Results for “cask” — Page 13

DOE OSTI2022

Mobile heat pipe cooled fast reactor system

A mobile heat pipe cooled fast nuclear reactor may be configured for transportation to remote locations and may be able to provide 0.5 to 2 megawatts of power. The mobile heat pipe cooled fast reactor may contain a plurality of heat pipes that are proximate to a plurality of fuel pins inside the reactor. The plurality of heat pipes may extend out of the reactor. The reactor may be configured to be placed in a standard shipping container, and may further be configured to be contained within a cask and attached to a skid for easier transportation.

McClure, Patrick Ray↗

DOE OSTI2022

The Hanford Lead Canister Collaboration

The Hanford Site, Management of the Cesium and Strontium Capsules (MCSC) Project is preparing to place cesium and strontium material into dry storage using stainless steel canisters that are housed inside vertical concrete casks. Part of the site’s aging management program is to use a spare canister system as a leading indicator canister to provide advance warning for signs of chloride induced stress corrosion cracking (CISCC) or any other kind of canister degradation that might take place over the system’s 300-year design life. The Hanford Lead Canister (HLC) will use electric heaters to simulate the decay heat of cesium and strontium to ensure the HLC simulates the environment and conditions the capsule-loaded systems will experience. In addition to being part of the site’s aging management program, the HLC is being made available for research and development that advances the industry regarding canister mitigation and repair (M&R). The HLC system closely resembles current spent nuclear fuel (SNF) dry storage systems, making the HLC useful for demonstrating technologies or performing research that is relevant to SNF canisters . Several organizations are collaborating to prepare the HLC, along with technical strategy, to perform the needed research and development activities, to include the Pacific Northwest National Laboratory (PNNL), Central Plateau Cleanup Company (CPCCo), and the Electric Power Research Institute (EPRI). In one example of this collaborative research, EPRI is planning to use the HLC to demonstrate in-situ M&R technologies under controlled conditions before the HLC is deployed at Hanford. There is a period of several years when the HLC will be available for canister M&R activity before it is deployed. This paper describes the development of the HLC, its role in the Hanford site’s MCSC Project aging management program, and some of the research activities that are planned for the HLC.

Klymyshyn, Nicholas A.↗

DOE OSTI2023

Modeling Non-UO2 Fuel With UNF-ST&DARDS

The U.S. Department of Energy’s Used Nuclear Fuel-Storage, Transportation & Disposal Analysis Resource and Data System (UNF-ST&DARDS) provides an easy-to-use interface to analyze irradiated UO2 fuel by allowing all analysis to be performed within the software and being able to store and use dozens of fuel assembly, canister, and cask designs [1]. However, performing these same analyses with non-UO2 fuel, such as UN or U3Si2, requires more user intervention in the process. This work uses UN, UN-ZrO2, and U3Si2 fuel to demonstrate how to perform criticality analyses in the current versions of UNF-ST&DARDS and how a non-UO2 fuel will compare to UO2. This work is part of a larger effort that also includes shielding and thermal analyses, but they will not be discussed.

Ivanusa, Pavlo↗

DOE OSTI2021

Experience with Inner Reflector Plug exchange in SNS

The Inner Reflector Plug (IRP) is a central component of the Spallation Neutron Source target monolith, which houses the mercury target and four liter-sized neutron moderator units. It is exposed to high-level radiation fields during routine operation and builds up significant activity. The IRP needs to be replaced due to moderator neutron poison and decoupler burn-out, which is used for shaping neutron pulses. The first IRP exchange took place in March 2018. The old IRP was extracted from the target monolith, providing space for the new one. It was split into three segments, each of which was handled separately. The lowest section of the IRP is the largest segment in size and in activity and is temporarily stored on-site for cool down before conduction post irradiation examination. In support of planning the replacement activities, a wide range of activation and transport analyses were performed. This included calculating isotope inventories and the radiation fields for each segment as it is extracted in storage casks, and the radiation field from the empty IRP pit in the target monolith. While the replacement was taking place, measurements were performed and later on compared to the calculations. During these studies, it was discovered that a significant contributor to the radiation field from the lower IRP segment is from photo-neutrons. Photo-nuclear physics was added to the analyses and calculated results compared well with measured dose rates.

