Low-enriched Uranium Core Research Articles

Californium-252 production at the High Flux Isotope Reactor − I: Validation study using campaign data

This paper presents a series of 252Cf production validation and code-to-code comparison studies performed based on data from the production campaigns at the High Flux Isotope Reactor (HFIR). These studies support efforts to convert HFIR from using highly enriched uranium (HEU) fuel to low-enriched uranium (LEU) fuel. HFIR must maintain its world-class performance and missions following this conversion, and because 252Cf is a vital neutron-emitting radioisotope used for a variety of high-impact applications (e.g., reactor startup, cancer treatment), the ability to efficiently produce 252Cf must be preserved. In this work, the HFIRCON, Shift, ORIGEN, and TCOMP codes were deployed, and several sets of data libraries were investigated to better understand the calculation codes and the data biases. As-loaded target composition data, as-run irradiation history data, and post-irradiation measurements from recent multi-cycle irradiation campaigns of the HEU core were used to validate and determine methodology biases. The findings demonstrated a good agreement, with results falling within 3 standard deviations of measurements. This paper lays the ground work for the second paper, which evaluates and compares 252Cf production and safety metrics with the HEU core and a proposed LEU core.

Californium-252 production at the High Flux Isotope Reactor - II: Comparison between the highly enriched uranium and a proposed low-enriched uranium core

This is the second paper on a 252Cf production study performed in support of efforts to convert the High Flux Isotope Reactor (HFIR) from highly enriched uranium (HEU) to low-enriched uranium (LEU) fuel. The first paper primarily focuses on validating computational tools and nuclear data. This companion paper evaluates another critical aspect: the 252Cf production capability with a proposed LEU core. HFIR must maintain its world-class performance and missions following conversion and because 252Cf is a vital, multipurpose neutron-emitting radioisotope, the ability to efficiently produce 252Cf must be preserved. In this study, the HFIRCON transport and depletion tool, several nuclear data libraries, and Campaign 78 data were used to compute 252Cf production, sensitivity, and safety metrics. Results indicate the 252Cf production and production rates are slightly higher with a 95MWth LEU core compared with those obtained with the 85MWth HEU core. Additionally, the target peak fission rate densities, discharge cumulative fission densities, and heat deposition rates with the LEU core are within a few percent of those calculated with the HEU core. The findings suggest HFIR’s 252Cf production capability can be effectively maintained with an LEU core without adversely affecting the safety metrics.

Taillefer—A Tool for Sensitivity Analysis and Uncertainty Propagation Studies for Steady-State Thermal-Hydraulic Simulations of Involute Fuel Element Research Reactors

Taillefer is a versatile Python tool for carrying out Sensitivity Analysis (SA) and uncertainty propagation (UP) studies based on Monte Carlo sampling. Developed with the primary goal of investigating sensitivities and uncertainties of steady-state thermal-hydraulic (SSTH) safety parameters of the high-performance research reactors Forschungs Neutronenquelle Heinz Maier-Leibnitz (FRM II) in Garching, Germany, and the Réacteur à Haut Flux (RHF) in Grenoble, France, it can also be used for a large variety of other modeling problems. The work presented here aims to explain the underlying mathematical background of SA and UP studies with Taillefer and to show some steps to verify these routines. Furthermore, a real-life application example is provided that demonstrates Taillefer’s use in SSTH analysis of the RHF. For this purpose, Taillefer is coupled to the external thermal-hydraulic software PLTEMP/ANL, which is one of the codes used at FRM II and RHF to access SSTH performance and safety parameters. Determining these crucial quantities is part of identifying possible low-enriched uranium (LEU) core designs that are suitable to replace the currently used highly enriched uranium fuels of the two reactors, supporting global nonproliferation efforts. Taillefer is a powerful tool in these conversion studies, as it increases the reliability of the LEU safety parameters by providing information about sensitivities and uncertainties in addition to the nominal values predicted by the thermal-hydraulic software.

Temperature feedback reactivity analysis for LEU-Fuelled SLOWPOKE-2 research reactor using DRAGON5 and DONJON5 codes

The purpose of this study is to provide a new deterministic model for the SLOWPOKE-2 nuclear research reactor at École Polytechnique de Montréal (EPM) with LEU (Low Enriched Uranium) core. Using the latest release of the code system DRAGON5 and DONJON5 and the cross-section data library ENDFB.VII rel.1 evaluation, the developed model is applied to simulate the neutronic behavior of the SLOWPOKE-2 research reactor. We studied the separate temperature effects of the main components of the core (i.e., fuel, coolant/moderator, beryllium reflector, and water reflector). The contribution of different physical phenomena to the RTC was assessed. The temperature reactivity feedback calculated using the deterministic approach based on the DRAGON5 and DONJON5 code system using the ENDF/B-VII.1 evaluated nuclear data library produced in the WIMD-D4 format is in good agreement. Therefore, this work proves the capability of DRAGON5 and DONJON5 codes, normally used for power reactors, to reliably simulate a low-power research reactor.

