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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Report on Field Test at INL Cask Farm of Single Detector Fast Neutron Spent Fuel Cask Verification System

Detecting diversion of spent fuel elements in dry storage casks is challenging due to the thick shielding used in cask construction. Measurements on top of the cask to map the underlying arrangement of the fuel elements and looking for anomalous changes over time has proven difficult to achieve using gamma rays due to the high scattering and attenuation from the thick steel structure, weakening information on the present or absence of fuel bundles. Simulations and laboratory experiments suggest that the high-energy neutron flux (>200 keV) measured directly above each fuel bundle is sufficient to produce a position map that enables detection of the present or absence of fuel bundles, and therefore diversion of a spent fuel bundle. A single-detector spent-fuel monitoring technique based on this principle was development at the Lawrence Livermore National Laboratory (LLNL). The INL Cask Farm in the INTEC technical area at Idaho National Laboratory (INL) offers the capability to test this technique on an MC-10 storage cask which has a distribution of full and empty fuel positions. An experimental test plan for the single-detector verification system was developed in consultation with INL personnel to be completed in FY2021. Due to travel advisories related to COVID-19, the experimental test plan was adapted to enable INL personnel to carry out the measurements in consultation with LLNL personnel following shipment of the LLNL system to INL. Field test measurements of the single detector verification system were successfully carried out at the INL cask farm on September 7-9, 2021. This document summarizes results of the field test.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Proposed Muon Radiography Method for Spent Fuel Cask Interrogation

Multiple scattering muon radiography is a technique that was developed at Los Alamos National Laboratory (LANL) to detect high Z material using cosmic ray muons as a radiation source. As a result,there has been considerable interest in using the technique to radiograph nuclear waste inside of storage casks. Recently,the technique has been used to interrogate the inside of a MC-10 dry fuel cask at Idaho National Laboratory (INL).This is the only technique that has been successful in imaging the interior of a spent fuel cask.The measurement was performed with the LANL Mini Muon Tracker (MMT), composed of two super-modules of drift tubes and readout electronics placed on each side of the fuel cask. The final result was a one dimensional histogram (projection) that showed areal scattering density within the cask binned according to horizontal position within the cask. While successful, the MMT is not designed for an optimal measurement of a fuel cask. The MMT only covers a limited solid angle around the cask and needs to be orientated so as to be aligned with the grid of stored fuel bundles. Even when properly aligned, the resulting areal scattering density projection indicates that a fuel bundle within a column is missing, but is ambiguous as to which row is lacking the fuel bundle. This paper discusses a potential new detector and analytical technique that may be used to better radiograph the nuclear material inside of a storage cask. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2018

Applications of Analytic Models to Spent Fuel Cask Analysis

The advent of high performance computing has changed the way scientists and engineers solve problems. Scientists previously would make a tremendous effort to simplify a complex problem to it’s essence by reducing geometry, neglecting changes and only preserving first-order effects. Now, scientist are able to leave the details in the problem and computationally arrive at a result. To develop the computational physics programs used by scientists, developers make assumptions and use approximation to reduce complex, continuous calculus to algebraic expressions which computers are able to solve.

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

Developing Design Criteria for Safeguards Seals for Spent Fuel Transportation Casks

Current designs for spent fuel transportation casks cannot ensure a cask's integrity during shipment, nor is there any verifiable means of maintaining continuity of knowledge (CoK) on a cask's contents. Spent fuel destined for encapsulation plants or geological repositories requires additional containment and surveillance (C/S) measures during shipment. Following final safeguards accountancy measurements on spent fuel assemblies, the shipment of verified assemblies will require unprecedented reliance on maintaining CoK on the fuel inside transport casks. Such increased reliance is due to the lack of reverification of spent fuel following encapsulation into disposal canisters and by meeting the requirement of dual C/S measures during such fuel shipments according to recommendations made by the Application of Safeguards to Geological Repositories (ASTOR) International Atomic Energy Agency (IAEA) expert group. By designing spent fuel transportation casks with effective seals integrated into their design, CoK can be more effectively maintained than by ad hoc C/S measures because seal integration ensures that a cask has not been tampered with. Externally applied seals might not be able to provide such assurance for currently designed spent fuel transportation casks, although some combination of seals, detectors, and/or a technology that can verify canister integrity might provide this assurance. This paper examines the design criteria for integrating safeguards seals into transportation casks and provides recommendations for near-term applications.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Classification and imaging of spent nuclear fuel dry casks using cosmic ray muons - 132

