The Function Test for the Reliability Confirmation of High Temperature and High Pressure Vessel for Irradiation

2010 ◽  
Author(s):  
Kook-Nam Park ◽  
Jong-Min Lee ◽  
Sung-Ho Ahn ◽  
Sang-Ik Wu ◽  
Young-Ki Kim
Keyword(s):  
High Pressure ◽  
High Temperature ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Temperature Cycling ◽  
Pressurized Water ◽  
Test Loop ◽  
Test Pressure ◽  
Test Fuel

The Fuel Test Loop (FTL) which is capable of an irradiation testing under a similar operating condition to those of PWR (Pressurized Water Reactor) and CANDU (CANadian Deuterium Uranium reactor) nuclear power plants has been developed and installed in HANARO, KAERI (Korea Atomic Energy Research Institute). It consists of In-Pile Section (IPS) and Out-of Pile System (OPS). The IPS, which is located inside the pool is divided into 3-parts; the in-pool pipes, the IVA (IPS Vessel Assembly) and the support structures. The test fuel is loaded inside a double wall, inner pressure vessel and outer pressure vessel, to maintain the functionality of the reactor coolant pressure boundary. The functional test and verification of the IVA were done through pressure drop, vibration, hydraulic and helium leakage tests. The brazing technique for the instrument lines has been checked for its functionality and performance. An IVA has been manufactured by a local technique and has finally been tested under high temperature and high pressure. The IVA and piping did not experience leakage, as the piping, flanges, and assemblies have been fully checked. Good data was obtained during the three cycle test which included a pressure test, pressure and temperature cycling, and constant temperature. A new concept of IVA, as easy attaching and separating test fuel at IVA will be designed thereafter.

The Capabilities and Limitations of Remote Visual Methods to Detect Service-Induced Cracks in Reactor Components

2006 ◽  
Author(s):  
Stephen E. Cumblidge ◽  
Steven R. Doctor ◽  
Michael T. Anderson
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Pacific Northwest National Laboratory ◽  
Nuclear Industry ◽  
Camera System ◽  
Pressurized Water ◽  
Water Reactors ◽  
Visual Testing ◽  
The U.S

Since 1977, the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Regulatory Research has funded a multiyear program at the Pacific Northwest National Laboratory (PNNL) to evaluate the reliability and accuracy of nondestructive evaluation (NDE) techniques employed for inservice inspection (ISI). Recently, the U.S. nuclear industry proposed replacing current volumetric and/or surface examinations of certain components in commercial nuclear power plants, as required by the ASME Boiler and Pressure Vessel Code Section XI, with a simpler visual testing (VT) method. The advantages of VT are that these tests generally involve much less radiation exposure and examination times than do volumetric examinations such as ultrasonic testing (UT). However, for industry to justify supplanting volumetric methods with VT, an analysis of pertinent issues is needed to support the reliability of VT in determining the structural integrity of reactor components. As piping and pressure vessel components in a nuclear power station are generally underwater and in high radiation fields, they need to be examined by VT from a distance with radiation-hardened video systems. Remote visual testing has been used by nuclear utilities to find cracks in pressure vessel cladding in pressurized water reactors, core shrouds in boiling water reactors, and to investigate leaks in piping and reactor components. These visual tests are performed using a wide variety of procedures and equipment. The techniques for remote visual testing use submersible closed-circuit video cameras to examine reactor components and welds. PNNL has conducted a parametric study that examines the important variables that affect the effectiveness of a remote visual test. Tested variables include lighting techniques, camera resolution, camera movement, and magnification. PNNL has also conducted a laboratory test using a commercial visual testing camera system to experimentally determine the ability of the camera system to detect cracks of various widths under ideal conditions.

