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2016 Vol. 37, No. S1

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Study on Overpressure Condition of Reactor Coolant System for Small Modular Reactor under Turbine Trip Accident
Chen Hongxia, Tian Haowen, Yu Na, Qiu Zhifang, Fang Hongyu, Guan Zhonghua
2016, 37(S1): 1-5. doi: 10.13832/j.jnpe.2016.S1.0001
Abstract:
The overpressure risk of small modular reactor under turbine trip accident is analyzed by RELAP5.To prevent RCS overpressure,the paper studies on the decreasing of the transient core power and primary overpressure protection.The results indicate that the RCS overpressure of the small modular reactor can be mitigated effectively with appropriate surge line flow area.
Analysis and Control of Fit Clearance between Main Bolt and Its Hole in Nuclear Reactor Pressure Vessel
Hu Yunfeng, Gou Yuan, Li Fengpei, Jiang Jiequan, Mu Huaming, Zhang Hongxiang, Bai Kai
2016, 37(S1): 6-8. doi: 10.13832/j.jnpe.2016.S1.0006
Abstract:
When the main-bolt screws into the reactor pressure vessel,the fit clearance of the main-bolt and its hole is related not only to the speed value of the bolts and the alignment accuracy,but also the effective removing of the main-bolt and the damage of the hole surface.The selection of the main-bolt’s spinning-speed and the specific values of centering requirements are given by the quantitative analysis of dthe eputy thread fit clearance.In order to obtain a stable ideal thread pair fit clearance,we should control the manufacturing of the thread and optimize the assembly process.
Study on Corrosion and Hydriding Performance and Model of Specifically Fuel Rod Cladding
Guo Xingkun, Tian Haowen, Zhou Yi, Zhang Minjie, Li Songling, Liu Songtao
2016, 37(S1): 9-11. doi: 10.13832/j.jnpe.2016.S1.0009
Abstract:
The corrosion and hydriding of zircaloy cladding are important factors for the characteristics of fuel rods.In this paper,the corrosion model and hydriding model of the fuel cladding used in a specifically reactor have been developed based on studying the existing models and considering the characteristics of the specifically fuel rod,and the new models have been validated using available irradiation data.
Design of Remote Control to Replace Electric Parts of Master-Slave Manipulator
Zhang Xianpeng, Jiang Changyu
2016, 37(S1): 12-13. doi: 10.13832/j.jnpe.2016.S1.0012
Abstract:
The control mode of domestic ZC series master-slave manipulator is transformed to remote control.The front,rear,left and right of the transformed manipulator can be flexibly controlled by using the remote controller,and the operator can be closer to the hot-cell window,thus to expand the scope of operation.
Research of Sludge Deposition on Tube Bundle for Steam Generator
Li Yang, Lin Xinru, Cao Nian, Niu Maozhi, Li Yong, Zan Yuanfeng
2016, 37(S1): 14-18. doi: 10.13832/j.jnpe.2016.S1.0014
Abstract:
The main reason for the corrosion of SG heat transfer tube is the deposition of sludge particles on the tube bundle.Numerical simulation was used to study the deposition characteristics of the sludge particles on the tube bundle,and the deposition characteristics of various size sludge particles on the different locations of the tube bundle.Studies have shown that low flow rate and backflow were the main reasons for the retention of sludge particles in the tube bundles.When the particle size is larger,the deposition of sludge particles in the bottom plate,flow distribution plate and support plate is more dispersed.With the particle size is decreased,more sludge particles are deposited on the central of the bottom plate,and on the barrel of the flow distribution plate,and on the central of the support plate.
Improvement of PWR online Sipping and Data Analysis Method
Li Ligang, Liao Zejun, Deng Zhixin
2016, 37(S1): 19-22. doi: 10.13832/j.jnpe.2016.S1.0019
Abstract:
Online sipping is an effective measure to screen a damaged fuel assembly in PWR,and it is an important part of the fuel assembly integrity management.This paper introduces the problems existing in online sipping calibration and data analysis method,analyzes the causes of the problems,and proposes improved methods.Practice shows that these improved methods reinforce the reliability of online sipping detection and judgment,solve the online sipping calibration problem of poor reliability,and solve the online sipping data analysis problem of low efficiency.
