scholarly journals CATHARE Assessment of PACTEL LOCA Experiments with Accident Management

2010 ◽  
Vol 2010 ◽  
pp. 1-8
Author(s):  
Luben Sabotinov ◽  
Heikki Purhonen ◽  
Vesa Riikonen
Keyword(s):  
Hydraulic System ◽  
Nuclear Power ◽  
Reasonable Agreement ◽  
Power Plants ◽  
Test Facility ◽  
Passive Safety ◽  
Work Package ◽  
Analytical Work ◽  
Accident Management ◽  
Physical Phenomena

This paper summarizes the analysis results of three PACTEL experiments, carried out with the advanced thermal-hydraulic system computer CATHARE 2 code as a part of the second work package WP2 (analytical work) of the EC project “Improved Accident Management of VVER nuclear power plants” (IMPAM-VVER). The three LOCA experiments, conducted on the Finnish test facility PACTEL (VVER-440 model), represent 7.4% cold leg breaks with combination of secondary bleed and primary bleed and feed and different actuation modes of the passive safety injection. The code was used for both defining and analyzing the experiments, and to assess its capabilities in predicting the associated complex VVER-related phenomena. The code results are in reasonable agreement with the measurements, and the important physical phenomena are well predicted, although still further improvement and validation might be necessary.

scholarly journals Simulation of MASPn Experiments in MISTRA Test Facility with COCOSYS Code

2008 ◽  
Vol 2008 ◽  
pp. 1-7 ◽  
Author(s):  
Mantas Povilaitis ◽  
Egidijus Urbonavičius
Keyword(s):  
Boundary Conditions ◽  
Nuclear Power ◽  
Power Plants ◽  
Volume Fraction ◽  
Characteristic Length ◽  
Initial Conditions ◽  
Test Facility ◽  
Work Package ◽  
Acceptable Agreement

An issue of the stratified atmospheres in the containments of nuclear power plants is still unresolved; different experiments are performed in the test facilities like TOSQAN and MISTRA. MASPn experiments belong to the spray benchmark, initiated in the containment atmosphere mixing work package of the SARNET network. The benchmark consisted of MASP0, MASP1 and MASP2 experiments. Only the measured depressurisation rates during MASPn were available for the comparison with calculations. When the analysis was performed, the boundary conditions were not clearly defined therefore most of the attention was concentrated on MASP0 simulation in order to develop the nodalisation scheme and define the initial and boundary conditions. After achieving acceptable agreement with measured depressurisation rate, simulations of MASP1 and MASP2 experiments were performed to check the influence of sprays. The paper presents developed nodalisation scheme of MISTRA for the COCOSYS code and the results of analyses. In the performed analyses, several parameters were considered: initial conditions, loss coefficient of the junctions, initial gradients of temperature and steam volume fraction, and characteristic length of structures. Parametric analysis shows that in the simulation the heat losses through the external walls behind the lower condenser installed in the MISTRA facility determine the long-term depressurisation rate.

An Integrated Experimental Test Facility to Support Development of the Passive Containment Cooling System of HPR1000

2020 ◽  
Author(s):  
Wei Li ◽  
Shuhong Du ◽  
Weiquan Gu ◽  
Nan Zhang ◽  
Ming Ding ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Cooling System ◽  
Heat Removal ◽  
Test Facility ◽  
Passive Safety ◽  
Significant Feature ◽  
Engineering System ◽  
Passive Safety System ◽  
Design And Manufacture

Abstract HPR1000 is an advanced nuclear power plant with the significant feature of an active and passive safety design philosophy, developed by the China National Nuclear Corporation. It is based on the large accumulated knowledge from the design, construction as well as operations experience of nuclear power plants in China. The passive containment cooling system (PCS) of HPR1000 is an important and innovative passive safety system to suppress the pressure in the containment during LOCA. In this paper, the detailed design process of PCS is reviewed, and an integrated experiment facility for the study on the coupling behavior between PCS and thermal hydraulic characteristics in the containment is described, and arrangement of measuring points including temperature, pressure, gas composition and so on are introduced in detailed. Also, the experimental energy released and energy vent to ensure the similarity of containment pressure response, thermal stratification and PCS heat removal is introduced. According to this versatile experiment facility can conduct real-engineering system test which is designed to support the PCS development. In addition, this valuable experience in the design and manufacture of integrated experiment facility can provide important technical support and guidance for the China next generation advanced PWR as well as safety related system.

The reactor research and test facility

2020 ◽  
pp. 69-79
Author(s):  
Andrey S. KIRILLOV ◽  
Aleksandr P. PYSHKO ◽  
Andrey A. ROMANENKO ◽  
Valery I. YARYGIN
Keyword(s):  
Measurement System ◽  
Nuclear Power ◽  
Power Plants ◽  
Test Facility ◽  
Physical Parameters ◽  
Pumping System ◽  
Current State ◽  
History Of ◽  
Special Measurement ◽  
Space Nuclear Power

The paper describes an overview of the history of development and the current state of JSC “SSC RF-IPPE” reactor research and test facility designed for assembly, research and full-scale life energy tests of space nuclear power plants with a thermionic reactor. The leading specialists involved in development and operation of this facility are represented. The most significant technological interfaces and upgrade operations carried out in the recent years are discussed. The authors consider the use of an oil-free pumping system as part of this facility during degassing and life testing. Proposed are up-to-date engineering solutions for development of the automated special measurement system designed to record NPP performance, including volt-ampere characteristics together with thermophysical and nuclear physical parameters of a ground prototype of the space nuclear power plant. Key words: reactor research and test facility, thermionic reactor, life energy tests, oil-free pumping system, automated special measurement system, volt-ampere characteristics.

