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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
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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.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
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 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)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Dry Cask Shipments of ATR Spent Nuclear Fuel at INL - 19600

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. (authors)

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DOE OSTI2026

USABILITY EXTENSION OF THE ŠKODA VPVR/M CASK FOR TRANSPORTING IRRADIATED FUEL ASSEMBLIES

New abstract from the final version being submitted now: This paper discusses the evolution and enhanced usability of the ŠKODA VPVR/M cask for the transport of irradiated fuel assemblies, particularly within the context of the demand for the delivery of newly appearing irradiated HEU fuel types for which the cask did not yet have a license to transport. Over time, the cask’s internal basket construction has demonstrated notable adaptability to accommodate various exotic HEU fuel types from research reactors of differing origins. The paper outlines sever-al custom internal baskets developed for specific fuel types, including those from Belarus and Serbia, as well as from Georgia, Uzbekistan, and MNSR cores, as well as a recently designed basket for MTR and TRIGA assemblies. The findings high-light the high flexibility and adaptability of the cask, supported by successful rede-signs and licensing efforts, underscoring its value for the safe and secure transport of nuclear material. Old abstract from the draft version that was already approved: The Russian Research Reactor Fuel Return (RRRFR) program, since its inception, has continuously used the ŠKODA VPVR/M Cask fleet designed for the repatriation of irradiated highly enriched uranium (HEU) fuel. As the program progressed (from shut-down and a quasi-abandoned reactor, and/or as it began to include fuels of Chinese and US origin), new challenges emerged for the transport Cask. These were fuel types that had not yet been licensed for the Cask. Although these requirements did not arise during the design of the basic ŠKODA VPVR/M Cask, as revealed by the retrospective analyses, the Internal Basket of the ŠKODA VPVR/M Cask gives a high degree of flexibility to accommodate additional fuel types. This paper provides a brief overview of the ŠKODA VPVR/M Cask, which holds a B(U) type license, and introduces the different types of Internal Baskets that have already been licensed to transport so-called exotic irradiated HEU fuel types, in addition to the original license. The paper presents a new Internal Basket design for accommodating MTR-type and TRIGA-type irradiated HEU fuel assemblies. This includes a detailed presentation of the design basis and the new MTR-TRIGA Internal Basket, as well as the licensing matters of the package under the name ŠKODA MTR-TRIGA Cask, and the conformity test (dry- and wet-run) operations made to verify compliance with the new Internal Basket. Then, as a summary, the usage record for the Cask fleet is presented, and finally, the paper concludes with the consolidated experiences gained during the utilization of the ŠKODA VPVR/M Cask fleet, emphasizing the high degree of Cask flexibility ensured by the Internal Basket’s construction.

42 - ENGINEERING↗
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 OSTI2018

Design and Development of the R79: A new Type B and IP-1-2-3 Transportation Cask for Historic Waste - 18570

In October 2014, Robatel began the design and development of a new Type B and IP-1-2-3 transportation cask, the R79. This cask is designed to transport 'historical' waste from nuclear power plants, nuclear research labs, and nuclear medical isotope production facilities. This 'historical' waste may consist of activated and/or contaminated components including irradiated metal from decommissioning activities, fuel assembly containers, consumables, dried residues, etc. It may also include dewatered ion exchange resins, dried concentrates (salt, metal oxide, sludge, etc.), and hydrogenous materials. The main isotopes in this waste are Co-60 and Cs-137. The cask is not designed to transport fissile material but a trace quantity of up to 22.5 g is allowed. For ease of loading and unloading, the cask is designed to handle waste prepackaged into standard drums. With a loading capacity of 300 kg (661 lb) and a volume of 350 L (12.3 ft{sup 3}), the R79 cask is well suited to carry small to medium-sized waste shipments. The cask body is constructed of 304L stainless steel, its impact limiters are filled with Robatel's FENOSOL{sup TM} foam, and the shielding material includes Robatel's PNT7{sup TM} compound for its thermal and neutron absorbing properties. The cask is loaded in a dry environment and is designed for road transportation. Some novel features were integrated into the design of this cask. With its impact limiters installed, the cask may be treated as a Type B container. Depending on the result of the gas generation analysis, the cask can be treated as either Type B(U) (no limit of the transport duration) or Type B(M) (transport duration limited to 2 weeks). With the impact limiters removed, the R79 cask functions as an industrial-type package (either IP-1, IP-2, or IP-3 depending on the content classification). This can be advantageous for a number of reasons. First, it takes less time for loading and unloading operations when the impact limiters are not installed. Secondly and more importantly, three industrial-type packages can be transported on the same conveyance compared to two Type B packages. These options allow the shipper more flexibility: if the content meets the lower radiation limits for an industrial type package, more waste can be transported on one shipment. If the content must be classified as Type B, the user only must install the impact limiters prior to shipment. The design and development of this cask was performed in 2015, with the first drop testing beginning in July 2015. The complete design documents were first submitted September 2015. Robatel completed the foam for the impact limiters in October 2016, with the completed impact limiters finished in March 2017. The end of manufacturing, delivery of the cask to our customer, and receipt of the regulatory certificate of approval is scheduled before the end of 2017. (authors)

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