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RSS2017 Vol. 29, No. 03
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2017,
29: 036001.
doi: 10.11884/HPLPB201729.160297
Abstract:
Transient calculation is one of the key parts in reactor safety analysis and system simulation. With the fast development of computer technology, now 3D spatial dependent kinetics calculation is feasible and has already been used in the system simulator. This paper reports a new transient nodal code based on the widely used non-linear iteration scheme. The new code applies the two-group CMFD and multi-group unified nodal method (UNM). The UNM nodal method applies equations of analytical nodal method (ANM) but eliminates the instability of ANM by mathematical transformation. It is highly accurate and also has good efficiency. In the thermal-hydraulic module, the 1D heat conduction equation is solved instead of the currently well used formulation form to give better accuracy. The three-nodes method based on the transverse leakage equation is applied to reduce the impact of control rod cusping on transient calculations. To test the performance of the new code, the Nuclear Energy Agency Committee on Reactor Physics(NEACRP) benchmark is calculated. Numerical results show that its accuracy is similar to that of other high performance nodal codes.
Transient calculation is one of the key parts in reactor safety analysis and system simulation. With the fast development of computer technology, now 3D spatial dependent kinetics calculation is feasible and has already been used in the system simulator. This paper reports a new transient nodal code based on the widely used non-linear iteration scheme. The new code applies the two-group CMFD and multi-group unified nodal method (UNM). The UNM nodal method applies equations of analytical nodal method (ANM) but eliminates the instability of ANM by mathematical transformation. It is highly accurate and also has good efficiency. In the thermal-hydraulic module, the 1D heat conduction equation is solved instead of the currently well used formulation form to give better accuracy. The three-nodes method based on the transverse leakage equation is applied to reduce the impact of control rod cusping on transient calculations. To test the performance of the new code, the Nuclear Energy Agency Committee on Reactor Physics(NEACRP) benchmark is calculated. Numerical results show that its accuracy is similar to that of other high performance nodal codes.
2017,
29: 036002.
doi: 10.11884/HPLPB201729.160302
Abstract:
The depletion calculation in reactor physics requires depletion chain data. However, the depletion chain that originates from evaluated nuclear data library is unnecessarily detailed for assembly and micro-depletion calculation, while conventional depletion chain compression methods are semi-empirical with limited application scope and accuracy. In this paper, a method that compresses depletion chain based on quantitative significance analysis of each nuclide and reaction channel is proposed. The significance analysis uses fine depletion chain computation results of representative problems as data source, and it is carried out by evaluating the influence on neutron absorption, production and number densities of target nuclides, induced by each basic unit compression. The method is applied to depletion chain compression in the context of PWR assembly calculation. Fine and compressed depletion chain computations of selected test cases are performed. The comparison among obtained numerical results show that, while preserving accuracy requirements, the proposed depletion chain compression method is capable of significantly reducing the complexity of the depletion chain, and the demanded storage and time savings are achieved consequently.
The depletion calculation in reactor physics requires depletion chain data. However, the depletion chain that originates from evaluated nuclear data library is unnecessarily detailed for assembly and micro-depletion calculation, while conventional depletion chain compression methods are semi-empirical with limited application scope and accuracy. In this paper, a method that compresses depletion chain based on quantitative significance analysis of each nuclide and reaction channel is proposed. The significance analysis uses fine depletion chain computation results of representative problems as data source, and it is carried out by evaluating the influence on neutron absorption, production and number densities of target nuclides, induced by each basic unit compression. The method is applied to depletion chain compression in the context of PWR assembly calculation. Fine and compressed depletion chain computations of selected test cases are performed. The comparison among obtained numerical results show that, while preserving accuracy requirements, the proposed depletion chain compression method is capable of significantly reducing the complexity of the depletion chain, and the demanded storage and time savings are achieved consequently.
2017,
29: 036003.
doi: 10.11884/HPLPB201729.160286
Abstract:
BEAVRS is an international benchmark model based on the commercial PWR reactor with full-core meticulous modeling from MIT, and it has detailed measurement data. BEAVRS is a large 3D reactor benchmark for verification consisted of a variety of enrichment fuels and control rod assemblies. In this paper, cosRMC(a Monte Carlo code independently developed by China) is adopted in the detailed modeling of BEAVRS benchmark to calculate the critical eigenvalues, full reactor power distribution and the control rod worth with different control rod assenmblies inserted in the hot zero power(HZP) status. The results are compared with those by the internationally renowned programs MCNP, OpenMC and MC21. In the HZP status, the critical eigenvalue error between cosRMC and MCNP is just 7.1 pcm; the eigenvalues with different control rod assemblies inserted are calculated and the error is less than 0.74% compared with the theoretical error 1.000, and the difference between the calculated control rod worth and the measured value is less than 100 pcm, and the calculation accuracy agrees well with that of similar software; also the calculated full reactor power distribution is compared with the measured value and the reasons of error are analysed. The feasibility and accuracy of cosRMC in the complex core meticulous modeling are verified preliminarily, which lays a foundation for future applications and improvement.
