Core Neutronic Parameters Research Articles

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.

Calculation of core neutronic parameters in electron accelerator driven subcritical TRIGA reactor

Abstract Investigating the dependence of reactor core neutronic parameters on the accelerator related parameters such as source multiplication coefficient (K s ), beam profile and electron beam energy (E e ), have a significant impact on the design of future ADS reactors and the cost of their construction. Using the MCNPX code, core neutronic parameters were calculated for three eigenvalues levels (k s ) such as axial and radial distributions of neutron flux, effective multiplication coefficient (K eff), photoneutron yield (Y n/e ), net neutron multiplication (M), energy gain (G), energy constant gain (G 0), power peaking factor (P max/P ave), and importance of neutron source (φ*) for the tungsten target in the core of TRIGA reactor which was driven by an electron accelerator. According to the results, when K s increases safety margin and the required accelerator current (I e ) decrease, but the parameters G and φ* increase. Also, using the parabolic spatial distribution for the electron source instead of the uniform spatial distribution, the parameters Y n/e , G and M increase. In addition, by increasing E e from 100 to 500 MeV, Y n/e and G are improved, and I e and accelerating power (p acc) decrease. Therefore, these results suggest that the dependence of reactor core neutronic parameters on accelerator related parameters is necessary to optimize the design cost of an ADS reactor.

A meshless local Petrov–Galerkin method for solving the neutron diffusion equation

The goal of this study is to solve the neutron diffusion equation by using a meshless method and evaluate its performance compared to traditional methods. This paper proposes a novel method based on coupling the meshless local Petrov–Galerkin approach and the moving least squares approximation. This computational procedure consists of two main steps. The first involved applying the moving least squares approximation to construct the shape function based on the problem domain. Then, the obtained shape function was used in the meshless local Petrov–Galerkin method to solve the neutron diffusion equation. Because the meshless method is based on eliminating the mesh-based topologies, the problem domain was represented by a set of arbitrarily distributed nodes. There is no need to use meshes or elements for field variable interpolation. The process of node generation is simply and fully automated, which can save time. As this method is a local weak form, it does not require any background integration cells and all integrations are performed locally over small quadrature domains. To evaluate the proposed method, several problems were considered. The results were compared with those obtained from the analytical solution and a Galerkin finite element method. In addition, the proposed method was used to solve neutronic calculations in the small modular reactor. The results were compared with those of the citation code and reference values. The accuracy and precision of the proposed method were acceptable. Additionally, adding the number of nodes and selecting an appropriate weight function improved the performance of the meshless local Petrov–Galerkin method. Therefore, the proposed method represents an accurate and alternative method for calculating core neutronic parameters.

Developing a Full-Core MCNP6 and RELAP5 Model of the European Pressurized Reactor Using NWURCS

The European pressurized reactor (EPR) is classified as a Generation III+ reactor. It differs from a conventional pressurized water reactor in many aspects, one of which is the core design. This evolutionary reactor lends itself to new fuel designs, such as thorium-based fuels. To perform new design calculations, a base case model needs to be established because the detailed models that are currently available are either proprietary or regulated. This paper therefore presents such a model based on the Monte Carlo method. This method is a valuable component of reactor neutronic calculations because geometry and materials can be accurately modeled.We modeled a full core of the EPR using MCNP6, in which the individual fuel pin geometry and material definitions were used together with radial and axial temperature characterization based on fuel assemblies considered as nodes. Data for both the neutronic and thermal-hydraulic models were mainly obtained from the U.S. EPR Final Safety Analysis Report (FSAR) [Rev. 5, AREVA (2013)].The neutronic and some thermal-hydraulic results were compared with data from the EPR FSAR. The following core neutronic parameters compared well with the FSAR data: the boron worth, axial flux distribution, neutron flux spectrum, reactivity coefficients, and control rod worth. However, the delayed neutron fraction showed a somewhat larger difference compared to the FSAR. Given this verification with the FSAR, confidence in the MCNP6 EPR model was therefore established. The model that we have developed serves as the basis for the follow-on study of introducing thorium in the EPR core.

