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
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Download: ML21145A252 (176)

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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 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

INDUSTRY NEEDS 2018
0.026
0.026

• Values in MTU 2019
  1.506
  1.532
  

2020
2.21
3.7

• Current fleet uses about 2021
  4.2
  7.9
  

2000 MTU/year 2022
3.7
11.6

• Letter to Secretary Perry 2023
  18.8
  30.4
  

July 5, 2018 2024
10.3
40.7

• Data from eight 2025
  12.4
  53.1
  

companies 2026
57.4
110.5

2027
73.6
184.1

• Not all ARs or advanced 2028
  108.1
  292.2
  

fuels need HALEU 2029
111.8
404.0

2030
185.5
589.5

QUESTIONS TO CONSIDER

• Will the fuel cycle process be similar to current fleet?

Mining C Conversion Enrich Fab b Reactor U3O8 UF6 UF6 FF

• 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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Validation and Role of Critical Experiments and Nuclear Data Presented to:

Presented by:

Technical Workshop on Transportation of High Assay Low-Enriched Uranium Bradley T. Rearden, Ph.D.

August 30-31, 2018 National Technical Director Nuclear Energy Institute Nuclear Data and Benchmarking Program August 30, 2018

Nuclear Data and Benchmarking Program Office of Nuclear Energy

• 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 ENDF/B-VII.0 JEFF-3.2 ENDF/B-VII.1 2 energy.gov/ne ener energy er gy.g gy.gov

.gov/n ov/ne

/ne

Abbreviated advanced reactor technology matrix (1/2)

Licensing Radial Reactor Companies Fuel / Thermal Fast Flowing Fuel Control action Coolant core Type Red = NRC Priority Enrichment spectrum Spectrum Fuel Form elements expected expansion Sodium Metallic External Oklo 2019 ~20%

heat pipes Castings drums HPR Sodium Westinghouse Thermal/ heat pipes External 2019 19.75% Oxide (eVinci) Epithermal (dual drums condenser)

TerraPower Metallic

~20% Sodium Internal rods (TWR) Rods SFR Metallic GE PRISM ~20% Sodium Internal rods Rods Oxide/

LFR Westinghouse 15-20% Lead Internal rods Nitride X-energy (Xe- External 2020s 15.5% Helium Pebbles TRISO 100) rods HTGR Areva (SC-

~20% Helium TRISO Internal rods HTGR)

External FHR Kairos 2020s ~17% FLiBe Pebbles TRISO rods 3 energy.gov/ne ener energy er gy.g gy.gov

.gov/n ov/ne

/ne

Abbreviated advanced reactor technology matrix (2/2)

Companies Licensing Radial Reactor Fuel / Thermal Fast Flowing Fuel Control Red = NRC action Coolant core Type Enrichment spectrum Spectrum Fuel Form elements Priority expected expansion Terrestrial Energy Molten 2019 ~5% Proprietary Salt Internal rod (IMSR) Salt Internal ZrH Thermal/ Molten Transatomic 2020s ~5% FLiBe Salt moderating Epithermal Salt rods MSR TerraPower Molten External 2020s ~20% Chloride salt Salt (MCFR) Salt rods?

Molten Elysium ~20% Chloride salt Salt Salt Molten FLiBe Energy Thorium FLiBe Salt Internal rods Salt Send updates to Brad - reardenb@ornl.gov 4 energy.gov/ne ener energy er gy.g gy.gov

.gov/n ov/ne

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Nuclear data is of fundamental importance in nuclear science and engineering Neutronics calculations rely on nuclear data for criticality, reactivity, power distributions, depletion, decay heat, and more.

5 energy.gov/ne ener energy er gy.g gy.gov

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Different reactor designs have different nuclear data needs 6 energy.gov/ne ener energy er gy.g gy.gov

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Nuclear data lifecycle 1.030 HEU-SOL-THERM-014 and -016 1.025 1.020 1.015 1.010 C/E Nuclear Data D Differential Data 1.005 1.000 Needs Measurements 0.995 0.990 0.985 v7 v7.1 v7.1-56 711-56 1-5 56 v7.1-252 v7 v7.1 1 25 1-25 2522 v7.1-200 v7 v7.1 1 200 1-20 200 ce_v7.1 cce_v7 e 7.1 Exp.. Unc. MG XS Unc.

Validation ation and Data Evaluation Applications (SAMMY)

(SCALE)

Nuclear N uc Data Evaluated Processing Nuclear Data (AMPX) Files (ENDF) 7 energy.gov/ne ener energy er gy.g gy.gov

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Cross section components:

Typically generated separately, then combined for distribution Thermal Resolved Scattering Resonances Fast Energy Range Unresolved Resonances 8 energy.gov/ne ener energy er gy.g gy.gov

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How are these general purpose libraries generated?

