Study on Change of Chromium Plating Thickness of Threads of Control Rod Drive Mechanism of Nuclear Power

2019 ◽  
Vol 956 ◽  
pp. 125-134
Author(s):  
Fang Wang ◽  
Song Xue ◽  
En Jiang
Keyword(s):  
Current Density ◽  
Nuclear Power ◽  
Power Plants ◽  
High Current Density ◽  
Root Diameter ◽  
Nuclear Radiation ◽  
Thickness Change ◽  
Control Rod ◽  
Drive Mechanism ◽  
Chromium Plating

Control rod drive mechanism is one of the key main components in nuclear power plants and serves in harsh environments such as high temperature, high pressure and nuclear radiation. In order to ensure the service life and to prevent biting when assembling, some of the threads of control rod drive mechanism need to be chromium plated. In view of the high demands of the same chromium plating thickness on all surfaces of the threads of control rod drive mechanism of nuclear power and non-uniformity in chromium plating thickness of threads due to poor throwing power of chromium plating solution, five representative kinds of threads of control rod drive mechanism were selected and the plating thickness change of the root diameter, pitch diameter and crest diameter of threads was studied in this paper by means of depositing different thickness chromium coating on the surface of threads. The experimental results show that thicker coating is deposited on the crest of thread because of high current density and thinner coating is deposited on the root of thread because of low current density, which can provide reference for specification of chromium plating thickness of thread products of control rod drive mechanism.

Dynamic Response Analysis of Control Rod Drive Mechanism Under the Stepping Impact Force for Nuclear Power Plants

2012 ◽  
Author(s):  
Xiaoyao Shen ◽  
Yongcheng Xie
Keyword(s):  
Dynamic Response ◽  
Nuclear Power ◽  
Power Plants ◽  
Structural Integrity ◽  
Impact Force ◽  
Response Analysis ◽  
Control Rod ◽  
Drive Mechanism ◽  
Finite Element Software ◽  
The Impact

The control rod drive mechanism (CRDM) is an important safety-related component in the nuclear power plant (NPP). When CRDM steps upward or downward, the pressure-containing housing of CRDM is shocked axially by an impact force from the engagement of the magnetic pole and the armature. To ensure the structural integrity of the primary coolant loop and the functionality of CRDM, dynamic response of CRDM under the impact force should be studied. In this manuscript, the commercial finite element software ANSYS is chosen to analyze the nonlinear impact problem. A nonlinear model is setup in ANSYS, including main CRDM parts such as the control rod, poles and armatures, as well as nonlinear gaps. The transient analysis method is adopted to calculate CRDM dynamic response when it steps upward. The impact loads and displacements at typical CRDM locations are successfully obtained, which are essential for design and stress analysis of CRDM.

The Effect of Elevated Temperatures and Nuclear Radiation on the Properties of Biological Shielding Concrete

2016 ◽  
Vol 677 ◽  
pp. 8-16 ◽  
Author(s):  
Jaroslava Koťátková ◽  
Jan Zatloukal ◽  
Pavel Reiterman ◽  
Jan Patera ◽  
Zbyněk Hlaváč ◽  
...  
Keyword(s):  
Gamma Radiation ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Elevated Temperatures ◽  
Nuclear Radiation ◽  
Containment Vessel ◽  
Negative Impacts

The paper reviews the so far known information about the properties of biological shielding concrete used in the containment vessel of nuclear power plants (NPP) and its behaviour when exposed to radiation. The damage of concrete caused by neutron and gamma radiation as well as by the accompanying generation of heat is described. However, there is not enough data for the proper evaluation of the negative impacts and further research is needed.

FIELD TEST OF WIRELESS SENSOR NETWORK IN THE NUCLEAR ENVIRONMENT

2014 ◽  
Vol 3 (01) ◽  
pp. 47-51 ◽  
Author(s):  
L. Li ◽  
Q. Wang ◽  
A. Bari ◽  
C. Deng ◽  
D. Chen ◽  
...  
Keyword(s):  
Electromagnetic Interference ◽  
Nuclear Power ◽  
Power Plants ◽  
Low Cost ◽  
Signal Propagation ◽  
Nuclear Radiation ◽  
Wireless Sensor ◽  
Wireless Signal ◽  
Nuclear Environment ◽  
Interference Tests

Wireless sensor networks (WSNs) are appealing options for the health monitoring of nuclear power plants due to their low cost and flexibility. Before they can be used in highly regulated nuclear environments, their reliability in the nuclear environment and compatibility with existing devices have to be assessed. In situ electromagnetic interference tests, wireless signal propagation tests, and nuclear radiation hardness tests conducted on candidate WSN systems at AECL Chalk River Labs are presented. The results are favourable to WSN in nuclear applications.

scholarly journals Review on the Change Law of the Properties and Performance of Polymer-Matrix Nuclear Radiation Shielding Materials Under the Coupling of Nuclear Radiation and Thermal Effects

2021 ◽  
Vol 9 ◽  
Author(s):  
Guang Hu ◽  
Weiqiang Sun ◽  
Yihong Yan ◽  
Rongjun Wu ◽  
Hu Xu
Keyword(s):  
Mechanical Properties ◽  
Polymer Matrix ◽  
Thermal Effects ◽  
Nuclear Power ◽  
Power Plants ◽  
Radiation Shielding ◽  
Nuclear Radiation ◽  
And Performance ◽  
Shielding Material

The polymer-matrix nuclear radiation shielding material is an important component of nuclear power plants. However, its mechanical properties and shielding performance gradually deteriorate due to the long-term synergy of nuclear radiation and thermal effects, which brings hidden dangers to the safe operation of the device. Based on this problem, this article makes a comprehensive review. First, the degradation of mechanical properties and shielding performance of polymer-matrix nuclear radiation materials in service is briefly described. Then, the research methods adopted by scholars to study the change law of properties and performance are introduced, and the main existing difficulties encountered by the study are summarized. Finally, the physical mechanism of the change of material properties is explained in detail, and a reference approach to solving the problem is proposed.

