Journal of Power and Vol. 2, No. 1, 2008 Energy Systems Development of RBMK-1500 Model for BDBA Analysis Using RELAP/SCDAPSIM Code*

Eugenijus USPURAS** and Algirdas KALIATKA** **Lithuanian Energy Institute 3 Breslaujos, Kaunas LT-44403, Lithuania E-mail: [email protected]

Abstract This article discusses the specificity of RBMK (channel type, boiling water, graphite moderated) reactors and problems of Reactor Cooling System modelling employing computer codes. The article presents, how the RELAP/SCDAPSIM code, which is originally designed for modelling of accidents in vessel type reactors, is fit to simulate the phenomena in the RBMK reactor core and RCS in case of Beyond Design Basis Accidents. For this reason, use of two RELAP/SCDAPSIM models is recommended. First model with described complete geometry of RCS is recommended for analysis of initial phase of accident. The calculations results, received using this model, are used as boundary conditions in simplified model for simulation of later phases of severe accidents. The simplified model was tested comparing results of simulation performed using RELAP5 and RELAP/SCDAPSIM codes. As the typical example of BDBA, large break LOCA in reactor cooling system with failure of emergency core cooling system was analyzed. Use of developed models allows to receive behaviour of thermal-hydraulic parameters, temperatures of core components, amount of generated hydrogen due to steam-zirconium reaction. These parameters will be used as input for other special codes, designed for analysis of processes in reactor containment.

Key words: RBMK-1500, Analysis of Beyond Design Basis Accidents, RELAP/SCDAPSIM

1. Introduction

Up to now the phenomena that could occur in case of a severe accident in RBMK reactors have not been analyzed in detail and there is no sufficient amount of literature available on this topic. On the other hand, the understanding of these phenomena would assist to develop strategies for the management of accidents. The employment of codes, like RELAP5/SCDAP, ATHLET-CD, ASTEC, etc, for analysis of BDBA in RBMK is not straightforward, because these codes are developed for analysis of processes in vessel type reactors. Here the discussion on challenges, which arise when RELAP/SCDAPSIM code is used for modelling of processes in channel-type RBMK reactors, is presented. These challenges are related with specificity of RBMK design. The analysis of typical example of BDBA in case of large break LOCA in reactor cooling system with failure of emergency core cooling system was performed for RBMK-1500 reactor of Ignalina NPP. Ignalina NPP is the only plant in Lithuania, which consists of two units, commissioned in December 1983 and August 1987. Both units are equipped with channel-type graphite-moderated reactors RBMK-1500. Unit 1 of Ignalina NPP was *Received 22 Aug., 2007 (No. 07-0385) shutdown for decommissioning at the end of 2004. [DOI: 10.1299/jpes.2.133]

133 Journal of Power and Vol. 2, No. 1, 2008 Energy Systems Abbreviations

ALS Accident Localization System BDBA Beyond Design Basis Accident BWR DS Drum Separator ECCS Emergency Core Cooling System FC Fuel Channel GDH Group Distribution Header LEI Lithuanian Energy Institute LOCA Loss-of-Coolant Accident LWR Light Water Reactor MCP Main Cooling Pump MSV Main Safety Valve NPP PWR Pressurized Water Reactor RBMK Russian abbreviation for “Large-power channel-type reactor” RCS Reactor Cooling System SDV-A Steam Discharge Valves to accident localization system SDV-C Steam Discharge Valves to Turbine Condensers SDV-D Steam Discharge Valves to Deaerators and to In-house Needs 2. Specificity of RBMK-type reactors

