scholarly journals Investigation Of Rod Control System Reliability Of Pwr Reactors

KnE Energy ◽  
2016 ◽  
Vol 1 (1) ◽  
Author(s):  
Syaiful Bakhri
Keyword(s):  
Control System ◽  
System Reliability ◽  
Nuclear Power ◽  
Power Plants ◽  
Thermal Power ◽  
Fault Tree ◽  
Reactor Core ◽  
Pressurized Water ◽  
Control Rod ◽  
Control Rods

<p class="NoSpacing1"><span lang="IN">The Rod Control System is </span>employed<span lang="IN"> to adjust the position of the control rods in the reactor core </span>which corresponds with <span lang="IN">the thermal power generated in the core </span>as well as <span lang="IN">the electric power generated in the turbine. In a Pressurized Water Reactor (PWR) type nuclear power plants, the control-rod drive </span>employs <span lang="IN">magnetic stepping-type mechanism. This </span>type of <span lang="IN">mechanism consists of a pair of circular coils and latch-style jack with the armature. When the </span>electric <span lang="IN">current </span>is <span lang="IN">supplied to the coils sequentially, the control-rods</span>, which <span lang="IN">are held on the drive shaft</span>, can be driven<span lang="IN"> up</span>ward<span lang="IN"> or down</span>ward<span lang="IN"> in increments. </span>This <span lang="IN">sequential current </span>c<span lang="IN">ontrol</span> drive<span lang="IN"> system is called the Control-Rod Drive Mechanism Control System (CRDMCS) or </span>known also as <span lang="IN">the Rod Control System (RCS). The p</span>urpose of this paper is to investigate the RCS reliability <span lang="IN">of APWR </span>using <span lang="IN">the Fault Tree Analysis (FTA)</span> method<span lang="IN"> since </span>the analysis of reliability which considers<span lang="IN"> the FTA</span> for common CRDM <span lang="IN">can </span>not <span lang="IN">be found</span> in <span lang="IN">any </span>public references. <span lang="IN">The FTA method is used to model the system reliability by developing the fault tree diagram of the system. </span>The<span lang="IN"> results show that the failure of the system is very dependent on the failure of most of the individual systems. However, the failure of the system does not affect the safety of the reactor, since the reactor trips immediately if the system fails. The evaluation results also indicate that the Distribution Panel is the most critical component in the system.</span></p>

Power Maximization of Fully Passive Lead-Bismuth Eutectic Cooled Small Modular Reactor

2014 ◽  
Author(s):  
Jaehyun Cho ◽  
Yong-Hoon Shin ◽  
Il Soon Hwang
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Natural Circulation ◽  
Reactor Core ◽  
Design Stage ◽  
Pressurized Water ◽  
Small Modular Reactor ◽  
Power Maximization ◽  
Lead Bismuth Eutectic ◽  
Water Reactors

Although the current Pressurized Water Reactors (PWRs) have significantly contributed to the global energy supply, PWRs have not been considered as a trustworthy energy solution owing to its several problems; spent nuclear fuels (SNFs), nuclear safety, and nuclear economy. In order to overcome these problems, lead-bismuth eutectic (LBE) fully passive cooling Small Modular Reactor (SMR) system is suggested. It is possible to not only provide the solution of the problem of SNFs through the transmutation feature of LBE coolant, but also increase the safety and economy through the concepts of the natural circulation cooling SMRs. It is necessary to maximize the advantages (safety and economy) of this type of Nuclear Power Plants for several applications in future. Accordingly, objective of the study is to maximize the reactor core power while the limitations of shipping size, materials endurance, long-burning criticality as well as safety under Beyond Design Basis Events must be satisfied. Design limitations of natural circulating LBE-cooling SMRs are researched and power maximization method is developed based on obtained design limitations. It is expected that the results are contributed to reactor design stage with providing several insights to designers as well as the methods for design optimization of other type of SMRs.

