Neutronic Parameters Research Articles

Evaluation of OpenMC calculations coupling with PHITS, FLUKA, and GEANT4 for ADS study

The Accelerator-Driven Subcritical system (ADS) is a subcritical reactor driven by the spallation neutrons generated in a spallation target under the bombardment of proton beam from an accelerator. To study ADS neutronics, the external spallation neutrons should be investigated carefully for coupling with a reactor core. This work focuses on the investigation into the OpenMC calculations by using different nuclear data libraries coupling with different high-energy Monte Carlo programs such as PHITS, FLUKA, and GEANT4 for ADS neutronics simulation. MCNPX direct calculations is performed as the baseline to compare to coupling-calculations. Firstly, the information of leaking neutrons from a spallation target are calculated with different codes and physical models to evaluate difference from the direct calculations to the coupling-calculations. Secondly, the neutronics characteristics of OECD-ADS model are calculated by using OpenMC calculations coupling with a surface-source from three high-energy codes to compare to MCNPX direct calculations. The neutronics parameters of mono-energy neutron for OECD-ADS evaluated from OpenMC and MCNPX with ENDF/B-VII.0 data library or only from OpenMC with different libraries are found to be in good agreement. The effect of high-energy neutrons is evaluated to be difference of around 25% in beam requirements by MCNPX direct calculations with INCL model and PHITS-OpenMC coupling-calculation with ENDF/B-VII.0 below 20 MeV. The differences in beam requirements between different coupling-calculations with the same library are compared. Moreover, the discrepancies in beam requirements from PHITS-OpenMC with JENDL-4.0/HE from 20 MeV to 200 MeV are around 15%. Finally, both neutron profiles and neutron spectra from coupling-calculations agree well with these from MCNPX direct calculations outside the coupling region of target.

The 2021 IAEA software intercomparison for k0-INAA

In order to establish the variation between results in mass fractions due to software implementation, as measured by the k0-method for INAA, the IAEA has organized a software intercomparison. A complete set of test spectra and associated information was assembled. Efficiency curves, neutron spectrum parameters, correction factors and mass fractions were calculated with the participating programs (k0-IPEN, k0-INRIM, k0-DALAT, k0-IAEA and KayWin) using identical peak areas. In this paper, we report on the observed discrepancies, causes, remedies and future software developments. The test data, as well as intermediate results and observed mass fractions of the certified reference material BCR-320R “channel sediment” are available through the IAEA on request. The variations in concentrations attributed to differences between the programs were initially found to be 5.6 and 7.9%, for certified and uncertified concentrations, respectively. After the certified concentrations had been made available to the participants and they had been allowed to improve their programs, the variations found were 2.7 and 3.4%, respectively. The main identified remaining causes of variation are differences in the procedures used for detector efficiency characterisation and neutron spectrum parameter determination.

Chromium-Coated Zirconium Cladding Neutronics Impact for APR-1400 Reactor Core

The accident-tolerant fuel concept involves replacing the conventional cladding system (zirconium) with a new material or coating that has specific thermomechanical properties. The aim of this study is to evaluate the neutronics performance of a chromium coating concept and design solutions. A Zircaloy–uranium fuel system (Zr–U) is currently used as a standard fuel system in pressurized water reactors around the world. This investigation presents the benefits of utilizing an alternative cladding material such as chromium coating and the effects on the thermal neutron parameters of the way in which the chromium coating is introduced in the reactor fuel. Among these significant benefits is an increase in the reactor fuel’s thermal conductivity, which improves reactor safety. Two types of fuel-cladding systems were investigated: Zircaloy–uranium (Zr–U) and Zircaloy–chromium (Zr–Cr–U) coating fuel systems. Neutronics analysis evaluations were performed for the selected fuel assemblies and a two-dimensional full core based on an APR-1400 reactor design. Neutronics analyses were performed for the application of the new fuel-cladding material systems using the reactor-physics Monte Carlo code Serpent 2.31.

