CAREM-25 RPV THERMAL REGIME EVALUATION DURING THE APPLICATION OF IN-VESSEL RETENTION STRATEGIES

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1 CAREM-25 RPV THERMAL REGIME EVALUATION DURING THE APPLICATION OF IN-VESSEL RETENTION STRATEGIES Lázaro E.Pomier Báez, Jorge H. Barón, Juan E. Núñez Mac Leod Instituto CEDIAC, Facultad de Ingeniería, Universidad Nacional de Cuyo Parque General San Martín, Ciudad Universitaria 5500 Mendoza, Mendoza, Argentina ABSTRACT The structural integrity of the reactor vessel is a key question in the analysis of the possibility for retaining the melted materials inside the pressure vessel as a severe accident management (SAM) strategy. The pressure of the system and the thermohydraulic behavior of relocated materials that determine the loads, stresses and the displacements of reactor vessel determine the vessel failure mode and the time until rupture. In-Vessel Retention (IVR) strategy analyses are carried on as part of the advanced design CAREM-25 SAM evaluations. One of the most promising initiatives in this area, is the development of an in-vessel metallic core catcher (IVCC) to arrest reactor vessel meltdown sequences during a severe accident. The concept working principle consists in the mixing of the catcher metallic material (so-called sacrificial material) with the corium relocating fragments in the reactor lower head after the initiation of relocation process. The catcher will limit the catchercorium mixture temperature by boiling the sacrificial material. For this purpose, a low-boiling point material is chosen. The analysis methodology presented in this paper is designed to evaluate the integrity of CAREM-25 pressure vessel during a severe accident sequence with complete core damage when the IVR strategy is employed. CAREM-25 is a multipurpose small advanced reactor design being developed by CNEA (Comisión Nacional de Energía Atómica) and INVAP S. E. in Argentina. keywords: advanced reactor, severe accident, core catcher, sam. I. INTRODUCTION CAREM-25 is a project for an advanced multipurpose small nuclear power plant of 100 MWt, (25 MW net electrical output), jointly developed by CNEA and INVAP S. E. in Argentina. The CAREM project is conceived with new generation design solutions and standing on the large worldwide experience accumulated in the safe operation of Light Water Reactors (LWR). The advanced nuclear reactor CAREM-25 includes several improvements not only related to the design basis events, but also to severe accident events. These features give the possibility to enhance the passive safety functions of the plant. The in-vessel evolution of a severe accident in a nuclear reactor is characterized, generally, by core uncovery and heat-up, core material oxidation and melting, molten material relocation and debris behavior in the lower plenum up to vessel failure. The in-vessel progression of the accident sequence involves a large number of physical and chemical phenomena that may depend on the accident sequence itself and the reactor type under consideration. The occurrence of such core meltdown accidents cannot be avoided for existing or future nuclear power plants (NPP) (vessel type reactors), despite the great efforts that have been made to reduce the likelihood of these events. SAM strategies are another important topic under consideration in the R&D (research and development) activities for the CAREM project. In this sense, the development of the CAREM NPP, considers the possibility of including a passive metallic in-vessel container (core catcher) in its design, to arrest the reactor pressure vessel meltdown sequence during a core damaging event, and therefore prevent its failure. The core catcher should be accompanied by an effective mean of heat removal that could guarantee the medium and long term cooling of the reactor vessel. The reactor cavity flooding is considered as a heat sink for the residual heat power transferred through the reactor lower head wall. This paper presents the methodology applied to evaluate the in-vessel retention strategy being developed for CAREM-25 design as a part of R&D activities and SAM. The calculation results confirm the validity of the approach. Nevertheless, further research is considered necessary because the remaining uncertainties associated to the involved phenomena.

