Experience in Chemical Decontamination of PWR Systems and Components

2009 ◽  
Author(s):  
Claude Steinkuhler ◽  
Koen Lenie ◽  
Reginald Coomans
Keyword(s):  
Critical Path ◽  
Heat Removal ◽  
Volume Control ◽  
Pressurized Water ◽  
Coolant Pump ◽  
Regenerative Heat ◽  
Reactor Coolant ◽  
Chemical Decontamination ◽  
Waste Production ◽  
Water Reactors

Tecnubel has recently performed various chemical decontamination of French and Belgian Pressurized Water Reactors (PWR) systems and components. The purpose of this paper is to present and compare these experiences. The objectives of these operation were the reduction of the general surface contamination together with the elimination of hot spots in Residual Heat Removal Systems (RHRS), Chemical and Volume Control Systems (CVCS) and Reactor Coolant Pumps (RCP). This reduction of contamination leads to the reduction of dosimetry to the maintenance personnel and allows the works on critical equipment. An additional challenge for three of these projects lay in the execution of a complicated operation on the critical path of a reactor refueling shutdown. The chemical decontamination were performed by circulating an adequate fluid in the systems or around the components. Since the contamination was generated at hot conditions during power operation, a redox attack on the surface was necessary. The EDF systems and components were decontaminated using a qualified EDF process of the EMMAC family. The Reactor Coolant Pump from the Belgian PWR was treated with the NITROX process, qualified by Westinghouse. The functions required by the decontamination system were very diverse and therefore an existing decontamination loop, which was previously developed for the decontamination of small circuits, was re-developed and adapted for bigger volumes by DDR Consult and Tecnubel. The results of five decontamination are presented and detailed in terms of efficiency and waste production. These projects were: the chemical decontamination of the RHRS of Flamanville 1 NPP, of the CVCS non regenerative heat exchanger at St Laurent des Eaux NPP, of the RHRS and CVCS of Bugey 2 NPP and of two RCP at the Westinghouse Belgian Service Center.

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.

Numerical and Experimental Research on the Fluid-Induced Forces of Clearance Flow in Canned Motor Reactor Coolant Pump

2019 ◽  
Vol 141 (6) ◽  
Author(s):  
Rui Xu ◽  
Yun Long ◽  
Yaoyu Hu ◽  
Junlian Yin ◽  
Dezhong Wang
Keyword(s):  
Large Mass ◽  
Stiffness Coefficient ◽  
Pressurized Water ◽  
Coolant Pump ◽  
Reactor Coolant ◽  
Various Boundary Conditions ◽  
Clearance Flow ◽  
Mass Coefficient ◽  
Cfd Method ◽  
Canned Motor

Reactor coolant pump (RCP) is one of the most important equipment of the coolant loop in a pressurized water reactor system. Its safety relies on the characteristics of the rotordynamic system. For a canned motor RCP, the liquid coolant fills up the clearance between the metal shields of the rotor and stator inside the canned motor, forming a long clearance flow. The fluid-induced forces of the clearance flow in canned motor RCP and their effects on the rotordynamic characteristics of the pump are numerically and experimentally analyzed in this work. A transient computational fluid dynamics (CFD) method has been used to investigate the fluid-induced force of the clearance. A vertical experiment rig has also been established for the purpose of measuring the fluid-induced forces. Fluid-induced forces of clearance flow with various whirl frequencies and various boundary conditions are obtained through the CFD method and the experiment. Results show that clearance flow brings large mass coefficient into the rotordynamic system and the direct stiffness coefficient is negative under the normal operating condition. The rotordynamic stability of canned motor RCP does not deteriorate despite the existence of significant cross-coupled stiffness coefficient from the fluid-induced forces of the clearance flow.

