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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 667 records · Page 37
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

Spent Nuclear Fuel Disposition at the Savannah River Site - 17153

For over forty years, the H-Canyon facility at the Savannah River Site (SRS) performed remotely operated radiochemical separations of irradiated targets and fuels to produce materials for national defense. Although the materials production mission has ended, the facility continues to play an important role in the stabilization and safe disposition of proliferable nuclear materials. (authors)

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DOE OSTI2015

High-level waste relocation requires use of new technology in aging nuclear facility - 15269

The West Valley Demonstration Project (WVDP) is preparing to relocate 278 canisters of vitrified high-level waste (HLW) from their current location inside the Main Plant Process Building (MPPB) to an outdoor pad for long-term, on-site storage. The waste was produced at the WVDP between 1996 and 2002 and was originally placed in interim storage inside the MPPB in anticipation of off-site shipment for disposal. Relocation of the canisters to a new on-site location is now necessary as part of WVDP Phase 1 Site Decommissioning work to enable pre-demolition activities inside the MPPB to proceed. The logistics of packaging and moving the HLW canisters within an aging facility to another on-site location has provided numerous technical challenges with subsequent innovations. Significant preparations have taken place to resolve structural inadequacies, including sub-surface soil subsidence issues, antiquated equipment, and facility reconfigurations to support up to 134 metric ton (148 ton) loads. To perform the transfer and storage of the HLW canisters, new equipment and fabrications were required including a robust storage system consisting of stainless steel canister overpacks within steel-reinforced concrete storage casks, three specialty cask transporters, a remotely operated welding system and construction of an outdoor engineered storage pad to store the 56 casks containing the relocated HLW canisters. Packaging and relocation of vitrified canisters is scheduled to begin in 2015. (authors)

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

Mobile Robots and Remote Characterisation Systems for Nuclear Decommissioning - 16028

This paper provides an updated summary of the research being conducted at the University of Manchester, UK, in support of the decommissioning plan for the Sellafield nuclear site in Cumbria, UK. The research is being conducted in collaboration with a number of companies including Sellafield Ltd, the National Nuclear Laboratory and Forth Engineering Ltd. The primary focus of the research is in support of the decommissioning of the legacy facilities on the Sellafield site, however the technology can be applied to other nuclear sites, both modern and legacy, as well as in other industries such as oil and gas, petrochemical and the marine industry. The progress on the development of novel characterisation platforms (mobile robots) and sensing systems is presented including the AVEXIS underwater inspection system, a robot spider for remote operations and algorithms for efficient compression of Lidar point cloud data. (authors)

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DOE OSTI2016

Cementitious Grouts for ILW Encapsulation - Hydration and Continuity of Supply within the UK - 16184

In the UK the preferred route to disposal for the majority of intermediate level nuclear waste (ILW) streams is encapsulation within a cementitious matrix, accounting for 83% of waste conditioned to date (NDA, 2013). The method is well-defined and produces a waste package that is both chemically and physically stable, providing a multi-layered barrier to the release of the waste species contained within. Due to the process of treatment, facilitated via remote operation, it is necessary for the grouting matrices to meet strict performance limits in terms of their flow, setting time and avoidance of the formation of bleed water or segregation. To this end, the limits on composition of the anhydrous materials (a mixture of ordinary Portland cement [OPC] and blast furnace slag [BFS]) are tight to ensure conformity. Whilst these limits on composition are well understood, their impact upon grout hydration, and hence performance, is poorly defined. This study investigates the effects of modifying the physical and chemical composition of both the OPC and the BFS components of the matrix in order to identify the sensitivity or otherwise of the system to ensure that a continuity of supply is maintained. A number of techniques, scanning electron microscopy - image analysis (SEM-IA), chemical shrinkage (measured via dilatometry), isothermal conduction calorimetry (ICC), thermogravimetric analysis (TGA) and X-ray diffraction (XRD), have been applied to assess the rate and degree of hydration (DoH) of a range of grouts prepared with OPC and BFS with a range of chemical and physical compositions. This has been coupled with subsequent engineering performance testing, with additional analysis to understand the impact of a fire scenario within a storage or disposal facility, to provide an indication towards the resilience of the waste packages to a potential catastrophic event. (authors)

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DOE OSTI2016

Decommissioning Experience at Rokkasho; Standing Type Manipulator: A1000S for Dismantling Tasks - 16434

