ASME Section III, Division 5, High Temperature Reactors High Temperature Materials
ASME Section III, Rules for Construction of Nuclear Facility Components - Division 5, High Temperature Reactors
22 GENERAL STUDIES OF NUCLEAR REACTORS↗
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Fuels and Materials Challenges for the Very High Temperature Reactor
High Temperature Gas-cooled Reactors (HTGRs) featuring particle fuel are a very mature technology. Commercial deployment occurred in the mid-1980's with the Fort St.Vrain and Thorium Hochtemperatur Reaktor by providing electricity to the grids in the United States and West Germany, respectively. The technology was then adopted by Japan and China with the operation of the High Temperature Test Reactor in Oarai, Japan and the High Temperature Reactor (HTR-10) in China. These commercial reactors typically operate with an outlet coolant temperature of 750 deg. C to 800 deg. C. Increasing the outlet temperature of the HTGR to even higher temperatures above 900 deg. C will improve the thermodynamic efficiency of the system and enable application of a new class of gas reactor, the very high temperature reactor (VHTR), to provide process heat, electricity, and hydrogen to a broad range of industrial customers with the attendant benefits of improved energy security and reduced CO{sub 2} emissions. However, the increase in coolant outlet temperature presents a number of technical challenges associated with fuels and materials. For over a decade, the U.S. Department of Energy has been sponsoring a broad program of research and development with a goal of addressing the technical fuels and materials challenges to support deployment of a VHTR that could safely and economically operate at outlet temperatures of 750 to 950 deg. C. In summary, fuels and materials challenges exist with deploying a VHTR with an outlet temperature of 950 deg. C in the areas of structural materials, graphite and TRISO fuel. However, R and D is underway or has recently been completed suggesting that technical solutions are possible. (authors)
11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
Transmutation Analysis of Enriched Uranium and Deep Burn High Temperature Reactors
High temperature reactors (HTRs) have been under consideration for production of electricity, process heat, and for destruction of transuranics for decades. As part of the transmutation analysis efforts within the Fuel Cycle Research and Development (FCR&D) campaign, a need was identified for detailed discharge isotopics from HTRs for use in the VISION code. A conventional HTR using enriched uranium in UCO fuel was modeled having discharge burnup of 120 GWd/MTiHM. Also, a deep burn HTR (DB-HTR) was modeled burning transuranic (TRU)-only TRU-O2 fuel to a discharge burnup of 648 GWd/MTiHM. For each of these cases, unit cell depletion calculations were performed with SCALE/TRITON. Unit cells were used to perform this analysis using SCALE 6.1. Because of the long mean free paths (and migration lengths) of neutrons in HTRs, using a unit cell to represent a whole core can be non-trivial. The sizes of these cells were first set by using Serpent calculations to match a spectral index between unit cell and whole core domains. In the case of the DB-HTR, the unit cell which was arrived at in this way conserved the ratio of fuel to moderator found in a single block of fuel. In the conventional HTR case, a larger moderator-to-fuel ratio than that of a single block was needed to simulate the whole core spectrum. Discharge isotopics (for 500 nuclides) and one-group cross-sections (for 1022 nuclides) were delivered to the transmutation analysis team. This report provides documentation for these calculations. In addition to the discharge isotopics, one-group cross-sections were provided for the full list of 1022 nuclides tracked in the transmutation library.
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Modeling of Prismatic High Temperature Reactors in Pronghorn
Pronghorn is a Multiphysics Object-Oriented Simulation Environment (MOOSE) based thermal-hydraulics code developed at Idaho National Laboratory (INL) for advanced reactor analysis. It has been previously applied to model Pebble-Bed High Temperature Reactors (High Temperature Reactor (HTR)s), Liquid-Metal Cooled Reactors, and Molten Salt Reactors, among others. This work applies the coarse-mesh thermal hydraulics capabilities in Pronghorn to model Prismatic-Core HTRs. In particular, the Oregon State University (OSU)’s High Temperature Test Facility (HTTF) is modeled with Pronghorn. The HTTF is a 1:4 height scaled-down facility of General Atomics’ Modular High Temperature Gas-cooled Reactor.
