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RSS2017 Vol. 29, No. 01
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2017,
29: 016001.
doi: 10.11884/HPLPB201729.160176
Abstract:
In the fast reactor analysis, one of the most important things is generation of homogenized cross sections starting from the evaluated nuclear data files. By using point-wise cross section directly during the generation of problem dependent ultrafine group cross section, the treatments of the elastic scattering resonance effect of intermediate atom weight nuclide and the interference effect of all nuclides would be more accurate. Therefore, the neutron spectrum obtained from solving neutron transport equation could be more accurate and it would be the weight function for group condensation. Comparing with Monte Carlo continuous energy calculation, the results obtained by using point-wise cross section directly show that the relative differences of ultrafine group cross sections and few group cross sections after group condensation are both less that 1%. Consequently, the core analysis could get better results.
In the fast reactor analysis, one of the most important things is generation of homogenized cross sections starting from the evaluated nuclear data files. By using point-wise cross section directly during the generation of problem dependent ultrafine group cross section, the treatments of the elastic scattering resonance effect of intermediate atom weight nuclide and the interference effect of all nuclides would be more accurate. Therefore, the neutron spectrum obtained from solving neutron transport equation could be more accurate and it would be the weight function for group condensation. Comparing with Monte Carlo continuous energy calculation, the results obtained by using point-wise cross section directly show that the relative differences of ultrafine group cross sections and few group cross sections after group condensation are both less that 1%. Consequently, the core analysis could get better results.
2017,
29: 016002.
doi: 10.11884/HPLPB201729.160181
Abstract:
The reactor reloading pattern optimization is a typical combinatorial optimization problem with a huge search space. It is very hard for traditional optimization algorithm to find the global optimal solution in such huge search space. However, for combinatorial optimization problem, the genetic algorithm (GA) provides a very effective solution by its excellent adaptive ability and optimization ability. This paper is focused on the reloading pattern optimization by using GA in a block-type high temperature gas cooled reactor(HTGR) and corresponding programs were written to realize this goal. To improve the calculation accuracy of core physics, the transport calculation with 26 groups is adopted in the core calculation, which will also be time-consuming. To make up for this shortcoming, the parallel optimization of GA is carried out. Finally, a refueling optimization benchmark in a small HTGR is constructed to test the optimization ability of GA. The results show that GA has a good optimization ability and computational stability for reloading pattern optimization in block-type HTGRs.
The reactor reloading pattern optimization is a typical combinatorial optimization problem with a huge search space. It is very hard for traditional optimization algorithm to find the global optimal solution in such huge search space. However, for combinatorial optimization problem, the genetic algorithm (GA) provides a very effective solution by its excellent adaptive ability and optimization ability. This paper is focused on the reloading pattern optimization by using GA in a block-type high temperature gas cooled reactor(HTGR) and corresponding programs were written to realize this goal. To improve the calculation accuracy of core physics, the transport calculation with 26 groups is adopted in the core calculation, which will also be time-consuming. To make up for this shortcoming, the parallel optimization of GA is carried out. Finally, a refueling optimization benchmark in a small HTGR is constructed to test the optimization ability of GA. The results show that GA has a good optimization ability and computational stability for reloading pattern optimization in block-type HTGRs.
2017,
29: 016003.
doi: 10.11884/HPLPB201729.160190
Abstract:
RMC is a self-developed Monte Carlo software for nuclear reactor analysis by Reactor Engineering Analysis Lab(REAL), Tsinghua University. On the basis of the self-developed subchannel modular(RMC-TH) and Monte Carlo CellTally the inner coupling interface is developed, which combines both input files and realizes the fast mesh correspondence process using the cell expansion technology for repeat structure with thermal hydraulics feedback, and it breaks through the threshold of geometrical variability and extensibility for coupled code. On-The-Fly Doppler broaden method is adopted considering the temperature effect on micro cross section, which only needs the 0K cross section library so that the memory cost can be apparently reduced. Steady state simulation analysis are performed on single rod and 1717 assembly model, and the results show the feasibility, accuracy and efficiency of the coupling methodology, so the technology roadmap and methodology foundation for the large scale and geometrically variable reactor steady-state as well as transient neutronics simulation with thermal hydraulic feedback and further neutronics-thermal hydraulics-depletion multi-physics simulation process are established.
RMC is a self-developed Monte Carlo software for nuclear reactor analysis by Reactor Engineering Analysis Lab(REAL), Tsinghua University. On the basis of the self-developed subchannel modular(RMC-TH) and Monte Carlo CellTally the inner coupling interface is developed, which combines both input files and realizes the fast mesh correspondence process using the cell expansion technology for repeat structure with thermal hydraulics feedback, and it breaks through the threshold of geometrical variability and extensibility for coupled code. On-The-Fly Doppler broaden method is adopted considering the temperature effect on micro cross section, which only needs the 0K cross section library so that the memory cost can be apparently reduced. Steady state simulation analysis are performed on single rod and 1717 assembly model, and the results show the feasibility, accuracy and efficiency of the coupling methodology, so the technology roadmap and methodology foundation for the large scale and geometrically variable reactor steady-state as well as transient neutronics simulation with thermal hydraulic feedback and further neutronics-thermal hydraulics-depletion multi-physics simulation process are established.
