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Idaho National Laboratory (INL) has recently compiled a user’s guide for the Light Weight Radioisotope Heater Unit (LWRHU). This guide gives an overview of the LWRHU to potential users. Information and data on mass properties, dimensions, thermal loads, and interfaces are discussed.
Idaho National Laboratory has completed irradiation of Np-237 targets in the Advanced Test Reactor’s (ATR) South Flux Trap (SFT) and I-7 positions. INL also progressed qualification of new ATR Gen 1 Np-237 targets for the North East Flux Trap (NEFT), inner A, and H positions. This paper gives an overview of operational and technical activities from March 2021 to February 2022.
A MS Excel tool was created to help track Pu-238 production, Np-237 feedstock use, estimate production scenarios, and coordinate planning for shipments. As-run estimates were used as a basis for beginning and end of life isotopic concentrations. Multiple target types were created to demonstrate a proof of concept while actual results were not yet available The initial goals of this production tool were: • Track unirradiated, currently irradiated, and irradiated targets at the Advanced Test Reactor (ATR) • Estimate production yields from irradiated targets in various positions for each cycle • Provide past, current, and future year quantities of Np-237 and Pu-238 at ATR, amounts shipped to ATR and amounts shipped from ATR • Estimate production from current and future target designs based on data from numerical modeling (values will be provided by Idaho National Laboratory (INL)) • After proof of principle, additional work is planned that will further enhance the tool with: o Incorporate the High Flux Isotope Reactor (HFIR) reactor at Oak Ridge National Laboratory (ORNL) for irradiation o Incorporate irradiations and target shipments from ORNL to INL and back to track Np-237 transfers across the Department of Energy (DOE) Complex
The INL FY 2021 Laboratory Overview is an opportunity for INL to share its achievements and plans for the future with the entire Laboratory, stakeholders, and the public. This document outlines how INL will advance its clean energy and national security objectives in the future and highlights FY 2020 accomplishments across the Laboratory in science and technology, community outreach, and operations.
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For the past seven decades the Lady Godiva benchmark (HEU-MET-FAST-001) has been the primary experiment used for 235 U nuclear data validation. The papers associated with large nuclear data library releases such as ENDF/B-VII.1 and ENDF/B-VIII.0 refer to it frequently. The reasons why HEU-MET-FAST-001 has been used as the primary validation experiment for 235 U nuclear data validation will be discussed in the following section. However, as discussed among the benchmark community, the standards associated with International Criticality Safety Benchmark Evaluation Project (ICSBEP) evaluations have changed throughout time; this is subject of the OECD/NEA WPEC (Working Party on Nuclear Criticality Safety) Subgroup 8. HEU-MET-FAST-001 is an older benchmark (issued during the inaugural year of ICSBEP in 1996 with only minor revisions occurring since then) and (along with many of the other benchmarks from this era) does not meet the standards for a modern benchmark. This work explores why HEU-MET-FAST-001 is useful for 235 U nuclear data validation and discusses other alternative validation experiments.
The Experiment for Unresolved Region Of Plutonium Actinides (EUROPA) is an integral critical experiment currently being designed to target the intermediate energy region of plutonium. Intermediate energies, those between 0.625 eV - 100 keV, contain the end of resolved and beginning of the unresolved resonance region, making experiments in this energy range prudent for validating the representation of cross sections in this transition region. Below is the preliminary design of the experiment to be performed at the National Criticality Experiments Research Center (NCERC). Experiment optimization required an exhaustive down selection of several moderating, absorbing, and reflecting materials using Particle Swarm Optimization (PSO) in order to achieve maximum sensitivity to the intermediate energy region.
