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Modelling and Simulation of Reactor Pressure Vessel Failure during Severe Accidents
KTH, School of Engineering Sciences (SCI), Physics, Nuclear Power Safety.ORCID iD: 0000-0003-0408-8807
2020 (English)Doctoral thesis, comprehensive summary (Other academic)
Abstract [en]

This thesis aims at the development of new coupling approaches and new models for the thermo-fluid-structure coupling problem of reactor pressure vessel (RPV) failure during severe accidents and related physical phenomena. The thesis work consists of five parts: (i) development of a three-stage creep model for RPV steel 16MND5, (ii) development of a thermo-fluid-structure coupling approach for RPV failure analysis, (iii) performance comparison of the new approach that uses volume loads mapping (VLM) for data transfer with the previous approach that uses surface loads mapping (SLM), (iv) development of a lumped-parameter code for quick estimate of transient melt pool heat transfer, and (v) development of a hybrid coupling approach for efficient analysis of RPV failure.

A creep model called ‘modified theta projection model’ was developed for the 16MND5 steel so that it covers three-stage creep process. Creep curves are expressed as a function of time with five parameters  (i=1~4 and m) in the new creep model. A dataset for the model parameters was constructed based on the available experimental creep curves, given the monotonicity assumption of creep strain vs temperature and stress. New creep curves can be predicted by interpolating model parameters from this dataset, in contrast to the previous method that employs an extra fitting process. The new treatment better accommodates all the experimental curves over the wide ranges of temperature and stress loads. The model was implemented into the ANSYS Mechanical code, and its predictions successfully captured all three creep stages and a good agreement was achieved between the experimental and predicted creep curves. For dynamic loads that change with time, the widely used time hardening and strain hardening models were implemented with a reasonable performance. These properties fulfil the requirements of a creep model for structural analysis.

A thermo-fluid-structure coupling approach was developed by coupling the ANSYS Fluent for the fluid dynamics of melt pool heat transfer and ANSYS Structural for structural mechanics of RPV. An extension tool was introduced to realize transient load transfer from ANSYS Fluent to Structural and minimize the user effort. Both CFD with turbulence models and the effective model PECM can be employed for predicting melt pool heat transfer. The modified theta projection model was used for creep analysis of the RPV. The coupling approach does not only capture the transient thermo-fluid-structure interaction feature, but also support the advanced models in both melt pool convection and structural mechanics to improve fidelity and facilitate implementation. The coupling approach performs well in the validation against the FOREVER-EC2 experiment, and can be applied complex geometries, such as a BWR lower head with forest of penetrations (control rod guide tubes and instrument guide tubes).

In the comparative analysis, the VLM and SLM coupling approaches generally have the similar performance, in terms of their predictability of the FOREVER-EC2 experiment and applicability to the reactor case. Though the SLM approach predicted slightly earlier failure times than VLM in both cases, the difference was negligible compared to the large scale of vessel failure time (~  s). The VLM approach showed higher computational efficiency than the SLM.

The idea of the hybrid coupling is to employ a lumped-parameter code for quick estimate of thermal load which can be employed in detailed structural analysis. Such a coupling approach can significantly increase the calculation efficiency which is important to the case of a prototypical RPV where mechanistic simulation of melt pool convection is computationally expensive and unnecessary. The transIVR code was developed for this purpose, which is not only capable of quick estimate of transient heat transfer of one- and two- layer melt pool, but also solving heat conduction problem in the RPV wall with 2D finite difference method to provide spatial thermal details for RPV structural analysis. The capabilities of transIVR in modelling two-layer pool heat transfer and transient pool heat transfer were demonstrated by calculations against the UCSB FIBS benchmark case and the LIVE-7V experiment, respectively. The transIVR code was then coupled to the mechanical solver ANSYS Mechanical for detailed RPV failure analysis. Validation against the FOREVER-EC2 experiment indicates the coupling framework successfully captured the vessel creep failure characteristics.

Abstract [sv]

