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All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Measurements of effective actinides neutron capture cross sections in a cadmium-filtered epithermal neutron spectrum

In this work, a total of 20 actinide samples were irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL): 233 U (1), 235 U (1), 236 U (1), 238 U (1), 237 Np (2), 239 Pu (1), 240 Pu (1), 242 Pu (2), 244 Pu (2), 241 Am (2), 243 Am (2), 244 Cm (2), and 248 Cm (2). The samples were loaded in a basket containing a 1.1-mm thick cadmium filter. The presence of the filter drastically changes the neutron spectrum in the samples and accentuates the reaction rates in the resonance region. Isotopic ratios were determined using Multi-Collector Inductively Coupled Plasma Mass Spectrometry (MC-ICP-MS) that allowed inferring effective capture cross sections. Detailed best-estimate MCNP models of the experiment using ENDF/B-VII.0 nuclear data were developed to calculate the effective neutron cross sections and, subsequently, compare them with the cross sections inferred from experimental results.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Discriminating Uranium Isotopes Using the Time-Emission Profiles of Long-Lived Delayed Neutrons

In nuclear nonproliferation and safeguards, detecting and accurately characterizing special nuclear material remains one of the greatest challenges. Uranium enrichment determination is typically achieved by measuring the ratio of characteristic γ -ray emissions from 235 U and 238 U. Fission also produces β -delayed neutrons, which have been used in the past to determine uranium enrichment from the time dependence of the long-lived delayed-neutron emission rate. Such measurements typically use moderated 3 He-tube detectors. We demonstrate an alternative measurement technique that employs a fast neutron active-interrogation probe and a scintillation detector to measure the enrichment of uranium using both the buildup and decay of β -delayed-neutron emission. Instead of 3 He tubes, a capture-based heterogeneous composite detector consisting of scintillating Li glass and polyvinyl toluene is constructed and used, offering a prospect to scale delayed-neutron measurements to larger detector sizes. Since the technique relies on the existing tabulated nuclear data, no calibration standards are required. It is shown that the buildup of delayed-neutron emission can be used to distinguish between uranium samples and infer the uranium enrichment level, with accuracy that rivals the method that employs the time-dependent decay of delayed-neutron emission.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2015

Mox powder calculation improvement: OECD Benchmark proposal and cristal - Rib results - 14228

In the framework of the WPNCS Expert Group on Uncertainty Analysis for Criticality Safety Assessment, we propose a blind numerical benchmark. Since the WPNCS Expert Group on Experimental Needs had highlighted the poor database of representative experiments of low-moderated MOX powders, this blind benchmark is focused on MOX wet powders Three PuO{sub 2} contents are proposed: 100%, 30% and 12.5%. Two Pu isotopic vectors are considered: a conservative one (0%, 71%, 17%, 11%, 1% for {sup 238}Pu up to {sup 242}Pu) and a realistic one obtained from reprocessed LWR 30 GWd/T fuels. For these 6 cases the powder moisture rate is 3%. The MOX fissile medium is a sphere surrounded by 20 cm water reflector. The required results are the calculated k{sub eff} and the associated prior uncertainty. Using benchmarks experiments, contributors should provide the projected-k{sub eff} and the corresponding posterior uncertainty. CRISTAL-V2 results based on JEFF3.1.1 nuclear data are presented. As an example, in the 30% PuO{sub 2} challenging case, the calculation uncertainty is decreased from 1160 pcm (1 σ) down to 450 pcm using EOLE and MASURCA critical Pu experiments. This uncertainty depends slightly on 'conservative' or 'realistic' Pu because the representativity factor does not vary significantly with the Pu vector. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Establishing Pb-203 production from electrodeposited Tl targets at Brookhaven National Laboratory