BEAVRS is an international benchmark model based on the commercial PWR reactor with full-core meticulous modeling from MIT, and it has detailed measurement data. BEAVRS is a large 3D reactor benchmark for verification consisted of a variety of enrichment fuels and control rod assemblies. In this paper, cosRMC(a Monte Carlo code independently developed by China) is adopted in the detailed modeling of BEAVRS benchmark to calculate the critical eigenvalues, full reactor power distribution and the control rod worth with different control rod assenmblies inserted in the hot zero power(HZP) status. The results are compared with those by the internationally renowned programs MCNP, OpenMC and MC21. In the HZP status, the critical eigenvalue error between cosRMC and MCNP is just 7.1 pcm; the eigenvalues with different control rod assemblies inserted are calculated and the error is less than 0.74% compared with the theoretical error 1.000, and the difference between the calculated control rod worth and the measured value is less than 100 pcm, and the calculation accuracy agrees well with that of similar software; also the calculated full reactor power distribution is compared with the measured value and the reasons of error are analysed. The feasibility and accuracy of cosRMC in the complex core meticulous modeling are verified preliminarily, which lays a foundation for future applications and improvement.
2017,
29: 036004.
doi: 10.11884/HPLPB201729.160433
Abstract:
Nuclear power software needs to give the uncertainty of calculation results, thus it can effectively evaluate the safety of nuclear power. Nuclear data are important sources of uncertainty in core calculation, and the sensitivity analysis is an important step for the uncertainty analysis of core parameter. The sensitivity analysis of nuclear data is carried out on the critical calculation in Qinshan Ⅱ core. Firstly, Monte Carlo procedures are used for the establishment of Qinshan Ⅱs first cycle model, then the sensitivity coefficient of nuclear data is calculated with Iterated Fission Probability method(IFP) based on continuous energy cross section library CENACE V1.0 developed by Chinese nuclear data center. By the analysis of the sensitivity coefficient with different temperature and different burnup, the important nuclear data to be concerned in the uncertainty analysis of nuclear power software would be found.
Nuclear power software needs to give the uncertainty of calculation results, thus it can effectively evaluate the safety of nuclear power. Nuclear data are important sources of uncertainty in core calculation, and the sensitivity analysis is an important step for the uncertainty analysis of core parameter. The sensitivity analysis of nuclear data is carried out on the critical calculation in Qinshan Ⅱ core. Firstly, Monte Carlo procedures are used for the establishment of Qinshan Ⅱs first cycle model, then the sensitivity coefficient of nuclear data is calculated with Iterated Fission Probability method(IFP) based on continuous energy cross section library CENACE V1.0 developed by Chinese nuclear data center. By the analysis of the sensitivity coefficient with different temperature and different burnup, the important nuclear data to be concerned in the uncertainty analysis of nuclear power software would be found.
2017,
29: 036005.
doi: 10.11884/HPLPB201729.160339
Abstract:
The traveling wave reactor (TWR) is an innovative nuclear system, whose discharge burnup is 400 GWd/tHM, about three to four times of existing fast reactor and six to eight times of pressurized water reactor. High discharge burnup forces the present codes to face the great challenges in precision.This paper studies the high burnup calculation characteristics of traveling wave reactor from energy spectrum, importance of fission production and calculation error accumulation of fuel consumption by KYLIN-1 code. The analysis results of typical traveling wave reactor hexagonal assembly show that the low enriched uranium assemblies have different energy spectra at beginning and end of life, which cause big calculation error ofkinf by existing code system. To ensure the correctness of the traveling wave reactor burnup calculations, the burn chain should contain 70 kinds of important fission product nuclides. The calculation error accumulation of fuel consumption is small when the burnup step is increased. The calculation error ofkinf is about 0.001% each burnup step.
The traveling wave reactor (TWR) is an innovative nuclear system, whose discharge burnup is 400 GWd/tHM, about three to four times of existing fast reactor and six to eight times of pressurized water reactor. High discharge burnup forces the present codes to face the great challenges in precision.This paper studies the high burnup calculation characteristics of traveling wave reactor from energy spectrum, importance of fission production and calculation error accumulation of fuel consumption by KYLIN-1 code. The analysis results of typical traveling wave reactor hexagonal assembly show that the low enriched uranium assemblies have different energy spectra at beginning and end of life, which cause big calculation error ofkinf by existing code system. To ensure the correctness of the traveling wave reactor burnup calculations, the burn chain should contain 70 kinds of important fission product nuclides. The calculation error accumulation of fuel consumption is small when the burnup step is increased. The calculation error ofkinf is about 0.001% each burnup step.
