High-resolution neutronics model for ^(238)Pu production in high-flux reactors

We proposed and compared three methods(filter burnup,single energy burnup,and burnup extremum analysis)to build a high-resolution neutronics model for 238Pu production in high-flux reactors.The filter burnup and single energy burnup methods have no theoretical approximation and can achieve a spectrum resolution of up to~1 eV,thereby constructing the importance curve and yield curve of the full energy range.The burnup extreme analysis method combines the importance and yield curves to consider the influence of irradiation time on production efficiency,thereby constructing extreme curves.The three curves,which quantify the transmutation rate of the nuclei in each energy region,are of physical significance because they have similar distributions.A high-resolution neutronics model for ^(238)Pu production was established based on these three curves,and its universality and feasibility were proven.The neutronics model can guide the neutron spectrum optimization and improve the yield of ^(238)Pu by up to 18.81%.The neutronics model revealed the law of nuclei transmutation in all energy regions with high spectrum resolution,thus providing theoretical support for high-flux reactor design and irradiation production of ^(238)Pu.

Neutronics analysis of a subcritical blanket system driven by a gas dynamic trap-based fusion neutron source for ^(99)Mo production

Gamma-emitting radionuclide ^(99m)Tc is globally used for the diagnosis of various pathological conditions owing to its ideal single-photon emission computed tomography (SPECT) characteristics.However,the short half-life of ^(99m)Tc (T_(1/2)=6 h)makes it difficult to store or transport.Thus,the production of ^(99m)Tc is tied to its parent radionuclide ^(99)Mo (T_(1/2)=66 h).The major production paths are based on accelerators and research reactors.The reactor process presents the potential for nuclear proliferation owing to its use of highly enriched uranium (HEU).Accelerator-based methods tend to use deuterium–tritium(D–T) neutron sources but are hindered by the high cost of tritium and its challenging operation.In this study,a new ^(99)Mo production design was developed based on a deuterium–deuterium (D–D) gas dynamic trap fusion neutron source (GDT-FNS) and a subcritical blanket system (SBS) assembly with a low-enriched uranium (LEU) solution.GDT-FNS can provide a relatively high-neutron intensity,which is one of the advantages of ^(99)Mo production.We provide a Monte Carlo-based neutronics analysis covering the calculation of the subcritical multiplication factor (k_(s)) of the SBS,optimization design for the reflector,shielding layer,and ^(99)Mo production capacity.Other calculations,including the neutron flux and nuclear heating distributions,are also provided for an overall evaluation of the production system.The results demonstrated that the SBS meets the nuclear critical safety design requirement (k_(s)<0.97) and maintained a high ^(99)Mo production capacity.The proposed system can generate approximately 157 Ci ^(99)Mo for a stable 24 h operation with a neutron intensity of 1×10^(14) n/s,which can meet 50%of China’s demand in 2025.

Neutronic analysis of silicon carbide cladding accident-tolerant fuel assemblies in pressurized water reactors

In resonance with the Fukushima Daiichi Nuclear Power Plant accident lesson, a novel fuel design to enhance safety regarding severe accident scenarios has become increasingly appreciated in the nuclear power industry. This research focuses on analysis of the neutronic properties of a silicon carbide(SiC) cladding fuel assembly, which provides a greater safety margin as a type of accident-tolerant fuel for pressurized water reactors. The general physical performance of SiC cladding is explored to ascertain its neutronic performance. The neutron spectrum, accumulation of ^(239)Pu, physical characteristics,temperature reactivity coefficient, and power distribution are analyzed. Furthermore, the influences of a burnable poison rod and enrichment are explored. SiC cladding assemblies show a softer neutron spectrum and flatter power distribution than conventional Zr alloy cladding fuel assemblies. Lower enrichment fuel is required when SiC cladding is adopted. However, the positive reactivity coefficient associated with the SiC material remains to be offset. The results reveal that SiC cladding assemblies show broad agreement with the neutronic performance of conventional Zr alloy cladding fuel. In the meantime, its unique physical characteristics can lead to improved safety and economy.

