In-Vessel Retention Analysis for PHWR Calandria under Severe Core Damage Accident Condition using ASTEC

Transcription

1 In-Vessel Retention Analysis for PHWR Calandria under Severe Core Damage Accident Condition using ASTEC O. S. Gokhale a*, P. Majumdar a, S. Belon b, D. Mukhopadhyay a, A. Rama Rao a a* Reactor Safety Division, Bhabha Atomic Research Center, Mumbai, India b Institut de Radioprotection et de Surete Nucleaire (IRSN), Cadarache, France Abstract The Severe Core Damage Accident (SCDA) scenario for Pressurized Heavy Water Reactors (PHWRs) are defined as accidents involving loss of core configuration typically induced by the disassembly of fuel channels inside horizontally placed cylindrical Calandria. Unlike in case of Light Water Reactors (LWR), the debris bed will be made of long pipe like structures submerged by the coolant at near saturation temperature. This debris configuration is widely different from debris of TMI-2 accident as well as in different molten fuel coolant interaction experiments. The moderator boil-off followed by the heat up of debris may lead to melting of debris, forming molten corium layers. However, this is strongly governed by the dimensions of the Calandria and fuel inventory and power. Light water present in the Vault surrounding the Calandria provides a large heat sink for decay heat removal from debris/corium. Addition of water into the Vault is one of the prescribed SAMG actions for Indian PHWRs to facilitate cooling of debris/corium. That paper presents the analysis carried out with the European Severe Accident code ASTEC V2.1 to assess the capability of Calandria to retain debris/magma under such SAMG action. The PHWR specific models of ASTEC facilitates the modelling of the horizontal reactor core and the heat up of different elements of reactor block like tube sheets and carbon steel balls of the end shield, baffle plate, octagonal flanges and liner tubes. That first adaptation of ASTEC to SCDA, is based on a modelling which take into account proper heat losses from debris bed to surroundings, which in turn, will predict a realistic heat up of debris and Calandria. The radiation heat transfer between debris/corium and the Calandria has been exclusively modeled with radiosity approach. The heat exchanged with End Shield which also acts as a heat sink have been accounted for this analysis. A simple heat transfer model has been developed to evaluate the external Calandria vessel cooling. The analysis predicts that the ex-vessel cooling of debris retained within the Calandria is found to be successful for 220 MWe and 540 MWe Indian PHWRs. In case of 220 MWe PHWR the heat up of the debris is even restricted to temperatures below the melting temperature. The Calandria shell temperature is predicted to be at saturation temperature of the vault water. For the investigated plant scenarios, the analysis illustrates that PHWR Calandria is capable of retaining the degraded core under severe accident condition. Further evolutions of ASTEC code to improve the prediction are expected. Modelling should evolve to represent detailed exchanges with the vault block (including thermal radiation) and to propose a model to describe complete scenario, starting by limited core damage accident (LCDA) and transition to SCDA phase. Keywords: ASTEC adaptation, In-Vessel Retention, PHWR * Corresponding author address: onkarsg@barc.gov.in (O. S. Gokhale)

