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DOE OSTI2023

A Plan to Qualify New Fuel for the High Flux Isotope Reactor for Material Minimization

The High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) is one of five high power research reactors the Office of Material Management and Minimization (M3) Program, Office of Conversion is working to convert from using highly enriched uranium (HEU) fuel to using low-enriched uranium (LEU) fuel. This effort stems from the primary objective within the U.S. Department of Energy (DOE) National Nuclear Security Administration (NNSA) to achieve permanent threat reduction by minimizing, and when possible, eliminating weapon-usable nuclear material around the world. Under M3’s Office of Conversion, the U.S. High Performance Research Reactor (USHPRR) Project is pursuing fuel qualification and licensing of LEU fuels to support the high-performance reactors. All high-performance reactors except HFIR will be converted to LEU monolithic uranium-molybdenum alloy fuel. HFIR will be evaluated for conversion to LEU using a uranium silicide fuel, namely, U3Si2-Al dispersion fuel. The mission of the USHPRR Project is to develop the technology needed to reduce, and eventually eliminate, worldwide use of HEU in civilian applications. The goal is to develop the technical means needed to use low enriched uranium (LEU) instead of HEU fuel in research and test reactors without significant penalties in performance, economics, or safety of the reactors. The USHPRR Project has four major elements, called Pillars: Fuel Qualification (FQ) managed at Idaho National Laboratory (INL), Fuel Fabrication (FF) managed at Pacific Northwest National Laboratory (PNNL), Reactor Conversion (RC) managed at Argonne National Laboratory (Argonne), and Cross-Cutting (CC) managed at Savannah River National Laboratory (SRNL). FQ is responsible for the qualification of the fuel type. RC is responsible for supporting reactor conversion analysis and overseeing licensing submittals leading to conversions of domestic reactors to LEU fuel. For the FQ effort, FQ (INL) worked in collaboration with RC (Argonne) and ORNL to develop the plan for the uranium silicide fuel qualification for HFIR. The resulting HFIR Fuel Qualification Plan provides the general approach for the USHPRR team to move the selected uranium silicide fuel design for HFIR conversion through qualification. Authorization and use in HFIR will be approved through the DOE’s Office of Science. Uranium silicide fuel was previously qualified in NUREG-1313 at an approximate maximum heat flux of 1.4 MW/m2 and a maximum fuel section temperature of about 130°C. In addition to the different regulator process utilized by DOE, these upper limits will be exceeded in HFIR; therefore, further testing will be necessary to ensure the fuel can meet HFIR qualification requirements. The HFIR fuel loading may exceed 4.8 gU/cm3 which was determined in the NUREG-1313 safety evaluation to be acceptable for use in non-power NRC-licensed reactors provided there exist no other safety considerations. In addition, the uranium silicide fuel will need to be qualified in a HFIR-specific design. This plan includes the currently available information from the USHPRR Project Functions and Requirements document and expands these requirements to ensure that planned tests have traceable results providing evidence that the requirements have been met. Data collection methods are discussed as well as the process to show that the requirements have been met. This document is designed to provide a pathway for researchers to obtain data necessary and at the appropriate quality level for HFIR fuel qualification

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2015

Automated Installation Verification of COMSOL via LiveLink for MATLAB

Verifying that a local software installation performs as the developer intends is a potentially time-consuming but necessary step for nuclear safety-related codes. Automating this process not only saves time, but can increase reliability and scope of verification compared to ‘hand’ comparisons. While COMSOL does not include automatic installation verification as many commercial codes do, it does provide tools such as LiveLink™ for MATLAB® and the COMSOL API for use with Java® through which the user can automate the process. Here we present a successful automated verification example of a local COMSOL 5.0 installation for nuclear safety-related calculations at the Oak Ridge National Laboratory’s High Flux Isotope Reactor (HFIR).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2023

A Plan to Qualify New Fuel for the High Flux Isotope Reactor for Material Minimization