2 II. CAREM-25 DESIGN FEATURES The CAREM design consist in an indirect cycle reactor of 100 MWt (25 MWe) with some distinctive features that greatly simplify the reactor plant and also contribute to a high level of safety. The main criteria used in the design of safety systems were simplicity, reliability, redundancy and passivity. The Fig. 1 shows a cross section view of reactor. III. CAREM-25 IVCC DESCRIPTION The main element in the IVR strategy developed for CAREM-25 reactor is the In-Vessel Metallic Core Catcher [1] concept. The IVCC is shown in Fig. 2. The device consists in two components. The lower and peripheral part is a zinc shield that covers the whole lower head of the reactor vessel. The shield is constructed from parts that can be individually attached or separated for easy installation, removal and inspection. The other part of IVCC (upper part) is constructed of a structured material with a low porosity. The function of this upper part, designated as zinc debris, is to mix well with the relocating corium. The configuration and texture of this upper part is still under study. Several solutions have been proposed, ranging from zinc pellets in a basket, honeycomb structures, to an array of zinc rods. The main objective of this solution is to provide sufficient contact area for the relocating core debris. Figure 1. CAREM-25 Cross Section View. The flow rate in the reactor primary system is achieved by the natural circulation of coolant. The pressure relief system has three relief valves to protect the integrity of the reactor pressure vessel against overpressure. The steam generators (SG) are of a Mini Helical vertical, once through design. The secondary system exits the steam generator with ample superheating. The fuel element cross section is hexagonal. Each fuel element contains 108 fuel rods, 19 guide thimbles, and one instrumentation thimble. The core has 61 fuel elements. To extinct the nuclear chain reaction and to keep the reactor in a sub-critical state, CAREM NPP has two different and independent safety systems activated by the reactor protection system or by the operator The emergency injection system prevents the core exposure in case of LOCA. The system injects borated water flooding the RPV. The system provides at least 48 hours of protection to the core after the accident initiation. CAREM containment system where the primary system, the reactor coolant pressure boundary and important ancillary systems are enclosed is a cylindrical concrete structure with an embedded steel liner. It is a pressure suppression type containment and assures that at least 48 hours after initiated the accident, without any external action, the pressure in its interior is kept below the maximum design value of 5 bar. Figure 2. CAREM-25 In-Vessel Core Catcher Schematic View. The IVCC working principle consists in the limitation of the catcher-corium mixture temperature by the boiling of the catcher material. In this way, the reactor vessel wall temperature is kept at the mixture temperature avoiding the thermal failure of the lower head. For this purpose, a low-boiling point material is needed. Several materials were analyzed from the thermodynamic properties point of view. Zinc was selected based on its thermal properties. It has a melting point of K, a boiling point of 1180 K, a relatively high thermal conductivity of 116 W/m/K, a fusion enthalpy of 6.67 kj mol -1, and a boiling enthalpy of kj mol -1. The zinc density is 7.14 g/cm 3, relatively low when compared to the uranium dioxide density (10.93 g/cm 3 ). The IVCC would also help to accomplish the following functions: The reduction (and eventual elimination) on the likelihood of steam explosions. The prevention of system re-criticality phenomena. The reduction of the fission products source term from the fuel fragments in the relocated corium.

3 IV. SAM IVR METHODOLOGY FOR CAREM-25 The SAM IVR strategy developed in the framework of severe accident R&D activities for CAREM-25 is shown in Fig. 3. Plant process Parameters Evaluation Timing of Melted Materials Relocation Process In-Vessel Core Catcher Dimensions Optimization Reactor Vessel Thermal Evaluation Ex-Vessel CHF Evaluation CAREM-25 Design Modifications Evaluation Figure 3. Design Modifications Evaluation Methodology. Considering that the proposed modifications are linked to a safety management strategy, which main objective is to prevent the accident to go beyond the reactor vessel limits, it is necessary to perform an evaluation of the major phenomena and events. The next sections provide information about every step of the general sequence. V. ANALYSIS OF PLANT PARAMETERS AND RELOCATION PROCESS CHRONOLOGY DURING A SEVERE ACCIDENT SEQUENCE The analysis of the plant parameters and relocation process chronology during a severe accident sequence was performed with the fully integrated computer code MELCOR [2]. MELCOR contains a number of physics packages and modules that model all essential phenomena and plant features. For the analysis it was selected, as the initial event leading to the severe accident sequence, the reactor pressure relief valve stuck open. The objective of this analysis is to study the events chronology inside the reactor vessel during the accident sequence to obtain the initial conditions for the Pre-ALGOR preprocessor, which purpose is to determine the proportions of IVCC. The Fig. 5 shows the nodalization employed to analyze the CAREM-25 with MELCOR. The analysis methodology is intended to demonstrate that the reactor lower head wall temperature will not reach values leading to its failure and the subsequent release of fission products. The methodology is based on the use of numerical simulation techniques such as the thermalhydraulic and finite elements analysis (FEA) codes. The calculation sequence is depicted in Fig. 4. Geometrical Data Operational Data Materials Properties Data Initial Conditions Boundary Conditions MELCOR PC Pre-ALGOR ALGOR PC CHF Limit Therlim-1 Therlim-2 RESULTS Figure 4. General Calculation Sequence. Figure 5. CAREM-25 Nodalization.