Syftet med denna avhandling är att utveckla nya metoder och nya modeller för att kunna koppla ihop termohydrauliska förlopp och uppkommen strukturrespons i botten på reaktortanken för förbättrad analys av reaktortankbrott under ett svårt haveri. Avhandlingen består av fem delar: (i) utveckling av en trestegsmodell för krypning i reaktortankens 16MND5-stål, (ii) utveckling av metoder för kopp ling av termohydrauliska förlopp och strukturrespons för studier av reaktortankbrott, (iii) jämförelse av prestanda av den nya metoden, som använder volume loads mapping (VLM) för dataöverföring, med den tidigare metoden, som använder surface loads mapping (SLM), (iv) utveckling av en lumped-parameter beräkningskod för snabb uppskattning av transient värmeöverföring i en smältpöl i botten på reaktortanken, och (v) utveckling av en hybrid kopplingsmetod för effektiv analys av reaktortankbrott.En krypmodell, kallad ‘modified theta projection model’, som behandlar krypprocessen i tre steg, utvecklades för 16MND5-stålet. Krypkurvor i den nya krypmodellen ges som en funktion av tid med fem parametrar 𝜃𝑖 (i=14 och m). Ett dataset för modellparametrar togs fram baserat på tillgängliga experimentalla krypkurvor, under anatagande om att kryptöjning är en monoton funktion av temperatur och spänning. Nya krypkurvor kan genereras genom att interpolera modellparametrar från detta dataset, i kontrast till den tidigare metoden som använder en speciell anpassningsprocess. Fördelen med det nya tillvägagångssättet är att den hanterar alla experimentella krypkurvor för ett brett spektrum av temperatur- och spänningsbelastningar. Modellen implementerades i ANSYS-koden och den korrekt behandlar alla tre krypsteg samt visar bra överenstämmelse mellan experimentella och beräknade krypkurvor. För analys av tidsberoende dynamiska belastningar implementerades välkända modeller av tids- och deformationshärdning med rimliga resultat. Dessa egenskaper uppfyller kravet på en krypmodell för strukturmekaniska analyser.En metod utvecklades för koppling mellan termohydrauliska förlopp och strukturmekanisk respons genom att koppla ihop ANSYS Fluent, som analyserar fluiddynamiska aspekter av värmeöverföring i en smältpöl, med ANSYS Structural, som analyserar strukturmekanisk respons av reaktortanksväggen. Ett speciellt verktyg togs fram för att möjliggöra överföring av transienta belastningar från ANSYS Fluent till ANSYS Structural och för att minimera användarinsatsen. Både CFD med turbulensmodeller och den effektiva PECM-modellen kan användas för att analysera värmeöverföring i en smältpöl. För krypanalysen av reaktortanken har ‘modified theta projection model’ använts. Den framtagna kopplingsmetoden har fördelen att den inte bara fångar de transienta aspekterna av interaktionen mellan termohydraulik och strukturmekanik men också stödjer avancerade modeller av konvektion i en smältpöl och strukturrespons och därigenom bidrar till förbättrad precision och underlättar implementeringen av modeller. Kopplingsmetoden fungerar bra när den valideras mot FOREVER-EC2 experimentet, och den kan användas i komplexa geometrier, som till exempel i botten på reaktortanken i en kokvattenreaktor (BWR), med en skog av genomföringar (ledrör för styrstavar och instrumentering).Jämförande beräkningar visar att de två kopplingsmetoderna, VLM-baserad och SLM-baserad, presterar lika bra, när det gäller hur de predikterar resultat av FOREVER-EC2 experimentet och deras lämplighet för reaktoranalyser. Även om SLM-metoden predikterar att tankbrott i båda fallen inträffar lite tidigare så är skillnaden mellan metoderna försumbar i jämförelse med tiden till tankbrott (~104 s). VLM-metoden var bättre än SLM-metoden när det gäller beräkningseffektiviteten.Idén om hybridkoppling går ut på att använda en lumped-parameter kod för snabb uppskattning av termiska belastningar, som senare kan användas i detaljerad strukturmekanisk analys. En sådan kopplingsmetod kan på ett signifikant sätt öka effektiviteten av beräkningar vilket är viktig i reaktorapplikationer där mekanistisk simulering av konvektion i en smältpöl är beräkningsintensiv och onödig. I detta syfte utvecklades koden transIVR. Koden kan dels ge en snabb uppskattning av transient värmeöverföring i en smältpöl, som består av en eller två materialskikt, och dels lösa värmeledningsproblem i reaktortanksväggen med 2D finita differensmetoden vilket genererar detaljerad information om termiska förhållanden som behövs för strukturmekaniska analyser. För att demonstrera att transIVR kan analysera dels värmeöverföring i en smältpöl bestående av två materialskikt och dels transient värmeöverföring i en smältpöl jämfördes transIVR prediktioner med UCSB FIBS-benchmarkfallet respektive LIVE-7V-experimentet. I nästa steg kopplades transIVR-koden till ANSYS Structural för att möjliggöra mer detaljerad analys av reaktortankbrott. Validering mot FOREVER-EC2 experimentet tyder på att den utvecklade kopplingsmetodiken korrekt återger tankbrott orsakat av krypning.Nyckelord: reaktortankbrott, koppling termohydraulik-strukturmekanik, modellering av krypning, värmeöverföring i smältpöl, svåra haverier, CFD, FEA, transIVR.

