A novel method for diagnosis and de-noising of control rod drive mechanism within floating nuclear reactor

The control rod drive mechanism (CRDM) is installed on the CRDM socket in reactor pressure vessel (RPV). Directed by Rod Control and Rod Position Indicating System (RGL), CRDM can impel the control rods move up and down in the nuclear reactor core, which implements the functions of reactor start-up, power regulation, power maintaining, normal reactor shutdown and abnormal (accident) shutdown. CRDM was developed by China Nuclear Power Research Institute (CNPRI). Several design improvements were conducted to solve the problems appeared in the operation of nuclear power station. Test bench was also set up and cold tests were carried out to investigate the characteristics of CRDM. The cold tests included lifting experiment, inserting experiment, rod drop experiment. And studies were carried out to analyze the signals of lifting coil, moving coil, stationary coil and the vibration signals. The test results show that the design of CRDM is reasonable and the operation is reliable.

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Crack Growth Safety Assessment of Overlay Weld on the Ω Seal for Control Rod Drive Mechanism

Volume 4: Nuclear Safety, Security, Non-Proliferation and Cyber Security; Risk Management ◽

10.1115/icone25-67615 ◽

2017 ◽

Author(s):

Jiacheng Luo ◽

Yong Zhang ◽

Pengzhou Li ◽

Juan Luo ◽

Lei Sun

Keyword(s):

Stress Corrosion ◽

Crack Growth ◽

Safety Assessment ◽

Nuclear Reactor ◽

Structural Integrity ◽

Safety Evaluation ◽

Structure Design ◽

Control Rod ◽

Drive Mechanism ◽

Weld Structure

As a key device, the Control Rod Drive Mechanism (CRDM) is of great important for the safety operation of the nuclear reactor, where the pressure shell of CRDM is exposed to the primary loop pressure, and is installed on the head of the reactor pressure vessel by the Ω seal. During the operation cycle of the reactor, cracks may be induced on the Ω seal weld by primary water stress corrosion cracking which is the major factor leading to leakage of coolant and impacting the safety of the reactor. Therefore, the Ω seal is required to be repaired to maintain structural integrity and safety by using the evaluated overlay weld technology. Based on fracture mechanics, this paper investigates the crack growth safety assessment of the overlay weld structure on the Ω seal of control rod drive mechanism, induced by the operation fatigue and stress corrosion, and the evaluation is completed by the standard code. The calculation method and assessment results can be a reference for the overlay weld structure design and safety evaluation of the Ω seal in CRDM.

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Analysis and Optimization Research on Latch Life of Control Rod Drive Mechanism Based on Approximate Model

Nuclear Engineering and Technology ◽

10.1016/j.net.2021.06.012 ◽

2021 ◽

Author(s):

Sitong Ling ◽

Wenqiang Li ◽

Tianda Yu ◽

Qiang Deng ◽

Guozhong Fu

Keyword(s):

Approximate Model ◽

Control Rod ◽

Drive Mechanism

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Development of a High Temperature and Radiation Resistant Linear Motion Servo Actuator System for Nuclear Reactor Control Rod Application

IRE Transactions on Nuclear Science ◽

10.1109/tns2.1962.4315909 ◽

1962 ◽

Vol 9 (1) ◽

pp. 193-203

Author(s):

J. S. Sulich ◽

M. H. Cardon

Keyword(s):

High Temperature ◽

Nuclear Reactor ◽

Linear Motion ◽

Reactor Control ◽

Control Rod ◽

Servo Actuator ◽

Radiation Resistant ◽

Actuator System ◽

Nuclear Reactor Control

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Modelling of primary water stress corrosion cracking (PWSCC) at control rod drive mechanism (CRDM) nozzles of pressurized water reactors (PWR)

Environment-Induced Cracking of Materials ◽

10.1016/b978-008044635-6.50053-4 ◽

2008 ◽

pp. 143-151

Author(s):

