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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 145 records · Page 8
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

A Scintillation-Fiber Based Cosmic-ray Muon Tomography System for Imaging Dry Storage Cask

Because of the growth of the nuclear power industry in the United States and the policy decision to ban reprocessing of commercial spent nuclear fuel, the spent fuel inventory at commercial reactor sites has been increasing. Spent nuclear fuel presents extra complexity of security and proliferation issues. These fuel assemblies are practically inaccessible for inspection purposes, as reopening a DSC would require special facilities and would be tremendously expensive. Cosmic-ray muon Tomography has been studied in the past for various applications and was demonstrated to be a good non-destructive assay technique for high-Z materials, such as spent nuclear fuel assemblies in dry storage casks (DSC). Compared to conventional imaging techniques, muon imaging has some advantages due to its negligible health risk and high penetrating power. In this paper, we report our simulation study on imaging a DSC using muon Tomography technique. The Point of Closest Approach (POCA) and the Maximum-Likelihood Expectation- Maximization (ML-EM) algorithms are implemented and evaluated. A scaled-down prototype muon tomography system is proposed and studied. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Remote Monitoring for Stress Corrosion Cracking in Dry Cask Storage Systems

Acoustic Emission monitoring is a technique that detects and records phenomena related to the elastic waves generated by the release of strain energy during crack propagation events. The sensors used for this method are highly sensitive and do not typically need to be in relatively close proximity to the indication when compared to methods such as ultrasonic or eddy current testing. This characteristic makes acoustic emission an attractive method for monitoring locations that are not readily accessible to active inspection methods, such as with in-situ examination of dry cask storage system canisters. Acoustic emission is unique among nondestructive evaluation (NDE) methods as emission is caused by the energy released from the object inspected. Most NDE methods require that energy be introduced to the part from an external source; which is not necessary with acoustic emission. Acoustic emission signals are created during crack growth and plastic deformation. EPRI has been investigating the feasibility of using acoustic emission as a complementary NDE method for monitoring the structural condition of dry fuel storage canisters. To assess this feasibility three first of a kind field trials were conducted over a one year period. The goals of the field trials were to determine the potential path forward for the method and to assess needs for further research and development. Why this matters: The study evaluated the efficacy of structural health monitoring methods to support ISFSI license renewal commitments to support aging management of welded stainless steel canisters. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Inspection of Dry Storage Casks Using PGAA and NAA Techniques

In this effort, an assessment of a chloride detection method is performed for inspecting the surfaces of dry cask storage system (DCSS) canisters, with focus on the implementation of a non-destructive nuclear analysis and methodology. Some DCSS canisters (especially those located in coastal environments) will be exposed to marine environments, which can cause atmospheric stress corrosion cracking (SCC). Chloride deposits on canister surfaces provide the corrosive environment necessary for atmospheric SCC with the residual stresses and materials properties conditions. The most susceptible regions on the canister surface are the weld areas, or where cold work has been performed, and may provide the necessary driving force for crack growth. The heat-affected zones (HAZs) near welds are also potentially susceptible to atmospheric SCC due to sensitization of the material caused by chromium depletion during the welding process. Information collected from the field and from laboratory studies has not been able to rule out the possibility of atmospheric chloride induced stress corrosion cracking (CISCC), although no occurrences of atmospheric SCC in DCSS canisters have been detected to-date. A number of researchers have demonstrated that localized corrosion can take place under such situations on materials commonly used for interim storage containers, including 304SS (Cook et al. 2010; Shirai et al. 2011) and 316SS (Tani et al. 2009). In this summary, the preliminary results of a neutron activation analysis (NAA) and prompt gamma activation analysis (PGAA) are performed using MCNP6, a highly sophisticated, multi-particle Monte Carlo simulation code (with the employment of the high performance computing system at the Idaho National Laboratory), and confirm the feasibility of using PGAA and NAA as an environmental inspection methodology for detection and monitoring of small quantities of chlorine on canister surfaces. The specific signature of gamma rays due reactions of sea salt constituents are documented and summarized in this paper. The environmental constituent for passive oxide film breaching is chlorine, which can cause pitting initiation and incubation time that may lead to cracking in stainless steel. Therefore, an assessment of chloride presence at a canister surface is essential before moving to other nondestructive analysis (NDA) methods for detection of pitting and cracking. Two nonlinear acoustic techniques will also be used in our integrated research project (IRP) research study that proves to be useful in the detection of SCC: nonlinear resonant ultrasound spectroscopy (NRUS), and time revisited elastic nonlinear diagnostic (TREND). These two techniques are complementary to marine atmospheric detection techniques. In this presentation only PGAA and NAA are described in detail. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Development of Friction Stir Processing for Repair of Nuclear Dry Cask Storage System Canisters

The Nuclear Regulatory Commission has identified chloride-induced stress corrosion cracking (CISCC) of austenitic stainless steel dry cask storage systems (DCSS) as an area of great concern. Friction Stir Processing (FSP) was used to repair laboratory-generated stress corrosion cracking (SCC) in representative stainless steel 304 coupons. Results of this study show FSP is a viable method for repair and mitigation CISCC. This paper highlights lessons learned and developed techniques relative to FSP development for crack repair in sensitized thick section stainless steel 304. These include: development of process parameters, welding at low spindle speed, use of weld power and temperature control and optimization of these controls. NDE and destructive analysis are also presented to demonstrate effectiveness of the developed methods for SCC crack repair.

Friction Stir↗
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 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↗