Traineeship Talks — Structure data evaluation at LLNL
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Physics - Nuclear physics and radiation physics↗
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ENDF/B-VIII.0: THE 8th Major Release of the ENDF/B Library Containing Cielo-Project Evaluations, New Standards and Thermal Scattering Data
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73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Nuclear data uncertainty quantification for the nuclide inventory of a Calvert Cliffs spent fuel sample
The impact of nuclear data cross section uncertainties and covariance matrices on the nuclide vector of spent nuclear fuel was investigated; This exercise was carried out for the Calvert Cliffs fuel assembly D047 benchmark available in the SFCOMPO database. Sample P irradiated in rod MKP109 for 4 cycles up to a burnup of approximately 44 GWd/MTU was selected for the analysis. Nuclear data uncertainties were taken from the most recent libraries released by evaluation projects JEFF, ENDF/B and JENDL, and were propagated using the SANDY code via a stochastic sampling approach. This paper provides a quantification of the uncertainty on the concentration of several actinides and fission products relevant for spent fuel management. Uncertainties generally below 5 % were predicted for the concentrations of most of the uranium, neptunium and plutonium isotopes relevant for SNF applications. Curium isotopes carry larger uncertainties that might exceed 10 %. The contribution of cross section uncertainties on the concentrations of fission products was found to be marginal with the exception of a few nuclides. These results can be significantly affected by the lack of evaluated covariance matrices for the capture cross section of several fission products. Burnup tracers such as {sup 148}Nd and {sup 137}Cs have negligible uncertainties because of the power normalisation imposed in every stochastic calculation.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Multi-step Monte Carlo calculations applied to nuclear reactor instrumentation - source definition and renormalization to physical values
Significant efforts have been made over the last few years in the French Alternative Energies and Atomic Energy Commission (CEA) to adopt multi-step Monte Carlo calculation schemes in the investigation and interpretation of the response of nuclear reactor instrumentation detectors (e.g. miniature ionization chambers - MICs and self-powered neutron or gamma detectors - SPNDs and SPGDs). The first step consists of the calculation of the primary data, i.e. evaluation of the neutron and gamma flux levels and spectra in the environment where the detector is located, using a computational model of the complete nuclear reactor core and its surroundings. These data are subsequently used to define sources for the following calculation steps, in which only a model of the detector under investigation is used. This approach enables calculations with satisfactory statistical uncertainties (of the order of a few %) within regions which are very small in size (the typical volume of which is of the order of 1 mm{sup 3}). The main drawback of a calculation scheme as described above is that perturbation effects on the radiation conditions caused by the detectors themselves are not taken into account. Depending on the detector, the nuclear reactor and the irradiation position, the perturbation in the neutron flux as primary data may reach 10 to 20%. A further issue is whether the model used in the second step calculations yields physically representative results. This is generally not the case, as significant deviations may arise, depending on the source definition. In particular, as presented in the paper, the injudicious use of special options aimed at increasing the computation efficiency (e.g. reflective boundary conditions) may introduce unphysical bias in the calculated flux levels and distortions in the spectral shapes. This paper presents examples of the issues described above related to a case study on the interpretation of the signal from different types of SPNDs, which were recently irradiated in the Jozef Stefan Institute TRIGA Mark II reactor in Ljubljana, Slovenia, and provides recommendations on how they can be overcome. The paper concludes with a discussion on the renormalization of the results from the second step calculations, to obtain accurate physical values. (authors)
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
Neutron Leakage Spectra Sensitivities for ICSBEP Benchmarks [Slides]
Abstract not provided.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Bare, Complex, and Potassium-Filled HEU Oralloy Annuli Benchmark Experiments
