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DOE OSTI2022

Insert Modeling in UNF ST&DARDS

The Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS) is a software tool that integrates a used nuclear fuel (UNF) or spent nuclear fuel (SNF) relational database and key analysis capabilities to simplify and automate numerous UNF management and fuel cycle–related activities. UNF-ST&DARDS is being developed for the US Department of Energy’s Office of Nuclear Energy Spent Fuel and Waste Disposition program. UNF-ST&DARDS provides an integrated framework that uses advanced modeling and simulation to predict the behavior of SNF over the timescales associated with permanent disposal in a geologic repository. After leaving the spent fuel pool, SNF is transferred to dry storage in a dual-purpose canister (DPC). DPCs are considered “dual purpose” because they are designed for both storage and transportation, removing the need to transfer the fuel to a separate transportation cask. However, much research has been conducted investigating the feasibility of directly disposing of DPCs in geologic repositories. Direct disposal of DPCs could reduce worker exposure during repackaging, reducing the amount of low-level waste from the discarded DPCs and potentially saving billions of dollars. Therefore, direct disposal of as-loaded DPCs is desirable if it can be done safely.

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DOE OSTI2021

Monitoring of helium gas leakage from canister storing spent nuclear fuel: Radiological consequences and management

Chloride-induced stress corrosion cracking (CISCC) of canisters has been identified as a potential degradation mechanism requiring ageing management — especially in a marine air environment. Methods for monitoring helium gas leakage from canisters have been developed on the basis of a phenomenon that the bottom temperature of the canister increases as the top temperature of the canister decreases when helium gas leaks during storage. In this paper, we consider the radiological consequences of CISCC based on two scenarios of helium gas leakage and propose effective management that includes a timeline for monitoring and mitigation. Monitoring helium gas leakage by measuring the difference in temperature between canister surfaces in the dry cask storage of spent fuel (SF) can detect leakage before SF undergoes oxidation as a result of air ingress and before country-specific regulatory limits of off-site doses are exceeded, thereby enabling mitigatory action.

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DOE OSTI2022

System Analysis Modeling and Intermodal Transportation for Commercial Spent Nuclear Fuel

The United States Department of Energy (DOE) Office of Nuclear Energy is applying knowledge and understanding in the areas of systems level engineering, analysis, and decision making to better inform the waste management pathways for U.S. spent nuclear fuel (SNF). Currently there are 93 operating and 23 shutdown commercial nuclear reactors in the United States. SNF at most of these locations is being stored in spent fuel pools, dry storage, or both [1, 2]. This paper initially goes over the basics of the agent-based simulation tool known as the Next Generation System Analysis Model (NGSAM) that has the capability to model interaction and movements of individual components or groups like casks, railcars etc. The next section covers some of the methods implemented in the Java Transportation Operations Model (JTOM) to facilitate the movement of assemblies, casks, railcars etc. and the various intermodal transfer options. This section also covers the current cask loading and intermodal transfer times implemented in NGSAM. The last section goes over the proposed values for intermodal transportation, and cask transfer/loading operations.

