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DOE OSTI2022

SSR COST – COmposite Structural Technologies for advanced reactor deployment (Final Scientific/Technical Report)

Moltex is designing a Generation 4 or Advanced nuclear reactor, the molten salt fast spectrum SSR-W. The design intent is to reduce the cost and complexity of the nuclear systems by relying on inherent safety characteristics, instead of engineered systems. One such characteristic as a natural tendency for power to decrease as temperature increases. Another characteristic that improves natural safety is the fact that the working fluids (molten salts) are at atmospheric pressure at operating temperatures. This significantly reduces the need for pressure boundary components and systems. Moltex is highly motivated to develop simple and economic reactors to reduce the carbon intensity and improve energy security of global electricity production. The COST project investigated the opportunities for reducing construction cost of advanced reactors with the adoption of Rigid Elastic Composite Civil works (RECC). The construction design and methods used in currently operational light water power plants would generally be suboptimal for the SSR-W since there is no requirement for civil structures to withstand high pressure accidents such as hydrogen combustion or steam explosion. In parallel to downselecting the most promising RECC-type technology, accident scenarios representing the largest credible challenge to the SSR-W civil and construction design were modelled. The capability of an SSR-W reactor constructed with and without the selected construction technology was evaluated and the consequential licensing challenges identified. A construction schedule and cost estimate of the SSR-W with and without the selected RECC construction technology was developed. Once the above data was collected, then Moltex was able to review the safety performance, license-ability and cost/schedule benefit of applying the preferred construction technology. The CAPEX/NPV benefit was quantified, and the licensing risks and opportunities were documented. Finally, the CAPEX/NPV benefit of the advanced construction technology was modest due to the suitability of traditional stick-built construction for the SSR-W design. However, previously unidentified benefits of the advanced construction technology were revealed during the studies that could credibly de-risk certain aspects of the design to be licensed. Since licensing (duration and delays) are very significant in new build project costs, the cost benefits may be significant. Further studies on the risk reduction of the construction technology are foreseen.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Assessment of the MAMMOTH and PRONGHORN Pebble Bed Reactor Modeling Capabilities for the HTR-PM Design

This assessment explores the current analysis methodologies of various external and MOOSE-based tools available at the Idaho National Laboratory to model pebble bed reactor multiphysics phenomena. The study involves three MOOSE-based, coupled codes: 1) MAMMOTH for reactor physics, 2) Rattlesnake for neutron transport and 3) Pronghorn for thermal fluids and strongly coupled conjugate heat transfer. These advanced tools are being developed under DOE-NE’s Nuclear Energy Advanced Modeling and Simulation program (NEAMS). They currently contain a collection of advanced modeling capabilities that allow high-fidelity simulations of advanced reactor concepts including gas and fluoride-salt cooled pebble bed reactors. Software development of these tools follows NQA-1 standards ensuring software quality that many of the currently deployed legacy codes lack. In this work, the MAMMOTH Pronghorn tools are compared to the legacy codes PEBBED-THERMIX-KONVEK. The external codes in use are LAMMPS for pebble mechanics and Serpent for cross section preparation. The coupling under MOOSE enables significant flexibility for modern engineering tools since MOOSE applications can be seamlessly coupled with various levels of resolution. This report focuses on low-resolution (r-z) and midresolution (3-D) homogenized geometries. The high-resolution Pebble Tracking Transport solution is not pursued herein and is postponed to future work. The results from this assessment show that the new MOOSE-based tools compare well with the results from legacy tools (r-z) and produce reasonable accuracy limited by the methods in the legacy tools. In the 3-D studies, the PJFNK-SPH methodology available in MAMMOTH allows the reproduction of Monte Carlo results with a transport-corrected diffusion solution. This has implications for the coupled solutions since without the SPH correction the temperature coefficient of reactivity is under-predicted and leads to lower temperature feedback and a significantly higher eigenvalue showing a difference of 2000 pcm.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Reactor System Demonstration with Cyber-Attack Scenarios Using CrowPis and Arduino Microcontrollers

