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DOE OSTI2025

TEM Characterization of Nanostructured Grade 91 Neutron Irradiated to 6 DPA at 300 °C and 500 °C

Previously, samples of coarse grained (CG, grain diameter > 1 µm), ultra-fine grained (UFG, 1 µm> grain diameter> 100 nm), and nanocrystalline (NC, 100 nm> grain diameter) grade 91 steel were neutron irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). Grade 91 is a ferritic/martensitic steel alloy with a nominal composition of Fe-9%Cr-1%Mo by weight, and it is being studied for applications in advanced nuclear reactor designs including as a fuel cladding material in gas cooled reactors. To assess the viability of this alloy for nuclear applications, the role of grain boundary size has not been sufficiently explored as a mechanism for developing radiation-tolerant materials. To achieve an UFG microstructure the material was processed using equal channel angular pressing (ECAP), and the NC microstructure was achieved through high pressure torsion (HPT). Such samples of various geometries for different types of characterization were irradiated to a dose of 6 displacements per atom (DPA) at separate temperature conditions of 300 °C and 500 °C. Thus, there were a total of 6 experimental conditions, three grain sizes (CG, UFG, and NC), each irradiated at two different temperature conditions (300 °C and 500 °C).

36 - MATERIALS SCIENCE↗

DOE OSTI2022

Versatile Test Reactor Conceptual Core Design

The VTR is a 300-MW(thermal) sodium-cooled fast reactor (SFR) designed for the specific purpose of delivering unique testing capabilities to enable the advancement of all reactor technologies. With its flux level, irradiation volume, and operational flexibility, the VTR will enable accelerated testing of materials, fuels, and various components needing irradiation testing. Proven SFR technologies and design approaches have been leveraged in designing the VTR core, ensuring the highest possible readiness level. This resulted in the VTR using ternary metallic fuel and delivering fast flux levels in excess of 4 x 10 15 n/cm 2 ∙ s over large useful volumes, corresponding to about 60 dpa/year in steel. As part of the design efforts, the VTR core performance has been determined for a representative configuration, ensuring that the reactivity control systems offer sufficient shutdown margins, that the core can be safely cooled in all situations, and that reactivity feedback coefficients are conducive to a favorable safety behavior. Furthermore, the incorporation of features such as fuel assembly storage in the shield region supports the flexible and reliable operation of the VTR. Additional design work has been ongoing as well. This includes thorough shielding performance evaluations to ensure safe operation of the VTR, verification and validation of the design tools used to achieve compliance with Nuclear Quality Assurance (NQA-1) requirements, early assessment of the impact of irradiation experiments on the core performance envelope and associated margins, and in-depth uncertainty quantification efforts to quantify the anticipated range of performance characteristics. An experimental program supporting the VTR core design has been set up, with the current focus being on thermal-hydraulic experiments. The purpose of this experimental program is to obtain confirmatory measurements to serve directly as part of the core design basis or as part of the validation cases supporting the simulation tools used.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2021

Mechanistic Source Term Determination

Understand the function and performance of the different barriers to radionuclide release in HTGRs, how these are incorporated into reactor design and safety analyses, and approaches to estimate radionuclide release under specific reactor conditions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI1994

Nuclear Safety [Vol. 35, No. 1, January-June 1994]

