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At least 397 records · Page 22
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

ENDF/B-VIII.1 Validation (Preliminary Release 2 + New 239 Pu)

This document should serve as a reference for validation testing of nuclear data files included in the second preliminary release of ENDF/B-VIII.1 and the newly evaluated 239 Pu nuclear data file to be included in the third preliminary release of ENDF/B-VIII.1. Experiment names listed in this report are based on International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook designations. Additional information on ICSBEP handbook designations can be found at https://www.oecd-nea.org/jcms/pl_20291/icsbep-handbook. The validation metrics used in this report include the effective neutron multiplication factor, k eff , and tritium production. All validation metrics were calculated using the radiation transport code Monte Carlo N-Particle (MCNP).

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Fast Neutron Evaluations for Advanced Lead Systems [Slides]

This presentation touches on fast neutron evaluations for advanced lead systems. This lecture covers project goals to evaluate lead (Pb) isotopes with a focus on improving predictive capabilities of fast systems. Additionally, the project sets out to benchmark available critical and shielding experiments with sensitivity to Pb and utilize RPI Quasi-Differential Scattering experiment for scattering reactions. This project will perform Resolved Resonance Region (RRR) and fast region evaluations for natural Pb isotopes with differential data on EXchange FORmat for experimental numerical nuclear reaction data (EXFOR) and ultimately will submit evaluations to the ENDF/B-VIII.1 library.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2024

HEU Pancake (Jemima) Plate Preliminary Characterization Report

The HEU pancake (Jemima) plates have been used in multiple International Criticality Safety Benchmark Evaluation Project (ICSBEP) evaluations, including: HEU-MET-FAST-072, HEU-MET-FAST-073, HEU MET-FAST-102, HEU-MET-INTER-006, HEU-MET-INTER-011, HEU-MET-MIXED-021, and IEU MET-FAST-025. This report only focuses on the physical dimension characterization, since concerns have been identified about reliable diameter and height measurements. Historically, height measurements with calipers and mass measurements have been performed for every plate. However, the combination of previous measurements (mass and caliper height measurements) with the drawing dimensions for the diameters led to unrealistic densities (>19 g/cm 3 ) or large ranges of densities for parts that were manufactured at the same time (17 g/cm 3 to >19 g/cm 3 ). Due to the oxidation of the plates, questions about the flatness of each plate and what gaps are introduced into the system have been discussed, since gaps tend to be one of the largest sources of uncertainty in stacked benchmark experiments. The purpose of this report is to characterize a subset of this commonly used fuel. A uniform method for how to define the HEU pancake plates for ICSBEP evaluations will also be proposed with a discussion on what measurements should be performed on the remaining HEU pancake plates in the NCERC inventory.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Idaho National Laboratory Team Visits NCERC

The DOE/NRC Criticality Safety for Commercial-Scale High-Assay Low-Enriched Uranium (HALEU) for Fuel Cycle and Transportation (DNCSH) initiative is a collaboration between the Department of Energy and the Nuclear Regulatory Commission focused on criticality safety benchmarking for processing and transportation applications related to HALEU and Advanced Reactors. The project includes design, siting and construction of a horizontal split table (HST), which has long been recognized as a need for validating advanced reactor systems.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

The Paucity Problem: Where Have All the Space Reactor Experiments Gone?

The Handbooks of the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPhEP) together contain a plethora of documented and evaluated experiments essential in the validation of nuclear data, neutronics codes, and modeling of various nuclear systems. Unfortunately, only a minute selection of handbook data (twelve evaluations) are of actual experimental facilities and mockups designed specifically for space nuclear research. There is a paucity problem, such that the multitude of space nuclear experimental activities performed in the past several decades have yet to be recovered and made available in such detail that the international community could benefit from these valuable historical research efforts. Those experiments represent extensive investments in infrastructure, expertise, and cost, as well as constitute significantly valuable resources of data supporting past, present, and future research activities. The ICSBEP and IRPhEP were established to identify and verify comprehensive sets of benchmark data; evaluate the data, including quantification of biases and uncertainties; compile the data and calculations in a standardized format; and formally document the effort into a single source of verified benchmark data. See full abstract in attached document.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Initial Application of TSUNAMI for Validation of Advanced Fuel Systems

