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At least 271 records · Page 15
All sourcesNASA NTRSDOE OSTI
DOE OSTI2020

Data Testing of Polyethylene Thermal Scattering Law with New Thermal Epithermal eXperiments (TEX) Plutonium Baseline Benchmark, PU-MET-MIXED-002

Five plutonium baseline experiments completed in September of 2018 from the Thermal Epithermal eXperiments (TEX) program have been accepted by the International Criticality Safety Benchmark Evaluation Project as PU-MET-MIXED-002 and will be included in the 2020 version of the handbook. The experiments were made up of stacked layers of plutonium Zero Power Physics Reactor (ZPPR) plates moderated by polyethylene (PE, chemical formula C2H4) to varying degrees to create configurations with five different neutron spectra, including one fast configuration, one thermal configuration, and three mixed configurations. Calculations of the benchmark show overprediction of k eff when using the most recent release of the Evaluated Nuclear Data File Version B (ENDF/B) library release (VIII.0) cross sections, particularly when compared to the previous ENDF/B-VII.1 results. One potential source of overprediction was hypothesized to be the new Molecular Dynamics (MD) generated PE Thermal Scattering Law (TSL). The overprediction was shown via calculation to not be caused by the new ENDF/B-VIII.0 thermal scattering law, and in fact the overprediction was mitigated by the new PE TSL.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Critical Experiments Targeting the Epithermal/Intermediate Cross Sections of Tantalum

Sandia National Laboratories (SNL) and Oak Ridge National Laboratory (ORNL) have collaborated to develop a capability to test the epithermal/intermediate cross sections of materials at the SNL critical experiments facility using the Seven Percent Critical Experiment (7uPCX) fuel. The Sandia Critical Experiments Program provides a specialized facility for performing water moderated and reflected critical experiments with UO 2 fuel rod arrays. The facility offers the ability to modify the core configuration and reactor tank to evaluate various reactor cores for pitch, moderator characteristics, and other criteria. A history of safe operations and flexibility in reactor core configuration has resulted in the completion of nine sets of critical benchmark experiments that have been documented in the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook. The experiment described here is expected to be evaluated for inclusion in the 2024 edition of the ICSBEP Handbook.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

IER 555: Godiva Benchmark Update CED-2 (Final Design Report)

The International Criticality Safety Benchmark Evaluation Project (ICSBEP) evaluation of the Godiva IV critical assembly, HEU-MET-FAST-086: GODIVA-IV DELAYED-CRITICAL EXPERIMENTS (HMF-086), was completed by Russ Mosteller. Five critical experiment configurations performed at the Los Alamos National Laboratory (LANL) Technical Area (TA)-18 were evaluated as acceptable benchmark cases. The five cases consist of four delayed critical configurations which differ in control rod positions and one prompt critical configuration. All cases calculated a lower $k_{eff}$ than measured by experiment. This data is referred to as the TA-18 Godiva IV benchmark in this report. In 2005, Godiva IV was disassembled for relocation to the Nevada Test Site (NTS), now Nevada National Security Site (NNSS), at the National Criticality Experiments Research Center (NCERC). Following the disassembly and subsequent reassembly and startup of Godiva IV at NCERC, additional information about the Godiva IV components was obtained. An errata note was added to the HMF-086 evaluation in the ICSBEP handbook to provide this new information until a revision to the benchmark evaluation could be performed. In addition to those corrections, there are differences between the Godiva IV assembly at TA-18 and the Godiva IV assembly at NCERC. These differences include both assembly-specific differences (differences in the safety block gap, differences in the control rod positions, a new NCERC Top Hat and contamination shield) as well as environmental differences, such as the size and shape of the experimental building where the assembly is located. An additional model with similar cases, referred to as the NCERC Godiva IV benchmark in this report, will be added to the revised HMF-086 to capture these additional differences. This will provide the best benchmark model of Godiva for use by those performing experiments at NCERC. The IER 555 CED-2 report documents the information that will be updated in the HMF-086 revision, both the corrections to the TA-18 Godiva IV benchmark and the subsequent changes to create a NCERC Godiva IV benchmark. It describes the measurements that will be performed for cases similar to those performed at TA-18. It also describes measurements that will be included in the evaluation as additional data.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Summary of LANL Critical Benchmark Comparison Study and Revisions for Cases Involving HEU, LEU, MIX, and Pu

