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.

Consideration of Relationship Between Decommissioning With Digital-Twin and Knowledge Management

2020 ◽  
Author(s):  
Yasuyoshi Taruta ◽  
Satoshi Yanagihara ◽  
Takashi Hashimoto ◽  
Shigeto Kobayashi ◽  
Yukihiro Iguchi ◽  
...  
Keyword(s):  
Knowledge Management ◽  
Power Plant ◽  
Nuclear Power Plant ◽  
Digital Technology ◽  
Nuclear Power ◽  
Next Generation ◽  
Digital Twin ◽  
Advanced Technologies ◽  
The World

Abstract Decommissioning of Nuclear Power Plant is a long-term project during which generations are expected to change. Therefore, it is necessary to appropriately transfer knowledge, technology and skills to the next generation. In recent years, in the world of decommissioning, attempts have been made to apply advanced technologies such as utilization of knowledge management and digital technology. This study describes adaptation in decommissioning from viewpoint of utilizing IT technology called digital twin and aspect of knowledge management.

scholarly journals Hypothetical Accident in Polish Nuclear Power Plant. Worst Case Scenario for Main Polish Cities

2019 ◽  
Vol 26 (1) ◽  
pp. 9-28
Author(s):  
Andrzej Mazur
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Atmospheric Transport ◽  
Radioactive Material ◽  
Second Phase ◽  
Pollution Transport ◽  
Worst Case ◽  
Hypothetical Accident

Abstract Poland is under threat of potential accidents in nuclear power plants located in its close vicinity, in almost all neighboring countries. Moreover, there are plans to establish a new nuclear power plant in Polish coast. In this paper the analysis of atmospheric transport of radioactive material released during a potential accident in the future nuclear power plant is presented. In the first part of study transport of radioactivity as seen from the long time perspective is analyzed. This involves trajectory analysis as a tool for describing the statistics of air pollution transport pattern and screening the meteorological situations for episode studies. Large sets of meteorological data for selected episodes were stored as a result of this process. Estimation of risk includes both analysis of the consequences and probability analysis of an occurrence of such situation. Episodes then were comprehensively studied in the second phase of the study, using the Eulerian dispersion model for simulation of atmospheric transport of pollutants. This study has proven that the time needed for reaction in case of (hypothetical) accident is enormously short.

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.

scholarly journals Assessment of Radioactive Gaseous Effluent Released From Nuclear Power Plant Ninh Thuan 1 under Scenario of Ines-Level 5 Nuclear Accident

2015 ◽  
Vol 25 (2) ◽  
pp. 165
Author(s):  
Nguyen Tuan Khai ◽  
Le Dinh Cuong ◽  
Do Xuan Anh ◽  
Duong Duc Thang ◽  
Trinh Van Giap ◽  
...  
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Emergency Preparedness ◽  
Nuclear Power ◽  
Meteorological Data ◽  
Nuclear Regulatory Commission ◽  
Nuclear Accident ◽  
Total Power ◽  
Model Calculations ◽  
Technical Parameters

Based on guides RG 1.109, RG 1.111 published by United States Nuclear Regulatory Commission (USNRC)~our~research~concentrates on assessing radiation doses caused by radioactive substances released from the nuclear power plant (NPP) Ninh Thuan 1 to the environment under scenario of an INES-level 5 nuclear accident caused by two incidents of the station black out (SBO) and loss of coolant accident (LOCA) using software~RASCAL4.3~provided by the Emergency Operations Center of USNRC. The plant Ninh Thuan 1 is assumed to use the VVER-1200 technology with a total power of 2400 MW\(_{e}\). The input data for the model calculations is based on building the accident scenario, the technical parameters of VVER-1200 technology and the meteorology. The meteorological data on dry and rainy seasons, which are typical for the Ninh Thuan region was also considered. The \(X/Q\) (s/m\(^{3}\)) quality and the maximum dose values were calculated within an area of 40 km radius from the NPP site, where \(X/Q\) (s/m\(^{3}\)) is the ratio of activity concentration to release rate. Based on the obtained results on dose distribution the necessary measures for nuclear emergency preparedness have been proposed according to the IAEA recommendations.

Study on Emergency Planning Zone Determination for CAP200 Small Modular Reactor

2018 ◽  
Author(s):  
Wang Xuan ◽  
Du Fenglei ◽  
Sun Dawei ◽  
Tang Te
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Source Term ◽  
Meteorological Data ◽  
Emergency Planning ◽  
Constraint Factor ◽  
Study Sites

Determination of the SMR emergency planning zone (EPZ) is one of the important external constraint factor of its marketing and application, which means that it is very important to formulate appropriate classification criteria and establish proper size range. In China, due to the requirement of “Criteria for emergency planning and preparedness for nuclear power plants: Part 1, The dividing of emergency planning zone.” (GB/T 17680.1-2008), for PWR nuclear power plant, its external plume EPZ should be within 7km–10km, and its internal plume EPZ should be within 3km∼5km. However, the scope of the standard for the emergency planning area is currently limited to conventional nuclear power plants, and for the current SMR, its emergency planning size is not included. In this paper, we will analyze the classification method of SMR EPZ based on the traditional Nuclear Power Plants feedback experience, including selection of source term, accident cutoff probability, determination method of the plume EPZ and the ingestion EPZ. Three typical nuclear power plant sites in China are chosen as CAP200 case study sites, including two inland nuclear power plant sites and one coastal site. The three sites can represent most of the meteorological and terrain characters of China nuclear power plants. According to the CAP200 source term and meteorological data of the sites, MACCS2 computer program is used to calculate the severe accidents consequence. Conclusions show that for the CAP200 SMR, the accident cutoff probability can be 1.0E−08 to 1.0E−07 per reactor per year, and its project dose exceeding probability in the three sites boundary is far below 30%, which directs that for CAP200 SMR, its plume and ingestion emergence planning zone is limited to the on-site area, and its off-site emergency response can be simplified.

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