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DOE OSTI2018

Conformal core cooling and containment structure

A nuclear reactor includes a pressure vessel and a nuclear reactor core comprising fissile material disposed inside the pressure vessel at the bottom of the pressure vessel. A secondary core containment structure includes a containment basket comprising insulation with a maximum stable temperature of at least 2200K cladded by steel. The bottom of the pressure vessel and the nuclear reactor core are disposed inside the containment basket with the containment basket spaced apart from the bottom of the pressure vessel by a clearance gap. The containment basket may comprise zirconia insulation cladded by steel. In some embodiments the clearance gap between the containment basket and the bottom of the pressure vessel is no larger than one meter. The secondary core containment structure may further comprise conduits arranged to inject water into the clearance gap between the containment basket and the bottom of the pressure vessel.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2018

Conformal core cooling and containment structure

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2017

The SAVY 4000 Container Storage Program at Los Alamos National Laboratory - 17313

Incidents involving release of nuclear materials stored in containers of convenience such as food pack cans, slip lid taped cans, paint cans, etc. has resulted in the U.S. Defense Nuclear Safety Board concerns over the lack of prescriptive performance requirements for interim storage of nuclear materials. Los Alamos National Laboratory (LANL) has shared in these incidents and in response proactively moved into developing a performance based storage container design, the SAVY-4000{sup R}. The SAVY-4000{sup R} is the first vented general use nuclear material container demonstrated to meet the requirements of DOE M 441.1-1, Nuclear Material Packaging Manual. The SAVY-4000{sup R} is an innovative and creative design demonstrated by the fact that it can be opened and closed in a few seconds without torque wrenches or other tools; has a built-in, fire-rated filter that prevents the build-up of hydrogen gas, yet retains 99.97% of plutonium particulates, and prevents release of material even in a 12 foot drop. Finally, it has been fire then drop tested and will reduce the risk to the public in the event of an earthquake/fire scenario. This will allow major nuclear facilities to credit the container towards source term Material at Risk (MAR) reduction. The container was approved for nuclear material storage in the TA-55 Plutonium Facility on March 15, 2011, and there are currently over 2000 SAVY-4000{sup R} containers available for use within the facility. A Safety Analysis Report was developed to demonstrate Manual 441.1-1 compliance and was approved April, 2014. The first four SAVY-4000{sup R} containers were packaged with plutonium on August 2, 2011. Key aspects of the SAVY-4000{sup R} vented storage container design will be discussed which include design description, qualification and testing, surveillance strategy, and the design life extension program as enhanced by surveillance activities with the intent to extend well beyond the current five year design life. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2019

Development of the B-25 Low Level Radioactive Waste Container in Taiwan - 1900x

Chinshan Nuclear Power Plant will be permanently stopped its operation in the middle of 2019 and will be the first nuclear power plant stepping into the decommissioning phase in Taiwan. According to Nuclear Reactor Facilities Regulation Act, a nuclear facility must be dismantled in 25 years after being permanently ceased operation. In its decommissioning plan, the B -25 low level radioactive waste container was one of the proposed containers used for this decommissioning task. However, the B-25 waste container was not used or developed in Taiwan before. To meet this need, Institute of Nuclear Energy Research developed a new type of B-25 waste container and will apply to the authority for its usage license. This article provided the information of the low level radioactive waste containers have been approved in Taiwan and described the design specifications of the developed B-25 waste container. Besides, the numerical evaluation and test results were also included in this article. Based on the evaluation and test results, the developed B-25 waste container satisfied the requirements of the storage container and industrial package type 2. It is believed that the usage license would be granted soon after applying it to the authority. (authors)

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DOE OSTI2021

A study on computational fluid dynamics modeling of a refrigerated container for COVID-19 vaccine distribution with experimental validation

