Advanced Visualization of the Radiation Cast by a Dry Spent Fuel Cask for Public Outreach
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Nelson, Noel [ORNL] (ORCID:0009000523710958)↗
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UNREVIEWED DISPOSAL QUESTION EVALUATION: Disposal of the Oak Ridge National Laboratory Cask Shipments of Resin Waste at the Area 5 Radioactive Waste Management Site, Nevada National Security Site, Nye County, Nevada
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Potential Research and Development Opportunities for Light Water Reactor Spent Nuclear Fuel at INL
This report documents the commercial spent nuclear fuel currently in storage at Idaho National Laboratory with a specific focus on the potential research and development (R&D) opportunities for this material. Commercial spent nuclear fuel is stored in dry casks on the INL site at the Idaho Nuclear Technology and Engineering Center (INTEC, formerly the Idaho Chemical Processing Plant or ICPP). This spent fuel was part of the DOE-industry cooperative dry cask demonstration project that began in 1983 and from historical fuel testing and treatment activities that had been performed at the Test Area North (TAN)-607 facility. Most of the fuel is “low burnup,” but given that it was removed from the respective reactors between 1970 and 1985, it was representative for its time. The majority of the fuel is of the 15×15 Westinghouse PWR variety, though additional fuel types (e.g., TN-BRP) and configurations are also stored. Potential R&D opportunities identified are (1) the REA-2023 cask, (2) the TN-BRP cask, and (3) the Castor V/21 cask. Fuel from the TN-24P may be an alternative to the intact assemblies in the Castor V/21. Specifically, the REA-2023 cask contains 9 rods that have been exposed to air since 2005. These rods have a detailed pre-loading characterization and thus any changes in corrosion or geometry can most likely be attributed to the storage environment. In addition, similar rods to the nine in REA-2023 reside in the Castor V/21 cask whose environment has been maintained with inert nitrogen. The Castor V/21 cask also contains intact PWR assemblies that could be used to perform limited confirmation of the pyroprocessing approach. An alternative to the use of the Castor assemblies for pyroprocessing is the consolidated fuel that is stored in the TN-24P, which would not require removal from the assembly grid since it has already been packaged in accessible two-piece shell strongback containers. The TN-BRP cask contains a number of Gd-doped, low burnup assemblies that could be of use to validate code packages used to justify burnup-up credit for boiling water reactors (BWRs). The INL dry casks are stored on the CPP-2707 pad at INTEC, adjacent to the CPP-603 Irradiated Fuel Storage Facility, which incorporates a dry hot cell that has the potential to receive a full scale commercial cask of the types identified. Several evaluations have been made regarding this task, including full scale mockup testing of the physical clearances required to remove and handle fuel assemblies and components. A proposed schedule of 14 months and cost of $5.3 million was the outcome of analysis performed by Fluor Idaho and is included as Appendix A & B. In addition to the spent fuel at INL, there is also a full-scale mockup assembly that was fabricated for the Protypical Consolidation Demonstration Project to validate fuel rod consolidation approaches. The assembly was fabricated by Westinghouse as a 15 x 15 PWR, using Zircaloy-4 tubing and steel end fittings in accordance with product specifications. The only departure from specification is the use of copper metal in lieu of uranium oxide pellets to achieve a weight that is within 10% of the standard.
