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DOE OSTI2025

Impact of recent ENDF nuclear data update, high initial enrichment and high burnup fuel on critical experiments applicability determination via the integral index c k for burnup credit validation

In 2012, NUREG/CR-7109 reported on the validation of burnup credit calculations involving major and minor actinides and major fission products which was investigated for pressurized and boiling water reactor (PWR and BWR) fuel enrichments up to 5 wt% 235 U and assembly-average burnups up to 60 GWd/MTU. Recently, there has been interest in increasing the maximum enrichment used in PWR fuel as high as 8 wt% 235 U and correspondingly increasing the maximum assembly-average burnups to approximately 75 GWd/MTU. These proposed increases in enrichment and burnup necessitate reinvestigation of the validation basis for k eff calculations for this expanded application space. Additionally, the 2012 study was performed by using the Evaluated Nuclear Data File (ENDF)/B-VII.0 nuclear data with the SCALE 6 covariance library, and the effects of using the newly released ENDF/B-VII.1 and ENDF/B-VIII.0 nuclear data and covariance libraries should be evaluated. In this work, published in NUREG/CR-7309 in 2025, the validation assessment was performed consistently with NUREG/CR-7109: modeling irradiated fuel assemblies in the Generic Burnup Credit (GBC)-32 cask defined in NUREG/CR-6747. The TSUNAMI-3D sequence was used to generate sensitivity data for the application model, and the data were compared with sensitivity data from select benchmark models. The integral parameter c k is the metric of similarity used in this study and is consistent with NUREG/CR-7109, where a c k value in excess of 0.8 indicates sufficient similarity for use in validation. A new set of benchmark experiments with sensitivity data has been assembled for this effort. The number of experiments with available sensitivity data is now 2,104, compared to 474 in NUREG/CR-7109. This increase was facilitated by the efforts of the Nuclear Energy Agency to generate sensitivity data for a majority of the experiments in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook to supplement the data available in the Oak Ridge National Laboratory (ORNL) Verified, Archived Library of Inputs and Data (VALID). The complete set of benchmarks considered here includes experiments for low-enriched uranium (LEU), intermediate enriched uranium (IEU), and a mixture of uranium and plutonium (MIX) from the ICSBEP Handbook and VALID, as well as ORNL models of the Haut Taux de Combustion (HTC) experiments and other potentially relevant models not included in VALID. The updated similarity study shows that none of the extended burnup and higher enrichment combinations considered show a significant decrease in the number of potentially applicable experiments, meaning sufficient critical experiments exist for the validation of BUC criticality safety calculations, with initial enrichments up to 8 wt% 235 U and burnups up to 80 GWd/MTU. Additionally, both the ENDF/B-VII.1 and ENDF/B-VIII.0 nuclear data libraries can be used for validation since the number of critical experiments applicable for validation increases for most cases with the most recent nuclear data compared to the previous one. As in previous BUC validation studies, the French HTC experiments are the most similar in a majority of the application cases studied, especially from representative discharge burnups ranging from 40 to 80 GWd/MTU. In conclusion, these results match the conclusions presented in NUREG/CR-7109 regarding validation of the primary actinides in BUC analyses.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

TEX-Hf Assemblies: Highly Enriched Uranium Plates with Hafnium Using Polyethylene Moderator and Polyethylene Reflector

This report documents the integral experiment evaluation and publication for IER-532 (TEX-Hf), Thermal/Epithermal eXperiments (TEX) with highly enriched uranium (HEU) fuel and hafnium (Hf), moderated and reflected by polyethylene. The design of TEX-Hf is a variation of IER-297 (TEX-HEU), with the inclusion of hafnium as a diluent material. The experiment and evaluation include seven configurations, all of which were acceptable as benchmark cases. These benchmark cases provide validation for hafnium in the thermal, intermediate, and fast neutron energy regimes by maximizing the sensitivity in k eff to the hafnium cross sections. The evaluation was reviewed and accepted by the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Technical Review Group on April 17, 2024, and was submitted to the ICSBEP in August 2024 following subgroup approval.

