Comments on "Acoustics, stability, and compensation in boiling water reactor pressure control systems"

1975 ◽  
Vol 20 (6) ◽  
pp. 819-820
Author(s):  
D. Chong
Keyword(s):  
Control Systems ◽  
Pressure Control ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Water Reactor ◽  
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.

Bounding Acoustic Load Development for Boiling Water Reactor Shroud and Shroud Supports

2020 ◽  
Author(s):  
Matthew Walter ◽  
Minghao Qin ◽  
Daniel Sommerville
Keyword(s):  
Pressure Vessel ◽  
Reactor Core ◽  
Reactor Pressure Vessel ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Water Reactor ◽  
Structural Evaluation ◽  
Loss Of Coolant Accident ◽  
Specific Analysis ◽  
Reactor Pressure

Abstract As part of the license basis of a nuclear boiling water reactor pressure vessel, a sudden loss of coolant accident (LOCA) event needs to be analyzed. One of the loads that results from this event is a sudden depressurization of the recirculation line. This leads to an acoustic wave that propagates through the reactor coolant and impacts several structures inside the reactor pressure vessel (RPV). The authors have previously published a PVP paper (PVP2015-45769) which provides a survey of LOCA acoustic loads on boiling water reactor core shrouds. Acoustic loads are required for structural evaluation of core shrouds; therefore, a defensible load is required. The previous research compiled plant-specific data that was available at the time. Since then, additional data has become available which will add to the robustness of the bounding load methodology that was developed. Investigations are also made regarding the shroud support to RPV weld, which was neglected from the previous study. This will allow a practitioner a convenient method to calculate bounding acoustic loads on all shroud and shroud support welds in the absence of a plant-specific analysis.

Recapturing Net Positive Suction Head Margins in Boiling Water Reactor Emergency Core Cooling Systems

2017 ◽  
Author(s):  
Alan J. Bilanin ◽  
Andrew E. Kaufman ◽  
Warren J. Bilanin
Keyword(s):  
Cooling System ◽  
Cooling Water ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Cooling Systems ◽  
Water Reactor ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Core Cooling ◽  
Reactor Pressure

Boiling Water Reactor pressure suppression pools have stringent housekeeping requirements, as well as restrictions on amounts and types of insulation and debris that can be present in the containment, to guarantee that suction strainers that allow cooling water to be supplied to the reactor during a Loss of Coolant Accident remain operational. By introducing “good debris” into the cooling water, many of these requirements/restrictions can be relaxed without sacrificing operational readiness of the cooling system.

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.

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