Reactor RBMK-1500 of Ignalina NPP is a boiling light water reactor with graphite moderator (1). The nominal thermal reactor power is 4800 MW, while electrical power is 1500 MW. After the Chernobyl accident the maximal allowed thermal power is set to 4200 MW i.e. Ignalina NPP is operating at the power below its nominal. Several important design features of RBMK-1500 are unique and extremely complex in respect to vessel type reactors: the fuel assemblies are loaded into individual channels rather than a single pressure vessel; the plant can be refuelled on-line; the neutron spectrum is thermalized by a massive graphite moderator. The specific features of the Ignalina NPP with RBMK-1500 reactor, compared with the vessel-type LWR, are described by Uspuras (2) and here are only shortly summarized. The schematic view of reactor cooling system and surrounding compartments, showing the location of the reactor core and its main components is presented in Fig. 1. Currently at Ignalina NPP the fuel of 2.4 %, 2.6 % and 2.8% enrichment with burnable erbium absorber is used. The fuel cladding is made of zirconium and niobium alloy. RBMK fuel assembly consists of two fuel bundles placed one above another. Each fuel assembly includes 18 fuel rods placed in two circles around the carrying rod. Each fuel assembly is placed in a separate fuel channel. The RBMK-1500 reactor has 1661 fuel channels as 1661 small reactor cores. The maximal power of single fuel channel is 3.75 MW, average – 2.53 MW. Due to such separation of reactor core the failure of fuel in one FC, e.g. as a result of loss of coolant, is not as dangerous as degradation of total reactor core in vessel type reactors. The fuel channels in RBMK-1500 are placed in a graphite stack, which is located in a reactor cavity filled with a helium and nitrogen mixture. Reactor cavity performs a function of confinement. In the reactor there are 1700 tons of graphite and in case of accident it could absorb decay heat from fuel. The fuel channel of RBMK reactor according to its function and location corresponds to a pressure vessel of vessel-type reactors. However, FC is not as strong as a vessel. If the FC wall heats up while the internal pressure is elevated, it may expand in radial direction (i.e., it may balloon) over the heated length until it contacts the graphite block(s). The constant temperature of 650°C may be taken as a conservative value of the acceptance

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criterion for the pressure range from 4 to 8 MPa (3). At pressures below 4.0 MPa, graphite blocks prevent tubes deforming to the point of rupture (3). The recent experiments performed in Russia showed that rupture of a single channel is not expected to cause a consequential rupture of the other fuel channels (4). Such design of channel type reactors allows to change the fuel assemblies on-line. This online refuelling puts specifics to the accidents during refuelling but the integral reactor core characteristic remains almost constant during the reactor operation. Thus, in RBMK reactors the effect of integral reactivity dependence on fuel burn-up is minimized. Comparing the RBMK with the vessel-type (BWR) reactors it is seen that these reactor types are quite similar in power per fuel quantity or fuel rod length (1). However, the specific power per core volume of the RBMK reactors is somewhat less, while the core heat capacity in RBMK due to large amount of graphite is high. These parameters have a certain impact on the operation of the reactor during accidents. Typical time of power increase due to reactivity initiated accidents in RBMK’s is being measured in seconds rather than in microseconds as for other LWR (3).

Fig.1 General view of NPP with RBMK reactor: 1 – graphite stack, 2 – fuel channel, 3–lower water piping, 4 – group distribution header, 5 – emergency core cooling pipes, 6 – MCP pressure pipes, 7 – main circulation pump, 8 – MCP suction pipes, 9 – pressure header, 10 – bypass pipes, 11 – suction header, 12 – downcomers, 13 – steam and water pipes, 14–steamlines, 15 – refuelling machine, 16 – drum separator

The control rods are made of boron carbide B4C and aluminium and are placed in the individual channels, which are independent from the channels with fuel bundles. The channels with control rods are cooled with water supplied by the system totally independent from the reactor cooling system. The pressure in channels with control rods is always close to atmospheric. The reactor coolant system consists of two loops, each having flow length of more than 200 m. Each loop has 830 vertical parallel fuel channels and numerous components, such as headers, pumps, valves, etc. Both RCS loops are interconnected via the steamlines and do not have a connection on the water part. This is different from vessel-type reactors where all fuel rods are placed in a single vessel. The total mass of water in the RCS is ~2000 tons. This mass related to 1 MW of power in RBMK-1500 (as well as accumulated energy there) is relatively larger in comparison to any other vessel-type reactor. Therefore in case of LOCA the pressure in the reactor cooling system of RBMK reduces slower than in RCS of vessel-type reactors.

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The RBMK-1500 is equipped with specific system of reinforced compartments, which is named Accident Localization System. This system is a pressure suppression type confinement with rather complex geometry (1). The specific feature of ALS is that in case of an accident the clean air, which fills the wetwell, is released to the environment and then the ALS is isolated. During this period the air from the drywell is pushed to wetwell and trapped there. The gas delay chamber in ALS consists of several sections sometimes called “labyrinth” to provide a long path for transport of fission products. The long path ensures better deposition of aerosols due to natural deposition processes. There is a special system for the cooling of the condensing pools. After initiation it simultaneously supplies the water to the sprays located in wetwell. As well in case of accident the operator has a possibility to connect another branch of this system by opening the valve and supply water to sprays located in the drywell. The total volume of ALS is ~48000 m3, from which ~20000 m3 is the drywell and ~28000 m3 is the wetwell. Most of the phenomena, which might occur during the severe accident progression in RBMK-1500 reactor are the same as for any other LWR. In case of loss of coolant and uncovering of fuel assemblies the process of heat up and relocation of fuel rods starts. The graphite stack and reactor cavity remains intact if the pressure in FCs is below 4 MPa. From there it follows that the high pressure melt ejection as well as direct containment heating, the phenomena very important for vessel-type reactors are not possible in RBMK. However, there are some specific phenomena, such as the interaction between fuel assemblies and fuel channel tubes, rupture of FCs, motion, fracture of graphite blocks, coolant – graphite interaction, melting of aluminium or aluminium oxide, materials which are included to control rod structure. 3. Development of RBMK-1500 MODEL using RELAP/SCDAPSIM code

