Elaboration of a Phenomena Identification Ranking Table (PIRT) for the modelling of In-Vessel Retention

doi:10.1016/j.anucene.2020.107617

Annals of Nuclear Energy

Volume 146, October 2020, 107617

https://doi.org/10.1016/j.anucene.2020.107617 Get rights and content

Highlights

•
Identification of important phenomena for IVR modelling.
•
All major severe accident codes included.
•
Uncertainty evaluation from sensitivity studies.
•
Importance of transient stratification and heat transfer.

Abstract

The strategy for In-Vessel Retention (IVR) of corium is currently considered for several new reactor designs in various countries. One of the issues for the demonstration of the success of this strategy is that there are significant uncertainties in physical modelling of corium in the lower plenum and its transient chemical and thermal interactions with the vessel leading to its significant ablation. Since the initial approaches developed in the nineties for the AP600 and Loviisa VVER-440 plant, knowledge about corium pools (thermochemistry and heat transfer characteristics) and mechanical behaviour of the vessel has improved and it is now possible to model more accurately this phenomenon. The H2020 European project IVMR (In-Vessel Melt Retention) addresses the issue of selecting and improving IVR models for safety evaluation. As a first step, a Phenomena Identification Ranking Table (PIRT) involving the relevant physical processes was developed.

The methodology followed to build this PIRT results from the consideration of a few principles: (a) Identifying and separating the risks with respect to which the importance of a physical process is evaluated. (b) Defining physical processes or parameters which can be considered as independent of the other ones. (c) Avoiding expert judgement, as much as possible, and using, instead, the results of previous sensitivity studies to estimate the impact of each physical process or parameter.

In order to obtain results representing the currently “shared” state of knowledge, code developers of the most widely used codes were consulted. It includes ASTEC, MELCOR, MAAP (EDF version), SOCRAT, ATHLET-CD integral codes and PROCOR, SIMPLE, HEFEST_URAN, and IVRSYS dedicated codes (specific to IVR calculations).

One of the positive outcomes of this PIRT is that the results show significant tendencies, with distinct groups of phenomena or parameters. This allows the identification of the uncertainties and of the phenomena/variables with the highest (or lowest) importance. Some dispersion in the results could also be noticed but it can be understood either because of the variability due to reactor design or the impact of models chosen by experts in the codes they have developed. When the dispersion cannot be explained, it indicates that the phenomenon is poorly understood and may deserve further research. Among the phenomena with highest importance, the heat transfers in the top metal layer and the chemical/thermal interaction with the oxide crust have been identified, as well as the transient formation and stratification of metal and oxide layers, including thermochemical peculiarities of the (U,Zr,Fe,O) system. Another phenomenon of highest importance is the mechanical behaviour of the thin ablated vessel wall, where elasticity, plasticity and creep all play a role. It is concluded that codes should include improved models for those phenomena in order to be able to provide a reasonable assessment of potential for IVR.

Introduction

A Phenomena Identification Ranking Table (PIRT) on in-vessel retention phenomenology was developed as part of the European IVMR project within the H2020 framework. Its first objective was to identify the main relevant physical phenomena, in order to give a priority ranking of development needs for codes within the work-package WP2.2 dedicated to modelling. Soon after the start of this action, it was decided to extend the scope of this PIRT to experts outside of the IVMR project in order to gather more technical opinions and obtain a consolidated evaluation in the end. The Korean institute KAERI was involved, as well as Russian organizations (IBRAE and Kurtchatov Institute) and the U.S. NRC and the Sandia National Laboratory (MELCOR code developers) due to their expertise in IVR phenomena. It should be noted that it is the first international effort dedicated to the identification of the most important models and parameters for IVR assessment. A more general PIRT was made previously during the EURSAFE project (Magallon et al., 2005) but it was not specific to IVR and not really intended as a list of model requirements, contrary to the present PIRT. We may also mention the work performed in the ASAMPSA2 project (Raimond et al., 2013), where the IVR issue was treated with identification of phenomena needing to be addressed (but without ranking).

At first, a specific methodology was proposed for this PIRT, with the main objective of being able to identify, as simply as possible, the main processes and variables, used in the safety demonstration, which have the greatest impact on the evaluation of risks and for which the uncertainties are significant, considering a severe accident (SA) scenario for a given design. The further identification of more detailed physical phenomena or properties with high uncertainties and impact is proposed in a second step.

