An advanced method for determination of loss of coolant accident in nuclear power plants

The analysis of the known results of RELAP5/V.3.2 simulation for loss of coolant & blackout accidents at WWER nuclear power plants showed that the design accident management strategies with design passive safety systems do not provide the necessary safety conditions for the maximum permissible temperature of fuel claddings, the minimum permissible level of coolant in the reactor and feed water in the steam generators. A conservative thermohydrodynamic model for a design and modernized blackout & loss-of-coolant accident management strategy at a nuclear power plant with WWER has been developed. Design passive safety systems carry out the design accident management strategy: pressurizer safety valves, secondary steam relief valves, and hydraulic reservoirs of the emergency core cooling system of the reactor. Promising afterheat removal passive systems and the reactor level and steam generator water level control systems carry out the modernized blackout & loss-of-coolant accident management strategy. The main conservative assumptions of the presented model of blackout & loss-of-coolant accidents: complete long-term failure of all electric pumps of active safety systems, the temperature of nuclear fuel in the central part of the fuel matrix is assumed as the maximum allowable one, effect of “run down” flow of a turbine feed pump and the coolant level in pressurizer on accident process is not considered. Computational modelling has found that violations of the safety conditions are over the entire range of leak sizes for the design blackout & loss-of-coolant accident management strategy. For the modernized blackout & loss-of-coolant accident management strategy, safety conditions are provided for 72 hours of the accident and more. The presented results of computational modelling of blackout accident management strategies for nuclear power plants can be used to modernize and improve symptom-informed emergency instructions and guidelines for the severe accident management at nuclear power plants with WWER. Application of the results of computational modelling of blackout accident management strategies is generally not substantiated for other types of reactor facilities. In this case, it is necessary to develop calculated models for blackout accident management taking into account the specifics of the structural and technical characteristics and operating conditions for safety related systems of nuclear power plants.

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Determination of the efficiency of electric power production by nuclear power plants using the methodology of full energy costs. Part 1. Energy costs for construction

The Problems of General Energy ◽

10.15407/pge2018.02.036 ◽

2018 ◽

Vol 2018 (2) ◽

pp. 36-44 ◽

Cited By ~ 2

Author(s):

V.D. Bilodid ◽

Keyword(s):

Electric Power ◽

Nuclear Power ◽

Power Plants ◽

Nuclear Power Plants ◽

Power Production ◽

Energy Costs ◽

Full Energy ◽

Electric Power Production

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

Volume 5: Advanced and Next Generation Reactors, Fusion Technology; Codes, Standards, Conformity Assessment, Licensing, and Regulatory Issues ◽

10.1115/icone25-67469 ◽

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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A constraint-based genetic algorithm for optimizing neural network architectures for detection of loss of coolant accidents of nuclear power plants

Neurocomputing ◽

10.1016/j.neucom.2018.09.014 ◽

2018 ◽

Vol 322 ◽

pp. 102-119 ◽

Cited By ~ 6

Author(s):

David Tian ◽

Jiamei Deng ◽

Gopika Vinod ◽

T.V. Santhosh ◽

Hissam Tawfik

Keyword(s):

Neural Network ◽

Genetic Algorithm ◽

Nuclear Power ◽

Power Plants ◽

Nuclear Power Plants ◽

Network Architectures ◽

Loss Of Coolant ◽

Neural Network Architectures

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Control Experiences Regarding Clearable Materials from Nuclear Power Plants and Nuclear Installations: Scaling Factors Determination and Measurements’ Acceptance Criteria Definition

Environments ◽

10.3390/environments6110120 ◽

2019 ◽

Vol 6 (11) ◽

pp. 120

Author(s):

Luca Albertone ◽

Massimo Altavilla ◽

Manuela Marga ◽

Laura Porzio ◽

Giuseppe Tozzi ◽

...

Keyword(s):

Nuclear Power ◽

Power Plants ◽

Nuclear Power Plants ◽

Fission Products ◽

Scaling Factors ◽

Activation Products ◽

Long Time ◽

Nuclear Installation ◽

Enriched Uranium

Arpa Piemonte has been carrying out, for a long time, controls on clearable materials from nuclear power plants to verify compliance with clearance levels set by ISIN (Ispettorato Nazionale per la Sicurezza Nucleare e la Radioprotezione - National Inspectorate for Nuclear Safety and Radiation Protection) in the technical prescriptions attached to the Ministerial Decree decommissioning authorization or into category A source authorization (higher level of associated risk, according to the categorization defined in the Italian Legislative Decree No. 230/95). After the experience undertaken at the “FN” (Fabbricazioni Nucleari) Bosco Marengo nuclear installation, some controls have been conducted at the Trino nuclear power plant “E. Fermi,” “LivaNova” nuclear installation based in Saluggia, and “EUREX” (Enriched Uranium Extraction) nuclear installation, also based in Saluggia, according to modalities that envisage, as a final control, the determination of γ-emitting radionuclides through in situ gamma spectrometry measurements. Clearance levels’ compliance verification should be performed for all radionuclides potentially present, including those that are not easily measurable (DTM, Difficult To Measure). It is therefore necessary to carry out upstream, based on a representative number of samples, those radionuclides’ determination in order to estimate scaling factors (SF), defined through the logarithmic average of the ratios between the i-th DTM radionuclide concentration and the related key nuclide. Specific radiochemistry is used for defining DTMs’ concentrations, such as Fe-55, Ni-59, Ni-63, Sr-90, Pu-238, and Pu-239/Pu-240. As a key nuclide, Co-60 was chosen for the activation products (Fe-55, Ni-59, Ni-63) and Cs-137 for fission products (Sr-90) and plutonium (Pu- 238, Pu-239/Pu-240, and Pu-241). The presence of very low radioactivity concentrations, often below the detection limits, can make it difficult to determine the related scaling factors. In this work, the results obtained and measurements’ acceptability criteria are presented, defined with ISIN, that can be used for confirming or excluding a radionuclide presence in the process of verifying clearance levels’ compliance. They are also exposed to evaluations regarding samples’ representativeness chosen for scaling factors’ assessment.

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