核动力工程

Safety Design Technology for Civil Engineering of Nuclear Island Building in Taishan Nuclear Power Plant

Hu Zhengchun, Liang Junfeng, Wan Zhijun, Cui Jianjun

2015, 36(S1): 1-4. doi: 10.13832/j.jnpe.2015.S1.0001

Abstract(11) PDF(0)

Abstract:
This paper describes the design features for the civil engineering of the nuclear island building in Taishan Nuclear Power Plant, elaborates the roles of the nuclear island building with European nuclear technology of third generation（EPR） in three basic security functions, including the reactivity control, cooling of nuclear fuel and radioactive substances containing to ensure the nuclear safety. The purpose is to enhance the public understanding of the multilayer security defense system of CEPR Nuclear Power Plant.

Design of CEPR Reactor Protection System

Li Guang, Liu Jianjun, Wan Lei, Liu Rui

2015, 36(S1): 5-8. doi: 10.13832/j.jnpe.2015.S1.0005

Abstract(10) PDF(0)

Abstract:
The paper presents the design concept, functions implemented, architecture, maintenance, periodic test, qualification and platform test of CEPR RPR system. Almost all the maintenance activities and periodic tests can be performed under the normal plant condition thanks to four redundancies design. Moreover, the intelligent tools are introduced for maintenance activities and periodic tests, which can greatly reduce the human error.

Optimization Analysis of Synthesized FSS Simulation Method Applied for DCS Design Evolution

Li Peijin, Zhou Weizhang, Xu Liangjun, Zhang Liqiang

2015, 36(S1): 9-13. doi: 10.13832/j.jnpe.2015.S1.0009

Abstract(10) PDF(0)

Abstract:
Due to the change of the process system design, the DCS design of the nuclear power plants is subject to a continuous evolution process, which brings constraint to FSS project implementation. This paper introduces a synthesized FSS simulation method which includes several general simulation methods（such as stimulation,emulation and simulation） to reduce such kind of constraint.

Automatic Diagnosis Used in CEPR State Oriented Procedures Emergency Operation Procedures

Lin Zhenhua, Huang Huiming, Shen Yunbin

2015, 36(S1): 14-16. doi: 10.13832/j.jnpe.2015.S1.0014

Abstract(11) PDF(0)

Abstract:
The diagnosis should be conducted before the nuclear power station enters into emergency operation procedure, and the diagnosis procedure will lead the operators to implement the emergency operation procedure. At present, all the domestic in-service nuclear power stations adopt the manual diagnosis, but for the 3rd generation units CEPR in Taishan nuclear power plant adopts the latest technique – the automatic diagnosis strategy. In this paper, a simple comparison of the two methods adopted for accident treatment in nuclear power stations is included, and the detail presentation about the logical flow of automatic diagnosis is emphasized. Besides, the feasibility and superiority of automatic diagnosis adopted in the nuclear power station is analyzed.

New Features of Turbine I＆C System in Taishan NPP

Liu Kangtai, Ren Zhenguo, Huang Xingyou, Ma Li

2015, 36(S1): 17-19. doi: 10.13832/j.jnpe.2015.S1.0017

Abstract(10) PDF(0)

Abstract:
Compared with CPR1000, the third generation CEPR nuclear power plant uses the third generation of control, protection and supervise system of turbine for the first time. Equipment redundancy, performance and logic design have been optimized, therefore the reliability of control and protection have been improved. This paper compares the two techniques, introduces and analyzes the new features of this third generation technique, for the purpose of supplying information for the operation and maintenance. Since the new system will be used by the following nuclear power stations, this paper will share the information of the new features with the same trade.

Configuration and Nuclear Safety of CEPR Multiple On-site Power Supply System

Chu Shaoxian, Yang Jicheng, Zhu Wangqiang, Shi Qing

2015, 36(S1): 20-23. doi: 10.13832/j.jnpe.2015.S1.0020

Abstract(10) PDF(0)

Abstract:
Taishan CEPR Nuclear Power Plant implements the world advanced third nuclear power technology. CEPR nuclear safety futures are specially integrated into the design of electrical power supply systems. The multiple on-site power sources configured in CEPR electrical power supply system indicate the characteristics of redundancy, diversity and separation, and satisfy the principle of Defense-in-Depth. In the design of NI safety power supply systems, the emergency power supply in different levels provide enough power supply under all Design Basic Conditions and Design Extension Conditions, for which they are fully demonstrated by the safety analysis under loss of off-site power supply systems.

Application of 4D Simulation in the Installation of Nuclear Island Primary Loop Equipment

Guo Xinwei, Liu Yu, Man Xiaoyu

2015, 36(S1): 24-25. doi: 10.13832/j.jnpe.2015.S1.0024

Abstract(7) PDF(0)

Abstract:
In order to minimize the potential risks during the first time application of new erection technology of "flip in the air" and "inclined introduction" for the nuclear island primary loop equipment, this paper presents a new approach of simulation to be used for the nuclear island primary equipment erection, which can perform dynamic simulation and process verification for primary equipment erection, and carry out collision inspection and space quantification between the equipment and the civil work structure of nuclear island building, and has been successfully applied for the installation of nuclear island primary equipment such as reactor vessel and Pressurizer.