Temperature Control Techniques for High Temperature Material Irradiation Test in HFETR
Liu Yang, Zhang Liang, Yang Wenhua
2016, 37(S1): 23-25. doi: 10.13832/j.jnpe.2016.S1.0023
Abstract:
Temperature control techniques for high temperature material irradiation test(MIT)in HFETR are summarized in this paper.Its feasibility and effectiveness are validated with the experimental results and simulation results with CFX code in a hastelloy MIT.
Analysis on Removing Ability of Decay Residual Heat of JNA、FAK System of Third and Fourth Unit in Tianwan Nuclear Power Plant
Luo Feng, Zeng Xiaokang
2016, 37(S1): 26-29. doi: 10.13832/j.jnpe.2016.S1.0026
Abstract(12) PDF(0)
Abstract:
Based on the heat load method,an analysis is performed for the removing ability of decay residual heat of core and spent fuel pool of core decay heat recovery circuit(JNA)、spent fuel pool cooling system(FAK) of The Third and Fourth Unit in Tianwan Nuclear Power Plant after the modification of fuel assembly.The results show that the removing ability of decay residual heat of JNA system of The Third and Fourth Unit in Tianwan Nuclear Power Plant could satisfy the need of removing of decay residual heat under normal condition,anticipated operating condition and accident condition.It is also pointed out that the ability of FAK system can satisfy the need of removing of decay residual heat of spent fuel pool.
Research Progress of HT-9 for Fast Reactors
Lyu Liangliang, Li Yuanming, Zhou Yi, Guo Zixuan, Guo Xingkun, Zhang Minjie
2016, 37(S1): 30-33. doi: 10.13832/j.jnpe.2016.S1.0030
Abstract:
HT-9 is ferritic/martensitic steel with Cr content of 12% and it is an excellent candidate material for fuel clad and duct in liquid metal fast reactors.The research progress of HT-9is introduced.The key application properties needed in liquid metal fast reactor environment are summarized and reviewed.The main problems of using HT-9 in lead cooled fast reactor are analyzed and the developing trend of HT-9 in future is commended.
Research of Lattice Style Tube Support Sheet Used by Steam Generator
Tan Guowei, Zhang Minjie, Deng Feng
2016, 37(S1): 34-37. doi: 10.13832/j.jnpe.2016.S1.0034
Abstract:
This paper researched the feature of lattice support structure,improved the structure of the dominant tube support sheet,and proposed a new tube support structure,which had all the advantages of lattice support structure and dominant tube support structure.Analysis shows that the new tube support structure in this paper is much better than the lattice support structure of Babcock& Wilcox Corporation,with vast potential application.
Diagnosis and Treatment of Low-Frequency Vibration of Steam Turbine Tilting Bearing
Wang Dacheng, Shi Qingfeng, Wang Miaomiao, Zhang Jinggan
2016, 37(S1): 38-42. doi: 10.13832/j.jnpe.2016.S1.0038
Abstract:
Unsteady low frequency vibration happened on the tilting bearing of a 600 MW nuclear steam turbine during the operation.Based on the vibration and load,the amount of oil,speed and other variables,and from the characteristics of vibration,it is ascertain that the low frequency vibration of a tilting pad bearing is mainly caused by the low frequency resonance resulted from the oil film instability of the thrust bearing and mis-installation of the vibration probe support,.and some suggestions are put forward to reduce or eliminate the low frequency vibration of the bearings.
Development of Detective Device for Control Rod Position Sensor in Nuclear Power Plants
Wang Yongchao, Zhang Bin, Dai Bo, Hu Dongmei
2016, 37(S1): 43-47. doi: 10.13832/j.jnpe.2016.S1.0043
Abstract:
To solve the control rod position sensor measurement problems in Tianwan nuclear power plant,the requirement of detective device for the control rod position sensor was proposed.In this paper,the design scheme and its technological process was completed based on the structure features and function requirements of the step-type control rod position sensor device,and the vessel simulating the working conditions,the pressure loop,the heating device and the control system were designed.The electric heating power,and the proportion,integral,differential setting parameters and pressure parameter of the control system were verified by the tests of heat capacity,automatic temperature control,and the pressure following control.Tests show that the function of each unit is normal.
