Design, Assembly, and Inspection of Advanced U-Bend/Antivibration Bar Configurations for PWR Steam Generators

1989 ◽  
Vol 111 (4) ◽  
pp. 371-377
Author(s):  
P. J. Langford
Keyword(s):  
Tube Bundle ◽  
Operating Experience ◽  
Development Program ◽  
Work Rate ◽  
Pressurized Water Reactor ◽  
Steam Generators ◽  
Flow Induced Vibration ◽  
Pressurized Water ◽  
Water Reactor ◽  
Vibration Tests

Tube vibration and wear potential in the U-bend region of pressurized water reactor (PWR) steam generators is reduced by enhanced design bases and fabrication procedures. Applicable vibration mechanisms are described and related to field experience to focus the development program which led to the enhancements. Technical bases were developed from flow-induced vibration tests and shaker tests in which the tube/antivibration bar (AVB) wear-producing forces and motions were characterized in terms of work rate. Fixtures to control weld shrinkage and instruments to measure tube/AVB fit-up were developed for fabrication. Assembly and application experience, including measurement of fabricated tube bundle/AVB fit-up, is summarized for several advanced steam generators. Implications for enhanced operating experience relative to conventional design configurations are noted.

Development of a Thermal Fatigue Screening and Evaluation Methodology for Pressurized Water Reactor Plants

2004 ◽  
Author(s):  
J. D. Keller ◽  
A. J. Bilanin ◽  
S. T. Rosinski
Keyword(s):  
Thermal Cycling ◽  
Development Program ◽  
Fatigue Cracking ◽  
Evaluation Methodology ◽  
Experimental Investigations ◽  
Pressurized Water Reactor ◽  
Primary Coolant ◽  
Modeling Tools ◽  
Pressurized Water ◽  
Water Reactor

Thermal cycling has been identified as a mechanism that can potentially lead to fatigue cracking in un-isolable branch lines attached to pressurized water reactor (PWR) primary coolant piping. A significant research and development program has been undertaken to understand the mechanisms causing thermal cycling and to develop models for predicting the thermal-hydraulic boundary conditions for use in piping structural and fatigue analysis. A combination of first-principles engineering modeling and scaled experimental investigations has been used to formulate improved thermal cycling modeling tools. This paper will provide an overview of the model development program, a summary of the supporting test program, and a description of the thermal cycling model structure. Benchmarking of the thermal cycling model against several PWR plant configurations is presented, demonstrating favorable comparison with cases where thermal stratification and cycling has been previously observed.

Analysis of PRHRS for an Integral Pressurized Water Reactor Using RELAP5/MOD3.4

2010 ◽  
Author(s):  
Xuhua Ye ◽  
Minjun Peng ◽  
Jiange Liu
Keyword(s):  
Natural Circulation ◽  
Heat Removal ◽  
Parameter Study ◽  
Pressurized Water Reactor ◽  
Steam Generators ◽  
Primary Coolant ◽  
Pressurized Water ◽  
Water Reactor ◽  
Pressure Peak ◽  
Residual Heat

An investigation on the thermal hydraulic characteristics of the passive residual heat removal system (PRHRS) which is used in an integral pressurized water reactor (INSURE-100) is presented in this paper. The main components of primary coolant system are enclosed in reactor vessel. Primary fluid flow circle is natural circulation. The PRHRS can remove the energy from the primary side as long as the residual heat exchanger (RHE) is submerged in the emergency cooldown tank (ECT). The parameter study is performed by considering the effects of an effective height between the steam generators and the RHE and a valve actuation time, which are useful for the design of the PRHRS. The mass flow in the PRHRS has been affected by the height difference between the steam generators and the RHE. The pressure peak of the primary side and PRHRS has been affected by the valve action time.

scholarly journals Experimental investigation of flow-induced vibration in the control rod guide cylinder of pressurized-water reactor

2019 ◽  
Vol 97 (1) ◽  
pp. 14-22
Author(s):  
Dao-Gang Lu ◽  
Hui-Min Zhang ◽  
Yuan-Peng Wang
Keyword(s):  
Experimental Investigation ◽  
Flow Rate ◽  
Experimental Results ◽  
Pressurized Water Reactor ◽  
Flow Induced Vibration ◽  
Pressurized Water ◽  
Control Rod ◽  
Water Reactor ◽  
Experiment Platform ◽  
Control Rods

An experiment was conducted to investigate flow-induced vibration (FIV) in the control rods of a pressurized-water reactor (PWR). Control rods and a guide cylinder (full scale compared with the real structure) were installed on an FIV experiment platform, with which flow distributions were simulated according to actual situations. The vibration displacements of the control rods were observed in different flow rate ranges of transverse and vertical flows. Several FIV characteristics of the control rods under different transverse and vertical flows were determined by analyzing the experimental results. A formula was also proposed to predict vibration.

Thermal Fatigue of Pressurized Water Reactor Coolant System Loop Drain Line Down Horizontal Elbow

2020 ◽  
Author(s):  
Yue Zou ◽  
Brian Derreberry
Keyword(s):  
Thermal Fatigue ◽  
Operating Experience ◽  
Fatigue Cracking ◽  
Screening Methods ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Water Reactor ◽  
Thermal Transients ◽  
Reactor Coolant ◽  
Coolant System

Abstract Thermal cycling induced fatigue is widely recognized as one of the major contributors to the damage of nuclear plant piping systems, especially at locations where turbulent mixing of flows with different temperature occurs. Thermal fatigue caused by swirl penetration interaction with normally stagnant water layers has been identified as a mechanism that can lead to cracking in dead-ended branch lines attached to pressurized water reactor (PWR) primary coolant system. EPRI has developed screening methods, derived from extensive testing and analysis, to determine which lines are potentially affected as well as evaluation methods to perform evaluations of this thermal fatigue mechanism for the U.S. PWR plants. However, recent industry operating experience (OE) indicate that the model used to predict thermal fatigue due to swirl penetration is not fully understood. There are limitations with the EPRI generic evaluation. In addition, cumulative effects from various thermal transients such as the reactor coolant system (RCS) sampling and excess letdown may also contribute to the failure of RCS branch lines. In this paper, we report direct OE from one of our PWR units where thermal fatigue cracking is observed at the RCS loop drain line close to the welded region of the elbow. A conservative analytical approach that takes into account the influence of thermal stratification, in accordance with ASME Section III Class 1 piping stress formula, is also proposed to evaluate the severity of fatigue damage to the RCS drain line, as a result of various transients. Finally, recommendations are made for future operation and inspection based on results of the evaluation.

Sludge collector performance in steam generators of pressurized water reactor

2000 ◽  
Vol 200 (1-2) ◽  
pp. 295-302 ◽  
Author(s):  
H Takamatsu ◽  
T Matsunaga ◽  
Robert M Wilson ◽  
T Kusakabe
Keyword(s):  
Pressurized Water Reactor ◽  
Steam Generators ◽  
Pressurized Water ◽  
Water Reactor
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