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Scaling and Design Analysis of a Sub-Scale, Reduced Temperature Stratified Flow Separate Effects Test Facility for the Experimental Investigation of Air Ingress in a High Temperature Reactor System
High temperature gas reactors (HTGR) are poised to spearhead the next generation of nuclear reactors, due to their passive safety characteristics and their applications flexibility. Able to provide base load power or process heat for industrial applications, HGTRs may be a critical component to reducing greenhouse gas emissions in the United States and abroad. However, increased interest and development has also highlighted areas requiring further investigation, one of which being the air-ingress accident. Air ingress may occur following a break in the coaxial cross duct and pressure vessel depressurization. The initial mode of ingress is hypothesized to be density driven cross flow as the cold air enters the lower plenum, at which point molecular diffusion and/or local buoyancy phenomena will dominate until a global natural circulation current is established. While computational simulations support this progression, little or no experimental data exists to validate and support this phenomenology. Therefore, it is necessary to generate such data; this research seeks to do so through the use of a sub-scale separate effects test facility - the Stratified Flow Separate Effects Test Facility (SFSETF). The goal of this facility is more specifically to focus on the transition to global natural circulation starting from air ingress after a double-ended guillotine break (DEGB) while neglecting depressurization. This paper will provide analyses performed in the design of this facility, including the geometric and hydraulic scaling analyses, safety analysis and instrumentation selection. (authors)
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Current status of the advanced high temperature reactor
The Advanced High Temperature Reactor (AHTR) is a design concept for a central station type [1500 MW(e)] Fluoride salt-cooled High-temperature Reactor (FHR) that is currently under development by Oak Ridge National Laboratory for the U. S. Dept. of Energy, Office of Nuclear Energy's Advanced Reactor Concepts program. FHRs, by definition, feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. The overall goal of the AHTR development program is to demonstrate the technical feasibility of FHRs as low-cost, large-size power producers while maintaining full passive safety. The AHTR design option exploration is a multidisciplinary design effort that combines core neutronic and fuel configuration evaluation with structural, thermal, and hydraulic analysis to produce a reactor and vessel concept and place it within a power generation station. The AHTR design remains at the notional level of maturity, as key technologies require further development and a logically complete integrated design has not been finalized. The present design space exploration, however, indicates that reasonable options exist for the AHTR core, primary heat transport path, and fuel cycle provided that materials and systems technologies develop as anticipated. (authors)
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Irradiation performance of AGR-1 high temperature reactor fuel
The AGR-1 experiment contained 72 low-enriched uranium oxide/uranium carbide TRISO coated particle fuel compacts in six capsules irradiated to burnups of 11.2 to 19.6% FIMA, with zero TRISO coating failures detected during the irradiation. The irradiation performance of the fuel including the extent of fission product release and the evolution of kernel and coating microstructures was evaluated based on detailed examination of the irradiation capsules, the fuel compacts, and individual particles. Fractional release of 110m Ag from the fuel compacts was often significant, with capsule-average values ranging from 0.01 to 0.38. Analysis of silver release from individual compacts indicated that it was primarily dependent on fuel temperature history. Europium and strontium were released in small amounts through intact coatings, but were found to be significantly retained in the outer pyrocarbon and compact matrix. The capsule-average fractional release from the compacts was 1 × 10 –4 to 5 × 10 –4 for 154 Eu and 8 × 10 –7 to 3 × 10 –5 for 90 Sr. The average 134 Cs fractional release from compacts was <3 × 10 –6 when all particles maintained intact SiC. An estimated four particles out of 2.98 × 10 5 in the experiment experienced partial cesium release due to SiC failure during the irradiation, driving 134 Cs fractional release in two capsules to approximately 10 –5 . Identification and characterization of these particles has provided unprecedented insight into the nature and causes of SiC coating failure in high-quality TRISO fuel. In general, changes in coating morphology were found to be dominated by the behavior of the buffer and inner pyrolytic carbon (IPyC), and infrequently observed SiC layer damage was usually related to cracks in the IPyC. Palladium attack of the SiC layer was relatively minor, except for the particles that released cesium during irradiation, where SiC corrosion was found adjacent to IPyC cracks. In conclusion, palladium, silver, and uranium were found in the SiC layer of irradiated particles, and characterization of these elements within the SiC microstructure is the subject of ongoing focused study.
