Chapter 4—The General Electric Company Surveillance Programs for Boiling Water Reactor Pressure Vessels

2009 ◽  
pp. 132-132-19
Author(s):  
JP Higgins ◽  
JN Kass
Keyword(s):  
Pressure Vessels ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
General Electric ◽  
General Electric Company ◽  
Water Reactor ◽  
Electric Company ◽  
Surveillance Programs ◽  
Reactor Pressure

Failure Probability Assessment for a Boiling Water Reactor Pressure Vessel Under Low Temperature Over-Pressure Event

2012 ◽  
Author(s):  
Hsoung-Wei Chou ◽  
Chin-Cheng Huang ◽  
Bo-Yi Chen ◽  
Hsien-Chou Lin ◽  
Ru-Feng Liu
Keyword(s):  
Low Temperature ◽  
Failure Probability ◽  
Pressure Vessel ◽  
Fracture Probability ◽  
Pressure Vessels ◽  
Reactor Pressure Vessel ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Water Reactor ◽  
Reactor Pressure

The fracture probability of a boiling water reactor pressure vessel for a domestic nuclear power plant in Taiwan has been numerically analyzed using an advanced version of ORNL’s FAVOR code. First, a model of the vessel beltline region, which includes all shell welds and plates, is built for the FAVOR code based on the plant specific parameters of the reactor pressure vessel. Then, a novel flaw model which describes the flaw types of surface breaking flaws, embedded weld flaws and embedded plate flaws are simulated along both inner and outer vessel walls. When conducting the fracture probability analyses, a transient low temperature over-pressure event, which has previously been shown to be the most severe challenge to the integrity of boiling water reactor pressure vessels, is considered as the loading condition. It is found that the fracture occurs in the fusion-line area of axial welds, but with only an insignificant failure probability. The low through-wall cracking frequency indicates that the analyzed reactor pressure vessel maintains sufficient stability until either the end-of-license or for doubling of the present license of operation.

The BWRVIP Integrated Surveillance Program

2004 ◽  
Author(s):  
Robert G. Carter ◽  
Timothy J. Griesbach ◽  
Timothy C. Hardin
Keyword(s):  
Research Program ◽  
Pressure Vessel ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Surveillance Program ◽  
Water Reactor ◽  
Weld Material ◽  
Surveillance Programs ◽  
Reactor Pressure ◽  
The U.S

Boiling Water Reactor (BWR) plants in the U.S. are designed with radiation surveillance programs. However, the surveillance materials in some plants do not necessarily represent the limiting plate and/or weld material of the reactor pressure vessel (RPV). Also, some plants do not have baseline data for the surveillance materials, which is needed to measure irradiation shift. In 1998 the BWR Vessel and Internals Project (BWRVIP) conceived the BWR Integrated Surveillance Program (ISP) to address these concerns. The ISP surveyed all BWR vessel limiting materials and all available BWR surveillance materials (including materials from a 1990s supplementary research program called the Supplemental Surveillance Program, or SSP). For each vessel limiting weld and limiting plate, a best representative surveillance material was assigned, based on heat number, similar chemistries, common fabricator, and the availability of unirradiated data. Many of the selected surveillance materials are good representatives for the limiting materials of multiple plants, so fewer capsules are required to be tested, reducing the overall cost of surveillance while also improving BWR fleet compliance with 10CFR50 Appendix H.

Failure Frequency Estimates for Boiling Water Reactor Pressure Vessels

2010 ◽  
Author(s):  
F. A. Simonen ◽  
S. R. Gosselin ◽  
J. E. Rhoads ◽  
A. T. Chiang
Keyword(s):  
Pressure Vessels ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Thermal Transients ◽  
Core Damage ◽  
Water Reactors ◽  
Vessel Rupture ◽  
Reactor Pressure

This paper reviews estimates of rupture frequencies for reactor pressure vessels (RPV) at boiling water reactor (BWR) nuclear power plants as reported in the literature. Results permit improved probabilistic risk assessments (PRA) for severe accidents that could cause core damage and/or challenge the capabilities of BWR reactor containment systems. Current and historical estimates of failure frequencies are considered for light water reactors in general and more specifically for BWR plants. The focus is on large ruptures that could give flow rates exceeding the rates associated with double ended breaks of large diameter recirculation piping. Rupture frequencies for BWR vessels as used for PRA evaluations have historically been assigned low values (e.g. 10−7 to 10−6 per vessel per year). The objective of the present work was to establish possible technical bases for more realistic values of rupture frequency (i.e. 10−8). Historical estimates from the early WASH-1400 reactor safety study were first reviewed and used as a point-of-reference. More recent estimates came from various sources such as a U.S. Nuclear Regulatory Commission expert elicitation process that estimated Loss-of-Coolant Accident (LOCA) frequencies. Other studies both by industry and by the USNRC have addressed rupture frequencies for BWR vessels subject to low-temperature-over-pressure (LTOP) events. On the other hand, recent comprehensive evaluations have focused mostly on RPV failure frequencies for pressurized water reactors (PWRs) caused by pressurized thermal shock events. An important consideration was that rupture frequencies for BWR vessels are believed to be lower than those for PWR vessels, because BWR vessels are less embrittled than PWR vessels and are subject to less severe thermal transients. The review concludes that prior studies support an estimate of 10−8 or less for BWR vessel rupture frequencies. Probabilistic fracture mechanics calculations for individual vessels accounting for plant specific conditions are recommended to support even lower estimated frequencies. Use of more realistic vessel rupture frequencies in a plant’s PRA provides an improvement in not only the perceived plants risk of core damage, but also provides better decision making for plant operation and maintenance activities in that a conservative initiating event treatment within a PRA can mask other initiating events of higher importance.

Stress analyses for reactor pressure vessels by the example of a product line '69 Boiling Water Reactor

Kerntechnik ◽  
2011 ◽  
Vol 76 (4) ◽  
pp. 225-230 ◽  
Author(s):  
L. Mkrtchyan ◽  
H. Schau ◽  
W. Wolf ◽  
W. Holzer ◽  
R. Wernicke ◽  
...  
Keyword(s):  
Pressure Vessels ◽  
Product Line ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Water Reactor ◽  
Reactor Pressure ◽  
Stress Analyses
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