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DOE OSTI2023

Neutron Absorber Plate Characterization Plan for Criticality Experiments Design

After being used in nuclear installations, depleted fuel can still be highly reactive and must be handled securely to prevent any radiological or criticality concerns. In particular, spent fuel from use in nuclear power reactors must be stored and transported in specifically designed containers using neutron absorber materials to prevent criticality. Various neutron absorber material types exist and are manufactured by various entities, as thoroughly described in the Handbook of Neutron Absorber Materials for Spent Nuclear Fuel Storage and Transportation Applications written by EPRI. Presently, one of the most modern and most widely used types of neutron absorber material contains particles of boron carbide, or B 4 C, embedded in aluminum matrix: Boralcan, manufactured by Rio Tinto. It is very important for the community to know as much as possible about such neutron absorber materials. Therefore, in the recent years, a US Department of Energy National Nuclear Security Administration–Nuclear Criticality Safety Program funded project initiated design of an experiment that places Boralcan neutron-absorbing plates in an established critical assembly using low-enriched uranium fuel at the Sandia Pulsed Reactor Facility/Critical Experiments (SPRF/CX) apparatus at Sandia National Laboratories. The goal of the experiment is to produce high-quality benchmark data to submit to the International Criticality Safety Benchmark Evaluation Project (ICSBEP), for use in validating calculational tools and nuclear data by criticality safety analysts. The project, named IER-554, is currently in its final design stage, following a successful preliminary design. In the work documented in the design study, ten critical configurations using Boralcan neutron absorber plates were designed, and the experiment was proven to be feasible, with a predicted low k eff uncertainty around 100 pcm. An overview of the modeled cutout of the critical assembly with a Boralcan plate is shown in Figure 1, representing one of the configurations planned for the critical experiments. Before the plates are inserted in the critical assembly, it is necessary to know more about their composition and uniformity. This summary focuses on the plate characterization plans. Each plate will undergo (1) neutron transmission measurements at different locations to determine the 10 B areal density and (2) an in-depth x-ray computed tomography (XCT) examination to obtain the exact Sizes and distribution of the B4C powder particles inside the plates. In parallel, plate modeling studies are performed with a goal to determine the validity of the currently used approximation of modeling the neutron absorber plates as a homogeneous mixture of Aluminum 1100 alloy and B4C— instead of explicitly modeling the B4C particles. By using the experimental 10 B areal density measurements, and the exact size and location of the B4C particles obtained by XCT, a plate model can theoretically be built that reproduces the plate with extremely high fidelity. The results of this modeling study could increase the confidence of the criticality safety community in its modeling methods when using this type of neutron absorber material, and the industry could use these validations to change the boron loading credit limits from the U.S. Nuclear Regulatory Commission standard review plan for dry cask storage of spent nuclear fuel. The modeling calculations are performed with SCALE 6.3.0 using the KENO V.a sequence for criticality calculations with the ENDF/B-VIII.0 continuous-energy cross section library.

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DOE OSTI2021

Skyshine Calculations for a Large Spent Nuclear Fuel Storage Facility with SCALE 6.2.3

The SCALE code system developed at Oak Ridge National Laboratory includes state-of-the-art capabilities for radiation source term and radiation transport simulations that can be used in numerous applications, including dose rate analyses of complex consolidated interim storage facilities (CISFs). A licensed CISF could be used to store tens of thousands of tonnes of spent nuclear fuel discharged from commercial power reactors using various cask and storage pad designs. A CISF design must comply with the regulatory requirements provided in 10 CFR Part 72, including requirements related to annual dose limits applicable to real individuals located beyond the area controlled by the licensee. Therefore, calculating a dose to the public is a necessary part of the licensing process for the construction of a CISF. These calculations are very challenging because of the complexity of the CISF design and the low magnitude of dose rate at large distances from the facility. This paper describes detailed far-field dose rate calculations performed for a proposed CISF using MAVRIC, the Monte Carlo radiation shielding sequence in SCALE 6.2.3, with automated variance reduction based on discrete ordinates calculations. The method presented in this paper uses a detailed Monte Carlo radiation transport simulation in one step from source to dose rate. A series of independent simulations was made using the complete site geometry (all casks present), but with only one cask containing radiation sources to obtain the dose rate maps produced by each storage cask. The CISF dose rate map was obtained by adding the dose rate maps produced by the independent individual cask simulations. Ample volumes of air and soil extending beyond the location of interest for dose rate calculation were included in the calculation model to properly simulate important radiation attenuation and scattering events that affect far-field dose rates. Furthermore, a comprehensive sensitivity study is included in this paper to illustrate the importance of selecting appropriate air volume, mass density, and composition for CISF skyshine dose rate calculations. Dry soil and soil containing water were analyzed to determine their effects on groundshine radiation.