Analysis of Destruction Laws of Double-Wall Encapsulated Insulation Configuration under LBLOCA Condition
Xing Dianchuan, Tang Ming, Wang Tao, Hou Fengwei, Luo Feng, Zhao Haijiang
2016, 37(S1): 48-51. doi: 10.13832/j.jnpe.2016.S1.0048
Abstract:
The insulation fiber debris,deduced from the LOCA fluid blasting,is a potential debris source blocking the sump strainer in the containment.The present work simulates the physical process and pressure field of LBLOCA by using of high-pressure air,moreover,destruction law of double-wall encapsulated insulation configuration is analyzed by replaying high speed video and accounting debris.The result indicates that the inner wall is the weakest part for double-wall encapsulated insulation.The modules separated from the pipe due to failure of connection mechanism subjected to high speed air impacting,and the insulation material fragments were released mainly through tearing the inner wall.The stainless-steel bands were proposed to reduce the debris by analyzing the failure mechanism of the insulation configuration.Finally,the insulation improvement method was verified by experiments.
Study on Adsorption Properties on Strontium and Cesium of HAP and HAP-AMP
Yang Bin, Li Bing, Zhang Jingsong
2016, 37(S1): 52-55. doi: 10.13832/j.jnpe.2016.S1.0052
Abstract:
The study about the effect of various factors on preparing dydration of five antimony oxide two(HAP) and hydration of five antimony oxide two-phosphorus ammonium molybdate(HAP-AMP) has been carried out in this paper.The best preparation process of HAP and HAP-AMP under experimental conditions is found;at the same time,the HAP of Sr2+,HAP-AMP for Sr2+,CS+ adsorption properties were investigated.Results show that under the experimental conditions,HAP of Sr2+ the maximum adsorption capacity can reach 0.73 mmol/g,and the maximum adsorption capacity of HAP-AMP for Sr2+,CS+ can be up to 0.47 mmol/g and 0.32mmol/g,respectively;at the same time,HAP-AMP for Sr2+,CS+ adsorption is with significant selectivity.
Qualification of Low Alloy Steel Forgings for Reactor Power Vessel
Yang Min, Luo Ying, Li Changxiang, Ma Shuli, Fu Qiang
2016, 37(S1): 56-58. doi: 10.13832/j.jnpe.2016.S1.0056
Abstract:
In order to solidify the manufacture process and associated key parameters of large low alloy steel forgings for reactor pressure vessel(RPV),an integrated technical qualification plan is established according to the RCC-M M140.This plan can evaluate the uniformity of the chemical composition,mechanical properties and metallographic structure of low alloy steel forgings for RPV.The plan has been implemented successfully during RPV manufacturing.
Prototype Test of Nuclear Class I Motor-Operated Globe Valves
Yu Haifeng, Xie Qingqing
2016, 37(S1): 59-63. doi: 10.13832/j.jnpe.2016.S1.0059
Abstract:
This paper presents the test purpose,contents,test apparatus,measurement parameters and test methods for the nuclear Class I motor-operated globe valves.The test results indicate that the valves,seats and stems are airtight without leakage,and opening and closing stroke is 11s under cold state before the test;the valves are opened and closed normally and the opening and closing stroke is 11s in the tests of hot cycle lifetime,hot cycle,heat-cold alternate and flow interruption capability;after the test,the valves and seats are airtight 10 min and 12 water droplets are leaked,the stems are airtight without leakage,and opening and closing stroke is 11s under cold state.All the test data have demonstrated that the valves meet the design requirements.