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Phenomena identification and ranking table study for thermal hydraulics for Advanced High Temperature Reactor
The Advanced High-Temperature Reactor (AHTR) is a conceptual design of a Fluoride salt-cooled High-temperature Reactor (FHR) utilizing 7 LiF-BeF 2 (66-34 mol%) as its primary coolant. To identify key phenomena that impose potential challenges on thermal hydraulics modeling and simulation of such a reactor system, a thermal hydraulics phenomena identification and ranking table (TH-PIRT) study was performed for the AHTR in a Department of Energy Nuclear Energy University Program (NEUP) integrated research project led by Georgia Institute of Technology. A panel of experts from regulators, industries, national laboratories, and academia was assembled for the study. In this paper, the TH-PIRTs identified by the panel for two events, namely, station blackout and simultaneous withdrawal of all control rods, are summarized and discussed in detail. In addition, the key phenomena that warrant further study and research for AHTR analysis are identified to support the validation of thermal hydraulics system-level analysis codes and computational fluid dynamics simulation tools, as well as future FHR reactor licensing.
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Linear variable differential transformer (LVDT)-based elongation measurements in Advanced Test Reactor high temperature irradiation testing
New materials are being considered for fuel, cladding and structures in next generation and existing nuclear reactors. These materials can undergo significant dimensional and physical changes during high temperature irradiations. Currently, such changes are determined by repeatedly irradiating a specimen for a specified period of time in the Advanced Test Reactor (ATR) and then removing it from the reactor for evaluation. The labor and time to remove, examine and return irradiated samples for each measurement make this approach very expensive. In addition, such techniques provide limited data and may disturb the phenomena of interest. To resolve these issues, an instrumented creep testing capability is being developed for specimens irradiated under pressurized water reactor coolant conditions in the ATR at the Idaho National Laboratory (INL). This paper reports the status of INL efforts to develop this testing capability. In addition to providing an overview of in-pile creep test capabilities available at other test reactors, this paper focuses on efforts to design and evaluate a prototype test rig in an autoclave at INL’s High Temperature Test Laboratory.
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An Experimental Test Facility to Support Development of the Fluoride Salt Cooled High Temperature Reactor
The need for high-temperature (greater than 600 C) energy exchange and delivery systems is significantly increasing as the world strives to improve energy efficiency and develop alternatives to petroleum-based fuels. Liquid fluoride salts are one of the few energy transport fluids that have the capability of operating at high temperatures in combination with low system pressures. The Fluoride Salt-Cooled High-Temperature Reactor design uses fluoride salt to remove core heat and interface with a power conversion system. Although a significant amount of experimentation has been performed with these salts, specific aspects of this reactor concept will require experimental confirmation during the development process. The experimental facility described here has been constructed to support the development of the Fluoride Salt Cooled High Temperature Reactor concept. The facility is capable of operating at up to 700 C and incorporates a centrifugal pump to circulate FLiNaK salt through a removable test section. A unique inductive heating technique is used to apply heat to the test section, allowing heat transfer testing to be performed. An air-cooled heat exchanger removes added heat. Supporting loop infrastructure includes a pressure control system; trace heating system; and a complement of instrumentation to measure salt flow, temperatures, and pressures around the loop. The initial experiment is aimed at measuring fluoride salt heat transfer inside a heated pebble bed similar to that used for the core of the pebble bed advanced high-temperature reactor. This document describes the details of the loop design, auxiliary systems used to support the facility, the inductive heating system, and facility capabilities.