O.F. Aly ◽

A.H.P. Andrade ◽

M. Mattar Neto ◽

M. Szajnbok ◽

H.J. Toth

Keyword(s):

Water Stress ◽

Stress Corrosion ◽

Stress Corrosion Cracking ◽

Corrosion Cracking ◽

Pressurized Water Reactors ◽

Pressurized Water ◽

Control Rod ◽

Drive Mechanism ◽

Primary Water ◽

Water Reactors

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Considerations of Material Selection for Control Rod Drive Mechanism of Reaktor Daya Eksperimental

Journal of Physics Conference Series ◽

10.1088/1742-6596/1198/3/032010 ◽

2019 ◽

Vol 1198 (3) ◽

pp. 032010

Author(s):

Sri Hastuty ◽

Petrus Zacharias ◽

M Awwaluddin ◽

Krismawan ◽

Putut Hery Setiawan ◽

...

Keyword(s):

Material Selection ◽

Control Rod ◽

Drive Mechanism ◽

Selection For

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Development of 3-D Simulation Analysis System for PWR Control Rod Drive Mechanism

Volume 1: Plant Operations, Maintenance and Life Cycle; Component Reliability and Materials Issues; Codes, Standards, Licensing and Regulatory Issues; Fuel Cycle and High Level Waste Management ◽

10.1115/icone14-89382 ◽

2006 ◽

Author(s):

Akio Tanaka ◽

Kensuke Futahashi ◽

Kiyoshi Takanabe ◽

Chikara Kurimura ◽

Jungo Kato ◽

...

Keyword(s):

Simulation Analysis ◽

Functional Evaluation ◽

Control Rod ◽

Drive Mechanism ◽

Virtual Analysis ◽

Analysis System ◽

D Model

3-D virtual analysis system to analyze the motion of Control Rod Drive Mechanism (CRDM) was developed. The analysis system consists of 3-D model established per the actual dimensions and interfaces of CRDM parts and a routine to calculate forces acting on the mechanism, and was verified by mock-up test using equipment same to actual product. The analysis system is useful for functional evaluation in maintenance or to factor out root causes in case of malfunction of CRDM.

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Effects of Weld Geometry on Residual Stress and Crack Driving Force for Centerhole Control Rod Drive Mechanism Nozzles: Part I — Weld Residual Stress

Volume 6: Materials and Fabrication ◽

10.1115/pvp2005-71077 ◽

2005 ◽

Cited By ~ 1

Author(s):

Wentao Cheng ◽

David L. Rudland ◽

Gery Wilkowski ◽

Wallace Norris

Keyword(s):

Residual Stress ◽

Residual Stresses ◽

Driving Force ◽

Nuclear Regulatory Commission ◽

Crack Driving Force ◽

Control Rod ◽

Drive Mechanism ◽

Weld Geometry ◽

Weld Residual Stress ◽

Regulatory Commission

The U.S. Nuclear Regulatory Commission (NRC) has undertaken a program to assess the integrity of control rod drive mechanism (CRDM) nozzles in existing plants that are not immediately replacing their RPV heads. This two-part paper summarizes some of the efforts undertaken on the behalf of the U.S.NRC for the development of detailed residual stress and circumferential crack-driving force solutions to be used in probabilistic determinations of the time from detectable leakage to failure. In this first paper, the finite element (FE) simulations were conducted to investigate the effects of weld geometry on the residual stresses in the J-weld for a centerhole CRDM nozzle. The variables of weld geometry included three weld heights (weld sizes) and three groove angles for each weld height while keeping the same weld size. The analysis results indicate that the overall weld residual stress decreases as the groove angle increases and higher residual stress magnitude is associated with certain weld height. The results also reveal that the axial residual stresses in the Alloy 600 tube are very sensitive to the weld height, and that the tube hoop stresses above the J-weld root increase with the increasing weld height.

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