Decommissioning is a technical process which includes various steps and processes from the clean-up of radioactive materials to the progressive demolition of the plant. One of these steps is the cutting of subassemblies, like tanks, pipes or walls in small pieces suitable for the sorting and disposal. Different solutions are offered for this cutting work and can be grouped as following: - Chip forming - sawing, drilling, milling, turning etc. - Shearing - punching, stamping, scissoring. - Abrading - Grinding, lapping, polishing; water-jet. - Heat - flame cutting, plasma cutting, laser cutting. - Electro-Chemical - Etching, Electrical discharge machining (EDM). Each method has its limitations in accuracy, cost, and effect on the material. In 2013, Waelischmiller Engineering won a contract for two remote handling systems dedicated to the dismantling cell at Rokkasho Nuclear Fuel Reprocessing Facility in Japan. For this project, Inconel and other metals needed to be cut. One of the main cutting methods was decided to use of an angle grinder handled by a robotic arm. Because the use of the overhead systems (bridges and cranes) was difficult for this installation, it was determined, that the best alternative was a power manipulator A1000 installed on the floor with a pre-positioning system. The project proceeded to go forward with the adaptation of the standard A1000 with integration of a control system in a Cartesian coordinate frame; resulting in a heavy duty arm; A1000S with 6 degrees of freedom. The heavy duty power arm A1000S was developed and designed for mechanical cutting with an angle grinder, especially for thick and hard material for example stainless steel and Inconel. Cutting a thick material with an angle grinder requires repeatable precise linear motion, in order not to damage the disc. The angle grinder should be handled with a manipulator which can cut in all directions in a linear motion, since the objects which should be cut are usually very large and fixed in any direction For these requirements, the Cartesian control function was added to our standard power manipulator type A1000. In some cases, the decommissioning work includes repeated access routes or fixed access points that the manipulator arm has to pass repeatedly. For example a fixed point where the waste materials are collected, or repeated transport route that the arm has to go along for each step. For this requirement, the A1000S has been equipped with the 'Go To-Mode'. It is a 'teach and playback mode' which has been developed especially for remote operating manipulators. Quick direct teach, quick play function and an intuitive touch-panel layout are some of the features. With the Go To-Mode high working efficiency was achieved. After rigorous testing at Waelischmiller premises; the manipulator was delivered to JNFL in Rokkasho. It was found that the obtained functionality and precision can be applicable not only for the cutting with a disc grinder, but also the laser cutting method, which requires precise trajectory control. This paper will provide: - A project description, - Presentation of solution process, - Technical description of the A1000S, - Future Applications of the robotic Arm A1000S. (authors)

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DOE OSTI2016

NitroJet{sup R}'s Verification Test of Contaminated Water Storage Tank Decontamination in Fukushima Daiichi Nuclear Power Plant - 16489

After the Fukushima Daiichi nuclear disaster, much radioactively contaminated water has generated and has been stored into tanks in the site. Therefore, these contaminated water storage tanks (flange-type) are severely required to be decontaminated in order to be dismantled or to be reused to store cleaner water with less radioactive liquid waste and exposure of workers. Then, IHI Corporation (hereinafter, IHI) have verified that NitroJet{sup R} had capability to decontaminate these tanks under governmental grant (the Validation of technologies for contaminated water management project in the FY2013 Supplementary Budget). For this verification, IHI achieved acquiring DF (decontamination factor) by fabricating radioactive partial mock-ups with Rb-86. Moreover, IHI have demonstrated remote operation of decontamination work by using tank's partially full-scale mock up. (authors)

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DOE OSTI2018

Savannah River Site Defense Waste Processing Facility Melt Cell Remote Cleanup - 18037

This paper documents experiences at Savannah River Site (SRS) related to both the modification and use of a commercial remote vehicle to remove years of process-related litter (glass and various objects) from a hot cell floor during a production outage. The outage was necessary to replace a Department of Energy Environmental Management (DOE-EM) remotely-operated glass melter that had vitrified waste for 14 years at the Defense Waste Processing Facility. Both chunks of waste glass and numerous items littered the floor of the cell all associated with the long production run. The presence of litter become problematic as it could eventually impede movement or placement of items necessary to continue facility operation. The cleanup also reduces radiation rates within the hot cell which is beneficial. (authors)

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DOE OSTI2018

Application of SIAL{sup R} Solidification Technology for Processing of Problematic Waste Streams - 18267