Nuclear Safety is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: THE CHERNOBYL ACCIDENT: 1 Chernobyl Accident Management Actions, A. R Sich 25 The IAEA-ASSET Approach to Avoiding Accidents is to Recognize the Precursors to Prevent Incidents, F. Reisch; ACCIDENT ANALYSIS: 36 A Review of the Available Information on the Triggering Stage of a Steam Explosion, D. F. Fletcher; 58 Analysis and Modeling of Flow-Blockage-Induced Steam Explosion Events in the High-Flux Isotope Reactor, R. P. Taleyarkhan, V. Georgevich, C. W. Nestor, U. Gat, B. L. Lepard, D. H. Cook, J. Freels, S. J. Chang, C. Luttrell, R. C. Gwaltney, and J. Kirkpatrick; 74 An Analysis of Disassembling the Radial Reflector of a Thermionic Space Nuclear Reactor Power System, M. S. El-Genk and D. V. Paramonov; CONTROL AND INSTRUMENTATION: 86 Standards for High-Integrity Software, D. R. Wallace, D. R. Kuhn, L M. Ippolito, and L. Beltracchi; DESIGN FEATURES: 98 Adoption of New Design Features for the Next Generation Nuclear Power Reactors, L. S. Tong; 114 Review of Nuclear Piping Seismic Design Requirements, G. C. Slagis and S. E. Moore; ENVIRONMENTAL EFFECTS: 128 PC-Based Probabilistic Safety Assessment Study for a Geological Waste Repository Placed in a Bedded Salt Formation, S. A. Khan; OPERATING EXPERIENCES: 142 Managing Aging in Nuclear Power Plants: Insights from NRC’s Maintenance Team Inspection Reports, A. Fresco and M. Subudhi; 150 Reactor Shutdown Experience, Compiled by J. W. Cletcher; 153 Selected Safety-Related Events, Compiled by G. A. Murphy; RECENT DEVELOPMENTS: 158 Reports, Standards, and Safety Guides, D. S. Queener; 168 Proposed Rule Changes as of Dec. 31, 1993; ANNOUNCEMENTS: 177 Symposium on Radioactive and Mixed Waste—Risk as a Basis for Waste Classification; 177 1995 Incineration Conference 178 Ninth Power Plant Dynamics Control and Testing Symposium; 174 The Authors

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI1989

Nuclear Safety [Vol. 30, No. 1, January-March 1989]

Nuclear Safety is a review journal that covers significant developments in the field of nuclear safety. Its scope includes the analysis and control of hazards associated with nuclear energy, operations involving fissionable materials, and the products of nuclear fission and their effects on the environment. Primary emphasis is on safety in reactor design, construction, and operation; however, the safety aspects of the entire fuel cycle, including fuel fabrication, spent-fuel processing, nuclear waste disposal, handling of radioisotopes, and environmental effects of these operations, are also treated. Table of Contents for this issue follows. GENERAL SAFETY CONSIDERATIONS: 1 International Conference on Thermal Reactor Safety: NUCSAFE 88 by A. P. Malinauskas and J. G. Pruett; ACCIDENT ANALYSIS: 18 Chernobyl Today: State of Research by A. Yu. Gagarinskii; CONTROL AND INSTRUMENTATION: 23 Problems Associated with Routine In-Plant Radioactive Effluent Monitoring Systems at U.S. Light-Water Reactors by Barry G. Wahlig, David M. Walker, Mahmoud R. Ghavi, and John M. Palms; DESIGN FEATURES: 34 Twentieth DOE/NRC Nuclear Air-Cleaning Conference by R. R. Bellamy, D. W. Moeller, and M. W. First, 46 Design Options for the Core Support Structure for Liquid-Metal-Cooled Reactor Plants by J. P. Burelbach, W. J. Kann, J. G. Saiveau, and R. W. Seidensticker; ENVIRONMENTAL EFFECTS: 53 The Radiological Consequences of the Chernobyl Accident by L. V. Konstantinov and A. J. González; WASTE AND SPENT FUEL MANAGEMENT: 70 Activities Related to Waste Management Compiled by E. G. Silver; OPERATING EXPERIENCES: 80 Fuel Performance Annual Report for 1986 by W. J. Bailey, 103 Safety Considerations Related to 12-Hour Shift Schedules by S. Gould, 105 Reactor Shutdown Experience Compiled by J. W. Cletcher, 108 Operating U.S. Power Reactors Compiled by E. G. Silver; RECENT DEVELOPMENTS: 130 General Administrative Activities Compiled by E. G. Silver, 140 Reports, Standards, and Safety Guides by D. S. Queener, 145 Status of Power-Reactor Licensing Activities Compiled by E. G. Silver, 153 Proposed Rule Changes as of Sept. 30, 1988; ANNOUNCEMENTS: 69 Northwestern University Short Course on Radiation Safety, 79 International Seminar on Fission Product Transport Processes in Reactor Accidents, 129 Harvard School of Public Health Offers Short Courses, 152 MIT Offers Summer Short Courses on Reactor Safety and Nodal Analysis, 152 Seminar on Use of PCs in Probabilistic Safety Analysis, 157 The Authors, 160 Indexes to Nuclear Safety, Volume 29.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2026

