Thermal Neutron Flux Distribution Research Articles

Neutronic Analysis of Accident-Tolerant Cladding Materials in 3D Full Core BEAVRS PWR Benchmark Using OpenMC Code

After the Fukushima Daiichi nuclear accident in 2011, the performance of nuclear fuel during accidents became a matter of great concern. To address this, a new type of fuel technology called accident-tolerant fuel (ATF) has been developed with the goal of enhancing the ability of light water reactors (LWRs) to withstand severe accident conditions. Iron-based alloys have been suggested as potential candidates for fuel cladding due to their favourable thermomechanical properties, lower reactivity with steam, and lower hydrogen generation. This study evaluates the neutronic performance of C26M (a 2nd generation nuclear grade FeCrAl alloy), APMT™, 310SS, and 304SS cladding materials by comparing them with Zircaloy-4 cladding in a 3D PWR core at the beginning of the cycle (BOC) using OpenMC code. The results revealed that the neutronic penalty varied for different alternative cladding materials where C26M exhibited the lowest neutronic penalty value of -12551 pcm, while 310SS demonstrated the highest with a value of -17855 pcm. Additionally, important parameters in the reactor core such as neutron spectrum, reactivity coefficients, boron worth, control rod bank worth, power distribution, and radial thermal neutron flux distribution are evaluated and discussed. The analysis results showed that C26M provided a significantly higher level of neutronic performance compared to APMT™, 304SS, and 310SS. Although this study primarily focused on the neutronic performance of PWRs at BOC, future research should encompass fuel depletion analysis to delve deeper into the potential of alternative cladding materials.

Neutron flux evaluation algorithm with a combination of Monte Carlo and removal-diffusion calculation methods for boron neutron capture therapy.

In Japan, the clinical treatment of boron neutron capture therapy (BNCT) has been applied to unresectable, locally advanced, and recurrent head and neck carcinomas using an accelerator-based neutron source since June of 2020. Considering the increase in the number of patients receiving BNCT, efficiency of the treatment planning procedure is becoming increasingly important. Therefore, novel and rapid dose calculation algorithms must be developed. We developed a novel algorithm for calculating neutron flux, which comprises of a combination of a Monte Carlo (MC) method and a method based on the removal-diffusion (RD) theory (RD calculation method) for the purpose of dose calculation of BNCT. We present the details of our novel algorithm and the verification results of the calculation accuracy based on the MC calculation result. In this study, the "MC-RD" calculation method was developed, wherein the RD calculation method was used to calculate the thermalization process of neutrons and the MC method was used to calculate the moderation process. The RD parameters were determined by MC calculations in advance. The MC-RD calculation accuracy was verified by comparing the results of the MC-RD and MC calculations with respect to the neutron flux distributions in each of the cubic and head phantoms filled with water. Comparing the MC-RD calculation results with those of MC calculations, it was found that the MC-RD calculation accurately reproduced the thermal neutron flux distribution inside the phantom, with the exception of the region near the surface of the phantom. The MC-RD calculation method is useful for the evaluation of the neutron flux distribution for the purpose of BNCT dose calculation, except for the region near the surface.

Discussing the possibility of using thorium-based fuels as an alternative fuel to uranium dioxide fuel for APR-1400 reactor

This work investigates the possibility of using thorium-based fuels as an alternative fuel for advanced power reactor APR-1400. MCNPX code version 2.7 with cross-section library ENDF.VII has been used to design an APR-1400 fuel assembly. This code has been used to study the neutronic performance of the proposed thorium-based Fuel types (0.944Th, U)O2, (0.955Th, 233U)O2, (0.934Th, rgPu)O2. The fuel burn-up parameters such as infinity multiplication factor (kinf), initial heavy metals concentrations, Minor actinides concentration and fission products concentration have been analyzed during 1500 effective full power days (FPDSs) for thorium-based Fuel types and compared with the common fuel. The analysis of the horizontal thermal power and neutron flux distribution provides valuable insights into the behavior and performance of the suggested Fuel types in the APR-1400 assembly. The analysis of the neutronic results ensures the viability of using the proposed thorium-based fuel types as an alternative fuel to UO2 because they achieved acceptable safety parameter values and provided a good power distribution through the fuel assembly compared to UO2.

