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

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At least 181 records · Page 10
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Radiation dose rate distributions of spent fuel dry casks estimated with MAVRIC based on detailed geometry and continuous-energy models

This study presents a detailed comparison of the dose rate distributions of a dry TN–32 fuel cask with two geometry models and two cross-sectional datasets. The accuracies of radiation dose rate estimation and computational efficiencies of each geometry model with two cross-sectional data-sets are compared. The use of automated variance reduction techniques can significantly improve the computational efficiency of such a realistic, deep penetration problem that involves radiation transport from different volumetric sources, thereby eliciting only a small statistical error. Monaco with Automated Variance Reduction using Importance Calculations (MAVRIC) is a computational sequence within the SCALE 6.2 code package based on consistent adjoint driven importance sampling (CADIS), a type of automated variance reduction technique. Homogenous and full fuel assembly models are built herein, and two nuclear cross-section libraries (V7–200N47G and continuous energy) are applied in this work. Based on the detailed comparisons, we found that neutron dose rate estimation is more dependent on geometry modeling than on cross-section data. For neutron-induced gamma rays, the dose rate distribution depends on both the spatial self-shielding effect and the cross-section library. The primary gamma rays respectively contribute to the total dose rate by ~91% and ~ 99% on side and top surfaces, and the dose rate accuracy is more dependent on the cross-section library than on geometry modeling. In terms of the computation efficiency and efforts spent on geometry modeling, the homogenous fuel assembly model with the MG library can produce an acceptable dose rate distribution, but the detailed fuel assembly model with the continuous-energy library is required for more precise dose rate estimation.

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2022

A risk-informed assessment tool to support review of license amendment requests for spent nuclear fuel dry cask storage systems

We report that while risk-informed safety evaluations are common methodologies used by the U.S. Nuclear Regulatory Commission (NRC), their application in the field of commercial spent nuclear fuel (SNF) dry storage has received limited attention. For instance, license amendment requests (LARs) for SNF dry cask storage systems (DCSSs) are evaluated according to U.S. NRC standard review plans (SRPs). However, risk-informed LAR review strategies could improve the understanding of critical system modifications, and lead to more predictable, efficient, and consistent LAR review processes. A methodology to develop a risk-informed LAR review tool is presented. Although focused on DCSSs, this methodology can be used to develop similar tools for a broad spectrum of safety-relevant systems. The end product (i.e., the tool) includes a tree diagram to help visualize the review risk significances assigned to a predefined set of system modifications and supporting rationale documentation. Instructions support correct tool application and provide the user with a path to incorporate currently unevaluated modifications in the tool structure. The demonstration of the methodology leads to the conclusion that risk-informed reviews of DCSS LARs are possible, although the tool precision could benefit from additional risk information. Further, the results indicate that many typical LARs involve low-risk or medium-risk modifications.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

An Alternative to the Continued Accumulation of Separated Plutonium in Japan: Dry Cask Storage of Spent Fuel

As of the end of 2017, Japan had a stockpile of 48 tons of separated plutonium. It will take more than a decade for Japan to convert most of that plutonium into “mixed-oxide” (MOX) fuel and load it into power reactors licensed to use such fuel. This is a very costly program. Including the cost of reprocessing, MOX fuel costs about ten times more than the low-enriched-uranium fuel that otherwise would be used by these reactors. Yet Japan has a policy to start separating more plutonium from spent reactor fuel as soon as the Rokkasho Reprocessing Plant can be put into operation – currently projected for 2021 after 24 years of delay due to a variety of technical problems and upgrades in safety requirements. Japan’s government remains committed to this costly and troubled program primarily because, pending the availability of additional spent-fuel storage space on or off site, reprocessing is seen as the only way to get spent fuel out of reactor pools that are getting full. Because of concerns about the safety of dense-packed spent-fuel pools, however, the prefectures and towns hosting Japan’s power reactors are beginning to accept on-site storage of spent fuel in dry casks. This is the option the United States and many other countries chose after they cancelled their reprocessing programs and their powerplant pools filled up. Its increasing availability in Japan, if backed by a determined policy to accelerate its introduction, should make it possible to cancel reprocessing in Japan as well.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Analytical Sensitivity Analysis of a Spent Nuclear Fuel Cask

