Evaluation of the Kinetics of Molten Pool Stratification in Case of In-Vessel Melt Retention Strategy

2018 ◽  
Author(s):  
L. Carénini ◽  
F. Fichot
Keyword(s):  
Heat Flux ◽  
Molten Pool ◽  
Metal Layer ◽  
Oxide Phase ◽  
Sensitivity Study ◽  
Heat Fluxes ◽  
Severe Accident ◽  
Total Power ◽  
Focusing Effect ◽  
The Impact

The In-Vessel Retention (IVR) strategy for Light Water Reactors (LWR) intends to stabilize and retain the core melt in the reactor pressure vessel. This type of Severe Accident Management (SAM) strategy has already been incorporated in the SAM guidance (SAMG) of several operating small size LWR (reactors below 500MWe, like VVER440) and is part of the SAMG strategies for some Gen III+ PWRs of higher power like the AP1000. One of the main issues for the demonstration of the success of the IVR strategy lies in the evaluation of the transient heat fluxes applied by the corium pool along the vessel wall. Indeed, these transient heat fluxes, during the corium pool stratification evolution, are expected to be higher than the steady-state ones, in particular due to the concentration of the heat flux in the top metal layer when it is thin (so called focusing effect). Another issue appears when a heavy metal is initially formed and rises later to the top (inversion of stratification): in such a situation, the metal goes through the oxide phase and accumulates a significant superheat which is likely to produce a high transient heat flux. Thus, it is of primary importance to be able to evaluate the duration of these transient peaks in order to evaluate the minimal residual vessel thickness after such fast transient ablation and draw conclusions about the vessel integrity. This paper first presents the phenomenology associated to the transient molten pool stratification and the model implemented in the severe accident integral code ASTEC (Accident Source Term Evaluation Code) to evaluate this kinetics. Then, evaluations are presented, based on a typical PWR reactor configuration. A sensitivity study is proposed to consider the impact of the main uncertainties on parameters which govern this kinetics. A particular focus is made on the physical phenomena driving the transient stratification of material layers in the corium pool and on the identification of critical situations with possible consequences in terms of vessel failure. The characteristic times of each individual process (chemistry, stratification, natural convection) are compared. In particular, the limiting cases of very fast chemistry or very slow chemistry are evaluated. This work is performed in the frame of the European H2020 project IVMR (In-Vessel Melt Retention) coordinated by IRSN. This project has been launched in 2015 and gathers 27 organizations with, as main objective, the evaluation of feasibility of IVR strategy for LWR (PWR, VVER, BWR) of total power 1,000MWe or higher.

A Revised Methodology to Assess In-Vessel Retention Strategy for High-Power Reactors

2018 ◽  
Author(s):  
F. Fichot ◽  
L. Carénini ◽  
W. Villanueva ◽  
S. Bechta
Keyword(s):  
Heat Flux ◽  
Steady State ◽  
High Power ◽  
Molten Pool ◽  
Metal Layer ◽  
Heat Fluxes ◽  
Standard Approach ◽  
Primary Circuit ◽  
Focusing Effect ◽  
Safety Margins

