核动力工程

Extended Low Power Operation Justification of Two Typical First Cycle for CPR1000

Bai Chengfei, Wang Xinxin, Cai Dechang

2014, 35(S2): 1-3. doi: 10.13832/j.jnpe.2014.S2.0001

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Abstract:
The ability of extended low power operation（ELPO）of two typical first cycles（with Pyrex or gadolinium） for CPR1000 are analyzed and demonstrated respectively. The results showed that with the appropriate operation limits, both of the two typical first cycles are feasible for ELPO.

Analysis of Reactor Physical Start-up Based on Neutron Multiplication Theory

Hao Jianli, Chen Wenzhen, Wang Shaoming

2014, 35(S2): 4-7. doi: 10.13832/j.jnpe.2014.S2.0004

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Based on the neutron multiplication theory, the physical start-up model is established. The processes of lifting control rod discontinuously are calculated by this model, and the results are compared with the analytic solution of the point-reactor kinetics. Based on this, the effects of sub-criticality, lifting control rod durations and speeds on the physical start-up are investigated, and it shows that the sub-criticality is an importance parameter in the physical start-up. The conclusions are very important to the safety analysis and operation of the ship nuclear reactors.

Calculation of Protection Parameter of Linear Power Density in Tianwan Nuclear Power Station

Li Youyi, Yao Jinguo, Li Zaipeng, Yang Xiaoqiang, Wang Han, Dong Chao, Ye Liusuo

2014, 35(S2): 8-11. doi: 10.13832/j.jnpe.2014.S2.0008

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Fifty-four neutron-temperature measuring channels（NTMCs） of in-core instrument system（ICIS） in Tianwan nuclear power station（TNPS） are divided into 4 groups, and the self-power nuclear detectors（SPNDs） current in channels of each group can be converted to power. The in-core power field can be obtained by extended calculation. Parameters such as the coefficient of converting current into power are calculated by ICIS server computer of upper level and transferred to ICIS server computer of lower level. The maximum linear power density of each fuel assembly（FA） is weighted average of linear power densities calculated by four NTMCs in surrounding impact area. The weight coefficient is relative to the distance from SPND to FA in surrounding impact area. This paper expounds the calculation method on protection parameter of linear power density by SPND’s current. This method is simple and timely response, and the error is less than 5.7%. It has been successfully used in the real-time online protection of TNPS operation units.

Application of Regularized Radial Basis Function Neural Network in Core Axial Power Distribution Reconstruction

PenG Xingjie, YinG Dongchuan, Li Qing, Wang Kan

2014, 35(S2): 12-15. doi: 10.13832/j.jnpe.2014.S2.0012

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The core axial power shape reconstruction method based on the regularized radial basis function neural network（r RBFNN） is proposed, and the axial power distribution can be reconstructed from 6-segment ex-core detector signals. The 7740 axial power distributions of ACP-100 modular reactor and the corresponding simulated ex-core detector readings are used to verify the r RBFNN reconstruction method. The results show that r RBFNN reconstruction method can reconstruct the core axial power distribution accurately, and this method is robust enough to deal with the inherent ill-posedness of the power distribution reconstruction problem.

Research of Neutron Time Characteristics of Subcritical Device

Bai Yun, Yang Bo, Gong Jian, PenG Xianjue

2014, 35(S2): 16-18. doi: 10.13832/j.jnpe.2014.S2.0016

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In the study of the method to diagnose the prompt neutron decay constant of the subcritical device by the Dense Plasma Focus source, the different neutron time characteristics are found in the subcritical device with different reflective layer, such as a metal reflector and a hydrogenous reflector. The different neutron time characteristics like k_eff, prompt neutron generation time, energy spectrum, and the different of prompt neutron decay constant within different energy segments and different geometric segments is analyzed by the Monte Carlo method. The results showed that, in the subcritical device with a hydrogenous reflective layer, within 10-6 s time scale, the leakage fast signal of the device can fully reflects the characteristics of the active region. In 10-6 ~10-4 s or much longer time, the shape of the leakage neutron spectrum and space spectrum may fully reflects the whole device.

Improvement of Characteristic Statistic Algorithm Reloading Optimization Program CSA for Actual Engineering Requirements

Liu Zhihong, Zhao Jing, SHi Xiuan, Zhang Ming, Gao Wei, Cai Dechang, Peng Lianghui

2014, 35(S2): 19-22. doi: 10.13832/j.jnpe.2014.S2.0019

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In order to use the characteristic statistic algorithm fuel management optimization program CSA in the actual nuclear power plant reloading design, the improvement of some special requirements in the practical engineering problems is needed. This paper introduces how to improve the original CSA optimization program for Ling’ao nuclear power plant and the validation results of improved program in the actual nuclear power plant reloading design problems. The problems include a core without burnable poison, a core with burnable poison, and an equilibrium-cycle core reloading design. The final results proved that after the corresponding improvement, the CSA program can be used to do reloading optimization for the actual reactors in the nuclear power plants.

Investigation of Transmutation Paths for Production of ²⁵²Cf in Reactors

Wu Mingyu, Zhang Qiang, Wang Shixi, Yang Yong, Yang Jiayin, Wang Jing

2014, 35(S2): 23-26. doi: 10.13832/j.jnpe.2014.S2.0023

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The generation and transmutation process of the transplutoniums in the conversion chains for 252 Cf production is calculated and analyzed. The production and conversion efficiency of the chains with different target nuclides and neutron spectrums including fast, thermal and moderated neutron spectrums is computed. The efficiency and the advantages and disadvantages of the conversion paths for 252 Cf under different neutron spectrums and neutron flux levels are analyzed. The multi-group depletion program STEP1.0 based on linear chain method is used to calculate the nuclide densities. The differences of the conversion efficiency caused by the neutron spectrum can be qualified more accurately with the multi-group cross section calculations. The higher conversion efficiency than the typical thermal neutron spectrum can be obtained by some moderated spectrum optimized for absorption reaction through the irradiation of some transplutoniums（242Pu, 244 Cm, 246Cm）.

Research of Neutron Fluence Computation Methods for MOX Fuelled Reactor Core

Tang Songqian, Tan Yi, Ying Dongchuan, Wei Shuping

2014, 35(S2): 27-30. doi: 10.13832/j.jnpe.2014.S2.0027

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Changes in the mechanical properties of reactor vessel materials result from the exposure to the fast neutron. The use of MOX fuel in LWRs presents different neutron characteristics, and it is worthy to study whether the present software can calculate the structural integrity of reactor components. This paper use MCNP, TORT and SCALE to calculate VENUS-2 benchmark, and the calculation shows that all this software can get reasonable result that can be used in the design. MCNP has the highest accuracy.

