Fluid Flow Field in the Annulus of Boiling Water (BWR) Nuclear Reactors Under Normal and Accident Conditions

2009 ◽  
Author(s):  
Raju Ananth ◽  
Karen Fujikawa ◽  
Jay Gillis
Keyword(s):  
Fluid Flow ◽  
Velocity Field ◽  
Flow Field ◽  
Pressure Vessel ◽  
Theoretical Study ◽  
Nuclear Reactors ◽  
Boiling Water ◽  
Flow Conditions ◽  
Accident Conditions ◽  
Reactor Pressure

This paper presents a theoretical study of the velocity field in the annulus formed between the Reactor Pressure Vessel (RPV) and the shroud of a Boiling Water Reactor (BWR) under normal and accident flow conditions. Simplified geometry and an ideal irrotational flow are assumed to solve the problem using velocity potentials.

Fluid-Structure Coupling Analysis for the Flow Distribution Device of Nuclear Reactor Internals

2013 ◽  
Author(s):  
Mingqian Zhang ◽  
Yuangang Duan ◽  
Xiaobing Ran ◽  
Yanwu Liu
Keyword(s):  
Fluid Flow ◽  
Flow Field ◽  
Pressure Vessel ◽  
Flow Distribution ◽  
Reactor Pressure Vessel ◽  
Turbulent Model ◽  
Pressurized Water ◽  
Fluid Structure ◽  
Ansys Cfx ◽  
Reactor Pressure

In order to better understand the stresses to which the flow distribution device (FDD) is subjected for the pressure fluctuation, we need to improve our knowledge of the fluid flow inside the reactor pressure vessel (RPV). The flow field of the reactor lower plenum which is associated with a typical pressurized water reactor (PWR) is simulated by using ANSYS CFX code. Calculations have been carried out from reactor pressure vessel inlet to the core outlet. Grid sizes of million nodes with the k-epsilon turbulent model have been used with a porous zone approach for the reactor core space. Predictions of the steady-state pressure and velocity field have been made. The results are compared with the scaled experiment data in order to verify the accurate description of the fluid flow. Based on this verified turbulent model, a sub-domain is extracted from the lower plenum for the transient two-way Fluid-Structure Interaction (FSI) simulation which is limited by the computer capability and computing time. This transient analysis of fluid-structure coupling system is conducted by using CFX and ANSYS in numeric calculation of flow field and structure, with an exchanging platform MFX-ANSYS/CFX which can transfer fluid pressure and structure displacement between computational fluid dynamics (CFD) and computational structure dynamics (CSD) grid systems. The loose coupling method is used to investigate the transient dynamic response of the flow distribution device which is immerged in the bottom plenum. Dynamic stress and strain of the flow distribution device are given and discussed. This analysis practice can be guidance for the optimization design of reactor and improve our understanding of reactor components flow induced vibration phenomena.

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.

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.

The Development of Radiation Embrittlement Models for U.S. Power Reactor Pressure Vessel Steels

2006 ◽  
Author(s):  
J. A. Wang ◽  
N. S. V. Rao ◽  
S. Konduri
Keyword(s):  
Pressure Vessel ◽  
Prediction Models ◽  
Irradiation Temperature ◽  
Reactor Pressure Vessel ◽  
Boiling Water ◽  
Radiation Embrittlement ◽  
Pressurized Water ◽  
Water Reactor ◽  
Water Reactors ◽  
Reactor Pressure

The information fusion technique is used to develop radiation embrittlement prediction models for reactor pressure vessel (RPV) steels from U.S. power reactors, including boiling water reactors and pressurized water reactors. The Charpy transition temperature-shift data is used as the primary index of RPV radiation embrittlement in this study. Six parameters—Cu, Ni, P, neutron fluence, irradiation time, and irradiation temperature—are used in the embrittlement prediction models. The results indicate that this new embrittlement predictor achieved reductions of about 49.5% and 52% in the uncertainties for plate and weld data, respectively, for pressurized water reactor and boiling water reactor data, compared with the Nuclear Regulatory Commission Regulatory Guide 1.99, Rev. 2. The implications of dose-rate effect and irradiation temperature effects for the development of radiation embrittlement models are also discussed.

Radiation Effects in Boiling Water Reactor Pressure Vessel Steels

1980 ◽  
Vol 102 (2) ◽  
pp. 177-186
Author(s):  
J. N. Kass ◽  
A. J. Giannuzzi ◽  
D. A. Hughes
Keyword(s):  
Pressure Vessel ◽  
Radiation Effects ◽  
Statistical Evaluation ◽  
Reactor Pressure Vessel ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Shelf Energy ◽  
Water Reactor ◽  
Pressure Vessel Steels ◽  
Reactor Pressure

Effect of neutron irradiation on notch toughness properties of Boiling Water Reactor pressure vessel steels was determined. Samples from several heats of plate, weld metal, and forgings were irradiated to three different fluence levels and tested. A statistical evaluation of the data was conducted to determine regression analysis mean decreases in upper shelf energy and increases in transition temperature versus fluence.

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