In order to investigate the severe accident phenomena in detail, at first the design of the reactor has to be well understood and then the appropriate computer codes have to be used. For the modelling of processes in reactor cooling system during DBA the RELAP5 and ATHLET codes have been used in Lithuanian Energy Institute since 1993. These codes, designed mostly for the analysis of thermal-hydraulic processes in vessel-type reactors, have been adapted to simulate RBMK reactors. During the code adaptation validation process (5) (6) it was shown that not only common phenomena for all type of reactors (Heat transfer in fuel assemblies prior to onset, during ant post CHF; Counter-current flow; MCP behaviour in one and two-phase conditions; Mixture level and steam separation in big volumes; Coolant blowdown through break and etc.) but also the specific phenomena for channel-type RBMK reactors (Flow instability in parallel steam generating channels; Radiation heat transfer between fuel assemblies and fuel channel tubes; Natural circulation development and degradation; Water entrainment from DS to steamlines and etc.) are modelled adequately in the developed RBMK-1500 model. However, in case of BDBA the specific phenomena related to fuel behaviour at high temperatures (Pellet cracking, pellet deformation, fuel relocation, melting; Cladding deformation due to collapse and or ballooning; Change of gap conductance; Wall thinning due to oxidation; Rapid and slow fission-gas driven fuel swelling; Cladding oxidation; Fuel–water/steam interaction and etc.) should be modelled. Since significant experience of usage of RELAP5 code was accumulated in LEI, the RELAP/SCDAPSIM/MOD3.2 code was selected for the analysis of processes in RBMK-1500 reactor and cooling system during BDBA. The RELAP/SCDAPSIM code (7) is an integrated, mechanistic computer code, which models the progression of severe accidents in light-water-reactor nuclear power plants. The entire spectrum of in-vessel severe accident phenomena, including reactor-coolant-system thermal-hydraulic response, core heatup, degradation and relocation, and lower-head thermal loads, are treated in this code in a unified framework for both boiling water reactors

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and pressurized water reactors. The RELAP/SCDAPSIM is being developed at Innovative Systems Software as part of the International SCDAP Development and Training Program. RELAP/SCDAPSIM/MOD3.2 uses the publicly available SCDAP/RELAP5/MOD3.2 (8) models, developed by the US Nuclear Regulatory Commission, in combination with proprietary advanced programming and numerical techniques, additional user options and advanced models. The application of RELAP/SCDAPSIM looks tempting, because thermal-hydraulic system in this code is modelled using RELAP5 part, and only fuel rods are modelled using SCDAP part. Unfortunately, the application of RELAP/SCDAPSIM code (7) for RBMK reactor is not straightforward. SCDAPSIM includes limitation that only one reactor core can be defined. Considering that each fuel channel of RBMK corresponds to pressure vessel, it is not possible to simulate multiple FCs in detail. As it was mentioned before, the fuel with erbium is used at Ignalina NPP. SCDAPSIM gives a possibility to define PWR or BWR fuel bundles with the user-defined fuel enrichment, but does not allow to include a plant specific core content. This might have an impact on the results. The performed analysis does not include the fission product release calculations. Also, usage of the RELAP/SCDAPSIM code does not enable to model the specific RBMK phenomena (Thermo mechanical interaction of the fuel bundle with the FC tube; Deformation and rupture of FC tubes; Deformation, motion, and fracture of graphite blocks; Bending of adjacent FC tubes due to forces imposed by the displacement of graphite bricks; Interaction of steam–water mixture with graphite blocks and metal structures) because this code, as well as other codes used for severe accident analysis around the world (MELCOR, ATHLET-CD, ASTEC, etc.), are developed mostly for the analysis of the vessel-type reactors (e.g. PWR and BWR). It should be mentioned that these limitations of RELAP/SCDAPSIM code were taken into account during the modelling. The general nodalization scheme of the RELAP/SCDAPSIM model is presented in Fig. 2. The model of the RCS consists of two loops, each of which corresponds to one loop of the actual circuit. Two DS in each RCS loop are modelled by generalized “separator” element (1).

