LEU Core Research Articles

Assessment of Axial Power Peaking Factors in GHARR-1 LEU Core: A Decadal Simulation Analysis

Assessment of Axial Power Peaking Factors in GHARR-1 LEU Core: A Decadal Simulation Analysis

Feasibility study for the adoption of multi-capsule irradiation protocol in the conduct of k0-based INAA using the GHARR-1 MNSR

The Ghana Research Reactor-1 (GHARR-1) is currently a 23 cm length LEU core Miniature Neutron Source Reactor (MNSR) with a 13 % U-235 enrichment having 335 fuel rods, 15 dummy rods and a central control rod for neutron regulation. It has 10 accessible irradiation channels of approximate length of 17 cm and formed part of the shim tray structure which was considered for active routine experiment in the present work. Unlike the routine characterization of the inner irradiation channel of the Ghana Research Reactor- 1 facility which involves mainly the introduction of a single irradiation capsule of length, 5 cm loaded with samples, the present study adopted a multi-capsule scheme in which three (3) capsules each of length 5 cm were introduced into the 23 cm long irradiation channel to scientifically interrogate the feasibility and hence provide a sound basis for extending and improving the accessible irradiation space by 60% during the utilization of the GHARR-1, especially for irradiation involving intermediate and long lived radionuclides. The objective is also to achieve the characterization of neutron spectrum and determine their spatial distribution for close to the full length of the irradiation channel. Validation protocols based on k0 method were developed through the analysis of some reference materials for a careful study of irradiation, decay and counting scheme which achieved optimum radionuclide selectivity. The overall approach adopted for the flux characterization involves the preparation of flux monitors, packaging of three (3) capsules each of both bare and cadmium cover samples, irradiation at bottom, middle and top spatial demarcation of the irradiation channel, with each demarcation being the region of the 5 cm length of the bottom, middle and top irradiation capsules. Sample counting was undertaken using the HPGe detector after the samples were allowed specific decay time after irradiation and prior to counting. Flux monitors were used for the flux characterization and reference materials were used for the validation protocol. Spectrum acquisition was made possible through the use of the Gamma Vision Software and spectrum parameters (thermal to epithermal neutron flux ration (f), epithermal neutron shaping factor (α), thermal, epithermal, and fast fluxes) were determined. Results obtained showed increasing f-value across the irradiation column as, 18.5 ± 1.7, 21.0 ± 2.2 and 23.0 ± 7.08 respectively from the bottom capsule to the top capsules. The corresponding epithermal neutron shaping factor (α-value) varied as, −0.096 ± 0.029, −0.18 ± 0.036 and −0.20 ± 0.06 from the bottom to the top capsule. The experimental results determined in the bottom, middle and top capsule irradiation column for thermal, epithermal and fast fluxes are 4.60 × 1011±2.5 × 1010, 2.49 × 1010±5.98 × 108, 9.24 × 1010±2.2 × 109; 4.21 × 1011±1.01 × 1010, 2.01 × 1010±4.82 × 108, 4.81 × 1010±1.15 × 109; and 3.90 × 1011±9.36 × 109, 1.65 × 1010±3.90 × 108, 4.82 × 1010±1.16 × 109 respectively. The validation protocol using standard reference materials and treating each capsule with separate reactor characterization parameters indicated respective z-score distribution within a 95% confidence interval.

WWR-K reactor LEU core design optimization for improving the experimental characteristics

The increase in demand for irradiation activities in the WWR-K reactor dictates the need to increase the number of irradiation positions in the core. Several core designs were considered by simulation with using MCNP6 code. As a result, core consisting of 22 FAs of the 1st type and 4 FAs of the 2nd type was chosen. An analysis of the calculation results showed that the new design of the core improves the experimental characteristics of the reactor and, at the same time, ensures nuclear safety. In particular, the number of irradiation positions in the fuel area increases from 3 to 5, in the new core design. The maximum thermal neutron flux increases to 2.1x1014 cm-2s−1. This paper presents the results of neutron analysis of the new WWR-K reactor core design in comparison with the existing.