Hundreds of critical experiments were constructed from Oak Ridge Alloy (oralloy) during the 1960s and 1970s at the Oak Ridge Critical Experiments Facility (ORCEF) in support of criticality safety operations at the Y-12 Plant. The purpose of these experiments included the evaluation of storage, casting, and handling limits and providing data for verification of calculation methods and cross sections for nuclear criticality safety applications. Recent contributions to the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook) include bare, complex, and potassium-filled annuli comprising multiple stacked units of highly enriched uranium (HEU), ~93.2 wt.% 235U, annuli, discs, and parallelepipeds. Prior contributions to the ICSBEP Handbook include bare (unreflected and unmoderated) cylinders; configurations of cylinders and annuli reflected with polyethylene, beryllium, or thin-graphite reflectors; GROTESQUE; and ORSphere. The newest contributions include a bare HEU 11-7 (11 inch outer diameter and 7 inch inner diameter) annulus capped at the top and bottom; bare HEU 13-7, 15-7, and 15-9 annuli; bare HEU complex annuli consisting of a 15-9 annulus containing either an off-center 7-inch-diameter cylinder, 5 × 5 inch parallelepiped, or split parallelepiped; and an HEU 13-7 annulus containing either empty stainless steel cans or potassium-metal-filled cans. The purpose of this latter pair of annulus configurations was to test the fast neutron cross sections of potassium as it was a candidate for coolant in some early space power reactor designs. A total of nine new configurations were evaluated as acceptable benchmark experiment measurements. Additional ORCEF oralloy experiments have also been identified for future benchmark evaluation.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Bare, complex, and potassium-filled HEU oralloy annuli Benchmark experiments - 14103
Hundreds of critical experiments were constructed from Oak Ridge Alloy (oralloy) during the 1960's and 1970's at the Oak Ridge Critical Experiments Facility (ORCEF) in support of criticality safety operations at the Y-12 Plant. The purpose of these experiments included the evaluation of storage, casting, and handling limits and providing data for verification of calculation methods and cross sections for nuclear criticality safety applications. Recent contributions to the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook) include bare, complex, and potassium-filled annuli comprising multiple stacked units of highly enriched uranium (HEU), ∼93.2 wt.% {sup 235}U, annuli, discs, and parallelepipeds. Prior contributions to the ICSBEP Handbook include bare (unreflected and unmoderated) cylinders; configurations of cylinders and annuli reflected with polyethylene, beryllium, or thin-graphite reflectors; GROTESQUE; and ORSphere. The newest contributions include a bare HEU 11-7 (11 inch outer diameter and 7 inch inner diameter) annulus oralloy-capped at the top and bottom; bare HEU 13-7, 15-7, and 15-9 annuli; bare HEU complex annuli consisting of a 15-9 annulus containing either an off-center 7-inch-diameter cylinder, 5 x 5 inch parallelepiped, or split parallelepiped; and an HEU 13-7 annulus containing either empty stainless steel cans or potassium-metal-filled cans. The purpose of this latter pair of annulus configurations was to test the fast neutron cross sections of potassium as it was a candidate for coolant in some early space power reactor designs. A total of nine new configurations were evaluated as acceptable benchmark experiment measurements. Additional ORCEF oralloy experiments have also been identified for future benchmark evaluation. (authors)
36 MATERIALS SCIENCE↗
A Discrete Hankel Transform Approach to Nuclear Data Processing for Fusion Applications
This study introduces advancements to the numerical solutions employed in the processing of nuclear data for fusion applications. It leverages the convolution theorem and Fourier transform techniques to enhance computational efficiency and broaden applicability. Building upon a previously reported discrete Hankel transform approach for Doppler broadening, this work refines the solution of convolution integrals central to these applications. The methodology provides a general and unified framework for evaluating any convolution operation, regardless of whether the underlying problem involves temperature effects in nuclear reactions. The applicability to the nuclear data processing for fusion is demonstrated by deriving the convolution integrals for some of the fusion-related quantities. As before, the convolution operation utilizes a Gaussian-based kernel; however, the discrete Hankel transform of order $𝛼$ = $\frac{1}{2}$ is now applied to the forward Fourier transform of the nonkernel argument, rather than the inverse Fourier transform. This modification eliminates the need for the integration of the nonkernel, cross section–based function, which is a step that posed challenges for certain pointwise cross-section representations. It also removes the requirement for cross-section linearization. Optimized for graphics processing unit architectures, the approach significantly improves computational performance. These advancements are currently under evaluation as the foundation for the next-generation thermonuclear data file processing codes being developed at Lawrence Livermore National Laboratory.