2017,
29: 016004.
doi: 10.11884/HPLPB201729.160199
Abstract:
The optimization of the core periodical physical flux map test could be done using the SOPHORA on-line core monitoring system developed by China General Nuclear Power Group. This system can perform the calculation of 3D online monitoring, core operation analysis in depth, core subsequent state simulation and prediction, the core operation strategy generation and optimization, and so on. Combining on-line theoretical calculation and instrument measurement, SOPHORA can get the best estimated measured power distribution. By coupling the self-reliant flux map processing code MAPLE and the physics experiment processing code HOLLY, one can analyse the feasibility of delay of RPN11 test interval. A single point ex-core calibration method has been developed in SOPHORA, which can reduce the frequency of carrying out RPN12 test. The results of sensitivity study of RPN11 and RPN12 test measurement process show that it is possible to extend the time interval of RPN11 test and to cancel some of the RPN12 test experiment.
The optimization of the core periodical physical flux map test could be done using the SOPHORA on-line core monitoring system developed by China General Nuclear Power Group. This system can perform the calculation of 3D online monitoring, core operation analysis in depth, core subsequent state simulation and prediction, the core operation strategy generation and optimization, and so on. Combining on-line theoretical calculation and instrument measurement, SOPHORA can get the best estimated measured power distribution. By coupling the self-reliant flux map processing code MAPLE and the physics experiment processing code HOLLY, one can analyse the feasibility of delay of RPN11 test interval. A single point ex-core calibration method has been developed in SOPHORA, which can reduce the frequency of carrying out RPN12 test. The results of sensitivity study of RPN11 and RPN12 test measurement process show that it is possible to extend the time interval of RPN11 test and to cancel some of the RPN12 test experiment.
2017,
29: 016005.
doi: 10.11884/HPLPB201729.160200
Abstract:
A new on-line core monitoring software system (CMSS), named SOPHORA, is developed by China General Nuclear Power Corporation (CGN). Based on the measured fixing incore detector (FID) readings and on-line core status parameters, the best estimated 3D core power distribution is reconstructed, and the safety relative parameters such as peaking local power margin and minimum DNB margin can be also provided. Pseudo simulated FID signals, which are generated from the core movable incore detector (MID) measurement data, are used as input for SOPHORA to verify and to validate the system accuracy. Comparison results of core power peaking factors between SOPHORA and flux map processing code MAPLE show that the SOPHORA three dimensional core power distribution reconstruction function is reasonable and acceptable.
A new on-line core monitoring software system (CMSS), named SOPHORA, is developed by China General Nuclear Power Corporation (CGN). Based on the measured fixing incore detector (FID) readings and on-line core status parameters, the best estimated 3D core power distribution is reconstructed, and the safety relative parameters such as peaking local power margin and minimum DNB margin can be also provided. Pseudo simulated FID signals, which are generated from the core movable incore detector (MID) measurement data, are used as input for SOPHORA to verify and to validate the system accuracy. Comparison results of core power peaking factors between SOPHORA and flux map processing code MAPLE show that the SOPHORA three dimensional core power distribution reconstruction function is reasonable and acceptable.
2017,
29: 016006.
doi: 10.11884/HPLPB201729.160208
Abstract:
Considering that the conventional Bondarenko-Iteration Method (BIM) would introduce much error in treating resonance interference effect, the Resonance-Interference-Factor Method (RIFM) and the Heterogeneous-Pseudo-Resonant-Isotope Method (HPRIM) was developed by researchers. Though these two methods can give relatively precise pin-averaged self-shielded cross sections by considering resonance interference effect, they cannot give exact spatial dependent self-shielded cross sections for coarsely modeling spatial self-shielding effect. In order to overcome this problem, the Pseudo-Resonant-Nuclide Subgroup Method (PRNSM) is proposed in this paper. The PRNSM takes into account the resonance interference effect by making resonance cross section table of the pseudo resonant nuclide on-line. The subgroup parameters of the pseudo resonant nuclide and the partial resonant nuclide are obtained by fitting method. The spatial dependent self-shielded cross sections are obtained by condensing the subgroup cross sections of the partial resonant nuclide with the subgroup flux which is obtained by solving subgroup fixed source problem of the pseudo resonant nuclide. The numerous numerical results show that the PRNSM can give both precise pin-averaged self-shielded cross sections and precise spatial dependent self-shielded cross sections for UO2 problems with different enrichments.