Neutronics and Accident Analysis of Th-U-MOX Loading in Pebble Bed High-temperature Reactors

Xia Bing, Li Fu, Wei Chunlin, Chen Fubing, Xu Xiaolin, Jing Yingqing

2014, 35(S2): 31-33. doi: 10.13832/j.jnpe.2014.S2.0031

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The neutronics and accidental features of HTR-PM with Th-U-MOX fuel loading are investigated. The parametric analysis is performed on the uranium fraction in MOX fuel and the C/HM ratio, for the optimization of Th-U-MOX loading. It is revealed that the requirement of natural uranium per energy production decreases when the C/HM ratio decreases. The optimized scheme reduces the uranium requirement for 8.5% and provides thorium utilization as low as 6%.

Burnup Calculations and Fuel Cycle Economics Analysis of Fusion-Fission Hybrid Reactor

Zu Tiejun, Wu Hongchun, Zheng Youqi, Cao Liangzhi

2014, 35(S2): 34-37. doi: 10.13832/j.jnpe.2014.S2.0034

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Based on the light water cooled pressure tube blanket, the refueling strategies have been established using the spent nuclear fuel discharged from pressurized water reactor and natural uranium oxide, and detailed burnup calculations have been carried out. The numerical results show that only about 80 tonnes of spent nuclear fuel or natural uranium are needed every 1500 EFPD（Equivalent Full Power Day） with a 3000 MWth output and tritium self-sufficiency（TBR＞1.20）. Based on the equilibrium fuel cycle, the fuel costs of electricity of the blanket using different fuel are calculated, and the results are 1.82×10-3$/（k W·h） for the case of spent nuclear fuel, and 1.35×10-3$/（k W·h） for the case of using natural uranium, respectively.

Analysis of Neutronics Characteristics of Small Modular Pb-Bi Cooled Reactor Core with Nitride Fuel

Yuan Xianbao, Cao Liangzhi

2014, 35(S2): 38-40. doi: 10.13832/j.jnpe.2014.S2.0038

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Based on the extensive investigation of small modular reactors in the world, a small modular Pb-Bi cooled reactor with nitride nuclear fuel（SMPBN） which is designed to meet the requirements for nuclear energy improvement is presented, and the neutronics characteristics of its core is analyzed in detail. The analysis concludes the following features: the bottleneck problem of the limited uranium resources in the nuclear power development can be fixed by using the spent-plutonium as the driven fuel and thorium as the fertile fuel; the using of Pu N and Th N as the fuels can improve the safety of the reactors and increase the breeding ratio; the liquid lead-bismuth selected as the coolant and reflector not only improves the ability of the natural circulation, but also enhances the reactor safety and improves the economy of neutron; the burn-up swing is almost zero in the whole lifetime due to the perfect inner breeding ability, and all of the important reactivity coefficients are negative to assure SMPBN inherent passive safety.

Conceptual Design of Fast Reactors Based on Once-Through and Closed Fuel Cycles

ZhenG Youqi, Wu Hongchun

2014, 35(S2): 41-43. doi: 10.13832/j.jnpe.2014.S2.0041

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Fast reactor is important for the sustainable development of nuclear energy due to its good breeding ability. In this paper, two conceptual designs of fast reactor using metal fuels are proposed, respectively, based on the once-through and closed fuel cycle. For the case of the closed fuel cycle, the reactor is designed to reach the breeding ratio of 1.4 with the double time of 11 years. Meanwhile, the case of once-through fuel cycle is designed to operate with long cycle. The in-core breeding and burn is realized by using the fuel shuffling for 38 years. Based on the two cases, the differences of designing fast reactor for different fuel cycle are compared in the view of core physical parameters, resource use ratio and economics.

Study on Long-Life Sodium Cooled Fast Reactor Concept with Radial Shuffling

Li Zhipeng, ZhenG Youqi, Cao Liangzhi

2014, 35(S2): 44-47. doi: 10.13832/j.jnpe.2014.S2.0044

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This paper proposes a core shuffling design for a long-life sodium cooled fast reactor. Based on the breed and burn strategy and by means of radial shuffling, the reactor can sustain critical for a long time without refueling. In this design concept, one-through fuel cycle is adopted for nuclear non-proliferation, heterogeneous core is considered to flatten the local power peak and to enhance the in-core breeding, inward-convergent shuffling is employed to extend the core life. Preliminary calculations had proved that the shuffling scheme is feasible. The core life can reach over 38 years and the discharged breed assemblies can be used as the driver fuel in another reactor. The key parameters of the core are within the acceptable range in current fast reactor designs.

Analysis Method of Dynamic Characteristics of Subcritical Reactor of ADS Induced by Accelerator Beam Transients

Yu Tao, Xie Jinsen, Liu Zijing

2014, 35(S2): 48-51. doi: 10.13832/j.jnpe.2014.S2.0048

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Accelerator beam transients induced by the instability of the High Power Proton Accelerator（HPPA） will lead to the rapid intensity change of the spallation neutron source, which will impact the safety of subcritical reactor of ADS. The dynamic characteristics of the subcritical reactor in the beam transients is an important issue of ADS safety. In this paper, based on the brief summation of previous research works, the primary calculation and analysis related to beam transients have been carried out on the ADS verification facility "VENUS1#". Based on the results of "VENUS1#", the real-time calling multi-mode parameters in ADS beam transients analysis is suggested.

Sensitivity Analyses of Parameters in an Accelerator Driven Minor Actinide Burner System

ZhOu Shengcheng, Wu Hongchun, Zheng Youqi, Li Xunzhao

2014, 35(S2): 52-55. doi: 10.13832/j.jnpe.2014.S2.0052

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Sensitivity analyses of coolant, subcriticality, fuel assembly geometry and thermal power for accelerator driven systems（ADS） are carried out using a home-developed steady-state core analysis code named LAVENDER. The numerical results show that the optimization based on single design target（e.g. beam current, minor actinide transmutation rate or thermohydraulic safety margin） may lead to the deterioration of other design targets and trade-offs should be made to balance the different design targets during the design of ADS.

Study on Basic Neutron Physical Parameters of Thorium-Uranium Used in Pebble Bed Fluoride Salt-Cooled High Temperature Reactor

Zhu Guifeng, Zou Yang, Xu Hongjie

2014, 35(S2): 56-59. doi: 10.13832/j.jnpe.2014.S2.0056

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By analyzing the energy production per neutron, neutron cumulative production per uranium and neutron cumulative sales per thorium in unit cell of Pebble Bed Fluoride Salt-cooled High Temperature Reactor（PB-FHR）, some basic neutron physical parameters are determined. For the optimal choice, C/Th ratio is about 80, and the discharge burnup of thorium is between 140~200 MW·d/kg（Th）; C/U ratio is between 400~600, and the discharge burnup of uranium is between 180~200 MW·d/kg（U）. It is beneficial to set the core radial layout as thorium-uranium-thorium pattern.