Nuclear science and engineering↗
Evaluation of Am–Li neutron spectra data for active well type neutron multiplicity measurements of uranium
Safeguarding nuclear material is an important and challenging task for the international community. One particular safeguards technique commonly used for uranium assay is active neutron correlation counting. This technique involves irradiating unused uranium with ( α,n) neutrons from an Am-Li source and recording the resultant neutron pulse signal which includes induced fission neutrons. Although this non-destructive technique is widely employed in safeguards applications, the neutron energy spectra from an Am-Li sources is not well known. Several measurements over the past few decades have been made to characterize this spectrum; however, little work has been done comparing the measured spectra of various Am-Li sources to each other. This paper examines fourteen different Am-Li spectra, focusing on how these spectra affect simulated neutron multiplicity results using the code Monte Carlo N-Particle eXtended (MCNPX). Two measurement and simulation campaigns were completed using Active Well Coincidence Counter (AWCC) detectors and uranium standards of varying enrichment. The results of this work indicate that for standard AWCC measurements, the fourteen Am-Li spectra produce similar doubles and triples count rates. Finally, the singles count rates varied by as much as 20% between the different spectra, although they are usually not used in quantitative analysis.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
The AME 2020 atomic mass evaluation (I). Evaluation of input data, and adjustment procedures*
Abstract This is the first of two articles (Part I and Part II) that presents the results of the new atomic mass evaluation, AME2020. It includes complete information on the experimental input data that were used to derive the tables of recommended values which are given in Part II. This article describes the evaluation philosophy and procedures that were implemented in the selection of specific nuclear reaction, decay and mass-spectrometric data which were used in a least-squares fit adjustment in order to determine the recommended mass values and their uncertainties. All input data, including both the accepted and rejected ones, are tabulated and compared with the adjusted values obtained from the least-squares fit analysis. Differences with the previous AME2016 evaluation are discussed and specific examples are presented for several nuclides that may be of interest to AME users.
74 ATOMIC AND MOLECULAR PHYSICS↗
Nuclear Data Sheets for A=244
Available information pertaining to the nuclear structure of ground and excited states for all known nuclei with mass numbers A=244 have been compiled and evaluated. The adopted level and decay schemes, as well as the detailed nuclear properties and configuration assignments based on experimental data are presented for these nuclides. When there are insufficient data, expected values from systematics of nuclear properties and/or theoretical calculations are quoted. Unexpected or discrepant experimental results are also noted. Information provided in 244 U is tentative as it from the the possible production in the study of 1960 Di03 of debris of the first large-scale thermonuclear test (Ivy Mike) of November 1, 1952 in the Pacific Ocean. No new data are available for 244 Np and the information for g.s Jp is tentative. More experiment work is needed for 244 Np as theoretical work suggest an isomer with Jp=7- and g.s with Jp=2+. In 244Pu, a new K isomer was identified and in addition rotational bands built on the isomer have been observed via the deep inelastic excitation. In 244 Am, specifically in the (d,p)transfer reaction and in the (n,g ) secondary g ’s, there are a large number of gammas that are unplaced. Since its last measurement in 1966, the electron capture dataset has been updated with a tentative partial level scheme for 244 Cm which also provided a new half life measurement for 244 Bk. In 244 Fm, a fusion evaporation experiment measured the half life and provided upper limit for the SF, a and EC branching. A summary and compilation of the discovery of various isotopes in this mass region is given in 2013 Fr02 ( 244 Np, 244 Pu, 244 Am, 244 Cm, 244 Bk, and 244 Cf), 2011Me01 ( 244 Es), and 2013 Th02 ( 244 Fm). This work supersedes the 2003 evaluation by Y.A. Akovali( 2003Ak04).
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Nuclear Data Sheets for A = 180
Spectroscopic data for all nuclei with mass number A = 180 have been evaluated, and the corresponding level schemes from radioactive decay and reaction studies are presented.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Evaluating 240 Pu Prompt Fission Neutron Spectra Including Chi-Nu Experimental Data
This report documents an evaluation of 240 Pu prompt fission neutron spectra (PFNS) including Chi-Nu experimental data. This work is a FY2025 Q4 NCSP (Nuclear Criticality Safety Program) milestone and named a “strategic priority”. This evaluation is new; it differs substantially from the most recent evaluation included in the ENDF/B library, which was for ENDF/B-VII.1 and then carried over unchanged to ENDF/B-VIII.1. The key difference between ENDF/B-VIII.1 and the new evaluation is that we have, for the first time, realistic experimental 240 Pu PFNS covering a broad incident and outgoing neutron energy range.
240Pu↗
Nuclear Data Sheets for A=248
Available information pertaining to the nuclear structure of ground and excited states for all known nuclei with mass numbers A=248 has been compiled and evaluated. Here, the adopted level and decay schemes, as well as the detailed nuclear properties and configuration assignments based on experimental data, are presented for these nuclides. When there are insufficient data, expected values from systematics of nuclear properties and/or theoretical calculations are utilized. Unexpected or discrepant experimental results are also noted. This work supersedes the 2014 evaluation by M.J. Martin (2014Ma86).