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Development of the Licensing Procedure and Regulatory Framework for the Spent Fuel Storage facility in Korea - 16423
The Public Engagement Commission on SNF management (PECOS) published the recommendation report on the spent fuel management of Korea in June, 2015. Based on this report, the future national policy for spent fuel management of Korea has been established, taking into consideration the national/international trends on policy and technology development. Regardless of the recommendation report of PECOS, Spent fuel storage casks are necessary in Korea. Therefore, various types of spent fuel storage casks are now being developed by Korea vendors. For example, Concrete storage casks and dual purpose metal casks are now under development by the KORAD (Korea Radioactive waste agency) for future use in the interim storage facility. Unfortunately, there is no independent licensing procedure for spent fuel storage casks in Nuclear Safety Act (NSA) because is the casks are considered an integral part of the interim storage facility safety equipment. Furthermore, the definition of radioactive waste in the NSA was changed recently. Now, spent fuel is separated from the definition of 'radioactive waste'. Spent fuel can only become radioactive waste by the decision through the meeting of the relative ministers of Korea government. For that reason, the licensing procedure for interim spent fuel storage facilities was separated from the current 'disposal facility, etc.' in accordance with the article 63 of NSA. The aim of this study is to develop the licensing procedure for the storage cask, and also to develop a revised draft of interim storage facility procedure within the NSA. Once this independent licensing procedure on the spent fuel storage cask is introduced to the NSA, it is expected that developers or operators can develop and commercialize the storage cask in Korea. In this study, an independent licensing procedure for the spent fuel storage facility was developed including the storage cask licensing procedure for the NSA. The procedure was composed of the permit for construction and operation of interim storage facility, criteria for permit, inspection criteria, periodic safety review, and decommissioning criteria, etc. The detailed contents of these requirements were also developed to introduce the regulatory framework for the interim storage facility. Particularly, periodic safety review and decommissioning criteria were developed in accordance with the recommendation of IRRS (International Regulatory Review Service) of IAEA. The developed licensing procedure for spent fuel storage facility will be introduced in NSA in response to the revised definition of NSA and the recommendation report of PECOS. (authors)
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Support of UJV Rez, a. s. for Shipments of Spent Nuclear Fuel Within the M3 Program - 19035
In 2005, UJV Rez, a. s. (UJV, the former Nuclear Research Institute Rez), Czech Republic joined the Russian Research Reactor Fuel Return (RRRFR) program under the US-Russian Global Threat Reduction Initiative (GTRI) (now Material Management and Minimization program, M{sup 3}). The primary goal of the program is to advance nuclear nonproliferation objectives encouraging eligible countries to convert their research reactors from highly enriched uranium (HEU) to low enriched uranium (LEU) fuel and eliminating stockpiles of HEU. UJV's daughter company, Research Centre Rez, operates the LVR-15 research reactor with output of 10 MWth, which has been in operation since 1957. After more than 50 years of operation of the reactor, a large amount of spent nuclear fuel (SNF) of Russian origin had been accumulated. The SKODA VPVR/M high capacity casks were used for SNF shipment from the LVR-15 research reactor back to the Russian Federation (RF). The cask has a capacity of 36 fuel assemblies, and 16 casks are available. This means that 576 FA can be transported in one shipment. Two SNF shipments of HEU and LEU SNF from UJV were realized in 2007 and 2013. After the shipments were completed, only LEU nuclear fuel remained on the territory of the Czech Republic. Proven to be an excellent and very sophisticated system for SNF transportation, the SKODA VPVR/M casks were further employed for SNF shipments from Bulgaria, Hungary, Poland, Ukraine, Belarus, Serbia, Vietnam, Uzbekistan and Georgia to Russia. Some shipments were carried out as air shipments with use of the TUK-145/C cask system developed in Russia that included the SKODA cask. In 2016, the new SKODA MNSR casks were developed on the base of the SKODA VPVR/M casks in the Czech Republic and currently are being employed in the shipments of HEU cores of Chinese Miniature Neutron Source Reactors (MNSR) to China. The TUK-145/C-MNSR cask system developed in Russia is used for air transportation. The shipment from Ghana was carried out in 2016; the shipment from Nigeria is under realization. Shipments from Syria, Iraq and Pakistan are planned. Sixteen shipments from eleven countries using a total of 109 SKODA casks have already been completed without any incident or accident and more than 3500 fuel assemblies or cores have been shipped to Russia and China. (authors)
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MELCOR Input Model for Spent Fuel Transportation and Storage Canister
This report documents the progress and current results of the MELCOR spent fuel cask input model. The MELCOR model is being developed to investigate aerosol transport and deposition given the aerosol physical phenomena models within MELCOR. To perform the analyses, a general portrayal of the MAGNASTOR® cask system has been employed; however, this system was selected based on available information to provide a reasonable representation of a spent fuel cask. The analytical results are not intended to characterize the performance of the MAGNASTOR® cask. Instead, the provided results are intended to enhance our general understanding of the aerosol behavior within casks and the validity of current models. The current model efforts are being performed to investigate hypothetical UO 2 release from failed fuel pins within a spent fuel cask. The existing MELCOR model of the MAGNASTOR® cask system has been adapted to permit future comparative analyses with the GOTHIC representation of the MAGNASTOR® cask. To support this comparison, the PNNL model characteristics that are unrelated to the aerosol modeling were applied to the MELCOR model. These characteristics included improved comparability of the axial fidelity, total spent fuel power, fuel pin axial power profile, and heat losses from cannister. The thermal-hydraulic solutions are improved within the capability of the MELCOR code and will permit better overall agreement with the GOTHIC results. Detailed results are presented on the thermal-hydraulic analysis of the MELCOR cask as well as characterization of UO 2 aerosol dispersion and deposition within the cask.