Application of SCALE code for reactor physics analysis of the as-built Nigeria miniature neutron source reactor operating with low enriched uranium fuel

This study presents a broad reactor physics analysis including, neutronic, burnup, and decay heat of the ‘as-built’ NIRR-1 LEU core performed using SCALE 6.2.3 code package. First, verification of neutronic parameters of the NIRR-1 was performed by modeling the reactor ‘as-built’ LEU core. Thereafter validation of the model was done by comparing calculated result with measured data during the commissioning of the reactor. The results showed that the estimated neutronic parameters using KENO-VI code of SCALE 6.2.3 code package are in close agreement with the measured data. Hence, KENO-VI can be used for detailed neutronic assessment of MNSRs. The total decay heat at the end of core life after operation for 216 EFPD was estimated to be 1620.15 W and was majorly due to the contribution from fission products which was estimated to be 1608.10 W. This value explains why MNSRs are cooled through natural convection of the coolant during operation and at shutdown conditions. Additionally, the burnup of the core was determined to be 0.53 wt% which is within the expected burnup for a typical MNSRs (less than1 wt%) with a core lifetime of about 200 EFPD (∼10 years). Furthermore, the mass of 239Pu buildup after 216 EFPD was far less than the IAEA category III nuclear material and indicates a huge advantage of research reactor conversion in reducing proliferation risk.

Heterogeneous model based burnup and its impact on integral parameters of IAEA MTR benchmark

Computational benchmarks are an important source for verifying the codes and modeling methodology in nuclear reactor analysis. The current study aims at utilizing the newly developed burnup capability of open source code OpenMC to perform analyses of the IAEA 10-MW MTR benchmark reactor. The whole core model developed for the benchmark is verified against the Serpent results from an earlier study. The effect of the volume-weighted homogenized (VWH) unit-cell model against a fully heterogeneous fuel assembly model to obtain burnup dependent number densities is determined. It was observed that the comparison of four factors (resonance escape, fast fission, thermal utilization, and reproduction factor) for two modeling options shows relative differences which increase with burnup. These differences translate to about a 350 pcm difference in the HEU case while the LEU core shows a relatively larger difference of about 500 pcm in integral core parameters. Root Mean Square (RMS) and maximum relative error in assembly power as large as 0.83 and 2 % respectively are seen for HEU and LEU cases. Although standard fuel element (SFE) and control fuel element (CFE) are geometrically significantly different, some studies have used VWH unit-cell based on SFE to generate and use number densities for CFE as well. The effect of this modeling choice is also analyzed. This led to a relative difference of more than 20 % in the concentration of Pu-239 for both HEU and LEU studies, with SFE based number density being overpredicted. Xe-135 concentration also shows about a 20 % difference for the HEU case. The difference in the whole core neutron multiplication factor, SFE vs SFE + CFE individual burnt fuel vector, is negligible. On the other hand, power RMS error and maximum relative difference for power fraction exhibit a maximum value of 0.84 and 2.24 respectively. This study can serve as a reference for future verification efforts using the IAEA 10-MW MTR benchmark reactor.

Assessment of neutronics performance of U3Si, U3Si2, and U–9Mo fuels for the support of MNSR conversion using SCALE 6.2.3 computational tools

In spite of the acceptability and production of Uranium Oxides (UO2) with Low Uranium Enrichment (LEU-UO2) fuel in commercial scale, the use of uranium–silicide (U–Si) in place of the UO2 is one of the promising concepts being proposed to increase the accident tolerance of nuclear fuels. In this study, SCALE 6.2.3 computational tools have been used to model the Nigerian Research Reactor-1 (NIRR-1) which is a typical Miniature Neutron Source Reactor (MNSR) and calculated the core neutronic parameters such as Clean Core Cold Excess Reactivity (CCER), Control Rod Worth (CRW), Shutdown Margin (SDM), Safety Reactivity Factor (SRF) and Neutron flux distribution for the Highly Enriched Uranium (HEU) and candidates LEU cores. The results showed a close agreement between the calculated and measured data. The bias between calculated and measured data was attributed to the computational limitation of this study that was caused by a limited number of particles histories used in criticality calculations. The neutronic performance of the U3Si2–Al, U3Si–Al and U9Mo–Al fuels indicated the suitability of these candidates LEU fuels for conversion of MNSRs from HEU to LEU. The observed reduction in the thermal flux of the candidate LEU fuels was attributed to higher inventory of U-238 in LEU fuels when compared with the HEU and further underscore the need to raise their thermal power by approximately 10% for efficient utilization of the MNSR in Neutron Activation Analysis.