This study covers developing and simulating nuclear reactor system using CrowPis and Arduino microcontrollers for demonstrating cyber-attack scenarios. The team was tasked with implementing more sensors and cybersecurity aspects to the reactor program that was created by last year’s high school interns. The team received the opportunity to collaborate and obtain advice from multiple university interns that helped us gain a better perspective of our project. Our mentor’s background in nuclear science was pivotal to our understanding what we could add to the reactor program to make it as realistic as possible. The first week of our internship was spent reading as much material as possible to gain an understanding of and the background for cyber-attacks and nuclear science. Nuclear science was a new horizon for each of the high school interns on the team, so spending this time in the beginning of our internship was crucial to our success. For the remaining portion of our internship, the team collectively did our best to implement as many sensors and use as much hardware as we could to make an accurate representation of a nuclear reactor in the program that was created. This internship was a big learning experience for everyone on the team. We all gained so many insights into the nuclear world and how it can benefit our lives, as well as how so many moving pieces are needed for it to work properly.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

Thermal Integration of Advanced Nuclear Reactors with a Reference Refinery, Methanol Synthesis, and a Wood Pulp Plant (Rev.1)

The present report is intended to provide process flow diagrams (PFDs) and energy and mass balance data sheets for a U.S. industrial sector subset with which nuclear heat and power could be integrated—a subset that includes the oil refining, methanol and pulp and paper industries. Coupling options for integrating nuclear energy into these industries are quantitatively outlined for reference systems, and future work will extend this analysis in greater detail. Opportunities for integrating small modular nuclear reactors (SMNRs) were investigated for each of the industrial process configurations. Aspen HYSYS and Cycle-Tempo models for a high-temperature gas-cooled reactor were developed to evaluate the proposed integration. This introductory evaluation provides a general description and assessment of the operating principles, reactor coolant core outlet temperature, and reactor size to be integrated with industry. The industrial processes of oil refining and the production of methanol, pulp and paper were simulated by using Aspen HYSYS, Aspen Plus, and the PRELIM (Petroleum Refinery Life Cycle Inventory Model) tool to develop process details. Cycle-Tempo models then extend the process modeling results to obtain net energy demands (e.g., heat, steam, and electricity) when accounting for process steam and waste heat recovery. This information is intended to foster the analysis of integrating an SMNR to decarbonize industrial facilities. The SMNR would provide reliable, competitive, and sustainable clean energy while reducing carbon emissions and other environmental impacts, such as water withdrawals, consumption, and contamination. The refining industry, exhibited in Figure ES1, is a leading consumer of fossil -fuel-based heat, power, and hydrogen in the U.S. industrial sector, generating over 164 million metric tons (MMT) of CO 2 emissions in 2023. The overall mass and energy pertaining to a generalized complex refinery in the United States is reflected in Figure ES1, along with energy metrics regarding integration with a nuclear power plant (NPP). Data sheets were developed to indicate the energy requirements for the overall refinery and each refinery process. The data sheet for the overall refinery is shown in Table ES2.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Overview of the US Department of Energy Light Water Reactor Sustainability Program

The US Department of Energy Light Water Reactor Sustainability Program is focused on the long-term operation of US commercial power plants. It encompasses two facets of long-term operation: (1) manage the aging of plant systems, structures, and components so that nuclear power plant lifetimes can be extended and the plants can continue to operate safely, efficiently, and economically; and (2) provide science-based solutions to the nuclear industry that support implementation of performance improvement technologies. An important aspect of the Light Water Reactor Sustainability Program is partnering with industry and the Nuclear Regulatory Commission to support and conduct the long-term research needed to inform major component refurbishment and replacement strategies, performance enhancements, plant license extensions, and age-related regulatory oversight decisions. The Department of Energy research, development, and demonstration role focuses on aging phenomena and issues that require long-term research and/or unique Department of Energy laboratory expertise and facilities and are applicable to all operating reactors. This paper gives an overview of the Department of Energy Light Water Reactor Sustainability Program, including vision, goals, and major deliverables.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

GROWTH OF THE INTERNATIONAL CRITICALITY SAFETY AND REACTOR PHYSICS BENCHMARK EXPERIMENT EVALUATION PROJECTS SINCE ICNC 2011

