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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Developments in Type B Casks activities within EnergySolutions Cask Division - 16554

At the end of 2008, the industry experienced sudden shortage of Type B Low Level Radioactive Waste (LLRW) transport casks available for transportation of nuclear waste within the United States of America. Due to retirement of previously certified Type B legacy packages and in order to meet the industry demand requirements, EnergySolutions (referred to as the Company) started fabrication of four additional 8-120B packages and added them in a fleet in late 2014, essentially, doubling their fleet of Type B casks. One 3-60B cask was delivered in Spring 2015 to support transportation of irradiated hardware. With growing Type B cask fleet and additional resource demand during maintenance/operations of newer and older casks, the Company had experienced several challenges. This paper will discuss lessons learned from manufacturing of new casks and address, in detail, the current developments associated with operations of Type A and B casks. The topics of discussion will include the Company current cask inventory, major developments associated with user interface and equipment improvements in 2015, issues and challenges the Company maintenance team faces in a daily operations of the casks as well as the latest licensing update for EnergySolutions owned Type B packages. Also, the paper will discuss the company strategic direction for improvement of the cask operations culture and user interface experience. This information shall benefit owners of LLRW cask fleet to improve efficiency of operations, customers experience and satisfaction. (authors)

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DOE OSTI2018

Rail-Cask Tests: Normal-Conditionsof- Transport Tests of Surrogate PWR Fuel Assemblies in an ENSA ENUN 32P Cask.

This report describes tests conducted using a full-size rail cask, the ENSA ENUN 32P, involving handling of the cask and transport of the cask via truck, ships, and rail. The purpose of the tests was to measure strains and accelerations on surrogate pressurized water reactor fuel rods when the fuel assemblies were subjected to Normal Conditions of Transport within the rail cask. In addition, accelerations were measured on the transport platform, the cask cradle, the cask, and the basket within the cask holding the assemblies. These tests were an international collaboration that included Equipos Nucleares S.A., Sandia National Laboratories, Pacific Northwest National Laboratory, Coordinadora Internacional de Cargas S.A., the Transportation Technology Center, Inc., the Korea Radioactive Waste Agency, and the Korea Atomic Energy Research Institute. All test results in this report are PRELIMINARY – complete analyses of test data will be completed and reported in FY18. However, preliminarily: The strains were exceedingly low on the surrogate fuel rods during the rail-cask tests for all the transport and handling modes. The test results provide a compelling technical basis for the safe transport of spent fuel.

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DOE OSTI2019

Dry cask radiation shielding validation and estimation of cask surface dose rate with MAVRIC during long-term storage

The dry storage cask became a semi-permanent solution for spent nuclear fuel since no permanent solution has been finalized over the past three decades. Considering the difficulty of deep geological repository projects in many aspects, prolonging the service life of dry casks longer than a few decades has become an option for spent nuclear fuel storage. It must be ensured that dry cask safety objectives, including radiation shielding, criticality safety, structure health, decay heat removal, and containment be fulfilled during the prolonged service life. This paper presents numerical modeling results of the radiation dose rate distribution on a TN-32 cask surface aged up to 300 years. The numerical modeling method was validated with measurement data, and the difference between the calculated results and measurement data in the total dose rate is 7%. The detailed radiation distribution variations over cask surfaces along the axial and radial directions for 300 years of storage were obtained, and their azimuthal angular dependence was established. The maximum total dose rate over the cask surfaces is less than 10 mrem/hour (9.258 ± 0.212 mrem/hour) after 40 years, and this outcome fulfills all dose rate requirements from 10 CFR 71.47(b). After 300 years of storage, the average total dose rates on the side surface and top surface are 0.150 ± 0.007mremmrem/hour and 0.057 ± 0.003mremmrem/hour, respectively, while the maxima are 0.601 ± 0.014 mrem/hour and 0.112 ± 0.006 mrem/hour, respectively. The variations in the fuel gamma ray, structure gamma ray, neutron and secondary gamma-ray contributions to the total dose rate are presented and analyzed in this work. The neutron and gamma-ray dose rate contributions by energy group are presented and analyzed at the cask top, mid-plane, and bottom at the selected storage times.

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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 OSTI2023

Verification of Spent Fuel Inside Dry Storage Casks by Cask Top Fast Neutron Mapping (FY2023 Mid-Year Report)

This project is developing a prototype scanner array verification system for detection of missing fuel assemblies in spent-fuel storage casks. The prototype consists of six fast-neutron scintillator detectors mounted to a linear actuator frame that is placed on the top of a spent fuel cask to scan across all fuel assembly positions. The scanner array was assembled and tested at LLNL in FY2022. A field test schedule has been requested at the Idaho National Laboratory (INL) Cask Farm site for FY2023. Note that the Cask Farm contractor determines this scheduling and not INL directly. Further system automation will be designed and implemented with the goal of obtaining a level of system operation that meets IAEA needs. This includes integration of the scanner array and data-acquisition control software into a single interface for operator use. In addition, commercial operators and the IAEA may have special requirements for portability, shipping, lifting, and installation. Prior to the Field Test at INL, the system will be operated at LLNL to exercise lifting procedure and linear actuators, monitor stability of detector energy and pulse-shape discrimination calibration, and test system software integration efforts. Following the Field Test, we will present results and discuss the technology with the IAEA. We will incorporate additional improvements to the system based on lessons learned from the field test and feedback from the IAEA. If successful, the technology can be transferred to the IAEA or other stakeholders for assessment.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

FY2024 Mid-Year Report: Verification of Spent Fuel Inside Dry Storage Casks by Cask Top Fast Neutron Mapping

This project is developing a prototype scanner array verification system for detection of missing fuel assemblies in spent-fuel storage casks. The prototype consists of six fast-neutron scintillator detectors mounted to a linear actuator frame that is placed on the top of a spent fuel cask to scan across all fuel assembly positions. The scanner array was assembled and tested at LLNL in prior years. A field test schedule has been requested at the Idaho National Laboratory (INL) Cask Farm site for FY2024. Following the Field Test, we will present results and discuss the technology with the IAEA. The IAEA may have special requirements for portability, shipping, lifting, and installation. We will incorporate additional improvements to the system based on lessons learned from the Field Test and feedback from the IAEA. If successful, the technology can be transferred to the IAEA or other stakeholders for assessment.

