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A lid system for a shielding cask is provided, the system comprising a plurality of contiguous panels, each panel is axially actuated by a first lift piston located laterally from the casks' longitudinal principal axis, and rotated about the cask by a ‘hinge’ piston at approximately 90 degrees along an arc from the first piston, the arc formed by a portion of the lip of the cask.
After the accident at Fukushima in 2011, 8 nuclear power plants (NPPs) in Germany were shut down immediately and parliament decided a phase-out of nuclear energy by 2022. As prerequisite for a quick dismantling of the shut-down NPPs, all fuel has to be removed from the spent fuel pools as soon as possible. For over 20 years, spent fuel assemblies from German NPPs have already been stored in CASTOR dual purpose casks by GNS. Already more than 1200 CASTOR casks are stored in dry interim storage facilities worldwide, and there is an additional demand for several hundred more casks for the remaining spent fuel assemblies in Germany until the final phase-out. Furthermore the phase-out decision has triggered the demand for a complementary solution for the damaged fuel rods that have been collected separately in the pools for disposal after the NPP shut-down. In 2011, GNS started to develop a disposal concept for these damaged fuel rods. For that purpose, GNS designed 2 quivers for damaged fuel rods to fit into square tubes of the basket of the CASTOR V/19 (PWR) and the CASTOR V/52 (BWR) respectively. The customisable internal baskets of the quivers facilitate the disposal of a large variety of nuclear inventory. Furthermore, the quiver features a robust design and a welded closure system. This design has been affirmed by a series of tests and qualification processes, supervised by the German authorities. This first of its kind solution ensures the dry interim storage of damaged fuel rods. The package design approval for the quiver for CASTOR V/19 is expected for 2016. This presentation will focus on the design of the Quivers in combination with the CASTOR V casks and describe the challenges during the licensing procedures. We will show the components of the Quiver and explain their mechanical characteristics during accident conditions of transport as well as the drying and welding process necessary to enable the long term storage of the Quivers. (authors)
This document serves to outline the testing of a Used Fuel Cask Identification Detector (CID) currently being designed under the DOE-NE MPACT Campaign. A bench-scale prototype detector will be constructed and tested using surrogate neutron sources. The testing will serve to inform the design of the full detector that is to be used as a way of fingerprinting used fuel storage casks based on the neutron signature produced by the used fuel inside the cask.
Uncertainty is a key metric in computational modeling that must be evaluated for results to have wide ranging applicability. A well characterized uncertainty range is ideal with clear error bars on results that can be presented to stakeholders. In the field of spent fuel cask modeling, this ideal has been historically difficult to achieve in practice because of the computationally intensive nature of the models used and the difficulty assigning reasonable uncertainties to quantities in as-built systems. The work in this report has been conducted to evaluate the overall state of uncertainty and sensitivity in spent fuel cask models and develop methodologies for evaluating these uncertainties. These methodologies must be practical for engineering applications. They should not require excessive computational resources or calendar time to achieve results. In engineering, the model must be on a scale such that it can be changed and adapted throughout a project as new information is discovered and project goals evolve. This report covers three major modeling task areas that provide an overview of the types of sensitivity and uncertainty present in a spent fuel storage and transportation system. Section 3 discusses sensitivity and uncertainty analysis in the effective thermal conductivity model for the fuel region and applies these results to a single assembly model. Section 4 shows sensitivity analysis of a full cask model in the TN-32B and Section 5 demonstrates the overall uncertainty workflow using Coolant Boiling in Rod Arrays – Spent Fuel Storage and STAR-CCM+ developed from the sensitivity work in the preceding sections.
Nuclear power plants have been supplying resilient and reliable electricity for decades, contributing to energy independence of the U.S.. However, nuclear waste management remains one of the most significant challenges in the industry. The safety of dry cask storage systems relies heavily on their thermal-hydraulic performance. Computational Fluid Dynamics (CFD) simulations are often used to demonstrate this performance and ensure that the system design meets safety standards. This study presents reduced numerical models for various types of dry cask systems. These numerical models can produce efficient and fast results based on the employed modeling strategies. Additionally, the study uses a novel approach to high-fidelity simulations to evaluate modeling assumptions in dry cask modeling. Large Eddy Simulations (LES) are used for this purpose, particularly in regions where fluid velocity is relatively high and the turbulence characteristics become important. Furthermore, the results of these high-fidelity simulations will enhance the interpretation of outcomes produced from a lower-fidelity CFD model.