The change of radial power factor distribution due to RCCA insertion at the first cycle core of AP1000

The using of a computer program for the PWR type core neutronic design parameters analysis has been carried out in some previous studies. These studies included a computer code validation on the neutronic parameters data values resulted from measurements and benchmarking calculation. In this study, the AP1000 first cycle core radial power peaking factor validation and analysis were performed using CITATION module of the SRAC2006 computer code. The computer code has been also validated with a good result to the criticality values of VERA benchmark core. The AP1000 core power distribution calculation has been done in two-dimensional X-Y geometry through ¼ section modeling. The purpose of this research is to determine the accuracy of the SRAC2006 code, and also the safety performance of the AP1000 core first cycle operating. The core calculations were carried out with the several conditions, those are without Rod Cluster Control Assembly (RCCA), by insertion of a single RCCA (AO, M1, M2, MA, MB, MC, MD) and multiple insertion RCCA (MA + MB, MA + MB + MC, MA + MB + MC + MD, and MA + MB + MC + MD + M1). The maximum power factor of the fuel rods value in the fuel assembly assumedapproximately 1.406. The calculation results analysis showed that the 2-dimensional CITATION module of SRAC2006 code is accurate in AP1000 power distribution calculation without RCCA and with MA+MB RCCA insertion.The power peaking factor on the first operating cycle of the AP1000 core without RCCA, as well as with single and multiple RCCA are still below in the safety limit values (less then about 1.798). So in terms of thermal power generated by the fuel assembly, then it can be considered that the AP100 core at the first operating cycle is safe.

Investigation of the influence of various reflector elements material types on the core nuclear parameters of the 2 MW Triga Mark-II research reactor using the Monte Carlo code

The influence of variation in the kind of material of neutron reflector elements on the core neutronic parameters of the Moroccan Triga Mark-II research reactor was examined. The comparison was made between three kinds of reflector element materials: graphite, beryllium, and beryllium oxide. MCNP5 Monte Carlo computations of the effective neutron multiplication factor (${k}_{{eff}})$ of the reactor according to the number of reflector material elements inserted in the reactor core for these three reflector elements materials families were carried out. The results found could be used in order to increase the ${k}_{{eff}}$ parameter without changing the size of the reactor.

Fuel loading pattern with 50% MOX fuel in HPR1000 core

Fuel loading pattern optimization with MOX (mixed oxide of plutonium and uranium) fuel is an important aspect for GIII LWR. For the HPR1000 core, an optimized MOX fuel is designed. Based on which, two loading patterns with 50% MOX fuel are designed. One is 1/4 core annually reloading pattern, and the other is 18-month reloading pattern. The core neutronic parameters are investigated by comparing with the 100% UO2 core. The MOX fuel has impacts on the nuclear characteristic parameters, but the 50% MOX core still satisfies the nuclear and safety criteria samilar to that considered in 100% UO2 core. The results of this study prove that the GIII LWR HPR1000 has excellent capability while loading with 50% MOX fuel.

Estimation of sensitivity coefficients of core characteristics based on reduced-order modeling using sensitivity matrix of assembly characteristics

ABSTRACTWe propose an estimation method of sensitivity coefficients of core neutronics parameters based on a multi-level reduced-order modeling approach. The idea is to use lower-level models to identify the dominant input parameter variations, constrained to the so-called active subspace, which are employed to determine the sensitivity coefficients of the core neutronic parameters. In our implementation, the lower-level model is represented by two-dimensional assembly calculations, which are employed in the preparation of the few-group cross-sections for core-wide calculations. The active subspace basis is estimated using the singular value decomposition of sensitivity matrices of assembly neutronics parameters. In numerical verification calculation, sensitivity coefficients of core characteristics for a typical three-loop PWR equilibrium-cycle are estimated using the proposed method and the direct method. Comparison of these two results shows that the proposed method well reproduces the results obtained by the direct method with lower calculation costs. Through the verification calculations, applicability of the proposed method to practical light water reactor analysis is confirmed.

