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 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
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
We present a theoretical model for detecting axions from neutron stars in a QCD phase of quark matter. The axions would be produced from a quark-antiquark pair uu¯or dd¯, in loop(s) involving gluons. The chi...We present a theoretical model for detecting axions from neutron stars in a QCD phase of quark matter. The axions would be produced from a quark-antiquark pair uu¯or dd¯, in loop(s) involving gluons. The chiral anomaly of QCD and the spontaneously broken symmetry are invoked to explain the non-conservation of the axion current. From the coupling form factors, the axion emissivities ϵacan be derived, from which fluxes can be determined. We predict a photon flux, which may be detectable by Fermi LAT, and limits on the QCD mass ma. In this model, axions decay to gamma rays in a 2-photon vertex. We may determine the expected fluxes from the theoretical emissivity. The sensitivity curve from the Fermi Large Area Telescope (Fermi LAT) would allow axion mass constraints for neutron stars as low as ma≤10−14eV 95% C.L. Axions could thus be detectable in gamma rays for neutron stars as distant as 100 kpc. A signal from LIGO GWS 170817 could be placed from the NS-NS merger, which gives an upper limit of ma≤10−10eV.展开更多
The acquisition of neutron time spectrum data plays a pivotal role in the precise quantification of uranium via prompt fission neutron uranium logging(PFNUL).However,the impact of the detector dead-time effect remains...The acquisition of neutron time spectrum data plays a pivotal role in the precise quantification of uranium via prompt fission neutron uranium logging(PFNUL).However,the impact of the detector dead-time effect remains paramount in the accurate acquisition of the neutron time spectrum.Therefore,it is imperative for neutron logging instruments to establish a dead-time correction method that is not only uncomplicated but also practical and caters to various logging sites.This study has formulated an innovative equation for determining dead time and introduced a dead-time correction method for the neutron time spectrum,called the“dual flux method.”Using this approach,a logging instrument captures two neutron time spectra under disparate neutron fluxes.By carefully selecting specific“windows”on the neutron time spectrum,the dead time can be accurately ascertained.To substantiate its efficacy and discern the influencing factors,experiments were conducted utilizing a deuterium-tritium(D-T)neutron source,a Helium-3(3He)detector,and polyethylene shielding to collate and analyze the neutron time spectrum under varying neutron fluxes(at high voltages).The findings underscore that the“height”and“spacing”of the two windows are the most pivotal influencing factors.Notably,the“height”(fd)should surpass 2,and the“spacing”twd should exceed 200μs.The dead time of the 3 He detector determined in the experiment was 7.35μs.After the dead-time correction,the deviation of the decay coefficients from the theoretical values for the neutron time spectrum under varying neutron fluxes decreased from 12.4%to within 5%.Similarly,for the PFNUL instrument,the deviation in the decay coefficients decreased from 22.94 to 0.49%after correcting for the dead-time effect.These results demonstrate the exceptional efficacy of the proposed method in ensuring precise uranium quantification.The dual flux method was experimentally validated as a universal approach applicable to pulsed neutron logging instruments and holds immense significance for uranium exploration.展开更多
The NEutron Detector Array(NEDA)is designed to be coupled to gamma-ray spectrometers to enhance the sensitivity of the setup by enabling reaction channel selection through counting of the evaporated neutrons.This arti...The NEutron Detector Array(NEDA)is designed to be coupled to gamma-ray spectrometers to enhance the sensitivity of the setup by enabling reaction channel selection through counting of the evaporated neutrons.This article presents the implementation of a double trigger condition system for NEDA,which improves the acquisition of neutrons and reduces the number of gamma rays acquired.Two independent triggers are generated in the double trigger condition system:one based on charge comparison(CC)and the other on time-of-flight(TOF).These triggers can be combined using OR and AND logic,offering four distinct trigger modes.The developed firmware is added to the previous one in the Virtex 6 field programmable gate array(FPGA)present in the system,which also includes signal processing,baseline correction,and various trigger logic blocks.The performance of the trigger system is evaluated using data from the E703 experiment performed at GANIL.The four trigger modes are applied to the same data,and a subsequent offline analysis is performed.It is shown that most of the detected neutrons are preserved with the AND mode,and the total number of gamma rays is significantly reduced.Compared with the CC trigger mode,the OR trigger mode allows increasing the selection of neutrons.In addition,it is demonstrated that if the OR mode is selected,the online CC trigger threshold can be raised without losing neutrons.展开更多