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A New Safeguards Technique for Spent Fuel Verification in Dry Storage by Using He-4 Detectors - 18262
It is demanded to find key issues and critical paths for verifying and monitoring a spent nuclear fuel cask or storage with a purpose of non-proliferation of nuclear weapon. However, unfortunately there are no reliable methods to verify the containment in a spent nuclear fuel cask or storage. Therefore, research and development (R and D) has been continuously required to develop new and alternative technologies for searching and verifying undeclared nuclear materials, activities, and facilities. Therefore, the present authors have developed a new safeguards technique of passively scanning a spent fuel cask or storage using pressurized He-4 scintillation neutron detectors. A new approach is to quickly verify content in a spent nuclear fuel cask or presence of all special nuclear materials in unknown packages by comparing experimental and computational results. For that, the present authors have developed a new algorithm and user-interface coupled ORIGEN-ARP and MCNP codes, and the prototype safeguards apparatus employing the array of He-4 fast neutron detectors has been designed, built, and tested. The feasibility test of this new apparatus has been performed with Cf-252 neutron sources and the laboratory scale model cask, minimized by a factor of ten compared to the actual dimension of a KN-21 cask. This technique has several advantages over current methods based on neutron and gamma detection; (1) incident neutron energy information is preserved, (2) disturbance of gamma-ray is excellently rejected (discriminated), and (3) much higher neutron counts and operational cost and space-effectiveness are expected since no moderator is required for fast neutrons. The final goal of this study will (1) conduct a prototype instrument and employ it to monitor a nuclear spent fuel cask or storage at a commercial nuclear power reactor facilities, (2) use spectral-specific measurement data to show proof of concept for measurable data successfully converted to quantifiable signatures, such as diversion/removal of fuel, and maintenance of fuel and cask integrity of fuel, and (3) develop a cask imaging capability. (authors)
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Nuclear Industry Input to the Development of Concepts for the Consolidated Storage of Used Nuclear Fuel - 13411
EnergySolutions and its team partners, NAC International, Exelon Nuclear Partners, Talisman International, TerranearPMC, Booz Allen Hamilton and Sargent and Lundy, have carried out a study to develop concepts for a Consolidated Storage Facility (CSF) for the USA's stocks of commercial Used Nuclear Fuel (UNF), and the packaging and transport provisions required to move the UNF to the CSF. The UNF is currently stored at all 65 operating nuclear reactor sites in the US, and at 10 shutdown sites. The study was funded by the US Department of Energy and followed the recommendations of the Blue Ribbon Commission on America's Nuclear Future (BRC), one of which was that the US should make prompt efforts to develop one or more consolidated storage facilities for commercial UNF. The study showed that viable schemes can be devised to move all UNF and store it at a CSF, but that a range of schemes is required to accommodate the present widely varying UNF storage arrangements. Although most UNF that is currently stored at operating reactor sites is in water-filled pools, a significant amount is now dry stored in concrete casks. At the shutdown sites, the UNF is dry stored at all but two of the ten sites. Various types of UNF dry storage configurations are used at the operating sites and shutdown sites that include vertical storage casks that are also licensed for transportation, vertical casks that are licensed for storage only, and horizontally orientated storage modules. The shutdown sites have limited to nonexistent UNF handling infrastructure and several no longer have railroad connections, complicating UNF handling and transport off the site. However four methods were identified that will satisfactorily retrieve the UNF canisters within the storage casks and transport them to the CSF. The study showed that all of the issues associated with the transportation and storage of UNF from all sites in the US can be accommodated by adopting a staged approach to the construction of the CSF. Stage 1 requires only a cask storage pad and railroad interface to be constructed, and the CSF can then receive the UNF that is in transportable storage casks. Stage 2 adds a canister handling facility, a storage cask fabrication facility and an expanded storage pad, and enables the receipt of all canistered UNF from both operating and shutdown sites. Stage 