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At least 577 records · Page 32

IER 488: MUSiC Accomplishments and Goals [Slides]

IER-488: Measurement of Uranium Subcritical and Critical (MUSiC) consists of measurements of bare, HEU systems covering a large range of reactivities. Multiple detector systems and analysis methods are being used to obtain detector-independent results. Massive amounts of data and information available. Substantial analysis has been completed for both critical and subcritical configurations. Work in progress is largely novel research and will result in several conference proceedings, journal publications, and a subcritical benchmark. There is still a large amount of data that has not been analyzed and much more to come from the analysis of the Measurements of Uranium Subcritical and Critical.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Implementation, capabilities, and benchmarking of Shift, a massively parallel Monte Carlo radiation transport code

This paper discusses the implementation, capabilities, and validation of Shift, a massively parallel Monte Carlo radiation transport package developed and maintained at Oak Ridge National Laboratory. It has been developed to scale well from laptop to small computing clusters to advanced supercomputers. Special features of Shift include hybrid capabilities for variance reduction such as CADIS and FW-CADIS, and advanced parallel decomposition and tally methods optimized for scalability on supercomputing architectures. Shift has been validated and verified against various reactor physics benchmarks and compares well to other state-of-the-art Monte Carlo radiation transport codes such as MCNP5, CE KENO-VI, and OpenMC. Some specific benchmarks used for verification and validation include the CASL VERA criticality test suite and several Westinghouse AP1000 ® problems. These benchmark and scaling studies show promising results.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Results of TRIPOLI-4 version 10 on the TRIPOLI-4 criticality validation suite, thermal spectra

This paper presents the results of the latest version (version 10) of TRIPOLI-4 on its criticality validation suite, thermal spectra part. The criticality validation suite is made of 148 selected ICSBEP benchmarks with various enrichments, fissile materials and spectra. The ICSBEP benchmarks computational models are prepared by the TRIPOLI-4 team and are run in the same conditions for each benchmark: initial source, number of neutrons per cycle, number of inactive and active cycles, k{sub eff} estimator and nuclear data. Comparisons are made with experimental k{sub eff}. On one case (PST-009), we have noted a discrepancy between the computational results and the experiments. This discrepancy is also found with other codes, which might indicate a possible issue with the benchmark. For all the other considered cases, uranium and plutonium solutions and fuel-rods-arrays, the agreement with the experiments is very good.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Grounding our Understanding of the Impacts of Boreal Forest Expansion on Shallow Cumulus Clouds with a Simple Modeling Framework

Abstract The expansion of the boreal forest poleward is a potentially important driver of feedbacks between the land surface and Arctic climate. A growing body of work has highlighted the importance of differences in evaporative resistance between different possible future Arctic land covers, which in turn alters humidity and cloudiness in the boundary layer, for these feedbacks. While thus far this problem has been studied primarily with complex Earth system models, we turn to a locally focused, idealized model capable of diagnosing and testing the sensitivity of first-order processes connecting vegetation, the atmospheric boundary layer, and low clouds in this critical region. This allows us to benchmark the mechanisms and results at the center of predictions from larger-scale simulations. A surface dominated by broadleaf trees, characterized by higher albedo and lower surface evaporative resistance, drives cooling and moistening of the boundary layer relative to a surface of needleleaf trees, characterized by lower albedo and higher surface evaporative resistance. Differences in evaporative resistance between these hypothetical Arctic vegetation covers are of equal importance to changes in albedo for the initial response of the boundary layer to boreal expansion, even with our idealized approach. However, compensation between the elevation of the lifting condensation level (LCL) and more rapid growth of the mixed layer over higher evaporative resistance surfaces can minimize changes in the favorability of shallow clouds over different land cover types under some conditions. We then perform two tests on the sensitivity of this compensating effect, to changes in water availability, represented first by a reduction in boundary layer humidity and then by both a reduction in humidity and soil moisture available to our vegetation surface. Finally, given the importance of this potential LCL–mixed-layer height compensation in our idealized modeling results, we look to determine its relevance in observational data from a field campaign in boreal Finland. These observations do confirm that such a coupling plays an important role in cumulus-topped boundary layers over a needleleaf forest surface. While our results confirm some underlying mechanisms at the center of prior work with Earth system models, they also provide motivation for future work to constrain the impact of boreal forest expansion. This will include both large eddy simulations to examine the impact of processes and feedbacks not resolved by a mixed-layer model, as well as a more systematic evaluation and comparison of relevant observations at the site in Finland and sites from prior boreal field campaigns. Significance Statement Clouds and vegetation are both important components of the climate system that interact across a range of scales. These interactions are central to understanding how changes at the land surface feedback on climate. For example, if a forest expands or recedes, diagnosing how that will impact clouds will determine whether you predict warming or cooling temperatures from that shift in the forest area. These predictions are often made with complex Earth system models, but we look to a more idealized representation of the land–atmosphere system to diagnose how shallow clouds should respond to changes in surface properties with different scenarios of boreal forest expansion at a more foundational level. This both grounds our understanding of previous analysis and provides helpful direction for future studies of this relevant and impactful land cover change.

