne 12 Decay data

- ENDF/B-VII.1

- Natural isotopic abundances (NIST database)

- ICRP 72 inhalation dose coefficients, EPA Report 12 on external exposure Neutron reaction cross section data

- JEFF 3.1/A special purpose activation file

- ENDF/B-VII.0, -VII.1 Fission product yields: ENDF/B-VII.0 Photon emission line-energy data

- Evaluated Nuclear Structure Data Files (ENSDF)

- ENDF/B-VII.1 Neutron emission libraries

- SOURCES 4C code

- Spontaneous fission decay and delayed neutron data

- Alpha stopping powers, (,n) cross sections, excitation levels Nuclear data for activation, depletion, and decay

energy.gov/ne 13 en en en en en en en en ener er er er er er er er ergy gy gy gy gy gy gy gy gy.g

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/ne 13 238U inelastic scattering cross section uncertainty differences between international libraries ENDF/B-VII.1 Europe Japan ENDF/B-VIII.0

energy.gov/ne 14 en en en en en en en en ener er er er er er er er ergy gy gy gy gy gy gy gy gy.g

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/ne 14 OECD Nuclear Energy Agency Uncertainty Analysis in Modeling sodium fast reactor study with ENDF/B-VII.1 uncertainties CE TSUNAMI: nominal values and uncertainties MET1000 MOX3600 nominal uncertainty nominal uncertainty Eigenvalue 1.0841(1) 1.49(1)%

1.0771(1) 1.52(1)%

CR worth 12081(11) pcm 2.81(1)%

4973(11) pcm 2.67(1)%

CE TSUNAMI: Top 3 contributors MET1000 MOX3600 Eigenvalue CR worth Eigenvalue CR worth U-238 inel.

U-238 inel.

U-238 inel.

U-238 inel.

Fe-56 inel.

Fe-56 inel.

U-238 cap.

Na-23 el.

Na-23 el.

Na-23 el.

Pu-239 cap.

U-238 chi MET1000 MOX3600

energy.gov/ne 15 en en en en en en en en ener er er er er er er er ergy gy gy gy gy gy gy gy gy.g

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/ne 15 Recent nuclear data developments of interest to the advanced reactor community

energy.gov/ne 16 en en en en en en en en ener er er er er er er er ergy gy gy gy gy gy gy gy gy.g

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/ne 16 Changes in graphite data ENDF/B-VII.0 (2006) to ENDF/B-VII.1 (2011)

Capture cross section increased from 3.36 mb to 3.86 mb: ~1,000 pcm ENDF/B-VIII.0 (2018)

New evaluations for thermal scatter based on molecular dynamics models from North Carolina State Includes data for crystalline and reactor-processed graphite HTTR loading ENDF-VII.0 C/E ENDF-VII.1 C/E Initial criticality 1.0165

1.0011

Full core 1.0097

1.0015

HTR-10 Configuration ENDF-VII.1 C/E ENDF-VIII.0 C/E First core 1.00267

1.00582

A. Hawari NC State HTR-10 Benchmark

energy.gov/ne 17 en en en en en en en en ener er er er er er er er ergy gy gy gy gy gy gy gy gy.g

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/ne 17 Library Code XS lib k

NSFP

ENDF/B-VII.1 KENO CE 1.6770(4)

(ref)

ENDF/B-VIII.0 KENO CE 1.6722(4)

í

HTR-10 pebble: KENO-VI eigenvalue comparison 1RWH7KHVWDWLVWLFDOXQFHUWDLQWLHVDUHJLYHQLQSDUHQWKHVHV

Differences between ENDF/B-VII.0 and VII.1: carbon capture Differences between ENDF/B-VII.1 and VIII.0: 235U and 238U Basis: ENDF 7.1 NWRDOO(1')SFP

But: graphite from ENDF 8.0

í7 But: 235U from ENDF 8.0

í702 But: 238U from ENDF 8.0 239 All ENDF 8.0

í438 Replace individual nuclides in ENDF/B-VII.1 calculation by ENDF/B-VIII.0 data:

HTR-10 fuel pebble

energy.gov/ne 18 en en en en en ener er er er er ergy gy gy gy gy gy.g

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/ne 18 Changes in 35Cl(n,p) cross section from ENDF/B-VII.0 to VII.1 ENDF/B-VII.1 ENDF/B-VII.0 JEFF-3.2 Simplified Molten Chloride Fast Reactor Data Library keff ENDF/B-VII.0 1.02993 +/- 0.00002 ENDF/B-VII.1 1.04924 +/- 0.00002 Reaction Sensitivity Cl-35 (n,p) Capture Reaction