Since the International Conference on Nuclear Criticality Safety (ICNC) 2011, the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPhEP) have continued to expand their efforts. Twenty countries have continued to participate in the ICSBEP since 2011; participating countries in the IRPhEP have grown from 16 in 2011 to 20 in 2015. Since ICNC 2011, the contents of the International Handbook of Evaluated Criticality Safety Benchmark Experiments have increased from 533 evaluations (approximately 58,000 pages), containing benchmark specifications for 4,552 critical or subcritical configurations to 561 evaluations (approximately 67,000 pages), containing benchmark specifications for 4,839 critical or subcritical configurations in the 2014 Edition of the ICSBEP Handbook. While the current number of criticality-alarm-placement/shielding benchmark configurations has remained at 24, the number of fundamental physics measurements relevant to criticality safety applications has increased from 200 to 207. Since ICNC 2011, the contents of the International Handbook of Evaluated Reactor Physics Benchmark Experiments have increased from 53 experimental series that were performed at 31 different reactor facilities to 143 experimental series that were performed at 50 different reactor facilities in the 2015 Edition of the IRPhEP Handbook. The status of the ICSBEP and the IRPhEP is discussed and benchmarks that have been more recently added to the handbooks are highlighted.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Benchmark Evaluation of Dounreay Prototype Fast Reactor Minor Actinide Depletion Measurements

Historic measurements of actinide samples in the Dounreay Prototype Fast Reactor (PFR) are of interest for modern nuclear data and simulation validation. Samples of various higher-actinide isotopes were irradiated for 492 effective full-power days and radiochemically assayed at Oak Ridge National Laboratory (ORNL) and Japan Atomic Energy Research Institute (JAERI). Limited data were available regarding the PFR irradiation; a six-group neutron spectra was available with some power history data to support a burnup depletion analysis validation study. Under the guidance of the Organisation for Economic Co-Operation and Development Nuclear Energy Agency (OECD NEA), the International Reactor Physics Experiment Evaluation Project (IRPhEP) and Spent Fuel Isotopic Composition (SFCOMPO) Project are collaborating to recover all measurement data pertaining to these measurements, including collaboration with the United Kingdom to obtain pertinent reactor physics design and operational history data. These activities will produce internationally peer-reviewed benchmark data to support validation of minor actinide cross section data and modern neutronic simulation of fast reactors with accompanying fuel cycle activities such as transportation, recycling, storage, and criticality safety.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Impact of Modifications in Continuous Recycling System for U/Pu and U/TRU with Fast and Thermal Reactors on Fuel-Cycle Performance and Waste Characterization

This paper studies the impact of modifications in continuous U/Pu and U/TRU recycling systems on fuel-cycle performance and waste characterization. The continuous recycling systems considered in this paper are multi-stage scenarios consisting of sodium-cooled fast reactors (SFR) and mixed-oxide (MOX) fueled pressurized water reactors (PWR). The study was performed by introducing a series of fuel-cycle cases that slightly differ in their fuel-cycle scenarios, core designs, and types of heavy metals (HM) recycled (Pu or TRU). The major exchange in the fuel-cycle scenarios considered here occurs in the PWR-MOX stage such that one set of cases recycles its discharged Pu/TRU and fabricate MOX fuel by topping Pu/TRU recovered from SFR blankets, and the other uses only recovered Pu/TRU from SFR blankets, and its discharged Pu/TRU is sent to SFR metallic fuel fabrication. All analyses performed in this paper also take into account 1% of annual growth in electric power production. Therefore, the design of reactor cores and fuel-cycle systems must be able to produce additional Pu/TRU in order to cover additional reactors built for the 1% growth in each year.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Statistical Analysis in Support of Control Fuel Temperature for the Advanced Gas Reactor Experiments