Considering that the conventional Bondarenko-Iteration Method (BIM) would introduce much error in treating resonance interference effect, the Resonance-Interference-Factor Method (RIFM) and the Heterogeneous-Pseudo-Resonant-Isotope Method (HPRIM) was developed by researchers. Though these two methods can give relatively precise pin-averaged self-shielded cross sections by considering resonance interference effect, they cannot give exact spatial dependent self-shielded cross sections for coarsely modeling spatial self-shielding effect. In order to overcome this problem, the Pseudo-Resonant-Nuclide Subgroup Method (PRNSM) is proposed in this paper. The PRNSM takes into account the resonance interference effect by making resonance cross section table of the pseudo resonant nuclide on-line. The subgroup parameters of the pseudo resonant nuclide and the partial resonant nuclide are obtained by fitting method. The spatial dependent self-shielded cross sections are obtained by condensing the subgroup cross sections of the partial resonant nuclide with the subgroup flux which is obtained by solving subgroup fixed source problem of the pseudo resonant nuclide. The numerous numerical results show that the PRNSM can give both precise pin-averaged self-shielded cross sections and precise spatial dependent self-shielded cross sections for UO2 problems with different enrichments.
2017,
29: 016007.
doi: 10.11884/HPLPB201729.160212
Abstract:
JMCT developed by Institute of Applied Physics and Computational Mathematics is a transport code based on Monte Carlo method. A series of benchmark experiments picked up from the international nuclear critical experiment database, were used to validate JMCT. The validation includes common fission nuclides(such as U and Pu), from lower enrichment to higher enrichment, their thermal neutron groups and fast neutron groups, and common neutron poison and reflector. The results of JMCT were compared with the results of experiments by MCNP and MONK. The correctness of JMCT is preliminarily verified.
JMCT developed by Institute of Applied Physics and Computational Mathematics is a transport code based on Monte Carlo method. A series of benchmark experiments picked up from the international nuclear critical experiment database, were used to validate JMCT. The validation includes common fission nuclides(such as U and Pu), from lower enrichment to higher enrichment, their thermal neutron groups and fast neutron groups, and common neutron poison and reflector. The results of JMCT were compared with the results of experiments by MCNP and MONK. The correctness of JMCT is preliminarily verified.
2017,
29: 016009.
doi: 10.11884/HPLPB201729.160226
Abstract:
Measurements of in-core detectors, as one of the most important inputs for a reactor core on-line power distribution monitoring system, seriously affect the on-line monitoring results. Therefore, the diagnosis and processing of detector failure are necessary. Harmonics expansion method is employed to the on-line monitoring calculation of reactor core power distribution. To diagnose failure of detectors, three methods direct method, comparison of measurements method and comparison of reconstructed responses methodare used in combination. Based on these methods, diagnosis and processing of detector failure have been added into NECP-ONION, an in-house online monitoring code developed by NECP Lab. Benchmark for Evaluation and Validation Reactor Simulations(BEAVRS) has been used for code validation, especially for the diagnosis and processing of detector failure. Numerical results show that, the combination of these three methods is not only effective to the complete failed condition, but also useful for measurements deviation of normal value. In addition, NECP-ONION is functional to distinguish detector failure or local power oscillation. This could reduce the misdiagnosis for detector failure. For the processing of detector failure, it has been observed that single detector failure or a small number of detectors failed has limit effect on the monitoring system NECP-ONION.
Measurements of in-core detectors, as one of the most important inputs for a reactor core on-line power distribution monitoring system, seriously affect the on-line monitoring results. Therefore, the diagnosis and processing of detector failure are necessary. Harmonics expansion method is employed to the on-line monitoring calculation of reactor core power distribution. To diagnose failure of detectors, three methods direct method, comparison of measurements method and comparison of reconstructed responses methodare used in combination. Based on these methods, diagnosis and processing of detector failure have been added into NECP-ONION, an in-house online monitoring code developed by NECP Lab. Benchmark for Evaluation and Validation Reactor Simulations(BEAVRS) has been used for code validation, especially for the diagnosis and processing of detector failure. Numerical results show that, the combination of these three methods is not only effective to the complete failed condition, but also useful for measurements deviation of normal value. In addition, NECP-ONION is functional to distinguish detector failure or local power oscillation. This could reduce the misdiagnosis for detector failure. For the processing of detector failure, it has been observed that single detector failure or a small number of detectors failed has limit effect on the monitoring system NECP-ONION.
2017,
29: 016008.
doi: 10.11884/HPLPB201729.160221
Abstract:
CTF (Coolant Boiling in Rod Arrays-Two Fluid) is a new sub-channel thermal/hydraulic simulation code developed by CASL (The Consortium for Advanced Simulation of Light Water Reactors) and PSU (Pennsylvania State University). It can solve steady or transient-state problems efficiently for both single assembly and full-core reactor. Thus, this code solves the computational efficiency and memory consumption problems effectively. First, CTF computes the BEAVRS benchmark in parallel with the domain decomposition technology. The power data which CTF uses are calculated by RMC(a Monte Carlo code for reactor core analysis). After calculating for 268 s, we get detailed fuel pin temperature, water temp and density output results. Accordingly, the efficiency and reliability of CTF is verified. On this basic work of CTF calculation for BEAVRS benchmark,the coupling between RMC and CTS for full-core problem will achieved soon.