-0.958 Pu-239 Nu-bar 0.603 U-238 Nu-bar 0.281 Na-23 Elastic Scatter Reaction 0.114 No data for FLiBe / FLiNaK thermal scattering

energy.gov/ne 19 en en en en en en en en ener er er er er er er er ergy gy gy gy gy gy gy gy gy.g

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/ne 19 Validation of methods and nuclear data for advanced applications

energy.gov/ne 20 en en en en en en en en ener er er er er er er er ergy gy gy gy gy gy gy gy gy.g

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/ne 20 Programmatic support for US leadership of the following projects:

- International Criticality Safety Benchmark Evaluation Project (ICSBEP)

- International Reactor Physics Benchmark Evaluation Project (IRPhEP)

Handbooks generated by these projects provide thousands of benchmark experiments from dozens of countries with an assessment of data integrity, quantification of experimental uncertainties, and thorough technical review with established deployment process Strong collaborations have been implemented with the Organisation for Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA)

International benchmark evaluation projects ICSBEP 22 contributing Countries

~69,000 pages

>5,000 approved benchmarks IRPhEP 21 contributing countries 50 reactor facilities 147 approved benchmarks

energy.gov/ne 21 en en en en en ener er er er er ergy gy gy gy gy gy.g

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/ne 21 0.97 0.98 0.99 1.00 1.01 1.02 1.03 1.04 C/E HEU-MET-FAST 0.97 0.98 0.99 1.00 1.01 1.02 1.03 1.04 C/E HEU-SOL-THERM 0.97 0.98 0.99 1.00 1.01 1.02 1.03 1.04 C/E LEU-COMP-THERM Computational Bias Experimental Uncertainty Cross-section Uncertainty 0.97 0.98 0.99 1.00 1.01 1.02 1.03 1.04 C/E MIX-COMP-THERM Computational bias for critical benchmarks

energy.gov/ne 22 en en en en en en en en ener er er er er er er er ergy gy gy gy gy gy gy gy gy.g

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/ne 22 Modern database of measured fuel compositions was expanded as part of a multi-year international collaboration. ORNL has coordinated this effort through the OECD/NEA Expert Group on Assay Data for Spent Fuel since 2009.

International Spent Nuclear Fuel Database SFCOMPO 2.0 provides a central repository of destructive assay data Databases maintained by OECD Nuclear Energy Agency Data Bank include:

í ICSBEP (Criticality safety database)

í IRPhEP (Reactor physics database)

í SFCOMPO (Spent fuel composition and decay heat database)

Data for PWR, BWR, AGR, MAGNOX, CANDU, RBMK, VVER-440, VVER-1000 fuels 44 reactors, 118 assemblies, 91 isotopes important to fuel cycle safety and WM 750 samples > 22,000 measurements Data essential for code validation and uncertainty analysis, integral nuclear data testing -- Energy and Security applications 239Pu data (all reactor types) 222 222 22 222 222 222 222 22 222 http://www.oecd-nea.org/sfcompo/

energy.gov/ne 23 en en en en en en en en ener er er er er er er er ergy gy gy gy gy gy gy gy gy.g

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/ne 23 5% < Hi-assay LEU < 20%

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/ne 24 NRC/NMSS perspectives on high assay fuel en en en en en en e er erg

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/ne 25 Example criticality validation process using the ES-4100 package Photos Courtesy of Jeff Arbital Y-12 National Security Complex Containment vessel

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/ne 26 ES-4100 w/ 20 w/o UF6 study:

Counteracting errors in ENDF/B-VII.1 - ENDF/B-VIII.0

-132 pcm

-238 pcm

-83 pcm

-65 pcm

+95 pcm

+216 pcm

+42 pcm 235U+238U evaluations ENDF-7.1 from ENDF-8.0 ENDF-8.0 1H ENDF-8.0 16O ENDF-8.0 235U ENDF-8.0 238U ENDF-7.1 1H ENDF-7.1 16O ENDF-7.1 235U ENDF-7.1 238U ENDF-7.1: keff = 0.86464 (8)

~450 pcm 235U+2338U l

ti

~450 pcm ENDF-8.0*

energy.gov/ne 27 en en en en en en en en ener er er er er er er er ergy gy gy gy gy gy gy gy gy.g

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/ne 27 Cross section changes ENDF/B-VII.1 - ENDF/B-VIII.0 OECD/NEA SG-46

energy.gov/ne 28 en en en en en en en en ener er er er er er er er ergy gy gy gy gy gy gy gy gy.g