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.

The low enriched uranium miniature neutron source reactor (LEU-MNSR) neutron spectrum characterization for k0-INAA

Ghana’s highly enriched uranium (90.2%) research reactor has been converted to a low enriched uranium (13.0%) type. Its neutron parameters have been characterized for the application of k0-instrumental neutron activation analysis. The magnitude of the LEU epithermal neutron shaping factor (α) at the irradiation site increased by about 25% as compared to the HEU, while the thermal neutrons decreased by about 5%. This indicates that the epithermal neutron spectrum is much dependent on the 1/E1+α. Similarly, the measured fast neutron flux increased by about 22%. Generally, the results show that the LEU core’s neutron spectrum is relatively harder, i.e., there is a marginal increase in high-energy neutrons. The reactor spectrum parameters have been used to validate the k0-IAEA software. Two certified reference materials (Carbonate rock GBW 07108 (GBW2) and NIST Standard Reference Material 1646a, Estuarine Sediment (EST)) and the synthetic multi-element standard (SMELS I, II & III) were analyzed based on a routine irradiation scheme. The process quality assurance was evaluated based on a u-score test. Most (80 %) of the elements determined were within |u-score| < 1.64, while a few were observed within 1.964 > |u-score| > 1.64. The results show a suitable characterization process and the general applicability of the k0-IAEA software.

Investigating Possibilities of fuelling NIRR-1 with Low Enriched Uranium Silicide-Aluminum Dispersion Fuels

In the last quarter of 2018, low enriched uranium dioxide fuel with zirconium alloy cladding was used successfully to convert the core of NIRR-1 from HEU to LEU fuel and the removed core returned to the country of origin. The objective of this study was to investigate the possibility of fueling the same system with alternate LEU fuel for future replacement of the current fuel, without any unacceptable compromise in reactor performance. Having more than one fuel options available for the same reactor system will present Nigeria an opportunity of making good economic decisions at the end of the cycle of the current LEU fuel. The performance of low enriched uranium silicide aluminum dispersion fuels in the core of NIRR-1 has been investigated and the results were identical with that of similar studies conducted elsewhere for generic MNSR system. Some of the calculated reactor parameters using this alternate LEU fuel were closely identical with that of the old HEU core. The computer software selected for this studies were the SCALE code system and the VENTURE PC. While the SCALE code system was employed to generate a properly averaged multigroup cross section library for the investigated LEU core models for NIRR-1 system, the VENTURE PC was utilized to give criticality information, few group fluxes and power density distributions within the core of the modeled system.Keywords— Reactor, Reactivity, Fuel, Enriched, Silicide

MCNP Simulation of Physics Parameters of Dispersion Fuels for Conversion of NIRR-1 to LEU

The Nigeria Research Reactor-1 (NIRR-1) is one of the Commercial Miniature Neutron Source Reactors (MNSRs) sited outside China and scheduled for conversion under the auspices of Reduced Enrichment for Research and Test Reactors (RERTR) program. Since 2006, the reduction in the fuel enrichment of MSNR facilities from greater than 90% HEU cores to less than 20% LEU cores has been embarked upon. Consequently in this work, the physics parameters of three dispersion LEU fuels, which include U3Si, U3Si2, and U9Mo enriched to 19.75% were determined by the MCNP code to investigate their suitability for the conversion of NIRR-1 to LEU. The following reactor core physics parameters were computed for the LEU fuel options: clean cold core excess reactivity (ρex), control rod (CR) worth, shut down margin (SDM), neutron flux distributions in the irradiation channels and kinetics data (i.e. effective delayed neutron fraction, βeff and prompt neutron lifetime, lf). Results are compared with experimental and calculated data of the current HEU core and indicate that it would be feasible to use any of the LEU options for the conversion of commercial MNSR in general and NIRR-1 in particular from HEU to LEU.