Coordinated Control of a Small Pressurized Water Reactor

2018 ◽  
Author(s):  
Peiwei Sun ◽  
Chong Wang
Keyword(s):  
Control System ◽  
Reactor Core ◽  
Reactor Power ◽  
Pressurized Water Reactors ◽  
Primary System ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Control Rod ◽  
Average Temperature ◽  
Water Reactors

Small Pressurized Water Reactors (SPWR) are different from those of the commercial large Pressurized Water Reactors (PWRs). There are no hot legs and cold legs between the reactor core and the steam generators like in the PWR. The coolant inventory is in a large amount. The inertia of the coolant is large and it takes a long time for the primary system to respond to disturbances. Once-through steam generator is adopted and its water inventory is small. It is very sensitive to disturbances. These unique characteristics challenge the control system design of an SPWR. Relap5 is used to model an SPWR. In the reactor power control system, both the reactor power and the coolant average temperature are regulated by the control rod reactivity. In the feedwater flow control system, the coordination between the reactor and the turbine is considered and coolant average temperature is adopted as one measurable disturbance to balance them. The coolant pressure is adjusted based on the heaters and spray in the pressurizer. The water level in the pressurizer is controlled by the charging flow. Transient simulations are carried out to evaluate the control system performance. When the reactor is perturbed, the reactor can be stabilized under the control system.

A Nodal Dynamic Model for Control System Simulation of the HTR-PM Reactor Core

2010 ◽  
Author(s):  
Zhe Dong ◽  
Xiaojin Huang ◽  
Liangju Zhang
Keyword(s):  
Control System ◽  
High Temperature ◽  
Power Distribution ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Pebble Bed ◽  
Nodal Model ◽  
High Temperature Gas

The modular high-temperature gas-cooled nuclear reactor (MHTGR) is seen as one of the best candidates for the next generation of nuclear power plants. China began to research the MHTGR technology at the end of the 1970s, and a 10 MWth pebble-bed high temperature reactor HTR-10 has been built. On the basis of the design and operation of the HTR-10, the high temperature gas-cooled reactor pebble-bed module (HTR-PM) project is proposed. One of the main differences between the HTR-PM and HTR-10 is that the ratio of height to diameter corresponding to the core of the HTR-PM is much larger than that of the HTR-10. Therefore it is not proper to use the point kinetics based model for control system design and verification. Motivated by this, a nodal neutron kinetics model for the HTR-PM is derived, and the corresponding nodal thermal-hydraulic model is also established. This newly developed nodal model can reflect not only the total or average information but also the distribution information such as the power distribution as well. Numerical simulation results show that the static precision of the new core model is satisfactory, and the trend of the transient responses is consistent with physical rules.

Dynamic fault tree analysis of auxiliary feedwater system in a pressurized water reactor

Kerntechnik ◽  
2021 ◽  
Vol 86 (2) ◽  
pp. 164-172
Author(s):  
R. A. Fahmy ◽  
R. I. Gomaa
Keyword(s):  
Power Plant ◽  
System Reliability ◽  
Nuclear Power ◽  
Fault Tree ◽  
Fault Tree Analysis ◽  
Cost Effective ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Tree Analysis ◽  
Dynamic Fault Tree

Abstract The safe and secure designs of any nuclear power plant together with its cost-effective operation without accidents are leading the future of nuclear energy. As a result, the Reliability, Availability, Maintainability, and Safety analysis of NPP systems is the main concern for the nuclear industry. But the ability to assure that the safety-related system, structure, and components could meet the safety functions in different events to prevent the reactor core damage requires new reliability analysis methods and techniques. The Fault Tree Analysis (FTA) is one of the most widely used logic and probabilistic techniques in system reliability assessment nowadays. The Dynamic fault tree technique extends the conventional static fault tree (SFT) by considering the time requirements to model and evaluate the nuclear power plant safety systems. Thus this paper focuses on developing a new Dynamic Fault Tree for the Auxiliary Feed-water System (AFWS) in a pressurized water reactor. The proposed dynamic model achieves a more realistic and accurate representation of the AFWS safety analysis by illustrating the complex failure mechanisms including interrelated dependencies and Common Cause Failure (CCF). A Simulation tool is used to simulate the proposed dynamic fault tree model of the AFWS for the quantitative analysis. The more realistic results are useful to establish reliability cantered maintenance program in which the maintenance requirements are determined based on the achievement of system reliability goals in the most cost-effective manner.

