The Effective Convectivity Model for Simulation of Molten Metal Layer Heat Transfer in a Boiling Water Reactor Lower Head

This paper is concerned with the development of approaches for assessment of core debris heat transfer and Control Rod Guide Tube (CRGT) cooling effectiveness in case of a Boiling Water Reactor (BWR) severe accident. We consider a hypothetical scenario with stratified (metal layer atop) melt pool in the lower plenum. Effective Convectivity Model (ECM) and Phase-Change ECM (PECM) are developed for the modeling of molten metal layer heat transfer. The PECM model takes into account reduced convection heat transfer in mushy zone and compositional convection that enables simulations of noneutectic binary mixture solidification and melting. The ECM and PECM are (i) validated against relevant experiments for both eutectic and noneutectic mixtures and (ii) benchmarked against CFD-generated data including the local heat transfer characteristics. The PECM is then applied to the analysis of heat transfer in a stratified heterogeneous debris pool taking into account CRGT cooling. The PECM simulation results show apparent efficacy of the CRGT cooling which can be utilized as Severe Accident Management (SAM) measure to protect the vessel wall from focusing effect caused by metallic layer.

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

Volume 1: Beyond Design Basis; Codes and Standards; Computational Fluid Dynamics (CFD); Decontamination and Decommissioning; Nuclear Fuel and Engineering; Nuclear Plant Engineering ◽

10.1115/icone2020-16236 ◽

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.

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Fuel heat transfer modelling in reduced-order boiling water reactor dynamics models

Annals of Nuclear Energy ◽

10.1016/s0306-4549(98)00013-9 ◽

1998 ◽

Vol 25 (16) ◽

pp. 1287-1300 ◽

Cited By ~ 10

Author(s):

T.H.J.J. Van der Hagen

Keyword(s):

Heat Transfer ◽

Boiling Water ◽

Boiling Water Reactor ◽

Reduced Order ◽

Water Reactor ◽

Reactor Dynamics ◽

Transfer Modelling ◽

Dynamics Models

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Parametric Study of Recriticality in a Boiling Water Reactor Severe Accident

Nuclear Technology ◽

10.13182/nt94-a34989 ◽

1994 ◽

Vol 107 (2) ◽

pp. 227-235 ◽

Cited By ~ 2

Author(s):

Bassam I. Shamoun ◽

Robert J. Witt

Keyword(s):

Parametric Study ◽

Boiling Water ◽

Severe Accident ◽

Boiling Water Reactor ◽

Water Reactor

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

Volume 2: Plant Systems, Construction, Structures and Components; Next Generation Reactors and Advanced Reactors ◽

10.1115/icone21-16681 ◽

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.

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