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All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

A Versatile Coupled Test Reactor Concept

A Versatile Coupled Test Reactor (VCTR) is a highly reconfigurable, sodium-cooled, coupled fast and thermal spectrum test reactor, which provides (1) a fast flux irradiation environment prototypical of potential fast reactor designs, (2) thermal and epithermal flux irradiation environments complementary to those of ATR and HFIR, and (3) other possibilities, including beam tubes for scientific experiments, irradiation vehicles for isotope production, etc. The ongoing design of a new versatile coupled thermal-fast test reactor at INL that can accommodate both fast and thermal irradiation, technology and safety features tests is described in this paper. A coupled reactor is defined as a reactor with two distinct spectral zones (Fast and Thermal), which are neutronically coupled to each other: some neutrons born in zone F cause fission in zone T and vice-versa. Only fast neutrons are allowed to go from one zone to the other, which is assured by a thermal neutron filter. A distinction of coupling can be made depending on the main purpose of using fuel assemblies with some level of neutron moderation (thermal zone) together with fuel assemblies with as little moderation as possible (fast zone). It can be used mainly to minimize the core power and fissile inventory, or it can be used mainly to provide additional control of the irradiation conditions in the thermal zone. The steady-state neutronics calculations were performed using the three-dimensional continuous-energy Monte Carlo reactor physics burnup calculation code Serpent2. Although, a few changes were necessary to support the Serpent code in calculating the coupling coefficients. Steady state and transient thermal analyses were performed with RELAP5-3D. RELAP5-3D is the latest in the RELAP5 code series developed at INL. One of the core configurations using only low enriched uranium fuel could provide both a high fast neutron flux and a high thermal neutron flux—respectively 3.5E15 n/s.cm2 and 1E15 n/s.cm2—in large volumes while maintaining core power below 300 MW. The use of plutonium in the fast zone provides more flexibility in fuel assembly design than when low enriched uranium is used, the designer has more degrees of freedom to work with; in particular it could provide additional control of the irradiation conditions in the thermal zone. The results showed that the current design fulfills most R&D needs requirements providing high thermal and fast fluxes and wide possibilities for diversified future reactor features test in appropriate loops, both in steady state and during transients.

coupled spectrum↗
DOE OSTI2016

VCTR: A Versatile Coupled Test Reactor Concept

The United States has currently no domestic fast neutron test and irradiation capabilities necessary to pursue a strong research and development program related to fast neutron reactors. This lack prevents the testing of fast reactor fuels in a prototypic environment, the testing of new safety features and of new coolants. A new irradiation test reactor is a costly investment that will operate for several decades and the initial specifications are crucial to ensure the relevance of such a facility in the long run. This new test reactor should be highly reconfigurable and allow operation as a fast test reactor or a thermal test reactor or a cou-pled fast-thermal test reactor. This report presents the on-going effort at Idaho National Laboratory (INL) to design a versatile coupled test reactor (VCTR—pronounced Victor) that could fulfill the R&D needs requiring high fast neutron fluxes but also high thermal neutron fluxes in configurations that could not be met in current thermal neutron test reactors such as ATR. The objective of this report is to summarize the efforts undertaken at INL from May 2015 to March 2016 aimed at understanding the main options for designing and building a versatile test reactor, with a primary mission to provide thermal and fast neutrons for irradiation of fuels and materials in a manner complementary to the ATR and HFIR reactors. Other missions can also be considered, including the validation of computer codes and the development of specific reactor technologies.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

Evaluation of Concepts for Mulitiple Application Thermal Reactor for Irradiation eXperiments (MATRIX)