Popova, Irina I.↗

DOE OSTI2023

Coupling SCALE with DAKOTA for Axial Burnup Profiles Assessment in Burnup Credit

This paper presents a computational study that demonstrates the application of the SCALE code system in conjunction with the Design Analysis Kit for Optimization and Terascale Applications (DAKOTA) for the analysis of key factors influencing the evaluation of burnup credit (BUC) in pressurized water reactors (PWRs). The primary objective of this analysis is to characterize the model by utilizing parameterization, uncertainty quantification, and optimization studies. Using this approach, we can comprehensively assess the system and conduct informed predictive studies. This study highlights the effectiveness of the SCALE code system integrated within the DAKOTA framework in terms of efficiency and capability. With the coupling of the burnup code ORIGAMI with the CSAS or TSUNAMI-3D sequence embedded in a DAKOTA analysis, we can characterize the factors that influence the k eff of PWR 17x17 spent nuclear fuel (SNF) in the GBC-32 computational benchmark cask for the assessment of BUC in criticality safety analysis. The coupling methodology used in this study is not exclusive to BUC analysis. However, the choice to apply this methodology to the BUC problem is particularly significant because of the diverse range of aspects it encompasses in nuclear criticality safety analyses. This problem presents a unique opportunity to explore and address multiple facets of such analyses related to BUC and illustrates the capability of the SCALE code system with DAKOTA. This analysis makes use of historical reference data for the axial burnup profile, where the entire space within the bounds is considered. Both SCALE and DAKOTA are currently integrated in the Nuclear Energy Advanced Modeling Simulation (NEAMS) Workbench code system, which has a user-friendly graphical interface that simplifies the setup of simulations and configuration of input parameters as well as the visualization of simulation results.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2020

OPTIMUS{sup TM} Packaging for Intermediate Level Wastes and Spent Fuel - 20129

NAC's OPTIMUS{sup TM} packaging is a family of Type B(U)F transportation packages designed for maximum flexibility and cost-efficiency to support shipment of a wide range of challenging wastes and materials. They are small, modular packaging designs meeting DOT weight limits for road transport. OPTIMUS{sup TM} contents include low level waste, mixed low-level waste, irradiated fuel waste and intermediate level waste. Contents can also include aerosol cans with compressed gas and liquefied gas propellants and standard DOT 3E lecture bottles for shipment to treatment facilities. OPTIMUS{sup TM} packaging provides a cost-effective solution for challenging and unique waste contents combined with design simplicity and operational flexibility meeting the requirements of both SSR-6 and 10 CFR 71 for Type B(U)F packaging. The modular design of the packaging components combined with the flexibility to add modified internals for enhanced packaging performance makes OPTIMUS{sup TM} the preferred solution for many radioactive material transportation needs. The OPTIMUS{sup TM} product line currently includes two packaging designs; OPTIMUS{sup TM}-L and OPTIMUS{sup TM}-H. OPTIMUS{sup TM}-L is a lightweight transportation packaging intended for low-activity wastes, whereas OPTIMUS{sup TM}-H is a larger, heavier package intended for high-activities wastes. Due to its small size and weight, six (6) to ten (10) OPTIMUS{sup TM}-L packages can be accommodated on a single legal-weight truck shipment, depending on the weight of the contents. For typical payloads, two (2) OPTIMUS{sup TM}-H packages can be accommodated on a single legal-weight truck shipment. The OPTIMUS{sup TM} packaging share the same cask containment vessel (CCV) in the OPTIMUS{sup TM}-H and OPTIMUS{sup TM}-L. The large cavity size of the CCV, which can accommodate a 416-liter (110-gallon) drum, combined with the small size, low weight and modularity of the OPTIMUS{sup TM} packaging provides unmatched flexibility. In this paper, NAC provides a technical overview of the OPTIMUS{sup TM} packaging and identifies the design features and technology advancements making the OPTIMUS{sup TM} a readily adaptable and flexible solution for packaging processing facilities, reactor and decommissioning wastes. Optimized shielding performance through the addition of internal shielding components and alternative shipping configurations are discussed. In addition, an overview of the first major deployment of the OPTIMUS{sup TM} packaging for the Whiteshell Laboratories Closure Project (WLCP) is discussed. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Optimization of the Post-Operational Phase on Two Belgian Multi-Unit Nuclear Power Plants: the Case of the Non-Fissile Irradiated Core Items - 20156

