The PRO-X program is actively supporting the design of nuclear systems by developing a framework to both optimize the fuel cycle infrastructure for advanced reactors (ARs) and minimize the potential for production of weapons-usable nuclear material. Three study topics are currently being investigated by Sandia National Laboratories (SNL) with support from Argonne National Laboratories (ANL). This multi-lab collaboration is focused on three study topics which may offer proliferation resistance opportunities or advantages in the nuclear fuel cycle. These topics are: 1) Transportation Global Landscape, 2) Transportation Avoidability, and 3) Parallel Modular Systems vs Single Large System (Crosscutting Activity).
This report documents the updated seismic shake table test plan. The report describes the shake table inputs (ground motions), test hardware, shake table facility, friction experiment, and proposed instrumentation.
This report is a preliminary test plan of the seismic shake table test. The final report will be developed when all decisions regarding the test hardware, instrumentation, and shake table inputs are made. A new revision of this report will be issued in spring of 2022. The preliminary test plan documents the free-field ground motions that will be used as inputs to the shake table, the test hardware, and instrumentation. It also describes the facility at which the test will take place in late summer of 2022.
Management of spent nuclear fuel and high-level radioactive waste consists of three main phases – storage, transportation, and disposal – commonly referred to as the back end of the nuclear fuel cycle. Current practice for commercial spent nuclear fuel management in the United States (US) includes temporary storage of spent fuel in both pools and dry storage systems at operating or shutdown nuclear power plants. Storage pools are filling to their operational capacity, and management of the approximately 2,200 metric tons of spent fuel newly discharged each year requires transferring older and cooler spent fuel from pools into dry storage. Unless a repository becomes available that can accept spent fuel for permanent disposal, projections indicate that the US will have approximately 136,000 metric tons of spent fuel in dry storage systems by mid-century, when the last plants in the current reactor fleet are decommissioned. Current designs for dry storage systems rely on large multi-assembly canisters, the most common of which are so-called “dual-purpose canisters” (DPCs). DPCs are certified for both storage and transportation, but are not designed or licensed for permanent disposal. The large capacity (greater number of spent fuel assemblies) of these canisters can lead to higher canister temperatures, which can delay transportation and/or complicate disposal. This current management practice, in which the utilities continue loading an ever-increasing inventory of larger DPCs, does not emphasize integration among storage, transportation, and disposal. This lack of integration does not cause safety issues, but it does lead to a suboptimal system that increases costs, complicates storage and transportation operations, and limits options for permanent disposal. This paper describes strategies for improving integration of management practices in the US across the entire back end of the nuclear fuel cycle. The complex interactions between storage, transportation, and disposal make a single optimal solution unlikely. However, efforts to integrate various phases of nuclear waste management can have the greatest impact if they begin promptly and continue to evolve throughout the remaining life of the current fuel cycle. A key factor that influences the path forward for integration of nuclear waste management practices is the identification of the timing and location for a repository. The most cost-effective path forward would be to open a repository by mid-century with the capability to directly dispose of DPCs without repackaging the spent fuel into disposalready canisters. Options that involve repackaging of spent fuel from DPCs into disposalready canisters or that delay the repository opening significantly beyond mid-century could add 10s of billions of dollars to the total system life cycle cost.
Currently, spent nuclear fuel (SNF) is stored in on-site independent spent-fuel storage installations (ISFSIs) at seventythree (73) nuclear power plants (NPPs) in the US. Because a site for geologic repository for permanent disposal of SNF has not been constructed, the SNF will remain in dry storage significantly longer than planned. During this time, the ISFSIs, and potentially consolidated storage facilities, will experience earthquakes of different magnitudes. The dry storage systems are designed and licensed to withstand large seismic loads. When dry storage systems experience seismic loads, there are little data on the response of SNF assemblies contained within them. The Spent Fuel Waste Disposition (SFWD) program is planning to conduct a full-scale seismic shake table test to close the gap related to the seismic loads on the fuel assemblies in dry storage systems. This test will allow for quantifying the strains and accelerations on surrogate fuel assembly hardware and cladding during earthquakes of different magnitudes and frequency content. The main component of the test unit will be the full-scale NUHOMS 32 PTH2 dry storage canister. The canister will be loaded with three surrogate fuel assemblies and twenty-nine dummy assemblies. Two dry storage configurations will be tested – horizontal and vertical above-ground concrete overpacks. These configurations cover 91% of the current dry storage configurations. The major input into the shake table test are the seismic excitations or the earthquake ground motions – acceleration time histories in two horizontal and one vertical direction that will be applied to the shake table surface during the tests. The shake table surface represents the top of the concrete pad on which a dry storage system is placed. The goal of the ground motion task is to develop the ground motions that would be representative of the range of seismotectonic and other conditions that any site in the Western US (WUS) or Central Eastern US (CEUS) might entail. This task is challenging because of the large number of the ISFSI sites, variety of seismotectonic and site conditions, and effects that soil amplification, soil-structure interaction, and pad flexibility may have on the ground motions.
