Engineering Papers⌕ Search

SEARCH · Engineering Papers

Results for “Nuclear data”

Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.

At least 1,099 records · Page 61
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Passive Safety Feature of KIPT Neutron Source Facility

Argonne National Laboratory (ANL) of the United States and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have cooperated on the development, design, and construction of a neutron source facility. The facility was constructed at Kharkov, Ukraine and its commissioning process is underway. The facility has an electron accelerator driven subcritical assembly. The electron beam power is 100 kW using 100 MeV electrons. Tungsten or natural uranium is the target material for generating neutrons driving the subcritical assembly. The subcritical assembly is composed of WWR-M2 Russian fuel assemblies with U-235 enrichment of 19.7 wt%, surrounded by beryllium reflector assembles and graphite blocks. The water coolant operates at room temperature during normal operation, when the facility is operating at full power using 100-KW electron beam power. The passive safety features of the facility are discussed in in this study. Monte Carlo computer code MCNPX was utilized in the analyses with ENDF/BVII.0 nuclear data libraries. Negative reactivity temperature feedback was consistently observed, which is important for the facility safety performance. Due to the design of WWR-M2 fuel assemblies, slight temperature increase and the corresponding density decrease in the water coolant would cause relatively large reactivity drop, which offset the reactivity gain by mistakenly loading an additional fuel assembly. The increase of fuel temperature also causes sufficiently large reactivity decrease. This enhances the facility safety performance because fuel temperature increase provides prompt negative reactivity feedback. (authors)

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2013

Parametric analyses of single-zone thorium-fueled molten salt reactor fuel cycle options

Analyses of fuel cycle options based on thorium-fueled Molten Salt Reactors (MSRs) have been performed in support of fuel cycle screening and evaluation activities for the United States Department of Energy. The MSR options considered are based on thermal spectrum MSRs with 3 different separations levels: full recycling, limited recycling, and 'once-through' operation without active separations. A single-fluid, single-zone 2250 MWth (1000 MWe) MSR concept consisting of a fuel-bearing molten salt with graphite moderator and reflectors was used as the basis for this study. Radiation transport and isotopic depletion calculations were performed using SCALE 6.1 with ENDF/B-VII nuclear data. New methodology developed at Oak Ridge National Laboratory (ORNL) enables MSR analysis using SCALE, modeling material feed and removal by taking user-specified parameters and performing multiple SCALE/TRITON simulations to determine the resulting equilibrium operating conditions. Parametric analyses examined the sensitivity of the performance of a thorium MSR to variations in the separations efficiency for protactinium and fission products. Results indicate that self-sustained operation is possible with full or limited recycling but once-through operation would require an external neutron source. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Role of (n,2n) reactions in transmutation of long-lived fission products

The conditions under which (n,γ) and (n,2n) reactions can help or hinder each other in neutron transmutation of long-lived fission products (LLFPs) are considered. Isotopic and elemental transmutation for the main long-lived fission products, {sup 79}Se, {sup 93}Zr, {sup 99}Tc, {sup 107}Pd, {sup 126}Sn, {sup 129}I, and {sup 135}Cs, are considered. The effect of (n,2n) reactions on the equilibrium amount of nuclei of the transmuted isotope and the neutron consumption required for the isotope processing is estimated. The aim of the study is to estimate the influence of (n,2n) reactions on efficiency of neutron LLFP transmutation. The code TIME26 and the libraries of evaluated nuclear data ABBN-93, JEF-PC, and JANIS system are applied. The following results are obtained: (1) The effect of (n,2n) reactions on the minimum number of neutrons required for transmutation and the equilibrium amount of LLFP nuclei is estimated. (2) It is demonstrated that, for three LLFP isotopes ({sup 126}Sn, {sup 129}I, and {sup 135}Cs), (n,γ) and (n,2n) reactions are partners facilitating neutron transmutation. The strongest effect of (n,2n) reaction is found for {sup 126}Sn transmutation (reduction of the neutron consumption by 49% and the equilibrium amount of nuclei by 19%).

