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At least 451 records · Page 25
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Coupling the Modeled Structural Transmissibility of a Used Nuclear Fuel Conveyance to Over the Road Data - 16215

The U.S. Department of Energy Office of Nuclear Energy (DOE-NE), Office of Fuel Cycle Technology, has established the Used Fuel Disposition Campaign (UFDC) to conduct research and development activities related to storage, transportation, and disposal of used nuclear fuel (UNF) and high-level radioactive waste (HLW). The mission of the UFDC is to identify alternatives and conduct scientific research and technology development to enable storage, transportation and disposal of used nuclear fuel and HLW generated by existing and future nuclear fuel cycles. The Storage and Transportation staff within the UFDC is responsible for addressing issues regarding the long-term or extended storage (ES) of UNF and its subsequent transportation. Current information is insufficient to determine the ability of UNF, including high-burnup fuel, to withstand shock and vibration loads that could occur when UNF is shipped by rail from nuclear power plant sites to a storage or disposal facility after extended storage. In order to make this determination, the magnitude of the transportation loads transmitted to the UNF must be quantified. Previous preliminary modeling work has shown how the structural transmissibility of the transport system can affect the magnitude of these loads and the importance of modeling all aspects of the transport system (i.e. rail car, transport cradle, cask, canister, and fuel). The work presented herein proposes a methodology for determining the structural transmissibility of a hypothetical transport system, and conditions existing over the road test (OTR) data to predict the performance of the hypothetical system. This method utilizes the modeled frequency response of the hypothetical system to scale the existing OTR acceleration data. This work will be relevant in creating models of a UNF conveyance during transport. As such, preliminary work in coupling the rail vehicle dynamics code NUCARS{sup C}, a general purpose multi-body rail vehicle dynamics simulation package developed by Transportation Technology Center Inc. (TTCI), to existing models of UNF during transport are also discussed herein. (authors)

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

Impact of modeling Choices on Inventory and In-Cask Criticality Calculations for Forsmark 3 BWR Spent Fuel

Simulation of boiling water reactor (BWR) fuel depletion poses a challenge for nuclide inventory validation and nuclear criticality safety analyses. This challenge is due to the complex operating conditions and assembly design heterogeneities that characterize these nuclear systems. Fuel depletion simulations and in-cask criticality calculations are affected by (1) completeness of design information, (2) variability of operating conditions needed for modeling purposes, and (3) possible modeling choices. These effects must be identified, quantified, and ranked according to their significance. This paper presents an investigation of BWR fuel depletion using a complete set of actual design specifications and detailed operational data available for five operating cycles of the Swedish BWR Forsmark 3 reactor. The data includes detailed axial profiles of power, burnup, and void fraction in a very fine temporal mesh for a GE14 (10×10) fuel assembly. The specifications of this case can be used to assess the impacts of different modeling choices on inventory prediction and in-cask criticality, specifically regarding the key parameters that drive inventory and reactivity throughout fuel burnup. This study focused on the effects of the fidelity with which power history and void fraction distributions are modeled. The corresponding sensitivity of the reactivity in storage configurations is assessed, and the impacts of modeling choices on decay heat and inventory are addressed.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

UNF-ST&DARDS Enhancements for RCCA Data in As-loaded Dual Purpose Cask Models

This report summarizes the work performed to enable detailed modeling of rod cluster control assembly (RCCAs) in dual purpose canisters (DPCs) in as-loaded configurations using the Used Nuclear Fuel – Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS). The goal of this project was to evaluate the reactivity impact that RCCAs have on k eff of DPCs with pressurized water reactor (PWR) fuel to potentially use the additional margin in future post-closure criticality safety analysis. This preliminary evaluation determined the number of DPCs that currently require compensatory actions prior to emplacement in a repository because of their high reactivity under post-closure criticality scenarios, which could be made acceptable by including the as-loaded RCCAs as specified in the Unified Database (UDB). This report briefly describes the modeling methods currently used within UNF-ST&DARDS and the modifications made to automate inclusion of the as-loaded RCCAs in the DPCs for post-closure criticality calculations for the loss of neutron absorber (NA) scenarios, or NA models. For the Zion site, the loss of basket scenarios, or degraded basket (DB) models, are also included. Discussion is also provided regarding the UDB data compared to available site-specific loading map data for the sites, which are specifically evaluated herein. This report compares the modifications to a similar evaluation for the Zion by site conducted by Walker as an initial validation. After the partial validation, all existing PWR sites with applicable NA models within the UDB were evaluated.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

