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All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Adjoint sensitivity analysis and data assimilation for verification of dry storage cask contents

Dry cask storage is a method for interim storage of spent fuel assemblies which contain fissile isotopes of uranium and plutonium. These can present a proliferation concern and consequently there is a need for non-destructive testing methods to verify a dry cask's contents for proliferation protection. We present an application of adjoint sensitivity analysis and data assimilation to a multigroup diffusion model of dry cask storage. Adjoint sensitivity analysis allows the efficient calculation of sensitivities for use in data assimilation to calibrate imprecisely known parameter values and data consistency tests to detect diversion scenarios. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Sensor Development for Liquid Water Detection in Dry Storage Casks: FY 2023 Status

Modeling efforts were undertaken in fiscal year (FY) 2023 to evaluate the feasibility and capability for sensing of liquid water inside canister-based dry cask storage systems (DCSSs). The focus was on the development of full-scale finite element models (FEMs) of ultrasound propagation through DCSS canister components. For these initial investigations, simulation of the baseplate component was targeted, envisioning water collection at the bottom of a vertically oriented canister. The effects of internal canister components (i.e., fuel basket, fuel assemblies) on ultrasound propagation in the baseplate component were evaluated in these efforts. The environment inside a DCSS confinement is intended to be inert and free of water to prevent potential corrosion of used fuel cladding or other internal hardware. Spent fuel assemblies are dried, after storage in water pools, to make sure water has been removed from assembly cavities. However, there is some uncertainty about the amount of residual water potentially left behind in a DCSS after drying processes, because water can become trapped in cavities or other small crevices in the surfaces formed by the fuel cladding, fuel assemblies, and other internal hardware components. Considering the complex space- and time-dependent temperature profiles in DCSSs, water may be in a liquid or gas phase depending on its location in the cask and how long the cask has been in storage. Evacuating most water and oxidizing agents contained within a canister is recommended by NUREG-1536 (NRC 2010), which covers DCSSs. As summarized by Salazar et al. (2020), existing guidance typically relies on achievable vacuum pumping pressures sustained over a hold period as a signal of dryness and water removal. However, time series data about water removal from full-scale commercial drying procedures are lacking (Hanson and Alsaed 2019). A review of drying specifications from several vendors led to the conclusion that if the specifications are followed correctly, the residual moisture in DCSSs should present an insignificant risk of cladding degradation (Knoll and Gilbert 1987). A more recent analysis concluded that much larger quantities of residual water could remain in DCSSs, but the amount would still not be expected to lead to significant corrosion of fuel cladding or other internal components (Jung et al. 2013). Industry drying procedures are mostly prescriptive in nature, and operational issues arising during the process could result in incomplete drying. A summary of operational issues and potential negative effects of residual water is provided by Salazar et al. (2019). Experimental efforts are ongoing to validate the extent of water removal in a DCSS based on drying procedures used at nuclear power plants through well-designed investigations of drying process efficacy and water retention (Durbin et al. 2021; Pulido et al. 2022a, Pulido et al. 2022b). This has been approached by simulating limited portions of a DCSS internal volume and fuel assemblies. So far, these efforts have focused on the effects of potential water trapping in the dashpot region of control rod guide tubes but are expected to be expanded to include the effects of other internal hardware features and failed fuel rod cladding. A method for detection and measurement of liquid water (Meyer et al. 2022; Meyer et al. 2021), in tandem with the drying information gained through experimental investigations, provides comprehensive bases for understanding the internal conditions of DCSSs and provides technical information to support licensing decisions.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Seismic Performance of Dry Casks Storage for Long- Term Exposure