SIAL{sup R} is a proven, safe, cost-effective and licensed waste solidification technology that has been used successfully for 20 years in Slovakia and the Czech Republic for processing of various problematic radioactive waste streams. This abstract presents and summarises the most interesting performance records of the SIAL{sup R} geopolymer solidification technology, the applied mobile devices as well as the characteristics of final products. Radioactive waste solidification into matrix SIAL{sup R} is based on polycondensation of inorganic aluminosilicate components with chemical composition similar to average cement. The main components of matrix SIAL{sup R} are natural inorganic raw materials containing SiO{sub 2}, Al{sub 2}O{sub 3}, CaO, MgO, Na{sub 2}O, K{sub 2}O, Fe{sub 2}O{sub 3}, and TiO{sub 2}. The treatment/solidification process is modified for individual radioactive streams with different compositions. To determine the most appropriate composition, the SIAL{sup R} application technology is laboratory tested on real samples of each individual waste stream identified for solidification. Then an appropriate mixture of aluminosilicate composition and other inorganic compounds is defined for a given solidification application. When determining the most appropriate mixture, other requirements are taken into consideration: maximum dose rate on the drum with solidified product, mass, volume, alpha radionuclide content in the final product, need for a waste's dewatering before solidification, dry matter content in the solidification product as well as other specific waste acceptance criteria (i.e., compressive strength, leachability, diffusion coefficients, radiation stability, biodegradability of matrix, distribution and partition coefficients, influence of frost and water presence to the product stability including long term physical stability in water, dustiness, flammability, time dependence of compression strength and water reduction, product homogeneity and also radiolytic G-value for hydrogen release, GH{sub 2}). The Amec Foster Wheeler baseline approach to implementing the SIAL{sup R} waste treatment process is structured on three steps - waste characterization, waste retrieval and pre-treatment and at last the solidification of pre-treated waste itself. The equipment used for the solidification process is relatively small, lightweight and designed for the specific application on site, directly 'in situ' or local to the tanks storing the sludge's/resins/other waste streams. This mitigates the problems normally experienced by transferring sticky, tight waste through long pipes. The equipment is normally 'tailor-made' to take into account the real space conditions and other existing facility requirements. At Bohunice and Dukovany sites, the facilities for encapsulation were designed to occupy existing areas of the plants in close proximity to the tanks where the waste was being stored. This allowed for the removal of the waste using Remote Operated Vehicles and for the waste to be immediately treated on site without the need for transportation thus saving both, time and resource. To date, SIAL{sup R} has been used successfully to immobilize approximately 2 x 10{sup 6} kg of waste which includes sludge, resin, crystalline borates and contaminated organic waste from Bohunice and Mochovce, Slovakia and approximately 300 m{sup 3} of spent ion exchange resins and sludge from tanks on site at the Dukovany NPP (Nuclear Power Plant) in the Czech Republic. The technology now has a track record of over 20 years which includes on-going research and development. The International missions WANO (the World Association of Nuclear Operators) and OSART (the Operational Safety Review Team) evaluated the SIAL{sup R} matrix technology at Dukovany NPP as an example of good practice. (authors)

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DOE OSTI2018

A Novel Mobile Examination and Remediation Facility for On-Site Remediation of Dry Storage Systems - 18389