3 provides a repackaging facility with a water-filled pool that provides flexibility for a range of repackaging scenarios. This includes receiving and repackaging 'bare' UNF into suitable canisters that can be placed into interim storage at the CSF, and enables UNF that is being received, or already in storage onsite, to be repackaged into canisters that are suitable for disposal at a geologic repository. The study used the 'Total System Model' (TSM) to analyze a range of CSF capacities and operating scenarios with differing parameters covering UNF pickup orders, one or more CSF sites, CSF start dates, CSF receipt rates and geologic repository start dates. The TSM was originally developed to model movement of UNF to the Yucca Mountain repository and was modified for this study to enable the CSF to become the 'gateway' to a future geologic repository. The TSM analysis enabled costs to be estimated for each scenario and showed how these are influenced by each of the parameters. This information will provide essential underpinning for a future Conceptual Design preparation. (authors)
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Using FENOSOL{sup TM} Phenolic Foam as a Shock Absorber for Type B Packaging - 17256
Type B and fissile material transportation packages are designed to withstand deformation from dynamic forces experienced during impacts in hypothetical accident conditions. A critical aspect is to ensure the containment system is not breached. The package design normally includes impact limiters made up of a shell filled with energy-absorbing material. Various types of materials are used as energy absorbers, including wood, foam, and honeycomb structures. Synthetic foam is an ideal material for impact limiters; it has isotropic material properties, it acts as a heat insulator during fire conditions, and it maintains structural integrity at high temperatures. In addition, the density of the foam may be controlled during the manufacturing process, allowing the cask designer flexibility to select the right density for each application. The foam properties can be used as inputs for numerical simulations of dynamic impacts. These simulations are a key component of the safety analysis report for Type B transportation casks. Robatel has recently invested in equipment to produce FENOSOL{sup TM} phenolic foam in our factory for use in our cask designs. This acquisition was made jointly between Robatel and the CEA, with Robatel ensuring the industrialization of the product. In the past decade, this foam was successfully integrated into more than 10 European cask designs, and there has been extensive in-laboratory characterization of its properties. FENOSOL{sup TM} has already been accepted by the French Nuclear Safety Authority (ASN) as a shock absorbing material for use in cask designs. Since the acquisition of the phenolic foam production process by Robatel, a new research and development program has been launched to provide more accurate information to our clients, extend the range of application of this foam, and to improve its properties. Robatel has built an extensive database of material characteristics through the use of a universal mechanical tester, a scanning electron microscope, and a nuclear magnetic resonance spectrometer. In addition, Robatel has successfully implemented FENOSOL{sup TM} in two recent cask designs. This included scale model drop testing of a 9 meter drop in three orientations, and a pin puncture drop test. Robatel is in the process of producing FENOSOL{sup TM} samples for testing according to ASTM standards. Various densities will be included in the study. The testing will include duro-meter hardness, specific heat, thermal conductivity, compressive strength and modulus, and tensile strength. Once the testing is complete, Robatel will compile the data in the form of product specification sheets for each density. These information sheets and test reports will be used to validate FENOSOL{sup TM} for use in future impact limiters for the North American Type B cask market. (authors)
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Vacuum drying of spent fuel storage canisters/casks - 'just how much water could be left in there, anyway?' - 15233
