scholarly journals Evaluation of Ultrasonic Time-of-Flight Diffraction Data for Selected Control Rod Drive Nozzles from Davis Besse Nuclear Power Plant

2011 ◽  
Author(s):  
Stephen E. Cumblidge ◽  
Anthony D. Cinson ◽  
Michael T. Anderson
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Diffraction Data ◽  
Nuclear Power ◽  
Time Of Flight ◽  
Control Rod ◽  
Time Of Flight Diffraction ◽  
Ultrasonic Time

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.

Analysis of Reasons for Non-Conformance of the Movable Latch Lock Plunger Releasing Current of Control Rod Drive Mechanism and Improvement Measures

2017 ◽  
Author(s):  
Yinhui Lan ◽  
Cuizhu He ◽  
Yuangang Duan ◽  
Feihua Liu
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Performance Test ◽  
Cold Test ◽  
Test Results ◽  
Suction Force ◽  
Control Rod ◽  
Drive Mechanism ◽  
Hot Test

As one of the most important equipment for reactivity control, Control Rod Drive Mechanism (CRDM), which is widely used in pressurized water reactor (PWR) nuclear power plant, has a series of important security functions. As an important component of the claw part in CRDM, the nonmagnetic shim materials have big influence on the movable latch lock plunger releasing current of CRDM. When cutting off the coil power, the nonmagnetic shim materials can block the magnetic circuit between the pole and latch lock plunger effectively and reduce the remanence suction force between magnetic pole and latch lock plunger, which can promote a quick latch lock plunger’s action and finish the step-jump and rod-release. In this paper, we introduce the background of non-conformance of the movable latch lock plunger releasing current of CRDM of PWR plant simply, and then we analyze the reasons of the non-conformance in detail, including a comprehensive analysis of various factors and a series of retest conclusions. Through specific analysis, the important role of nonmagnetic shims in CRDM and its big influence on movable latch lock plunger releasing current are proved. Based on research and test results, we show our optimized measures on the movable latch lock plunger releasing current in detail, from the perspective of technical specification for raw materials and improved processing technology during production. At last, one latch unit which occurred the movable latch lock plunger releasing current non-conformance and experienced a 1.7 million steps performance test is used in the following performance verification test. Being installed in this latch mechanism, our nonmagnetic shims finished an integral series of cold test, hot test and cold test after hot test. All of the test results meet our design requirements and all of the releasing current under cold condition is better than the results of 1.7 million steps performance test. Especially, the independently developed nonmagnetic shims improved the movable latch lock plunger releasing current significantly. In addition, other current results of hot test are equal to that of 1.7 million steps performance test. In conclusion, these optimized measures may not only provide data for solving the problem completely, but also provide reference for the manufacture of nonmagnetic shim materials in the future nuclear power plant.

scholarly journals R&D on a Nonlinear Dynamics Analysis Code for the Drop Time of the Control Rod

2017 ◽  
Vol 2017 ◽  
pp. 1-6
Author(s):  
Daogang Lu ◽  
Yuanpeng Wang ◽  
Qingyu Xie ◽  
Huimin Zhang ◽  
Muhammed Ali
Keyword(s):  
Nonlinear Dynamics ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Response Analysis ◽  
Safe Operation ◽  
Control Rod ◽  
Long Time ◽  
Drop Time ◽  
Dynamics Response

Whether the control rod can drop down in time is one of the important guarantees for the safe operation of the nuclear power plant. The drop-down process of the control rod is very complicated. For a long time, the researchers have done a lot of work on that, but it is hard to consider all the nonlinear factors. This paper considers the main factors together. Based on the theoretical analysis, we developed the nonlinear dynamics response analysis software for the nuclear power plant, which can be used to calculate the rod’s drop-down time. Compared with the results of the experiments, the software we developed proves to be applicable and reliable.

scholarly journals Control Rod Modeling and Worth Calculation for a Typical 1100 MWe Nuclear Power Plant Using WIMS/D4 and CITATION