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2021

Progress on the reevaluation and validation of the n+233U neutron cross sections

The set of 233 U resonance parameters of the ENDF/B-VIII.0 nuclear data library was adopted from the previous ENDF/B-VII.1 evaluation using the external levels to update the thermal values. Adoption of IAEA 2017 thermal standards ( σ f = 533.0 ± 2.2 b, σ c = 44.9 ± 0.9 b, and ν ‾ tot = 2.487 ± 0.011 ) and of the IAEA-recommended thermal-neutron induced prompt fission neutron spectrum (PFNS) with average PFNS energy of 2.030 ± 0.013 MeV requires a re-evaluation of 233 U neutron cross sections in the resolved resonance region. A newly produced evaluation is being tested on benchmarks carefully selected from the Handbook of International Criticality Safety Benchmark Experiments (ICSBEP) which are highly sensitive to 233 U data. An important goal of this work was to eliminate the strong negative gradient of the calculated effective multiplication factors with respect to the epithermal fission fraction observed in the validation of the ENDF/B-VIII.0 library for those assemblies. A significant improvement in integral performance of critical 233 U solutions is observed for the newly proposed evaluation. Further work addressing the fast neutron region is needed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Physical particularities of nuclear reactors using heavy moderators of neutrons

In nuclear reactors, thermal neutron spectra are formed using moderators with small atomic weights. For fast reactors, inserting such moderators in the core may create problems since they efficiently decelerate the neutrons. In order to form an intermediate neutron spectrum, it is preferable to employ neutron moderators with sufficiently large atomic weights, using {sup 233}U as a fissile nuclide and {sup 232}Th and {sup 231}Pa as fertile ones. The aim of the work is to investigate the properties of heavy neutron moderators and to assess their advantages. The analysis employs the JENDL-4.0 nuclear data library and the SCALE program package for simulating the variation of fuel composition caused by irradiation in the reactor. The following main results are obtained. By using heavy moderators with small neutron moderation steps, one is able to (1) increase the rate of resonance capture, so that the amount of fertile material in the fuel may be reduced while maintaining the breeding factor of the core; (2) use the vacant space for improving the fuel-element properties by adding inert, strong, and thermally conductive materials and by implementing dispersive fuel elements in which the fissile material is self-replenished and neutron multiplication remains stable during the process of fuel burnup; and (3) employ mixtures of different fertile materials with resonance capture cross sections in order to increase the resonance-lattice density and the probability of resonance neutron capture leading to formation of fissile material. The general conclusion is that, by forming an intermediate neutron spectrum with heavy neutron moderators, one can use the fuel more efficiently and improve nuclear safety.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

TREAT neutronics analysis and design support, Part I: Multi-SERTTA

Experiment vehicle design is necessary in preparation for Transient Reactor Test (TREAT) facility restart and the resumption of transient testing to support Accident Tolerant Fuel (ATF) characterization and other future fuels testing requirements. Currently the most mature vehicle design is the Multi-SERTTA (Static Environments Rodlet Transient Test Apparatuses), which can accommodate up to four concurrent rodlet-sized specimens under separate environmental conditions. Robust test vehicle design requires neutronics analyses to support design development, optimization of the power coupling factor (PCF) to efficiently maximize energy generation in the test fuel rodlets, and experiment safety analyses. Calculations were performed to support analysis of a near-final design of the Multi-SERTTA vehicle, the design process for future TREAT test vehicles, and establish analytical practices for upcoming transient test experiments. Models of the Multi-SERTTA vehicle containing typical PWR-fuel rodlets were prepared and neutronics calculations were performed using MCNP6.1 with ENDF/B-VII.1 nuclear data libraries. Calculation of the PCF for reference conditions of a PWR fuel rodlet in clean water at operational temperature and pressure provided results between 1.10 and 1.73 W/g-MW depending on the location of the four Multi-SERTTA units with the stack. Basic changes to the Multi-SERTTA secondary vessel containment and support have minimal impact on PCF; using materials with less neutron absorption can improve expected PCF values, especially in the primary containment. An optimized balance is needed between structural integrity, experiment safety, and energy deposition in the experiment. Type of medium and environmental conditions within the primary vessel surrounding the fuel rodlet can also have a significant impact on resultant PCF values. The estimated reactivity insertion worth into the TREAT core is impacted more by the primary and secondary Multi-SERTTA vehicle structure with the experiment content and contained environment having a near negligible impact on overall system reactivity. Additional calculations were performed to evaluate the peak-to-average assembly powers throughout the TREAT core, as well as the nuclear heat generation for the various structural components of the Multi-SERTTA assembly. Future efforts include the evaluation of flux collars to shape the PCF for individual Multi-SERTTA units during an experiment such as to achieve uniformity in test unit environmental conditions impacted by the non-uniform axial flux/power profile of TREAT. Upon resumption of transient testing, experimental results from both the Multi-SERTTA and Multi-SERTTA-CAL will be compared against calculational results and methods for further optimization and design strategies. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Release of Evaluated 235 U(n,f) Average Prompt Fission Neutron Multiplicities Including the CGMF Model

