Email alert
RSS2022 Vol. 34, No. 2
Recommend Articles
Display Method:
2022,
34: 026001.
doi: 10.11884/HPLPB202234.210402
Abstract:
Monte Carlo method is becoming the first choice of particle transport simulation methods because of its strong capability of complicated geometry treatment, perfected nuclear data and universal property. Monte Carlo softwares have been rapidly developing with the development of computers and at present, they are widely used in applications sa radiation shielding, criticality safety analysis, post-treatment of spent fuel, radioactivity waste treatment, decommissioning retire of nuclear facilities, emergence of nuclear accident, nuclear well-logging and nuclear medical science, etc. This paper provides a brief introduction of the development of Monte Carlo simulations in the past and at present and perspectives about its future.
2022,
34: 026002.
doi: 10.11884/HPLPB202234.210372
Abstract:
To avoid inaccurate predicted values of zero power physics test (ZPPT) parameters, which will affect the commissioning and startup of the nuclear power plant, the numerical reactor can be used to accurately predict the ZPPT parameters. Based on the codes of the CAP1400 numerical reactor system, including the Monte Carlo code JMCT, the deterministic high-fidelity code NECP-X and advanced neutronics code SCAP-N, the CAP1400 first core was modeled to realize high-fidelity simulation of the ZPPT parameters. Numerical results show that the calculation results of the three high-fidelity simulation codes are in good agreement. Taking JMCT as a reference, the absolute deviations of NECP-X and SCAP-N for the gray control bank worth are within ±8×10−5, the relative deviations for the black control bank worth are within ±3%, the relative deviations for the total worth of all the black control banks are within ±1%, and the relative deviations for the assemble relative power are within ±2.5%. The ZPPT parameters of the CAP1400 first core are accurately predicted, which can effectively support the commissioning and startup of the CAP1400 reactor.
To avoid inaccurate predicted values of zero power physics test (ZPPT) parameters, which will affect the commissioning and startup of the nuclear power plant, the numerical reactor can be used to accurately predict the ZPPT parameters. Based on the codes of the CAP1400 numerical reactor system, including the Monte Carlo code JMCT, the deterministic high-fidelity code NECP-X and advanced neutronics code SCAP-N, the CAP1400 first core was modeled to realize high-fidelity simulation of the ZPPT parameters. Numerical results show that the calculation results of the three high-fidelity simulation codes are in good agreement. Taking JMCT as a reference, the absolute deviations of NECP-X and SCAP-N for the gray control bank worth are within ±8×10−5, the relative deviations for the black control bank worth are within ±3%, the relative deviations for the total worth of all the black control banks are within ±1%, and the relative deviations for the assemble relative power are within ±2.5%. The ZPPT parameters of the CAP1400 first core are accurately predicted, which can effectively support the commissioning and startup of the CAP1400 reactor.
2022,
34: 026003.
doi: 10.11884/HPLPB202234.210366
Abstract:
The MCNP program and the ENDF/B-VII.1, ENDF/B-VIII.0 nuclear data were used to calculate the criticality respect to the critical benchmarking suits of 119 models for nuclear data testing. The benchmarking suits consists of 233U, high-enriched uranium, intermediate enriched uranium, low-enriched uranium and Pu critical configurations in fuel type, covering the spectra of fast, intermediate, and thermal neutrons. With ENDF/B-VIII.0 nuclear data, the average and standard deviation of the relative deviation between the calculation keff and the benchmarking keff were −68.98×10−5 and 407.88×10−5 respectively, and the tallies that the ratio of deviation to experimental uncertainty within 1σ, among 1σ–3σ and beyond 3σ were 107, 9 and 3, respectively. The resultant data were compared to the calculation results with ENDF/B-VII. 1 in the same computation conditions, which indicates that the ENDF/B-VIII.0 nuclear data was more accurate, and more consistent with the benchmarking results for criticality calculation.
The MCNP program and the ENDF/B-VII.1, ENDF/B-VIII.0 nuclear data were used to calculate the criticality respect to the critical benchmarking suits of 119 models for nuclear data testing. The benchmarking suits consists of 233U, high-enriched uranium, intermediate enriched uranium, low-enriched uranium and Pu critical configurations in fuel type, covering the spectra of fast, intermediate, and thermal neutrons. With ENDF/B-VIII.0 nuclear data, the average and standard deviation of the relative deviation between the calculation keff and the benchmarking keff were −68.98×10−5 and 407.88×10−5 respectively, and the tallies that the ratio of deviation to experimental uncertainty within 1σ, among 1σ–3σ and beyond 3σ were 107, 9 and 3, respectively. The resultant data were compared to the calculation results with ENDF/B-VII. 1 in the same computation conditions, which indicates that the ENDF/B-VIII.0 nuclear data was more accurate, and more consistent with the benchmarking results for criticality calculation.
2022,
34: 026004.
doi: 10.11884/HPLPB202234.210378
Abstract:
The fuel assemblies bow with different stresses in the PWR reactor core and this phenomenon can significantly affect the local neutron moderation in the reactor core. Based on the CGN’s in-house software package PCM, which is composed of the fuel assembly neutron cross section software PINE and the core design software COCO, this study developed a specific model for fuel assembly bowing to analyze the impact on local power distribution, and compared the results with that of the Monto Carlo code JMCT. The study shows that the results of fuel assembly bowing model in PCM have good consistency with those of JMCT, and that PCM is valid in fuel assembly bowing analysis. The fuel assembly bowing has a significant impact on core local power distribution. This phenomenon shall be specially considered in the PWR reactor design.
