scholarly journals (Severe accident technology of BWR (Boiling Water Reactor) reactors)

1989 ◽  
Author(s):  
L. Ott
Keyword(s):  
Boiling Water ◽  
Severe Accident ◽  
Boiling Water Reactor ◽  
Water Reactor

Enhancement of Plant Availability by Outage Shortening in European Advanced Boiling Water Reactor (EU-ABWR)

2013 ◽  
Author(s):  
Kazuhiro Kamei ◽  
Kazuyoshi Kataoka ◽  
Kazuto Imasaki ◽  
Noboru Saito
Keyword(s):  
Boiling Water ◽  
Severe Accident ◽  
Boiling Water Reactor ◽  
Plant Availability ◽  
Control Rod ◽  
Water Reactor ◽  
Spray System ◽  
Maintenance And Inspection ◽  
Reactor Pool ◽  
Safety Systems

European Advanced Boiling Water Reactor (EU-ABWR) is developed by Toshiba. EU-ABWR accommodates an armored reactor building against Airplane Crash, severe accident mitigation systems, the N+2 principle in safety systems, the diversity principle and a large output of 1600 MWe. These features enable EU-ABWR’s design objectives and principles to be consistent with the requirements in the Finnish utility and the safety requirements of Finnish YVL guide. By adopting Scandinavian outage processes, the Plant Availability is aimed to be greater than 95%. ABWRs have an excellent design potential to acheive short outage duration (e.g., shortening of maintenance and inspection duration by applying Fine Motion Control Rod Drive and Reactor Internal Pump). In addition, the EU-ABWR applies following key design improvements to reduce a refueling outage duration; a) Direct Reactor Pressure Vessel (RPV) Head Spray System, b) Self-standing Control Rods and c) Water shielding reactor pool. In this paper, coolability of RPV due to application of the Direct RPV Head Spray System is also verified with numerical evaluations by Computation Fluid Dynamics (CFD) analysis.

A sensitivity study of MELCOR nodalization for simulation of in-vessel severe accident progression in a boiling water reactor

2019 ◽  
Vol 343 ◽  
pp. 22-37 ◽  
Author(s):  
Yangli Chen ◽  
Huimin Zhang ◽  
Walter Villanueva ◽  
Weimin Ma ◽  
Sevostian Bechta
Keyword(s):  
Sensitivity Study ◽  
Boiling Water ◽  
Severe Accident ◽  
Boiling Water Reactor ◽  
Water Reactor

Thermo-Mechanical Analysis of Instrumentation Guide Tube Failure During a Severe Accident in a Nordic Boiling Water Reactor

2020 ◽  
Author(s):  
Ying Yue ◽  
Walter Villanueva ◽  
Hongdi Wang ◽  
Dingqu Wang
Keyword(s):  
Finite Element ◽  
Finite Element Model ◽  
Failure Modes ◽  
Element Model ◽  
Boiling Water ◽  
Severe Accident ◽  
Boiling Water Reactor ◽  
Guide Tube ◽  
Water Reactor ◽  
Water Reactors

Abstract Vessel penetrations are important features of both pressurized water reactors and boiling water reactors. The thermal and structural behaviour of instrumentation guide tubes (IGTs) and control rod guide tubes (CRGTs) during a severe accident is vital in the assessment of the structure integrity of the reactor pressure vessel. Penetrations may fail due to welding failure, nozzle rupture, melt-through, etc. It is thus important to assess the failure mechanisms of penetrations with sufficient details. The objective of this paper is to assess the timing and failure modes of IGTs at the lower head during a severe accident in a Nordic boiling water reactor. In this study, a three-dimensional local finite element model was established using Ansys Mechanical that includes the vessel wall, the nozzle, and the weld joint. The thermo-mechanical loads of the finite element model were based on MELCOR results of a station blackout accident (SBO) combined with a large-break loss-of-coolant accident (LBLOCA) including an external vessel cooling by water as a severe accident management strategy. Given the temperature, creep strain, elastic strain, plastic strain, stress and displacement from the ANSYS simulations, the results showed the timing and failure modes of IGTs. Failure of the IGT penetration by nozzle creep is found to be the dominant failure mode of the vessel. However, it was also found that the IGT is clamped by the flow limiter before the nozzle creep, which means that IGT ejection is unlikely.

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.

Application of Nontransfer Type Plasma Heating Technology for Core-Material-Relocation Tests in Boiling Water Reactor Severe Accident Conditions

2018 ◽  
Vol 4 (2) ◽  
Author(s):  
Yuta Abe ◽  
Ikken Sato ◽  
Toshio Nakagiri ◽  
Akihiro Ishimi ◽  
Yuji Nagae
Keyword(s):  
Plasma Heating ◽  
Boiling Water ◽  
Severe Accident ◽  
Core Material ◽  
Experimental Program ◽  
Small Scale ◽  
Boiling Water Reactor ◽  
Water Reactor ◽  
X Ray ◽  
Heating Technology

A new experimental program using nontransfer (NTR) type plasma heating is under consideration in Japan Atomic Energy Agency (JAEA) to clarify the uncertainty on core-material relocation (CMR) behavior of boiling water reactor (BWR). In order to confirm the applicability of this new technology, authors performed preparatory plasma heating tests using small-scale test pieces (107 mm × 107 mm × 222 mm (height)). An excellent perspective in terms of applicability of the NTR plasma heating to melting high melting-temperature materials such as ZrO2 has been obtained. In addition, molten pool was formed at the middle height of the test piece indicating its capability to simulate the initial phase of core degradation behavior consistent with the real UO2 fuel PHEBUS fission products (FP) tests. Furthermore, application of electron probe micro-analyzer (EPMA), scanning electron microscope (SEM)/energy dispersive X-ray spectrometry (EDX), and X-ray computed tomography (CT) led to a conclusion that the pool formed consisted mainly of Zr with some concentration of oxygen which tended to be enhanced at the upper surface region of the pool. Based on these results, an excellent perspective in terms of applicability of the NTR plasma heating technology to the severe accident (SA) experimental study was obtained.

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