Chemical Behavior of Aqueous Iodine at the Forsmark Boiling Water Reactor Nuclear Power Plants

1997 ◽  
Vol 120 (1) ◽  
pp. 48-56 ◽  
Author(s):  
Rikard Malmbeck ◽  
Gunnar Skarnemark
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Chemical Behavior ◽  
Water Reactor

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.

New generation of boiling-water-reactor nuclear power plants of increased safety

Atomic Energy ◽  
1992 ◽  
Vol 73 (1) ◽  
pp. 558-563 ◽  
Author(s):  
Yu. I. Mityaev ◽  
Yu. I. Tokarev ◽  
I. N. Sokolov ◽  
�. �. Pakkh ◽  
V. I. Abramov
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Water Reactor ◽  
New Generation

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

2017 ◽  
Author(s):  
Koji Nishino ◽  
Yoshitaka Tsutsumi ◽  
Kazuyoshi Yonekura ◽  
Nobuo Kojima ◽  
Yukio Watanabe ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Central Research ◽  
Relief Valve ◽  
Water Reactor ◽  
Seismic Tests ◽  
Main Steam

The functional requirements of Main Steam Safety Relief Valves (SRVs) provided in 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 also, it is necessary to determine the practical seismic operability limits. We used a resonant shaking table in the Central Research Institute of the Electric Power Industry (CRIEPI) [1], which is capable of seismic tests at acceleration levels that have been unachievable until now, and in seismic tests carried out on an SRV, we obtained results confirming that validated seismic operability was possible even at response accelerations as high as 20×9.8 m/s2. The seismic operability results obtained for this SRV will be applied to a fragility analysis of seismic PRA.

Computerized Automatic Start-Up Control for Boiling Water Reactor Nuclear Power Plants

1984 ◽  
Vol 17 (2) ◽  
pp. 2219-2224
Author(s):  
K. Hirata ◽  
T. Tojo ◽  
Y. Murata ◽  
K. Niki ◽  
H. Nakai ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Water Reactor ◽  
Start Up

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.

scholarly journals Calculation of Potential Radiation Doses Associated with Predisposal Management of Dismantled Steam Generators from Nuclear Power Plants

2020 ◽  
Vol 12 (12) ◽  
pp. 5149
Author(s):  
Ga Hyun Chun ◽  
Jin-ho Park ◽  
Jae Hak Cheong
Keyword(s):  
Nuclear Power ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Radiation Doses ◽  
Steam Generators ◽  
Pressurized Water ◽  
Water Reactor ◽  
General Pathway ◽  
Heavy Water Reactor ◽  
Past Experiences

Although the generation of large components from nuclear power plants is expected to gradually increase in the future, comprehensive studies on the radiological risks of the predisposal management of large components have been rarely reported in open literature. With a view to generalizing the assessment framework for the radiological risks of the processing and transport of a representative large component—a steam generator—12 scenarios were modeled in this study based on past experiences and practices. In addition, the general pathway dose factors normalized to the unit activity concentration of radionuclides for processing and transportation were derived. Using the general pathway dose factors, as derived using the approach established in this study, a specific assessment was conducted for steam generators from a pressurized water reactor (PWR) or a pressurized heavy water reactor (PHWR) in Korea. In order to demonstrate the applicability of the developed approach, radiation doses reported from actual experiences and studies are compared to the calculated values in this study. The applicability of special arrangement transportation of steam generators assumed in this study is evaluated in accordance with international guidance. The generalized approach to assessing the radiation doses can be used to support optimizing the predisposal management of large components in terms of radiological risk.

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