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 mul...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.展开更多
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 th...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.展开更多
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 si...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.展开更多
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 an...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.展开更多
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 singl...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.展开更多
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 an...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.展开更多
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 u...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.展开更多
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. M...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.展开更多
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 a...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.展开更多
China HC-SB TBM is designed as 3×3 sub-modules which making its structure to become robust but much complex. HC-SB TBM box is mounted inside a 20 cm of frame for a 1/4 of 1TER test port with 66.4 cm in width, an...China HC-SB TBM is designed as 3×3 sub-modules which making its structure to become robust but much complex. HC-SB TBM box is mounted inside a 20 cm of frame for a 1/4 of 1TER test port with 66.4 cm in width, and 44.5 cm in height, and 67 cm in depth. There have two 2 cm of gaps on side faces. A frame thickness around TBM module is 20 cm. In module of HC-SB TBM, LiaSiO4 pebble bed is used as tritium breeder zone. The packing factor of LiaSiO4 pebble bed is 0.59. The concentration of Li6 is 80%. Be pebble bed is used as neutron multiplication zone. The packing factor for Be pebble bed is selected as 0.8. All structural materials are Eurofer in which are helium cooling channel with diameter 0.6 cm of circular cross section. All sub-modules have a common FW that is independently cooled. There are only 4 Li4SiO4 regions which is lower volume ratio compared to Be. It is an advantage because Be has a better heat conductivity than Li4SiO4.展开更多
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-l...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.展开更多
The Local Monte Carlo(LMC)method is used to solve the problems of deep penetration and long time in the neutronics calculation of the radial neutron camera(RNC)diagnostic system on the experimental advanced supercondu...The Local Monte Carlo(LMC)method is used to solve the problems of deep penetration and long time in the neutronics calculation of the radial neutron camera(RNC)diagnostic system on the experimental advanced superconducting tokamak(EAST),and the radiation distribution of the RNC and the neutron flux at the detector positions of each channel are obtained.Compared with the results calculated by the global variance reduction method,it is shown that the LMC calculation is reliable within the reasonable error range.The calculation process of LMC is analyzed in detail,and the transport process of radiation particles is simulated in two steps.In the first step,an integrated neutronics model considering the complex window environment and a neutron source model based on EAST plasma shape are used to support the calculation.The particle information on the equivalent surface is analyzed to evaluate the rationality of settings of equivalent surface source and boundary.Based on the characteristic that only a local geometric model is needed in the second step,it is shown that the LMC is an advantageous calculation method for the nuclear shielding design of tokamak diagnostic systems.展开更多
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(HC...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.展开更多
The Molten Salt Reactor （MSR）, one of the ‘Generation Ⅳ＇ concepts, is a liquid-fuel reactor, which is different from the conventional reactors using solid fissile materials due to the flow effect of fuel salt. Th...The Molten Salt Reactor （MSR）, one of the ‘Generation Ⅳ＇ concepts, is a liquid-fuel reactor, which is different from the conventional reactors using solid fissile materials due to the flow effect of fuel salt. The study on its neutronics considering the fuel salt flow, which is the base of the thermal-hydraulic calculation and safety analysis, must be done. In this paper, the theoretical model on neutronics under steady condition for a single-liquid-fueled MSR is conducted and calculated by numerical method. The neutronics model consists of two group neutron diffusion equations for fast and thermal neutron fluxes, and balance equations for six-group delayed neutron precursors considering the flow effect of fuel salt. The spatial discretization of the above models is based on the finite volume method, and the discretization equations are computed by the source iteration method. The distributions of neutron fluxes and the distributions of the delayed neutron precursors in the core are obtained. The numerical calculated results show that, the fuel salt flow has little effect on the distribution of fast and thermal neutron fluxes and the effective multiplication factor; however, it affects the distribution of the delayed neutron precursors significantly, especially the long-lived one. In addition, it could be found that the delayed neutron precursors influence the neutronics slightly under the steady condition.展开更多