The 30 cm drop is the remaining NRC normal conditions of transport (NCT) regulatory requirement (10 CFR 71.71) for which there are no data on the response of spent fuel. While obtaining data on the spent fuel is not a direct requirement, it allows for quantifying the risk of fuel breakage resulting from a cask drop from a height of 30 cm or less. Because a full-scale cask and impact limiters are very expensive, 3 consecutive drop tests were conducted to obtain strains on a full-scale surrogate 17x17 PWR assembly. The first step was a 30 cm drop of a 1/3 scale cask loaded with dummy assemblies. The second step was a 30 cm drop test of a full-scale dummy assembly. The third step was a 30 cm drop of a full-scale surrogate assembly. The results of this final test are presented in this paper. The test was conducted in May 2020. The acceleration pulses on the surrogate assembly were in good agreement with the expected pulses derived from steps 1 and 2. This confirmed that during the 30 cm drop the surrogate assembly experienced the same conditions as it would have if it had been dropped in a full-scale cask with impact limiters. The surrogate assembly was instrumented with 27 strain gauges. Pressure paper was inserted between the rods within the two long and two short spacer grid spans in order to register the pressure in case of rod-to-rod contact. The maximum observed peak strain on the surrogate assembly was 1,724 microstrain at the bottom end of the assembly. The pressure paper sheets from the two short spans were blank. The pressure paper sheets from the two long spans, except a few middle ones, showed marks indicating rod-to-rod contact. The maximum estimated contact pressure was 4,100 psi. The longitudinal bending stress corresponding to the maximum observed strain value (calculated from the stress-strain curve for low burnup cladding) was 22,230 psi. Both values are significantly below the yield strength of the cladding. The major conclusion is that the fuel rods will maintain their integrity following a 30 cm drop inside of a transportation cask.
The 30 cm drop is the remaining NRC normal conditions of transport (NCT) regulatory requirement (10 CFR 71.71) for which there are no data on the actual surrogate fuel. While obtaining data on the actual fuel is not a direct requirement, it provides definitive information which aids in quantifying the risk of fuel breakage resulting from a cask drop from a height of 30 cm or less. The 30 cm drop test with the full-scale surrogate assembly conducted in May 2020 was the last step needed for quantifying the strains on the surrogate assembly rods under NCT. The full-scale surrogate assembly used in the 2020 30 cm drop test was built using a new 17x17 Pressurized Water Reactor (PWR) Westinghouse skeleton filled with the copper rods and 3 zircaloy rods from the full-scale surrogate assembly used in the Multi-Modal Transportation Test (MMTT). Felt pads were attached to the surrogate assembly bottom prior to the 30 cm drop to adequately represent the effects of the impact limiters and the cask. Note that felt "programming material" has been used extensively in past drop tests and is known to be a good material for programming a desired shock pulse. The felt pad configuration was determined during a previous series of tests reported in. The acceleration pulses observed on the surrogate assembly during the test were in good agreement with the expected pulses. This confirmed that during the 30 cm drop the surrogate assembly experienced the same conditions as it would if it was dropped in the cask with the impact limiters.