The fuel assemblies bow with different stresses in the PWR reactor core and this phenomenon can significantly affect the local neutron moderation in the reactor core. Based on the CGN’s in-house software package PCM, which is composed of the fuel assembly neutron cross section software PINE and the core design software COCO, this study developed a specific model for fuel assembly bowing to analyze the impact on local power distribution, and compared the results with that of the Monto Carlo code JMCT. The study shows that the results of fuel assembly bowing model in PCM have good consistency with those of JMCT, and that PCM is valid in fuel assembly bowing analysis. The fuel assembly bowing has a significant impact on core local power distribution. This phenomenon shall be specially considered in the PWR reactor design.
2022,
34: 026005.
doi: 10.11884/HPLPB202234.210300
Abstract:
For the accurate measurement of environmental X/γ radiation dose rate, the conversion from energy spectrum to dose rate is used. In this paper, the energy spectrum and dose value of the electrically cooled high-purity germanium detector (HPGe detector) in the energy range of 0.01−1.5 MeV are obtained by Monte Carlo simulation. And the method “Conversion of complete spectra without deconvolution” is used for energy spectrum-dose conversion study. It is found that the dose rate calculated by Conversion of complete spectra without deconvolution method is in good agreement with the simulated dose rate. Experimental verification was carried out in the standard radiation field (Co-60, Cs-137) of the environmental γ radiation air absorbed dose of the China Institute of Metrology, the result shows that in the energy range of 0.01−1.5 MeV, the error between the dose rate value obtained by energy spectrum-dose conversion and the standard dose rate value is less than ±10%. The study shows that it is feasible to solve the energy spectrum-dose (rate) conversion coefficient by the method “Conversion of complete spectra without deconvolution”.
For the accurate measurement of environmental X/γ radiation dose rate, the conversion from energy spectrum to dose rate is used. In this paper, the energy spectrum and dose value of the electrically cooled high-purity germanium detector (HPGe detector) in the energy range of 0.01−1.5 MeV are obtained by Monte Carlo simulation. And the method “Conversion of complete spectra without deconvolution” is used for energy spectrum-dose conversion study. It is found that the dose rate calculated by Conversion of complete spectra without deconvolution method is in good agreement with the simulated dose rate. Experimental verification was carried out in the standard radiation field (Co-60, Cs-137) of the environmental γ radiation air absorbed dose of the China Institute of Metrology, the result shows that in the energy range of 0.01−1.5 MeV, the error between the dose rate value obtained by energy spectrum-dose conversion and the standard dose rate value is less than ±10%. The study shows that it is feasible to solve the energy spectrum-dose (rate) conversion coefficient by the method “Conversion of complete spectra without deconvolution”.
2022,
34: 026006.
doi: 10.11884/HPLPB202234.210309
Abstract:
To perform realistic core calculations, few-group neutron cross-sections library by functions of burn-up and thermal hydraulics parameters should be prepared in advance. Traditional deterministic parameterization process is based on the macro cross-section. However, this method should consider the historical effect of some physical states, which increases the number of calculation branches. Thus, this paper proposes a new parameterization process based on the micro cross-section of nuclide. This method effectively eliminates the historical impact. Therefore, only burn-up and material temperature need to be considered. The calculation process is performed by RMC code. Firstly, all of material nuclides micro cross-section and nuclides density is calculated. Then, macro cross-section is obtained by the micro cross-section lib. To test the method accuracy, a self-design pressurized water reactor model is built. The test results agree well with the reference results calculated by RMC full core calculation.
To perform realistic core calculations, few-group neutron cross-sections library by functions of burn-up and thermal hydraulics parameters should be prepared in advance. Traditional deterministic parameterization process is based on the macro cross-section. However, this method should consider the historical effect of some physical states, which increases the number of calculation branches. Thus, this paper proposes a new parameterization process based on the micro cross-section of nuclide. This method effectively eliminates the historical impact. Therefore, only burn-up and material temperature need to be considered. The calculation process is performed by RMC code. Firstly, all of material nuclides micro cross-section and nuclides density is calculated. Then, macro cross-section is obtained by the micro cross-section lib. To test the method accuracy, a self-design pressurized water reactor model is built. The test results agree well with the reference results calculated by RMC full core calculation.
2022,
34: 026007.
doi: 10.11884/HPLPB202234.210327
Abstract:
The relationship between 131I activity in the thyroid and radiation detection results is related to the thyroid geometry, detection distance and other factors, which is a key parameter for estimating the 131I content in the thyroid and its possible irradiation damage. We developed an application for gamma radiation measurement simulation of NaI detectors based on the MCPT radiation transport numerical simulation algorithm library. Then several physical models with different volumes of thyroid containers and different detection distances were established. Finally, the detection efficiency of detectors were obtained under different measurement states by Monte Carlo numerical calculations. The results of the numerical simulation are consistent with that of the theoretical calculation when the distance between the thyroid-like container and the detector is large, which proves that the application could be used to quantitatively analyze the detection efficiency of NaI detectors. It is indicated that the detection efficiency for the detector with small distances is significantly influenced by the size of the thyroid-like container and the distance. The detection efficiency table given by the simulations provides a basis for an in-depth experimental study.
