Investigation of critical safety function “Heat sink” at low power and cold condition for Kozloduy Nuclear Power Plant WWER-1000/V320

2012 ◽  
Vol 40 (1) ◽  
pp. 221-228 ◽  
Author(s):  
M. Andreeva ◽  
M.P. Pavlova ◽  
P.P. Groudev
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Low Power ◽  
Heat Sink ◽  
Nuclear Power ◽  
Safety Function ◽  
Cold Condition

Research on Electric and I&C Equipment Safety Function Classification of Nuclear Power Plant

2021 ◽  
Author(s):  
Wang Yuqi ◽  
Sun Qian
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Function Analysis ◽  
System Component ◽  
Design Work ◽  
Equipment Safety ◽  
Safety Function

Abstract Classification of System, Component and Structure (SSC) is the base as well as high level demand of nuclear power plant. Equipment classification including electric and Instrument and Control (I&C) equipment is the precondition of correct design regulation and standard. Safety function classification is key pass of electric and I&C equipment classification. This paper researches the method of nuclear power plant electric and I&C equipment safety function classification. Firstly from view of function, it explains the importance of function classification. Then function analysis and classification of equipment is implemented by design order. Lastly from view of accident analysis, function classification is validated, and a complete approach of function classification is formed. The purpose of this paper is the NPP electric and I&C equipment safety function classification as an example, to study and summarize the method of the electric and I&C equipment safety function classification, and to provide the basis for specific items design work according to design requirements. At the same time, a practical method is provided for other similar NPP electric and I&C equipment classification work. The electric and I&C equipment function classification of nuclear power plant satisfy the basic principles requirement of relative nuclear power rules and codes. It provides an important basis of equipment classification for next nuclear power plants.

Ultimate Heat Sink Sizing Requirements on a Month-to-Month Basis

2009 ◽  
Author(s):  
Dong Zheng ◽  
Allen T. Vieira ◽  
Julie M. Jarvis ◽  
George P. Emsurak
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Heat Sink ◽  
Nuclear Power ◽  
Meteorological Data ◽  
Meteorological Condition ◽  
Water Requirement ◽  
Next Generation ◽  
Worst Case ◽  
Automated Methods

The Ultimate Heat Sink (UHS) of a nuclear power plant is a complex cooling water system which serves the plant during normal and accident conditions. For some next generation nuclear plants, the UHS sizing is a major design and licensing analysis task. The analysis involves detailed modeling of the transient heat loads and the selection of worst-case meteorological data for the plant site. The UHS sizing requirements for a representative next generation nuclear power plant are evaluated on a month-to-month basis. This paper assesses the UHS water requirement for each month of year. The UHS analysis for a representative next generation nuclear plant with mechanical draft cooling towers and a water basin is used to determine the maximum evaporation of the basin for the worst-case meteorological data on a month-to-month basis. To size the cooling tower basin, automated methods have been developed which determine the highest evaporative losses from the basin and highest basin temperature over a 30-day design basis accident period. This paper also evaluates the month-to-month basin temperature changes. This assessment is done for a representative next generation nuclear power plant and considers the monthly historical meteorological data over 45 years. The result of this assessment of monthly UHS water requirement is of interest in assessing the margin in the UHS design. This monthly assessment is also useful in demonstrating that the automated methods used to establish the limiting 30-day meteorological condition are indeed accurate. In addition, these results may be useful in helping to plan plant maintenance activities.

Research on Operating Strategy of Total Loss of Essential Service Water System With RHR Not Connected in PWR Unit

2020 ◽  
Author(s):  
Shengtao Zhang ◽  
Ke Yi
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Heat Sink ◽  
Nuclear Power ◽  
Cooling System ◽  
Water System ◽  
Total Loss ◽  
Operating Strategy ◽  
Service Water ◽  
Essential Service

Abstract Essential Service Water System (WES) is part of the nuclear power plant cooling system which provides the final heat sink for nuclear power plants. Therefore, WES must operate stably, safely and reliably for a long time. The total loss of WES accident is a design extended condition and will result in the loss of the final heat sink of the unit. The consequences of the accident are severe. In order to deal with the accident quickly and effectively and ensure the safety and economics of the power plant in accident condition, it’s necessary to formulate corresponding treatment strategy to deal with the transient. This paper developed a strategy for dealing with the total loss of WES with Residual Heat Removal System (RHR) not connected condition in Generation III nuclear power plant. The structure of the WES system and the types of failures that may occur are analyzed, and thus the symptoms of the faults are obtained and the entry conditions for the operating strategy are determined. The effect of faults on unit equipment and safety functions and the impact on nuclear steam supply system (NSSS) control are analyzed in this paper. Combined with the unit design, the system and equipment for controlling and mitigating related safety functions are analyzed, and the mitigation method and the fallback strategy of the fault are determined. Thereby a complete operating strategy of total loss of WES with RHR not connected is obtained. In addition, this paper analyzes and evaluates the operating strategy by simulating thermal hydraulic calculation for the first time. The results show that without staff intervention Component Cooling System (WCC) temperature reached 55°C limits after running a few minutes. Based on the intervention of the operating strategy proposed in this paper, WCC temperature reached the 55°C limits when the unit was operated at about 4 hours and 55 minutes. The result shows that and the strategy can effectively alleviate the failure and provide sufficient intervention time for the operator to bring the unit to a safe state.