Neutronic Calculation Analysis for CN HCCB TBM-Set

Using the Monte Carlo transport code MCNP, neutronic calculation analysis for China helium cooled ceramic breeder test blanket module （CN HCCB TBM） and the associated shield block （together called TBM-set） has been carried out based on the latest design of HCCB TBM-set and C-lite model. Key nuclear responses of HCCB TBM-set, such as the neutron flux, tritium production rate, nuclear heating and radiation damage, have been obtained and discussed. These nuclear performance data can be used as the basic input data for other analyses of HCCB TBM-set, such as thermal-hydraulics~ thermal-mechanics and safety analysis.

Neutronics analysis of a stacked structure for a subcritical system with LEU solution driven by a D-T neutron source for~(99)Mo production

The utilization of neutrons markedly affects the medical isotope yield of a subcritical system driven by an external D-T neutron source.The general methods to improve the utilization of neutrons include moderating multiplying,and reflecting neutrons,which ignores the use of neutrons that backscatter to the source direction.In this study,a stacked structure was formed by assembling the multiplier and the low-enriched uranium solution to enable the full use of neutrons that backscatter to the source direction and further improve the utilization of neutrons.A model based on SuperMC was used to evaluate the neutronics and safety behavior of the subcritical system,such as the neutron effective multiplication factor,neutron energy spectrum,medical isotope yield,and heat deposition.Based on the calculation results,when the intensity of the neutron source was 59×10^(13)n/s,the optimized design with a stacked structure could increase the yield of ^(99)Mo to182 Ci/day,which is approximately 16% higher than that obtained with a single-layer structure.The inlet H_(2)O coolant velocity of 1.0 m/s and initial temperature of 20℃ were also found to be sufficient to prevent boiling of the fuel solution.

A neutronic experiment to support the design of an Indian TBM shield module for ITER

A shield module is associated with an Indian Test Blanket Module (TBM) in ITER to limit the radiation doses in port inter-space areas.The shield module is made of stainless steel plates and water channels.It is identified as an important component for radiation protection because of its radiation exposure control functionality.The radiation protection classification leads to more assurance of the component design.In order to validate and verify the design of the shield module,a neutronic laboratory-scale experiment is designed and executed.The experiment is planned by considering the irradiation under a neutron source of 14 MeV and yields of 1010 n s-1.The reference neutron spectrum of the ITER TBM shield module has been achieved through optimization of the neutron source spectrum by a combination of steel and lead materials.The neutron spectrum and flux are measured using a multiple foil activation technique and neutron dose-rate meter LB 6411 (He-3 proton recoil counter with polyethylene),respectively.The neutronic design simulation is assessed using MCNP5 and FENDL 2.1 crosssection data.The paper covers neutronic design,irradiation and the outcome of the experiment in detail.

Neutronic analysis of ITER radial x-ray camera

The radial x-ray camera(RXC) is designed to measure the poloidal profile of plasma x-ray emission with high spatial and temporal resolution. The RXC diagnostic system consists of internal camera module and external camera module that view the core region and outer region through the vertical slots of the diagnostic first wall and diagnostics shield module of the equatorial port plug. To ensure the normal performance of the silicon photodiode array detectors of the cameras in the hard neutron irradiation environment in ITER tokamak, it is necessary to calculate neutron flux, radiation damage and the nuclear heating of the silicon photodiode array detectors and simulate the radiation maps of the range of RXC. This work estimated the nuclear environment of RXC based on Monte Carlo N-particle transport code, plasma scenarios of ITER tokamak and the RXC-integrated ITER CLITE model. The neutron flux of silicon photodiode array detectors and the lifetime of the silicon photodiode detector in the camera were calculated. The neutronic analysis results show that the shielding design has achieved the effect as expected and is able to guarantee the normal work of the detector during the ITER deuterium–deuterium phase without replacement, three detectors of the external camera can be operated during the whole deuterium–tritium phase without replacement.