The Load Rejection Transient Analysis of Chinshan NPP (BWR/4) Using TRACE/PARCS

2014 ◽  
Author(s):  
Jhih-Jhong Huang ◽  
Hsiung-Chih Chen ◽  
Jong-Rong Wang ◽  
Lih-Yih Liao ◽  
Chunkuan Shih ◽  
...  
Keyword(s):  
Control System ◽  
Nuclear Power ◽  
Thermal Power ◽  
Reactor Core ◽  
Second Step ◽  
System Response ◽  
Recirculation Flow ◽  
Fuel Assemblies ◽  
Flow Control System ◽  
Main Steam

Chinshan Nuclear Power Plant (NPP) is the first boiling water reactor (BWR) NPP in Taiwan. It has two units of BWR/4 reactor made by GE Company and each rated thermal power was 1775 MW without power uprate (now its rated thermal power is 1805 MW after power uprate). This research focuses on the development of the Chinshan NPP TRACE (TRAC/RELAP Advanced Computational Engine)/PARCS (Purdue Advanced Reactor Core Simulator) model. The model is done in two steps: The first step is the development of a TRACES/PARCS model of Chinshan NPP which includes the vessel, fuel assemblies, the main steam lines and important control systems (such as the feedwater control system, recirculation flow control system, etc.). Key parameters (such as power, feedwater flow rate, reactor dome temperature, etc.) were identified to refine the model further in the frame of a steady state analysis. The second step is development of TRACE/PARCS model for the load rejection transient. In order to check the system response of the Chinshan NPP TRACE/PARCS model, this study uses the load rejection transient results of startup tests to benchmark the analysis results of Chinshan NPP TRACE/PARCS model. The trends of TRACE/PARCS analysis results were consistent with the startup test data. It indicates that there is a respectable accuracy in the Chinshan NPP TRACE/PARCS model for the load rejection transient.

Experimental Investigations of Control Rod Drive Mechanism

2013 ◽  
Author(s):  
Guangyao Lu ◽  
Zhaohui Lu ◽  
Wenyuan Xiang ◽  
Yonghong Lv ◽  
Wenyou Huang ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Nuclear Reactor ◽  
Reactor Core ◽  
Experimental Investigations ◽  
Control Rod ◽  
Drive Mechanism ◽  
Start Up ◽  
Power Regulation ◽  
Set Up ◽  
Control Rods

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.

scholarly journals Early quantification of reliability for a safety critical and control system: a case study of reactor core cooling system

2018 ◽  
Vol 7 (2.12) ◽  
pp. 248
Author(s):  
Vinay Kumar ◽  
Suraj Gupta ◽  
Anil Kumar Tripathi
Keyword(s):  
Control System ◽  
Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Cooling System ◽  
Reactor Core ◽  
Safety Critical ◽  
Core Cooling ◽  
And Control

Using Probabilistic Reliability analysis for Quantifying reliability of a system is already a common practice in Reliability Engineering community. This method plays an important role in analyzing reliability of nuclear plants and its various components. In Nuclear Power Plants Reactor Core Cooling System is a component of prime importance as its breakdown can disrupt Cooling System of power plant. In this paper, we present a framework for early quantification of Reliability and illustrated with a Safety Critical and Control System as case study which runs in a Nuclear Power Plant.  

scholarly journals INTELLIGENT TECHNIQUES FOR ANOMALY DETECTION IN NUCLEAR REACTORS

2021 ◽  
Vol 247 ◽  
pp. 21011
Author(s):  
George Ioannou ◽  
Thanos Tagaris ◽  
Georgios Alexandridis ◽  
Andreas Stafylopatis
Keyword(s):  
Signal Processing ◽  
Neutron Flux ◽  
Nuclear Power ◽  
Power Plants ◽  
Reactor Core ◽  
Machine Learning Techniques ◽  
Full Potential ◽  
Pressurized Water ◽  
Learning Techniques ◽  
Water Reactors

The safe operation of nuclear power plants is highly dependent on the ability of quickly and accurately identifying possible anomalies and perturbations in the reactor. Operational defects are primarily diagnosed by detectors that capture changes in the neutron flux, placed at various points inside and outside of the core. Neutron flux signals are subsequently analyzed with signal processing techniques in an effort to be better described (have their higher-order characteristics uncovered, locate transient events, etc). To this end, the application of intelligent techniques may be extremely beneficial, as it may assist and extend the current level of analysis. Besides, the combination of signal processing methodologies and machine learning techniques in the framework of nuclear power plant data is an emerging topic that has yet to show its full potential. In this context, the current contribution attempts at introducing intelligent approaches and more specifically, deep learning techniques, in neutron flux signal analysis for the identification of perturbations and other anomalies in the reactor core that may affect its operational capabilities. The obtained results of an initial stage of analysis on neutron flux signals captured at pressurized water reactors are encouraging, underlying the robustness and the potential of the proposed approach.