• 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 9 energy.gov/ne ener energy er gy.g gy.gov

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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 -16 p.c.m.

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 +275 p.c.m.

for very different reasons. This replacement study shows -122 p.c.m.

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.

+522 p.c.m. -638 p.c.m.

-14 p.c.m.

• E. Bauge et al., Eur. Phys. J. A (2012) 48: 113 We do not know if either evaluation is correct but both get the correct answer.

10 energy.gov/ne ener energy er gy.g gy y.g

.go go ov/n ov/ne

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

Angular Distribution Proposed Final V. Sobes - ORNL 11 energy.gov/ne ener energy er gy.g gy.gov

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Nuclear data for activation, depletion, and decay

• 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 12 energy.gov/ne ener energy er gy.g gy.gov

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

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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)%

MET1000 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 MOX3600 14 energy.gov/ne ener energy er gy.g gy.gov

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Recent nuclear data developments of interest to the advanced reactor community 15 energy.gov/ne ener energy er gy.g gy.gov

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Changes in graphite data ENDF/B-VII.0 (2006) ENDF/B-VIII.0 (2018) to ENDF/B-VII.1 (2011)

• New evaluations for thermal

• Capture cross section increased from scatter based on molecular 3.36 mb to 3.86 mb: ~1,000 pcm dynamics models from North Carolina State ENDF-VII.0 ENDF-VII.1 HTTR loading C/E C/E

• Includes data for crystalline Initial criticality 1.0165
  1.0011
  and reactor-processed Full core 1.0097
  1.0015
  

graphite HTR-10 ENDF-VII.1 ENDF-VIII.0 Configuration C/E C/E First core 1.00267
1.00582

HTR-10 Benchmark A. Hawari NC State 16 energy.gov/ne ener energy er gy.g gy.gov

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HTR-10 pebble: KENO-VI eigenvalue comparison Library Code XS lib k N SFP ENDF/B-VII.1 KENO CE 1.6770(4) (ref)

ENDF/B-VIII.0 KENO CE 1.6722(4) í 

1RWH7KHVWDWLVWLFDOXQFHUWDLQWLHVDUHJLYHQLQSDUHQWKHVHV

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

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

• Differences between ENDF/B-VII.0 and VII.1: carbon capture HTR-10 fuel pebble

• Differences between ENDF/B-VII.1 and VIII.0: 235U and 238U 17 energy.gov/ne ener energy er gy.g gy.gov

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Changes in 35Cl(n,p) cross section from ENDF/B-VII.0 to VII.1 Data Library keff ENDF/B-VII.0 1.02993 +/- 0.00002 ENDF/B-VII.1 1.04924 +/- 0.00002 ENDF/B-VII.0 JEFF-3.2 Simplified Molten Chloride Fast Reactor ENDF/B-VII.1 Reaction Sensitivity Cl-35 (n,p) Capture Reaction -0.958 Pu-239 Nu-bar 0.603 No data for FLiBe / FLiNaK thermal scattering U-238 Nu-bar 0.281 Na-23 Elastic Scatter Reaction 0.114 18 energy.gov/ne ener energy er gy.g gy.gov

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Validation of methods and nuclear data for advanced applications 19 energy.gov/ne ener energy er gy.g gy.gov

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International benchmark evaluation projects

• 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)

ICSBEP IRPhEP

• 22 contributing Countries

• 21 contributing countries

• ~69,000 pages

• 50 reactor facilities

• >5,000 approved benchmarks

• 147 approved benchmarks 20 energy.gov/ne ener energy er gy.g gy.gov

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

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International Spent Nuclear Fuel Database SFCOMPO 2.0 provides a central repository of destructive assay data 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.

http://www.oecd-nea.org/sfcompo/

• 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 239Pu

• 44 reactors, 118 assemblies, 91 isotopes important data (all reactor types) 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 22 2 energy.gov/ne ener energy er gy.g gy.gov

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5% < Hi-assay LEU < 20%

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

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NRC/NMSS perspectives on high assay fuel 24 energy.gov/ne en ener erggy.g gy g .gov

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Example criticality validation process using the ES-4100 package Photos Courtesy of Jeff Arbital Containment vessel Y-12 National Security Complex 25 energy.gov/ne ener energy er gy.g gy.gov

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

Counteracting errors in ENDF/B-VII.1 - ENDF/B-VIII.0 ENDF-7.1: keff = 0.86464 (8)

ENDF-7.1 238U ENDF-8.0 1H

+42 -132 pcm pcm ENDF-8.0 16O ENDF-7.1 235U

+216 -238 pcm ~4 450 p

~450 cm pcm pcm 238 235U+23 38U evaluations l ti ENDF-7.1 from ENDF-8.0 ENDF-8.0 235U ENDF-7.1 16O

