核动力工程

Development of Subgroup Method 2D Arbitrary Geometry Reso-nance Calculation Code SUGAR

LIU Qing-jie, WU Hong-chun, CAO Liang-zhi, CHEN Qi-chang

2010, 31(S2): 1-4,15.

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Conventional resonance calculation methods are developed based on the equivalence theory which are only appropriate for the simple regular geometry.It is not capable of dealing resonance calculation for complicated geometries.In this paper,the subgroup resonance calculation method is developed based on the subgroup theory and subgroup neutron transport equation.Resonance calculation is realized for 2D arbi-trary geometry by introducing sophisticated transport solver.Therefore the spatially dependent resonance multi-group cross-sections can be obtained directly by subgroup transport calculation.A computation code SUGAR is developed according to this method and verified by numerical samples.Good accuracy is observed to prove this method and code is very promising for complicate resonance calculation in two dimensional ar-bitrary geometries.

Subgroup Method for Resonance Self-Shielding Calculation and Research on Resonance Interference Effect

HUANG Shi-en, WANG Kan, YAO Dong

2010, 31(S2): 5-8,20.

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We combined the subgroup method and the method of characteristics for neutron resonance self-shielding calculation,and programmed a resonance calculation code SGMOC.This code is based on the WIMSD format multi-group data library.The numerical results of SGMOC are in good agreement with those of MCNP,and thus SGMOC has high calculation precision and geometry flexibility.Based on the SGMOC code,we studied two modification methods for the resonance interference effect calculation.The conditional probability method had a correction effect about 30~180pcm for k_inf calculation of UO₂ fuel cell.The method by virtue of NJOY code had a correction effect about 20~130pcm for k_inf calculation of UO₂ fuel cell.

Study on Application of Resonance Self-Shielding Calculation Method Applying Wavelets Scaling Function Expansion in Resonance Self-Shielding Calculation Problem with Temperature Distribution

YANG Wei-yan, WU Hong-chun, CAO Liang-zhi, LIU Qing-jie

2010, 31(S2): 9-15.

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The radial power distribution in the fuel cell and the continuous-energy spectrums in the fuel cell with a temperature distribution are obtained by the wavelets scaling function expansion method.Wave-lets scaling function expansion continuous-energy self-shielding method is developed,in order to solve the nuclide resonance problem with multiple resonances in complicate geometry.It has been validated and veri-fied by comparison to Monte Carlo calculations.In this method,the continuous-energy cross-sections are processed by NJOY,while the multi-group nuclear data library is jeff31 issued by IAEA.However,for dif-ferent temperature problems,because of Doppler effect,the continuous-energy data library is different,and should be updated by applying NJOY.The calculation efficiency is a problem needed to be improved.There-fore,in this paper,the interpolation is utilized to obtain the continuous-energy cross-sections for other tem-peratures between those given temperatures.Also the precision of the temperature interpolation is discussed.Finally,the differences of continuous-energy spectrums,reaction rates and k_inf results are presented and compared with MCNP calculation.

Study on Acceleration Method for Wavelets Scaling Function Expansion Resonance Calculation

ZU Tiejun, WU Hong-chun, CAO Liang-zhi, YANG Wei-yan, LIU Guo-ming

2010, 31(S2): 16-20.

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The continuous-energy resonance self-shielding calculation method based on the wavelets scaling function expansion method is a valuable potential method to solve the complex resonance problem.Daubechies’ wavelets scaling function expansion method is used to discretize the energy variable of neutron flux within the resonant energy range.In this method,the neutron transport equation is transformed to a set of coefficients equations of wavelets scaling function expansion.Calculation of the coefficients is so time-consuming that a powerful neutron transport calculation method is needed for better calculation efficiency.In this paper,the discrete direction probability method（DDPM） is employed as a tool for solving the wavelet scaling function expansion coefficients.The numerical results indicate that it is capable of computing the coefficients with high computational efficiency and precision.

Three-Dimensional Modular Characteristics Method

LIU Zhou-yu, WU Hong-chun, CAO Liang-zhi, CHEN Qi-chang, LI Yun-zhao

2010, 31(S2): 21-24.