SDV-A MSV I-III SDV-A MSV I-III

1 To Turbines and SDV-C 15 2 Feed- SDV-D SDV-D Feed- water water

14 13 18 3 Reactor 16 78 12 17

4

From 11 From 5 ECCS ECCS

6910

Fig.2 Ignalina NPP model (complete model) nodalization diagram: 1 - DS, 2 - downcomers, 3 - MCP Suction Header, 4 - MCP suction piping, 5 - MCPs, 6 - MCP pressure piping, 7-bypass line, 8 - MCP Pressure Header, 9 - GDHs, 10 - lower water piping, 11 - reactor core inlet piping, 12 - reactor core piping, 13 - reactor core outlet piping, 14 – steam and water piping, 15 - steamline, 16 – artificial elements for modelling of core slumping and bypass, 17 - valve for break modelling, 18 – model of compartments, which surround the RCS pipelines

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All downcomers are represented by a single equivalent pipe (2), further subdivided into a number of control volumes. The pump suction header (3) and the pump pressure header (8) are represented as RELAP5 “branch” (7) (8) elements. Three operating MCPs are represented by one equivalent “pump” element (5) with check and throttling-regulating valves. The throttling-regulating valves are used for coolant flow rate regulation through the core. These valves are modelled by employing “servo valve” elements. The normalized flow area versus normalized stem position is described in the RELAP5 model. The bypass line (7) between the pump suction header and the pump pressure header is modelled with the manual valves closed. This is in agreement with a modification performed at the Ignalina NPP. All fuel channels of the left core pass (this reactor cooling loop is assumed as intact) are represented by a few equivalent channels (12) operating at specific power and coolant flow. The group of 20 distribution headers (9) with connecting pipelines is modelled by RELAP5 “branch” component. The pipelines of the water communications (10) are connected to each FC. Each of these components represents the quantity of pipes appropriate to the number of elements in the corresponding FC in the core. The vertical parts of the FC (13) above the reactor core are represented by RELAP5 components “pipes”. The pipelines of the steam-water communications (14) are connecting the fuel channels with DS. The right reactor cooling loop is assumed as affected (the loop where the rupture occurs). The steam separated in the separators is directed to turbines via steamlines (15). Two turbine control valves regulate steam supply to the turbines. The control of these valves was modelled by “servo valve” elements based on the algorithm of steam pressure regulators used at Ignalina NPP. There are four steam discharge valves (SDV-C) in each loop of the RCS to direct the steam to the condensers of the turbines. The pressure of the steam is also controlled, and peaks of pressure are eliminated by two steam discharge valve (SDV-A) and 12 main safety valves (MSV), discharged the steam to pressure suppression pool of the ALS tower. The model also takes into consideration steam mass flow rate through the steam discharge valve to the deaerator (SDV-D) for in-house needs. All models of steam discharge valves are connected to the “time dependent” elements, which define boundary conditions in turbine condensers or ALS pressure suppression pool. The feedwater injection into the DS is simulated explicitly using RELAP5 “pipe”, “junction”, “volume” and “pump” elements (not presented here). For the modelling of breaks in pipelines of reactor cooling system, the “valve” element (17) is used. This valve is connected to the volume (18), which represents the compartments covered reactor cooling system pipelines. The reactor core is described using SCDAP elements. The fuel channels with fuel assemblies are divided into a few (depending on the needs of modelling) equivalent groups according to power and coolant flow rate values. For the modelling of fuel the “BWR fuel” elements are used in SCDAP part. Fuel channels and square profile 0.25 x 0.25 m graphite blocks are modelled by cylindrical “shroud model” with the equivalent perimeter of inner shroud surface and thickness. Each core structure is modelled using 16 axial nodes of 0.5 m length each. The fuel rod is modelled using eight radial nodes, five to represent the fuel pellet, one for the gap region and two for the cladding. The fuel channels and graphite columns are modelled also using eight radial nodes. Two of these radial nodes are used for the fuel channel wall, two for the gap and graphite rings region and four for the graphite column. For the description of material properties (specific heat, thermal conductivity, density) of the gap and graphite rings region and of the graphite column, the user specified material properties cards are used in SCDAP part. The last layer in “shroud” element, which models the fuel channel, is described as the zirconium tube. Such approach allows to model the coolant and fuel channel (which is made from of zirconium with 2.5% Niobium alloy) interaction process. It is assumed that approximately 95% of generated fission and decay power is generated in the fuel, and 5% in the graphite stack. The philosophy of RELAP/SCDAPSIM code is that there is one single reactor core