The In-Vessel Retention (IVR) strategy for Light Water Reactors (LWR) intends to stabilize and isolate corium and fission products in the reactor pressure vessel and in the primary circuit. This type of Severe Accident Management (SAM) strategy has already been incorporated in the design and SAM guidances (SAMGs) of several operating small and medium capacity LWRs (reactors below 500 MWe, e.g. VVER440) and is part of the SAMG strategies for some Gen III+ PWRs of higher power such as the AP1000 or the APR1400. However, the demonstration of IVR feasibility for high power reactors requires using less conservative models as the safety margins are reduced. In Europe, the IVMR project aims at providing new experimental data and a harmonized methodology for IVR. A synthesis of the methodology applied to demonstrate the efficiency of IVR strategy for VVER-440 in Europe (Finland, Slovakia, Hungary and Czech Republic) was made. It showed very consistent results, following quite comparable methodologies. The main weakness was identified in the evaluation of the heat flux that could be reached in transient situations, e.g. under the “3-layers” configuration, where the “focusing effect” may cause higher heat fluxes than in steady-state (due to transient “thin” metal layer on top). Analyses of various designs of reactors with a power between 900 and 1300 MWe were also made. Different models for the description of the molten pool were used: homogeneous, stratified with fixed configuration, stratified with evolving configuration. The last type of model provides the highest heat fluxes (above 3 W/m2) whereas the first type provides the lowest heat fluxes (around 500 kW/m2) but this model is not realistic due to the immiscibility of molten steel with oxide melt. Obviously, there is a need to reach a consensus about best estimate practices for IVR assessment to be used in the major codes used for safety analysis, such as ASTEC, MELCOR, SOCRAT, MAAP, ATHLET-CD, SCDAP/RELAP, etc. Despite the model discrepancies, and leaving aside the unrealistic case of homogeneous pool, the average calculated heat fluxes can reach, in many cases, values which are well above 1 MW/m2. This could reduce the residual thickness of the vessel considerably and threaten its strength and integrity. Therefore, it is clear that the safety demonstration of IVR in high power reactors requires a more careful evaluation of the situations which can lead to formation of either a very thin top metal layer provoking the focusing effect or significantly overheated metal, e.g. after oxide and metal layer inversion. Both situations are illustrated in this paper. The demonstration also requires an accurate thermo-mechanical analysis of the ablated vessel. The standard approach based on “yield stress” (plastic behaviour) is compared with more detailed calculations made on realistic profiles of ablated vessels. The validity of the standard approach is discussed. The current approach followed by many experts for IVR is a compromise between a deterministic analysis using the significant knowledge gained during the last two decades and a probabilistic analysis to take into account large uncertainties due to the lack of data for some physical phenomena, e.g. associated with molten pool transient behaviour, and due to excessive simplifications of models. A harmonization of the positions of safety authorities on the IVR strategy is necessary to allow decision making based on shared scientific knowledge. Some elements that might help to reach such harmonization are proposed in this paper, with a preliminary revision of the methodology that could be used to address the IVR issue. In the proposed revised methodology, the safety criterion is not based on a comparison of the heat flux and the Critical Heat Flux (CHF) profiles as in the current approaches but on the minimum vessel thickness reached after ablation and the maximum pressure load that is applied to the vessel during the transient. The main advantage of this revised criterion is in consideration of both steady-state and transient loads on the RPV. Another advantage is that this criterion is more straightforward to be used in a deterministic approach.

The Impact of Transient Behavior of Corium in the Lower Head of a Reactor Vessel for In-Vessel Melt Retention Strategies

2016 ◽  
Author(s):  
L. Carénini ◽  
F. Fichot
Keyword(s):  
Nuclear Power ◽  
Metal Layer ◽  
Heat Fluxes ◽  
Severe Accident ◽  
Total Power ◽  
Light Water Reactors ◽  
Light Water ◽  
Accident Management ◽  
Water Reactors ◽  
The Impact