This report documents results obtained for revisions made to cases involving Highly Enriched Uranium (HEU), Intermediate Enriched Uranium (IEU), a mixture of Pu and Uranium (MIX), as well as Pu cases. A previous summary of revisions for HEU an Pu cases was reported and additional investigations into four cases originally presented therein uncovered further revisions which led to better agreement with other transport codes, those cases are updated in this report. The summary of all cases reported in Reference 2 is updated in this report. In addition, a previous summary of revisions for LEU and MIX was reported, a summary of those revisions in reproduced in this report for a comprehensive summary of changes to benchmarks beginning in fiscal year 2020 to current date. The report focuses on the changes made to LANL benchmarks modeled with MCNP6 using ENDF/B-VII.1 nuclear data that appeared to have discrepant results when compared with results of other codes. Feedback was used to pinpoint review of benchmark input files and to revise them when necessary. This report documents the results of review and revision of specific benchmarks highlighted as possibly discrepant in the comparison study. In addition, there is an effort tied to this work involving collaboration between LANL XCP and NCS Divisions in the development of a shared review/revision procedure and use of a new benchmark repository. LANL has a benchmark library of critical experiments from the International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook modeled for use with MCNP. This collection is now over 1100 benchmarks, referred to as the Whisper-1.1 library because it is used with the sensitivity/uncertainty package, Whisper, which supports nuclear criticality safety validation and is released with MCNP6.2. The collection, originally created several decades ago, is a combination of smaller collections, which has been revised and expanded, by various groups at LANL over the years. The original authors are no longer at the laboratory and little formal documentation of review and revision of these benchmarks exists today. A branch of the benchmark collection was already the subject of a formal review undertaken by the LANL NCS Division and expanded to include XCP Division.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Delayed Critical Enriched Uranium Metal, 7-in.-diam. Cylinder with Thin Stainless Steel Top and Bottom Reflectors

Three 7 in. diameter highly enriched uranium (HEU; 93.17 wt % 235 U) metal cylinders were assembled on the vertical assembly machine of the Oak Ridge Critical Experiments Facility (ORCEF) with thin stainless steel reflectors (0.5, 1, and 2 in. thick) on the top and bottom. These experiments were performed during January 13–25, 1965, and used seven operational days at ORCEF. Before these experiments, unreflected and unmoderated 7 in. diameter HEU metal cylinders had been assembled to delayed criticality at ORCEF, and the results were benchmarked in HEU-MET-FAST-051. In addition to the critical experiments, the prompt neutron decay constant was measured via the Rossi-α technique, and those results are also presented in this report. To achieve near-delayed-critical systems, 5 × 5 × 0.03125 in. rectangular uranium metal parts were used on top of 7 in. diameter cylinders inside the top reflector. The delayed critical configuration from the three experiments described herein are acceptable for use as criticality safety benchmark experiments for the International Criticality Safety Benchmark Evaluation Program (ICSBEP), coordinated by the Nuclear Energy Agency, once the uncertainty analysis is completed. Based on previous ICSBEP benchmarks with this enriched uranium metal at ORCEF, the uncertainty in $k_{eff}$ is expected to be as low as ±0.0002. The prompt neutron time decay data could be the basis of an International Reactor Physics Evaluation Program benchmark.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

S/U Comparison Study with a Focus on USLs

Under a DOE Nuclear Criticality Safety Program (NCSP) task involving Analytical Methods, three Laboratories collaborated in a comparison of results obtained from Sensitivity/Uncertainty (S/U) packages relevant to validation of transport codes. The task involves Institut de Radioprotection et de Sûreté Nucléaire (IRSN), Los Alamos National Laboratory (LANL), and Oak Ridge National Laboratory (ORNL) comparing results of MORET 5/MACSENS V3.0, MCNP6.2/Whisper-1.1, and SCALE 6.2.3/TSUNAMI/USLSTATS respectively. All Monte Carlo transport code results utilize ENDF/B-VII.1. Four cases from the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook) were selected as application models: HEU-MET-FAST-013-001, HEU-SOL-THERM 001-008, PU-MET-FAST-022-001, and PU-SOL-THERM 001-001. Ultimately, comparison is made between Upper Subcritical Limits (USLs) obtained using each code package for each application case. Since differences exist in whether packages take into account margin of subcriticality (MOS), the USL may be computed using bias and bias uncertainty, also known as the calculational margin (CM) in ANSI/ANS 8.24. Application of portions of MOS to the USL for nuclear data uncertainty of and potential code margin is referred to as USL herein. In either case, additional MOS is considered for actual application cases.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