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/ne 28 Sensitivity of keff to nuclear data quantifies how important each cross section is for application of interest

energy.gov/ne 29 en en en en en ener er er er er ergy gy gy gy gy gy.g

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/ne 29 Role of Sensitivity and Uncertainty Analysis in Validation

• Clearly identifies processes that are important to validate

- Materials, Nuclides, Reactions, Energy

• Assists with challenging areas of applicability where few or no similar experiments are available

• Premise of S/U-based validation

- Computational biases are primarily caused by errors in the cross-section data

- Errors are bounded by cross-section uncertainties represented in covariance data

energy.gov/ne 30 en en en en en en en en ener er er er er er er er ergy gy gy gy gy gy gy gy gy.g

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/ne Example application of S/U methods:

Safety assessment for transportation of burned nuclear fuel Point-wise neutron cross-section data: ~60,000 data points per nuclide keff Simplified neutron transport model of fuel pin Problem-specific multi-group neutron cross-section data:

238 data points per nuclide Explicit 3D neutron transport model of shipping cask

energy.gov/ne 31 en en en en en en en en ener er er er er er er er ergy gy gy gy gy gy gy gy gy.g

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/ne 31 Sensitivities of keff of a shipping cask to cross section data

energy.gov/ne 32 en en en en en ener er er er er ergy gy gy gy gy gy.g

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/ne 32

• ENDF/B-VII.1 contains data for 187 isotopes.

• SCALE 6.1 data retained for ~215 missing nuclides.

• Modified ENDF/B-VII.1 239Pu nubar, 235U nubar, H capture, and several fission product uncertainties, with data contributed back to ENDF/A repository.

• Fission spectrum (chi) uncertainties processed from ENDF/B-VII.1 and from JENDL 4.0 (minor actinides).

• No uncertainties available for scattering secondary particle energy/angular distributions Uncertainties in nuclear data SCALE 6.2 covariance library

energy.gov/ne 33 en en en en en ener er er er er ergy gy gy gy gy gy.g

.g

.g

.g

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.gov ov ov ov ov ov/n

/n

/n

/n

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/ne 33 S/U analysis to identify important processes Application specific

• Overall uncertainty: 0.52% N/k

energy.gov/ne 34 en en en en en en en en ener er er er er er er er ergy gy gy gy gy gy gy gy gy.g

.g

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.gov ov ov ov ov ov ov ov ov/n

/n

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/ne 34 Identify and analyze benchmark experiments to quantify bias in application 34 34 34 34 34 34 34 34 34 34

energy.gov/ne 35 en en en en en en en en ener er er er er er er er ergy gy gy gy gy gy gy gy gy.g

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.gov ov ov ov ov ov ov ov ov/n

/n

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/ne 35

• Quantifies overall similarity potential sources of bias in keff between design application and benchmark experiment.

Correlation coefficient (ck)

(a.k.a. representativity factor) ck V ae 2

V aV e Covariance between Experiment (e) and Application (a) due to all nuclides and reactions Standard deviations for Application (a) and Experiment (e) due to all nuclides and reactions

energy.gov/ne 36 y.g

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.gov ov ov ov ov ov/n

/n

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/ne 36 36 36 36 36 36 36 36 36 36 36 36 36 36 36 36 36 36 36 366 en en enn en en en en en en enn en en en en en enn en en en en en en en en en en en en enn en en en en en en en en en e er er err er er er err er er err er er err er err er err er err er er er err err er e gy gy gy gy gy gy gy gy gy gy gy gy gy gyy gy gyy gy gy gy gy g

NUCLEAR CRITICALITY EXPERIMENTS APPLICATION APPLICATION Code Validation: Identification of laboratory experiments that are similar to the targeted application

energy.gov/ne 37 en en en en en en en en ener er er er er er er er ergy gy gy gy gy gy gy gy gy.g

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.gov ov ov ov ov ov ov ov ov/n

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/ne 37 Similarity as independent parameter for trending analysis Biased keff for Application (bias is this intercept - 1.0)

Confidence band (uncertainty in bias)

Positive Bias Adjustment Gap in experimental data

energy.gov/ne 38 en en en en en en en en ener er er er er er er er ergy gy gy gy gy gy gy gy gy.g

.g

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.gov ov ov ov ov ov ov ov ov/n

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/ne 38 Regulatory basis for validation applicability ISG-10 ck 

recommended Biased keff for Application (bias is this intercept - 1.0)

Confidence band (uncertainty in bias)