We present new data on the^(63)Cu(γ,n)cross-section studied using a quasi-monochromatic and energy-tunableγbeam produced at the Shanghai Laser Electron Gamma Source to resolve the long-standing discrepancy between e...We present new data on the^(63)Cu(γ,n)cross-section studied using a quasi-monochromatic and energy-tunableγbeam produced at the Shanghai Laser Electron Gamma Source to resolve the long-standing discrepancy between existing measurements and evaluations of this cross-section.Using an unfolding iteration method,^(63)Cu(γ,n)data were obtained with an uncertainty of less than 4%,and the inconsistencies between the available experimental data were discussed.Theγ-ray strength function of^(63)Cu(γ,n)was successfully extracted as an experimental constraint.We further calculated the cross-section of the radiative neutron capture reaction^(62)Cu(n,γ)using the TALYS code.Our calculation method enables the extraction of(n,γ)cross-sections for unstable nuclides.展开更多
Pulse pile-up is a problem in nuclear spectroscopy and nuclear reaction studies that occurs when two pulses overlap and distort each other,degrading the quality of energy and timing information.Different methods have ...Pulse pile-up is a problem in nuclear spectroscopy and nuclear reaction studies that occurs when two pulses overlap and distort each other,degrading the quality of energy and timing information.Different methods have been used for pile-up rejection,both digital and analogue,but some pile-up events may contain pulses of interest and need to be reconstructed.The paper proposes a new method for reconstructing pile-up events acquired with a neutron detector array(NEDA)using an one-dimensional convolutional autoencoder(1D-CAE).The datasets for training and testing the 1D-CAE are created from data acquired from the NEDA.The new pile-up signal reconstruction method is evaluated from the point of view of how similar the reconstructed signals are to the original ones.Furthermore,it is analysed considering the result of the neutron-gamma discrimination based on charge comparison,comparing the result obtained from original and reconstructed signals.展开更多
The superconducting magnet system of a fusion reactor plays a vital role in plasma confinement,a process that can be dis-rupted by various operational factors.A critical parameter for evaluating the temperature margin...The superconducting magnet system of a fusion reactor plays a vital role in plasma confinement,a process that can be dis-rupted by various operational factors.A critical parameter for evaluating the temperature margin of superconducting magnets during normal operation is the nuclear heating caused by D-T neutrons.This study investigates the impact of nuclear heat-ing on a superconducting magnet system by employing an improved analysis method that combines neutronics and thermal hydraulics.In the magnet system,toroidal field(TF)magnets are positioned closest to the plasma and bear the highest nuclear-heat load,making them prime candidates for evaluating the influence of nuclear heating on stability.To enhance the modeling accuracy and facilitate design modifications,a parametric TF model that incorporates heterogeneity is established to expedite the optimization design process and enhance the accuracy of the computations.A comparative analysis with a homogeneous TF model reveals that the heterogeneous model improves accuracy by over 12%.Considering factors such as heat load,magnetic-field strength,and cooling conditions,the cooling circuit facing the most severe conditions is selected to calculate the temperature of the superconductor.This selection streamlines the workload associated with thermal-hydraulic analysis.This approach enables a more efficient and precise evaluation of the temperature margin of TF magnets.Moreover,it offers insights that can guide the optimization of both the structure and cooling strategy of superconducting magnet systems.展开更多
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.