fluoride salt reactor FLiNaK↗
Phenomena Identification and Ranking Table (PIRT) study for metallic structural materials for advanced High-Temperature reactor
The Fluoride High-Temperature Reactor (FHR) technology promises many benefits including passive safety, proliferation-resistant waste forms, and improved economics. However, selection of reliable structural materials and identification of the possible degradation mechanisms for these is important for the licensure and the safe operation of FHRs. In order to address this task, the Georgia Tech led Integrated Research Project (IRP) hosted a Phenomena Identification and Ranking Table (PIRT) panel of experts to address degradation mechanisms and other materials related issues of importance to the FHRs. Futhermore, materials, ones that come in contact with FLiBe or FLiNaK molten salts or other related environments like high temperature steam etc., were considered in this PIRT. Focus of this PIRT was the metallic alloys, especially the ones that are permitted for the construction of elevated temperature Class A components by the ASME code. Degradation mechanisms considered in this PIRT included chemical degradation, mechanical degradation, radiation degradation, and synergistic effect of these mechanisms that may negatively impact operations or cause some safety concerns for the major structural components of FHRs. Main components which were considered included vessel and primary piping, primary heat exchangers, steam generator vessel, steam generator tubes, intermediate loop piping, valves and pumps. Welds in all structural components were identified as an important class of material, which varies in composition and properties, and needs more attention. Importance of impurity control in molten fluorides considered for FHR was highlighted throughout PIRT panel discussions. As a result, this paper gives a summary of important results from the PIRT panel discussions and report.
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Experimentally Validated Numerical Models of Non-Isothermal Turbulent Mixing in High Temperature Reactors
In the Very High Temperature Reactor (VHTR), the non-uniform heat generation and flow maldistribution of coolant in the core generates conditions leading into the lower plenum with many jets at different temperatures and velocities. The hope is that the combined flow becomes fully mixed in a short distance so that the fluctuating temperatures in the mixing region do not cause unwanted thermal stresses in the supporting structures. These cyclic thermal stresses can cause degradation of the wall microstructure including crack initiation and growth, ultimately leading to the fatigue failure in the lower plenum, and can be accelerated by the high temperature environment in the VHTR. The focus of this project was to provide experimental data for future validation of computational efforts aimed at capturing the turbulent mixing in the lower plenum. Computational fluid dynamics (CFD) is also used as part of this effort to assess the requirements to model fundamental flows with application to the lower plenum as well at the applied flow conditions that mimic the experimental facilities. These efforts will lay the groundwork for future incorporation of stress analysis as well as fatigue and failure studies for multiple VHTR components. An additional emphasis is placed on reduced order models (ROM) using canonical flows, which will help to balance computational expense and accuracy and enable future work in uncertainty quantification. The flow in the lower plenum is extremely complex with hundreds of cooling jets entering and then being forced to change directions by 90° while navigating through an array of cylindrical support posts. Generating experimental data for such a complicated flow is difficult, not to mention that the computational requirements to generate an adequate temporal and spatial sampling is extremely expensive and time consuming. Therefore, the ultimate focus of this research is a “unit cell” which consists of six jets in the presence of a crossflow, and seven cylindrical support posts. It is scaled to achieve similar ranges of key dimensionless numbers. This serves to mimic flow conditions in different regions of the lower plenum by adjusting velocities, temperatures, and orientations, and represents a compromise between the complexity inside the full-scale application and the ideal conditions of a more fundamental study. In addition to the unit cell experiments and simulations, the complex flows in the VHTR lower plenum are broken down into a number of simplified (in some cases even fundamental) flows, two of which are turbulent jets and flow past cylinders. Each of these are studied, computationally and experimentally for isothermal and non-isothermal conditions. Flows where trusted experimental data is available in the literature (e.g., turbulent free jet), additional experiments are not needed. The result of these efforts is a database of experimental data and multiple computational studies, which provide an outlook to the requirements, needed to capture turbulent mixing flow physics The ROM results provide assessments for appropriate resource-light models that balance accuracy with reduced computational burdens.