The process of moving spent fuel assemblies from wet storage in the spent fuel pool to dry storage in an inert gas environment within a canister or cask typically involves two main stages. In the first stage, the bulk of the water is removed by a draining or siphoning operation, conducted with pressure inside the canister maintained at 1.3 to 2 atm. This process removes most of the water in the cask cavity, but residual water can be left behind, in three main regions. The largest region, typically, is the bottom of the cavity below the end of the siphon tube. A second general class of locations, also located at the bottom of the cask cavity, consists of the 'dashpots' at the ends of the various non-fuel tubular structures within the fuel assemblies. For PWR assemblies, these are the control rod guide sleeve tubes and instrumentation thimbles. For BWR assemblies, these are the water rods of various designs and control blade ends in some configurations. A third class of locations consists of porous material within the cavity that can absorb water, such as thin ceramic neutron poison plates, and ostensibly intact fuel rods that nevertheless might have pinholes or fine cracks in the cladding that could allow pool water to creep inside the fuel rod. To remove residual water from these locations, the second stage of drying operations utilizes vacuum drying, by reducing the pressure sufficiently to convert the water to vapor that can then be pumped out of the canister cavity. In this operation, the cask cavity pressure is slowly reduced from the nominal condition of (1.3-2.0 atm (1000-1520 torr), to a final target pressure below 3 torr (0.004 atm). This is typically accomplished in three to five stages, with prescribed 'hold' points defined in procedures or in the Technical Specifications, to preclude sudden pressure spikes due to flashing of water to steam, or development of large local temperature gradients in the fuel rods or other cask internal components. This paper discusses the various locations in the system where residual water may 'hang up' in the cask, and presents a discussion of the physical basis for the effectiveness of vacuum drying for removing it. Results and conclusions based on realistic qualitative and quantitative evaluations of the transient heat transfer behavior of the system are presented, based on detailed modeling of a range of spent fuel canister designs and decay heat loads, most notably the 24P Dry Shielded Canisters (DSCs) stored at the Rancho Seco Independent Spent Fuel Storage Installation (ISFSI), the NAC-MPC modules at Maine Yankee, the 24P DSCs stored at Calvert Cliffs, and the Holtec HI-STORM modules at Hope Creek. The evaluations presented here show that ice formation within a spent fuel canister is highly unlikely, due to the availability of a large thermal source term in the form of the decay heat load of the spent fuel. The formation of ice is an operational concern, since it mainly affects operation of equipment external to the canister (i.e., the hoses, valves, and fittings of the vacuum drying equipment.) Evaluations are presented for quantifying the amount of residual water that could remain in the cask cavity, even after successful drying operations, since vacuum drying is a limit process, and by definition, cannot remove all of the water from the cavity, no matter how long the operation goes on, even if all of it is in vapor form. (authors)
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Twenty years of dry interim storage of spent nuclear fuel (SNF) in Germany- History and developments - 15575
According to article 6 of the German Atomic Energy Act, storage of nuclear fuel requires a license granted by the competent authority in this field, which is the Federal Office of Radiation Protection. In 1983 a license to store spent nuclear fuel (SNF) and later expanded for vitrified high level waste (HAW) for the interim storage facility in Gorleben was granted. Dry Interim storage at the NPP sites in its actual form started in 2002 in the interim storage facility next to the NPP Lingen. Since this time each NPP erected its own storage facility and three central storage facilities have been built. All of these facilities use dry storage in metallic casks. The actual storage licenses for SNF and the HAW (which is stored in Gorleben and Rubenow) are limited to 40 years starting with the first cask emplacement. As no opening of the cask is intended during the storage period the condition of the inventory and the cask before loading has to be known in detail. This requires for instance detailed knowledge of the history of the SNF loaded as well as supervision of the manufacturing process of the cask. Within the last twenty years several mishaps influenced the regulatory point of view. During a loading campaign in a boiling water reactor traces of pool water remained in the gasket slot and during the drying procedure crystalline boron salts prevented the required tightness. After that event a long-term experiment was started and is still going on which studies the effects of water and minor damages to the gasket. As well changes in the loading procedure have been introduced. The pressure switch used to monitor the pressure in the space between primary and secondary lid has a self surveillance function. This indicated a malfunction 23 times in the last twenty years. Examinations