The Department of Energy (DOE), in partnership with its national laboratories and the National Aeronautics and Space Administration (NASA), is responsible to produce Pu-238 isotope in the United States for use in space exploration. Major activities in the DOE complex are focused at Idaho National Laboratory (INL) and Oak Ridge National Laboratory (ORNL). INL is responsible for storing Np-237 feedstock, irradiation qualification in the Advanced Test Reactor (ATR), and irradiating targets containing Np-237 to produce Pu-238. ORNL is responsible for target design, target fabrication, irradiation qualification in HFIR, and processing of targets to extract Pu-238 heat source material. A key part of the program lifecycle is moving irradiated targets from INL to ORNL. The BEA Research Reactor (BRR) cask was identified as a potential shipping cask for the transport of both unirradiated and irradiated targets between the project sites. This paper will discuss the production at INL and shipment of Pu-238 to ORNL using the BRR cask.
For the characterization of spent fuel and for the safe direct disposal of SNF we present: a) a method of numerical assessment of their radionuclides inventory that is derived from the initial fuel characteristics and individual burn-up / cooling time history of that particular fuel assembly, and b) how the numerical information can be experimentally assessed using muon tomography as a nondestructive assay of large and heterogeneous SNF casks. We combine both complementary methods, i.e. numerical assessment and muon tomography (MT) experimental assay, to provide a powerful tool for verifying the nuclear inventory without compromising the casks or radioactive material containers. Here, we shall present the numerical assessment technique and preliminary but very promising experimental results, and explain the scope of a versatile and mobile MT-detection system for the identification of nuclear material inside large casks containing spent fuel assemblies and/or other large volume heterogeneous material containers for which the European Commission has recently granted R and D funding within the CHANCE project to be developed. (authors)
To ensure criticality safety and maintain a high level of security of Used Nuclear Fuel (UNF), it is necessary to perform detailed Particle Transport simulations. The commonly used technique is the 3-D, continuous energy Monte Carlo method that is time-consuming and, for eigenvalue calculations, may have difficulties with the source convergence. Wenner and Haghighat and Dufek demonstrated that in criticality problems with high dominance ratio and/or under-sampling possibilities, the fission matrix (FM) methodology should be used in order to be able to obtain unbiased solution. The RAPID (Real-time Analysis for Particle transport and In-situ Detection) code system is based on the Multi-stage Response-function Transport (MRT) methodology, and uses the FM formulation to determine eigenvalue, subcritical multiplication and fission density distribution of nuclear systems. This paper uses a benchmark problem similar to the GBC-32 Cask to evaluate the accuracy and performance of RAPID against the results of a reference MCNP solution. We performed a detailed sensitivity analysis on a single-assembly model in order to determine the MCNP eigenvalue parameters necessary to obtain a reliable reference solution. These results provided a guide for the selection of an appropriate set of parameters for the eigenvalue calculation for the whole cask. The paper is organized as follows: initially, the RAPID code system is briefly discussed; then, the benchmark problem is introduced. The methodology used to analyze results is proposed and some of the results of different cases are analyzed to identify the best set of eigenvalue parameters for a reference MCNP calculation. The selected case is then compared with the RAPID calculation for both a single assembly and a full cask model. In the end some conclusions and future work perspectives are drawn.