Here, we report nuclear science and engineering is a field increasingly dominated by computational studies resulting from increasingly powerful computational tools. As a result, analytical studies, which previously pioneered nuclear engineering, are increasingly viewed as secondary or unnecessary. However, analytical solutions to reduced-fidelity models can provide important information concerning the underlying physics of a problem and aid in guiding computational studies. Similarly, there is increased interest in sensitivity analysis studies. These studies commonly use computational tools. However, providing a complementary sensitivity study of relevant analytical models can lead to a deeper analysis of a problem. This work provides the analytical sensitivity analysis of the one-dimensional (1D) cylindrical mono-energetic neutron diffusion equation using the forward sensitivity analysis procedure (FSAP) developed by Cacuci. Further, these results are applied to a reduced-fidelity model of a spent nuclear fuel cask, demonstrating how computational analysis might be improved with a complementary analytic sensitivity analysis.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Computational Fluid Dynamic Modeling of Dry Cask Simulator with Crosswind

The purpose of this study is to create a STAR-CCM+ model of a Belowground Vertical Dry Cask Simulator (BVDCS) at Sandia National Laboratories (SNL) and validate the model with SNL’s experimental results. The BVDCS consists of a single boiling water reactor assembly fitted with electric heaters encompassed by a containment vessel and shell to represent a belowground spent nuclear fuel (SNF) dry storage system. Blowers are located near the inlet and outlet of the BVDCS to simulate crosswind conditions. In addition to the experimental results, the STAR-CCM+ model developed for this study is compared with a previous computational fluid dynamics (CFD) model in a different software program, which is used as a software-to-software benchmark. The experimental results provide a dataset to compare the STAR-CCM+ model results for a variety of different conditions. The main objective is to validate and improve STAR-CCM+ CFD models for spent nuclear fuel storage systems with explicitly modeled external environments and “wind driven” crossflows. These CFD models aide in the study of external particle deposition in spent nuclear fuel storage systems, which is important to predicting the significance of chloride induced stress corrosion cracking (CISCC). In addition to experimental comparison, a sensitivity analysis study is performed using the STAR-CCM+ model. The sensitivity analysis provides a quantitative assessment of the sensitivity of various parameters. This helps provide information on various parameters that are of particular importance to constructing a model representative of real life systems. The STAR-CCM+ model compared well to the experimental results showing similar responses to changes in cross wind flow, and a number of parameters are identified for model improvement.

Jensen, Ben J.↗
DOE OSTI2015

Dry Cask Storage Verification with Muon Radiography [Slides]

Cosmic ray muons can penetrate significant shielding that defeats typical radiographic probes. Preliminary results on identifying partial defects in dry cask contents are promising. Further modeling and data analysis are underway.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Creation of a Geant4 Muon Tomography Package for Imaging of Nuclear Fuel in Dry Cask Storage

This is the final report of the NEUP project “Creation of a Geant4 Muon Tomography Package for Imaging of Nuclear Fuel in Dry Cask Storage”, DE-NE0000695. The project started on December 1, 2013 and this report covers the period December 1, 2013 through November 30, 2015. The project was successfully completed and this report provides an overview of the main achievements, results and findings throughout the duration of the project. Additional details can be found in the main body of this report and on the individual Quarterly Reports and associated Deliverables of the project, uploaded in PICS-NE.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Experimental Determination and Modeling of Used Fuel Drying by Vacuum and Gas Circulation for Dry Cask Storage

More than 120 drying tests were conducted using a mock fuel assembly with depleted uranium rods and heater rods to simulate decay heat. These tests followed standard industry practice for vacuum and forced helium drying (FHD). Both single effect tests evaluating a single fuel assembly or cask feature and combined effect tests were conducted. Single effect tests evaluated drying of a specific, known amount of water in a feature. Combined effect tests evaluated specific features and followed the flooding of the chamber, dewatering, and blowdown procedures before beginning the drying procedure. Specific simulated fuel assembly and canister features evaluated included PWR dashpot, BWR water rod, failed fuel rod, spacer disc, and Boral sheet. Ceria pellets were fabricated and used in the failed fuel rod to avoid issues of uranium contamination. A method of monitoring water content in the canister gas and gas stream was developed using optical emission spectroscopy (OES) in addition to monitoring of relative humidity in the chamber and in the vacuum lines.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Advanced Concepts for Dry Storage Cask Thermal-Hydraulic Testing