Effect of Microbiology Ecology in Main Seal Water System of Power Plants
Zhang Wei, Bian Chunhua, Chen Huan, Xu Guangzhi, Zhong Xiaoping, He Luping, Yan Guoquan
2016, 37(S1): 64-67. doi: 10.13832/j.jnpe.2016.S1.0064
Abstract:
Microbiology influenced corrosion(MIC) happens in the main pump seal water system in nuclear power plants.For systematical study of the microbiology ecology in this system,many numbers of water samples were selected and their microbiology composition was tested.Results of microbial culture indicate that the total number of colonies from the source of water is the highest.MPN method test result indicates that the iron bacteria is the main type in this system,including Ralstonia,Acinetobacter,Pseudomonas and other bacteria that hard to cultured.Study reveals that the blocking phenomenon in the main pump seal water system is caused by many types of bacteria,and to reduce the effects of MIC,microbiology ecology must be destroyed.Regular monitoring and disinfection could decrease the risk of blocking.
Several Important Safety Issues Considered in Design of Land Reactor Test Engineering
Qin Legang, Zhao Yulong, Tang Bin
2016, 37(S1): 68-70. doi: 10.13832/j.jnpe.2016.S1.0068
Abstract:
Construction purpose for the land reactor test engineering and its difference with the design of nuclear power plants are briefly described in this paper.Several considerations and further explanations with regard to important safety issues for the design of land reactor test engineering are given,including safety objective,emergency plan,regulations and standards,defense in depth,beyond design basis accident,and radioactive material confinement.
Preliminary Research on Multi-Reactor Accident Radiation Dose Evaluation Methodology in Near-Site
Peng Haicheng, Zhang Yan, Fang Sheng, Diao Fei, Lu Chunhong
2016, 37(S1): 71-74. doi: 10.13832/j.jnpe.2016.S1.0071
Abstract:
Multi-reactor accident source terms have the characteristics of multi-point simultaneous dispersion,stronger direct radiation exposure,combination of building radiation shielding effect and reflection effect,and more complex radioactive material concentration computing.Current radiation dose evaluation methodology and systems can not fully satisfy the need of near-site dose evaluation and emergency under multi-reactor accident.It has been proved that small scale wind field and atmospheric dispersion are key elements affecting plume exposure dose to emergency workers,thus the complex terrain and building effect can not be neglected by the atmospheric dispersion model.Through the analysis of main Chinese multi-reactor distribution and site condition,CFD has been recommended as near-site fluent field computation method.For the multi-reactor dose field computing,conservative method proves the virtue of computing time saving and strong operability,while realistic method needs much time with more precision,and it should be selected based on the nuclear accident emergency demand.
A Methodology Study for Deep Penetration Shielding Calculations of Research Reactors Based on MCNP Code
Zhang Yin, Liu Caixia, Zhang Li, Zhou Qi, Han Guosheng, Wei Shuang
2016, 37(S1): 75-79. doi: 10.13832/j.jnpe.2016.S1.0075
Abstract:
MCNP code was used to execute the verification calculations of the shielding design for a research reactor in this paper.Various techniques including geometry splitting,source term simplification,multi-group cross-section and energy cutoff were introduced to optimize the calculation.The result shows that the output value from MCNP coincides well with that from the deterministic methods.Besides,a significant increasing of the computing efficiency up to 70% and a small variance of the run output lower than 10% were obtained as a result of the application of the aforesaid techniques.Moreover,a conclusion can be drawn from the analysis that the method proposed to deal with the deep penetration problems in this paper is found to be applicable to the shielding calculations of other research reactors with thick shielding.
Critical Safety Analysis of Nuclear Fuel Temporary Storage
Liu Caixia, Zhang Yin, Yi Lei, Wang Jinrong, Han Guosheng, Gong Dianrong
2016, 37(S1): 80-83. doi: 10.13832/j.jnpe.2016.S1.0080
Abstract:
The critical safety of nuclear fuel elements from a research reactor and two NPPs under normal and accidental conditions was analyzed through the calculation of keff by MCNP.Through the calculation,an upper threshold of the enrichment degree which could be stored in the facility was obtained,the effects of distance and water density on keff were analyzed;and several fuel pellets storage schemes were compared.The calculation results indicate that,under normal or accidental conditions the critical safety is acceptable;the enrichment threshold is 7.44%;keff approximated linear increases and quadratic decreases as the enrichment degree increases,and approximately linear decreases as the distance increases;fuel pellets with different enrichment degrees should be districted in a certain area according to their enrichment degrees,and fuel pellets with comparative high enrichment degrees should not be stored in the center of the storage area.