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A new three-dimensional modular characteristics method is proposed in this paper.With this method,the problem geometry is cut into many cuboid cells,from which all the typical cells are selected,and only the characteristic lines of the typical cells are saved to significantly reduce the memory requirement.Corresponding to this technique,new angular quadrature sets are derived.Numerical results demonstrate that this new technique is accurate and efficient to reduce the requirement of memory.

Discontinuous Finite Element Method for Neutron Transport Equations on No Matching Mesh

WEI Jun-xia, YANG Shu-lin, WANG Shuang-hu, SHEN Wei-dong

2010, 31(S2): 25-28,42.

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This paper constructed a no matching mesh for neutron transport equations under 2-D cylindrical geometry,and designed a new sweep algorithm and a new method for computing the mesh boundary flows for neutron transport equations on no matching mesh.The new methods can assure the neutron transport equations to be solved successfully using the discontinuous finite element method on no matching mesh.

Coupled P_N-DP_N Method for Solving the Neutron Transport Equation of Planar Geometry

ZHENG Zheng, WU Hong-chun, CAO Liang-zhi, LI Yun-zhao, CHEN Qi-chang

2010, 31(S2): 29-33.

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One of the difficulties in spherical harmonics（P_N） method is to deal with the discontinuity of neutron flux near interfaces or void boundaries.To solve this problem,Yvon proposed the expansion of the spherical harmonics over each half of the angular range,thus yielding the double spherical harmonics（DP_N） method.The purpose of this study is to combine the P_N method and the DP_N method to solve the time-independent isotropic neutron transport equation in one-dimensional planar geometry.Numerical results from several test problems are presented to demonstrate that great improvements in accuracy can be obtained using this coupled P_N-DP_N method.

Modified Time Discrete Scheme for Time-Dependent Neutron Transport Equation

HONG Zhen-ying, YUAN Guang-wei, FU Xue-dong, YANG Shu-lin

2010, 31(S2): 34-37.

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The classical time discrete scheme does not consider the self adaptive time step and the physical quantity about differential quantity for time variable exits numerical oscillation.Therefore,the numerical precision is very poor.In this paper the time discrete scheme is studied for the spherical time-dependent neutron transport equation.The modified time discrete scheme is constructed for self adaptive time step.The numerical results show that the modified time discrete scheme is very simple and has more accurate numerical precision.The iteration number for modified time discrete scheme is lower than that of the conventional exponential method and diamond difference.Furthermore,the differential curve on time about flux is more smooth than that of the exponential method and diamond difference.

Direct Solutions for Inverse Transport Problems

CHENG Yu-xiong, CAO Liang-zhi, WU Hong-chun, WANG Meng-qi, CHEN Qi-chang

2010, 31(S2): 38-42.

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According to the direct exposure from the flash radiographic image,the direct method for inverse Boltzmann transport equation is presented.The line absorption coefficients and interface locations of object are reconstructed directly at the same time.The algorithm uses ART,Tikhonov,TSVD and TV regularization methods because of the ill-posedness of inverse problems.Numerical results with appropriate regularization parameter agree well with the reference value and avoid amplifying the scatter noise.

Monte Carlo Code for Reactor Analysis-RMC2.0

LI Ze-guang, WANG Kan, SHE Ding, XU Qi, LIU Yu-xuan

2010, 31(S2): 43-47.

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This paper describes a newly developed Monte Carlo code RMC2.0 used for reactor analysis which is applicable to continuous energy and arbitrary geometry.RMC2.0 employs ACE format library and includes two different tracking methods（ray-tracking and delta-tracking）.Some techniques such as uniform energy grid and hash-table method are adopted to improve the performance of RMC2.0.Criticality benchmarks and typical reactor examples are calculated for validation,which shows that RMC2.0 performs well in both accuracy and efficiency compared with MCNP5.Based on Message Passing Interface（MPI）,RMC2.0 succeeds in parallel computation and represents a high speed-up.

Research on MCNP Bootstrapping Calculation Method

YANG De-feng, CHENG He-ping

2010, 31(S2): 48-53.

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Monte Carlo method lacks accuracy,or even fails to get a result when dealing with compli-cated models or cases with very low possibility.At this time we may consider dividing the model into several parts（bootstrapping） and calculate the problem in separate steps.In this paper,the application of bootstrap-ping is discussed,including recording surface source,selecting boundary geometry.A few bootstrapping ex-amples show that the best estimated results can be obtained using SSW/SSR card and reflective/vacant boundary.