CTF (Coolant Boiling in Rod Arrays-Two Fluid) is a new sub-channel thermal/hydraulic simulation code developed by CASL (The Consortium for Advanced Simulation of Light Water Reactors) and PSU (Pennsylvania State University). It can solve steady or transient-state problems efficiently for both single assembly and full-core reactor. Thus, this code solves the computational efficiency and memory consumption problems effectively. First, CTF computes the BEAVRS benchmark in parallel with the domain decomposition technology. The power data which CTF uses are calculated by RMC(a Monte Carlo code for reactor core analysis). After calculating for 268 s, we get detailed fuel pin temperature, water temp and density output results. Accordingly, the efficiency and reliability of CTF is verified. On this basic work of CTF calculation for BEAVRS benchmark,the coupling between RMC and CTS for full-core problem will achieved soon.
2017,
29: 016010.
doi: 10.11884/HPLPB201729.160235
Abstract:
The program of nuclear medical ship, funded by the cancer healing program of IAEA, was proposed according to the existing Miniature Neutron Source Reactor (MNSR) technology. The nuclear medical ship equipped with Boron Neutron Capture Therapy(BNCT) device was developed, which opened a new scope of nuclear science application. The reactor of nuclear medical ship was designed in accordance with In-hospital Neutron Irradiator mark I (IHNI-1). In IHNI-1, heavy concrete was used as shielding material, and the reactor pool was cylindrical. Whereas, the volume and total weight of the reactor were too large to meet the ships requirement. After the design and optimization of the reactors shielding system by M-C method, stainless steel and B-polyethylene were chosen as shielding materials, and square pool was substituted by compact cylindrical pool. The result shows that on the premise of guaranteeing safety, shielding systems mass and volume were cut down greatly, i. e., this design can meet the requirement of nuclear medical ship.
The program of nuclear medical ship, funded by the cancer healing program of IAEA, was proposed according to the existing Miniature Neutron Source Reactor (MNSR) technology. The nuclear medical ship equipped with Boron Neutron Capture Therapy(BNCT) device was developed, which opened a new scope of nuclear science application. The reactor of nuclear medical ship was designed in accordance with In-hospital Neutron Irradiator mark I (IHNI-1). In IHNI-1, heavy concrete was used as shielding material, and the reactor pool was cylindrical. Whereas, the volume and total weight of the reactor were too large to meet the ships requirement. After the design and optimization of the reactors shielding system by M-C method, stainless steel and B-polyethylene were chosen as shielding materials, and square pool was substituted by compact cylindrical pool. The result shows that on the premise of guaranteeing safety, shielding systems mass and volume were cut down greatly, i. e., this design can meet the requirement of nuclear medical ship.
2017,
29: 016011.
doi: 10.11884/HPLPB201729.160238
Abstract:
Compared with conservative accident analysis method, best estimate plus uncertainty (BEPU) accident analysis method can gain realistic analysis results and safety margins, and improve the nuclear power plant economy and flexibility of operation with the reasonable security assurance. The rod ejection accident (REA) analysis models under the condition of hot full power (HFP) and hot zero power (HZP) were established by using the best estimate code RELAP5-3D according to the design features of CNP600. The dominant phenomena and processes were identified through REA phenomena identification and ranking table (PIRT), and the important input parameters which have significant impacts on the key safety parameters were selected. The DAKOTA code was used to generate the samples of the important uncertainty input parameters by taking the Latin hypercube sampling (LHS) approach and the one sided tolerance upper limits of the key safety parameters were calculated through non-parametric method. In both cases the results obtained show that the maximum average fuel pellet enthalpy, peak fuel pellet temperature, peak cladding temperature, peak system pressure during accident transient meet the REA acceptance criteria. In both cases the one sided tolerance upper limits of peak nuclear power obtained by the non-parametric method are reasonable, and the one sided tolerance upper limits of maximum average fuel enthalpy obtained have considerable safety margins compared with the traditional REA conservative analysis values.
Compared with conservative accident analysis method, best estimate plus uncertainty (BEPU) accident analysis method can gain realistic analysis results and safety margins, and improve the nuclear power plant economy and flexibility of operation with the reasonable security assurance. The rod ejection accident (REA) analysis models under the condition of hot full power (HFP) and hot zero power (HZP) were established by using the best estimate code RELAP5-3D according to the design features of CNP600. The dominant phenomena and processes were identified through REA phenomena identification and ranking table (PIRT), and the important input parameters which have significant impacts on the key safety parameters were selected. The DAKOTA code was used to generate the samples of the important uncertainty input parameters by taking the Latin hypercube sampling (LHS) approach and the one sided tolerance upper limits of the key safety parameters were calculated through non-parametric method. In both cases the results obtained show that the maximum average fuel pellet enthalpy, peak fuel pellet temperature, peak cladding temperature, peak system pressure during accident transient meet the REA acceptance criteria. In both cases the one sided tolerance upper limits of peak nuclear power obtained by the non-parametric method are reasonable, and the one sided tolerance upper limits of maximum average fuel enthalpy obtained have considerable safety margins compared with the traditional REA conservative analysis values.