Technoeconomic Design Optimization for Fast Reactors. Part II: Impact of Technoeconomic Constraints on Optimal Design

There is a current drive toward optimizing reactors, particularly small/micro reactors to minimize cost and maximize performance. Previous work has investigated the development of technoeconomic workflows for the design optimization of pool-type fast reactors that aim to deploy into district energy grids. Initial scoping studies verified that the workflow was capable of capturing design trends throughout a variety of design configurations and problem formulations while remaining sufficiently flexible. In this paper, this methodology is applied to understand how cost functions and technoeconomic constraints can drive optimal reactor design. Specifically, the UPu10Zr-fueled fast reactor model from Part I is adapted to include changes in the fissile content limits, control rod worth limits, control rod drive cost, and assumed fuel form. In the case of constraint relaxation at fixed power (fissile content and control rod worth limits), cost sensitivities of 5% to 10% were uncovered. Multi-objective optimization at varying reactor power levels with individualized control rod drives for each assembly (as opposed to one operational and one safety drive) increased cost by $\$10$ to $\$25$ million and substantially altered the optimal core geometry, favoring geometries with substantially fewer control rod placements relative to baseline optimization. Finally, a multi-objective optimization was performed at varying power levels with the fuel form overhauled to metallic, high-assay low-enriched uranium–based U10Zr with more refined fuel cost models. In the case of uranium fueling, the costs increased by at least $50 million relative to the baseline case. Furthermore, economic fuel zoning and lower reactivity swing cores were recovered. Each case serves to demonstrate the value of applying technoeconomic workflows to initial reactor design scoping studies to better understand the trade-off for a proposed concept between different design options.

Argonne Reactor Computation (ARC) codes↗

DOE OSTI2021

Fate and transport of unruptured tri-structural isotropic (TRISO) fuel particles in the event of environmental release for advanced and micro reactor applications

Advanced nuclear reactor designs and advanced fuel types offer safety features that may reduce environmental consequences in an accident scenario when compared to conventional reactors and fuels. One advanced reactor fuel is tri-structural isotropic (TRISO) fuel particles which are approximately 0.9 mm in diameter. TRISO particle mobility, assuming the particle is unruptured and the encapsulated radionuclides are contained, was explored through a theoretical examination of mobility through atmospheric, soil and groundwater, surface water, and non-human biota transport pathways. TRISO particles are too large and dense to travel in the atmosphere except under extreme conditions. TRISO particles are too large to penetrate most soil profiles and so cannot be transported with groundwater. TRISO particles will settle out of the water column in surface waters but the transport will depend on the energy of the body of water and likelihood of extreme dispersion events. TRISO particles could be transported by non-human biota. The size of TRISO particles could allow them to be intentionally moved by non-human biota if they are ingested as a gastrolith or if they are mistaken for something in the organism’s diet. Generally, TRISO particles will have reduced environmental mobility compared to releases of radionuclides in the event of an accident in a conventional nuclear reactor. The extent of transport has implications in emergency planning zone designations and other considerations for licensing and deploying TRISO-fueled reactors. Finally, further research and experimental work exploring TRISO particle mobility is required to understand the full environmental mobility of TRISO particles.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗

DOE OSTI2020

Model MC&A for Pebble Bed Reactors (Technical Direction No. 5 Task 2.6 Letter Report)

In preparation for non-light water reactor (non-LWR) activities, US Nuclear Regulatory Commission (NRC) staff are advancing risk-informed and performance-based licensing approaches and addressing key policy issues. One non-LWR reactor concept is a pebble bed reactor (PBR). This reactor design uses spherical fuel elements (pebbles) that are continually added to and removed from the reactor core. The free movement of the fuel in this design presents new challenges for material control and accounting (MC&A) programs. Therefore, an assessment of MC&A program features and measures for a PBR was performed to help NRC staff develop associated MC&A regulations or regulatory guides. The current regulatory framework for non-LWR fuel cycles excludes support for licensing reviews for MC&A programs for PBRs. Licensing reviews of an MC&A program for PBRs can be facilitated by (1) a model MC&A program for a PBR based on identification and assessment of MC&A program features and recommended measures for a reference PBR and (2) a methodology for assessing MC&A performance that can help assess different MC&A program features and measures. This report supports the NRC’s non-LWR Vision and Strategy Near-Term Implementation Action Plans.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2022