of all events indicated no systematic failure, but a weak point seems to be the plug-in connection. The manufacturer introduced some modifications to improve the system. Introduction of FEM methods in safety analysis to amend the existing methods of testing allowed a new view on and in the cask safety analysis. Progress in this field is closely connected to the available computing capacity and stipulates experts to ongoing improvement. After expiring of the storage license which will be the case for the first cask loaded in 1992 and stored in Ahaus 2032 an extension or a new license can only be granted if the at this time relevant requirements are met. This will require detailed knowledge about the happenings and accidents during the previous storage period. All experience of the periodic safety assessment (PSA) which started in 2013 and accompanying experiments conducted till now and during the licensed storage time will be needed. (authors)
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A novel application of cosmic rays for spent nuclear fuel monitoring - 15738
Monitoring nuclear waste and controlling spent nuclear fuel at its source is one of the main strategies to minimize the risks of nuclear proliferation and reduce potential homeland threats. Storage of high level waste from a fission reactor typically involves placing spent fuel in sealed dry casks. After the spent nuclear fuel has been placed inside the dry cask, the cask is welded, not allowing for visual inspection. Conventional methods for examining the interior of materials e.g., X-rays, are limited by the fact that they cannot penetrate very dense well-shielded objects. Instead, cosmic ray muons have the potential to allow for assessment of spent nuclear fuel stored within sealed dense dry casks. Objectives: 1.Exploit the ability of cosmic ray muons to penetrate high density materials allowing the distribution of materials within the object to be inferred from the muon tracks. 2. Develop a monitoring system that tracks the path of cosmic ray muons, including their deflection and attenuation, to identify the contents of a closed dry cask. Methods: - Monte-Carlo simulations using Geant4; - Bayesian decision theory for signal classification; - Wavelet analysis for noise cancellation; - Chaotic Neural Networks for signal encryption. Results: 1. Preliminary statistical analysis proved that the method is feasible and has the ability to identify the presence of spent nuclear fuel in a dry cask. 2. Using monoenergetic muons, a dry cask with one fuel assembly missing can be correctly identified with error rate <1%. Broader merit: Once completed, the proposed project for cosmic ray muon transport will enable a new and potentially transformative approach towards efficient, inexpensive and remote cosmic ray muon applications in archaeology, medicine and safeguarding of nuclear materials.
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MECHANICAL SHOCK AND VIBRATION ANALYSIS OF SPENT NUCLEAR FUEL CARRIED BY THE ATLAS RAILCAR
Researchers at Pacific Northwest National Laboratory have completed a structural-dynamic analysis of spent nuclear fuel subjected to the mechanical shock and vibration environment that is anticipated during normal conditions of transport in casks carried by the Atlas railcar. The Atlas railcar is a new railcar design that is being developed specifically for the purpose of carrying spent nuclear fuel casks. The analysis used best-estimate railcar dynamics models of the Atlas railcar and considered 17 different spent nuclear fuel transportation cask systems, representing the current fleet of cask options. This work used NUCARS, a specialized railcar dynamics explicit finite element code to calculate railcar dynamic response to prescribed speeds and track configurations. The railcar dynamics models provided cask transient motion for a wide range of speeds and track conditions, generating a relatively large database of potential cask motion. All of the cask motion transients were then applied as loading conditions to LS-DYNA structural-dynamic models of a single fuel rod. The analyses predict that the Equipos Nucleares S.A./U.S. Department of Energy (ENSA/DOE) multimodal transportation test of 2017 provided a relatively stronger vibration environment than is expected from the Atlas railcar. This paper describes the analysis methods, the analysis results, and compares the results of the Atlas transportation analysis to the test results and analyses of the ENSA/DOE multimodal transportation test of 2017.
Klymyshyn, Nicholas A.↗
HI-STORM Overpack and MPC-32 Thermal-Hydraulic Model with MOOSE Framework
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.
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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.
