KOURA project

SAFIR2014 mid-term seminar 21.3.2013 Elina Syrjälahti VTT Technical Research Centre of Finland

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Contents

Overview of the KOURA project 3D thermal hydraulics Reactor dynamics of boiling water reactors Future development

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Overview of KOURA project

KOURA (Kolmiulotteiset Reaktorianalyysit) is a four-year project Extent approx. 2 person-years per year 9 researcher in 2012

Fundamental objective is to have a truly independent transient

calculation system which can be utilized by the safety authority and other end-users for safety analyses Two main objectives in 2011-2014

To supplement VTT’s code system with three-dimensional thermal hydraulics modelling To enhance VTT’s boiling water reactor modelling capabilities

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3D thermal hydraulics: PORFLO

PORFLO is 3D thermal hydraulics code, developed at VTT Designed to analyse multiphase flow problems related to NPP

safety analysis VTT’s own code can be coupled to other codes, e.g. to reactor

dynamics codes Mainly targeted at applications where 3D phenomena are

significant Code utilizes the concept of porous medium to model structural

features that are not represented explicitly in a computational mesh Fuel bundles Internal structures of reactor pressure vessel (RPV)

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3D thermal hydraulics: new unstructured PORFLO

PORFLO was rewritten in 2011 Main improvements:

Ability to handle unstructured meshes Conservation laws integrated on unstructured

meshes (hexahedra, prisms, pyramids, tetrahedral)

No staggered grid -> velocity calculated at cell centres

Grid can be fitted to structures that can be represented with affordable number of cells, e.g. cylindrical walls of RPV

Parallelization Utilization of general data format (CGNS)

PORFLO can read same mesh as other CFD codes

EPR RPV mesh

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PORFLO EPR RPV model

Results with structured mesh and earlier PORFLO version presented in SAFIR2010 final seminar New model: Body fitted-mesh with about

250 000 cells, quarter of RPV Based on publicly available data Single-phase stationary simulation results

are expectable

T (°C)

Temperature on the vertical planes defined by the inlet and outlet nozzles. New PORFLO, unstructured mesh.

296 °C

350°C

Temperature with the structured mesh and the old PORFLO code

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Reactor dynamics of boiling water reactors

Main objective is the enhancement of VTT’s boiling water reactor capabilities Further development of BWR features of internally coupled

TRAB3D/SMABRE and validation of the code Increase knowledge of the dynamics of the boiling water

reactors Training of reactor dynamics specialists on using TRAB3D for

BWR analyses

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Reactor dynamics of BWR: Internal coupling of TRAB3D/SMABRE

TRAB3D models neutronics of reactor core with 3D nodal model System code SMABRE models all hydraulics Heat transfer can be calculated with either code TRAB3D includes also models for rest of the BWR RPV

thermal hydraulics using one-dimensional channels has been used also alone for BWR analysis Internal coupling

Extends modelling capabilities to transients with reversed flow More versatile for modelling of different type of reactors

Development and validation of BWR features has been restarted during SAFIR2014, during SAFIR2010 focus was on PWR reactors

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Reactor dynamics of BWR: Internal coupling of TRAB3D/SMABRE

Validation of the code Olkiluoto 1

overpressurization transient (measured data) Separate disturbances of

the Olkiluoto 1 load rejection test , whole transient in 2013 (measured data) PWR features

EPR control rod ejection

EPR pump seizure

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Validation of internal coupling of TRAB3D/SMABRE

Results of the overpressurization transient are very promising Steam line dynamics has not yet been implemented in the SMABRE

model

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Validation of internal coupling of TRAB3D/SMABRE

Calculation of the overpressurization transient revealed that radial fuel rod model had been erroneously implemented from TRAB3D to SMABRE

Results shown in this presentation calculated with simpler average fuel rod model Localization of the problem with fuel rod

model improved results also in other cases, as an example EPR control rod ejection test

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Reactor dynamics of BWR: OECD/NEA benchmark

BWR stability benchmark was launched in 2011 Based on the stability event at the Oskarshamn-2 in February 1999

In 2012, TRAB-3D model for Oskarshamn-2 plant has been

constructed

Creation of a new model from scratch

is laborious, but a very useful lesson in modelling of BWRs Test calculations have been done, but

actual stability event will be calculated in 2013 due to the schedule of the benchmark Stand-alone TRAB3D

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Reactor dynamics of BWR: OECD/NEA benchmark

Geometry of TRAB3D circuit model for Oskarshamn-2 Results of the test calculations

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Plans for the near future

PORFLO will be further developed, verified and validated In 2013 focus on fuel assembly level and closure laws

Reactor dynamics codes will be improved Defects found in TRAB3D/SMABRE validation calculations Development of the pin power reconstruction model has

already started within the KOURA project Modelling of the fuel heat transfer models will be improved by

utilizing results of the PALAMA project Initialization of fuel rod at burn-up FINIX fuel behaviour model

In future, PORFLO will be coupled with reactor dynamics codes

In 2013 only simple data transfer in form of files

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Summary

KOURA is a four-year project Two main areas

3D thermal hydraulics: focus on the development and applications of the PORFLO code Reactor dynamics: focus on the BWR modelling

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VTT creates business from technology