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DOE OSTI2020

Dual Purpose Canister Reactivity and Groundwater Absorption Analyses

The current spent nuclear fuel (SNF) management strategy includes reliance on dry storage. Utilities are meeting their interim storage needs on an individual basis with use of large-capacity dry storage casks, with a current focus on meeting existing storage and transportation requirements, as disposal requirements are not currently available. These casks are commonly known as dual-purpose (i.e., storage and transportation) canisters (DPCs). However, a small percentage of single-purpose (storage only) systems is also being used to meet storage needs. These are included under the “DPC” heading. This report investigates the postclosure criticality safety aspects of DPCs. Placing large, heavy waste packages containing DPCs into a repository for direct disposal has not yet been implemented domestically or internationally. Therefore, direct disposal of DPCs represents new engineering and scientific challenges. Some of the engineering challenges that have already been addressed include handling and placement, use of ramps vs. shafts, use of hoists, use of transport equipment, and thermal management. Additionally, some studies have been conducted in the past regarding the feasibility of direct disposal from a criticality analysis perspective and have concluded that while possible, demonstrating subcriticality over the disposal time period is a challenge. The alternative to direct disposal of DPCs into a repository is to repackage the SNF into different canisters. The direct disposal of DPCs without cutting them open and repackaging is appealing because it could be more cost-effective, reduce the complexity of fuel management operations both in and outside reactor facilities, and result in less cumulative worker dose during interim storage and handling before eventual disposal in a deep geologic repository.

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DOE OSTI2022

Dual Purpose Canister Reactivity and Groundwater Absorption Analyses

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DOE OSTI2022

Next Generation System Analysis Model: Recently Added Features and Future Plans

To better enable informed decision-making regarding the back-end of the nuclear fuel cycle, the Integrated Waste Management System (IWMS) program within the U.S. Department of Energy, Office of Nuclear Energy (DOE-NE) has been sponsoring the development and application of system analysis tools capable of analyzing various system options for the management of spent nuclear fuel (SNF) and high-level radioactive waste (HLW).With these tools, IWMS architecture analyses are being conducted to support the future deployment of a comprehensive nuclear waste management system that considers all major back-end aspects of the nuclear fuel cycle (i.e., transportation, storage, and disposal). The Next Generation System Analysis Model (NGSAM) is an agent-based simulation software tool expressly designed to be capable of modeling features within various IWMS architectures. NGSAM imports data from Oak Ridge National Laboratory (ORNL)’s unified database (e.g., historic assembly information, thermal profiles for assembly heat, and at-reactor dry storage loadings) to ensure that each simulation initializes with a realistic representation of the state of commercial SNF in the U.S. Recent major enhancements implemented into NGSAM in the period since NGSAM was last presented at the WM2019 conference include: • Tracking of railroad escort car acquisition and buffer car acquisition • The addition of heavy haul truck (HHT) and barge routes for some sites, as well as support for user-defined inter-modal routes • Updates to the logic that checks the transportation cask thermal limit maps prior to package transport • An allocation method that predicts when reactor sites would pack assemblies from their spent fuel pools for dry storage, and prioritizes shipments directly from the pools of those reactor sites in the preceding periods (before the predicted loadings to dry storage), thus reducing the number of casks loaded into dry storage at reactor sites • The addition of “reactor site family” operational limits to restrict the number of loads of SNF taken from the pool and from dry storage at a given reactor site each year • Added support for multiple canister loading map options and packages with multiple compatible transportation overpacks • Updates to the handling of non-commercial SNF, including a new database that contains data to support the updates • Updates to allow analysis of hypothetical scenarios which include repackaging at reactor sites, e.g., for possible comparative analysis with other scenarios • Implementation of additional output reports, or modification of existing ones • The ability to generate and implement user edits via the NGSAM website • The ability to model loading SNF from pool storage at an interim storage facility (ISF) into dry storage at the ISF • The ability to model consolidating SNF from different existing storage containers at a DOE site into the same DOE standard canister • The ability to model transferring SNF casks from one transportation mode to another, e.g., from HHT to rail, referred to as transloading. These new features have improved NGSAM capabilities and users’ experience with the model. Preliminary NGSAM requirements for modeling advanced reactor fuels, reprocessing, treatment, and conditioning were considered, and this paper describes them at a high level. Other nuclear fuel cycle system analysis tools developed under sponsorship of DOE-NE, like the VISION code developed at Idaho National Laboratory, might be better suited for initial high-level analysis of those technologies and advanced fuel cycles. As technologies are developed and system concepts evolve, NGSAM could provide value by providing more detailed modeling of transport, storage, and disposal of spent fuel and wastes from advanced reactors and advanced fuel cycles at the fuel element and waste container level.

