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All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Impacts of LEU+ and HBU Fuel on Decay Heat and Radiation Source Term

Using low-enriched uranium with 235 U enrichment slightly greater than 5.0 wt % (LEU+) fuel is desirable for light water reactors (LWR) because it can enable longer cycles and/or smaller fresh fuel batches. This report analyzes the impact of LEU+ on back-end considerations, such as its impact on the spent fuel pool, and on the thermal and shielding performance of dry storage systems. For both pressurized water reactor (PWR) and boiling water reactor LEU+ fuel, this report concludes that additional cooling time or regionalized loading plans are required to maintain the same level of thermal and shielding performance of dry storage systems loaded with LEU fuel. For PWR fuel, this report concludes that the impact of LEU+ fuel on the time-to-boil of a spent fuel pool (SFP) is minimal and does not require considerable changes in SFP management.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

LOCA and LOFA Safety Margin Analysis for U-10Mo Monolithic LEU Fuel Conversion of the MURR{sup R} Reactor

The University of Missouri Research Reactor (MURR{sup R}) is a multi-disciplinary research and education facility providing a broad range of analytical and irradiation services to the research community and the commercial sector. The source of neutrons for the experimental facilities of MURR{sup R} is a compact 10 MW core fueled with highly-enriched uranium (HEU) fuel. The continued presence of HEU fuel in civilian installations such as research and test reactors poses a threat to national and international security. Therefore, most research and test reactors, both domestic and international, have started a program of conversion to the use of low-enriched uranium (LEU) fuel. MURR{sup R} is one of five U.S. high-performance research and test reactors (USHPRRs) that are actively collaborating with the U.S. National Nuclear Security Agency (NNSA) Office of Material Management and Minimization (M3) to develop a suitable LEU fuel replacement for the currently required HEU fuel. A new type of high-density LEU fuel based on an alloy of uranium and molybdenum is currently undergoing qualification testing, and is expected to allow the conversion of all USHPRRs. As a part of the LEU conversion of MURR{sup R}, an analysis of core behavior under a set of accident conditions for a core fueled with an LEU monolithic alloy of uranium and 10 wt% molybdenum (U-10Mo) has been conducted. This paper presents the newest results for the loss-of-coolant accidents (LOCAs) and loss-of-flow accidents (LOFAs). Other classes of accidents have also been investigated but were not included here. Namely, reactivity insertion accidents (RIAs), fuel failure during operations accidents (FFAs), and fuel handling accidents (FHAs)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

A Low-Enriched Uranium (LEU) option for the conversion of FRM II

The Forschungs-Neutronenquelle Heinz Maier-Leibnitz (FRM II) is Germany’s most powerful research reactor and uses Highly Enriched Uranium (HEU) fuel enriched at 93 %. The Technical University of Munich (TUM) operates the reactor and has been mandated to convert the FRM II to a lower-enrichment fuel. In preparation for the lower-enrichment fuel downselection, planned for 2023, TUM is evaluating multiple conversion scenarios. In this paper, it is demonstrated that a conversion of FRM II to Low-Enriched Uranium (LEU) is scientifically possible when using the novel monolithic U-10Mo fuel system, which is currently in the qualification process. As a research reactor, safety criteria, neutron flux distribution and cycle length are key metrics. To begin, neutronic and thermal-hydraulic models were created based on current technical drawings and relevant material properties. Specifically tailored computational methods and coupling schemes have also been developed to properly evaluate all relevant characteristics for the conversion of the FRM II reactor. Using these models and methods, a systematic parameter study was performed to explore the defined design space. The results of that study indicate that many LEU designs using monolithic U-10Mo appear to be viable to convert FRM II to LEU. One example of an LEU solution for the FRM II reactor that fulfills the stated safety, compatibility and scientific performance requirements is discussed in more detail.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Thermal-hydraulic analyses of MIT reactor LEU transition cycles