The relationship between 131I activity in the thyroid and radiation detection results is related to the thyroid geometry, detection distance and other factors, which is a key parameter for estimating the 131I content in the thyroid and its possible irradiation damage. We developed an application for gamma radiation measurement simulation of NaI detectors based on the MCPT radiation transport numerical simulation algorithm library. Then several physical models with different volumes of thyroid containers and different detection distances were established. Finally, the detection efficiency of detectors were obtained under different measurement states by Monte Carlo numerical calculations. The results of the numerical simulation are consistent with that of the theoretical calculation when the distance between the thyroid-like container and the detector is large, which proves that the application could be used to quantitatively analyze the detection efficiency of NaI detectors. It is indicated that the detection efficiency for the detector with small distances is significantly influenced by the size of the thyroid-like container and the distance. The detection efficiency table given by the simulations provides a basis for an in-depth experimental study.
2022,
34: 026008.
doi: 10.11884/HPLPB202234.210352
Abstract:
To verify the accuracy and applicability of JMCT program, 56 models are selected from the International Handbook of Evaluated Criticality Safety Benchmark Experiments. Models are established in JLAMT, and the effective multiplication factors are calculated by JMCT and compared with the reference values and MCNP calculation results. The results show that the calculated values of JMCT are generally within three standard deviations of the reference values and the relative deviations between the JMCT calculation results and the reference values are generally less than 2%. The study has verified the accuracy and applicability of JMCT program to a certain extent, and it provides a reference for the subsequent application and verification of JMCT.
To verify the accuracy and applicability of JMCT program, 56 models are selected from the International Handbook of Evaluated Criticality Safety Benchmark Experiments. Models are established in JLAMT, and the effective multiplication factors are calculated by JMCT and compared with the reference values and MCNP calculation results. The results show that the calculated values of JMCT are generally within three standard deviations of the reference values and the relative deviations between the JMCT calculation results and the reference values are generally less than 2%. The study has verified the accuracy and applicability of JMCT program to a certain extent, and it provides a reference for the subsequent application and verification of JMCT.
2022,
34: 026009.
doi: 10.11884/HPLPB202234.210399
Abstract:
Based on tritium production mechanism and the tritium calculation model in China General Nuclear Power Corporation (CGN), the historic tritium discharges in lots of nuclear power plants around the world have been comprehensively gathered and analyzed. It is recognized that the activation of boric acid in the primary loops and activation of Be in secondary neutron source (40% of annual tritium discharge in CGN’s operating pressurized water reactor units) are the predominant origins, and the tritium release from intact fuel rods with zirconium alloy can be neglected. It is found that use of double encapsulated secondary neutron source or cancellation of secondary neutron source is the only way to reduce significantly the tritium discharges in the pressurized water reactors.
Based on tritium production mechanism and the tritium calculation model in China General Nuclear Power Corporation (CGN), the historic tritium discharges in lots of nuclear power plants around the world have been comprehensively gathered and analyzed. It is recognized that the activation of boric acid in the primary loops and activation of Be in secondary neutron source (40% of annual tritium discharge in CGN’s operating pressurized water reactor units) are the predominant origins, and the tritium release from intact fuel rods with zirconium alloy can be neglected. It is found that use of double encapsulated secondary neutron source or cancellation of secondary neutron source is the only way to reduce significantly the tritium discharges in the pressurized water reactors.
2022,
34: 026010.
doi: 10.11884/HPLPB202234.210264
Abstract:
To analyze the characteristics of burnable poisons used in the micro gas-cooled reactor, this paper investigates how the kinds and layout of burnable poisons influence the reactor characteristics, such as reactivity and lifetime, based on the long-lifetime core (15 MW-20 a) model, short-lifetime core (5 MW-1 a) model and longer-lifetime core (5 MW-3−10 a) model, using Monte Carlo procedure. The results show that, as for the long-lifetime core, the monolithic type Er2O3 can reduce the core excess reactivity effectively with a certain reactivity punishment at the end of lifetime, and the monolithic B4C can reduce the core excess reactivity excellently with a better power distribution if an inhomogeneous distribution layout of B4C is applied, and the separated type B4C makes the characteristic curve of burn-up flatter. As for the short-lifetime core, the separated type Gd2O3 is an appropriate choice while the B4C with usual layouts is not, but it has a good behavior if the B4C disperses in the core graphite as the burnable poison. As for the longer-lifetime core, the separated type Gd2O3 not only can control the core excess reactivity effectively, but also can guarantee the core have the inherent safety of automatic shutdown only by the negative temperature feedback. These results will instruct the follow-up development of micro gas-cooled reactor devices.
To analyze the characteristics of burnable poisons used in the micro gas-cooled reactor, this paper investigates how the kinds and layout of burnable poisons influence the reactor characteristics, such as reactivity and lifetime, based on the long-lifetime core (15 MW-20 a) model, short-lifetime core (5 MW-1 a) model and longer-lifetime core (5 MW-3−10 a) model, using Monte Carlo procedure. The results show that, as for the long-lifetime core, the monolithic type Er2O3 can reduce the core excess reactivity effectively with a certain reactivity punishment at the end of lifetime, and the monolithic B4C can reduce the core excess reactivity excellently with a better power distribution if an inhomogeneous distribution layout of B4C is applied, and the separated type B4C makes the characteristic curve of burn-up flatter. As for the short-lifetime core, the separated type Gd2O3 is an appropriate choice while the B4C with usual layouts is not, but it has a good behavior if the B4C disperses in the core graphite as the burnable poison. As for the longer-lifetime core, the separated type Gd2O3 not only can control the core excess reactivity effectively, but also can guarantee the core have the inherent safety of automatic shutdown only by the negative temperature feedback. These results will instruct the follow-up development of micro gas-cooled reactor devices.