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with a few equivalent fuel assemblies and one lower head, where the melted core material is collected. Unfortunately, it is not the case for RBMK reactor. Because each fuel assembly is placed inside an individual fuel channel, in case of core overheating, the fuel will be melting within the boundary of their fuel channel. To avoid this RELAP/SCDAPSIM code limitation, the artificial channel for core bypass was added in the RBMK-1500 model (16, see Fig. 2). This artificial channel with junctions and volumes (described by “pipe”, “junctions” and “time dependent volumes”) was used for describing conditions for radial spreading of core melt (it was mentioned that this phenomenon is not present in RBMK) and receiving any slumped core material in the SCDAP model. The developed RBMK-1500 model allows to analyze the processes in reactor cooling system in case of BDBA, including overheating and failure of fuel rods in fuel assemblies, steam – zirconium reaction, when steam is contacting the fuel rods and fuel channel tube. This model can be used till fuel melting starts, because the channel blockage phenomenon will not be modelled adequately. Also the modelling of deformation and rupture of fuel channel tubes, as well as deformation, motion, and fracture of graphite blocks are not possible. Thus, only scenarios, where fuel channels remain intact, can be analyzed using this model. For the analysis of later phases of severe accidents, where the melting of fuel and channel blockage starts, the simplified model with two single fuel channels was developed (see Fig. 3). In this simplified RELAP/SCDAPSIM model two separate paths are simulated. The left path consists of one fuel channel (3), where the fuel assemblies, fuel channel and surrounding graphite column are described using RELAP5 heat structure elements. The right path simulates an identical fuel channel (4), where the fuel rods are described by SCDAP element “BWR fuel” and fuel channel with graphite column are described using SCDAP “shroud” element. When RBMK-1500 is operating at maximal allowed thermal power 4200 MW, the power level of separate fuel channels may vary from 0.9 up to 3.75 MW. However, the channels with average power level are in majority. Thus, the elements (3) and (4) simulate single channels of initial average power 2.55 MW. The coolant flow rate in separate channels, pressure in GDH (5) and DS (2) are described by elements, in which the change of pressure, temperature and coolant flow rate are defined as boundary conditions. These boundary conditions are obtained from the calculations with RELAP/SCDAPSIM code using the detailed model of RBMK-1500, described above. The use of two RELAP5 and SCDAP elements with identical boundary conditions allows to observe the influence of specific severe accident phenomena (modelled in SCDAP elements) on behaviour of fuel rods and FC parameters.

Fig.3 Scheme of simplified RBMK-1500 model using RELAP/SCDAPSIM: 1, 2– DS and steamlines; 3 – model of FC, developed using RELAP5 heat structure elements; 4 – model of FC, developed using SCDAP elements; 5 –GDH

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For the approximate assessment of consequences in case of severe accidents in RBMK, we can assume that coolant conditions and fuel behaviour in all fuel channels will be similar. Therefore the results of SCDAPSIM analysis using simplified model (temperatures of core elements, generation of hydrogen) can be extrapolated to the all 1661 fuel channels in the reactor. 4. RELAP/SCDAPSIM model testing

The RELAP/SCDAPSIM was tested by employing a model, which is presented in Fig. 3. The simplified two paths model with identical fuel channels, described using RELAP5 heat structure elements and special SCDAP elements (see Fig. 3) allows to investigate the behaviour of SCDAP elements (fuel and shroud), to gain understanding of limits and uncertainties. The testing was performed using FCs heat-up process analysis. It was assumed that the coolant in fuel channels is lost within the first 200 seconds. It leads to core uncovery and heat-up of core components: fuel pellets, claddings, FC walls and graphite columns. In RBMK the heat is generated not only in the fuel, but also in the graphite – moderator. Initially the temperature of graphite is higher than fuel cladding and FC wall temperatures, because fuel assemblies and fuel channels are cooled by coolant. When the fuel, cladding and channel wall temperatures exceed the graphite temperature, the graphite stack starts work as heat sink. In Fig. 4 the behaviour of temperatures of FC components is presented. As it is shown in presented figure, there is a good agreement between temperatures of heat structures calculated using RELAP5 and SCDAP during the steady state, except fuel pellet temperature. As it is presented in Fig. 4, the peak fuel pellet centerline temperature, calculated using RELAP5 heat structure elements, is a little higher in comparison with the temperature, calculated using SCDAP elements. These inadequacies can be explained that in RELAP5 calculation the real RBMK-1500 data were used (thermal conductivity and volumetric heat capacity of fuel pellets, and cladding, gas composition in the gap between fuel and cladding). In SCDAP calculation, the “BWR-type” fuel rods were assumed. After the beginning of the accident, within the first 600 seconds the behaviour of temperatures, calculated using RELAP5 and SCDAP codes, demonstrated good agreement (see Fig. 4). Later, after 600 seconds at higher temperature level the steam – zirconium reaction starts. Due to this reaction the hydrogen and additional heat are generated. The additional generated heat leads to the significant fuel, cladding and channel wall temperatures increase in fuel channel, which is modelled using SCDAP elements.