Heat deposition analysis for the High Flux Isotope Reactor’s HEU and LEU core models

The High Flux Isotope Reactor at Oak Ridge National Laboratory is an 85MWth pressurized light-water-cooled and -moderated flux-trap type research reactor. The reactor is used to conduct numerous experiments, advancing various scientific and engineering disciplines. As part of an ongoing program sponsored by the US Department of Energy National Nuclear Security Administration Office of Material Management and Minimization, studies are being performed to assess the feasibility of converting the reactor’s highly enriched uranium fuel to low-enriched uranium fuel. To support this conversion project, reference models with representative experiment target loading and explicit fuel plate representation were developed and benchmarked for both fuels to (1) allow for consistent comparison between designs for both fuel types and (2) assess the potential impact of low-enriched uranium conversion. These high-fidelity models were used to conduct heat deposition analyses at the beginning and end of the reactor cycle and are presented herein.This paper (1) discusses the High Flux Isotope Reactor models developed to facilitate detailed heat deposition analyses of the reactor’s highly enriched and low-enriched uranium cores, (2) examines the computational approach for performing heat deposition analysis, which includes a discussion on the methodology for calculating the amount of energy released per fission, heating rates, power and volumetric heating rates, and (3) provides results calculated throughout various regions of the highly enriched and low-enriched uranium core at the beginning and end of the reactor cycle.These are the first detailed high-fidelity heat deposition analyses for the High Flux Isotope Reactor’s highly enriched and low-enriched core models with explicit fuel plate representation. These analyses are used to compare heat distributions obtained for both fuel designs at the beginning and end of the reactor cycle, and they are essential for enabling comprehensive thermal hydraulics and safety analyses that require detailed estimates of the heat source within all of the reactor’s fuel element regions.

Radioactivity and decay heat analysis of the MNSR with HEU and LEU fuels using the GETERA code

The knowledge of the decay heat of nuclear fuel is necessary for performing the reactor safety analysis. In this study, analysis of the radioactivity, decay heat and the isotopic mass variation with time post the shutdown of the Miniature Neutron Source Reactor (MNSR) for the potential LEU fuels (U3Si2-Al, U3Si-Al, U9Mo-Al, 19.75% enriched and UO2, 12.6% enriched) and the existing HEU fuel (UAl4-Al, 90% enriched) have been performed using the GETERA code. The results showed that the total radioactivities were 2.13E05, 2.58E05, 2.59E05, 2.13E05 and 2.57E05 Ci, and the total decay heats were 1379.9, 1677.0, 1677.2, 1380.6 and 1672.6 W instantly after the reactor shutdown for the LEU and HEU-UAl4 cores respectively. Finally, good agreements were noticed between the analytical, ORIGEN-2 and GETERA results.

Radioactivity and decay heat analysis of the MNSR with HEU and LEU fuels using the GETERA code

The knowledge of the decay heat of nuclear fuel is necessary for performing the reactor safety analysis. In this study, analysis of the radioactivity, decay heat and the isotopic mass variation with time post the shutdown of the Miniature Neutron Source Reactor (MNSR) for the potential LEU fuels (U3Si2-Al, U3Si-Al, U9Mo-Al, 19.75% enriched and UO2, 12.6% enriched) and the existing HEU fuel (UAl4-Al, 90% enriched) have been performed using the GETERA code. The results showed that the total radioactivities were 2.13E05, 2.58E05, 2.59E05, 2.13E05 and 2.57E05 Ci, and the total decay heats were 1379.9, 1677.0, 1677.2, 1380.6 and 1672.6 W instantly after the reactor shutdown for the LEU and HEU-UAl4 cores respectively. Finally, good agreements were noticed between the analytical, ORIGEN-2 and GETERA results.