There are four completed tests among a series of Advanced Gas Reactor (AGR) irradiation experiments that are being conducted within the Advanced Reactor Technologies (ART) Fuel Development and Qualification Program. The main objectives of the fuel experimental campaigns are to provide the necessary data on fuel performance to support fuel process development, qualify a fuel design and fabrication process for normal operation and hypothetical accident conditions, and support development and validation of fuel performance and fission product transport models and codes.1 Since this tristructural isotropic fuel is qualified to be used in high temperature gas-cooled reactors (HTGRs), it is essential that the testing conditions of fuel irradiations are designed to mimic those in the conceptual gas reactors. As a result, temperatures of the AGR test fuel have to be maintained within a specified range that is typical of the HTGRs. In each AGR capsule, the test fuel temperature is controlled independently using a variable fraction neon and helium gas mixture that fills the gap volume for each capsule. The temperature as measured by a thermocouple (TC) terminated in the graphite holder (which houses fuel compacts) is used as a feedback for the gas flow controllers. Direct temperature measurement for fuel compacts is not possible because contact between a TC head and fuel particle could lead to undesirable failures. Therefore, an ABAQUS based finite element thermal model was created to predict the daily average fuel temperatures during the entire irradiation period.2,3 It is well-known that extended exposure to neutron irradiation and high temperatures leads to changes in geometries of fuel compacts and graphite holder, which result in changes of the gas gap leading to a change in the relationship between TC and fuel temperatures. In addition, TCs are also expected to drift and sometimes fail under irradiation conditions. These facts make maintaining test fuel temperature within a specified range challenging. This paper describes how the regression analysis methods are used in conjunction of calculated fuel temperatures from early irradiation periods (when irradiation effect is relatively small) to effectively control fuel temperature for the remaining irradiation, when irradiation effect became profound. This is particularly important in the event of all TCs in a capsule fail.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

TRANSIENT ANALYSES OF THE VERSATILE TEST REACTOR USING RELAP5-3D

The DOE Office of Nuclear Energy established the Versatile Test Reactor (VTR) program in February 2017 in response to reports outlining the need for a fast spectrum test reactor. In this framework, Idaho National Laboratory (INL), as the proposed hosting site for the VTR, is expanding its capabilities leveraging on its staff nuclear expertise on experiments design and reactors operation, codes development and assessment, advanced safety analyses. One of the key points of this strategy is to develop capabilities necessary to perform transient analyses in order to understand the plant behavior during upset conditions and appropriately quantify margins. In this context system thermal-hydraulic models of the VTR are being developed using the SAS4A/SASSYS-1 code developed at Argonne National Laboratory as well as the INL code RELAP5-3D. The former provides reference analyses whereas the latter could provide an independent assessment of the reference analyses if deemed necessary by the project. This paper focuses on the RELAP5-3D modeling, highlighting the different solutions adopted for correctly analyzing sodium fast reactor behavior. After showing the modeling activities, we present a set of reference transient analyses for protected and unprotected transients, highlighting the intrinsic safety characteristics of the VTR design.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

NSUF Research Reactor Fitness Workshop Results

The Nuclear Science User Facilities has been working with the National Organization of Test, Research, and Training Reactors to collect input from the US university reactor community on their current and future needs. The effort started with a web survey in March to collect initial data. NSUF held a follow-up workshop in July at the Center for Advanced Energy Studies in Idaho Falls. We identified four areas of interest in the university research reactor community that can affect future operations and sustainability of the reactors: capital infrastructure, regulatory burden, staffing and knowledge transfer, and utilization and relevancy. Members of the community presented challenges and best practices from their facilities. Discussions were held to create and distill the list of challenges for each area. A working group was formed for each area to continue the discussion and prioritize the challenges. NSUF and the working groups intend to create a final recommendations report by the end of the year. This panel session will discuss the workshop and summarize the discussion in the four topic areas. The audience is encouraged to add to the discussion.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

POST-IRRADIATION EXAMINATION AND SAFETY TESTING OF AGR-2 PEBBLE BED MODULAR REACTOR FUEL AT OAK RIDGE NATIONAL LABORATORY