The Massachusetts Institute of Technology Reactor (MITR) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. In the framework of non-proliferation policy, research and test reactors have started a program to convert highly enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. A new type of LEU fuel based on a high-density alloy of uranium and molybdenum (U-10Mo) is expected to allow conversion. A transition core plan, from 22 fresh LEU fuel elements to 24 in an equilibrium state, is proposed. This study evaluates the thermal-hydraulic safety margins of each transition cycle and state (i.e. beginning, middle and end). The STAT7 and RELAP5 codes are used in the study. STAT7 provides an integrated platform for statistical propagation of uncertainties in determining steady-state thermal-hydraulic margins. A RELAP5 model was created to verify MITR's STAT7 model, and provide transient analyses. The steady-state analyses determined that, during the entire transition from a fresh to equilibrium core, the Limiting System Safety Settings (LSSS) has a 10.2% margin to the minimum power at which onset of nucleate boiling (ONB) is precluded at 3σ confidence. The loss of flow (LOF) transient, which is the most limiting anticipated transient for the MITR, was simulated. With a conservative initial state of the reactor (operating at μ + 3σ power level and LSSS mass flow rate), the maximum calculated centerline fuel temperature was 115 degrees C, which is a significantly lower value than the blistering limit of 350 degrees C. The cladding wall temperatures did not exceed the ONB temperature throughout the transient. Therefore, during LOF, no nucleate boiling is expected, precluding any critical heat flux trigger. Overall, this study indicates that the proposed LEU transition core specifications have significant margin to thermal-hydraulic limits, during steady-state operation and the LOF transient.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2026

Fuel Fabrication Specification Impact Analysis for NBSR LEU Conversion

As part of a national initiative to enhance nuclear security and reduce proliferation risks, significant efforts have been undertaken by the National Nuclear Security Administration Material Management and Minimization Office of Reactor Conversion Program to convert U.S. high performance research reactors (USHPRRs) from the use of highly enriched uranium (HEU) to low-enriched uranium (LEU), including the National Bureau of Standards Reactor (NBSR). The current plan is to procure LEU fuel assemblies from commercial fabricators according to fuel specifications tailored for each USHPRR. The analysis conducted at Brookhaven National Laboratory was part of an effort to identify the sources of uncertainty in the fuel specifications that may impact the performance of the NBSR core after its conversion and, in particular, to assess the range of acceptable tolerance limits from the perspective of core safety and reactor performance. Using the stochastic neutronics code MCNP 6.2, the variations in important NBSR neutronics characteristics were analyzed as a function of the specification parameters independently and in combination. The important NBSR specification parameters analyzed were the fuel isotopic composition, the amount of impurity content in cladding, the fuel plate thickness, and the fuel element 235U mass loading. The range of variation of each specification parameter was based on the technical specification limit or available as-fabricated assay data and uncertainties. The NBSR neutronics characteristics selected for analysis were the reactor reactivity characteristics at equilibrium and the equilibrium fuel cycle length. Results show that with variations in the fabrication parameters of the as-fabricated U-10Mo fuel within the specification limitations, the excess reactivity of the NBSR LEU core remains well below the 15% Δk/k technical specification limit, and the shutdown margin is always significantly greater than the required 0.68% Δk/k. This ensures that the NBSR can be operated safely and reliably shut down for all analyzed cases within the specified fabrication limits after the LEU conversion. In the prototypic case, the fuel cycle length was 1.5 days longer than the targeted 38.5 days. In a credible worst-case scenario, where all low-reactivity parameters were combined, the fuel cycle length was reduced to 35.5 days, which is still considered manageable for reactor operations. Variations in cycle length are primarily driven by changes in 235U loading, with other parameters having secondary effects.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

Numerical Simulations of Flow-Induced Deflections in MITR LEU Fuel Plate Due to Channel Size Disparity