Application of Three-Dimensional Discrete Ordinatesand Monte Carlo Coupling Method on Fast Neutron Fluence Rate Calculation

ZhenG Zheng, Wu Hongchun, Cao Liangzhi, Mei Qiliang, Li Hui

2014, 35(S2): 60-63. doi: 10.13832/j.jnpe.2014.S2.0060

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A three-dimensional discrete ordinates and Monte Carlo coupling method（the SN-MC coupling method） is applied on Pressurized Water Reactor（PWR） pressure vessel Fast Neutron Fluence rate（FNF） calculation, and the results of the SN-MC coupling method are compared with those of the discrete ordinates method（the SN method）, the Monte Carlo method（the MC method） and the measurements. Numerical results show that the difference between the calculated FNF by the SN-MC coupling method and the measured FNF are less than 20%, which satisfies the requirement of the difference for FNF calculated methods in Regulatory Guide（RG） 1.19 promulgated by Nuclear Regulatory Commission（NRC）. The speed of the SN-MC coupling method is 2~10 times faster than that of the MC method when the two methods obtain the same tally errors.

Calculation of Neutron Source Strength from Secondary Neutron Source in PWRs

JinG Futing, Xiao Feng, Liu Jiajia, Tan Yi, Lyu Huanwen

2014, 35(S2): 64-66. doi: 10.13832/j.jnpe.2014.S2.0064

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Based on the theory of neutron release from the secondary neutron source, the calculation method for the neutron source strength was built. MCNP was used to calculate the reaction rate and neutron flux in the secondary neutron source assembly, and the average radioactive capture cross section of 123 Sb and the transform coefficient from radioactive activity to neutron source strength were obtained.

Application of Monte Carlo Method in Fast Reactor Assembly Homogeneous Constant Calculation

Du Xianan, Wu Hongchun, ZhenG Youqi

2014, 35(S2): 67-70. doi: 10.13832/j.jnpe.2014.S2.0067

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In the fast reactors, the scattering resonance of intermediate weight nuclide and the space-coupling effects are significant. To solve these problems, this study uses Open MC based on Monte Carlo method to generate the few group constants and combines the deterministic method to do the core calculation. For this analysis, 2D R-Z homogeneous core model was used for the generation of the constants. The assembly calculation shows that the scattering resonance in high energy region can be well dealt with. Both k_eff and power distribution have a discrepancy of less than 1% from the reference solution. Therefore, this hybrid method is suitable for the fast reactor analysis.

Analysis of Sensitivity and Uncertainty of Calculated k_eff to Nuclear Data on China Experimental Fast Reactor

Yang Jun, Yu Hong, Xu Li, Hu Yun

2014, 35(S2): 71-75. doi: 10.13832/j.jnpe.2014.S2.0071

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In order to analyze the nuclear data uncertainty of k_eff for CEFR, the formula of sensitivity to k_eff was derived based on the first order perturbation theory, and the multi-group diffusion nodal method was selected to solve the formula. A sensitivity and uncertainty analysis code named SUAPH was developed, based on Neutronics Analysis System for Fast Reactor（NAS）, and verified. Based on the available covariance matrixes, the sensitivity and uncertainty analysis of k_eff of the first loading on CEFR was done, and the uncertainty value of calculated k_eff due to nuclear data was about 2.02%.

Mechanism Analysis of Difference between the Contribution of Nuclear Data to k_eff Uncertainty in HTR-10 and HTR-PM

Hao Chen, Guo Jiong, Li Fu, Wang Lidong

2014, 35(S2): 76-79. doi: 10.13832/j.jnpe.2014.0076

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Based on the TSUNAMI-3D-K5 model of SCALE6.1 code, the complete 3d models of sphere bed high temperature gas cool experimental reactor HTR-10 and sphere bed model high temperature gas cool reactor HTR-PM core have been established to carry out the research of keffuncertainty by using the 44 groups covariance matrix of nuclear data built-in the SCALE6.1. Through the uncertainty and sensitivity analysis of different nuclear reactions in HTR-10 and HTR-PM, the mechanism of difference between the contributions of nuclear data to the keffuncertainty has been understood. The results show that the difference in neutron leakage due to the difference between the core sizes, the difference between material content of 239 Pu and 235 U and the average fission neutrons are the main reason for the difference between the contributions of nuclear data to the uncertainty of k_eff.

Nuclear Data Requirement Analysis of TMSR Facility Basing on S/U Analysis

Wang Wenming, Zhang Huanyu, Wu Haicheng, Liu Ping

2014, 35(S2): 80-82. doi: 10.13832/j.jnpe.2014.S2.0080

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The sensitivity of k_eff of MSRE facility to the critical nuclear data is calculated by Monte Carlo perturbation method. Based on the multi-group covariance data library developed by China Nuclear Data Center, the uncertainty of nuclear data introduction is analyzed by S/U method, and the importance of the nuclear data is listed in order.

Preliminary Study of Sensitivity Analysis Based on Iterated Fission Probability Method

Qiu Yishu, Liang Jingang, Wang Kan

2014, 35(S2): 83-86. doi: 10.13832/j.jnpe.2014.S2.0083

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In order to investigate the analysis methodology for k_eff sensitivity to nuclear data in continuous-energy Monte Carlo codes, the theory about calculating the eigenvalue sensitivity coefficients to nuclear data is introduced. And the foundation of the Iterated Fission Probability（IFP） method which is widely used for calculating the adjoint flux in continuous-energy Monte Carlo codes and approaches for tallying the adjoint-weighted reaction rates are discussed. Finally, with the continuous-energy Reactor Monte Carlo code RMC, sensitivity analysis is performed on a polyethylene sphere criticality benchmark based on the IFP method and the results are compared to those calculated by SCALE.

Analysis of Eigenvalue Implicit Sensitivity Based on Subgroup Resonance Self-Shielding Technique

Liu Yong, Cao Liangzhi, Wu Hongchun, Zu Tiejun

2014, 35(S2): 87-90. doi: 10.13832/j.jnpe.2014.S2.0087

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The direct effect of the problem-specific multigroup cross sections on the eigenvalue, namely the explicit sensitivity, is obtained efficiently based on the classical perturbation. The expression of resonance self-shielding multigroup cross section sensitivity coefficients with respect to non-resonance nuclide cross sections are deduced based on the generalized perturbation theory for the subgroup resonance self-shielding technique. The indirect effect of non-resonance nuclide on the eigenvalue through resonance self-shielding procedure, namely the implicit sensitivity, can be calculated based on the foresaid two results. The relative importance of the implicit sensitivity is discussed in comparison with the explicit sensitivity.

