Surge Protection of the Reactor Coolant Pump Motors in Nuclear Power Plants

2021 ◽  
Author(s):  
Choong-Koo Chang ◽  
Erastus Mwongela Musyoka
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Coolant Pump ◽  
Reactor Coolant ◽  
Surge Protection

A Study on Improvement Through Analysis of Cold Hydrostatic Test Cases in Korea Nuclear Power Plants

2010 ◽  
Author(s):  
Ki Sang Song ◽  
Kyeong Sik Chae
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Volume Control ◽  
Test Sequence ◽  
Coolant Pump ◽  
Reactor Coolant ◽  
Water Reactors ◽  
Under Construction ◽  
Design Pressure

The objective of this study is to analyze the effectiveness of the Cold Hydrostatic Test (CHT) process and determine the optimum method of completing a CHT through case studies in the Korea nuclear power plants. In this study, all the 9 CHT cases, performed for the past sixteen years (1993 to 2009) in Korea nuclear power plants, will be examined and evaluated. There are twenty (20) operating Units and eight (8) Units under construction at 3 nuclear facility sites in Korea. Among the 20 Units, only 4 Units at the Wolsong site are pressurized heavy water reactors (PHWR), the others are pressurized light water reactors (PWR). CHT is based on the requirements of ASME NB-6200 & NC-6220. CHT is a mandatory test to verify integrity of weld points and interfaces associated with the equipment and pipes of the Reactor Coolant System (RCS) pressure boundary. The design pressure of the RCS is 2,500psia (175.8 kg/cm2 a). The major steps of test sequence of a CHT is RCS filling, venting, heat up, pressurization and inspection. Reactor Coolant Pump (RCP) operation is utilized as thermal input to raise RCS temperature over 120 °C. The Chemical and Volume Control System (CVCS) Charging Pump, or temporary hydro pump is used to pressurize the RCS. CHT requires pressure to be raised and maintained more than 10 minutes at 1.25 times of design pressure, and then be depressurized and inspected at the design pressure of 2,500psia (175.8 kg/cm2 a). According to the analyzed results of the CHT cases, all CHTs were successfully conducted but there are several items which need to revised and modified for increased effectiveness of the CHT. These items include pressurizer manway gasket leakage, improper process of the procedure and others. In conclusion, the results of this study will be used to prevent similar errors and improve the effectiveness of the CHT for future nuclear power plants projects in Korea.

Structural integrity evaluation for interference-fit flywheels in reactor coolant pumps of nuclear power plants

2005 ◽  
Vol 19 (11) ◽  
pp. 1988-1997 ◽  
Author(s):  
June-soo Park ◽  
Ha-cheol Song ◽  
Ki-seok Yoon ◽  
Taek-sang Choi ◽  
Jai-hak Park
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Structural Integrity ◽  
Interference Fit ◽  
Reactor Coolant

The Application of Concurrent Installation of the RCS Cross Over Leg and the Reactor Vessel Internals

2010 ◽  
Author(s):  
K.-W. Park ◽  
J.-H. Bae ◽  
S.-H. Park
Keyword(s):  
Feasibility Study ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Reactor Vessel ◽  
Pressurized Water ◽  
Coolant Pump ◽  
Reactor Coolant ◽  
Study Results ◽  
Under Construction

The reactor vessel internals (RVI) of a pressurized water reactor (PWR) must be installed precisely in the reactor vessel (RV) according to the requirements for levelness, orientation and vertical alignments for its proper functions and structural integrity. For the precise installation, deformation of the RV should be controlled during the RVI installation. Traditionally, the RVI has been installed in the RV after the completion of welding work for large bore pipings in the reactor coolant system (RCS). To reduce installation time, the concurrent installation of the RVI and RCS pipings is investigated. This paper describes the feasibility study on the concurrent installation including the Finite Element Method (FEM) analyses of the RV deformation due to the welding and heat treatment of the pipings. Based on the feasibility study results, the optimum schedule of the RVI installation in parallel with the installation of the cross-over leg pipings (reactor coolant pump inlet pipings) and confirmation measurement locations are developed. Thereby the concurrent installation will be applied to the nuclear power plants under construction in Korea, and it is expected to reduce installation period of 2 months compared to the traditional sequential installation method.