2017,
29: 036006.
doi: 10.11884/HPLPB201729.160357
Abstract:
Large equipment models need tremendous amount of three-dimensional solid entities to build, when using CSG method. For example, Dayawan reactor whole core model has tens of thousands of solid entities. A large-scale visual modeling software JLAMT for Monte-Carlo particle transport calculation is developed by the Institute of Applied Physics and Computational Mathematics (IAPCM). In this paper, for the particle transport program oriented visual modeling, a program module for hierarchical modeling and management is developed and plugged in JLAMT, which allows users to separate a complex model into different layers and edit part of the model in a certain layer. Therefore, it improves the performance and efficiency of three-dimensional modeling of complex models for the Monte Carlo particle transport calculation.
Large equipment models need tremendous amount of three-dimensional solid entities to build, when using CSG method. For example, Dayawan reactor whole core model has tens of thousands of solid entities. A large-scale visual modeling software JLAMT for Monte-Carlo particle transport calculation is developed by the Institute of Applied Physics and Computational Mathematics (IAPCM). In this paper, for the particle transport program oriented visual modeling, a program module for hierarchical modeling and management is developed and plugged in JLAMT, which allows users to separate a complex model into different layers and edit part of the model in a certain layer. Therefore, it improves the performance and efficiency of three-dimensional modeling of complex models for the Monte Carlo particle transport calculation.
2017,
29: 036007.
doi: 10.11884/HPLPB201729.160383
Abstract:
Coupled neutronics and thermal-hydraulics calculation can enhance the computational accuracy in reactor simulation. With the development of computational capability, the coupled Monte-Carlo neutron transportation and thermal-hydraulics calculation plays an important role in reactor core design and reactor safety. This paper gives a short introduction of the Monte Carlo code JMCT, the sub-channel code COBRA-EN and their coupling scheme. The coupled codes are applied on the 33 fuel bundles model and the UO2 assembly simulation. The calculated results are discussed, especially, the convergence criteria and quantification of correlation between statistical uncertainty and coupled error.
Coupled neutronics and thermal-hydraulics calculation can enhance the computational accuracy in reactor simulation. With the development of computational capability, the coupled Monte-Carlo neutron transportation and thermal-hydraulics calculation plays an important role in reactor core design and reactor safety. This paper gives a short introduction of the Monte Carlo code JMCT, the sub-channel code COBRA-EN and their coupling scheme. The coupled codes are applied on the 33 fuel bundles model and the UO2 assembly simulation. The calculated results are discussed, especially, the convergence criteria and quantification of correlation between statistical uncertainty and coupled error.
2017,
29: 036008.
doi: 10.11884/HPLPB201729.160179
Abstract:
The reactor pressure vessel (RPV) neutron flux calculation method for MOX fuel core was preliminarily researched, especially the core source calculation method for MOX fuel core. Based on three-dimensional discrete ordinate code TORT, the RPV fast neutron flux was calculated for CAP1400 reaction core carrying 50% MOX fuel, and the difference between MOX fuel core and UO2 fuel core was compared and analyzed. The results demonstrate that CAP1400 reaction core carrying 50% MOX fuel can satisfy the request of the radiation safety of RPV. For radiation shielding optimization, the suggestion is that much attention should be paid to the core loading pattern selection for the pivotal position.
The reactor pressure vessel (RPV) neutron flux calculation method for MOX fuel core was preliminarily researched, especially the core source calculation method for MOX fuel core. Based on three-dimensional discrete ordinate code TORT, the RPV fast neutron flux was calculated for CAP1400 reaction core carrying 50% MOX fuel, and the difference between MOX fuel core and UO2 fuel core was compared and analyzed. The results demonstrate that CAP1400 reaction core carrying 50% MOX fuel can satisfy the request of the radiation safety of RPV. For radiation shielding optimization, the suggestion is that much attention should be paid to the core loading pattern selection for the pivotal position.
2017,
29: 036009.
doi: 10.11884/HPLPB201729.160294
Abstract:
Zero power physics test (ZPPT) is implanted after the nuclear reactor approach to criticality to make sure the core performance is in compliance with the nuclear design. However, traditional equipment used in ZPPT is bulk in volume and the measurement limit is narrow and the precision is low to adapt new test method. In the Physics Start-up Analysis System(PSAS) developed by China Nuclear Power Research Institute, the reactivity measurement method, software and hardware design, low-level current measurement range alternative, data process and transmission are optimized to enhance the measurement capability and adaptability. The test results on research reactors and Yangjiang Nuclear Power Plants unit 3 verify that PSAS is suitable for the pressurized water reactor physics test.
Zero power physics test (ZPPT) is implanted after the nuclear reactor approach to criticality to make sure the core performance is in compliance with the nuclear design. However, traditional equipment used in ZPPT is bulk in volume and the measurement limit is narrow and the precision is low to adapt new test method. In the Physics Start-up Analysis System(PSAS) developed by China Nuclear Power Research Institute, the reactivity measurement method, software and hardware design, low-level current measurement range alternative, data process and transmission are optimized to enhance the measurement capability and adaptability. The test results on research reactors and Yangjiang Nuclear Power Plants unit 3 verify that PSAS is suitable for the pressurized water reactor physics test.