scholarly journals A decision-support approach to severe accident management in nuclear power plants

2020 ◽  
pp. 1-12
Author(s):  
Marko Bohanec ◽  
Ivan Vrbanić ◽  
Ivica Bašić ◽  
Klemen Debelak ◽  
Luka Štrubelj
Keyword(s):  
Decision Support ◽  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Severe Accident ◽  
Accident Management

scholarly journals A modern data measurement system to study and test thermionic heat to electricity converters

2018 ◽  
Vol 4 (3) ◽  
pp. 179-183
Author(s):  
Andrey Kirillov ◽  
Valeriy Yarygin
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Test System ◽  
Temperature Fields ◽  
Test Facility ◽  
Measurement Systems ◽  
Data Measurement ◽  
Functional Capabilities ◽  
Thermionic Nuclear Power ◽  
A Current

Studies and tests are conducted to determine the performance of thermionic nuclear power plants (TNPP) a stage in which is pre-irradiation testing of laboratory thermionic converters (TIC) with flat and cylindrically shaped electrodes using test facilities fitted with automated data measurement systems (DMS). The TIC volt-ampere characteristics (VAC) are measured in the DMS jointly with the measured test section and experimental test facility temperature fields. The structure and the characteristics of a DMS based on products from ICP DAS Co., Ltd are presented. A developed VAC measurement program providing the operator with a convenient graphic interface and enabling adjustment of the measurement parameters has been considered. The VAC recording errors in the process of measurements have been determined using TIC simulators. The error in the VAC diffusion portion on a simulator (with a current of less than 3 A) is not more than 1%. Thanks to the use of modern components, the developed DMS offers extended functional capabilities for measuring the thermocouple signals in an experimental electrophysical test facility. The DMS structure provides for the convenience of scaling (through a larger number of measuring channels) and makes it possible to add modules from other manufacturers. The experience of operating this DMS will be used to develop the DMS for an in-pile test system designed for similar functions.

scholarly journals Seismic Vulnerability Assessment of Site-Vicinity Infrastructure for Supporting the Accident Management of a Nuclear Power Plant

2017 ◽  
Vol 2017 ◽  
pp. 1-7 ◽  
Author(s):  
T. J. Katona ◽  
A. Vilimi
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Disaster Management ◽  
Nuclear Power ◽  
Power Plants ◽  
Public Utility ◽  
Qualitative Assessment ◽  
Design Basis ◽  
Accident Management ◽  
Plant Site

Nuclear power plants shall be designed to resist the effects of large earthquakes. The design basis earthquake affects large area around the plant site and can cause serious consequences that will affect the logistical support of the emergency actions at the plant, influence the psychological condition of the plant personnel, and determine the workload of the country’s disaster management personnel. In this paper the main qualitative findings of a study are presented that have been performed for the case of a hypothetical 10−4/a probability design basis earthquake for the Paks Nuclear Power Plant, Hungary. The study covers the qualitative assessment of the postearthquake conditions at the settlements around the plant site including quantitative evaluation of the condition of dwellings. The main goal of the recent phase of the study was to identify public utility vulnerabilities that define the outside support conditions of the nuclear power plant accident management. The results of the study can be used for the planning of logistical support of the plant accident management staff. The study also contributes to better understanding of the working conditions of the disaster management services in the region around the nuclear power plant.

Issues Associated With the Development of Severe Accident Management Guidelines for NPPs in China

2013 ◽  
Author(s):  
Zhiyi Yang ◽  
Yimin Chong ◽  
Chun Li ◽  
Jian Deng ◽  
Xianhong Xu ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Severe Accident ◽  
Regulatory Body ◽  
Control Room ◽  
Technical Approach ◽  
Management Guidelines ◽  
Accident Management ◽  
Term Monitoring

After Fukushima nuclear accident, the Severe Accident Management Guidelines (SAMGs) are required according to the policy of the regulatory body in China. Most nuclear power plants (NPPs) in China adopt the technical approach of generic-SAMG of the Westinghouse Owner Group, consisting of severe accident control room guideline (SACRG), diagnostic flow chart (DFC), severe accident guideline (SAG), severe challenge status tree (SCST), severe challenge guideline (SCG), technical support center (TSC) long term monitoring guideline, and SAMG termination guideline (SAEG). A number of issues have been identified during the development of the SAMGs for M310+ NPPs, which is a dominant reactor type in China. The paper discussed these issues and identified some considerations for their resolution.

Analysis of the Core Exit Temperature and the Peak Cladding Temperature During a SBLOCA: Application to a Scaled-Up Model

2016 ◽  
Vol 2 (2) ◽  
Author(s):  
Andrea Querol ◽  
Sergio Gallardo ◽  
Gumersindo Verdú
Keyword(s):  
Time Delay ◽  
Large Scale ◽  
Test Facility ◽  
The Core ◽  
Accident Management ◽  
Loss Of Coolant ◽  
Exit Temperature ◽  
Scale Test ◽  
Versus Peak ◽  
Physical Phenomena

During loss-of-coolant accidents (LOCAs), operators may start accident management (AM) actions when the core exit temperature (CET) measured by thermocouples exceeds a certain value. However, a significant time delay and temperature discrepancy in the superheat detection were observed in several facilities. This work is focused on clarifying CET thermocouple responses versus peak cladding temperature (PCT) and studying if the same physical phenomena are reproduced in two TRACE5 models with different geometry (a large-scale test facility (LSTF) and a scaled-up LSTF) during a pressure vessel (PV) upper head small break LOCA (SBLOCA). Results obtained show that the delay between the core uncover and the CET excursion is reproduced in both cases.

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