ECCS Operability With One or More Subsystem(s) Inoperable

2002 ◽  
Author(s):  
Stephen R. Swantner ◽  
James D. Andrachek
Keyword(s):  
Nuclear Power ◽  
Cooling System ◽  
Technical Specification ◽  
Safety Analysis ◽  
Design Flow ◽  
Random Component ◽  
Loss Of Function ◽  
Plant Safety ◽  
Technical Specifications ◽  
Core Cooling

Plant Technical Specifications are issued by the US NRC to ensure that safe nuclear power plant operation is maintained within the assumptions for parameters and Structures, Systems, and Components (SSCs) made in the plant safety analysis reports. The Technical Specifications are made up of Limiting Conditions for Operation (LCOs), which are the minimum set of requirements that must be met based on the assumptions of the safety analysis, Actions, which are the remedial or compensatory actions that must be taken if the LCO is not met, and Surveillance Requirements, that demonstrate that the LCO is met. The Technical Specification Actions contain Completion Times (CTs) which are the time within which remedial actions must be taken, in the event that the LCO is not met. The Improved Standard Technical Specifications (ISTS) for Westinghouse plants are contained in NUREG-1431, Revision 2. Condition A of Technical Specification 3.5.2 (ECCS- Operating) in NUREG-1431, Revision 2, allows components to be taken out of service for up to 72 hours, as long as 100% of the ECCS flow equivalent to a single Operable ECCS train exists. Condition A would allow, for example, the A train low head safety injection (LHSI) and the B train high head safety injection (HHSI) pumps to be taken out of service (for 72 hours) as long as it could be demonstrated that the remaining components could provide 100% train equivalent flow capacity. The “cross-training” allowed by this Condition in the ISTS provides flexibility when performing routine pre-planned preventive maintenance and testing, as well as during emergent corrective maintenance and testing associated with random component inoperabilities. Without this flexibility, a unit would have to initiate a plant shutdown within 1 hour, if component(s) were inoperable in different trains. In order to implement this flexibility, the various combinations of components in opposite trains must be evaluated to determine whether 100% of the ECCS flow equivalent to a single Operable ECCS train exists with those components out of service. This evaluation ensures that the safety analysis assumption associated with one train of emergency core cooling system (ECCS) is still preserved by various combinations of components in opposite trains. An ECCS train is inoperable if it is not capable of delivering design flow to the reactor coolant system (RCS). Individual components are inoperable of they are not capable of performing their design function, or support systems are not available. Due to the redundancy of trains and the diversity of subsystems, the inoperability of one component in a train does render the ECCS incapable of performing its function. Neither does the inoperability of two different components, each in a different train, necessarily result in a loss of function for the ECCS. The intent of Condition A is to maintain a combination of components such that 100% of the ECCS flow equivalent to a single Operable ECCS train remains available. This allows increased flexibility in plant operations under circumstances when components in the required subsystem may be inoperable, but the ECCS remains capable of delivering 100% of the required flow equivalent. This paper presents a methodology for identifying the minimum set of components necessary for 100% of the ECCS flow equivalent to a single Operable ECCS train. An example of the implementation of this methodology is provided for a typical Westinghouse 3-loop ECCS design.

RISK INFORMED DECISION MAKING USING MULTIOBJECTIVE OPTIMIZATION TECHNIQUE

2009 ◽  
Vol 16 (06) ◽  
pp. 483-493 ◽  
Author(s):  
G. SRINIVAS ◽  
A. K. VERMA ◽  
A. SRIVIDYA
Keyword(s):  
Multiobjective Optimization ◽  
Nuclear Power ◽  
Cooling System ◽  
Optimization Technique ◽  
Technical Specification ◽  
Solution Space ◽  
Specific Level ◽  
Safety Margins ◽  
Regulatory Bodies ◽  
Technical Specifications