Most of the spent nuclear fuel (SNF) in dry storage in the United States is safely and securely stored in dual-purpose (i.e., storage and transportation) canisters [1]. However, some of the SNF is stored in storage-only (i.e., non-transportable) packages, or packages that may become nonconforming for continued storage or transportation for a variety of reasons. Currently, the capability to remediate dry storage systems (DSSs) for SNF that are not transportable or that need to be repaired to maintain confinement boundary integrity during storage service does not exist without on-site access to a spent fuel pool. Hence, there is a potential future need for a mobile facility with remediation capabilities that can be used at multiple storage installations. These capabilities would permit needed repairs to an existing DSS to support continued storage or for the dry transfer of fuel to a new conveyance suitable for transport, if necessary. This is a technical paper that does not take into account the contractual limitations under the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste (Standard Contract) (10 CFR Part 961). Under the provisions of the Standard Contract, DOE does not consider spent nuclear fuel in canisters to be an acceptable waste form, absent a mutually agreed to contract amendment. To the extent discussions or recommendations in this paper conflict with the provisions of the Standard Contract, the Standard Contract provisions prevail. A novel design concept for a Mobile Examination and Remediation Facility (MERF) is proposed with functions and capabilities to remediate a DSS. The proposed MERF would be a multi-purpose facility, remotely operated and computer-controlled with semi-automated equipment to allow trained operators to safely perform remediation activities while keeping worker radiation dose as low as reasonably achievable. It would include a shielded rotating platform with support towers and would use commercial off- the-shelf examination and repair equipment. A slip ring arrangement underneath the rotating platform would allow for the power supply without impeding the continuous rotation and would simplify cable management. The support towers would be mounted on the rotating platform, and each tower would have vertical and horizontal positioning capability for axial and radial tool location. The rotating platform would be attached to the main MERF structure. Capabilities would include visual, surface, and volumetric examinations to fully characterize physical conditions on canister surface before and after a repair; repair techniques, e.g., laser beam welding, friction stir welding, or cold spray, to perform needed remediation activities; and associated equipment for surface preparation such as cutting and grinding equipment. The proposed design would be simple, modular, robust, economically viable, and easily transportable and would have a small footprint. The modular design would present several advantages including compatibility with a variety of different DSSs, on-site assembly, and transportability; it also would be built in interchangeable sections to facilitate decontamination. Further, shielding would be supplemented through interface with the existing transfer cask and storage overpack. Robustness and canister stability would be achieved using pneumatic holders and steel support structures designed with vibration absorbers to reduce impacts and/or a 'pendulum' effect during a seismic event. Thus, the proposed MERF would be designed to: (i) minimize handling of an affected DSS to the extent practicable, (ii) provide radiation shielding and minimize dose to operators and the public during remediation activities, and (iii) enable examination, repair, and repackaging. This paper describes the proposed MERF design concept and capabilities, main components, equipment, different modes of operation, and operating procedures. Structural and safety assessments would need to be performed before final design, licensing, and deployment. The MERF is evaluated with respect to regulatory impact (first of a kind, prior experience), operational impact (safety, security, staffing, ongoing operations, constraints), worker dose (low, medium, high), cost (low, medium, high), schedule (duration, time sequence), and technology readiness level. (authors)

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DOE OSTI2018

Development of Non Destructive Assay Approach for Characterization of Spent Fuel Debris at Fukushima - 18543

Fukushima is considering options for decommissioning the damaged reactor. One of the technical challenges is the accurate characterization of fissile isotopes in the spent fuel debris. This is due to the fissile content, high gamma and neutron emission and uncertainties regarding the fuel burnup and unknown, heterogeneous mixture of materials that may be present. These requirements lead to significant challenges for potential measurement approaches with regard to the requirements for shielding and remote operation that result from plant and operational constraints. Mirion Technologies (MTKK) has been working to perform feasibility and concept design studies for a non-destructive assay (NDA) system that could be used to characterize the fissile content of retrieved samples of fuel debris. The main advantage of using an NDA technique is that the entire volume of the sample is interrogated so the method is not reliant upon assumptions of representativeness of samples retrieved and sent for laboratory analysis. The proposed technique comprises a combination of gamma spectroscopy and combined passive / active neutron interrogation with advanced data analysis algorithms in order to determine the maximum available data on the spent fuel properties and obtain a full radionuclide characterization of each sample. This includes quantification of individual fissile nuclides with validated uncertainties. In this paper we describe the NDA techniques and systems that are proposed, their benefits and limitations, and how they may be implemented into a full-scale system for fuel element debris characterization. We describe the studies that were performed based on experience of real systems and research supplemented with extensive computer modelling work to represent the expected fuel element debris properties. The work benefits from Mirion's extensive global physics and engineering expertise in designing, supplying and operating a wide range of NDA systems for diverse applications and in different challenging physical and operational environments. We describe how the advanced and innovative physics techniques that have resulted from Mirion's collective experience in these areas, can be deployed in an NDA system. The final approach that has been developed provides a means for determining the masses of the individual fissile isotopes with robust uncertainties and capable of implementation into an NDA system that meets the expected requirements for an operational plant. We focus on the techniques that have been developed and describe how they are of wider relevance and interest for projects that include challenges for characterisation of fuel residues. (authors)

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DOE OSTI2019

Radiochemical Characterization of PWR Control Rods - 19030

NRG has, in cooperation with the German partner Hoefer and Bechtel, performed a radiochemical characterization of the control rods from a German Pressurized Water Reactor. To perform the cutting and sampling of the spent control rods at the reactor site NRG has developed a dedicated, remotely operated tool. This tool enabled the sampling of both the stainless steel tube and the absorber material from the upper, middle and lower parts of selected rods. The samples were transported from the reactor site to NRG's laboratories at Petten, the Netherlands, were they were analysed chemically and radiologically. Due to the different chemical constitution of the control rod tubes and the absorber, the procedures for the chemical separation of the different compounds, prior to the radiological characterization, varied. The route for the stainless steel tube samples turned out more complex than that for the absorber samples: there are more nuclides to be determined, and therefore more treatment steps and component separations were necessary. Subsequently to the component separations, the radiochemical characterization was performed using alpha and gamma spectrometry and the Liquid Scintillation Counting (LSC) technique. The results of the measurements in combination with simulations performed with a reactor physics code provided for a better understanding about the distribution of radionuclides in absorber rods, both in the rod tubes and the absorber core. In addition, the results of the measurements can be used to validate reactor physics codes. (authors)