One of the main goals of severe accident management strategies is to mitigate radiological releases to people and environment. To choose the most appropriate strategy, one needs to know the probability of its success taking into account the associated uncertainties. In the field of corium and debris behavior and coolability, research programs are still on going and the possibilities to efficiently cool and retain corium and debris inside the Reactor Pressure Vessel (RPV) then inside the containment are difficult to evaluate. This leads to uncertainties in safety assessments particularly when margins to RPV or containment failure are too weak. In Vessel Melt Retention (IVMR) strategies for Light Water Reactors (PWR, BWR, VVER) intend to stabilize and retain the core melt in the RPV (as it happened during the TMI-2 accident). This would reduce significantly the threats to the last barrier (the containment) and therefore reduce the risk of release of radioactive elements to the environment. This type of Severe Accident Management (SAM) strategy has already been incorporated recently in the SAM guidance (SAMG) of several operating medium size Light Water Reactors (reactor below 500MWe (like VVER440)) and is part of the SAMG strategies for some Gen III+ PWRs of higher power like the AP1000. A European project coordinated by IRSN and gathering 23 organizations (Utilities, Technical Support Organizations, Nuclear Power Plant vendors, Research Institutes…) has been launched in 2015 with as main objective the evaluation of feasibility of IVMR strategies for Light Water Reactors (PWR, VVER, BWR) of total power around 1000MWe (which represent a significant part of the European Nuclear Power Plants fleet). This paper intends to show how it is possible to introduce transient evolutions of the stratified corium pool in the evaluation of the heat flux profile along the vessel wall. Indeed, due to chemical reactions in the U–Zr–O–Fe molten pool, separation between non-miscible metallic and oxide phases may occur, modifying the thermal load applied to the RPV. If stabilized stratified corium configurations are well defined and modeled, transient evolutions of material layers in the corium pool are still difficult to predict. The evaluations presented are based on calculations performed with the severe accident integral code ASTEC (Accident Source Term Evaluation Code) for a typical PWR reactor. The modeling of transient evolution of corium layers leads to configurations with a thin light metal layer on top of the oxidic one, increasing the so called “focusing effect” (intense heat fluxes on the RPV walls adjacent to the top metal layer). A sensitivity study on some uncertain parameters is proposed to evaluate the impact on the kinetics of layers inversion. Depending on the duration of these transient heat fluxes, the mechanical strength of the RPV could be challenged.

Analysis of Focusing Effect of Light Metallic Layer in Stratified Molten Pool Under IVR-ERVC Condition

2015 ◽  
Vol 1 (2) ◽  
Author(s):  
Xi Wang ◽  
Xu Cheng
Keyword(s):  
Heat Flux ◽  
Molten Pool ◽  
Fluid Dynamic ◽  
Metal Layer ◽  
Local Heat ◽  
Metallic Layer ◽  
Parameter Method ◽  
Focusing Effect ◽  
Local Heat Flux ◽  
Reactor Pressure

The main failure mechanism of in-vessel corium retention through external reactor vessel cooling (IVR-ERVC) happens when the local heat flux through reactor pressure vessel (PRV) wall exceeds the critical heat flux (CHF). High local heat flux in the molten pool is usually caused by the metallic layer focusing effect due to stratification. In this paper, depending on experimental correlations, both the lump parameter method and computational fluid dynamic method are used to investigate the mechanism of focusing effect. The concentration factor varying with the height of metallic layer is studied. The results show that the lump parameter method probably overestimates the wall heat flux of metal layer.

scholarly journals Modelling of Severe Accident and In-Vessel Melt Retention Possibilities in BWR Type Reactor

2018 ◽  
Vol 2018 ◽  
pp. 1-14 ◽  
Author(s):  
Mindaugas Valinčius ◽  
Tadas Kaliatka ◽  
Algirdas Kaliatka ◽  
Eugenijus Ušpuras
Keyword(s):  
Heat Transfer ◽  
Heat Flux ◽  
Pressure Vessel ◽  
Heat Fluxes ◽  
Reactor Pressure Vessel ◽  
Severe Accident ◽  
Focusing Effect ◽  
Accident Management ◽  
Lower Head ◽  
Reactor Pressure