Evaluation Method of Response Reliability During an Accident and its Applicability to Fast Reactor Plants

2018 ◽  
Author(s):  
Masaaki Suzuki ◽  
Kazuyuki Demachi ◽  
Shigeru Takaya ◽  
Yoshitaka Chikazawa
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Fast Reactor ◽  
Nuclear Power ◽  
Evaluation Method ◽  
Severe Accident ◽  
Evaluation Procedure ◽  
System Safety ◽  
Recovery Capacity ◽  
Safety Function

In this study, we develop new indices that evaluate the response capability — response margin and response reliability — of a nuclear power plant in case of occurrence of a severe accident by considering the ability of the nuclear power plant to recover system safety functions, which may have been affected by the accident, to the minimum safety values. Further, we demonstrate a specific evaluation procedure for our proposed indices. While performing the evaluation, we consider margins with respect to a minimum safety function level and a time constraint along with their temporal changes. Additionally, for a simplified fast reactor (FR) plant model, we conduct a trial assessment of the recovery capacity of the system to ensure system safety based on the concept of new indices. Further, the applicability of the response reliability evaluation method to FR plants is discussed based on the viewpoints of reflecting the characteristics of each reactor type.

Performance of a Spray Pond for Nuclear Power Plant Ultimate Heat Sink

1977 ◽  
Vol 99 (4) ◽  
pp. 650-656
Author(s):  
V. E. Schrock ◽  
G. J. Trezek ◽  
L. R. Keilman
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Thermal Performance ◽  
Heat Sink ◽  
Nuclear Power ◽  
Field Tests ◽  
Computer Code ◽  
Loss Of Coolant Accident ◽  
Loss Of Coolant ◽  
Transient Thermal

Spray ponds have become an attractive method of providing the “ultimate heat sink”, i.e., the assured means of dissipating heat from a nuclear power plant. Two redundant spray ponds were the choice for this service in the Rancho Seco Nuclear Generating Station owned by Sacramento Municipal Utility District. This paper describes the results of full scale field tests of the Rancho Seco ponds which were conducted to verify the thermal performance, drift loss characteristics, and the capability to sustain the cooling requirements for a period of 30 days following a loss-of-coolant accident (LOCA). Correlations of local and average nozzle efficiency and of the drift loss are presented. A computer code was developed for the transient thermal performance of the pond. After verification the code was used to predict performance following LOCA under adverse meteorological conditions based on weather records.

Optimization of Test Interval of Ignalina Nuclear Power Plant Auxiliary Feedwater Pumps

2009 ◽  
Author(s):  
Roman Voronov ◽  
Robertas Alzbutas
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Failure Modes ◽  
Cooling System ◽  
Model Parameters ◽  
Testing Interval ◽  
Safety Function ◽  
Core Cooling ◽  
Safety Systems

Some safety systems of the Ignalina Nuclear Power Plant (NPP) operate in standby mode. An equipment of such systems is periodically tested and that allows timely detect and repair equipment failures. The periodic testing is an important measure of ensuring systems’ operability and reliability. However, during the test and repair the equipment cannot perform it’s safety function, therefore too often testing decreases the availability of the system. This paper describes the mathematical model that represents how availability and reliability of the systems and their components depend on testing interval, taking into account different failure modes of the equipment. This model allows to find the optimal testing interval for the safety. As an example, the auxiliary feedwater pumps, that are a part of the Ignalina NPP Reactor Emergency Core Cooling System, are analysed. The model parameters calculation is based on Ignalina NPP data regarding pumps operation and failure as well as on general Nordic NPPs reliability data (T-Book) appling Bayesian approach for parameters updating. The analysed safety system is a redundant system that consists of six pumps and other equipment. Therefore a model for multiple components failure was developed. The model accounts for actual operational requirements of the system. The results of this model are compared with usually used binominal model.

A Study About Safety I&C System Software V&V in Nuclear Power Plant

2016 ◽  
Author(s):  
Pengfei Gu ◽  
Shengchao Wang ◽  
Weihua Chen ◽  
Suyuan Yu
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Software Verification ◽  
Third Party ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Safety Function ◽  
Software Classification

Since the digital instrument and control (I&C) system was used in nuclear power plant (NPP), software Verification and Validation (V&V) is becoming more and more important for ensuring the safety function to be correctly implemented by the safety system. According to the classification of different safety functions of I&C system, with the use of digital technology, software classification needs further discussion to determine. And different classification of software performs different testing. Software V&V processes can be divided into different stages by reference to the standard IEEE 1012-2004, and each stage focuses on different contents. In China, software V&V on the NPP has been started from LingAo Phase II project and strongly done on other CPR1000 projects, which is a Generation II + pressurized water reactor. Based on the practice about YangJiang units 5 and 6 projects, combined with the relevant laws and regulation standards, this study summarizes the characteristics of the independent third party software V&V, analyzes the key points and methods of V&V activities in the software development process. As a result, it is also benefit to the design, operation and maintenance of safety I&C System as technical reference.

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