2017,
29: 016012.
doi: 10.11884/HPLPB201729.160243
Abstract:
The three dimension multi-group neutron transport Monte-Carlo program MCMG-Ⅱ and the multi-group point burnup program STEP1.0 based on linear chain method are adopted in the transport-burnup coupling system. The coupling information between the burnup and transport program includes neutron spectrum, source strength, reaction rates, initial nuclide density and its variation and the fission products. The interpreted script language Python is used to implement the coupling system. The coupling information can be transferred accurately and smoothly between the programs by its powerful text processing functions. It can also accomplish the step by step transport-burnup calculations automatically according to the total power, radiation time and timing step numbers with its interpreted, interactive characters and perform the graphic processing of the results. The encapsulation and partitioning strategy has been used in the architecture design of the system. The coupling system can do the high fidelity burnup calculations with more comprehensive reference to the factors such as geometry, neutron spectrum and radiation time automatically.
The three dimension multi-group neutron transport Monte-Carlo program MCMG-Ⅱ and the multi-group point burnup program STEP1.0 based on linear chain method are adopted in the transport-burnup coupling system. The coupling information between the burnup and transport program includes neutron spectrum, source strength, reaction rates, initial nuclide density and its variation and the fission products. The interpreted script language Python is used to implement the coupling system. The coupling information can be transferred accurately and smoothly between the programs by its powerful text processing functions. It can also accomplish the step by step transport-burnup calculations automatically according to the total power, radiation time and timing step numbers with its interpreted, interactive characters and perform the graphic processing of the results. The encapsulation and partitioning strategy has been used in the architecture design of the system. The coupling system can do the high fidelity burnup calculations with more comprehensive reference to the factors such as geometry, neutron spectrum and radiation time automatically.
2017,
29: 016013.
doi: 10.11884/HPLPB201729.160260
Abstract:
A tightly coupled multi-physics model for molten salt reactor(MSR) system involving the reactor core and the rest of the primary loop has been developed and employed in an in-house developed computer code TANG-MSR. In this paper, the computer code is used to simulate the behavior of steady state and transient state of our redesigned thorium molten salt reactor(TMSR). The simulation results demonstrate that the models employed in TANG-MSR can capture major physics phenomena in MSR and the redesigned TMSR has excellent performance of safety and sustainability.
A tightly coupled multi-physics model for molten salt reactor(MSR) system involving the reactor core and the rest of the primary loop has been developed and employed in an in-house developed computer code TANG-MSR. In this paper, the computer code is used to simulate the behavior of steady state and transient state of our redesigned thorium molten salt reactor(TMSR). The simulation results demonstrate that the models employed in TANG-MSR can capture major physics phenomena in MSR and the redesigned TMSR has excellent performance of safety and sustainability.
2017,
29: 016014.
doi: 10.11884/HPLPB201729.160261
Abstract:
Core layout and depletion features of thorium molten salt reactor(TMSR) are analyzed using the self-developed lattice and core code SONG/TANG-MSR. Based on the results of previous lattice optimization research, fuel salt that contains no BeF2 is adopted, BeO is introduced as moderator instead of graphite, and SiC is chosen as cladding material. According to further lattice analysis at core level, the optimized three-zone layout of core is obtained, with high breeding ratio, negative power coefficient, flat temperature distribution. Thereafter, the depletion calculation and analysis is performed considering the online processing of fuel salt. The results show that the core has high breeding ratio, short doubling time, and long term stable operation. Thus, the design of TMSR has been greatly improved with increased breeding ability and sufficient safety margins.
Core layout and depletion features of thorium molten salt reactor(TMSR) are analyzed using the self-developed lattice and core code SONG/TANG-MSR. Based on the results of previous lattice optimization research, fuel salt that contains no BeF2 is adopted, BeO is introduced as moderator instead of graphite, and SiC is chosen as cladding material. According to further lattice analysis at core level, the optimized three-zone layout of core is obtained, with high breeding ratio, negative power coefficient, flat temperature distribution. Thereafter, the depletion calculation and analysis is performed considering the online processing of fuel salt. The results show that the core has high breeding ratio, short doubling time, and long term stable operation. Thus, the design of TMSR has been greatly improved with increased breeding ability and sufficient safety margins.
2017,
29: 016015.
doi: 10.11884/HPLPB201729.160264
Abstract:
Based on comprehensive screening, lattice parameters including fuel composition, moderator material, lattice size and structure are redesigned for thorium molten salt reactor (TMSR) applying SONG/TANG-MSR codes. Innovatively, a fuel composition without BeF2 is adopted to increase the solubility for actinides (ThF4, UF4), and BeO is introduced as moderator to improve neutron economy. Moreover, SiC cladding is employed to separate fuel and moderator. Meanwhile such ceramic cladding contributes to maintain structure stability and improve resistance to radiation. With these lattice parameters, TMSR can achieve a high breeding ratio and maintain a negative power coefficient as well. Based on this lattice design, TMSR can have excellent performance of safety and sustainability.
