Current Status of the Melcor Nodalization for Atucha I Nuclear Power Plant. Zárate. S.M. and Valle Cepero, R.

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1 Current Status of the Melcor Nodalization for Atucha I Nuclear Power Plant Zárate. S.M. and Valle Cepero, R. presentado en USNRC Cooperative Severe Accident Research Program (CSARP), Maryland, EE. UU., 7-9 mayo 2001

2 CURRENT STATUS OF THE MELCOR NODALIZATION FOR ATUCHA I NUCLEAR POWER PLANT Zárate. S.M. and Valle Cepero, R. Nuclear Regulatory Authority Argentina OUTLINE Goal of presentation Development of a model for Atucha I with MELCOR Status Future Activities GOAL OF PRESENTATION Brief of overview of Argentina NRA activity with MELCOR code The following activities, with MELCOR code, have been continued in NRA as part of the project on Severe Accidents. In 1997 a global agreement was signed with NRC, therefore MELCOR is used at Argentina since The first works were in order to make well know the range of physical phenomena and thermal hydraulics response that MELCOR code models and learns to apply it. DEVELOPMENT OF A MODEL FOR ATUCHA I WITH MELCOR Objective is to analyze accident progression in Atucha I with MELCOR. A model for Atucha I is being developed in two parts: One of them involves reactor pressure vessel and reactor coolant system The other involves containment aspects. Atucha I has a particular design (prototype). The specific features are laid in both Reactor Pressure Vessel and Containment. In order to evaluate the response of the model: Two accidental sequences will be considered: Station Blackout (High Pressure Scenario) LB LOCA (Low Pressure Scenario) Up to now the model is being improved under a Station Blackout accidental sequence. 725

3 GENERAL PLANT DESCRIPTION Atucha I Nuclear Power Plant, has been designed and constructed by SIEMMES Company of Erlangen, Germany. The Atucha I, object of this study, is a nuclear power plant with a pressure heavy water reactor (PHWR) which is fuelled with natural uranium as well as slightly enriched uranium, reactor uses heavy water like coolant and moderator, it is on line refueled. The thermal power of 1170 MW, with nominal output electrical power of 357 MW. The reactor coolant system has two parallel loops, each of them having a circulation pump, a vertical steam generator with U tubes and the necessary connecting pipes. One of loops has a pressurizer with safety valves. Both primary system equipment and the accumulators within the emergency core coolant system are enclosed in a containment building. REACTOR COOLANT SYSTEM AND MODERATOR SYSTEM REACTOR PRESSURE VESSEL 2 - STEAM GENERATORS 3 - REACTOR COOLANT PUMPS 4 - PRESSURIZER 5 - MODERATOR PUMPS 6 - MODERATOR COOLERS 7 - EMERGENCY COOLING SYSTEM INLET 8 - PRESSURE AND INVENTORY CONTROL SYSTEM 9 - SHUTDOWN COOLING SYSTEM (MODERATOR) 10 - SECONDARY SIDE SAFETY VALVES 11 - PRESSURIZER RELIEF TANK 12 - PRIMARY SIDE SAFETY VALVES 13 - SECONDARY INLET LIGHT WATER 14 - RESIDUAL HEAT REMOVAL SYSTEM 15 - SERVICE COOLING WATER SYSTEM FOR PLANT SECURED 726

4 REACTOR REACTOR PRESSURE VESSEL AND INTERNALS (1) Reactor Pressure Vessel. (2) Lower Filler Pieces. (3) Moderator Tank. (4) Moderator Piping. (5) Coolant Channel. (6) Control Rod Guide Tube. (7) Boron Inlet Nozzle. (8) Heavy Water Outlet Nozzle. (9) Moderator Tank Closure Head. (10) Reactor Pressure Vessel Closure Head. (11) Upper Filler Pieces. (12) In Core-Instrumentation Nozzle. (13) Heavy Water Inlet Nozzle. The reactor pressure vessel (RPV) and the reactor coolant system are connected by the nozzles number 8, and 13. The reactor coolant flows inside RPV by the nozzle 13, in a downwards direction, between RPV and Moderator Tank. Thus RPV and Moderator Tank form the annulus for the in - flowing coolant. The reactor coolant reaches the lower plenum and flows inside the coolant channels in an upward direction through the 253 channels where the fuel assembly is located. The lower filler pieces and the upper filler pieces are provided in the RPV in order to reduce the heavy water inventory required. The upper filler pieces constitute the biological shield during shutdown of reactor. The control rod is moved into the tube guide 6. The tube guide is arranged slanting to make possible the refuelling procedure. The reactor contains 24 black (absorbers made of hafnium) and 5 grey steel control rods. One of the main design features of the RPV in Atucha -I is the moderator tank. The moderator system consists of two identical loops operating in parallel. Each loop comprises a moderator cooler, a moderator pump, and the interconnecting piping with valves. The moderator system performs various functions depending on the operating mode of the reactor 1- During normal operation the moderator system maintains the moderator at lower temperature than that of the reactor coolant. 2- During shutdown of the reactor, the moderator system is switched over to the residual heat removal position by means of the moderator valves. 3- During emergency core cooling the moderator serves a high-pressure core re-flooding and coolant system. The emergency core cooling position is similar to that of the residual heat 727