Background: Promising developments in Pb-212 radiopharmaceutical therapies have increased demand for Pb-203 diagnostic agents. Building on previous work from various isotope production facilities, this study optimized Pb-203 production from electrodeposited Tl targets at Brookhaven National Laboratory (BNL). The additional supply of Pb-203 may help meet growing preclinical and clinical demands. Results: Two Tl targets were irradiated at the Brookhaven Linac Isotope Producer facility with 30 ± 1 MeV protons, measured using previously published cross section data. Distribution coefficients for Pb Resin in acetate media were investigated for both Na + and K + cations, where potassium acetate was ~ 4 times more effective at stripping Pb from the Pb Resin. The Tl electrodeposition was optimized to deposit 350 mg of Tl (~ 60 mg/cm 2 ) on Au backing in under 6 h. The proposed separation process was completed in < 1.5 h and achieved > 98% and 92 ± 3% recovery of Tl and Pb, respectively, with an overall Tl-Pb separation factor of 6 × 10 5 . The experimentally measured half-life of Pb-203 was 52.4 ± 0.7 h, agreeing with 51.93 ± 0.02 h reported by the National Nuclear Data Center. The radioisotopic purity of the Pb fraction at 24 h post end of bombardment (EOB) from a 24 h irradiation was 66% Pb-203, 28% Pb-201, and 6% Pb-200. Following chemical separation, the Pb-203 produced in this work (21 MBq Pb-203 EOB) achieved apparent molar activities of 10 ± 5 and 0.9 ± 0.5 GBq/µmol for [ 203 Pb]Pb-DOTAM and [ 203 Pb]Pb-DO3A, respectively, decay corrected to EOB. Data derived from this work suggests BNL can produce > 10’s GBq Pb-203 with > 99% radiochemical and radioisotopic purity from Tl-205 for worldwide distribution. Conclusions: The production and separation of Pb-203 from natural Tl target material was successfully demonstrated at BNL. Existing methods were adapted and optimized for the facilities at BNL. Results from this work will guide future large-scale Pb-203 production opportunities at BNL for clinical applications.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2017

Nuclear Forensics and Radiochemistry: Reaction Networks

In the intense neutron flux of a nuclear explosion the production of isotopes may occur through successive neutron induced reactions. The pathway to these isotopes illustrates both the complexity of the problem and the need for high quality nuclear data. The growth and decay of radioactive isotopes can follow a similarly complex network. The Bateman equation will be described and modified to apply to the transmutation of isotopes in a high flux reactor. A alternative model of growth and decay, the GD code, that can be applied to fission products will also be described.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2026

On the uncertainties in helium generation predictions for fission and fusion alloys

With ongoing advances in fusion and advanced fission reactors, quantifying irradiation effects in materials is critical. Transmutation-induced helium in cladding and structural materials can drive swelling and embrittlement, thereby reducing these components’ lifespans. Yet most studies ignore the considerable uncertainties in predicting helium generation rates. In this work, we created a code wrapper, F-SCATTER, that automatically performs simulations in FISPACT-II. We used this tool to investigate potential variance in helium generation rate, or He/dpa, calculations based on deviations in alloy composition, irradiating neutron flux spectrum, computational methodology, and nuclear data sources. We used 12 wt% Cr HT9 steel as the reference case and observed a 6.5%–98.3% He/dpa spread based on compositional variation within a single chemical specification, a 1.8%–11.5% He/dpa variation upon the incorporation of a 15% artificial uncertainty in flux at each energy, and a He/dpa difference as high as 231% when using ENDF/B-VIII.0 versus TENDL-2021 data libraries. Similar results were found for other prominent iron-based alloys, including Grade 91, castable nano-structured alloy, and 316H—where additional variations exist based on reactor type (e.g. thermal, fast, or fusion) and alloying elements such as carbon, nitrogen, and nickel. Based on the simulated results, we conclude that a significant part of the heat-to-heat variability in swelling responses of Fe-based alloys can be driven by impurity content in alloy compositions, and, therefore, chemical control should be a key element in supply chain design for advanced nuclear energy systems. Furthermore, we provide critical recommendations on best practices for evaluating and reporting helium production and lattice damage rates when computing predictions with multiphysics programs such as FISPACT-II.

FISPACT-II↗
DOE OSTI2026

IER-620 CED-3b: Experiment Execution Summary for the Pulsed-Neutron Die-Away Experiments (PNDA) with Propylene Glycol and Mobilmet 423

There is a strong need for new benchmarks to validate neutron thermal scattering laws (TSLs). Lawrence Livermore National Laboratory (LLNL) has designed a Pulsed-Neutron Die Away (PNDA) testbed for this purpose. The experiment has a deuterium-tritium (D-T) neutron generator that impinges a 10 -4 s, mono-energetic pulse of 14.1 MeV neutrons on a target sample. After the pulse, the neutron population moderates and establishes a thermal equilibrium within the sample, with a fundamental spatial mode and characteristic decay-time eigenvalue, ⍺. The ⍺ eigenvalue can be extracted from the experimental measurements of the time-dependent neutron flux coming off the surface of the sample and can then be used as an integral parameter (similar to k eff ) to validate nuclear data involved with neutron migration, thermalization, and absorption. For moderating materials and geometric configurations, the ⍺ eigenvalue is heavily dependent on thermal neutron scattering of the target material. For that reason, a PNDA experiment can have a higher sensitivity to TSLs than is commonly available with the k eff parameter in critical experiments.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2026