A key issue with the distribution of vaccines to prevent COVID-19 is the temperature level required during transport, storage, and distribution. Typical refrigerated transport containers can provide a temperature-controlled environment down to -30 °C. However, the Pfizer vaccine must be carefully transported and stored under a lower temperature between -80 °C and -60 °C. One way to provide the required temperature is to pack the vaccine vials into small packages containing dry ice. Dry ice sublimates from a solid to a gas, which limits the allowable transport duration. This can be mitigated by transporting in a -30 °C refrigerated container. Moreover, because the dry ice will sublimate and thereby release CO 2 gas into the transport container, monitoring the CO 2 concentration within the refrigerated container is also essential. In the present work, a 3D computational fluid dynamics model was developed based on a commercially available refrigerated container and validated with experimental data. The airflow, temperature distribution, and CO 2 concentration within the container were obtained from the simulations. The modeling results can provide guidance on preparing experimental setups, thus saving time and lowering cost, and also provide insight into safety precautions needed to avoid hazardous conditions associated with the release of CO 2 during vaccine distribution.

59 BASIC BIOLOGICAL SCIENCES↗

DOE OSTI2017

Containment Analysis Capability of UNF-ST&DARDS

The Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS) methodology to perform automated containment analyses for potential transportation packages based on canister loading map information is described in this paper, and its capability is illustrated with example results. The allowable leakage rate is calculated with the procedures provided in ANSI N14.5-2014 and NUREG/CR-6487, which were adapted for containment analysis of a transportation package containing fuel assemblies with different nuclear characteristics (e.g., initial enrichment, burnup, and cooling time) and clad integrity (intact or damaged). UNF-ST&DARDS applies different source term calculation methodologies for low-burnup fuel (LBF) (i.e., <45 GWd/tonne U) assemblies and high-burnup fuel (HBF) (i.e., ≥45 GWd/tonne U) assemblies. The LBF radionuclide activities are based on actual fuel assembly burnup, initial enrichment, and cooling time. Bounding radionuclide activities based on a fuel pellet burnup value of 65 GWd/tonne U and actual fuel assembly cooling time are used for HBF assemblies. The fraction of failed fuel rods and the release fractions for the contributors to releasable source terms recommended in NUREG-1617 are used in the containment analysis regardless of fuel assembly burnup. However, UNF-ST&DARDS supports different parameter values for LBF and HBF assemblies. Finally, the containment analysis methodology for as-loaded transportation packages is presented in detail, and the UNF-ST&DARDS containment analysis capability is illustrated with results for simulated transportation packages containing spent nuclear fuel canisters in dry storage at selected sites.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Water Resistant Container Technical Basis Document for the TA-55 Criticality Safety Program

Criticality safety at TA-55 relies on nuclear material containers that are water resistant to prevent significant amounts of water from coming into contact with fissile material in the event of a fire that causes a breach of glovevbox confinement and subsequent fire water ingress. A “water tight container” is a container that will not allow more than 50ml of water ingress when fully submerged, except when under sufficient pressure to produce structural discontinuity. There are many types of containers, welded containers, hermetically sealed containers, filtered containers, etc.

61 RADIATION PROTECTION AND DOSIMETRY↗

DOE OSTI2020

Pressurizing Hagan and SAVY containers to 30-psig (air) to measure the release of analytical cerium oxide test powder

In response to an ESS surrounding the storage of sealed sources at TA-55, personnel from RP-SVS (Radiation Protection Services) and ORI-2 (Operational Readiness & Execution) were approached by ES-55 (Facility System Engineering) to provide technical testing of known nuclear material storage containers. Testing criteria were determined by SB-PF (Safety Basis for Plutonium Facilities). Tests were conducted in conjunction with multiple related projects at the Aerosol Sciences Laboratory (TA-03-0130-0103). An experimental system was quickly developed to deliver a pressure pulse (30-psig) that mimics a sealed source burst scenario. A series of twelve tests was conducted. Six tests were done with two different 5QT SAVY-4000 containers, where each SAVY filter-lid combination was subjected to three successive test insults. Three tests were done with a (0.375" diameter filter) 8Q.T Hagan container, and three tests with a (0.625" diameter filter) 8QT Hagan container. An unused, fresh container filter was used for each tested Hagan container, each receiving only one test insult per filter. Cerium oxide (CeO 2 ) powder was loaded (100 grams per test) into a nozzle in the tested Hagan and SAVY containers, and the nozzle was hard-plumbed to a ball valve and a pressure source. This system was installed into the Los Alamos RRFMC (Respirable Release Measurement Chamber), which is an integrated multipurpose aerosol wind tunnel that satisfies NQA-1 subpart 2.4 for R&D work. The ball valve was fitted with a mechanical linkage for operation from outside the wind tunnel. An aerodynamic particle sizer counted the particle concentrations and size distributions of released aerosol. Respirable aerosol released during the tests was measured and a correction factor for wind tunnel flowrate and internal duct deposition was applied.