Positive Bias Adjustment Gap in experimental data

energy.gov/ne 39 en en en en en ener er er er er ergy gy gy gy gy gy.g

.g

.g

.g

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.gov ov ov ov ov ov/n

/n

/n

/n

/n

/ne 39 Regulatory basis for fission product burnup credit September 2012

 
















energy.gov/ne 40 en en en en en ener er er er er ergy gy gy gy gy gy.g

.g

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.gov ov ov ov ov ov/n

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/ne 40

• Nuclear data and validation studies:

- Gap analysis for nonLWR (ORNL - Sobes/Bostelmann)

- Investigation of HA-LEU transportation validation basis (ORNL -

Rearden/Scaglione/Marshall/Clarity/Holcomb)

• Nuclear data generation:

- Investigation and generation of application driven covariance data (ORNL - Sobes)

- Improvements of nuclear data for depletion, activation, and decay (ORNL - Wieselquist)

- New measurement of 238U (n,n) with associated uncertainties (LBNL - Bernstein)

• International benchmarking activities:

- Multi-Physics Experimental Data, Benchmark, and Validation (ORNL - Valentine)

- International Physics Benchmark Programs: ICSBEP and IRPhEP (INL - Bess)

• University projects:

- Generation of thermal scattering data for graphite (N.C. State, X-energy, ORNL)

- Generation of thermal scattering sensitivity/uncertainty capabilities (U. Michigan, ORNL)

Nuclear Data and Benchmarking Program Initial Activities

energy.gov/ne 41 en en en en en en en en ener er er er er er er er ergy gy gy gy gy gy gy gy gy.g

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ORNL is managed by UT-Battelle, LLC for the US Department of Energy Integral Experiments in the United States -

Cost and Process Douglas G. Bowen, Ph.D.

Nuclear Data and Criticality Safety Group Leader Reactor and Nuclear Systems Division Oak Ridge National Laboratory Nuclear Criticality Safety Program Execution Manager INL-NEI Technical Workshop on Transportation of HALEU August 31, 2018

22 Background / History

• Defense Nuclear Facilities Safety Board (DNFSB)

Recommendations 93-2 and 97-2:

93-2 (3/23/1993): Need for a general-purpose critical experiment capability that will ensure safety in handling and storage of fissionable material.

97-2 (5/19/1997): Need for improved criticality safety practices and programs to alleviate potential adverse impacts on safety and productivity of DOE operations.

• 97-2 encompassed ongoing DOE activities of 93-2 while broadening scope to address important cross-cutting safety activities needed to ensure NCS throughout the Complex.

• DOE Implementation Plan for Board Recommendation 93-2 and 97-2 resulted in establishment of the US Nuclear Criticality Safety Program (NCSP)

3 NCSP Organization and Overview

• Mission

- Provide sustainable expert leadership, direction and the technical infrastructure necessary to develop, maintain and disseminate the essential technical tools, training and data required to support safe, efficient fissionable material operations within the Department of Energy.

• Vision

- Continually improving, adaptable and transparent program that communicates and collaborates globally to incorporate technology, practices and programs to be responsive to the essential technical needs of those responsible for developing, implementing and maintaining nuclear criticality safety.

44 NCSP Technical Program Elements Analytical Methods (AM) - 15% of budget Maintain and improve the Production Codes and Methods for Criticality Safety Engineers (MCNP/SCALE, NJOY/AMPX)

Nuclear Data (ND) - 13% of budget Perform Measurements of Basic Nuclear (Neutron) Physics Cross-Sections and Generate New Evaluated Cross-Section Libraries and Covariance Data for Use in Production Criticality Safety Codes Information Preservation and Dissemination (IPD) - 4% of budget Protects Valuable Analyses and Information Related to Criticality Safety (includes ICSBEP)

Integral Experiments (IE) - 52% of budget Critical and Subcritical Experiments at the Critical Experiments Facility (CEF) at the Device Assembly Facility (DAF) in Nevada and Sandia National Laboratory Pulse Reactor Facility-provides integral tests of codes and data Training and Education (TE) - 6% of budget Web-based training modules and 1- & 2-week Hands-On Criticality Safety courses for Criticality Safety Engineers, Line Management, and Oversight Personnel Technical Support (TS) - 10% of budget Managerial and technical support TS - Technical Support MT - Management team TMs - Task managers CSSG - Criticality Safety Support Group CSCT - Criticality Safety Coordinating Team NDAG - Nuclear Data Advisory Group

55 Current NCSP Work Sites FY2019 NCSP Budget: $26.8 million

66

• National Laboratories Argonne (ANL)

Brookhaven (BNL)

Lawrence Livermore (LLNL)

Los Alamos (LANL)