2022,
34: 026011.
doi: 10.11884/HPLPB202234.210265
Abstract:
The radial power distribution of UO2 pellet is one of the significant parameters in the fuel rod design of nuclear power plant. The research on the calculation of radial power distribution of the UO2 pellet with the Monte-Carlo particle transport software JMCT coupling the burnup analysis software JBURN is presented. The calculation model, the set-up of the transport and burnup as well as the calculation method of the fission power, and capture power are introduced in this paper. The calculation results show that the result of JMCT-JBURN accord well with that of the reference calculation, and the maximum difference is less than 3%, which proves the applicability of the calculation method and software. This method can be used for engineering design.
The radial power distribution of UO2 pellet is one of the significant parameters in the fuel rod design of nuclear power plant. The research on the calculation of radial power distribution of the UO2 pellet with the Monte-Carlo particle transport software JMCT coupling the burnup analysis software JBURN is presented. The calculation model, the set-up of the transport and burnup as well as the calculation method of the fission power, and capture power are introduced in this paper. The calculation results show that the result of JMCT-JBURN accord well with that of the reference calculation, and the maximum difference is less than 3%, which proves the applicability of the calculation method and software. This method can be used for engineering design.
2022,
34: 026012.
doi: 10.11884/HPLPB202234.210298
Abstract:
Fuel rod design verification is the evaluation process of fuel rod safety performance during operation in reactor, in which the uncertainty of input parameters has important effect on evaluation results. To study the uncertainty systematically, fuel rod performance analysis software has been coupled with Dakota software to carry out fuel rod design verification, the results of nonparametric Monte Carlo and Latin Hypercube Sampling have been compared with those of traditional method. It turns out that the fuel rod inner pressure criterion is vulnerable to be under challenge for the reason of input uncertainty under consideration by traditional method. The defects can be made up by statistical nonparametric sampling, by which a larger safety margin is obtained, and a theoretical basis for fuel rod safety and economic performance enhancement is provided. Meanwhile, the temperature calculation result obtained by two sampling methods can be more referential compared with traditional method. For the cladding corrosion and strain criterion, the results of sampling methods and traditional method show no significant difference, for the reason that the uncertain input parameters are selected suitably. In conclusion, the statistical method based on nonparametric sampling can be more practically significant for safety performance evaluation of fuel rod in operation.
Fuel rod design verification is the evaluation process of fuel rod safety performance during operation in reactor, in which the uncertainty of input parameters has important effect on evaluation results. To study the uncertainty systematically, fuel rod performance analysis software has been coupled with Dakota software to carry out fuel rod design verification, the results of nonparametric Monte Carlo and Latin Hypercube Sampling have been compared with those of traditional method. It turns out that the fuel rod inner pressure criterion is vulnerable to be under challenge for the reason of input uncertainty under consideration by traditional method. The defects can be made up by statistical nonparametric sampling, by which a larger safety margin is obtained, and a theoretical basis for fuel rod safety and economic performance enhancement is provided. Meanwhile, the temperature calculation result obtained by two sampling methods can be more referential compared with traditional method. For the cladding corrosion and strain criterion, the results of sampling methods and traditional method show no significant difference, for the reason that the uncertain input parameters are selected suitably. In conclusion, the statistical method based on nonparametric sampling can be more practically significant for safety performance evaluation of fuel rod in operation.
2022,
34: 026013.
doi: 10.11884/HPLPB202234.210301
Abstract:
The dispersed fuel element is often used in small integrated PWR. The fuel core is composed of a large number of small fuel microspheres mixed in the metal matrix, and the volume heterogeneity exists between the fuel particles and the matrix. If the local space self-screen effect of the dispersion fuel can not be effectively treated, it will bring some deviation to the calculation of physical property parameters. The traditional deterministic method neglects the double-heterogeneity of dispersive fuel. In this paper, the explicit modelling and the random medium program based on SuperMC are used to establish the volumetric homogenization model and grain model to verify double-heterogeneity. The traditional reactivity-equivalent physical transformation (RPT) model of the dispersive plate-fuel was established to correct the calculation deviation. The results show that the random media program combined with SuperMC can simulate the calculation of transport and burnup of various types of particles, and can well analyze and solve the double-heterogeneity problem.
The dispersed fuel element is often used in small integrated PWR. The fuel core is composed of a large number of small fuel microspheres mixed in the metal matrix, and the volume heterogeneity exists between the fuel particles and the matrix. If the local space self-screen effect of the dispersion fuel can not be effectively treated, it will bring some deviation to the calculation of physical property parameters. The traditional deterministic method neglects the double-heterogeneity of dispersive fuel. In this paper, the explicit modelling and the random medium program based on SuperMC are used to establish the volumetric homogenization model and grain model to verify double-heterogeneity. The traditional reactivity-equivalent physical transformation (RPT) model of the dispersive plate-fuel was established to correct the calculation deviation. The results show that the random media program combined with SuperMC can simulate the calculation of transport and burnup of various types of particles, and can well analyze and solve the double-heterogeneity problem.
2022,
34: 026014.
doi: 10.11884/HPLPB202234.210312
Abstract:
In a pressurized water reactor nuclear power plant, the compression force of the fuel assembly, coolant flow, radiation creep, burnup and other factors will cause the bowing of the fuel assembly. The bowing of the fuel assembly affects the distribution of the water gap between the assemblies. It affects the slowing behavior of neutrons and the heat transfer performance of the core, which in turn affects the operating parameters of the reactor core. This paper discusses the cause and mechanism, influence and consequences of assembly bowing (including the influence on core power distribution, core radial power tilt, nuclear enthalpy rise hot channel factor, heat flux hot channel factor and other parameters), and uses Monte Carlo software JMCT to verify the correctness of the calculation program of component bowing PCM. Finally, through the deterministic calculation program, a simulation analysis of the assemblies bowing of the CPR1000 nuclear power plant is carried out. The calculation results show that: at certain fuel consumption, as the water gap increases or decreases, the fuel assembly power will increase or decrease, the power distribution of the core will tilt, affecting the safe operation of nuclear power plants.