Neutronic and thermal-hydraulic analysis of fission molybdenum-99 production at Tehran Research Reactor using LEU plate targets

Efficient and safe production of molybdenum-99 (99Mo) radiopharmaceutical at Tehran Research Reactor (TRR) via fission of LEU targets is studied. Neutronic calculations are performed to evaluate produced 99Mo activity, core neutronic safety parameters and also the power deposition values in target plates during a 7 days irradiation interval. Thermal-hydraulic analysis has been also carried out to obtain thermal behavior of these plates. Using Thermal-hydraulic analysis, it can be concluded that the safety parameters are satisfied in the current study. Consequently, the present neutronic and thermal-hydraulic calculations show efficient 99Mo production is accessible at significant activity values in TRR current core configuration.

Sensitivity Analysis of Core Neutronic Parameters in Electron Accelerator-driven Subcritical Advanced Liquid Metal Reactor

Calculation of the core neutronic parameters is one of the key components in all nuclear reactors. In this research, the energy spectrum and spatial distribution of the neutron flux in a uranium target have been calculated. In addition, sensitivity of the core neutronic parameters in accelerator-driven subcritical advanced liquid metal reactors, such as electron beam energy (E e ) and source multiplication coefficient (k s ), has been investigated. A Monte Carlo code (MCNPX_2.6) has been used to calculate neutronic parameters such as effective multiplication coefficient (k eff ), net neutron multiplication (M), neutron yield (Y n / e ), energy constant gain (G 0 ), energy gain (G), importance of neutron source (4*), axial and radial distributions of neutron flux, and power peaking factor (P max /P ave ) in two axial and radial directions of the reactor core for four fuel loading patterns. According to the results, safety margin and accelerator current (Ie) have been decreased in the highest case of ks, but G and 4* have increased by 88.9% and 21.6%, respectively. In addition, for LP1 loading pattern, with increasing Ee from 100 MeV up to 1 GeV, Y n / e and G improved by 91.09% and 10.21%, and I e and P acc decreased by 91.05% and 10.57%, respectively. The results indicate that placement of the NpePu assemblies on the periphery allows for a consistent k eff because the NpePu assemblies experience less burn-up.

Effect of burnup on the neutronic parameters of ITU TRIGA Mark II research reactor

This study aims to estimate burnup of the fuel elements for the Istanbul Technical University TRIGA Mark II Research and Training Reactor using a Monte Carlo-based burnup-depletion code. Effect of burnup on the core neutronic parameters, effective core multiplication factor, fast/epithermal/thermal neutron fluxes, and core-average neutron spectrum, and incoming neutron spectrum of the piercing beam port (PBP), is investigated at the Beginning of Life (BOL) and End of Life (EOL). Operational data peculiar to a selected operation sequence, which contains positions of CRs, power level of the reactor, material temperatures and latest core map, are used to determine the current fuel burnup of fuel elements at the time under consideration. A specific operation sequence is selected for the analysis. Furthermore, all control rods are considered fully withdrawn to assess the excess reactivity. Results are obtained using MONTEBURNS2 with ENDFB/V-II.1 neutron/photon library for a full power of 250 kW. Neutron cross-section libraries at the full-power operating temperatures are generated using NJOY. From the results, the calculated burnup values of the core at the sequence considered and EOL are found to be 420 MWh and 560 MWh, respectively. Remaining excess reactivity is calculated to be less than 0.3 $. It is observed that core average thermal neutron flux reduces by 1 % while the fast and epithermal neutron fluxes remain almost unchanged.