2 1. Introduction The 8 th European Review Meeting on Severe Accident Research -ERMSAR-2017 A typical Pressurized Heavy Water Reactor (PHWR) consists of 19 or 37 element fuel bundles arranged in horizontally placed Pressure Tube (PT). Typically a single PT houses 12 fuel bundles (approximatively 1m length). The PT is housed inside a Calandria Tube (CT) and the gap between PT and CT is filled with CO 2 which is monitored with Annual Gas Monitoring System. The assembly of fuel bundles, PT and CT forms a channel. The channels are surrounded by low pressure moderator contained in large diameter vessel called calandria. Under normal operating conditions the moderator temperature is maintained with the help of moderator cooling system. The calandria is closed on either ends by end shields (ES). The ES are large structures with high thermal inertia that consist of Carbon Steel balls enclosed between two tube sheets. Moreover balls and sheets are cooled by ES cooling water system (figure 6). The whole calandria is kept inside a vault which is filled with water under normal operating conditions. Then the calandria is submerged in the vault water. A cut section of the core of PHWR is shown on figure 1. Detailed description of PHWR design features can be found elsewhere [1]. In PHWRs, severe accident is categorized into two distinct phases, namely a) Limited Core Damage accident (LCDA) and b) Severe Core Damage accident (SCDA), based on the severity of the damage. The extent of the core damage depends on the availability of the moderator heat sink. The moderator cooling system remove the decay heat from the CT surface and maintains the channel integrity even if the fuel bundle has failed and the PT geometry has deformed. The other extreme scenario, SCDA, occurs when moderator cooling system is not available. In such case, moderator boil-off will occur and channels will get exposed progressively from top to steam. This leads to their rapid heat up accompanied by oxidation. At sufficiently high temperature, the channels would sag, disassemble and relocate at the bottom of the Calandria. The integrity of the Calandria depends on the availability of water in-calandria vessel and on the external cooling of calandria by the Vault water. Details of typical postulated severe accident scenario can be found elsewhere [2]. The ASTEC severe accident integral code is developed and maintained by the French Institut de Radioprotection et de Sûreté Nucléaire (IRSN) with contributions specific to PHWR model from the Bhabha Atomic Research Center (BARC). The recent major version V2.1 [3] carried additional features to model and to simulate channel type vertical and horizontal reactors. New models have been introduced in this version to tackle the core asymmetries that arise in PHWRs during accidents beyond design basis [4]. The elaborated application of the PHWR module is available in the paper [5] for a specific accident in a typical 220 MWe Indian PHWR consisting in a LOCA with loss of ECCS and power supply. It may be noted that the module is still limited to accident scenario wherein the core channels are intact and well cooled by the surrounding moderator. This paper describes the further development work carried on ASTEC code for simulating SCDA in PHWRs. The models involved in the geometrical adaptation of ASTEC to represent horizontal cylindrical core configuration typical for a PHWR and the development of heat transfer models specific for such configuration are discussed in details. The models are used to assess the capability of the calandria to retain debris/corium under Severe Accident Management Guidelines (SAMG) for 220 MWe and 540 MWe PHWR. 2. Geometrical Adaptation of ASTEC LOWERPLE for PHWR In the simulation of the late phase of a core damage accident in Pressurized Water Reactors (PWR) using ASTEC, the debris formed are classically contained by a semi-ellipsoidal lower head and associated plenum which is represented by a specific mesh modelling (so called "LOWERPLE" mesh described in [6]). Whereas, in case of PHWRs, the debris are contained by the calandria itself, whose shape is a wide cylinder of several meters of diameter. The proposed modelling consists to represent calandria as a cylindrical lower plenum using actual description of debris beds and corium available in ASTEC code. That solution was made possible as most of phenomena modelled for corium relocated in PWR lower plenum are relevant for late phase of SCDA for PHWR. The models designed for molten pool in case of PWRs can be used in some extent with a proper adaptation of the shape as for PHWRs, corium layers fill some parallelepiped shape. To account for the different geometry of PHWRs, the lower plenum specific modelling of ICARE module of ASTEC V2.1 has been modified Representation of Calandria and Debris/Corium The Calandria of PHWR is symmetric about the vertical axis. Hence only one half (one side of the axis) is represented using a 2D meshed cylindrical object (so called LOWERPLE macro component shown in Figure 2). As shown in Figure 3, each element of calandria wall is described by one thermal node, one composition and four different faces on which physical phenomena can be evaluated (conduction, convection, radiation, etc ). For instance, the upper and lower faces interact with adjacent elements of the calandria wall by conductive heat Indicate here the SESSION name and the Paper N 604

3 transfer whereas the most internal and the most external faces interact respectively with moderator/debris/magma and vault water. figure 1: Cut section of a typical PHWR core As for PWR classical lower plenum modelling, the relocated materials can be described by up to 3 corium layers and two debris beds (respectively below and above the corium layers). Figure 4 shows the element (so called component) representing the layer of magma/debris in the 2D meshing of the calandria. The layer element has an internal (FI) and an external (FE) faces for heat exchanges with internal structures and calandria respectively. Similarly it has top and bottom faces for heat exchanges with neighboring layers (either debris or corium). The layer element is provided with an additional area (represented by A on figure 5) for heat exchange with end shield elements (circular element which close the cylindrical calandria). As the end shields have a big thermal inertia and are water cooled, their role in energy storage and sink have to be modelled using a specific description. figure 2 Vertical half of Calandria discretized in radial and axial direction figure 3 Schematic view of a calandria element and associated faces for interaction Severe Accident Scenarios Session - paper N 604