In a pressurized water reactor nuclear power plant, the compression force of the fuel assembly, coolant flow, radiation creep, burnup and other factors will cause the bowing of the fuel assembly. The bowing of the fuel assembly affects the distribution of the water gap between the assemblies. It affects the slowing behavior of neutrons and the heat transfer performance of the core, which in turn affects the operating parameters of the reactor core. This paper discusses the cause and mechanism, influence and consequences of assembly bowing (including the influence on core power distribution, core radial power tilt, nuclear enthalpy rise hot channel factor, heat flux hot channel factor and other parameters), and uses Monte Carlo software JMCT to verify the correctness of the calculation program of component bowing PCM. Finally, through the deterministic calculation program, a simulation analysis of the assemblies bowing of the CPR1000 nuclear power plant is carried out. The calculation results show that: at certain fuel consumption, as the water gap increases or decreases, the fuel assembly power will increase or decrease, the power distribution of the core will tilt, affecting the safe operation of nuclear power plants.
2022,
34: 026015.
doi: 10.11884/HPLPB202234.210350
Abstract:
Active neutron multiplicity counting method could provide the multiplicity distribution of neutrons multiplied from one fission event induced by an external neutron source in nuclear components. The equivalent mass of material in uranium component can be obtained by substituting the neutron multiplicity counting into the mass inversion equation with the known characteristic parameters. Therefore, this method is widely used in nuclear material accounting, nuclear security measurement and arms control verification. In this paper, Mont Carlo neutron transport calculation was used to simulate the experiment of the uranium mass estimation by classical point model equation. Based on this, an improved equation is proposed, which could significantly reduce the effect of source-sample coupling and neutron back irradiation on the accuracy of uranium mass estimation for the experiment. The validity of the mass equation is tested by numerical method, and the deviation between real mass and estimated mass obtained by the original point model equation and the improved mass equation is compared. The active neutron multiplicity counting simulation calculation is realized based on the original neutron transport simulation software. A model of active neutron multiplicity counting detection system and 32 models of uranium metal hemispherical shell are established. The active neutron multiplicity counts of DT source and AmLi source with different distances from uranium components are simulated and the estimated mass of the object are obtained with the simulated results. The numerical simulation results show that the average deviation of the improved uranium component mass inversion is reduced to less than 10%.
Active neutron multiplicity counting method could provide the multiplicity distribution of neutrons multiplied from one fission event induced by an external neutron source in nuclear components. The equivalent mass of material in uranium component can be obtained by substituting the neutron multiplicity counting into the mass inversion equation with the known characteristic parameters. Therefore, this method is widely used in nuclear material accounting, nuclear security measurement and arms control verification. In this paper, Mont Carlo neutron transport calculation was used to simulate the experiment of the uranium mass estimation by classical point model equation. Based on this, an improved equation is proposed, which could significantly reduce the effect of source-sample coupling and neutron back irradiation on the accuracy of uranium mass estimation for the experiment. The validity of the mass equation is tested by numerical method, and the deviation between real mass and estimated mass obtained by the original point model equation and the improved mass equation is compared. The active neutron multiplicity counting simulation calculation is realized based on the original neutron transport simulation software. A model of active neutron multiplicity counting detection system and 32 models of uranium metal hemispherical shell are established. The active neutron multiplicity counts of DT source and AmLi source with different distances from uranium components are simulated and the estimated mass of the object are obtained with the simulated results. The numerical simulation results show that the average deviation of the improved uranium component mass inversion is reduced to less than 10%.
2022,
34: 026016.
doi: 10.11884/HPLPB202234.210356
Abstract:
The numerical simulation program of neutron activation products and their radiation characteristics is an important tool to study the activation effect of materials. A numerical simulation program capable of simulating the neutron activation effect of materials was developed through the secondary development of the JMCT software, and it was named “neutron activation simulation program”, which is expected to be applied to the research of arms control verification and nuclear security. The correctness of the neutron activation simulation program in calculating the neutron transport and activation inside nuclear warheads was tested, and the calculation results are accurate. The activation effect of concrete floors under the irradiation of the fission neutrons from fissile materials was studied based on the neutron activation simulation program, and the calculation results furthermore prove the functions of the neutron activation simulation program.
The numerical simulation program of neutron activation products and their radiation characteristics is an important tool to study the activation effect of materials. A numerical simulation program capable of simulating the neutron activation effect of materials was developed through the secondary development of the JMCT software, and it was named “neutron activation simulation program”, which is expected to be applied to the research of arms control verification and nuclear security. The correctness of the neutron activation simulation program in calculating the neutron transport and activation inside nuclear warheads was tested, and the calculation results are accurate. The activation effect of concrete floors under the irradiation of the fission neutrons from fissile materials was studied based on the neutron activation simulation program, and the calculation results furthermore prove the functions of the neutron activation simulation program.
