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At least 1,189 records · Page 66
All sourcesNASA NTRSDOE OSTI
DOE OSTI2013

State-of-the-art on instant release of fission products from spent nuclear fuel

Within the EURATOM FP7 Collaborative Project 'Fast / Instant Release of Safety Relevant Radionuclides from Spent Nuclear Fuel (CP FIRST-Nuclides)', a State-of-the-Art Report was prepared. The fast / instant release fraction (IRF) is defined as a fraction of the inventory of radionuclides that may be rapidly released from the fuel and fuel assembly materials at the time of canister breaching. In the context of safety analysis for a repository, the time span for mobilization of this fraction can be considered instantaneously, even if the process takes some time in experiments. Radionuclides contributing to the fast release are fission gases (Kr and Xe), easily soluble elements such as cesium and iodine, and other elements which are hardly incorporated in the UO{sub 2} crystal lattice. The present contribution summarizes the results obtained from published studies focused on rapid release experiments carried out with different spent nuclear fuel (SNF), samples, sizes, techniques (batch and flow-through), and durations. A total of 80 experiments cover the study of UO{sub 2} SNF from pressure water reactors (PWR) of different initial enrichments and burn-up, while 20 experiments were performed with UO{sub 2} SNF from boiling water reactors (BWR) and 8 with MOX fuel. (authors)

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DOE OSTI2020

Workshop to Plan R&D Support of Fuel/Basket Modification for Direct Disposal of Future DPCs

By 2030 about half of all spent nuclear fuel (SNF) arising from the current fleet of commercial power plants will be in dual-purpose canisters (DPCs), which are designed for storage and transportation but not for disposal. As an alternative to complete repackaging of the fuel for disposal, considerable cost savings and lower worker dose could be realized by directly disposing of this SNF in DPCs. The principal technical consideration is criticality control in a geologic repository, because the DPCs are large and depend on neutron absorbing basket components for criticality control. Neutron absorbing materials are generally aluminum-based, and under disposal conditions can degrade after a few hundred years contact with ground water. Simple modifications to the SNF assemblies or the DPC baskets could help to achieve direct disposal, and this is one of the approaches being studied to address the possibility of disposal criticality (SNL 2020a). Five fuel/basket modification concepts have been proposed (SNL 2020b) and a virtual workshop was conducted to solicit review and feedback on these concepts. The proposed solutions are: 1) zone loading of DPCs to limit reactivity, 2) replacing absorber plates with advanced neutron absorbing (ANA) material, 3) adding disposal control rods to pressurized water reactor (PWR) assemblies, 4) rechanneling boiling water reactor (BWR) assemblies with ANA material, and 5) basket insert plates (chevron inserts) made from ANA material. The presentations from the workshop are provided in this report, and the workshop discussions are summarized. This information includes prioritization of the proposed fuel/basket modification solutions, and prioritization of the associated model development, validation testing, and quality assurance activities. Information documented in this report will help to steer research and development efforts at Sandia National Laboratories, Oak Ridge National Laboratory, and Idaho National Laboratory that support the U.S. Department of Energy, Office of Nuclear Energy, Spent Fuel and Waste Science and Technology program

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DOE OSTI2020

DOE Standard Canister Loading Methodology Poster

Over 250 different types of spent nuclear fuel (SNF) are managed by the U.S. Department of Energy (DOE) at the Idaho National Laboratory (INL). The current configuration of SNF at INL is not transportable off-site. In the Yucca Mountain Repository License Application, the SNF was to be loaded into DOE Standard Canisters which are capable of storage, transportation, and eventual disposal using the proper overpacks. The DOE Standard Canister will allow INL to place the SNF in a Road-Ready configuration in preparation for transportation off-site. The focus of this work is development of a general loading methodology for the DOE Standard Canister. This methodology is to be general enough that it will apply to all SNF to be loaded into the DOE Standard Canister. The main sections of the methodology to be outlined are prerequisites before loading (including any issues with materials such as inserts or baskets), loading of SNF, and canister closure and quality assurance (including welding and inspection). Each subsection will analyze and provide the bases for each step.