Measurement of the 75 As ⁢(𝑛, 2⁢𝑛) cross section at 14.1 MeV at the National Ignition Facility

Here, a new measurement of the 75 As (𝑛, 2⁢𝑛) measurement was performed using the 14.1 MeV neutron pulse produced by fusion experiments at the National Ignition Facility (NIF). The target material consisted of GaAs foils encapsulated in aluminum irradiation containers, along with Au monitor foils, and positioned in the NIF chamber during high neutron yield shots. After irradiation, the GaAs and Au foils were counted with high purity germanium detectors to assess the production of (𝑛, 2⁢𝑛) activation products to determine the flux and the cross section of interest. The measured cross section was 966 ± 71 mb at 14.1 ± 0.37 MeV. This work provides a proof of concept for a platform for performing “ride-along” cross section measurements at NIF to contribute to existing nuclear data as well as measure new cross sections in the future.

and nuclear chemistry↗
DOE OSTI2016

Measurement of the 169 Tm(n,3n) 167 Tm cross section and the associated branching ratios in the decay of 167 Tm [Measurement of the 169 Tm(n,3n) 167 Tm cross section and the branching ratios in the decay of 167 Tm]

The cross section for the 169 Tm(n,3n) 167 Tm reaction was measured from 17 to 22 MeV using quasimonoenergetic neutrons produced by the 2 H(d,n) 3 He reaction. This energy range was studied to resolve the discrepancy between previous (n,3n) cross-section measurements. In addition, the absolute γ-ray branching ratios following the electron-capture decay of 167 Tm were measured. Furthermore, these results provide more reliable nuclear data for an important diagnostic that is used at the National Ignition Facility to estimate the yield of reaction-in-flight neutrons produced via the inertial-confinement-fusion plasma in deuterium-tritium capsules.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2019

Investigation of Application Specific k eff Sensitivities to Existing and New Subcritical and Critical Experiments

Integral subcritical and critical experiments are essential in providing validation of computational codes which improve confidence in the predicted k eff results of specific applications. This work discusses the relevance of existing integral subcritical and critical experiments to applications far reaching across the nuclear complex. These applications include Travelling Wave Reactors (TWR), Kilopower Reactor Using Stirling Technology (KRUSTY), and plutonium reflected by tantalum. The primary goal of this work is to improve knowledge of energy-dependent cross-section sensitivities for specific applications, identify nuclear data gaps in existing experiments, enable an optimized design of new experiments to address these gaps, and design practical new experiments that could be executed at the National Criticality Experiments Research Center (NCERC) that have a high-value to these applications.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

On Integral Benchmarks for Resolving TSL Conflicts for ENDF/B-VIII.1 Release [Slides]

This presentation touches on polystyrene ORNL evaluation in TSL review process for ENDF/B-VIII.1. the presentation found excellent agreement with INS and transmission measurements. Additionally, PS and PE evaluations with different phonon spectra, similar total xs., calculate keff similarly. This presentation found for critical benchmarks. While extremely useful for validation of nuclear data at all energies, critical benchmarks are not the best tool to provide a definitive answer on conflicting TSLs. In conclusion this presentation found INS and transmission measurements need to be the basis of validation of TSLs.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

High Flux Isotope Reactor Core Analysis-Challenges and Recent Enhancements in Modeling and Simulation

A concerted effort over the past few years has focused on enhancing the core depletion models for the High Flux Isotope Reactor (HFIR) as part of a comprehensive study for designing a HFIR core that would use low-enriched uranium (LEU) fuel. A HFIR core depletion model that is based on current state-of-the-art methods and nuclear data was needed for use as a reference for the design of an LEU fuel for HFIR and to improve the basis for analyses that support HFIR’s current operation with high-enriched uranium (HEU) fuel. This paper summarizes the recent improvements in modeling and simulation for HFIR core analyses, with a focus on core depletion models.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2017