Oak Ridge (ORNL)

Pacific Northwest (PNNL)

Sandia (SNL)

• Sites Nevada National Security Site (NNSS)

Savannah River (SRNL)

Y-12

• Universities Rensselaer Polytechnic Institute (RPI)

Georgia Institute of Technology (Ga Tech)

North Carolina State University (NCSU)

Massachusetts Institute of Technology (MIT)

University of Florida (Gainesville) (UF)

University of Tennessee (Knoxville) (UTK)

US DOE NCSP Contributors U.K.: AWE (JOWOG-30)

France:

+/-

IRSN (Formal MOU with NCSP)

+/-

CEA (Nuclear Data)

Belgium: Institute for Reference Materials and Measurements (IRMM) differential nuclear data measurements OECD/NEA

+/-

ICSBEP

+/-

WPEC

+/-

WPNCS

,QWHUQDWLRQDO3DUWQHUV 86&RQWULEXWRUV

77 NCSP Integral Experiments

• NCSP integral measurements are performed at Sandia National Laboratories (SNL) and National Criticality Experiments Research Center (NCERC), currently operated by Los Alamos National Laboratory NCERC is located at the Nevada National Security Site (NNSS) inside the Device Assembly Facility (DAF)

• Types of experiments that can be performed Subcritical Rocky Flats shells, BeRP ball, Np-237 sphere, TACS shells, etc.

Critical/Delayed Supercritical NCERC: Planet, Comet, Godiva IV, Flattop Sandia: Sandia Pulse Reactor critical assembly (2 fuel types, currently)

Prompt Supercritical NCERC: Godiva IV (< 300 deg. C pulse)

DAF/NCERC SNL/TA-V/SPR Facility keff Subcritical Regime Delayed Critical Regime Prompt Supercritical Regime Keff<1.0 keff=1 keff=1+

88 NCSP Critical Assemblies NCERC - Np-237 Sphere NCERC - BeRP Ball NCERC - Flattop NCERC - Godiva IV NCERC - TACS SNL - BUCCX - U(4.31)/Fission Product Experiments NCERC - Planet NCERC - Comet Sandia National Laboratory NCERC/DAF SNL - 7uPCX - U(6.9) UO2 rods

99 Overview of the NCSP CEdT Process

• Experimental phases

- CED experiment proposal is submitted

- CED preliminary design of the experiment

- CED final design of the experiment

- CED-3 CED-3a - Schedule/cost/procurement/installations/etc.

CED-3b - experiment execution

- CED-4 CED-4a - summary of experimental data collected during the experiment to ensure it met requirements CED-4b - publish final laboratory report or formal critical experiment benchmark report

• Each experiment is assigned a team of experts to provide support

• The experiments take years to complete and are dependent upon the regulatory environment, critical experiment assembly availability, availability of trained operators, etc.

CEdT Manual

Roles & Responsibilities

Guidance for

Completing the experimental phases

Obtaining approval from the NCSP Manager

Requesting schedule/scope baseline changes

Technical conflict resolutions

Using the NCSP experiment database

Requesting a new experiment

10 10 Costs to Design and Perform Critical Experiments CEdT Phase Gate Description Cost (k$)

(low)

Cost (k$)

(high)

Ž Comments CED-1 Preliminary Design

$ 75

$ 150 3-12 months Depends significantly on the complexity of the experiment CED-2 Final Design

$ 100

$ 250 6-12 months CED-3 CED-3a Costs estimated for procurements and procedure development; resource loaded schedule developed; component fabrication

$ 50

$ 300 3-6 months Material procurements, reactor safety committee approvals, safety basis changes, and procedure reviews can be expensive CED-3b Experiment execution

$ 100

$ 1,000 3-6 months Approximate costs per site:

SNL - $45k/week; NCERC -

$75k/week CED-4 CED-4a Process experimental data; Begin to document final report

$ 50

$ 250 3-6 months CED-4b Publish final report

$ 50

$ 150 6-12 months Sponsor report or an evaluation for the International Handbook of Evaluated Criticality Safety Benchmark Experiments Total Estimated Cost

$ 425

$ 2,100 24-54 months

11 11 Experimental Cost Discussion

• Sandia Example (6.9% Fuel Benchmark)

- Experiments for ICSBEP handbook Series of 19 configurations

- Experimental duration and costs Phase Date/Duration Cost (x$1,000)

CED-0 Late 2012 CED-1 3/2013 80 CED-2 9/2013 75 CED-3a 1/2014 200 CED-3b 9/2014 195 CED-4a 9/2015 243 CED-4b Total Cost Duration ~3 yr.