4 The model consists of eight control volumes (CV) (7 for the reactor vessel and internals, and 1 CV for the containment system) representing CAREM-25 major hydrodynamic characteristics. The stuck open valve is situated at the top of the reactor vessel. Another valve, at the bottom of the vessel, represents the reactor lower head failure when conditions for that are reached. The chronology of events during he accident sequence is shown in the Table 1. TABLE 1. Chronology of Events Event Time (s) SRV stuck open starts 0 Upper plenum water level lost 1764 Core uncovers (top of active zone) Gap release in Ring Gap release in Ring Oxidation starts in Ring Oxidation starts in Ring Oxidation starts in Ring Gap release in Ring Core uncovers (bottom of active zone) Core support plate fail in ring Reactor vessel dryout Reactor equals cont. Pressure Core support plate fail in ring Core support plate fail in ring Lower head fail in ring Debris ejection to cavity starts Simulation ended VI. DETERMINATION OF IVCC DIMENSIONS The FEA code employed in the methodology is not able to evaluate the melting process of zinc. To overcome this situation, in the variants with core catcher it was necessary to evaluate the final state (decay heat power) of the corium-zinc mixture in the vessel after zinc meltdown. Furthermore, before setting up a FEA model it is necessary to determine the IVCC dimensions. To determine the dimensions of the in-vessel core catcher and the residual heat power of the corium after the interaction with the core catcher material the specific code PRE-ALGOR [3] has been designed. VII. REACTOR LOWER HEAD INTEGRITY ANALYSIS The potential for the thermal failure of CAREM vessel to occur was evaluated based on ALGOR [4] system code. ALGOR had passed NRC (Nuclear Regulatory Commission) inspection. ALGOR's compliance with federal regulations for quality in the nuclear industry assures that all of our customers will receive finite element analysis (FEA) and Mechanical Event Simulation software of the highest possible quality. The nodalization schemes for CAREM lower head analyses with ALGOR only consider the lower part of the vessel, the corium mass, the crust between vessel and corium or between corium and core catcher (when included), and the core catcher. Fig. 7 shows the FEA model. After the core support plate failure in different places, the total amount of molten material in the lower head is about 7766 kg as shown in Fig. 6. This mass represents 96% of the material composition of the intact CAREM core. Figure 7. Cladding Mass Variation in Core Ring 1. Figure 6. Lower Head Mass Inventory. These results are the initial conditions for the following step of the methodology. The crust is modeled with a thin layer (5 mm thickness). Due to the uncertainties in the formation of this crust layer, there are variants considering it, and without its inclusion, during simulations. The model takes advantaged of geometry symmetrical layout. The following conservative boundary conditions were assumed in the FEA analysis with ALGOR:

5 Convection and radiation heat transfer is only considered from the vessel outer wall. Two different scenarios were considered for the cavity: dry and wet conditions. The cavity temperature is assumed at 448 K in dry conditions. In wet conditions (flooded cavity), cavity temperature corresponds to the saturation temperature of water (373 K) at atmospheric pressure conditions. Inside the mass of the corium, only conduction heat transfer is considered. Boundary temperature values were adopted from MELCOR results. Decay heat power is uniformly distributed among the mass of corium. After the zinc meltdown, stratification was considered. This situation is considered conservative from the point of view of heat transfer between corium and zinc. To consider the uncertainty associated to crust formation, every variant was evaluated considering (CRUST) and