One of the severe accident management strategies for nuclear reactors is the melted corium retention inside the reactor pressure vessel. The work presented in this article investigates the application of in-vessel retention (IVR) severe accident management strategy in a BWR reactor. The investigations were performed assuming a scenario with the large break LOCA without injection of cooling water. A computer code RELAP/SCDAPSIM MOD 3.4 was used for the numerical simulation of the accident. Using a model of the entire reactor, a full accident sequence from the large break to core uncover and heat-up as well as corium relocation to the lower head is presented. The ex-vessel cooling was modelled in order to evaluate the applicability of RELAP/SCDAPSIM code for predicting the heat fluxes and reactor pressure vessel wall temperatures. The results of different ex-vessel heat transfer modes were compared and it was concluded that the implemented heat transfer correlations of COUPLE module in RELAP/SCDAPSIM should be applied for IVR analysis. To investigate the influence of debris separation into oxidic and metallic layers in the molten pool on the heat transfer through the wall of the lower head the analytical study was conducted. The results of this study showed that the focusing effect is significant and under some extreme conditions local heat flux from reactor vessel could exceed the critical heat flux. It was recommended that the existing RELAP/SCDAPSIM models of the processes in the debris should be updated in order to consider more complex phenomena and at least oxide and metal phase separation, allowing evaluating local distribution of the heat fluxes.

Reflood Oxidation Model Based on Reaction-Diffusion Equations

2010 ◽  
Author(s):  
Xiaoqiang He ◽  
Hongxing Yu ◽  
Guangming Jiang
Keyword(s):  
Metal Layer ◽  
Reaction Diffusion ◽  
Reaction Diffusion Equations ◽  
Diffusion Equations ◽  
Severe Accident ◽  
Model Based ◽  
Oxidation Model ◽  
The Impact ◽  
Relap5 Code ◽  
Good Agreement

An important accident management measure in PWRs is the injection of water to cool the degrading core, in which process the temperature and hydrogen production will significantly increase due to enhanced oxidation after shattering of zircaloy fuel rod. This phenomenon can be described by Zr oxidation model and shattering model. The process of Zr oxidation is usually represented by parabolic rate correlations. But, after consumption of primary β-Zr, or in steam starvation conditions, the correlation approach is restricted. Besides, using this approach, it is impossible to obtain detailed oxygen distribution in the cladding which impacts the detailed mechanical behavior, such as shattering of cladding. The shattering of cladding is mainly contributed by deep cracks penetrating the oxide layer as well as the adjacent metallic. In SCDAP/RELAP5, the shattering criterion is relevant to the thickness of β-Zr, the cladding temperature, and the cooldown rate. After shattering of cladding, the oxide scale is simply removed. This shattering criterion deviates from the experiment of Chung and Kassner when the maximum cladding temperature exceeds 1560 K, and the model can’t reveal the impact of the cladding surface temperature before cooldown on cladding conditions after shattered. An oxidation model based on reaction-diffusion equations at the temperature range from 923K to 2098K is developed in this study. By comparison with experimental data, the model shows reasonable results. Based on the oxidation model, the advanced shattering criterion is adopted, and a new empirical model to describe the cladding conditions after shattered is proposed. In present shattering model, R(T, m), which is the ratio between the area of new crack surfaces in the metal layer and the area of outer cladding surface, is the function of T (the temperature of the cladding surface before cooldown) and m (the thickness of the metal layer). With the help of single-rod QUENCH experiment, the preliminary expression of R(T, m) is obtained, and the results are in a good agreement qualitatively with the observation in this experiment. Further activities should focus on the impact of m and T on R(T, m), which needs more detailed single-rod experiments. Those developed models can be implemented into the SCDAP/RELAP5 code easily and used in the severe accident analysis in the future.

scholarly journals Uncertainty analysis of computational methods for deriving sensible heat flux values from scintillometer measurements

2009 ◽  
Vol 2 (3) ◽  
pp. 1383-1417 ◽  
Author(s):  
P. A. Solignac ◽  
A. Brut ◽  
J.-L. Selves ◽  
J.-P. Béteille ◽  
J.-P. Gastellu-Etchegorry ◽  
...  
Keyword(s):  
Heat Flux ◽  
Classical Method ◽  
Sensible Heat Flux ◽  
Bowen Ratio ◽  
Heat Fluxes ◽  
Eddy Correlation ◽  
Time Averaging ◽  
Sensible Heat ◽  
Ratio Measurement ◽  
The Impact