Automated reactor physics analysis framework of High Flux Isotope Reactor low-enriched uranium silicide dispersion fuel designs

The High Flux Isotope Reactor (HFIR) is a versatile research reactor that provides one of the highest steady-state neutron fluxes of any reactor in the world. The HFIR reactor physics team investigated the conversion of the current 93 wt% highly enriched uranium U3O8-Al dispersion fuel to a 19.75% low-enriched uranium (LEU) U3Si2-Al dispersion fuel. The team continuously develops a Python module to streamline the analysis steps required for an LEU core design to ensure reproducible and agile design iteration. The Python module automates the data processing between analysis steps and automates the input perturbation for branch calculations and design changes. The automated framework has proven to significantly increase the efficiency and reproducibility of the reactor physics team to design High Flux Isotope Reactor (HFIR) LEU cores and thoroughly analyze performance metrics, safety metrics, and thermal safety margins. Consequently, the team can now respond rapidly to fuel fabrication engineer and thermal-hydraulic-structural analyst requests.Numerous combinations of LEU fuel designs are explored, of which two LEU fuel designs are presented in this paper: a low density silicide design, and a high-density silicide design. Results show that both designs meet or exceed safety and performance metrics with exception for minor differences caused by the hardened spectrum from LEU.

The low-enriched uranium core design of a MW heat pipe cooled reactor

The heat pipe cooled reactor usually has compact core and fast neutron spectrum, which requires highly enriched fuel. For better nuclear non-proliferation performance, a low-enriched uranium reactor core design, UPR-SL, has been proposed on the basis of the existing heat pipe cooled reactor core design UPR-S in this paper. A high-density UN fuel was used to replace the original UO2 fuel to reduce the fuel enrichment. The core diameter of 1 m and the 1 MW thermal power were inherited from UPR-S. The new reactor has a total of 109 heat pipes, 436 fuel rods, and 88 moderator rods in the active zone. The core length and control system were redesigned using the SARAX code system. The lifetime, power distribution, control rod worth and reactivity feedback were analyzed. The results show that using high-density fuel, increasing the axial length of the active core and introducing local neutron moderation into the active core are effective to reduce the fuel enrichment for a heat pipe cooled reactor with no degradation of core performance. The softened neutron spectrum brought stronger feedback to the core.

Numerical investigation of heat transfer characteristics of moderator assembly employed in a low-enriched uranium nuclear thermal propulsion reactor

The design of a nuclear thermal propulsion (NTP) reactor based on low-enriched uranium (LEU) requires additional moderator elements in the core to physically meet the critical requirements. This design softens the core energy spectrum and can provide more thermal neutrons for the fission reaction, but the heat transfer characteristics between the fuel and moderator assembly are more complex. Aiming at the typical LEU unit design, the heat transfer mathematical model is established using the principle of heat flow diversion and superposition. The model adopts the heat transfer relationship based on STAR-CCM+ simulation rather than the empirical expression used in the past literature to improve the applicability of the model. The heat transfer coefficients in the proposed model are evaluated under different Reynolds numbers and thermal power. The deviations between the proposed model and CFD simulation are analyzed. The results show that the calculation of the heat transfer coefficient between the proposed model and the CFD simulation maintains a good consistency, most of which are within 10%. It may provide a reliable and conservative temperature estimation model for future LEU core design.

Radiological consequence analysis for hypothetical accidental release from Nigerian Research Reactor-1