Research on Monte Carlo Methods for Particle Transport in Media with Continuously Varying Cross-Sections

LI Ze-guang, ZHANG Xi-si, WANG Kan

2010, 31(S2): 53-57,87.

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In the traditional Monte Carlo method,the cross-sections are assumed to be constant,but this Monte Carlo technique is no longer applicable in continuous materials where the material cross-sections vary over the particle flight paths.Three Monte Carlo methods are developed to solve this kind of problem in this paper,namely Sub-stepping method,Delta-tracking method and Direct Numerical method.Based on the results from different methods to calculate the particle transport in media with continuously varying cross-sections,the comparison of three methods is made and Delta-tracking method is used as the main method to solve this kind of problem.And also an improved Delta-tracking method is posted in this paper to make this method more powerful in solving problems where the material cross-sections vary sharply over the particle flight paths.

Analytic Basis Function Expansion Nodal Method for Multi-Group Neutron Diffusion Calculation in Three-Dimensional Triangular Prism Geometry

WANG Kun-peng, WU Hong-chun, CAO Liang-zhi, WANG Chang-hui

2010, 31(S2): 58-62.

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An analytic basis function expansion nodal method for directly solving the two-group neutron diffusion equation in the triangular geometry without transverse integration was proposed.The distributions of homogeneous neutron flux within a node were expanded into a set of analytic basis functions satisfying the diffusion equation at any point in a triangular node for each group.To improve the nodal coupling relations and computational accuracy,nodes were coupled with each other with both the zero-and first-order partial neutron current moments across all the three interface of the triangle at the same time.To simplify the derivation,coordinate conversion was used to transform the arbitrary triangle to regular triangle.With a new sweeping scheme developed for triangular geometry,the response matrix technique was used to solve the nodal diffusion equations iteratively.Based on the proposed model,the code ABFEM-3T was developed.Both rectangular and hexagonal assembly benchmark problems were calculated to validate the accuracy of the program.The numerical results for the series of benchmark problems show that both the core multiplication factor and nodal power distribution are predicted accurately.So this method can be used in complex unstructured neutron diffusion problems.

A Series of Power Reconstruction Methods Based on Solving Diffusion Equation Approximately in Node

ZHAO Wen-bo, HU Yong-ming, WANG Kan, YAO Dong

2010, 31(S2): 63-67.

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Based on approximately solving diffusion equation in node,a series of reconstruction methods are developed for NGFMN code.These methods are categorized by hyperbolic functions,boundary conditions and the order of Legendre polynomials utilized in equation solving.2D IAEA and 2D BIBLIS benchmarks have been solved.The numerical results are compared with those from CITATION,and the reconstruction method with smaller errors is selected.

Development of Deterministic Code System for GFR Depletion Calculation

ZHANG Jing-yu, SUN Yu-liang, LI Fu, YAN Jian-qiu

2010, 31(S2): 68-72.

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In order to solve the depletion problem of Gas-Cooled Fast Reactor（GFR） quickly and accurately,a deterministic code system is developed.Based on the general compution framework of depletion process,some codes for data conversion,cross section regulation,depletion interpolation and automatic control,are programmed and used to couple the transport code WIMS and diffusion code CITATION.The simulation of the depletion process of one GFR model shows that the deterministic code system is feasible,and it is capable to analyze the neutronics characteristics during the depletion process of GFR.

Nodal SP₃ Method for Solving Neutron Transport Equation

LI Yun-zhao, WU Hong-chun, CAO Liang-zhi, YAO Dong

2010, 31(S2): 73-78,91.

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This paper establishes a nodal method to solve the two coupled equations transformed from the transport equation by SP₃ method.In the nodal method,diffusion equation is discretized into average fluxes in the nodes and partial currents on the nodal boundaries.Nodal response relationship is obtained by expanding the detailed nodal flux into a sum of exponential functions.Numerical results demonstrates that the nodal SP₃ method can treat polygonal meshes in different shapes with good parallel property,and that it can provide a similar accuracy while runs about six times faster than the nodal-S_N（N=4） method DNTR under the same mesh.