without considering the presence of this layer (NO CRUST) between the reactor vessel wall and corium. The calculation variants were the following: Without IVCC and Dry Cavity. Without IVCC and Wet Cavity. With IVCC and Dry Cavity. With IVCC and Wet Cavity. Every calculation variant is coded as follows: NCA: IVCC not considered & cavity atmosphere is air. NCH2O: IVCC not considered & cavity is flooded. CA: IVCC considered & cavity filled with air. CH2O: IVCC considered & cavity is flooded. SS: steady state calculation. TS: transient state calculation. CRUST: crust layer is modeled. NO CRUST: crust layer is not modeled. DCH: decay heat power. 175: Air cavity temperature (448 K). To establish a criterion about the possible vessel damage during the accident sequence, three levels of temperature were adopted. The first value is 811 K (538 C). The value determines when the strength of steel decreases rapidly as its temperature exceeds 811 K. This is the maximum operational limit temperature established by ASME [5] for pressure vessels. The second level of damage is considered in the interval between K ( C). This is the transition temperature from ferritic to austenitic steel. This transition could produce two different failure modes [6] in the vessel depending on reactor cooling conditions, the global vessel failure or the local vessel failure. The third limit value is 1600 K (1323 C), the fusion temperature of SA-533B steel. The ALGOR calculations for the reactor lower head show the following results. For all variants with the cavity in dry condition, it is expected ablation of the inner wall of reactor vessel (see Table 2). Resulting temperatures of 1513 K (with crust) and 1519 K (no crust) are close to steel melting temperature. TABLE 2. Without IVCC and Dry Conditions Case NCA175SS + CRUST NCA175SS + NO CRUST NCA175TS + CRUST NCA175TS + NO CRUST These results confirm that vessel might lost its integrity if no measures are provided to avoid the increase of temperature. When the cavity is flooded (wet condition) and the core catcher is not simulated, temperatures in the lower head wall do not increase over 740 K (NCH2OSS- CRUST t RPV = 720 K, NCH2OSS-NO CRUST t RPV = 733 K) as it is shown in Table 3. These results confirm the effectiveness of the cavity flooding strategy. TABLE 3. Without IVCC and Wet Conditions Variante NCH2O5SS + CRUST NCH2OSS + NO CRUST NCH2OTS + CRUST NCH2OTS + NO CRUST Temperature values are far below SA533B steel melting point, when core catcher is simulated as depicted in Table 4. For the variant with dry condition and no crust (CA175ST-NO CRUST), the maximum calculated value was 1303 K. In the case of variant CA175ST-CRUST, the maximum temperature value was 966 K. TABLE 4. With IVCC and Dry Conditions Variante CA175SS + CRUST CA175SS + NO CRUST CA175TS + CRUST CA175TS + NO CRUST The combination of a flooded cavity and the core catcher brought the vessel wall temperatures below 700 K (see Table 5). These results confirm the effectiveness of this dual strategy to arrest a reactor vessel meltdown sequence. TABLE 5. With IVCC and Wet Conditions Variante CH2OSS + CRUST CH2OSS + NO CRUST CH2OTS + CRUST CH2OTS + NO CRUST 693.0

6 VIII. EX-VESSEL CHF EVALUATION To remove the remaining heat power from the vessel once the core catcher and corium have interacted, it is necessary to guarantee an appropriate heat sink. The cavity flood is being considered to fulfill this role. It is expected that high heat fluxes will appear between the water and the vessel. Unfortunately, the heat flux is limited by the critical heat flux phenomena (CHF). To evaluate the thermal limit two FORTRAN codes, THERLIM-1 [7] and THERLIM-2 [8] were assembled. They are based on the ULPU [9] and SULTAN [10] correlations, respectively. The results presented offer a valuable information to evaluate the conditions in which the process of heat exchange takes place in the cavity of