Removal of bond sodium from Fermi-1 blanket assemblies using a melt-drain-evaporate process
Not Available
Herrmann, Steven D. (ORCID:0000000293734913)↗
Engineering topics
Jarrell, Josh J.
Publications and source records attributed to Jarrell, Josh J..
Development of a MOOSE thermal model of the MPC-32 canister and HI-STORM overpack
Nuclear power is a significant source of electricity in the United States, but the average age of nuclear power plants is around 40 years old. The safe management of the spent nuclear fuel (SNF) is a key aspect of the back-end of the nuclear fuel cycle. Spent fuel dry storage systems are becoming a popular and effective solution in this area, given the absence of a final disposal system. The spent fuel cask system (dry cask method) provides a feasible solution to maintain spent fuel for 60 years before final disposal. Dry cask storage has many characteristics that make it attractive. It fulfills the safety requirements of the Nuclear Regulatory Commission (NRC) while providing modularity and flexibility to contractors. The HI-STORM overpack and MPC-32 canister are the main parts of the HI-STORM 100 dry cask storage system. These components remove heat from the system using natural circulation, requiring no human intervention. This is the characteristic that provides passive heat removal and low maintenance features in dry cask storage systems. To develop a thermal model for a dry cask storage system, the physics behind the system should be defined clearly. There are two natural circulation loops in the system; circulation of helium cools down the nuclear assemblies in the MPC, while circulation of air cools down the walls of the MPC. This work aims to develop a thermal model of the MPC-32 canister and HI-STORM overpack using the Multiphysics Object-Oriented Simulation Environment (MOOSE). MOOSE is an open-source framework developed by Idaho National Laboratory (INL) for multiscale, multiphysics simulations. In this study, we will investigate and demonstrate the thermal-hydraulics modeling capabilities of the MOOSE framework, including natural circulation, heat transfer, and porous flow.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Technical Basis for Extended Dry Storage of Aluminum-Clad Spent Nuclear Fuel
This report addresses previously identified technical knowledge and technology gaps and presents the key findings of the multiyear, EMTD-funded research efforts with the goal of evaluating whether ASNF can be safely stored dry for many decades and whether changes to existing ASNF management practices are appropriate. The findings formulate the technical basis for the extended dry storage of ASNF, leading to the main conclusion: All completed work to close the identified knowledge gaps on extended (>50 years) ASNF dry storage supports the conclusion that extended ASNF dry storage is both safe and viable.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
A High-Assay Low-Enriched Uranium Fuel Transportation Concept
The uranium 235U enrichment commonly used in fuel production for U.S. light water nuclear reactors typically does not exceed 5 wt%. In contrast, many of the currently investigated advanced reactor concepts demand fuel with higher enrichments. This includes high-assay low-enriched uranium (HALEU), characterized by a 235U enrichment of 5 to 20 wt%. The necessity of HALEU transportation in the fuel production cycle leads to new challenges caused by various technical and regulatory hurdles. Current U.S. Nuclear Regulatory Commission–approved transportation package designs for UF6 with enrichments above 5 wt% provide relatively small payloads [=116 kg (250 lb)]. Furthermore, in accordance with 10 CFR 71.55, package design activities for fissile material enriched above 5 wt% need to consider water infiltration in the containment as part of the criticality safety evaluations. This study presents a transportation package concept for HALEU advanced nuclear reactor fuel with a significantly higher payload of up to 376 kg (830 lb) of fissile material per package and up to 1881 kg (4149 lb) of HALEU per legal weight truck. The anticipated chemical form of the transported material is UO2 downblended from available highly enriched uranium. The concept utilizes a combination of existing transportation packaging, 18 inner canisters, and a novel basket design that includes a borated aluminum flux trap. Criticality and shielding evaluations; fundamental structural, confinement, and thermal assessments; and studies on package operations are presented. The results of this study build significant confidence in the technical feasibility of a high-capacity HALEU transportation package concept while demonstrating the concept’s potential to meet U.S. regulatory requirements.