The High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) is one of five high power research reactors the Office of Material Management and Minimization (M3) Program, Office of Conversion is working to convert from using highly enriched uranium (HEU) fuel to using low-enriched uranium (LEU) fuel. This effort stems from the primary objective within the U.S. Department of Energy (DOE) National Nuclear Security Administration (NNSA) to achieve permanent threat reduction by minimizing, and when possible, eliminating weapon-usable nuclear material around the world. Under M3’s Office of Conversion, the U.S. High Performance Research Reactor (USHPRR) Project is pursuing fuel qualification and licensing of the high-performance reactors to operate with LEU fuels. All high-performance reactors except HFIR will be converted to LEU monolithic uranium-molybdenum alloy fuel. HFIR will be evaluated for conversion to LEU using a uranium silicide fuel, namely, U3Si2-Al dispersion fuel.The mission of the USHPRR Project is to develop the technology needed to reduce, and eventually eliminate, worldwide use of HEU in civilian applications. The goal is to develop the technical means needed to use low enriched uranium (LEU) instead of HEU fuel in research and test reactors without significant penalties in performance, economics, or safety of the reactors. The USHPRR Project has four major elements, called Pillars: Fuel Qualification (FQ) managed at Idaho National Laboratory (INL), Fuel Fabrication (FF) managed at Pacific Northwest National Laboratory (PNNL), Reactor Conversion (RC) managed at Argonne National Laboratory (Argonne), and Cross-Cutting (CC) managed at Savannah River National Laboratory (SRNL). FQ is responsible for the qualification of the fuel type. RC is responsible for supporting reactor conversion analysis and overseeing licensing submittals leading to conversions of domestic reactors to LEU fuel. For the FQ effort, FQ (INL) worked in collaboration with RC (Argonne) and ORNL to develop the plan for the uranium silicide fuel qualification for HFIR.The resulting HFIR Fuel Qualification Plan provides the general approach for the USHPRR team to move the selected uranium silicide fuel design for HFIR conversion through qualification. Authorization and use in HFIR will be approved through the DOE’s Office of Science. Uranium silicide fuel was previously qualified in NUREG-1313 at an approximate maximum heat flux of 1.4 MW/m2 and a maximum fuel section temperature of about 130°C. In addition to the different regulator process utilized by DOE, these upper limits will be exceeded in HFIR; therefore, further testing will be necessary to ensure the fuel can meet HFIR qualification requirements. The HFIR fuel loading may exceed 4.8 gU/cm3 which was determined in the NUREG-1313 safety evaluation to be acceptable for use in non-power NRC-regulated reactors provided there exist no other safety considerations. In addition, the uranium silicide fuel will need to be qualified in a HFIR-specific design. This plan includes the currently available information from the USHPRR Project Functions and Requirements document and expands these requirements to ensure that planned tests have traceable results providing evidence that the requirements have been met. Data collection methods are discussed as well as the process to show that the requirements have been met. This document is designed to provide a pathway for researchers to obtain data necessary and at the appropriate quality level for HFIR fuel qualification.

Shokes, Tamara↗

DOE OSTI2014

HFIR Dissolution Nuclear Criticality Analysis Information

The purpose of this analysis is to demonstrate the safety of High Flux Isotope Reactor (HFIR) dissolution from a criticality safety perspective and to establish any Criticality Safety Limits (CSLs) that may be necessary. HFIR fuel consists of two annular elements that are designed to be nested together with a beryllium reflector in the reactor configuration. HFIR cores are currently stored in L-Area as separate inner elements and outer elements, each on a unique aluminum carrier. HFIR elements will be shipped to H-Canyon in the 70-ton Cask Railcar and loaded into the dissolver with the HFIR insert installed. The HFIR insert consists of a small well for the inner fuel element and a large well for the outer fuel element.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗

DOE OSTI2012

RELAP5 model of the high flux isotope reactor with low enriched fuel thermal flux profiles

The High Flux Isotope Reactor (HFIR) currently uses highly enriched uranium (HEU) fabricated into involute-shaped fuel plates. It is desired that HFIR be able to use low enriched uranium (LEU) fuel while preserving the current performance capability for its diverse missions in material irradiation studies, isotope production, and the use of neutron beam lines for basic research. Preliminary neutronics and depletion simulations of HFIR with LEU fuel have arrived to feasible fuel loadings that maintain the neutronics performance of the reactor. This article illustrates preliminary models developed for the analysis of the thermal-hydraulic characteristics of the LEU core to ensure safe operation of the reactor. The beginning of life (BOL) LEU thermal flux profile has been modeled in RELAP5 to facilitate steady state simulation of the core cooling, and of anticipated and unanticipated transients. Steady state results are presented to validate the new thermal power profile inputs. A power ramp, slow depressurization at the outlet, and flow coast down transients are also evaluated. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2026