4 figure 4: 3D representation of an element representing the debris/magma layer figure 5: Schematic view of a debris/magma element 2.2. Representation of END SHIELD figure 6 : Cut view of the end shield of PHWR calandria figure 7 : Cross section of the end shield of a PHWR calandria As discussed in section 1, calandria of PHWRs is closed on both the ends with disc like structures called end shields. The specific ASTEC mesh for lower plenum does not have a provision to build an element representing those PHWRs structures. ASTEC modifications were carried out to facilitate the definition of the end shields geometry including the peculiarity of its design. The end shield is composed of two shielding wall which contains carbon steel (CS) pebbles cooled down by water. In ASTEC model, it was chosen to represent end shield by a specific set of radial elements (up to three), each of them associated to a thermal node. Those elements represent respectively the inner plate, the CS pebbles bed and the outer plate of a typical PHWR end shield as shown on figure 6. To obtain a precise description of the heat sink through the end shields, several geometrical and design details can be provided by the user. For instance, user defined parameters allows to define respectively the number and the equivalent diameter of penetrations through the end shield for passage of coolant channels as shown in figure 7. Another example concerns the description of CS balls enclosed in the end shield which are modeled similarly to a debris bed taking into account the adequate thermal conductivity. Thus, the porosity of the CS balls layer and their diameter are defined by the user. Each layer composing the end shield elements has one inner and one outer heat exchange surface represented by FI and FE faces respectively. The Indicate here the SESSION name and the Paper N 604

5 areas are calculated according to the calandria geometry (and meshing) and taking into account penetrations of the channels. For the porous layer representing the CS balls, an additional area of exchange (FH) is defined to represent the surface area of the CS balls and used to represent inner cooling by end shield water flow. 3. Management of Heat Exchanges within calandria Associated to the specific meshing and elements to represent the PHWRs calandria, the heat exchanges were described taking into account conduction, convection and radiation. The development consisted mainly to adapt the conductive, convective and radiative heat exchanges in the ASTEC lower plenum mesh to suit specific needs of PHWR configuration Heat Conduction in calandria Each layer of the end shield defined in the LOWERPLE mesh is represented by a single thermal node. Conduction between these nodes is evaluated using classical conduction between solid layers. The CS balls layer, if defined in, is treated taking into account the thermal effective conductivity of the equivalent debris bed Convective Heat Exchanges in calandria The innermost layer of the end shields element, the calandria wall and the upper layer of the debris are expected to be in contact with the steam and remaining water present within calandria. The convective heat exchanges between the steam/water and these surfaces are incorporated into the model by taking into account the correct area available for heat exchanges. That area was obtained taking into account the level of the debris and the water level within the calandria checking the submergence of debris layers in water Convective/Conductive Heat Exchanges within Debris/Magma Layers The innermost layer of the end shield and calandria wall is expected to be in contact with debris and corium. The convective or conductive heat exchanges among the layers in contact and between end shields or calandria and the debris/corium layers are evaluated using the same principle that in classical ASTEC PWR configuration. The model manages dynamically the contact and surface area available for heat exchanges as the appearance of convective phenomena (according to viscosity of molten corium). The area available for exchanges were obtained based on the level of the magma/debris and the area of calandria wall or end shield in contact with those layers. The correlations used for the calculation of heat transfer coefficients do not depend on any parameters which are specific to the geometry of the layers or lower head wall. These correlations were established from the BALI-experiments [7] whose geometry is based on a cylindrical slice as shown in figure. A correlation was derived and extrapolated to the hemispherical geometry, like in case of PWR and is used by default in ASTEC V2.1 code. Of course, the original correlation from the BALI experiment (applicable to the experiment geometry) was suitable for the horizontal cylindrical calandria configuration, like in case of PHWRs. figure 8 : Schematic view of BALI experiment Setup [7] Severe Accident Scenarios Session - paper N 604