MSR Safety Adequacy Demonstration US Contributions Towards a Common Goal

Molten Salt Reactors (MSRs) are currently under development worldwide. The US government has been sponsoring development of MSR safety adequacy evaluation information and tools. The US is a member of the Generation IV International Forum (GIF) MSR provisional System Steering Committee (pSSC). MSR pSSC collaboration is limited to exchange of open-fundamental information. Information employed to assess reactor safety adequacy needs to remain open to external scrutiny. US entities continue to measure fuel salt thermophysical and thermochemical properties, which have now been released into an open database. Non-US entities also continue to measure and openly publish fuel salt thermophysical and thermochemical properties. US entities continue to perform MSR accident progression experiments and simulations, the results of which are being reported in the open, scientific literature. The demonstrations have not yet been at sufficient scale to necessitate international collaboration.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Development and Validation of a Simulation Tool to Predict the Combined Structural, Electrical, Electrochemical and Thermal Responses of Automotive Batteries

Battery safety is one of the most important design factors of electrified vehicles (EVs) since battery failure may lead to a catastrophic consequence. To improve the safety of battery systems in EVs, battery behaviors under various abuse conditions should be well understood. One way to examine battery response under extreme conditions is to conduct abuse tests in different scenarios. They can provide first-hand pass/fail information on battery safety, but they are usually expensive and time-consuming, which is inconvenient during design iterations and optimization. On the other hand, computational modeling can predict the battery behaviors in a more efficient and cost-effective way, which becomes an important tool to evaluate battery safety. Modeling battery responses and failure is challenging since it involves multiple coupled physical processes, needs to consider complex material properties and is computationally intensive. This project aims to develop a practical simulation tool to predict the combined structural, electrical, electrochemical, and thermal responses of automotive batteries to crash-induced crush and short circuit and validate it for conditions relevant to automotive crash. Advanced material constitutive models will be utilized to capture the mechanical response of cell components. Moreover, methods will be developed to reduce computational complexity of the model and allow battery safety simulations to extend from cell-level to module/pack level with affordable computational resources.

25 ENERGY STORAGE↗
DOE OSTI2022

Review of SCALE Validations Applicable to Spent Nuclear Fuel Shielding Calculations

This report presents a review of publicly available documents providing validations of SCALE code system capabilities for depletion and shielding calculations. The validation studies primarily used information available in the Spent Fuel Isotopic Composition Database, the Shielding Integral Benchmark Archive and Database, and the International Criticality Safety Benchmark Evaluation Project Handbook. Validation results based on radiochemical assay data and shielding benchmark experiments relevant to spent nuclear fuel shielding applications are summarized in this report. The information summarized in this report supports pressurized water reactor and boiling water reactor spent fuel storage and transportation safety analyses.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS2012

Long Duration Head-Down Tilt Bed Rest Studies: Safety Considerations Regarding Vision Health

Visual symptoms reported in astronauts returning from long duration missions in low Earth orbit, including hyperopic shift, choroidal folds, globe flattening and papilledema, are thought to be related to fluid shifts within the body due to microgravity exposure. Because of this possible relation to fluid shifts, safety considerations have been raised regarding the ocular health of head-down tilt (HDT) bed rest subjects. HDT is a widely used ground ]based analog that simulates physiological changes of spaceflight, including fluid shifts. Thus, vision monitoring has been performed in bed rest subjects in order to evaluate the safety of HDT with respect to vision health. Here we report ocular outcomes in 9 healthy subjects (age range: 27-48 years; Male/Female ratio: 8/1) completing bed rest Campaign 11, an integrated, multidisciplinary 70-day 6 degrees HDT bed rest study. Vision examinations were performed on a weekly basis, and consisted of office-based (2 pre- and 2 post-bed rest) and in-bed testing. The experimental design was a repeated measures design, with measurements for both eyes taken for each subject at each planned time point. Findings for the following tests were all reported as normal in each testing session for every subject: modified Amsler grid, red dot test, confrontational visual fields, color vision and fundus photography. Overall, no statistically significant differences were observed for any of the measures, except for both near and far visual acuity, which increased during the course of the study. This difference is not considered clinically relevant as may result from the effect of learning. Intraocular pressure results suggest a small increase at the beginning of the bed rest phase (p=0.059) and lesser increase at post-bed rest with respect to baseline (p=0.046). These preliminary results provide the basis for further analyses that will include correlations between intraocular pressure change pre- and post-bed rest, and optical coherence tomography measurements of the retina.