Updated neutronics analyses of a water cooled ceramic breeder blanket for the CFETR

The water cooled ceramic breeder（WCCB） blanket employing pressurized water as a coolant is one of the breeding blanket candidates for the China Fusion Engineering Test Reactor（CFETR）.Some updating of neutronics analyses was needed, because there were changes in the neutronics performance of the blanket as several significant modifications and improvements have been adopted for the WCCB blanket, including the optimization of radial build-up and customized structure for each blanket module. A 22.5 degree toroidal symmetrical torus sector 3 D neutronics model containing the updated design of the WCCB blanket modules was developed for the neutronics analyses. The tritium breeding capability, nuclear heating power, radiation damage,and decay heat were calculated by the MCNP and FISPACT code. The results show that the packing factor and^6 Li enrichment of the breeder should both be no less than 0.8 to ensure tritium self-sufficiency. The nuclear heating power of the blanket under 200 MW fusion power reaches201.23 MW. The displacement per atom per full power year（FPY） of the plasma-facing component and first wall reach 0.90 and 2.60, respectively. The peak H production rate reaches150.79 appm/FPY and the peak He production reaches 29.09 appm/FPY in blanket module #3.The total decay heat of the blanket modules is 2.64 MW at 1 s after shutdown and the average decay heat density can reach 11.09 kW m^（-3） at that time. The decay heat density of the blanket modules slowly decreases to lower than 10 W m^（-3） in more than ten years.

Neutronic calculations of the China dual-functional lithium–lead test blanket module with the parallel discrete ordinates code Hydra

The China dual-functional lithium–lead test blanket module(DFLL-TBM) is a liquid Li Pb blanket concept developed by the Institute of Nuclear Energy Safety Technology of the Chinese Academy of Sciences for testing in ITER to validate relevant tritium breeding and shielding technologies. In this study, neutronic calculations of DFLL-TBM were carried out using a massively parallel three-dimensional transport code, Hydra, with the Fusion Evaluated Nuclear Data Library/MG. Hydra was developed by the Nuclear Engineering Computational Physics Lab based on the discrete ordinates method and has been devoted to neutronic analysis and shielding evaluation for nuclear facilities. An in-house Monte Carlo code(MCX) was employed to verify the discretized calculation model used by Hydra for the DFLL-TBM calculations. The results showed two key aspects:(1) In most material zones,Hydra solutions are in good agreement with the reference MCX results within 1%, and the maximal relative difference of the neutron flux is merely 3%, demonstrating the correctness of the calculation model;(2) while the current DFLL-TBM design meets the operation shielding requirement of ITER for 4 years, it does not satisfy the tritium self-sufficiency requirement. Compared to the two-step approach, Hydra produces higher accuracies as it does not rely on the homogenization technique during the calculation process. The parallel efficiency tests of Hydra using the DFLL-TBM model also showed that this code maintains a high parallel efficiency on O(100) processors and, as a result, is able to significantly improve computing performance through parallelization. Parameter studies have been carried out by varying the thickness of the beryllium armor layer and the tritium breeding zone to understand the influence of the beryllium layer and breeding zone thickness on tritium breeding performance. This establishes a foundation for further improvement in the tritium production performance of DFLL-TBM.

Implementation of high-fidelity neutronics and thermal–hydraulic coupling calculations in HNET

To perform nuclear reactor simulations in a more realistic manner,the coupling scheme between neutronics and thermal-hydraulics was implemented in the HNET program for both steady-state and transient conditions.For simplicity,efficiency,and robustness,the matrixfree Newton/Krylov(MFNK)method was applied to the steady-state coupling calculation.In addition,the optimal perturbation size was adopted to further improve the convergence behavior of the MFNK.For the transient coupling simulation,the operator splitting method with a staggered time mesh was utilized to balance the computational cost and accuracy.Finally,VERA Problem 6 with power and boron perturbation and the NEACRP transient benchmark were simulated for analysis.The numerical results show that the MFNK method can outperform Picard iteration in terms of both efficiency and robustness for a wide range of problems.Furthermore,the reasonable agreement between the simulation results and the reference results for the NEACRP transient benchmark verifies the capability of predicting the behavior of the nuclear reactor.

Neutronics analysis for MSR cell with different fuel salt channel geometries

The neutronic properties of molten salt reactors(MSRs)differ from those of traditional solid fuel reactors owing to their nuclear fuel particularity.Based on the Monte-Carlo N particle transport code,the effects of the size and shape of the fuel salt channel on the neutron physics of an MSR cell are investigated systematically in this study.The results show that the infinite multiplication factor(k?)first increases and then decreases with the change in the graphite cell size under certain fuel volume fraction(FVF)conditions.For the same FVF and average chord length,when the average chord length is relatively small,the k?values for different fuel salt channel shapes agree well.When the average chord length is relatively large,the k?values for different fuel salt channel shapes differ significantly.In addition,some examples of practical applications of this study are presented,including cell selection for the core and thermal expansion displacement analysis of the cell.