Abstract. The use of scintillometers to determine sensible heat fluxes is now common in studies of land-atmosphere interactions. The main interest in these instruments is due to their ability to quantify energy distributions at the landscape scale, as they can calculate sensible heat flux values over long distances, in contrast to Eddy Correlation systems. However, scintillometer data do not provide a direct measure of sensible heat flux, but require additional data, such as the Bowen ratio (β), to provide flux values. The Bowen ratio can either be measured using Eddy Correlation systems or derived from the energy balance closure. In this work, specific requirements for estimating energy fluxes using a scintillometer were analyzed, as well as the accuracy of two flux calculation methods. We first focused on the classical method (used in standard software). We analysed the impact of the Bowen ratio according to both time averaging and ratio values; for instance, an averaged Bowen ratio (β) of less than 1 proved to be a significant source of measurement uncertainty. An alternative method, called the "β-closure method", for which the Bowen ratio measurement is not necessary, was also tested. In this case, it was observed that even for low β values, flux uncertainties were reduced and scintillometer data were well correlated with the Eddy Correlation results.

scholarly journals Experimental Study on Stratification Morphology of the Molten Pool During Severe Accident

2021 ◽  
Vol 9 ◽  
Author(s):  
Yang Li ◽  
Houjun Gong ◽  
Yunwen Hu ◽  
Shengxing Yang ◽  
Yong Li ◽  
...  
Keyword(s):  
Stainless Steel ◽  
Oxide Layer ◽  
Molten Pool ◽  
Metal Layer ◽  
Atomic Ratio ◽  
Severe Accident ◽  
Main Element ◽  
Reactor Accident ◽  
Oxidation Degree ◽  
Experimental Parameters

Stratification morphology of a molten pool under severe reactor accident was investigated by the CESEF experimental facility. The experimental scale was 5,000 g, the atomic ratio of U/Zr was 1.5, the content of stainless steel was 10%, and the oxidation degree of Zr was 40–100%. It was shown that the molten pool was obviously stratified within the range of experimental parameters; one was a metal layer, and the other was an oxide layer. The layered morphology of the molten pool was different with the composition of different corium. With the decrease in the Zr oxidation degree, the metal layer moved downward in the molten pool, and the molten pool would overturn. The main elements in the oxide layer were U, Zr, and O, and the content of stainless steel was low. The main element in the metal layer was stainless steel and contained a certain amount of U and Zr.

Lumped Capacitance and 3D CFD Conjugate Heat Transfer Modelling of an Automotive Turbocharger

2015 ◽  
Author(s):  
R. Burke ◽  
C. Copeland ◽  
T. Duda ◽  
M. A. Reyes-Belmonte
Keyword(s):  
Heat Transfer ◽  
High Speed ◽  
Weak Link ◽  
Three Dimensional ◽  
Sensitivity Study ◽  
Heat Fluxes ◽  
Strategy Development ◽  
Inlet Temperature ◽  
Compression Process ◽  
The Impact