Currently, spent nuclear fuel (SNF) is stored in onsite independent spent fuel storage facilities (ISFSIs), which is a dry storage facility, at 55 nuclear power plant sites. The majority of SNF in dry storage is in welded metal canisters (2,917 canisters at the end of 2019). The canisters are loaded for storage in storage overpacks (vertical casks or horizontal storage modules) and placed on outdoor concrete pads. Because the SNF will be stored at ISFSIs for an extended period of time, there is growing concern with regards to the behavior of the SNF within these dry storage systems during earthquakes. To address these concerns, the SFWST program is considering conducting an earthquake shaker table test. The goal of this test is to determine the strains and accelerations on fuel assembly hardware and cladding during earthquakes of different magnitudes to better quantify the potential damage an earthquake could inflict on spent nuclear fuel rods. The seismic integrity of the storage system has been addressed in the past by the US Nuclear Regulatory Commission and is not the focus of this potential test. Instead the DOE would benefit from knowing the condition of the fuel cladding from storage, transportation, to disposal so that it can ascertain repository performance for the fuel and packaging in its final state. A seismic event is part of the possible loading events that the fuel could experience in its lifetime. This report proposes several earthquake shaker table tests with different degrees of complexity. Alternative 1 was defined in the FY20 work scope. Alternatives 2 and 3 were recently developed to take advantage of the NUHOMS 32PTH dry storage canister that may be available in FY21 for this test at a minimum cost to the project. The selection of the alternative(s) will depend on the available budget and the SFWST program priorities for the near future.
The goal of this transportation analysis (TA) is to update the 2008 TA in order to evaluate the impacts associated with the transportation of transuranic (TRU) waste from waste generator sites to the Waste Isolation Pilot Plant (WIPP) facility and from waste generator sites to the Idaho National Laboratory (INL).
The DOE R&D program under the Spent Fuel Waste Science Technology (SFWST) campaign has made key progress in modeling and experimental approaches towards the characterization of chemical and physical phenomena that could impact the long-term safety assessment of heat-generating nuclear waste disposition in deep clay/shale/argillaceous rock. International collaboration activities such as heater tests and postmortem analysis of samples recovered from these have elucidated key information regarding changes in the engineered barrier system (EBS) material exposed to years of thermal loads. Chemical and structural analyses of sampled bentonite material from such tests has as well as experiments conducted on these are key to the characterization of thermal effects affecting bentonite clay barrier performance and the extent of sacrificial zones in the EBS during the thermal period. Thermal, hydrologic, and chemical data collected from heater tests and laboratory experiments has been used in the development, validation, and calibration of THMC simulators to model near-field coupled processes. This information leads to the development of simulation approaches (e.g., continuum vs. discrete) to tackle issues related to flow and transport at various scales of the host-rock and EBS design concept. Consideration of direct disposal of large capacity dual-purpose canisters (DPCs) as part of the back-end SNF waste disposition strategy has generated interest in improving our understanding of the effects of elevated temperatures on the EBS design. This is particularly important for backfilled repository concepts where temperature plays a key role in the EBS behavior and long-term performance. This report describes multiple R&D efforts on disposal in argillaceous geologic media through development and application of coupled THMC process models, experimental studies on clay/metal/cement barrier and host-rock (argillite) material interactions, molecular dynamic (MD) simulations of water transport during (swelling) clay dehydration, first-principles studies of metaschoepite (UO2 corrosion product) stability, and advances in thermodynamic plus surface complexation database development. Drift-scale URL experiments provides key data for testing hydrological-chemical (HC) model involving strong couplings of fluid mixing and barrier material chemical interactions. The THM modeling focuses on heater test experiments in argillite rock and gas migration in bentonite as part of international collaboration activities at underground research laboratories (URLs). In addition, field testing at an URL involves in situ analysis of fault slip behavior and fault permeability. Pore-scale modeling of gas bubble migration is also being investigated within the gas migration modeling effort. Interaction experiments on bentonite samples from heater test under ambient and elevated temperatures permit the evaluation of ion exchange, phase stability, and mineral transformation changes that could impact clay swelling. Advances in the development, testing, and implementation of a spent nuclear fuel (SNF) degradation model coupled with canister corrosion focus on the effects of hydrogen gas generation and its integration with Geologic Disposal Safety Assessment (GDSA). GDSA integration activities includes evaluation of groundwater chemistries in shale formations.