Meteorology & Atmospheric Sciences↗

SFCOMPO Database of Spent Nuclear Fuel Assay Data – the Next Frontier

SFCOMPO is the world’s largest database for measured spent nuclear fuel assay data. An international effort coordinated by the Nuclear Energy Agency (NEA) resulted in a significant expansion of the database and its release online in 2017 as a downloadable application. The SFCOMPO Technical Review Group (TRG) was recently formed under the direction of NEA’s Nuclear Science Committee/Working Party on Nuclear Criticality Safety and was mandated to maintain and further coordinate the development of SFCOMPO. This TRG is currently focused on (1) critical evaluation of the experimental assay data by independent experts and (2) development of benchmarks and benchmark models that can be applied to validate burnup codes. This will improve the quality and documentation of the experimental datasets and enable their use by the international community to support code validation for design and safety analysis of spent nuclear fuel transportation, storage, and repository applications. It follows the precedent and draws on the experience gained from similar NEA efforts in the International Reactor Physics Experiment Evaluation Project and the International Criticality Safety Benchmark Experiment Project. Ongoing SFCOMPO evaluations have served as a test bed to develop templates for documenting evaluations, develop review guidance, improve approaches for a global uncertainty analysis, and devise a strategy focused on providing practical information of highest value to the user community. The current effort, status, and associated challenges are discussed.

Ilas, Germina↗

High Multiplication Neutron Noise Measurements Using the 7uPCX Assembly

The Seven Percent Critical Experiment, or 7uPCX, is a system that mimics the physics of light water nuclear reactor systems by using uranium dioxide fuel pins at an enrichment of approximately 7%. 7uPCX is often used for benchmarking and nuclear data purposes. As part of a collaboration between Lawrence Livermore National Laboratory, Los Alamos National Laboratory, Sandia National Laboratories, and the Institut de Radioprotection et de Sûreté Nucléaire, measurements were performed in support of a high-multiplication, subcritical benchmark candidate for a thermal system. The measurements were completed by making subcritical configurations using the fuel loading pattern of a well-documented configuration of the Seven Percent Critical Experiment at Sandia National Laboratories, which exists as a benchmark in the International Criticality Safety Benchmark Evaluation Project handbook. These measurements, which aimed to capture multiplications ranging from approximately 10 to 1,000, also serve as an intercomparison between both the fielded detector systems and analysis methodologies with the goal of better characterizing the detectors and their ability to capture the state of the criticality these types of systems. Los Alamos National Laboratory’s measurements for this collaboration were made with five linked helium-3 based neutron multiplicity detectors placed on the periphery of the reactor tank beyond the infinite reflector thickness of water, and four organic scintillators placed in dry-wells in-reactor near the edge of the upper fuel grid plate. This work captures the Los Alamos National Laboratory measurements, the Rossi-α and Feynman-Y results, provides an intercomparison between the results of the 3 He neutron detectors and the organic scintillators, compares to simulation where possible, and expands on future work.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

EBR-II Reactor Physics Benchmark Evaluation Report

This report provides a reactor physics benchmark evaluation with associated uncertainty quantification for the critical configuration of the April 1986 Experimental Breeder Reactor II Run 138B core configuration.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Multiple Critical Unresolved Region Integral Experiment (MCURIE): A Proposed Integral Critical Experimental Framework for Unresolved Region and Intermediate Energies