Radiological dispersion study is a key element in safety analysis report (SAR) of every nuclear facility for the purpose of emergency response planning. In this work, computational approach was used to determine the total effective dose and ground deposition at critical positions onsite and offsite of the Nigerian Research Reactor-1 (NIRR-1) facility which will be useful in the ongoing development of final SAR for NIRR-1 Low Enriched Uranium (LEU) core. In the methodology used, NIRR-1 LEU core was depleted with TRITON module of SCALE 6.2.3 code and the fission inventory in the core was calculated after a continuous operation at full power of 231.931MWD/MTU for 918 Effective Full Power Days (EFPD) at an operation regime of 3 h per day, 3 days per week and 48 weeks per year. Hot Spot was employed for atmospheric transport and dose calculations with consideration of different accidental scenarios in which 20%, 30%, 60% and 100% gaseous inventory was hypothetically released into the atmosphere. From the results obtained, the total effective dose to maximum exposed workers at 10 m and maximum exposed members of public at 300 m from the reactor were 3.10mSvand0.43mSv respectively for the worst-case scenario with 100% release while the maximum ground deposition was 5.5×106Bq/m2 with corresponding maximum ground shine dose rate of 7.5×10−4mSv/hr. This results are at least one order of magnitude below the dose limits recommended by the International Commission on Radiological Protection (ICRP) and indicate that the present LEU core of NIRR-1 is unlikely to cause any detectable health effect on workers and members of public in the event that 100% of its gaseous inventory is released into air in the environment. Hence it could unequivocally be said that the population is safe from the operation of NIRR-1 in its present location.

The Low-Enriched Uranium Core Design of a Mw Heat Pipe Cooled Reactor

The Low-Enriched Uranium Core Design of a Mw Heat Pipe Cooled Reactor

Comparative analysis of core life-time for the NIRR-1 HEU and LEU cores

In this study, the end of lifetime for the Nigeria Research Reactor-1 high enriched uranium and the low enriched uranium fuelled core were analysed. The SCALE 6.2.3 computational sequences, CSAS6 and TRITON coupled with KENO-VI code were employed for Eigen value and depletion calculations respectively. The study was carried out to demonstrate the advantage of conversion of miniature neutron source reactor from high enriched uranium to the low enriched uranium fuel. The results obtained showed that the clean cold core excess reactivity values of the high enriched uranium and the low enriched uranium fuelled cores were comparable with the experimental data and indicate the suitability of the computational approach used in this study. Similarly, the integral worth of top beryllium shims was greater in the high enriched uranium core in comparison with the low enriched uranium fuelled core and could be attributed to the increased neutron leakage caused by hardening of the neutron spectrum due to higher inventory of U-238 in the low enriched uranium core. Furthermore, the lifetime of the low enriched uranium fuelled core was found to be higher when compared to the high enriched uranium core and could also be attributed to the higher loading of U-235 and U-238 in the low enriched uranium core. The longevity of the low enriched uranium fuelled core in comparison to the high enriched uranium indicate an advantage for its longer time of utilization in neutron activation analysis and underscore the importance of conversion from high enriched uranium to low enriched uranium fuel.

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.

Startup Test Plan and Predictions for Highly Enriched Uranium to Low-Enriched Uranium Fuel Conversion at the University of Missouri Research Reactor

Nonpower reactors licensed by the U.S. Nuclear Regulatory Commission require a startup test plan as part of any facility modification to verify operability prior to resumption of operations. In order to support conversion of the University of Missouri Research Reactor from the use of highly enriched uranium to low-enriched uranium (LEU) fuel, a startup test plan has been devised to measure certain reactor physics parameters for the initial all-fresh LEU core licensing documentation that will be submitted. These parameters include the approach to critical, primary coolant void coefficient of reactivity, flux trap void coefficient of reactivity, determination of flux trap sample reactivity worth, radial and axial thermal neutron flux mapping, control blade worth calibration, primary and pool coolant temperature coefficient of reactivity, and flux mapping of experimental positions. In this paper, predictions for these parameters made using the Monte Carlo N-Particle Version 5 (MCNP5) radiation transport code are reported. These predictions will support the startup tests by providing a baseline set of expectations and additional insight into the performance of the LEU core.

Steady-State Safety Analysis of Ghana Research Reactor-1 With Low-Enriched-Uranium Core

Abstract Research reactors all over the world are expected to operate within certain safety margins just like pressurized water reactors and boiling water reactors. These safety margins mainly include onset of nucleate boiling ratio (ONBR), departure from nucleate boiling ratio (DNBR), and flow instability ratio (FIR) in addition to the maximum clad or fuel temperature and saturation temperature or boing point of the coolant inside the core of the reactor. This study carried out steady-state safety analysis of the Ghana Research Reactor-1 (GHARR-1) with low enriched uranium (LEU) core. Monte Carlo N-particle (MCNP) code was used to obtain radial and axial power peaking factors used as inputs in the preparation of the input file of plate temperature code of Argonne National Laboratory (PLTEMP/ANL code), which was then used to obtain the mentioned safety parameters of GHARR-1 with LEU core in this study. The data obtained on the ONBR were used to obtain the initiation of nucleate boiling boundary data with respect to the active length of the reactor core for various reactor powers. The obtained results for LEU core were also compared with that of the high enriched uranium (HEU) core. The results obtained show that the 34 kW GHARR-1 with LEU core is safe to operate just as the previous 30 kW HEU core was safe to operate.