The Design Features of Integrated Modular Water Reactor (IMR)

2004 ◽  
Author(s):  
Takashi Kanagawa ◽  
Masashi Goto ◽  
Shuji Usui ◽  
Tadahiko Suzuta ◽  
Akimi Serizawa ◽  
...  
Keyword(s):  
Control System ◽  
Water System ◽  
Heat Removal ◽  
Volume Control ◽  
Circulation System ◽  
Primary System ◽  
Main Theme ◽  
Two Phase ◽  
Water Reactor ◽  
Reactor Coolant

Small-to-medium-sized (300–600MWe) reactors are required for the electric power market in the near future (2010–2030). The main theme in the development of small-to-medium-sized reactor is how to realize competitive cost against other energy sources. As measures to this disadvantage, greatly simplified and downsized design is needed. From such point of view, Integrated Modular Water Reactor (IMR), which electric output power is 350 MWe, adopts integrated and high temperature two-phase natural circulation system for the primary system. In this design, main coolant pipes, a pressurizer, and reactor coolant pumps are not needed, and the sizes of a reactor vessel and steam generators are minimized. Additionally, to enhance the economy of the whole plant, fluid system, and Instrumentation & Control system of IMR have also been reviewed to make them simplest and smallest taking the advantage of the IMR concept and the state of the art technologies. For example, the integrated primary system and the stand-alone direct heat removal system make the safety system very simple, i.e., no injection, no containment spray, no emergency AC power, etc. The chemical and volume control system is also simplified by eliminating the boron control system and the seal water system of reactor coolant pumps. In this paper, the status of the IMR development and the outline of the IMR design efforts to achieve the simplest and smallest plant are presented.

Experimental Research on the Fluid Induced Forces of Clearance Flow in Canned Motor Reactor Coolant Pump

2018 ◽  
Author(s):  
Rui Xu ◽  
Yaoyu Hu ◽  
Yun Long ◽  
Junlian Yin ◽  
Dezhong Wang
Keyword(s):  
Large Mass ◽  
Stiffness Coefficient ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Coolant Pump ◽  
Reactor Coolant ◽  
Various Boundary Conditions ◽  
Clearance Flow ◽  
Mass Coefficient ◽  
Canned Motor

Reactor coolant pump is one of the key equipment of the coolant loop in a pressurized water reactor system. Its safety relies on the characteristics of the rotordynamic system. For a canned motor reactor coolant pump, the liquid coolant fills up the clearance between the metal shields of the rotor and stator inside the canned motor, forming a clearance flow. The fluid induced forces of the clearance flow in canned motor reactor coolant pump and their effects on the rotordynamic characteristics of the pump are experimentally analyzed in this work. A vertical experiment rig has been established for the purpose of measuring the fluid induced forces of the clearance. Fluid induced forces of clearance flow with various whirl frequencies and various boundary conditions are obtained through the experiment. Results show that clearance flow brings large mass coefficient into the rotordynamic system and the direct stiffness coefficient is negative under the normal operating condition. The rotordynamic stability of canned motor reactor coolant pump does not deteriorate despite the existence of significant cross-coupled stiffness coefficient from the fluid induced forces of the clearance flow.

Use of PRA in the Design of the Westinghouse AP1000 Plant

2009 ◽  
Author(s):  
Robert J. Lutz ◽  
James H. Scobel ◽  
Richard G. Anderson ◽  
Terry Schulz
Keyword(s):  
Cooling Water ◽  
Heat Removal ◽  
Accident Prevention ◽  
Severe Accident ◽  
Pressurized Water ◽  
Core Damage ◽  
Service Water ◽  
Containment Shell ◽  
Water Reactors ◽  
Accident Scenarios

Probabilistic Risk Assessment (PRA) has been an integral part of the Westinghouse AP1000, and the former AP600, development programs from its inception. The design of the AP1000 plant is based on engineering solutions to reduce or eliminate many of the dominant risk contributors found in the existing generation of Pressurized Water Reactors (PWRs). Additional risk reduction features were identified from insights gained from the AP1000 PRA as it evolved with the design of the plant. These engineered solutions include severe accident prevention features that resulted in a significant reduction in the predicted core damage frequency. Examples include the removal of dependencies on electric power (both offsite power and diesel generators) and cooling water (service water and component cooling water), removal of common cause dependencies by using diverse components on parallel trains and reducing dependence on operator actions for key accident scenarios. Engineered solutions to severe accident consequence mitigation were also used in the AP1000 design based on PRA insights. Examples include in-vessel retention of molten core debris to eliminate the potential for ex-vessel phenomena challenges to containment integrity and passive containment heat removal through the containment shell to eliminate the potential for containment failure due to steam overpressure. Additionally, because the accident prevention and mitigation features of the AP1000 are engineered solutions, the traditional uncertainties associated with the core damage and release frequency are directly addressed.

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