Optimization of CEPR NI Mechanical Equipment Material Design

Xiao Kaihua, Ning Fangmao, Lei Yawei, Li Zhongliang

2015, 36(S1): 26-29. doi: 10.13832/j.jnpe.2015.S1.0026

Abstract(6) PDF(0)

Abstract:
Considering the special requirements, optimization and selection principles for the mechanical material in nuclear power plants, the optimization of material in the CEPR unit is introduced in the terms of aging, fatigue, and corrosion degrading, to provide the reference for the selection of reliable, economic and safe material for domestic nuclear power units during the design stage.

Test Plan for Digital Control System in Third Generation PWR Project

Zhang Jian, Liu Shun, Zou Xiangyang, Wang Yi

2015, 36(S1): 30-33. doi: 10.13832/j.jnpe.2015.S1.0030

Abstract(10) PDF(0)

Abstract:
Based on the I ＆ C architecture of third generation PWR, this paper presents the test plan, and justifies the reasonability. After the test on various subsidiary instrumentation and control systems, the system integration test is carried out. During the test, the test completeness and reasonability was taken into account.

Overall Design of Instrumentation and Control for Third Generation PWR NPP

Zhang Jian, Li Guang, Zhang Congjie, Liu Shun

2015, 36(S1): 34-36. doi: 10.13832/j.jnpe.2015.S1.0034

Abstract(13) PDF(0)

Abstract:
This paper focus on the content and practice of the overall design of Third Generation PWR instrument and control. It mainly presents the methodology of overall I＆C design and the detailed implementation rules, and therefore describes how the methodology and rules are applied for the third generation pressurized water reactor project. And also, it presents the problems faced in the licensing process and provides experience for future projects.

Schedule and Impacted Factors for Taishan CEPR Integral Commissioning

Yu Weiming, Zhang Ke, Chen Shiji, Wu Jiabin

2015, 36(S1): 37-39. doi: 10.13832/j.jnpe.2015.S1.0037

Abstract(6) PDF(0)

Abstract:
Considering the problems in the preliminary test stages for the process systems of electrical, instrumentation and control and part of the nuclear island in Taishan CEPR NPP, combined the overall commissioning schedule and prerequisites, the potential risk during commissioning phase is analyzed, and the measures to reduce the effect of these risks on the overall commissioning schedule are provided.

Structure Features and Design Qualification of AFA3G LE Fuel Assembly

Yao Bo, Xiang Wenxin, Wang Peng, Ye Chen

2015, 36(S1): 40-43. doi: 10.13832/j.jnpe.2015.S1.0040

Abstract(8) PDF(0)

Abstract:
This paper introduces the design features of AFA3 G LE fuel assembly, and compares the difference to AFA3 G fuel assembly. Furthermore, the qualification tests for AFA3 G LE are briefly introduced.

Analysis of Design Features for CEPR Unit Overhauling

Wang Jun, Jia Guoan

2015, 36(S1): 44-46. doi: 10.13832/j.jnpe.2015.S1.0044

Abstract(15) PDF(0)

Abstract:
Analysis of the overall strategy for various overhauling of CEPR as well as the duration of the main activities in each overhauling window found that the new design features, and new tools and materials contributed to the short outage duration for CEPR unit overhauling. This paper points out the main difficulties faced by the short duration of overhauling for Taishan CEPR units.

Studies on Cross Connection between CEPR Auxiliary Power Supply Systems

Shi Qing, Yang Jicheng, Su Wanhua, Wang Yan

2015, 36(S1): 47-50. doi: 10.13832/j.jnpe.2015.S1.0047

Abstract(14) PDF(0)

Abstract:
Taishan CEPR is the third- generation advanced pressure water reactor. According to the redundancy requirements of safety consumers, the auxiliary power supply system in the nuclear island sets up as 4 independent trains. Different cross-connection between four trains can make sure that the safety consumers work as the redundancy requirement, and provide lighting and power for maintenance, which is the condition for improving the maintenance efficiency. Mechanical interlocking is the method to manage the switch-over between the cross-connection. "Special condition" can make sure the consistence between cross-connection and CEPR separation rule and operation technical specification.

CEPR Core Design and Fuel Management in Taishan Nuclear Power Plant

Wang Yonggang, Xiang Wenxin, Yao Bo, Lyu Yunjiang

2015, 36(S1): 51-53. doi: 10.13832/j.jnpe.2015.S1.0051

Abstract(15) PDF(0)

Abstract:
Taishan nuclear power plant（TNPS） has adopted CEPR technology, the 3rdgeneration PWR. Larger core scale provides lower linear power density and higher neutron economy. Mode T used in the control rods can improve the efficiency and reduce the wear of control rods. Using enriched boron optimizes the coolant chemistry. 18-month fuel management is introduced from the initial core. This feature improves the fuel economy in maximum.

Application of RSE-M 2010 in In-Service Inspection Program of CEPR Units

Shao Chunbing, Xiang Wenxin, Ning Fangmao, Xiao Kaihua

2015, 36(S1): 54-58. doi: 10.13832/j.jnpe.2015.S1.0054

Abstract(8) PDF(0)

Abstract:
The application of RSE-M2010 in the in-service inspection program of CEPR units is introduced in this paper, in terms of the latest evolution of RSE-M2010 and the CEPR design and construction, and the limitation and the solution are analyzed, to provide the new ideas to establish the specification for the in-service inspection of Chinese nuclear pressure equipment.

