scholarly journals Station Black-Out Analysis with MELCOR 1.8.6 Code for Atucha 2 Nuclear Power Plant

2012 ◽  
Vol 2012 ◽  
pp. 1-17 ◽  
Author(s):  
Analia Bonelli ◽  
Oscar Mazzantini ◽  
Martin Sonnenkalb ◽  
Marcelo Caputo ◽  
Juan Matias García ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Heavy Water ◽  
Nuclear Power ◽  
Failure Time ◽  
Thermal Power ◽  
Radiation Heat Transfer ◽  
Primary System ◽  
Water Release ◽  
Plant Safety

A description of the results for a Station Black-Out analysis for Atucha 2 Nuclear Power Plant is presented here. Calculations were performed with MELCOR 1.8.6 YV3165 Code. Atucha 2 is a pressurized heavy water reactor, cooled and moderated with heavy water, by two separate systems, presently under final construction in Argentina. The initiating event is loss of power, accompanied by the failure of four out of four diesel generators. All remaining plant safety systems are supposed to be available. It is assumed that during the Station Black-Out sequence the first pressurizer safety valve fails stuck open after 3 cycles of water release, respectively, 17 cycles in total. During the transient, the water in the fuel channels evaporates first while the moderator tank is still partially full. The moderator tank inventory acts as a temporary heat sink for the decay heat, which is evacuated through conduction and radiation heat transfer, delaying core degradation. This feature, together with the large volume of the steel filler pieces in the lower plenum and a high primary system volume to thermal power ratio, derives in a very slow transient in which RPV failure time is four to five times larger than that of other German PWRs.

Various MTC Studies of TRACE with RCS Pressure Predictions under ATWS for Maanshan

2013 ◽  
Vol 284-287 ◽  
pp. 1151-1155
Author(s):  
Che Hao Chen ◽  
Jong Rong Wang ◽  
Hao Tzu Lin ◽  
Chun Kuan Shih
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Thermal Power ◽  
Nuclear Regulatory Commission ◽  
The United States ◽  
Graphic User Interface ◽  
Pressurized Water ◽  
Plant Safety ◽  
Regulatory Commission

The objective of this study is to utilize TRACE (TRAC/RELAP Advanced Computational Engine) code to analyze the reactor coolant system (RCS) pressure transients under ATWS (Anticipated Transient Without Scram) for Maanshan PWR (Pressurized Water Reactor) in various MTC (Moderator Temperature Coefficient) conditions. TRACE is an advanced thermal hydraulic code for nuclear power plant safety analysis, which is currently under development by the United States Nuclear Regulatory Commission (USNRC). A graphic user interface program named SNAP (Symbolic Nuclear Analysis Package), which processes inputs and outputs for TRACE is also under development. Maanshan nuclear power plant (NPP) is the only Westinghouse PWR in Taiwan. The rated core thermal power of Maanshan with MUR (Measurement Uncertainty Recapture) is 2822 MWt. In document SECY-83-293, all initializing events were classified as either turbine trip or non-turbine trip events and their ATWS risks were also evaluated according to these two events. Loss of condenser vacuum (LOCV) and Loss of normal feedwater (LONF) ATWS were identified as limiting transients of turbine trip and non-turbine trip events in this study. According to ASME Code Level C service limit criteria, the RCS pressure for Maanshan NPP must be under 22.06 MPa. Furthermore, we select the LOCV transient to analyze various MTC effects on RCS pressure variations.

scholarly journals An Investigation on the Possible Radioactive Contamination of Environment during a Steam-Line Break Accident in a VVER-1200 Nuclear Power Plant

2019 ◽  
Vol 14 (2) ◽  
pp. 299-311
Author(s):  
Abid Hossain Khan ◽  
Angkush Kumar Ghosh ◽  
Md Sumon Rahman ◽  
S M Tazim Ahmed ◽  
C L Karmakar
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Thermal Power ◽  
Steam Line ◽  
Radioactive Elements ◽  
Time Duration ◽  
Plant Safety ◽  
Safety Features ◽  
Safety Systems

In this work, the possibility of contamination of environment by radioactive elements due to a steam-line break accident has been investigated for a VVER-1200 type nuclear power plant. Personal Computer Transient Analyzer (PCTRAN) has been used to generate the response data of the plant safety systems numerically for an accidental condition like such. A break of 1000 cm2 in the A-loop of the steam line has been considered. A break of the size is considered a “Large Break”, which is believed to be responsible for multiple serious accidents in the past. Also, it has also been assumed that off-site AC power supply is unavailable. Simulations were run for time duration of 300 seconds since most of the safety features of the plant should respond within 50 seconds from the initiation of the accident. Results show that SCRAM is initiated within 22.5 seconds from the emergence of the break, which limited the peak core thermal power to around 105% of the nominal value. The peak temperatures of fuel elements and fuel cladding are recorded to be around 1850oC and 620oC respectively, which are both within the safety limits. The pressure inside reactor pressure vessel has not undergone any significant changes, showing no sign of failure. Again, the pressure inside the reactor containment building is kept within 2.5 bar by the safety systems, indicating that there is no possibility of containment failure due to over-pressure. Finally, the readings from radiation monitor show that there is no noticeable release of radioactive elements to the environment during the accident. Therefore, it may be concluded that the release of radioactive elements in the surrounding environment during a steam-line break accident is very unlikely provided that the plant safety systems are fully functional.