Cromwell, Ronita L.↗
DOE OSTI2020

Transition Core Planning and Safety Analyses in Support of LEU Fuel Conversion of the University of Missouri Research Reactor (MURR)

The University of Missouri Research Reactor (MURR®) is one of six U.S. High Performance Research Reactors (USHPRR), including one critical facility, that is working with the National Nuclear Security Administration (NNSA) Office of Material Management and Minimization (M3) Reactor Conversion Program to convert from highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel. The M3 Reactor Conversion USHPRR Project objectives include the development of LEU fuel element designs that will ensure safe reactor operations and to maintain the existing experimental performance of each facility. The work is being conducted through many inter-related activities being completed by four Project Pillars: Fuel Qualification (FQ), Fuel Fabrication (FF), Reactor Conversion (RC), and Cross Cutting (CC). A new type of LEU fuel based on an alloy of uranium-10 wt% molybdenum (U-10Mo) is expected to allow the conversion of those USHPRR, like MURR, requiring higher density fuels. The very-high-density LEU U-10Mo monolithic fuel is currently undergoing irradiation testing and post-irradiation examination under a planned and documented fuel qualification effort. The FQ Pillar will document fuel property and fuel performance data and qualify the fuel for use in these reactors. The FF Pillar is fabricating fuel for ongoing and future irradiation tests, as well as conducting fabrication demonstrations to validate or update preliminary fabrication assumptions. The FF Pillar is also working to develop and install commercial manufacturing capacity with the U-10Mo monolithic fuel to produce prototypic fuel. Working with the RC Pillar at Argonne, MURR has progressed through a preliminary fuel element design using preliminary data for the proposed monolithic alloy of U-10Mo. Analyses were completed in previous work that found for typical equilibrium operations with the preliminary LEU fuel element design, in conjunction with a power uprate to 12 MW and appropriate changes to the MURR Limiting safety system settings (LSSS), MURR will have adequate margins to safety for steady-state operations and postulated transient accidents and will have experimental performance in key locations that meets or exceeds current operations with HEU fuel. The purpose of this work is to develop a sequence of transition cycles that will enable MURR to transition from operation with the reactor core loaded with fresh LEU fuel elements only to typical equilibrium operations with mixed-burnup cores following conversion while meeting operational requirements on safety and experimental performance. It is expected that the use of fresh LEU fuel at conversion and subsequent low burnup of the LEU fuel elements that will initially be available for use following conversion will result in critical control blade positions that will substantially change the axial power distribution in the core and the neutron flux available in key experimental locations relative to equilibrium LEU operations. Given the constraints of MURR safety margins, operational practices, and production and research, a novel method has been developed to identify a transition sequence that minimizes the time MURR operates atypically compared to the current prototypic cycles using HEU fuel. The proposed transition sequence moves quickly to the same sort of equilibrium cycles for the LEU fuel that have already been evaluated in documented preliminary safety analyses. Although shifting the neutron flux peak to the lower half of the core during initial cycles with LEU at 12 MW reduces the experiment performance in some key locations relative to current HEU operations at 10 MW, all LEU cores provide an average performance that meets or exceeds that of HEU. An LEU cycle is reached that meets or exceeds the level of experimental performance predicted for current HEU and equilibrium LEU operations in more than 450 key locations identified by a reactor specialist at MURR by the 23rd cycle following conversion and that afterwards will enable MURR to consistently meet its experimental performance requirements. The proposed transition sequence only requires the fabrication of 34 fresh LEU elements in the first year of operation and does not exceed the anticipated availability of fresh elements that can be produced by the fuel fabricator. By the third year after conversion, 22 fresh LEU elements will be required each year, which is the same as expected for equilibrium LEU operations and the same as current operations with HEU fuel. The proposed transition sequence thus combines a relatively short time period before equilibrium burnup is achieved, a temporary increase of fuel elements needed annually relative to typical operations that are within the production capabilities of the fuel fabricator, and demonstrates comparable experimental performance of the LEU cores relative to current HEU operations. Further measures may be taken to reduce any initial experimental performance penalty even further, where possible, by repositioning certain experiments to leverage the increased performance in the lower axial experimental positions in the initial cycles following conversion or leaving the experiments in the irradiation facilities longer in order to achieve the required neutron fluence. This analysis may require refinement depending on the experimental facilities in use at the time of conversion. Nonetheless, the results presented here, including the experimental performance, core burnup, and critical control blade positions throughout the transition cycles, show that the proposed transition cycle fuel management patterns are consistent with what is expected and desired for MURR operation with LEU U-10Mo fuel. Detailed core power distributions from the neutronics models were also used to evaluate safety margins during steady-state operations for the selected transition cycles and the equilibrium LEU core. It is shown that there are adequate safety margins for both steady-state operations and postulated accident scenarios. For the steady-state operations with the preliminary LEU fuel element design the analysis predicts at least 2.49 MW margin to the onset of flow instability at the LSSS power of 15 MW. Considering the LSSS power is 125% of full license power, the margin to OFI is sufficient. In addition, the critical heat flux ratio at LSSS power is well above the requirement of CHFR > 2.0 from NUREG-1537 for all considered cases. For postulated transient accidents, the minimum margin to the fuel temperature safety limit is at least 109 °C. In summary, the proposed sequence of core loadings for MURR operations following conversion to LEU fuel and a power uprate to 12 MW provides sufficient safety margins for both steady-state operations and postulated transient accidents during a proposed sequence of transition cycles to equilibrium operations. Analysis has shown that there are some local experimental performance penalties during the initial cycles. Although there are local shifts in the experimental performance, on average all LEU cores at 12 MW have equal or higher performance than HEU at 10 MW. Temporary adjustments are being planned that will produce suitable experimental performance during these cycles. The results indicate that for the equilibrium LEU core the experimental performance exceeds that of current HEU operations in all key locations while also demonstrating sufficient safety margins.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Out-of-Pile Test of LVDT-Based Creep Test Rig at PWR Prototypical Conditions