42 ENGINEERING↗

DOE OSTI2025

Final Design for Thermal/Epithermal eXperiments (TEX) with Lithium Absorbers to Provide Validation Benchmarks for Y-12 Electrorefining Facility

One of the main goals of the Thermal/Epithermal eXperiments (TEX) project is to use existing Nuclear Criticality Safety Program (NCSP) assets to create critical experiment plutonium and uranium test beds for materials important to criticality safety that have insufficient benchmark evaluations. The plutonium test bed experiments were completed in 2018 and are published in the 2020 edition of the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook. The uranium test bed assemblies were completed in 2023 and accepted in the 2024 edition of the ICSBEP Handbook. The Nuclear Criticality Safety (NCS) group at Y-12 National Security Complex has identified programmatic need for validation cases for uranium electrorefining operations at Y-12. The electrorefining operation credits lithium enriched in 6 Li in addition to 35 Cl as absorbers in the design criticality safety evaluation for precluding criticality under upset conditions in the large and geometrically unfavorable electro-refiner. There is, however, inadequate experimental validation for the 6 Li absorbers. As an extension of the TEX uranium test bed, TEX-Cl critical experiments were performed with sodium chloride salt to address the 35 Cl thermal absorption as well as other validation needs at Los Alamos National Laboratory (LANL). These experiments were completed in 2024 and accepted into the 2025 ICSBEP Handbook. To continue the methodology used in TEX-Cl, TEX-Li aims to accomplish the same. The overall design of both experiments was to use commercially available, high purity, salts with polyethylene moderator and HEU plates to configure a critical assembly. Three experiments are planned for TEX-Li using encapsulated lithium carbonate (Li 2 CO 3 ), with natural 6 Li abundance. For all experiments, the highly enriched uranium (HEU) Jemima plates will be used as fissile material. Multiple layers will be stacked together with encapsulated Li 2 CO 3 alternated with polyethylene in standard configurations. Standard stacking was found to be optimal in matching the different sensitivity profiles provided by the Y-12 models. Three configurations are proposed with varying polyethylene moderation and a constant 1/4” absorber thickness. The first uses 11 layers of 5/4” polyethylene, the second uses 9 layers of 3/4” polyethylene, and the third uses 10 layers of 1/2” polyethylene. Calculations showed that some alternative forms of lithium-based materials provided slightly less-optimal sensitivity profiles when compared to lithium carbonate but come with other drawbacks. These alternatives included lithium aluminate (LiAlO 2 ), Aluminum-2050 alloy, Aluminum-8090, Aluminum-2095, lithium hydride (LiH), and lithium fluoride (LiF). Lithium aluminate and aluminum-2050 provided comparable sensitivity profiles when compared to lithium carbonate and can be used instead if lithium carbonate cannot be readily procured. After a broad material study, lithium carbonate outperformed any alternative material with a balance in affordability and workability. The assessment of experimental uncertainties of the non-absorber and absorber components was predicted to be 0.00089 and 0.00093 Δk eff , respectively. The largest uncertainties may be reduced with precision dimensional inspection of the components. Many of the parts and equipment for IER 575 have already been fabricated or procured for previous projects and therefore do not contribute significantly to the overall cost of this experiment. This includes the Jemima plates and Comet critical assembly machine, which are existing NCSP assets, as well as the aluminum platen and polyethylene reflector rings, which were fabricated and authorized for the TEX experiment involving HEU with polyethylene. Lithium carbonate containers will be procured by LANL and will be filled by LLNL. The total material costs for TEX-Li experiments are estimated to be on the order of $\$$47,400. Precision inspection, including dimensional, mass, density, and impurity, is recommended for all components for an estimated cost of $\$$12,000.

35Cl↗

DOE OSTI2026

Final Design for Thermal/Epithermal eXperiments (TEX) with Lithium Absorbers to Provide Validation Benchmarks for Y-12 Electrorefining Facility (IER 575, CED-2 Report)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2025

An entropy-based debiasing approach to quantifying experimental coverage for novel applications of interest in the nuclear community

This manuscript proposes a novel information-theoretic approach to the quantification of experimental relevance, i.e., coverage, to achieve optimal data assimilation results for nuclear engineering applications. Specifically, this work posits the need for a new metric, called coverage (q C ) of an application’s quantity of interest, i.e., eigenvalue or power peaking for an advanced reactor concept, defined herein as the theoretically maximum achievable reduction in the quantity’s uncertainty given measurements from a pool of experiments in a manner that is independent of the data assimilation procedure employed. Currently, reduction in a quantity’s uncertainty is strongly biased by the underlying assumptions of the assimilation procedure to account for the under-determined nature of such problems and the similarity criterion employed to identify relevant experiments. To address this challenge, this work has developed a coverage metric, q C , based on mutual information, which establishes a new conceptual framework for assessing coverage, one that is independent of the model parameters and responses degree of variations in both the experimental and application domains, i.e., linear vs non-linear, and their prior uncertainty distributions, i.e., Gaussian vs. non-Gaussian. The q C is an entropic measure capable of addressing coverage for general nonlinear problems with non-Gaussian uncertainties and inclusive of the measurement uncertainties from multiple experiments. Numerical experiments from manufactured analytical problems as well as a set of benchmarks from the ICSBEP handbook are employed to demonstrate its theoretical and practical performance as compared to the c k -based experiment selection methodology, commonly employed in the neutronic community. The manuscript then employs other well-known adaptations to existing data assimilation methodologies for nonlinear and non-Gaussian problems capable of achieving the coverage posited by q C .