A number of advanced reactor concepts and accident-tolerant fuels (ATFs) are under development or in the initial stages of deployment and testing in the nuclear power industry. Many of these concepts incorporate chemical or material forms that have not traditionally been used in the nuclear industry and may present challenges for nuclear criticality safety validation. Sensitivity/uncertainty methods are ideally suited for assessing the applicability of existing critical experiments to these potentially challenging applications. This paper presents two case studies recently completed at Oak Ridge National Laboratory (ORNL) which investigated the use of the TSUNAMI tools within the SCALE code system to perform such assessments.Many advanced reactor concepts include uranium enrichments above 5 wt% 235U but below the 20 wt% limit of low-enriched uranium (LEU). This material, referred to as high assay LEU (HA-LEU), poses difficulties across the entire infrastructure of nuclear power, including enrichment, transportation, fabrication, and storage. A potentially representative transportation package was selected to perform a criticality safety validation assessment for UF6 containing HA-LEU. Sensitivity data were generated using the TSUNAMI-3D sequence, and an assessment of the applicability of a range of critical experiments was performed using the TSUNAMI-IP tool. The critical experiments were drawn from the set of available sensitivity data files distributed with the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook. The results indicate that a sufficient number of available experiments may be available to perform validations for these systems.Many different cladding materials and fuel dopants are included in ATFs to mitigate the consequences of a loss of coolant accident. Some of these systems have already been fabricated and introduced as lead test assemblies in commercial reactors, while others are still being developed. Several concepts were considered for pressurized-water reactors (PWRs) and boiling-water reactors (BWRs) in comparison to the current Zircaloy and UO2 fuel system. In most of these cases, the ATF materials did not significantly impact the number of applicable experiments for validation. Some fuel and cladding materials were identified, however, which may reduce the number of available experiments and which will increase the nuclear data–induced uncertainty in neutron multiplication.

Marshall, William (B.J.)↗
DOE OSTI2019

General Handling Calculations for the Modified Hospital Can

Containers are evaluated using a Monte-Carlo N-Particle transport code (MCNP). The code calculated keff is compared to critical experiment benchmarks to determine a bias. Additional margin of safety is added to the bias to establish an upper subcritical limit (USL). Modeled cases are considered subcritical if the neutron multiplication factor plus two sigma remains below the USL. Net weight limits allow for any amount of fissile/moderator mixture to exist so long as it meets the weight requirement, therefore all mixture types must be evaluated. Cases that exceed the USL must either be controlled against or have justification why that configuration cannot reasonably exist. Current and legacy loadings and container dimensions must be taken into consideration.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2012

Design and producing of fine-group cross section library HENDL3.0/FG for subcritical system

To improve the accuracy of the neutron analyses for subcritical system with thermal fission blanket, a coupled neutron and photon (315 n + 42{gamma}) fine-group cross section library HENDL3.0/FG based on ENDF/B-VII, JEFF3.1 and JENDL3.3 was produced by FDS team. In order to test the availability and reliability of the HENDL3.0/FG data library, shielding and critical safety benchmarks were performed with VisualBUS code. The testing results indicated that the discrepancy between calculation and experimental values of nuclear parameters fell in a reasonable range. It showed that the nuclear data library had accuracy and availability. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Comparison Study of Upper Subcritical Limits Derived Using Sensitivity/Uncertainty Tools Case Studies of Benchmarks and Applications