The hydromechanical stability of the fuel plates in parallel coolant channels of a Materials Testing Reactor (MTR) fuel element design is of great importance to the safety of research and test reactors. Previous analytical, experimental, and numerical efforts focused on parallel channels with the same or similar size; also, in the prior numerical simulations, the fuel plate was often assumed to be perfectly flat. This work presents the results of a fluid-structure interaction simulation performed to evaluate the flow-induced deflections of the fuel plates in the low-enriched uranium (LEU, <20 wt% 235 U) fuel element design for the conversion (from highly enriched uranium) of the Massachusetts Institute of Technology Reactor (MITR-II, also referred to as MITR). Various manufacturing and assembly tolerances of the MITR LEU elements are considered in the analysis, and the effects of channel size disparity, nonideal plate shape, and flow rate uncertainty are investigated. Results show that, for all cases analyzed, the deflection occurs toward the larger channel, and the change in any channel stripe remains small (less than 0.021 mm) compared to fabrication tolerances. In addition to simulation work, a hydraulic performance test of the MITR LEU fuel element is currently planned to support conversion to the use of LEU fuel.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Steady-State Thermal-Hydraulic Analysis of the LEU-Fueled Dalat Nuclear Research Reactor

This paper presents results of steady-state thermal-hydraulic analysis for the designed working core of the Dalat Nuclear Research Reactor (DNRR) using the PLTEMP/ANL code. The core was designed to be loaded with 92 low-enriched uranium (LEU) VVR-M2 fuel bundles (FBs) and 12 beryllium rods surrounding a neutron trap at the core center, for replacement of the previous core with 104 high-enriched uranium (HEU) VVR-M2 FBs. Before using this code for thermohydraulic analysis of the designed LEU working core, it was validated by comparing calculation results with experimental data collected from the HEU working core of the DNRR. The discrepancy between calculated results and measured data was at the maximum about 0.8°C and 1.5°C of fuel cladding and outlet coolant temperatures, respectively. In the design calculation, thermohydraulic safety was confirmed through evaluation of the fuel cladding and coolant temperatures, as well as of other safety parameters such as Departure from Nucleate Boiling Ratio (DNBR) and Onset of Nucleate Boiling Ratio (ONBR). The calculation results showed that, in normal operation conditions at full nominal thermal power of 500 kW without uncertainty parameters, the maximum fuel cladding temperature of the hottest FB was about 90.4°C, which is lower than its limit value of 103°C, the minimum DNBR was 32.0, which is much higher than the recommended value of 1.5, and the minimum ONBR was 1.43, which is higher than the recommended value of 1.4 for VVR-M2 LEU fuel type. When the global and local hot channel factors were taken into account, the maximum temperature of fuel cladding at the hottest FB was about 98.4 °C, for global only, and 114.3°C, for global together with local hot channel factors. The calculation results confirm the safety operation of the designed LEU core loaded with 92 fresh VVR-M2 FBs.

Nguyen, Kien-Cuong↗
DOE OSTI2017

Gas generation during waste treatment of acidic solutions from the dissolution of irradiated LEU targets for 99 Mo production

The goal of the Reduced Enrichment for Research and Test Reactors Program is to limit the use of high-enriched uranium (HEU) in research and test reactors by substituting low-enriched uranium (LEU) wherever possible. The work reported here documents our work to develop the calcining technologies and processes that will be needed for 99 Mo production using LEU foil targets and the Modified Cintichem Process. The primary concern with the conversion to LEU from HEU targets is that it would result in a five- to six-fold increase in the total uranium. This increase results in more liquid waste from the process. We have been working to minimize the increase in liquid waste and to minimize the impact of any change in liquid waste. Direct calcination of uranium-rich nitric acid solutions generates NO 2 gas and UO 3 solid. We have proposed two processes for treating the liquid waste from a Modified Cintichem Process with a LEU foil. One is an optimized direct calcination process that is similar to the process currently in use. The other is a uranyl oxalate precipitation process. The specific goal of the work reported here was to characterize and compare the chemical reactions that occur during these two processes. In particular, the amounts and compositions of the gaseous and solid products were of interest. A series of experiments was carried out to show the effects of temperature and the redox potential of the reaction atmosphere. The primary products of the direct calcination process were mixtures of U 3 O 8 and UO 3 solids and NO 2 gas. The primary products of the oxalate precipitation process were mixtures of U 3 O 8 and UO 2 solid and CO 2 gas. Higher temperature and a reducing atmosphere tended to favor quadrivalent over hexavalent uranium in the solid product. These data will help producers to decide between the two processes. In addition, the data can be used to design appropriate off-gas systems for pilot and production facilities.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Fuel Swelling and Creep Analysis for a MURR LEU U-10Mo Monolithic Plate