Development of Neutronic and Thermal-Hydraulic Couple Code of AHRs and Preliminary Investigation on Bubble Distribution

WANG Liang-zi, YAO Dong, WANG Kan

2010, 31(S2): 79-82.

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The absence of radiolytic-gas bubbles strongly affects the steady operation of Aqueous Ho-mogeneous Reactors（AHRs）,and because of the buoyancy and collision effects,the bubble behavior and their steady distribution in the reactor is always hard to predict.Now,computation methods of AHRs always use a function to express the bubble distribution approximately.Related investigation shows that in CFX,using Multiple Size Group（MUSIG） model,the multi-phase simulation results predict the bubble behavior well,and tally with the experimental results of an AHR.This paper firstly introduces the neutronic and ther-mal-hydraulic couple code FMCAHR_CFX,and in the last chapter,FMCAHR_CFX is used for a preliminary analysis of a simple AHR.

Analysis of Physical Performance of Two-Row Hexagonal Fuel Assembly for SCWR

WANG Lian-jie, QIN Dong, LI Qing, YAO Dong

2010, 31(S2): 83-87.

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A new hexagonal fuel assembly（FA） design which has two rows of fuel rods between the hexagonal moderator channels is proposed for the thermal supercritical water cooled reactor（SCWR）.The two-row hexagonal fuel assembly concept is first and foremost considered for the performance of uniform moderation and sufficient moderation,and with the respect of structural performance and thermal-hydraulic performance.The physical performance of the two-row hexagonal FA with acceptable configuration is investigated.The results show clearly that a better balance between uniform moderation and sufficient moderation can be obtained in the two-row hexagonal fuel assembly.The design objective of the local power peaking factor less than 1.10 can be achieved under the condition of using the same enrichment fuel and containing no burnable poison.

An Innovative Algorithm for Reactor Core Loading Pattern Optimization

GONG Zhao-hu, YAO Dong, WANG Kan

2010, 31(S2): 88-91.

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To solve the reactor loading pattern optimization problem in realistic nuclear power plants,this paper presents a new optimization method-Interval Bound Algorithm,and introduces the origin of the algorithm and describes its implementation procedure.Based on this algorithm,the corresponding parallel loading pattern optimization code of IBALPO is developed using Fortran95.Finally,this code system is used to solve a realistic and difficult reload problem to show its performance,and a better pattern is obtained compared to the one searched by experienced engineers.

Monitoring of In-Core Power Distribution by Ex-Core Detectors

LI Fu, ZHOU Xu-hua, WANG Deng-ying, GUO Jiong, LUO Zheng-pei

2010, 31(S2): 92-96.

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For most reactors,the ex-core ion-chambers are the only real-time nuclear detectors.The necessity and the possibility to monitor the in-core power distribution are discussed in this paper,and the methodology to monitor the in-core power distribution via ex-core detector readings based on the harmonics synthesis method and detector spatial response function are proposed,and the preliminary numerical results are presented for the low-temperature heating reactors and high-temperature gas-cooled reactors,therefore the feasibility is demonstrated.

Harmonics Expansion Method for On-line Monitoring of PWR Power Distribution

WANG Chang-hui, WU Hong-chun, CAO Liang-zhi

2010, 31(S2): 97-101.

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The power distribution of PWRs can be expended by harmonics.Combined with the detector readings,the harmonics can be used to reconstruct the reactor power distribution on-line with high speed.The accuracy of reconstruction is related to the precision of neutron detectors and the similarity of the reference case to the real one.But it is difficult to choose the reference case because the condition of the reactor is much complex.In this paper,some simulations of the reactor are done to generate a data library of harmonics for on-line monitoring of the PWR power distribution.Measurement from DAYABAY nuclear power station is used to verify the accuracy of this method.The result shows that the relative errors between the measurement and the reconstruction result is less than 3%,which satisfies the requirement of engineering.

Study on Core Control Strategy for Initial Cycle Utilizing Gadolinium Burnable Poisons

LIU Qi-wei, YU Ying-rui

2010, 31(S2): 102-106.