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DOE OSTI2023

Continued Development and Advanced Testing of DPC Filler Cements (on FY22 R&D and Demonstration Activities) (Progress Report)

Commercial generation of energy by nuclear power plants in the United States (U.S.) has produced thousands of metric tons of spent nuclear fuel (SNF), the disposal of which is the responsibility of the U.S. Department of Energy (DOE). Utilities typically utilize the practice of storing this SNF in dual-purpose canisters (DPCs). DPCs were designed, licensed, and loaded to meet Nuclear Regulatory Commission (NRC) requirements that preclude the possibility of a criticality event during SNF storage and transport, but were not designed or loaded to preclude the possibility of a criticality event during the regulated post-closure period following disposal, which could be up to 1,000,000 years (Price, 2019). There are several options being investigated that could facilitate the disposal of SNF stored in DPCs in a geologic repository (Hardin et al., 2015; SNL 2020b; SNL 2021b). These include: (1) repackage the SNF into canisters that are designed to prevent criticality during the regulated post-closure period following disposal, but with an increased disposal cost estimated at approximately $\$$20B in United States dollars (USD) (Freeze et al., 2019); (2) analysis of the probability and consequences of criticality from the direct disposal of DPCs during a 1,000,000-year post-closure period in several geologic disposal media (Price, 2019); and (3) filling the void space of a DPC with a material before its disposal that significantly limits the potential for criticality over the post-closure regulatory period. This report further investigates the third option, filling DPC already containing SNF with a material to limit the potential for criticality over the post-closure regulatory period.

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DOE OSTI2016

Spent Nuclear Fuel Management and Disposal in Finland - Status and Lessons Learned at the End of 2015 - 16507

In November 2015 the Finnish Government approved and the Parliament ratified Posiva's 2012 License Application for Construction of a mined, deep geological repository for the spent nuclear fuel (SNF) at the Olkiluoto site in the Eurajoki municipality based upon recommendations provided by the Finnish Radiation Safety Authority (STUK), international expert groups and long-standing local support. The preceding siting process began in 1983 and it included disposal-concept selection (KBS-3V), 'consent-based' site selection and large-scale in-situ tests conducted in an underground research laboratory, the ONKALO, situated ∼440 m below the mean Sea level immediately adjacent to the planned repository at the Olkiluoto site. At the end of 2015, the Finnish SNF repository is projected to open in 2024 as the first of its kind. Four related lessons learned in Finland during the past 30 years are: 1. The selection of potentially-suitable sites for geological disposal of SNF proved to be more difficult and more time consuming than initially envisioned. 2. The siting of nuclear facilities has evolved from being mainly a radionuclide containment and isolation challenge to becoming a critical moving-target, socio-political challenge governing progress with majority local acceptance being a fundamental imperative. 3. Public and political acceptance and support are strongly linked to the competency, credibility and transparency of the involved organisations and the maturity/pedigree of the proposed disposal concept. 4. Timely sharing lessons learned and conducting selective collaborations with other countries proved to be successful keys to timely and cost-effectively obtain and maintain public and political acceptance and support. (authors)

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DOE OSTI2014

Investigation of Coupled Processes and Impact of High Temperature Limits in Argillite Rock