In this work, finite element analysis (FEA) is used to model and simulate the thermo-mechanical behavior of a selected LEU MURR fuel plate under typical irradiation conditions of the proposed LEU MURR core. The proposed LEU fuel element for MURR conversion has 23 curved fuel plates. The fuel element design and analysis are described in Reference [7]. The safety analysis for calculating the margin to flow instability that had elements with a mixture of burnups, which is typical for MURR operations. In addition to the application of engineering hot channel factors to account for variations in the fuel element manufacturing that could affect the coolant channel gaps, the analysis assumed a constriction of the coolant channel gaps due to an increase in the fuel plate thicknesses during the element lifetime. Estimates of the local plate thickness increase at end-of-life (EOL) due to fuel swelling and creep were developed from experimental data presented in References [8] and [9]. The maximum local plate thickness increase from that analysis was calculated for fuel plate 23 in an EOL element. Furthermore, the minimum margin to flow instability for the reference LEU cores was calculated to occur in coolant channel 23 of an EOL element, which is the coolant channel between plates 22 and 23. Therefore,the work performed here evaluated plate 23 only, as this was found to be the most limiting plate.

Mohamed, Walid↗
DOE OSTI2019

Recovery of Mo in LEU-Modified Cintichem Process at Elevated Mo Concentrations

Argonne National Laboratory with support from DOE’s National Nuclear Security Administration Office of Material Management & Minimization (M 3 ) is helping SHINE Medical Technologies in developing domestic production of Mo-99. SHINE is planning to produce Mo-99 by fission of low enriched uranium-235 (LEU) in a subcritical aqueous solution using accelerator-based neutron generation. SHINE’s subcritical system can contain ~250L of LEU solution that is processed through a recovery titania-based column to recover Mo, while the remaining LEU solution can be available for another irradiation. The Mo-containing solution is then fed into another titania-based concentration column to further concentrate the solution to allow for the final purification step using the LEU modified Cintichem (LMC) process. The LMC process is a relatively small-scale process that was designed to work with 50mL volumes. By concentrating the Mo solution that feeds into the LMC process from ~250L to ~50mL, the actual chemical concentration of Mo becomes important. It is estimated that up to ~15mM Mo concentration could be present in the LMC feed. Since the LMC process was originally proposed for purification of Mo after dissolution of irradiated U solid targets, the chemical concentration of Mo was not a concern, due to presence of only trace amounts of Mo.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Controlling Pu behavior on Titania: Implications for LEU Fission-Based Mo-99 Production

Molybdenum-99 is the parent isotope of the most widely used isotope, technetium-99m, in all diagnostic nuclear medicine procedures. Due to proliferation concerns associated with the use of highly enriched uranium (HEU), the preferred method of fission-based Mo-99 production uses low enriched uranium (LEU) targets. Using LEU versus HEU for Mo-99 production produces similar to 30 times more Pu-239, due to neutron capture on U-238 to produce Np-239, which ultimately decays to Pu-239 (t(1/2) = 24,110 yr). Argonne National Laboratory is supporting a potential US Mo-99 producer in their efforts to produce Mo-99 from an LEU solution. In order to mitigate the generation of large volumes of greater-than-class-C (GTCC) low level waste (Pu-239 concentrations greater than 1 nCi/g), we have focused our efforts on the separation chemistry of Pu and Mo with a titania sorbent in sulfate media. Results from batch and column experiments show that temperature and acid wash concentration can be used to control Pu behavior on titania.

low enriched uranium↗
DOE OSTI2017

JM-1 SLOWPOKE-II Research Reactor Supporting Analysis Report for the Conversion from HEU to LEU