This report documents an evaluation of the average prompt fission neutron multiplicity, $\overline{v}_p$, of 235 U from 200 keV to 15 MeV that is a potential release candidate for the upcoming U.S. nuclear data library, ENDF/B-VIII.1. This evaluation had to be re-done from "scratch", as the input to the $\overline{v}_p$ evaluation of the previous library, ENDF/B-VIII.0, was lost. That means that all available experimental data were re-analyzed and uncertainties were re-estimated. Another major difference to ENDF/B-VIII.0 is that this evaluation includes model information from the Hauser-Feshbach fission fragment decay code CGMF, while ENDF/B-VIII.0 is based purely on experimental data. CGMF links several fission quantities with each other; $\overline{v}_p$ is predicted by assumptions made on, e.g., pre-neutron emission yields as a function of mass, the total kinetic energy, or spin and parity of fission fragments. This allows to perform two types of validation for the new 235 U $\overline{v}_p$: On the one hand, one can employ evaluated CGMF parameters obtained from fitting to experimental 235 U $\overline{v}_p$ to predict yields as a function of mass, the average total kinetic energy, or the mean energy of the prompt fission neutron spectrum. These model-predicted values can then be compared to experimental and evaluated data. The model-predicted fission-observable values using evaluated parameters obtained here are reasonably close to experimental data indicating the evaluated 235 U(n,f) $\overline{v}_p$ are physical. On the other hand, one can validate 235 U $\overline{v}_p$ with respect to integral responses such as fast ICSBEP critical assemblies or LLNL pulsed spheres. LLNL pulsed-sphere neutron-leakage spectra are minimally impacted by the new 235 U $\overline{v}_p$ as these experimental data are shape data and the $\overline{v}_p$ would mostly lead to a change in normalization of the data as the spheres are relatively thin (0.7 and 1.5 mean-free path) and, thus, mostly depend on 235 U $\overline{v}_p$ from 12-15 MeV. The change in the predicted effective neutron multiplication factor, k eff , of selected ICSBEP critical assemblies, however, is large compared to values using ENDF/B-VIII.0 and experimental k eff : The average bias is 108 pcm across all studied k eff values versus 12 pcm for ENDF/B-VIII.0. A reasonable performance in simulating keff (mean bias of 14 pcm) can be retained by tweaking 235 U $\overline{v}_p$ from 3-5 MeV, and combining it with a recent 235 U PFNS evaluation that is also a ENDF/B-VIII.1 release candidate.