The current legal framework in Belgium foresees the progressive phase out of nuclear power between October 2022 (Doel 3) and December 2025 (Doel 2). Upon its definitive shutdown, each unit of the Tihange and Doel sites will enter a Post-Operational Phase (POP) and be prepared for its Decontamination and Decommissioning (D and D). Prior to obtaining the D and D license, the Operator Electrabel is legally required to remove any non-fissile irradiated core items stored in the deactivation pools. The non-fissile irradiated core items consist essentially of control rods, poison rods and source thimbles as well as thimble plugs and foreign materials irradiated during operation: - Their significant content in highly radiant radionuclides (up to 6 TBq of Co-60 per kg of irradiated material) renders all existing operational waste management processes inadequate due to insufficient biological shielding; - Their high concentrations in long-lived radionuclides call for their disposal in a geological repository for which no final design nor waste acceptance criteria are expected prior to 2050. Uncertainties in the Belgian energy supply and security, however, require the Operator to be prepared for a partial nuclear phase out, where one or more units would benefit from lifetime extension while the remaining units would undergo decommissioning. The present paper aims at presenting how Electrabel, in partnership with Tractebel, addressed this challenge by maximizing the use of synergies within the respective sites as well as between both sites themselves, all the while accounting for site specificities. The most recent results and state of progress of the project will be detailed and the first lessons learned will be shared. The project has been split in multiple tasks and phased as follows: - An inventory phase aimed at mapping the contents, origin, composition and history of the non-fissile irradiated core items; - A pre-characterization phase based on neutron activation models; - A waste sorting phase aimed at separating waste forms for which an evacuation route exists from those for which such route does not exist; - A feasibility phase aimed at exploring all possible scenarios for the management of non-fissile irradiated core items and identifying the optimal feasible solution for each site; - A preparation phase (currently ongoing), developing further the optimal solution and ensuring that back-up solutions are available for any foreseeable change of context (licensing issue, modification in the nuclear phase-out program, etc.) and initiating early contacts with potential subcontractors for segmentation works and cask manufacturers, as well as the Belgian regulatory body and waste management agency. This phase also foresees the investigation of destructive and non-destructive radiological measurements to support the detailed characterization of the waste forms; - A realization phase (future work). (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

An Overview of R and D on Retrievability and Retrieval Technology in Germany - 20462