Validation of Analysis Models on Relocation Behavior of Molten Core Materials in Sodium-Cooled Fast Reactors Based on the Melt Discharge Experiment

2020 ◽  
Author(s):  
Kai Igarashi ◽  
Ryoji Onuki ◽  
Takaaki Sakai ◽  
Shinya Kato ◽  
Ken-ichi Matsuba ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Large Scale ◽  
Heat Removal ◽  
Early Discharge ◽  
Core Material ◽  
Bottom Plate ◽  
Guide Tube ◽  
Control Rod ◽  
Analysis Models

Abstract In order to improve the safety of nuclear power plants, it is necessary to make sure measures against their severe accidents. Especially, in the case of a sodium-cooled fast reactor (SFR), there is a possibility of significant energy release due to formation of a large-scale molten fuel pool accompanied by re-criticality in the event of a core disruptive accident (CDA). It is important to ensure in-vessel retention that keeps and confines damaged core material in the reactor vessel even if the CDA occurs. CDA scenario initiated by Unprotected Loss Of Flow (ULOF), which is a typical cause of core damage, is generally categorized into four phases according to the progression of core-disruptive status, which are the initiating, early-discharge, material-relocation and heat-removal phases for the latest design in Japan. During the material-relocation phase, the molten core material flows down mainly through the control rod guide tube and is discharged into the inlet coolant plenum below the bottom of the core. The discharged molten core material collides with the bottom plate of the inlet plenum. Clarification of the accumulation behavior of molten core material with such a collision on the bottom plate is important to reduce uncertainties in the safety assessment of CDA. In present study, in order to make clear behavior of core melt materials during the CDAs of SFRs, analysis was conducted using the SIMMER-III code for a melt discharge simulation experiment [1] in which low-melting-point alloy was discharged into a shallow water pool. This report shows the validation results for the melt behavior by comparing with the experimental data.

A New Safety Analysis Method of Control Rod Ejection Accident in PWR NPP Based on the Failure of Causal Relationship

2018 ◽  
Author(s):  
Shiyu Yan ◽  
Hua Liu ◽  
Zhaohui Liu ◽  
Xiaohua Yang ◽  
Meng Li ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Safety Analysis ◽  
Pressurized Water ◽  
Control Rod ◽  
Drive Mechanism ◽  
Cause And Effect ◽  
Simulation Calculation ◽  
Causal Logic

In view of control rod ejection accident of the traditional pressurized water reactor, the safety thought of the design phase is to validate reliability and availability of DCS I&C in the severe accidents. Now the most important and effective means is simulation calculation and analysis. It is applied for the imaginary accident of the nuclear power plant by using computer software. The new safety analysis steps based on the analysis of cause-and-effect logic failure: firstly, the composition and working principle of control rod drive mechanism is analyzed; secondly, a list of factors-the dynamics and structure, environmental reasons, the function of the control rod drive mechanism and status analysis-are all taken into account, the initial cause of failure modes with causal logic analysis is carried out; thirdly, based on cause-and-effect logic failure, the prevention and improvement measures of accidents, the new criterion of design are put forward. The advantages of cause-and-effect logic failure safety analysis: 1.be based on causal logic. 2. the system aspects is added compared with the past method that is only based on simulation calculation and analysis of the hypothetical accident, the accident the transient process of the key security parameters as the acceptance criteria. 3. The verification and audit of the lack of safety design criteria, completeness of design content, sufficiency problem are performed before the simulated calculation and analysis. 4. The coverage of safety analysis is expanded. Some good advices are provided for the design, operation and maintenance of nuclear power plant.

scholarly journals Laser Ultrasonic System for Surface Crack Visualization in Dissimilar Welds of Control Rod Drive Mechanism Assembly of Nuclear Power Plant

2014 ◽  
Vol 2014 ◽  
pp. 1-10 ◽  
Author(s):  
Yun-Shil Choi ◽  
Hyomi Jeong ◽  
Jung-Ryul Lee
Keyword(s):  
Wave Propagation ◽  
Nuclear Power ◽  
Ultrasonic Waves ◽  
Full Scale ◽  
Head Part ◽  
Visualization System ◽  
Control Rod ◽  
Drive Mechanism ◽  
Dissimilar Weld ◽  
Ultrasonic Propagation

In this paper, we propose a J-groove dissimilar weld crack visualization system based on ultrasonic propagation imaging (UPI) technology. A full-scale control rod drive mechanism (CRDM) assembly specimen was fabricated to verify the proposed system. An ultrasonic sensor was contacted at one point of the inner surface of the reactor vessel head part of the CRDM assembly. Q-switched laser beams were scanned to generate ultrasonic waves around the weld bead. The localization and sizing of the crack were possible by ultrasonic wave propagation imaging. Furthermore, ultrasonic spectral imaging unveiled frequency components of damage-induced waves, while wavelet-transformed ultrasonic propagation imaging enhanced damage visibility by generating a wave propagation video focused on the frequency component of the damage-induced waves. Dual-directional anomalous wave propagation imaging with adjacent wave subtraction was also developed to enhance the crack visibility regardless of crack orientation and wave propagation direction. In conclusion, the full-scale specimen test demonstrated that the multiple damage visualization tools are very effective in the visualization of J-groove dissimilar weld cracks.

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