1500 1000 a) start of Zr c) C o 800 oxidation 1000 600 start of Zr SCDAP 500 SCDAP oxidation 400 RELAP5 Temperature, RELAP5 0 200 -2000 0 2000 4000 6000 -2000 0 2000 4000 6000

1500 1000 b) d) C o 800 1000 600 500 start of Zr SCDAP SCDAP 400

Temperature, oxidation RELAP5 RELAP5 0 200 -2000 0 2000 4000 6000 -2000 0 2000 4000 6000 Time, s Time, s Fig.4 Behaviour of temperatures of core components in case of fast loss of coolant flow: a) - peak fuel pellet temperature, b) - peak fuel cladding temperature, c) - peak fuel channel tube wall temperature, d) - peak graphite column temperature

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The performed testing of RELAP/SCDAPSIM code and developed model shows that the main phenomena are adequately modelled, and developed models are suitable for analysis of processes in reactor and its cooling system in case of BDBA. 5. Analysis of large break LOCA with ECCS failure

Large break LOCA (break of pipeline with the biggest diameter – MCP pressure header) without pumped water injection to the core was selected for the analysis. Such type of accident leads to the fastest core uncovery and core damage. Nevertheless, all the severe accident relevant phenomena (core uncovery, fuel cladding failure, melting of spacers, etc.) occur and can be analyzed. In the Level 1 Probabilistic Safety Analysis of Ignalina NPP the frequency of such accident sequence is estimated 3.7·10-7 /year. However, considering large diversity and redundancy of the emergency core cooling system at NPP such frequency seems too conservative. Due to failure of pumped injection only ~180 m3 of water from ECCS hydro-accumulators could be injected to the reactor core. A few minutes after beginning of the accident the reactor core components start to heat up. However, the graphite, which is a moderator in RBMK-1500 reactor, is an efficient heat sink and slows down the heat up of the core. Power density in RBMK reactors is less compared to vessel-type reactors, therefore the heat-up of RBMK reactor is slower compared to other light water reactors. The analysis was performed using both (complete and simplified) models. At first the RBMK-1500 model with full reactor cooling system, presented in Fig. 2, was used. The process starts at time moment t = 0.0, when the break valve in right (affected) RCS loop (element 17 in Fig. 2) opens. The break flow area is two times higher as cross-section of MCP pressure header and it is equal 1.27 m2. The steam-water mixture from the RCS is discharged into the volume (18), which models ALS. The flow rate of discharged coolant is presented in Fig. 5. Before the accident the water is injected from MCP pressure header through GDH into fuel channels. Initially the pressure in GDH is higher than pressure in DS. After the break, within the first seconds the pressure in broken MCP header decreases down to atmospheric. The coolant in pipelines, connecting GDH’s changes the flow direction, which leads to the closure of GDH check valve. After the close of check valves the coolant through the break is discharged only from MCP side. The reactor scram system is activated due to pressure increase in reinforced compartments of ALS and all control rods are inserted within 12 – 14 seconds. The decay heat was calculated using SCALE code, taking into account the average condition of reactor core – fuel 2.6 % enrichment, containing 0.5% of Erbium as burnable poison power, and 15 MW days/kg burnup (see Fig. 6).

4000 30000 W s 3000 Reactor scram system 20000 actuation 2000

Flow rate, rate, kg/ Flow 10000 1000 Reactor power, M

0 0 1E-06 0 1E-05 0.0001 0.001 0.01 0.1 1 10 1E-06 0 1E-05 0.0001 0.001 0.01 0.1 1 10 Time, h Time, h

Fig.5 Discharge of coolant through the break Fig.6 Reactor power behaviour

Due to large break LOCA the pressure in RCS decreases (see Fig. 7) and water boils in all primary circuit. The ECCS pumps are failed, and only a small amount of water-steam