Optimization Design of Radiation Protection in CEPR Nuclear Power Plant

Ren Xueming, Ye Jianqiang, Ma Boyang

2015, 36(S1): 59-62. doi: 10.13832/j.jnpe.2015.S1.0059

Abstract(9) PDF(0)

Abstract:
This paper introduces the optimization design of radiation protection in CEPR unit, from the aspects including radiological zoning, source term control, maintenance optimization and room design. It demonstrates that the optimization design of radiation protection in CEPR is feasible and applicable, comparing to current PWRs, the anticipated annual collective dose per unit is in an advanced level.

Design Features and Engineering Practice of Radiation Monitoring System in CEPR Nuclear Power Plant

Peng Xuewei, Wang Ensheng, Ye Jianqiang, Ma Boyang

2015, 36(S1): 63-66. doi: 10.13832/j.jnpe.2015.S1.0063

Abstract(12) PDF(0)

Abstract:
It presents equipment selection, monitoring point arrangement of local gamma dose rate, and internal and external power supply design features of the radiation monitoring system in Taishan CEPR Nuclear Power Plant. Meanwhile, in order to easily operate and maintain the radiation monitoring system, the achievable engineering upgrading plan of the instrument control architecture is presented.

Key Techniques of CEPR Full Scope Simulator

Zhou Weizhang, Bian Xiushi, Deng Jijie, Cui Hao

2015, 36(S1): 67-70. doi: 10.13832/j.jnpe.2015.S1.0067

Abstract(11) PDF(0)

Abstract:
A third generation NNP technology, CEPR, is adopted in Taishan Nuclear Power Plant with the advanced digital control system（DCS） and other new techniques. Corresponding new features have developed for CEPR FSS. A combined using of DCS simulation and emulation is adopted for DCS database evolution. Core model COX3 D which calculates neutron flux distribution inside each assembly is used to improve fidelity of SPND simulation. A DCS initial condition maintenance tool was developed for DCS database tuning and updating. An integrated FSS monitor system has been built for effective monitor of trainee’s operation activities.

Construction of Double Containment of CEPR Reactor Building

Zhang Xiangke, Ma Guolu

2015, 36(S1): 71-74. doi: 10.13832/j.jnpe.2015.S1.0071

Abstract(8) PDF(0)

Abstract:
The design of the double containment for the reactor building is one of the major evolutions of the third-generation CEPR technology for improving nuclear safety, for which the structural design is more complex, and the construction is more difficult and the quality and schedule control is key to the project. To overcome the challenges of the new design and organize the construction in a proper and efficient way is of great significance to the self-reliance and economics of CEPR nuclear power technology. Based on the experiences gained from similar projects both home and abroad, the Taishan nuclear power project phase I adopted the innovative measures of steel liner design optimization, modularized construction and the construction sequence of the inner containment before the outer containment. Thanks to these efforts, the unit 1 from first concrete to dome lifting only took 24 months, with both quality and schedule superior to similar projects. This paper summarizes these experiences and gives some suggestions of improvement, so as to serve as a reference for the construction of similar projects in the future.

Study on Radiation Zone Adjustment of Reactor Plant during Refueling Outage in Qinshan Phase II NPP

Ke Haipeng, Wang Chuan, Ceng Jinzhong, Liu Jie

2015, 36(S1): 75-77. doi: 10.13832/j.jnpe.2015.S1.0075

Abstract(9) PDF(0)

Abstract:
In this paper, based on the analysis of the change of radiation level in the ten refueling outage of unit 1 reactor plant in Qinshan phase II nuclear power plant, the adjustment of radiation zoning in the reactor building during the refuel outage is proposed, in terms of the optimization of radiation protection and exposure dose control. The adjustment results show that the reduced radiation risk of reactor plant is visible, and most of the radiation zone can be degraded to yellow zone or green zone.

Analysis of Operation of Reactor Coolant Pump Shaft Seal System in Fangjiashan NPP

Liao Xiangling

2015, 36(S1): 78-81. doi: 10.13832/j.jnpe.2015.S1.0078

Abstract(15) PDF(0)

Abstract:
This paper describes the ANDRITZ coolant pump shaft seal system of Fangjiashan nuclear power plant by using the operation data obtained in the functional experiments. The paper also analyzes the operational risk of seal injection water, the effects on the running main pump when the main valve status is changed, and, in addition, the effects on the starting or running main pump when each level of main shaft seal system’s parameters is changed. The results show that the seal injection water has certain operational risks, while the main system pressure has little effect on the high or low pressure leakage of the seal injection water. The main system pressure should be higher than 2.75 MPa in order to make it safe when the main pump is started.