2022,
34: 026017.
doi: 10.11884/HPLPB202234.210362
Abstract:
The inherent safety characteristic of high temperature gas-cooled reactor (HTGR) is universally acknowledged. On account of the double-heterogeneity of the coated fuel and the irregular core pattern of the prismatic HTGR, a code system based on Monte-Carlo homogenization and deterministic transport concept was used as a nuclear design code system for prismatic HTGR. An investigation on homogenization model and preliminary validation of this code system is conducted based on VHTRC benchmark which is a prismatic HTGR criticality assembly. The validation results indicate that group constants generation based on Monte-Carlo homogenization and appropriate energy group, then modified by super homogenization (SPH) method further, ensures high precision of homogeneous core calculation.
The inherent safety characteristic of high temperature gas-cooled reactor (HTGR) is universally acknowledged. On account of the double-heterogeneity of the coated fuel and the irregular core pattern of the prismatic HTGR, a code system based on Monte-Carlo homogenization and deterministic transport concept was used as a nuclear design code system for prismatic HTGR. An investigation on homogenization model and preliminary validation of this code system is conducted based on VHTRC benchmark which is a prismatic HTGR criticality assembly. The validation results indicate that group constants generation based on Monte-Carlo homogenization and appropriate energy group, then modified by super homogenization (SPH) method further, ensures high precision of homogeneous core calculation.
2022,
34: 026018.
doi: 10.11884/HPLPB202234.210364
Abstract:
With the support of the modelling conversion platform cosVMPT, the 360° model is needed to solve this inaccuracy from asymmetry. The detailed structure of all main components for China Fusion Engineering Test Reactor (CFETR) has been developed. The whole model consists of Central Solenoid (CS), Vacuum Vessel (VV), Port (16 upper/lower ports, 6 equatorial ports in which 2 oblique are ones used for NBI beam channel), Thermal Shielding (VVTS & CTS), Toroidal and Poloidal Field Coils (TFC & PFC), Cryostat and the House Building. The water-cooled ceramic breeder (WCCB) blanket and water cooled divertor with full detailed structure is inserted, while spare ports are filled with shielding material. The preliminary calculation has been performed by MCNP code combining with the advanced “on-the-fly” GVR method, which was introduced to generate the global weight window to accelerate the neutron transport, and the reliable neutron and photon flux map was obtained. This work realizes the application of cosVMPT platform and OTF method in complicated fusion reactor, and it validates their robustness. In addition, the results of 360° model has been compared to that of sector model, to verify the validity and applicability of using sector model to simplify the modelling and calculation.
With the support of the modelling conversion platform cosVMPT, the 360° model is needed to solve this inaccuracy from asymmetry. The detailed structure of all main components for China Fusion Engineering Test Reactor (CFETR) has been developed. The whole model consists of Central Solenoid (CS), Vacuum Vessel (VV), Port (16 upper/lower ports, 6 equatorial ports in which 2 oblique are ones used for NBI beam channel), Thermal Shielding (VVTS & CTS), Toroidal and Poloidal Field Coils (TFC & PFC), Cryostat and the House Building. The water-cooled ceramic breeder (WCCB) blanket and water cooled divertor with full detailed structure is inserted, while spare ports are filled with shielding material. The preliminary calculation has been performed by MCNP code combining with the advanced “on-the-fly” GVR method, which was introduced to generate the global weight window to accelerate the neutron transport, and the reliable neutron and photon flux map was obtained. This work realizes the application of cosVMPT platform and OTF method in complicated fusion reactor, and it validates their robustness. In addition, the results of 360° model has been compared to that of sector model, to verify the validity and applicability of using sector model to simplify the modelling and calculation.
2022,
34: 026019.
doi: 10.11884/HPLPB202234.210388
Abstract:
Heat pipe cooled nuclear reactor has unique advantages, such as passive heat transfer technology, modular design and inherent safety. It has been extensively used in deep space, deep sea and decentralized electricity markets. Recently, Los Alamos National Laboratory designed a 5 MWth heat pipe cooled nuclear reactor for decentralized electricity markets. This paper aimed to provide an appropriate reference for the choice of types of monolithic core in the engineering design of heat pipe cooled nuclear reactor. With the help of a Monte Carlo simulation program RMC, reactivity, neutron spectrum, breeding performance and burnup evolution were explored on the heat pipe cooled nuclear reactor with monolith materials SS-316, SiC and Mo-14Re, respectively. The results show the reactor with monolith materials SS-316, Mo-14Re and SiC can achieve a 10-year operation when the enrichment of the initially loaded 235U is about 19.35%, 28.80% and 17.10%, respectively. The SiC monolith core requires the least initial 235U enrichment. Besides, the core with SiC monolith can produce the highest mass of fissile 239Pu and 241Pu (about 11.91 kg) and minor actinides (about 92.08 g) for a 10-year operation. Thus, SiC monolith is recommended for the pipe cooled nuclear reactor.
Heat pipe cooled nuclear reactor has unique advantages, such as passive heat transfer technology, modular design and inherent safety. It has been extensively used in deep space, deep sea and decentralized electricity markets. Recently, Los Alamos National Laboratory designed a 5 MWth heat pipe cooled nuclear reactor for decentralized electricity markets. This paper aimed to provide an appropriate reference for the choice of types of monolithic core in the engineering design of heat pipe cooled nuclear reactor. With the help of a Monte Carlo simulation program RMC, reactivity, neutron spectrum, breeding performance and burnup evolution were explored on the heat pipe cooled nuclear reactor with monolith materials SS-316, SiC and Mo-14Re, respectively. The results show the reactor with monolith materials SS-316, Mo-14Re and SiC can achieve a 10-year operation when the enrichment of the initially loaded 235U is about 19.35%, 28.80% and 17.10%, respectively. The SiC monolith core requires the least initial 235U enrichment. Besides, the core with SiC monolith can produce the highest mass of fissile 239Pu and 241Pu (about 11.91 kg) and minor actinides (about 92.08 g) for a 10-year operation. Thus, SiC monolith is recommended for the pipe cooled nuclear reactor.