Based on comprehensive screening, lattice parameters including fuel composition, moderator material, lattice size and structure are redesigned for thorium molten salt reactor (TMSR) applying SONG/TANG-MSR codes. Innovatively, a fuel composition without BeF2 is adopted to increase the solubility for actinides (ThF4, UF4), and BeO is introduced as moderator to improve neutron economy. Moreover, SiC cladding is employed to separate fuel and moderator. Meanwhile such ceramic cladding contributes to maintain structure stability and improve resistance to radiation. With these lattice parameters, TMSR can achieve a high breeding ratio and maintain a negative power coefficient as well. Based on this lattice design, TMSR can have excellent performance of safety and sustainability.
2017,
29: 016016.
doi: 10.11884/HPLPB201729.160306
Abstract:
In order to simulate the complex structure fuel assembly, Nuclear Power Institute of China(NPIC) has developed an advanced neutron transport lattice code, named KYLIN-2. The subgroup method is adopted in KYLIN-2 to treat resonance problems. And method of characteristics(MOC) is used to solve neutron transport equation and generalized coarse mesh finite difference(GCMFD) method is used to accelerate the calculation. In order to solve the depletion equation, CRAM method and PPC method are used. The graphical input and display interface are developed to make sure the KYLIN-2 code can be used easily by engineers. The numerical results show that the KYLIN-2 can calculate the PWR assembly accurately.
In order to simulate the complex structure fuel assembly, Nuclear Power Institute of China(NPIC) has developed an advanced neutron transport lattice code, named KYLIN-2. The subgroup method is adopted in KYLIN-2 to treat resonance problems. And method of characteristics(MOC) is used to solve neutron transport equation and generalized coarse mesh finite difference(GCMFD) method is used to accelerate the calculation. In order to solve the depletion equation, CRAM method and PPC method are used. The graphical input and display interface are developed to make sure the KYLIN-2 code can be used easily by engineers. The numerical results show that the KYLIN-2 can calculate the PWR assembly accurately.
2017,
29: 016017.
doi: 10.11884/HPLPB201729.160309
Abstract:
The mechanical shim (MSHIM) operation mode and the load following technology of the large-scale passive advanced PWR core use the relevant design of the AP1000 nuclear power plant as reference, but differ from it due to its higher power and unique control rod loading pattern. Therefore, it is extremely necessary to analyze the applicability of evaluation code for the MSHIM operation mode, the performance of the MSHIM load following operation mode, the characteristics of the start-up and restart processes based on the design of large-scale passive advanced PWR. In this paper, the applicability of MSHIM operation mode for the AP1000 calculation code to the large-scale passive advanced PWR is firstly evaluated based on the analysis of the methodology for the PWRs MSHIM operation mode and the analysis of the difference between AP1000 and the large-scale passive advanced PWR. Secondly, the typical 100%-70%-100% and 100%-70%-100% load following operation modes of both the initial and the equilibrium cycles are calculated and analyzed based on the neutronics modeling of the large-scale passive advanced PWR core. In addition, the operation capability of the initial start-up and restart processes in MSHIM mode of the initial and equilibrium cycles are simulated and analyzed quantitatively. The numerical results show that MSHIM mode has the certain ability for load following in both initial and equilibrium cycles without adjusting the boron concentration, but for start and restart conditions it is required to adjust the concentration of boron.
The mechanical shim (MSHIM) operation mode and the load following technology of the large-scale passive advanced PWR core use the relevant design of the AP1000 nuclear power plant as reference, but differ from it due to its higher power and unique control rod loading pattern. Therefore, it is extremely necessary to analyze the applicability of evaluation code for the MSHIM operation mode, the performance of the MSHIM load following operation mode, the characteristics of the start-up and restart processes based on the design of large-scale passive advanced PWR. In this paper, the applicability of MSHIM operation mode for the AP1000 calculation code to the large-scale passive advanced PWR is firstly evaluated based on the analysis of the methodology for the PWRs MSHIM operation mode and the analysis of the difference between AP1000 and the large-scale passive advanced PWR. Secondly, the typical 100%-70%-100% and 100%-70%-100% load following operation modes of both the initial and the equilibrium cycles are calculated and analyzed based on the neutronics modeling of the large-scale passive advanced PWR core. In addition, the operation capability of the initial start-up and restart processes in MSHIM mode of the initial and equilibrium cycles are simulated and analyzed quantitatively. The numerical results show that MSHIM mode has the certain ability for load following in both initial and equilibrium cycles without adjusting the boron concentration, but for start and restart conditions it is required to adjust the concentration of boron.
2017,
29: 016018.
doi: 10.11884/HPLPB201729.160337
Abstract:
The development of advanced thorium-based nuclear system raises new requirements on nuclear data. The multi-group data file of critical nuclides in the thorium-uranium recycle is the foundation of physical design, analysis and calculation of the reactor core. Based on authoritative nuclear data processing code NJOY, this paper obtains a WIMS format multi-group cross section data files through processing the ENDF/B-VII.1 evaluation nuclear data file, uses the specific update maintenance procedure WILLIE to get a WIMS format data file, and conducts a series of critical benchmarks on the data file using the multi-group reactor core calculation code WIMSD5B. The results show that the computed results of the WIMS file based on the processing of ENDF/B-VII.1 are basically the same as those of the latest WIMS-D file published on the websites of the WIMS-D library updating project (WLUP) with higher accuracy and reliability than those of the shipped WIMS-D file of the WIMSD5B code. Furthermore, the average deviation of the new WIMS file performing in the validation of 16 thorium-uranium cycle benchmarks is 0.225 3% smaller than that of the old WIMS file.