+95

-83 pcm pcm

-65 pcm ENDF-7.1 1H ENDF-8.0 238U ENDF-8.0*

26 energy.gov/ne ener energy er gy.g gy.gov

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Cross section changes ENDF/B-VII.1 - ENDF/B-VIII.0 OECD/NEA SG-46 27 energy.gov/ne ener energy er gy.g gy.gov

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Sensitivity of keff to nuclear data quantifies how important each cross section is for application of interest 28 energy.gov/ne ener energy er gy.g gy.gov

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

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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 Simplified neutron transport model of fuel pin keff Problem-specific multi-group Explicit 3D neutron neutron cross-section data:

transport model of 238 data points per nuclide 30 energy.gov/ne ener energy er gy.g gy.gov

.gov/n ov/ne

/ne shipping cask

Sensitivities of keff of a shipping cask to cross section data 31 energy.gov/ne ener energy er gy.g gy.gov

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Uncertainties in nuclear data SCALE 6.2 covariance library

• 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 32 energy.gov/ne ener energy er gy.g gy.gov

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S/U analysis to identify important processes Application specific

• Overall uncertainty: 0.52% N/k 33 energy.gov/ne ener energy er gy.g gy.gov

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Identify and analyze benchmark experiments to quantify bias in application 34 energy.gov/ne ener energy er gy.g gy.gov

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Correlation coefficient (ck)

(a.k.a. representativity factor)

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

Covariance between 2 Experiment (e) and Application (a)

V ae due to all nuclides and reactions ck V aV e Standard deviations for Application (a) and Experiment (e) due to all nuclides and reactions 35 energy.gov/ne ener energy er gy.g gy.gov

.gov/n ov/ne

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Code Validation: Identification of laboratory experiments that are similar to the targeted application APPLICATION NUCLEAR CRITICALITY EXPERIMENTS 36 6 energy.gov/ne en ene errg gyy.g

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Similarity as independent parameter for trending analysis Gap in experimental data Biased keff for Application (bias is this intercept - 1.0)

Confidence band (uncertainty in bias)

Positive Bias Adjustment 37 energy.gov/ne ener energy er gy.g gy.gov

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Regulatory basis for validation applicability ISG-10 ck 

recommended Gap in experimental data Biased keff for Application (bias is this intercept - 1.0)

Confidence band (uncertainty in bias)

Positive Bias Adjustment 38 energy.gov/ne ener energy er gy.g gy.gov

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Regulatory basis for fission product burnup credit September 2012

 

  
  



 

39 3 9 energy.gov/ne ener energy er gy.g gy.gov

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Nuclear Data and Benchmarking Program Initial Activities

• 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) 40 energy.gov/ne ener energy er gy.g gy.gov

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

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

Background / History

• Defense Nuclear Facilities Safety Board (DNFSB)

Recommendations 93-2 and 97-2:

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

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) 2

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.

3

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 TS - Technical Support

• Technical Support (TS) - 10% of budget MT - Management team TMs - Task managers

- Managerial and technical support CSSG - Criticality Safety Support Group CSCT - Criticality Safety Coordinating Team 4 NDAG - Nuclear Data Advisory Group

Current NCSP Work Sites FY2019 NCSP Budget: $26.8 million 5

US DOE NCSP Contributors 86&RQWULEXWRUV ,QWHUQDWLRQDO3DUWQHUV

• National Laboratories U.K.: AWE (JOWOG-30)

- Argonne (ANL)

France:

- Brookhaven (BNL)

- Lawrence Livermore (LLNL)

+/- IRSN (Formal MOU with NCSP)

- Los Alamos (LANL) +/- CEA (Nuclear Data)

- Oak Ridge (ORNL) Belgium: Institute for Reference Materials and

- Pacific Northwest (PNNL) Measurements (IRMM) differential nuclear

- Sandia (SNL) data measurements

• Sites

- Nevada National Security Site (NNSS) OECD/NEA

- Savannah River (SRNL) +/- ICSBEP

- Y-12 +/- WPEC

• Universities +/- WPNCS

- 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) 6

NCSP Integral Experiments DAF/NCERC

• 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 SNL/TA-V/SPR Facility

- 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)

Delayed Critical Regime keff=1+

keff Subcritical Regime Prompt Supercritical Regime 7 Keff<1.0 keff=1

NCSP Critical Assemblies Sandia National Laboratory NCERC/DAF SNL - BUCCX - U(4.31)/Fission Product Experiments NCERC - Np-237 Sphere NCERC - BeRP Ball NCERC - TACS NCERC - Godiva IV NCERC - Flattop SNL - 7uPCX - U(6.9) UO2 rods NCERC - Planet NCERC - Comet 8