Retrievability is a term that is included in most radioactive waste management programs around the world. Although national definitions vary, the overall understanding of retrievability concerns the ability to recover waste packages from the repository mine after their emplacement. Different countries may implement retrievability in very different ways, ranging from a built-in reversibility into the emplacement process to the stipulation that retrievability may in no way impede passive safety in the post-closure phase. Germany takes a middle course such that retrievability in a HLW repository (repository for high level radioactive waste) may have no significant detrimental effect on passive safety. In Germany, the current siting process considers rock salt, clay rock, and crystalline rock as potential host rocks. Therefore, Research and Development (R and D) has been investigating repository concepts and retrievability in all host rocks. After introduction of retrievability in 2010, existing repository concepts were modified to facilitate retrieval. The changes made comprised, for example, equipping boreholes with steel liners, developing new technologies, and modifying existing emplacement devices. Apart from retrieval of HLW, retrieval of other wastes from underground repositories that were designed and operated without retrievability in mind, poses major technical and scientific challenges. Currently, studies are under way to investigate the feasibility and costs of partial retrieval of waste from an underground repository for hazardous and highly toxic waste in France, Stocamine. With regard to retrieval, the rock-mechanical conditions are deteriorating rapidly, so time is of the essence. In the Asse II mine in Germany, about 126,500 waste casks with low and intermediate level waste await retrieval. The repository mine suffers from difficult rock-mechanical conditions and inflow of brine, locally into emplacement areas. The Federal Company for Radioactive Waste Disposal (BGE) is legally required to retrieve all wastes from the Asse. Due to the specific challenges, R and D is needed to develop technical solutions for the safe retrieval from each of the emplacement chambers. R and D on retrievability and retrieval technology does not only address challenges in radioactive or toxic waste retrieval but may also help to germinate innovation to better master complex underground situations in general, e.g. in deep mining, tunneling, or repository construction. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2024

Fabrication and Testing of DOE Standard Canister Closure Leak Test Assembly

DOE manages over 300 types of spent nuclear fuel (SNF), many of which are located at the Idaho National Laboratory (INL) site. Managing this large variety of SNF for storage, transportation, and disposal poses a challenge to DOE. The Idaho Cleanup Project and INL are collaborating on the DOE SNF Road-Ready Demonstration (“Road-Ready Demonstration”), which will develop and demonstrate the designs, technology, processes, and regulatory framework for packaging DOE-managed SNF for “road-ready dry storage.” Road-ready dry storage is an SNF management concept in which SNF is packaged into dry, sealed canisters that are then placed in onsite storage in anticipation of later transport and disposition. The forward-looking goal of the Road-Ready Demonstration is to establish the foundation for a large-scale road-ready dry storage program at the INL site. In support of establishing a large-scale road-ready dry storage program at the INL site, the Road-Ready Demonstration will first package Fort St. Vrain SNF currently in dry storage at INL into several DOE Standard Canisters (DOESCs). These DOESCs will in turn be loaded into another containment similar to commercial multi-purpose canisters. This multi-purpose canister will then be compatible with a transportation or storage system, such as a storage cask for interim storage or transportation package for offsite transport. These DOESCs will remain sealed over the course of their storage, transportation, and applicable disposal functions. The closure process for the DOESC will include fuel and basket loading, welding, inspection, leak testing, and, if needed, repair. As a follow-up to previous discussions on the design of the DOE Closure Leak Test Assembly (LTA), this report describes recent fabrication and testing efforts performed at INL. DOESCs are sealed by two sequential gas tungsten arc welds, both of which are performed by remotely operated and semiautomatic welding systems. The first weld is a circumferential pipe weld that completes the assembly of the canister body and lid assembly. The second and final closure weld attaches the vent plug to the vent socket via a butt joint. After the second weld is performed, the welds are helium leak tested using an evacuated envelope technique. The LTA was designed for both remote and manual operation. This report describes the fabrication and performance testing associated with the evacuated envelope technique. INL staff designed, fabricated, and tested the LTA at INL facilities. This testing included establishing technique and system sensitivities in accordance with ASME and American National Standards Institute N14.5 requirements. Forthcoming work will cover such areas as design optimization, process and personnel qualification, and implementation in Road-Ready Demonstration operations.