A novel approach for managing the excess reactivity at the beginning of the fuel cycle of VVER-1200

This work discusses various fuel pellet designs loaded with different types and concentrations of burnable absorbers to suppress the excess reactivity at the early stage of the fuel life in the VVER-1200 fuel assembly. MCNPX, V. 2.7 with the CINDER 90 fuel depletion code was utilized to analyze the neutronic characteristics of the proposed cases. Concentric shells (CS) of integral burnable absorbers are investigated in the fuel tube. Gadolinium and Erbium at different concentrations are investigated in the CS at different operation cycles to manage the excess reactivity. Gadolinium and boron as traditional burnable absorbers (BAs) are investigated in the standard fuel assembly to compare their results with the suggested cases. The infinity multiplication factor (kinf), the concentrations of the fuel composition, reactor-grade plutonium (rgPu), minor actinides (MAs) and some important fission products poisoned were studied with burnup. The radial power and thermal neutron flux distribution through one-six of the VVER-1200 assembly have been analyzed to distinguish the optimum BA and design for flattening the power profile. The neutronic analysis indicates that using Er2O3 in a concentric shell of a fuel pellet is an interesting method for reactivity management.

Development of optimization method for uniform dose distribution on superficial tumor in an accelerator-based boron neutron capture therapy system.

To treat superficial tumors using accelerator-based boron neutron capture therapy (ABBNCT), a technique was investigated, based on which, a single-neutron modulator was placed inside a collimator and was irradiated with thermal neutrons. In large tumors, the dose was reduced at their edges. The objective was to generate a uniform and therapeutic intensity dose distribution. In this study, we developed a method for optimizing the shape of the intensity modulator and irradiation time ratio to generate a uniform dose distribution to treat superficial tumors of various shapes. A computational tool was developed, which performed Monte Carlo simulations using 424 different source combinations. We determined the shape of the intensity modulator with the highest minimum tumor dose. The homogeneity index (HI), which evaluates uniformity, was also derived. To evaluate the efficacy of this method, the dose distribution of a tumor with a diameter of 100mm and thickness of 10mm was evaluated. Furthermore, irradiation experiments were conducted using an ABBNCT system. The thermal neutron flux distribution outcomes that have considerable impacts on the tumor's dose confirmed a good agreement between experiments and calculations. Moreover, the minimum tumor dose and HI improved by 20 and 36%, respectively, compared with the irradiation case wherein a single-neutron modulator was used. The proposed method improves the minimum tumor volume and uniformity. The results demonstrate the method's efficacy in ABBNCT for the treatment of superficial tumors.

Intensity-modulated irradiation for superficial tumors by overlapping irradiation fields using intensity modulators in accelerator-based BNCT.

The distribution of the thermal neutron flux has a significant impact on the treatment efficacy. We developed an irradiation method of overlapping irradiation fields using intensity modulators for the treatment of superficial tumors with the aim of expanding the indications for accelerator-based boron neutron capture therapy (BNCT). The shape of the intensity modulator was determined and Monte Carlo simulations were carried out to determine the uniformity of the resulting thermal neutron flux distribution. The intensity modulators were then fabricated and irradiation tests were conducted, which resulted in the formation of a uniform thermal neutron flux distribution. Finally, an evaluation of the tumor dose distribution showed that when two irradiation fields overlapped, the minimum tumor dose was 27.4Gy-eq, which was higher than the tumor control dose of 20Gy-eq. Furthermore, it was found that the uniformity of the treatment was improved 47% as compared to the treatment that uses a single irradiation field. This clearly demonstrates the effectiveness of this technique and the possibility of expanding the indications to superficially located tumors.

Searching for the most optimum burnable absorbers (BAs) for AP-1000 from the neutronic, thermal-hydraulic, and solid mechanics points of view

Because of the importance of absorber materials in regulating the reactivity of AP-1000, several materials with varied designs have been investigated to find the most optimum absorber materials from the neutronic, thermal-hydraulic, and solid mechanics perspectives. In this work, ZrB2, Er2O3, (SiO2-B2O3) and (Al2O3-B4C) were investigated as burnable absorbers (BAs) and H3BO3, Ag–In–Cd, and SS-304 as non-burnable absorbers. For the investigated BAs, the main neutronic parameters such as the fuel burn-up, axial thermal neutron flux distribution and the pin by pin distribution have been simulated. For the non-burnable absorbers, the control rod worth (CRW) and the boron concentration coefficient have been calculated. In the thermal-hydraulic analysis, the temperature distribution of the fuel, clad, coolant and the minimum departure from nucleate boiling ratio (MDNBR) have been studied. The BAs’ effects on solid mechanics analysis have emerged on both the von-Mises stress and the maximum fuel outer surface displacement.