The JM-1 research reactor (SLOWPOKE-2 type) at the International Centre for Nuclear and Environmental Sciences (ICENS) of the University of West Indies, in Kingston (Jamaica), was successfully converted from HEU to LEU fuel on September 2015. As part of the work to produce the corresponding Safety Analysis Report (SAR), which was to be reviewed and approved by the Jamaican regulator in order to provide authorization for the reactor commissioning, a detailed safety analysis was carried out. This analysis, involving a variety of calculations, is summarized in the present report; describing the basic methodology and the obtained results. The report covers both the neutronic and thermal hydraulic aspects of the analysis, for which MCNP and PLTEMP/ANL codes, respectively, were used. On one hand, the HEU core was modeled and analyzed, subsequently comparing the obtained results with the available experimental data. On the other hand, the modifications involved with the change to an LEU core were introduced into the models and the results examined to ensure that the differences in the expected behavior did not negatively affect the reactor safety or performance. In fact, the LEU core was found to present significant advantages in different areas. Consequently, the present report provided a solid basis for the SAR.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Modifications to LEU-Modified Cintichem Process to Allow Processing of Fission-Produced 99 Mo from Up to 750 g of Irradiated Uranium Targets

Eden Radioisotopes, LLC is developing a TRIGA* -sized, 2 MW all-target reactor system licensed from Sandia National Laboratories. This all-target reactor uses up to 60 Argonne-designed annular LEU-foil targets to produce fission-produced 99 Mo. The Eden low-enriched uranium (LEU) target contains ~137 g of uranium and will be irradiated at up to 30 kW, with a target burnup determined by the number of targets processed each week. The annular target utilizes metallic uranium foils 400 mm long by 72 mm wide by 250±25 -µm-thick covered by a thin (10 µm on each face) nickel fission barrier. The foil is placed between two 35 mil aluminum tubes that form the outer and inner walls of the cladding cylinder, which is hydroformed to provide high thermal conductivity within the target. Eden expects to process between three and five targets per process batch, to maintain the process batch 99 Mo inventory at or below 5000 curies to maintain high 99 Mo recovery, within a few hours of discharge from the reactor. Eden expects to meet full U.S. demand by processing 30 targets per week, with each target operated for two weeks at 30 kW. To meet less than full U.S. demand, both the target power and the number of targets processed weekly will be adjusted. With increased 99 Mo production and given the size of U targets processed in a single batch (up to ~750g-U), it is important to identify whether any modifications are needed to the LEU-modified Cintichem process.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Adapting LEU-Modified Cintichem Process for Processing Up to 750 Grams of Uranium Targets

Argonne National Laboratory (Argonne) developed a low-enriched uranium (LEU)-modified Cintichem (LMC) process to enable production and purification of fission-produced molybdenum-99 (Mo-99) from LEU foil targets. This separation scheme starts with nitric acid (HNO 3 ) dissolution of irradiated LEU foil targets that yields an approximate final volume of 2 L. The dissolved uranium solution containing fission products, including Mo-99, is then treated by loading the solution on a titania-based recovery column to separate Mo-99 from bulk uranium. After several washing steps, Mo and other fission products are stripped from the titania column using sodium hydroxide (NaOH). The eluate is then acidified using nitric acid and subsequently processed for final purification using the LMC process. Uranium and most fission products are not adsorbed on the titania column but are eluted (i.e., removed using a solvent) during the wash steps; they can be purified and recycled into new uranium targets.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2024

Thermo-Mechanical Analysis of Irradiated MURR LEU Fuel Plates

The University of Missouri Research Reactor (MURR®) is a multi-disciplinary research and education facility providing a broad range of analytical, materials science, and irradiation services to the research community and the commercial sector. MURR is one of five U.S. high performance research reactors (USHPRR), plus one critical facility, actively collaborating with the National Nuclear Security Administration (NNSA) Material Management and Minimization (M 3 ) Reactor Conversion Program to convert from the use of highly enriched uranium (HEU, ≥ 20 wt% U 235) to low-enriched uranium (LEU, < 20 wt% U-235) fuel. All USHPRR, including MURR, completed designs with a new type of very high-density LEU fuel based on an alloy of uranium and 10-weight percent molybdenum (U-10Mo) for conversion to LEU fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Preliminary Thermo-Mechanical Analysis of Irradiated MURR LEU Fuel Element