235U↗
DOE OSTI2022

Correlation Calculations for the Russian Pu Metal Fast Experiments

Nuclear criticality experiments are often conducted in campaigns with multiple variations. These experiments reuse the same basic components, like the fuel, moderator, or positioning machines. The components have uncertainties in their geometry and composition that propagate to models of the experiments. Shared components create shared uncertainty between the $k_{eff}$ of benchmarks. The shared uncertainty is commonly quantified with a covariance, or correlation coefficient. These covariances can impact criticality safety and nuclear data validation applications. While benchmark evaluations tabulate an experiment’s uncertainty, they often lack a detailed calculation of correlations between experiments. Even some very commonly used benchmarks, like the Russian Pu Metal Fast (PMF) experiments, have missing correlations. This paper presents our approach to calculate the correlations for five of the Russian PMF experiments. The experiments share hemispherical Pu shells that induce a correlation between modeled $k_{eff}$ values. We estimated the correlations with simplified and detailed models of the experiments through linear-perturbation theory. The correlations between the experiments vary significantly between the detailed vs. simplified models. We also investigate how the correlations affect validation metrics of the experiments.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2019

Development and Maintenance of NJOY

NJOY is the leading nuclear data processing code, which is developed at Los Alamos National Laboratory. NJOY21 is the current production version. It replaces the legacy, Fortran-based, NJOY2016 which is a direct descendent of the original version released in 1977. NJOY21 is the next-generation version of NJOY that is being completely modernized. It has been released with an Open Source license and has no export control restrictions. More information about NJOY21 can be found in Section 2. This report serves to document the development and maintenance plan for NJOY-– including legacy NJOY and NJOY21. This document is constantly being updated as the development of NJOY proceeds. The source of this document can be found in a git repository at LANL.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2015

Reassessment of uncertainties for the PU-MET-INTER-002 benchmark - 14432

Uncertainty for the PU-MET-INTER-002 benchmark of ZPR-6/10 was unusually large for a ZPR assembly and was dominated by the uncertainties in matrix tube pitch and composition uncertainties for plutonium, graphite and stainless steel plates. The original uncertainty analysis for PU-MET-INTER-002 was limited by computer resources, software and the nuclear data available in 2002, all of which have been improved since. An as-built Monte Carlo model of ZPR-6/10 was used to quantify components of the transformation from the detailed model to a simplified RZ benchmark in 2002. Compositional uncertainties were estimated based upon TWODANT calculations of the RZ benchmark model (20 group, S{sub 8} - P{sub 1}). MC{sup 2}-2 was used with END/B-V.2 data to generate homogenized cross section data for TWODANT. The MC{sup 2}-2 calculations did not account for lattice cell spatial heterogeneity which was driven by the manual process required to generate ∼80 modifications to the deterministic model and by the software limitations of MC{sup 2}-2. Improvements to DIF3D-VARIANT, development of the MC{sup 2}-3 and BuildZPRModel codes, and better computer resources make accurate 3D transport calculations of reactivity changes practical today. BuildZPRModel imports the as-built geometry and creates a 'drawer' homogenized DIF3D model where each radial DIF3D mesh is the exact size of a single ZPR matrix tube and drawer. A MC{sup 2}-3 input is constructed for each drawer to produce homogenized cross section data where the spatial cell heterogeneity is captured by solving a 1D MOC transport calculation on the ultrafine energy group level. Collapsing 3D drawers into 1D MC{sup 2}-3 drawer models follows conventional procedures and is done by BuildZPRModel. MCNP6 was used to re-compute relevant geometry components of the uncertainty using ENDF/B-VII.0 data with the as-built model. Changes in the geometry components compared with earlier results are insignificant. DIF3D calculations of the compositional uncertainties were performed using 33, 70, 116, 230, 425, and 703 group energy structures and selectively combined with P{sub 3}-P{sub 7} transport approximations and P{sub 1} and P{sub 3} scattering kernels. 3D diffusion theory calculations were also performed for comparison. Many composition uncertainty components were essentially unchanged with the updated methodology. However, the dominant experimental uncertainty for PU-MET-INTER-002 is the composition of stainless steel. The new calculations reduce the stainless steel uncertainty from ∼115 pcm to ∼70 pcm which significantly reduces the total uncertainty for PU-MET-INTER-002. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Position-Dependency of Fuel Assembly Homogenization in a Pressurized Water Reactor