The Critical Unresolved Region Integral Experiment (CURIE) critical experiment was performed at the National Criticality Experiments Research Center (NCERC) at the Device Assembly Facility (DAF) at the Nevada Nuclear Security Sites (NNSS) in 2020. The objective of CURIE was to improve the quality of integral nuclear data in the uranium-235 ( 235 U) unresolved resonance region (URR) by performing benchmark integral experiments that were sensitive to the URR energy ranges. The CURIE experiment was evaluated for the International Criticality Safety Benchmark Evaluation Project (ICSBEP) handbook and the benchmark evaluation was accepted in 2022. The URR is a region within the intermediate neutron energy range (the intermediate energy ranges from 0.7 eV to 100 keV). The observed resonance structure in neutron cross sections is due to discrete energy levels in the nucleus and are characterized by resonance parameters. In the URR region, the resonance parameters are only partially resolved as the resolution of the experiment techniques becomes comparable to the average width of the resonances themselves, and the resonances are so close to one another that the structure cannot be determined empirically. The ENDF/B-VIII.0 and JEFF-3.3 nuclear data libraries define the URR as beginning at 2.25 keV and continuing until 25 keV. There are minimal intermediate neutron energy benchmarks available in the ICSBEP benchmark handbook, and aside from CURIE there are none that are highly sensitive in the URR energy region. There is a current need for additional experiments sensitive to the URR energy region. There is a new proposed subgroup for the Organization for Economic Co-operation and Development and Nuclear Energy Agency (OECD NEA) Working Party on International Nuclear Data Evaluation Cooperation (WPEC), so any new models or information will need experimental validation and testing. The NEA Working Party on Nuclear Criticality Safety (WPNCS) recent experimental needs and priority list includes intermediate energy 235 U and 238 U experiments, as does the recent Integral Experiments to Address Nuclear Criticality Safety Needs Meeting (May 2023). A variant on the original CURIE experiment called the Multiple Critical Unresolved Region Integral Experiment (MCURIE) is proposed to provide further investigation of the intermediate and URR energy region for uranium. MCURIE will utilize existing fuels at NCERC but use alternative moderators and reflectors to modify the neutron absorption, scattering, and fission spectra of the experiments, allowing for precise targeting of nuclear data sensitivities. The overall goal of MCURIE is to develop a framework for designing and performing integral benchmark experiments with high sensitivities in the intermediate and URR energy regions for nuclear data validation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Verification of MCNP5-1.60 and MCNP6-Beta-2 for Criticality Safety Applications

To verify that both MCNP5-1.60 and MCNP6-Beta-2 are performing correctly for criticality safety applications, several suites of verification/validation benchmark problems were run in early 2012. Results from these benchmark suites were compared with results from previously verified versions of MCNP5. The goals of this verification testing were: (1) Verify that MCNP5-1.60 works correctly for nuclear criticality safety applications, producing the same results as for the previous verification performed in 2010; (2) Determine the sensitivity to computer roundoff using different Fortran-90 compilers for building MCNP5 and MCNP6, to support moving to current versions of the compilers; and (3) Verify that MCNP6-Beta-2 works correctly for nuclear criticality safety applications, producing the same results as for MCNP5-1.60. This provides support for eventual migration of users and applications to MCNP6. The current production version of MCNP5 included in the RSICC release package is MCNP5-1.60. This version was first distributed by RSICC in October 2010. While there were subsequent RSICC distributions of the MCNP package in July 2011 and February 2012, no changes were made to MCNP5-1.60. The RSICC release package in February 2012 included both MCNP5-1.60 and the current beta version of MCNP6, MCNP6-Beta-2. MCNP6 is the merger of MCNP5 and MCNPX capabilities. The current release of MCNP6 available from RSICC as of February 2012 is MCNP6-Beta-2. This version includes all of the features for criticality safety calculations that are available in MCNP5-1.60, and many new features largely unrelated to nuclear criticality safety calculations. This release is a 'beta' release to allow intermediate and advanced users to begin testing the merged code in their field of expertise. It should not be used for production calculations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

LANL-SNL Collaboration on NCS Validation.

During 2016, nuclear criticality safety (NCS) practitioners from SNL and code developers from LANL collaborated in several areas of interest to the DOE/NNSA Nuclear Criticality Safety Program (NCSP). This collaboration involved. Testing of the preliminary release of the MCNP6- Whisper methodology, with feedback to the developers, Sharing of the benchmark catalogs (the collection of MCNP input files and benchmark results), with 1101 cases from LANL and 866 cases from SNL, Comparison and analysis of 357 benchmarks common to both catalogs, Investigation of the impact of the different benchmark catalogs on sensitivity-uncertainty based NCS validation results from MCNP6-Whisper, Investigation of the impact of randomized selections from the benchmark catalog on sensitivity-uncertainty based validation results from MCNP6-Whisper. Investigation of the use of MCNP6-Whisper in selecting benchmarks for use in NCS validation for unique, nonstandard, legacy fuel applications. This paper summarizes the collaboration work and initial results. It must be noted that the results described herein are preliminary and need further research and detailed analysis. However, the initial results are very interesting, and it is important to share them with the NCSP community.