In addition to the first goal of code development focused on integration and implementation of models having the most significant impact on risk evaluation and the identification of those with limited impact, other benefits are also expected from this PIRT. The identification of phenomena for which complementary tests are needed will be very important in the design of new experiments. For safety analysts, this PIRT can also provide a synthesis of the main phenomena and uncertainties to be considered in the IVR evaluation.

It is important to keep in mind that the evaluated “knowledge” corresponds to what is implemented in the models of widely used codes, therefore it is what we may call the “common knowledge”: in this sense, the paper is also a critical review of this “common knowledge”. The idea is not to identify what should be included for the most detailed modelling of in-vessel retention but what the essential and necessary models are for a severe accident code, considering the degree of accuracy commonly associated with such codes. Of course, some simplifications or assumptions are made when using severe accident codes but it is important to realize that such codes have been the basis and are still the basis for evaluations of severe accidents consequences and mitigation strategies. Therefore, the identification of important models for IVR, using severe accident codes, remains crucial.

This paper intends to present the main results and outcomes of this work. First, the methodology used to develop this PIRT on IVR phenomena is described. Then, the main results are presented and discussed.

Section snippets

Methodology

The methodology for the development of the IVR PIRT results from the consideration of a few principles:

• Identifying and separating the risks with respect to which the importance of a physical process is evaluated;

• Defining physical processes or parameters which can be considered as independent of the other ones;

• Avoiding expert judgement, as much as possible, and using, instead, the results of previous analytical and experimental analyses including sensitivity studies to estimate the impact

Evaluations of uncertainties and impacts for the macro-variables

The uncertainties have been evaluated by each participant for each macro-variable. The results presented in Fig. 2 are the average values obtained. The macro-variables are grouped depending on what they are related to, i.e., the corium pool, the External Reactor Vessel Cooling (ERVC), the vessel wall or the Steam Explosion (SE) in case of vessel failure. The level of 50% is arbitrarily marked with the red line at the top of Fig. 2 and macro-variables above that level are identified as having

Ranking of parameters impacting the risks of excessive heat flux and ablation under transient situations

The detailed ranking of the parameters, which impact the risks of excessive heat flux in stabilized or in transient situations (risks 2 and 3) is presented in Table 3. The parameters are listed only when their overall impact calculated using Eq. (3) is above 10%.

Since uncertainties have been evaluated first for the macro-variables only, it is useful at this stage to distinguish in the list of significant phenomena:

•
The ones associated with high uncertainties, which correspond to the main issues

Risk of vessel failure

The risk of vessel failure for a given ablated profile of the vessel wall was also investigated. Results are presented in Fig. 5 and based on the following references mentioned by the participants Chu et al., 1999, Fichot et al., 2018, Filippov et al., 2016, Lichachev et al., 1994, Mao et al., 2015, Mao et al., 2016, and Rempe et al. (1993). Regarding this risk, all the parameters appear with non-negligible impact and remain within 10–20%. This indicates that it was not enough investigated and

Conclusions

A PIRT on IVR phenomena was developed, based on the evaluations performed by the developers of the main SA codes and selected experts of IVR. It is the first time that such work is performed and the outcomes allow the identification of the main uncertainties and of the important phenomena or parameters and physical properties for the modelling of IVR. The phenomena or parameters with negligible impact were also identified.

The most important identified needs are:

•
New models should be included in

CRediT authorship contribution statement

F. Fichot: Conceptualization, Methodology, Investigation, Formal analysis, Supervision, Writing - original draft. L. Carénini: Conceptualization, Methodology, Investigation, Formal analysis, Resources, Software, Supervision, Writing - original draft. N. Bakouta: Investigation, Formal analysis, Resources, Software, Writing - review & editing. H. Esmaili: Investigation, Formal analysis, Resources, Software, Writing - review & editing. L. Humphries: Investigation, Formal analysis, Resources,

Declaration of Competing Interest

The authors declare that they have no known competing financial interests or personal relationships that could have appeared to influence the work reported in this paper.

Acknowledgments

The authors wish to acknowledge the contributions of R. Lee, M. Salay from US-NRC and R. Gauntt from SNL.

This work was partly funded by the European Project H2020 IVMR No. 662157.