Depletion benchmark for a high-assay low-enriched uranium fuel experiment in the advanced test reactor

Reactor physics depletion benchmarks for high-assay low-enriched uranium (HALEU) fuel are limited in number. In particular, there is limited data for HALEU benchmarks for U-10Mo (uranium-10% molybdenum) plate fuel that is being developed for use in the United States’ high performance research reactors including the Advanced Test Reactor (ATR), Advanced Test Reactor Critical Facility (ATR-C), High Flux Isotope Reactor (HFIR), Massachusetts Institute of Technology Reactor (MITR), University of Missouri Research Reactor (MURR), National Bureau of Standards Reactor (NBSR). These six reactors currently operate with highly enriched uranium dispersed fuel in an aluminum matrix. In support of conversion to a HALEU fuel, qualification of U-10Mo formed into a monolithic foil is being performed. Fuel qualification involves irradiating fuel specimens in the ATR. The irradiation tests provide an opportunity to benchmark depletion capabilities of reactor physics codes in support of the ATR operation, as well as develop benchmarks that can be used by other institutions to benchmark other reactor physics codes. This paper documents the development of a benchmark model of the irradiation of the ATR Full-size plate In center flux trap Position 7 (AFIP-7) experiment using the depletion codes MC21 and Advanced Dimensional Depletion for Engineering of Reactors (ADDER).

Nielsen, Joseph W. [Idaho National Laboratory (INL↗

DOE OSTI2020

The Local Seismoacoustic Wavefield of a Research Nuclear Reactor and Its Response to Reactor Power Level

In this article, we describe the seismoacoustic wavefield recorded outdoors but inside the facility fence of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (Tennessee). HFIR is a research nuclear reactor that generates neutrons for scattering, irradiation research, and isotope production. This reactor operates at a nominal power of 85 MW, with a full-power period between 24 and 26 days. This study uses data from a single seismoacoustic station that operated for 60 days and sampled a full operating reactor cycle, that is, full-power operation and end-of-cycle outage. The analysis presented here is based on identifying signals that characterize the steady, that is, full-power operation and end-of-cycle outage, and transitional, that is, start-up and shutdown, states of the reactor. We found that the overall seismoacoustic energy closely follows the main power cycle of the reactor and identified spectral regions excited by specific reactor operational conditions. In particular, we identified a tonal noise sequence with a fundamental frequency around 21.4 Hz and multiple harmonics that emerge as the reactor reaches 90% of nominal power in both seismic and acoustic channels. We also utilized temperature measurements from the monitoring system of the reactor to suggest links between the operation of reactor’s subsystems and seismoacoustic signals. We demonstrate that seismoacoustic monitoring of an industrial facility can identify and track some industrial processes and detect events related to operations that involve energy transport.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2020

ORNL Post-irradiation Examination (LAMDA) Final Report on Characterization of Irradiated Grade 92 Ferritic-Martensitic Steels

Irradiation effects on ferritic-martensitic steel Grade 92 were investigated in this project. Samples from two heats of Grade 92, i.e., G92-2b and NF616, were selected, which were irradiated in the High Flux Isotope Reactor (HFIR) and the Advanced Test Reactor (ATR), respectively. Samples from two heats of ferriticmartensitic steel T91, irradiated in the ATR, were also selected as references for the Grade 92 samples. The selected G92-2b samples were irradiated to 0.5–14.7 displacements per atom (dpa) at two temperature regimes: 400–497 °C and 683–720°C. The selection of the high irradiation temperatures up to ~720°C is to test the performance at such upper bound temperatures, which would provide insights for the upper bound applicable temperature for Grade 92. The selected NF616 and T91 samples were irradiated to 3–8.2 dpa at 292–448°C. Except for T91 heat2 samples in 3-mm diameter discs, all the other selected samples are of type SS-J2 miniature tensile specimens. Two Nuclear Science User Facilities (NSUF), i.e., the Irradiated Materials Examination and Testing (IMET) hot cell facility and the Low Activation Materials Design and Analysis (LAMDA) facility, were used in the Post-Irradiation Examination (PIE) activities for the selected samples, which include: Tensile tests at room temperature (IMET); Fractography of the tensile fracture surfaces (LAMDA: scanning electron microscopy and energy; dispersive spectroscopy); Vickers hardness measurements (LAMDA); and Microstructural characterization (LAMDA: electron backscatter diffraction, focused-ion beam, and transmission electron microscopy).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2025