New and improved materials are being considered to support the existing nuclear reactors and future next-generation reactors. The materials can significantly degrade and limit their properties in harsh reactor environments. To accurately understand the material’s degradation, real-time data is required under prototypic-irradiation conditions. Additionally, understanding the creep behavior of materials under harsh environments is essential for an evaluation of safety concerns. To provide these capabilities, the Idaho National Laboratory’s (INL’s) High Temperature Test Laboratory (HTTL) has developed several instrumented test rigs to obtain real-time data from specimens in well-controlled pressurized water reactor (PWR) coolant conditions in the Materials and Test Reactors (MTRs). This technical report focuses on INL’s efforts to evaluate and enhance the former prototype creep test rig that relied on linear variable differential transformers in laboratory settings. Specifically, the test rig is capable of detecting changes in the length of tensile specimen, which is useful for measuring thermal expansion and creep loading.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2023

Progress Towards the Reality of Low-Temperature Critical Experiments with Low-Temperature TEX

Integral benchmarks for criticality safety do not exist below room temperature. Incredible support for below room temperature benchmarks is present across the international community. Validation of low-temperature neutron cross sections is necessary to verify any operation at temperatures below room temperature which is typically observed in environments far from the equator. For example, a fissile material transportation truck may routinely observe ambient temperatures down to -40°C, which is the lower temperature bound of the normal conditions of transportation defined in the United States Title 10 Code of Federal Regulations §71.71c2. Similarly, the IAEA, which contains 176 member states, has the same requirement for subcriticality for fissile material transportation. Additionally, sub-room temperature benchmarks can validate newly produced cross sections, that include novel thermal scattering laws, from North Carolina State University. A six-case experiment designed based off the TEX-HEU baseline experiments has been proposed utilizing a chilled vacuum chamber capable of cooling the HEU stack down to -50°C. The six proposed configurations span the entire neutron fission energy range and can be tested at intermediate temperatures between room temperature and -40°C. Of the six configurations, the fastest case utilizes nearly 135 kg of HEU. Understanding the system is required prior to critical experiment operations. In order to test the functionality of the system, testing with surrogate fuel materials has been conducted. The following discusses the design of the low temperature TEX (LT-TEX) experiment as well as the progress made via surrogate testing towards a reality in which low-temperature integral benchmarks exist in the International Criticality Safety Benchmark Evaluation Project (ICSBEP).

