CORE HEATUP PREDICTION DURING SB LOCA WITH RELAP5/MOD3.2.2 GAMMA

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1 International Conerence Nuclear Enery in Central Europe Hoteli Bernardin, Portorož, Slovenia, September -, www: tel.:+ 5 57, ax:+ 5 5 Nuclear Society o Slovenia, PORT, Jamova 9, SI- Ljubljana, Slovenia CORE HEATUP PREDICTION DURING SB LOCA WITH RELAP5/MOD.. GAMMA ABSTRACT Iztok Parzer, Borut Mavko Jože Stean Institute Reactor Enineerin Division Jamova 9, SI- Ljubljana, Slovenia Iztok.Parzer@ijs.si, Borut.Mavko@ijs.si Stojan Petelin University o Ljubljana Faculty o Maritime Studies and Transport Pot pomorščakov, SI- Portorož, Slovenia Stojan.Petelin@ijs.si The paper ocuses on the phenomena leadin to core uncoverin and heatup durin the SB LOCA and the ability o RELAP5/MOD.. to predict core overheatin. The code prediction has been compared to the three experiments, one conducted on the separate eect test acility NEPTUN in Switzerland and the other two conducted on two interal test acilities, PMK- in Hunary and PACTEL acility in Finland. In the case o a series o boilo experiments perormed on the NEPTUN test acility the inluence o the two correlations available in MOD.. or determinin interphase dra has been studied. In the case o IAEA-SPE- experiment simulation on PMK- acility the main oal o the analysis was to study the adequate modelin o the hexaonal core channel with 9-rod bundle and the phenomena durin the core uncoverin. The third analyzed experiment, OECD-ISP-, was perormed on PACTEL acility to study dierent natural circulation modes durin SB LOCA. The analysis also ocused on the inal stae o this SB LOCA experiment, when core dryout and heatup was observed due to radual emptyin o the primary system. Followin the experience the appropriate modelin options have been used to achieve better representation o the important phenomena durin the SB LOCA. INTRODUCTION Certain deiciencies in the RELAP5/MOD..x code [] have revealed in the past ew years, concernin the prediction o reactor core heatup in the late phases o LOCA. Accordin to previous experience in modelin separate and interal eect tests ([], [], [], [5], [] and [7]) it was assumed that it should be suicient to set up an adequate simulation model and make the appropriate selection o interphase dra and heat transer correlation to successully model the LOCA phenomena with RELAP5/MOD.. []. Special attention was paid to the core reion model in order to capture the phenomena durin core uncoverin and heatup correctly and to prove the above hypothesis..

2 . ERIMENT. Series o Boil-o Experiments on NEPTUN NEPTUN is a separate eect test acility in Switzerland. The three analyzed boil-o experiments were conducted at.5 MPa with initial subcoolin o K at power levels: exp.no. 57 with. kw (base case), exp.no. 5 with. kw (increased power) and exp.no. 5 with.5 kw (reduced power). EPRI and Bestion interacial dra correlation were tested and intercompared.. IAEA-SPE- on PMK- PMK- is a sinle-loop scaled-down -loop model (:7) o Hunarian Paks NPP (- loop VVER-). It operates at ull scaled power, ull primary and secondary pressure and ull scaled primary low. The IAEA-SPE- experiment was a 7. % LOCA (medium), considerin a beyond desin scenario with HPSI pumps unavailable and hydro-accumulators in action. LPSI pumps were activated durin core heatup. Secondary emerency bleed and eed was established to assure stable secondary heat sink.. ISP- on PACTEL PACTEL is a three-loop scaled-down model (:5) o Finnish Loviisa NPP (-loop VVER-). It operates at decay heat (.7 % scaled-down) power level, reduced primary pressure and at natural circulation low. Durin the analyzed ISP- experiment the primary system was radually emptied. Seven drainins, each o about 9 % o the initial primary inventory, were completed. Steady state was always achieved between successive drainins. Core heatup was observed ater 7 th drainin. Secondary pressure was kept constant and secondary inventory maintained by manual level reulation. THEORETICAL BACKGROUND It was assumed that the correct prediction o core dryout and heatup depends mostly on the choice o adequate correlation in bubbly and bubbly-slu vertical low reime. In the recent version o RELAP5/MOD.. two dierent interphase dra correlations are used. For bubbly and slu low reimes the drit lux approach is used and or all the other low reimes the dra coeicient method is selected. The drit lux model speciies the distribution coeicient and the vapor drit velocity. These two quantities must be converted into a constitutive relation or the interacial rictional orce per unit volume in order to compute the lobal interacial riction coeicient FI: F C v v, () i i R R where C i is an unknown coeicient and v R is the relative velocity between the phases. Within the context o the drit lux model, the relative velocity between the phases in not the dierence between the phasic velocities but is a weihted dierence between the phase velocities, iven by vr Cv C v, () where C is iven by the drit lux correlation and C is iven by Proceedins o the International Conerence Nuclear Enery in Central Europe, Portorož, Slovenia, Sept. -,

