COSINE Validation Experiment Plan: 1st Phase

2013 ◽  
Author(s):  
Lifang Liu ◽  
Xiaoliang Fu ◽  
Zheng Du ◽  
Nan Yu ◽  
Yanhua Yang
Keyword(s):  
Power Plant ◽  
Nuclear Power ◽  
System Analysis ◽  
Channel Code ◽  
Thermal Hydraulics ◽  
Plant Design ◽  
Code System ◽  
International Database ◽  
Valuable Data ◽  
Validation Experiment

COSINE (COre and System INtegrated Engine for design and analysis) is the software platform for nuclear power plant design and analysis being developed in China, which contains three thermal-hydraulics codes and three physics codes in the first phase of this project. Three thermal-hydraulics codes in COSINE are sub-channel code, system analysis code, and containment analysis code. Separate experiments data and integral experiments data are needed to validate the models and analyze the uncertainty of these codes. Some data are obtained from international database and organizers. At the same time, an experiment plan has been made by SNPSDC to meet the COSINE thermal-hydraulics codes validation requirements. In the 1st phase of this plan, eight separate effects facilities are being constructed to get some valuable data from January 2011 to December 2014, and the test contents has been programmed. The facilities, contents and the schedules of the tests of COSINE validation experiment plan-I are detailed in this paper.

Engineered Security Measures of Nuclear Power Plants

2003 ◽  
Author(s):  
F. J. Moody
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Human Error ◽  
Mechanical Damage ◽  
Thermal Hydraulics ◽  
Plant Design ◽  
Security Measures ◽  
Safety Features

Even in the absence of depraved terrorist threats, nuclear plants have been designed to respond safely to postulated accidents. Redundant safety features are built into plants to trigger safe shutdown and containment of possible accidents. The defined accidents range from minor leakage and operator errors to a complete loss of coolant from the reactor. Post-accident scenarios are postulated by experts in reactor and containment thermal-hydraulics, and all other sciences embraced by nuclear power plant design. The probability of failure is determined for all engineered safety systems. Then analytical and experimental programs are employed to predict the long term post-accident thermal-hydraulic state of a plant and its effect on the environment. The postulated accidents and safety system responses include effects resulting from mechanical damage and component malfunctions, such as pipe ruptures and the failure of pumps and valves. The initiating causes can be material failure, human error, and environmental effects from earthquakes, floods, and other severe acts of nature. It is prudent to build on an already established safety and accident technology to include the effects of external, planned attacks on a nuclear plant. This process includes “matching wits” with the minds of those who plot evil, and reinforcing protective security barriers where potential vulnerabilities are detected. Hard questions to ask and answer are, “What are the greatest potential security threats to a nuclear power plant? What possible human activity could make them happen? How can they be prevented?” Reactor and containment thermal-hydraulics contributes significantly to answering these questions.

Transient Validation of Areva’s New Arcadia™ Code System

2014 ◽  
Author(s):  
Sebastian Kuch ◽  
Mario Leberig ◽  
Richard Brock ◽  
Florian Reiterer ◽  
Michael Riedmann ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Leading Edge ◽  
Safety Analysis ◽  
Thermal Hydraulics ◽  
Code System ◽  
The Core ◽  
Wide Range ◽  
Fully Coupled

AREVA has developed a new leading edge code suite to meet the challenges arising from increasing expectations in nuclear power plant availability and fuel performance while satisfying stricter safety requirements. ARCADIA™ [1] is an advanced 3D coupled neutronics/thermal-hydraulics/thermal-mechanics code system for Light Water Reactor (LWR) fuel assembly and core design calculations as well as safety analysis, using a new software architecture allowing for nodal and pin-by-pin calculation capability. ARCADIA™ was licensed by the US Nuclear Regulatory Commission (NRC) for applications for PWR UO2 cores in 2013. It is on the way to be licensed in other countries for AREVA customers. ARCADIA™ contains the steady-state and transient core-simulator ARTEMIS™ [2] for core design and coupled transient safety analysis. ARTEMIS™ can be used in a coupled mode with S-RELAP5 and CATHARE 2 to allow fully coupled transient analysis, combining the sophisticated 3D core model of ARTEMIS™ with the proven system thermal-hydraulics of S-RELAP5 and CATHARE 2 including a detailed simulation of the Instrumentation and Control (I&C). This allows simulating complex transients affecting the core as well as the primary and secondary side including I&C signals and responses. For the validation of ARTEMIS™ a comprehensive set of validation cases was selected, including international benchmarks and measurements covering various classes of transients. These cases include a ‘Load Rejection to station service’ event at a German 1300 MWe plant, where a wide range of system and core parameters was measured that allow the validation of the fully coupled code system. Another validation case is a nodal recalculation of the core behavior during the pump shaft break transient that occurred in the Gösgen nuclear power plant in 1985 [3]. The paper will provide representative example results for the abovementioned validation cases.

Valves Maintainability Study on Nuclear Power Plant Design Stage

2017 ◽  
pp. 307-313
Author(s):  
Xiaohu Yang ◽  
Jiao Deng ◽  
Hong Jiang ◽  
Lifei Yang ◽  
Wen Chen
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Design Stage ◽  
Plant Design

Nuclear Power Plant Design

1969 ◽  
Vol 7 (2) ◽  
pp. 181-181
Author(s):  
T. V. Sheehan
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Plant Design

Verification of Accumulator Model in COSINE

2013 ◽  
Author(s):  
Huie Sha ◽  
Hao Zhang ◽  
Zheng Du ◽  
Yiqiang Xiong ◽  
Yanhua Yang
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Liquid Flow ◽  
Nuclear Power ◽  
Mathematic Model ◽  
Wall Friction ◽  
Plant Design ◽  
Software Platform ◽  
Gravity Effects ◽  
Isothermal System

The objective of COSINE (COre and System INtegrated Engine for design and analysis) project is to develop a software platform which is used for nuclear power plant design and safety analysis. The system code (SYST) is a part of COSINE code. In this paper, the mathematic model of accumulator is established. The model is based on following assumptions: nitrogen above liquid level in accumulator is represented in idea gas equation, water is modeled as an isothermal system. The model for liquid flow include inertia, wall friction, form loss, and gravity effects. Several cases were calculated under different conditions, and the results were compared with RELAP5. It shows that COSINE results agree well with RELAP5.

Multi-Pressure Surface Condenser Performance Evaluation: A Case Study

2004 ◽  
Author(s):  
Komandur S. Sunder Raj
Keyword(s):  
Power Plant ◽  
Nuclear Power ◽  
Performance Modeling ◽  
Plant Design ◽  
Turbine Generator ◽  
Modeling Tools ◽  
Pressure Surface ◽  
Design Basis ◽  
Predicting Performance

Surface condensers for power plant applications are generally specified and designed following turbine-condenser optimization studies. The turbine manufacturer provides turbine-generator performance data (thermal kit) at the very outset of plant design when the condenser is usually a black box and not much is known about its design. The turbine-generator guarantee would then be based on a specified condenser pressure that may or may not be attainable once the condenser is actually specified and designed. The condenser pressure used for the turbine performance guarantee might assume a single-pressure condenser while the actual design might be a multi-pressure condenser. In order to properly predict and monitor the performance and conduct diagnostics on a multi-pressure condenser, it is important to understand the design basis and develop an accurate model using performance modeling tools. The paper presents a multi-pressure condenser case study for a 600 Mwe nuclear power plant. The paper discusses the design basis used, interface between the turbine and condenser, use of a performance modeling tool for predicting performance, determining capacity losses attributable to the condenser and conducting diagnostics.

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