Corrosion Behavior of Si3N4 Ceramics under High-Temperature and High-Pressure Water Condition

2007 ◽  
Vol 26-28 ◽  
pp. 259-262 ◽  
Author(s):  
Weon Ju Kim ◽  
Seok Min Kang ◽  
Ji Yeon Park
Keyword(s):  
High Pressure ◽  
High Temperature ◽  
Grain Boundary ◽  
Corrosion Behavior ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Boundary Phase ◽  
Pressure Water ◽  
Si3n4 Ceramics

Silicon nitride (Si3N4) ceramics have been considered for various components of nuclear power plants such as mechanical seal of reactor coolant pump (RCP), guide roller for control rod drive mechanism (CRDM), and seal support, etc. Corrosion behavior of Si3N4 ceramics in high-temperature and high-pressure water must be elucidated before they can be considered for components of nuclear power plants. In this study, the corrosion behaviors of Si3N4 ceramics at hydrothermal condition (300°C, 9.0 MPa) were investigated in pure water. The grain-boundary phase was preferentially corroded and the corrosion reaction was controlled by the diffusion of the reactive species and/or products through the corroded layer. Results of this study imply that the variation of sintering aids and/or the control (e.g., crystallization) of the grain-boundary phase are necessary to increase the corrosion resistance of Si3N4 ceramics in high-temperature water.

Experience in Dismantling and Packaging of Pressure Vessel and Core Internals

2010 ◽  
Author(s):  
Peter Pillokat ◽  
Jan Hendrik Bruhn
Keyword(s):  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Hydraulic Press ◽  
Cylindrical Part ◽  
Water Jet ◽  
Pressurized Water ◽  
Vacuum Suction ◽  
Water Reactor ◽  
Water Jet Cutting

Nuclear Company AREVA is proud to look back on versatile experience in successfully dismantling nuclear components. After performing several minor dismantling projects and studies for nuclear power plants, AREVA completed the order for dismantling of all remaining Reactor Pressure Vessel internals at German Boiling Water Reactor Wuergassen NPP in October ’08. During the onsite activities about 121 tons of steel were successfully cut and packed under water into 200l- drums, as the dismantling was performed partly in situ and partly in an underwater working tank. AREVA deployed a variety of different cutting techniques such as band sawing, milling, nibbling, compass sawing and water jet cutting throughout this project. After successfully finishing this task, AREVA dismantled the cylindrical part of the Wuergassen Pressure Vessel. During this project approximately 320 tons of steel were cut and packaged for final disposal, as dismantling was mainly performed by on air use of water jet cutting with vacuum suction of abrasive and kerfs material. The main clue during this assignment was the logistic challenge to handle and convey cut pieces from the pressure vessel to the packing area. For this, an elevator was installed to transport cut segments into the turbine hall, where a special housing was built for final storage conditioning. At the beginning of 2007, another complex dismantling project of great importance was acquired by AREVA. The contract included dismantling and conditioning for final storage of the complete RPV Internals of the German Pressurized Water Reactor Stade NPP. Very similar cutting techniques turned out to be the proper policy to cope this task. On-site activities took place in up to 5 separate working areas including areas for post segmentation and packaging to perform optimized parallel activities. All together about 85 tons of Core Internals were successfully dismantled at Stade NPP until September ’09. To accomplish the best possible on-site performance and to achieve a minimization of the applied collective dose rates, each on-site activity was previously planned in detail and personnel exercised each task at original size mock ups under most realistic onsite conditions. Planning was especially focused on an optimized size minimization and packaging concept to reduce the number of filled waste packages. The segmentation of components strictly followed a sophisticated cutting and packaging concept developed under consideration of possible cutting techniques, the resulting geometry and logistical conditions. Therefore, segments were post processed by hydraulic press and band saw in order to minimize their volume, where applicable.

The Failure Analysis and Processing of Digital Reactor Protection System

2013 ◽  
Author(s):  
Jun Zhao ◽  
Xing Zhou ◽  
Jin Hu ◽  
Yanling Yu
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Phase 1 ◽  
Protection System ◽  
Pressurized Water Reactor ◽  
Communication Failure ◽  
Pressurized Water ◽  
Power Failure ◽  
Self Test ◽  
Reactor Protection System

The Qinshan Nuclear Power Plant phase 1 unit (QNPP-1) has a power rating of 320 MWe generated by a pressurized water reactor that was designed and constructed by China National Nuclear Corporation (CNNC). The TELEPERM XS I&C system (TXS) is to be implemented to transform analog reactor protection system (RPS) in QNPP-1. The paper mainly describes the function, structure and characteristic of RPS in QNPP-1. It focuses on the outstanding features of digital I&C, such as strong online self-test capability, the degradation of the voting logic processing, interface improvements and CPU security. There are some typical failures during the operation of reactor protection system in QNPP-1. The way to analyze and process the failures is different from analog I&C. The paper summarizes typical failures of the digital RPS in the following types: CPU failure, communication failure, power failure, Input and output (IO) failure. It discusses the cause, risk and mainly processing points of typical failure, especially CPU and communication failures of the digital RPS. It is helpful for the maintenance of the system. The paper covers measures to improve the reliability of related components which has been put forward effective in Digital reactor protection system in QNPP-1. It will be valuable in nuclear community to improve the reliability of important components of nuclear power plants.