The multi-principal element characteristic of high-entropy alloys has revolutionized the conventional alloy design concept of single-principal element,endowing them with excellent mechanical properties.However,owing t...The multi-principal element characteristic of high-entropy alloys has revolutionized the conventional alloy design concept of single-principal element,endowing them with excellent mechanical properties.However,owing to this multi-principal element nature,high-entropy alloys exhibit complex deformation behavior dominated by alternating and coupled deformation mechanisms.Therefore,elucidating these intricate deformation mechanisms remains a key challenge in current research.Neutron diffraction(ND)techniques offer distinct advantages over traditional microscopic methods for characterizing such complex deformation behavior.The strong penetration capability of neutrons enables in-situ,real-time,and non-destructive detection of structural evolution in most centimeter-level bulk samples under complex environments,and ND allows precise characterization of lattice site occupations for light elements,such as C and O,and neighboring elements.This review discussed the principles of ND,experiment procedures,and data analysis.Combining with recent advances in the research about face-centered cubic high-entropy alloy,typical examples of using ND to investigate the deformation behavior were summarized,ultimately revealing deformation mechanisms dominated by dislocations,stacking faults,twinning,and phase transformations.展开更多
To generate a neutron beam exhibiting a Maxwellian energy distribution with narrow emission angles for measuring the neutron capture reaction rates of the s-process nuclides,a monoenergetic 3.4 MeV proton beam produce...To generate a neutron beam exhibiting a Maxwellian energy distribution with narrow emission angles for measuring the neutron capture reaction rates of the s-process nuclides,a monoenergetic 3.4 MeV proton beam produced by the tandem-accelerator in the China Institute of Atomic Energy was utilized.The proton beam was first transmitted through a 60.5μm aluminum foil and then impinged on a natural LiF target to produce neutron beam via^(7)Li(p,n)7Be reaction.The quasi-Gaussian energy distribution of protons in the LiF target resulted in neutron energy spectra that agreed with a Maxwellian energy distribution at kT=(22±2)keV,which was achieved by integrating neutrons detected within an emission angle of 65.0°±2.6°using a ^(6)Li glass detector positioned at 65°relative to the proton beam direction.The narrow angular spread of the Maxwelliandistributed neutron beam enables direct measurement of neutron capture cross-sections for most s-process nuclides,overcoming previous experimental limitations associated with broad angular distributions.展开更多
At high count rates,pile-up events involving neutron and gamma signals result in inaccurate neutron counting and distortions in the energy spectrum.Additionally,a bipolar cusp-like pulse shaping algorithm based on an ...At high count rates,pile-up events involving neutron and gamma signals result in inaccurate neutron counting and distortions in the energy spectrum.Additionally,a bipolar cusp-like pulse shaping algorithm based on an unfolding synthesis technique was proposed.This algorithm exhibits a narrow pulse shape,and the parallel design of the dual algorithms enables the recovery of pile-up signal amplitudes while preserving the distinct characteristics of neutron and gamma signals.The simplicity of the algorithm facilitates real-time neutron/gamma discrimination on an FPGA,allowing the energy spectra to be updated with each incoming signal.Furthermore,the algorithm can be readily tailored to various experimental conditions by adjusting the decay time constants.Multi-objective optimization reduces the need for manual parameter tuning by rapidly identifying the optimal parameters.Testing with a^(241)Am-Be neutron source and a NaIL scintillator yielded a figure of merit(FoM)value of 2.11 and produced a clear energy spectrum even at high count rates.展开更多
In recent years,there have been fewer missions to detect neutrons in low Earth orbits(LEO),and the data obtained have been extremely limited.Studying the distribution of the neutron energy spectrum in LEO satellites t...In recent years,there have been fewer missions to detect neutrons in low Earth orbits(LEO),and the data obtained have been extremely limited.Studying the distribution of the neutron energy spectrum in LEO satellites through detection can help solve three major scientific problems:the source of particles in the inner radiation belt,information on solar-accelerated particles,and the proportion of neutrons from different sources in near-Earth space.The detection efficiency and accuracy of neutrons are affected by charged and primary particles in the environment and secondary neutrons produced by the spacecraft itself,which has been a hot research topic.The neutron spectrometer developed in this study adopts two combinations of 15 silicon detectors in terms of detector type and arrangement,which are used for neutron detection via the nuclear reaction method and recoil proton method,respectively,in which a 