The Advanced Test Reactor (ATR) is a high power density test reactor specializing in fuel and materials irradiation. For more than 45 years, the ATR has provided irradiations of materials and fuels testing along with radioisotope production. Originally operated primarily in support of the Offcie of Naval Reactors (NR), the mission has gradually expanded to cater to other customers, such as the DOE Office of Nuclear Energy (NE), private industry, and universities. Unforeseen circumstances may lead to the decommissioning of ATR, thus leaving the U.S. Government without a large-scale materials irradiation capability to meet the needs of its nuclear energy and naval reactor missions. In anticipation of this possibility, work was performed under the Laboratory Directed Research and Development (LDRD) program to investigate test reactor concepts that could satisfy the current missions of the ATR along with an expanded set of secondary missions. This work can be viewed as an update to a project from the 1990’s called the Broad Application Test Reactor (BATR). In FY 2012, a survey of anticipated customer needs was performed, followed by analysis of the original BATR concepts with fuel changed to low-enriched uranium. Departing from these original BATR designs, four concepts were identified for further analysis in FY2013. The project informally adopted the acronym MATRIX (Multiple-Application Thermal Reactor for Irradiation eXperiments). This report discusses analysis of the four MATRIX concepts along with a number of variations on these main concepts. Designs were evaluated based on their satisfaction of anticipated customer requirements and the “Cylindrical” variant was selected for further analysis of options. This downselection should be considered preliminary and the backup alternatives should include the other three main designs. The baseline Cylindrical MATRIX design is expected to be capable of higher burnup than the ATR (or longer cycle length given a particular batch scheme). The volume of test space in IPTs is larger in MATRIX than in ATR with comparable magnitude of neutron flux. In addition to the IPTs, the Cylindrical MATRIX concept features test spaces at the centers of fuel assemblies where very high fast flux can be achieved. This magnitude of fast flux is similar to that achieved in the ATR A-positions, however, the available volume having these conditions is greater in the MATRIX design than in the ATR. From the analyses performed in this work, it appears that the Cylindrical MATRIX design can be designed to meet the anticipated needs of the ATR replacement reactor. However, this statement must be qualified by acknowledging that this design is quite immature, and therefore any requirements currently met must be re-evaluated as the design matures. Also, some of the requirements were not strictly met, but are believed to be achievable once features to be added later are designed.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Assessment of Low Enriched Uranium Fueled Core Configurations for the Versatile Test Reactor

The Versatile Test Reactor (VTR) Program was initiated in FY-2017. This multi-laboratory effort supported by the U.S. Department of Energy has the long-term objective of completing a Versatile Test Reactor concept with associated cost and schedule estimate that enables a fully informed DOE decision within 3 years, to have a fully operating VTR by 2026 if acquisition is authorized, and to enable accelerated testing of advanced reactor fuels and materials required for advanced reactor technologies. The VTR concept uses sodium as coolant, features a core power of 300 MWth, and employs reactor-grade plutonium as fissile material in a U-Pu-10Zr fuel alloy. The reference core layout provides 10 assembly locations for high-flux irradiation testing with peak fast flux in excess of 4x1015 n/cm2-s. In order to establish a comparison point with other types of fuel, alternative core configurations were investigated using Low Enriched Uranium (LEU) fuel. The objective of this study was to modify the reference VTR core such as to use LEU fuel and assess the performance tradeoff, while retaining the same core envelope.

Stauff, Nicolas E.↗
DOE OSTI2016

Advanced Reactor Fuels Irradiation Experiment Design Objectives

This report summarizes the objectives of the current irradiation testing activities being undertaken by the Advanced Fuels Campaign relative to supporting the development and demonstration of innovative design features for metallic fuels in order to realize reliable performance to ultra-high burnups. The AFC-3 and AFC-4 test series are nearing completion; the experiments in this test series that have been completed or are in progress are reviewed and the objectives and test matrices for the final experiments in these two series are defined. The objectives, testing strategy, and test parameters associated with a future AFC test series, AFC-5, are documented. Finally, the future intersections and/or synergies of the AFC irradiation testing program with those of the TREAT transient testing program, emerging needs of proposed Versatile Test Reactor concepts, and the Joint Fuel Cycle Study program’s Integrated Recycle Test are discussed.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Assessment of the Technical Maturity of Generation IV Concepts for Test or Demonstration Reactor Applications, Revision 2