Depletion Benchmark of the AFIP-7 Experiment in the Advanced Test Reactor

Reactor physics depletion benchmarks for low-enriched uranium fuel are limited in number. In particular, there is very limited data for LEU benchmarks for U-10Mo (Uranium-10% Molybdenum) plate fuel developed for use in U.S. high-performance research reactors (USHPRR). USHPRR includes the Advanced Test Reactor (ATR), Advanced Test Reactor Critical Facility (ATR-C), High Flux Isotope Reactor (HFIR), University of Missouri Research Reactor (MURR), Massachusetts Institute of Technology Reactor (MITR), and National Bureau of Standards Reactor (NBSR) at the National Institute of Science and Technology. These reactors are fueled with high-enriched uranium dispersed fuel in a silicon/aluminum matrix. In support of conversion to a HALEU fuel, qualification of U-10Mo formed into a monolithic foil is being performed. Fuel qualification involves irradiated fueled specimens in the ATR. The irradiation tests provide an opportunity to benchmark depletion capabilities of reactor physics codes in support of the ATR operation, as well as develop benchmarks that can be used by other institutions to benchmark other reactor physics codes. This report documents the development of a benchmark model of the irradiation of the ATR Full -size plate In center flux trap Position 7 (AFIP-7) experiment.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2018

Alternate Design Concepts for the High Flux Isotope Reactor using Low-Enriched Uranium Fuel Systems: Scoping Analysis

The Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor (HFIR) is a multi-mission research reactor dedicated mainly to isotope production, neutron scattering experiments, material irradiation, testing and activation. In the framework of non-proliferation policies, the international community is currently aiming to reduce or even eliminate, when and where possible, the use of Highly Enriched Uranium (HEU: 235 U/U ≥ 20 wt. %) in civilian applications. In that context, most research reactors worldwide still operating with HEU fuel are actively engaged in an effort to convert to Low Enriched Uranium fuel system (LEU: 235 U/U < 20 wt. %). Within the US, the High Performance Research Reactor (USHPRR) fleet (which includes HFIR) is expected to convert to LEU fuel using the UMo “monolithic” fuel system currently under development. ORNL is responsible for the development of a LEU design for HFIR. In the past years, Argonne National Laboratory (ANL) offered technical assistance to the ORNL LEU design team whenever deemed appropriate. Due to their complex nature, acceptable HFIR LEU designs identified so far, all based on the UMo monolithic fuel system, would require a substantial fabrication development effort before being proven commercially viable. In order to mitigate technical risks and reduce cost/schedule uncertainties, it appears appropriate to study alternate design solutions and alternate fuel systems as possible backup options for the HFIR conversion to LEU fuel. In that context, ANL has been exploring alternate design concepts for HFIR. Complementing the ORNL design activities, the goal is to find the best design solution for HFIR in term of cost, performance and safety. The present report summarizes the findings regarding two types of alternate conceptual design: U 3 Si 2 designs and un-contoured UMo design. Based on available tools and methods, preliminary analysis show that both concepts could probably meet performance and steady-state safety requirements but element fabrication would remain complex and would require a non-trivial fabrication development effort.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2025

A methodology for corrosion testing ODS steels in liquid tin under reactor conditions in HFIR

Liquid metal–based divertor concepts have promising attributes for application in fusion reactors. Liquid tin is being considered as one of those liquid metals due to its excellent thermophysical properties. However, tin is extremely corrosive to steels at elevated temperatures, and the coupled effects of corrosion with neutron irradiation for application in fusion reactors have not been quantified. Researchers at Oak Ridge National Laboratory (ORNL) have designed and irradiated a series of capsules in the High Flux Isotope Reactor (HFIR) to help address gaps in the literature with respect to coupled corrosion and neutron irradiation effects. The capsules are filled with solid tin designed to melt due to the gamma heating in HFIR and allow interaction with the specimens. The capsules contain a SiC thermometer for post-irradiation temperature verification. Five capsules were irradiated in HFIR for 10.5 days, accumulating 0.93 dpa in the FeCrAl specimens. Post-irradiation temperature verification was performed on the SiC thermometers, and temperatures were observed to be higher than what was expected from thermal models. An explanation for the model underprediction is given herein. In conclusion, the capsules described within this article are the first to irradiate liquid metal within the HFIR flux trap and demonstrated the safety basis for continued liquid coolant studies in HFIR.