The main goal of this study is to evaluate the long-term seismic performance of freestanding and anchored Dry Storage Casks (DSCs) using experimental tests on a shaking table, as well as comprehensive numerical evaluations that include the cask-pad-soil system. The study focuses on the dynamic performance of vertical DSCs, which can be designed as free-standing structures resting on a reinforced concrete foundation pad, or casks anchored to a foundation pad. The spent nuclear fuel (SNF) at nuclear power plants (NPPs) is initially stored in fuel-storage pools to control the fuel temperature. After several years, the fuel assemblies are transferred to DSCs at sites contiguous to the plant, known as Interim Spent Fuel Storage Installations (ISFSIs). The regulations for these storage systems (10 CFR 72) ensure adequate passive heat removal and radiation shielding during normal operations, off-normal events, and accident scenarios. The integrity of the DSCs is important, even if the overpack does not breach, because eventually the spent fuel-rods need to be shipped either to a reprocessing plant or a repository. DSCs have been considered as a temporary storage solution, and usually are licensed for 20 years, although they can be relicensed for operating periods of up to 60 years. In recent years, DSCs have been reevaluated as a potential mid-term solution, in which the operating period may be extended for up to 300 years. At the same time, recent seismic events have underlined the significant risks DSCs are exposed. The consideration of DCSs for storing spent fuel for hundreds of years has created new challenges. In the case of seismic hazard, longer-term operating periods not only lead to larger horizontal accelerations, but also increase the relative effect of vertical accelerations that usually are disregarded for smaller seismic events. These larger seismic demands could lead to casks sliding and tipping over, impacting the concrete pad or adjacent casks. The casks may also slide and collide with other casks or structural components. Also, the different DSC components may impact each other during these events. This study provides a comprehensive evaluation of DSCs subjected to these extreme demands, including the effect of vertical accelerations, and soilstructure interaction.

36 MATERIALS SCIENCE↗
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)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Aging Management Considerations for Dry Cask Storage Canisters - 18472

Dry cask storage systems (DCSSs) are used to store used nuclear fuel (UNF) at independent spent fuel storage installations (ISFSIs) prior to final geologic storage or reprocessing. As storage durations increase beyond the twenty-year period commonly implemented for licensing the storage sites, it is important to consider aging related degradation issues for DCSS components. A large portion of the DCSS designs used throughout the world employ a ventilated concrete overpack with a welded stainless steel canister that confines the UNF. This paper explores key issues related to aging management for these canister components, including: (1) canister environment, (2) inspection limitations, and (3) aging management program guidance and requirements. While the austenitic stainless steel used for the canisters are generally resistant to most forms of environmental attack, it is possible for conditions to arise that lead to pitting and even stress corrosion cracking of the canister systems. A large number of studies related to canister degradation have been performed by industry groups and government researchers, both in the US and worldwide. These advances in the collective knowledge provide insight into risk factors for DCSS. Such information is beneficial to improve ISFSI siting decisions and ensure successful relicensing of existing ISFSIs. This paper highlights relevant research and also discusses applicability to degradation of other equipment and containers used to process and isolate UNF and other nuclear waste streams from the public and the environment. Inspection of the canister surface is an important tool to investigate and evaluate the potential for degradation. Existing DCSS designs typically do not offer ready access to the canister surface in order to perform the inspection. Further complicating inspection efforts are the surface conditions present on the sheltered canisters after extended storage times, including atmospheric debris such as dust and pollen. Given ongoing development efforts, inspection capabilities and associated limitations should be assessed at multiple points in time over the course of an aging management program to incorporate new advances in the field. This paper explores the limitations of currently available capabilities in inspection and delivery, and provides insight into the gaps remaining to be filled to permit an effective inspection program. Recommendations and requirements related to programs for aging management of canisters have been published or are in progress from a variety of groups, including industry groups, regulatory authorities, and Codes and Standards. An effective aging management program should consider the technical bases of these various inputs and maintain compliance with requirements. Given multiple sources, decisions will be required when considering conflicting guidance, and this paper will include discussion of the relative merits of the various approaches. (author)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2022