Place, publisher, year, edition, pages
Sweden: KTH Royal Institute of Technology, 2020. , p. 53
Series
TRITA-SCI-FOU ; 2020:13
Keywords [en]
RPV failure, thermo-fluid-structure coupling, creep modelling, melt pool heat transfer, severe accident, CFD, FEA, transIVR
National Category
Energy Engineering
Identifiers
URN: urn:nbn:se:kth:diva-273366ISBN: 978-91-7873-540-2 (print)OAI: oai:DiVA.org:kth-273366DiVA, id: diva2:1430414
Public defence
2020-06-12, Via Zoom: https://kth-se.zoom.us/webinar/register/WN_TwfxbuXLT9iekTlDv4ZUWg, Du som saknar dator/datorvana kan kontakta weimin@kth.se för information., Stockholm, 09:30 (English)
Opponent
Supervisors
Available from: 2020-05-19 Created: 2020-05-14 Last updated: 2020-06-05Bibliographically approved
List of papers
1. A modified theta projection model for creep behavior of RPV steel 16MND5
Open this publication in new window or tab >>A modified theta projection model for creep behavior of RPV steel 16MND5
2020 (English)In: Journal of Materials Science & Technology, ISSN 1005-0302, Vol. 47, p. 231-242Article in journal (Refereed) Published
Abstract [en]

During a hypothetical severe accident of light water reactors, the reactor pressure vessel (RPV) could fail due to its creep under the influence of high-temperature corium. Hence, modelling of creep behavior of the RPV is paramount to reactor safety analysis since it predicts the transition point of accident progression from in-vessel to ex-vessel phase. In the present study we proposed a new creep model for the classical French RPV steel 16MND5, which is adapted from the “theta-projection model” and contains all three stages of a creep process. Creep curves are expressed as a function of time with five model parameters θi(i=1−4  and  m). A model parameter dataset was constructed by fitting experimental creep curves into this function. To correlate the creep curves for different temperatures and stress loads, we directly interpolate the model’s parameters θi(i=1−4  and m) from this dataset, in contrast to the conventional “theta-projection model” which employs an extra single correlation for each θi(i=1−4 and m), to better accommodate all experimental curves over the wide ranges of temperature and stress loads. We also put a constraint on the trend of the creep strain that it would monotonically increase with temperature and stress load. A good agreement was achieved between each experimental creep curve and corresponding model’s prediction. The widely used time-hardening and strain-hardening models were performing reasonably well in the new method.

Place, publisher, year, edition, pages
Elsevier, 2020
Keywords
16MND5 steel, Creep modelling, Tertiary stage, Reactor pressure vessel, Theta projection model
National Category
Mechanical Engineering Metallurgy and Metallic Materials
Identifiers
urn:nbn:se:kth:diva-273363 (URN)10.1016/j.jmst.2020.02.016 (DOI)2-s2.0-85081013789 (Scopus ID)
Note

QC 20200515

Available from: 2020-05-14 Created: 2020-05-14 Last updated: 2020-05-25Bibliographically approved
2. Validation of a thermo-fluid-structure coupling approach for RPV creep failure analysis against FOREVER-EC2 experiment
Open this publication in new window or tab >>Validation of a thermo-fluid-structure coupling approach for RPV creep failure analysis against FOREVER-EC2 experiment
Show others...
2019 (English)In: Annals of Nuclear Energy, ISSN 0306-4549, E-ISSN 1873-2100, Vol. 133, p. 637-648Article in journal (Refereed) Published
Abstract [en]

The failure of reactor pressure vessel (RPV) during a severe accident of light water reactors is a thermal fluid-structure interaction (FSI) problem which involves melt pool heat transfer and creep deformation of the RPV. The present study is intended to explore a reliable coupling approach of thermo-fluid-structure analyses which will not only be able to reflect the transient thermal FSI feature, but also apply the advanced models and computational platforms to melt pool convection and structural mechanics, so as to improve simulation fidelity. For this purpose, the multi-physics platform of ANSYS encompassing Fluent and Structural capabilities was employed to simulate the fluid dynamics and structural mechanics in a coupled manner. In particular, the FOREVER-EC2 experiment was chosen to validate the coupling approach. The natural convection in melt pool was modeled with the SST turbulence model with a well-resolved boundary layer, while the creep deformation for the vessel made of 16MND5 steel was analyzed with a new three-stage creep model (modified theta projection model). A utility tool was introduced to transfer the transient thermal loads from Fluent to Structural which minimizes the user effort in performing the coupled analysis. The validation work demonstrated the well-posed capability of the coupling approach for prediction of the key parameters of interest, including temperature profile, total displacement of vessel bottom point and the evolution of wall thickness profile in the experiment. Ltd. All rights reserved.