27μm-thick^(6)LiF conversion layer is used for thermal neutron detection up to 0.4 eV and a 300μm-thick high-density polyethylene conversion layer is used for fast-neutron detection up to 14 MeV and below.The design of the detector set can also remove the influence of primary charged particles and secondary neutrons in the detection environment to a certain extent,thereby improving the accuracy of neutron detection.In this study,the neutron spectrometer hardware,firmware,software design,and basic performance of the front-end readout chip SKIROC2A were tested.The readout circuit of each channel baseline ADC code was less than 17;thus,the channel consistency was good.The RMS noise of the channel baseline was only 7.1 mV and exhibited good stability.The maximum number of events that could be processed per second is 75.The overall power consumption was 3 W,the weight was 792 g,and the volume was less than 1 dm^(3).Furthermore,the neutron spectrometer was tested for principle and detection efficiency using various neutron sources,such as ^(241)Am-Be neutron source,2.5 MeV neutron beam,and 14 MeV neutron beam,and the experiments were analyzed with corresponding simulations.The experimental data and simulation results were in good agreement and met the design requirements.The intrinsic detection efficiency of the probes used in the neutron spectrometer was 1.05%for 14 MeV fast neutrons.展开更多
From an engineering feasibility standpoint, what level of performance metrics can be ultimately achieved when designing a reactor using well-established nuclear fuels and structural materials that have already undergo...From an engineering feasibility standpoint, what level of performance metrics can be ultimately achieved when designing a reactor using well-established nuclear fuels and structural materials that have already undergone irradiation testing? The irradiation capability, which hinges on parameters like neutron flux level, irradiation channels' volume, and fuel cycle duration, is a core indicator for high-flux reactors. We propose a conceptual design of an ultra-high flux fast reactor(UFFR) with strong irradiation capability, which utilizes U-20Pu-10Zr alloy fuel and employs lead-bismuth as the coolant. The maximum neutron flux in the core reaches 1.32×10^(16) cm^(-2)s^(-1), while the average neutron flux in the irradiation channels attains 1.19×10^(16) cm^(-2)s^(-1). The volume of the central irradiation channel exceeds 10000 cm^(3), and the fuel cycle duration is 165 d, placing all its performance indicators among the top in the world. Based on the analyses of reactor physics and thermalhydraulics, it has been demonstrated that all reactivity coefficients are negative and all physical parameters meet the design criteria, ensuring the inherent safety of UFFR. An assessment of the irradiation capability has been carried out based on californium-252(^(252)Cf) production, indicating that the irradiation capability of UFFR surpasses that of the high flux isotope reactor(HFIR). The yield of ^(252)Cf from UFFR is 14.39 times that of HFIR, and its nuclei conversion rate is 3.21 times that of HFIR.展开更多
The ultracold neutron(UCN)transport code,MCUCN,designed initially for simulating UCN transportation from a solid deuterium(SD_2)source and neutron electric dipole moment experiments,could not simulate UCN storage and ...The ultracold neutron(UCN)transport code,MCUCN,designed initially for simulating UCN transportation from a solid deuterium(SD_2)source and neutron electric dipole moment experiments,could not simulate UCN storage and transportation in a superfluid^(4)He(SFHe,He-Ⅱ)source accurately.This limitation arose from the absence of an^(4)He upscattering mechanism and the absorption of^(3)He.And the provided source energy distribution in MCUCN is different from that in SFHe source.This study introduced enhancements to MCUCN to address these constraints,explicitly incorporating the^(4)He upscattering effect,the absorption of^(3)He,the loss caused by impurities on converter wall,UCN source energy distribution in SFHe,and the transmission through negative optical potential.Additionally,a Python-based visualization code for intermediate states and results was developed.To validate these enhancements,we systematically compared the simulation results of the Lujan Center Mark3 UCN system by MCUCN and the improved MCUCN code(iMCUCN)with UCNtransport simulations.Additionally,we compared the results of the SUN1 system simulated by MCUCN and iMCUCN with measurement results.The study demonstrates that iMCUCN effectively simulates the storage and transportation of ultracold neutrons in He-Ⅱ.展开更多
基金supported by the Natural Science Foundation of Anhui Province(No.1808085MA10)Anhui Provincial Key R&D Program(No.202104g0102007)the National Natural Science Foundation of China(No.21805283)。
文摘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.
基金This work was supported by the Chinese TMSR Strategic Pioneer Science and Technology Project(No.XDA02010000)the Frontier Science Key Program of Chinese Academy of Sciences(No.QYZDYSSW-JSC016)the Shanghai Sailing Program(No.Y931021031).
文摘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.
基金supported by the China Postdoctoral Science Foundation(No.2021M703045)the National Natural Science Foundation of China(No.12075067)the National Key R&D Program of China(No.2018YFE0180900).