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DOE OSTI2022

A risk-informed assessment tool to support review of license amendment requests for spent nuclear fuel dry cask storage systems

We report that while risk-informed safety evaluations are common methodologies used by the U.S. Nuclear Regulatory Commission (NRC), their application in the field of commercial spent nuclear fuel (SNF) dry storage has received limited attention. For instance, license amendment requests (LARs) for SNF dry cask storage systems (DCSSs) are evaluated according to U.S. NRC standard review plans (SRPs). However, risk-informed LAR review strategies could improve the understanding of critical system modifications, and lead to more predictable, efficient, and consistent LAR review processes. A methodology to develop a risk-informed LAR review tool is presented. Although focused on DCSSs, this methodology can be used to develop similar tools for a broad spectrum of safety-relevant systems. The end product (i.e., the tool) includes a tree diagram to help visualize the review risk significances assigned to a predefined set of system modifications and supporting rationale documentation. Instructions support correct tool application and provide the user with a path to incorporate currently unevaluated modifications in the tool structure. The demonstration of the methodology leads to the conclusion that risk-informed reviews of DCSS LARs are possible, although the tool precision could benefit from additional risk information. Further, the results indicate that many typical LARs involve low-risk or medium-risk modifications.

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DOE OSTI2023

Friction Experiment Data Analysis

The Spent Fuel Waste Disposition (SFWD) program under the U.S. Department of Energy (DOE) is planning a seismic shake table test of full-scale dry storage systems of spent nuclear fuel (SNF) 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 during representative earthquakes. A concrete layer will be installed on the shake table before the test to simulate conditions representative of an ISFSI pad. In the shake table tests with the vertical cask, the cask will be free-standing because this is representative of all, except two, ISFSIs in the U.S. with vertical dry storage casks. The static and dynamic friction coefficients between the steel bottom of the cask and the concrete layer on the shake table are important parameters that will affect cask behavior during the test. These parameters must be known for the pre- and post-test modelling, data analysis, and model validation. The friction experiment was performed at the Engineering Department of the University of New Mexico (UNM) to determine the friction coefficients between a steel plate with the same finish as the bottom of the vertical cask manufactured for the test and different concrete surfaces. In this experiment the steel plate was fixed and the concrete sample was pulled over the plate with a constant displacement rate using an MTS machine. This allowed for collecting continuous horizontal force data over the length of the steel plate. Four displacement rates and three vertical loads were used. The tests were performed with four concrete blocks with different degrees of the surface roughness - light sandblast, light to medium sandblast, medium bush hammer, and heavy sandblast. The total number of tests was 48. The data were used to calculate static and dynamic friction coefficients.

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DOE OSTI2022

Verification of Spent Fuel Inside Dry Storage Casks by Fast/Epi-thermal Neutron Mapping (FY2022 NA-241 Annual Report)

Development of a spent-fuel verification method based upon positional measurement of fast neutrons emitted from the top surface of the dry storage casks. A proof-of-principle field test on a partially loaded spent-fuel cask will be conducted by measuring the fast neutron flux on the top of the cask as a function of position using a single detector. If a useful verification signature is demonstrated, a multidetector scanning array will be built and tested to demonstrate a practical rapid verification system.