2022,
34: 021001.
doi: 10.11884/HPLPB202234.220012
Abstract:
Diode pumped metastable rare gas laser (DPRGL) is a potential candidate for single aperture scaled extremely high energy lasers. In the current studies, the metastable atoms were generated in confined discharge gaps, which limited the high power output. In this paper, a DPRGL in a plasma jet is firstly proposed and demonstrated. A high concentration of metastable argon atoms (about 1014 cm−3) is produced in the plume of an atmospheric plasma jet, which separates the discharge and gain regions. By using a 811 nm linewidth narrowed diode laser as pumping source, based on a “pump, laser and flow” mutually orthogonal configuration, the system realizes 912 nm laser output. The plasma jet type DPRGLs have the potential to break through the limitations of the existing schemes, and significantly improve the power scaling ability.
2022,
34: 021002.
doi: 10.11884/HPLPB202234.210251
Abstract:
Due to themerits of high quality, high efficiency, high robustness and compact size, the fiber laser systems have developed rapidly in the past 20 years, and have been widely used. However, there are still some factors (such as nonlinear effect, thermal effect, transverse mode instability (TMI), etc.) that limit the scaling of output power of fiber lasers. As one of the main limitations, Stimulated Raman Scattering (SRS) effect not only reduces the fiber lasers’ efficiency, but also increases the risk of device damage with the backward-Stokes light. In addition, recent studies have shown that SRS effect in few-mode fiber lasers can result in TMI and another type of mode distortion, which is on a much slower time scale. Thereby, from the aspect of fiber design and laser systems, amounts of mitigation strategies have been carry out to suppress this nonlinear effect. In this paper, besides the concise description of SRS-induced mode distortion, the suppression strategies in the past decade are also summarized from the aspect of the fiber laser system optimization.
2022,
34: 022001.
doi: 10.11884/HPLPB202234.210382
Abstract:
The reliability of optical streak camera in the environment of extreme radiation and electromagnetic field has been investigated on “Chenguang” Accelerator. The image of a series of pulse laser was used to indicate the scanning status of the streak camera. The influence of different manners of radiant and electromagnetic shield has been explored. At last, the conditions of this kind of streak camera working normally in these extreme environments were gained in this experiment: enhancing radiation shield to reduce the equivalent air absorbed dose of single shot near the streak camera to 6 μGy, reducing the outer covering gaps of the streak camera, using photoelectric trigger and so on.
The reliability of optical streak camera in the environment of extreme radiation and electromagnetic field has been investigated on “Chenguang” Accelerator. The image of a series of pulse laser was used to indicate the scanning status of the streak camera. The influence of different manners of radiant and electromagnetic shield has been explored. At last, the conditions of this kind of streak camera working normally in these extreme environments were gained in this experiment: enhancing radiation shield to reduce the equivalent air absorbed dose of single shot near the streak camera to 6 μGy, reducing the outer covering gaps of the streak camera, using photoelectric trigger and so on.
2022,
34: 022002.
doi: 10.11884/HPLPB202234.210339
Abstract:
The interaction between intense laser and plasma can produce various parametric instability processes and nonlinear effects. In this paper, the nonlinear equation satisfied by the transverse field envelope is derived by Karpman method, and the solitary wave solution is obtained in one-dimensional case. It is found that the half width of the solitary wave is inversely proportional to the amplitude of the laser beam. The influence of various parameters in magnetized plasma on the half width of the solitary wave is analyzed. The results show that for the case of right-hand circularly polarized laser, the half width of soliton decreases with the increase of electron number density, and increases with the increase of magnetic field intensity. In the case of left-hand circularly polarized laser, the half width of soliton increases with the increase of electron number density, and decreases with the increase of magnetic field intensity.
The interaction between intense laser and plasma can produce various parametric instability processes and nonlinear effects. In this paper, the nonlinear equation satisfied by the transverse field envelope is derived by Karpman method, and the solitary wave solution is obtained in one-dimensional case. It is found that the half width of the solitary wave is inversely proportional to the amplitude of the laser beam. The influence of various parameters in magnetized plasma on the half width of the solitary wave is analyzed. The results show that for the case of right-hand circularly polarized laser, the half width of soliton decreases with the increase of electron number density, and increases with the increase of magnetic field intensity. In the case of left-hand circularly polarized laser, the half width of soliton increases with the increase of electron number density, and decreases with the increase of magnetic field intensity.
2022,
34: 023001.
doi: 10.11884/HPLPB202234.210420
Abstract:
To reduce the weight of the radome and meet the high-power application requirement, a high-power built-in curved meander-line polarization conversion radome is proposed. The radome has the curved meander-line polarization conversion plate inside the dielectric sealing cover, and through joint optimization of the arc-shaped broken wire grid unit structure and the dielectric sealing cover structure, the weight is reduced while the electric field on the metal grid is reduced, which is more conducive to engineering applications in the high-power field. In response to the application requirements of the C-band high-power linear polarization helical array antenna, a high-power polarization conversion radome with a center frequency of 4.3 GHz was optimized and loaded on the high-power helical array antenna. And the radiation characteristics and power capacity were tested. The simulation and experimental results are consistent. The radome can convert circularly polarized waves to linearly polarized waves. Its insertion loss is 0.2 dB, its axial ratio is 20 dB, and its power capacity reaches 48 MW.