Feasibility Study on the Initial Kartini Reactor Core Using Plate Type Fuel Elements

The plate type fuel element conversion is proposed to solve a supply problem of TRIGA standard rod type fresh fuel in the long term and to extend the lifetime by reducing the dependence of buying imported elements. The plate type fuel is an alternative since the Indonesian industry has been able to produce such fuel elements. The change of core configuration is expected to improve the reactor performance for irradiation facilities and fuel element lifetime. The SRAC2006 is used to perform neutronic calculations while the nuclear fuel lifetime is calculated by SWAT. This study begins with performing a core properties comparison of UZrH1.6 as the current fuel material and U3Si2-Al as the fuel material candidate. The results show that the Kartini reactor core is possible to load U3Si2-Al as the fuel material and makes higher excess reactivity compared to the current fuel material. Furthermore, U3Si2-Al in the plate type element geometry is variedly arranged in the new reactor core configuration to optimize the neutronic core parameters. The new core configuration is composed of 10 standard fuel elements, 4 fuel control elements, and the graphite material baffle that is located between the active core and annular reflector for serves as an additional reflector. The configuration produced sufficient core excess reactivity and adequate shutdown margin. It also produced negative temperature feedback reactivity and power peaking factor that fulfilled the safety requirements. Improvement of new reactor core performance was obtained by more irradiation facilities, higher thermal neutron flux, and longer maximum estimated burn up compared to the current core configuration.

Investigation of gamma ray, electron, and neutron interaction parameters of some topological insulating materials

The gamma ray, neutron, and electron radiation interaction parameters were calculated for some topological insulating (TI) materials such as Sb2Te3, Bi2Se3, Bi2Te2Se and Bi2Te3 at various energies. The gamma ray interaction parameters such as mass attenuation coefficient, half value layer, tenth value layer, effective atomic number, equivalent atomic number, exposure buildup factor, and energy absorption buildup factor were calculated in the energy range of 0.015–15 MeV at penetration depths of 1–40 mfp by using the five-parameter Geometric Progression (G-P) fitting approximation method. The buildup factors were found to be dependent on the energy of the incident photon, penetration depth, and material elemental composition, and they were also found to be higher for Sb2Te3 than for other TI materials throughout the energy region of investigation. The neutron interaction parameters, such as fast neutron removal cross section, were calculated for the selected topological insulator materials using Phy-X/PSD software and found to be higher for Bi2Te3 than for the other three materials. The electron stopping power values were calculated for the selected topological insulator materials in the energy range of 0.01 MeV to 1000 MeV using the ESTAR program. It was found that the stopping power values decreased slowly with increasing electron energy in the lower energy region, remained almost constant in the medium energy region, and increased rapidly in the higher energy region.

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.

Design and Neutronic, Thermal-Hydraulic Analysis of DSCF Assembly for a SMR and Investigation of the Effect on the Thermal Power Uprate

Small modular reactors (SMRs) can be a significant option for developing countries with low energy demand. Due to the lack of sufficient experience in the field of SMRs, extensive research should be done on SMRs to improve the performance of these systems. Using dual surface-cooled fuel (DSCF) is one of the methods that can increase the performance of SMRs. In this study, for the first time the core of a NuScale reactor (as a SMR) is designed based on DSCF without any change in core dimensions by analyzing neutronic, thermal-hydraulic, and natural circulation parameters. In addition, according to the departure from nucleate boiling ratio, the uprate of the thermal power in a reactor using DSCF is investigated. For this purpose, typical solid fuels as well as DSCFs under clean-cold and full-power conditions are primarily modeled for the four different lattices that maintain the same assembly dimensions, mass, and enrichment fuels as the original fuel assembly. The effective multiplication factor, and power peaking factor, as important neutronic parameters, are calculated. Then the departure from nucleate boiling, pressure drop, velocity, and temperature distribution calculations, as important thermal-hydraulic and natural circulation parameters, are accomplished via a computational fluid dynamics code. The best core configuration of DSCF for the NuScale core is determined based on comparing the neutronic, thermal-hydraulic, and natural circulation parameters of various lattices and typical solid fuels. Regarding the final result, a DSCF assembly configuration, called a 12 × 12 assembly, is suggested.