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DOE OSTI2024

Pretest Modeling A Spent Nuclear Fuel Seismic Shake Test

The U.S. Department of Energy Spent Fuel and Waste Science and Technology (SFWST) program is planning to conduct a series of full-scale shake table tests to simulate hypothetical earthquake conditions and record the response of surrogate spent nuclear fuel (SNF) assemblies in a dry canister storage system mockup. The shake table motions will represent a range of hypothetical earthquake conditions at hypothetical locations in the continental U.S. to generally define the range of mechanical loads that SNF can be expected to experience during extended dry storage periods. This paper describes the pretest predictions made with LS-DYNA models of the mockup storage systems. The test will use two dry storage system configurations, a mockup vertical concrete cask (VCC) and a mockup horizontal storage module (HSM). The test will use a production-quality canister and basket. Within the canister will be four instrumented fuel assemblies with fuel rods containing surrogate mass and 28 instrumented dummy assemblies that are intended to match the mass and outer dimension of a fuel assembly. The finite element models include models of the VCC and HSM on the shake table to calculate the system level dynamic responses and separate single fuel assembly models to calculate stress and strain in fuel assembly components. Both types of models include nonlinear behavior like rod-to-rod contact and the ability for VCC’s to rock and slide. This paper presents the expected response of the VCC, HSM, and fuel assemblies to the shake table testing that is planned to start in April of 2024. The earthquake conditions represent seismic hazards in the 2,000-to-20,000-year return period range. The test data is expected to confirm the expectation that fuel rod cladding will remain intact, fuel assembly structural components like guide tubes will remain intact, and no significant VCC sliding or tipping will occur in the range of conditions to be tested with the shake table.

Klymyshyn, Nicholas A.↗

DOE OSTI2021

Dual Purpose Canister Reactivity and Groundwater Absorption Analyses (Rev. 7)

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DOE OSTI2020

Verification of Spent Fuel Inside Dry Storage Casks by Fast/Epi-thermal Neutron Mapping: FY2020 Annual Report to NA241 (Annual Report)

Lawrence Livermore National Laboratory has embarked upon development of a novel methodology for verification of spent fuel inside dry storage casks. The issue of verification of the integrity of spent fuel inside dry storage has been one of the most challenging technical problems facing IAEA for many decades. The difficulty arises because of the massive amount of shielding between the inner spent fuel and the outer surface where measurements can be performed. Despite many efforts, detection of useful gamma rays or neutrons from the outer surface has so far proven unsuccessful.

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DOE OSTI2023

Fabrication and Testing of DOE Standard Canister Closure Leak Test Assembly – 24139

DOE manages over 300 types of spent nuclear fuel (SNF), many of which are located at the Idaho National Laboratory (INL) site. Managing this large variety of SNF for storage, transportation, and disposal poses a challenge to DOE. The Idaho Cleanup Project and INL are collaborating on the DOE SNF Road-Ready Demonstration (“Road-Ready Demonstration”), which will develop and demonstrate the designs, technology, processes, and regulatory framework for packaging DOE-managed SNF for “road-ready dry storage.” Road-ready dry storage is an SNF management concept in which SNF is packaged into dry, sealed canisters that are then placed in onsite storage in anticipation of later transport and disposition. The forward-looking goal of the Road-Ready Demonstration is to establish the foundation for a large-scale road-ready dry storage program at the INL site. In support of establishing a large-scale road-ready dry storage program at the INL site, the Road-Ready Demonstration will first package Fort St. Vrain SNF currently in dry storage at INL into several DOE Standard Canisters (DOESCs). These DOESCs will in turn be loaded into another containment similar to commercial multi-purpose canisters. This multi-purpose canister will then be compatible with a transportation or storage system, such as a storage cask for interim storage or transportation package for offsite transport. These DOESCs will remain sealed over the course of their storage, transportation, and applicable disposal functions. The closure process for the DOESC will include fuel and basket loading, welding, inspection, leak testing, and, if needed, repair. As a follow-up to previous discussions on the design of the DOE Closure Leak Test Assembly (LTA), this report describes recent fabrication and testing efforts performed at INL. DOESCs are sealed by two sequential gas tungsten arc welds, both of which are performed by remotely operated and semiautomatic welding systems. The first weld is a circumferential pipe weld that completes the assembly of the canister body and lid assembly. The second and final closure weld attaches the vent plug to the vent socket via a butt joint. After the second weld is performed, the welds are helium leak tested using an evacuated envelope technique. The LTA was designed for both remote and manual operation. This report describes the fabrication and performance testing associated with the evacuated envelope technique. INL staff designed, fabricated, and tested the LTA at INL facilities. This testing included establishing technique and system sensitivities in accordance with ASME and American National Standards Institute N14.5 requirements. Forthcoming work will cover such areas as design optimization, process and personnel qualification, and implementation in Road-Ready Demonstration operations.