Bayesian data assimilation↗

DOE OSTI2026

Evaluation of Sandia NCS Benchmark Suite Updates

The Sandia Nuclear Criticality Safety (NCS) program’s benchmark suite was recently updated. This suite is used to ensure that NCS calculations using computer-based neutron transportation codes have an established baseline comparison of calculated versus known experimental results. The Evaluated Nuclear Data File (ENDF) version used for the MCNP models in the suite was changed from ENDF/B-VII.1 to ENDF/B-VIII.0. Additionally, relevant thermal scattering law data libraries (TSLs) were updated. The sensitivity of the calculational bias of each benchmark model to these changes is discussed. Implementation of the ENDF/B-VIII.0 library and updated TSLs results in improvements to bias distribution in the intermediate enriched uranium, plutonium, and mixed uranium–plutonium (IEU, PU, and MIX) fissionable material benchmark categories, but a small bias increase in low- and high-enriched uranium categories (LEU and HEU, respectively). The results also highlight the sensitivity of the benchmarks, with average lethargy of neutrons causing fission energies (EALF) in the intermediate energy range to ENDF/B library changes. The most numerous bias changes were observed in the thermal energy region when transitioning from ENDF/B-VII.1 to ENDF/B-VIII.0. In conclusion, most of the unique bias changes observed in MCNP 6.3.0 between the two nuclear data libraries were in the LEU-COMP-THERM evaluation subset.

ICSBEP↗

DOE OSTI2026

Expanded Intercomparison of Nuclear Data Libraries Using Jupiter and Jupiter High-240 Experiments

There is a limited availability of plutonium experiments with sensitivity to lead in the ICSBEP (International Handbook of Evaluated Criticality Safety Benchmark Experiments) Handbook. The Jupiter and Jupiter High-240 experiments were performed at the National Criticality Experiments Research Center as a collaborative effort between Los Alamos National Laboratory and the Japan Atomic Energy Agency to assess lead void coefficients in a plutonium-lead system containing weapons- and reactor-grade plutonium, respectively. Concurrent with benchmark development, an intercomparison of calculations using different nuclear data libraries has been performed to assess the usability of the experimental data for nuclear data adjustment in a “softer-that-fast” neutron energy spectrum. Eigenvalue calculations using MCNP with the ENDF/B-VIII.0 and TENDL-2021 nuclear data libraries calculate closest to the benchmark values for Jupiter. Calculations using JENDL-5 and ENDF/B-VIII.1 match best with the Jupiter High-240 values. Lead void worth calculations using the various nuclear data libraries are all within 3σ of their respective measured values. Perturbation studies between ENDF/B-VIII.0 and ENDF/B-VIII.1 demonstrate an approximate increase in calculated eigenvalues for the Jupiter series experiments by ~240 pcm for plutonium (mostly 239 Pu) and ~120 pcm for lead accompanied by a decrease contributed by ~113 pcm for copper and ~13 pcm for stainless steel. Nuclear data sensitivities and uncertainties investigated using Whisper show slightly lower sensitivity to scatter than a lead-reflected plutonium sphere but greater sensitivity to neutron capture. The sensitivities between Jupiter and Jupiter High-240 for lead are very similar for both ENDF/B-VIII.0 and ENDF/B-VIII.1 nuclear data. These benchmarks are more sensitive to neutron capture in lead than other plutonium benchmark experiments and would be useful for both lead and 240 Pu validation. In conclusion, with the high degree of compensating effects between copper, lead, and plutonium cross sections, additional isolated Pb-Pu and Cu-Pu benchmarks would be beneficial in improving these nuclear data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Verified, Archived Library of Inputs and Data (VALID) Online Repository [Slides]

VALID continues to serve as a library of high-quality models used for evaluating SCALE and nuclear data. Large number of cases are in the pipeline. Online repository is available for use. Future plans include simplifying the process for adding cases, including additional models outside the ICSBEP Handbook (HTC models, DNCSH models, International Handbook of Evaluated Reactor Physics Benchmark Experiments [IRPhE], SINBAD, etc.), and increasing the availability of models and results for external users.