97 MATHEMATICS AND COMPUTING↗

LANL-SNL Collaboration on NCS Validation

During 2016, nuclear criticality safety (NCS) practitioners from SNL and code developers from LANL collaborated in several areas of interest to the DOE/NNSA Nuclear Criticality Safety Program (NCSP). This collaboration involved: - Testing of the preliminary release of the MCNP6- Whisper methodology, with feedback to the developers, - Sharing of the benchmark catalogs (the collection of MCNP input files and benchmark results), with 1101 cases from LANL and 866 cases from SNL, - Comparison and analysis of 357 benchmarks common to both catalogs, - Investigation of the impact of the different benchmark catalogs on sensitivity-uncertainty based NCS validation results from MCNP6-Whisper, - Investigation of the impact of randomized selections from the benchmark catalog on sensitivity-uncertainty based validation results from MCNP6-Whisper. - Investigation of the use of MCNP6-Whisper in selecting benchmarks for use in NCS validation for unique, nonstandard, legacy fuel applications. This paper summarizes the collaboration work and initial results. It must be noted that the results described herein are preliminary and need further research and detailed analysis. However, the initial results are very interesting, and it is important to share them with the NCSP community.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Polymer Compositions in Critical Experiments: Possibly Not What You Think

The Chlorine Worth Study (CWS) Experiment was performed at the National Criticality Experiments Research Center (NCERC) in December 2021. Its goal was to provide a validation benchmark experiment with chlorine and plutonium in the thermal neutron spectrum. This purpose is necessary to reduce the margin of subcriticality in aqueous chloride operations at the Los Alamos National Laboratory Plutonium Facility (PF-4). Reducing the margin of subcriticality will enable higher throughput, required to meet NNSA mission needs. The experiment used layers of plutonium plates, polyethylene (HDPE), aluminum, and polyvinyl chloride (PVC) or chlorinated polyvinyl chloride (CPVC). They were optimized to match plutonium-chloride solutions of 30 g/L, 300 g/L, and 600 g/L. Upon completion of the experiment, a International Criticality Safety Benchmark Evaluation Program (ICSBEP) report was immediately started. The compositions of all materials were assumed pure unless additional information was known (such as for the plutonium plates). During the benchmark analysis, the assumed CPVC composition was questioned. The follow-on work led to lessons learned on compositions in benchmarks. Materials, and specifically polymers, are often much more complex than a basic chemical formula. The CWS experiment is used as an example in this paper to document the lessons learned.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Nuclear data uncertainty and sensitivity analysis of the VHTRC benchmark using SCALE

The Very High Temperature Critical Assembly (VHTRC) experiment represents one of the few data sets available for the validation of HTGR lattice physics. Within the framework of the IAEA Coordinated Research Project on HTGR Uncertainty Analysis in Modeling, uncertainty and sensitivity analyses of this graphite-moderated facility are performed as the validation reference to the prismatic MHTGR-350 fuel block calculations. Nominal multi-group and continuous-energy criticality calculations with the KENO-VI Monte Carlo code of the SCALE code package are compared with the continuous-energy Monte Carlo Code Serpent. Good agreement with a maximum difference of 250 pcm is obtained. The experimental data set, however, differs by several hundred pcm with the results of both codes when using the ENDF-VII.0 library. When using ENDF/B-VII.1 data, this difference is reduced to a few hundred pcm and the calculations lie within the experimental error bars. Uncertainties of the VHTRC multiplication factors due to uncertainties in nuclear data are determined with SAMPLER/KENO-VI of SCALE 6.2 and with continuous-energy TSUNAMI of SCALE 6.2. For all experimental configurations, the obtained uncertainty is found to be 0.58% when using ENDF/B-VII.0 data. The top contributor to this uncertainty is the average number of neutrons per fission event of U-235. With ENDF/B-VII.1 data, uncertainties of about 0.66% are obtained in particular due to an increased uncertainty of U-235 nu-bar in the latest ENDF release. In conclusion, when considering these nuclear data uncertainties, the obtained 1s uncertainty intervals are overlapping with the experimental error bars.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Estimation of uncertainties for subcritical benchmark measurements - 14387