6 Hence this BALI-correlation was incorporated to calculate the heat transfer coefficient between the corium and the calandria wall for PHWR applications, instead of the default BALI-integrated 3D correlation applicable to PWR Development of model for Radiative Heat Transfer within calandria specific for PHWR During SCDA phase of Severe Accidents in PHWRs, the temperature difference between the top of debris bed or of the corium pool (mainly formed by molten debris) and the calandria wall is significant as the calandria is externally cooled by vault water. The radiative heat transfer is, thus, dominant within the calandria and plays a major role in the energy removal from the molten corium through calandria to the vault water. A specific radiative model was developed based on the net radiation enclosure method to facilitate the calculation of radiative heat exchanges between different elements within the calandria. The following components participate in the radiative heat exchanges within the calandria: The top most magma/debris layer The components of calandria wall, not submerged in water or corium/debris layers taking into account the symmetry along the vertical axis The portion of the innermost layer of the end shield, not submerged in water or corium/debris layers The symmetry of the calandria along vertical axis is taken into consideration for the view factors calculation. The model also checks the possible formation or disappearance of corium/debris layers during the transient and updates the list of participating walls whenever required. figure 10 Two finite rectangle at an angle figure 9 Two distant finite rectangle planes with an angle [9] The RADPHWR model uses two approaches for calculation of view factors between the participating surfaces. If an end shield is modelled within the calandria, a finite surface approach is used for view factors calculation, wherein, the calandria is assumed to be of finite length closed on either ends with end shield. The length is provided by the user in the discretization of the calandria vessel. The view factors between walls and the top most corium/debris layer are then calculated based on the formulae for finite flat plates at an angle (figure 9). The second option is to use an infinite surface approach, in case the end shields are not modelled. Thus, the calandria of the PHWR and the magma layers are assumed to be of infinite length (thermal configuration is then 2D). The view factors for calandria wall and the top most corium/debris layer are then calculated based on the formulae for infinite flat plates at an angle (Figure 10). That last option allows to perform 2D simple calculation (useful for design studies for instance) Modification of CESAR Heat Exchange Coefficient model for PHWR specific Ex-vessel Heat Transfer Correlations It s reminded that the calandria vessel is initially submerged by vault water during normal operating Thus, it is necessary to model external cooling of Calandria to be able to achieve realistic simulation of SCDA in PHWRs. The work presented in this paper corresponds to an extension of ASTEC for SCDA in PHWR which was done ahead in time to the availability of the new external reactor vessel cooling model in ASTEC (available Indicate here the SESSION name and the Paper N 604

7 from the version 2.1.1). As a consequence, such model was not used to represent the external cooling of calandria by the vault water. The external cooling was assessed using an explicit coupling with a thermalhydraulic control volume. Moreover, the Calandria being horizontally placed and cylindrical in shape, specific set of correlations were required for the calculation of heat transfer correlation on the external surface of the calandria. These heat transfer correlations were not available in the classical correlations used by CESAR module for horizontal or vertical wall configuration. It corresponds for example to convective heat transfer on an inclined wall configuration (for which certain geometry dependent parameters are required such as inclination angle). To estimate the PHWRs ex-vessel cooling, the CESAR heat exchange coefficient model was, thus, modified to include these set of specific correlations. Following Table enlists the heat transfer correlations used by CESAR to evaluate the PHWR ex-vessel heat transfers: Table 1: List of correlations used for the PHWR specific Ex-vessel Cooling Heat Transfer Regime CHF Natural Convection to Water Natural Convection to Steam Nucleate Boiling Film Boiling Correlation INEEL Correlation McAdam's Correlation Churchill-Chu's Correlation Forster-Zuber's Correlation Bromley's Correlation and Breen and Westwater Correlation Analysis of Station Blackout Scenario in 540 MWe PHWRs with SAMG Injection into Vault To assess the capabilities of newly developed models and modifications carried out with regard to SCDA simulation of PHWRs, analysis was carried out to simulate Station Blackout Scenario with SAMG action of water injection into Vault. The plant parameters used in this analysis are typical to an Indian PHWR of 540 MWe capacity. The analysis starts with the end state of Limited Core Damage Accident (LCDA), i.e., the SBO has progressed to a state, where, all the channels within the core are collapsed to the Calandria bottom. The Calandria moderator level has dropped to half of the vessel height. The collapsed channels are assumed to form a debris bed represented by lower debris layer in the lower plenum mesh of ASTEC. figure 11 In-calandria moderator mass evolution (modelled by ASTEC lower plenum water) figure 12 Calandria rupture disk flow rate Severe Accident Scenarios Session - paper N 604