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DOE OSTI2019

Synroc Waste Treatment Facility for Fission-based Molybdenum-99 Production - 19335

ANSTO is building a nuclear waste treatment plant that represents a significant step change in the treatment of intermediate-level waste (ILW) from Molybdenum-99 nuclear medicine production. This industrial scale facility shall deploy ANSTO Synroc technologies; a wasteform tailored for immobilizing fission products, an integrated process line for converting liquid waste into free-flowing powder, automated canister filling and sealing, and a hot isostatic pressing (HIP) system designed for remote operation. The HIP represents the last processing step to consolidate the wasteform into a final durable solid compact with the required performance characteristics. This paper shall outline the various phases of development in realising the final plant design as well as timelines to progress through cold- and hot-commissioning. Further, plans for storage and final disposal of the HIPed canister product will be discussed. (authors)

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DOE OSTI2020

Removal of Metal Oxides from Reactor Metals/Alloys by Laser Ablation

Storage of low and intermediate level radioactive waste in a deep geological repository (L and ILW DGR): - Large volumes of metallic waste = increased production of hydrogen gas. - Large fraction of the metallic wastes are out-of-core reactor components. - Carbon steel feeder pipes. - Radioactive contamination is concentrated within the surface oxides. - Removal of surface contaminants would lead to volume reduction of metallic waste that is stored in L and ILW DGR. - Current decontamination techniques generate large volumes of secondary waste. Objectives: Oxide removal from carbon steel using a ytterbium fiber laser (1064 nm) with a particular focus on identifying the key laser parameters that affect the removal efficiency. Advantages of laser ablation: Generate low volumes of waste, Flexibility of deployment, Remote operation. Grow an oxide film on carbon steel with the incorporation of cobalt. Optimize laser cleaning technology for the removal of surface oxides. Incorporate an oxide collection system for the ablated particles. Conclusions: Objective 1 - Simulated inactive oxides on carbon steel, using Co{sup 2+} as a surrogate for Co radionuclide. Objective 2 - Optimized laser parameters (200 MW/cm{sup 2}, 3 scans, 15 μm overlap); Complete removal of cobalt from surface and bulk metal with optimized cleaning settings. Objective 3 - Glass fiber media H14 HEPA filters had highest collection yield; High oxide collection yields (∼100%). Future Work: Further testing of the liquid trap will verify the efficacy of the HEPA filter and determine the amount of ablated oxides that pass through. Testing of the laser ablation system needs to be done on other reactor materials along with the incorporation of different transition metal cations. Improvements to the oxide collection system are required to improve collection yields.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Life cycle Management of 324 Project High-Activity, Mixed/Low-Level Radioactive Waste - 20506

U.S. DOE contractor CH2M Hill Plateau Remediation Company (CHPRC) manages the 324 Building and is preparing to remotely excavate and disposition the radioactive soil beneath the building to allow further deactivation of the building. The 324 Building is a non-reactor Category 2 Nuclear Facility located in the 300 Area of the Hanford Site. Records indicate that in October 1986, approximately 516 liters of a concentrated liquid waste stream containing cesium-137 ({sup 137}Cs) and strontium-90 ({sup 90}Sr) spilled onto the floor of a hot cell, B-Cell, in the 324 Building. The spill contained an estimated 1.3 million curies (Ci) of radioactivity. A breach in the sump of B-Cell was discovered in November 2009 during characterization of the soil under the building. Exposure rates in excess of 10,000 R/hr were detected through a system of access pipes installed under the hot cell, and an estimated 1.557 E+05 Ci of {sup 137}Cs and 6.842 E+04 Ci of {sup 90}Sr are in the soil beneath B-Cell. The magnitude of the soil contamination below B-Cell represents one of the most challenging remediation activities in the DOE complex. The objective of CHPRC's 324 Project is to remotely excavate and disposition the highly contaminated soil under the hot cell so building deactivation and stabilization activities can resume. Depending on the volume and radioactivity of contaminated soil under B-Cell, the material will be dispositioned either at the Hanford Site's Environmental Restoration Disposal Facility (ERDF) or in grout monoliths that will be created in the hot cells (A, C and D) next to B-Cell during future building demolition activities. The current project scope includes installing structural supports under the walls of B-Cell and using remotely operated equipment to remove debris from the hot cell, remove debris mixed with grout on the floor, saw through and remove the cell floor and liner, and excavate contaminated soil under B-Cell to a depth of up to 3.66 meters (12 feet). Addition or removal of debris in A-Cell may be required to support B-Cell cleanout and excavation activities. The base approach is to disposition debris and soil bins in the Radiochemical Engineering Cell (REC) monoliths, and subsequently disposition the monoliths during building demolition. (authors)