The development of advanced thorium-based nuclear system raises new requirements on nuclear data. The multi-group data file of critical nuclides in the thorium-uranium recycle is the foundation of physical design, analysis and calculation of the reactor core. Based on authoritative nuclear data processing code NJOY, this paper obtains a WIMS format multi-group cross section data files through processing the ENDF/B-VII.1 evaluation nuclear data file, uses the specific update maintenance procedure WILLIE to get a WIMS format data file, and conducts a series of critical benchmarks on the data file using the multi-group reactor core calculation code WIMSD5B. The results show that the computed results of the WIMS file based on the processing of ENDF/B-VII.1 are basically the same as those of the latest WIMS-D file published on the websites of the WIMS-D library updating project (WLUP) with higher accuracy and reliability than those of the shipped WIMS-D file of the WIMSD5B code. Furthermore, the average deviation of the new WIMS file performing in the validation of 16 thorium-uranium cycle benchmarks is 0.225 3% smaller than that of the old WIMS file.
2017,
29: 016019.
doi: 10.11884/HPLPB201729.160334
Abstract:
Radiation heating rate is one of the important input data for the performance research, thermal hydraulic and mechanical analysis of in-core special materials such as gray rods, therefore evaluation of radiation heating rate of in-core special materials is important for the research of new in-core materials. In order to improve the calculation precision of neutron and photon radiation heating rates, we developed a source code for a 3D Monte Carlo(MC) code, produced cross sections of special materials, established CAP1400 model, simulated and calculated radiation heating rate of special materials in CAP1400 by using MC code. The MC code calculates the radiation heating rate of special materials in CAP1400 more accurately, and it can give radiation heating rate distribution, which supports domestic design of in-core special materials such as gray rods and related components.
Radiation heating rate is one of the important input data for the performance research, thermal hydraulic and mechanical analysis of in-core special materials such as gray rods, therefore evaluation of radiation heating rate of in-core special materials is important for the research of new in-core materials. In order to improve the calculation precision of neutron and photon radiation heating rates, we developed a source code for a 3D Monte Carlo(MC) code, produced cross sections of special materials, established CAP1400 model, simulated and calculated radiation heating rate of special materials in CAP1400 by using MC code. The MC code calculates the radiation heating rate of special materials in CAP1400 more accurately, and it can give radiation heating rate distribution, which supports domestic design of in-core special materials such as gray rods and related components.
2017,
29: 016020.
doi: 10.11884/HPLPB201729.160336
Abstract:
Compared with the 18-month refueling, 24-month fuel cycle can reduce the times of overhaul, improve the load factor, and increase generating capacity. The core cycle length in a 177 assembly PWR meets 24-month refueling requirements by increasing the enrichment of the fresh fuel assemblies and the refueling fuel assembly components, considering the nominal and actual 24-month refueling. The economic evaluation was performed based on fuel management results, considering power generation revenue and cost including fuel cycle cost, refueling and overhaul cost, spent fuel storage costs. Preliminary results show that the 177-assembly PWR with 88 fresh fuel assemblies enriched at 4.95% is possible to achieve nominal 24-month refueling, and the average discharge burnup of fuel is about 48 GWd/tU. The cycle length with 104 fuel assemblies can achieve the actual 24 refueling cycle length, and neutron parameters meet the relevant safety limit requirements. It is shown that the 24-month refueling in the 177 assembly PWR is feasible, and its economy in the case of the high load factor is approximately similar with that of 18 months fuel cycle.
Compared with the 18-month refueling, 24-month fuel cycle can reduce the times of overhaul, improve the load factor, and increase generating capacity. The core cycle length in a 177 assembly PWR meets 24-month refueling requirements by increasing the enrichment of the fresh fuel assemblies and the refueling fuel assembly components, considering the nominal and actual 24-month refueling. The economic evaluation was performed based on fuel management results, considering power generation revenue and cost including fuel cycle cost, refueling and overhaul cost, spent fuel storage costs. Preliminary results show that the 177-assembly PWR with 88 fresh fuel assemblies enriched at 4.95% is possible to achieve nominal 24-month refueling, and the average discharge burnup of fuel is about 48 GWd/tU. The cycle length with 104 fuel assemblies can achieve the actual 24 refueling cycle length, and neutron parameters meet the relevant safety limit requirements. It is shown that the 24-month refueling in the 177 assembly PWR is feasible, and its economy in the case of the high load factor is approximately similar with that of 18 months fuel cycle.