2017,
29: 036010.
doi: 10.11884/HPLPB201729.160376
Abstract:
Fuel loading pattern optimization with MOX (mixed oxide of plutonium and uranium) fuel is an important aspect for GIII LWR. For the HPR1000 core, an optimized MOX fuel is designed. Based on which, two loading patterns with 50% MOX fuel are designed. One is 1/4 core annually reloading pattern, and the other is 18-month reloading pattern. The core neutronic parameters are investigated by comparing with the 100% UO2 core. The MOX fuel has impacts on the nuclear characteristic parameters, but the 50% MOX core still satisfies the nuclear and safety criteria samilar to that considered in 100% UO2 core. The results of this study prove that the GIII LWR HPR1000 has excellent capability while loading with 50% MOX fuel.
Fuel loading pattern optimization with MOX (mixed oxide of plutonium and uranium) fuel is an important aspect for GIII LWR. For the HPR1000 core, an optimized MOX fuel is designed. Based on which, two loading patterns with 50% MOX fuel are designed. One is 1/4 core annually reloading pattern, and the other is 18-month reloading pattern. The core neutronic parameters are investigated by comparing with the 100% UO2 core. The MOX fuel has impacts on the nuclear characteristic parameters, but the 50% MOX core still satisfies the nuclear and safety criteria samilar to that considered in 100% UO2 core. The results of this study prove that the GIII LWR HPR1000 has excellent capability while loading with 50% MOX fuel.
2017,
29: 036011.
doi: 10.11884/HPLPB201729.160172
Abstract:
Power control banks-turbine load test (RGL04 test) is to be performed to verify the accuracy of calibration curve by quickly reducing the power before nuclear power plants connecting to power grid. Overheating phenomenon occurred during the tests in CPR1000 units of nuclear power plants in Ningde and Yangjiang units. This paper analyzes and evaluates the causes and the factors that lead to overheating during the RGL04 test. The overheat of the core can be predicted by calculating the theoretical value of isothermal temperature coefficient. A sound advice to avoid large superheat is given at the end of this paper.
Power control banks-turbine load test (RGL04 test) is to be performed to verify the accuracy of calibration curve by quickly reducing the power before nuclear power plants connecting to power grid. Overheating phenomenon occurred during the tests in CPR1000 units of nuclear power plants in Ningde and Yangjiang units. This paper analyzes and evaluates the causes and the factors that lead to overheating during the RGL04 test. The overheat of the core can be predicted by calculating the theoretical value of isothermal temperature coefficient. A sound advice to avoid large superheat is given at the end of this paper.
2017,
29: 036012.
doi: 10.11884/HPLPB201729.160338
Abstract:
In the field of nuclear arms control, non-proliferation is an significant area, particularly the proliferation of weapon usable nuclear materials, which is concerned by many countries. We used burnup calculating code MCORGS to simulate the relation between the burnup of spent fuel in PWR and Pu isotopic composition. We built the model of pin-cell with 20 average axial zones, calculated its axial burnup distribution, and got the axial burnup and Pu-239 isotopic mass concentration distribution in different axial positions in PWR. Based on burnup calculation we found that Pu-239 isotopic concentrations in different axial position varied greatly. Further more, we simulated VVER1000 and PWRs 1717 assembly. The simulation shows that the radial burnups in an assembly were also different. There are much different burnups in different positions of nuclear reactor, which result in many low burnup zones in spent fuel. This kind of spent fuel of LWR might brings serious nuclear proliferation to international community, and should be supervised more strictly.
In the field of nuclear arms control, non-proliferation is an significant area, particularly the proliferation of weapon usable nuclear materials, which is concerned by many countries. We used burnup calculating code MCORGS to simulate the relation between the burnup of spent fuel in PWR and Pu isotopic composition. We built the model of pin-cell with 20 average axial zones, calculated its axial burnup distribution, and got the axial burnup and Pu-239 isotopic mass concentration distribution in different axial positions in PWR. Based on burnup calculation we found that Pu-239 isotopic concentrations in different axial position varied greatly. Further more, we simulated VVER1000 and PWRs 1717 assembly. The simulation shows that the radial burnups in an assembly were also different. There are much different burnups in different positions of nuclear reactor, which result in many low burnup zones in spent fuel. This kind of spent fuel of LWR might brings serious nuclear proliferation to international community, and should be supervised more strictly.
2017,
29: 036013.
doi: 10.11884/HPLPB201729.160191
Abstract:
Ex-core detector response function represents the contribution to the detector tally from each reactor core assembly, reflects the relationship between reactor core power distribution and detector tally. In this paper, the adjoint transport calculation by three-dimensional discrete coordinate method (SN) program TORT was researched, and a post-process code to achieve the conversion between three-dimensional adjoint flux and PWR fuel assembly response functions was developed; Based on the CAP1400 reactor model, its ex-core detector response functions were analyzed, and the results match well with the TORT forward calculation results. Through the research of this paper, three-dimensional distribution calculation of PWR plant ex-core detector response functions was achieved.
