Monitoring the outside environment of a nuclear power station with a boiling water type reactor

1970 ◽  
Vol 29 (1) ◽  
pp. 699-702
Author(s):  
V. A. Knyazev ◽  
P. I. Kotikov ◽  
V. G. Laptev ◽  
Yu. V. Chechetkin
Keyword(s):  
Nuclear Power Station ◽  
Nuclear Power ◽  
Boiling Water ◽  
Water Type ◽  
Power Station ◽  
Type Reactor

Probabilistic Fracture Analysis for Boiling Water Reactor Pressure Vessels Subjected to Low Temperature Over-Pressure Event

2010 ◽  
Author(s):  
Hsoung-Wei Chou ◽  
Chin-Cheng Huang ◽  
Bo-Yi Chen ◽  
Ru-Feng Liu ◽  
Hsien-Chou Lin
Keyword(s):  
Low Temperature ◽  
Nuclear Power Station ◽  
Nuclear Power ◽  
Probability Of Failure ◽  
Pressure Vessels ◽  
Boiling Water ◽  
Power Station ◽  
Circumferential Welds ◽  
Water Reactor ◽  
Reactor Pressure

With the development of probabilistic fracture mechanics (PFM) methods in recent years, the risk-informed approach has gradually been used to evaluate the structural integrity and reliability of the reactor pressure vessels (RPV) in many countries. For boiling water reactor (BWR) pressure vessels, it has been demonstrated that it is not necessary to perform the inservice inspections of beltline circumferential welds to maintain the required safety margins because their probability of failure is orders of magnitude less than that of beltline vertical welds, thus may well reduce the associated substantial cost and person-rem exposure. In Taiwan, however, the inservice inspections of shell welds still have to be performed every ten years per ASME Boiler and Pressure Vessel Code, Section XI inspection requirements for a BWR type Chinshan nuclear power station. In this work, a very conservative PFM model of FAVOR code consistent with that USNRC used for regulation is built with the plant specific parameters concerning the beltline shell welds of RPVs of Chinshan nuclear power station. Meanwhile, a hypothetical transient of low temperature over-pressure (LTOP) event which challenges the BWR RPV integrity most severely is also assumed as the loading condition for conducting the PFM analyses. Further, the effects of performance of inservice inspection are also studied to determine the benefit of the costly inspection effort. The computed low probability of failure indicates that the analyzed RPVs can provide sufficient reliability even without performing any inservice inspection on the circumferential welds. It also indicates that performing the inservice inspections can not promote the compensating level of safety significantly. Present results can be regarded as the risk incremental factors compared with the safety regulation requirements on RPV degradation and also be helpful for the regulation of BWR plants in Taiwan.

Management of water and chemicals at nuclear power station with channel type reactor and nuclear steam superheat

1971 ◽  
Vol 30 (2) ◽  
pp. 183-186 ◽  
Author(s):  
O. T. Konovalova ◽  
T. I. Kosheleva ◽  
V. V. Gerasimov ◽  
L. S. Zhuravlev ◽  
G. A. Shchapov
Keyword(s):  
Nuclear Power Station ◽  
Nuclear Power ◽  
Power Station ◽  
Type Reactor ◽  
Steam Superheat ◽  
Channel Type

Status of Pre-Operating Tests for the Latest ABWR in Hamaoka Nuclear Power Unit No. 5

2004 ◽  
Author(s):  
S. Yukinori ◽  
Y. Ohta ◽  
G. Yanase ◽  
H. Yoshida ◽  
Y. Yamamoto ◽  
...  
Keyword(s):  
Electric Power ◽  
Nuclear Power Station ◽  
Nuclear Power ◽  
Water System ◽  
Boiling Water ◽  
Boiling Water Reactors ◽  
Power Station ◽  
Reactor Building ◽  
Water Reactors ◽  
Station Unit

Kashiwazaki-Kariwa Nuclear Power Station Unit 6 and Unit 7 (K6 & K7) are the first advanced boiling water reactors (ABWR) in the world. Toshiba has been engaged in the design and construction of boiling water reactors (BWR) for Hamaoka Nuclear Power Station Unit 5 (H5) of Chubu Electric Power Co., Inc. and Higashidori Nuclear Power Station Unit 1 of Tohoku Electric Power Co., Inc. H5 has features of the simplified reactor coolant recirculation system including a reactor internal pump (RIP) and reduction in plant startup time using a fine-motion control rod drive (FMCRD). The H5 has been improved through additional modifications such as a reduction in the number of adjustable speed drives (ASDs) required by RIPs. This was managed through multi drive and the elimination of a shaft seal from FMCRDs by applying magnetic coupling. The basic performance of the modified FMCRD (S-FMCRD) has been verified by the joint study of Japanese operating utilities and vendors. In addition, various other tests are performed on the H5 to validate its actual component design during the manufacturing stage and pre-operation stage. Construction of H5 began in March 1999. Bedrock inspection was completed in May 2000, the reactor pressure vessel (RPV) installation was finished in July 2002, and power receiving was completed in December 2002. Currently, H5 is scheduled to start commercial operation in January 2005. Pre-operating tests began at the end of 2002. The utility systems test, like the MUWP (makeup water system), HVAC (heating, ventilation and air conditioning), etc, began two months prior to power-receiving, followed by both the RCCW (reactor building closed cooling water system) and RCWS (reactor building cooling sea water system) tests. The primary system pre-operating test, which is usually performed at the peak of the pre-operating test period, was completed by mid-February, 2004. Each test schedule was controlled and each test followed the predefined test plan. On the other hand, S-FMCRDs and control rods (CRs) were installed, followed by an RPV overpressure integrity inspection test conducted in September 2003. Following this test, RIPs and related control system confirmation tests was performed Subsequently, FMCRDs pre-operation test was planned to confirm performance levels. The modified design concepts, such as S-FMCRDs and multi drive ASDs for RIPs etc, were scheduled for verification during this phase.

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