文摘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.
基金supported by the National Special Project for Magnetic Confined Nuclear Fusion Energy(Grant Nos 2013GB108004,2014GB119000,and 2015BG108002)
文摘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.
基金supported by Natural Science Foundation of China (No. 12305190)Lingchuang Research Project of China National Nuclear Corporation (CNNC)the Science and Technology on Reactor System Design Technology Laboratory
文摘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.
基金This work was supported by the Chinese Academy of Sciences and the National Natural Science Foundation of China No.10175068.
文摘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.
文摘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.
文摘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.
基金supported by the National Special Project for Magnetic Confined Nuclear Fusion Energy(Nos.2013GB108004,2014GB122000,and2014GB119000)National Natural Science Foundation of China(No.11175207)
文摘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.
文摘China HC-SB TBM is designed as 3×3 sub-modules which making its structure to become robust but much complex. HC-SB TBM box is mounted inside a 20 cm of frame for a 1/4 of 1TER test port with 66.4 cm in width, and 44.5 cm in height, and 67 cm in depth. There have two 2 cm of gaps on side faces. A frame thickness around TBM module is 20 cm. In module of HC-SB TBM, LiaSiO4 pebble bed is used as tritium breeder zone. The packing factor of LiaSiO4 pebble bed is 0.59. The concentration of Li6 is 80%. Be pebble bed is used as neutron multiplication zone. The packing factor for Be pebble bed is selected as 0.8. All structural materials are Eurofer in which are helium cooling channel with diameter 0.6 cm of circular cross section. All sub-modules have a common FW that is independently cooled. There are only 4 Li4SiO4 regions which is lower volume ratio compared to Be. It is an advantage because Be has a better heat conductivity than Li4SiO4.
基金supported by Anhui Provincial Key R&D Program (202104g0102007)Hefei Municipal Natural Science Foundation (2022011)+2 种基金Collaborative Innovation Program of Hefei Science CenterChinese Academy of Sciences(2022HSC CIP024)International Partnership Program of Chinese Academy of Sciences (116134KYSB20200001)。
文摘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.
基金support and help in this research.This work was supported by Users with Excellence Program of Hefei Science Center CAS(No.2020HSC-UE012)Comprehensive Research Facility for Fusion Technology Program of China(No.2018-000052-73-01-001228)National Natural Science Foundation of China(No.11605241)。
文摘The Local Monte Carlo(LMC)method is used to solve the problems of deep penetration and long time in the neutronics calculation of the radial neutron camera(RNC)diagnostic system on the experimental advanced superconducting tokamak(EAST),and the radiation distribution of the RNC and the neutron flux at the detector positions of each channel are obtained.Compared with the results calculated by the global variance reduction method,it is shown that the LMC calculation is reliable within the reasonable error range.The calculation process of LMC is analyzed in detail,and the transport process of radiation particles is simulated in two steps.In the first step,an integrated neutronics model considering the complex window environment and a neutron source model based on EAST plasma shape are used to support the calculation.The particle information on the equivalent surface is analyzed to evaluate the rationality of settings of equivalent surface source and boundary.Based on the characteristic that only a local geometric model is needed in the second step,it is shown that the LMC is an advantageous calculation method for the nuclear shielding design of tokamak diagnostic systems.
文摘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.
基金National Nature Science Foundation of China (10575079)
文摘The Molten Salt Reactor （MSR）, one of the ‘Generation Ⅳ＇ concepts, is a liquid-fuel reactor, which is different from the conventional reactors using solid fissile materials due to the flow effect of fuel salt. The study on its neutronics considering the fuel salt flow, which is the base of the thermal-hydraulic calculation and safety analysis, must be done. In this paper, the theoretical model on neutronics under steady condition for a single-liquid-fueled MSR is conducted and calculated by numerical method. The neutronics model consists of two group neutron diffusion equations for fast and thermal neutron fluxes, and balance equations for six-group delayed neutron precursors considering the flow effect of fuel salt. The spatial discretization of the above models is based on the finite volume method, and the discretization equations are computed by the source iteration method. The distributions of neutron fluxes and the distributions of the delayed neutron precursors in the core are obtained. The numerical calculated results show that, the fuel salt flow has little effect on the distribution of fast and thermal neutron fluxes and the effective multiplication factor; however, it affects the distribution of the delayed neutron precursors significantly, especially the long-lived one. In addition, it could be found that the delayed neutron precursors influence the neutronics slightly under the steady condition.