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DOE OSTI2021

Evaluation of Safety Standards for Fuel System and Fuel Container Integrity of Alternative Fuel Vehicles

There are two Federal Motor Vehicle Safety Standards (FMVSS) in place that specify requirements for integrity of the fuel system and fuel container on compressed natural gas (CNG) fuel vehicles. These are FMVSS Nos. 303, “Fuel system integrity of compressed natural gas vehicles,” and FMVSS No. 304, “CNG fuel container integrity.” At this time, no FMVSS are defined for the fuel system and fuel container integrity of propane and liquefied natural gas (LNG) vehicles or fuel system integrity requirements for heavy-duty CNG vehicles on the road. However, there are voluntary industry design standards and best practices, as well as regulations defined in other countries. FMVSS No. 303 specifies requirements for the integrity of the CNG fuel system of light-duty vehicles and school buses, and FMVSS No. 304 specifies requirements for fuel container integrity on all CNG vehicles. FMVSS No. 304 applies to containers used for vehicle propulsion, whereas containers used to transport CNG and transportation of CNG containers are regulated by the DOT Pipeline and Hazardous Materials Safety Administration. FMVSS Nos. 303 and 304 are performance-based standards to consistently test and validate equipment and are not design restrictive. Despite the increasing number of natural gas and propane medium- and heavy-duty vehicles on the road, there are no FMVSS fuel system integrity requirements beyond light-duty and school buses for CNG vehicles and no FMVSS fuel system integrity requirements for propane vehicles. NHTSA is researching fuel system safety for medium- and heavy-duty natural gas and propane vehicles to update FMVSS Nos. 303 and 304. NHTSA is also researching current best practices and standards for high pressure fuel tanks in motor vehicles as they may apply to FMVSS No. 304.

30 DIRECT ENERGY CONVERSION↗

DOE OSTI2017

Modeling of Radiolytic Production of Oxidants Relevant to Corrosion of a Used Fuel Container in Deep Geologic Repository Environments - 17641

The Canadian used nuclear fuel disposal concept involves a deep geological repository using a multi-barrier system. A key barrier is the used fuel container (UFC). It consists of an inner vessel of carbon steel, which provides the structural strength to withstand repository loads, and an outer coating of copper acting as an external corrosion barrier. Relative to other design concepts, coated containers have thinner walls that provide less shielding for γ-radiation, raising the possibility of radiolysis-induced corrosion on the exterior of the container. Also, any trapped aerated water inside the container or in the bentonite clay may condense on the container surface, be exposed to γ-radiation and influence the container corrosion. Understanding the effect of radiation on container corrosion is one key factor to consider in the assessment of the integrity and longevity of the UFC. Humid air radiolysis produces HNO{sub 3} as the dominant oxidizing species. A conservative hemispherical water droplet model was developed to predict the accumulation of HNO{sub 3} and its impact on corrosion. Radiolysis of groundwater in the presence of Cl{sup -} mainly produces H{sub 2}O{sub 2} and HOCl. The concentration of these species is <10 mM and their sum remains almost constant. Radiation exposure tests showed radiation-induced corrosion on carbon steel and Cu. A systematic study of the Cu corrosion and the corrosion rate as a function of [H{sub 2}O{sub 2}], [HOCl], pH and γ-radiation is on-going. (authors)