Ex-core detector response function represents the contribution to the detector tally from each reactor core assembly, reflects the relationship between reactor core power distribution and detector tally. In this paper, the adjoint transport calculation by three-dimensional discrete coordinate method (SN) program TORT was researched, and a post-process code to achieve the conversion between three-dimensional adjoint flux and PWR fuel assembly response functions was developed; Based on the CAP1400 reactor model, its ex-core detector response functions were analyzed, and the results match well with the TORT forward calculation results. Through the research of this paper, three-dimensional distribution calculation of PWR plant ex-core detector response functions was achieved.
2017,
29: 036014.
doi: 10.11884/HPLPB201729.160298
Abstract:
The fast, accurate and highly automatic power regulation ability in a large range is a main research objective of advanced nuclear power plant control strategy. In the current PWR plant, the mechanical shim control and operation strategy is a good method that could meet all above requirements. In mechanical shim, control rods are mainly used to control core operation. Thus the rod worth has a great influence on operation ability. This paper has studied the influence of increased gray control rod worth on operation ability. The results show that increased gray control rod worth and improvement of rod worth axial distribution could enhance load follow operation ability obviously. The analysis results of this paper have a great reference meaning to the design improvement of rod cluster control assembly and the research on advanced PWR nuclear power plant control strategy.
The fast, accurate and highly automatic power regulation ability in a large range is a main research objective of advanced nuclear power plant control strategy. In the current PWR plant, the mechanical shim control and operation strategy is a good method that could meet all above requirements. In mechanical shim, control rods are mainly used to control core operation. Thus the rod worth has a great influence on operation ability. This paper has studied the influence of increased gray control rod worth on operation ability. The results show that increased gray control rod worth and improvement of rod worth axial distribution could enhance load follow operation ability obviously. The analysis results of this paper have a great reference meaning to the design improvement of rod cluster control assembly and the research on advanced PWR nuclear power plant control strategy.
2017,
29: 036015.
doi: 10.11884/HPLPB201729.160189
Abstract:
Nuclear criticality safety is one of the most important security issues in the process of development of nuclear industry, in which evaluation and analysis of critical accidents is the foundation of shielding design and emergency plan after critical accidents. Therefore it has important research significance and engineering value. CAACS is a program developed to analyze criticality accidents of solution system. It can calculate the number of fission, fission power and temperature change with time of the criticality accident and so on. CAACS has been verified by the benchmark experiments. This paper presents the analysis and calculation of two criticality accidents, and compares the calculated results with the accident estimates. The result shows that, the calculation results of CAACS concur with the accident estimates, the study could provide technical means for the design of fuel reprocessing plants, and also lay a foundation for the subsequent research of critical transient.
Nuclear criticality safety is one of the most important security issues in the process of development of nuclear industry, in which evaluation and analysis of critical accidents is the foundation of shielding design and emergency plan after critical accidents. Therefore it has important research significance and engineering value. CAACS is a program developed to analyze criticality accidents of solution system. It can calculate the number of fission, fission power and temperature change with time of the criticality accident and so on. CAACS has been verified by the benchmark experiments. This paper presents the analysis and calculation of two criticality accidents, and compares the calculated results with the accident estimates. The result shows that, the calculation results of CAACS concur with the accident estimates, the study could provide technical means for the design of fuel reprocessing plants, and also lay a foundation for the subsequent research of critical transient.
2017,
29: 036016.
doi: 10.11884/HPLPB201729.160355
Abstract:
With the development of nuclear power industry, spent fuel of nuclear power plant is increasing. The disposal of minor actinides(MA) of spent nuclear fuel in nuclear power plants is not only an important process of recycling nuclear fuel, but also key step in the closed cycle. If MA can be disposed properly, the utilization rate of fuel can be improved, and MA can be turned into useful isotopes such as radionuclide fuel cell isotopes and neutron source isotopes and so on. Internationally accepted method of disposal is transmutation, but the difficulties of MA transmutation are how to select the type of transmutation reactor and how to improve the transmutation rate. Because the pressurized water reactor (PWR) is the most mature type of reactor and the main commercial operation reactor type, PWR has the most possibility to transmutate MA at this stage. Therefore, we did the research of the transmutation characteristics of MA with MCNP in PWR. Studying the design, layout and supplementation of MA transmutation rod and the influence on the effective multiplication factor, we explored the best transmutation MA design and technology directions in PWR. The results would lay the theoretical foundation for MA transmutation in PWR.