Nanofluid applications in pressurized water reactors: A review

The main purpose of this paper was to review nanofluid applications as the main coolant of pressurized water reactors (PWRs) with both solid and dual-cooled annular fuel. In this regard, first, theoretical and experimental models for calculating thermo-physical properties of nanofluids including density, specific heat, thermal conductivity, viscosity, and heat transfer coefficient were reviewed and the effects of different variables such as particle volume concentration, particle size, shape, and temperature on these parameters were determined. In the following, research related to neutronic, thermal-hydraulic, and safety investigations of nanofluids as the main coolant of PWRs with solid and dual-cooled annular fuel were reviewed. Corresponding methods, codes, and results were discussed and compared. Based on studies, it was concluded that nanofluids' effects on key neutronic parameters (keff, neutron flux and radial and axial power peaking factor) depend on the used nanoparticle, volume concentration, and nanoparticle deposition on the cladding, while nanofluids can increase heat transfer, critical heat flux, and minimum departure from nuclear boiling ratio (MDNBR). Also, it was found that nanofluids with low volume concentration in reactor core with dual-cooled annular fuel can replicate the same advantages. Furthermore, using dual-cooled annular fuel instead of solid fuel increases MDNBR margin and reduces coolant flow which leads to a more compact design of the reactor core and a reduction of capital cost. Finally, incoming challenges and future works were presented.

Integrated analysis to investigate the viability of using Thorium-based fuel instead of traditional fuel in CANDU reactor

This work gives the nuclear operator an integrated insight into the possibility of operating a CANDU reactor with thorium-based fuel types. MCNPX code was used to design 3D model of CANDU bundle. Thorium mixed with 235U, 233U and rgPu were investigated as a fuels in separate fuel cycles and their results were compared with the traditional fuel (UO2). The main neutronic and safety parameters such as infinity multiplication factor (kinf), neutron generation time (Λ), breeding ratio, Doppler reactivity coefficient (DRC), effective delayed neutron fraction (βeff) and moderator temperature coefficient (MTC) have been studied for the investigated fuels. Due to the significant impact of the rgPu and minor actinides (MAs) on the environment, their concentrations have been investigated with burn-up for the proposed fuels to reach the optimum one from the neutronic and economic point of view.

A feasibility study of SMART reactor power performance optimizations-part 1: Steady-state and burn-up analysis

SMART is an integral small pressurized water nuclear reactor design with a rated power output of 100 MWe from 330 MWth, but it needs a higher power output for the United Kingdom energy market. This study applies Monte Carlo code OpenMC to build a full-core model and innovatively adjust the simulation coefficients to approach the reactor operating conditions. The analysis results point out the reasonable optimization’s technical direction. The model’s sensitivity to ENDF and JEFF nuclear data libraries and spatial division is tested and verified. Then it performs a series of simulations to obtain the core’s neutronic parameters, such as neutron energy and spatial distributions, effective neutron multiplication factor keff and its variation versus depletion. The analysis found that the initially designed core’s keff is 1.22906, and the temperature reactivity defect is 11612 pcm. In 1129 full-power operating days, the keff will decrease to 0.99126, and the reactor depletes 8.524 × 1026235U atoms. However, the outermost fuel assemblies’ 235U depletion rate is lower than 45% in this extended refuelling cycle, and their ending enrichment is higher than 2.4%. That means the fuel economy of the original design’s two-batch refuelling scheme core layout is insufficient. Improving the thermal neutron fluence in these assemblies may optimize the SMART power performance effectively.

Study of the Allegro Core Performance

The presented paper is related to introduction of the design and neutronic characterization of the start-up core developed for Gas cooled Fast Reactor (GFR) demonstrator. Slovak University of Technology in Bratislava joined the project ALLEGRO in last decade within the consortium of middle-European institutions. In the development plan of the GEN IV GFR the ALLEGRO demonstrator is one of the most necessary steps. The ALLEGRO reactor is small helium cooled 75 MWth thermal power unit. Its main objective is to demonstrate the key GFR technologies and to perform tests of innovative materials. The reactor core is based on the standard and MOX pin type fuel in the first phase of the project. The active core of a large GFR 2400 makes use of ceramic materials, but in the first ALLEGRO core MOX fuel will be used. It will be mainly to demonstrate the viability of the technology and to acquire necessary experimental data for further research. In the presented works are identified the main discrepancies between ALLEGRO and GFR 2400 designs, the sensitivity analysis was performed for both reactors. Neutronic characterization is aimed to determination of the standard neutronic parameters using conventional computational systems. The results of sensitivity and uncertainty calculations are presented in conjunction with similarity analysis.