Calculation of Depletion Sensitivity Coefficient Based on First-Order Generalized Perturbation Theory

Yang Chao, Cao Liangzhi, Wu Hongchun, Zu Tiejun

2014, 35(S2): 91-93. doi: 10.13832/j.jnpe.2014.S2.0091

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A code for calculating the depletion sensitivity coefficient of atomic densities is developed on the basis of a first-order generalized perturbation theory. The sensitivity coefficient of atomic density of 244 Cm at 50GWd/T due to the change of cross section is analyzed, and the results show that the sensitivities to the fission cross section of 235 U and 239 Pu and the capture cross section of 238 U, 240 Pu, 241 Pu, 242 Pu and 243 Am are large; the energy dependent sensitivity of 244 Cm to the capture of 243 Am in thermal and resonance regions is much larger than that in the fast region, therefore, the accuracy improvement in the thermal and resonance regions should take a higher-priority than that in the fast region.

Verification of CosMC Based on VENUS-2 Critical Benchmark

Hu Jiaju, Ma Xubo, Chen Yixue, Yu Hui, Quan Guoping

2014, 35(S2): 94-97. doi: 10.13832/j.jnpe.2014.S2.0094

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When a reactor core contains a large number of multiple cycles and higher burnup plutonium fuel, the core neutronics characteristics will change. To verify the accuracy of current database and existing nuclear software, OECD / NEA proposed VENUS-2 benchmark experiments. Cos MC is designed to perform the Monte Carlo reactor calculation that can handle the complex geometry. In this paper, the latest database and Cos MC is adopted to calculate VENUS-2 benchmark, and the results were compared with other software. The results show that: the calculation results of Cos MC agree well with the experimental results and the measured values. It is demonstrated that it is feasible for Cos MC to calculate the critical problem of the reactor core containing MOX fuel.

Verification and Validation of Lattice Code ROBIN-1.7

Chen Guohua, Huang Yong, Jiang Xiaofeng, Wang Tao, Zhang Shaohong

2014, 35(S2): 98-101. doi: 10.13832/j.jnpe.2014.S2.0098

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Verification and validation of lattice code ROBIN-1.7 is performed by using international public benchmark problems, which include critical experiments data, depletion benchmark problems of OECD NEA and other neutron transport-depletion calculations with MCNP code. The result（reactivity, pin power distribution, isotope concentration） shows that independent modules of ROBIN-1.7 such as resonance treatment module, neutron transport module and depletion module are developed and integrated correctly. It is also demonstrated that ROBIN-1.7 has reached the industry level for PWR application.

Validation of Core Analysis System ORIENT 1.0

Wang Tao, Jiang Xiaofeng, Lyu Dong, Chen Guohua, Huang Yong, Zhang Shaohong

2014, 35(S2): 102-104. doi: 10.13832/j.jnpe.2014.S2.0102

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Code validations are performed by comparing the calculated results against the measured data from three types of operating reactors. A total of 59 reactor cycles are evaluated and the results compared not only include the parameters measured at the startup physics test stage for zero- and low-power conditions but also that measured at the normal operation stage for full-power conditions. Validation results demonstrate that the neutronic models incorporated in ORIENT system are of high quality and the system is highly acceptable for performing routine neutronic analyses at PWR nuclear power plants.

Preliminary Study on Strategy of Verification of Reactor Monte Carlo Code CosMC

MA Xubo, Yao Yuan, Hu Jiaju, Wu Jun, Chen Yixue, Yu Hui, Quan Guoping

2014, 35(S2): 108-111. doi: 10.13832/j.jnpe.2014.S2.0108

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Software verification and validation is of great significance to ensure the software quality, reasonable and efficient. Verification and validation strategy is the key point of the verification and validation process. In this study, we proposed the PIRT table of the verification and validation for Cos MC. The reactor critical computing functions and the accuracy of the neutron flux calculated by Cos MC were verified by using of well-known international critical C5G7 benchmark. The results agreed well with the results calculated by MCNP.

SONG—Description of Multi-Functional Lattice Code

Si Shengyi, Chen Qichang, Zhao Jinkun, Bei Hua

2014, 35(S2): 112-115. doi: 10.13832/j.jnpe.2014.S2.0112

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SONG, a multi-functional lattice code focusing on the core analysis of next-generation reactor systems, adopted Method Of Characteristics（MOC） to solve the neutron transport equation with arbitrary geometry. It can simulate flexibly the two-dimensional fuel subassembly model or core model with rectangular lattice or hexagonal lattice, which consist of pin-type or plate-type fuel elements. To meet the demands of calculation on thorium-uranium and uranium-plutonium fuel cycle with extra-long lifetime and extra-deep burnup, the exponential matrix method is adopted to solve the problem of extended fission products and actinides chain. Verification shows that SONG satisfies the anticipated performance and function requirements.

Development of NECP-SARAX1.0 Code for Fast Reactor Physics Calculation

ZhenG Youqi, Wu Hongchun, NeCP team

2014, 35(S2): 116-118. doi: 10.13832/j.jnpe.2014.S2.0116

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A new code NECP-SARAX1.0 is developed for the nuclear design and steady analysis of the fast reactor core based on the special features of fast reactor physics. The continuous energy cross section library from ENDF/BVII is used. The OPENMC code is adopted to generate the multi-group cross sections. The SN-nodal method transport is applied in the core calculation based on the unstructured mesh modeling. It is suitable for both the critical and subcritical core calculation. The perturbation method is used to calculate the Doppler coefficients. Numerical results show that the new code is of high precision. Compared with direct MCNP calculation, the difference of k_eff is around 100×10^-5.

SONG—Development of Resonance Modules

Chen Qichang, Si Shengyi, Bei Hua, Zhao Jinkun

2014, 35(S2): 119-122. doi: 10.13832/j.jnpe.2014.S2.0119

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The development of lattice code SONG is mainly for the research and development of new type reactors. The resonance calculation modules in the SONG code are applicable for pin-type and plate-type fuel elements, including Stamm’ler method and Spatially Dependent Dancoff Method（SDDM）. The Stamm’ler method is a traditional resonance calculation method based on equivalence theory, which is simple and efficient but unable to treat the space-dependent resonance shielding. Based on Stamm’ler method, SDDM method evaluates further the escape probability of different regions, thus to estimate the space-dependent self-shielding effect. Current theory of Stamm’ler method and SDDM are both for pin-type fuel element. While in the SONG code, the methods were extended for plate-type fuel element, and general resonance modules were developed. Numerical results indicate that the resonance theoretical derivation is appropriate and the accuracy of resonance modules is satisfied.