With the development of nuclear power industry, spent fuel of nuclear power plant is increasing. The disposal of minor actinides(MA) of spent nuclear fuel in nuclear power plants is not only an important process of recycling nuclear fuel, but also key step in the closed cycle. If MA can be disposed properly, the utilization rate of fuel can be improved, and MA can be turned into useful isotopes such as radionuclide fuel cell isotopes and neutron source isotopes and so on. Internationally accepted method of disposal is transmutation, but the difficulties of MA transmutation are how to select the type of transmutation reactor and how to improve the transmutation rate. Because the pressurized water reactor (PWR) is the most mature type of reactor and the main commercial operation reactor type, PWR has the most possibility to transmutate MA at this stage. Therefore, we did the research of the transmutation characteristics of MA with MCNP in PWR. Studying the design, layout and supplementation of MA transmutation rod and the influence on the effective multiplication factor, we explored the best transmutation MA design and technology directions in PWR. The results would lay the theoretical foundation for MA transmutation in PWR.
2017,
29: 036017.
doi: 10.11884/HPLPB201628.160291
Abstract:
The 3-D Monte Carlo transport code JMCT is developed by Software Center for High Performance Numerical Simulation independently and JLAMT is its pre-processing visualized modeling tool. Benchmarks including BW, KRITZ, BEAVRS et al were simulated by JMCT and MCNP5 codes, in addition, the effective multiplication factors and tally results were compared with the MCNP5 code and measured data separately. Effective multiplication factor calculations between two codes agree well with each other within 89.1 pcm for the KRITZ2 benchmark, the relative difference in power distribution between JMCT and MCNP5 is below 2% mostly, average difference is about 1%; as for BEAVRS benchmark, the maximum differences of the core radial power distribution of JMCT compared to MCNP5 and measurements are 7.06% and 16.6% respectively, JMCT control rod bank worth calculation results show good agreement of JMCT with MCNP5 and measurement.
The 3-D Monte Carlo transport code JMCT is developed by Software Center for High Performance Numerical Simulation independently and JLAMT is its pre-processing visualized modeling tool. Benchmarks including BW, KRITZ, BEAVRS et al were simulated by JMCT and MCNP5 codes, in addition, the effective multiplication factors and tally results were compared with the MCNP5 code and measured data separately. Effective multiplication factor calculations between two codes agree well with each other within 89.1 pcm for the KRITZ2 benchmark, the relative difference in power distribution between JMCT and MCNP5 is below 2% mostly, average difference is about 1%; as for BEAVRS benchmark, the maximum differences of the core radial power distribution of JMCT compared to MCNP5 and measurements are 7.06% and 16.6% respectively, JMCT control rod bank worth calculation results show good agreement of JMCT with MCNP5 and measurement.
2017,
29: 036018.
doi: 10.11884/HPLPB201729.160194
Abstract:
Nuclear reactors will produce a large number of neutrons when the plant is in operation. The neutrons could have strong effect on in-core materials and generate active products which could cause destructive effects on staff. Therefore, the high precision and high efficiency calculation of material neutron activation has a significant value for reactor radiation protection. Based on Bateman equation which is used in traditional Transmutation Trajectory Analysis (TTA), in this paper, limit operation is used to derive the expression of generalized TTA in which repeated eigenvalues are allowed. In this way, the restriction of similar decay constants are removed. Meanwhile, the backtracking algorithm is included for automatically searching nuclide linear chains and the computational efficiency is improved visibly by using the improved TTA. On this foundation, an activation calculation code named ITACT(Improved TTA ACTIVATION) is developed. Finally, this paper combines ITACT with EAF-2007 database to study the activation of cladding material in PWR reactor and first wall material in fusion reactor. Compared with the European general activation code FISPACT, for long life nuclei, the results are in good agreement with each other. But for short life nuclei, ITACT gets higher calculation accuracy, which verifies its feasibility and accuracy.
Nuclear reactors will produce a large number of neutrons when the plant is in operation. The neutrons could have strong effect on in-core materials and generate active products which could cause destructive effects on staff. Therefore, the high precision and high efficiency calculation of material neutron activation has a significant value for reactor radiation protection. Based on Bateman equation which is used in traditional Transmutation Trajectory Analysis (TTA), in this paper, limit operation is used to derive the expression of generalized TTA in which repeated eigenvalues are allowed. In this way, the restriction of similar decay constants are removed. Meanwhile, the backtracking algorithm is included for automatically searching nuclide linear chains and the computational efficiency is improved visibly by using the improved TTA. On this foundation, an activation calculation code named ITACT(Improved TTA ACTIVATION) is developed. Finally, this paper combines ITACT with EAF-2007 database to study the activation of cladding material in PWR reactor and first wall material in fusion reactor. Compared with the European general activation code FISPACT, for long life nuclei, the results are in good agreement with each other. But for short life nuclei, ITACT gets higher calculation accuracy, which verifies its feasibility and accuracy.