Structural Integrity Evaluation of High Strength Anchor Bolt for Supporting Steam Generator and Reactor Coolant Pump in Nuclear Power Plants

2013 ◽  
Author(s):  
Eun-Mo Lim ◽  
Nam-Su Huh ◽  
Hee-Jin Shim ◽  
Chang-Kyun Oh ◽  
Hyun-Su Kim
Keyword(s):  
Steam Generator ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
High Strength ◽  
Nominal Stress ◽  
Coolant Pump ◽  
Anchor Bolt ◽  
Reactor Coolant ◽  
Anchor Bolts

In Korea, a fitness-for service evaluation for assuring structural integrity of high strength anchor bolts which support nuclear components such as steam generator and reactor coolant pump, has been one of the important issues in nuclear industry. The main failure mechanism of high strength anchor bolts supporting nuclear components might be degradation due to stress corrosion cracking and brittle fracture. In the present work, the structural integrity of high strength anchor bolts which are used to support steam generator and reactor coolant pump of one of the Korean older vintage nuclear power plants is evaluated by adopting a procedure proposed by Electric Power Research Institute (EPRI) based on an elastic fracture mechanics concept. In this EPRI’s procedure, an accurate estimation of nominal stress acting on the cross section of the bolt is a crucial element since a structural integrity of an anchor bolt is evaluated in the EPRI’s procedure using this nominal stress incorporating reference flaw factors reflecting effects of stress concentration due to bolt thread and reference sized surface crack. In this context, detailed elastic finite element stress analyses are firstly performed on the anchor bolt assemblies to come up with nominal stress in the cross-section of anchor bolt. As for loading condition, bolt pretention as well as normal and faulted loads of the anchor bolts were considered. In addition, the structural integrity of the anchor bolts is demonstrated by comparing nominal stresses of anchor bolts with the maximum allowable stresses obtained by using the EPRI’s reference flaw factors and critical fracture toughness. Furthermore, the accuracy of EPRI’s reference flaw factors which are derived on the assumption that reference sized surface crack is existed on the thread roots is investigated using 3-dimensional elastic finite element fracture mechanics analyses.

General Overview of CVCS Design for the TSN Project

2010 ◽  
Author(s):  
Antonio Ciriello ◽  
Man Liu
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Water Injection ◽  
Reactor Design ◽  
Auxiliary System ◽  
Volume Control ◽  
Primary System ◽  
Reactor Coolant ◽  
Design Concepts

This paper resumes the results of the collaboration between AREVA and CNPDC during the past two years for performing and achieving the basic design of EPR™ reactor CVCS system for the TSN NPP. The CVCS (Chemical and Volume Control System) is an essential auxiliary system of the PWR technology based nuclear power plants all around the world. In the EPR™ reactor design, as it is also in similar nuclear power plants, this auxiliary system has well determined functions, which are: reactivity control, reactor coolant volume control, coolant chemistry control, primary system main pumps seal water injection as well as the pressurizer auxiliary spray regulation for the Reactor Coolant System. In the EPR™ reactor design, the CVCS is mainly an operational system and only some valves and instrumentations take part at some specific safety functions, (e.g. Reactivity Control, Containment of Radioactive Substances). In the first part of this paper a general introduction to the EPR™ reactor CVCS technology, including the related safety functions and detailed operational functions of CVCS, is presented. In the TSN EPR™ reactor CVCS design, the system is divided into eight sections, (defined from RCV1 to RCV8). The corresponding detailed description of these sections, including their functions, structure and main components, as they have been implemented in the EPR™ reactor CVCS design for the TSN NPP, is then presented in the second part of this paper. In addition some specific design features for EPR™ reactor CVCS system for the TSN NPP, such as the hydrogenation station technology, are also focused in this paper. The reference power plant, concerning the CVCS design, for the TSN NPP is the FA3 NPP, but different design concepts have been implemented in the TSN NPP with regards to the coolant purification section (RCV2), and the coolant filtering in the reactor coolant pumps seal injection and leak-off lines.

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