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mixture is supplied into the core due to fast boiling of water in reactor cooling system. This phenomenon and the supply of water from hydro-accumulators allow to cool the core components for a short time span. As it is presented in Fig. 8, after approximately 200–300 seconds. there is no water supply in to the fuel channels of affected RCS loop. Thus, the failure of all long-term ECCS subsystems will lead to the reactor core overheating in long-term. The behaviour of fuel, cladding, FC wall and graphite temperatures for fuel channels in both RCS loops is presented in Figs 9, 10. It is necessary to note that the analysis was performed for the case when water supply for cooling of reactor control rods is terminated at the beginning of the accident. Such assumption leads to more conservative results. If the circuit for reactor control rods cooling remains available, the water of this circuit removes some part of heat from the core. In this case the reactor core heat-up process will be slow down. As it is seen from presented figures, the heat-up process of fuel pellets, claddings, FC walls and graphite columns in affected RCS loop starts earlier, comparing to intact loop. It is because the fuel channels in intact RCS loop are longer time period cooled by water (see Fig. 8). The heat-up of fuel starts after approximately 15 – 20 minutes in the affected RCS loop. After approximately 40 minutes from beginning of the accident the coolant supply to fuel channels of intact loop is terminated. Thus, the heat-up process in intact loop starts approximately 30 minutes later, comparing with affected loop.

channel from intact loop 10 DS-affected GDH-intact channel from affected loop 8 GDH-affected 6 s Water supply from ECCS 6 Closure of GDH 4 hydroacumulators 4 check valve

Pressure, MPa 2 Flow rate,Flow kg/ 2 Start of MCP pressure header break 0 0 1E-06 0 1E-05 0.0001 0.001 0.01 0.1 1 10 1E-06 0 1E-05 0.0001 0.001 0.01 0.1 1 10 Time, h Time, h

Fig.7 Pressure behaviour in DS and GDH of Fig.8 Supply of water in the single fuel intact and affected RCS loops channel of affected and intact RCS loops 1200 1200 Affected RCS loop Affected RCS loop 1000 1000 C C o o 800 800 600 600 Intact RCS loop 400 400 Intact RCS loop Temperature, Temperature, Temperature, 200 fuel 200 FC wall cladding graphite 0 0 -2 0 2 4 6 8 -202468 Time, h Time, h

Fig.9 Behaviour of peak fuel and cladding Fig.10 Behaviour of peak FC wall and temperatures in affected and intact loop, graphite column temperatures in affected and calculated using complete RCS model intact loop, calculated using complete RCS model

As it is shown in Figs 9, 10, 8 hours after the beginning of the accident the temperatures of reactor core components in both RCS loops are close together. Thus, for more conservative estimation, it is sufficient to analyze the processes only in one channel with average power, assuming that coolant conditions and fuel behaviour in all fuel channels will be similar. This assumption allows to use a simplified RELAP/SCDAPSIM model. The results of analysis later can be extrapolated to the all 1661 fuel channels in the reactor. In Figs 11 – 13 the parameters, calculated using both (simplified and complete RCS)

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models are presented. The initial conditions: (1) coolant flow rate at FC inlet, (2) pressure in GDH and DS, (3) coolant temperatures in GDH and DS, were assumed from complete RCS model calculations. Eight hours after the beginning of the accident the coolant flow rate through the fuel channels is equal to zero, pressure in GDH and DS – atmospheric. The conditions were averaged for intact and affected loops. This explains why the heat-up process starts a little later in calculations using simplified model (see Figs 11, 12). 8 hours after the beginning of the accident, the temperatures of core components are higher in simplified model. This is because zero coolant flow rate was assumed in a simplified model during a long time span. In complete RCS model some very small oscillations of coolant flow rate are observed. This leads to softer conditions in complete model for a long period of time (after 8 hours).

1800 1800 Full model 1500 1500 C C o o 1200 1200 Full model Simplified model Simplified model 900 900 600 600 Temperature, Temperature,

fuel Temperature, 300 300 FC wall cladding graphite 0 0 -2 0 2 4 6 8 10 12 14 16 18 20 22 24 26 -20 2 4 6 8 101214161820222426 Time, h Time, h Fig.11 Behaviour of peak fuel and cladding Fig.12 Behaviour of peak FC wall and temperatures calculated using simplified RCS graphite column temperatures calculated and complete (for affected loop) models using simplified RCS and complete (for affected loop) models

The simulation using simplified model (see Fig. 3) runs approximately 10 times faster, comparing with complete RCS model (see Fig. 2). The simplified model allows to simulate the hydrogen generation for a long period of time (Fig. 13). As it is presented in Fig. 13, approximately 0.8 kg of hydrogen is generated in one single fuel channel during 24 hours. After 24 hours from beginning of accident the hydrogen generation is terminated, because of the absence of steam in fuel channels. It is necessary to note that the heat removal by circuit of reactor control rods cooling was not taken into account.