Used Nuclear Fuel Storage, Transportation and Disposal Analysis Resource and Data System (UNF-ST[and]DARDS) is a comprehensive data and calculation capability for UNF safety assessments currently being developed at Oak Ridge National Laboratory through a collaborative effort of multiple national laboratories and industry participants. The purpose of UNF-ST[and]DARDS is to provide a unified domestic UNF system of data and analysis capabilities in support of U.S. Department of Energy waste management and fuel cycle-related objectives. UNF-ST[and]DARDS is designed to provide safety assessments for UNF storage, transportation, and disposal systems based on actual characteristics of the as-loaded UNF (e.g., fuel assembly burnup, enrichment, decay time, cladding integrity, etc.). The cask-specific safety evaluations allow quantification of realistic safety margins and conditions and provide the technical basis for addressing concerns related to potential fuel failure (e.g., for high-burnup fuel). This paper describes the UNF-ST[and]DARDS methodology for calculating radiation dose rates for U.S. Nuclear Regulatory Commission (NRC) licensed casks based on actual UNF canister contents. Realistic dose rate values based on actual canister contents may be used in place of bounding dose rate values to support development of repackaging operations procedures, evaluation of radiation-related transportation risks, and communication with stakeholders. UNF-ST[and]DARDS capabilities have been recently expanded to include automated radiation source term calculations and dose rate analyses of as-loaded transportation casks. This unique capability determines realistic dose rate values based on actual canister UNF contents that may be used in radiation dose assessments to demonstrate safe transportation of commercial UNF. (authors)
Spent fuel transfer cask is used to transfer a spent fuel from the reactor tank to the spent fuel storage or for spent fuel inspection. Typically, the cask made from steel cylinders that are either welded or bolted closed. The cylinder is enclosed with additional steel, concrete, or other material to provide radiation shielding and containment of the spent fuel. This paper will discuss the Conceptual Design of fuel transfer cask for Reactor TRIGA Puspati (RTP)
Dry cask storage is a method for interim storage of spent fuel assemblies which contain fissile isotopes of uranium and plutonium. These can present a proliferation concern and consequently there is a need for non-destructive testing methods to verify a dry cask's contents for proliferation protection. We present an application of adjoint sensitivity analysis and data assimilation to a multigroup diffusion model of dry cask storage. Adjoint sensitivity analysis allows the efficient calculation of sensitivities for use in data assimilation to calibrate imprecisely known parameter values and data consistency tests to detect diversion scenarios. (authors)
Aluminum-clad spent nuclear fuel will be transported for processing in the 70-ton nuclear fuel element cask from L Basin to H-canyon. During transport these fuels would be expected to experience high temperature aqueous corrosion from the residual L Basin water that will be present in the cask. Cladding corrosion losses during transport were calculated for material test reactor (MTR) and high flux isotope reactors (HFIR) fuels using literature and site information on aqueous corrosion at a range of time/temperature conditions. Calculations of the cladding corrosion loss were based on Arrhenius relationships developed for aluminum alloys typical of cladding material with the primary assumption that an adherent passive film does not form to retard the initial corrosion rate. For MTR fuels a cladding thickness loss of 33 % was found after 1 year in the cask with a maximum temperature of 260 {degrees}C. HFIR fuels showed a thickness loss of only 6% after 1 year at a maximum temperature of 180 {degrees}C. These losses are not expected to impact the overall confinement function of the aluminum cladding.
At present, the United States does not have a designated disposal site for used nuclear fuel — the nation therefore faces the prospect of extended long‐term storage (i.e., >60 years) and deferred transportation of used fuel at operating and decommissioned nuclear power plant sites. Under current U.S. federal regulations contained in Title 10 of the Code of Federal Regulations (CFR) 72.42, the initial license term for an Independent Spent Fuel Storage Installation (ISFSI) must not exceed 40 years from the date of issuance (20 years before the rule change in 2010 that became effective in 2011). The U.S. Nuclear Regulatory Commission (NRC) may renew licenses at the expiration of the license term upon application by the licensee for a period not to exceed 40 years. Applications for ISFSI license renewals must include the following: 1. Time‐limited aging analyses (TLAAs) that demonstrate that structures, systems, and components (SSCs) important to safety will continue to perform their intended function for the requested period of extended operation and 2. A description of the aging management program (AMP) for the management of issues associated with aging that could adversely affect SSCs important to safety. In addition, the application must include design basis information as documented in the most recently updated final safety analysis report, as required by 10 CFR 72.70. Information contained in previous applications, statements, or reports filed with the Commission under the license may be incorporated by reference, provided that these references are clear and specific. The NRC has issued the “Standard Review Plan for Renewal of Spent Fuel Dry Cask Storage System Licenses and Certificates of Compliance,” NUREG‐1927, under which NRC may renew a specific license or a Certificate of Compliance (CoC) for a term not to exceed 40 years. Both the license and the CoC renewal applications must contain revised technical requirements and operating conditions (fuel storage, surveillance and maintenance, and other requirements) for the ISFSI and dry cask storage system (DCSS) that address aging effects that could affect the safe storage of the used fuel. The information contained in the license and CoC renewal applications will require NRC review to verify that the aging effects on the SSCs in DCSSs/ISFSIs are adequately managed for the requested period of extended operation. To date, all of the ISFSIs across the United States with more than 1,500 dry casks loaded with used fuel have initial license terms of 20 years — three ISFSIs (Surry, H.B. Robinson, and Oconee) have received their renewed licenses for 40 years, and two other ISFSIs (Calvert Cliffs and Prairie Island) have applied for license renewal for 40 years. In addition, renewal of the CoC for the VSC‐24 system is pending approval from NRC.