Criticality analysis for safety-critical software in nuclear power plant distributed control system

Kerntechnik ◽  
2021 ◽  
Vol 86 (5) ◽  
pp. 343-352
Author(s):  
J. Cui ◽  
Y. Cai ◽  
Y. Wu
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Application Process ◽  
Software Module ◽  
Critical System ◽  
Safety Critical ◽  
Plant Safety ◽  
Safety Integrity Level ◽  
Criticality Analysis

Abstract Software criticality analysis examines the degree of contribution that each individual failure mode of a software component has on the reliability of software. Higher safety integrity levels are assigned to software modules whose failures cause an unacceptable impact on the operation of the system, and these levels require the implementation of more rigorous software quality assurance measures as defined in IEEE Std 1012 and in the customer’s system requirements specification. In this paper, a novel software criticality analysis method is proposed, the results of which can be used to guide the development of newly developed software and the procurement of Commercial-Off-The-Shelf (COTS) software. The software structure is first analyzed and the software is divided into modules according to their functions. Then the criticality levels of software components are preliminarily classified by means of a safety criticality preliminary analysis tree, followed by their verification through the software hazard and operability analysis (HAZOP). Finally, the target Safety Integrity Level (SIL) of each software module is determined based on its criticality level and the overall safety objective (i. e., SIL) of the system it resides in. As an example, this proposed method is applied to a nuclear power plant safety-critical system to demonstrate the detail application process and to verify the feasibility of the method. Compared with the existing software criticality analysis methods, this method has better operability and verifiability, and can be utilized as a technical guidance for the software criticality analysis of nuclear power plant digital control systems.

Design of the Communication Independence for ACPR1000 Nuclear Power Plant Digital Safety I&C System

2017 ◽  
Author(s):  
Sun Na ◽  
Shi Gui-lian ◽  
Xie Yi-qin ◽  
Li Gang ◽  
Jiang Guo-jin
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Communication Networks ◽  
Nuclear Power ◽  
Protection System ◽  
Plant Safety ◽  
Safety And Reliability ◽  
Digital Protection ◽  
Architecture And Design ◽  
Reactor Protection System

Communication independence is one of the key criteria of digital safety I&C system design. This paper mainly analyzes the requirements for communication independence in safety regulations and standards, and then introduces the architecture and design features, including communication failure processing measures, of communication networks of ACPR1000 nuclear power plant safety digital protection system based on FirmSys platform developed by CTEC. The communication design meets the regulations requirements and effectively improves the safety and reliability of the system, and it is successfully applied in reactor protection system (RPS) of Yang Jiang nuclear power plant unit 5&6. In addition this design can provide reference for communication designs of other NPPs and industries.

A Simulation of Small Break Loss of Coolant Accident (SB-LOCA) in AP1000 Nuclear Power Plant Using RELAP5-MV

2017 ◽  
Author(s):  
Eltayeb Yousif ◽  
Zhang Zhijian ◽  
Tian Zhao-fei ◽  
A. M. Mustafa
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Power Systems ◽  
Nuclear Power ◽  
Power Plants ◽  
Computational Simulation ◽  
Primary System ◽  
Transient Time ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant

To ensure effective operation of nuclear power plants, it is very important to evaluate different accident scenarios in actual plant conditions with different codes. In the field of nuclear safety, Loss of Coolant Accident (LOCA) is one of the main accidents. RELAP-MV Visualized Modularization software technology is recognized as one of the best estimated transient simulation programs of light water reactors, and also has the options for improved modeling methods, advanced programming, computational simulation techniques and integrated graphics displays. In this study, transient analysis of the primary system variation of thermo-hydraulics parameters in primary loop under SB-LOCA accident in AP1000 nuclear power plant (NPP) is carried out by Relap5-MV thermo-hydraulics code. While focusing on LOCA analysis in this study, effort was also made to test the effectiveness of the RELAP5-MV software already developed. The accuracy and reliability of RELAP5-MV have been successfully confirmed by simulating LOCA. The calculation was performed up to a transient time of 4,500.0s. RELAP5-MV is able to simulate a nuclear power system accurately and reliably using this modular modeling method. The results obtained from RELAP5 and RELAP5-MV are in agreement as they are based on the same models though in comparison with RELAP5, RELAP5-MV makes simulation of nuclear power systems easier and convenient for users most especially for the beginners.

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