scholarly journals INVESTIGATION OF PITTING CORROSION ON THE MALCOMIZED STAINLESS STEEL CONTROL ROD PUSH RODS IN THE LA CROSSE BOILING WATER REACTOR.

1970 ◽  
Author(s):  
Keyword(s):  
Stainless Steel ◽  
Pitting Corrosion ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Control Rod ◽  
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.

Optimization of Boiling Water Reactor Control Rod Patterns Using Linear Search

1984 ◽  
Vol 67 (1) ◽  
pp. 38-45 ◽  
Author(s):  
Takashi Kiguchi ◽  
Kazuyori Doi ◽  
Takaharu Fukuzaki ◽  
Byorn Frogner ◽  
Chan Lin ◽  
...  
Keyword(s):  
Boiling Water ◽  
Boiling Water Reactor ◽  
Linear Search ◽  
Reactor Control ◽  
Control Rod ◽  
Water Reactor

scholarly journals Inhibition of Radioactive Contamination of Stainless Steel in Boiling Water Reactor

1983 ◽  
Vol 20 (10) ◽  
pp. 871-873 ◽  
Author(s):  
Takashi HONDA ◽  
Akira MINATO ◽  
Masakiyo IZUMIYA ◽  
Katsumi OHSUMI ◽  
Hideo MATSUBAYASHI
Keyword(s):  
Stainless Steel ◽  
Radioactive Contamination ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Water Reactor

Mechanical Testing of Roll Repaired Boiling Water Reactor Control Rod Drive Housings

2007 ◽  
Author(s):  
Fernando Corchon ◽  
In˜aki Gorrochategui ◽  
Sam Ranganath
Keyword(s):  
Thermal Cycling ◽  
Predictive Model ◽  
Surface Condition ◽  
Testing Machine ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Control Rod ◽  
Water Reactor ◽  
Head Penetration ◽  
Wall Thinning

Cracking and occasional leaks have been reported in some Boiling Water Reactor (BWR) control rod drive (CRD) stub tubes. Roll expansion of the housing against the Reactor Pressure Vessel (RPV) bottom head penetration has been used successfully to provide a leak barrier. The recently approved ASME Code Case N-730 “Roll expansion of Class 1 Control Rod Drive (CRD) Bottom Head Penetrations in BWRs, Section XI, Division 1” provides the specific criteria for the application of roll expansion. The minimum roll band length in the Code Case was based on the requirement that the roll joint capability exceed the scram forces on the CRD. The roll joint capability was based on a simplified analytical model with assumed friction factors. The predictive model was then compared with the results of extensive testing on mockups. This paper describes the results of the testing that has been performed to determine the load capability of roll repairs for different roll band lengths, material combinations (stainless steel and Alloy 600), percent wall thinning, thermal cycling and surface condition. The mock-ups were rolled using procedures and rolling equipment similar to those used in actual plant application. The mock-ups were tested in a testing machine by applying a ‘push force’ on the housing. In addition to measuring the force using a load cell, strain gages were also used to measure the strains on the housings. LVDTs were used to monitor the displacement during the test. The results showed that the resistance of the rolled joint (i.e. the load capability) is proportional to the roll length. The load capability was not a strong function of wall thinning or thermal cycling. It was strongly affected by the surface condition (e.g. oxidation) and the housing material yield strength. The predictive model was consistent with the test results and confirmed that the roll expansion joint has substantial load capability. Thus, the roll joint is not only a leak barrier, but also a structural load-carrying joint that is sufficient to resist the upward scram loads on CRDs.

A Mathematical Method for Boiling Water Reactor Control Rod Programming

1985 ◽  
Vol 71 (3) ◽  
pp. 568-579 ◽  
Author(s):  
Shinji Tokumasu ◽  
Michihiro Ozawa ◽  
Hiroshi Hiranuma ◽  
Michiro Yokomi
Keyword(s):  
Mathematical Method ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Reactor Control ◽  
Control Rod ◽  
Water Reactor

Characterization of Oxide Films on Stainless Steel Surfaces of Boiling Water Reactor Primary Cooling Systems

1987 ◽  
Vol 95 (4) ◽  
pp. 257-265 ◽  
Author(s):  
Kazumi Asahi ◽  
Masao Kitamura ◽  
Eishi Ibe ◽  
Yamato Asakura ◽  
Shunsuke Uchida
Keyword(s):  
Stainless Steel ◽  
Oxide Films ◽  
Boiling Water ◽  
Boiling Water Reactor ◽  
Cooling Systems ◽  
Water Reactor ◽  
Steel Surfaces
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