2017,
29: 036019.
doi: 10.11884/HPLPB201729.160399
Abstract:
In order to compare core characteristics of conventional fast reactor with travelling wave reactor, a conceptual neutronic design of conventional fast reactor called HBFR (High Burnup Fast Reactor) with maximum burnup up to 300 000 MWd/tHM was given. In order to decrease the burnup reactivity swing, a refueling strategy which refuel only one-sixth of fuel assembly was chosen. The NAS code was used to analyse three different operating conditions: cold room temperature, hot standby and full-power conditions. Some core parameters such as criticality, power distribution, DPA characters, temperature and power reactivity, control rod worth, etc. are calculated. The results show that the maximum burnup of fuel assembly is 300 000 MWd/tHM, the average burnup is about 219 000 MWd/tHM and the burnup reactivity swing is 3.7%k/k which can be controlled by regulation rods. The design of HBFR can meet the design objectives and design limits and provide data to compare with TWR effectively.
In order to compare core characteristics of conventional fast reactor with travelling wave reactor, a conceptual neutronic design of conventional fast reactor called HBFR (High Burnup Fast Reactor) with maximum burnup up to 300 000 MWd/tHM was given. In order to decrease the burnup reactivity swing, a refueling strategy which refuel only one-sixth of fuel assembly was chosen. The NAS code was used to analyse three different operating conditions: cold room temperature, hot standby and full-power conditions. Some core parameters such as criticality, power distribution, DPA characters, temperature and power reactivity, control rod worth, etc. are calculated. The results show that the maximum burnup of fuel assembly is 300 000 MWd/tHM, the average burnup is about 219 000 MWd/tHM and the burnup reactivity swing is 3.7%k/k which can be controlled by regulation rods. The design of HBFR can meet the design objectives and design limits and provide data to compare with TWR effectively.
2017,
29: 036020.
doi: 10.11884/HPLPB201729.160400
Abstract:
Reactor shielding calculations is the foundation for assessing nuclear plant safety performance. It is also an important method to provide guidelines for constructions and operations of nuclear plant. JSNT is a massively parallel discrete ordinates transport code developed by CAEP software center for high performance numerical simulation, with high simulation precision and efficiency. In this paper, a commercial PWR has been modeled and simulated using JSNT code. Flux distributions are obtained and compared with experimental measuring results. Comparisons show that results obtained by both S8 and S16 calculations could satisfy engineering demand. Moreover, in comparison to S8 calculation, S16 could improve the calculation precision significantly and decrease relative errors of some detectors under 1%.
Reactor shielding calculations is the foundation for assessing nuclear plant safety performance. It is also an important method to provide guidelines for constructions and operations of nuclear plant. JSNT is a massively parallel discrete ordinates transport code developed by CAEP software center for high performance numerical simulation, with high simulation precision and efficiency. In this paper, a commercial PWR has been modeled and simulated using JSNT code. Flux distributions are obtained and compared with experimental measuring results. Comparisons show that results obtained by both S8 and S16 calculations could satisfy engineering demand. Moreover, in comparison to S8 calculation, S16 could improve the calculation precision significantly and decrease relative errors of some detectors under 1%.
2017,
29: 036021.
doi: 10.11884/HPLPB201729.160326
Abstract:
The incident energy of thermal neutron ranges from 110-5 eV to 5 eV, which is comparable to energy of thermal motion of nuclide in the reactor, the reactions between neutron and target have vastly different characteristic in this energy range. The considerable effect of chemical bond, crystal lattice structure and thermal motion need to be taken into account. Distinct theoretical and processing methods have to be employed for different materials to obtain thermal scattering data used for transport codes. Correlative processing methods for thermal scattering data are introduced in this paper, Thermc module used for processing thermal scattering data was developed based on some of these methods. In addition, thermal scattering data result from Thermc module of RXSP and Thermr module of NJOY were compared in detail, macro examination was also carried out base on benchmark, results of comparison and examination demonstrate that data from Thermc are accurate.
The incident energy of thermal neutron ranges from 110-5 eV to 5 eV, which is comparable to energy of thermal motion of nuclide in the reactor, the reactions between neutron and target have vastly different characteristic in this energy range. The considerable effect of chemical bond, crystal lattice structure and thermal motion need to be taken into account. Distinct theoretical and processing methods have to be employed for different materials to obtain thermal scattering data used for transport codes. Correlative processing methods for thermal scattering data are introduced in this paper, Thermc module used for processing thermal scattering data was developed based on some of these methods. In addition, thermal scattering data result from Thermc module of RXSP and Thermr module of NJOY were compared in detail, macro examination was also carried out base on benchmark, results of comparison and examination demonstrate that data from Thermc are accurate.
2017,
29: 036022.
doi: 10.11884/HPLPB201729.160192
Abstract:
KYCORE is a radial MOC and axial SN coupled neutron transport software developed by China Institute of Nuclear Power. Its highly accurate coupling between 2D MOC and 1D SN is realized by angular flux, which is one of the most accurate methods applying to 3D neutron transport calculation by now. The calculation is accelerated by CMFD for engineering application. This article introduces the 2D/1D coupling and acceleration theory. The accuracy and efficiency of the KYCORE 3D neutron transport calculation are verified by comparison with the Monte Carlo program.