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
NASA NTRS2004

Development of Risk Assessment Matrix for NASA Engineering and Safety Center

This paper describes a study, which had as its principal goal the development of a sufficiently detailed 5 x 5 Risk Matrix Scorecard. The purpose of this scorecard is to outline the criteria by which technical issues can be qualitatively and initially prioritized. The tool using this score card has been proposed to be one of the information resources the NASA Engineering and Safety Center (NESC) takes into consideration when making decisions with respect to incoming information on safety concerns across the entire NASA agency. The contents of this paper discuss in detail each element of the risk matrix scorecard, definitions for those elements and the rationale behind the development of those definitions. This scorecard development was performed in parallel with the tailoring of the existing Futron Corporation Integrated Risk Management Application (IRMA) software tool. IRMA was tailored to fit NESC needs for evaluating incoming safety concerns and was renamed NESC Assessment Risk Management Application (NAFMA) which is still in developmental phase.

Malone, Roy W., Jr.↗
DOE OSTI2022

Sensitivity/Uncertainty Comparison Study Involving IRSN, LANL, and ORNL Tools to Support Validation

Under a DOE Nuclear Criticality Safety Program (NCSP) task involving Analytical Methods, three Laboratories collaborated in a comparison of results obtained from Sensitivity/Uncertainty (S/U) packages relevant to validation of transport codes. The task involves Institut de Radioprotection et de Sûreté Nucléaire (IRSN), Los Alamos National Laboratory (LANL), and Oak Ridge National Laboratory (ORNL) comparing results of MORET 5/MACSENS V3.0, MCNP6.2/Whisper-1.1, and SCALE 6.2.3/TSUNAMI/USLSTATS respectively. All Monte Carlo transport code results utilize nuclear data from ENDF/B-VII.1 evaluation. This study examines five cases from the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook) selected as application models: IEU-MET- FAST-002-001, LEU-COMP-THERM-001-001, LEU-SOL-THERM-004-001, MIX-COMP- THERM-001-001, and U233-SOL-THERM-001-001. This is a continuation of a previous study to examine Pu and HEU cases: HEU-MET-FAST-013-001, HEU-SOL-THERM-001-008, PU-MET- FAST-022-001, and PU-SOL-THERM-001-001. Ultimately, comparison is made between Upper Subcritical Limits (USLs) obtained using each code package for each application case. Since differences exist in whether packages take into account margin of subcriticality (MOS), the USL is computed using only bias and bias uncertainty, also known as the calculational margin (CM) in ANSI/ANS-8.24. Results comparison appears to show that benchmark selection has a greater influence on the USL than the method used for calculation of bias and bias uncertainty.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

A Comparison of RESRAD and GoldSim Models for Assessing Radiological Dose for a RCRA Landfill - 20418

US Ecology Idaho (USEI) operates a Resource Conservation and Recovery Act (RCRA) Subtitle C, Hazardous Waste Treatment, Storage, and Disposal (TSD) Facility on a 640-acre property in Owyhee County, Idaho. USEI accepts a wide variety of RCRA-exempt low activity radioactive wastes, including naturally occurring radioactive material (NORM) and technologically enhanced NORM (TENORM) (USEI 2009). The performance of the landfill with regard to radiological operating permit requirements was previously assessed using a model constructed with the residual radioactivity (RESRAD) computer program, developed by Argonne National Laboratory for the U.S. Department of Energy (DOE). RESRAD supports the evaluation of several environmental transport pathways related to a radionuclide-contaminated soil source term, but it was not conceived as a model for the evaluation of landfill radiological performance. A more comprehensive and realistic model is desired in order to better support submissions to both the State of Idaho's Department of Environmental Quality as well as the U.S. Nuclear Regulatory Commission. A Performance Assessment (PA) computer model was developed using GoldSim software for the USEI RCRA Subtitle C landfill. The RESRAD computer model was specifically developed for calculating soil cleanup criteria and radiological dose and cancer risk from residual radioactive material in soil. To evaluate site-specific conditions, users may select from among a number of available environmental transport and exposure pathways and modify 'default' parameter values. In GoldSim, a user must construct the model 'from scratch.' However, this modeling is facilitated by a number of specialized elements available in GoldSim to support a radiological mass transport model. These include elements for defining radiological decay and ingrowth, container failure and radiological release, advective and diffusive transport, and other processes. Advantages afforded by the use of RESRAD for modeling radiological dose for a RCRA Subtitle C landfill include ease of use and reasonable flexibility in specifying site-specific conditions. GoldSim allows for considerably more flexibility and site-specificity than RESRAD, including evaluation of potentially relevant environmental transport processes not supported in RESRAD. The probabilistic modeling capabilities of GoldSim also far exceed those of RESRAD. The pros and cons of RESRAD and GoldSim for this modeling problem are explored with a focus on identifying approaches and critical factors in identifying the appropriate platform. The capabilities of RESRAD and GoldSim for mathematically modeling the disposal system will be contrasted and compared. For example, the existing RESRAD radiological safety assessment evaluated potentially complete exposure pathways related to infiltration to groundwater and upwards diffusion of radon. Additional transport pathways identified in the USEI Idaho facility Conceptual Site Model include deposition of radon decay products in cover material, root uptake of radionuclides in disposed wastes by native plants, and mixing of cover material by burrowing animals. Results of the two models and how the different transport and dose pathways affect results will be discussed. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