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Selected Profiles of Burnup and Isotopic Concentrations in Pebble Bed High Temperature Reactor Fuel
Pebble Bed High Temperature Reactors (PB-HTRs) feature a unique fuel loading, burnup, and discharge process that is markedly different that a light water reactor or any other batch-loaded reactor that uses fixed fuel elements. During PB-HTR operation, small spherical fuel elements are dropped onto the top of the bed at small and regular intervals, 'trickle' through the core, and pass out of the core via a discharge chute at the bottom. Partially burnt pebbles are circulated back to the top of the core for another pass through the flux field. The process continues until the pebble achieves a specified discharge burnup. This results in high fuel utilization relative to a comparable batch-loaded fuel assembly which may or may not achieve full burnup depending upon where it is placed in the core. On the other hand, the radial and axial variation in the flux field and the random placement of each element on top of the bed mean that there will be a distribution of the final burnup and isotopic concentrations in spent fuel. This leads to some uncertainty in burnup analysis that has consequences for burnup monitoring (and thus efficient operation), source term quantification, and material accountability. The PEBBED code was developed specifically for the design and analysis of PB-HTRs. Although it is still under development, the code captures the important physics of cores fueled by dropping in, and if so configured, recirculating small graphite spheres in which the Tri-Isotropic (TRISO) coated particle fuel is embedded. In this study, the capabilities of the code for computing the isotopic contents of PBR fuel were exploited to examine and compare the variability of selected isotopes, in this case U-235, Pu-239, and Cs-137, in the fuel discharged from different PBR concepts. Results indicate the sensitivity of certain isotopes to the core geometry, mainly the presence of an inner reflector. (authors)
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Advanced Reactor Technologies: Very High Temperature Reactor Research and Development Quarterly Report
Major accomplishments for the months of January, February and March 2020 in the fields of Fuels Development Qualification, High Temperature Materials Development, Graphite and GRC Modeling and Validation.
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Advanced Reactor Technologies: Very High Temperature Reactor Research and Development Quarterly Report April, May, June 2020
Report of activity for the Fiscal Year 2020 third quarter
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Secondary Heat Exchanger Design and Comparison for Advanced High Temperature Reactor
The goals of next generation nuclear reactors, such as the high temperature gas-cooled reactor and advance high temperature reactor (AHTR), are to increase energy efficiency in the production of electricity and provide high temperature heat for industrial processes. The efficient transfer of energy for industrial applications depends on the ability to incorporate effective heat exchangers between the nuclear heat transport system and the industrial process heat transport system. The need for efficiency, compactness, and safety challenge the boundaries of existing heat exchanger technology, giving rise to the following study. Various studies have been performed in attempts to update the secondary heat exchanger that is downstream of the primary heat exchanger, mostly because its performance is strongly tied to the ability to employ more efficient conversion cycles, such as the Rankine super critical and subcritical cycles. This study considers two different types of heat exchangers—helical coiled heat exchanger and printed circuit heat exchanger—as possible options for the AHTR secondary heat exchangers with the following three different options: (1) A single heat exchanger transfers all the heat (3,400 MW(t)) from the intermediate heat transfer loop to the power conversion system or process plants; (2) Two heat exchangers share heat to transfer total heat of 3,400 MW(t) from the intermediate heat transfer loop to the power conversion system or process plants, each exchanger transfers 1,700 MW(t) with a parallel configuration; and (3) Three heat exchangers share heat to transfer total heat of 3,400 MW(t) from the intermediate heat transfer loop to the power conversion system or process plants. Each heat exchanger transfers 1,130 MW(t) with a parallel configuration. A preliminary cost comparison will be provided for all different cases along with challenges and recommendations.
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Systems and methods for enhancing isolation of high-temperature reactor containments
A high-temperature containment-isolation system for transferring heat from a nuclear reactor containment to a high-pressure heat exchanger is presented. The system uses a high-temperature, low-volatility liquid coolant such as a molten salt or a liquid metal, where the coolant flow path provides liquid free surfaces a short distance from the containment penetrations for the reactor hot-leg and the cold-leg, where these liquid free surfaces have a cover gas maintained at a nearly constant pressure and thus prevent high-pressures from being transmitted into the reactor containment, and where the reactor vessel is suspended within a reactor cavity with a plurality of refractory insulator blocks disposed between an actively cooled inner cavity liner and the reactor vessel.
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