42 ENGINEERING↗

DOE OSTI2024

Drop Analysis of Department of Energy Standard Canister with Fort Saint Vrain SNF

DOE manages over 300 types of SNF, most of which are located at the INL site. The Idaho Cleanup Project and INL are collaborating on the Road-Ready Capability Demonstration Project, which will develop and demonstrate the designs, technology, processes, and regulatory framework for packaging DOE-managed SNF at the INL site for ?road-ready dry storage?. ?Road-ready dry storage? is a SNF management concept where SNF is packaged into dry and sealed canisters, which are then placed in on-site storage in anticipation of later transportation. The forward-looking goal of the Demonstration is establishing the foundation for a large-scale road-ready dry storage program at the INL site. The Demonstration will first package Fort Saint Vrain SNF currently stored at INL into several DOE Standard Canisters. These Standard Canisters will then be loaded into another commercial transportation or storage containment system (e.g., storage cask or transportation package). The Standard Canister is a class of standardized canisters designed for containing the large variety of DOE-managed SNF during interim storage, transportation and/or disposal at a geological repository. One critical aspect of road-ready dry storage is the ability to license the DOE Standard Canisters and its associated transportation package to 10 CFR 71. Depending on the SNF and transportation strategy, the Standard Canisters may have to maintain structural integrity under normal conditions of transport and hypothetical accident scenarios (i.e., drop events). The Standard Canisters have been tested and analyzed under various SNF loading configurations and accident drop events in support of the Idaho Spent Fuel Facility and other DOE programs. However, no analysis has been completed to support the recent Demonstration. This analysis will consider the Ø0.5 m × 5.1 m Standard Canister under drop scenario(s) considered in previous INL tests and analyses, including the 9 m drop at 80 degree off vertical. However, this analysis will consider the more recent Fort Saint Vrain loading configurations proposed for the Demonstration. This analysis will performed using strain-based acceptance criteria established by the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section III, Division 3. It will be compared to previous analyses and form the foundation of further formal calculations that will be used to support licensing efforts of the road-ready dry storage system at INL.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Neutron Absorber Plate Characterization Plan for Criticality Experiments Design

After being used in nuclear installations, depleted fuel can still be highly reactive and must be handled securely to prevent any radiological or criticality concerns. In particular, spent fuel from use in nuclear power reactors must be stored and transported in specifically designed containers using neutron absorber materials to prevent criticality. Various neutron absorber material types exist and are manufactured by various entities, as thoroughly described in the Handbook of Neutron Absorber Materials for Spent Nuclear Fuel Storage and Transportation Applications written by EPRI. Presently, one of the most modern and most widely used types of neutron absorber material contains particles of boron carbide, or B 4 C, embedded in aluminum matrix: Boralcan, manufactured by Rio Tinto. It is very important for the community to know as much as possible about such neutron absorber materials. Therefore, in the recent years, a US Department of Energy National Nuclear Security Administration–Nuclear Criticality Safety Program funded project initiated design of an experiment that places Boralcan neutron-absorbing plates in an established critical assembly using low-enriched uranium fuel at the Sandia Pulsed Reactor Facility/Critical Experiments (SPRF/CX) apparatus at Sandia National Laboratories. The goal of the experiment is to produce high-quality benchmark data to submit to the International Criticality Safety Benchmark Evaluation Project (ICSBEP), for use in validating calculational tools and nuclear data by criticality safety analysts. The project, named IER-554, is currently in its final design stage, following a successful preliminary design. In the work documented in the design study, ten critical configurations using Boralcan neutron absorber plates were designed, and the experiment was proven to be feasible, with a predicted low k eff uncertainty around 100 pcm. An overview of the modeled cutout of the critical assembly with a Boralcan plate is shown in Figure 1, representing one of the configurations planned for the critical experiments. Before the plates are inserted in the critical assembly, it is necessary to know more about their composition and uniformity. This summary focuses on the plate characterization plans. Each plate will undergo (1) neutron transmission measurements at different locations to determine the 10 B areal density and (2) an in-depth x-ray computed tomography (XCT) examination to obtain the exact Sizes and distribution of the B4C powder particles inside the plates. In parallel, plate modeling studies are performed with a goal to determine the validity of the currently used approximation of modeling the neutron absorber plates as a homogeneous mixture of Aluminum 1100 alloy and B4C— instead of explicitly modeling the B4C particles. By using the experimental 10 B areal density measurements, and the exact size and location of the B4C particles obtained by XCT, a plate model can theoretically be built that reproduces the plate with extremely high fidelity. The results of this modeling study could increase the confidence of the criticality safety community in its modeling methods when using this type of neutron absorber material, and the industry could use these validations to change the boron loading credit limits from the U.S. Nuclear Regulatory Commission standard review plan for dry cask storage of spent nuclear fuel. The modeling calculations are performed with SCALE 6.3.0 using the KENO V.a sequence for criticality calculations with the ENDF/B-VIII.0 continuous-energy cross section library.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Development of Shielding Benchmarks Using the Godiva IV Assembly