Nesaraja, Caroline D. [Oak Ridge National Laborato↗
Predicting Correlation Coefficients for Monte Carlo Eigenvalue Simulations
Monte Carlo methods are most often considered as a reference for neutron transport simulations since very limited approximations are made about nuclear data and system geometry. To report uncertainty of any tally evaluated as generation averages, the sample variance is divided by the number of active generations, which is based on the assumption that the neutron generations are independent. Correlation effects between neutrons in multiplying systems, particularly when performing power iteration to evaluate eigenvalues have been observed in previous work. Neglecting the correlation effect results in an underestimate of uncertainty reported by Monte Carlo calculations. Previous work has also proposed methods to predict the underestimation ratio. Yamamoto et al expanded the fission source distribution with diffusion equation modes, performed numerical simulation of the AR (autoregressive) process of the expansion coefficients and used the correlation of the AR process to predict that of the Monte Carlo eigenvalue simulation. Sutton applied the discretized phase space (DPS) approach to predict the underestimation ratio but the method cannot predict the ratio when one neutron generates offspring in different phase space regions or generates a random number of offspring. This paper presents a method to predict the correlation effect with the model of multi-type branching processes (MBP). The method requires simulations for one generation of neutrons without knowing the source distribution and can predict the underestimation ratio for the cases where the traditional DPS approach does not work. The generation-to-generation correlation determines the convergence rate of active generations, the bias of variance estimator for each generation and the underestimation ratio of variance estimator for tallies averaged over active generations. The generation-to-generation correlation is characterized by the Auto-Correlation Coefficients (ACC) between tallies from different generations. (authors)
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
Validating nuclear models: electron data to interpret neutrino physics
Nuclear models in Monte Carlo event generators are crucial for interpreting neutrino scattering experiments, but are dicult to verify using only neutrino data due to incomplete data for some energy ranges. In particular, we seek to constrain the GENIE neutrino event generator's models. To overcome the uncertainties inherent to neutrino experiments, it is useful to take advantage of more controllable electron scattering experiments to evaluate current models and place constraints on future models. In this preliminary study on the feasibility of this strategy, we evaluate variables sensitive to the choice of nuclear model for the GENIE default model and data from the Jeerson Labs CLAS experiment. We nd some disagreement between them, likely explained by the the use of a dated default and background known to be remaining in the CLAS data that is not accounted for.
72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
Source Term Analysis of Xenon (STAX): An effort focused on differentiating man-made isotope production from nuclear explosions via stack monitoring
An overview of the hardware and software developed for the Source Term Analysis of Xenon (STAX) project is presented which includes the data collection from two stack monitoring systems installed at medical isotope production facilities, infrastructure to transfer data to a central repository, and methods for sharing data from the repository with users. STAX is an experiment to collect radioxenon emission data from industrial nuclear facilities with the goal of developing a better understanding of the global radioxenon background and the effect industrial radioxenon releases have on nuclear explosion monitoring. The final goal of this work is to utilize collected data along with atmospheric transport modeling to calculate the contribution of a peak or set of peaks detected by the International Monitoring System (IMS) to provide desired discriminating information to the International Data Centre (IDC) and National Data Centers (NDCs). Types of data received from the STAX equipment are shown and collected data was used for a case study to predict radioxenon concentrations at two IMS stations closest to the Institute for RadioElements (IRE) in Belgium. The initial evaluation of results indicate that the data is very valuable to the nuclear explosion monitoring community.
07 ISOTOPE AND RADIATION SOURCES↗
Neutron-induced fission cross section measurements for uranium isotopes {sup 236}U and {sup 234}U at LANSCE
A well established program of neutron-induced fission cross section measurement at Los Alamos Neutron Science Center (LANSCE) is supporting the Fuel Cycle Research program (FC R and D). The incident neutron energy range spans from sub-thermal up to 200 MeV by combining two LANSCE facilities, the Lujan Center and the Weapons Neutron Research facility (WNR). The time-of-flight method is implemented to measure the incident neutron energy. A parallel-plate fission ionization chamber was used as a fission fragment detector. The event rate ratio between the investigated foil and a standard {sup 235}U foil is converted into a fission cross section ratio. In addition to previously measured data new measurements include {sup 236}U data which is being analyzed, and {sup 234}U data acquired in the 2011-2012 LANSCE run cycle. The new data complete the full suite of Uranium isotopes which were investigated with this experimental approach. Obtained data are presented in comparison with existing evaluations and previous data.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