Nuclear Power Plant operations are guided by Limiting conditions of operations (LCO's) laid out in the document referred to as the Technical Specifications (TS). This Technical Specification is a legitimate framework approved by the Regulatory Bodies for the Safe operations of the Nuclear Power plants. In the past, the regulatory bodies used a deterministic approach as the basis for making decisions on safety issues and organizing the activities that they carry out. This was done by applying high level criteria such as the need to provide defence in depth and adequate safety margins. However with the availability of detailed Plant Specific Level-1 Probabilistic Safety Assessment (PSA), these limiting conditions need to be reviewed/revised based on the analysis results. This review of the LCO's is not a trivial exercise if the entire solution space of the variables defining the variables has to be investigated. This paper reviews the case for revision of Surveillance test frequencies of the Emergency Core Cooling System injection valves, using the multiobjective optimization technique.

Application of TRACE and FRAPTRAN in the Spent Fuel Pool of Chinshan Nuclear Power Plant

2013 ◽  
Vol 479-480 ◽  
pp. 543-547
Author(s):  
Jong Rong Wang ◽  
Hao Tzu Lin ◽  
Wan Yun Li ◽  
Shao Wen Chen ◽  
Chun Kuan Shih
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Cooling System ◽  
Safety Issue ◽  
Water Injection ◽  
Safety Analysis ◽  
Spent Fuel ◽  
Spent Fuel Pool ◽  
Study Case

In the nuclear power plant (NPP) safety, the safety analysis of the NPP is very important work. In Fukushima NPP event, due to the earthquake and tsunami, the cooling system of the spent fuel pool failed and the safety issue of the spent fuel pool generated. In this study, the safety analysis of the Chinshan NPP spent fuel pool was performed by using TRACE and FRAPTRAN, which also assumed the cooling system of the spent fuel pool failed. There are two cases considered in this study. Case 1 is the no fire water injection in the spent fuel pool. Case 2 is the fire water injection while the water level of the spent fuel pool uncover the length of fuel rods over 1/3 full length. The analysis results of the case 1 show that the failure of cladding occurs in about 3.6 day. However, the results of case 2 indicate that the integrity of cladding is kept after the fire water injection.

RELIABILITY BASED OPTIMIZATION OF TECHNICAL SPECIFICATION OF FRONTLINE SYSTEMS OF NUCLEAR POWER PLANTS USING MULTI-OBJECTIVE APPROACH

2012 ◽  
Vol 19 (01) ◽  
pp. 1250002
Author(s):  
G. SRINIVAS ◽  
A. K. VERMA ◽  
A. SRIVIDYA ◽  
SANJAY KUMAR KHATTRI
Keyword(s):  
Genetic Algorithm ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Technical Specification ◽  
Multi Objective Optimization ◽  
Potential Core ◽  
Multi Objective ◽  
Genetic Algorithm Approach ◽  
Technical Specifications

Technical Specifications define the limiting conditions of operation, maintenance and surveillance test requirements for the various Nuclear Power plant systems in order to meet the safety requirements to fulfill regulatory criteria. These specifications impact even the economics of the plant. The regulatory approach addresses only the safety criteria, while the plant operators would like to balance the cost criteria too. The attempt to optimize both the conflicting requirements presents a case to use Multi-objective optimization. Evolutionary algorithms (EAs) mimic natural evolutionary principles to constitute search and optimization procedures. Genetic algorithms are a particular class of EA's that use techniques inspired by evolutionary biology such as inheritance, mutation, natural selection and recombination (or cross-over). In this paper we have used the plant insights obtained through a detailed Probabilistic Safety Assessment with the Genetic Algorithm approach for Multi-objective optimization of Surveillance test intervals. The optimization of Technical Specifications of three front line systems is performed using the Genetic Algorithm Approach. The selection of these systems is based on their importance to the mitigation of possible accident sequences which are significant to potential core damage of the nuclear power plant.