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DOE OSTI2016

Dry Canister Storage System Inspection and Robotic Delivery System Development

EPRI is working to develop and demonstrate technologies capable of inspecting welded stainless steel canisters used in dry cask storage systems in situ, without the need to remove them from their concrete casks. This involves the coupling of specialized cameras and inspection devices with small remotely operated vehicles (ROVs). The objective is to demonstrate the capability to identify any presence of (or precursors to) stress corrosion cracking (SCC) and other anomalies that may require additional investigation. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2024

FORCE Regression Testing

Via programs including the Light Water Reactor Sustainability and Integrated Energy Systems, the U.S. Department of Energy has invested in the Framework for Optimization of ResourCes and Economics (FORCE) software framework (Idaho National Laboratory 2024a) for the technical and economic analysis of nuclear-integrated energy systems (IES). Nuclear IES expand the use of nuclear from traditional baseload electricity generation to a flexible and adaptive source of combined heat and power. Nuclear heat can be used in the production of a variety of energy currencies such as hydrogen and ammonia as well as other heat applications including water desalination and district heating. FORCE is designed with the intent to provide interconnected analysis tools that enable the accurate technical and economic assessment of specific nuclear IES configurations for individual energy markets. FORCE consists of three main analysis pathways: HYBRID (Idaho National Laboratory 2024b), which contains high-resolution physical models for IES; Holistic Energy Resource Optimization Network (HERON) (Idaho National Laboratory 2024c), which analyzes IES long-term economic viability; and Optimization of Real-time Capacity Allocation (ORCA) (Idaho National Laboratory 2024d), designed for real-time control of IES via digital twins and optimal decision making, including autonomous and remote operation research. Development of the FORCE ecosystem is guided by three pillars: capability, which assures that the computational requirements of IES analysis are met by the software tools; reliability, which provides for consistent code performance and expected behaviors; and accessibility, which lowers the barrier to entry for using the software and accelerates analysis by users beyond the FORCE primary developers. Reliability of the FORCE ecosystem is established according to the American Nuclear Society?s Nuclear Quality Assurance (NQA-1) program [American Society of Mechanical Engineers 1982], with specific levels of software quality assurance (SQA) within NQA-1 applied to each software tool in FORCE. As the tools within FORCE have matured, some integration algorithms to accurately connect the software tools for holistic analysis have been developed and deployed within the FORCE software repository. In accordance with NQA-1 standards, regression tests are required to guarantee the software performs consistently even when new capabilities are added to the software. In this report, we document the deployment of both unit tests, which test the consistent behavior of small pieces of the FORCE code base, as well as integration tests, which test the consistent performance of full use cases for the FORCE integration algorithms. We further document the encapsulation of these tests within a test harness, which collectively checks for each successful test completion on demand. Finally, we document the automation of the test harness using GitHub Actions [GitHub 2024], which require all tests succeed before any new capability or other changes can be added to the FORCE integration software

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2025

E-198A Proton Nucleon Scattering

During the past year the E-198 spectrometer was completed.This involved the construction and implementation of 15 MWPCplanes, a remotely operated liquid helium transport system forthe LBL superconducting magnets and a room temperature gas jettarget. The experimental setup is indicated in the first attachedfigure.

43 PARTICLE ACCELERATORS↗
DOE OSTI2025

HotCell Components

To support operations in the HFEF Hot Cell and the FCF Hot Cell, multiple components were designed to support the Bottle Cutter Project and the Sodium Loop Project. These components were designed and analyzed to ensure they can function in their environment and be operated remotely. FEA analysis were performed, prototypes were made, and parts were manufactured to complete this project. This poster documents these parts and test results.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