the reactor when this is flooded. By means of the values predicted it is possible to evaluate the thermal reliability of the reactor vessel during the execution of the flood strategy. In order to show the utility of these results the following example is presented (see Table 6). To simplify the analyses the values predicted by the ULPU correlation are taken. The task consists of making an evaluation of the departure from nucleate boiling (DNBR) for the case of the variant NCH2OSS+NO CRUST. This variant considers the flood of the cavity and a value of the temperature of 693 K was obtained at the internal wall of the vessel. È [degrees] TABLE 6. DNBR Evaluations q ULPU [kw/m 2 ] DNBR-1 [-] DNBR-2 [-] The DNBR is calculated as the quotient between predicted correlation value to the actual heat flux value through the wall. The value of the heat flux, q Actual, obtained in the case of variant NCH2OSS+NO CRUST was 150 [kw/m 2 ] (DNBR-1) considering the conditions of the referred variant. The results of the analysis demonstrate that for these conditions a good margin is present until the crisis of the boiling in the surface of the vessel. Nevertheless, if the value of the heat flux increases up to 400 [kw/m 2 ] (DNBR-2) no longer exists an appreciable margin, particularly in the zones near the pole of the lower plenum of the reactor vessel (È= 0 degrees). Generally, the margin until the boiling crisis must be between 1,25 and 1,3 [11]. IX. CONCLUSIONS The in-vessel retention concept based on the use of a metallic core catcher has been presented. The developed methodology guarantees, from satisfactory results, that the novel proposal of the internal metallic zinc container is feasible to achieve the retention of the melted material inside the vessel of the nuclear reactor CAREM-25. The methodology has demonstrated that the temperature of the wall of the vessel, when using the invessel metallic container and the measures of accident management (flood of the reactor cavity), does not reach values that can lead to their failure and the subsequent release of radioactive material. The tools of numerical simulation on which the methodology of applied analysis is sustained, altogether with the applied conservative conditions, offer a suitable safety margin to the obtained results. REFERENCES [1] J. H. Barón, Conceptual Design of a Metallic In- Vessel Core Catcher, ICONE 8, ASME, Baltimore, USA [2] R. O. Gauntt, R. K. Cole, et. al., MELCOR MELCOR Computer Code Manuals, NUREG/CR-6119, Revision 1, U.S. Nuclear Regulatory Commission, [3] L. Pomier, PRE-ALGOR- Preprocessor for ALGOR System Code, CEDIAC Internal Report, [4] ALGOR Finite Element Analysis System Software, Revision 12.05a Win. February, [5] ASME, Use of SA-533 Grade B, Class 1 Plate and SA-508 Class 3 Forgings and Their Weldments for Limited Elevated Temperature Service, Sect. III, Division 1, Cases of ASME Boiler and Pressure Vessel Code, Case N-499, Approved December 16, [6] F. E. Haskin, et. al., Perspectives on Reactor Safety, Chapter 3.5- Molten Pours onto the Lower Head, NUREG/CR-6042, Rev. 1, November, [7] L. Pomier, THERLIM-1 Program to evaluate CHF Limits in the Outer Surface of Reactor Vessel Using ULPU Correlation, CEDIAC Internal Report, [8] L. Pomier, THERLIM-2 Program to evaluate CHF Limits in the Outer Surface of Reactor Vessel Using SULTAN Correlation, CEDIAC Internal Report, [9] T. G. Theofanous, In-Vessel Retention as a Severe Accident Management Strategy, In-vessel Core Debris Retention and Coolability Workshop proceedings, Garching near Munich, Germany, 3-6 march, [10] S. Rougé, et. al., Reactor Vessel External Cooling for Corium Retention SULTAN Experimental Program and Modelling with CATHARE Code, In-vessel Core Debris Retention and Coolability Workshop proceedings, Garching near Munich, Germany, 3-6 march, [11] L. S. Tong, J. Weisman, Thermal Analysis of Pressurized Water Reactors, ANS, 1970.