Shale and clay-rich geological formations have been considered as potential host rock for geological disposal of high-level radioactive waste throughout the world, because of their low permeability, low diffusion coefficient, high retention capacity for radionuclides, and capability to self-seal fractures. Permeabilities arising from the primary porosity of these rock types are low because of the extremely fine-grained constituents and the very small pores associated with the fine-grained material. Permeability values are typically less than 10 microdarcies (10-17 m2 or about 10-10 m/s equivalent hydraulic conductivity) and are often two to four orders of magnitude lower. While fractures can occur, clay and shale often demonstrate the tendency to self-seal fractures, which reduces the effects of fractures on bulk permeability. This occurs as a result of swelling from the increased water potential in fractures, the generation of fracture infilling materials and mineral precipitation, and plastic deformation of the rock (Mazurek et al., 2003). The low permeability of clay and shale rock are well-known in the hydrogeology community where these rock types represent aquitards that severely limit groundwater movement, and in petroleum geology, where they act as caprocks limiting the rise of buoyant petroleum fluids. Other favorable characteristics of clay/shale rock are the strong sorptive behavior for many radionuclides and (in saturated systems) the low flow rates, which typically lead to reducing conditions because of the lack of oxygen transport from the surface. Clay and shale rock also act to chemically buffer the effects of materials introduced through repository construction, operation, and emplaced materials (Arcos et al., 2008).

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DOE OSTI2020

Investigation of Cold Spray as a Dry Storage Canister Repair and Mitigation Tool

The lack of a permanent spent nuclear fuel repository means utilities are faced with extending licenses for dry cask storage systems (DCSS) canisters far beyond their original 20-year license period. DCSS canisters are made from austenitic stainless steels (SSs) and fabricated using fusion welds. High heat input and melting associated with fabrication welds causes material around the weld to be susceptible to chloride-induced stress corrosion cracking (CISCC). The purpose of the work reported here is to investigate the use of cold spray for repair and mitigation of CISCC in DCSS canisters to ensure integrity far beyond their original license period. This report presents an analysis of cold spray for canister life extension, a summary of related efforts, relevant technology down selections, work accomplished, and future efforts relative to use of cold spray for repair and mitigation of CISCC in DCSS canisters.

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DOE OSTI2019

Oxidation of Spent Fuel During Fuel Handling Accidents Test Plan

The transfer of commercial spent nuclear fuel (CSNF) from transport or dry storage cask into final repository cask represents a scenario where fuel elements are potentially at risk to exposure to air and elevated temperatures. In the event of an accident during handling operations, the facility cooling may cease to function, and the temperature of the fuel would begin to rise due to decay heat generation. CSNF is generally stored in an inert environment, but handling operations could be performed in an air environment. If any damaged fuel elements are present during this postulated accident the fuel could begin to oxidize from UO 2 to U 3 O 8 . If the oxidation event is severe, volume expansion from oxidation could result in large-scale cladding failure and large releases of CSNF debris into the transfer facility. Depending on the design of the fuel handling facility, any SNF release could have an impact on off-site dose.

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DOE OSTI2017

Extended Storage Considerations for Dry Cask Storage Systems Using Welded Stainless Steel Canisters - 17436

Dry storage of used nuclear fuel is likely to continue for several decades due to the lack of a permanent repository. The Electric Power Research Institute (EPRI) conducted a Failure Modes and Effects Analysis (FMEA) in order to identify degradation mechanisms that are relevant to Dry Cask Storage Systems (DCSSs) which store used fuel within a welded stainless steel canister housed inside a vented concrete cask. Chloride-Induced Stress Corrosion Cracking (CISCC) was identified as the potential degradation mechanism that would be most likely to challenge the confinement function of these canisters. This paper describes research efforts aimed at identifying canisters with CISCC prior to loss of confinement and understanding the consequences should such a loss occur. Key areas of research include development of aging management guidance with recommendations for the scope, frequency, and coverage for canister inspections, demonstration of non-destructive evaluation capabilities, and evaluation of the potential consequences in terms of dose to site personnel and the general public should cracking develop and grow through a canister wall. (authors)

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DOE OSTI2016

NWMO Canadian used nuclear fuel repository - 16314

A brief introduction to Canada's Used Fuel Repository Plan, including the status of site selection and the conceptual design. Canada's plan for the long-term care of used nuclear fuel is known as Adaptive Phased Management (APM). It emerged from dialogue with Canadians and experts and best meets the key priorities considered important by citizens. The Federal Government selected APM as Canada's plan in June 2007. The plan is consistent with the long-term management approach adopted by other countries with nuclear power programs, such as Finland, Sweden, Switzerland, the United Kingdom and France. Canada's plan is both a technical method and a management system. The project is designed to be implemented in phases over many decades. New learning and social priorities will be incorporated into Canada's plan for the long-term management of used nuclear fuel, driving refinement and adaptation of the plan throughout its implementation. The major phases for the APM project are described.