8 As the debris continue to generate decay heat, the moderator within the Calandria boils off. The decrease in the moderator mass and the flow rate transient through the Calandria rupture disk are shown in Figure 11 and Figure 12 respectively. With complete moderator boil off, the only heat sink available for the decay heat is the water in the vault and the end shields. The vault water and the end shield water reach saturation temperature and boil off begins with rupture disk opening. Figure 13 shows the water levels in the Calandria, end shield and vault. The flow rates through End shield opening and vault opening are shown in Figure 14. The SAMG action consisting in a water injection into the vault is started as soon as the moderator level drops below top of calandria at a specified flow rate [8] as shown in Figure 15. In actual practice, it s reasonable to consider that this action will be initiated much earlier. There is sufficient time margin of more than 16 hours available from the core collapse to initiate that operator action. Since the calandria remains submerged in vault water the calandria temperature remains near saturation as shown in Figure 16. figure 13 Vault, end shield and core water level evolutions figure 14 Flowrates through end shield rupture disk and vault rupture disk figure 15 Vault water injection flowrate transient figure 16 Calandria wall temperature evolution Indicate here the SESSION name and the Paper N 604

9 The rise in the core temperature at different times is shown in Figure 17 to Figure 20. Figure 21 shows the temperature evolution for different heat structures connected to the end shield structure. Figure 22 shows the temperature evolution of debris. The debris temperature stabilizes towards end of the transient suggesting effective heat removal by the vault water. No melting of debris is observed. figure 17 Initial configuration with cold debris bed at 10000s figure 18 Rise of the temperature of debris figure 19 Debris temperature at s figure 20 Final configuration and temperature of debris (no melting observed) Severe Accident Scenarios Session - paper N 604

10 K time (s) figure 21 Temperature evolution of the end shield structures figure 22 Debris layer temperature evolution with 4. Analysis of Station Blackout Scenario in 220 MWe PHWRs with SAMG Injection into Vault The 220 MWe PHWRs have lower power density in the calandria as compared to the 540 MWe PHWRs. Design details of 220 MWe PHWRs can be found elsewhere [10]. The SBO analysis for 220 MWe also shows that injection of water into the vault as a SAMG action is able to fill up the vault as shown in Figure 23. Since the Calandria remains submerged in water, there is no rise in the temperature of the Calandria wall as shown in Figure 24. figure 23 Vault, End Shield and Core water level evolutions figure 24 Calandria wall temperature evolutions figure 26 Debris bed temperature evolution Indicate here the SESSION name and the Paper N 604