Neutron transport methods used to establish subcriticality require validation by comparison to critical experiments considered to be benchmarks. Whisper is a sensitivity/uncertainty analysis tool developed to assist with the task of validation in nuclear criticality safety. Details on the Whisper methodology can be found in References 1-3 on the MCNP ® reference collection website at https://mcnp.lanl.gov. Whisper-1.0 was originally developed in 2014 and used to assist with nuclear criticality safety validation at Los Alamos National Laboratory. Whisper was upgraded in 2016 to Whisper-1.1 and prepared for release with MCNP6.2. Whisper contains a library of over 1100 critical experiment benchmarks and quantifies neutronic similarity of an application to benchmarks in the library. Using highest similarity benchmarks, Whisper computes a calculational margin (CM) encompassing of the worst-case bias and bias uncertainty at a 99% confidence level for each application. In addition, portions of the margin of subcriticality (MOS) for nuclear data uncertainty and potential code errors are computed. The baseline upper subcritical limit (USL) computed by Whisper is comprised of the CM, MOS nuclear data , and MOS code errors . The Whisper baseline USL is absent a portion of the MOS due to the area of application, which is applied based upon judgment by the criticality safety analyst. An objective of this paper is to present the baseline USL, CM and portions of the MOS as computed by Whisper for comparison with similar sensitivity/uncertainty tools, such as those used by IRSN and ORNL. The initial comparison involves four critical experiment benchmarks: HEU-MET-FAST-013-001, HEU-SOLTHERM-001-008, PU-MET-FAST-022-001, AND PU-SOL-THERM-001-001, which have been: 1. modeled independently by LANL, IRSN, and ORNL based upon information provided in the ICSBEP Handbook, 2. are common in S/U libraries for LANL, IRSN, and ORNL, 3. span a range of energy spectrum and fissile material, and 4. taken as applications for the purposes of this study and therefore excluded from use as a benchmark for calculating the upper subcritical limit. Results presented in this paper have been computed using covariance data for all isotopes in ENDF/BVII.0 using a 44-group energy structure. Benchmarks in the Whisper library were run in MCNP6.2 using 100,000 neutrons per cycle, skipping 100 cycles for 500 active cycles. Reference 10 also compares the results for baseline USL with an order of magnitude greater neutrons, using the same total number of cycles with 1,000,000 neutrons per cycle. Subsequent to the results presented in Reference 10 changes were made to the benchmark library, as discussed in Reference 11. Newer results using the revised benchmark library are presented herein.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2013

235U Resolved Resonance Evaluation for Benchmark Calculations in the Intermediate Energy Region

The Working Party on International Nuclear Data Evaluation Co-operation (WPEC) subgroup 29 (SG 29) was established to investigate an issue with the 235U capture cross-section in the energy range 0.1 to 2.25 keV. The WPEC criticality calculation results indicated an overestimation of the 235U capture cross-section of 10% or more. To understand and solve the problem, a recommendation was made to perform new capture cross-section measurements followed by a resonance evaluation. Hence, time of-flight capture cross-section measurements were done at the Rensselaer Polytechnic Institute (RPI) and at the Los Alamos National Laboratory (LANL). These new measurements were used together with the computer code SAMMY to reevaluate the 235U resonance parameters in the energy range from thermal to 2.25 keV. The impact of the new evaluation in benchmark calculations was done for the critical benchmark sensitive to the 0.1 to 2.25 keV energy range. The purpose of this work is to describe the 235U SAMMY evaluation and present the critical benchmark results.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2023

SCALE 6.2 Lattice Physics Performance Assessment

The US Nuclear Regulatory Commission relies on the lattice physics analysis capabilities of the SCALE code system to perform confirmatory licensing analyses. Either SCALE lattice physics code—TRITON/NEWT or Polaris—can be used to generate cross section data used by the PARCS nodal core simulator for full-core neutronics calculations. This report presents an assessment of the accuracy of SCALE lattice physics codes for preparation of lattice physics data that are used to support simulator codes such as the NRC’s PARCS, for UO 2 -mixed oxide (MOX)/Zr fueled light water reactor (LWR) analyses. Due to the nature of lattice physics calculations, critical reactor experiment benchmarks cannot be modeled in explicit detail in a lattice physics code. However, this limitation does not mean that these measurement data are not usable for lattice physics studies. Therefore, either geometry approximations or axial buckling must be implemented to determine the critical water height. These modeling limitations have led to development of a three-phase assessment strategy. In the first phase, selected critical experiment benchmarks are modeled using the SCALE 3D continuous-energy (CE) Monte Carlo (MC) code KENO, which is the most rigorous neutron transport method available in SCALE, with no approximations in the spatial, angle, or energy treatments. Biases and statistical uncertainties in quantities of interest such as k eff and pin power distributions are determined by comparing CE KENO results to experimental data. This first phase in the accuracy assessment is to demonstrate and establish the use of CE KENO as a reference solution for the second phase. In the second phase, 14 numerical test suites are used to compare SCALE lattice physics calculations with CE KENO as a reference solution. In the third phase, SCALE lattice physics depletion calculations are performed, and the spent fuel isotopic results are compared with available radiochemical assay measurements. Isotopic measurement comparisons provide quantitative assessment of isotopic density distribution predictions with the depletion models in SCALE lattice physics codes. This report documents results for all test suites. The assessment was performed using standard production techniques unless otherwise noted. Both TRITON/NEWT and Polaris exhibited acceptable accuracy for most test cases. For the few test cases in which acceptable accuracy criteria were not met, further code and data development are planned. The computer codes used in this assessment are as follows: SCALE 6.2 rev19189 (pre-release of 6.2.1) was used for CE KENO, TRITON/NEWT and Polaris (PWR only) calculations; SCALE 6.2.2 was used for Polaris calculations to address several updates in support of support boiling water reactor (BWR) lattice geometries. For this work, there is no difference between SCALE 6.2.1 and SCALE 6.2.2 beyond the inclusion of the Polaris BWR analysis capability; The ENDF/B-VII.1 continuous energy and 252 group libraries deployed in SCALE 6.2 were generated with AMPX 6.2 that is distributed with SCALE.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