Neutronics Optimization of LiPb-He Dual-Cooled Fuel Breeding Blanket for the Fusion-Driven sub-critical System

The concept of the liquid Li17Pb83 and Helium gas dual-cooled Fuel Breeding Blanket (FBB) for the Fusion-Driven sub-critical System (FDS) is presented and analyzed. Taking self-sustaining tritium (TBR >1.05) and annual output of 100 kg or more fissile 239Pu (FBR > 0.238) as objective parameters, and based on the three-dimensional Monte Carlo neutron-photon transport code MCNP/4A, a neutronics-optimizated calculation of different cases was carried out and the concept is proved feasible. In addition, the total breeding ratio ( BR = TBR + FBR ) is listed corresponding to different cases.

Neutronic Analysis of Generic Heavy Water Research Reactor Core Parameters to Use Standard Hydride Fuel

This research presents neutronic calculations for heavy water research reactor core substituting hydride fuel for uranium dioxide fuel. The aim of this research is feasibility analysis of reactor utilization with its original design using a new proposed fuel and changing the coolant and moderator circuit to light water. The required group constants for the CITATION code will be calculated using WIMSD-4 code. Neutronic calculations such as multiplication factors, radial and axial power peaking factor and fuel burn-up calculations are carried out by the CITATION code.

Preliminary performance analysis and optimization based on 1D neutronics model for Indian DEMO HCCB blanket

India,under its breeding blanket R&D program for DEMO,is focusing on the development of two tritium breeding blanket concepts;namely the lead-lithium-cooled ceramic breeder and the helium-cooled ceramic breeder(HCCB).The study presented in this paper focuses on the neutronic design analysis and optimization from the tritium breeding perspective of the HCCB blanket.The Indian concept has an edge-on configuration and is one of the variants of the helium-cooled solid breeder blanket concepts proposed by several partner countries in ITER.The Indian HCCB blanket having lithium titanate(Li2TiO3)as the tritium breeder and beryllium(Be)as the neutron multiplier with reduced-activation ferritic/martensitic steel structure aims at utilizing the low-energy neutrons at the rear part of the blanket.The aim of the optimization study is to minimize the radial blanket thickness while ensuring tritium self-sufficiency and provide data for further neutronic design and thermal-hydraulic layout of the HCCB blanket.It is found that inboard and outboard blanket thicknesses of 40 cm and 60 cm,respectively,can give a tritium breeding ratio(TBR)>1.3,with 60%6Li enrichment,which is assumed to be sufficient to cover potential tritium losses and associated uncertainties.The results also demonstrated that the Be packing fraction(PF)has a more profound impact on the TBR as compared to 6Li enrichment and the PF of Li2TiO3.

Neutronics Optimization of Tritium Breeding Blan-ket for the FDS

Neutronics optimization calculations have been performed for the tritium breed-ing blankets with solid ceramic breeder Li2O and 1iquid eutectic breeder Lil7Pb83, respectively,based on a 2-D geometrical configuration using the Monte Carlo neutron-photon transport codeMCNP/4B. The effects of beryllium, 6Li enrichment and various structural materials on TritiumBreeding Ratio have been systematically analyzed.

Study of Criticality Safety and Neutronic Performance for a 348-Fuel-Pin Ghana Research Reactor-1 LEU Core Using MCNP Code

The National Nuclear Research Institute of the Ghana Atomic Energy Commission is undertaking steps to convert the Ghana Research Reactor-1 from HEU Core to LEU. The proposed LEU core consists of 12.5% enriched UO2 fuel elements clad in Zircaloy-4 alloy. This is done in collaboration with Reduced Enrichment for Research and Test Reactor. The versatile MCNP code was used to analyse the neutronics parameters given in the SAR of HEU core, thereby characterizing the core. Subsequently, the LEU core was indentified with necessary changes to the HEU MCNP model. It was ascertained that the reactivity for the LEU core with the same number of fuel pins as the HEU was inadequate, hence the fuel pins were increased from 344 to 348. The neutron flux at the irradiation sites was found to be below the nominal value at full power for the LEU and hence the nominal power was increased to 34 kW for a nominal flux value of 1 × 1012 n/cm2.s. The parameters investigated for the HEU and LEU are shown in this paper.