Subcritical measurements are used for many applications including nonproliferation, treaty verification, in-situ monitoring for criticality, and reactor subcriticality monitoring. Most analysis methods involve correlated neutron detection, based upon the property that fission events can create multiple neutrons that are born at nearly the same time (on a scale of 10{sup -13} or 10{sup -14} seconds). Many different time-correlated methods have been used since the 1950's and are still widely utilized today. This particular work describes an uncertainty analysis of measured data using the Hage-Cifarelli formalism of the Feynman Variance-to-Mean method. The Feynman Variance-to-Mean method was introduced by Feynman et al. in 1956 and is based upon the differences between the detector count data and the data that would be expected from measurements taken from a Poissonian source. The Hage-Cifarelli formalism relates moments of the Feynman histograms to several parameters of a system: leakage multiplication (M{sub L}), spontaneous fission rate (F{sub S}), (α,n) neutron emission rate (S{sub α}), detector efficiency (ε), and the moments of the number of neutrons emitted per fission (ν). The first part of the uncertainty method described in this work will show how to determine uncertainties of the first (R{sub 1}) and second (R{sub 2}) reduced factorial moments of the Feynman histograms. The second part of the method will describe how to determine the uncertainties in the other parameters given uncertainties in the reduced factorial moments; in particular, this work will focus on determining the uncertainties in M{sub L} and F{sub S}. This work will utilize measured data from a recent benchmark evaluation with a 4.5 kg sphere of α-phase plutonium to validate the uncertainty analysis. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

Hardware functional obfuscation with ferroelectric active interconnects

Existing circuit camouflaging techniques to prevent reverse engineering increase circuit-complexity with significant area, energy, and delay penalty. In this paper, we propose an efficient hardware encryption technique with minimal complexity and overheads based on ferroelectric field-effect transistor (FeFET) active interconnects. By utilizing the threshold voltage programmability of the FeFETs, run-time reconfigurable inverter-buffer logic, utilizing two FeFETs and an inverter, is enabled. Judicious placement of the proposed logic makes it act as a hardware encryption key and enable encoding and decoding of the functional output without affecting the critical path timing delay. Additionally, a peripheral programming scheme for reconfigurable logic by reusing the existing scan chain logic is proposed, obviating the need for specialized programming logic and circuitry for keybit distribution. Our analysis shows an average encryption probability of 97.43% with an increase of 2.24%/ 3.67% delay for the most critical path/ sum of 100 critical paths delay for ISCAS85 benchmarks.

42 ENGINEERING↗

MC 2 -3: Multigroup Cross Section Generation Code for Fast Reactor Analysis

This paper presents the methods and performance of the MC2 -3 code, which is a multigroup cross-section generation code for fast reactor analysis, developed to improve the resonance self-shielding and spectrum calculation methods of MC2 -2 and to simplify the current multistep schemes generating region-dependent broad-group cross sections. Using the basic neutron data from ENDF/B data files, MC2 -3 solves the consistent P1 multigroup transport equation to determine the fundamental mode spectra for use in generating multigroup neutron cross sections. A homogeneous medium or a heterogeneous slab or cylindrical unit cell problem is solved in ultrafine (2082) or hyperfine (~400 000) group levels. In the resolved resonance range, pointwise cross sections are reconstructed with Doppler broadening at specified temperatures. The pointwise cross sections are directly used in the hyperfine group calculation, whereas for the ultrafine group calculation, self-shielded cross sections are prepared by numerical integration of the pointwise cross sections based upon the narrow resonance approximation. For both the hyperfine and ultrafine group calculations, unresolved resonances are self-shielded using the analytic resonance integral method. The ultrafine group calculation can also be performed for a two-dimensional whole-core problem to generate region-dependent broad-group cross sections. Verification tests have been performed using the benchmark problems for various fast critical experiments including Los Alamos National Laboratory critical assemblies; Zero-Power Reactor, Zero-Power Physics Reactor, and Bundesamt für Strahlenschutz experiments; Monju start-up core; and Advanced Burner Test Reactor. Verification and validation results with ENDF/B-VII.0 data indicated that eigenvalues from MC2 -3/DIF3D agreed well with Monte Carlo N-Particle5 MCNP5 or VIM Monte Carlo solutions within 200 pcm and regionwise one-group fluxes were in good agreement with Monte Carlo solutions.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