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DOE OSTI2014

A FRAMEWORK TO DEVELOP FLAW ACCEPTANCE CRITERIA FOR STRUCTURAL INTEGRITY ASSESSMENT OF MULTIPURPOSE CANISTERS FOR EXTENDED STORAGE OF USED NUCLEAR FUEL

A multipurpose canister (MPC) made of austenitic stainless steel is loaded with used nuclear fuel assemblies and is part of the transfer cask system to move the fuel from the spent fuel pool to prepare for storage, and is part of the storage cask system for on-site dry storage. This weld-sealed canister is also expected to be part of the transportation package following storage. The canister may be subject to service-induced degradation especially if exposed to aggressive environments during possible very long-term storage period if the permanent repository is yet to be identified and readied. Stress corrosion cracking may be initiated on the canister surface in the welds or in the heat affected zone because the construction of MPC does not require heat treatment for stress relief. An acceptance criteria methodology is being developed for flaw disposition should the crack-like defects be detected by periodic Inservice Inspection. The external loading cases include thermal accident scenarios and cask drop conditions with the contribution from the welding residual stresses. The determination of acceptable flaw size is based on the procedure to evaluate flaw stability provided by American Petroleum Institute (API) 579 Fitness-for-Service (Second Edition). The material mechanical and fracture properties for base and weld metals and the stress analysis results are obtained from the open literature such as NUREG-1864. Subcritical crack growth from stress corrosion cracking (SCC), and its impact on inspection intervals and acceptance criteria, is not addressed.

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DOE OSTI2020

Grimsel Test Site - A Successful International Underground Research Laboratory for Many Decades - 20429

For more than 35 years, Nagra and its partners from around the world have been conducting underground research projects at the Grimsel Test Site (GTS, www.grimsel.com) to contribute to the development and confirmation of safe geological disposal concepts and for the characterization of suitable host rock formations. Over the years, the results of this internationally recognized research program have been, and continue to be, incorporated directly into exploration programs, modelling, safety, and engineering feasibility studies on options for deep geological repositories. Each project of the GTS program involves field-testing, laboratory studies, design and modelling tasks, and integrates all scientific and technical aspects. Each project phase is planned with a duration of three to five years to facilitate practical and administrative aspects and allow flexibility for updating the overall project plans with the latest findings. Scientific and engineering interaction among the different projects is ensured via an international steering committee meeting. Hosting an IAEA level C radiation- controlled zone, which allows use of radionuclides, including actinides such as thorium, uranium, neptunium, plutonium and americium, in in-situ experiments is one of the reasons why GTS also developed as a center of excellence for work with radioactive tracers under realistic in-situ boundary conditions. Last year, a new five-year program (2019 to 2023) started which includes projects with a planning horizon of decades. The new five-year program includes a new phase of in-situ experiments using radionuclides such as migration experiments in the Colloid Formation and Migration project (CFM), the Long-Term Diffusion experiment (LTD) and the newly established C-14 and I-129 Migration in cement project (CIM). The 'High Temperature effects on Bentonite' (HotBENT) project is starting in the current phase and is studying the effects of elevated temperatures (>175 deg. C) on bentonite materials. As a generic underground research laboratory (URL) it is expected that the GTS will provide in the coming years a platform for international collaboration, knowledge development and knowledge transfer for the next generation of scientists and engineers in the area of radioactive waste disposal and geosciences. A key role regarding knowledge transfer and training is provided by the well-established Grimsel Training Center (GTC), which (beside many URL related issues) also covers many general aspects of radioactive waste management. In this paper we provide an overview of the current program at the GTS, focusing on the experiments that study the migration of radionuclides through engineered barrier materials and the geosphere. (authors)