FeCrAl↗

DOE OSTI2024

Fast neutron irradiation capability in existing thermal test reactors

In today’s nuclear industry, momentum towards the design, licensing, and construction of advanced nuclear demonstration plants, including fast reactors, is at a remarkably high level. However, there are currently no dedicated fast spectrum irradiation test facilities in the United States to support the development of fast spectrum technologies. As a result, a unique situation is developing where most of these plants will likely be designed by leveraging historic nuclear material technologies, but where the further optimization and advancement is impeded by the lack of fast neutron irradiation test facilities. While these circumstances present a challenge, there are some near-term opportunities that, if seized, can still help develop advanced fast reactor materials to a meaningful level of readiness to support future commercial fast reactors. Here, in this paper, we assess the feasibility of using thermal neutron filtering materials in existing experiment positions in the Advanced Test Reactor (ATR) at Idaho National Laboratory and the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory to simulate fast reactor test environments for nonfuel test specimens. Items investigated include the incident neutron flux (both fast and thermal), the total neutron fluence and cumulative atom displacements, helium production rate due to thermal neutron capture in nickel, and the potential impact that the thermal neutron filter material has on the cycle length of a given reactor. It is concluded that while HFIR provides the highest fast flux of all the options investigated, it is limited in the amount of thermal neutron filtering material that can be introduced into an experiment position without significantly affecting the operation of the reactor. Irradiation in Outboard-A positions in the ATR was found to be the most realistic near-term experiment avenue due to having ample space for several capsules in a moderately fast flux.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2016

Thermal Safety Analyses for the Production of Plutonium-238 at the High Flux Isotope Reactor

There has been a considerable effort over the previous few years to demonstrate and optimize the production of plutonium-238 ( 238 Pu) at the High Flux Isotope Reactor (HFIR). This effort has involved resources from multiple divisions and facilities at the Oak Ridge National Laboratory (ORNL) to demonstrate the fabrication, irradiation, and chemical processing of targets containing neptunium-237 ( 237 Np) dioxide (NpO 2 )/aluminum (Al) cermet pellets. A critical preliminary step to irradiation at the HFIR is to demonstrate the safety of the target under irradiation via documented experiment safety analyses. The steady-state thermal safety analyses of the target are simulated in a finite element model with the COMSOL Multiphysics code that determines, among other crucial parameters, the limiting maximum temperature in the target. Safety analysis efforts for this model discussed in the present report include: (1) initial modeling of single and reduced-length pellet capsules in order to generate an experimental knowledge base that incorporate initial non-linear contact heat transfer and fission gas equations, (2) modeling efforts for prototypical designs of partially loaded and fully loaded targets using limited available knowledge of fabrication and irradiation characteristics, and (3) the most recent and comprehensive modeling effort of a fully coupled thermo-mechanical approach over the entire fully loaded target domain incorporating burn-up dependent irradiation behavior and measured target and pellet properties, hereafter referred to as the production model. These models are used to conservatively determine several important steady-state parameters including target stresses and temperatures, the limiting condition of which is the maximum temperature with respect to the melting point. The single pellet model results provide a basis for the safety of the irradiations, followed by parametric analyses in the initial prototypical designs that were necessary due to the limiting fabrication and irradiation data available. The calculated parameters in the final production target model are the most accurate and comprehensive, while still conservative. Over 210 permutations in irradiation time and position were evaluated, and are supported by the most recent inputs and highest fidelity methodology. The results of these analyses show that the models presented in this report provide a robust and reliable basis for previous, current and future experiment safety analyses. In addition, they reveal an evolving knowledge of the steady-state behavior of the NpO 2 /Al pellets under irradiation for a variety of target encapsulations and potential conditions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2021