Validation of 3D discrete ordinate transport code with the practical engineering shielding calculation benchmark problems

This paper introduces extended validation works on JSNT-S code based on two practical engineering shielding calculation problems, the PWR Pressure Vessel Fluence Calculation Benchmark (PWR benchmark) and H. B. Robinson Unit 2 Pressure Vessel Benchmark (HBR-2 benchmark). The benchmarks put forward new challenges to the JSNT-S code, including the complexity of neutron source calculation, strong attenuation of the flux in the shielding material and modeling of complex shield structure geometry. SourceTrans code is developed to convert core pin-by-pin power distribution into neutron source, visual modeling code JLAMT is used to build 3Dfinemodels. Neutronic parameters provided in the benchmark report are validated, including the 47 group neutron flux spectrum, E > 1 MeV flux, E > 0.1 MeV flux, rates of displacement per atom (DPA), activation reaction rate and so on. The results show that JSNT-S code has good accuracy in simulating the practical engineering shielding calculation, and extend the reference solutions for benchmark problems.

Advanced Multi-Modeling of PWR Dynamics and Deep Learning based Computational Tool in SIMULINK and LabVIEW

The reactivity monitoring, prediction, and investigation is the most important parameter to ensure the safety and reliable operation of a nuclear power plant. This parameter is gained further importance in Pressurized Water Reactor (PWR) due to more sophisticated reactivity insertion mechanisms and innovative reactor core fuel loading scheme. Based on deterministic internal and external dynamics and neutronics analysis of Advanced PWR, all the reactivity feedback effects such as Doppler effect, moderator effect, control rod effect, liquid boron effect and reactor poisons effect are investigated, modeled and stochastically optimized using deep artificial intelligence. Advance Pressurized Water Reactor (APWR) of 600 MWe rating (AP-600) is used as a reference reactor model and based on the dynamics of AP-600, an Advanced Pressurized Water Reactor Dynamics and Intelligent Stochastic Optimization based Deterministic Neutronics Simulation (APD-ISO-DNS) Code is developed in the hybrid SIMULINK andLabVIEW environments. AP-600 reactor model is fine-tuned and adjusted for 300 MWe PWR (P-300) and 1070 MWe Advanced Chinese PWR (ACP-1000) using neutronics parameters and operational dynamic data of operating PWR nuclear power plants in Pakistan. Various load reduction transient experiments are conducted and dynamic transient simulations of P-300, AP-600 and ACP-1000 are evaluated in SIMULINK and in LabVIEW environments and found as per design basis.

SERPENT CODE POINT KINETIC PARAMETERS ANALYSIS: A NEXUS FOR NIGERIA RESEARCH REACTOR – 1 CORE CONVERSION USING ENRICHED URANIUM DIOXIDE FUEL

The International Atomic Energy Agency (IAEA) requires that all test and research reactors operating on Higher Enriched Uranium (HEU) should be converted to Low Enriched Uranium (LEU) for safety and security purposes. Nigeria having a Miniature Neutron Source Reactor (MNSR) has been long interested in fuel technological research not just to develop the area but also to meet with resolution on the nuclear treaty set out by the global nuclear regulatory body. In this study, reactor kinetic parameters such as effective delayed neutron fractions, prompt neutron lifetime as well as mean neutron generation time were analysed for Nigerian Research Reactor-1 (NIRR-1). Serpent Monte Carlo code 1.17 is used in the analysis. For delayed neutron parameters determination, we used fission probability iteration under one averaged generation time and neutron population rate.The calculated values for delayed neutron were recorded as analogue prompt and implicit prompt neutron lifetime, reproduction time and emission time are in the order of 3×10-7 (s), in agreement with the calculated data from the nuclear data libraries and some literature.The result for delay neutron fraction and other time-based parameters support the fuel core conversion for NIRR-1.The computational and pictorial results obtained from Serpent code simulation described well the transient behavior of the delayed neutron in this reactor.The analytical results also spelled out the relevance and compatibility of low enriched uranium dioxide fuel over higher enriched type.The result of this study conforms with other results obtained from similar reactors but with different Monte Carlo codes and with higher enriched uranium

Research of the spatial effect of PWR fuel pin with disperse particles

Fully ceramic microencapsulated (FCM) fuel is one of the most popular accident tolerant fuels (ATF) and it has a particular spatial structure, made of SiC matrix embedded with TRISO fuel particles. Fuel pin embedded with disperse burnable poison particles, has similar spatial structure. They are both double-heterogeneous and they have much unique neutronic characteristic. In this paper, efforts are made to analyze how the spatial structure of the disperse fuel particles and burnable poison particles affects their neutronic behaviors in detail, from pin-cell level, which is also referred as spatial effect. Many important neutronic parameters are discussed, including infinite multiplication factor (Kinf), four factors (fast fission factor ∊, resonance escape probability factor p, thermal utilization factor f and thermal reproduction factor η), neutron flux level, nuclides number density. They are compared with an equivalent homogeneous structure and the corresponding differences would reveal the unique spatial effect of this kind of double-heterogeneous structure. Besides, sub-regions analyses are also carried on. Finally, it is proved that the double-heterogeneous spatial structure would bring much influence on neutronic behaviors, and the spatial effect of disperse burnable poison particles would be even more serious.