The purpose of this report is to review technical issues relevant to the performance evaluation of dry storage systems during vacuum drying and long-term storage operations. It also provides updates on experimental components under development that are vital for pursuing advanced studies. Validation of the extent of water removal in a multi-assembly dry storage system using an industrial vacuum drying procedure is needed, as operational conditions leading to incomplete drying may have potential impacts on the fuel, cladding, and other components in the system. Water remaining in canisters/casks upon completion of vacuum drying can lead to cladding corrosion, embrittlement, and breaching, as well as fuel degradation. Therefore, additional information is needed to evaluate the potential impacts of water retention on extended long-term dry storage. A general lack of data and experience modeling the drying process necessitates the testing of advanced concepts focused on the simulation of industrial vacuum drying. Smaller-scale tests that incorporate relevant physics and well-controlled boundary conditions are necessary to provide insight and guidance to the modeling of prototypic systems undergoing drying processes. This report describes the development and testing of waterproof, electrically-heated spent fuel rod simulators as a proof of concept to enable experimental simulation of the entire dewatering and drying process. This report also describes the preliminary development of specially-designed, unheated mock fuel rods for monitoring internal rod pressures and studying water removal from simulated failed fuel rods. A variety of moisture monitoring instrumentation is also being considered and will be downselected for the tracking of dewpoints of gas samples. The effects of cladding oxidation and crud on water retention in dry storage systems can be explored via separate effects tests (SETs) that would measure chemisorbed and physisorbed water content on cladding samples. The concepts listed above will be incorporated into an advanced dry cask simulator with multiple fuel assemblies in order to account for important inter-assembly heat-transfer physics. Plans are described for harvesting up to five full-length 5x5 laterally truncated assemblies from commercial 17x17 PWR skeleton components with the goal of constructing this simulator.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Verification of Spent Fuel Inside Dry Storage Casks by Fast/Epi-thermal Neutron Mapping: FY2020 Annual Report to NA241 (Annual Report)

Lawrence Livermore National Laboratory has embarked upon development of a novel methodology for verification of spent fuel inside dry storage casks. The issue of verification of the integrity of spent fuel inside dry storage has been one of the most challenging technical problems facing IAEA for many decades. The difficulty arises because of the massive amount of shielding between the inner spent fuel and the outer surface where measurements can be performed. Despite many efforts, detection of useful gamma rays or neutrons from the outer surface has so far proven unsuccessful.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

COBRA-SFS Version 6.2.1 User Guide: A Thermal Hydraulic Analysis Code for Spent Fuel Storage and Transportation Casks

This document is an updated and reduced version of the complete documentation of the COBRA-SFS thermal analysis code (COolant Boiling in Rod Arrays – Spent Fuel Storage) (Michener et al. 1995). This manual provides the analyst with input instructions and fundamental guidelines for contrasting models using COBRA-SFS. COBRA-SFS is designed to analyze the dry storage and transportation casks that house spent or used nuclear fuel. The primary focus and application is on square light water reactor fuel assemblies, however the code also has the capability to model hexagonal assemblies and is will suited to any geometry that can be defined as an extruded 2-D map.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Microreactor Assembly Transportation Cask Model Description for Criticality Safety Validation Basis Assessment

Criticality safety analyses are completed on a transportation cask used for microreactor assembly shipment to provide an example of model and analysis to industry for reproducing this type of study on their microreactor fuel shipment. The fuel assembly considered is based on a gas-cooled microreactors (GC-MR), which utilizes HALEU fuel in the form of TRISO particles and utilizes various design options considered in industry designs. Various versions of this GC-MR assembly were studied, with and without YH 2 moderator, providing similar conclusions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Reassessment of Dose Contributions from the Paducah Cask Special Waste Form Buried in the E-Area Low-Level Waste Facility

During a recent revision to the groundwater and inadvertent human intruder screening analysis report (Aleman and Hamm, 2023), weaknesses within the screening logic were observed resulting in the addition of U-235 being brought back into the list of radionuclides requiring inventory limits within the Paducah Cask special waste form (SWF). In addition, while reviewing the earlier vadose zone transport analysis efforts, an inconsistency was observed in the initial U-235P inventory that was supplied to the Consolidated Waste Tracking System (CWTS) back in 2009.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2026

Microreactor Core Transportation Cask Model Description for Criticality Safety Validation Basis Assessment (Rev. 1)