ORPHEE research reactor: 3D core depletion calculation using Monte-Carlo code TRIPOLI-4®

ORPHEE is a research reactor located at CEA Saclay. It aims at producing neutron beams for experiments. This is a pool-type reactor (heavy water), and the core is cooled by light water. Its thermal power is 14MW. ORPHEE core is 90cm height and has a cross section of 27×27cm2. It is loaded with eight fuel assemblies characterized by a various number of fuel plates. The fuel plate is composed of aluminium and High Enriched Uranium (HEU). It is a once through core with a fuel cycle length of approximately 100 Equivalent Full Power Days (EFPD) and with a maximum burnup of 40%. Various analyses under progress at CEA concern the determination of the core neutronic parameters during irradiation. Taking into consideration the geometrical complexity of the core and the quasi absence of thermal feedback for nominal operation, the 3D core depletion calculations are performed using the Monte-Carlo code TRIPOLI-4® (Both and Peneliau, 1996; Trama et al., 2012; Brun, 2011). A preliminary validation of the depletion calculation was performed on a 2D core configuration by comparison with the deterministic transport code APOLLO2 (Sanchez, 2009). The analysis showed the reliability of TRIPOLI-4® to calculate a complex core configuration using a large number of depleting regions with a high level of confidence.

Sensitivity analysis of core neutronic parameters in accelerator driven subcritical reactors

In this paper, sensitivity of the ADSRs core neutronic parameters to the accelerator related parameters such as beam profile, source multiplication coefficient (ks) and proton beam energy (Ep) has been investigated. TRIGA reactor has been considered as the case study of the problem. Monte Carlo code MCNPX has been used to calculate neutronic parameters such as: effective multiplication coefficient (keff), net neutron multiplication (M), spallation neutron yield (Yn/p), energy constant gain (G0), energy gain (G), importance of neutron source (φ*), axial and radial distributions of neutron flux and power peaking factor (Pmax/Pave) in two axial and radial directions of the reactor core for three eigen values levels (ks) including: 0.91, 0.97 and 0.99. According to the results, using a parabolic spatial distribution instead of a uniform spatial distribution increases the relative differences of spallation neutron yield, net neutron multiplication and energy gain by 4.74%, 4.05% and 10.26% respectively. In consequence the required accelerator current (Ip) will be reduced by 7.14% to preserve the reactivity. Although safety margin is decreased in highest case of ks, but energy gain increases by 93.43% and the required accelerator current and importance of neutrons source decrease by 48.3% and 2.64% respectively. In addition, increasing Ep from 115MeV up to1GeV, improves spallation neutron yield and energy gain by 2798.71% and 205.12% and decreases the required accelerator current and power by 96.83% and 72.44%, respectively. Therefore, our results are indicative of the fact that investigating sensitivity of the core neutronic parameters to the accelerator related parameters are necessary in order to optimally design a cost-efficient ADSR.

Neutronic evaluations for MNSR research reactor core conversion from HEU to LEU

We studied enrichment reduction evaluations for Iranian Miniature Neutron Source Reactor (MNSR) preserving prior reactivity limitations. Utilization of high density Low Enriched Uranium (LEU) fuels such as UO2, U3Si2 and U9Mo have been addressed from neutronics point of view.The Monte Carlo code MCNP4C has been used for studying the reactor core neutronic parameters. The feasibility of candidate LEU fuels have been validated via the MNSR reference fuel parameters.The results show that the average neutron flux in irradiation channels of LEU core configurations is reduce within 7%. According to the calculated reactivity parameters and thermal neutron flux in irradiation channels, this reduction would not affect the reactor safety as well as its applicability.

