Application of LAPUR6 to Lungmen ABWR Stability Analysis

2013 ◽  
Vol 284-287 ◽  
pp. 1146-1150 ◽  
Author(s):  
Hao Tzu Lin ◽  
Jong Rong Wang ◽  
Chun Kuan Shih
Keyword(s):  
Stability Analysis ◽  
Nuclear Power ◽  
Thermal Power ◽  
Core Stability ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Design Features ◽  
Water Reactor ◽  
The Core ◽  
Stability Performance

Lungmen nuclear power plant (NPP) is the first ABWR (Advanced Boiling Water Reactor) in Taiwan and still under construction. It has two identical units with 3,926 MWt rated thermal power each and 52.2×106 kg/hr rated core flow. The core has 872 bundles of GE14 fuel, and the steam flow is 7.637×106 kg/hr at rated power. According to the chapter 4 of Lungmen NPP FSAR (Final Safety Analysis Report), the design features of Lungmen NPP improve the core stability performance and assure that it is more stable than the current BWR (Boiling Water Reactor) NPP in the normal operating regions. In this research, the LAPUR6 stability analysis of Lungmen NPP is performed in order to versify the design features of Lungmen NPP which causes the more stable than the current BWR NPPs. The analysis results of LAPUR6 indicate that the design features of Lungmen NPP can improve the core stability performance effectively and result in the more stable than the current BWR NPPs.

Experience With Extended Power Uprate at Leibstadt Nuclear Power Plant

2008 ◽  
Author(s):  
Robert Engel ◽  
Karl Zichanowicz
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Structural Damage ◽  
Nuclear Power ◽  
Thermal Power ◽  
Operating Experience ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Water Reactor ◽  
Extended Power

Since the 1970s power uprates have been employed to enhance the electricity output of nuclear power plants. Extended power uprate is defined as an increase of reactor thermal power in excess of 7% up to about 20% of the original plant licensed thermal power. Such power uprates generally require significant modification of major plant equipment and were initially implemented at boiling water reactor plants approximately 15 years ago. The early experience with extended power uprates was very positive. However, in recent years an increased number of failures of nuclear power plant components totally or partly caused by those power uprates have been reported. This paper presents the issues and operating experience with extended power uprate at the Leibstadt Boiling Water Reactor Nuclear Power Plant in Switzerland concerning the mechanical and electrical equipment. Failures such as fretting-wear and fatigue damage due to an increased vibration level, structural damage due to thermal overheating as well as shortened maintenance intervals due to elevated temperature and increased pump speed are reported. As a summary, the failures of mechanical and electrical components caused by extended power uprate experienced at the plant during the last years are related to plant reliability and have not adversely impacted plant safety.

Reactor Core Stability Analysis for a Boiling Water Reactor by a Transient Model

1984 ◽  
Vol 65 (3) ◽  
pp. 374-382 ◽  
Author(s):  
Shigeaki Tsunoyama ◽  
Tohru Mitsutake ◽  
Shigeo Ebata ◽  
Shirley A. Sandoz
Keyword(s):  
Stability Analysis ◽  
Reactor Core ◽  
Core Stability ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Transient Model ◽  
Water Reactor

Severe Accident Context Quantification for Long-Term Station Blackout in Boiling Water Reactor Nuclear Power Plants

2018 ◽  
Vol 4 (2) ◽  
Author(s):  
Gueorgui I. Petkov ◽  
Monica Vela-Garcia
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Boiling Water ◽  
Severe Accident ◽  
Boiling Water Reactor ◽  
Water Reactor ◽  
Accident Scenario ◽  
Station Blackout ◽  
Fukushima Daiichi

The realistic study of dynamic accident context is an invaluable tool to address the uncertainties and their impact on safety assessment and management. The capacities of the performance evaluation of teamwork (PET) procedure for dynamic context quantification and determination of alternatives, coordination, and monitoring of human performance and decision-making are discussed in this paper. The procedure is based on a thorough description of symptoms during the accident scenario progressions with the use of thermo-hydraulic (TH) model and severe accident (SA) codes (melcor and maap). The opportunities of PET procedure for context quantification are exemplified for the long-term station blackout (LT SBO) accident scenario at Fukushima Daiichi #1 and a hypothetic unmitigated LT SBO at peach bottom #1 boiling water reactor (BWR) reactor nuclear power plants (NPPs). The context quantification of these LT SBO scenarios is based on the IAEA Fukushima Daiichi accident report, “State-of-the-Art Reactor Consequence Analysis” and TH calculations made by using maap code at the EC Joint Research Centre.

Evaluation of the Activated Radioactivity of Turbine Equipments in BWR

2010 ◽  
Author(s):  
Masato Watanabe ◽  
Motonori Nakagami
Keyword(s):  
Cross Sections ◽  
Nuclear Power ◽  
Neutron Transport ◽  
Group Structure ◽  
Track Detector ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Main Stream ◽  
Water Reactor ◽  
Main Steam

The activated radioactivity of turbine equipments irradiated by neutron originating from 17N in the main stream is evaluated for an introduction of clearance system to boiling-water reactor (BWR) plant. The 17N, main neutron source is generated by 17O(n, p)17N reaction in the core region. The evaluation results clarified that the activated radioactivity of the turbine equipment is extremely small comparing to the clearance level. The feature of the evaluation is as follows. (1) Actual radioactive concentration of the 17N in the main steam in Hamaoka nuclear power station unit 5 (Hamaoka-5) which is an advanced boiling-water reactor (ABWR) was measured with solid-state track detector (SSTD). The 17N concentration is used for the neutron transport calculation as initial neutron sources. (2) The turbine equipments were modeled as two-dimensional geometry for DORT code. (3) Activation cross-sections for major nuclides subject to the clearance evaluation were based on JENDL3.3 on 175 energy group structure (VITAMIN-J). (4) Minor nuclides subject to the clearance evaluation were calculated with ORIGEN-S code.

Seismic Test Results of the Main Steam Isolation Valve for Japanese Boiling Water Reactor Nuclear Power Plants

2020 ◽  
Author(s):  
Hideaki Itabashi ◽  
Yoshitaka Tsutsumi ◽  
Koji Nishino ◽  
Shin Kumagai
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Shaking Table ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Central Research ◽  
Water Reactor ◽  
Seismic Tests ◽  
Main Steam

Abstract The functional requirements of Main Steam Isolation Valves (MSIVs) provided in the Boiling Water Reactor (BWR) nuclear power plants in Japan have been previously evaluated via seismic tests and so forth. However, since the response acceleration has increased in line with a recent reassessment of standard earthquake ground motions, it is necessary to evaluate seismic operability with respect to high acceleration. In addition, from the viewpoint of equipment fragility in seismic PRA, it is necessary to determine practical seismic operability limits. We used a resonant shaking table in the Central Research Institute of the Electric Power Industry (CRIEPI), which is capable of seismic tests at acceleration levels previously unachievable, and in seismic tests carried out on an MSIV, we obtained results confirming that validated seismic operability was possible even at response accelerations as high as 15 × 9.8 m/s2. The seismic operability results obtained for this MSIV will be applied to a fragility analysis of seismic PRA.

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