Self-powered Through-wall communication for dry cask storage monitoring

Many nuclear facilities, such as spent fuel storage dry casks and nuclear reactor pressure vessels, are entirely sealed by metal layers to prevent harmful radiation. For safety and security operations, the temperature, pressure, radiation, and humidity inside the vessel needs to be closely monitored. However, no practical technology is currently available to realize the through-wall data communication and monitoring for these vessels due to the inside harsh environment of high temperature and nuclear radiation. In this paper, an innovative self-powered wireless through-wall data communication system for the nuclear environment is presented, which demonstrates a successful solution to such challenges. The presented system is composed of four modules, i.e., energy harvester with power management circuits, ultrasound wireless communication using high-temperature piezoelectric transducers, electronic circuits for sensing and data transmission, and radiation shielding for electronics. Here, constitutive functions of each module were firstly designed and followed by the system integration. Experiments were conducted subsequently to validate the designed functions and evaluate the performance of the integrated system. Results showed that the average power of over 40 mW was harvested from the thermal flow inside the nuclear spent fuel canisters which could provide enough energy to operate the sensing and data communication systems. The gamma radiation test results showed that the thermoelectric energy harvester and ultrasound transceivers can withstand radiation dosing over 100 Mrad. Furthermore, temperature shock tests demonstrated that the entire system including the shielded electronics can survive and maintain their functionalities at temperatures as high as 195°C. Under the in-lab mocked-up high temperature conditions and radiation shielding, the proposed system is foreseen to survive and operate stably for fifty years inside a nuclear spent fuel canister, and send the frequency modulated data out of the canister for 3 s in every 10 min.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Evaluation of Neutron Flux on the Surface of a PWR Dry Storage Cask - 17574

Since the construction of a dry storage facility for spent nuclear fuel is expected in Korea, several methods have been studied for assuring the safety of pressurized-water reactor (PWR) spent fuels in dry storage casks. Existing evaluation methods are mostly based on detection technology of gamma-rays. However, the gamma-ray detection methods have high uncertainties due to background radiation. Therefore, we developed an evaluation method for the spent nuclear fuel monitoring based on neutron detection. In this study, we evaluated neutron flux on the surface of a PWR dry storage cask using computational simulation. Neutron release rate and energy spectra from a Westinghouse (WH) 17x17 type spent fuel with 10 years cooling time were calculated using ORIGEN-ARP code. MCNP code was used to simulate geometry and material properties of the spent fuel assemblies and casks. Neutron flux after 10 years of cooling time of the spent fuels was calculated on the surface of top, side, and bottom of the casks. The metal cask and the concrete cask, called as KORAD-21 casks which newly developed by Korea Radioactive Waste Agency (KORAD), were selected for this study. As a result, for the metal cask, average neutron flux were 4.79x10{sup 2} neutrons/cm{sup 2} sec, 5.29x10{sup 2} neutrons/cm{sup 2} sec, and 1.47x10{sup 3} neutrons/cm{sup 2} sec on the surface of top, side, and bottom of the cask, respectively. For the concrete cask, average neutron flux on the surface of top, side and bottom of the casks were 5.07x10{sup 1} neutrons/cm{sup 2} sec, 1.18x10{sup 1} neutrons/cm{sup 2} sec, and 1.46x10{sup 3} neutrons/cm{sup 2} sec, respectively. The study results can be applied to evaluate characteristics of PWR spent fuels and the neutron flux of dry storage cask. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Computed Tomography of Dry Cask Storage for Used Fuel