2-step nuclear reactor core analysis has long been established as the conventional calculational procedure for the pressurized water reactor (PWR) due to its efficiency. This is because the procedure significantly avoids the potentially huge computational burden arising from an explicit geometrical modeling and consideration of over ∼ 500 energy groups. It instead approaches the problem with spatial homogenization and group condensation: a 2-D lattice transport calculation to homogenize the fuel assembly cross-sections and a 3-D whole core nodal diffusion. One notes that the few-group constant generation using deterministic method could be complicated due to its limited flexibilities in geometrical modeling and inherent drawbacks of the multi-group approximation. Furthermore, accuracy of the deterministic few-group constant generation can especially be compromised during the resonance treatment calculation. As such, an alternative lattice calculation using the Monte Carlo (MC) approach, in place of the deterministic fuel assembly homogenization (few-group constant generation), has previously been proposed. During the assembly homogenization, directions and magnitudes of the current at boundaries are unknowns. Therefore, zero net current boundary conditions (infinite lattice) are instead used as the assembly node. The assumption is quite reasonable considering heterogeneities inside the node are more dominant than heterogeneities due to nodal location in the core. In fact, this approximation has successfully been applied to many LWR core analyses to reasonable accuracy. It should nonetheless be noted that the net current zero boundary assumption effectively ignores position-dependency of the fuel assembly homogenization in the general 2-step method. As such, there are a number of well-known researches on the extended assembly homogenization techniques, which introduce correction factors to compensate effects of the assembly environment, such as Generalized Equivalence Theory (GET), Super-homogenization method (SPH), and color set calculation method. Even though the assembly environment effect has long been acknowledged, its effect has not properly been investigated due to the limited computing power in the past and due to the reasonable accuracy of the extended assembly homogenization techniques. Nonetheless, quantification of position-dependency of the assembly homogenization in the core could still potentially offer additional value in terms of calculation accuracy to the 2-step reactor analysis. It is upon this motivation that the paper is prepared. In this research, position-dependent few-group constants were calculated and compared against those of infinite lattices. All calculations were completed using the MC Serpent2 code with ENDF/B7.0 nuclear data library. In this study, position-dependency of the fuel assembly homogenized cross sections in a PWR core has been investigated through comparison of the conventional 2-step reactor analysis with the reference high fidelity Monte Carlo simulation. It was found that the homogenized two-group constants are clearly position-dependent as the discrepancies between infinite lattice and whole-core calculations are quite significant especially in a rodded assembly. As such, the corresponding correction required due to the assembly environment effect is not negligible compared to the correction required due to assembly heterogeneity itself. It is therefore quite clear that position-dependency of the group cross sections should be considered in the 2-step procedure for a more accurate PWR core analysis. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

New Measurements to Resolve Discrepancies in Evaluated Model Parameters of 181 Ta

To resolve discrepancies in evaluated cross sections among major nuclear data libraries, energy-differential neutron transmission and radiative capture yield of 181 Ta were measured from 0.15 to 100 keV using multiple sample thicknesses. The new measurements provide resolution such that the resolved resonance region (RRR) can be evaluated up to at least 2.5 keV and the unresolved resonance region can be evaluated up to at least 100 keV. The transmission and capture yield measurements were modeled using resonance parameters from three major libraries to assess the predictive capability of each. It was found that JENDL-5.0 performed best in the RRR. Because of the poor performance of the U.S. ENDF/B evaluation, it is recommended that ENDF/B be reevaluated for 181 Ta.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Impact of new prompt fission neutron spectrum evaluations and the associated cross-correlations on critical assemblies using MCNP6 - 14385

Two recent re-evaluations of the neutron-induced prompt fission neutron spectrum (PFNS) of uranium and plutonium actinides have included some cross-correlated uncertainties that have not been released in a new nuclear data library. The first re-evaluation included cross-isotope correlations of the PFNS within each of the uranium and plutonium suites of actinides below the second-chance fission threshold. The second, more recent re-evaluation includes incident energy-dependent correlations of the {sup 239}Pu(n,f) PFNS. Both of these new evaluations and their cross-correlations, given the importance of these actinide quantities and because the impact of the incident energy-dependent correlations has never been studied previously, could have an impact on a variety of criticality safety applications. For example, the ke uncertainty due to the uranium PFNS uncertainties in the Big Ten critical assembly is underpredicted by ∼10% when the cross-isotope correlations in the uranium PFNS are ignored. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
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 OSTI2015