The present work explores alternative approaches for in situ monitoring with near real-time data analysis and imaging of dry casks using muons. Monitoring can provide useful signals about the cask contents in a yes/no decision format faster and with fewer muons than traditional imaging. Imaging can be performed as a second step in this process if specific information about the location of the fuel assemblies is needed. To explore and analyze these two modalities, simulations of muon scattering and displacement were performed for various scenarios, including vertical and horizontal fully loaded dry casks, half-loaded dry casks, dry casks with one row of fuel assemblies missing, dry casks with one fuel assembly missing, and empty dry casks. This work demonstrates monitoring can provide fast and reliable indication about the contents of the cask. For 100,000 muons, a few minutes of measurement time with a 3.6 m{sup 2} detector at a 45 deg. zenith angle, 90% detection probability for missing fuel assemblies can be achieved with a false alarm rate lower than 1%. If a positive signal is detected, then longer measurements can be performed and coupled with muon imaging algorithms to determine the location of missing nuclear material. In these cases, transmission muon radiography can produce images of acceptable resolution for all scenarios. In the horizontal dry cask case, all scenarios can be identified and separated adequately. In the vertical dry cask case, similar conclusions can be drawn, but because of shadowing from other fuel assemblies, the location of a missing fuel assembly cannot be identified in the 2-D transmission map. However, views from different directions may resolve this. (authors)

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

Radiation Templates of Spent Fuel in Casks

BNL and INL propose to perform a scoping study, using heavily collimated gamma and fast neutron detectors, to obtain passive radiation templates of dry storage casks containing spent fuel. The goal is to demonstrate sufficient spatial resolution and sensitivity to detect a missing fuel assembly. Such measurements, combined with detailed modeling and decay corrections should provide confidence that the cask contents have not been altered, despite loss of continuity of knowledge (CoK). The concept relies on the leakage of high energy gammas and neutrons through the shielding of the casks. Tests will emphasize organic scintillators with pulse shape discrimination, but baseline comparisons will be made to high purity germanium (HPGe) and collimated moderated 3He detectors deployed in the same locations. Commercial off-the-shelf (COTS) detectors and data acquisition electronics will be used with custom-built collimators and shielding.

36 MATERIALS SCIENCE↗
DOE OSTI2022

Uncertainty in Thermal Modeling of Spent Nuclear Fuel Casks

Uncertainty is a key metric in computational modeling that must be evaluated for results to have wide ranging applicability. A well characterized uncertainty range is ideal with clear error bars on results that can be presented to stakeholders. In the field of spent fuel cask modeling, this ideal has been historically difficult to achieve in practice because of the computationally intensive nature of the models used and the difficulty assigning reasonable uncertainties to quantities in as-built systems. The work in this report has been conducted to evaluate the overall state of uncertainty and sensitivity in spent fuel cask models and develop methodologies for evaluating these uncertainties. These methodologies must be practical for engineering applications. They should not require excessive computational resources or calendar time to achieve results. In engineering, the model must be on a scale such that it can be changed and adapted throughout a project as new information is discovered and project goals evolve. This report covers three major modeling task areas that provide an overview of the types of sensitivity and uncertainty present in a spent fuel storage and transportation system. Section 3 discusses sensitivity and uncertainty analysis in the effective thermal conductivity model for the fuel region and applies these results to a single assembly model. Section 4 shows sensitivity analysis of a full cask model in the TN-32B and Section 5 demonstrates the overall uncertainty workflow using Coolant Boiling in Rod Arrays – Spent Fuel Storage and STAR-CCM+ developed from the sensitivity work in the preceding sections.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

PWR commercial reactor criticals as validation for spent fuel transport cask burnup credit - 14489

The burnup credit approach for a spent fuel transportation cask (SFTC) with a high density basket for 32 PWR fuel assemblies required credit not only for actinides but also for fission products, otherwise the required minimum burnups of fuel assemblies would have been too high to be practical. At the time of the development of the burnup credit methodology for this condition, no established approach existed to quantitatively validate credit for fission products. A validation approach was therefore developed specifically for this application, based on Commercial Reactor Criticals (CRCs). In this approach, hot critical state points of PWR plants were evaluated using the same codes that were used for the burnup credit evaluations for the transportation cask, namely CASMO-4 and MCNP4. A total of 36 of those state-points were analyzed, from 5 plants with two different assembly types. For each axial section of each fuel assembly, a separate CASMO-4 depletion analysis was performed, which required a total of about 4000 individual CASMO runs. The spent fuel compositions were then extracted from the CASMO runs and transferred to MCNP, where each core was modeled, considering the existing 1/8 symmetry. The core operating conditions were taken from work previously performed and documented for the Yucca Mountain project. The results of the MCNP calculations were then analyzed to establish bias and bias uncertainty values. Additional studies were performed to compare the in-core condition for the fuel with the condition that exists for the fuel in the spent fuel transport cask. The paper presents and discusses the results of the calculations, and the analysis of those results to establish bias and bias uncertainty values. The paper also presents results of the studies comparing in-core with transport cask conditions, and discusses the resulting limitations of this approach. (author)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Shielding analysis of TN-32 spent fuel dry cask with SCALE