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Small Waste in a Large Package - Meeting Requirements for Packaging, Transport and Disposal - 18092
A small amount of Class C Mixed Low Level Waste (MLLW) required packaging and disposition to an off-site facility. The AREVA TN RAM cask, designed and licensed for shipping Type B waste packages, was chosen as the most appropriate shipping solution. Due to packaging, shipping and disposal requirements, the standard TN RAM cask could not be utilized without making some modifications. Unique solutions and approaches had to be developed by the AREVA design team to complete the project within the anticipated budget and schedule. The general work scope was to characterize, remove from a pool, profile, package, certify and transport four irradiated components to an approved off-site facility. The material makeup of the components (three stainless steel plates and a lead counterweight) resulted in a MLLW classification which required special packaging and disposal methods. The AREVA design team worked with the disposal facility to identify all requirements associated with disposal of Class B/C MLLW. After defining the packaging and disposal requirements, it was determined that unique spacers and void fillers would be required. The design solution consisted of a liner insert as well as axial and radial spacers filled with inert material. Remaining voids in the liner insert were filled with recycled glass beads to fulfill the requirement for maximum 10% void space in the waste package. The axial spacers also served to center the waste load in the shipping cask to comply with the AREVA TN RAM Cask NRC license. Concentration averaging and waste consolidation were used to reach the Class C MLLW determination for shipping and disposal. This disposal method for Class C waste met the requirements of macro-encapsulation. Therefore, no additional treatment action was required when disposing of Class C MLLW. AREVA's solution provides a model for meeting macro-encapsulation requirements while also meeting void space requirements. This solution provides a model for how to adapt large shipping casks and liners to smaller waste packages that would not normally be a good fit for traditional casks. (authors)
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First Road / River Barge Security Transport of SNF under the New German Transport Security Guidelines - 19102
A transfer of 342 spent fuel assemblies from NPP Obrigheim to NPP Neckarwestheim, both located alongside the river Neckar in the south of Germany, is required to promote the decommissioning process of NPP Obrigheim, which started in 2008 after the NPP had been shut down three years before. The spent fuel storage facility in Neckarwestheim offers sufficient storage space to accept the 342 SNF assemblies, stored in fifteen dry casks type CASTOR{sup R} 440/84 mvK. After several transport feasibility studies had been conducted, Daher Nuclear Technologies GmbH was awarded the commission to plan, prepare and conduct the SNF transport. Since the combined road / river barge transport promised to be the most suitable mode of transportation Daher Nuclear Technologies started to prepare its security concept, a prerequisite to apply for and attain the transport license mandatory for transporting spent nuclear fuel in Germany. In March 2014 the company applied for the transport license, which, after more than three years and numerous amendments to the security concept, was finally granted to Daher Nuclear Technologies on May 16, 2017 by the German Federal Office for the Safety of Nuclear Waste Management (successor organization of the BfS). Simultaneously to the license application phase the transport equipment - amongst others the heavy-duty trucks and trailers, the pusher tugs, the river barge and transport frames - had to be provided and modified to meet the intensified requirements of the New German Transport Security Guidelines Road / Rail as well as several other legal regulations referring to the transport of dangerous goods by road and inland waterways (e.g. ADR - European Agreement concerning the International Carriage of Dangerous Goods by Road and ADN - European Agreement concerning the International Carriage of Dangerous Goods by Inland Waterways). The planning process involved various meetings with representatives of different authorities and administrative bodies incl. accredited experts as well as subcontractors. Tremendous training efforts had to be undertaken to train more than 100 transport participants regarding security measures, health, safety and environmental measures, as well as the proper handling of radioactive material. This also included the training of the security staff, that had to be qualified in compliance with the new Transport Security Guidelines Road and Rail. Finally, in June 2017, when all preparations where complete, all requirements met, and all permits obtained, the first ever security SNF transport on a German river was performed. Three laden CASTOR{sup R} 440/84 mvK transport and storage casks were transported in one batch via heavy-duty truck-trailer from the NPP Obrigheim to the NPP's own harbor, to be rolled on to the river barge, that would carry the three casks upstream the river Neckar to the approx. fifty kilometer distant NPP Neckarwestheim. The journey on the river took about eleven hours, in which several bridges and locks had to be passed. The entire transport was escorted by qualified security staff and auxiliary police forces on waterways and alongside on roads. It attracted nationwide public attention, due to its pioneer status and the fact, that there hadn't been any SNF or HLW shipments in Germany since 2011. After the convoy safely arrived at NPP Neckarwestheim's pier the three laden CASTOR{sup R} 440/84 mvK casks were rolled off the river barge via heavy-duty truck-trailers to be stored in the on-site interim storage facility. This first transport was a success and was followed by four additionally successful security transports in September, October, November and December 2017. The public attention receded significantly during the transport campaign, which ended on December 19, 2017 with the transfer of the last three of fifteen casks to the on-site interim storage facility in Neckarwestheim. With the last cask safely transferred to the interim storage facility Daher Nuclear Technologies successfully completed the transport campaign on schedule, meeting one of EnBW's fundamental objectives to finish the transfer of the 342 SNF assemblies by the end of 2017. (authors)