Sensitivity Coefficient Evaluation of an Accelerator-Driven System Using ROM-Lasso Method

We propose the use of reduced-order modeling to improve the sensitivity coefficient evaluation method based on Lasso-type penalized linear regression. In this method, cross sections of interest are uniformly randomly sampled, and corresponding perturbed core analyses are performed. The sensitivity coefficients of the higher-dimensional model are expanded by the active subspace (AS) attained by the lower-dimensional model, and the expansion coefficients are estimated by the Lasso regression. In addition, AS bases can be flexibly chosen according to neutronics parameters of interest. We conducted a verification calculation for an accelerator-driven system and clarified that the proposed method successfully reduces the calculation cost by a couple of orders of magnitude compared with the direct method. The proposed method can be used to practically evaluate the sensitivity coefficients of various parameters.

Thick target neutron spectral yield and dose measurements with [formula omitted]C (p, n) system at intermediate proton energies between 8–20 MeV

The study of neutron parameters from a thick natural graphite target is of immense importance for obtaining the background neutron energy and yield corrections in case of proton irradiation of low Z targets, where graphite is used predominantly as the supporting target material. There are few measurements with thick graphite target reported in literature at proton energies up to 12 MeV and some at higher energies beyond 20 MeV. The present study focuses on the thick target neutron yield, dose and spectra measurements from thick natC(p,n) system at intermediate proton energies between 8–20 MeV. The experimental measurements were performed at 0°and 90°with respect to the incident beam direction. The generated data will complement the discontinuity in the neutron yield and spectral measurements, in the proton energy range of 12–20 MeV. These dosimetric estimates in the proton energy range of 8–20 MeV will be significant towards the personal radiation protection of occupational workers and neutron shielding evaluations.

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.

Neutronics analyses of COOL blanket for CFETR

In this study, an advanced liquid blanket based on a liquid PbLi technology is proposed, and it is called a supercritical CO2 (S-CO2) cooled Lithium-Lead (COOL) blanket. As a candidate for the Chinese Fusion Engineering Testing Reactor (CFETR), it aims to achieve high tritium breeding and shielding performance. It uses S-CO2 as the coolant for the structural components and PbLi as the tritium breeder and neutron multiplier. To evaluate the neutronics performances of this blanket, a detailed 22.5° three-dimensional neutronics model of CFETR was built using the cosVMPT conversion tool. To enhance the shielding capability of the blanket, a shielding block was designed and added at the back of the manifold to moderate neutrons. Evaluation results of the neutron damage of toroidal field coils (TFC) components confirmed that the blanket meets shielding requirements. In addition, the main neutronics parameters including neutron wall loading (NWL), tritium breeding ratio (TBR), nuclear heat, and irradiation damage, were calculated using MCNP5. The results indicated that TBR was 1.205, and the total nuclear heat was 1191.64 MW under the fusion power of 1.5 GW. The peak NWL was 1.97 MW/m2, and the maximum DPA and gas production were 14.80 DPA/FPY, 132.65 He appm/FPY, and 631.30 H appm/FPY.

Neutronic studies for the second Egyptian research reactor ET-RR-2

Abstract The start up physics studies SPS of the Second Egyptian Research Reactor is presented in this work. The SPS consists of the approach to criticality, the power and flux distribution of the first critical core, criticality, power and flux distribution of the working core. The working core is the beryllium-reflected core with cobalt loaded for irradiation. Some of the neutronic parameters are determined, these parameters are the power peaking factor, power distribution, the excess reactivity for both the water and beryllium reflector cores, the core excess reactivity with the cobalt loaded, the reactivity worth of the cobalt irradiation box. Flux distribution for one fast, three eipthermal and one thermal energy group are also determined. All the above neutronic parameters are calculated by using the lattice cell code WIMSD4 and the two dimensional diffusion code DIXY2.