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)

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DOE OSTI2019

Dry Cask Shipments of ATR Spent Nuclear Fuel at INL

The High Load Charger (HLC) cask will be replacing the Advanced Test Reactor (ATR) spent fuel transfer cask that has been used at Idaho National Laboratory (INL) to wet-transfer spent ATR fuel elements from ATR to Idaho Nuclear Technology and Engineering Center (INTEC) in order to support future plans at both facilities. INTEC intends to discontinue use of the CPP-666 storage pools in compliance with the Idaho Settlement Agreement and the basins will no longer be available to store spent ATR fuel elements. In order to support the core-internals change out in 2021 for ATR, spent ATR fuel elements must be removed from the canal and shipped to INTEC. INL contractors will use the HLC cask for safe, dry transfer of spent Advanced Test Reactor (ATR) fuel elements between the ATR Complex managed by Battelle Energy Alliance, LLC and the Idaho Nuclear Technology and Engineering Center (INTEC) managed by Fluor Idaho, LLC. The HLC cask (also known as the High Load Charging Cask or High Load Charger) is a lead-shielded, stainless steel transfer container originally designed in 1954 by the Idaho Nuclear Corporation to meet the radiation and contamination controls requirements of the time for shipment of spent nuclear fuel. The cask has been modified over the years and is currently used for wet-to-dry transfers at INTEC. It is approved only for out-of-commerce transfers. The cask will arrive at the ATR where eight spent ATR fuel elements will be loaded into two, newly-designed buckets. Each bucket is a quarter section, right-cylinder shape that has four compartments. The cask will be lifted out of the ATR canal allowing most of the water to drain from the cask. Processing the 12,500 kg cask will be completed in ATR and then transported on a tractor trailer to INTEC where the loaded buckets will be removed from the cask. The loaded buckets will undergo a drying process before placing the four buckets into a canister for long-term storage in the Irradiated Fuel Storage Facility (IFSF) at INTEC. A new transportation plan that was submitted to, and approved by, the U.S. Department of Energy to provide the cask safety basis, which describes the hazard analysis and accident analysis for the cask during transport. Supporting accident analysis include criticality, thermal, dose, and drop analyses. Of particular interest for this cask was the drop analysis from which was derived the speed limit for the transport. The drop analysis addressed drops in air onto an unyielding surface and in water onto an energy-absorbing pad for the cask in the ATR facility. These analyses are typically applied as the bounding drop scenarios during transport; however, this resulted in a safety basis speed limit of 12.9 kph. In order to more accurately represent the in-transit accident scenario, LS-DYNA software was procured and the model was updated to more closely represent the potential drop heights and impact surfaces of asphalt and soil for the cask while in transit. This analysis resulted in a safety basis speed limit of 48 kph. This safety basis speed limit is more practical to implement and control.

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DOE OSTI2018

Leakage Rate Testing Considerations and Implementation of ANSI N 14.5-2014 Standard within EnergySolutions Type (B) cask fleet - 18629

In March 2016, Nuclear Regulatory Commission (NRC) staff issued Information Notice 2016-04: 'ANSI N14.5-2014 Revision and leakage rate testing considerations'. American National Standards Institute (ANSI) N14.5, 'American National Standard for Radioactive Materials-Leakage Tests on Packages for Shipment', applies to leakage rate testing of Low Level Radioactive Waste (LLRW) Type B packages operated under 10CFR71. In their information notice NRC is encouraging Type B cask license holders and licensees (cask users) to adapt a new version of ANSI N 14.5 standard. Typically, leakage rate testing for Type B packages (or casks) is performed at different stages of cask development (design and fabrication leak tests) and operations (periodic/maintenance and pre-shipment leak tests) with the purpose to reasonably assure that the packaging will contain radioactive material under both normal and hypothetical accident conditions of transport described in 10CFR71. The standard was updated to provide a modern, robust and standardized approach for performance of leak rate testing on Type B casks by the industry stakeholders. Although implementation of new version of the leak testing standard may not be problematic for a modern Type B package designs, it may be extremely challenging to apply its principles to the existing legacy Type B casks which were designed/fabricated more then 20-30 years ago based on older requirements. Majority of such casks are not 'leak test' friendly. Some of legacy Type B packages may not readily available to meet newer ANSI N 14.5 standard requirements and will require implementation of alternative compensatory measures in lieu of the standard's requirements. In addition, the new standard mandates implementation of additional requirements for leak rate testing performance and personnel qualifications during routine maintenance of Type B casks by owner and during operations of casks by users. The focus of the paper will be to discuss ANSI N 14.5-2014 requirements and provide specific examples on how these requirements will impact the current leak rate test practices. This will be illustrated based on several Type B cask models operated within EnergySolutions fleet (both newer and legacy cask designs). Finally, the paper will cover considerations of the cask closure design features that have to be considered by Type B cask designers in order to ensure robustness of leak testing during design, fabrication and, most importantly, operation of a package. This information shall benefit both owners and users of Type B casks with respect to improvement of the cask designs and Type B cask operational experience. (author)

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