Finalizing the {sup 63}Cu and {sup 65}Cu Resonance Evaluations for the ENDF/B-VIII Release

The resolved resonance region evaluation for the only two natural isotopes of copper ({sup 63}Cu and {sup 65}Cu) was carried out at the Oak Ridge National Laboratory (ORNL) in support of the US Department of Energy (DOE) Nuclear Criticality Safety Program (NCSP). The particular interest in the isotope of copper arose because it was linked with poor performance of the Zeus series of integral benchmarks from the International Criticality Safety Benchmark Evaluation Project (ICSBEP). Good agreement between the calculated and experimentally measured values for the ICSBEP benchmarks is essential to the NCSP because the ICSBEP benchmarks are used for validation purposes of criticality safety calculations. The particular set of four Zeus experiments had a highly-enriched uranium core with a thick copper-reflector resulting in a large sensitivity of the criticality of these systems to the cross section of copper in the intermediate energy range. Sobes et al. presented the effect of the newly resolved resonance region evaluations for the two isotopes of copper on the Zeus integral benchmarks, as well as 23 other benchmarks from the ICSBEP during the 2016 ANS Annual Meeting. The results presented in Ref. were calculated with the ENDF/B-VII.1 cross section libraries used for all other materials. However, it is suggested that the resolved resonance region evaluations for the isotopes of copper should not be judged only on their performance in the integral benchmark calculations especially when the ENDF/B-VII.1 cross sections are used for all of the other isotopes. Three reasons are provided for this. First and foremost, a new evaluation is an assessment of the physics of the nucleus of the particular isotope, and the goal of an application agnostic nuclear data evaluation is to improve evaluation by fitting state-of-the art R-Matrix formalism to highly precise experimental cross section measurements. The improved performance in the integral benchmarks should only come as a consequence of the improvement in the physics of the evaluation. Secondly, there is some uncertainty in the nuclear data evaluation of the cross section of all isotopes, so there could be some error in evaluating another isotope used in calculating the integral benchmark experiment. Such an error may have been having a compensating effect, in combination with the previous evaluations for the isotopes of copper. Thirdly, with the two isotopes of copper proposed for inclusion in the upcoming release of the ENDF/B-VIII library, performance of the new resonance evaluation for copper should be evaluated based on the calculation of the integral benchmarks with all of the evaluations proposed for inclusion in the ENDF/B-VIII release. These benchmarking results using the ENDF/B-VIII evaluations as they converge closer to their final form will be presented at the conference. This summary stresses the importance of the physics of the evaluation and reviews some of the steps taken in finalizing a resolved resonance region evaluation. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Employing Whisper in selecting benchmarks for traditional validation methods