Modeling of Thermal Hydraulics Features of Top Water Reflood Experiment PARAMETER-SF3 Using SOCRAT/V2 Code

2009 ◽  
Author(s):  
Arcadii E. Kisselev ◽  
Valerii F. Strizhov ◽  
Alexander D. Vasiliev ◽  
Vladimir I. Nalivayev ◽  
Nikolay Ya. Parshin
Keyword(s):  
Experimental Data ◽  
Nuclear Power ◽  
Cooling System ◽  
Severe Accident ◽  
Numerical Code ◽  
Thermal Hydraulics ◽  
Vver Fuel ◽  
Design Basis ◽  
Best Estimate ◽  
Core Cooling

The PARAMETER-SF3 test conditions simulated a severe LOCA (Loss of Coolant Accident) nuclear power plant sequence in which the overheated up to 1700÷2300K core would be reflooded from the top and the bottom in occasion of ECCS (Emergency Core Cooling System) recovery. The test was successfully conducted at the NPO “LUTCH”, Podolsk, Russia, in October 31, 2008, and was the third of four experiments of series PARAMETER-SF. PARAMETER facility of NPO “LUTCH”, Podolsk, is designed for studies of the VVER fuel assemblies behavior under conditions simulating design basis, beyond design basis and severe accidents. The test bundle was made up of 19 fuel rod simulators with a length of approximately 3.12 m (heated rod simulators) and 2.92 m (unheated rod simulator). Heating was carried out electrically using 4-mm-diameter tantalum heating elements installed in the center of the rods and surrounded by annular UO2 pellets. The rod cladding was identical to that used in VVERs: Zr1%Nb, 9.13 mm outside diameter, 0.7 mm wall thickness. After the maximum cladding temperature of about 1900K was reached in the bundle during PARAMETER-SF3 test, the top flooding was initiated. The thermal hydraulic and SFD (Severe Fuel Damage) best estimate numerical complex SOCRAT/V2 was used for the calculation of PARAMETER-SF3 experiment. The counter-current flow limitation (CCFL) model was implemented to best estimate numerical code SOCRAT/V2 developed for modeling thermal hydraulics and severe accident phenomena in a reactor. Thermal hydraulics in PARAMETER-SF3 experiment played very important role and its adequate modeling is important for the thermal analysis. The results obtained by the complex SOCRAT/V2 were compared with experimental data concerning different aspects of thermal hydraulics behavior including the CCFL phenomenon during the reflood. The temperature experimental data were found to be in a good agreement with calculated results. It is indicative of the adequacy of modeling the complicated thermo-hydraulic behavior in the PARAMETER-SF3 test.

Hybrid heat pipe based passive in-core cooling system for advanced nuclear power plant

2015 ◽  
Vol 90 ◽  
pp. 609-618 ◽  
Author(s):  
Yeong Shin Jeong ◽  
Kyung Mo Kim ◽  
In Guk Kim ◽  
In Cheol Bang
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Heat Pipe ◽  
Nuclear Power ◽  
Cooling System ◽  
Core Cooling

The SCORPIO-VVER Core Monitoring and Surveillance System for VVER Type of Reactors

2014 ◽  
Author(s):  
Jozef Molnar ◽  
Radim Vocka
Keyword(s):  
Surveillance System ◽  
Nuclear Power ◽  
Power Plants ◽  
Slovak Republic ◽  
Technical Specification ◽  
Training Simulator ◽  
History Of ◽  
Monitoring And Surveillance ◽  
Technical Specifications ◽  
European Nuclear