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DOE OSTI2018

Experience by BAM in the Transport Package Safety Assessment of German Waste Containers - 18469

Transport and storage containers for low and mid-level radioactive waste are getting more and more of relevance due to the nuclear phase out decision in 2011. For higher activities Type B(U) approved waste containers will be needed for the shut down and dismantling phases of NPPs. It is expected that large quantities are required in the near future. German waste containers are generally approved for transport and interim storage and are also intended for the final disposal in the Konrad repository. BAM is involved in the authority licensing of transport package designs in Germany. In this context, BAM is responsible for the assessment of safe containment, mechanical, thermal and quality management issues. BAM also operates test facilities and performs drop as well as thermal tests during package licensing procedures. This paper summarizes our experience in the transport package design assessment of Type B(U) waste containers. A general overview of the approval process, the requirements and approaches BAM applies are described. Some examples are used to illustrate different aspects and technical issues we are addressing during the package assessment. In the first part of the paper the specific design aspects of German waste containers are described. Here, a general overview is given e.g. about closure systems, impact limiter designs, and the handling concepts. Furthermore, the wide range of radioactive content and their physical behavior including the impact on the packages assessment are described. The second part is focused on questions about the licensing and assessment process of German waste containers from the BAM authority point of view. The general approaches for the strategy of demonstration are outlined on the basis of the test conditions according to IAEA Regulations SSR-6. Furthermore, particular issues of the mechanical and thermal assessment with respect to the specific test conditions are discussed. For accident conditions of transport, aspects to be mentioned are e.g. the assessment of the lid bolts, the axial gap applied between content and lid, and the thermal behavior of the wood filled impact limiter after the fire test. However, issues of the assessment for routine and normal conditions of transport are addressed in this paper, too. The intention of this paper is introduce recent approval procedure experience in Germany, describing technical evaluation issues and so reduce rounds of questions during applications. (authors)

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DOE OSTI2023

On the Durability of Tin‐Containing Perovskite Solar Cells

Abstract Tin (Sn)‐containing perovskite solar cells (PSCs) have gained significant attention in the field of perovskite optoelectronics due to lower toxicity than their lead‐based counterparts and their potential for tandem applications. However, the lack of stability is a major concern that hampers their development. To achieve the long‐term stability of Sn‐containing PSCs, it is crucial to have a clear and comprehensive understanding of the degradation mechanisms of Sn‐containing perovskites and develop mitigation strategies. This review provides a compendious overview of degradation pathways observed in Sn‐containing perovskites, attributing to intrinsic factors related to the materials themselves and environmental factors such as light, heat, moisture, oxygen, and their combined effects. The impact of interface and electrode materials on the stability of Sn‐containing PSCs is also discussed. Additionally, various strategies to mitigate the instability issue of Sn‐containing PSCs are summarized. Lastly, the challenges and prospects for achieving durable Sn‐containing PSCs are presented.

14 SOLAR ENERGY↗

DOE OSTI2025

An approach for spent nuclear fuel containment integrity verification using gas tagging

Verification of containment integrity is required for spent nuclear fuel (SNF) managed by the commercial nuclear industry and U.S. Department of Energy (DOE), especially after extended storage. Certain SNF storage systems, such as the DOE road-ready dry storage system, hold several packaged containments within a welded over-canister. These packaged containments are called Department of Energy Standard Canisters (DOESCs). DOESC leakage identification is challenging because their containment boundary cannot be accessed for testing and their contents (i.e., SNF and fill gas) are often similar. There are concerns that this could result in costly characterization and repackaging operations of DOE road-ready dry storage systems if compromised DOESCs are suspected. Here, to address these concerns, this paper presents an approach for applying a gas tagging process using xenon to uniquely identify compromised inaccessible containments following extended storage. The containments considered for this application are seven DOESCs, each packaged within a single over-canister. Two different SNF loading configurations from the Advanced Test Reactor and Fort Saint Vrain nuclear power plant are considered. These configurations are used to represent research reactor aluminum-clad spent nuclear fuel (ASNF) and TRi-structural ISOtropic (TRISO) SNF types. Results for this application show that for ASNF and TRISO type fuels for which the selected fuels are representative, the volume of taggant required at loading is determined primarily by the lower detection limit and leak rate of taggant from a compromised DOESC, rather than the amount of fission-generated xenon in the loaded fuel. While the application presented is suited for larger leaks, smaller leaks could be detected by modifying certain design parameters. This gas tagging approach can also be applied to other DOE containments and advanced reactor SNF storage systems.