Eidelpes, Elmar↗
Technology Development for Dry Storage of Aluminum-Clad Spent Nuclear Fuel - 20490
A candidate disposition pathway for the > 13 MTHM of aluminum-clad spent nuclear fuel (ASNF), owned and managed by the U.S. Department of Energy, is the drying and placement of the SNF into sealed-canister dry storage, with the ASNF-in- canisters 'road-ready' for transportation to and final direct disposal in a repository waste package. Technical information gaps in fuel drying, and fuel dry storage behavior, have previously challenged the declaration of technology readiness for drying and placement of this fuel into the DOE Standard Canister design for > 50 years of safe dry storage. The principal technical information gaps included: i) characterization and thermal dehydration behavior of aluminum (oxy)hydroxide films attendant on the cladding due to film formation during reactor operation and during post-discharge up to long-term wet storage histories; and ii) G-value data to enable estimation of the radiolytic gas generation from the cladding with its (oxy)hydroxide films. Thus, the oxide films on the ASNF challenged the safety of a sealed storage canister with thermal and radiolytic decomposition of the waters on the films that can lead to corrosion, pressurization, and flammability issues. These gaps in the technical information base have largely been closed. This paper discusses the investigations at the Idaho National Laboratory (INL) and the Savannah River National laboratory (SRNL), and outlines the pending technology development work for input to an engineering design to enable a road-ready dry storage system for ASNF. The ASNF inventory considered for road-ready dry storage is stored at the Savannah River Site (SRS) and at the Idaho Nuclear Technology and Engineering Center (INTEC) at the INL. The ASNF inventory in the SRS L Basin is from foreign and domestic research reactors (FRR and DRR), and is diverse in terms of design, irradiation, and post-reactor-discharge storage conditions; these factors yield a range of characteristics of cladding oxide films on ASNF. Mixed aluminum (oxy)hydroxide (boehmite and bayerite/gibbsite) films, non-uniform in thicknesses up to a maximum local thickness bounded by 25 μm, were observed on ASNF materials removed from wet storage in the L Basin and in non-sealed dry storage at the SRS following reactor service and a long-term (up to 40+ years) interim storage history. The ASNF inventory at INTEC, in both wet and vented dry storage, is predominantly from the Advanced Test Reactor (ATR), but it also includes DRR and FRR fuel. To address a profound behavior of these films, radiolysis testing of aluminum specimens with mixed type boehmite/bayerite oxide films was performed to develop basic data on G-values for production of radiolytic hydrogen under dry storage conditions with nominal relative humidity, temperature, and cover gas. Modeling and simulation of canister internal environments with postulated inventories of oxide films provides estimation of the evolution of the conditions of the canister loaded with ASNF. Simultaneous Thermal Analysis (TGA/DSC) of hydrated oxide powders, and laboratory-scale aluminum specimens with a bayerite film (∼10 μm) using TGA methods, inform drying time/temperature conditions to be used for the ASNF. A demonstration project is recommended for Verification and Validation of the drying and storage of the ASNF. Remaining major tasks leading up to the hot demonstration include scale-up radiolysis testing and scale-up drying testing. Engineering design with the information from the technology program will establish the safety basis and enable long-term (> 50 years) dry storage compatibility with ASNF in the DOE Standard Canister pending its transportation to and disposal in a repository. This full capability would show the ASNF-in-canister storage system to be road-ready. (authors)
12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
UO 2 HALEU Transportation Package Evaluation and Recommendations
This M3FT-19IN030205042 milestone report presents the current status on investigations related to the development of HALEU transportation capabilities in the U.S. The report initially reviews potential short-term HALEU production options and describes an expected HALEU composition. It then discusses the regulatory bases for HALEU transportation and identifies two compliant packaging designs: (1) the TN Americas TN-LC and (2) the NAC International OPTIMUS™-L. To modify these designs for HALEU transportation, an inner canister concept, and modular metal and foam basket concepts were developed, and this is described in the report. The expected activity, criticality tendency, radiation level, and decay heat of the package content are evaluated considering the specific system characteristics. Based on the results of the investigations, the research team consisting of INL, PNNL, and ORNL staff concluded that, given the anticipated source material composition, the OPTIMUS™-L, or a comparable design, is the preferred candidate. This is due to its larger HALEU payload per LWT, lighter weight, and simpler handling procedures compared to the TN-LC. Thus, ongoing research efforts for the scope of this project will focus on more thorough evaluations of the OPTIMUS™-L packaging for HALEU transportation. A tentative timeline is presented for the remaining work that is planned for FY 2020. The goal is the development of F&Rs that provide guidance to a potential HALEU transportation package vendor on critical specifications by August 2020.
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