Analysis of Thread Corrosion Damage in Primary Manway of Steam Generator and Application of Thread Wire

Liu Qiang, Zhang Xiaoguang

2015, 36(S1): 82-84. doi: 10.13832/j.jnpe.2015.S1.0082

Abstract(8) PDF(0)

Abstract:
As the thread in the primary manway of the steam generator of Qinshan phase II nuclear power plant was damaged, the reason was analyzed in terms of material, stretching and equipment maintenance. Mechanical properties of the wire thread insert were studied and the thread is repaired by the thread wire technology. The test result shows that the stretching of the bolt installed with wire thread insert meets the requirement, and the engineering application of the thread wire technique on the steam generator can provide a reference for solving the similar problems in other nuclear power plants.

Analysis and Processing of Failure of ATE Pump Mechanical Seal for Qinshan NPP 3#, 4# Units

Wang Yunxi, Zhang Qiang, Xu Shuqing, Chen Jun, Han Chao

2015, 36(S1): 85-88. doi: 10.13832/j.jnpe.2015.S1.0085

Abstract(8) PDF(0)

Abstract:
The structural of the ATE pump for Qinshan NPP 3# and 4# Units is analyzed, and the operation parameters and the design criteria of the mechanical seal is contrasted. Four factors that may lead to the seal failure which refer to too large seal face line speed and pressure and insufficient auxiliary flushing water flow and pressure are theoretically analyzed. It shows that the too large seal face pressure of the mechanical seal is the root cause which causes the mechanical seal burned. According to the conclusion, a set of re-mounted, balanced cartex-sn28/110-00 mechanical seal replaces the bulk-type and unbalanced mechanical seal M74N4/110-G6, which completely eliminates the frequently burned mechanical seal of the ATE pump.

Experimental Analysis of Water Supply of Auxiliary Feed Water System after Breakup of One Feed Water Line

Deng Antao

2015, 36(S1): 89-92. doi: 10.13832/j.jnpe.2015.S1.0089

Abstract(13) PDF(0)

Abstract:
This test is one of the projects specified in the commissioning program for Fangjiashan nuclear power units, which is mainly to verify that when the water is supplied by the auxiliary feed water system with one broken feedwater line, the total flow from the break shall not be more than 250 m³/h, and the feed water flow to each of two intact steam generator（SG） shall not be below 45 m³/h. The difficulty lies in the fact that there will be no break in the test, and no flow data can be measured directly. A model is established to analyze the operation point of the auxiliary feedwater pump during the breakup accident, and fit the unknown parameters in the model with the data from normal tests. And then we can calculate the water supply flows to the breakup and normal lines with all the operation point data. Analysis shows that the water supply data meets the requirements of the acceptance criteria.

Water Supply Facility under Emergency Condition and Safety Analysis of Qinshan Nuclear Power Plant Unit 1

Fu Rongzhen, Ma Mingze, Xiao Yanjun, hu Qiang

2015, 36(S1): 93-98. doi: 10.13832/j.jnpe.2015.S1.0093

Abstract(15) PDF(0)

Abstract:
One safety improvement considering Fukushima accident is proposed for Qinshan Nuclear Power Plant unit-1. Under the sever accident condition, the reactor residual heat of the reactor core is removed by supplying the emergency water to the primary and secondary loop with the mobile pump, and add water supply channel, interface, water resource and mobile pump. The effect of added facilities on the safety of normal operation is analyzed. The potential safety problem resulted from the emergency water supply is discussed, and the improvement or mitigation measures is proposed.

Research of Reliability Statistical Method for Equipment of Fangjiashan Nuclear Power Plant

Gao Weiguang

2015, 36(S1): 99-102. doi: 10.13832/j.jnpe.2015.S1.0099

Abstract(8) PDF(0)

Abstract:
This paper describes the calculation method of reliability statistics on Random Censoring. Using Fangjiashan N1-EAM information application platform and based on reasonable assumptions, the classification statistics is carried out on the data of the equipment defect, and the statistical methods for reliability data of Fangjiashan Unit is discussed by examples.

Outage Analysis and Maintenance Strategy of Processing Instrument System of Qinshan Nuclear Power Plant II

Nie Wei, Xu Linyan, Chen Pujie, Liu Linjuan

2015, 36(S1): 103-106. doi: 10.13832/j.jnpe.2015.S1.0103

Abstract(9) PDF(0)

Abstract:
The essay introduces the brief analysis of aging defects occurred in the Qinshan Nuclear Power Plant II. Combining with experiences of domestic and overseas nuclear power plants,the maintenance strategies such as attaching greater importance to the operating ambient of equipment cabinets, preventative maintenance, optimizing spare parts purchasing, upgrading and updating equipments, which guarantee reliability of the processing instrument system and security of nuclear power units.

Discussion For Some Questions in Debugging of Compressed Air System of Nuclear Power Plant

Lin Zhong

2015, 36(S1): 107-110. doi: 10.13832/j.jnpe.2015.S1.0107

Abstract(9) PDF(0)

Abstract:
Compressed air system provides compressed air what the whole nuclear power plant needed, and needs to meet the operation and Maintenance of the power facilities of all plant. With the experience during debugging of the compressed air system of Fangjiashan Nuclear practical, demonstrated the Reliability of the overall design of the system in terms of supply capacity of the system, isolate control and packing capacity at accident conditions; At the same time it has made a number of field-proven reasonably practical measures.