2022 ◽  
Vol 2022 ◽  
pp. 1-13
Author(s):  
Izza Shahid ◽  
Nadeem Shaukat ◽  
Amjad Ali ◽  
Meer Bacha ◽  
Ammar Ahmad ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Power Distribution ◽  
Nuclear Power ◽  
Operating Conditions ◽  
Pressurized Water ◽  
Control Rod ◽  
Operational Conditions ◽  
The Core ◽  
Safety Parameters

A typical 1100 MWe pressurized water reactor (PWR) is a second unit installed at the coastal site of Pakistan. In this paper, verification analysis of reactivity control worth by means of rod cluster control assemblies (RCCAs) for startup and operational conditions of this typical nuclear power plant (CNPP) has been performed. Neutronics analysis of fresh core is carried out at beginning of life (BOL) to determine the effect of grey and black control rod clusters on the core reactivity for startup and operating conditions. The combination of WIMS/D4 and CITATION computer codes equipped with JENDL-3.3 data library is used for the first time for core physics calculations of neutronic safety parameters. The differential and integral worth of control banks is derived from the computed results. The effect of control bank clusters on core radial power distribution is studied precisely. Radial power distribution in the core is evaluated for numerous configurations of control banks fully inserted and withdrawn. The accuracy of computed results is validated against the reference values of Nuclear Design Report (NDR) of 1100 MWe typical CNPP. It has been observed that WIMS-D4/CITATION shows its capability to effectively calculate the reactor physics parameters.

scholarly journals Determination of the boron concentration of OPR 1000 reactor critical state using the Cosi OPR1000 system

2016 ◽  
Vol 19 (4) ◽  
pp. 241-248
Author(s):  
Son An Nguyen ◽  
Nguyen Trung Tran ◽  
Tuan Quoc Tran ◽  
Cuong Quang Ly ◽  
Lan Thi Ha Le ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Critical State ◽  
Nuclear Power ◽  
Boron Concentration ◽  
Nuclear Fission ◽  
Bisection Method ◽  
Control Rod ◽  
Operating Data

In the operation of a nuclear power plant (NPP), to adjust the capacity of NPP is necessary. When the NPP capacity is changed the nuclear fission is also changed. The methods used in changing the capacity of NPP include: changing the boron concentration, changing the position of the control rod groups, and changing the boron concentrations and the position of the control rod groups together. This report presents some results of the research, measurement boron concentrations when nuclear power plans OPR1000 critically state in the cases of ARO, ARI SB, ARI R1, R5 = 191 cm on the basis of the bisection method in the boron concentrations adjustment. The experiment is performed on core the simulator for OPR 1000 nuclear power plant. The results in the 4 cases were similar with NPP operating data using OPR1000 reactor.

Advanced Pipe Replacement Procedure for a Defective CRDM Housing Nozzle Enables Continued Normal Operation of a Nuclear Power Plant

2006 ◽  
Author(s):  
Geoff Gilmore ◽  
Andrew Becker
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Training Programs ◽  
Repair Process ◽  
Normal Operation ◽  
Repair System ◽  
Control Rod ◽  
Replacement Procedure ◽  
Pipe Replacement

During the 2003 outage at the Ringhals Nuclear Plant in Sweden, a leak was found in the vicinity of a Control Rod Drive Mechanism (CRDM) housing nozzle at Unit 1. Based on the ALARA principle for radioactive contamination, a unique repair process was developed. The repair system includes utilization of custom, remotely controlled GTAW-robots, a CNC cutting and finishing machine, snake-arm robots and NDE equipment. The success of the repair solution was based on performing the machining and welding operations from the inside of the SCRAM pipe through the CRDM housing since accessibility from the outside was extremely limited. Before the actual pipe replacement procedure was performed, comprehensive training programs were conducted. Training was followed by certification of equipment, staff and procedures during qualification tests in a full scale mock-up of the housing nozzle. Due to the ingenuity of the overall repair solution and training programs, the actual pipe replacement procedure was completed in less than half the anticipated time. As a result of the successful pipe replacement, the nuclear power plant was returned to normal operation.

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