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 CEdT Manual Roles & Responsibilities Guidance for

• The experiments take years to complete and are Completing the experimental phases Obtaining approval from the NCSP dependent upon the regulatory environment, critical Manager Requesting schedule/scope baseline experiment assembly availability, availability of trained changes operators, etc. Technical conflict resolutions Using the NCSP experiment database Requesting a new experiment 9

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

Description Ž Comments Gate (low) (high)

CED-1 Preliminary Design $ 75 $ 150 3-12 months Depends significantly on the CED-2 Final Design $ 100 $ 250 6-12 months complexity of the experiment Material procurements, reactor Costs estimated for procurements and safety committee approvals, procedure development; resource CED-3a $ 50 $ 300 3-6 months safety basis changes, and loaded schedule developed; procedure reviews can be CED-3 component fabrication expensive Approximate costs per site:

CED-3b Experiment execution $ 100 $ 1,000 3-6 months SNL - $45k/week; NCERC -

$75k/week Process experimental data; Begin to CED-4a $ 50 $ 250 3-6 months document final report CED-4 Sponsor report or an evaluation for the International Handbook CED-4b Publish final report $ 50 $ 150 6-12 months of Evaluated Criticality Safety Benchmark Experiments Total Estimated Cost $ 425 $ 2,100 24-54 months 10

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 11

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 150 200 CED-3a 10/2017 - 65 (Component Fabrication)

- 125 (Procedure Dev.)

CED-3b In progress 600 (est.)

(2018)

CED-4a TBD 250 (est.)

CED-4b Total Cost 7+ years 1,300 so far 12

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.

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

13

Questions 14

Overview of Criticality Analysis Validation Presented by:

John M. Scaglione Reactor and Nuclear Systems Division Oak Ridge National Laboratory ORNL is managed by UT-Battelle, LLC for the US Department of Energy

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) 2

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 Traditional Advanced reactors

• Fuel forms to focus on UO3 Triso

- Powder

- Pellets UO2 Metal

- Rods UF6 Oxide

- Fuel assemblies

• Configuration development considers both Molten salt normal conditions of transport and hypothetical accident conditions

- Demonstrate under all credible transport conditions that the system is subcritical 3

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.

4

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) 5

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 6

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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 8

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 9

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 10

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 11

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

Standard UF6 cylinder data Model # Nominal Maximum Fill limit Model # Nominal Maximum Fill limit diameter enrichment (lb. UF6) diameter enrichment (lb. UF6)

(in.) (wt% 235U) (in.) (wt% 235U) 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 14

Kaolite-insulated packages ES-4100 4 5-in.dia 58-in. tall 20 gal DPP-3 DPP-1 ES-3100 18-in. dia MD-2 14-in. dia 5-in. dia 30-in. tall DPP-2 17-in. dia 29-in. tall 31-in. tall 33 gal 12-in. dia 24-in. tall 19 gal 2.6 gal 17-in. tall 23 gal 9 gal Courtesy of Jeff Arbital 15 Y-12 National Security Complex

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

ES-4100 design features ES-4100

• 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 17

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 18

Selection of applicable critical experiments using similarity assessment Plot of Ck by LCE group 1.0 0.9 Ck is a 0.8 correlation 0.7 coefficient 0.6 indicating how 0.5 similar an Ck 0.4 experiment is to 0.3 an application 0.2 model 0.1 0.0

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-0.2 0 10 20 30 40 50 60 70 80 90 100 235 Enrichment (wt% U) 20

Summary of applicable critical benchmarks

/&(VE\JURXS Application system Number of applicable critical experiments Package Enrichment Total

/ BU ICF ICI ICM ICT IMF IMI IMM IST LCF LCM LCT LST L-Met-T L-Misc-T 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 21

Trend analysis 1.025 k(x) - weighted using initial USLSTATS, USL-1 1.020 Normalized keff values enrichment Application 1.015 1.010 1.005 1.000 k-eff 0.995 0.990 0.985 0.980 0 5 10 15 20 25 30 3 Enrichment (wt. %)

22

Trend 1.025 k(x) - weighted analysis 1.020 using EALF USLSTATS, USL-1 1.015 Normalized keff values Application 1.010 1.005 1.000 k-eff 0.995 0.990 0.985 0.980 0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1 EALF (eV) 23

Trend analysis 1.025 k(x) - weighted using ck 1.020 USLSTATS, USL-1 similarity Normalized keff values coefficient 1.015 Application 1.010 1.005 1.000 k-eff 0.995 0.990 0.985 0.980 0.8 0.82 0.84 0.86 0.88 0.9 0.92 0.94 0.96 0.98 1 ck 24

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 25

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

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 27