The Nuclear Criticality Safety Program at Sandia National Laboratories (SNL) uses a collection of benchmarks developed from the International Handbook of Evaluated Criticality Safety Benchmark Experiments for validation of MCNP. In early 2016, an effort began to expand the benchmark collection to cover a broader range of materials and neutron spectra. The program undertook the task of updating current benchmarks to reflect best practices from Los Alamos National Laboratory (LANL) for MCNP calculations. Spring of 2016, LANL shared the Whisper code with SNL in a collaborated effort to investigate application of the code to activities at SNL. Whisper is a statistical package developed to support nuclear criticality safety validation efforts. It measures how well benchmarks correlate to a system or process and calculates a USL. Whisper has begun to develop into an asset for the SNL Nuclear Criticality Safety team to determine and reassure benchmarks selected for validation truly correlate process parameters (i.e., fall within an acceptable Area of Applicability). SNL's use of traditional statistical methods, with the abilities of Whisper to identify correlating benchmarks, has facilitated in ensuring confidence in validation and USL determination. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Sensitivities and correlations of critical experiments due to uncertainties of system parameters and nuclear data - 14416

With increasing demand on accuracy of criticality calculations, correlations between benchmark experiments due to system parameter uncertainties become of interest in bias determination. The impact of different interpretations of system parameter uncertainties on the calculation of covariances and correlations of k{sub eff} is discussed in this article. Following the suggestion of the OECD/NEA EGUACSA Benchmark phase IV [1], 21 experiments from the ICSBEP international Handbook of Evaluated Criticality Safety Benchmarks [2] are used for this analysis (LEU-COMP-THERM series 07, exp. 01-04 and 39, exp. 01-17). The experiments share certain assembly and fuel features and were carried out in the same lab. Therefore, correlations in the calculation of k{sub eff} appear. Correlation coefficients between experiments due to nuclear data uncertainties are calculated with TSUNAMI [3] and show high values for most experiments. For the correlation analysis due to system parameter uncertainties mainly two scenarios are considered: In scenario A all uncertain parameters are varied mutually for all fuel rods, representing a systematic deviation of the actual experimental setup from the design value. This analysis is performed for all experiments listed above. In scenario E certain parameters like the outer cladding diameter are varied for each fuel rod individually, representing a random scattering of the varied parameter. Scenarios B, C and D represent intermediate scenarios between A and E, which are discussed shortly. Uncertainty and sensitivity calculations of k{sub eff} on the uncertain experimental parameters based on Monte Carlo sampling are presented and discussed. The calculations were performed with SUnCISTT (Sensitivities and Uncertainties in Criticality Inventory and Source Term Tool [4]) and SCALE 6.1.2 [3]. For the 21 EGUACSA proposed experiments the outer cladding diameter is the leading effect for the uncertainty of k{sub eff} for scenario A. In contrast, for scenario E the impact of the cladding diameter on k{sub eff} cancels out due to rod wise variations leaving the fuel diameter as the leading effect. The correlation analysis based on system parameter uncertainties indicates a significant dependence of the correlation coefficients on the interpretation of the uncertainties. It shows high values for scenario A, if the neutronic features are similar (EALF, H/U). For scenario E the overall correlations are significantly smaller and variations between experiments decrease. The impact of the determined correlations on bias calculations of application cases will be discussed by the authors in the accompanying paper 'Impact of correlated data in validation procedures'. (authors)

42 ENGINEERING↗
DOE OSTI2022

Updates and Validation for the n+ 63,65 Cu Cross Sections [Abstract]

The neutron induced total, elastic, and capture cross sections of 63,65 Cu isotopes were selected for evaluation in the resolved and unresolved resonance energy ranges by the National Criticality Safety Program to resolve discrepancies related to benchmark performance. This is especially evident for the series of ZEUS benchmarks in which copper is used as a reflector. Because copper is also used as structural material in both fission and fusion reactors, the need to address benchmark discrepancies linked to nuclear data deficiencies is a task of primary importance. The aim of this work is to describe the steps of evaluation work towards a consistent improvement of the benchmark performance. The R-matrix analysis with the SAMMY code focused on the 63 Cu(n,γ) reaction channel between 100-300 keV coupled to unresolved resonance region parameters up to 650 keV to fit average cross section data from a recent experiment. Due to the high sensitivity of many benchmarks to elastic scattering angular distribution data, especially for the 65 Cu isotope, the impact of these data was tested by generating Legendre coefficients from both resonance parameters and the Hauser-Feshbach model. Guided by the findings of Shaw et al., the performance of the current evaluation for 65 Cu was compared to that of ENDF/B-VII.1 and ENDF/B-VIII.0 by testing the reactivity coefficients corresponding to the validation suite of experimental criticality benchmarks for thermal, intermediate, and fast systems taken from the International Criticality Safety Benchmark Experiments Project Handbook. The benchmark performance is especially sensitive to 63 Cu(n,γ) and 65 Cu elastic scattering for neutron energies in the 100–500 keV region, whereas 100 keV is the upper limit of the resolved resonance region in the ENDF/B-VIII.0 evaluations for 63,65 Cu. The results highlight the need to handle the transition from the resolved resonance region to the high energy region carefully.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2022