The SCORPIO-VVER core monitoring and surveillance system has proved since the first installation at Dukovany NPP in 1999 to be a valuable tool for the reactor operators and reactor physicists. It is now installed on four units of Dukovany NPP (EDU, Czech Republic), on two units of Bohunice NPP (EBO, Slovak Republic) replacing the original Russian VK3 system and on the full scale plant training simulator at the Centre for training and education of the reactor operators and reactor physicist in Trnava (Slovak Republic). By both Czech and Slovak nuclear regulatory bodies the system was licensed as a Technical Specification Surveillance tool. Since it’s first installation, the development of SCORPIO-VVER system continues along with the changes in VVER reactors operation. The system is being adapted according the utility needs and several notable improvements in physical modules of the system were introduced. The most significant changes were done in support of the latest optimized Gd bearing fuel assemblies, improvements in the area of core design (neutron physics, core thermal hydraulics and fuel thermal mechanics), adaptation of the system to up-rated unit conditions (uprated power up to 107%), in design and methodology of the limit and technical specifications checking and improvements in the predictive part of the system. After the currently finished upgrades the SCORPIO-VVER is still in focus of Central European nuclear power plants with the roadmap of upgrades and modifications up to 2016. This paper shortly describes the system’s main functions, the history of implementation at the VVER-440 type of reactors and deals with the system’s future upgrades and plans to meet the latest requirements of efficient and safety NPP operation.

Application of TRACE and CFD in the Spent Fuel Pool of Chinshan Nuclear Power Plant

2011 ◽  
Vol 145 ◽  
pp. 78-82 ◽  
Author(s):  
Jong Rong Wang ◽  
Hao Tzu Lin ◽  
Yung Shin Tseng ◽  
Chun Kuan Shih
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Cooling System ◽  
Safety Issue ◽  
Safety Analysis ◽  
Spent Fuel ◽  
Energy Research ◽  
Important Work ◽  
Spent Fuel Pool

In the nuclear power plant (NPP) safety, the safety analysis of the NPP is very important work. In Fukushima NPP event, due to the earthquake, the cooling system of the spent fuel pool failed and the safety issue of the spent fuel pool generated. After Fukushima NPP event, INER (Institute of Nuclear Energy Research, Atomic Energy Council, R.O.C.) performed the safety analysis of the spent fuel pool for Chinshan NPP which also assumed the cooling system of the spent fuel pool failed. The geometry of the Chinshan NPP spent fuel pool is 12.17 m × 7.87 m × 11.61 m and the initial condition is 60 ¢J / 1.013 × 105 Pa. In general, the NPP safety analysis is performed by the thermal hydraulic codes. The advanced thermal hydraulic code named TRACE for the NPP safety analysis is developing by U.S. NRC. Therefore, the safety analysis of the spent fuel pool for Chinshan NPP is performed by TRACE. Besides, this safety analysis is also performed by CFD. The analysis result of TRACE and CFD are similar. The results show that the uncovered of the fuels occur in 2.7 days and the metal-water reaction of the fuels occur in 3.5 days after the cooling system failed.

Experience Gained From the Living PSA Project for Dukovany NPP

2014 ◽  
Author(s):  
Stanislav Hust’ák
Keyword(s):  
Nuclear Power ◽  
Model Development ◽  
Technical Specification ◽  
Event Analysis ◽  
The Czech Republic ◽  
Plant Safety ◽  
Risk Monitoring ◽  
Process Event ◽  
Operational Modes ◽  
Procedure Development

PSA specialists in ÚJV Řež, a. s. maintain a Living Probabilistic Safety Assessment (Living PSA) program for Dukovany Nuclear Power Plant (NPP), a four-unit VVER-440 plant type, which is operated in the Czech Republic. This project has been established as a broad framework for all plant activities related to risk assessment and as a support for risk-informed decision making carried out at this plant. In addition to recommendations for design and operation measures in order to increase the plant safety, it provides a basis and platform for all PSA applications at Dukovany NPP. The Living PSA model for Dukovany NPP is an integrated model representing the complete scope of Level 1 and Level 2 PSA for all plant operational modes. It produces the unit specific outputs for any Dukovany NPP unit. The RiskSpectrum® PSA software has been used for development and quantification of the PSA model. It is continuously updated and extensively used for various PSA applications at Dukovany NPP (e.g. risk monitoring, evaluation of plant Technical Specification changes, support for procedure development and training process, event analysis, etc.). The paper focuses on the important features of the Living PSA project for Dukovany NPP. It also discusses the broad experience gained during model development and update as well as possible future enhancements.