07 - ISOTOPES AND RADIATION SOURCES↗

DOE OSTI2014

SU-E-T-273: Radiation Shielding for a Fixed Horizontal-Beam Linac in a Shipping Container and a Conventional Treatment Vault

Purpose: A fixed horizontal-beam linac, where the patient is treated in a seated position, could lower the overall costs of the treatment unit and room shielding substantially. This design also allows the treatment room and control area to be contained within a reduced space, such as a shipping container. The main application is the introduction of low-cost, high-quality radiation therapy to low- and middle-income regions. Here we consider shielding for upright treatments with a fixed-6MV-beam linac in a shipping container and a conventional treatment vault. Methods: Shielding calculations were done for two treatment room layouts using calculation methods in NCRP Report 151: (1) a shipping container (6m × 2.4m with the remaining space occupied by the console area), and (2) the treatment vault in NCRP 151 (7.8m by 5.4m by 3.4m). The shipping container has a fixed gantry that points in one direction at all times. For the treatment vault, various beam directions were evaluated. Results: The shipping container requires a primary barrier of 168cm concrete (4.5 TVL), surrounded by a secondary barrier of 3.6 TVL. The other walls require between 2.8–3.3 TVL. Multiple shielding calculations were done along the side wall. The results show that patient scatter increases in the forward direction and decreases dramatically in the backward direction. Leakage scatter also varies along the wall, depending largely on the distance between the gantry and the wall. For the treatment room, fixed-beam requires a slightly thicker primary barrier than the conventional linac (0.6 TVL), although this barrier is only needed in the center of one wall. The secondary barrier is different only by 0–0.2 TVL. Conclusion: This work shows that (1) the shipping container option is achievable, using indigenous materials for shielding and (2) upright treatments can be performed in a conventional treatment room with minimal additional shielding. Varian Medical Systems.

07 ISOTOPE AND RADIATION SOURCES↗

DOE OSTI2013

Characterization Data Package for Containerized Sludge Samples Collected from Engineered Container SCS-CON-210

This data package contains the K Basin sludge characterization results obtained by Pacific Northwest National Laboratory during processing and analysis of four sludge core samples collected from Engineered Container SCS-CON-210 in 2010 as requested by CH2M Hill Plateau Remediation Company. Sample processing requirements, analytes of interest, detection limits, and quality control sample requirements are defined in the KBC-33786, Rev. 2. The core processing scope included reconstitution of a sludge core sample distributed among four to six 4-L polypropylene bottles into a single container. The reconstituted core sample was then mixed and subsampled to support a variety of characterization activities. Additional core sludge subsamples were combined to prepare a container composite. The container composite was fractionated by wet sieving through a 2,000 micron mesh and a 500-micron mesh sieve. Each sieve fraction was sampled to support a suite of analyses. The core composite analysis scope included density determination, radioisotope analysis, and metals analysis, including the Waste Isolation Pilot Plant Hazardous Waste Facility Permit metals (with the exception of mercury). The container composite analysis included most of the core composite analysis scope plus particle size distribution, particle density, rheology, and crystalline phase identification. A summary of the received samples, core sample reconstitution and subsampling activities, container composite preparation and subsampling activities, physical properties, and analytical results are presented. Supporting data and documentation are provided in the appendices. There were no cases of sample or data loss and all of the available samples and data are reported as required by the Quality Assurance Project Plan/Sampling and Analysis Plan.

core sampling↗

DOE OSTI2018

An Assessment of MELCOR 2.1: Containment Thermal-Hydraulic Tests in the Heissdampfreaktor (HDR) Facility