文摘We present a theoretical model for detecting axions from neutron stars in a QCD phase of quark matter. The axions would be produced from a quark-antiquark pair uu¯or dd¯, in loop(s) involving gluons. The chiral anomaly of QCD and the spontaneously broken symmetry are invoked to explain the non-conservation of the axion current. From the coupling form factors, the axion emissivities ϵacan be derived, from which fluxes can be determined. We predict a photon flux, which may be detectable by Fermi LAT, and limits on the QCD mass ma. In this model, axions decay to gamma rays in a 2-photon vertex. We may determine the expected fluxes from the theoretical emissivity. The sensitivity curve from the Fermi Large Area Telescope (Fermi LAT) would allow axion mass constraints for neutron stars as low as ma≤10−14eV 95% C.L. Axions could thus be detectable in gamma rays for neutron stars as distant as 100 kpc. A signal from LIGO GWS 170817 could be placed from the NS-NS merger, which gives an upper limit of ma≤10−10eV.
基金supported by the National Natural Science Foundation of China(No.42374226)Jiangxi Provincial Natural Science Foundation(Nos.20232BAB201043 and 20232BCJ23006)+2 种基金Nuclear Energy Development Project(20201192-01)National Key Laboratory of Uranium Resource Exploration-Mining and Nuclear Remote Sensing(ECUT)(2024QZ-TD-09)Fundamental Science on Radioactive Geology and Exploration Technology Laboratory(2022RGET20).
文摘The acquisition of neutron time spectrum data plays a pivotal role in the precise quantification of uranium via prompt fission neutron uranium logging(PFNUL).However,the impact of the detector dead-time effect remains paramount in the accurate acquisition of the neutron time spectrum.Therefore,it is imperative for neutron logging instruments to establish a dead-time correction method that is not only uncomplicated but also practical and caters to various logging sites.This study has formulated an innovative equation for determining dead time and introduced a dead-time correction method for the neutron time spectrum,called the“dual flux method.”Using this approach,a logging instrument captures two neutron time spectra under disparate neutron fluxes.By carefully selecting specific“windows”on the neutron time spectrum,the dead time can be accurately ascertained.To substantiate its efficacy and discern the influencing factors,experiments were conducted utilizing a deuterium-tritium(D-T)neutron source,a Helium-3(3He)detector,and polyethylene shielding to collate and analyze the neutron time spectrum under varying neutron fluxes(at high voltages).The findings underscore that the“height”and“spacing”of the two windows are the most pivotal influencing factors.Notably,the“height”(fd)should surpass 2,and the“spacing”twd should exceed 200μs.The dead time of the 3 He detector determined in the experiment was 7.35μs.After the dead-time correction,the deviation of the decay coefficients from the theoretical values for the neutron time spectrum under varying neutron fluxes decreased from 12.4%to within 5%.Similarly,for the PFNUL instrument,the deviation in the decay coefficients decreased from 22.94 to 0.49%after correcting for the dead-time effect.These results demonstrate the exceptional efficacy of the proposed method in ensuring precise uranium quantification.The dual flux method was experimentally validated as a universal approach applicable to pulsed neutron logging instruments and holds immense significance for uranium exploration.
基金supported by MICIU MCIN/AEI/10.13039/501100011033Spain with Grant PID2020-118265GB-C42,-C44,PRTR-C17.I01Generalitat Valenciana,Spain with Grant CIPROM/2022/54,ASFAE/2022/031,CIAPOS/2021/114 and by the EU NextGenerationEU,ESF funds.This work was also supported by the National Science Centre(NCN),Poland(Grant No.2020/39/D/ST2/00466).
文摘The NEutron Detector Array(NEDA)is designed to be coupled to gamma-ray spectrometers to enhance the sensitivity of the setup by enabling reaction channel selection through counting of the evaporated neutrons.This article presents the implementation of a double trigger condition system for NEDA,which improves the acquisition of neutrons and reduces the number of gamma rays acquired.Two independent triggers are generated in the double trigger condition system:one based on charge comparison(CC)and the other on time-of-flight(TOF).These triggers can be combined using OR and AND logic,offering four distinct trigger modes.The developed firmware is added to the previous one in the Virtex 6 field programmable gate array(FPGA)present in the system,which also includes signal processing,baseline correction,and various trigger logic blocks.The performance of the trigger system is evaluated using data from the E703 experiment performed at GANIL.The four trigger modes are applied to the same data,and a subsequent offline analysis is performed.It is shown that most of the detected neutrons are preserved with the AND mode,and the total number of gamma rays is significantly reduced.Compared with the CC trigger mode,the OR trigger mode allows increasing the selection of neutrons.In addition,it is demonstrated that if the OR mode is selected,the online CC trigger threshold can be raised without losing neutrons.