The United States Department of Energy (DOE) commissioned a study the suitability of different advanced reactor concepts to support materials irradiations (i.e. a test reactor) or to demonstrate an advanced power plant/fuel cycle concept (demonstration reactor). As part of the study, an assessment of the technical maturity of the individual concepts was undertaken to see which, if any, can support near-term deployment. A Working Group composed of the authors of this document performed the maturity assessment using the Technical Readiness Levels as defined in DOE’s Technology Readiness Guide . One representative design was selected for assessment from of each of the six Generation-IV reactor types: gas-cooled fast reactor (GFR), lead-cooled fast reactor (LFR), molten salt reactor (MSR), supercritical water-cooled reactor (SCWR), sodium-cooled fast reactor (SFR), and very high temperature reactor (VHTR). Background information was obtained from previous detailed evaluations such as the Generation-IV Roadmap but other technical references were also used including consultations with concept proponents and subject matter experts. Outside of Generation IV activity in which the US is a party, non-U.S. experience or data sources were generally not factored into the evaluations as one cannot assume that this data is easily available or of sufficient quality to be used for licensing a US facility. The Working Group established the scope of the assessment (which systems and subsystems needed to be considered), adapted a specific technology readiness scale, and scored each system through discussions designed to achieve internal consistency across concepts. In general, the Working Group sought to determine which of the reactor options have sufficient maturity to serve either the test or demonstration reactor missions.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Design Studies for a Multiple Application Thermal Reactor for Irradiation Experiments (MATRIX)

The Advanced Test Reactor (ATR) is a high power density test reactor specializing in fuel and materials irradiation. For more than 45 years, the ATR has provided irradiations of materials and fuels testing along with radioisotope production. Should unforeseen circumstances lead to the decommissioning of ATR, the U.S. Government would be left without a large-scale materials irradiation capability to meet the needs of its nuclear energy and naval reactor missions. In anticipation of this possibility, work was performed under the Laboratory Directed Research and Development (LDRD) program to investigate test reactor concepts that could satisfy the current missions of the ATR along with an expanded set of secondary missions. A survey was conducted in order to catalogue the anticipated needs of potential customers. Then, concepts were evaluated to fill the role for this reactor, dubbed the Multi-Application Thermal Reactor Irradiation eXperiments (MATRIX). The baseline MATRIX design is expected to be capable of longer cycle lengths than ATR given a particular batch scheme. The volume of test space in In-Pile-Tubes (IPTs) is larger in MATRIX than in ATR with comparable magnitude of neutron flux. Furthermore, MATRIX has more locations of greater volume having high fast neutron flux than ATR. From the analyses performed in this work, it appears that the lead MATRIX design can be designed to meet the anticipated needs of the ATR replacement reactor. However, this design is quite immature, and therefore any requirements currently met must be re-evaluated as the design is developed further.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Comparison of Subdivided Versus Full-Length Fuel Pin Burnup in the ATF-2 Safety Analysis

The Accident Tolerant Fuels (ATF) campaign, under the Fuel Cycle Research and Development (FCRD) Advanced Fuels program, was developed after the Fukushima Daiichi Nuclear Power Plant disaster in 2011. Under ATF, fuel and cladding concepts are to be developed and tested that may more readily withstand an accident similar to the one that occurred at Fukushima Daiichi. ATF is a combination of industry, universities, and national laboratories in an effort to develop advanced accident tolerant cladding and fuel concepts. The end goal of the ATF program is to insert a lead test rod (LTR) or lead test assembly (LTA) into an operating commercial power light water reactor by 2022. In order to meet this 2022 insertion goal, ATF is split into three separate phases: ATF-1, ATF-2, and ATF-3. The first phase of the ATF program, ATF-1, is the concept feasibility study and early irradiation testing. Feasibility, fabrication, and out-of-pile testing are happening at universities and laboratories all over the United States. Irradiation testing of these concepts is happening at the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). Early irradiation testing involves testing clad and/or fuel in drop-in capsules in the ATR. Currently, these capsules are in the relatively low-flux positions in the ATR, but some newer concepts and concepts that require higher burnup will be moved to the higher-flux positions later on during irradiation. Currently, Oak Ridge National Laboratory (ORNL), Westinghouse Electric Company, Areva, and General Electric (GE) have inserted capsules for ATF-1, although more concepts are planned. The second phase of ATF, ATF-2, is the focus of this paper. ATF-2 is the loop testing of fuel pins with new cladding and fuel concepts to simulate pressurized water reactor (PWR) and boiling water reactor (BWR) environments. ATF-2 will concurrently be happening at the ATR and at the Halden Reactor in Norway. The ATF-2 irradiation at the ATR will only simulate a PWR environment. ATF-2 will be inserted into the ATR pressurized water loop, Loop 2A, in the Center Flux Trap (CFT), where chemistry, temperature, and pressure can all be controlled separately from the primary coolant system and other loops in ATR. Loop 2A will be borated, pressurized to 2200 psi nominal, and have an average water temperature of 300 deg. C. A borated loop with fueled rodlets is extremely rare for Nuclear Energy experiments in the ATR, so ATF-2 will demonstrate new and unique capabilities for INL. ATF-3 is the testing of fresh and irradiated ATF concepts at the Transient Reactor Test (TREAT) Facility. The TREAT Reactor is planned for restart, re-establishing U.S. transient testing capabilities after nearly 20 years of non-operations. TREAT will test ATF concepts under transient conditions, such as a reactivity insertion accident. In addition to fresh fuel pins, TREAT will use from ATF2 and do transient testing on those irradiated pins. The information TREAT will provide to the vendors to support fuel licensing will be valuable, showing that these concepts do enhance the accident tolerance of the fuel in commercial reactors. (author)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Development of The DOME Shield Model For The NRIC Virtual Test Bed