Deterministic modelling of the SPERT IV reactor transients using the multi-physics capability of WIMS

The ANSWERS WIMS reactor physics code is being developed for whole core multi-physics modelling. The established neutronics capability for lattice calculations has recently been extended to be suitable for whole core modelling of Small Modular Reactors (SMRs). A whole core transport, SP3 or diffusion flux solution is combined with fuel assembly resonance shielding. An integrated thermal hydraulic solver permits temperature and density variations to feedback to the neutronics calculation. This capability can be applied to both steady state and time dependent transient problems. Nuclear reactor design and safety case development requires assessment of a range of reactor transients, to inform both normal operation limitations and accident scenario analysis. This paper presents new methodology developed in WIMS to couple the core neutronics to the integrated core thermal hydraulics solver for the simulation of reactor transients in whole core models. To support the validation of the multi-physics capability of WIMS, this capability has been applied to the reactivity insertion transient experiments performed in the plate fuelled SPERT-IV reactor. This study employs WIMS using the whole core solver MERLIN, which calculates the time-dependent flux distribution and magnitude, coupled to ARTHUR, which solves for the time-dependent thermal hydraulics solution, and provides thermal feedback via the cross sections generated by GEOM, which performs resonance shielding calculations and generates cross section data for the plate geometry. This spatial kinetics model, with dynamic cross section generation, allows for variations in the temperature and neutron flux profile with time.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2024

An Autonomous Critical Data Extrapolator for the AGN-201m

Nuclear nonproliferation serves as a key goal, being undertaken by the International Atomic Energy Agency (IAEA). To recognize proliferation there are two pathways that states, who intend to use nuclear material for malicious purposes can take, diversion can misuse. Diversion is when fissile nuclear material is declared to the IAEA for non-weapon purposes, but then covertly removed. If the source of nuclear material, that is not declared and not fissionable, is placed inside the reactor core to create fissile material used to create weapons then the state is using the second pathway of proliferation, misuse. With the emerging development in areas of simulation and machine learning the creation of virtual models of reactor systems, digital twins, serve as a potential method to identify proliferation through detecting anomalous behavior in the reactor. A digital twin for a physical nuclear reactor has never been developed, as digital twins serve as an emerging technology. To investigate the process for development and use of a digital twin for a nuclear reactor Idaho State University’s AGN-201m serves as the nuclear reactor used for development of this digital twin. A data acquisition system has been installed to the reactor system allowing for the transfer of collected data from a reactor operation to Idaho National Laboratory’s Deeplynx data warehouse. When utilizing data to train reactor physics and machine learning models, a significant challenge encountered is the initial state of the data. Nuclear proliferation will have the capacity to be detected when the reactor immediately starts up, nor will it occur after the reactor shuts down. Generally, it will be detected when the reactor is operating at some desired power over a sufficient period for that specific reactor design. For the AGN-201m this will be when the reactor is critical (generally 1 mW or above) for a timespan that is within or less than the range of a regular business day. Datasets sent to Deeplynx have had to be manually cut to when the reactor is critical based on plots of power levels. This method is inefficient and laborious, especially when using multiple datasets at once to train a model. To provide a more streamlined approach an automated critical data extrapolator is developed, with capabilities of recognizing when the reactor operation first reaches criticality, and when the reactor undergoes a SCRAM and is shutdown.

99 GENERAL AND MISCELLANEOUS↗

DOE OSTI2025

Development and Evaluation of Embrittlement Resistant Alloys for Advanced LWR Cladding

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2025

Development and Evaluation of Embrittlement Resistant Alloys for Advanced LWR Cladding