One dimensional wave-action engine models have become an essential tool within engine development including stages of component selection, understanding system interactions and control strategy development. Simple turbocharger models are seen as a weak link in the accuracy of these simulation tools and advanced models have been proposed to account for phenomena including heat transfer. In order to run within a full engine code, these models are necessarily simple in structure yet are required to describe a highly complex 3D problem. This paper aims to assess the validity of one of the key assumptions in simple heat transfer models, namely, that the heat transfer between the compressor casing and intake air occurs only after the compression process. Initially a sensitivity study was conducted on a simple lumped capacity thermal model of a turbocharger. A new partition parameter was introduced αA, which divides the internal wetted area of the compressor housing into pre and post compression. The sensitivity of heat fluxes to αA was quantified with respect to the sensitivity to turbine inlet temperature (TIT). At low speeds, the TIT was the dominant effect on compressor efficiency whereas at high speed αA had a similar influence to TIT. However, modelling of the conduction within the compressor housing using an additional thermal resistance caused changes in heat flows of less than 10%. Three dimensional CFD analysis was undertaken using a number of cases approximating different values of αA. It was seen that when considering a case similar to αA=0, meaning that heat transfer on the compressor side is considered to occur only after the compression process, significant temperature could build up in the impeller area of the compressor housing, indicating the importance of the pre-compression heat path. The 3D simulation was used to estimate a realistic value for αA which was suggested to be between 0.15 and 0.3. Using a value of this magnitude in the lumped capacitance model showed that at low speed there would be less than 1% point effect on apparent efficiency which would be negligible compared to the 8% point seen as a result of TIT. In contrast, at high speeds, the impact of αA was similar to that of TIT, both leading to approximately 1% point apparent efficiency error.

Estimating Eddy Heat Flux from Float Data in the North Atlantic: The Impact of Temporal Sampling Interval

2007 ◽  
Vol 24 (5) ◽  
pp. 923-934 ◽  
Author(s):  
Brian S. Chinn ◽  
Sarah T. Gille
Keyword(s):  
Heat Flux ◽  
North Atlantic ◽  
Sampling Interval ◽  
Heat Fluxes ◽  
Temporal Sampling ◽  
The North ◽  
Eddy Heat Flux ◽  
Daily Sampling ◽  
The North Atlantic ◽  
The Impact

Abstract Acoustically tracked float data from 16 experiments carried out in the North Atlantic are used to evaluate the feasibility of estimating eddy heat fluxes from floats. Daily float observations were bin averaged in 2° by 2° by 200-db-deep geographic bins, and eddy heat fluxes were estimated for each bin. Results suggest that eddy heat fluxes can be highly variable, with substantial outliers that mean that fluxes do not converge quickly. If 100 statistically independent observations are available in each bin (corresponding to 500–1000 float days of data), then results predict that 80% of bins will have eddy heat fluxes that are statistically different from zero. Pop-up floats, such as Autonomous Lagrangian Circulation Explorer (ALACE) and Argo floats, do not provide daily sampling and therefore underestimate eddy heat flux. The fraction of eddy heat flux resolved using pop-up float sampling patterns decreases linearly with increasing intervals between float mapping and can be modeled analytically. This implies that flux estimates from pop-up floats may be correctable to represent true eddy heat flux.

Risk-Informed Approach to Debris Bed Coolability Issue

2012 ◽  
Author(s):  
Sergey E. Yakush ◽  
Nazar T. Lubchenko ◽  
Pavel Kudinov
Keyword(s):  
Heat Flux ◽  
Surrogate Model ◽  
Thermal Power ◽  
Local Heat ◽  
Distribution Functions ◽  
Heat Fluxes ◽  
Severe Accident ◽  
Accident Risk ◽  
Dimensional Simulation ◽  
Accident Conditions

Coolability of an ex-vessel debris bed in severe accident conditions is considered from the risk perspective. The concept of “load versus capacity” is employed to quantify the probability of failure (local dryout). Possible choices of “load” and “capacity” in terms of heat fluxes, thermal power or melt mass are discussed. Results of Monte Carlo simulations of distribution functions for the local heat flux and the dryout heat flux at the debris bed top point (defined as the extensions of one-dimensional counterparts) are presented. A surrogate model for the dryout heat flux is developed by the generalization of two-dimensional simulation results. Dryout probabilities are obtained under the conservative assumptions (neglecting the coolability improvement due to side ingress of water into a non-flat debris bed), and from the surrogate model. Outlook is given for the prospective development of the risk-informed approach to debris bed coolability in the context of comprehensive severe accident risk analysis.

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