Investigation of Benchmark $k$ eff Sensitivity and Uncertainty for 239 Pu fission in Specific Energy Ranges

Nuclear data at intermediate energies (from 1 to 100s of keV) are evaluated based on scarce differential data and theory unable to capture physics’ expected structure. There is also a lack of integral data. This is a known deficiency and is challenging to address. Calculated effective multiplication factor, k eff , values for intermediate energy experiments are ~25× further from experiment than for fast energies and are often well outside the experimental uncertainties. The goal of the PARADIGM (PARallel Approach of Differential and InteGral Measurements) project is to significantly re duce the uncertainties of intermediate energy nuclear data for 239 Pu. To this end, PARADIGM simultaneously optimizes experiments at both the Los Alamos Neutron Science Center (LANSCE) and National Criticality Experiments Research Center (NCERC). The combined set of data will inform new intermediate-energy nuclear data. By execution of differential and integral experiments, establishment of new theory, and undertaking nuclear data evaluation in parallel, the timeline to deliver improved nuclear data to users will be reduced significantly that is to three years. For the PARADIGM project, it was decided to optimize an integral experiment for two neutron energy ranges, within the full intermediate energy range. The low energy range goes from 1 to 30 keV, while the higher energy range goes from 30 to 600 keV. This work focuses on nuclear data sensitivities and uncertainties for 239 Pu fission for existing experiments in the International Criticality Safety Benchmark Evaluation Project (ICSBEP). When designing new experiments, it is important to understand what benchmarks currently exist. For a more traditional experiment design (in which a specific application model(s) exists), comparisons would be made between the application model(s) and existing benchmarks. For PARADIGM, there is no specific application model, but instead the specific nuclear data reaction and energy ranges of interest can be explored for existing benchmarks.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2022

Evaluation of Oak Ridge National Laboratory Health Physics Research Reactor Operation Data for Critical Benchmark Creation

The Oak Ridge National Laboratory (ORNL) Health Physics Research Reactor (HPRR) was a research reactor designed and built at ORNL in 1961. The critical assembly used a highly enriched uranium and molybdenum alloy as the fuel and could be operated in steady-state or burst modes. The HPRR has recently been the object of an investigation to create a criticality benchmark. Such benchmarks are very important, as they are used primarily to show the accuracy of newly developed modeling codes and to help experimental validation and reactor licensing. The evaluated experiments considered in this paper were carried out between 1974 and 1986 from various HPRR activities such as steady-state subcritical, steady-state critical, and burst prompt super-critical operations of the reactor for dosimetry, irradiation, or training purposes. By using the HPRR experimental logbook information and the as-built drawings of the critical assembly, a highly detailed model of the HPRR was created with SCALE 6.2.4/KENO-VI, and a first version of a critical benchmark of the HPRR was developed following the International Criticality Safety Benchmark Evaluation Project (ICSBEP) guidelines for thorough description and uncertainty/sensitivity quantification. Unfortunately, in most of the evaluated experiments, the obtained difference between calculated and experimental k eff is around 1,000 pcm, corresponding to a relative error of approximately 1%, beyond the quality standards of the ICSBEP recommending a relative error below 0.1%. Moreover, the derived experimental uncertainty is high, around 4% relative, mainly due to the U-Mo fuel density uncertainty, but also from numerous other factors. For these reasons, the creation of a valuable critical benchmark from HPRR operation data is thus far compromised. In this paper, the different steps of the experiments’ evaluation are summarized, and the reasons for the experimental/calculation discrepancies and potential ways to solve them are explored. This paper also aims to remind us always to exercise considerable care when performing experimental work, and to record all the data possible for potential future uses.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