The University of Missouri Research Reactor (MURR) is a multi-disciplinary research and education facility providing a broad range of analytical, materials science, and irradiation services to the research community and the commercial sector. MURR is one of five U.S. high performance research reactors (USHPRR), plus one critical facility, that is actively collaborating with the National Nuclear Security Administration (NNSA) Material Management and Minimization (M3) Office of Reactor Conversion and Uranium Supply to convert from the use of highly enriched uranium (HEU, ≥ 20 wt% U-235) to low-enriched uranium (LEU, < 20 wt% U-235) fuel. A new type of very high-density LEU fuel based on an alloy of uranium and 10-weight percent molybdenum (U-10Mo) is expected to allow the conversion of some USHPRR, including MURR, to LEU fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

HFIR LEU High Density Silicide Dispersion Optimized Design Steady-State Heat Transfer Analyses

Steady-state heat transfer simulations of the Oak Ridge National Laboratory High Flux Isotope Reactor (HFIR) with the low-enriched uranium (LEU) high-density silicide dispersion Optimized fuel design were performed to support comprehensive performance and safety metric studies concerning this design. The LEU Optimized design operates at 95 MW to maintain HFIR’s current highly enriched uranium (HEU) core performance level at 85 MW. Full cycle Mode 1 full flow Case 1 (inlet temperature), Case 2 (flux-to-flow), and Case 3 (inlet pressure) safety limit analyses were performed to assess the margins to critical heat flux. Under the prescribed conditions, this LEU design meets the safety limit and limiting control setting requirements outlined in HFIR’s documented safety analysis; however, the safety margins are less than those for the 85 MW HEU core, and several assumptions were made where fuel fabrication and qualification data are currently lacking for the silicide fuel design. Effects of changes to pertinent fuel fabrication assumptions and uncertainty factors on thermal safety margins were also evaluated, showing that the margins are sensitive to many of these parameters. Power and pressure perturbations were also performed, indicating that significant steady-state thermal margins could be gained by increasing the coolant inlet pressure.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

COMSOL Results for the Nominal Steady-State Operation of the Proposed 95-MW LEU Silicide Core for HFIR Conversion

Engineering design studies are being performed to determine the feasibility of converting the High Flux Isotope Reactor (HFIR) from highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel at Oak Ridge National Laboratory. This activity is sponsored by the Office of Reactor Conversion and Uranium Supply (ORCUS) under the auspices of the US Department of Energy National Nuclear Security Administration’s Office of Material Management and Minimization. HFIR is a very high flux, pressurized, light water–cooled and moderated, flux trap–type research reactor with a core made of involute shaped U 3 O 8 /Al cermet fuel plates and coolant channels. HFIR currently operates at a thermal power of 85 MW and supports key national and international missions in neutron scattering, isotope production, materials/fuels irradiation, neutron activation analysis, gamma irradiation, and neutrino research. Advanced multiphysics computational fluid dynamics models have been developed in the COMSOL Multiphysics software to simulate the steady-state operating conditions for the proposed low-and high-density LEU U 3 Si 2 -Al (uranium silicide dispersion) fuel designs. The COMSOL models for HFIR inner and outer fuel element models incorporate various essential inputs and physics such as spatially dependent nuclear heat deposition, multilayer heat conduction, conjugate heat transfer, turbulent flows (using Reynolds-averaged Navier Stokes turbulence models), structural mechanics (thermal–structural interactions and fuel swelling), and oxide layer build-up. This report presents the best-estimate thermal hydraulics results for the low- and high-density optimized silicide LEU core designs at 95 MW steady-state nominal operation.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