Place, publisher, year, edition, pages
PERGAMON-ELSEVIER SCIENCE LTD, 2019
Keywords
Reactor pressure vessel, Creep failure, Thermal fluid-structure interaction, Computational fluid dynamics, Computational structural mechanics, Coupled analysis
National Category
Energy Engineering
Research subject
Energy Technology; Energy Technology
Identifiers
urn:nbn:se:kth:diva-260983 (URN)10.1016/j.anucene.2019.06.067 (DOI)000484649800061 ()2-s2.0-85068784394 (Scopus ID)
Note

QC 20191010

Available from: 2019-10-10 Created: 2019-10-10 Last updated: 2020-05-14Bibliographically approved
3. Comparative Analysis of Reactor Pressure Vessel Failure using Two Thermo-Fluid-Structure Coupling Approaches
Open this publication in new window or tab >>Comparative Analysis of Reactor Pressure Vessel Failure using Two Thermo-Fluid-Structure Coupling Approaches
(English)Manuscript (preprint) (Other academic)
Abstract [en]

Reactor pressure vessel (RPV) failure analysis is a thermo-fluid-structure coupled probleminvolving the molten pool heat transfer and RPV structural failure. It is of great importance tothe qualification of severe accident mitigation strategies as well as the general assessment ofsevere accident risk in light water reactors. Two coupling approaches, i.e. volume loadsmapping (VLM) and surface loads mapping (SLM), are widely employed to resolve thisproblem. The present study was performed to compare the performance of the two couplingapproaches by simulating vessel failures in the FOREVER-EC2 Experiment and a postulatedsevere accident of a reference Boiling Water Reactor (BWR). Both simulations of theexperiment using the two different approaches showed good agreements with experimental datain terms of total deformation in vertical direction of bottom point and wall thickness changes.The spatial distributions of creep strain and total deformation were also similar between bothsimulations. For the reference BWR case, good agreements were achieved between the couplingapproaches in predicting the maximum creep strain and the total deformation in verticaldirection of bottom point. The deformation rate of the vessel wall was slow at the early stage,but increased with time, and ultimately accelerated, which resulted in a drastic deformation andfinal vessel failure. Similar spatial distributions of creep strain and total deformation were alsopredicted by both simulations. Although the SLM approach predicted the deformation beganslightly earlier than what the VLM approach did, the difference between the failure timespredicted by the two approaches were negligible compared to the large scale of vessel failuretime (~10^4s). Generally speaking, the VLM and SLM coupling approaches have the similarperformance, in terms of their predictability of experiment and applicability to reactor case,though the VLM approach was computationally more efficient.

Keywords
reactor pressure vessel, vessel integrity, creep failure, thermo-fluid-structure interaction, coupled analysis
National Category
Energy Engineering Applied Mechanics
Identifiers
urn:nbn:se:kth:diva-273364 (URN)
Note

QC 20200529

Available from: 2020-05-14 Created: 2020-05-14 Last updated: 2020-05-29Bibliographically approved
4. Development of a Lumped-Parameter Code for Efficient Assessment of In-Vessel Melt Retention Strategy of LWRs
Open this publication in new window or tab >>Development of a Lumped-Parameter Code for Efficient Assessment of In-Vessel Melt Retention Strategy of LWRs
(English)Manuscript (preprint) (Other academic)
Abstract [en]

Motivated by efficient assessment of an in-vessel melt retention (IVR) strategy employed insome designs of light water reactors (LWRs), a lumped-parameter code named as transIVR wasdeveloped in the present study, which has the features as (i) quick estimates of transient meltpool heat transfer with reasonable accuracies; and (ii) two-dimensional representation of heatconduction in the RPV wall and crust surrounding the melt pool, facilitating the coupledthermo-mechanical analysis of the reactor pressure vessel (RPV) under thermal loads. ThetransIVR code was first benchmarked against the UCSB FIBS case of two-layer configurationand then validated against the LIVE-7V experiment; both showed the code’s capability inpredicting transient and two-layer melt pool heat transfers. Finally, the code was coupled withANSYS Mechanical to simulate the RPV failure experiment FOREVER-EC2, and the resultsshowed acceptable agreements with those of both the previous CFD approach and theexperiment. Hence, the transIVR code is an efficient tool suitable to address related safetyconcerns of LWRs, including IVR efficacy and vessel integrity .

Keywords
Melt pool heat transfer, RPV failure, coupled analysis, severe accident, IVR
National Category
Energy Engineering
Identifiers
urn:nbn:se:kth:diva-273365 (URN)
Note

QC 20200529

Available from: 2020-05-14 Created: 2020-05-14 Last updated: 2020-05-29Bibliographically approved

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