The Nuclear Criticality Safety Program (NCSP) is developing a shielding benchmark using the Godiva IV assembly as a source. Its present status is reviewed herein. Even eight decades into the nuclear era, substantial work remains to develop a database of shielding benchmarks to support future nuclear development. A nuclear simulation is only as good as its supporting data, inputs, and validation basis. Uncertainty in these areas is addressed using conservatism, which adds margin and, occasionally, cost. In many shielding situations, high accuracy is not necessary because additional material is not particularly expensive. After all, 1–2 cm of lead often reduces the gamma ray dose substantially. However, in certain areas, conservatism can add unnecessary cost. These areas include mobile shielding applications such as casks, ships, microreactors, and spacecraft, where weight and, thus, margin is expensive. Although these characteristics are side benefits for NCSP shielding benchmark development, the main driver is enabling more reliable placement of criticality accident alarm systems (CAASs) in nuclear material facilities, such as those dedicated to the production of advanced reactor fuels. CAAS placement relies on more than accurate data and code validation. It also relies on sufficiently accurate materials specifications, geometry specifications, and a well-defined, alarm-producing baseline accident. All these things require tacit knowledge and understanding of the problem being evaluated. Benchmarks can help ensure this understanding.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2023

Fabrication and Testing of DOE Standard Canister Closure Leak Test Assembly – 24139

DOE manages over 300 types of spent nuclear fuel (SNF), many of which are located at the Idaho National Laboratory (INL) site. Managing this large variety of SNF for storage, transportation, and disposal poses a challenge to DOE. The Idaho Cleanup Project and INL are collaborating on the DOE SNF Road-Ready Demonstration (“Road-Ready Demonstration”), which will develop and demonstrate the designs, technology, processes, and regulatory framework for packaging DOE-managed SNF for “road-ready dry storage.” Road-ready dry storage is an SNF management concept in which SNF is packaged into dry, sealed canisters that are then placed in onsite storage in anticipation of later transport and disposition. The forward-looking goal of the Road-Ready Demonstration is to establish the foundation for a large-scale road-ready dry storage program at the INL site. In support of establishing a large-scale road-ready dry storage program at the INL site, the Road-Ready Demonstration will first package Fort St. Vrain SNF currently in dry storage at INL into several DOE Standard Canisters (DOESCs). These DOESCs will in turn be loaded into another containment similar to commercial multi-purpose canisters. This multi-purpose canister will then be compatible with a transportation or storage system, such as a storage cask for interim storage or transportation package for offsite transport. These DOESCs will remain sealed over the course of their storage, transportation, and applicable disposal functions. The closure process for the DOESC will include fuel and basket loading, welding, inspection, leak testing, and, if needed, repair. As a follow-up to previous discussions on the design of the DOE Closure Leak Test Assembly (LTA), this report describes recent fabrication and testing efforts performed at INL. DOESCs are sealed by two sequential gas tungsten arc welds, both of which are performed by remotely operated and semiautomatic welding systems. The first weld is a circumferential pipe weld that completes the assembly of the canister body and lid assembly. The second and final closure weld attaches the vent plug to the vent socket via a butt joint. After the second weld is performed, the welds are helium leak tested using an evacuated envelope technique. The LTA was designed for both remote and manual operation. This report describes the fabrication and performance testing associated with the evacuated envelope technique. INL staff designed, fabricated, and tested the LTA at INL facilities. This testing included establishing technique and system sensitivities in accordance with ASME and American National Standards Institute N14.5 requirements. Forthcoming work will cover such areas as design optimization, process and personnel qualification, and implementation in Road-Ready Demonstration operations.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2025