References (34)

V.I. Almjashev et al.
Experimental study of transient phenomena in the three-liquid oxidic-metallic corium pool
Nucl. Eng. Des.
(2018)
S.V. Bechta et al.
VVER vessel steel corrosion at interaction with molten corium in oxidizing atmosphere
Nucl. Eng. Des.
(2009)
L. Carénini et al.
Modeling issues related to molten pool behavior in case of In-Vessel Retention Strategy
Ann. Nucl. Energy
(2018)
R. Le Tellier et al.
Transient stratification modelling of a corium pool in a LWR vessel lower head
Nucl. Eng. Des.
(2015)
D. Magallon et al.
European expert network for the reduction of uncertainties in severe accident safety issues (EURSAFE)
Nucl. Eng. Des.
(2005)
J.F. Mao et al.
Creep and damage analysis of reactor pressure vessel considering core meltdown scenario
Procedia Eng.
(2015)
J.F. Mao et al.
Creep deformation and damage behavior of reactor pressure vessel under core meltdown scenario
Int. J. Press. Vessels Pip.
(2016)
Ozrin
Distribution of fission product residual decay heat in stratified core melt of LWR and its influence on sidewall heat flux
Nucl. Eng. Des.
(2013)
J.M. Seiler et al.
Consequences of material effects on in-vessel retention
Nucl. Eng. Des.
(2007)
Belon S., Bouillet C., Bonneville H., Andrews N., Faucett C., Insight of Core Degradation Simulation in Integral Codes...

Bonnet J.M., Seiler J.M., Thermal hydraulic phenomena in corium pool: the BALI experiment. In: Proceedings of ICONE7...

Bonnet, J.M., Seiler, J.M., 2000. In-vessel corium pool thermalhydraulics for bounding cases. In: Proc. of salt test...

Carénini, L., Fichot, F., Bakouta, N., Filippov, A., Le Tellier, R., Viot, L., Melnikov, I., Pandazis, P., Main...

Carénini, L., Fichot, F., 2016. The Impact of Transient Behavior of Corium in the Lower Head of a Reactor Vessel for...

J.-S. Cho et al.

Enhanced natural convection in a metal layer cooled by boiling water

Nucl. Technol.

(2004)

Chu, T.Y., Pilch, M.M., Bentz, J.H., Ludwigsen, J.S., Lu, W.-Y., Humpries, L.L., 1999. Lower Head Failure Experiment...

Esmaili, H., Khatib-Rahbar, M., Analysis of In-Vessel Retention and Ex-Vessel Fuel Coolant Interaction for AP1000, U.S....

Cited by (7)