基金National Key R&D Program of China(2023YFB3711904,2022YFA1603801)National Natural Science Foundation of China(12404230,52471181,52301213,52130108,52471005)+2 种基金National Nature Science Foundation of Zhejiang Province(LY23E010002)Open Fund of the China Spallation Neutron Source,Songshan Lake Science City(KFKT2023B11)Guangdong Basic and Applied Basic Research Foundation(2022A1515110805,2024A1515010878)。
文摘The multi-principal element characteristic of high-entropy alloys has revolutionized the conventional alloy design concept of single-principal element,endowing them with excellent mechanical properties.However,owing to this multi-principal element nature,high-entropy alloys exhibit complex deformation behavior dominated by alternating and coupled deformation mechanisms.Therefore,elucidating these intricate deformation mechanisms remains a key challenge in current research.Neutron diffraction(ND)techniques offer distinct advantages over traditional microscopic methods for characterizing such complex deformation behavior.The strong penetration capability of neutrons enables in-situ,real-time,and non-destructive detection of structural evolution in most centimeter-level bulk samples under complex environments,and ND allows precise characterization of lattice site occupations for light elements,such as C and O,and neighboring elements.This review discussed the principles of ND,experiment procedures,and data analysis.Combining with recent advances in the research about face-centered cubic high-entropy alloy,typical examples of using ND to investigate the deformation behavior were summarized,ultimately revealing deformation mechanisms dominated by dislocations,stacking faults,twinning,and phase transformations.
基金National Natural Science Foundation of China(12125509,11961141003,12275361,U2267205,12175152,12175121)National Key Research and Development Project(2022YFA1602301)Continuous-support Basic Scientific Research Project。
文摘To generate a neutron beam exhibiting a Maxwellian energy distribution with narrow emission angles for measuring the neutron capture reaction rates of the s-process nuclides,a monoenergetic 3.4 MeV proton beam produced by the tandem-accelerator in the China Institute of Atomic Energy was utilized.The proton beam was first transmitted through a 60.5μm aluminum foil and then impinged on a natural LiF target to produce neutron beam via^(7)Li(p,n)7Be reaction.The quasi-Gaussian energy distribution of protons in the LiF target resulted in neutron energy spectra that agreed with a Maxwellian energy distribution at kT=(22±2)keV,which was achieved by integrating neutrons detected within an emission angle of 65.0°±2.6°using a ^(6)Li glass detector positioned at 65°relative to the proton beam direction.The narrow angular spread of the Maxwelliandistributed neutron beam enables direct measurement of neutron capture cross-sections for most s-process nuclides,overcoming previous experimental limitations associated with broad angular distributions.
基金supported by the National Natural Science Foundation of China(NSFC)(No.12075308)。
文摘At high count rates,pile-up events involving neutron and gamma signals result in inaccurate neutron counting and distortions in the energy spectrum.Additionally,a bipolar cusp-like pulse shaping algorithm based on an unfolding synthesis technique was proposed.This algorithm exhibits a narrow pulse shape,and the parallel design of the dual algorithms enables the recovery of pile-up signal amplitudes while preserving the distinct characteristics of neutron and gamma signals.The simplicity of the algorithm facilitates real-time neutron/gamma discrimination on an FPGA,allowing the energy spectra to be updated with each incoming signal.Furthermore,the algorithm can be readily tailored to various experimental conditions by adjusting the decay time constants.Multi-objective optimization reduces the need for manual parameter tuning by rapidly identifying the optimal parameters.Testing with a^(241)Am-Be neutron source and a NaIL scintillator yielded a figure of merit(FoM)value of 2.11 and produced a clear energy spectrum even at high count rates.