793

12 12 Experimental Cost Discussion

• NCERC Example (LLNL Pu TEX Experiments)

- Experiments for ICSBEP handbook Series of 10 experiments

+/- Five baseline thermal, intermediate, and fast experiments

+/- Five with a tantalum layer to test cross sections

- Experimental duration and costs Phase Date/Duration Cost (x$1,000)

CED-0 5/2011 CED-1 9/2012 100 CED-2 11/2014 150 CED-3a 10/2017 200

- 65 (Component Fabrication)

- 125 (Procedure Dev.)

CED-3b In progress (2018) 600 (est.)

CED-4a TBD 250 (est.)

CED-4b Total Cost 7+ years so far 1,300

13 13 Experimental Cost Discussion

• NCERC Example (Extreme)

KRUSTY Critical Experiment NNSA/NASA collaboration CEdT Team consisted of LANL personnel and the NNSS M&O operator Phase Durations and Costs Phase Duration CED-1 1 yr.

CED-2 1.5 yr.

CED-3a 7 mo.

CED-3b 3 mo.

CED-4a 1.5 yr.

expected CED-4b Total Cost Duration ~3 yr.

14 Questions

ORNL is managed by UT-Battelle, LLC for the US Department of Energy Overview of Criticality Analysis Validation Presented by:

John M. Scaglione Reactor and Nuclear Systems Division Oak Ridge National Laboratory

22 Methods using sensitivity and uncertainty (S/U) analysis to assess similarity of models are available in existing computer codes

• The International Handbook of Evaluated Criticality Safety Benchmark Experiments (IHECSBE) contains ~5,000 laboratory critical experiments performed at various critical facilities around the world

• Computational tools are available to survey the critical experiments and use a mathematics-based approach to select benchmarks that are applicable to the application model of interest (e.g.,

transportation package model)

• Techniques are available to fill in gaps using cross section data uncertainty (NUREG/CR-7109)

33 Performance of criticality calculations requires detailed knowledge of the application system (package and contents) and modes for reconfiguration

• Parameters important for nuclear criticality safety control include materials, mass, geometry, density, enrichment, reflection, moderation, concentration, interaction, neutron absorption, and volume

• Fuel forms to focus on Powder Pellets Rods Fuel assemblies

• Configuration development considers both normal conditions of transport and hypothetical accident conditions Demonstrate under all credible transport conditions that the system is subcritical Traditional Advanced reactors UO3 Triso UO2 Metal UF6 Oxide Molten salt

44 Criticality safety analyses are performed to show that a proposed fuel transport configuration meets applicable requirements

- 10 CFR 71.55 general requirements for fissile material packages:

a package used for the shipment of fissile material must be so designed and constructed and its contents so limited that it would be subcritical if water were to leak into the containment system, or liquid contents were to leak out of the containment system so that, under the following conditions, maximum reactivity of the fissile material would be attained:

1) The most reactive credible configuration consistent with the chemical and physical form of the material;

2) Moderation by water to the most reactive credible extent; and

3) Close full reflection of the containment system by water on all sides, or such greater reflection of the containment system as may additionally be provided by the surrounding material of the packaging.

55 Calculated results frequently do not exhibit exact agreement with expectations

• The computational method is the combination of the computer code, the data used by the computer code, and the calculational options selected by the user

• Criticality safety evaluations require validation of the calculational method with critical experiments that are as similar as possible to the safety analysis models and for which the keff values are known

• The goal of this validation is to establish a predictable relationship between calculated results and reality

- A quantitative understanding of the difference or bias between calculated and expected results

- Uncertainty in this difference (bias uncertainty)

66 The traditional approach to criticality validation is to compute bias and bias uncertainty values through comparisons with critical experiments

• Trending analyses are typically used in these comparisons

• The difference between the expected and calculated values of the effective neutron multiplication factor, keff, of a critical experiment is considered the computational bias for that experiment

• The uncertainty in the bias is established through a statistical analysis of the trend

77 Criticality analysis process

• Develop application model and identify metrics that define it

• Select appropriate benchmark experiments

• Calculate bias and uncertainty

• Process is agnostic to application model RCA = Radiochemical assay LCE = laboratory critical experiment