A Burnup Credit Approach for Margin Estimation of Loaded Boiling Water Reactor Canisters in UNF-ST&DARDS

This paper discusses the modeling approach applied in the Used Nuclear Fuel Storage, Transportation, & Disposal Analysis Resource and Data System (UNF-ST&DARDS) to take credit for the reduced reactivity associated with the burnup of boiling water reactor (BWR) fuel for canister-specific as loaded criticality analysis. Burnup credit is routinely used for the analysis of pressurized water reactor (PWR) spent nuclear fuel (SNF) storage (wet), transportation, and disposal. The fresh fuel assumption has historically been applied to BWR SNF criticality analyses for dry storage and transportation, and the peak reactivity method of burnup credit has been applied to pool storage. The fresh fuel assumption ignores the credit for gadolinium burnable absorbers in BWR fuel, as well as the burnup of the fuel, and is therefore conservative. This work employs the framework traditionally used for PWR burnup credit analyses and applies available data and recently published regulatory guidance to justify a new approach to model the as-loaded burnup and enrichment of SNF assemblies in currently loaded dry casks. The following features of this criticality analysis approach are presented in this paper: (1) selection of axial burnup profiles for BWR fuel from publicly available sources using methods derived from recently published research, (2) justification of the selected axial burnup profiles as relevant to all types of BWR fuel, (3) justification for modeling the fuel assemblies with a uniform axial and radial enrichment, and (4) justification for modeling the axial void profile and control blade insertion during depletion. This analysis approach was used to evaluate 153 already-loaded BWR casks at 8 sites as a function of time in UNF ST&DARDS. Finally, the results of the as-loaded criticality safety margin assessments are presented. The results show the k eff calculated for storage and transportation cases range from 0.71 to 0.83, and the eigenvalues calculated for the disposal cases range from 0.85 to 1.03, though the highest cases are dominated by very limiting hypothetical damaged fuel assumptions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Shielding Analysis Capability of UNF-ST&DARDS

The Used Nuclear Fuel Storage, Transportation and Disposal Analysis Resource and Data System (UNF-ST&DARDS) is used to perform dose rate calculations for spent nuclear fuel (SNF) transportation packages based on the actual physical and nuclear characteristics (i.e., assembly design, burnup, initial enrichment, and cooling time) of the as-loaded SNF. Nuclear fuel data, transportation package model templates, and SNF canister loading map information residing within the tool facilitate automated generation of SCALE input files for radiation source term and dose rate calculations. Transportation package specific models developed for UNF-ST&DARDS dose rate analyses are described in detail. UNF-ST&DARDS dose rate analyses were performed in this paper for over 400 SNF canisters from 16 sites in their designated transportation casks. For simplicity, representative dose rate calculation results are presented as a function of time (i.e., selected calendar years between 2020 and 2100) for 73 SNF canisters in dry storage at four sites. For these canisters, the projected maximum dose rate values at 2 m from the lateral surfaces of the vehicle under normal conditions of transport (NCT) would vary between 1.9 and 11.5 mrem/h in 2020. Five SNF canisters will exceed the regulatory dose rate limit of 10 mrem/h at 2 m in 2020, and the analyzed SNF canisters will comply with regulatory dose rate limits by 2030. An analysis of the impact on the dose rate of fuel failure and reconfiguration during transportation indicated that the maximum dose rate for hypothetical accident conditions will be unaffected, and the NCT maximum dose rate at 2 m would have a maximum increase by a factor of 1.7 for a representative pressurized water reactor package and by a factor of 2.6 for a representative boiling water reactor package relative to intact fuel. Finally, analysis of the actual heat loading and the dose rate at 2 m from the lateral surface of the vehicle for the five SNF canisters exceeding the NCT regulatory dose rate limit of 10 mrem/h in 2020 showed that the dose rate could be more limiting with respect to regulatory requirements than the heat loading; i.e., the canister transportability date may be evaluated based on the transportation package’s external dose rate.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