Sensitivity analyses of the use of different reflector materials on the neutronics parameters of Tehran research reactor

In this paper, the effect of changes in neutron reflector type on neutronics parameters of Tehran research reactor is analyzed. In this study, at first, calculations for the HEU and LEU fuel configurations of the reactor core in which the light water is used as a neutron reflector in the core is done. Then, by using the reflectors such as graphite, beryllium and heavy water, changes on the neutronic parameters are examined. The results show that by altering the reflector, at HEU core configuration (compared with LEU), more changes appear in parameters such as neutron multiplication factor, average fast and thermal neutron flux, excess reactivity and shut down margin. Moreover, at LEU core configuration, changes are tangible specifically on parameters of cycle length and power peaking factor. In addition, the evaluated fuel temperature coefficient of reactivity is greater at HEU than LEU, while the temperature alteration of fuels represented greater influence on reactivity at LEU configuration.

A Comparative Study of the HEU, LEU Fuels: Assessing the Effect of the MNSR Fuel Burnup on the Effective Multiplication Factor and Neutron Fluxes

The GETERA and MCNP4C codes have been used to perform the core conversion analysis of the MNSR from the existing HEU to four potential LEU fuels. The effect of fuel burnup on the effective multiplication factor (keff) and the neutron fluxes in the irradiation sites was also investigated. Results showed that the “keff“ dropped sharply in the first days of the reactor operation time (Xe effect), then decreased slowly due to the fuel consumption. The axial thermal, epithermal and fast neutron flux distributions in the irradiation sites exhibited similar flux patterns for all types of fuel. The values of the axial thermal neutron flux were slightly lower in the LEU fuel. For the HEU fuel, the fuel residence time in the core was shorter than that for the LEU fuel. For both HEU and LEU cores, the changes in the neutron fluxes in the irradiation sites with burnup were negligible, so no burnup effect on the neutron fluxes was observed.

Monte Carlo modeling of the Pakistan Research Reactor-1 (PARR-1)

Pakistan Research Reactor-1 (PARR-1) is a swimming pool type research reactor established in 1965. Originally, it was designed to operate at 5MW utilizing 93% Highly Enriched Uranium (HEU) fuel. Due to proliferation concerns, the reactor core was converted to LEU core in 1992, utilizing Uranium Silicide (U3Si2-Al) fuel having 235U enrichment of 19.99% and with an up graded operating power of 9MW. The first LEU core was operated at full reactor power (9MW) composed of only fresh fuel and is termed as first high power core. The objective of this article is to develop the detailed model of this first high power core of PARR-1 employing Monte Carlo N-Particle (MCNP5) computer code. The MCNP5 is equipped with the cross-section library ENDF B-VI. To validate the MCNP model, the simulated values of initial criticality with control rod positions, excess reactivity, shutdown margin and combined control rod worth are compared with the published work. The relative error in criticality is 2.6% (with control rod positions 51.6% out of the core), 2.7% in excess reactivity, 6% in the shutdown margin and 3.3% for combined control rod worth has been observed. This model shall be applied to predict the power profile, peaking factors and feedback reactivity coefficients of the reactor in future.

Prompt neutron lifetime calculations for the NIRR-1 reactor

Prompt neutron lifetime calculations have been performed for the NIRR-1 reactor HEU and LEU cores using the 1/v insertion and the Adjoint flux weighing methods. Results of calculations obtained for the HEU and LEU cores are respectively 57.3+-0.8 and 47.5+-0.7 for the 1/v insertion and 56.9+-0.3 and 46.3+-0.5 for the Adjoint flux. There is a good agreement seen between the two methods for both cores. The prompt neutron lifetime was observed to be shorter in the LEU than for the HEU as expected. However, the Adjoint flux weighing method seemed to be the easiest method in calculating the prompt neutron lifetime for NIRR-1.