Application of Thermal Hydraulic Code SOCRAT/V2 to Top Water Reflood Experiment PARAMETER-SF3

2009 ◽  
Author(s):  
Alexander D. Vasiliev
Keyword(s):  
Experimental Data ◽  
Nuclear Power ◽  
Cooling System ◽  
Severe Accident ◽  
Numerical Code ◽  
Thermal Hydraulics ◽  
Vver Fuel ◽  
Design Basis ◽  
Best Estimate ◽  
Core Cooling

The PARAMETER-SF3 test conditions simulated a severe LOCA (Loss of Coolant Accident) nuclear power plant sequence in which the overheated up to 1700–2300K core would be reflooded from the top and the bottom in occasion of ECCS (Emergency Core Cooling System) recovery. The test was successfully conducted at the NPO “LUTCH”, Podolsk, Russia, in October 31, 2008, and was the third of four experiments of series PARAMETER-SF. PARAMETER facility of NPO “LUTCH”, Podolsk, is designed for studies of the VVER fuel assemblies behavior under conditions simulating design basis, beyond design basis and severe accidents. The test bundle was made up of 19 fuel rod simulators with a length of approximately 3.12 m (heated rod simulators) and 2.92 m (unheated rod simulator). Heating was carried out electrically using 4-mm-diameter tantalum heating elements installed in the center of the rods and surrounded by annular UO2 pellets. The rod cladding was identical to that used in VVERs: Zr1%Nb, 9.13 mm outside diameter, 0.7 mm wall thickness. After the maximum cladding temperature of about 1900K was reached in the bundle during PARAMETER-SF3 test, the top flooding was initiated. The thermal hydraulic and SFD (Severe Fuel Damage) best estimate numerical complex SOCRAT/V2 was used for the calculation of PARAMETER-SF3 experiment. The counter-current flow limitation (CCFL) model was implemented to best estimate numerical code SOCRAT/V2 developed for modeling thermal hydraulics and severe accident phenomena in a reactor. Thermal hydraulics in PARAMETER-SF3 experiment played very important role and its adequate modeling is important for the thermal analysis. The results obtained by the complex SOCRAT/V2 were compared with experimental data concerning different aspects of thermal hydraulics behavior including the CCFL phenomenon during the reflood. The temperature experimental data were found to be in a good agreement with calculated results. It is indicative of the adequacy of modeling the complicated thermo-hydraulic behavior in the PARAMETER-SF3 test.

Missile Safety Analysis of Nuclear Steam Turbine

2012 ◽  
Author(s):  
Rong Chen ◽  
Feng Liang ◽  
Wen Xiang Hua ◽  
Pu Ning Jiang ◽  
Xiao Zhong He
Keyword(s):  
Steam Turbine ◽  
Nuclear Power ◽  
Turbine Unit ◽  
Safety Analysis ◽  
Low Pressure ◽  
Turbine Rotor ◽  
High Energy ◽  
Fracture Mechanisms ◽  
Low Pressure Turbine ◽  
Plant Safety

This paper introduced the basic turbine missile safety analysis process for the Shanghai Turbine Plant’s nuclear power station. The turbine-generator unit is not considered as part of a nuclear power plant’s safety related equipment, but there is still the possibility that a high-energy missile created by failure of the low-pressure turbine disks could penetrate the stationary barriers of the turbine (especially the last stage of the low-pressure turbine), and damage plant safety-related equipment. The failure mechanisms of the turbine failure that could hypothetically generate missiles were described. Rotor damage due to stress corrosion cracking with brittle fracture and ductile fracture and destructive over-speed protection control system failure were discussed as the main elements to evaluate the safety of the turbine unit. Rotor stress analysis using FEA was used to evaluate the strength of the turbine rotor. Key factors that influence fracture mechanisms were discussed. Crack initiation and growth to the critical size was estimated. The strike probability of the turbine missile was accessed to assure the safety of the nuclear safety related equipment from a steam turbine unit generated missile.

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