Mutual Standby and Debug of EAS and RIS under H4 Condition

Li Zheng

2015, 36(S1): 111-113. doi: 10.13832/j.jnpe.2015.S1.0111

Abstract(9) PDF(0)

Abstract:
After the loss of coolant accident（LOCA） for preventing low pressure safety injection pump or a security shell spray pump function complete failure（H4） beyond design basis accident, design the H4 pipeline, in H4 conditions using still available low pressure safety injection pump or containment spray pump to realize the functions of the reactor core, long-term cooling. Of H4 condition under the safety injection system（RIS） and the containment spray system（EAS） reserve for test, by selecting features: low and high performance of the low pressure safety injection pump and low characteristic of the containment spray pump to verify the parameters in accident conditions can still meet the requirements, at the same time verified to the reactor coolant system（RCP） cooling system, thermal injection, outlet flow of the pump meet the Qinshan nuclear power plant expansion project（Fang Shan nuclear power project） debugging safety standards in the outline.

Commissioning of Main Feed Water Control Valves after Replacement

Wang Changzheng

2015, 36(S1): 114-117. doi: 10.13832/j.jnpe.2015.S1.0114

Abstract(7) PDF(0)

Abstract:
This essay introduces commissioning tests after replacement of main feed water control valve of operating unit, analyzes risks of commissioning tests and provides counter-measures. It also mainly stresses problems emerged from commissioning tests of main feed water control valve as well as their treatment.

Experimental Analysis of RRA System Performance at Atmospheric Pressure

Xu Zhaoping

2015, 36(S1): 118-121. doi: 10.13832/j.jnpe.2015.S1.0118

Abstract(14) PDF(0)

Abstract:
Residual Heat Removal System（RRA） can be used to maintain the reactor coolant temperature at cold shutdown conditions to offer the time required for the refueling and maintenance operations. In the maintenance in the cold shutdown condition（with the fuel in the reactor）, this system will keep the reactor coolant temperature less than 60 centigrade for the repairman to get into the manhole of any one of the steam generators and maintain the reactor cooling. In this test, it requires to prove that in this condition the reactor coolant does not enter the steam generator bottom head to result in additional staff radioactive contamination when the primary loop high level alarm occurs. The residual heat removal pump is required to run without cavitation even at its designed maximum flow rate when the primary loop low level alarm occurs. The level of the primary loop should change slowly to avoid the instrument false when the RCP300 MN is put into use during the maintenance cold shutdown condition.

Fault Analysis and Improvement of Regional Power Monitor Loop for No.2 Shutdown System of Heavy Water Reactor

Xu Qinghua

2015, 36(S1): 122-124. doi: 10.13832/j.jnpe.2015.S1.0122

Abstract(11) PDF(0)

Abstract:
No.2 shutdown system of CANDU-6 heavy water reactor in Qinshan III Nuclear Power Plant tripped several times due to the failure of the regional power monitor loop. Typical three channel trip failures were analyzed. The amplifier and potentiometer in the regional power monitor loop were analyzed theoretically and practically. Potentiometer failure contributed to the channel trip. Abrasion of contact patch and resistance thread because of long-term adjust is the main reason of potentiometer, according to the analysis of the failure potentiometer. Maintenance strategies such as periodic check and potentiometer replacement are developed, and the reliability of the regional power monitor loop is improved.

I＆C Commissioning Strategies with Delayed Supply of DCS Equipments for Fangjiashan Nuclear Power Project

Xin Huimin

2015, 36(S1): 125-129. doi: 10.13832/j.jnpe.2015.S1.0125

Abstract(14) PDF(0)

Abstract:
On condition that the supply schedule of the Digital Contral System（DCS） equipments were delayed for Fangjiashan nuclear power project unit 1, series of strategies were adopted in I＆C commissioning to ensure the project progress, including the confirmation of DCS minimum system, separated supply of software and hardware, utilization of the PRE vision（prerequisite） in cover-opening functional tests of nuclear auxiliary system, adoption of temporary scheme in hydrostatic testing, and implementation of DEN on Site using domestic technology. Risk analysis, feasibility study and plan formulation were conducted for the strategies. The project practice indicated that the adopted strategies in I＆C commissioning were effective, and helpful for the control of project progress.

Analysis of In-Service Inspection Capability for SCC Cracks of INCONEL-600 Nickel-Base Alloy Welds

Xia Weiming, He Ziang, Xu Feng, Sun Jun

2015, 36(S1): 130-134. doi: 10.13832/j.jnpe.2015.S1.0130

Abstract(12) PDF(0)

Abstract:
Taking the weld inspection of Inconel-600 Ni-based alloy weld in Qingshan 300 MW NPP as an example, the disadvantages of current in-service inspection method and inspection frequency in the inspection of the SCC cracks of INCONEL-600 alloy welds are analyzed, and the corresponding solutions to Inconel-600 Ni-based alloy weld inspection are presented.

Seismic Margin Assessment for Qinshan Ⅱ Nuclear Power Plant

Yang Ning, hu Bin

2015, 36(S1): 135-138. doi: 10.13832/j.jnpe.2015.S1.0135

Abstract(9) PDF(0)

Abstract:
After the accident of Fukushima NPP, NNSA requires that Seismic Margin Assessment（SMA） should be implemented on all operational NPPs in China mainland. This paper describes the SMA process and result of 650 MW PWR unit in Qinshan II Nuclear Power Plant. It is shown that the SMA of 650 MW PWR unit satisfies the requirement of NNSA and complies with the EPRI-NP-6041 as well as the seismic safety margin.