0.000030 1

a) g b) 0.8 s 0.000020 0.6

0.4 0.000010 Flow rate, kg/ 0.2 Amount of hydrogen, k hydrogen, of Amount

0.000000 0 0 5 10 15 20 25 30 0 5 10 15 20 25 30 Time, h Time, h

Fig.13 Flow rate (a) and total amount (b) of generated hydrogen in one single fuel channel with average power, calculated using simplified RCS model 6. Conclusions

The article presents the specific and BDBA phenomena of RBMK reactors and approaches for modelling of these phenomena using RELAP/SCDAPSIM code. To avoid RELAP/SCDAPSIM code limitations, use of two models is recommended. First, complete RCS model with artificial channel for core bypass during radial spreading of core melt and artificial volume for receiving any slumped core material, allows to analyze the processes in reactor cooling system in case of BDBA, including overheating and failure of fuel in fuel assemblies, steam–zirconium reaction, when steam is contacting the fuel rods

143 Journal of Power and Vol. 2, No. 1, 2008 Energy Systems

and fuel channel tube. This model can be used till the beginning of fuel melting, because the channel blockage phenomenon will be modelled not adequately. Second, the simplified single fuel channel model should be used for the analysis of later phases of severe accidents, where the melting of fuel and channel blockage starts. At the moment, developed model does not enable to simulate the core melt relocation and cooling ability of slumped debris on the lower metal structure below the reactor core. The application of special SCDAPSIM model for this simulation is foreseen in the future. The boundary conditions for the simplified model are estimated from the calculations with complete (detailed) model of RBMK-1500. Assuming that coolant conditions and fuel behaviour in all fuel channels will be similar, the results of analysis using the simplified model (temperatures of core elements, generation of hydrogen) can be extrapolated to all 1661 fuel channels in the reactor. Received parameters (amount of generated hydrogen and non-condensable gases) should be used as input for other special codes, designed for analysis of processes in containment (ALS). Using the developed models the analysis of large break LOCA with loss of pumped water injection to RCS is presented in the article. The performed analysis provided information regarding code acceptability for the severe accident analysis in RBMK reactor and assessment of the timing of key events. These results were used during the development of severe accident management guidelines for RBMK-1500 at Ignalina NPP. Acknowledgments

This analysis was performed in the frames of the PHARE project No. 2003/5812.04.02 “Support for VATESI and its TSOs in assessment of beyond design basis accidents for RBMK-1500 reactors” and project “Development of Manual on Management of Beyond Design Basis Accidents at Ignalina NPP” coordinated by the consortium Jacobsen Engineering Ltd from UK and Scientec from USA and sponsored by the UK Department of Trading and Industry. The authors of the article would like to express their gratitude to the sponsors, coordinators and Ignalina NPP for support of the performed analysis. References

(1)Almenas K., Kaliatka A., Uspuras E., Ignalina RBMK-1500. A Source Book. Extended and Updated Version (1998), p. 198, Lithuanian Energy Institute, Kaunas, Lithuania. (2)Uspuras E., Urbonavicius E. and Kaliatka A., Specific features of the RBMK-1500 reactor and BDBA management, Energetika, No. 3, (2005), pp. 1-9. (3)Safety Reports Series No. 43, Accident analysis for nuclear power plants with graphite moderated boiling water RBMK reactors, International Atomic Energy Agency, (2005), p. 59, IAEA, Vienna. (4)Medvedeva N., Timkin S., Andrejev A., Zhilko V., Peshkov I., Marciniuk D., Poshtovaja O., Analysis of piping behaviour in the graphite stack of RBMK-1000 in case of single technological channel rupture, Annual report, Elektrogorsk research centre on safety of nuclear power plants, (2004), pp. 11-21, Elektrogorsk, (in Russian). (5)Kaliatka A., Uspuras E., Benchmark analysis of main circulation pump trip events at the Ignalina NPP using RELAP5 code, Nuclear Engineering and Design. Vol. 202, (2000), pp. 109-118. (6)Urbonas R., Uspuras E., Kaliatka A., State-of-the-art computer code RELAP5 validation with RBMK-related separate phenomena data, Nuclear Engineering and Design. Vol. 225, (2003), pp. 65-81. (7)Allison C. M., Wagner R. J., RELAP5/SCDAPSIM/MOD3.2 (am+) Input Manual Supplemental, Innovative Systems Software, (2001) LLC, http://www.relap.com. (8)SCDAP/RELAP5 Development Team, SCDAP/RELAP5/MOD3.2 Code Manual, Vol. 1-5 (1998), NUREG/CR-6150, INEL-96/0422.

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