The primary challenge preventing the deployment of reactor designs that leverage wrought FeCrAl as an advanced light-water reactor (LWR) cladding is irradiation hardening and embrittlement. Wrought FeCrAl alloys experience a loss of post-irradiation ductility and fracture toughness under low-temperature neutron irradiation (< 350°C) resulting from the combined effects of dislocation loop formation and the irradiation-enhanced precipitation of Cr-rich alpha-prime precipitates throughout the microstructure. Over the past decade, significant improvements in our understanding regarding the effect of Cr and Al content have been enabled through neutron irradiations, but even optimized wrought alloys such as C26M remain vulnerable to brittle failure during storage, transportation, and handling following irradiation. This report summarizes multiple irradiation campaigns initiated over the past 7 years and provides recent insights into the effect of (1) minor alloying composition, (2) alloy processing strategy, (3) crystal structure, and (4) the use of advanced oxide dispersion strengthened (ODS) alloys. The first irradiation campaign irradiated body-centered cubic FeCrAl alloy C26M with and without elements such as Mo and Y. It also included face-centered cubic alumina-forming austenitic (AFA) alloys that have a steeper neutronic penalty but that have far superior high-temperature strength than wrought FeCrAl. Finally, this campaign compared conventional wrought C26M with the same material produced using modern powder metallurgical hotisostatic pressing (PM-HIP). The second irradiation campaign, initiated in 2018, irradiated wrought and FeCrAl-ODS materials up to 50 dpa to assess their irradiation resistance to high-doses. The results of this work indicate that wrought C26M suffers significant degradation following neutron irradiation at LWR-relevant temperatures regardless of Mo and Y content. The PM-HIP variant appears to be more resilient to irradiation-induced ductility loss in comparison with wrought variants but still loses almost all ductility by 8 dpa. Notably, the AFAs retained at least 5% total elongation after irradiation at the same dose and temperature condition, although additional deterioration may be expected at higher fluence levels. FeCrAl-ODS materials show the greatest promise as a transformative longer-term accident-tolerant fuel (LT-ATF) cladding material. The extruded 106ZY10C alloy retained 10% total elongation after 16 dpa irradiation and retained 8% total elongation after 50 dpa. Additional challenges with FeCrAl-ODS alloys also remain, including optimization of end cap joining methodologies, enhancement of fracture toughness, scaling production to prove economic viability, and optimizing post-pilger heat treatments to maximize ductility and irradiation resistance. This series of irradiations demonstrates the efficacy of the LWR testbed available within the United States for the rapid irradiation and down-selection of LT-ATF candidates. As the country rapidly accelerates its timeline for the deployment of advanced reactor concepts, the effective utilization of the High Flux Isotope Reactor for separate-effects style irradiations should continue to be prioritized to answer the final questions pertaining to LT-ATF candidates necessary for the deployment of advanced boiling water reactors and small modular reactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Optimization of Graded Bed Reactors for Syngas to Olefin (STO) Processes

Reactor designs with mixed catalysts play an important role to transform a multiple reactor system to single-shot reactors. In addition to savings in capital and ease of implementation, single-shot reactors are useful to break equilibrium limitations, therefore increasing the yield and selectivity of desired product. However, the nonlinear, highly exothermic and poorly conditioned nature of mixed-catalyst systems makes it difficult for commercial process simulation and optimization tools to optimize these systems. Here, this study describes the zone and partial moving finite element approaches and their applications to o-xylene oxidation and syngas to olefin (STO) processes. 35% and 5% increases are achieved in the product yield for o-xylene oxidation and STO problems, respectively. Shown with these two examples, this tool is very promising in process intensification for other reactive systems.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2022

Optimization Modeling for Advanced Syngas to Olefin Reactive Systems

Reactor designs with mixed catalysts play an important role in transforming a multiple reactor system to single-shot reactors. In addition to savings in capital and ease of implementation, single-shot reactors are useful to break equilibrium limitations, thereby increasing the yield and selectivity of desired product as shown in previous studies. However, the nonlinear and highly exothermic nature of mixed-catalyst systems makes it difficult for commercial process simulation and optimization tools to optimize these systems. This study describes the development and application of optimization strategies for mixed-catalyst, single-shot reactors for syngas to olefin (STO) processes. Finding the optimal catalyst distribution is challenging and requires advanced solution strategies for singular optimal control problems, which are poorly conditioned and often lead to flat response surfaces. The graded bed and partial-moving finite-element approaches are used to find the optimal catalyst distribution that maximizes the olefins yield. A 1.3% increase in the yield is observed from one zone to three zones. The yield further improves from three zones to the exact infinite dimensional solution by 0.2%. This improvement can be realized in practice by changing only the catalyst distribution, without any extra investment. Lastly, the results suggest that a mixed-catalyst single shot reactor bed can be applied to other reaction mechanisms to increase reactor performance.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2022