Delayed Critical Enriched Uranium Metal, 7-in.-diam. Cylinder with Thin Stainless Steel Top and Bottom Reflectors

Three 7 in. diameter highly enriched uranium (HEU; 93.17 wt % 235 U) metal cylinders were assembled on the vertical assembly machine of the Oak Ridge Critical Experiments Facility (ORCEF) with thin stainless steel reflectors (0.5, 1, and 2 in. thick) on the top and bottom. These experiments were performed during January 13–25, 1965, and used seven operational days at ORCEF. Before these experiments, unreflected and unmoderated 7 in. diameter HEU metal cylinders had been assembled to delayed criticality at ORCEF, and the results were benchmarked in HEU-MET-FAST-051. In addition to the critical experiments, the prompt neutron decay constant was measured via the Rossi-α technique, and those results are also presented in this report. To achieve near-delayed-critical systems, 5 × 5 × 0.03125 in. rectangular uranium metal parts were used on top of 7 in. diameter cylinders inside the top reflector. The delayed critical configuration from the three experiments described herein are acceptable for use as criticality safety benchmark experiments for the International Criticality Safety Benchmark Evaluation Program (ICSBEP), coordinated by the Nuclear Energy Agency, once the uncertainty analysis is completed. Based on previous ICSBEP benchmarks with this enriched uranium metal at ORCEF, the uncertainty in $k_{eff}$ is expected to be as low as ±0.0002. The prompt neutron time decay data could be the basis of an International Reactor Physics Evaluation Program benchmark.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2013

The ORSphere Benchmark Evaluation and Its Potential Impact on Nuclear Criticality Safety

In the early 1970’s, critical experiments using an unreflected metal sphere of highly enriched uranium (HEU) were performed with the focus to provide a "very accurate description…as an ideal benchmark for calculational methods and cross-section data files." Two near-critical configurations of the Oak Ridge Sphere (ORSphere) were evaluated as acceptable benchmark experiments for inclusion in the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook). The results from those benchmark experiments were then compared with additional unmoderated and unreflected HEU metal benchmark experiment configurations currently found in the ICSBEP Handbook. For basic geometries (spheres, cylinders, and slabs) the eigenvalues calculated using MCNP5 and ENDF/B-VII.0 were within 3 of their respective benchmark values. There appears to be generally a good agreement between calculated and benchmark values for spherical and slab geometry systems. Cylindrical geometry configurations tended to calculate low, including more complex bare HEU metal systems containing cylinders. The ORSphere experiments do not calculate within their 1σ uncertainty and there is a possibility that the effect of the measured uncertainties for the GODIVA I benchmark may need reevaluated. There is significant scatter in the calculations for the highly correlated ORCEF cylinder experiments, which are constructed from close-fitting HEU discs and annuli. Selection of a nuclear data library can have a larger impact on calculated eigenvalue results than the variation found within calculations of a given experimental series, such as the ORCEF cylinders, using a single nuclear data set.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Chlorine Worth Study Nuclear Data

The Chlorine Worth Study (CWS) was a series of experiments that took place at the National Criticality Experiments Research Center (NCERC), operated by Los Alamos National Laboratory (LANL). The focus of the experiments was to develop new integral benchmarks for the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook with high sensitivity to chlorine in the thermal neutron energy region and which match sensitivities of aqueous chloride operations at LANL. This work discusses the experiment and how the experimental and simulated results using different nuclear data libraries compare to each other.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