KYCORE is a radial MOC and axial SN coupled neutron transport software developed by China Institute of Nuclear Power. Its highly accurate coupling between 2D MOC and 1D SN is realized by angular flux, which is one of the most accurate methods applying to 3D neutron transport calculation by now. The calculation is accelerated by CMFD for engineering application. This article introduces the 2D/1D coupling and acceleration theory. The accuracy and efficiency of the KYCORE 3D neutron transport calculation are verified by comparison with the Monte Carlo program.
2017,
29: 036023.
doi: 10.11884/HPLPB201729.160186
Abstract:
This paper studies and proposes a mechanism-based calculation method of secondary neutron source (SNS) intensity of nuclear reactor, and calculated the SNS intensity of the reactor of a certain nuclear power plant at the refueling outage of the second cycle. In order to verify the calculated SNS intensity, when the step 2 of the refueling process is finished, the counting rate of the ex-core source range detector is calculated and compared with the measured data. The comparison result shows good agreement and indicates the correctness of the SNS intensity calculation result and the reasonableness of the proposed calculation method.
This paper studies and proposes a mechanism-based calculation method of secondary neutron source (SNS) intensity of nuclear reactor, and calculated the SNS intensity of the reactor of a certain nuclear power plant at the refueling outage of the second cycle. In order to verify the calculated SNS intensity, when the step 2 of the refueling process is finished, the counting rate of the ex-core source range detector is calculated and compared with the measured data. The comparison result shows good agreement and indicates the correctness of the SNS intensity calculation result and the reasonableness of the proposed calculation method.
2017,
29: 036024.
doi: 10.11884/HPLPB201729.160333
Abstract:
This paper studies the application of solving high temperature reactor (HTR) neutron diffusion equation with Jacobian-free Newton-Krylov (JFNK) method. Results show that when solving neutron diffusion equation, the relative residual norm of JFNK method decreases slowly at the beginning. Then the rate of convergence become faster and finally reaches a relatively stable value. This feature is conducive to a high-accuracy solution. In the test of two kinds of additional equations, the neutron diffusion equation with flux normalization condition has a better nonlinear convergence behavior. However, due to the longer computational time in solving linear equations, its total computational time is more than the one with k expression. More efficient preconditioning methods should be studied to improve linear equations.
This paper studies the application of solving high temperature reactor (HTR) neutron diffusion equation with Jacobian-free Newton-Krylov (JFNK) method. Results show that when solving neutron diffusion equation, the relative residual norm of JFNK method decreases slowly at the beginning. Then the rate of convergence become faster and finally reaches a relatively stable value. This feature is conducive to a high-accuracy solution. In the test of two kinds of additional equations, the neutron diffusion equation with flux normalization condition has a better nonlinear convergence behavior. However, due to the longer computational time in solving linear equations, its total computational time is more than the one with k expression. More efficient preconditioning methods should be studied to improve linear equations.
2017,
29: 036025.
doi: 10.11884/HPLPB201729.160424
Abstract:
The burnup calculation of reactors is related to the fuel management of the reactor, and directly affect the economic evaluation of the core, therefore how to calculate burnup characteristics of reactors faster and better is an important part of the research. With the development of the reactor, its geometry becomes more and more complex. Some of the existing one-dimensional or two-dimensional burnup coupled program, due to restrictions on the geometry processing, are difficult to meet the requirements of advanced reactors design and analysis. In this paper, we combine the advantages of MCNP which is good at dealing with complex geometry, and FISPACT which could process radionuclides very comprehensively. The FISPACT and MCNP were coupled for burnup calculation, and the coupled program was verified by calculating the examplesIAEA ADS-benchmark and fusion reactor example. The calculation results of keff and tritium breeding ratio(TBR) by the coupled program are in good agreement with the standard results, and the errors are within acceptable ranges.
The burnup calculation of reactors is related to the fuel management of the reactor, and directly affect the economic evaluation of the core, therefore how to calculate burnup characteristics of reactors faster and better is an important part of the research. With the development of the reactor, its geometry becomes more and more complex. Some of the existing one-dimensional or two-dimensional burnup coupled program, due to restrictions on the geometry processing, are difficult to meet the requirements of advanced reactors design and analysis. In this paper, we combine the advantages of MCNP which is good at dealing with complex geometry, and FISPACT which could process radionuclides very comprehensively. The FISPACT and MCNP were coupled for burnup calculation, and the coupled program was verified by calculating the examplesIAEA ADS-benchmark and fusion reactor example. The calculation results of keff and tritium breeding ratio(TBR) by the coupled program are in good agreement with the standard results, and the errors are within acceptable ranges.