UNF-ST&DARDS Unified Database and the Automatic Document Generator

The U.S. Department of Energy, Office of Nuclear Energy is sponsoring development of a database to store information related to spent nuclear fuel (SNF) in support of its Spent Fuel and Waste Disposition efforts. This database, referred to as the Unified Database (UDB), is part of a larger engineering analysis tool, the Used Nuclear Fuel Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS). The UDB provides a comprehensive, controlled source of SNF information, including dry cask attributes, assembly data, economic attributes, transportation infrastructure attributes, potential future facility attributes, and federal government radioactive waste attributes. There are a number of existing and envisioned data reports that can be expected to use data stored within the UDB; however, previously, there was not a streamlined method to couple the database to such data reports. Therefore, to streamline the creation of these reports, two methods were developed to generate documents from information in the database automatically. The first method used Java and LaTeX for automatically generating the report, and the second method used the Python programming language along with Sphinx, a Python documentation generator. There are some advantages and disadvantages to both approaches, but both methods produced equally high-quality, automatically generated reports that were directly coupled to the database. Finally, this study describes data currently available in the UDB; explains the two different methods for automatically generating reports from these data; and shows examples of inline text, figures, and tables automatically generated using both approaches.

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

Planning for Receipt of Mk-18A Recovered Nuclear Material at Oak Ridge National Laboratory

Oak Ridge National Laboratory (ORNL) ships and receives a wide variety of radioisotopes. These shipments will include the Mark-18A high-burnup targets, which are being processed at Savannah River National Laboratory to recover valuable isotopes that will then be shipped to ORNL for programmatic use. The Mark-18A targets will be processed to recover two products: a plutonium oxide stream and an oxide material that will contain the americium, curium, and lanthanides. A Type A shipping package has been designed, fabricated, and certified to transport the americium/curium/lanthanide oxide product in special form capsules. The package was designed with both gamma and neutron shielding for contents reading up to 5,000 R/hour. The plutonium stream will be shipped to ORNL in a Model 9975 Type B transport cask. A new content envelope for the Type B cask is being developed. This paper discusses the preparation underway at ORNL to receive the packages and transfer the material into hot cell processing and storage facilities for subsequent use in research and development applications and as feedstock for heavy actinide production.

Patton, Bradley↗
DOE OSTI2019

Perspectives on Previous DOE Transportation Acquisition Efforts - 19049

The Department of Energy (DOE) is planning for future large-scale transport of commercial Spent Nuclear Fuel (SNF) and High-Level Radioactive Waste (HLW) to eventual disposal and/or storage facilities. Various options for procuring transportation services are being evaluated. To inform this work, a review of past comments received on DOE transportation acquisition activities in the early 1990's through 2016 was conducted. Through the course of data collection, staff analyzed comments submitted from ten DOE initiatives from DoE's Office of Civilian Radioactive Waste Management (OCRWM) and DoE's Office of Nuclear Energy (DOE-NE). This paper summarizes and outlines key themes arising from these comments. Past issues identified by a wide variety of commenters (nuclear industry and cask vendors, nuclear utilities and purchasers, states, Tribes, local governments, rail carriers, nongovernmental organizations, trade organizations, engineering firms, and members of the public) were examined. These issues concerned DoE's activities to make maximum use of the private sector to transport and store SNF and HLW. The key themes analyzed include: Uniqueness of the Shipping Campaign, Views on Privatization, Industry Risk and Privatization, Unclear Roles and Responsibilities, and Responsibilities Recommended for DOE to Maintain. This paper concludes by discussing lessons that can be learned from analyzing these public comments and what that may mean moving forward in the U.S. radioactive waste management program. (authors)

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

Cosmic Ray Muon Tomography

This presentation discusses muons and their application in detectors, imaging and simulation.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2015

Impact of modeling choices on inventory and in-cask criticality calculations for FORSMARK 3 BWR spent fuel - 14441

Simulation of boiling water reactor (BWR) fuel depletion poses a challenge for nuclide inventory validation and nuclear criticality safety analyses. This challenge is due to the complex operating conditions and assembly design heterogeneities that characterize these nuclear systems. Fuel depletion simulations and in-cask criticality calculations are affected by (1) completeness of design information, (2) variability of operating conditions needed for modeling purposes, and (3) possible modeling choices. These effects must be identified, quantified, and ranked according to their significance. This paper presents an investigation of BWR fuel depletion using a complete set of actual design specifications and detailed operational data available for five operating cycles of the Swedish BWR Forsmark 3 reactor. The data includes detailed axial profiles of power, burnup, and void fraction in a very fine temporal mesh for a GE14 (10x10) fuel assembly. The specifications of this case can be used to assess the impacts of different modeling choices on inventory prediction and in-cask criticality, specifically regarding the key parameters that drive inventory and reactivity throughout fuel burnup. This study focused on the effects of the fidelity with which power history and void fraction distributions are modeled. The corresponding sensitivity of the reactivity in storage configurations is assessed, and the impacts of modeling choices on decay heat and inventory are addressed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