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88 Acceptance criterion kp + 'kp + i + 'ki + + 'kE + 'kx + 'km klimit kp is the calculated multiplication factor of the model for the system being evaluated kp is an allowance for statistical or convergence uncertainties, or both, in the determination of kp, material and fabrication tolerances, uncertainties due to geometric or material representation limitations of the models used in the determination of kp is the bias that results from the calculation of the benchmark criticality experiments using a particular calculation method and nuclear cross section data k is bias uncertainty that includes statistical or convergence uncertainties, or both, in the computation of,

uncertainties in the benchmark criticality experiments, uncertainty in the bias resulting from application of the linear least-squares fitting technique to the critical experiment results, and a tolerance interval multiplier to yield a single-sided 95% probability and 95% confidence level kx is a supplement to and k that may be included to provide an allowance for the bias and uncertainty from nuclide cross section data that might not be adequately accounted for in the benchmark criticality experiments used for calculating km is a margin for unknown uncertainties and is deemed adequate to ensure subcriticality of the physical system being modeled. This term is typically referred to as an administrative margin klimit is the upper limit on the keff value for which the system is considered acceptable

99 Selection of critical experiments

• The critical experiments and the safety basis model need to use the nuclear data in a similar energy-dependent manner; otherwise, an incorrect bias could be generated

• Historically, similarity has been left largely to professional judgment using qualitative and integral quantitative comparisons to select critical experiments

- Qualitative parameters considered might include

• fissionable, moderating, and neutron-absorbing materials present;

• type of geometry (e.g., fuel pin lattices);

• type of neutron reflection (i.e., bare, water reflected, steel reflected, etc.);

• qualitative characterization of the energy dependence of the neutron flux as thermal, intermediate, or fast

- Quantitative parameters include

• Energy of average lethargy of a neutron causing fission (EALF)

• ratio of moderating nuclei to fissile nuclei (e.g., H/X)

• fuel enrichment

• lattice fuel pitch

10 10 Sensitivity/uncertainty (S/U) tools can be used to assess application and critical experiment model similarity with a quantifiable metric

• Uncertainty analysis is performed for the safety analysis (application) model and for each candidate critical experiment model Uncertainty analysis results rely heavily on the cross-section uncertainty data in the covariance data file Sensitivity is the fractional change in keff due to a fractional change in a nuclear data value or S (¨k/k)/(¨/)

• Energy-dependent keff uncertainties for each application model and each critical experiment are compared, producing a correlation coefficient (ck) for each application/experiment model pair A high ck value of near 1 for an application/critical experiment pair indicates that both models have similar sensitivities to the same nuclear data and consequently should have similar biases Low ck values indicate that the two systems differ significantly and may have significantly different biases

11 11 In many instances there are nuclides in the application model for which there are few or no appropriate critical experiments available

• Historically, when a particular material could not be evaluated in a safety analysis model, the material was either removed or a ¨k penalty was used based on engineering judgment

• NUREG/CR-7109 provides a validation approach for nuclides that lack experimental data (e.g., minor actinides and structural materials) for criticality safety evaluations

- The approach is based on the uncertainty in keff due to nuclear data uncertainties

- Model-specific sensitivity data, which are in units of (¨k/k)/(¨/), can be used to translate nuclear data uncertainties, which are in units of ¨/, into uncertainty in the model keff value

12 12 Plots of computational and experimental uncertainty

• The plot suggests that the nuclear data uncertainties are overestimated

• It also demonstrates the relative merits of analytical techniques that can be used to address validation gaps using nuclear data uncertainties

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ORNL is managed by UT-Battelle, LLC for the US Department of Energy Example application of process

14 14 Standard UF6 cylinder data Model #

Nominal diameter (in.)

Maximum enrichment (wt% 235U)

Fill limit (lb. UF6)

Model #

Nominal diameter (in.)

Maximum enrichment (wt% 235U)

Fill limit (lb. UF6) 1S 1.5 100.0 1.0 48F 48 4.5 27,030 2S 3.5 100.0 4.9 48Y 48 4.5 27,560 5A 5.0 100.0 54.9 48T 48 1.0 20,700 5B 5.0 100.0 54.9 48O 48 1.0 26,070 8A 8.0 12.5 255.0 48OM Allied 48 1.0 27,030 12A 12.0 5.0 460.0 48OM 48 1.0 26,070 12B 12.0 5.0 460.0 48H, 48HX 48 1.0 27,030 30B, 30C 30.0 5.0 5,020.0 48G 48 1.0 26,840 48A, 48X 48.0 4.5 21,030.0 Source: ANSI N14.1-2012