Analysis of Neutronics and Thermal-Hydraulic Behavior in a Fuel Pin of Pressurized Water Reactor (PWR)

This paper presents a comparative analysis of different parameters such as enthalpy, moderator temperature, moderator density, flow velocity, pressure, and fuel temperature profile at the fuel pin cell level of PWR. Moreover, in this paper pitches to fuel pin radius ratio are varied from 2.3 to 4. The methods and implementation strategy are such that the coupled neutronic and thermal-hydraulic analysis is executed in a fully one dimensional (1D) manner. The thermal hydraulic is based on moderator/coolant mass and enthalpy equation together with one group diffusion equation for fuel pin. Modelling of fuel pin cell and subchannel is executed in two steps. First, the governing equations are derived assuming that all the parameters appearing in the equations are temperature independent. Fuel pin centerline temperature and radially averaged temperature equations are derived from Fourier laws of thermal conductivity. Finally, diffusion coefficient, fission cross-section and absorbing cross-section are evaluated with respect to the fuel pin temperature. The outcome will be helpful for further neutronics and thermal analysis of PWR. Thermal hydraulics parameter varies the maximum 30 percentage from the lowermost value.

Neutronics Comparison Analysis of the Water Cooled Ceramics Breeding Blanket for CFETR

China Fusion Engineering Test Reactor（CFETR） is an ITER-like fusion engineering test reactor that is intended to fill the scientific and technical gaps between ITER and DEMO.One of the main missions of CFETR is to achieve a tritium breeding ratio that is no less than 1.2to ensure tritium self-sufficiency.A concept design for a water cooled ceramics breeding blanket（WCCB） is presented based on a scheme with the breeder and the multiplier located in separate panels for CFETR.Based on this concept,a one-dimensional（1D） radial built breeding blanket was first designed,and then several three-dimensional models were developed with various neutron source definitions and breeding blanket module arrangements based on the 1D radial build.A set of nuclear analyses have been carried out to compare the differences in neutronics characteristics given by different calculation models,addressing neutron wall loading（NWL）,tritium breeding ratio（TBR）,fast neutron flux on inboard side and nuclear heating deposition on main in-vessel components.The impact of differences in modeling on the nuclear performance has been analyzed and summarized regarding the WCCB concept design.

5-17 Neutronic Research on Gas-cooled Travelling Wave Fast Reactor

In order to improve the utilization of uranium resources and alleviate nuclear waste disposal pressure, a schemeof 500MWth Gas-cooled Travelling Wave Fast Reactor (GTWFR) has been designed. The MCNPX and CINDER90code have been used to calculate and analysis the neutronic parameters in GTWFR scheme within 25 burning years.The results show that the GTWFR can keep self-sustaining within 20 years without refueling using once-throughfuel cycle. The system has promising neutronics performance, as it has flat power distribution, deep burnup, andcan remain stable in ultra-long period in consequence of the transmutation of breeding wave.

Neutronic analysis of deuteron-driven spallation target

Deuteron-driven spallation targets have garnered attention recently because they can provide high-energy neutrons to transmute long-lifetime fission products.In this study,the Geant4 toolkit was used to simulate the inter-action between a deuteron beam at 500 MeV and a com-posite target composed of alternating lead-bismuth eutectic(LBE)and water.The water was used because it may be employed as a target coolant.The energy spectrum,neu-tron yield,average energy,and total energy of the emitted neutrons were calculated for different thicknesses and thickness ratios between the LBE and water.For a constant target thickness,the neutron yield increases with an increasing thickness ratio of LBE to H 2 O,while the aver-age energy of the emitted neutrons decreases with an increasing in the aforementioned thickness ratio.These two aspects support the use of a pure target,either LBE or water.However,with an increasing LBE-to-H 2 O thickness ratio,the total energy of the emitted neutrons increases and then decreases.This result supports the addition of water into the LBE target.The angular distributions of the emitted neutrons show that the rear of the target is suit-able for loading nuclear waste containing minor actinides and long-lifetime fission products.