The demand for dry cask storage for used nuclear fuel is on the rise due to the continuous operation of currently existing nuclear plants which are reaching or have reached the capacity of their used fuel pools. Dry cask storage allows used fuel that has already been cooled in the used fuel pool for at least five years to be surrounded by an inert gas inside a specially designed and licensed cask. Currently, 75 of the 78 U.S.A reactor sites either have or are pursuing an independent spent fuel storage installation (ISFSI) for dry cask fuel storage [1] with over 2,000 casks already loaded [2]. Cask loading will only continue to increase until a final repository solution is operational or the U.S. begins to recycle fuel. The challenge of monitoring the condition of the used fuel inside the cask without exposing the used fuel has been apparent since the 1980s. The objective of the University of Florida (UF) project is ultimately to investigate the feasibility of imaging the used fuel assemblies’ location and evaluating the inner structural condition using emission source computed tomography (CT). A combination of radiation transport models will be used to simulate gamma-rays and neutrons traversing through the used fuel, structure, and cask body materials. In the simulations, multiple linear detector arrays will record the radiation outside the cask wall and will be placed at various angles in order to model rotating a linear detector array around the entire cask. The combination of data from all the detectors will give the planar radiation projection onto the detectors, more commonly called the Radon transform, of the structure of used fuel assemblies. This project can be processed with filtered back projection or other more advanced reconstructive imaging algorithms to recreate the fuel structures. A majority of the historical work done for used fuel has been on verification, diversion detection, and burnup credit validation in order to meet regulatory guidelines. One of the more commonly used systems for single assembly verification is the Fork Detector developed at Los Alamos National Laboratory in 1988 [3]. A group at the Idaho National Laboratory developed a Compton Dry-Cask Imaging Scanner that has the ability to detect empty vs full storage positions but it suffered a setback when transported to the Doel Nuclear Power Plant near Antwerp, Belgium; the system was unable to detect full energy peaks because of the additional scattering caused by a thick ballistic steel shield [4]. As detectors, computational power, and modeling improves, a few groups have leveraged these advancements and started to investigate CT hoping, that despite the difficulties, the technique can give information on fuel condition. So far the groups have mostly limited the tomographic work to a per assembly basis [5] [6] [7], rather than the full dry cask as the University of Florida plans to investigate. The downsides of CT versus simpler verification and validation are the amount of data required and the time it takes to acquire that data. Given that dry casks are intended to sit undistributed for up to 60 years, a scanning time of several weeks to several months is not unreasonable if such systems do not require human operators beyond the initial setup.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Multiple-Angle Muon Radiography of a Dry Storage Cask

A partially loaded dry storage cask was imaged using cosmic ray muons. Since the cask is large relative to the size of the muon tracking detectors, the instruments were placed at nine different positions around the cask to record data covering the entire fuel basket. We show that this technique can detect the removal of a single fuel assembly from the center of the cask.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Reviewing Spent Nuclear Fuel Dry Cask Storage Programs to Identify Challenges for an Integrated Waste Management System - 17535

Under the Nuclear Waste Policy Act (NWPA), the U.S. Department of Energy (DOE) has responsibility for removing commercial spent nuclear fuel (SNF) from nuclear power plant sites and disposing of the SNF in a deep geologic repository. However, the methods used by commercial nuclear power plants for onsite storage of their SNF define the starting conditions for DOE efforts to develop a waste management system that includes packaging, transportation, storage, and disposal of commercial SNF. The NWPA (as amended) also charges the U.S. Nuclear Waste Technical Review Board (Board) with evaluating the technical and scientific validity of DOE activities related to the management and disposal of SNF and high-level radioactive wastes (HLW). DOE's efforts related to developing an integrated program for storing, transporting, and disposing of SNF and HLW were the focus of a public meeting held by the Board on August 24, 2016, in Washington, D.C. To prepare for the public meeting, the Board's staff compiled information on the dry cask storage systems in service at nuclear power plant sites in order to identify issues that would need to be addressed by DOE in planning to transport SNF away from the sites. This paper is based on a briefing presented to the Board members by the Board's staff before the public meeting. Post-meeting findings and recommendations of the Nuclear Waste Technical Review Board are documented in the Board letter to the Department of Energy dated December 8, 2016. A note about terminology: in this paper, the phrase 'dry cask storage system' is generic and applies to a complete system used for dry storage of commercial SNF.Within these storage systems, commercial SNF is sealed inside a container called either a cask or a canister depending upon the specific type of storage system used. This distinction is discussed in more detail in the body of this paper. When referring to the containers holding SNF in particular systems, this paper uses the terms 'casks' or 'canisters,' as appropriate. When discussing the collective group of all containers, this paper uses the phrase 'casks and canisters'. Dry cask storage of SNF is a widespread practice throughout the U.S. commercial nuclear industry. As of August 1, 2016, more than 2300 storage casks and canisters were loaded at U.S. commercial nuclear power plant sites. This number is expected to increase by more than 200 casks and canisters annually, reaching a total of approximately 4500 by 2025. Multiple vendors have developed their own unique designs of dry cask storage and transportation systems. The majority of storage system designs (75 percent) feature SNF loaded in welded stainless steel canisters that are stored in ventilated concrete overpacks that allow for convection cooling. Some older dry cask storage system designs do not include a welded stainless steel canister; instead, SNF is loaded directly into a heavily shielded metal cask to minimize radiation exposure to workers and prevent atmospheric release of radionuclides. These types of systems are sealed by O-rings compressed under a bolted lid. After loading, dry cask storage systems are located on concrete pads at Independent Spent Fuel Storage Installations (ISFSI) at nuclear power plant sites and will remain there until the time when they will be transported to a consolidated interim storage facility or a deep geologic repository. Eventual transportation of canistered SNF will include removing the stainless steel canisters from the concrete overpacks, possibly repackaging the SNF, and loading the canisters into specially designed transportation casks for shipment by truck, rail, or barge. In the case of the non-canistered casks, some are approved for transportation and can be shipped after impact limiters are installed on the casks. SNF will need to be removed from casks not approved for transportation and transferred to transport-approved casks. This paper highlights five issues and challenges presented by the nuclear industry's use of dry cask storage systems that will need to be addressed by DOE in developing an integrated waste management system. One example, among the five, is that several dry-storage casks and canisters have received Nuclear Regulatory Commission certificates of compliance for storage only and do not have certificates of compliance for transportation. This will pose a challenge when the SNF must be moved to a consolidated interim storage facility or a deep geologic repository. Another example is that, when loaded with SNF, the total weight of new, larger transportation cask designs will exceed 142 metric tonnes (156 tons) which, when combined with the weight of the rail-car, will exceed the gross weight limit for unrestricted operation on some rail lines. This will also affect future transportation planning for shipments of SNF. All five issues and challenges are discussed in detail in the body of the paper. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Sensitivity analysis for best-estimate thermal models of vertical dry cask storage systems