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Based on the present MODE-G control strategy,an improved new core control strategy is developed for the initial cycle utilizing gadolinium burnable poisons.With the relaxed axial offset control（RAOC） procedure to limit the core axial offset（AO） in a band region,load follow operations are simulated under the improved core control strategy and the simulation results are compared with those obtained using MODE-G control strategy.Additionally,a preliminary evaluation of the impact of the an ejected rod with the improved core control strategy is performed.It is shown that the improved core control strategy can control the core axial power distribution more satisfactorily and effectively.

Progress of Optimization and Modification of China ITER Helium-Cooled Solid Breeder TBM

ZHANG Guo-shu, FENG Kai-ming

2010, 31(S2): 107-110.

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In this paper,scientific and experimental values of DEMO breeding blanket（DEMO-BB） on International Thermonuclear Experimental Reactor（ITER）,functional objectives and general testing strategy,and etc.for Testing Blanket Module（TBM） on ITER are summarized.The design procedures are introduced and discussed emphatically from 2003 to 2010,such as the progress of general structure,the optimization of neutronics,thermo-hydraulics and electro-magnetic structure for China Helium Coolant Solid Breeder（HCSB） TBM.Finally,some important design experiences are summarized.

Fuel Management and Economics Analysis on M310 Core with 30% MOX Fuels

JIA Shi-zhuang, CAO Liang-zhi, WU Hong-chun, ZHENG You-qi, FU Xian-gang, YANG Jue

2010, 31(S2): 111-115.

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This paper proposes the in-core fuel management of M310 core with 30% MOX fuels and investigates the impact on the key core wide nuclear design parameters.The paper also does some economics analysis of the corresponding fuels.The study shows that loading MOX fuel assembly in M310 is feasible technically.Economics analysis model is established,and the fuel cost for the proposed core is analyzed.The result shows that though the introduction of MOX fuel greatly increases the fuel cost,there exists the possibility to decrease the cost.

Physical Design of a Thorium Fuel-based Long Life Reactor Core

YU Gang-lin, WANG Kan

2010, 31(S2): 116-120.

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The design objectives of a long life reactor include the high fuel burnup and full power natural circulation,which increases the difficulty of reactor physical design: a higher conversion ratio to achieve zero reactivity swing with burnup and increase the P/D ratio to achieve natural circulation.It is a hard mission for the U-Pu fuel long life reactor design.This paper describes a spent Pu-Thorium fuel and lead-bismuth coolant based long life core design,taking the advantages of thorium fuel in fast neutron spectrum.The MCBurn code is chosen to perform the physical calculations..

Fission-Fusion Mixed Neutron Field in Fission Reactor

DENG Yong-jun, YUAN Shu, LI Run-dong, FENG Qi-jie

2010, 31(S2): 121-124,127.

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In this paper,the physical designs about Fission-fusion mixed neutron field were introduced.The local mixed neutron fields which use the LiD and uranium target as the converter material were converted from the high-intensity reactor thermal and fission fast neutron fields.The structure of the conversion target and the material of neutron spectrum adjustment were optimal designed through the MCNP code.The structure of the converter target,the neutron spectrum of the Fission-fusion mixed neutron field and the heating rate distribution were obtained.

Development of Multi-Group Library for Fusion-Fission Hybrid Reactors

YI Wei-wei, HU Ze-hua, LI Mao-sheng

2010, 31(S2): 125-127.

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According to the calculation requirement of fusion-fission hybrid reactors and the relevant studies,a new 187-group constants library has been developed.This library containing more than 100 nuclides with basic data selected from ENDF/B-VI.8 can be used to calculate the neutron transport equation.The weight flux,group structures,thermal neutron transport and the Bondarenko option is considered during the process.Some calculations and comparative analyses are performed based on a series of existing benchmark experiments using ANISN.The result indicates that the library satisfies the basic requirements on calculation for designing the hybrid energy reactors.

Treatment and Improvement of Equivalent Cross Sections in Control Rod Region for High Temperature Gas-cooled Reactor

GUO Jiong, LI Fu, WANG Deng-ying

2010, 31(S2): 128-131,135.