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Probability of Pinching Failure in PWR SNF after Long-Term Storage
A probabilistic assessment was performed on pinching failure of PWR SNF rod cladding under accidental SNF cask transport conditions. a set of 3,000 rod pinching scenarios was analyzed, with different geometrical, material, and loading parameters, using a Latin Hypercube Sampling (LHS). The investigated accidents include train derailments leading to cask collisions or cask drops. Pinching failure was related to cladding offset strain, defined as the plastic deformation interval in radial cladding direction normalized with respect to the outer cladding diameter. Finite element (FE) models were used to compute SNF rod deformations and the offset strain demand OSD in each of the assessed pinching scenarios. The numerical models account for the contribution of the degraded fuel pellets to rod pinching stiffness. The offset strain capacity OSC was computed using regression and interpolation models based on ring compression (RCT) test results, and account for hydride-related effects on cladding ductility. The results show a low probability of pinching failure. Hydride-related cladding embrittlement affects the structural integrity of PWR SNF rods after long-term dry storage of up to 300 years. Radial hydride patterns or large amounts of circumferentially oriented hydrides reduce the cladding ductile deformation capability when pinched at low material temperatures. The probability of hydride-related pinching failure of PWR cladding is estimated after long-term dry storage to address: How do PWR SNF rod conditions like hydrogen content, vacuum drying peak cladding hoop stress σ{sub θ}, and cladding material temperature affect cladding ductility? Is hydride-related cladding embrittlement a material degradation mechanism that affects the structural integrity of PWR SNF rods under pinching loading? What are the controlling parameters for deformation demands in PWR SNF rods under pinching loading? Is the pinching load amplitude an important parameter that controls the probability of pinching failure? How do parameters like pellet-cladding gap or fuel pellet degradation affect the pinching loading capacity of PWR SNF rods? Statistical modelling of PWR SNF rod conditions in U.S. inventory: A pseudo rod database was generated based on EIA database. Analysis of train accident scenarios: a probability density function (pdf) of a pinching load distribution P{sub 9m} of a 9-m cask drop is linearly scaled to generate a more representative: distribution P{sub AS}. The new distribution considers train collision velocities, cask drops for on- and off-bridge accidents, and cask impact orientation. LHS generates 3,000 rod pinching scenarios based on nine independent geometrical, material, and loading parameters. Probabilistic FE Analysis: The LHS realizations are used to generate 3,000 FE models. The FE analysis is conducted for P{sub 9m}, and P{sub AS}. The workflow is automatized. The OSC is computed with regression and interpolation models, as a function of the hydrogen content C{sub H}, vacuum drying peak σ{sub θ}, and SNF temperature T{sub FR}: The regression data was obtained in RCTs on PWR cladding. The OSD is derived from the cladding deformations in the FE models. Cladding failure was identified via a comparison of OSC and OSD. A 300 and a 30 years-of-dry-storage duration is investigated: The dry-storage period affects the expected SNF temperature and consequently, the OSC and the OSD. The material temperature LHS population is adapted to account for a SNF temperature variation. The probability of hydride-related, ductile cladding failure during a severe transportation accident is not significantly affected by the pinching load amplitude: The probability is slightly reduced from a range of less than 0.0333% to 0.1667% for P{sub 9m}, to a range of less than 0.0333% to 0.1333% for P{sub AS}, depending on the vacuum drying peak cladding hoop stress modelling assumptions. Further, the SNF dry-storage duration does not significantly affect the probability of ductile pinching failure. Based on available train accidents data, the pinching failure probability for a 5,000 km long transportation route ranges from less than 1.62 E-7 to a value of 6.47 E-7, depending on the vacuum drying peak cladding hoop stress modelling assumptions. The pinching load amplitude is not a controlling parameter for the probability of ductile cladding pinching failure, and does not control the cladding deformations once pellet-cladding contact is established, The pellet-cladding gap width, however, plays a key role. Fuel pellet degradation only affects the deformation capacity, if the pellet stiffness is significantly reduced (more than 90%). Only highly embrittled cladding is likely to fail. The deformation demands on the cladding under pinching loading are small.
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