Method for Resonance Elastic Scattering Correction in Deterministic Method

He Qingming, Cao Liangzhi, Wu Hongchun, Zu Tiejun

2014, 35(S2): 123-126. doi: 10.13832/j.jnpe.2014.S2.0123

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The asymptotic scattering kernel is applied in NJOY to solve the slowing down equation, where the resonance elastic scattering is neglected. This assumption contributes considerable errors to eigenvalue and Doppler Coefficient（DC）. To take this effect into account in deterministic method, the Doppler Broadening Rejection Correction（DBRC） method is employed to correct the free gas model of MCNP, which is used to generate the resonance integral tables instead of NJOY. The infinite medium multiplication factor and DC of Light Water Reactor（LWR） are analyzed based on the subgroup method and the results are compared with those of MCNP. The numerical results show that the correction method proposed in this paper can be used to consider the resonance elastic scattering and promote the precision of the deterministic method.

SONG—Development of Transport Modules

Chen Qichang, Si Shengyi, Zhao Jinkun, Bei Hua

2014, 35(S2): 127-130. doi: 10.13832/j.jnpe.2014.S2.0127

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Method Of Characteristics（MOC） and Coarse-Mesh Finite Difference（CMFD） were adopted for the transport and accelerate calculation in the SONG code, which is capable of direct lattice calculation in library group structure. The geometry modules of SONG code implement the lattice based modular ray tracing for hexagonal or square lattice with pin-type or plate-type fuel elements. By modular programming and scientific data structure designing, the geometrical independent transport module and expansible geometry modules were developed. Geometry modules for different shape lattices share the uniform data interface with the transport module. The numerical results indicate that SONG code has the ability of treating the multiform lattices, and is with well performance in terms of precision, efficiency and stability.

Development of 2D/1D Fusion Transport Code Based on Modularity MOC

Liang Liang, Wu Hongchun, ZhenG Youqi, Li Yunzhao

2014, 35(S2): 131-134. doi: 10.13832/j.jnpe.2014.S2.0131

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With a variety of new reactors being proposed, the demand for whole-core heterogeneous computing is increasingly urgent. Based on the needs and feasibility considerations, 2D/1D fusion method for solving the whole-core three-dimensional heterogeneous equation is raised. The international community has developed a number of such codes. This paper develops a 2D/1D fusion code, MOCHA₂D1D which is based on 2D modularity MOC, applying Sn difference method for 1D calculation. After verification, the accuracy meets the requirements of reactor physics calculation.

Tiger-3D: 2D/1D Coupled Whole-Core Transport Code Based on Large-Scale Parallel Computation

Wu Wenbin, Li Qing, Wang Kan

2014, 35(S2): 135-139. doi: 10.13832/j.jnpe.2014.S2.0135

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2D/1D coupled method can obtain good accuracy and efficiency for 3D transport equation. Generally, parallel features of axial planes and MOC angles are adopted in many codes. However, large-scale parallel computation is not realized. In this paper, 2D/1D coupled whole-core transport method was studied within the framework of 3D CMFD, in which spacial domain-decomposed matrix MOC was radially adopted while finite difference diffusion was axially adopted. By combining the parallel features of both directions of the 2D/1D coupled method, the code Tiger-3D with large-scale parallel computation was developed based on MPI. Numerical results demonstrated that Tiger-3D has good accuracy and high efficiency.

Single Assembly Pin-by-Pin Homogenization Model for Handling Axial 3D Heterogeneity Effect

Lyu Dong, Yu Lulin, Han Yu, Wang Dezhong

2014, 35(S2): 140-142. doi: 10.13832/j.jnpe.2014.S2.0140

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Abstract:
Traditional and advanced homogenization model based on single assembly and color-set/full core transport solution can not represent the axial heterogeneity of Light Water Reactor（LWR）. The paper proposed a new 3D homogenization model based on the single assembly pin-by-pin solution to provide the coarse mesh parameters for 3D nodal calculation. Numerical result from BWR mini-core benchmark shows that the 3D effect on core axial power distribution can be reproduced accurately.

Research of Ray Effect Mitigation Methods in Discrete Ordinates Method

Chen Mengteng, Zhang Bin, Chen Yixue, Hu Ye, Zhang Penghe, Zhao Jingchang, Zang Qiyong

2014, 35(S2): 143-146. doi: 10.13832/j.jnpe.2014.S2.0143

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When calculating the large volume of void regions and strongly absorbing media models by discrete ordinates method, the ray effect problems may arise, namely resulting in the flux spatial distortion such as the ripple like distribution. The nature and cause of ray effects are described and analyzed. Because the solution of the transport equation by the discrete ordinates method is conducted only along a few specific directions, and the original continuous angular variables are changed to the finite number of discrete directions, the quadrature sets are incapable of approximating the scalar flux from the discrete angular fluxes. Through research and analysis of the mainstream methods, the first collision source method is selected as major research direction. The multi-dimensional ray effect mitigation program RAY is developed, and the calculation of benchmark problem shows that RAY can mitigate the ray effects effectively.

Reconstruction and Parallelization of 3D SN Program for Neutron/Photon Transport

Cheng Tangpei, Lei Wei, Zhong Bin, Shen Huayun, Wei Junxia, Deng Li

2014, 35(S2): 147-150. doi: 10.13832/j.jnpe.2014.S2.0147

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The application of the neutron/photon transport simulation program based on 3D SN method is limited by the intensive computational requirement and prohibitive computational time. Based on the geometry domain partition method, this paper reports our efforts on the reconstruction and parallelization of the legacy SN program by redesigning the data structures, the functional module as well as the compute flow on J Adaptive Structured Meshes applications Infrastructure（JASMIN）. The correctness of the reconstructed program is proved by modeling the fixed-source and k-eigenvalue benchmark problems. In the strong scalability test, the parallel efficiency of the reconstructed program can reach up to 37.07% on 512 CPU cores when modeling the k-eigenvalue problem with 100 million degree-of-freedom.

Development of Burnup Program Based on Chebyshev Rational Approximation Method

Han Wenjing, Zhang Jingyu, Liu Jianzheng, Chen Yixue

2014, 35(S2): 151-154. doi: 10.13832/j.jnpe.2014.S2.0151

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A burnup code ABURN was developed based on Chebyshev rational approximation method（CRAM） and EAF database. This paper calculates the fusion reactor first wall and UO₂ burnup problems. Compared with the European FISPACT, ABURN can achieve the same accuracy as FISPACT. Due to the CRAM method, this program has a high flexibility in terms of burnup steps, which verifies the feasibility and accuracy of ABURN program.