Measurements of the 107Ag neutron capture cross sections with PHWT at the CSNS Back-n facility

2022 ◽  
Author(s):  
X. X. Li ◽  
L. X. Liu ◽  
W. Jiang ◽  
J. Ren ◽  
H. W. Wang ◽  
...  
Keyword(s):  
Cross Section ◽  
Cross Sections ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Astrophysics ◽  
Reaction Cross ◽  
Pressurized Water ◽  
Resonance Parameters ◽  
Control Rod ◽  
The Cross

Abstract Silver indium cadmium (Ag-In-Cd) control rod is widely used in pressurized water reactor nuclear power plants, and which is continuously consumed in a high neutron flux environment. The mass ratio of 107Ag in Ag-In-Cd control rod is 41.44%. To accurately calculate the consumption value of the control rod, a reliable neutron reaction cross section of the 107Ag is required. Meanwhile, 107Ag is also an important weak r nuclei. Thus, the cross sections for neutron induced interactions with 107Ag are very important both in nuclear energy and nuclear astrophysics. The (n, γ) cross section of 107Ag has been measured in the energy range of 1-60 eV using a back streaming white neutron beam line at China spallation neutron source. The resonance parameters are extracted by an R-matrix code. All the cross section of 107Ag and resonance parameters are given in this paper as datasets. The datasets are openly available at https://www.scidb.cn/s/aaUJbu.

Lessons Learned in Implementation of Guidelines for Quantifying Support System Initiating Event Frequencies

2008 ◽  
Author(s):  
Rupert A. Weston ◽  
Ashley J. Mossa
Keyword(s):  
Support System ◽  
Nuclear Power ◽  
Power Plants ◽  
Fault Tree ◽  
Lessons Learned ◽  
Pressurized Water ◽  
Common Cause Failures ◽  
Areas Of Concern ◽  
Pilot Implementation ◽  
American Society

The pilot implementation results for Regulatory Guide 1.200 identified four probabilistic risk assessment (PRA) technical elements that required additional guidance. One of these elements involved the use of fault tree technique to quantify the frequencies of support system initiating events (SSIEs). To address this technical element, guidelines were developed by the Electric Power Research Institute (EPRI) to provide a common industry approach for addressing the identification and quantification of SSIEs. The EPRI guidelines were issued as an interim report to allow trial use and pilot implementation by the industry prior to finalizing the guidelines. These interim guidelines provide an industry-consensus approach for addressing areas of concern in the development of support system initiating event models to ensure that the associated supporting requirements of the American Society of Mechanical Engineers (ASME) PRA Standard for internally initiated events are satisfied. A Pressurized Water Reactor Owners Group (PWROG) pilot implementation of the EPRI interim guidelines was conducted to determine whether the pilot participants have adequately addressed all areas of concern in the development of SSIE models. To determine this, a SSIE model currently used was selected by each of four the pilot participants and subject to detail review to demonstrate whether these models meet the expectations of the EPRI interim guidelines. The EPRI interim guidelines identified the areas of concern to be addressed in using fault tree technique to develop and quantify SSIE models. The guidelines addressed several areas of concern including the treatment of passive failures, the assignment of an appropriate mission time for primary and secondary failures, treatment of common cause failures (CCFs) between running and standby equipment, and consideration of all combinations of CCFs. The PWROG pilot implementation of the interim guidelines summarized the lessons learned and provided feedback to EPRI for consideration in finalizing the guidelines. In addition to the compilation of lessons learned, the PWROG implementation of the EPRI interim guidelines identified existing practices used to develop fault tree models for quantifying SSIE frequencies. Such practices did not necessarily follow a common approach and did not fully meet the expectations of the interim guidelines. Detailed reviews of the SSIE models currently in use at nuclear power plants (NPPs) for the pilot participants demonstrated that the elements of evaluation described in the interim guidelines were not addressed consistently among the PWROG pilot participants. Recommended improvements were identified and incorporated in the SSIE models to meet the expectations of the EPRI interim guidelines. The re-quantification of SSIE frequencies based on the recommended improvements, demonstrated that by not adequately addressing all elements in the evaluation, the SSIE frequency may be under-estimated.

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