基金supported by the National Key Research and Development Program(Nos.2023YFA1606901 and 2022YFA1602400)National Natural Science Foundation of China(Nos.U2230133,12275338,and 12388102)Open Fund of the CIAE Key Laboratory of Nuclear Data(No.JCKY2022201C152).
文摘We present new data on the^(63)Cu(γ,n)cross-section studied using a quasi-monochromatic and energy-tunableγbeam produced at the Shanghai Laser Electron Gamma Source to resolve the long-standing discrepancy between existing measurements and evaluations of this cross-section.Using an unfolding iteration method,^(63)Cu(γ,n)data were obtained with an uncertainty of less than 4%,and the inconsistencies between the available experimental data were discussed.Theγ-ray strength function of^(63)Cu(γ,n)was successfully extracted as an experimental constraint.We further calculated the cross-section of the radiative neutron capture reaction^(62)Cu(n,γ)using the TALYS code.Our calculation method enables the extraction of(n,γ)cross-sections for unstable nuclides.
基金partially supported by MICIU MCIN/AEI/10.13039/501100011033Spain with grant PID2020-118265GB-C42,-C44,PRTR-C17.I01+1 种基金Generalitat Valenciana,Spain with grant CIPROM/2022/54,ASFAE/2022/031,CIAPOS/2021/114the EU NextGenerationEU,ESF funds,and the National Science Centre (NCN),Poland (grant No.2020/39/D/ST2/00466)
文摘Pulse pile-up is a problem in nuclear spectroscopy and nuclear reaction studies that occurs when two pulses overlap and distort each other,degrading the quality of energy and timing information.Different methods have been used for pile-up rejection,both digital and analogue,but some pile-up events may contain pulses of interest and need to be reconstructed.The paper proposes a new method for reconstructing pile-up events acquired with a neutron detector array(NEDA)using an one-dimensional convolutional autoencoder(1D-CAE).The datasets for training and testing the 1D-CAE are created from data acquired from the NEDA.The new pile-up signal reconstruction method is evaluated from the point of view of how similar the reconstructed signals are to the original ones.Furthermore,it is analysed considering the result of the neutron-gamma discrimination based on charge comparison,comparing the result obtained from original and reconstructed signals.
基金the National Natural Science Foundation of China(Nos.52222701,52077211,and 52307034).
文摘The superconducting magnet system of a fusion reactor plays a vital role in plasma confinement,a process that can be dis-rupted by various operational factors.A critical parameter for evaluating the temperature margin of superconducting magnets during normal operation is the nuclear heating caused by D-T neutrons.This study investigates the impact of nuclear heat-ing on a superconducting magnet system by employing an improved analysis method that combines neutronics and thermal hydraulics.In the magnet system,toroidal field(TF)magnets are positioned closest to the plasma and bear the highest nuclear-heat load,making them prime candidates for evaluating the influence of nuclear heating on stability.To enhance the modeling accuracy and facilitate design modifications,a parametric TF model that incorporates heterogeneity is established to expedite the optimization design process and enhance the accuracy of the computations.A comparative analysis with a homogeneous TF model reveals that the heterogeneous model improves accuracy by over 12%.Considering factors such as heat load,magnetic-field strength,and cooling conditions,the cooling circuit facing the most severe conditions is selected to calculate the temperature of the superconductor.This selection streamlines the workload associated with thermal-hydraulic analysis.This approach enables a more efficient and precise evaluation of the temperature margin of TF magnets.Moreover,it offers insights that can guide the optimization of both the structure and cooling strategy of superconducting magnet systems.