The NAC-SMP - a Next Generation Transformation Package - 18391

NAC International Inc. (NAC) is developing a Small Multi-Purpose (SMP) Type B(U)F-96 transportation cask designed with flexibility and cost-efficiency objectives to support shipment of a wide range of wastes with fissile material contents. The NAC-SMP package is a small, modular design that meets US DOT weight limits for road transport. Contents can be accommodated in multiple configurations including standard drums of up to 416 liters (110 U.S. gallons) in capacity. The package will also accommodate unique waste contents and fuel designs where cost-effective packaging options have been a challenge. The NAC-SMP development goals include simplicity of design and operational flexibility to meet Type B requirements. In particular, the cask body is designed to satisfy the dimensional and mass constraints, shielding requirements, and general transport requirements for a variety of contents under 10 CFR 71 and IAEA SSR-6 regulations. Flexibility is assured by the cask's modular design, which consists of multiple vessels to provide containment, shielding and overpack capability. The containment boundary design permits adaptability to content requirements by relying on interchangeable internal components for secondary containment, supplemental shielding and shoring, as required. In this paper, NAC provides an overview of the NAC-SMP cask and identifies the design features and technology advancements that makes the cask a readily adaptable and flexible solution for packaging US DOE and other orphan wastes, and to enable the integration of these wastes into the waste management system process and disposition infrastructure. (authors)

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

The transportation factor in site evaluations for spent nuclear fuel storage and disposal facilities - 15460

New facility siting studies will be necessary to fulfill the Blue Ribbon Commission (BRC) on America's Nuclear Future 2012 recommendation to develop consolidated interim storage (CIS) for spent nuclear fuel in the next decade. Future site evaluations for used nuclear fuel (UNF) storage and disposal facilities must address transportation requirements, especially access to mainline railroads, early in the site selection process. The U.S. Department of Energy (DOE) in 2013 proposed a strategy for managing spent nuclear fuel in accordance with the BRC recommendations by establishing a pilot interim storage facility by 2021 and a consolidated interim storage facility by 2025. The authors demonstrate that currently available geographic information systems and already adopted siting criteria (defining favorable and unfavorable conditions) allow the identification of the transportation portion of a storage facility siting process to begin without delay. Ready access to the mainline rail system is absolutely critical since existing rail access, or easily constructed access, is a highly favorable condition for successful siting. Despite widespread rail abandonments over the past three decades, vast areas of the United States retain mainline rail service or could connect to existing mainlines by building new rail spurs of less than 25-50 miles. This precondition for transportation planning is critical. Planners need to start with the beginning - availability of rail transportation to and from the interim storage facility/facilities. A second step in transportation screening would be to evaluate those potential site locations within 50 miles of existing mainline railroads using other appropriate variables. This effort could begin with the application of the 1984 repository siting guidelines for transportation (10 CFR 960.5-2-7)[1]. Repository surface facilities have essentially the same transportation requirements as storage facilities. The guidelines in 10 CFR 960 provide a readily available template for evaluating sites based on transportation risks, impacts and costs. Initial site screening, after considering proximity to mainline railroads as the precursory step, should evaluate: 1) proximity to highly populated areas; 2) proximity to interstate highways; 3) proximity to environmentally sensitive areas and lands protected by federal, state, or tribal governments; and 4) proximity to military installations, hazardous materials facilities, or other facilities that might impact transportation safety and security.Several challenges exist which prevent effective transportation planning. The diversity of canister and cask types and SNF characteristics and conditions, the increasingly limited rail access at the shipping sites and the ease with which public attention can be marshalled against a waste management program, all suggest that future efforts need to identify potential storage sites that reduce transportation impacts and costs by reducing near-site transportation construction and reduce total system transportation impacts. This can most easily be accomplished by reducing shipping distances in a process that is easily defined and implemented. The paper concludes that the DOE should develop a site screening and evaluation program with transportation as a central consideration. (authors)