The data from the multi-modal transportation test conducted in 2017 demonstrated that the inputs from the shock events during all transport modes (truck, rail, and ship) were amplified from the cask to the spent commercial nuclear fuel surrogate assemblies. These data do not support common assumption that the cask content experiences the same accelerations as the cask itself. This was one of the motivations for conducting 30 cm drop tests. The goal of the 30 cm drop test is to measure accelerations and strains on the surrogate spent nuclear fuel assembly and to determine whether the fuel rods can maintain their integrity inside a transportation cask when dropped from a height of 30 cm. The 30 cm drop is the remaining NRC normal conditions of transportation regulatory requirement (10 CFR 71.71) for which there are no data on the actual surrogate fuel. Because the full-scale cask and impact limiters were not available (and their cost was prohibitive), it was proposed to achieve this goal by conducting three separate tests. This report describes the first two tests — the 30 cm drop test of the 1/3 scale cask (conducted in December 2018) and the 30 cm drop of the full-scale dummy assembly (conducted in June 2019). The dummy assembly represents the mass of a real spent nuclear fuel assembly. The third test (to be conducted in the spring of 2020) will be the 30 cm drop of the full-scale surrogate assembly. The surrogate assembly represents a real full-scale assembly in physical, material, and mechanical characteristics, as well as in mass.
The work for Step 1 performed at Sandia National Laboratories and reported in Section 7 has been updated to incorporate new data and to conduct new simulations using a new larger base case domain. The new simulations also include statistical analysis for different fracture realizations. A sensitivity analysis was also conducted to the study of the effect of domain size. A much larger mesh was selected to minimize boundary effects. The DFN model was upscaled to the new base case domain and the much larger domain to generate relevant permeability and porosity fields for each case. The calculations updated for Step 2 are described in Section 12.1. New calculations have also been conducted to model the flooding of the CTD and the resulting pressure recovery. The modeling includes matching of pressure and chloride experimental data at the six observation locations in Well 12M133. The modeling was done for the 10 fracture realizations. The Step 2 recovery simulations are described in Section 12.2. The Step 2 work is summarized in Section 12.3.
This report presents a comparative analysis of spent nuclear fuel management options to support the U.S. Department of Energy (DOE). Specifically, a set of scenarios was constructed to represent a range of possible combinations of alternative spent fuel management approaches. Analyses were performed to provide simple and credible estimates of relative costs to the U.S. government and to the nuclear utilities for moving forward with each scenario. The analyses of alternatives and options related to spent nuclear fuel management presented in this report are based on technical and programmatic considerations and do not include an evaluation of relevant regulatory and legal considerations (e.g., needs for new or modified regulations or legislation). This report has been prepared for informational and comparison purposes only and should not be construed as a determination of the legal permissibility of specific alternatives and options. No inferences should be drawn from this report regarding future actions by DOE. To the extent this report conflicts with provisions of the Standard Contract, those provisions prevail.
This report describes the Shaker Table Test conducted on September 12, 2018, at the Dynamic Certification Laboratories (DCL) in Sparks, Nevada. This report satisfies Milestone M3SF-19SN010202021 Shaker Table Test, Sandia National Laboratories (SNL) Work Package (Parent WBS # 1.08.01.02.02; Work Package #SF-19SN01020202). The Shaker Table Test is related to the Multi-Modal Transportation Test (MMTT) conducted in 2017. During the MMTT, accelerations and strains were measured on the transportation platform, ENsa UNiversal (ENUN) 32P dual-purpose rail cask, cradle, basket, and three surrogate 17x17 pressurized water reactor (PWR) assemblies (one from SNL, one from Spain and one from Korea).
U.S. knowledge in deep geologic disposal in crystalline rock is advanced and growing. U.S. status and recent advances related to crystalline rock are discussed throughout this report. Brief discussions of the history of U.S. disposal R&D and the accumulating U.S. waste inventory are presented in Sections 3.x.2 and 3.x.3. The U.S. repository concept for crystalline rock is presented in Section 3.x.4. In Chapters 4 and 5, relevant U.S. research related to site characterization and repository safety functions are discussed. U.S. capabilities for modelling fractured crystalline rock and performing probabilistic total system performance assessments are presented in Chapter 6.