Study and Verification of Burnup Calculation Strategy in Lattice Physics Code SimFA

Liu Tingting, Lin Xusheng, Yang Senquan, Xie Zhengquan, Luo Fanghui

2014, 35(S2): 155-158. doi: 10.13832/j.jnpe.2014.S2.0155

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Abstract:
Advanced analysis code on fuel assembly homogenization, Sim FA, has been developed by CNPO to improve the accuracy of few group cross section generation for the further use in the core code embedded in full-scope NPP control room simulators. In this paper, the burnup calculation strategy adopted in Sim FA, i.e. the Extended Predictor-Corrector Method incorporating the Linear Rate Method is presented, with the emphasis on Gadolinium-bearing assembly case. In comparison with the conventional burnup calculation methods, Sim FA demonstrates a better performance considering both the calculation efficiency and accuracy. The burnup calculation strategy implemented in Sim FA enables a fairly accurate burnup calculation with sufficiently large step size, and is therefore considered as a suitable way for the gadolinium-bearing assembly burnup calculation.

Study on Thermal Neutron Scattering Data for ~7LiH

Wang Jia, Song Hongzhou, Ye Tao, Hu Zehua, SuN Weili

2014, 35(S2): 159-163. doi: 10.13832/j.jnpe.2014.S2.0159

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Based on the theory of thermal neutron scattering, a new code named Sirius was developed to produce the thermal scattering law data. Frequency distributions of 7Li and 1H bound in 7Li H were calculated by first-principles with the frozen-phonon approach. These frequency distributions were utilized in Sirius code to get the thermal scattering law data for 7Li H. Theoretical analysis indicated that the thermal scattering law data is reasonable.

Effect of Neutron Absorber Layout Scheme for Spent Fuel Storage on Critical Safety

Zhao Jun, Han Song, Su Genghua, SHi Xiuan, Cai Dechang

2014, 35(S2): 164-166. doi: 10.13832/j.jnpe.2014.S2.0164

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Abstract:
Based on the burn-up credit method, the effect of the soluble boron concentration in the storage on k_eff was studied by APOLLO and MCNP5 program. The results showed that the change rate of k_eff resulted from the change of the concentration of soluble boron in the storage reduced with the increasing of the fuel enrichment, approximately in linear. Mutual interference effects of the neutron poison in the storage were the main contribution to the absorber value. The neutron poison value has a liner relationship with the distance of the BSS plate. The improved layout scheme of neutron poison in accordance with the energy spectrum distribution of neutron outside the spent fuel assemblies can improve the critical security and economy of the spent fuel storage systems.

Criticality Safety Design for High Density Spent Fuel Storage Rack

Yang Haifeng, Huo Xiaodong, Yi Xuan, SHao Zeng

2014, 35(S2): 167-169. doi: 10.13832/j.jnpe.2014.S2.0167

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Abstract:
Based on the design characteristics, operational strategies, as well as the complicated fuel assembly designs, factors leading to a harder neutron spectrum are considered and a combination of operational parameters is identified to obtain the conservative isotopic inventories. Criticality analysis models are built, and factors such as the end effect and the credible accident conditions are studied in detail. Finally, a primary（criticality） design of the high density spent fuel storage rack is developed which meets the most updated（criticality） criteria and can be used in NPP projects.

Development and Validation of Cylindrical Solution System Criticality Accident Analysis Program

Yu Miao, Huo Xiaodong, Liu Guoming, SHao Zeng, Yi Xuan, Yang Haifeng

2014, 35(S2): 170-172. doi: 10.13832/j.jnpe.2014.S2.0170

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Abstract:
Analysis and evaluation for criticality accident of the cylindrical solution system has important academic significance and value of the engineering. In order to realize the analysis and evaluation of this system, this paper studies the mechanism of the development of the nuclear criticality accidents, and developes a simulation and analysis program CAACS for the criticality accidents of the cylindrical solution system independently, which provides the technical means of accident analysis for the transformation and construction of subsequent commercial fuel reprocessing plants plant, and lays the foundation for subsequent critical transient studies.

Design and Processing of Multi-Group Cross Section Library for SONG

Bei Hua, Zhao Jinkun, Chen Qichang, Si Shengyi

2014, 35(S2): 173-175. doi: 10.13832/j.jnpe.2014.S2.0173

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Abstract:
Multi-functional lattice code SONG is now being developed, to meet the research demand of new type reactors. So it is necessary to consider the new specifications on these reactor in fuel, structure, coolant, modulator, spectrum and depletion depth, in building the new multi-group cross section library. In the presence of these new characteristics, the library is specially designed in dealing with burp-up chain, energy structure, reaction path, resonance parameters and so on. Using the Evaluated Nuclear Data File（ENDF）, the nuclide date processing code（NJOY）, nuclide date auxiliary processing code（RUNBAT） and library management code（MANLIB）, the multi-group cross section library is built. With this library, a series of verification work is preliminarily carried out. The calculated results show that the library is reliable and meets the requirement of lattice code

Validation with CENDL-NP Thermal Criticality Benchmark

Wu Haicheng, Tan Yingcan, Zhang Huanyu, Jin Yongli, Wang Wenming, Liu Ping

2014, 35(S2): 176-178. doi: 10.13832/j.jnpe.2014.S2.0176

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Abstract:
To solve the overestimation of k_eff in the thermal criticality experiment using CENACE-1.0 data library, the evaluated nuclear data library CENDL-NP-1.0 were validated with several kinds of thermal criticality benchmarks. By analyzing the trends and the relativities among the k_eff results, it is found that the direct contributions to the overestimation of k_eff calculation results for the metal thermal installation with high uranium enrichment are the（n,f）,（n,γ） cross sections and the nubar of 235 U in thermal energy region.

Application of Matrix Exponential Rational Approximation in Point Reactor Kinetics

Cai Yun, Li Qing, Wang Kan

2014, 35(S2): 179-182. doi: 10.13832/j.jnpe.2014.S2.0179

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Abstract:
The solution of point reactor kinetics is approximated by the rational approximation of the matrix exponential function. Three rational approximating methods which are Padé approximation, Quadrature Rational Approximation Method and Chebyshev Rational Approximation Method have been studied. And Richardson extrapolation technique has been used to improve the accuracy. The calculation results demonstrated that when the large positive reactivity is inserted, Quadrature and Chebyshev Rational Approximation Method are with better effects and stability, while Padé approximation with extrapolation may face the stability problem.

Development of Neutron Kinetic Code for Molten Salt Reactors

Zhuang Kun, Cao Liangzhi, ZhenG Youqi, Wu Hongchun

2014, 35(S2): 183-185. doi: 10.13832/j.jnpe.2014.S2.0183

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This study establishes the suitable dynamic models for molten salt reactors considering the effects of fuel flow on the distribution of delayed neutron precursors and then develops a new code named MOREL. Some MSRE experimental data from Oak Ridge National Laboratory（ORNL） are chosen to verify the code, especially the DNP model, and the numerical results indicate that MOREL can be used for the analysis of the molten salt reactors.