Reference Fuel Development for Non-Aluminum Spent Nuclear Fuel Management

The Savannah River Site (SRS) L Area Facility provides for the safe receipt, storage, handling, and shipping of spent nuclear fuel (SNF) and has received more than 47,000 SNF assemblies since 1964. In order to consolidate receipt and storage analysis, L Area criticality safety has historically used several aluminum-clad “reference fuels” to establish bounding storage, handling, and cask loading limits and then applied a reactivity comparison approach to demonstrate that candidate fuels may be processed under the reference fuel limits. Currently, only aluminum-clad SNF from off-site research reactors are transferred on-site. The Accelerated Basin De-Inventory (ABD) program will begin the removal of bundled Non-Aluminum Spent Nuclear Fuel (NASNF) from the L Area disassembly basin for dissolution and disposition, which has driven the need for a non-aluminum reference fuel. This paper discusses the process and results of creating a new fictional homogenous NASNF highly enriched uranium reference fuel, “MITZ,” to be used in nuclear criticality safety evaluations for upcoming SNF disposition operations. Data demonstrating the relationship between neutron multiplication behavior and fuel spacing is generated for the new reference fuel MITZ as well as existing reference fuels representing several types of SNF assemblies.

Reference Fuel↗

DOE OSTI2025

Reference Fuel Development for Non-Aluminum Spent Nuclear Fuel Management

The Savannah River Site (SRS) L Area Facility provides for the safe receipt, storage, handling, and shipping of spent nuclear fuel (SNF) and has received more than 47,000 SNF assemblies since 1964. In order to consolidate receipt and storage analysis, L Area criticality safety has historically used several aluminum-clad “reference fuels” to establish bounding storage, handling, and cask loading limits and then applied a reactivity comparison approach to demonstrate that candidate fuels may be processed under the reference fuel limits. Currently, only aluminum-clad SNF from off-site research reactors are transferred on-site. The Accelerated Basin De-Inventory (ABD) program will begin the removal of bundled Non-Aluminum Spent Nuclear Fuel (NASNF) from the L Area disassembly basin for dissolution and disposition, which has driven the need for a non-aluminum reference fuel. This paper discusses the process and results of creating a new fictional homogenous NASNF highly enriched uranium reference fuel, “MITZ,” to be used in nuclear criticality safety evaluations for upcoming SNF disposition operations. Data demonstrating the relationship between neutron multiplication behavior and fuel spacing is generated for the new reference fuel MITZ as well as existing reference fuels representing several types of SNF assemblies.

Reference Fuel↗

DOE OSTI2024

New systems in MOOSE

The Multiphysics Object-Oriented Simulation Environment (MOOSE) serves as a common library of classes between applications developed for advanced reactor analysis, fusion device engineering, spent fuel cask analysis, geochemistry studies, among other fields. These applications drive the development of the framework to meet their needs. Systems in MOOSE group capabilities that share a common purpose and generally common code. They can be leveraged by all downstream applications, providing extensive code re-use and shared maintenance. They facilitate the discovery by new users of the classes meeting at least partially their needs, and offer the same opportunities for customization as other systems. The addition of a new system to MOOSE opens new ways of solving or discretizing nonlinear problems, of performing distributed postprocessing, and a plethora of other needs. While new systems can be introduced in downstream applications rather than at the framework level, the framework team monitors common needs across the community and often triggers their addition. Documentation, training material, development needs can be centralized, limiting duplicated work across the community. The last three years have seen a large expansion in the capabilities of MOOSE. The supporting role of the framework in the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program has created numerous feature requests to support neutronics, thermal hydraulics, computational fluid dynamics and thermo-mechanics simulations in the Griffin, SAM, Pronghorn and Bison applications respectively. Similarly, laboratory-directed research and development (LDRD) projects in additive manufacturing, high-Reynolds flow simulations, structure optimization also necessitate an expansion of the framework capabilities. This summary reports on the new systems created in MOOSE, their design, their capabilities and some of the relevant interfaces.