5 removal, but additionally, water is injected into the hot legs of the reactor coolant loops and into the upper plenum of the RPV. The residual heat removal chain connected to the moderator coolers during emergency core cooling is the same as during residual heat removal. All systems of the residual heat removal chain are of a consistent two - loop desing. THE PRESSURIZER The pressurizer system is connected to one reactor coolant loop and basically comprises the pressurizer with the electric heaters, the surge line, the spray line with the safety valves. The function of the pressurizer system is to maintain the appropriate pressure in the reactor coolant system in order to prevent boiling of the coolant under all operating conditions (principle of the pressurized reactor), and to avoid or limit the pressure variations caused by volume fluctuations during load changes. The pressurizer is partly filled with satured water and partly with steam. Its available volume is 40 m 3. It has three relief and safety valves, whose set point is 125 kgf/cm 2 for one of them and 132 kgf/cm 2 for the other two. NODALIZATION SCHEME FOR THE REACTOR COOLANT SYSTEM WITH A MELCOR General Nodalization of the RCS with the Control Volume Package (CVH). 728

6 The basic plant nodalization can be seen in the figure, concerning the control volume hydrodynamics package (CVH) of the pressure vessel and reactor coolant system. The pressure vessel was divided into nine volumes, three of this volumes belong to moderator system (CV140, 141, 142), and the other six represented the downcomer (CV111,112,113), lower plenum (CV120), core (CV130), upper plenum (CV150) as well as the necessary communications among them, with positive flows, in accordance with the usual directions in non accidental situations. THE MODERATOR SYSTEM In spite of the fact the moderator system is practically independent we must take into account that the reactor coolant system and the moderator system are connected by the pressure equalization openings of the moderator tank closure head. Therefore in a first phase of the nodalization the moderator system has been considered of independent manner from another control volume. In a second phase of the nodalization, the moderator system will be linked to control volume number 150 which represents the upper plenum. In this case, the area for the flow path will be the real area, between moderator tank and upper plenum. At the moment, the moderator tank has been divided in one volume. The volumes 140 and 142 are representing part of the inlet/outlet moderator system pipe. Four tabular functions allow the moderator inlet/outlet simulate with its respective flows and temperature. THE PRESSURIZER It was modelated as a unique volume. This volume is connected from its inferior part to one of the hot legs through a flow path. The coolant volume in the surge line is added to the pressurizer volume. A flow path links the pressurizer with quench tank. The break of the protection membrane of the quench tank, is produced at 5 kgf/cm 2. In this case the coolant falls into sumps of the containment. THE STEAM GENERATOR The primary side of the steam generator has been divided in 5 control volumes, two of them are representing the collectors cameras of inlet and outlet and the others three are representing the U - tube where the primary coolant flow. By the secondary side, three control volumes have been considered, one of them is representing pipe of feedwater part and the third volume is representing the steam collector. The models is very similar to the moderator system model, but in this case two tabular function have been considered in order to simulate the inlet of mass and energy, and others two tabular function which are representing the steam collector and so these functions are simulating the extraction. In the case of simulating a Station Blackout sequence the function which is representing the feedwater injection must be stopped when the electric energy is lost. NODALIZATION SCHEME FOR THE CORE PACKAGE The figure displays the nodalization required by the COR package. It can be observed a disposition in three radial rings and fourteen axial levels, implicating two hydrodynamic volumes CV120 (lower plenum) and CV130 (active zone). 729

7 Core Nodalization One level is representing the lower plenum, which is composed by a unique material (carbon steel with austenitic platinum plating) corresponding to the filler piece. Other level is representing to the metallic plate which serves as the lower fixing level for the coolant channels and the control rod guide tubes. Due to this plate, the coolant flow is divided in two parts, one of them flow inside the coolant channel in an upward direction and the other part flows under this plate. The other twelve axial levels are representing the moderator tank height. The moderator tank accommodates all coolant channels and each of them contains one fuel bundle column. The fuel assemblies are bundles of 36 closely packed fuel rod which are arranged in 4 concentric rings having 1, 6, 12 and 18 fuel rods each, plus an additional structural rod located in the external ring. Each fuel rod consists of a stack of uranium dioxide pellets enclosed by a thin walled zircaloy 4 canning tube. Each fuel assembly, together with the filler body and the closure plug, forms the fuel bundle column. Therefore, the axial levels three and fourteen contains inside its corresponding cell only zircaloy structures but the cells of the rest of the axial level which are representing the active zone are composed by uranium dioxide and zircaloy of the fuel cladding. Then, a group of the geometric parameter has been annexed, such as outer radio of the uranium dioxide pellets and gap between fuel and cladding. In the case of the others structures, has been annexed the thickness corresponding to the filler pieces, the metallic plate in the bottom of the moderator tank and the corresponding to the guide tube of the control rods. The thickness for the lower head has been declared too. 730