Simulation-based studies on graphite absorption properties for ASTRA critical experiments

This work is focused on the improvement of the graphite impurity content characterization for the ASTRA critical facility. The equivalent boron content for the graphite reflectors of the ASTRA facility reported in the benchmark specifications was estimated using the results of the exponential experiments performed by graphite manufacturer. Because raw measured data are not available for these experiments, we tried to reassess the equivalent boron content for the ASTRA graphite reflectors using the results of analogous exponential experiments performed earlier on the empty RBMK critical facility at the National Research Center «Kurchatov Institute» and further simulation of the experiments utilizing the Monte Carlo code Serpent with ENDF/B-VII.0, ENDF/B-VII.1, and ENDF/B-VIII.0 cross-section libraries. The same grade graphite was used for manufacturing the graphite blocks for the RBMK critical facility and the ASTRA critical facility. The Monte Carlo simulations confirmed the accuracy of the exponential measurements performed for RBMK graphite characterization. It was also shown that the simulation must reproduce the graphite absorption cross-section estimated using the exponential experiments for the agreement between the measured and calculated distribution of thermal neutron flux. Therefore, the equivalent boron content should be assessed individually for each cross-section library used in the simulation. In case of the ASTRA facility the carbon capture cross section of 3.36mb at 0.0253 eV (ENDF/B-VII.0) was used to deduce the equivalent boron content of 1 ppmwt from the measured results. Thermal capture cross-section of carbon was significantly increased in the ENDF/B-VII.1 and other recent cross-section libraries. Therefore, when calculating the ASTRA benchmark with these cross-section libraries the equivalent boron content reported in the benchmark specification should be reassessed.

Searching for managing the reactivity and increasing the fuel cycle life in the PWR by an untraditional method

This study aims to find an economic method to decrease the excess reactivity and increase the fuel cycle length. The neutronic feasibility to replace the 238U with 232Th and the traditional assembly with Blanket-Seed assembly was analyzed. MCNPX code version 2.7 was used to design a three-dimension model of PWR to simulate the neutronic characteristics of the investigated fuel types and designs. The radial thermal neutron flux and the radial thermal power distribution through the assembly have been studied. The radial thermal neutron flux and thermal power distribution provide an insight view about the proposed fuel types and designs. The effect of the proposed fuel types and design on the infinity multiplication factor (Kinf), the breeding fissile concentration, the reactor-grade plutonium (rgPu), the minor actinides (MAs), conversion ratio (CR) and the concentration of the most radioactive fission products have been analyzed and compared with that of the standard fuel. From the neutronic point of view, the results showed the economic feasibility of using 232Th in regulating the initial excess reactivity.

Determination of thermal neutron flux distribution at the rotary rack served for elemental concentration analysis using the k0-INAA method

The accuracy of elements concentration determination using the k0-standardization method directly depends on irradiation and measurement parameters including Non-1/E epithermal neutron flux distribution shape α (ϕ epi ≈1/E1+α ) , thermal-to-epithermal neutron flux ratio f, efficiency ε, peak area… In the case of the irradiation position at the rotary rack of the Dalat Nuclear Research Reactor (DNRR), the difference of thermal neutron flux between the bottom (3.54x1012 n.cm-2.s-1) and the top (1.93x1012 n.cm-2.s-1) of the 15 cm aluminum container is up to 45%. Therefore, it is necessary to accurately determine above-mentioned parameters in the sample irradiation position. The present paper deals with the determination of the distribution of thermal neutron flux along the sample irradiation container by using 0.1% Au–Al wire activation technique. The thermal neutron flux was then used to calculate the concentration of elements in the Standard Reference Material 2711a and SMELS type III using k0-INAA method at different positions in the container. The obtained results with the neutron flux correction were found to be in good agreement with the certified values. In conclusion, the proposed technique can be applied for activation analyses without sandwiching flux monitors between samples during irradiations.

Design of Coded Neutron Source Based on a RFQ Accelerator by MCNP Simulation

In this paper, the influence of non-ideality factors of thermal neutron source for coded neutron source imaging has been studied by numerical calculation,in order to establish the design standard of coded neutron source. Then, the pinhole imaging detector (pinhole) of MCNP has been used to simulate the imaging process of the coded neutron source imaging, and the parameters such as the uniformity of the spatial distribution of the thermal neutron flux at the imaging plane and the uniformity of the angular emission of the coded neutron source are studied. The design results show that the spatial distribution of thermal neutron flux is less than 13% and the angular distribution is less than 10% with the optimized structure of moderator including a 4 cm thick polyethylene disk and a wedge of neutron absorbing layer closed to the junction surface of collimator and water tube.