Development of Coupled 3-D Neutronics/Thermal-Hydraulics Code for SCWR Core Transient Analysis

Wang Lianjie, Zhao Wenbo, Chen Bingde, Yao Dong, Yang Ping

2014, 35(S2): 186-189. doi: 10.13832/j.jnpe.2014.S2.0186

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Abstract:
A coupled three dimensional neutronics/thermal-hydraulics code STTA（SCWR Three dimensional Transient Analysis code） is developed for SCWR core transient analysis. Nodal Green’s Function Method based on the second boundary condition（NGFMN_K） is used for solving the transient neutron diffusion equation. The SCWR sub-channel code ATHAS is integrated into NGFMN_K by the serial integration coupling approach. The NEACRP-L-335 PWR benchmark problem and SCWR rod ejection problems are studied to verify STTA. Numerical results show that the PWR solution of STTA agrees well with the reference solutions, and the SCWR solution is reasonable. The coupled code can be well applied to the core transients and accidents analysis with 3-D core model during both subcritical pressure and supercritical pressure operation.

Simultaneous Solution of Neutron/Thermal-Hydraulic Coupled System Based on Finite Difference Newton-Krylov Method

Zhang Han, Guo Jiong, ZhOu Xiafeng, Fan Kai, Wang Lidong, Li Fu

2014, 35(S2): 190-193. doi: 10.13832/j.jnpe.2014.S2.0190

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Abstract:
The Newton-Krylov method is utilized to solve the Neutron/Thermal-Hydraulic coupled system in nuclear reactors which convergence rate is higher than the traditional method. Finite difference method is employed to calculate the Jacobian matrix to avoid the analytical expressions of Jacobian matrix. The sparse pattern of the Jacobian matrix is considered to build a higher efficient algorithm compared with the dense Jacobian matrix. The numerical result shows that the advanced Newton-Krylov method is more efficient than the traditional method for the simplified reactor model.

Research of Simulation Program FRET for Fuel Rod Resonance Effective Temperature Mechanism

Yin Qiang, Chai Xiaoming, Tu Xiaolan, Pan Junjie

2014, 35(S2): 194-196. doi: 10.13832/j.jnpe.2014.S2.0194

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Abstract:
Power distribution in the fuel rod is simulated by Monte Carlo program, and the thermodynamic behavior of the fuel rod is simulated by the fuel rod performance analysis program. Model to calculate the effective temperature of the fuel rods resonance is established, and the simulation program FRET is developed to simulate the mechanism for the fuel rod resonance effective temperature. The comparison with the calculation results from the SMART program in the SCIENCE program package showed that the FRET program can accurately simulate and calculate the resonance effective temperature of the PWR without the burnable poison fuel rods. The FRET program does not require the user to input the weighted experience data of the fuel effective temperature and can be used to calculate different types of fuel rods, which is with more extensive applications, compared with the related calculation program used in the engineering.

Research and Development of Radial Power Density Distribution Model for FUPAC

Tu Xiaolan, Chai Xiaoming, Yin Qiang, Liu Dong, Lu Wei

2014, 35(S2): 197-199. doi: 10.13832/j.jnpe.2014.S2.0197

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This paper presents Part Parameters Separation（PPS） method. The method solves the APS problem effectively. To improve the practicability in using PPS method, the sensitivity affecting the radial power distribution is analyzed in this paper to determine the key parameters. On this basis, this paper establishes the radial power density distribution model, which is used in Fuel Rod Performance Analysis Code（FUPAC）. The model is validated with the measured data and the KYLIN-1 calculated results. The results show that the model has a high precision and a wide range of engineering practicality, and it is suitable for various types of fuels, such as UO₂ fuel, UO₂-Gd₂O₃ fuel and IFBA fuel.

Modeling of PWR Fuel Rod Irradiation Behaviour

Zhou Yi, Chen Ping, Zhang Lin, Li Wenjie, XinG Shuo, Guo Xingkun, Liu Zhenhai, Tu Xiaolan

2014, 35(S2): 200-202. doi: 10.13832/j.jnpe.2014.S2.0200

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PWR fuel rod irradiation behaviour models have been developed, including thermal model, mechanical model, fission gas release model, cladding irradiation growth and corrosion model. Validation of models has been done by comparing with the test data. Fuel rod irradiation behaviour models can be used for the fuel rod structure integrity evaluation and performance analysis code development.

Feasibility Study on Fuel Management with Dual-Enrichment 18 Months Refueling

Zhang Shixun, Zou Tingting, Zhang Hong, Pan Hui

2014, 35(S2): 203-206. doi: 10.13832/j.jnpe.2014.S2.0203

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Abstract:
Aiming at the problems of PWR plants with 18 months refueling model, such as low batch discharge burn up and low fuel utilization, two types of fuel assemblies enriched in 4.45% and 4.95% are adopted for the fuel management with 18 months refueling in this paper. The results show that, by optimizing the layout of the fuel assemblies and burnable poison, dual-enrichment 18 months fuel cycle can meet the design criteria of current 18 months cycle fuel management. In the same cycle length, it can achieve a higher average discharge burn up and a fewer fuel cost to get a better economy benefit.

Development of Calculation Code for Fission Product in PWR Primary Loop

Xu Zhilong, Wu Xiaochun, Wan Haixia, Li Long, SHao Jing, Liu Lili, Zhang Jing

2014, 35(S2): 207-210. doi: 10.13832/j.jnpe.2014.S2.0207

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In order to develop the calculation code with the intellectual property right for the source term of Fission Product in Primary Loop（FPPL）, the generation, release and migration of FPPL is studied, and the calculation model for each of the above processes has been developed, and a complete set of calculation method for FPPL has been established. Based on this, a calculation code for FPPL has been developed and validated.

Application of a 3D Discrete Ordinates Program in Heating Rate Calculation for CAP1400 Nuclear Power Plant Internals

Ding Qianxue, Wang Mengqi, Li Hui, Mei Qiliang

2014, 35(S2): 211-214. doi: 10.13832/j.jnpe.2014.S2.0211

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The heating rate of CAP1400 reactor internals is calculated with 3D discrete coordinate（SN） program（TORT）, and the results are compared and verified with the results of MCNP and DORT. The problem solved in the reactor shielding design is fixed-source problem, SORCERY（developed by the Westinghouse company） is used to transfer the core power distribution（pin by pin） to 3D source distribution. Because of the large CAP1400 reactor model, a great deal of computer resources will be expended in the production of 3D fixed source, and the limit of array dimensions in SORCERY will be exceeded, so, another auxiliary program PSOR is developed. And TORT could be used in the CAP1400 large scale problem.