Achievements of recent research on severe accidents at cea/irense in support of future nuclear fission technology
2024, Nuclear Engineering and Design
In the current context of nuclear revival under the constraint of global warming and an ever-increasing world demand for energy, it is essential to take all precautions in the event of a serious accident during the design phase and to review the severe accident mitigation features of existing plants. The current nuclear revival is accompanied by the development of prototypes based on alternative concepts to light water reactors, such as the Sodium-cooled Fast Reactor (SFR), High Temperature Reactor (HTR), Molten Salt Reactor (MSR), which are also under consideration as reactors of reduced power output, or Small Modular Reactors (SMR). This paper summarizes the main outcomes and achievements of the last five years of experimental and numerical research conducted at CEA/IRESNE into prospective severe accidents within Pressurized Water Reactors (PWR) and SFRs.
Development and validation of the MAVR-TA code for analyzing the release and transport of fission products during a severe accident at a VVER NPP. Part 3 – Modelling of the release of fission products from the melt
2023, Nuclear Engineering and Design
The MAVR-TA code is designed for modelling of fission product release from fuel and melt, as well as fission products transport in the primary circuit and in the containment during a severe accident at a NPP with VVER. The third part of the work provides information on the models of the MAVR-TA code related to modelling of fission products release from the melt due to evaporation from the melt surface and due to molten core concrete interaction. The paper describes the models and approximations implemented in the code, as well as experimental tests used to validate the models of fission product release in the MAVR-TA code. The results obtained during validation are described and analyzed.
Three-dimensional numerical simulation of droplet formation by Rayleigh–Taylor instability in multiphase corium
2021, Nuclear Engineering and Design
Citation Excerpt :
This whole regime, called the stratification transitory regime, must be finely modeled because the main risk for the IVR comes from the “focusing effect”, i.e., the concentration of the heat flux at the lateral surface of the top metallic layer (Le Tellier et al., 2015b). As discussed in Fichot et al. (2020) and further illustrated in Carénini et al. (2020) by code-to-code benchmarks, large uncertainties still exist in the quantitative assessment of such stratification transients and they have a strong impact on the overall evaluation of vessel wall melt-through under IVR conditions. In order to reduce this uncertainty, the few existing integral models that describe stratification transients in severe accident codes are in need for validation or parameter calibration.
During a severe accident in a nuclear reactor, the melting of the core may lead to the formation of a multiphase liquid pool (corium) in the vessel lower head. The heat transfer at the boundary with the vessel is affected by diffusive and convective mass fluxes. In particular, the development of Rayleigh–Taylor instabilities influence the thickness of the top metallic layer and therefore the “focusing effect” of the heat flux, which is the main risk for the vessel integrity. We use a Cahn–Hilliard pseudo-binary model to describe the uranium/oxygen/zirconium/iron mixture. The diffusion and the convection are governed by the Cahn–Hilliard equation and the Navier–Stokes equations under the Boussinesq approximation. In this work, the model is isothermal and the buoyancy force is only due to the gradient of chemical composition. The model is solved in three dimensions with a pseudo-spectral code. The initial configuration consists of a light layer of iron-rich fluid above a heavy layer of uranium/oxygen/zirconium mixture. A thin layer of heavier metallic phase lays at the interface and eventually triggers a Rayleigh–Taylor instability. The metallic phase forms a plume which falls downward and then breaks up into droplets due to the Rayleigh-Plateau instability. The phenomenon is alimented by diffusion which generates the heavy metallic phase at the interface. The droplet formation observed in an experiment of corium stratification transient from the literature is qualitatively captured. The mobility, the viscosity and the surface tension are shown to have an influence on the mass transfer.
A two-phase mathematical model to describe the dissolution of corium crust by molten steel
2021, Nuclear Engineering and Design
This paper deals with the study of dissolution phenomena in corium crust, in a severe accident. In case of an In-Vessel Retention (IVR) strategy, it is expected that the corium pool will be surrounded by an oxide crust, which will be in contact with molten steel from the top, and along the sides. CORDEB experiments have shown that this crust becomes permeable, having an impact on the thickness of molten steel layer above it. A thin molten steel layer atop crust may lead to an excessive heat flux to the Reactor Pressure Vessel (RPV), resulting in a rupture or melt-through. Corium being a mixture of ( $U, Zr, O$ ) atoms, undergoes a quaternary interaction with Fe atoms. Such an interaction plays an important role in the stability (or not) of the crust. A thermochemical study of this quaternary system shows that it can be reduced to a pseudo-binary in each phase: $(U, Zr) + O$ in the oxide phase and $(U, Zr) + Fe$ in the metal phase. Therefore, in this paper, dissolution of a binary two-phase porous region by liquid, is studied by means of an up-scaled mathematical model. The model is derived by volume averaging transport equations – Mass, Momentum and Specie – over a Representative Elementary Volume (REV). Final system of Partial Differential Equations (PDEs) is closed by deriving empirical relations for the effective diffusivities and for the mass transfer coefficients. Model is solved numerically to study the progress of dissolution, facilitating, molten steel filtration through the two-phase region. A qualitative comparison has also been made with CORDEB experiment. Results without convection show that the model is able to demonstrate the creation of liquid metal channels similar to the ones observed in the CORDEB findings. Further, the result with pressure gradient driven convection shows the possibility of complete crust dissolution, which is due to the convection enhanced dissolution in the corium crust.
ASTECV2.2b and SUNSET uncertainty and sensitivity analyses on IVMR WWER1000 Test Case
2023, Proceedings of Science
High-Temperature Characterization of Melted Nuclear Core Materials: Investigating Corium Properties Through the Case Studies of In-Vessel and Ex-Vessel Retention
2022, Frontiers in Energy Research

View all citing articles on Scopus

View full text

Elaboration of a Phenomena Identification Ranking Table (PIRT) for the modelling of In-Vessel Retention

Highlights

Abstract

Introduction

Section snippets

Methodology

Evaluations of uncertainties and impacts for the macro-variables

Ranking of parameters impacting the risks of excessive heat flux and ablation under transient situations

Risk of vessel failure

Conclusions

CRediT authorship contribution statement

Declaration of Competing Interest

Acknowledgments

Nucl. Eng. Des.

Nucl. Eng. Des.

Ann. Nucl. Energy

Nucl. Eng. Des.

Nucl. Eng. Des.

Procedia Eng.

Int. J. Press. Vessels Pip.

Nucl. Eng. Des.

Nucl. Eng. Des.

Enhanced natural convection in a metal layer cooled by boiling water

Nucl. Technol.