Microcalorimetry for High-Resolution X-Ray Spectroscopy

Magnetic Microcalorimeters (MMCs) are gamma-ray detectors with an energy resolution 10x higher than high-purity germanium detectors. They can increase the accuracy of non-destructive analysis of nuclear materials, enable the detection of new isotopes (e.g. Pu-242 of U-236), and improve nuclear data in cases where Ge detectors are limited by line overlap. MMCs consist of a magnetic sensor operated at temperatures below 50 mK, and they infer gamma-ray energies from the change in magnetization due to the temperature increase after gamma-ray absorption. The goal of this project is to further increase the energy resolution and sensitivity of MMC gamma detectors.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2024

Scattering Calorimeter FY24 Deliverable Report

A simulation-based method has been developed to prototype new detector designs for nuclear data measurements utilizing neutron scattering. This method uses representative physics inputs for signal and background generation, full detector resolution smearing benchmarked by experimental data, and a neutron beam timing simulation to produce analyzable output like a physical measurement. A test case has been studied using a hybrid time-of-flight calorimeter detector for scattering cross-section measurements on 239 Pu with 1-5 MeV incident monoenergetic neutrons. Data analysis methods have been developed to perform event-level particle reconstruction and reaction channel discrimination. This analysis has been used to estimate the capability of the test detector to perform simultaneous scattering and fission cross section measurements, as well as its ability to provide neutron spectra and particle angular information.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2022

Validation of MCNP Critical Benchmarks Models of Highly Enriched Uranium Cylinders

A new centralized repository of high-quality MCNP models of critical benchmark experiments is currently under development at Los Alamos National Laboratory (LANL). The benchmark experiments are described in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook, and the initial set of benchmark models are derived from the Whisper Suite provided with MCNP6.2. This effort is a collaboration among the Nuclear Criticality Safety, Nuclear Data, and Monte Carlo code development/application organization at LANL. The objective is to create a current single LANL benchmark collection that includes the latest ICSBEP revision that has a formal review and revision process, is contained in an open- source repository, and utilizes new Python tools for improved input and output file review.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Lilith: An Enduring Plutonium Critical Assembly [Slides]

A design of an enduring Pu critical assembly has been produced. This is very important to DOE/NNSA missions. This assembly would result in advances/improvements in nuclear data validation, analytical methods validation, dosimetry, reactor kinetics, materials, and more.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Template for estimating uncertainties of measured neutron-induced fission cross-sections

A template for estimating uncertainties (unc.) of measured neutron-induced fission, (n,f), cross-sections (cs) is presented. This preliminary template not only lists all expected unc. sources but also supplies ranges of unc., estimates for correlations between unc. of the same and different experiments which can be used if the information is nonexistent. If this template is applied systematically when estimating experimental covariances for an evaluation, it may help in pinpointing missing unc. for individual datasets, identifying unreasonably low unc., and estimating correlations between different experimental datasets. Thus, a detailed unc. estimate – usually, a time-intensive procedure – can be undertaken more consistently and efficiently. As an example, it is shown that unc. and correlations of 239 Pu(n,f) by Merla et al. [Proceedings of the Conference on Nuclear Data for Science and Technology 1991 Jülich (Springer-Verlag, Berlin, 1992) pp. 510–513], which are questionably low in the GMA database underlying the neutron cs standards evaluations, are distinctly larger at 14.7 MeV and more strongly correlated if this template is used for reestimating the associated covariances.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2021

Development of a Fast-Spectrum Self-Powered Neutron Detector for Molten Salt Experiments in the Versatile Test Reactor

Abstract—The self-powered neutron detector (SPND) is a widely used flux monitor in thermal nuclear reactors. Although this is a mature technology, the current state of the art is tuned for a thermal neutron spectrum, so many of the devices currently in use lack sensitivity to fast neutrons. Because current in SPNDs is produced through nuclear reactions with the neutron flux inside a reactor, sensitivity in SPNDs is determined by the neutron cross section of the neutron-sensitive portion of the detector, termed the emitter. This neutron cross section drops by orders of magnitude between thermal and fast neutron energies for many emitters in currently used SPNDs, with a corresponding drop in current from the detector. This paper discusses efforts to develop a fast-spectrum self-powered neutron detector (FS-SPND) that is sensitive to neutrons with energies ranging from 0.025 eV up to 1 MeV. An in-depth analysis of Evaluated Nuclear Data File (ENDF)/B-VII.1 neutron-capture cross sections was performed, and four new materials were identified that are suitable emitter candidates for use in measuring fast neutrons. All four materials are stable mid-shell nuclei in the region between doubly magic 132Sn and 208Pb. Each candidate was simulated with the Geant4 Monte Carlo simulation toolkit to optimize overall detector efficiency.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