Preliminary Study on Dynamic Rod Worth Measurement Technology in CNPRI

Peng Sitao, Wang Yinan, Li Wen, Fu Xuefeng, Lu Haoliang, Guan Yu, Li Jinggang

2014, 35(S2): 215-217. doi: 10.13832/j.jnpe.2014.S2.0215

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This paper introduces the study on DRWM in China Nuclear Power Technology Research Institute（CNPRI）. Firstly, a code named sim ADRC is developed to repeat the method which adopted by DRWM module of ADRC. Then the DRWM correct coefficients calculation module is developed based on SCIENCE. The module is verified by the analog experiment based on sim ADRC and a set of history RPN current data of DRWM experiments. Result shows that the new developed module works well, with the maximum bank worth error only 5.7%, which is less than the acceptance criterion 10%.

Development of Computerized Reactor Start-up Instrument

Hong Jingyan, Li Yiguo, Zhang Jinhua, PenG Dan, Wu Xiaobo, Hao Qian

2014, 35(S2): 218-220. doi: 10.13832/j.jnpe.2014.S2.0218

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A set of computerized reactor start-up equipment is developed, which consists of the detector, data acquisition circuit, computer and data process software. The data acquisition circuit is made up of plus amplifier and plus counter, multi-range-ultra-low current amplifier, analog to digital conversion circuit and data transfer circuit. This equipment can be used in the experiment of subcritical extrapolation, supercritical interpolation, value of central control rod and value of radial fuel element. A test has performed in the zero power experiment of Miniature Neutron Source Reactor with LEU, and some results have been illustrated.

Development of High Resolution Particle Transport Monte Carlo Code JMCT

Deng Li, Lei Wei, Li Gang, Zhang Baoyin, ShangGuan Danhua, Hu Zehua, MA Yan

2014, 35(S2): 221-223. doi: 10.13832/j.jnpe.2014.S2.0221

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3-D Monte Carlo neutron and photon transport code JMCT has been developed based on the JCOGIN toolbox. The viewdata is equipped in pre-processor and post processor. The domain decomposition and the parallel computation about particle（MPI） and spatial domain（Open MP） have been realized. The full-core pin-mode from Chinese Daya Bay Nuclear Power Station is simulated. The detail pin-power distribution and k_eff result are shown in this paper. The validity of JMCT has been proved.

Research of On-the-fly Doppler Broadening Based on Stochastic Sampling Algorithm

Yang Feng, Liang Jingang, Yu Ganglin, Wang Kan, Li Wanlin

2014, 35(S2): 224-227. doi: 10.13832/j.jnpe.2014.S2.0224

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This paper introduces an on-the-fly Doppler broadening method considering the effect of thermal motion based on the stochastic sampling algorithm, at every collision sets. The method only requires the cross sections at a temperature of 0K, but not the the temperature distribution in the model material, and consequently to eliminate the need of Doppler processing codes. At first, this paper implements the Doppler module to validate the SDB method on the microscopic level, then combining the Reactor Monte Carlo code（RMC）, implements its on-the-fly Doppler broadening function to validate the SDB method on the macroscopic level. On the basis of current study, the capability of the SDB method to deal with temperature reactivity effect is proved.

Domain Decomposition of Combinatorial Geometry Monte Carlo Simulation for Memory Overload Full-Core Nuclear Reactor

Li Gang, Lei Wei, Zhang Baoyin, Deng Li, MA Yan, Li Rui, SHangGuan Danhua, Fu Yuanguang, Hu Xiaoli

2014, 35(S2): 228-230. doi: 10.13832/j.jnpe.2014.S2.0228

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With the development of computer technology and the increasing requirement on nuclear reactor physics, the full-core reactor model contains too many cells and tallies to be loaded for Monte Carlo transport simulation on a single core processor. A domain decomposition of combinatorial geometry Monte Carlo transport is presented in this paper, to simulate the models overload the CPU memory. The tree-based decomposition and asynchronous communication of particle information between domains are also described in the paper. A full-core reactor model from Daya Bay nuclear power station is simulated to verify the domain decomposition algorithms. The domain decomposition provides an available way for the coupled transport and burn-up simulation of the full-core reactor model.

Research of Full Core Burnup Calculations Based on Tally Data Decomposition in RMC

Liang Jingang, Qiu Yishu, Wang Kan, Chai Xiaoming, Qiang Shenglong

2014, 35(S2): 231-234. doi: 10.13832/j.jnpe.2014.S2.0231

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Insufficient memory is the bottleneck problem for large scale transport simulation using Monte Carlo methods. When doing reactor burnup analysis, excessive reaction cross sections are required to be tallied in transport step, thereby the scale of depletion is restricted by the memory storage of computers. To address this problem, a combined parallel method is proposed and implemented in Reactor Monte Carlo code RMC. Tally data is distributed in parallel processes by using tally data decomposition algorithm, which is coupled with the parallel point depletion module. Full core benchmark tests are carried out. The results illustrate that the memory footprint are reduced evidently by using the combined parallel method. It is demonstrated that the data decomposition methods are effective to realize the full core burnup calculations.

Verification on Criticality Calculation of RMC with BEAVRS Full-Core Benchmark

Tang Xiao, Liang Jingang, Wang Kan, Ge Panhe, Li Wanlin

2014, 35(S2): 235-238. doi: 10.13832/j.jnpe.2014.S2.0235

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BEAVRS is a new benchmark based on 1960’s the United States commercial reactor with detailed core construction parameters and operating measurement data for the verification of reactor analysis tool. In this paper, RMC, developed by the Department of Engineering Physics of Tsinghua University for reactor physics analysis, is adopted in the modeling of BEAVRS benchmark, to calculate a series important parameters such as the critical effective multiplication factor under different boron concentration and control rod step number, the control rod worth, the temperature coefficient and U235 fission rate in the instrument tube. The result goes on well with the measured data and the results from similar software, which verified the accuracy and reliability of RMC in the reactor criticality analysis.

Sophisticated Modeling and Calculation of Daya Bay Nuclear Power Station Reactor Based on JMCT

Fu Yuanguang, Ma Yan, ShangGuan Danhua, Li Gang, Li Shu, Zhang Baoyin, Deng Li

2014, 35(S2): 239-241. doi: 10.13832/j.jnpe.2014.S2.0239

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Based on the general Monte Carlo neutron photon coupled transport code JMCT, the sophisticated model of Daya Bay Nuclear Power Station reactor was built. Meanwhile, the k_eff and some local tallies were calculated. The results were compared with the same models calculated by the Monte Carlo particle transport code MCNP, and it showed that relative errors were less than 10-3, which verifies the capability of JMCT in dealing with the large-scale complex geometry models.

2014 Vol. 35, No. S2