8 The particular configuration of the control rods in Atucha which are made of hafnium is not considered by MELCOR which permits only two option B4C for the BWRs and Ag-In -Cd for the PWRs. Up to now the control rods have not been considered because its mass is negligible. HEAT STRUCTURES FOR THE MODERATOR TANK One of main problems for models the Atucha -I reactor is the great mass of heavy water inside of the moderator tank. In this case and taking into account in normal operating exist a thermal transfer from channels to moderator tank, a thermal transfer through a heat structure located between hydrodynamic volume which is representing to the core and the hydrodynamic volume which is representing to the moderator tank was considered. Heat Structure of one channel related to axial levels of the core model, COR package. 731

9 HEAT STRUCTURES OF THE REACTOR PRESSURE VESSEL In this case the heat structures have been declared taking into account that materials considered into core has not been considered as heat structure in order to avoid overestimating the mass of the corresponding material. For the structures which are representing the outer wall of the reactor vessel, the option of a insulated boundary condition, was applied. Some many heat structures corresponding to reactor pressure vessel of CNA I. 732

10 RELATIONSHIP BETWEEN COR AND HEAT STRUCTURE PACKAGES In the record CORZjj02 the user has to declare two fields, one of them is the field IHSA which specifies the number corresponding to the heat structure located in the outer radial boundary for an axial level determined. Thereinafter the data are introduced through heat structure package. In the case of the PWR the heat structure located in the outer radial boundary is the corresponding to the baffle. In the case of the Atucha-I are the walls of the 16 channels arranged in the outer ring have been considered as heat structures to be introduced in the field IHSA. Structures between fuel and reactor vessel, 1- fuel, 2- gap, 3-cladding, 4- channel wall, 5- part of the moderator, 6- foils, 7- moderator, 8- wall of the moderator tank, 9- downcomer, 10- reactor vessel wall. HEAT STRUCTURES FOR STEAM GENERATOR, PRESSURIZER AND PRIMARY PIPES In this cases the heat structures have been modeled in the usual mode. At the moment the model considers only a loop but the flow path areas and volume have been duplicated. BUILDING AND STRUCTURES ARRANGEMENTS The Reactor Building, the Reactor Auxiliary Building and the Fuel Storage Building constitute the Controlled Area in which all systems assigned to the nuclear section are installed. The rest of the buildings are located in the Conventional Section of the nuclear power plant. The engineered safety features such as containment SPRAY system or Hydrogen Control System are not present in CNA I. 733

11 The Reactor Building contains the reactor, the reactor coolant system, the moderator system and associated equipment. REACTOR BUILDING Double containment type (KWU) REACTOR BUILDING 1- Reactor pressure vessel 2 - Steam generator 3 - Reactor coolant pump 4 - Pressurizer 5 - Moderator cooler 6 - Refueling machine travelling gear 7 - Refueling machine 8 - Tilter 9 - Fuel transfer tube 10 - Containment 11 - Reactor building 12 - Annulus Spherical Steel Shell (inside, designed to internal pressure). Compartments: Steam Generator Reactor Cavity Fuelling Machine Pressurizer Concrete Shell (outside) It is formed by a cylindrical reinforced concrete shield with a hemispherical top enclosure and is founded on a base slab. Its available volume is m 3 734

12 CONTAINMENT INPUT DESK Arrangement: CVH = 18 FL = 32 HS = 64 NODALIZATION SCHEME FOR THE REACTOR BUILDING The pictures 3, 4 and 5 show the basic reactor building nodalization, concerning the control volume hydrodynamics package (CVH). The heat structures represent different walls and floors. The spherical steel shell has been divided in a cylindrical wall and one hemispherical dome over it. The concrete shell has been divided in a cylindrical wall and one hemispherical dome over it. Some flow paths are always open but other of them will open up at differential pressures value such as 0.05 MPa, 0.08MPa, In order to simulate the break of the containment two flow paths, which will open at determined pressure (6 atm), have been considered. Pictures 6, 7 and 8 show some Flow Paths schematic representation. FUTURE ACTIVITIES The future activities will be focuses in the following points: Improve the capability of the proposed model. Apply this model to analyze sequences, which have been identified, by CNA I PSA, as more important accidental sequences for the risk. Analysis hydrogen production during in vessel phase and Corium Concrete Interaction. Analysis hydrogen behavior into containment building (Containment failure characteristics). Evaluate the possibility of application mitigation technique, such as igniters, recombiners or combination of these mitigation devices (Dual concept) for the control of hydrogen. Analysis radionuclide behaviours. To determine the Source Term for the accidental sequences. 735

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18 RESULTS OF THE STABILIZATION FOR s OF THE ACCIDENT TIME 741

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