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DOE OSTI2016

Intermediate Scale Testing Recommendation Report

A summary of recommendations for near-term intermediate-scale testing pertaining to a salt repository is provided in this report. Each proposal was asked to implement a phased progression, initiating with test plan production in FY 2017 and early-stage testing, if possible. Beyond 2017, testing is anticipated to progress to an underground setting and involve intermediate-scale field activities. Each test concept was presented at the June 6th 2016 meeting in Las Vegas NV and a team of DOE-NE, DOE-EM, and National Laboratory staff discussed the rnerits of each proposal. Discussions among managers and researchers in the weeks following the meeting led to selection of a path forward for phased testing that includes a series of small diarneter borehole tests designed to illuminate thermomechanical processes and potential vapor and brine transport. These tests are intended to be implemented at the WIPP facility and involve collaboration between SNL, LANL, and LBL. This document summarizes the test concepts generated by the te s of researchers and decisions made subsequent to the June 6th meeting.

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DOE OSTI2019

Dual-Purpose Canister Filling Demonstration Project Progress Report

This report discusses the initial progress made at the Oak Ridge National Laboratory to support direct disposal of dual-purpose canisters (DPCs) using filler materials to demonstrate that the probability of criticality in DPCs during disposal to be below the probability for inclusion in a repository performance assessment. In the initial phase of a multi-phase effort that will result in a full-scale demonstration, a computational fluid dynamics (CFD) model was developed to gauge the filling process and to uncover any unforeseen issues. The initial filling simulations of the lower region (mouse holes) of a prototypic DPC show successful removal of the inner space voids and smooth, even progression of the liquid level. In the initial phase, flow through a pipe that is similar to the drain pipe in a DPC will be investigated separately to gain valuable insight of flow regime inside a pipe. The initial experimental setups for validating the computational filling model have been designed, and the various assembly parts are being procured. The experience gained from the initial experiments will be applied to the next steps toward a full-scale demonstration and to the validation of multiphysics filling simulation models.

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DOE OSTI2015

French management model for sustainable transportation of used fuel - 15376

Safe and efficient transportation of radioactive material is key to the success of our industry. Used fuel transports are inherently complex due to the nature of the material and the high visibility. Only a few countries have developed comprehensive used fuel transportation systems that require a wide range of equipment, services and expertise. In the many countries, such as the Unites States, where interim consolidated storage, recycling or geological repository are not available yet, similar systems will have to be implemented. For more than 50 years AREVA TN has safely shipped more than 7,000 used fuel transport casks. The transportation system that was initially developed in the 1970's has been adapted and enhanced over the years to meet more restrictive regulatory requirements and evolving customer needs, and to address public concerns. The numerous 'lessons learned' have offered data and guidance that have allowed for also efficient and consistent improvement over the decades. Based on AREVA TN extensive expertise, this paper will describe the different phases and milestones that need to be met to set up, license and operate a successful used fuel transportation management system. Transportation of spent fuel in France could not be performed effectively and efficiently without strong collaboration with Electricite de France (EDF) who plays a critical role as the shipper. EDF will provide a nuclear operator's perspective on the most critical aspects of a successful used fuel transportation program. (authors)

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DOE OSTI2013

Long-term monitoring experiences at the HADES underground lab and its relevance for radwaste repository monitoring