Transient Studies on Low-Enriched-Uranium Core of Ghana Research Reactor–1 (GHARR-1)

Following the core conversion of Ghana’s miniature neutron source reactor (MNSR) from highly enriched uranium (HEU) to low-enriched uranium (LEU), there has been a change in the fuel composition, fuel, clad, and other reactor core parameters. Since the allowable core power in a nuclear reactor is limited by thermal considerations, this study presents transient analysis of the LEU core of Ghana Research Reactor−1 (GHARR-1). The transient study has been carried out using the Monte Carlo N-Particle code version 5 (MCNP5) and the Program for the Analysis of Reactor Transients (PARET)/Argonne National Laboratory (ANL) computational tools. The behavior of the reactor core at normal and accident conditions of large reactivity insertions was studied. Transient results obtained for accidental large reactivity insertions of 6.71 mk indicated that boiling might occur in the coolant because under such large reactivity insertions, the coolant temperature was close to the saturation temperature of the coolant. The results show that boiling will not occur in the core for other reactivity insertions of 1.94, 2.1, 2.99, 3.87, and 4.0 mk considering that the outlet coolant temperatures obtained are far below the saturation temperature of 100°C at a pressure of 1 atm. The clad and fuel meat temperatures obtained for all the reactivity insertions are far below the melting points of Zircaloy-4 clad material and UO2 fuel. The results of the power profiles obtained show that the reactor is inherently safe even under large reactivity insertion conditions. The results obtained were found to agree well with the available experimental results. Comparison of the results of the LEU core with the previous HEU core has shown that temperature rise in the LEU core is lower than that in the HEU core under reactor transient conditions.

Design optimization methods for high-performance research reactor core design

Common objectives of high-performance research reactors include neutron scattering, materials irradiation, and isotope production. The design of a core that optimizes these multiple objectives often presents a multidimensional (i.e., in design space), multiobjective optimization problem. The developed systematic approach discussed herein draws on response surface methodology to validate and leverage a surrogate model to serve in place of high-fidelity computational analyses. Optimization and design analysis methods leverage this surrogate model to provide a flexible tool for generating optimized designs and understanding the impact of design decisions on desired metrics. In applications to High Flux Isotope Reactor (HFIR) low-enriched uranium (LEU) core designs, neutronic, isotopic evolution, and thermal hydraulic analyses are used to generate key performance and safety metrics for assessing the feasibility and fitness of given designs. Three optimized designs that consider different desired metrics and constraints (e.g., key metric weighting and fabrication constraints) are presented, providing potential design options that satisfy the requirements for HFIR’s conversion from high enriched uranium to LEU fuel.

A commissioning rod assembly for the Jamaican SLOWPOKE-2 nuclear reactor.

Within the framework of the Material Management and Minimization Conversion Program of the U.S. Department of Energy's National Nuclear Security Administration, the Argonne National Laboratory approved the manufacturing of a Jamaica's SLOWPOKE-2 nuclear reactor (JM-1) mock-up, reactor removal tools, and commissioning rods by Polytechnique Montreal. This mock-up reactor was then used to practice dry runs of the JM-1 high enriched uranium core removal activities and fresh low enriched uranium core loading operations of the JM-1 reactor conversion. One of the most crucial elements in the commissioning of a new reactor core is the commissioning rod assembly. Hence, this paper presents the design, the fabrication, the calibration, and the dry runs performed at Polytechnique Montreal with commissioning rod assembly. The work was then resumed at Jamaica's International Centre for Environmental and Nuclear Sciences.

Fuel – clad chemical interaction evaluation of the TREAT reactor conceptual low-enriched-uranium fuel element