基金supported by the National Natural Science Foundation of China(NSFC)(Nos.42225405 and U2106202)。
文摘In recent years,there have been fewer missions to detect neutrons in low Earth orbits(LEO),and the data obtained have been extremely limited.Studying the distribution of the neutron energy spectrum in LEO satellites through detection can help solve three major scientific problems:the source of particles in the inner radiation belt,information on solar-accelerated particles,and the proportion of neutrons from different sources in near-Earth space.The detection efficiency and accuracy of neutrons are affected by charged and primary particles in the environment and secondary neutrons produced by the spacecraft itself,which has been a hot research topic.The neutron spectrometer developed in this study adopts two combinations of 15 silicon detectors in terms of detector type and arrangement,which are used for neutron detection via the nuclear reaction method and recoil proton method,respectively,in which a 27μm-thick^(6)LiF conversion layer is used for thermal neutron detection up to 0.4 eV and a 300μm-thick high-density polyethylene conversion layer is used for fast-neutron detection up to 14 MeV and below.The design of the detector set can also remove the influence of primary charged particles and secondary neutrons in the detection environment to a certain extent,thereby improving the accuracy of neutron detection.In this study,the neutron spectrometer hardware,firmware,software design,and basic performance of the front-end readout chip SKIROC2A were tested.The readout circuit of each channel baseline ADC code was less than 17;thus,the channel consistency was good.The RMS noise of the channel baseline was only 7.1 mV and exhibited good stability.The maximum number of events that could be processed per second is 75.The overall power consumption was 3 W,the weight was 792 g,and the volume was less than 1 dm^(3).Furthermore,the neutron spectrometer was tested for principle and detection efficiency using various neutron sources,such as ^(241)Am-Be neutron source,2.5 MeV neutron beam,and 14 MeV neutron beam,and the experiments were analyzed with corresponding simulations.The experimental data and simulation results were in good agreement and met the design requirements.The intrinsic detection efficiency of the probes used in the neutron spectrometer was 1.05%for 14 MeV fast neutrons.
基金supported by the National Natural Science Foundation of China (Grant No.12575180)the Lingchuang Research Project of China National Nuclear Corporation (CNNC)。
文摘From an engineering feasibility standpoint, what level of performance metrics can be ultimately achieved when designing a reactor using well-established nuclear fuels and structural materials that have already undergone irradiation testing? The irradiation capability, which hinges on parameters like neutron flux level, irradiation channels' volume, and fuel cycle duration, is a core indicator for high-flux reactors. We propose a conceptual design of an ultra-high flux fast reactor(UFFR) with strong irradiation capability, which utilizes U-20Pu-10Zr alloy fuel and employs lead-bismuth as the coolant. The maximum neutron flux in the core reaches 1.32×10^(16) cm^(-2)s^(-1), while the average neutron flux in the irradiation channels attains 1.19×10^(16) cm^(-2)s^(-1). The volume of the central irradiation channel exceeds 10000 cm^(3), and the fuel cycle duration is 165 d, placing all its performance indicators among the top in the world. Based on the analyses of reactor physics and thermalhydraulics, it has been demonstrated that all reactivity coefficients are negative and all physical parameters meet the design criteria, ensuring the inherent safety of UFFR. An assessment of the irradiation capability has been carried out based on californium-252(^(252)Cf) production, indicating that the irradiation capability of UFFR surpasses that of the high flux isotope reactor(HFIR). The yield of ^(252)Cf from UFFR is 14.39 times that of HFIR, and its nuclei conversion rate is 3.21 times that of HFIR.
基金the National Key R&D Program of China(No.2024YFE0110001)the National Natural Science Foundation of China(U1932219)the Mobility Programme endorsed by the Joint Committee of the Sino-German Center(M0728)。
文摘The ultracold neutron(UCN)transport code,MCUCN,designed initially for simulating UCN transportation from a solid deuterium(SD_2)source and neutron electric dipole moment experiments,could not simulate UCN storage and transportation in a superfluid^(4)He(SFHe,He-Ⅱ)source accurately.This limitation arose from the absence of an^(4)He upscattering mechanism and the absorption of^(3)He.And the provided source energy distribution in MCUCN is different from that in SFHe source.This study introduced enhancements to MCUCN to address these constraints,explicitly incorporating the^(4)He upscattering effect,the absorption of^(3)He,the loss caused by impurities on converter wall,UCN source energy distribution in SFHe,and the transmission through negative optical potential.Additionally,a Python-based visualization code for intermediate states and results was developed.To validate these enhancements,we systematically compared the simulation results of the Lujan Center Mark3 UCN system by MCUCN and the improved MCUCN code(iMCUCN)with UCNtransport simulations.Additionally,we compared the results of the SUN1 system simulated by MCUCN and iMCUCN with measurement results.The study demonstrates that iMCUCN effectively simulates the storage and transportation of ultracold neutrons in He-Ⅱ.