A cooperative research and development agreement (CRADA) has been established between Pebble Bed Modular Reactor (PBMR) State Owned Company (SOC) Ltd. of the Republic of South Africa and Battelle Energy Alliance, LLC (BEA), the managing and operating contractor of Idaho National Laboratory (INL). The objective of this agreement is to conduct safety tests and destructive examination on compacts containing tristructural isotropic (TRISO) coated particle fuel that was provided by Necsa (for PBMR and title transferred to the Department of Energy [DOE]) and irradiated in the Advanced Test Reactor (ATR) in Capsule 4 as part of the Advanced Gas Reactor (AGR)-2 experiment performed under a previous CRADA (08-CR-12). The proposed safety testing and post-test examination will assist with assessing fuel performance during high-temperature reactor accident conditions. Due to different capabilities, some work will be performed by DOE’s Oak Ridge National Laboratory (ORNL) under a separate agreement. This statement of work describes the work scope to be performed by ORNL in support of the CRADA. Based on consultation with PBMR and ORNL, INL may modify the scope of work to be performed at ORNL. In particular, data from initial tests and examinations may result in scope modification as necessary.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2011

Neutron flux measurements in the side-core region of Hunterston B advanced gas-cooled reactor

The core restraints of advanced gas-cooled reactors are important structural components that are required to maintain the geometric integrity of the cores. A review of neutron dosimetry for the sister stations Hunterston B and Hinkley Point B identified that earlier conservative assessments predicted high thermal neutron dose rates to key components of the restraint structure (the restraint rod welds), with the implication that some of them may be predicted to fail during a seismic event. A revised assessment was therefore undertaken [Thornton, D. A., Allen, D. A., Tyrrell, R. J., Meese, T. C., Huggon, A.P., Whiley, G. S., and Mossop, J. R., 'A Dosimetry Assessment for the Core Restraint of an Advanced Gas Cooled Reactor,' Proceedings of the 13. International Symposium on Reactor Dosimetry (ISRD-13, May 2008), World Scientific, River Edge, NJ, 2009, W. Voorbraak, L. Debarberis, and P. D'hondt, Eds., pp. 679-687] using a detailed 3D model and a Monte Carlo radiation transport program, MCBEND. This reassessment resulted in more realistic fast and thermal neutron dose recommendations, the latter in particular being much lower than had been thought previously. It is now desirable to improve confidence in these predictions by providing direct validation of the MCBEND model through the use of neutron flux measurements. This paper describes the programme of work being undertaken to deploy two neutron flux measurement 'stringers' within the side-core region of one of the Hunterston B reactors for the purpose of validating the MCBEND model. The design of the stringers and the determination of the preferred deployment locations have been informed by the use of detailed MCBEND flux calculations. These computational studies represent a rare opportunity to design a flux measurement beforehand, with the clear intention of minimising the anticipated uncertainties and obtaining measurements that are known to be representative of the neutron fields to which the vulnerable steel restraint components are exposed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Neutron damage reduction in a traveling wave reactor

Traveling wave reactors are envisioned to run on depleted or natural uranium with no need for enrichment or reprocessing, and in a manner which requires little to no operator intervention. If feasible, this type of reactor has significant advantages over conventional nuclear power systems. However, a practical implementation of this concept is challenging as neutron irradiation levels many times greater than those in conventional reactors appear to be required for a fission wave to propagate. Radiation damage to the fuel and cladding materials presents a significant obstacle to a practical design. One possibility for reducing damage is to soften the neutron energy spectrum. Here we show that using a uranium oxide fuel form will allow a shift in the neutron spectrum that can result in at least a three fold decrease in dpa levels for fuel cladding and structural steels within the reactor compared with the dpa levels expected when using a uranium metal fuel. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Model biases in high-burnup fast reactor simulations

A new code system called the Advanced Reactor Modeling Interface (ARMI) has been developed that loosely couples multiscale, multiphysics nuclear reactor simulations to provide rapid, user-friendly, high-fidelity full systems analysis. Incorporating neutronic, thermal-hydraulic, safety/transient, fuel performance, core mechanical, and economic analyses, ARMI provides 'one-click' assessments of many multi-disciplined performance metrics and constraints that historically require iterations between many diverse experts. The capabilities of ARMI are implemented in this study to quantify neutronic biases of various modeling approximations typically made in fast reactor analysis at an equilibrium condition, after many repetitive shuffles. Sensitivities at equilibrium that result in very high discharge burnup are considered ( and >20% FIMA), as motivated by the development of the Traveling Wave Reactor. Model approximations discussed include homogenization, neutronic and depletion mesh resolution, thermal-hydraulic coupling, explicit control rod insertion, burnup-dependent cross sections, fission product model, burn chain truncation, and dynamic fuel performance. The sensitivities of these approximations on equilibrium discharge burnup, k{sub eff}, power density, delayed neutron fraction, and coolant temperature coefficient are discussed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Calculation of heating values for the high flux isotope reactor