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DOE OSTI2021

Joint Work Plan for DPC Filler Testing

The US generates approximately 2,000 Metric Tons of Heavy Metal (MTHM) of commercial spent fuel (SNF) each year and currently stores ~ 85,000 MTHM of commercial SNF at 70+ reactor sites, the disposal of which is the responsibility of the US Department of Energy (DOE). SNF is initially stored in spent nuclear fueld pools (SFPs). SFPs were initially constructed by US utilities for temporary fuel storage, but with no final disposal pathway available, SFPs are reaching capacity. To allow continued operation of the nation's commercial nuclear reactor fleet, utilities started transferring SNF from SFPs (wet storage) to dry cask storage systems, typically using dual-purpose (storage and transportation) canisters (DPCs). And while DPCs were designed, licensed and loaded to meet Nuclear Regulatory Commission (NRC) requirements that preclude the possibility of a critical event during SNF storage and transport, they 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). DPC filler option criteria are detailed and materials that exhibit these attributes are explored. This document is an update of the SNL Joint Workplan on Filler Investigations for DPCs.

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DOE OSTI2021

Plan for Development and Application of Risk Assessment Approach for Transportation Package Approval of an MNPP for Domestic Highway Shipment

For nuclear reactors, Probabilistic Risk Assessment (PRA) has been conducted since the 1970s [e.g., see WASH-1400 (NRC 1975), NUREG-1150 (NRC 1990), and NUREG-1935 (Chang et al. 2012)]. PRA has also been used to assess a dry cask storage system at a nuclear power plant [see NUREG-1864 (NRC 2007)]. PRA techniques have also been applied to the transportation of spent nuclear fuel, most notably in NUREG/CR-4829 (Fischer et al. 1987), NUREG/CR-6672 (Sprung et al. 2000), and NUREG-2125 (NRC 2014). Additional guidance is provided in International Atomic Energy Agency (IAEA) IAEA-TECDOC-1346 (2003). Transportation PRA was also used in the evaluations of transportation impacts in reports such as the Repository Final Environmental Impact Statement (DOE 2002) and the Repository Final Supplemental Environmental Impact Statement (DOE 2008).

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DOE OSTI2022

Remote Monitoring Systems/Remote Data Transmission for International Nuclear Safeguards.

The use of remotely transmitted data from a nuclear facility under international nuclear safeguards to an inspectorate headquarters has been rapidly growing since inception as its value in reducing inspection effort and cost is demonstrated. There are opportunities for further growth moving forward including (1) the number of spent fuel casks in dry interim storage are increasing, leading to strain on inspection resources and potentially increased radiation exposure to inspectors, (2) the frequency of encapsulating spent nuclear fuel for final disposal in geological repositories occurs at a rate that may lead to the need for on-site inspectors unless systems can be developed to remotely transmit data, and (3) new facility types such as small modular reactors may rely heavily on remotely transmitted data due in part to remote locations of operation and mobility. Challenges need to be addressed too and include (1) hesitancy to implement remote data transmission by states, (2) data collection, transmission, security, and analysis, and (3) reliable power and communications. This report examines the evolution, equipment deployed, status, and opportunities/challenges of remote data transmission moving forward.

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