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For the HTR core physical calculation,the modeling of strong absorber region in diffusion theory is a well-known problem,even with the homogenization treatment.The problem is how to take the advantage of high performance computing for solving diffusion equation,and at the same time to keep accuracy as transport theory.The method of equivalent cross section is proposed based on the combined transport and diffusion theory.The homogenization method is used to estimate the homogeneous cross section by volume-weighted in the control rod region.The modified diffusion constant of the control rod region is derived from the fact that the net current at the interface of the control rod region calculated by diffusion theory must be equal to that by transport theory.These equivalent cross sections can be directly used in the diffusion equation and there is no need to adjust the diffusion solver.The method of the equivalent cross section has been used to the region of the 2-dimensional transport calculation.And improvement is proposed to divide the mesh of the absorber region in the process of the calculation of the equivalent cross section.It is verified in different mesh configurations.The numerical result demonstrated the good performance and improvement of this method.

Elastic Recoil Detection of Helium Content in Nuclear Material by Heavy Ion

LIU Chao-zhuo

2010, 31(S2): 132-135.

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The retention capacity of helium in reactor materials is one of the key factors to assess their safety performance.In this paper,the basic principle and experimental method of elastic recoil detection（ERD） analysis were presented that the energy spectra of helium can been obtained by elastic recoil detection of ion beam.And the analysis of helium content and depth distribution in thin film can been achieved.With 12MeV carbon ion as a probe,helium in zirconium film was elastically recoiled forward.The content and depth distribution of helium were obtained by analyzing the energy spectra.The helium content of film material annealed at different temperatures was detected for comparison.

Experimental Measurement of Neutron Noise Time Interval Distribution and Theoretical Research on Calculation Method

CHEN Li-gao, YU Gang-lin, WANG Kan, LI Cheng-long, YANG Xin

2010, 31(S2): 136-140,158.

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Experimental research on the measurement of the reactor dynamic parameters is carried out by the time interval distribution method（Babala method）.This paper describes how to acquire the distribution of the lengths of intervals between neutron pulse counts using MSP430 Single-Chip Computer and how to transfer the data to the PC.The interval distribution of neutron signals on the sub-critical facility at Depart-ment of Engineering Physics in Tsinghua University is measured,and the results are analyzed,to study the calculation method for the eigen value α based on the measured data.Further more,Feynman method and Rossi-α method are used to calculate the value of α,which are compared with the Babala method.The results indicate that Babala method for dynamic parameter measurement could reduce the measurement time and im-prove the accuracy of measurement results.This paper also gives out the statistical distribution difference be-tween the source neutrons and reactor neutrons.

Reactor Reloading and Start-up without Neutron Source in Qinshan NPP

LIAO Ze-jun, KONG De-ping

2010, 31(S2): 141-144.

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In the 5th and 11th fuel cycle of Qinshan NPP,the core was loaded with an over decayed neutron source,and without irradiation activated neutron source respectively,which caused a blind area of ex-core nuclear detecting instrumentation.This paper introduces the Chinese regulatory requirements and the corresponding measurements which Qinshan NPP took to apply the reactor start-up without neutron source,and shows that the safety of reactor start-up without neutron source can be ensured.

Benchmark CENDL-3.1 with International Handbook of Evaluated Criticality Safety Benchmark Experiments

ZHANG Hua, LIU Ping, WU Hai-cheng

2010, 31(S2): 145-149.

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Nearly one thousand assemblies were selected from the International Criticality Safety Benchmark Evaluation Project（ICSBEP） in the International Handbook of Evaluated Criticality Safety Benchmark Experiments.These cases cover the fast,inter-and thermal spectrum.Criticality calculations were done by using Monte Carlo code MCNP4C.The ACE files were produced with nuclear data processing code NJOY99.259.The calculated results of effective multiplication factor k_eff with CENDL-2.1 and CENDL-3.1 libraries were compared with the experimental results.The systematic validation shows that the data from CENDL-3.1 give better results of k_eff values than those from CENDL-2.1.

3D Calculation Analysis of Activation and Shutdown Dose Rate for a PWR Nuclear Power Plant

HAN Jing-ru, CHEN Yi-xue, WANG Ji-liang, QUAN Guo-ping, LU Dao-gang

2010, 31(S2): 150-153.

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Based on 3D Monte Carlo method,the rigorous 2-step（R2S） method has been used for the design analysis of a pressurized-water reactor（PWR） for the first time.The work preliminary validated the FISPACT-2007 application feasibility in a PWR.3D model of a minitype PWR was built.Residual radioactivity and shutdown dose rates were performed after the PWR operating 40 years using the R2S method.The results indicated that the R2S method can be used in the parts activation analysis and decommissioning strategy establishment of the PWR nuclear power plant.