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

Chute Silo ILW Project Berkeley - 16649

Magnox, a Tier 1 Site Licensed Company (SLC), is the management and operations contractor responsible for the decommissioning of twelve of the Nuclear Decommissioning Authority's (NDA) nuclear sites in the UK. Aquila Nuclear Engineering is a specialist Tier 3, turnkey engineering company based in the UK, delivering mission critical engineering solutions to Europe's nuclear and nuclear medicine industries. Berkeley Nuclear Power Station, Berkeley, England started generating electricity in 1962, ceasing operation 27 years later in 1989. The Chute Silo ILW Project at this Reactor Site had been set up to retrieve irradiated control rods and charge chutes disposed of within a subterranean vault (silo). The project involved in-silo size reduction of waste in to 30-35 waste baskets. The waste basket was then to be retrieved via a pintle grab and hoist mechanism in to a Transfer Flask (through a silo and a Transfer Flask gamma gate). The Transfer Flask was then to transit along existing crane rails, out of the silo building and over the top of an Import/Export Facility. A Ductile Cast Iron Container (DCIC) was to be imported into Import/Export Facility and inspected/prepared to receipt the waste basket from the Transfer Flask. Once filled, the DCIC was then to be sealed and exported to an on-site storage facility. Due to an issue with the Magnox Tier 2 Framework holder in delivering a cost effective and viable solution for their Chute Silo ILW Retrieval Project, Magnox took the ground breaking decision to 'self-perform'. This involved dividing the larger project into several smaller packages. Two of these packages - for the shielded Transfer Flask and the Import/Export facility - were put out for competitive tendering directly to the Tier 3 Supply Chain. The work completed by the Tier 2 Company was shared with the Tier 3's, but Magnox expressly requested the Tier 2 proposals were scrutinised and that far more cost effective, robust and simpler solutions were provided for within the Tier 3 Tenders. Aquila won both work packages and delivered them with significant and demonstrable cost savings, on time and under budget. This was achieved by the collaborative arrangements between Aquila and Magnox, the methodologies of both company's project teams and the innovation and 'fitness for purpose' of Aquila's design pragmatism. The original Tier 2 Framework holder's price to Magnox for the Chute Silo ILW Project was between UK Pounds 2-3 M. Aquila delivered the equivalent scope for under UK Pounds 1 M. Aquila's original programme was maintained despite the award of additional scope and (Magnox) delays on site. Aquila came under the original target price for each discrete phase of the project. The above was achieved because of: - the collaborative arrangements between Aquila and Magnox; - the methodologies of both company's project teams; - the design features put forward by Aquila (welcomed and adopted by Magnox). In 2014 Aquila Nuclear Engineering received formal recognition of its achievements on this project by winning an NDA Supply Chain Award for 'Best example of supplier 'going the extra mile''. Points highlighted in the award were: - improved customer focus, leading to stronger relationships - encouraging challenge and innovation - continuous suggestions for improvement - UK Pounds 1 million savings and three months programme reduction.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2014

Effects of Multiple Drying Cycles on High-Burnup PWR Cladding Alloys

The purpose of this research effort is to determine the effects of canister/cask vacuum drying and storage on radial hydride precipitation in high-burnup (HBU) pressurized water reactor (PWR) cladding alloys during cooling for a range of peak drying-storage temperatures, internal gas pressures, and hoop stresses. The HBU PWR cladding alloys have a wide range of hydrogen contents and varying hydride morphology after in-reactor service. Radial hydrides are a potential embrittlement mechanism for HBU cladding subjected to hoop-stress loading, which may be significant during normal cask transport. Ring compression tests (RCTs), which simulate pinch-type loading at grid spacers, are used to determine cladding ductility as a function of RCT temperature and the ductile-to-brittle transition temperature (DBTT). Previous tests were conducted with pressurized and sealed cladding rodlets heated to 400°C (the NRC ISG-11, Rev. 3, limit for all fuel burnups under normal conditions of storage and short-term loading operations) and cooled slowly at 5°C/h. Following this drying-storage simulation, the DBTT for HBU M5 ® decreased from 80°C to 70°C to <20°C as the peak hoop stress decreased from 140 MPa to 110 MPa to 90 MPa. Under similar drying-storage conditions, the DBTT for HBU ZIRLO™ decreased from 185°C to 125°C to 20°C to <20°C as the peak cladding hoop stress decreased from 140 MP to 110 MPa to 90 MPa to 80 MPa. These tests were conducted with a single heating-cooling cycle as simulation of the vacuum drying process with one drying cycle. (ISG-11, Rev. 3 allows <10 drying cycles, with cladding temperature variations that are less than 65°C each.)