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2026

A Mobile Hot Cell for Conditioning Disused Sealed Radioactive Sources for Storage or Transportation

An innovative Mobile Hot Cell (MHC) has been developed for conditioning Disused Sealed Radioactive Sources (DSRS) category 1 and 2 for storage or transportation. The MHC is designed to provide both Radiological and Biological containment with a maximum capacity of 1000 Ci Co60 or 5000Ci Cs137 source and can be transported via standard cargo containers. This project has been supported through the National Nuclear Safety Administration (NNSA) Offsite Source Recovery Program (OSRP). The project is intended for the international community rather than domestic although domestic use is a possibility. Many countries have significant stockpiles of these devices that are often stored in less-than-optimal circumstances. This necessitates that these devices be addressed expeditiously, and the sources secured. The MHC utilizes robotics, automation, and other non-traditional methods for disassembling, characterizing, and packaging these sources that have reached end of life or are otherwise not needed. These innovative approaches are necessary to facilitate an expedited timeline to efficiently and safely secure these sources in a non-proliferation effort. Conditioning efforts include disassembling the device such as a teletherapy head used for cancer treatment, or blood/research irradiators such that the radioactive sources may be removed safely. The sources are then characterized. Leak checks are performed, dimensions are verified, and serial numbers are confirmed. Upon completion, the sources are typically placed into a Standard Forms Capsule which is seal welded closed. It is leak tested and placed into a Long-Term Storage Shield (LTSS) which can either be secured for storage directly or loaded into an appropriate cask for transportation. Further innovations include multiple deployment scenarios that include a full deployment of MHC components, deployment of the MHC automation internal components to an existing hot cell, deployment of minimally required MHC components and incorporation of sand for shielding, and integration of the MHC for Silo Storage, or Bore Hole Storage efforts. The MHC has evolved from a very specific use case to a “Swiss Army Knife” type of a tool in that it can be readily adapted to a large variety of situations. Innovative approaches such as the use of robotics, Computer Numeric Control (CNC) machining centers, automated welding equipment, HDMI Cameras, and LED lighting are some of the developed technologies incorporated into the MHC design. Shielding is accomplished with a steel walled Base Box which is surrounded by four nesting doll shield shells which when combined limits the external dose rate to 5mr/hr when a 1000 Ci Co60 source is exposed inside.

99 - GENERAL AND MISCELLANEOUS↗

DOE OSTI2025

Analysis of Radiological Release From Fueled Irradiation Experiments During Manual Handling (Slides)

Irradiation experiments are manually handled at the Advanced Test Reactor (ATR) by qualified operators using long handled tools. Standard handling evolutions include insertion and removal from the reactor vessel, transfer to/from a storage location or cask in the ATR canal, and handling/reconfigurations at a canal working tray. Such routine handling has the potential to result in mechanical damage to the experiment boundary which is credited to retain fission products. Since damage can occur due to operator error, this is an anticipated occurrence. This work determines the radiological consequences to receptors inside the ATR facility, as well as public receptors. Given the wide variety of fuel types tested at the ATR, a generic approach to the analysis is taken. The radiological inventory is determined to bound a variety of fuel types (e.g., ceramic and molten fuel matrices) and fission powers on the U-235 enrichment spectrum. The source term analysis considers different release fractions to bound different fuel types and burnups. The postulated handling events occur underwater; thus, retention of the released isotopic content is considered within the canal water column. Retention of radionuclides in the water column is determined using a modified approach of United States (US) Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.183. Radiological dose to the facility receptor is determined using a compartment release model. Dose to the public receptor is determined using atmospheric dispersion models using site specific atmospheric conditions with the use of the Radiological Safety Analysis Computer (RSAC) program version 7.2.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