Application of Burnup Credit Method in Critical Safety Analysis

YANG Bo

2010, 31(S2): 154-158.

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The project is a part of STC partnership of CEA-CAEA Co-operation Protocol in the peaceful uses of nuclear energy.This paper mainly presents BUC（BURNUP CREDIT） method and related computer code package（APOLLO-2,DARWIN and CRISTAL）.By using these codes,sensitivity analysis for main physics parameters（including fuel temperature,moderator temperature,critical boron concentration and etc.） of fuel assembly in different irradiation conditions is calculated.The result shows that the neutron multiplica-tion factor will increase as the increasing of temperature and boron concentration.In order to consider the BUC method conservatively,the isotopic correction factor and its calculation uncertainty are taken into ac-count.Then,the paper lists the comparison of the results between calculation and experiment.Finally,this paper gives the analysis of BUC method according to five different axial burnup profiles.

Neutron Spectrum Adjustment Code 2NP for Surveillance Capsule in Nuclear Power Plants

SUN Zheng, SHEN Feng, BAI Ning

2010, 31(S2): 159-161.

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Neutron spectrum adjustment is the effective method to estimate the neutron fluence in the nuclear power plant reactor pressure vessel（RPV）.The aim of this work is to develop the neutron spectrum adjustment code 2NP based on the generalized least squares method（GLSQM）,which will be used in neutron fluence estimation of RPV.The accuracy of the code 2NP have been proven by calculating the benchmark ANO issued by IAEA.

Simulation Method for Running-in Phase of HTR-10

XIA Bing, LI Fu, WU Zong-xin

2010, 31(S2): 162-166.

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The simulation method of the running-in phase of the HTR-10,the 10MW pebble-bed high-temperature gas-cooled reactor,has been demonstrated in this paper,based on the operation data of the HTR-10 and using VSOP,the HTGR physics analysis program.The major idea of the simulation includes the precise tracing on the power history and the fuel loading-unloading history of the core,the description of the spatial distribution of the compositions within the core at different moments,the determination of the detail simulation scheme on the fuel cycling,the effective control poisons added into the model according to the history of the control rod position,and the thermo-hydraulic feedback calculations based on the real inlet thermal conditions.

Effective Yield Analysis of 14 MeV Neutrons Induced by Thermal Neutrons in ~⁶LiD

LUO Yong, YE Bin, LI Quan-wei, PENG Feng

2010, 31(S2): 167-170.

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The principle of thermal neutron converting to 14MeV neutron is briefly introduced.A calculation and analysis model for 14MeV neutron effective production is established.The probability for the thermal neutron absorbed by 6Li and for T not leaked in ⁶LiD are calculated.The cross sections for T and D fusion reaction and T and 6Li fusion reaction are proposed,and the curves for the effective yields of 14 MeV neutron with the thickness of ⁶LiD is given.The results show that when ⁶LiD thickness is between 0 and 0.5mm,the effective yields of 14MeV with ⁶LiD thickness almost augment linearly,and then gradually remain at a set value;and when ⁶LiD thickness is 0.7mm and 1.0mm,the effective yields of a thermal neutron converted to 14 MeV neutron is 3.18×10^-4 and 3.53 ×10^-4,respectively.

High-Efficiency Monte Carlo Simulation Method on Energy Deposition of Photon and Electron

QIU You-heng, YING Yang-jun, WANG Min, CHEN Xing-liang

2010, 31(S2): 171-174.

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MCNP program counts the energy deposition of photon and electron by the *F8 method.The computational efficiency of *F8 is very low.With a high cutoff energy of electron,the efficiency of *F8 will be greatly increased,but the precision may be decreased.And the appropriate cutoff energy is unknown.A new calculation method and speedup technique of photon and electron energy deposition are introduced in this paper.The new method merely counts the energy loss of electron collision,whether a electron will be cutoff or not rest with the electron energy and position in the speedup technique.Compared to the method used in MCNP program,the calculation efficiency is increased and the precision is very high in our new method.It is the most important character that the cutoff energy of electron is selected automatically in the speedup technique.

2010 Vol. 31, No. S2