Miniature neutron source reactor burnup calculations using IRBURN code system

Fuel consumption of Iranian miniature neutron source reactor (MNSR) during 15years of operation has been investigated. Reactor core neutronic parameters such as flux and power distributions, control rod worth and effective multiplication factor at BOL and after 15years of irradiation has been calculated.The Monte Carlo-based depletion code system IRBURN has been used for studying the reactor core neutronic parameters as well as the isotopic inventory of the fuel during burnup. The precision and accuracy of the implemented model has been verified via validation the results for neutronic parameters in the MNSR final safety analysis report.The results show that keff decreases from 1.0034 to 0.9897 and the total U-235 consumption in the core is about 13.669g after 15years of operational time. Finally, our studying shows the consumption rate of MNSR is about 1%.

Analysis of core life-time and neutronic parameters for HEU and potential LEU/MEU fuels in a typical MNSR

For the standard HEU and various potential LEU and MEU cores of MNSR systems, a comparative neutronic study has been carried out using the cluster option in the WIMS-D/4 and RZ-model in the CITATION code. For the HEU core, the computed values of excess reactivity, control rod worth, shutdown margin and the value of equilibrium xenon worth are in good agreement with the corresponding values found in literature. The 235U enrichment required for the LEU UO2–Zr core found as 12.5% is in good agreement with the 12.6% found in literature. Compared with 1100g total uranium loading for the HEU core, the corresponding requirement for UO2, U3Si–Al, U3Si2–Al and U–9Mo–Al cores have been found as 10733.5, 6319, 5515 and 7549g respectively. For the HEU core, the computed 18.98mk value of the total worth of top Be shim plates is in good agreement with the corresponding experimental value of 18.5mk reported in FSAR. The corresponding computed values for LEU and MEU cores are 17.28mk for UO2–Zr fuel, 17.47mk for UO2–Al, 18.37mk for U3Si2–Al. Compared with 21.2year value for the HEU core, 30.6years cycle length for UO2–Zr LEU core has been found.

Burn up calculations for the Iranian miniature reactor: A reliable and safe research reactor

Presenting neutronic calculations pertaining to the Iranian miniature research reactor is the main goal of this article. This is a key to maintaining safe and reliable core operation. The following reactor core neutronic parameters were calculated: clean cold core excess reactivity ( ρ ex ), control rod and shim worth, shut down margin (SDM), neutron flux distribution of the reactor core components, and reactivity feedback coefficients. Calculations for the fuel burnup and radionuclide inventory of the Iranian miniature neutron source reactor (MNSR), after 13 years of operational time, are carried out. Moreover, the amount of uranium burnup and produced plutonium, the concentrations and activities of the most important fission products, the actinide radionuclides accumulated, and the total radioactivity of the core are estimated. Flux distribution for both water and fuel temperature increases are calculated and changes of the central control rod position are investigated as well. Standard neutronic simulation codes WIMS-D4 and CITATION are employed for these studies. The input model was validated by the experimental data according to the final safety analysis report (FSAR) of the reactor. The total activity of the MNSR core is calculated including all radionuclides at the end of the core life and it is found to be equal to 1.3 × 10 3 Ci. Our investigation shows that the reactor is operating under safe and reliable conditions.

Annular Fast Reactor Cores with Low Sodium Void Worth for TRU Burning

Annular sodium-cooled fast reactor cores [600 MW(electric)] with low sodium void worth are developed for burning transuranic nuclides discharged from light water reactors. The several core design variants are developed by changing the core configuration, the core height, the fuel assembly design and type of nonfuel assemblies in the core, and their core performance parameters including safety-related reactivity coefficients are analyzed and inter-compared. The study focuses on the core neutronic parameters without going into the detailed safety and material compatibility studies. The study shows that the several cores of the annular type can be designed to have low sodium void worth, high transmutation capability, and all the negative temperature reactivity coefficients except for the positive one related to coolant expansion that can be compensated for by the reactivity coefficients by the fuel axial expansion and the fuel Doppler effects under the off-normal events, which increase temperatures. Of the cores considered, the use of a larger central control region and fuel assemblies with high coolant flow area in the core boundaries is found to be the most effective and simple way to achieve low sodium void worth and high transmutation capability.