Study on Safety Regulatory Elements of Nuclear Reactor Decommissioning
Zhang Hong, Zhang Liang, Zhang Qibin, Xu Jian, Huang Qingyong
2016, 37(S1): 84-87. doi: 10.13832/j.jnpe.2016.S1.0084
Abstract:
Based on the analysis of the safety characteristics and the main hazards existed probably for nuclear reactor decommissioning,and the consideration of the experiences and lessons obtained from relevant accidents/incidents and good practices,safety regulatory elements have been studied and related proposals are provided,which are beneficial for improving the direction and effectiveness of nuclear safety regulatory processes,and promoting the safe implementation for the decommissioning projects of nuclear reactors.
Discussion on Seismic Fortification Requirements for Nuclear Fuel Cycle Facility
Zhao Yulong, Xu Jianhua, Kong Qingjun, Li Tao
2016, 37(S1): 88-92. doi: 10.13832/j.jnpe.2016.S1.0088
Abstract:
The relevant seismic fortification standards and requirements on both domestic and international nuclear fuel cycle facilities were briefly described in this paper.Taking the facility potential risk as the starting point of seismic fortification and fully considering the specific characteristics of nuclear fuel cycle facilities extensively,this paper discussed the key factors of seismic fortification.In addition,taking the seismic fortification practice in domestic nuclear facilities into account,the requirements and suggestions on typical nuclear fuel cycle facility were made,and the international relevant seismic fortification requirements were compared for the purpose of reaching a clear unified agreement on seismic fortification criterion,to offer suggestions for nuclear fuel cycle facility design and evaluation to some extent.
Uncertainty Analysis of Severe Accident Results with MACCS Model
Zhang Liang, Huang Qingyong, Zhang Qibing, Xu Jian, Zhang Hong, Xie Jianlun
2016, 37(S1): 93-95. doi: 10.13832/j.jnpe.2016.S1.0093
Abstract:
In order to understand the effect of the input parameters on the estimation results from MACCS model,a station blackout(SBO) accident in a nuclear power plant is taken as an example,and the uncertainty of a severe accident consequence estimated by MACCS model is analyzed using the Latin hypercube sampling techniques.The results of the uncertainty analysis show that:the uncertainty factors of the early individual dose caused by the accident are all about 2.9(95% confidence limited),and the uncertainty factors of the long-term individual dose are all about 1.7(95% confidence limited).We can find that:when used to estimate the accident consequence,the MACCS code has an uncertainty within 3(95% confidence limited) caused by the input variables.
Seismic Analysis of Typical Equipments in High Level Radioactive Vitrification
Wang Xiaorong, Zhao Yulong, Zhang Xiaowei, Sun Dequan, Huang Xingrong
2016, 37(S1): 96-98. doi: 10.13832/j.jnpe.2016.S1.0096
Abstract:
This paper elaborates respectively the computation input,numerical models,software,methods,load combination,stress evaluation criteria,items and codes between China and Germany about typical equipments of the high level radioactive vitrification projects,and analyzes the coherences and differences at the same time.Results indicate that both have some safety margin,although they use different codes and softwares.
Treatment of Safety Issues in Radioactive Liquid Waste Storage Facility Design of Research Reactors
Wang Jing, Li Tao, Xu Jianhua, Zhao Yulong
2016, 37(S1): 99-102. doi: 10.13832/j.jnpe.2016.S1.0099
Abstract:
In the nuclear safety evaluation of research reactors,there are differences in the amount and activity concentration of radioactive liquid waste produced by research reactors owing to the variation of their types and operation modes.Moreover,the standards available for research reactors are still lacked and relevant standards on nuclear power plants and nuclear fuel reprocessing plants can only be taken as references.Therefore,the treatment of design safety issues need to be adjusted according to specific facilities.In this paper,the proper safety requirements for research reactor liquid waste storage facilities are proposed by the discussion of several treatments of key safety problems in the evaluation.