Criticality safety analyses are completed on transportation casks used for microreactor whole core shipment to provide examples of models and analyses to industry, regulators, and nuclear community at large to be used in verification and validation analyses of similar applications. The microreactors considered are based on a Gas-Cooled Microreactor (GCMR) and a Heat-Pipe Microreactor (HPMR), both utilize HALEU fuel in the form of TRISO particles and various other design options considered in industry microreactor designs. Variant design options of GCMR and HPMR were also investigated to provide a wider application range for each technology. Criticality safety analyses for the GCMR and HPMR packages were performed using the CSAS6 sequence of SCALE 6.3.2 with the ENDF/B-VII.1-based continuous energy neutron libraries. Different scenarios were investigated, including normal operation and water flooded conditions to represent nominal and hypothetical accident scenarios. Sensitivity and similarity analyses are also performed using the TSUNAMI sequence of SCALE 6.3.2, and the similarity analysis uses all the experiments from the ICSBEP Handbook with Intermediate and mixed Enriched Uranium (IEU) and Low Enriched Uranium (LEU) systems, together with additional experiments that are sponsored by the DNCSH program. Many experiments were found marginally similar to the GCMR and HPMR models, with similarity index (ck) values greater than 0.8 but less than 0.9. Among all the experiments analyzed, no case has a ck value greater than 0.9, indicating that additional critical experiments might be needed to further validate the criticality safety models for microreactors transport packages.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Disposition of Chicago Pile 5 (CP-5) Converter Tubes in the 10-160B Cask

This paper will focus on the unique characterization, packaging, and transportation issues associated with the disposition of the two CP-5 Converter Tube assemblies from Argonne National Laboratory. The converter tubes were constructed of combinations of HEU and alloys of zirconium, and were part of the original research facilities attached to the CP-5 reactor during operating evolutions. These assemblies were heavily irradiated during their operational lifetime, and were segregated from the balance of irradiated test specimens when the reactor was deactivated and slated for Decontamination and Demolition (D&D). In addition, the substantial contribution of fissile material to the assemblies’ inventory made the potential disposition pathways extremely challenging. As a result, these items became part of Argonne’s legacy “nuclear footprint”, and were added to the Nuclear Footprint Reduction Project scope for disposition. The Project was responsible for the size reduction and characterization of these items, as well as the ultimate disposition. After negotiating a disposal pathway for these tubes, there were significant transportation issues that required a small team to overcome, in order to successfully ship these items to the Nevada National Security Site (NNSS). The Project team at Argonne, technical support from transportation specialists, licensing support from the 10-160B license owner, the Savanah River National Lab (SRNL) Packaging Certification Team (PCT, and the DOE EM-33 staff contributed to license and safety analysis report amendments that eventually authorized the shipment of the material. The paper will identify the organizations, and the specific actions, required to successfully make three “one of a kind” shipments of irradiated test specimen material. This will include the unique packaging configurations, contents modification for the cask license (via the Amendment process), criticality evaluations, and associated review and approval processes.

CP-5↗
DOE OSTI2018

Muon-Computed Tomography using POCA Trajectory for Imaging Spent Nuclear Fuel in Dry Storage Casks

Cosmic ray muon-computed tomography (μCT) is a new imaging modality with unique characteristics that can be particularly important for applications in nuclear proliferation detection and international treaty verification. Using cosmic ray muons for nuclear security presents several potential advantages. Among others, muons are generated naturally in the atmosphere, can penetrate high-density materials, and are freely available. No radiological sources are required and consequently there is no associated radiological dose. Recently, the feasibility of using muons for imaging spent nuclear fuel stored in shielded casks has been explored and has been proved beneficial. However, challenges in μCT imaging include low muon flux of ~10,000 muons/m2/min, the effects of multiple Coulomb scattering (MCS) blurring the image, and inefficiency in being able to use all recorded muons for imaging. In this paper, we argue that the use of muon tracing should produce tomographic muon images with improved quality – or more quickly for the same image quality – compared to the case where conventional methods are used. In our paper, we report on the development and assessment of a novel muon tracing method for μCT. The proposed method back projects the muon’s scattering angle into each pixel crossed by its PoCA trajectory then forward projects the variance of the scattering angle in each pixel to detector bins along the muon’s incident horizontal direction. Two scenarios were simulated to assess the expected detection capability of this proposed method. GEANT4 was used to model the main characteristics of 1-60 GeV muons through matter. The simulated images showed an expected improvement in resolution and a reduced the reliance on the muon momentum information compared to a more conventional muon tomography method.

Liu, Zhengzhi↗