The Transient Reactor Test (TREAT) facility resides at the Materials and Fuels Complex (MFC) at Idaho National Laboratory (INL). The TREAT reactor is currently undergoing design and engineering studies for its conversion from a high enriched uranium (HEU) to a low-enriched uranium (LEU) core. The conceptual design of the LEU fuel element identified two main design differences compared with the HEU fuel element; namely, it will contain four times more fissionable material in its graphite matrix and distinct nuclear-grade Zirconium alloy, as Zircaloy-3 was used in the HEU fuel assembly and is not commercially available currently. These design changes may impact the magnitude of chemical interaction between fuel and cladding materials during physical contact under expected TREAT operation conditions and, therefore, was evaluated through a combination of experimental testing and thermodynamic modeling in order to determine implications for the fuel assembly. In this study, two potential cladding material types, Zircaloy-4 or Zr-1Nb alloys, were evaluated, and it was found for both material types that the extent of interaction and specific chemical reactions are minimal and no detrimental effect on the overall cladding properties is observed. The resulting interaction layer of 3–6 μm was measured after a 2-week exposure at 820 °C. The thermodynamic analysis was extended to temperatures beyond the TREAT reactor operation and accident conditions in order to give some insight that may be of interest for other reactor systems as the High Temperature Gas Reactors (operation above 1000 °C) and for Nuclear Reactor Severe Accident phenomenology study where the UO2 fuel could reach temperatures over 2800 °C and melt.

Reactors as a Source of Antineutrinos: Effects of Fuel Loading and Burnup for Mixed-Oxide Fuels

In a conventional light water reactor loaded with a range of uranium and plutonium-based fuel mixtures, the variation in antineutrino production over the cycle reflects both the initial core fissile inventory and its evolution. Under the assumption of constant thermal power, we calculate the rate at which antineutrinos are emitted from variously fueled cores, and the evolution of that rate as measured by a representative ton-scale antineutrino detector. We find that antineutrino flux decreases with burnup for Low Enriched Uranium cores, increases for full mixed-oxide (MOX) cores, and does not appreciably change for cores with a MOX fraction of approximately 75%. Accounting for uncertainties in the fission yields, in the emitted antineutrino spectra, and the detector response function, we show that the difference in core-wide MOX fractions at least as small as 8% can be distinguished using a hypothesis test. The test compares the evolution of the antineutrino rate relative to an initial value over part or all of the cycle. The use of relative rates reduces the sensitivity of the test to an independent thermal power measurement, making the result more robust against possible countermeasures. This rate-only approach also offers the potential advantage of reducing the cost and complexity of the antineutrino detectors used to verify the diversion, compared to methods that depend on the use of the antineutrino spectrum. A possible application is the verification of the disposition of surplus plutonium in nuclear reactors.

High-Fidelity Modeling and Simulation for a High Flux Isotope Reactor Low-Enriched Uranium Core Design

A high-fidelity model of the High Flux Isotope Reactor (HFIR) with a low-enriched uranium (LEU) fuel design and a representative experiment loading has been developed to serve as a new reference model for LEU conversion studies. With the exception of the fuel elements, this HFIR LEU model is completely consistent with the current highly enriched uranium HFIR model. Results obtained with the new LEU model provide a baseline for analysis of alternate LEU fuel designs and further optimization studies.The newly developed HFIR LEU model has an explicit representation of the HFIR-specific involute fuel plate geometry, including the within-plate fuel meat contouring, and a detailed geometry model of the fuel element side plates. Such high-fidelity models are necessary to accurately account for the self-shielding from 238U and the depletion of absorber materials present in the side plates. In addition, a method was developed to account for fuel swelling in the high-density LEU fuel plates during the depletion simulation. Calculated time-dependent metrics for the HFIR LEU model include fission rate and cumulative fission density distributions, flux and reaction rates for relevant experiment locations, point kinetics data, and reactivity coefficients.

Estimate of radiation release from MIT reactor with un-finned LEU core during Maximum Hypothetical Accident

The Massachusetts Institute of Technology Reactor (MITR-II) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. At 6 MW, it delivers neutron flux and energy spectrum comparable to light water reactor (LWR) power reactors in a compact core using highly enriched uranium (HEU) fuel. In the framework of non-proliferation policy, the international community aims to minimize the use of HEU in civilian facilities. Within this context, research and test reactors have started a program to convert 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 the conversion of U.S. domestic high performance reactors like MITR. The current study focuses on the impacts of MITR Maximum Hypothetical Accident (MHA), which is also the Design Basis Accident (DBA), with LEU fuel. The MHA for the MITR is postulated to be a coolant flow blockage in the fuel element that contains the hottest fuel plate. It is assumed that the entire active portion of five fuel plates melts. The analysis shows that, within a 2-h period and by considering all the possible radiation sources and dose pathways, the overall off-site dose is 302.1 mrem (1 rem = 0.01 Sv) Total Effective Dose Equivalent (TEDE) at 8 m exclusion area boundary (EAB) and a higher dose of 392.8 mrem TEDE is found at 21 m EAB. In all cases the dose remains below the 500 mrem total TEDE limit goal based on NUREG-1537 guidelines.