Design of Irradiation Facilities at Grid E-1 of Plate Type Research Reactor Bandung

Design of Irradiation Facilities at Grid E-1 of Plate Type Research Reactor Bandung. Plate Type Research Reactor Bandung (PTRRB) core design is one of the result of PTRRB research programs. In the previous study the irradiation facilities at grid E-1 has not been designed and also distribution of thermal, epithermal and fast neutron flux at grid E-1 has not been studied. Since that data is very important especially in radioisotope production and neutron beam tube analysis, therefore in this study irradiation facilities at grid E-1will be designed. Previous PTRRB core design is a base for designing irradiation facilities at grid E-1. Considering geometrical of grid E-1 and aluminum tube dimension there are three possibilities aluminum tube configuration. The configurations are configuration 1, 2 and 3. Each configuration was modelled as arrangement of four aluminum tubes and each tube filled by four aluminum irradiation capsules. That configuration was starting point to made MCNP PTRRB reactor core model so there are three MCNP PTRRB reactor core model. MCNP PTRRB reactor core model is needed because MCNP software are computer program for calculating excess reactivity and neutron flux distribution at grid E-1. Result excess reactivity calculation of three configuration indicate that after installing irradiation tube excess reactivity is lower than of limit excess reactivity value 10.9 % of neutronic safety criteria of PTRRB design. Based on neutronic safety criteria, the three configuration is accepted for irradiation facilities PTRRB. Neutron flux calculation result of three configuration reveals that the highest neutron flux is located at capsule no II and III. Profile of thermal neutron flux, epithermal neutron flux and fast neutron flux of three configurations are similar. Neutron flux of thermal, epithermal and fast neutron of three configuration are slightly different. The calculation result reveal that highest thermal neutron flux at grid E-1 is 2.70 × 1013(n/cm 2.sec) at configuration 2. Based on neutronic safety criteria and thermal neutron flux, configuration 2 is appropriate for irradiation facilities of PTRRB.

Characteristics of Thermal Neutron Flux Distribution in a Phantom Irradiated by Epithermal Neutron Beam from Double Layer Beam Shaping Assembly (DBSA)

A simulation study on the Double-layer Beam Shaping Assembly (DBSA) system has been carried out. This study used fast neutron beam resulting from reactions of 30 MeV protons with beryllium target. The MCNPX code was utilized to design the DBSA and the phantom as well as to calculate neutron flux on the phantom. The distribution of epithermal neutron flux and gamma in the DBSA and phantom were computed using the PHITS code. The spectrum of radiation beams generated by the DBSA shows the characteristics that the typical epithermal neutron flux of 1.0 x109 n/(cm2.s), the ratio of epithermal to the thermal and fast neutron flux of 344 and 85, respectively and the ratio of gamma dose to the epithermal neutron flux of 1.82 x 10-13 Gy.cm2. The test of epithermal neutron beams irradiation on the water phantom shows that epithermal neutrons are thermalized and penetrate the phantom up to 12 cm in depth. The maximum value of neutron flux is 1.1 x 109 n/(cm2.s) at a depth of 2 cm in phantom.

Measurement of Thermal and Epithermal Neutron Flux Distribution in the Torus Hall of LHD using Activation Method in the First Deuterium Experiment Campaign

Measurement of Thermal and Epithermal Neutron Flux Distribution in the Torus Hall of LHD using Activation Method in the First Deuterium Experiment Campaign

Modeling of the Dalat Nuclear Research Reactor (DNRR) with the Serpent 2 Monte Carlo code

This paper presents a model development of the Dalat Nuclear Research Reactor (DNRR) loaded with low enriched uranium (LEU) fuel using the Serpent 2 Monte Carlo code. The purpose is to prepare the DNRR Serpent 2 model for performing fuel burnup calculations of the DNRR as well as for generating multi-group neutron cross sections to be further used in the kinetics calculations of the DNRR with a 3D reactor kinetics code. The DNRR Serpent 2 model was verified through comparing with the MCNP6 criticality calculations under different reactor conditions. The parameters to be compared include the effective neutron multiplication factor, radial and axial powerdistributions, and thermal neutron flux distributions. The comparative results generally show a good agreement between Serpent 2 and MCNP6 and thus indicate that the DNRR Serpent 2 model can be used for further calculations of the DNRR.