Alteration of Air Conditioning Units in Simulator Building of Fangjiashan Nuclear Power Plant

Zhang Ganxing, Zhang Guoqing

2015, 36(S1): 139-142. doi: 10.13832/j.jnpe.2015.S1.0139

Abstract(9) PDF(0)

Abstract:
During the commissioning of the thermostatic and humid static air conditioning unit（THACU）, four THACUs in MCR and computer room are turned on and off frequently, which will threaten the safety and life span of the units. Meanwhile, when the four THACUs are in operation, the heater is running at 100% power to keep the environment under a constant temperature, which causes a serious waste of electrical power. Through the alteration, the problems are solved and the THACUs can operate properly and stably.

Inspection, Analysis and Optimization Management for Control Rod of Qinshan Ⅱ NPP

Wang Lingbin, Shi Zhonghua

2015, 36(S1): 143-146. doi: 10.13832/j.jnpe.2015.S1.0143

Abstract(8) PDF(0)

Abstract:
There is little experiences in the control rod operation of domestic nuclear power plants, and no control rod regulation or standard is available for the change of the control rods. In order to understand the control rod cladding, Qinshan II NPP checked the control rod cladding in service by nondestructive testing methods such as the ultrasonic and eddy current.,The reliable technical data for the operation of the control rods is obtained, to provide a basic for the adjustment and change of the control rod assemblies. In this paper, through the analysis and evaluation of the typical defect mechanism for the control rod assemblies, several measures to optimize the management of the control rods are proposed, to provide as a reference for the management of the control rods of other nuclear power plants.

Cause Analysis and Disposal Measures for Site Radiation Dose Increased When Transferring Fuel Assembly in Spent Fuel Pool

Shi Zhonghua, Deng Zhixin, Liao Zejun, Zhang Xuhui, Wang Lingbin

2015, 36(S1): 147-150. doi: 10.13832/j.jnpe.2015.S1.0147

Abstract(14) PDF(0)

Abstract:
During the unloading of the fuel assembly of two units in Qinshan phase II NPP, when the fuel assembly was transferred in the spent fuel pool, the site radiation dose increased. By the checking and analysis of the fuel assembly damage, the detection of the activated corrosion products, the calculation of the thickness of the water shielding layer above the fuel assemblies, and the measurement of the water shield capability and the local radiation shielding of the spent fuel pool, the root causes for the increasing of site radiation dose are found: during the unloading of the fuel assemblies, the water of the loading well is drained to the transportation pool, to meet the requirement for the fuel assembly unload conditions, and due to the loss of the shielding effect of the loading well water, the shielding of the spent fuel pool near the loading well became weaker. So when the fuel assembly is moved in this area, the radiation dose increased. Corresponding measures are taken to effectively avoid the reoccurrence of this event.

Reason Analysis and Corresponding Strategy for Cooling Water Intake Blockage at Nuclear Power Plants

Ruan Guoping

2015, 36(S1): 151-154. doi: 10.13832/j.jnpe.2015.S1.0151

Abstract(8) PDF(0)

Abstract:
Based on the investigation and analysis of intake blockage events on the global units in service, the cause of the cooling water intake blockage is analyzed, and the corresponding experience feedbacks and improvements are put forward by strengthening the identification and analysis of initiating conditions, reinforcing the design review and change, enhancing the maintenance strategy and control, and improving the response capability. Operation status of the intake of Qinshan units are analyzed and evaluated comprehensively, and the improved methods and preventive measures are put forward to prevent the invasion of aquatic organisms such as water hyacinth and breeding of sea creatures, and strengthen the maintenance and operation management of rotating screens.

Analysis of Manual Synchronization Surge Current of Emergence Diesel Generator in Qinshan Nuclear Power Plant

Zhou Renhuai

2015, 36(S1): 155-158. doi: 10.13832/j.jnpe.2015.S1.0155

Abstract(9) PDF(0)

Abstract:
In order to decrease the surge current of the generator when synchronized, the voltage frequency and phase position difference of the generator should satisfy the standard. Based on the manual synchronization of the emergence diesel generator, the influence of voltage frequency and phase position difference are analyzed briefly. The result shows that the phase position difference is key factor. The paper provides the assessment and improvement advice.

Configuration of Mobile Diesel Generator Vehicle for Nuclear Power Plants in Qinshan Region

Zhou Guohua, Lu Zhongbin, Xu Hongsheng

2015, 36(S1): 159-162. doi: 10.13832/j.jnpe.2015.S1.0159

Abstract(9) PDF(0)

Abstract:
Based on the typical successful path and the required load capacity for the removal of residual heat for the nuclear power plants in Qinshan region, three different methods are adopted to calculate the diesel generator capacity and analyze the selection. Selection of the diesel with smaller capacity and the generator with larger capacity can satisfy the allowable bus voltage drop when the plant directly starts the large capacity motor. It solves the difficult problem that the large capacity diesel cannot be incorporated into a mobile diesel generator vehicle. This paper analyzes the main considerable factors for a mobile diesel generator vehicle, such as the emergency and test selection switch.