The isotopes of Zr with A = [90, 91, 92, 94] make up more than 97% of naturally occurring Zr and are important to many nuclear applications such as nuclear reactors. One of the attractive qualities of naturally occurring Zr isotopes is that they have a low σγ/σt ratio at most neutron energies, such that they improve the neutron economy in reactors by preferentially scattering neutrons rather than absorbing them. This same quality also presents a challenge to measuring the capture cross section, σγ, of Zr isotopes. The ENDF/B-VIII.0 library has a relative uncertainty of approximately 10-20% for incident neutron energies < 0.1 MeV, and uncertainty greater than 20% for energies > 0.1 MeV for the majority of natural Zr isotopes. This motivated the Nuclear Criticality Safety Program (NCSP) to embark on a campaign to accurately measure and evaluate these isotopes of Zr. Here we will show energy-dependent neutron capture cross section measurements for the first enriched sample to be measured: 90 Zr. The measurements of isotopically enriched samples are being carried out at the Geel Electron Linear Accelerator (GELINA) facility of the Joint Research Center - Geel (JRC-Geel) of the European Union. As isotopic enrichment is a costly process we are careful not to activate any of the samples, as this may hinder future radiation-sensitive measurements. The activation analysis is presented in a report by Brown et al.. Once we were satisfied that the Zr samples would not be activated by the measurements, the 90 Zr sample was fabricated at Oak Ridge National Laboratory (ORNL) and shipped to GELINA. The dimensions of the cylindrical sample are approximately 0.12 cm thick and a radius of 2.5 cm. Since 90 Zr is not chemically reactive to air, bare metallic samples were employed. The sample was measured at a flight path (FP) length of 60 m, using four C 6 D 6 detectors on FP14. The final paper will include experimental details and measured cross section data for 90 Zr compared to current evaluated nuclear data libraries.
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.
The evaluation of 239 Pu neutron resonance parameters coupled to neutron multiplicities $\overline{v}_p$ is of particular importance to investigate the ($\mathcal{n, γf}$) reaction in which a $\mathcal{γ}$-ray emission occurs before the scission of the compound nuclear. This reaction has offered one explanation for the fluctuations in the measured values of $\overline{v}_p$ In this regard, the competition between ($\mathcal{n, γf}$) reaction and the direct fission process can be also included in the R matrix analysis of fission and capture measured data. The goal of this work is the coupled evaluation of the $\mathcal{n}$+ 239 Pu resonance parameters and related neutron multiplicities by ensuring the adoption of thermal neutron constants recently evaluated at the International Atomic Nuclear Energy as well as the recommended (thermal-neutron) induced prompt neutron fission spectrum (PFNS). Moreover, this new set of physical evaluated quantities should also guarantee the agreement for high-leakage solution benchmarks while keeping the good performance of large thermal solution assemblies.
A one-dimensional finite-element model based on the MOOSE framework has been constructed, and overall transport estimates are predicted based on available diffusivities and sorption isotherms in the extant literature. The predicted transport profiles are compared to experimental results, where such results are available, including total inventory analysis of gamma emitters, liquid scintillation results for Sr-90 concentrations, radial deconsolidation leach-burn-leach within compacts, tomographic reconstructions based on detailed gamma scanning and destructive physical sampling of several rings. The magnitude of expected transport to the sink ring is generally predicted by the model for silver and cesium, however, modeled detailed concentration profiles are not consistent with experimental results. Strontium and europium transport is not well characterized by model results.
Methods and systems for the separation of hydrogen isotopes from one another are described. Methods include utilization of a hydrogen isotope selective separation membrane that includes a hydrogen isotope selective layer (e.g., graphene) and a hydrogen ion conductive supporting layer. An electronic driving force encourages passage of isotopes selectively across the membrane at an elevated separation temperature to enrich the product in a selected hydrogen isotope.
The MCNP-ORIGEN Activation Automation (MOAA) tool couples MCNP and ORIGEN-S, allowing to support ATR and TREAT experiments by predicting their composition and activity before their irradiation takes place. MOAA is developed and maintained by the Irradiation Experiment Neutronics Analysis group (C-150) at INL. This poster presents the motivation behind the development of MOAA, its main workflow, as well as the software quality assurance (SQA) efforts required by LWP-13620, "Managing Information Technology Assets."