Current practice for commercial spent nuclear fuel management in the United States of America (US) includes storage of spent fuel in both pools and dry storage cask systems at nuclear power plants. Most storage pools are filled to their operational capacity, and management of the approximately 2,200 metric tons of spent fuel newly discharged each year requires transferring older and cooler fuel from pools into dry storage. In the absence of a repository that can accept spent fuel for permanent disposal, projections indicate that the US will have approximately 134,000 metric tons of spent fuel in dry storage by mid-century when the last plants in the current reactor fleet are decommissioned. Current designs for storage systems rely on large dual-purpose (storage and transportation) canisters that are not optimized for disposal. Various options exist in the US for improving integration of management practices across the entire back end of the nuclear fuel cycle.
The Mizunami Underground Research Laboratory is located in the Tono area (Central Japan). Its main purpose is providing a scientific basis for the research and development of technologies needed for deep geological disposal of radioactive waste in fractured crystalline rocks. The current work is focused on the experiments in the research tunnel (500 m depth). The collected tunnel and borehole data were shared with the participants of DEvelopment of COupled models and their VALidation against EXperiments (DECOVALEX) project. This study describes how these data were used to (1) develop the fracture model of the granite rocks around the research tunnel and (2) validate the model.
Experimental hydrology data from the Mizunami Underground Research Laboratory in Central Japan have been used to develop a site-scale fracture model and a flow model for the study area. The discrete fracture network model was upscaled to a continuum model to be used in flow simulations. A flow model was developed centered on the research tunnel, and using a highly refined regular mesh. In this study development and utilization of the model is presented. The modeling analysis used permeability and porosity fields from the discrete fracture network model as well as a homogenous model using fixed values of permeability and porosity. The simulations were designed to reproduce hydrology of the modeling area and to predict inflow of water into the research tunnel during excavation. Modeling results were compared with the project hydrology data. Successful matching of the experimental data was obtained for simulations based on the discrete fracture network model.
An ENUN 32P cask supplied by Equipos Nucleares S.A. (ENSA) was transported 9,600 miles by road, sea, and rail in 2017 in order to collect shock and vibration data on the cask system and surrogate spent fuel assemblies within the cask. The task of examining 101,857 ASCII data files – 6.002 terabytes of data (this includes binary and ASCII files) – has begun. Some results of preliminary analyses are presented in this paper. A total of seventy-seven accelerometers and strain gauges were attached by Sandia National Laboratories (SNL) to three surrogate spent fuel assemblies, the cask basket, the cask body, the transport cradle, and the transport platforms. The assemblies were provided by SNL, Empresa Nacional de Residuos Radiactivos, S.A. (ENRESA), and a collaboration of Korean institutions. The cask system was first subjected to cask handling operations at the ENSA facility. The cask was then transported by heavy-haul truck in northern Spain and shipped from Spain to Belgium and subsequently to Baltimore on two roll-on/roll-off ships. From Baltimore, the cask was transported by rail using a 12- axle railcar to the American Association of Railroads’ Transportation Technology Center, Inc. (TTCI) near Pueblo, Colorado where a series of special rail tests were performed. Data were continuously collected during this entire sequence of multi-modal transportation events. (We did not collect data on the transfer between modes of transportation.) Of particular interest – indeed the original motivation for these tests – are the strains measured on the zirconium-alloy tubes in the assemblies. The strains for each of the transport modes are compared to the yield strength of irradiated Zircaloy to illustrate the margin against rod failure during normal conditions of transport. The accelerometer data provides essential comparisons of the accelerations on the different components of the cask system exhibiting both amplification and attenuation of the accelerations at the transport platforms through the cradle and cask and up to the interior of the cask. These data are essential for modeling cask systems. This paper concentrates on analyses of the testing of the cask on a 12-axle railcar at TTCI.