11 figure 25 Temperature evolution of the end shield structures Figure 25 and Figure 26 show a decreasing trend in the temperatures of end shield heat structures and debris suggesting effective heat removal by vault water. 5. Conclusions The development and the extension of ASTEC V2.1 code to carry out the SCDA late phase analysis for PHWR is described in the present paper. It is demonstrated through the analysis of a plant case that the code is capable of modeling the PHWR SCDA scenario. In particular, a first model dealing with the PHWR specific in-calandria radiation model dedicated to late phase configuration and the external-calandria cooling heat transfer have been incorporated in ASTEC. It facilitates the simulation of SCDA phase and allows testing of the possible SAMG actions to insure the in-calandria melt retention. The analysis carried out for 220 MWe and 540 MWe with SAMG action of water injection into vault demonstrates that the calandria remains intact. The debris temperature stabilizes in both the cases indicating successful removal of decay heat through vault water. The present work is pursued by efforts of BARC, in the frame of collaboration with IRSN, to extend the capabilities of ASTEC code. Some improvements done at the end of 2016 concerned the evaluation of heat exchanges during SCDA and in particular the thermal exchanges at the external side of the calandria vessel. For instance, the radiative heat exchanges between uncovered part of calandria and vault wall was evaluated during boil off of the vault water (after vault rupture disk rupture). Another improvement consisted to use the external reactor vessel cooling model available in the latest version of ASTEC V2.1 to evaluate the convective cooling of the calandria. Thus, the behavior during SCDA would be modelled more precisely. One major evolution which is expected in further version of the code will link the current description for LCDA and the new SCDA modelling. That work has already started and aims to propose a first modelling of the transition phase corresponding to the simulation of LCDA, core disassembly and progressive transition to SCDA configuration in a full plant scenario. Of course many uncertainties remain on the core disassembly, on the shape and state of debris formed by core disassembly and their relocation to the bottom of calandria. These uncertainties impact the possible evaluation of SCDA and transition phase. The proposed evolution of ASTEC code do not pretend to answer fully to the simulation challenges represented by that complex phase but will certainly provide a first simulation of a complete PHWR transient. Severe Accident Scenarios Session - paper N 604

12 Abbreviations: The 8 th European Review Meeting on Severe Accident Research -ERMSAR-2017 BFFLEP CESAR Baffle Plate ASTEC code thermalhydraulic module MSHIELD LWR Shield Balls Light Water Reactor CS Carbon Steel OS Outer Shell CT Calandria Tube PHWR Pressurized Heavy Water Reactor DP Diaphragm Plate PT Pressure Tube ES End Shield PWR Pressurized Water Reactor ESHIELD IS External Tube Sheet Inner Shell SAMG Severe Accident Management Guidelines ISHIELD Inner Tube Sheet SBO Station Black Out LOWERPLE Specific mesh for SCDA lower plenum of ASTEC code SS Severe Core Damage Accident Stainless Steel References: [1] 2008, Analysis of Severe Accidents in Pressurized Heavy Water Reactors, IAEA TECDOC 1594, IAEA, Vienna. [2] 2013, Benchmarking of Severe Accident Computer Codes for Heavy Water Reactor Application, IAEA TECDOC 1727, IAEA, Vienna. [3] P. Chatelard, S. Belon, L. Bosland, L. Carénini, O. Coindreau, F. Cousin, C. Marchetto, H. Nowack, L. Piar, L. Chailan, Main modelling features of the ASTEC V2.1 major version, Annals of Nuclear Energy, Volume 93, July 2016, Pages [4] P. Majumdar, B. Chatterjee, H.G. Lele, G. Guillard, F. Fichot, ASTEC adaptation for PHWR limited core damage accident analysis, Nuclear Engineering and Design, Volume 272, June 2014, Pages [5] P. Majumdar, H.G. Lele, S. Belon, F. Fichot, Analysis of Large break LOCA with simultaneous loss of power supply in 220 MWe India PHWR using ASTEC, 6th European Review meeting on Severe Accident Research (ERMSAR-2013), Avignon (France), Palais des Papes, 2-4 October, [6] L. Carénini, J. Fleurot, F. Fichot, Validation of ASTEC V2 models for the behaviour of corium in the vessel lower head, Nuclear Engineering and Design, Volume 272, June 2014, Pages [7] J. M. Bonnet, Thermal hydraulic phenomena in corium pools: the BALI experiment, JAERI-conf, [8] Maurya et.al., Provision of improved inventory lone-up (Hookup) schemes in 700 MWe PHWR project for severe accident management based on operational experience of earlier units, IAEA-CN- 158/33. In: International Conference on Topical Issues in Nuclear Installation Safety, Mumbai, India [9] Gross, U., Spindler, K., and Hahne, E., 1981, "Shape factor equations for radiation heat transfer between plane rectangular surfaces of arbitrary position and size with rectangular boundaries," Lett. Heat Mass Transfer, vol. 8, pp [10] S. S. Bajaj and A. R. Gore, The Indian PHWR, Nuclear Engineering Design, Vol. 236, pp , Indicate here the SESSION name and the Paper N 604