MELCOR is being developed at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission. MELCOR is a fully integrated code (encompassing the reactor coolant system and the containment building) that models the progression of postulated accidents in light water reactor power plants. It provides a capability for independently auditing analyses submitted by reactor manufacturers and utilities. In order to assess the adequacy of containment thermal - hydraulic modeling incorporated in the MELCOR code, a key containment test facility was analyzed. This report documents MELCOR code calculations for simulating steam - water blowdown tests performed in the Heissdampfreaktor ( HDR) de-commissioned containment facility located near Frankfurt , Germany . These tests are a series of blowdown experiments in a large scaled test facility ; including some tests with the addition of hydrogen release which are intended to simulate a variety of postulated break s inside large containment buildings. The key objectives of this MELCOR assessment are to study: (1) the expansion and transport of high energy steam - water releases, (2) heat and mass transfer to structural passive heat sinks, and (3) containment gas mixing and stratification. Moreover, MELCOR results are compared to the CONTAIN code for the same tests.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2021

Alternative Liquid DSA Containers for Sludge Slurry Storage in the SRNL Shielded Cells

A pending revision of the Documented Safety Analysis (DSA) for the Savannah River National Laboratory (SRNL) defines “DSA containers'’ for liquid sludge slurries as containers not susceptible to a flashing spray release in a fire and requires that a list of DSA containers be maintained. High Level Waste (HLW) samples are typically received from the Savannah River Site (SRS) tank farm after retrieval from the storage tanks. Glass bottles are preferred for long-term storage of radioactive liquid sludge slurry samples but are not on the DSA container list. An evaluation has been conducted to identify vessel lids for the glass bottles that would vent prior to pressurization and liquid superheating in a fire event. Low melting Field's Metal eutectic alloy with a melting point of 62°C has been identified as a preferred material of construction for vessel lids for this application because the material is resistant to puncture and is expected to have good chemical compatibility and radiation stability in the shielded cells environment. A test vessel was designed to evaluate the alloy and confirm that it would melt before the water in the vessel reaches the boiling point. The tests confirmed: 1) that the reported melting point of the alloy was accurate, 2) that a top fashioned from this material will melt creating an open atmosphere in the vessel headspace prior to sample boiling under moderate and fast heating rates, and 3) that nonuniform heating of the bottle from the bottom also results in melting of the alloy prior to boiling. Based on the results, neither an engulfing fire with standing or toppled bottles or a fire producing localized heat at the vessel bottom would result in flashing spray conditions. It is recommended that glass vessels of various volumes ranging from 125 mL to 1 gallon with modified tops containing the specified Field's Metal eutectic alloy which are appropriately designed to avoid flashing spray material releases be added to the DSA container list. These vessels will be utilized in the Shielded Cells for the storage of radioactive liquid samples of sludge slurries. Use of these vessels will also be advantageous for other liquid samples containing accountable amounts of nuclear material throughout other SRNL laboratories. Similarly constructed paraffin vessel lids could also potentially be added to the approved list, though additional testing would be required.

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DOE OSTI2023

Simulations of Criticality Control Overpack Container Compaction at the Waste Isolation Pilot Plant

Criticality Control Overpack (CCO) containers are being considered for the disposal of defense-related nuclear waste at the Waste Isolation Pilot Plant (WIPP). At WIPP, these containers would be placed in underground disposal rooms, which will naturally close and compact the containers closer to one another over several centuries. This report details simulations to predict the final container configuration as an input to nuclear criticality assessments. Each container was discretely modeled, including the plywood and stainless steel pipe inside the 55-gallon drum, in order to capture its complex mechanical behavior. Although these high-fidelity simulations were computationally intensive, several different material models were considered in an attempt to reasonably bound the horizontal and vertical compaction percentages. When exceptionally strong materials were used for the containers, the horizontal and vertical closure respectively stabilized at 43:9 % and 93:7 %. At the other extreme, when the containers completely degraded and the clay seams between the salt layers were glued, the horizontal and vertical closure reached respective final values of 48:6 % and 100 %.

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