scholarly journals Control blade history reactivity and pin power effects evaluated with Westinghouse BWR nodal core simulator POLCA8

2019 ◽  
Vol 7 (3A) ◽  
Author(s):  
Petri Forslund Guimaraes
Keyword(s):  
Cross Sections ◽  
Critical Condition ◽  
Prediction Accuracy ◽  
Predictive Accuracy ◽  
Boiling Water ◽  
Simultaneous Presence ◽  
Cbh Model ◽  
Control Rod ◽  
Water Reactor ◽  
Control Rods

The so-called “Control Blade History” (CBH) effect has always posed a serious challenge for any nodal core simulator in performing Boiling Water Reactor (BWR) core analyses. In this paper a method to handle such CBH effects is proposed based on the idea of interpolating lattice physics data between two extreme cases with regard to CBH, namely, the case with the control rod always inserted during depletion and the case with the control rod never inserted during fuel irradiation. In POLCA8, the latest upgrade of the Westinghouse BWR nodal core simulator POLCA, one applies the methodology to macroscopic cross sections, discontinuity factors, pin powers and detector constants. Overall, the proposed CBH model performs very well in terms of predictive accuracy of reactivity and pin powers although simultaneous presence of control rods (CRs) and burnable absorbers (BAs) still poses a challenge due to some observed interference of their impact on reactivity. Applying the CBH model for pin power reconstruction is particularly promising and provides excellent prediction accuracy in the vicinity of the CR and at the point of CR withdrawal being the most challenging and critical condition with regard to CBH.

Multi-Group Staggered Tests for Highly Redundant Systems

2020 ◽  
Author(s):  
Shota Soga
Keyword(s):  
Boiling Water ◽  
Test Scheme ◽  
Redundant System ◽  
Common Cause ◽  
Control Rod ◽  
Water Reactor ◽  
Unique System ◽  
Common Cause Failures ◽  
Reactor Protection System ◽  
Control Rods

Abstract A reactor protection system (RPS) in a boiling water reactor (BWR) is a unique system since it represents high redundancy compared with other safety-related systems in the BWR. Owning to its high redundancy, NUREG-5500 vol.3 showed that unreliability of the General Electric BWR4-type RPS is dominated by common-cause failures (CCFs) in its highly redundant components. Since the staggered test scheme can reduce the period of latent CCF states of redundant standby components, the scheme is a good candidate to improve the reliability of the RPS. However, the scheme is impracticable for the RPS because of its high redundancy. In this paper, a new concept of a “multi-group staggered test scheme” for a highly redundant system is proposed. In the multi-group staggered test scheme, components are arranged into several groups, and all components in a group are tested simultaneously. This paper presents a preliminary example of the new scheme that is applied to control rod notch tests, in which all withdrawn control rods are moved at least one notch. It is shown that the new scheme can reduce the CCF risk in RPS.

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

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.

scholarly journals Development of Dynamic Insertion Model of Control Rods of a Boiling Water Reactor under Seismic Events

2011 ◽  
Vol 77 (774) ◽  
pp. 319-328 ◽  
Author(s):  
Yuichi KOIDE ◽  
Hirokuni ISHIGAKI ◽  
Hiromitsu MATSUNAGA ◽  
Naoki FUKUSHI ◽  
Tomomi SHIRAKI
Keyword(s):  
Boiling Water ◽  
Boiling Water Reactor ◽  
Seismic Events ◽  
Water Reactor ◽  
Control Rods

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
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