Loading requirements for dry cask storage of spent nuclear fuel are driven primarily by decay heat capacity limitations, which themselves are determined through recommended limits on peak cladding temperature within the cask. This study examines the relative sensitivity of peak material temperatures within the cask to parameters that influence both the stored fuel residual decay heat as well as heat removal mechanisms. Here, these parameters include the detailed reactor operating history parameters (e.g., soluble boron concentrations and the presence of burnable poisons) as well as factors that influence heat removal, including non-dominant processes (such as conduction from the fuel basket to the canister and radiation within the canister) and ambient environmental conditions. By examining the factors that drive heat removal from the cask alongside well-understood factors that drive decay heat, it is therefore possible to make a contextual analysis of the most important parameters to evaluation of peak material temperatures within the cask.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Detection of Missing Assemblies and Estimation of the Scattering Densities in a VSC-24 Dry Storage Cask with Cosmic-Ray-Muon-Based Computed Tomography

We report that highly energetic, cosmic-ray muons can penetrate a dry storage cask and yield information about the material inside it by making use of the physics of multiple Coulomb scattering. Work by others has shown this information may be used for verification of dry storage cask contents after continuity of knowledge has been lost. In our modeling and simulation approach, we use ideal planar radiation detectors to record the trajectories and momentum of both incident and exiting cosmic ray muons; this choice allows us to demonstrate the fundamental limit of the technology for a particular measurement and reconstruction method. In a method analogous to computed tomography with the attenuation coefficient replaced by scattering density, we apply a filtered back projection algorithm in order to reconstruct the geometry in modeled scenarios for a VSC-24 concrete-walled cask. We also report on our attempt to estimate material-specific information. A scenario where one of the middle four spent nuclear fuel assemblies is missing—undetectable with a simple PoCA-based approach—is expected to be detectable with a CT-based approach. Moreover, a trickier scenario where one or more assemblies is replaced by a dummy assembly is put forward. Lastly, in this case, we expect that this dry storage cask should be found to be not as declared based on our simulation and reconstruction results.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Field Test Report Neutron Scintillator Array Dry Storage Cask Scanner FY2024