HFIR SiC-SiC Composite Clad Tube Bowing Test: Pre-Irradiation Characterization

This report describes the pre-irradiation characterization of the SiC-SiC composite clad tube bowing experiment that will be irradiated in the High Flux Isotope Reactor (HFIR). There are concerns that SiCSiC fuel cladding in light water reactors could undergo bowing because of the non-uniform fast neutron flux profiles. This experiment is combined with the SiC-SiC channel box experiment, which has similar concerns regarding control blade movements in boiling water reactors. The HFIR experiment aims to validate the predicted deformation and stresses in the composite clad specimen after the exposure of fast neutron flux gradients. Significant radial fast neutron flux gradients that exist in the permanent reflector of HFIR were thoroughly characterized by using detailed 3D neutronic calculations. The 3D displacement damage dose rate profile and the resulting volumetric swelling in SiC were used as inputs to structural analyses to determine the predicted deformation that will affect the specimens. The tube specimens were characterized by using traditional dimensional inspection and surface profilometry to provide detailed information regarding the pre-irradiation condition as the first step to evaluate the radiation-induced deformation. Furthermore, fine engraving markers were inscribed along all outer surfaces of the specimen and mapped by using a digital microscope and a multidimensional stage. This allows marker spacings to be accurately measured and compared with similar measurements that will be made post-irradiation to provide local radiation-induced strain mapping. The experiment successfully completed two HFIR cycle irradiations, cycles 492 and 493, which finished on July 25, 2021.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2020

Fuel Conversion Efforts at the High Flux Isotope Reactor – a 2020 Status Update

The High Flux Isotope Reactor (HFIR) provides one of the world’s highest steady-state neutron fluxes in the world for neutron scattering experiments focused on impactful scientific discovery, as well as materials irradiation studies and production of medical, industrial, and research isotopes. Efforts are ongoing to convert HFIR from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel while maintaining or enhancing current performance and safety margin, thus sustaining HFIR’s mission portfolio and reactor-based neutron science leadership. This paper presents a status update on the HFIR fuel conversion efforts.

Chandler, David↗

DOE OSTI2023

Fuel Conversion Efforts at the High Flux Isotope Reactor – a 2023 Status Update

Sizemore, Carol↗

DOE OSTI2017

Combining RHF and HFIR Disposition Campaigns – Analysis, Opportunity, and Lessons Learned

Savannah River Nuclear Solutions has tasked its H-Canyon facility to prepare for disposition of used Oak Ridge National Laboratory High Flux Isotope Reactor (HFIR) cores. HFIR cores consist of an inner and outer assembly of highly enriched uranium fuel plates. The two elements are annular, nest inside each other, and the fuel form is a curved plate held between two collars. H-Canyon has a dissolver insert specifically designed for HFIR cores and has performed this type of campaign before between 1976 and 1988. At the H-Canyon facility, a dissolver is a large stainless steel pot loaded with a nitric acid solution into which uranium-alloy fuel is dissolved. The dissolver uses a device called an insert to feed the fuels to the acid solution in a controlled manner. The mission will be to dissolve a number of HFIR cores, purify the uranium, and down blend it to low enriched uranium suitable for commercial power reactors.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

Needs and Requirements for Future Research Reactors (ORNL Perspectives)

The High Flux Isotope Reactor (HFIR) is a vital national and international resource for neutron science research, production of radioisotopes, and materials irradiation. While HFIR is expected to continue operation for the foreseeable future, interest is growing in understanding future research reactors features, needs, and requirements. To clarify, discuss, and compile these needs from the perspective of Oak Ridge National Laboratory (ORNL) research and development (R&D) missions, a workshop, titled “Needs and Requirements for Future Research Reactors”, was held at ORNL on May 12, 2015. The workshop engaged ORNL staff that is directly involved in research using HFIR to collect valuable input on the reactor’s current and future missions. The workshop provided an interactive forum for a fruitful exchange of opinions, and included a mix of short presentations and open discussions. ORNL staff members made 15 technical presentations based on their experience and areas of expertise, and discussed those capabilities of the HFIR and future research reactors that are essential for their current and future R&D needs. The workshop was attended by approximately 60 participants from three ORNL directorates. The agenda is included in Appendix A. This document summarizes the feedback provided by workshop contributors and participants. It also includes information and insights addressing key points that originated from the dialogue started at the workshop. A general overview is provided on the design features and capabilities of high performance research reactors currently in use or under construction worldwide. Recent and ongoing design efforts in the US and internationally are briefly summarized, followed by conclusions and recommendations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