Deploying Isotope Identification Machine Learning Models Project #: 22-142 Year 1 of 3
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Why Lady Godiva Should be Replaced as the Default Validation Experiment for U-235 Nuclear Data
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.
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Design Optimization for EUROPA Critical Experiment
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.
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Measurement of ( n, γ ) cross section of 90 Zr [Abstract]
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.
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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.
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239 Pu R -matrix Analysis and Neutron Multiplicities in the Neutron Energy Region up to a few keVs [Abstract]
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.
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Comparison of AGR-3/4 Fission Product Transport Model to Measurements and Extraction of Diffusivities via Analytical Fits
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.
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Hydrogen isotope separation methods and systems
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.
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Nuclear Quality Assurance Certification of the MCNP-ORIGEN Activation Automation tool
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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Supercritical Fluid Separation and Purification of Rare Earth Elements, particularly Lanthanides including 177-Lu, to Lower Energy Consumption and Reduce Waste
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Idaho National Laboratory’s (INL) Role in Radioisotope Power Systems (RPS)
Overview of INL's support of Radioisotope Power Systems. Given to Mount Science and Energy Museum Association
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Characterisation of Specimens from a Former Soviet Thorium Processing Facility to Support Site Remediation
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.
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Determining Cs-137 Background body burdens for Wild Pigs At Savannah River Site
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.
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Fission Product Gas Monitoring During Fuel Drying Operations - 20094
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)
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Identification and Quantification of Beta-Emitters in Waste Packages by Segmented Gamma-Scanning - 20114
A commonly applied method for the non-destructive characterization of the radioactive inventory of waste packages (e.g. 200-l drums) is segmented gamma-scanning. The gamma-emitting inventory is identified via their characteristic lines in the measured spectra. The corresponding net peak areas allow for quantification applying appropriate models. The identification and quantification of beta-emitters present in the waste package require further evaluation. While some of them emit gamma-rays, too, which can be used for identification and quantification, pure beta-emitters contribute to the gamma-spectra only by their Bremsstrahlung-spectra. {sup 32}P, {sup 90}Sr and {sup 90}Y are typical representatives of these isotopes. A method is developed for identification and quantification of these pure beta-emitters in radioactive waste packages. Origin of the procedure is the evaluation of the measured gamma-spectrum. After the identification of all lines present in the spectrum (characteristic gamma-lines, single- and double-escape-lines, X-ray-lines, annihilation peak etc.) this information is used together with calibration data of the detection system and further a-priori information like matrix composition, cross-section data from databases etc. to simulate the spectra without the Bremsstrahlung-components by an appropriate method. This may be Monte Carlo simulations or, like in our case, a computer code taking into account mathematical descriptions of the physical effects like photo effect, (multiple) Compton scattering, backscattering etc. and the attenuation effects of the waste matrix. In any case, the resulting spectrum must simulate the real spectrum without its Bremsstrahlung-component as good as possible. Subtracting the simulated from the measured spectrum then results in the Bremsstrahlung-components caused by the beta-emitters. Prior to evaluation, a set of Bremsstrahlung-spectra is calculated and stored for different beta-emitters in different matrices using the Bethe-Heitler equation. This data set can be extended for new conditions, if necessary, and is allocated in a data base for future application. Applying a non-linear optimization using the generalized reduced gradient method for comparing the calculated with the derived Bremsstrahlung-spectra results in the identification of the beta-emitter and a factor, which is proportional to its activity. The proportionality is determined by previous calibration measurements. The application of the method is demonstrated by measurements on a 200-l mock-up drum and a real waste package. (authors)
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A Study on Radionuclides Selection for DCGLs Development for Site Release of KRR Units 1 and 2 site in Korea - 20256
The selection of Derived Concentration Guideline Levels (DCGLs) in nuclear facilities decommissioning is important because they serve as criteria for site characterization and final status surveys. DCGLs should be derived for KAERI TRIGA Mark II and III research reactor site clearance and prior to the work, a radionuclides list for DCGLs consideration was created. Radionuclides selection for DCGLs was performed according to the procedure proposed by the US NRC based on the characterization survey related to the decommission site. By evaluating the dose contribution of each radionuclide using D and D computer code, radionuclides with a dose contribution fraction of 0.1% or more for the entire radionuclide were selected. As a result, the selected radionuclides were beta radionuclides H-3, C-14 and Sr-90 and gamma radionuclides Co-60, Cs-137, Eu-152 and Eu-154. The selected radionuclides can be presented for consultation with domestic regulatory agencies in the future and will be used as the technical base of DCGLs for the final status survey. (authors)
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Evaluation of the Sum-of-Fractions Methodology for Water and Polyethylene Moderated Systems
Sum-of-Fractions is a method intended to assure a subcritical margin for aqueous solutions and slurries of fissionable isotopes. The method indicates that a system is subcritical if the sum of the ratios of the mass of each isotope in a mixture to its individual minimum subcritical mass limit is less than or equal to one. The basis of the Sum-of Fractions has historically been derived from allowances given in ANSI/ANS-8.15-1981. However, the allowance was removed in ANSI/ANS-8.15-2014 due to a lack of technical basis. A methodology was developed to assess the validity of using the Sum-of-Fractions for water or polyethylene moderated systems for the following nuclides: 232 U, 233 U, 234 U, 235 U, 237 Np, 236 Pu, 238 Pu, 239 Pu, 240 Pu, 241 Pu, 242 Pu, 241 Am, 242 mAm, 243 Am, 242 Cm, 243 Cm, 244 Cm, 245 Cm, 246 Cm, 247 Cm, 249 Cf, and 251 Cf. The methodology uses available benchmark data for mixtures of 233 U, 235 U, and 239 Pu to establish the calculational margin, and a mass limit reduction to establish the margin of subcriticality. Water or polyethylene moderated and reflected mixtures containing the nuclides are evaluated with SCALE 6.2.4. Including the calculational margin, subcritical mass limits for each nuclide were computed for optimally water or polyethylene moderated and fully reflected systems. These masses were used to create nuclide mixtures in which the sum of the mass to subcritical mass limit ratios is one. The various nuclide mixtures were modeled over a range of moderation and demonstrate the k eff does not exceed the calculational margin. For additional assurance of subcriticality, a significant mass reduction is applied to each computed minimum critical mass of the nuclides without adequate benchmark data consistent with the method in ANSI/ANS-8.15-2014.
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Conceptual Spacer Design for the ATR GEN I Target for Pu-238 Production in the Advanced Test Reactor at Idaho National Laboratory
The initial target design used for Pu-238 production at Idaho National Laboratory was designed by Oak Ridge National Laboratory to optimize the production of Pu-238 in the High Flux Isotope Reactor (HFIR) and are referred to as HFIR GEN II targets. To take advantage of the Advanced Test Reactor’s (ATR) taller active core region a redesign of the HFIR GEN II targets was needed. It was proposed to stack two HFIR GEN II targets nose to nose about the core center line; however, this resulted in excessive neutron and photon heating in the pellets located in the center. This peak heating was not desirable so three alternative designs were investigated for the ATR GEN I targets. The python-based code, MCNP to ORIGEN2 in Python (MOPY), was used to calculate the heating rates after 40 days of irradiation to capture the effects of each configuration. The purpose of this paper is to document the details of these conceptual design calculations and comparisons for the ATR GEN I targets.
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