Manufacturing License Review of Military Nuclear Safety Mechanical Equipment—Mockup-Manufacture and Review Key Points
Zhang Xiaowei, Wang Xiaorong, Sun Dequan, Li Tao, Huang Xingrong
2016, 37(S1): 103-106. doi: 10.13832/j.jnpe.2016.S1.0103
Abstract:
The mockup manufacture is one important part of the manufacturing license review for military nuclear safety mechanical equipment.In this paper,it was introduced that the key points in mockup manufacture review such as mockup selection principles,standard specifications,key process and test contents.Moreover,the test standards,items and sequence which were of necessity in the identification test of active mechanical equipment mockup of various kinds were elaborated in this paper,to provide references for similar equipment review afterwards.
Safety Review and Design Alteration for Low Temperature Reactor Control Room
Xu Jianhua, Qin Legang, Sun Dequan, Zhao Yulong
2016, 37(S1): 107-110. doi: 10.13832/j.jnpe.2016.S1.0107
Abstract:
The reactor control room is the most concentrated area of the human-machine interface,and also the location of most mis-operation.Improving the control room design is one of the key factors to improve the reactor safety.Based on the characteristics of the low temperature reactor,safety review is conducted according to the regulation and standards related to nuclear power plants,and making reference to the design rule and function requirements of the low temperature reactor.According to the safety review requirements,designers changed the related design of the control room for optimization,so as to improve the safety of facilities.
Methodology of Developing the Emergency Action Levels of Uranium Conversion Facility
Wu Jing, Dong Bo, Ma Wencai, Huang Jing
2016, 37(S1): 111-114. doi: 10.13832/j.jnpe.2016.S1.0111
Abstract:
We analyzed the safety features and potential risks of the uranium conversion plant.The analytical method of the UF6 or HF leakage accident was presented.We gave the justification of working out the EAL(Emergency Action Level) of the uranium plant on the basis of investigations of the practice taken by International Atomic Energy Agency(IAEA) and US Department of Energy(DOE).The rules of assessing emergency states and the risk features were also considered in the research.
Study on Methodology for Nuclear Accidents Control and Integrated Emergency Actions of Advanced Pressurized Water Reactors
Wang Linbo, Gou Feng, Dong Bo, Zhang Qiang, Zhang Huanchao, Ding Tongwei
2016, 37(S1): 115-117. doi: 10.13832/j.jnpe.2016.S1.0115
Abstract:
With the strategy of defense in depth,a methodology for nuclear accidents control and integrated emergency actions of the advanced pressurized water reactors is proposed.The integrated actions there should be included:less probabilities of occurrence of a large amount radiation release;consummate the bylaws for the nuclear power plant operation and accident management;emergency preparation and response between different organization levels;national and collectivize wrecking abilities construction.
Nuclear Safety Analysis of First Order Liquefier-Condenser in Uranium Purification and Conversion Plant
Dong Bo, Wu Jing, Ma Wencai, Ding Tongwei
2016, 37(S1): 118-121. doi: 10.13832/j.jnpe.2016.S1.0118
Abstract:
We estimated the safety features and the typical problems of the first order Liquefier-Condenser according to the laws and standards of the nuclear safety,including anti-seismic,alleviating of leakage accident,accurate weighing and so on.We made several critical recommendations of designing according to the points mentioned,in order to enhance the intrinsic safety of the plant.
Nuclear Criticality Safety Analysis for UF6 Conversion to Uranium Metal
Zhang Qiang, Wang Linbo, Ding Tongwei, Huang Jing
2016, 37(S1): 122-126. doi: 10.13832/j.jnpe.2016.S1.0122
Abstract:
Using the current nuclear criticality safety standards in China and MCNP4C code,the nuclear critical safety analysis and evaluation were carried out for UF6 conversion to uranium metal production line.The nuclear critical benchmark experiment data for international publication were selected,and the deflection and subcritical limiting value with MCNP4C code in the evaluated system were confirmed.The neutron effective multiplication factor for the copper post in the natural condition and accident condition were calculated and analyzed by taking conservative assumptions,and the production line was evaluated according to the comparison of nuclear criticality safety standards.The results show that the subcritical control parameters or the maximum neutron effective multiplication factor is less than the corresponding subcritical limits,and it is in the state of subcritical safety.