The method for calculating the probability of brittle destruction of NPP equipment in different operating modes with postulated defectiveness

2021 ◽  
Vol 14 (1) ◽  
pp. 34-39
Author(s):  
D. A. Kuzmin ◽  
A. Yu. Kuz’michevskiy
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Power Plants ◽  
Reactor Pressure Vessel ◽  
Practical Significance ◽  
Reactor Unit ◽  
Operating Modes ◽  
Brittle Destruction ◽  
Reactor Pressure

The destruction of equipment metal by a brittle fracture mechanism is a probabilistic event at nuclear power plants (NPP). The calculation for resistance to brittle destruction is performed for NPP equipment exposed to neutron irradiation; for example, for a reactor plant such as a water-water energetic reactor (WWER), this is a reactor pressure vessel. The destruction of the reactor pressure vessel leads to a beyond design-basis accident, therefore, the determination of the probability of brittle destruction is an important task. The research method is probabilistic analysis of brittle destruction, which takes into account statistical data on residual defectiveness of equipment, experimental results of equipment fracture toughness and load for the main operating modes of NPP equipment. Residual defectiveness (a set of remaining defects in the equipment material that were not detected by non-destructive testing methods after manufacturing (operation), control and repair of the detected defects) is the most important characteristic of the equipment material that affects its strength and service life. A missed defect of a considerable size admitted into operation can reduce the bearing capacity and reduce the time of safe operation from the nominal design value down to zero; therefore, any forecast of the structure reliability without taking into account residual defectiveness will be incorrect. The application of the developed method is demonstrated on the example of an NPP reactor pressure vessel with a WWER-1000 reactor unit when using the maximum allowable operating loads, in the absence of load dispersion in different operating modes, and taking into account the actual values of the distributions of fracture toughness and residual defectiveness. The practical significance of the developed method lies in the possibility of obtaining values of the actual probability of destruction of NPP equipment in order to determine the reliability of equipment operation, as well as possible reliability margins for their subsequent optimization.

The New Generation of High-Pressure Heaters for Steam Turbine Units at Nuclear Power Plants

2021 ◽  
Vol 68 (4) ◽  
pp. 285-294
Author(s):  
V. M. Zorin ◽  
A. S. Shamarokov ◽  
S. B. Pustovalov
Keyword(s):  
High Pressure ◽  
Steam Turbine ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
New Generation

Development of Long Rotating Blade Applied in the New Generation Nuclear Steam Turbine

2012 ◽  
Author(s):  
Kai Cheng ◽  
Zeying Peng ◽  
Gongyi Wang ◽  
Xiaoming Wu ◽  
Deqi Yu
Keyword(s):  
Steam Turbine ◽  
Nuclear Power ◽  
Power Plants ◽  
Structure Design ◽  
Good Reliability ◽  
Pressurized Water ◽  
Height Range ◽  
Water Reactor ◽  
Rotating Blade ◽  
Last Stage

In order to meet the high economic requirement of the 3rd generation Pressurized Water Reactor (PWR) or Boiling Water Reactor (BWR) applied in currently developing nuclear power plants, a series of half-speed extra-long last stage rotating blades with 26 ∼ 30 m2 nominal exhaust annular area is proposed, which covers a blade-height range from 1600 mm to 1900 mm. It is well known that developing an extra long blade is a tough job involving some special coordinated sub-process. This paper is dedicated to describe the progress of creating a long rotating blade for a large scaled steam turbine involved in the 3rd generation nuclear power project. At first the strategy of how to determine the appropriate height for the last-stage-rotating-blade for the steam turbine is provided. Then the quasi-3D flow field quick design method for the last three stages in LP casing is discussed as well as the airfoil optimization method. Furthermore a sophisticated blade structure design and analyzing system for a long blade is introduced to obtain the detail dimension of the blade focusing on the good reliability during the service period. Thus, except for CAD and experiment process, the whole pre-design phase of the extra-long turbine blade is presented which is regarded as an assurance of the operation efficiency and reliability.

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