The objective of the University of Florida Non Destructive Evaluation (NDE) team is to use Emission Source Tomography to determine and verify the spent fuel location and condition in TN-32 dry casks. In order to investigate this technique, a radiation analysis of TN-32 cask will be performed with MAVRIC control module of SCALE 6.1 and MCNP6 code. The radiation source term from active fuel regions and fuel assembly structure components is calculated, and the MAVRIC geometry model is demonstrated in this paper. Finally, the procedure and future work is listed in this paper. And the shielding analysis result will be presented in ANS winter conference. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

MODELING A SPENT NUCLEAR FUEL CASK SEISMIC TEST

The US 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 canister dry storage system mockup. The shake table motions will represent a range of hypothetical earthquake conditions at hypothetical locations in the continental US to generally define the range of mechanical loads that SNF can be expected to experience during extended dry storage periods. The earthquake conditions will represent seismic hazards in the 2,000-to-20,000-year return period range. The test will use instrumented pressurized water reactor fuel assemblies with surrogate mass inside the fuel rods instead of radioactive fuel pellets. Dummy assemblies with similar mass and dimensions to actual fuel assemblies will occupy the rest of the fuel assembly locations in a SNF canister that holds 32 fuel assemblies. The canister will be located inside a mockup vertical concrete cask. Instrumentation will record the motion of the major components of the complex dynamic system, and strain gauges will be used to record the cladding strain at select locations. Preparations for the test require modeling predictions to identify the range of response of the system and to help select specific earthquake cases to be simulated on the shake table from a large set of potential cases. This paper describes the pretest nonlinear finite element modeling efforts that have been completed to date, including cask system level modeling and fuel assembly modeling in LS-DYNA. The cask system level models are critical for anticipating sliding or tipping of an unanchored cask during the test. The fuel assembly model is needed to estimate the range of cladding strain response and fuel assembly structural response to be expected during the test and cladding strain measurements will be key metrics for model validation and the development of modeling best practices after the test is concluded.

Klymyshyn, Nicholas A.↗
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 OSTI2015

Criticality Safety Analysis Of As-loaded Spent Nuclear Fuel Casks

The final safety analysis report (FSAR) or the safety analysis report (SAR) for a particular spent nuclear fuel (SNF) cask system documents models and calculations used to demonstrate that a system meets the regulatory requirements under all normal, off-normal, and accident conditions of spent fuel storage, and normal and accident conditions of transportation. FSAR/SAR calculations and approved content specifications are intended to be bounding in nature to certify cask systems for a variety of fuel characteristics with simplified SNF loading requirements. Therefore, in general, loaded cask systems possess excess and uncredited criticality margins (i.e., the difference between the licensing basis and the as-loaded calculations). This uncredited margin could be quantified by employing more detailed cask-specific evaluations that credit the actual as-loaded cask inventory, and taking into account full (actinide and fission product) burnup credit. This uncredited criticality margin could be potentially used to offset (1) uncertainties in the safety basis that needs to account for the effects of system aging during extended dry storage prior to transportation, and (2) increases in SNF system reactivity over a repository performance period (e.g., 10,000 years or more) as the system undergoes degradation and internal geometry changes. This paper summarizes an assessment of cask-specific, as-loaded criticality margins for SNF stored at eight reactor sites (215 loaded casks were analyzed) under fully flooded conditions to assess the margins available during transportation after extended storage. It is observed that the calculated k eff margin varies from 0.05 to almost 0.3 Δk eff for the eight selected reactor sites, demonstrating that significant uncredited safety margins are present. In addition, this paper evaluates the sufficiency of this excess margin in applications involving direct disposal of currently loaded SNF casks.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Test Plan for Cask Identification Detector

This document serves to outline the testing of a Used Fuel Cask Identification Detector (CID) currently being designed under the DOE-NE MPACT Campaign. A bench-scale prototype detector will be constructed and tested using surrogate neutron sources. The testing will serve to inform the design of the full detector that is to be used as a way of fingerprinting used fuel storage casks based on the neutron signature produced by the used fuel inside the cask.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