As several advanced reactor concepts are maturing, test beds are needed to accelerate the demonstration and deployment of these advanced nuclear technologies. The National Reactor Innovation Center (NRIC) is building new or enhancing existing US Department of Energy infrastructure to support testing of components and systems. Demonstration of Microreactor Experiments (DOME) will utilize the Experimental Breeder Reactor-II (EBR-II) dome containment structure to host reactor demonstrations. A reactor supplemental shielding is needed so that DOME dose requirements are met. To accelerate the confirmatory analysis required for the reactor demonstration, the NRIC Virtual Test Bed (VTB) is developing a virtual model of the DOME shield that will be made available on the VTB public repository. This will allow developers to leverage advanced modeling and simulation tools to ensure their reactor demonstration concept will meet dose requirement and the limit concrete temperature in the shield during steady state and transient operation conditions. This paper presents the model developed for the DOME shield using open-source tools: MOOSE heat transfer module, Monte Carlo code OpenMC, and Cardinal to calculate the DOME shield temperature distribution during steady state

22 - GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

ATF-2 Ramp Conceptual Design Report

The following report documents the conceptual design for the ATF-2 Ramp experiment. The experiment aims to fill the in-pile irradiation testing gap to conduct integral ramp testing, which was created by the closure of R2, Osiris, and Halden test reactors. The concept involves the simultaneous ramping of three fuel pins using a power axial locator mechanism in the Loop-2A testing facility in the center flux trap of the advanced test reactor (ATR). The three fuel pins are ramped in individual coolant channels containing a prototypic PWR environment. Rod failure will be detected using a fuel rod elongation sensor attached to the fuel pin upper end cap. An axial stack of concentric hafnium and zirconium shrouds will be used to shape the flux around the test pins to create different power levels in each pin and to ensure the peak power location of each pin remains in the center of the test rod. Monte Carlo simulations are used to demonstrate the viability of this design concept.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Graphite Moderation in Fast Reactors

With the renewed interest in designing and building a fast test reactor in the U.S., several new reactor concepts have been emerging. Although the main mission of a fast test reactor would be to enable high fast flux irradiation capability, there could be added value for such a fast test reactor to also offer thermal neutron irradiation capabilities. As such, several of these concepts considered the use of a moderator material in a Sodium-cooled Fast Reactor (SFR), which is claimed to also offer benefits with respect of mass of fuel required. The work discussed in this paper focuses on exploring and assessing the benefits and challenges resulting from making use of thermalized neutrons to reduce the critical fuel mass. The analyses are performed for a graphite moderating material, using a simple core model to easily identify the phenomena observed. It is found that it is necessary to reduce the fuel enrichment near the moderated regions in order to avoid unacceptable power peaking. In the end, use of moderating materials outside a SFR core does not provide significant benefit in term of fuel volume reduction, but allows for a noticeable reduction of the mass of fissile material needed. However, it also implies a number of additional challenges which are highlighted in this paper.

core optimization↗
DOE OSTI2022

Multi-Modal Characterization of Nuclear Fuels and Materials at the Idaho National Laboratory Materials & Fuels Complex