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Overview of INL's support of Radioisotope Power Systems. Given to Mount Science and Energy Museum Association
The former Prydniprovski Chemical Plant (PChP) is a large ex-soviet nuclear materials processing site of roughly 260 hectares, located in central Ukraine. PcHP was in operation between 1948 and 1991 for the purposes of processing uranium and thorium ore into ore concentrate. During its operational lifetime, tens of millions of tonnes of radioactive waste was generated on the site. A consequence of these activities, coupled with the mismanagement of the resulting radioactive materials and by-products, is that approximately 45% of the southern part of the site has become radioactively contaminated. This area comprises some 100-200 buildings and structures, many of which are in a state of disrepair and dereliction. Around 20 of these buildings are heavily radioactively contaminated, with radiation levels from a few μSv/hr to a maximum of 4 mSv/hr reported. The European Commission is currently working with the Ukraine Ministry of Energy and Coal Use and the Nuclear Regulator to improve radiological safety and security at the site. In order for these remediation works to proceed, the site management needs to characterise the radioactive materials that are held there. As part of this activity, the team from the University of Bristol have undertaken a survey of one of the most heavily contaminated buildings, collecting a selection of in situ material samples. A suite of complimentary materials characterisation techniques (e.g. X-ray Fluorescence, Raman Spectroscopy, etc) have been deployed in order to determine the radionuclides and composition of the compounds present. Characterisation of these samples will allow the operators of the chemical plant to go forward with remediation works within the building.
Savannah River Site (SRS) is a unique U.S. Department of Energy facility in that most of the 310-square miles of land has not be industrialized since operations began in 1953. This has allowed abundant, undisturbed vegetation structure to exists as well as an increase in herd sizes among wildlife populations onsite. Cs-137 is widespread across SRS due to the production of nuclear materials, but is mostly from global atmospheric fallout from nuclear weapons testing. Due to most of SRS being undisturbed land, the concentrations of Cs-137 are higher on site than off site. The SRS began conducting annual deer harvests that were open to the public beginning in 1965. All animals harvested during these hunts are monitored for Cs-137, which is measured using a Sodium Iodide detector. There has been an effective method for calculating the background Cs-137 body burden for the deer population at the SRS, developed by Gaines and Novak using data collected during hunts and gamma overflight data to determine the areas contaminated by site activities. For animals that have a Cs-137 concentration above the established background level, a dose is assigned to the hunter. By applying the same method for wild pigs using data collected from hunts, gamma overflight, and developing a site-specific home range of wild pigs, we were able to calculate the Cs-137 background body burden concentrations of wild pigs on site. Objectives: Develop site-specific home range for wild pigs at SRS; Determine the average concentrations of Cs-137 in wild pigs from background; Use Gaines and Novak method for determining background Cs-137 body burdens wild pigs on SRS. Methods and Materials: Development of site-specific home range of wild pigs was determined to be 2.5 miles using: Four studies conducted between 1969 and 1992 for the SRS wild pig population. (Mayer and Brisbin) The background compartments were determined using: Gamma overflight data showing contaminated areas; Site-specific home range of the wild pigs. Results and Conclusion: All data were generated through the JMP software to provide a histogram and summary statistics. The estimate of an upper tolerance limit for the background contamination with 95% confidence is 1.965 pCi/g for the wild pigs. This means that for wild pigs on SRS from the background compartments are expected to have Cs-137 concentrations of less than 1.965 pCi/g. The current background for the deer is 2.59 pCi/g. For the background compartments Cs-137 concentrations from 2012-2018: Mean: 0.815 pCi/g, Maximum: 2.341 pCi/g, Minimum: -0.704 pCi/g.
The goal of the project was to devise and implement a method of accurately qualifying any noble gas emitted during fuel drying operations (part of the dry fuel storage process) in order to provide the data needed to validate the off-site dose calculations. The imputes of the project was the NRC publishing Information Notice 18-01, 'Noble Fission Gas Releases During Spent Fuel Cask Loading Operations' in February 2018. The information notice describes some industry events that occurred during vacuum drying operations as part of a dry fuel campaign. The solution was devised by Fort Calhoun Station staff in partnership with Mirion Technologies technical staff. The monitoring was accomplished by directing all the effluent from the fuel storage cask vacuum drying machine through a sample chamber containing a compact ISOCS characterized CZT based gamma spectrometer connected to a Mirion Data Analyst. This resulted near real-time quantification of the radionuclides of concern. (authors)