2017,
29: 016021.
doi: 10.11884/HPLPB201729.160353
Abstract:
By the linearization method, the mathematic model of point reactor can be simplified. The simplified point reactor model could be further converted to the transfer function with the Laplace transfer tool. Until now, the point reactor transfer function is still used as the basic means for researching the nuclear reactor control system and the relative engineering items. However, through the analysis, it is found that the linearizd point reactor model based on the hypothesis that reactivity is zero when reactor is at the stable state has its own problems. This paper gives the concise analysis of the problems. The analysis method is the combination of theory and experiment. Theoretically, the focus is to compare the time domain results and the frequency domain results by means of the system dynamics principles. Experimentally, based on the engineering parameters, the focus is to compare the time domain computer simulation results and the frequency domain computer simulation results. Through the comparison analysis, the problems are exposed clearly. The paper gives the model modification access for the problems accordingly.
By the linearization method, the mathematic model of point reactor can be simplified. The simplified point reactor model could be further converted to the transfer function with the Laplace transfer tool. Until now, the point reactor transfer function is still used as the basic means for researching the nuclear reactor control system and the relative engineering items. However, through the analysis, it is found that the linearizd point reactor model based on the hypothesis that reactivity is zero when reactor is at the stable state has its own problems. This paper gives the concise analysis of the problems. The analysis method is the combination of theory and experiment. Theoretically, the focus is to compare the time domain results and the frequency domain results by means of the system dynamics principles. Experimentally, based on the engineering parameters, the focus is to compare the time domain computer simulation results and the frequency domain computer simulation results. Through the comparison analysis, the problems are exposed clearly. The paper gives the model modification access for the problems accordingly.
2017,
29: 016022.
doi: 10.11884/HPLPB201729.160391
Abstract:
The boric acid concentration measured from boron meter and chemical analysis deviates from each other in the process of reactor unit operation. In this paper, we use the experimental data from QSⅡ nuclear power plant to analyze the influence of 10B abundance on the deviation. We find out that the deviation will reach a maximum value during the MOL, and some advice on the operation of the plant is proposed according to this phenomenon. Whats more, we calculate the boric acid concentration through ORIENT and verify its reliability by comparison with chemical analysis. As a consequence, we can predict and monitor the changes of boric acid concentration in the core of reactor more accurately.
The boric acid concentration measured from boron meter and chemical analysis deviates from each other in the process of reactor unit operation. In this paper, we use the experimental data from QSⅡ nuclear power plant to analyze the influence of 10B abundance on the deviation. We find out that the deviation will reach a maximum value during the MOL, and some advice on the operation of the plant is proposed according to this phenomenon. Whats more, we calculate the boric acid concentration through ORIENT and verify its reliability by comparison with chemical analysis. As a consequence, we can predict and monitor the changes of boric acid concentration in the core of reactor more accurately.
2017,
29: 016023.
doi: 10.11884/HPLPB201729.160411
Abstract:
This paper studies the performance analysis and design consideration of the supercritial water reactor(SCWR) fuel assembly(FA). The SCWR fuel design goal is also studied. A comparative analysis is made for different fuel assembly design on the neutronics/thermal- hydraulics performance and the implementability of structure. The study shows that the square FA with big water rod adopting the simple structure of combined square FA with single water rod can provide not only sufficient and uniform moderation of fuel but also a simple split-flow of the moderator and coolant. This design shows a good neutronics and thermal-hydraulics performance and implementability, which is a good choice for SCWR fuel assembly design.
This paper studies the performance analysis and design consideration of the supercritial water reactor(SCWR) fuel assembly(FA). The SCWR fuel design goal is also studied. A comparative analysis is made for different fuel assembly design on the neutronics/thermal- hydraulics performance and the implementability of structure. The study shows that the square FA with big water rod adopting the simple structure of combined square FA with single water rod can provide not only sufficient and uniform moderation of fuel but also a simple split-flow of the moderator and coolant. This design shows a good neutronics and thermal-hydraulics performance and implementability, which is a good choice for SCWR fuel assembly design.
2017,
29: 016024.
doi: 10.11884/HPLPB201729.160454
Abstract:
The JMCT software is used to calculate the gamma-ray response functions for NaI crystal. The complete process of all the photon-electron coupled transportation is simulated by JMCT. The gamma-ray energy deposition spectrum and detector response functions are obtained and discussed. It is found that the results calculated by JMCT are the same as the results calculated by MCNP, and correspond well to the results calculated by Berger, which shows that the detector response functions simulated by JMCT are correct and satisfying. JMCT will play a more and more important role in the field of the experimental nuclear physics and nuclear analysis technology.
The JMCT software is used to calculate the gamma-ray response functions for NaI crystal. The complete process of all the photon-electron coupled transportation is simulated by JMCT. The gamma-ray energy deposition spectrum and detector response functions are obtained and discussed. It is found that the results calculated by JMCT are the same as the results calculated by MCNP, and correspond well to the results calculated by Berger, which shows that the detector response functions simulated by JMCT are correct and satisfying. JMCT will play a more and more important role in the field of the experimental nuclear physics and nuclear analysis technology.