Application of Vacuum Devices in NPP Primary Loop

Song Zhengchi

2015, 36(S1): 163-165. doi: 10.13832/j.jnpe.2015.S1.0163

Abstract(12) PDF(0)

Abstract:
Vacuum Devices by liquid ring vacuum pump, resulting in a more appropriate degree of vacuum in a loop, and combined gravitational water-filled way, the air steam generator U-tube out of the reach of a water-filled exhaust loop purpose. By applying this set of devices dramatically increased the primary circuit water-filled exhaust efficiency, save time, and be able to effectively optimize the debugging process and overhaul processes.

Study on Commission of Equipment Cooling Water System in Nuclear Power Plants

Li Heng

2015, 36(S1): 166-169. doi: 10.13832/j.jnpe.2015.S1.0166

Abstract(12) PDF(0)

Abstract:
Equipment Cooling Water System（RRI） is an important auxiliary system of nuclear power plants, and its users cover the fields of nuclear island and the periphery of nuclear power plant facilities systems（BOP）system. It must be ensured that the Equipment Cooling Water System can operate stably under normal operation condition and various accident conditions. Hence, equipment performance and logic functions of RRI must be verified one by one. This paper focuses on fluctuating tank test, motor pump test, flux adjustment test, and automatic switch test of different serials, analyzes and judges the problems occurred in the tests, and discusses the flux distribution method, hole-digging calculation method and the influence on the flux by the reverse installation of flow orifices, and the way to solve these problems.

Experimental Study on Seismic Qualification of Van and Coiler System for Nuclear Power Plants

Hua Xia, Liu Linlin

2015, 36(S1): 170-172. doi: 10.13832/j.jnpe.2015.S1.0170

Abstract(7) PDF(0)

Abstract:
Vacuum Devices by liquid ring vacuum pump, resulting in a more appropriate degree of vacuum in a loop, and combined gravitational water-filled way, the air steam generator U-tube out of the reach of a water-filled exhaust loop purpose. By applying this se

Application Status Analysis of Specimen Reconstitution Technique

Mo Huajun, Sun Kai

2015, 36(S1): 173-176. doi: 10.13832/j.jnpe.2015.S1.0173

Abstract(5) PDF(0)

Abstract:
This paper describes the existing cases of specimen reconstitution, and summarizes the process and characteristics of arc stud welding, electron-beam welding, laser welding and surface active joint. A comparative analysis of the advantages and disadvantages of these welding methods has been carried out, and the quality monitoring method in the process of reconstitution and after reconstitution is listed. The application of reconstitution technique in nuclear engineering is analyzed, the gap between our existing experiences and international advanced level is analyzed and the direction of improvement is pointed out. It is proved that arc stud welding and electron-beam welding are the most widely used techniques, and the laser welding is available for any type of specimen.

Research on Damages Caused by Valve Quenching Cracks

He Ziang, Chen Shu, Jiang Shenghan, Yin Kaiju, Chen Yong

2015, 36(S1): 177-179. doi: 10.13832/j.jnpe.2015.S1.0177

Abstract(9) PDF(0)

Abstract:
Work carried out with chemical analysis, microstructure analysis, crack fracture analysis and Rockwell hardness analysis find out that the materials chemistry and Rockwell hardness of the value meet the design requirements. The spool crack result of the ribbon of segregation in course of the size and the uneven distribution of carbides materials in the forging rolling producing is quenching crack.

Qualification of Steam Generator Alloy I-690TT U-Bend Tubes in ACP1000 Nuclear Power Plants

Li Lei, He Gening, Zhang Fuyuan, Huang Wei, Huo Meng, hou Ye

2015, 36(S1): 180-183. doi: 10.13832/j.jnpe.2015.S1.0180

Abstract(8) PDF(0)

Abstract:
In order to validate the whole manufacturing process and key technical parameters of nuclear steam generator heat transfer tubes and stabilize the properties during the tube mass production, an integrated technical qualification plan is established. This plan can evaluate the homogeneity of the chemical, mechanical and metallurgical properties, and can verify the effectiveness of non-destructive inspection methods of I-690 TT U-bend tubes. The plan has been implemented successfully in the domestic manufacture of ACP1000 steam generator tubes.

Experimental Investigation on Heat Transfer Characteristic of Saturated Boiling in Round Tube

Luo Feng

2015, 36(S1): 184-186. doi: 10.13832/j.jnpe.2015.S1.0184

Abstract(14) PDF(0)

Abstract:
An experiment on saturated heat transfer characteristic in round channel was carried on using water. The dimension of round tube isφ9.5 mm×1 mm with a 1000 mm heating length. The dimensionless Re_lo number and Bo number were used as important parameters to predict heat transfer coefficient in round channel. Based on the two dimensionless numbers, the data were used to develop a new correlation. The correlation was then tested against the experimental data, the error was about ±20%.