Thermal neutron profile inside J-PARC main ring tunnel

Abstract To estimate the beam losses and activation in the J-PARC main ring (MR) tunnel, thermal neutron distribution was measured using gold foil activation detectors with the Cd difference method. Hundred pairs of bare and Cd-covered gold foils were prepared to cover the entire area of the MR tunnel. The relative and absolute activities of the gold foils were measured using an imaging plate and an HP-Ge detector for some reference foils, respectively. A total of 97% of the thermal neutrons were detected in the injection section, where a beam collimation system was installed to localize the beam loss in the MR tunnel. The thermal neutron fluxes were compared with the calculation results from a Monte Carlo simulation code, PHITS. The calculation results were generally in good agreement with the experimental thermal neutron flux distribution. Furthermore, we investigated a simple empirical formula, ϕ th = C Q ∕ S , to estimate the thermal neutron flux in the accelerator room, where Q is the neutron source intensity and S is the total surface area of the room. A coefficient C of 1.61 ± 0.15 was deduced in the MR tunnel, which is within the error range of the previous value measured in the CERN CHARM facility using a 24 GeV/c proton beam despite the different projectile energies and room shapes. The γ dose rates at the same positions as the gold foils were measured as well. A strong correlation was observed between the thermal neutron fluxes and γ dose rates with respect to the surrounding concrete types. This result suggests that the dose rate on the concrete wall can be used to estimate the distribution of the thermal neutron fluxes and the amount of beam loss in the MR.

Design of an irradiation rig using screen method for silicon transmutation doping at the Dalat research reactor

The neutron transmutation doping of silicon (NTD-Si) at research reactors has beensuccessfully implemented in many countries to produce high-quality semiconductors. In the late 1980s, NTD-Si has been tested at the Dalat Nuclear Research Reactor (DNRR) but the results have been limited. Therefore, the design and testing of an irradiation rig for NTD-Si at the DNRR are necessary to have a better understanding in order to apply the NTD-Si in a new research reactor of the Research Centre for Nuclear Science and Technology (RCNEST), which has planned to be built in Viet Nam. This paper presents the design and testing of a new irradiation rig using screen method for testing NTD-Si at the DNRR. The important parameters in the rig such as neutron spectrum and thermal neutron flux distribution were determined by both calculation using MCNP5 computer code and experiment. The aluminum ingots, which have similar neutronic characteristics with silicon ingots, were irradiated in the rig to verify the appropriate design. The uniformity of thermal neutron flux in the rig is less than 5% in axial and 2% in radial directions, respectively. However, the thermal/fast flux ratio of the irradiation rig is 4.38/1 would affect target resistivity of testing Silicon ingots after irradiation.

Fuel management code system against experimental measurements of the CMRR reactor

The China Mianyang Research Reactor (CMRR), with a thermal power of 20 MW, is an important high performance research reactor operated in China. In this paper, we introduced the fuel management (FM) computation toolbox and confirmatory experiments performed at this reactor. The toolbox system utilizes two roadmaps, the deterministic strategy as the standard method and the Monte Carlo (MC) strategy as the reference way. Interface glue-code packages for both roadmaps are developed to link the existing kernel codes to perform the input-file generation, data manipulation, and output data processing for fuel cycle analysis. In particular, a new MC based coupling strategy is proposed called the “burnup related nuclides list” method which can perform the FM engineering problem with high accuracy and speed. For the aspect of confirmatory experiments, we emphasize on the methods and instruments, including the control rods (CRs) critical position, the CR worth curve, and the thermal neutron flux distribution. We have developed a reactivity meter to perform the CRs worth measurement using the inverse kinetics method, and employed the Dy-Al foils to perform the thermal neutron flux distribution measurements. All three sets of experiments are performed after each refueling. With these data accumulated over years, our FM toolbox system is profoundly validated for various combinations of fuel burnup and core arrangement. It can be concluded that our FM toolbox and strategies, both the deterministic and the MC way, are satisfactory for the FM mission of our reactor.

Accuracy of Batan-3DIFF and MCNP6 codes for thermal neutron flux distribution at the irradiation position of RSG-GAS reactor

This paper presents the accuracy of the neutron diffusion and Monte Carlo methods in determining the axial thermal neutron flux in the irradiation facility. The calculations are compared with the activation method. The Au-foils are inserted close to the targets, the Low Enrichment Uranium (LEU) electroplating and TeO2. A new cell model for the thin layer of the LEU target is proposed. The MCNP calculation results are in very good agreement with the experimental core parameters, such as excess reactivity and total control rod worth. However, the neutron diffusion method code, Batan-3DIFF, has a higher relative difference of 10.14%. The proposed multi-zone homogenised model can improve the criticality calculation by 124% compare to the homogenised cell. For the axial thermal neutron distribution, the MCNP gives very satisfying results within the standard deviation of experimental results. However, the Batan-3DIFF code has an average relative difference of 7%.

Accuracy of Batan-3DIFF and MCNP6 codes for thermal neutron flux distribution at the irradiation position of RSG-GAS reactor

Accuracy of Batan-3DIFF and MCNP6 codes for thermal neutron flux distribution at the irradiation position of RSG-GAS reactor