Discussion on Safety Regulation Major Concerns of Sodium-Cooled Fast Breeder Reactor
Ding Tongwei, Dong Bo, Zhang Qiang, Wang Linbo, Huang Jing
2016, 37(S1): 127-130. doi: 10.13832/j.jnpe.2016.S1.0127
Abstract:
Sodium-cooled fast breeder reactor(SFR)is recognized as one of the most mature and most promising commercial fast reactors.However,due to the aspects of materials,coolant safety and economics,SFR in China is still in the operation stage of experimental reactor at present.Due to the lack of SFR’s regulations,standards,technology and experience of the safety regulation,the regulatory work is facing great challenges.In this paper,SFR is compared with PWR.The characteristics of SFR and safety regulatory are combined to propose 15 items of major regulatory concerns,such as core,system and equipment,and a series of related suggestions is given.
Quantitative Analysis on Typical Initial Events of Nuclear Fuel Cycle Facilities
Li Feng, Wang Xiaorong, Qin Legang, Zhang Huanchao, Zhao Mi
2016, 37(S1): 131-134. doi: 10.13832/j.jnpe.2016.S1.0131
Abstract:
The characteristics of initial events(IEs) for nuclear fuel cycle facilities were analyzed.According to the model of IEs and failure mechanism of IEs for hardware of nuclear fuel cycle facilities,the method for nuclear fuel cycle IEs quantification was developed.For IEs caused by complex factors,a fictitious model with the same process characteristics was analyzed as an example.The frequency of some IEs,such as static spark,pipe plugs leakage,leakage caused by corrosion,solvent detector failure,were calculated through PSA analysis software.
Study on Flood Control Standard for Nuclear Fuel Cycle Facilities
Kong Qingjun, Xie Maolin, Sun Dequan, Zhao Yulong, Wang Jing
2016, 37(S1): 135-138. doi: 10.13832/j.jnpe.2016.S1.0135
Abstract:
After Fukushima nuclear accident in Japan,all the countries in the world are more and more concerning the potential impact of external events on nuclear facilities,including the potential impact of the flood.However,we are lacking of a reasonable flood control standard for nuclear fuel cycle facilities in China currently,which will not only hinder the flood protection for nuclear facilities,but also affect the safety of nuclear facilities directly.Based on the potential risks of nuclear facilities,referring to the existing four classes of nuclear facilities risk classification,considering the importance of nuclear facilities,economic losses and environmental consequences caused by flood,this paper proposes a flood control standard for each risk categories of nuclear facilities.
Hazards Analysis for Decommissioning of HWRR Core Structure
Zhang Huanchao, Gou Feng, Dong Bo, Wang Linbo, Zhang Yu
2016, 37(S1): 139-141. doi: 10.13832/j.jnpe.2016.S1.0139
Abstract:
According to the structure of HWRR,the dismantling sequence and the decommissioning process,the main logic diagram(MLD) and failure mode and effect consequence analysis(FMECA) were carried out separately to identify the potential hazards in the decommissioning activities of HWRR.The key potential hazards were identified and categorized.
Safety Measures of US Navy Nuclear Powered Ship Reactor Plant
Lan Yang, Zhang Yue
2016, 37(S1): 142-144. doi: 10.13832/j.jnpe.2016.S1.0142
Abstract:
The U.S.Nuclear Powered Warships have safely operated for more than 50 years without experiencing any reactor accident or any release of radioactivity that hurt human health or had an adverse effect on marine life.Naval reactors have an outstanding record of over 151 million miles safely steamed on nuclear power,and they have amassed over 6500 reactor-years of safe operation.Based on general safety principles and safety requirements,safety regulation and safety technology of the US naval reactor plant,the paper studies the safety measures of the US Nuclear Powered ship reactor plant.