To support more rigorous analysis on global security issues at Sandia National Laboratories (SNL), there is a need to develop realistic data sets without using "real" data or identifying "real" vulnerabilities, hazards or geopolitically embarrassing shortcomings. In response, an interdisciplinary team led by subject matter experts in SNL's Center for Global Security and Cooperation (CGSC) developed a hypothetical case description. This hypothetical case description assigns various attributes related to international SNF transportation that are representative, illustrative and indicative of "real" characteristics of "real" countries. There is no intent to identify any particular country and any similarity with specific real-world events is purely coincidental. To support the goal of this report to provide a case description (and set of scenarios of concern) for international SNF transportation inclusive of as much "real-world" complexity as possible -- without crossing over into politically sensitive or classified information -- this SAND report provides a subject matter expert-validated (and detailed) description of both technical and political influences on the international transportation of spent nuclear fuel.
The experimental breeder reactor (EBR-II) used fuel with a layer of sodium surrounding the uranium-zirconium fuel to improve heat transfer. Disposing of EBR-II fuel in a geologic repository without treatment is not prudent because of the potentially energetic reaction of the sodium with water. In 2000, the US Department of Energy (DOE) decided to treat the sodium-bonded fuel with an electrorefiner (ER), which produces metallic uranium product, a metallic waste, mostly from the cladding, and the salt waste in the ER, which contains most of the actinides and fission products. Two waste forms were proposed for disposal in a mined repository; the metallic waste, which was to be cast into ingots, and the ER salt waste, which was to be further treated to produce a ceramic waste form. However, alternative disposal pathways for metallic and salt waste streams may reduce the complexity. For example, performance assessments show that geologic repositories can easily accommodate the ER salt waste without treating it to form a ceramic waste form. Because EBR-II was used for atomic energy defense activities, the treated waste likely meets the definition of transuranic waste. Hence, disposal at the Waste Isolation Pilot Plant (WIPP) in southern New Mexico, may be feasible. This report reviews the direct disposal pathway for ER salt waste and describes eleven tasks necessary for implementing disposal at WIPP, provided space is available, DOE decides to use this alternative disposal pathway in an updated environmental impact statement, and the State of New Mexico grants permission.
One of the major challenges of simulating flow and transport in the far field of a geologic repository in crystalline host rock is related to reproducing the properties of the fracture network over the large volume of rock with sparse fracture characterization data. Various approaches have been developed to simulate flow and transport through the fractured rock. The approaches can be broadly divided into Discrete Fracture Network (DFN) and Equivalent Continuum Model (ECM). The DFN explicitly represents individual fractures, while the ECM uses fracture properties to determine equivalent continuum parameters. We compare DFN and ECM in terms of upscaled observed transport properties through generic fracture networks. The major effort was directed on making the DFN and ECM approaches similar in their conceptual representations. This allows for separating differences related to the interpretation of the test conditions and parameters from the differences between the DFN and ECM approaches. The two models are compared using a benchmark test problem that is constructed to represent the far field (1 × 1 × 1 km3) of a hypothetical repository in fractured crystalline rock. The test problem setting uses generic fracture properties that can be expected in crystalline rocks. The models are compared in terms of the: 1) effective permeability of the domain, and 2) nonreactive solute breakthrough curves through the domain. The principal differences between the models are mesh size, network connectivity, matrix diffusion and anisotropy. We demonstrate how these differences affect the flow and transport. We identify the factors that should be taken in consideration when selecting an approach most suitable for the site-specific conditions.
In response to the expansion of nuclear fuel cycle (NFC) activities -- and the associated suite of risks -- around the world, this project evaluated systems-based solutions for managing such risk complexity in multimodal and multi-jurisdictional international spent nuclear fuel (SNF) transportation. By better understanding systemic risks in SNF transportation, developing SNF transportation risk assessment frameworks, and evaluating these systems-based risk assessment frameworks, this research illustrated interdependency between safety, security, and safeguards risks is inherent in NFC activities and can go unidentified when each "S" is independently evaluated. Two novel system-theoretic analysis techniques -- dynamic probabilistic risk assessment (DPRA) and system-theoretic process analysis (STPA) -- provide integrated "3S" analysis to address these interdependencies and the research results suggest a need -- and provide a way -- to reprioritize United States engagement efforts to reduce global nuclear risks. Lastly, this research identifies areas where Sandia National Laboratories can spearhead technical advances to reduce global nuclear dangers.