During two weeks of Field Testing at the Idaho National Laboratory INTEC Cask Farm in July and August 2024, the LLNL Dry Storage Cask Scanner Array was lifted on top of an MC-10 dry storage fuel cask and operated to acquire neutron and gamma-ray data from the 24 fuel bundle positions. Neutron and gamma-ray data acquisition scans across the top of the cask of varying dwell times were performed July 15-18, 2024 and August 19-22, 2024 to evaluate the ability of the scanner data to reveal asymmetries in the fuel positions that reflect asymmetries in the MC-10 cask fuel bundle loading. The MC-10 cask 24 position fuel bundle loading at the INTEC Cask Farm is well documented, including the locations of six empty fuel bundle positions. This loading presents an opportunity to test the ability of the scanner system to detect diversion of spent fuel bundles as well as to validate the MC-10 cask MCNP modeling. The cask scanner array consists of six Stilbene crystal scintillator detectors and a linear actuator frame that moves the six detectors across the MC-10 dry storage cask to obtain data above each of the 24 fuel bundle positions. The detectors are connected to a pulse-shape discrimination data acquisition system capable of generating separate neutron and gamma-ray spectra for each detector and for each scan position. From the prior single detector Field Test in 2021 and iteration with MCNP modeling, the neutron and gamma-ray data were analyzed in multiple energy regions to identify an analysis method that would provide the strongest and most consistent signature of the asymmetric MC-10 cask fuel loading1 . From both the 2021 Field Test and the current Field Test results, the neutron capture gamma-ray count rate around 2.2 MeV provides the strongest signature of the asymmetric MC-10 cask fuel loading and has qualitative agreement with MCNP calculations. Counting all gamma-rays produces a similar signature. Neutrons emerging from the cask top are moderated and captured by the hydrogen in the polyethylene moderator and scintillator detector, producing a 2.2 MeV gamma ray which is seen in the scintillator gamma-ray spectrum. The count rate in the 2.2 MeV gamma-ray region is ~50 c/s, which is ~1000x higher than the ~0.05 n/s rate in the > 4MeV neutron region, and ~50x greater than the ~1 n/s rate in the neutrons > 500 keV region. Analysis of the 2.2 MeV neutron-capture Compton-scattered gamma-rays produces a statistically significant signature of the INTEC Cask Farm MC-10 asymmetric fuel loading. MCNP simulations indicate that the average neutron energy spectrum offers the potential to detect a large asymmetry from several missing bundles as well as individual missing fuel bundles. Testing this feature will require measurements on a cask with single missing elements.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Cosmic ray muon computed tomography of spent nuclear fuel in dry storage casks

Radiography with cosmic ray muon scattering has proven to be a successful method of imaging nuclear material through heavy shielding. Of particular interest is monitoring dry storage casks for diversion of plutonium contained in spent reactor fuel. Using muon tracking detectors that surround a cylindrical cask, cosmic ray muon scattering can be simultaneously measured from all azimuthal angles, giving complete tomographic coverage of the cask interior. This article describes the first application of filtered back projection algorithms, typically used in medical imaging, to cosmic ray muon scattering imaging. The specific application to monitoring spent nuclear fuel in dry storage casks is investigated via GEANT4 simulations. With a cylindrical muon tracking detector surrounding a typical spent fuel cask, simulations indicate that missing fuel bundles can be detected with a statistical significance of ~18σ in less than two days exposure and a sensitivity at 1σ to a 5% missing portion of a fuel bundle. Finally, we discuss potential detector technologies and geometries.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2019

Verification of dry storage cask loading using monoenergetic photon sources

Safeguarding the contents of spent nuclear fuel dry storage casks requires the verification of cask contents, especially if continuity of knowledge has been lost. The high attenuation of the cask’s thick shielding and stored fuel assemblies makes verification difficult using conventional photon radiography. Thomson-scattering quasi-monoenergetic photon sources offer narrow angular and energy spread for increased penetration. This paper demonstrates the advantages of such a source by examining effects of source properties on transmission through thick, highly attenuating objects. This information is then applied to calculations and simulations of transmission for a variety of photon sources and cask configurations. Here, the performance of the source is compared to traditional bremsstrahlung sources and other advanced methods, in the context of typical nuclear fuel dry storage casks. It is shown that the use of a Thomson scattering based quasi-monoenergetic photon source enables determination of fuel assembly occupancies faster and more accurately than other methods.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