Calculating the amount of energy released by a fission reaction (fission Q value) and the heating rate distribution in a nuclear reactor is an important part of the safety analysis. However, these calculations can become very complex. One of the codes that can be used for this type of analyses is the Monte Carlo transport code MCNP5. Currently it is impossible to calculate the Q value and heating rate disposition for delayed beta and delayed gamma particles directly from MCNP5. The purpose of this paper is to outline a rigorous method for indirectly calculating the Q values and heating rates in the High Flux Isotope Reactor (HFIR), based on previous similar studies carried out for very high-temperature reactor configurations. This method has been applied in this study to calculate heating rates for the beginning of cycle (BOC) and end-of-cycle (EOC) states of HFIR. In addition, the BOC results obtained for HFIR are compared with corresponding results for the Advanced Test Reactor. The fission Q value for HFIR was calculated as 200.2 MeV for the BOC and 201.3 MeV for the EOC. It was also determined that 95.1% and 95.4% of the heat was deposited within the HFIR fuel plates for the BOC and EOC models, respectively. This methodology can also be used for heating rate calculations for HFIR experiments. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Preliminary neutronics design of china lead-alloy cooled demonstration reactor (CLEAR-III) for nuclear waste transmutation

China Lead-Alloy cooled Demonstration Reactor (CLEAR-III), which is the concept of lead-bismuth cooled accelerator driven sub-critical reactor for nuclear waste transmutation, was proposed and designed by FDS team in China. In this study, preliminary neutronics design studies have primarily focused on three important performance parameters including Transmutation Support Ratio (TSR), effective multiplication factor and blanket thermal power. The constraint parameters, such as power peaking factor and initial TRU loading, were also considered. In the specific design, uranium-free metallic dispersion fuel of (TRU-Zr)-Zr was used as one of the CLEAR-III fuel types and the ratio between MA and Pu was adjusted to maximize transmutation ratio. In addition, three different fuel zones differing in the TRU fraction of the fuel were respectively employed for this subcritical reactor, and the zone sizes and TRU fractions were determined such that the linear powers of these zones were close to each other. The neutronics calculations and analyses were performed by using Multi-Functional 4D Neutronics Simulation System named VisualBUS and nuclear data library HENDL (Hybrid Evaluated Nuclear Data Library). In the preliminary design, the maximum TSRLLMA was {approx}11 and the blanket thermal power was {approx}1000 MW when the effective multiplication factor was 0.98. The results showed that good performance of transmutation could be achieved based on the subcritical reactor loaded with uranium-free fuel. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Development of strong-sense validation benchmarks for the fluoride salt-cooled high-temperature reactor

The Fluoride salt-cooled High-temperature Reactor (FHR) is a class of reactor concepts currently under development for the U. S. Dept. of Energy. The FHR is defined as a Generation IV reactor that features low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. Recent experimental work using simulant fluids have been performed to demonstrate key 'proof of principle' FHR concepts and have helped inform the reactor design process. An important element of developing FHR technology is to sufficiently validate the predictive accuracy of the computer codes used to model system response. This paper presents a set of thermal-hydraulics experiments, defined as Strong-Sense Benchmarks (SSB's), which will help establish the FHR validation domain for simulant fluid suitability. These SSB's are more specifically designed to investigate single-phase natural circulation which is the dominant mode of FHR decay heat removal during off-normal conditions. SSB s should be viewed as engineering reference standards and differ from traditional confirmatory experiments in the sense that they are more focused on fundamental physics as opposed to reproducing high levels of physical similarity with the prototypical design. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