Large Eddy Simulation of Random Pebble Bed Using the Spectral Element Method

The high-fidelity simulations are performed with Nek5000/NekRS to show the importance of various phenomena associated with fluid flow and heat transfer through pebble beds. As a result of this work, Kairos Power and NEAMS have a better understanding of the behavior of pebble bed geometries. Kairos Power will leverage the data generated and conclusions reached as part of this investigation to help accelerate the deployment of their reactor design.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2021

ACCRUE—An Integral Index for Measuring Experimental Relevance in Support of Neutronic Model Validation

A key challenge for the introduction of any design changes, e.g., advanced fuel concepts, first-of-a-kind nuclear reactor designs, etc., is the cost of the associated experiments, which are required by law to validate the use of computer models for the various stages, starting from conceptual design, to deployment, licensing, operation, and safety. To achieve that, a criterion is needed to decide on whether a given experiment, past or planned, is relevant to the application of interest. This allows the analyst to select the best experiments for the given application leading to the highest measures of confidence for the computer model predictions. The state-of-the-art methods rely on the concept of similarity or representativity, which is a linear Gaussian-based inner-product metric measuring the angle—as weighted by a prior model parameters covariance matrix—between two gradients, one representing the application and the other a single validation experiment. This manuscript emphasizes the concept of experimental relevance which extends the basic similarity index to account for the value accrued from past experiments and the associated experimental uncertainties, both currently missing from the extant similarity methods. Accounting for multiple experiments is key to the overall experimental cost reduction by prescreening for redundant information from multiple equally-relevant experiments as measured by the basic similarity index. Accounting for experimental uncertainties is also important as it allows one to select between two different experimental setups, thus providing for a quantitative basis for sensor selection and optimization. The proposed metric is denoted by ACCRUE, short for Accumulative Correlation Coefficient for Relevance of Uncertainties in Experimental validation. Using a number of criticality experiments for highly enriched fast metal systems and low enriched thermal compound systems with accident tolerant fuel concept, the manuscript will compare the performance of the ACCRUE and basic similarity indices for prioritizing the relevance of a group of experiments to the given application.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2023

Engagement opportunities in OECD NEA benchmark development

A myriad of opportunities is available to collaborate via international benchmark exercises and experimental data preservation activities. Many such opportunities abound under the auspices of the Nuclear Science Committee of the Organisation for Economic Co-operation and Development Nuclear Energy Agency (NEA). Key projects and activities of relevance to the development of advanced reactors design include the International Criticality Safety Benchmark Evaluation Project (ICSBEP), the International Reactor Physics Experiment Evaluation Project (IRPhEP), the International Assay Data of Spent Nuclear Fuel Database (SFCOMPO), the Shielding Integral Benchmark and Archive Database (SINBAD), and The International Experimental Thermal HYdraulicS Database (TIETHYS), and various cooperative benchmark exercises. Interested participants are encouraged to contact the leadership and secretariat of the various Technical Working Groups and Working Parties to become more engaged. This paper provides a summary of the current benchmark exercises and experimental databases available for international participation.

96 KNOWLEDGE MANAGEMENT AND PRESERVATION↗

DOE OSTI2022

Investigations of Advanced Measurement Methods for High-Temperature Heat-Pipe Experiments

Various types of microreactor designs are under consideration, and among those, the heat-pipe cooled reactor designs are currently being actively investigated at Idaho National Laboratory (INL). In order to support non-nuclear testing and demonstration of technology under the Department of Energy (DOE) Microreactor Program (MRP), Single Primary Heat Extraction and Removal Emulator (SPHERE) was designed and constructed to perform high-temperature heat pipe (HP) experiments at INL. Based on the SPHERE experimental setup, this summary paper conducts the preliminary investigations among the available literature and summarizes some potential advanced measurement techniques for operando high-temperature heat pipe which can be implemented to support MRP. The primary objective for the extensive literature research is to investigate the possibilities and feasibilities for designing and construct an advanced experimental test facility with the aim of producing high-fidelity high-resolution heat pipe data during its operations.

42 ENGINEERING↗