To reduce the weight of the radome and meet the high-power application requirement, a high-power built-in curved meander-line polarization conversion radome is proposed. The radome has the curved meander-line polarization conversion plate inside the dielectric sealing cover, and through joint optimization of the arc-shaped broken wire grid unit structure and the dielectric sealing cover structure, the weight is reduced while the electric field on the metal grid is reduced, which is more conducive to engineering applications in the high-power field. In response to the application requirements of the C-band high-power linear polarization helical array antenna, a high-power polarization conversion radome with a center frequency of 4.3 GHz was optimized and loaded on the high-power helical array antenna. And the radiation characteristics and power capacity were tested. The simulation and experimental results are consistent. The radome can convert circularly polarized waves to linearly polarized waves. Its insertion loss is 0.2 dB, its axial ratio is 20 dB, and its power capacity reaches 48 MW.
2022,
34: 023002.
doi: 10.11884/HPLPB202234.210406
Abstract:
The composition and waveform of the calibration platform were defined in GJB151B CS115 for pulsed conducted sensitivity test, but the specific requirements for each device (pulse generator, current injection probe, calibration clamp, etc.) in the calibration platform are still not clear. To solve this problem, a time domain circuit model of the calibration platform was developed according the previous research on circuit simulation of pulsed current injection. By adjusting the parameters respectively, influences on the rise time and fall time of the calibration waveform caused by the inner inductance of the pulse generator and equivalent elements of the injection probe are simulated. Finally, the technical indicators for each device in the platform are obtained. This work is an essential supplement to GJB151B CS115 and is helpful to set up CS115 test platform for conducted immunity of electronic equipment test.
The composition and waveform of the calibration platform were defined in GJB151B CS115 for pulsed conducted sensitivity test, but the specific requirements for each device (pulse generator, current injection probe, calibration clamp, etc.) in the calibration platform are still not clear. To solve this problem, a time domain circuit model of the calibration platform was developed according the previous research on circuit simulation of pulsed current injection. By adjusting the parameters respectively, influences on the rise time and fall time of the calibration waveform caused by the inner inductance of the pulse generator and equivalent elements of the injection probe are simulated. Finally, the technical indicators for each device in the platform are obtained. This work is an essential supplement to GJB151B CS115 and is helpful to set up CS115 test platform for conducted immunity of electronic equipment test.
2022,
34: 024001.
doi: 10.11884/HPLPB202234.210392
Abstract:
CT tube is the core component of CT equipment. CT tube adopts dynamic magnetic focusing technology to facilitate the realization of small focus when there is large current output. Flying focal spot technology can record each projection of each scan from multiple angles, improve the sampling rate, and greatly improve the image definition and quality. CST software is used to simulate the CT tube based on dynamic magnetic focusing. When the CT tube outputs high current of 1 A, the key parameters of the dual magnetic quadrupole lens are optimized to get the target size of focus of 0.2 mm×0.6 mm and the function of flying focus.
CT tube is the core component of CT equipment. CT tube adopts dynamic magnetic focusing technology to facilitate the realization of small focus when there is large current output. Flying focal spot technology can record each projection of each scan from multiple angles, improve the sampling rate, and greatly improve the image definition and quality. CST software is used to simulate the CT tube based on dynamic magnetic focusing. When the CT tube outputs high current of 1 A, the key parameters of the dual magnetic quadrupole lens are optimized to get the target size of focus of 0.2 mm×0.6 mm and the function of flying focus.
2022,
34: 024002.
doi: 10.11884/HPLPB202234.210439
Abstract:
In particle accelerators, the impedance of vacuum components is an important cause of beam instabilities. Smaller emittance is required in the new generation storage-ring-based light source, where smaller vacuum chamber is needed and the impedances are increased correspondingly.Therefore, the impedance of the vacuum components should be evaluated accurately and well optimized during the design stage. Impedance measurement is an important way to check the impedance model. The coaxial wire method is commonly used for laboratory bending measurements. For small-aperture vacuum components, the measurement error caused by the inner conductor is studied in this paper. For narrow-band impedances, pillbox cavity is used to study the influence of different inner conductor radius on the measurement results. As benchmarks, different numerical simulations, such as wake field, scattering parameters and eigenmodes, are carried out. The results show that the simulations are in good agreement with the measurement. It is also found that there exist deviations between the measurement and the existing theories on the dependence of the resonant frequency shift on the inner conductor radius. For non-narrowband impedances, we measured the impedance of a strip line kicker and studied the influence of the inner conductor, which verified the validity of the coaxial wire method.
In particle accelerators, the impedance of vacuum components is an important cause of beam instabilities. Smaller emittance is required in the new generation storage-ring-based light source, where smaller vacuum chamber is needed and the impedances are increased correspondingly.Therefore, the impedance of the vacuum components should be evaluated accurately and well optimized during the design stage. Impedance measurement is an important way to check the impedance model. The coaxial wire method is commonly used for laboratory bending measurements. For small-aperture vacuum components, the measurement error caused by the inner conductor is studied in this paper. For narrow-band impedances, pillbox cavity is used to study the influence of different inner conductor radius on the measurement results. As benchmarks, different numerical simulations, such as wake field, scattering parameters and eigenmodes, are carried out. The results show that the simulations are in good agreement with the measurement. It is also found that there exist deviations between the measurement and the existing theories on the dependence of the resonant frequency shift on the inner conductor radius. For non-narrowband impedances, we measured the impedance of a strip line kicker and studied the influence of the inner conductor, which verified the validity of the coaxial wire method.