15Property "ANSI code" (as page type) with input value "ANSI N14.1-2012</br></br>15" contains invalid characters or is incomplete and therefore can cause unexpected results during a query or annotation process. 15 Kaolite-insulated packages ES-4100 4 5-in.dia 58-in. tall 20 gal DPP-3 18-in. dia 30-in. tall 33 gal ES-3100 5-in. dia 31-in. tall 2.6 gal MD-2 17-in. dia 24-in. tall 23 gal DPP-1 14-in. dia 29-in. tall 19 gal DPP-2 12-in. dia 17-in. tall 9 gal Courtesy of Jeff Arbital Y-12 National Security Complex

16 16 Example criticality validation process using the ES-4100 package Photos Courtesy of Jeff Arbital, Y-12 National Security Complex Containment vessel

17 17 ES-4100 design features

• Multi-pack: 4 containment vessels (CVs) per drum

• CV inner dimensions: 5.0-in. dia x 58 in. tall

• Outer drum size: 34.0-in. dia x 71 in. tall

• Insulation: Kaolite 1600

• Neutron absorber: 277-4 cast ceramic w/B4C

• Gross weight: approximately 2,000 lb

- Less than gross weight of four 6M-110s

• Content weight allowance: 4 x 88 lb

- Over 350 lb of content weight ES-4100

18 18 Allowable contents

• University of Missouri Research Reactor (MURR) fuel

• Massachusetts Institute of Technology (MIT) reactor fuel

• Loose Advanced Test Reactor (ATR) fuel rods

• Materials Test Reactor (MTR)-type fuel elements and components

• Foreign Research Reactor (FRR) fuels

• Other fuels

• 1,000 g 235U per CV limit

- Typical US pressurized water reactor (PWR) fuel assembly has ~23,000 g 235U

- Typical US boiling water reactor (BWR) fuel assembly has ~8,700 g 235U

19 19 Selection of applicable critical experiments using similarity assessment Ck is a correlation coefficient indicating how similar an experiment is to an application model

-0.2

-0.1 0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 Ck ICF ICI ICM ICT IMF IMI IMM IST LCF LCM LCT L-Met-T L-Misc-T LST Plot of Ck by LCE group

20 20 Ck trended with enrichment

-0.2

-0.1 0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 0

10 20 30 40 50 60 70 80 90 100 Ck Enrichment (wt% 235U)

21 21 Summary of applicable critical benchmarks Application system Number of applicable critical experiments Package Enrichment

/ BU ICF ICI ICM ICT IMF IMI IMM IST LCF LCM LCT L-Met-T L-Misc-T LST Total ES4100 Evaluated 1

2 6

76 29 3

1 63 1

5 1,157 79 48 113 1,584 Ck > 0.9 0

0 0

0 0

0 0

19 0

0 52 0

7 95 173 Ck > 0.8 0

0 0

0 0

0 0

63 0

0 472 4

46 113 698

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22 22 0.980 0.985 0.990 0.995 1.000 1.005 1.010 1.015 1.020 1.025 0

5 10 15 20 25 30 35 k-eff Enrichment (wt. %)

k(x) - weighted USLSTATS, USL-1 Normalized keff values Application Trend analysis using initial enrichment

23 23 0.980 0.985 0.990 0.995 1.000 1.005 1.010 1.015 1.020 1.025 0

0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1

k-eff EALF (eV) k(x) - weighted USLSTATS, USL-1 Normalized keff values Application Trend analysis using EALF

24 24 0.980 0.985 0.990 0.995 1.000 1.005 1.010 1.015 1.020 1.025 0.8 0.82 0.84 0.86 0.88 0.9 0.92 0.94 0.96 0.98 1

k-eff ck k(x) - weighted USLSTATS, USL-1 Normalized keff values Application Trend analysis using ck similarity coefficient

25 25 Criticality (keff) validation summary

• Validate criticality calculational method using available critical experiment data and appropriate statistical analysis techniques

• Uncertainty in keff due to nuclear data uncertainties can be used to cover validation gaps

• If new critical experiments are needed, a process exists to ensure that the critical experiment is designed to fill the gaps using existing computational tools

• The fuel form and the packages internal design are important for development of appropriate design basis configurations and selection of applicable benchmarks

• Note that it is also required to demonstrate that the fuel can be stored safely after use in the reactor (10 CFR 50)

- The same criticality experiments may or may not be applicable

- Any new experiment design should also consider storage conditions to maximize range of applicability

ORNL is managed by UT-Battelle, LLC for the US Department of Energy BACKUP

27 27 All nuclear data used in criticality calculations have some error

• Sources of error include

- the type of data

- the experimental apparatus and procedure used to measure the data

- the quality and amount of measured data

- nuclear models used to fill in data gaps

- the evaluation technique used to combine measured and modeled data and resolve conflicting data

- conversion of the data into formats suitable for use in the computational method