Irradiation Surveillance of RPV for Unit 1＆2 of QinshanⅡNPP

Jiang Guofu, Li Guoyun, Luan Xingfeng, Zhang Haisheng, Huang Juan, Yang Xu, Cao Jiebao, Sun Kai

2015, 36(S1): 187-190. doi: 10.13832/j.jnpe.2015.S1.0187

Abstract(10) PDF(0)

Abstract:
Tests for all 8 irradiation capsules of unit 1＆2 of QinshanⅡNPP were finished. Variation of strength, elongation and impact toughness of RPV steels with different fast neutron fluence was acquired. Upper shelf energy and ductile-to-brittle transition temperature（DBTT） shift were evaluated. Compared with non-irradiated materials, irradiation embrittlement effects were found in all specimens in irradiation capsules. DBTT shifts were lower than predicted value by FIS formula. Results showed that embrittlement effect of RPV steels for both units was relatively in low range.

Study on Calculation Method for Heat Transfer of Cooling Pipe in Pressurizer Relief Tank

Zhang Minjie, Tian Haowen, Mao Huihui, Chen Shu, Tian Yu, Liu Songtao, Gong Junyong

2015, 36(S1): 191-193. doi: 10.13832/j.jnpe.2015.S1.0191

Abstract(8) PDF(0)

Abstract:
In the design of the improved nuclear power plants, there are many changes in the system parameters of the pressurizer relief tank. As an important component for the heat transfer system in the pressurizer relief tank, the cooling pipe must be redesigned based on the heat transfer calculation. This paper analyzes the complexity of the heat transfer calculation, and gives a simplified model and relevant calculation method.

Mechanism Analysis and Improvement Measures for ThicknessReducing of Pipe behind Restriction Orifice in Deaerating System in Nuclear Power Plants

Tian Haowen, Zhang Tengfei, Ceng Xiaokang

2015, 36(S1): 194-197. doi: 10.13832/j.jnpe.2015.S1.0194

Abstract(8) PDF(0)

Abstract:
The pipe-wall thickness in a straight section, which is located behind a restriction orifice in deaerating system in 1# unit of Tianwan Nuclear Power Plant, reduced from 6.6 mm to 2.5 mm in a refueling cycle. It is prone to cause leakages that lead to a potential safety hazard for operation. This paper analyzes the location of abrasion and the mechanism of cavitation happened in this thickness-reducing pipe behind the restriction orifice, and comprehensively discusses the design deficiencies in the orifices of the deaerating system and the advantages of replacing the single-stage orifice by the multistage orifices. After the improvement, the operation requirements of water-supply pipelines in the deaerating system can be satisfied, and the erosion rate of pipelines can be effectively reduced.

Research on Irradiation Effects of Domestic RPV Materials

Mo Huajun, Liu Xiaosong, Li Guoyun, Pan Longxuan

2015, 36(S1): 198-200. doi: 10.13832/j.jnpe.2015.S1.0198

Abstract(11) PDF(0)

Abstract:
Reactor safety is mainly influenced by irradiation embrittlement and fracture toughness decreasing of reactor pressure vessel（RPV） steel. To evaluate RPV safety,it is important to research irradiation effect of RPV steels under high neutron fluence. Programs about the embrittlement research of Chinese domestic RPV materials by the irradiation tests are described in this paper. The results indicated that there is irradiation embrittlement and fracture toughness decreasing in high neutron exposure for Chinese domestic RPV materials under the operation condition.

Development of Modular Tube-Type Irradiation Rig

Liu Yang, Tong Mingyan, Yang Wenhua, Xu Bin

2015, 36(S1): 201-203. doi: 10.13832/j.jnpe.2015.S1.0201

Abstract(8) PDF(0)

Abstract:
In order to process the irradiation test in the high gamma-ray flux region, a modular tube-type irradiation rig（MTIR） is developed. The MTIR is with the functions of adjusting and measuring, and can cool the test section in the high temperature and gamma-ray flux region. The verification tests prove that the MTIR can satisfy the irradiation test requirements, and efficiently uses the inner radiation holes of the High Flux Engineering Test Reactor（HFETR）, to shorten the radiation test period for the material, and improve the HFETER radiation capability.

Development of Digital Protection System for China Mianyang Research Reactor

Chen Yi, Zhang Yang, Wang Mingshan, Huang Xiaojin, Huang Wen, Yao Jian, He Fang

2015, 36(S1): 204-206. doi: 10.13832/j.jnpe.2015.S1.0204

Abstract(14) PDF(0)

Abstract:
The digital protection system of China Mianyang Research Reactor has been developed, which is based on the parallel processing of MCU and FPGA, three checking channels, the partial logical, double judge. The analysis of the system shows that it is reliable and safe, any single fault can not hinder the implementation of the system’s safety function, and the failure probability is less than 10^-5. It has been running for four years to prove that it meets the requirement of China Mianyang Research Reactor.

Application of High-Power Laser Decontamination Technology in Nuclear Facility Decommissioning

Fan Kai, Zhao Yu, Zhang Yongling, Dai Bo

2015, 36(S1): 207-210. doi: 10.13832/j.jnpe.2015.S1.0207

Abstract(13) PDF(0)

Abstract:
In this study, the development status and applications of high-power laser decontamination technology at home and abroad are systematically investigated and summarized, as well as its characteristics and application occasions. A new high-power laser decontamination device is proposed. Its basic theory and system are discussed. The developing direction and prospect in China are also summed up.

2015 Vol. 36, No. S1