In the frame of its radwaste disposal research programme, SCK-CEN started the construction of the HADES underground research facility in 1980. Including several extensions and a comprehensive experimental programme, it has provided a lot of experience on monitoring. Monitoring is performed for many reasons: construction follow-up, field characterisation, investigation of phenomena, and model validations - in which the underground lab offers the opportunity for up-scaling conventional laboratory set-ups. Construction monitoring has allowed to develop and optimise the underground construction techniques in a previously poorly known environment, resulting in a well-mastered application of mechanised methods for gallery construction with minimal damage to the host formation. Access to this formation also allows its characterisation, both geotechnical, geological and geochemical, and the detailed investigation of phenomena such as fracturing and oxidation. Finally, instrumented set-ups allow to test various numerical models by comparing the observations with the predicted behaviour. The specific conditions of the underground laboratory put particular requirements to the sensors. These conditions include the long-term nature of many set-ups - typically several years to decades, the inaccessibility of many sensors after installation, high mechanical and water pressures, and corrosion. Combined with the fact that many sensors are custom made, obtaining and maintaining a fully functional instrumented set-up can be challenging. A lot of experience has therefore been gained which is very valuable when designing the monitoring of radwaste repositories - and it has allowed us to determine the critical success factors for monitoring. Engineers tend to look at this first from a technical viewpoint - and there are many technical aspects indeed that determine the reliability of monitoring. A first one is the combination of different observations ('redundancy') which can be implemented by the use of several sensors, different sensor principles, different (coupled) parameters, and the combination of point measurements with geophysical techniques. Cabling is also a critical issue as it is often considered as the primary enemy of barrier integrity. Minimal cabling techniques, such as distributed fibre optic monitoring and wireless signal transmission, therefore get increasing attention. Also the interpretation of the monitoring data - in particular those that are perceived as 'wrong' or 'unexpected', needs sufficient attention. The long-term experience has however also shown that the design of a monitoring programme must look beyond the technical part. In particular for long-term applications, issues such as data management and record keeping are vital to guarantee success in this. (authors)

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DOE OSTI2018

Dual-Purpose Canister Filling Demonstration Project Progress Report, Rev. 1

This report discusses the initial progress made at the Oak Ridge National Laboratory to support direct disposal of dual-purpose canisters (DPCs)using filler materials to demonstrate that the probability of criticality in DPCs during disposal to be below the probability for inclusion in a repository performance assessment. In the initial phase of a multi-phase effort that will result in a full-scale demonstration, a computational fluid dynamics (CFD) model was developed to gauge the filling process and to uncover any unforeseen issues. The initial filling simulations of the lower region (mouse holes) of a prototypic DPC show successful removal of the inner space voids and smooth,even progression of the liquid level. In the initial phase, flow through a pipe that is similar to the drain pipe in a DPC will be investigated separately to gain valuable insight of flow regime inside a pipe. The initial experimental setups for validating the computational filling model have been designed,and the various assembly parts are being procured. The experience gained from the initial experiments will be applied to the next steps toward a full-scale demonstration and to the validation of multiphysics filling simulation models.

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DOE OSTI2013

Blue Ribbon Commission, Yucca Mountain Closure, Court Actions - Future of Decommissioned Reactors, Operating Reactors and Nuclear Power - 13249

Issues related to back-end of the nuclear fuel cycle continue to be difficult for the commercial nuclear power industry and for the decision makers at the national and international level. In the US, the 1982 NWPA required DOE to develop geological repositories for SNF and HLW but in spite of extensive site characterization efforts and over ten billion dollars spent, a repository opening is nowhere in sight. There has been constant litigation against the DOE by the nuclear utilities for breach of the 'standard contract' they signed with the DOE under the NWPA. The SNF inventory continues to rise both in the US and globally and the nuclear industry has turned to dry storage facilities at reactor locations. In US, the Blue Ribbon Commission on America's Nuclear Future issued its report in January 2012 and among other items, it recommends a new, consent-based approach to siting of facilities, prompt efforts to develop one or more geologic disposal facilities, and prompt efforts to develop one or more consolidated storage facilities. In addition, the March 2011 Fukushima Daiichi accident had a severe impact on the future growth of nuclear power. The nuclear industry is focusing on mitigation strategies for beyond design basis events and in the US, the industry is in the process of implementing the recommendations from NRC's Near Term Task Force. (authors)

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