The Idaho National Laboratory (INL) leads cutting-edge research pertaining to the advancement of nuclear reactor technologies, including nuclear fuels and materials. The INL Irradiated Materials Characterization Laboratory (IMCL), Electron Microscopy Laboratory (EML), and future Sample Preparation Laboratory (SPL) are available to the nuclear research community to assess the behavior of nuclear fuels and materials, efficiently and comprehensively characterizing from the engineering to atomistic scale. The IMCL is a unique, 12,000-square-foot facility located at the INL Materials and Fuels Complex designed for analysis of irradiated materials. The facility operates advanced characterization instruments that are sensitive to vibration, temperature, and electromagnetic interference in modular radiological shielding and confinement systems, granting researchers the ability to assess the microstructural, chemical, mechanical and thermophysical properties of nuclear materials, especially irradiated fuels. The EML is dedicated to advanced characterization of materials with optical and electron microscopy tools, including scanning electron microscopy/focused-ion beam (SEM/FIB) and transmission electron microscopy (TEM). Upon construction, the SPL will be a 3 story, 49,000 sq. ft facility, that is the most modern reactor structural materials testing and analysis facility in the world, designed to investigate reactor structural materials in support of life-extension programs and development of advanced reactor concepts, including mechanical testing and advanced characterization capabilities. This presentation will showcase some of the main capabilities available at both IMCL, EML, and SPL, specifically illustrating how these characterization techniques are incorporated into multi-modal characterization work scopes to elucidate the degradation of nuclear structural materials and irradiated fuels.

36 MATERIALS SCIENCE↗
DOE OSTI2017

A Summary of the Department of Energy's Advanced Demonstration and Test Reactor Options Study

An assessment of advanced reactor technology options was conducted to provide a sound comparative technical context for future decisions by the Department of Energy (DOE) concerning these technologies. Strategic objectives were established that span a wide variety of important missions and advanced reactor technology needs were identified based on recent Department of Energy and international studies. A broad team of stakeholders from industry, academia and government was assembled to develop a comprehensive set of goals, criteria and metrics to evaluate advanced irradiation test and demonstration reactor concepts. Point designs of a select number of concepts were commissioned to provide a deeper technical basis for evaluation. The technology options were compared on the bases of technical readiness and the ability to meet the different strategic objectives. Using the study’s evaluation criteria and metrics, an independent group of exerts from industry, universities and national labs scored each of the point designs. Pathways to deployment for concepts of varying technical maturity were estimated for the different demonstration systems with regard to cost, schedule and possible licensing approaches. Furthermore, this study also presents the tradeoffs that exist among the different irradiation test reactor options in terms of the ability to conduct irradiations in support of advanced reactor research and development (R&D) and to serve potential secondary missions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

A Summary of the Department of Energy’s Advanced Demonstration and Test Reactor Options Study

An assessment of advanced reactor technology options was conducted to provide a sound comparative technical context for future decisions by the U.S. Department of Energy (DOE) concerning these technologies. Strategic objectives were established that span a wide variety of important missions, and advanced reactor technology needs were identified based on recent DOE and international studies. A broad team of stakeholders from industry, academia, and government was assembled to develop a comprehensive set of goals, criteria, and metrics to evaluate advanced irradiation test and demonstration reactor concepts. Point designs of a select number of concepts were commissioned to provide a deeper technical basis for evaluation. The technology options were compared on the bases of technical readiness and the ability to meet the different strategic objectives. Using the study's evaluation criteria and metrics, an independent group of experts from industry, universities, and national laboratories scored each of the point designs. Pathways to deployment for concepts of varying technical maturities were estimated for the different demonstration systems with regard to cost, schedule, and possible licensing approaches. This study also presents the trade-offs that exist among the different irradiation test reactor options in terms of the ability to conduct irradiations in support of advanced reactor research and development and to serve potential secondary missions. The main findings of the study indicate the following: (1) for industrial process heat supply, a high-temperature gas-cooled reactor is the best choice because of the high outlet temperature of the reactor and its strong passive and inherent safety characteristics; (2) for resource utilization and waste management, a sodium-cooled fast reactor (SFR) is best because of the use of a fast flux to destroy actinides; (3) to realize the advantages of a promising but less-mature technology, a fluoride salt-cooled high-temperature reactor and a lead-cooled fast reactor fare about the same; (4) for fulfilling the needs of a materials test reactor, a SFR is considered best because of its ability to produce high fast flux, incorporate test loops, and provide additional large volumes for testing.