KENO↗

DOE OSTI2025

TEX-Chlorine: Highly Enriched Uranium with Chloride Absorbers to Provide Validation Benchmarks for Y-12 Electrorefining Facility

This report documents the final benchmark of the TEX-Chlorine (IER-499) Thermal Epithermal eXperiments (TEX) with highly enriched uranium with chlorine absorbers and high-density polyethylene reflectors and moderators. TEX-Chlorine is a variation of the TEX-HEU baseline assembly with the addition of sodium chloride absorber plates. This evaluation contains three experimental configurations that were performed on Comet at NCERC between July and August 2024. The three configurations, which were acceptable as benchmark cases, spanned from thermal (first two configurations) to fast (third configuration). All three cases were reviewed and accepted by the ICSBEP TRG in April 2025 and was submitted to the ICSBEP in August 2025 after receiving subgroup approval.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

DOE OSTI2025

Microreactor Assembly Transportation Cask Model Description for Criticality Safety Validation Basis Assessment

Criticality safety analyses are completed on a transportation cask used for microreactor assembly shipment to provide an example of model and analysis to industry for reproducing this type of study on their microreactor fuel shipment. The fuel assembly considered is based on a gas-cooled microreactor (GC-MR), which utilizes HALEU fuel in the form of TRISO particles and utilizes various design options considered in industry designs. Various versions of this GC-MR assembly were studied, with and without YH2 moderator, providing similar conclusions. The shipment cask design is revised based on an existing design ES-3100, developed by Y-12 for the transport of highly enriched uranium (HEU), but is enlarged to hold the GC-MR fuel assembly. Criticality safety analysis for the cask/GC-MR fuel assembly package was performed using the CSAS6 sequence of SCALE6.3.2, utilizing the ENDF/B-VII.1 based continuous energy neutron library, and the analysis strictly follows the guideline from NRC reference reports. Different scenarios, e.g. normal operation, undamaged cask with water flooded, damaged cask with optimal water moderation, have been analyzed and it could be concluded the package would always have a large margin of subcriticality even packed in an infinite array. Sensitivity and similarity analyses are also performed using the TSUNAMI sequence of SCALE6.3.2, and the similarity analysis uses all the experiments from the ICSBEP Handbook with Intermediate and Mixed Enriched Uranium (IEU) and Low Enriched Uranium (LEU) systems together with additional ones that are sponsored by the DNCSH program. These similarity analyses indicate that dry cases have no similar benchmark experiments (ck values greater than 0.8), which may become problematic if more assemblies are shipped together (or a fully loaded core is shipped) and margin to criticality is reduced. However, the damaged cask models with flooded assemblies exhibited similarities to many experiments with ck values greater than 0.8.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2025

Increased Fidelity and Associated Computational cost of Detailed Integral Experiment Benchmarks [Slides]

It does not seem like the system is significantly more sensitive to diameters of components near the center of the core. Intuitively it is, but was not detectable with simulations run to a Monte Carlo k eff uncertainty of 0.00002. The system is more sensitive to heights of components near the center of the core. Most (if not all) Zeus style benchmarks have perturbed core component heights individually.

42 ENGINEERING↗

DOE OSTI2026

Microreactor Assembly Transportation Cask Model Description for Criticality Safety Validation Basis Assessment (Rev. 3)

Criticality safety analyses are completed on a transportation cask used for microreactor assembly shipment to provide an example of model and analysis to industry for reproducing this type of study on their microreactor fuel shipment. The fuel assembly considered is based on a gas-cooled microreactor (GC-MR), which utilizes HALEU fuel in the form of TRISO particles and utilizes various design options considered in industry designs. Various versions of this GC-MR assembly were studied, with and without YH 2 moderator, providing similar conclusions. The shipment cask design is revised based on an existing ES-3100 design, developed by Y-12 for the transport of highly enriched uranium (HEU), but is enlarged to hold the GC-MR fuel assembly. Criticality safety analysis for the cask/GC-MR fuel assembly package was performed using the CSAS6 sequence of SCALE6.3.2, utilizing the ENDF/B-VII.1 based continuous energy neutron library, and the analysis strictly follows the guideline from NRC reference reports. Different scenarios, e.g. normal operation, undamaged cask with water flooded, damaged cask with optimal water moderation, have been analyzed and it could be concluded the package would always have a large margin of subcriticality even packed in an infinite array. Sensitivity and similarity analyses are also performed using the TSUNAMI sequence of SCALE6.3.2, and the similarity analysis uses all the experiments from the ICSBEP Handbook with Highly Enriched Uranium (HEU), Intermediate and Mixed Enriched Uranium (IEU) and Low Enriched Uranium (LEU) systems, together with additional ones that are sponsored by the DNCSH program, and selected IRPhEP experiments using TRISO fuel and graphite moderator. These similarity analyses indicate that the dry nominal design has no similar benchmark experiments (ck values greater than 0.8), which may become problematic if more assemblies are shipped together (or a fully loaded core is shipped) and margin to criticality is reduced. However, the cask models with flooded assemblies exhibited similarities to many experiments with c k values greater than 0.8.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