Cumulative fission yields of short-lived isotopes under natural-abundance-boron-carbide-moderated neutron spectrum

The availability of gamma spectroscopy data on samples containing mixed fission products at short times after irradiation is limited. Due to this limitation, data interpretation methods for gamma spectra of mixed fission product samples, where the individual fission products have not been chemically isolated from interferences, are not well-developed. The limitation is particularly pronounced for fast pooled neutron spectra because of the lack of available fast reactors in the United States. Samples containing the actinide isotopes 233, 235, 238U, 237Np, and 239Pu individually were subjected to a 2$ pulse in the Washington State University 1 MW TRIGA reactor. To achieve a fission-energy neutron spectrum, the spectrum was tailored using a natural abundance boron carbide capsule to absorb neutrons in the thermal and epithermal region of the spectrum. Our tailored neutron spectrum is unique to the WSU reactor facility, consisting of a soft fission spectrum that contains some measurable flux in the resonance region. This results in a neutron spectrum at greater than 0.1 keV with an average energy of 70 keV, similar to fast reactor spectra and approaching that of 235U fission. Unique fission product gamma spectra were collected from 4 minutes to 1 week after fission using single-crystal high purity germanium detectors. Cumulative fission product yields measured in the current work generally agree with published fast pooled fission product yield values from ENDF/B-VII, though a bias was noted for 239Pu. The present work contributes to the compilation of energy-resolved fission product yield nuclear data for nuclear forensic purposes.

Finn, Erin C.↗
DOE OSTI2017

A Novel Methodology For Generating Thermal Scattering Cross Sections and Uncertainties

With the rise in interest of GEN-IV reactor systems, newer, and more accurate thermal scattering nuclear data are needed. Both GEN-IV reactors and current light water reactors applying for license extensions require high fidelity cross sections and uncertainties to better thermal margins and thereby maximizing energy production and improving economics and safety. Thermal moderator data play a key role in both reactor physics and nuclear criticality safety analyses. Due to the lack of uncertainties data for thermal scattering material, there is no way to quantify the effects of thermal scattering uncertainties in quantities of interest (e.g., k eff ) in reactor or criticality safety systems.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
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 OSTI2026

Westcott g factors extended to arbitrary neutron energy spectra

Westcott 𝑔 factors are used in Neutron Activation Analysis (NAA) and Prompt Gamma-ray Activation Analysis (PGAA) to evaluate the impact of non-1∕𝑣 behavior in the neutron-capture cross sections of certain nuclei on activation product yields. This non-1∕𝑣 behavior arises from the presence of neutron resonances in the neutron- capture cross sections that overlap with the source neutron spectrum at low (< 5 eV) energies. Historically, Westcott 𝑔 factors that have been cataloged for NAA and PGAA applications are the result of calculations that assume a Maxwellian neutron flux distribution with a given temperature. In this work, we use this approach with updated neutron-capture cross sections from the Evaluated Nuclear Data File, version VIII.1 (ENDF/B-VIII.1) to tabulate Westcott 𝑔 factor values for a broad range of Maxwellian distribution temperatures, comparing the results against currently-available 𝑔 factors from International Atomic Energy Agency tables and other sources. Here, it was discovered during this analysis that the use of guided thermal and cold-neutron beams at certain facilities necessitates an approach for evaluating Westcott 𝑔 factors based on arbitrary non-Maxwellian spectra. In this paper, we present an approach for calculating 𝑔 factors with user-specified neutron spectra, and we demonstrate these methods to obtain Westcott 𝑔-factors for guided- and cold-neutron beams at the Budapest Research Reactor and the Forschungsreaktor München II reactor. As part of this work, open-source software has been developed that can be used to perform these calculations for applications in PGAA and NAA experiments.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