Billone, M. C.↗
DOE OSTI2021

Modeling Environmental Effects on Ventilated Spent Fuel Storage Systems

This report describes newly developed external environment wind effects models of spent nuclear fuel (SNF) dry storage systems. The primary purpose of these wind effects models is to better understand particle deposition on SNF canisters in the context of chloride-induced stress corrosion cracking. The goal of this effort is to further the understanding and improve the wind effects model of the Nuclear Horizontal Modular Storage (NUHOMS ® ) Advanced Horizontal Storage Module–High Seismic (AHSM-HS) storage system in support of the planned Canister Deposition Field Demonstration (CDFD) project (Durbin et al. 2021, Suffield et al. 2021). The steps in this wind effects investigation begin with a validation case of comparing experimental data with a STAR-CCM+ computational fluid dynamics (CFD) model of the Belowground Vertical Dry Cask Simulator (BVDCS) with the external environment explicitly modeled. Next, a test case is performed with the AHSM-HS models comparing solar-loading strategies for a standalone and wind effects model in STAR-CCM+ and a standalone model in ANSYS Mechanical Parametric Design Language (APDL). The final portion of this study compares results of the validation exercise with applications of wind effects models for two horizontal storage systems, a site specific NUHOMS ® horizontal storage module and a NUHOMS ® AHSM-HS.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Tank Side Cesium Removal IXC-150 Cask Detonation Report

A set of experiments were conducted at Los Alamos National Laboratory’s firing point 88 to determine the blast pressure and effects of a stoichiometric gas mixture of Hydrogen (H 2 ) and Nitrous Oxide (N 2 O) on the IXC-150 Storage Pad Vent Stack assemblies. The Lab was asked to design, build, and test the vent stack assemblies using a reaction gas volume of 340 in 3 that mimics the Tank Side Cesium Removal (TSCR) Ion Exchange Column (IXC) assembly. The experimental system with vent stack assembly was designed with input from the Washington River Protection Services (WRPS) to ensure an accurate test setup. The tests were conducted at LANL on the newly built filter/gas assembly at the end of April 2021 timeframe, using WRPS supplied vent stack assemblies and internal HEPA filters. The objective of these tests was to observe that the structural integrity of the vent stack assemblies was maintained and determine the resulting blast overpressures at a distance of 16 and 32 inches from the assembly. All work was performed under the LANL Quality Assurance Program (SD330), using a graded application of ASME NQA-1- 2008/NQA-1a-2009. To ensure that all client quality assurance (QA) expectations were addressed, evaluation of the end data needs was performed and the appropriate controls applied for this work.

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

THERMAL MODELING OF HANFORD LEAD CANISTER’S HEATER BENCH TESTS

A computational fluid dynamics (CFD) model was built to simulate planned testing of heater assemblies for the Hanford Lead Canister (HLC) project. The HLC is a canister storage system that will contain heaters to simulate the decay heat of nuclear material and provide the canister storage system with environmental conditions equivalent to the operating conditions on a dry storage pad. The HLC will be equipped with long-term data collection and monitoring systems to provide an early warning of corrosion, pitting, cracking, or other signs of canister degradation that might threaten the integrity of the containment boundary over the potentially long term of dry storage. An important part of the HLC development is to confirm the function and ability of the electric heater assemblies that were specially designed to provide heating similar to the decay heat of nuclear material contained within the canister storage system. Heater bench testing is planned for early 2022 in a test configuration that does not include the canister. The goal of the bench testing is to verify that the heaters can replicate the decay heat of a canister with nuclear material and to validate the thermal models, which are critical to understanding the HLC’s thermal environment, including the local air flow within the canister storage system. Testing of the heater assemblies inside the canister system are planned in the future to validate canister level thermal models, and rigorous pre-deployment testing of the complete HLC cask and canister system is intended to be completed before the HLC is deployed in the 2025-2026 timeframe. This study presents the pre-test temperature predictions of the bench testing. A description of the heater assemblies and planned bench testing is presented. The model was developed with the commercial CFD code STAR-CCM+. An uncertainty analysis was run with the CFD model to determine the uncertainty in the temperature predictions and provide a range over which the predicted temperatures are expected to vary. The uncertainty analysis was performed by coupling STAR-CCM+ with the software Dakota, which provides advanced parametric analyses, including quantification of margins and uncertainty with computational models. This work is expected to provide insight into SNF canister behavior.

Suffield, Sarah R.↗