A comparative study of the neutron flux spectra in the MNSR irradiation sites for the HEU and LEU cores using the MCNP4C code

A comparative study for fuel conversion from the HEU to LEU in the Miniature Neutron Source Reactor (MNSR) has been performed in this paper using the MCNP4C code. The neutron energy and lethargy flux spectra in the first inner and outer irradiation sites of the MNSR reactor for the existing HEU fuel (UAl4–Al, 90% enriched) and the potential LEU fuels (U3Si2–Al, U3Si–Al, U9Mo–Al, 19.75% enriched and UO2, 12.6% enriched) were investigated using the MCNP4C code. The neutron energy flux spectra for each group was calculated by dividing the neutron flux by the width of each energy group. The neutron flux spectra per unit lethargy was calculated by multiplying the neutron energy flux spectra for each energy group by the average energy of each group. The thermal neutron flux was calculated by summing the neutron fluxes from 0.0 to 0.625eV, the fast neutron flux was calculated by summing the neutron fluxes from 0.5MeV to 10MeV for the existing HEU and potential LEU fuels. Good agreements have been noticed between the flux spectra for the potential LEU fuels and the existing HEU fuels with maximum relative differences less than 10% and 8% in the inner and outer irradiation sites.

Neutronic analysis for core conversion (HEU–LEU) of the low power research reactor using the MCNP4C code

Abstract Comparative studies for conversion of the fuel from HEU to LEU in the miniature neutron source reactor (MNSR) have been performed using the MCNP4C code. The HEU fuel (UAl4-Al, 90% enriched with Al clad) and LEU (UO2 12.6% enriched with zircaloy-4 alloy clad) cores have been analyzed in this study. The existing HEU core of MNSR was analyzed to validate the neutronic model of reactor, while the LEU core was studied to prove the possibility of fuel conversion of the existing HEU core. The proposed LEU core contained the same number of fuel pins as the HEU core. All other structure materials and dimensions of HEU and LEU cores were the same except the increase in the radius of control rod material from 0.195 to 0.205 cm and keeping the outer diameter of the control rod unchanged in the LEU core. The effective multiplication factor (keff ), excess reactivity (ρ ex ), control rod worth (CRW), shutdown margin (SDM), safety reactivity factor (SRF), delayed neutron fraction (β eff ) and the neutron fluxes in the irradiation tubes for the existing and the potential LEU fuel were investigated. The results showed that the safety parameters and the neutron fluxes in the irradiation tubes of the LEU fuels were in good agreements with the HEU results. Therefore, the LEU fuel was validated to be a suitable choice for fuel conversion of the MNSR in the future.

Calculation and comparison of xenon and samarium reactivities of the HEU, LEU core in the low power research reactor

Comparative studies for the conversion of the fuel from HEU to LEU in the Miniature Neutron Source Reactor (MNSR) have been performed using the MCNP4C and GETERA codes. The precise calculations of 135Xe and 149Sm concentrations and reactivities were carried out and compared during the MNSR operation time and after shutdown for the existing HEU fuel (UAl4–Al, 90% enriched) and the potential LEU fuels (U3Si2–Al, U3Si–Al, U9Mo–Al, 19.75% enriched and UO2, 12.6% enriched) in this paper using the MCNP4C and GETERA codes. It was found that the 135Xe and 149Sm reactivities did not reach their equilibrium reactivities during the daily operating time of the reactor. The 149Sm reactivities could be neglected compared to 135Xe reactivities during the reactor operating time and after shutdown. The calculations for the UAl4–Al produced the highest 135Xe reactivity in all the studied fuel group during the reactor operation (0.39mk) and after the reactor shutdown (0.735mk), It followed by U3Si–Al (0.34mk, 0.653mk), U3Si2–Al (0.33mk, 0.634mk), U9Mo–Al (0.3mk, 0.568mk) and UO2 (0.24mk, 0.448mk) fuels, respectively. Finally, the results showed that the UO2 was the best candidate for fuel conversion to LEU in the MNSR since it gave the lowest 135Xe reactivity during the reactor operation and after shutdown.