Advanced reactors↗
DOE OSTI2023

HydroGEN Seedling: High-Temperature Reactor Catalyst Material Development for Low-Cost and Efficient Solar-Driven Sulfur-Based Processes

The HyS process, driven by solar power, has great potential to reach high-efficiency and low-cost hydrogen production without greenhouse gas emissions. The high-temperature section of the HyS cycle, which operates the catalytic decomposition of sulfuric acid into sulfur dioxide, oxygen, and water, is a fundamental part of the cycle affecting the overall plant efficiency and cost. Therefore, a high-performance catalyst (i.e., low cost, high catalytic activity, and low degradation catalyst) is of critical importance to achieve high efficiency and low hydrogen cost. Research and development has highlighted that a Pt-based monometallic catalyst had unacceptable catalytic activity and performance degradation for a high-efficiency and low-cost hydrogen production process. A high-efficiency solar receiver-reactor system, which incorporates the new catalyst, also needs to be developed to achieve the required plant efficiency and cost. Greenway Energy (GWE) and the University of South Carolina (USC), partnering with HydroGEN node laboratories Idaho National Laboratory (INL), Savannah River National Laboratory (SRNL), and National Renewable Energy Laboratory (NREL), propose the development of a new catalyst formulation, included in a novel solar receiver-reactor concept, to be tested experimentally in the last part of the project.

08 HYDROGEN↗
DOE OSTI2023

Draft Summary Report on Irradiated Fuel Handling and Management for LOTUS

The National Reactor Innovation Center (NRIC) has conceptualized the design of the Laboratory for Operation and Testing in the United States (LOTUS) test bed to provide the United States Department of Energy (DOE) with the infrastructure necessary to make advanced reactor designs available for commercial developers. LOTUS will provide a test bed to developers with the capabilities of supporting a wide range of experiment design possibilities. Upon completion of the developers’ operations and experiments within the test bed, the irradiated fuel, reactor components, and other experiment materials must be removed. Idaho National Laboratory (INL) possesses significant capabilities for radioactive material handling such as casks, carts, and forklifts. However, given the unique environment presented by the NRIC-LOTUS test bed, located inside the Zero Power Physics Reactor (ZPPR) Perimeter Intrusion Detection and Assessment System (PIDAS) area at the Materials and Fuels Complex (MFC) and the complexity of novel removal activities of recently operated reactor experiments through the new proposed access tunnel. The efficacy was not apparent for existing equipment to provide all the needed capability. To bridge the potential gaps in cask designs, storage, and transportation, NRIC requested the development of trade studies for the transfer, handling, and storage requirements of irradiated fuel salts and other radioactive materials. NRIC directed Boston Government Services, LLC (BGS) to perform the trade studies and develop a report analyzing alternatives. In addition to the BGS reports, the Idaho National Lab’s (INL), provided by the first potential user’s Advanced Reactor Development (ARD) team, prepared a feasibility study for the storage of specific irradiated fuel within the existing ZPPR vault.This summary report is intended to present the trade studies, options, and alternatives that were investigated. The maturity level of LOTUS, the reactor, fuel salt containers, gloveboxes, and reactor testing campaign and concept of operations were not at a level sufficient to base critical decisions on. This report is not intended to present a final recommendation. The final recommendations for fuel storage location, transport, handling equipment, and operations will be made in FY 2024 and will be based on known materials, test campaign requirements, funding, and final analysis of the fuel and equipment to be used.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Conceptual Design of a Pumped Water Capsule for TREAT

This study performed mechanical design, thermal hydraulic, and fuel performance investigations to determine whether an existing large water capsule design could be modified to create forced convection boundary conditions on nuclear fuel specimens in the Transient Reactor Test facility. This concept was found to be viable and offers several advantages compared to a static water environment for a variety of transient power shape categories. The pumped capsule is not foreseen to be superior to a past concept for a larger water loop, but the pumped capsule approach is recommended for further engineering and deployment based on its high cost-to-value ratio primarily as a stop gap capability until a larger water loop can be realized.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