The bolted closure joint of a bare spent fuel cask is susceptible to age-related degradation and potential loss of confinement function under long-term storage conditions. Elastomeric seals, a component of the joint typically used to facilitate leak testing of the primary seal that includes the metallic seal and bolting, is susceptible to degradation over time by several mechanisms, principally via thermo-oxidation, stress-relaxation, and radiolytic degradation under time and temperature condition. Irradiation and thermal exposure testing and evaluation of an ethylene-propylene diene monomer (EPDM) elastomeric seal material similar to that used in the CASTOR® V/21 cask for a matrix of temperature and radiation exposure conditions relevant to the cask extended storage conditions, and development of semiempirical predictive models for loss of sealing force is in progress. A special insert was developed to allow Compressive Stress Relaxation (CSR) measurements before and after the irradiation and/or thermal exposure without unloading the elastomer. A condition of the loss of sealing force for the onset of leakage was suggested. The experimentation and modeling being performed could enable acquisition of extensive coupled aging data as well as an estimation of the timeframe when loss of sealing function under aging (temperature/radiation) conditions may occur.
Aluminum-clad spent nuclear fuel will be transported for processing in the 70-ton nuclear fuel element cask from L Basin to H-canyon. During transport these fuels would be expected to experience high temperature aqueous corrosion from the residual L Basin water that will be present in the cask. Cladding corrosion losses during transport were calculated for material test reactor (MTR) and high flux isotope reactors (HFIR) fuels using literature and site information on aqueous corrosion at a range of time/temperature conditions. Calculations of the cladding corrosion loss were based on Arrhenius relationships developed for aluminum alloys typical of cladding material with the primary assumption that an adherent passive film does not form to retard the initial corrosion rate. For MTR fuels a cladding thickness loss of 33 % was found after 1 year in the cask with a maximum temperature of 263 °C. HFIR fuels showed a thickness loss of only 6% after 1 year at a maximum temperature of 180 °C. These losses are not expected to impact the overall confinement function of the aluminum cladding.
Aluminum-clad spent nuclear fuel will be transported for processing in the 70-ton nuclear fuel element cask from L Basin to H-canyon. During transport these fuels would be expected to experience high temperature aqueous corrosion from the residual L Basin water that will be present in the cask. Cladding corrosion losses during transport were calculated for material test reactor (MTR) and high flux isotope reactors (HFIR) fuels using literature and site information on aqueous corrosion at a range of time/temperature conditions. Calculations of the cladding corrosion loss were based on Arrhenius relationships developed for aluminum alloys typical of cladding material with the primary assumption that an adherent passive film does not form to retard the initial corrosion rate. For MTR fuels a cladding thickness loss of 33% was found after 1 year in the cask with a maximum temperature of 263 °C. HFIR fuels showed a thickness loss of only 6% after 1 year at a maximum temperature of 180 °C. These losses are not expected to impact the overall confinement function of the aluminum cladding.
On September, 2015, an inspection was performed on the TN-32B cask that will be used for the high-burnup demonstration project. During the survey, wooden cribbing that had been placed within the cask eleven years earlier to prevent shifting of the basket during transport was removed, revealing two areas of residue on the aluminum basket rails, where they had contacted the cribbing. The residue appeared to be a corrosion product, and concerns were raised that similar attack could exist at more difficult-to-inspect locations in the canister. Accordingly, when the canister was reopened, samples of the residue were collected for analysis. This report presents the results of that assessment, which determined that the corrosion was due to the presence of the cribbing. The corrosion was associated with fungal material, and fungal activity likely contributed to an aggressive chemical environment. Once the cask has been cleaned, there will be no risk of further corrosion.
BNL and INL propose to perform a scoping study, using heavily collimated gamma and fast neutron detectors, to obtain passive radiation templates of dry storage casks containing spent fuel. The goal is to demonstrate sufficient spatial resolution and sensitivity to detect a missing fuel assembly. Such measurements, combined with detailed modeling and decay corrections should provide confidence that the cask contents have not been altered, despite loss of continuity of knowledge (CoK). The concept relies on the leakage of high energy gammas and neutrons through the shielding of the casks. Tests will emphasize organic scintillators with pulse shape discrimination, but baseline comparisons will be made to high purity germanium (HPGe) and collimated moderated 3He detectors deployed in the same locations. Commercial off-the-shelf (COTS) detectors and data acquisition electronics will be used with custom-built collimators and shielding.