3 C i C.. () The drit lux model also speciies that the relative velocity v R can be written as the ratio o the vapor drit velocity and the liquid volume raction. v R v j () and the coeicient C i is then determined to be ( ρ ρ ) Ci. (5) v j The lobal interacial riction coeicient FI becomes FI F + F i i ρ ρ ρ m ( C v C v ). () The so-called EPRI correlation is used to determine the distribution coeicient C in RELAP5 code []: C L, (7) K + ( K )( ) r where L max min exp * (, j, ) * (, j, ) ( C ) exp ( C ) or C otherwise i < 7 () and urther the coeicients C, K, L and exponent r depend on pressure, phasic densities and Reynolds number. In the last ew versions o RELAP5 code a new Bestion correlation has been introduced to describe the behavior o the interphase dra coeicient: ( ) 5 ρ Ci, (9) D where D stands or the hydraulic diameter and the C distribution coeicient, was simply set to. []. Proceedins o the International Conerence Nuclear Enery in Central Europe, Portorož, Slovenia, Sept. -,

4 . RESULTS OF SIMULATION WITH RELAP5/MOD.. GAMMA Three input models have been developed or the three test acilities. The nodalization o the NEPTUN acility [] relects only the vertical test section channel with the inserted rod bundle and consists o volumes, connected with junctions. The nodalization o the sinle-loop interal test acility PMK- consists o volumes inter-connected with junctions. The nodalization o the three-loop interal test acility PACTEL is the larest amon the three and consists o volumes and junctions. Special attention has been paid to the core bundle model in each test acility. The interphase riction model has been chosen to be bundle interphase riction, as described in chapter. The special rod bundle heat transer option or parallel low has also been chosen in all simulations. Where applicable (IAEA-SPE-) the break low model has been tuned to the experimental data as close as possible to provide the correct primary inventory and coolant distribution around the primary loop. The correct primary coolant mass and its distribution were supposed to be crucial or the correct prediction o core dryout and heatup.. Boil-o Experiments on NEPTUN Amon the simulations o boil-o tests at the NEPTUN acility only the base case (exp.no. 57) results are shown here. It can be observed that when Bestion correlation was chosen or interphase riction prediction, the simulation results are closer to the experimental data. Liquid level in the test section, expressed in terms o pressure dierence across the test section (Fiure ) was predicted considerably better by usin Bestion correlation, as well as the timin o heatup onset and the maximum rod temperature reached at the end o the test (Fiure ) has been matched closely with this choice o interphase dra correlation. (kpa) reachin saturation pressure dierence in test section -Bestion Fiure : NEPTUN exp.no.57 (base case) pressure dierence in test section 9 claddin temperature at elevation.97 m ( o C) 7 5 heatup heater o reachin saturation -Bestion Fiure : NEPTUN exp.no.57 (base case) rod claddin temperature Proceedins o the International Conerence Nuclear Enery in Central Europe, Portorož, Slovenia, Sept. -,

5 .5. IAEA-SPE- on PMK- The results o the IAEA-SPE- simulation indicate that the primary pressure response (Fiure ) has been predicted by RELAP5 code very closely. No siniicant dierences have been observed in break low prediction (Fiure ) reardless o the interphase riction correlation choice. In spite o the correct primary inventory prediction the primary inventory was not distributed correctly in the later phase o the transient. Certain amount o liquid remained trapped in the hot le loop seal, aectin the liquid level in the core (Fiure 5). Aain no dierences were observed between the two simulations usin EPRI or Bestion interphase dra correlation. Nevertheless, the timin o core heatup and the maximum claddin temperature reached in the period when the core stayed uncovered (Fiure ), was predicted very well in both simulations. upper plenum pressure (.75 m ) (MPa) accumulator injection Fiure : IAEA-SPE- on PMK- upper plenum pressure interated break low (k) -Bestion Fiure : IAEA-SPE- on PMK- interated break low In spite o the correct primary inventory prediction the primary inventory was not distributed correctly in the later phase o the transient. Certain amount o liquid remained trapped in the hot le loop seal, aectin the liquid level in the core (Fiure 5). Aain no dierences were observed between the two simulations usin EPRI or Bestion interphase dra correlation. Nevertheless, the timin o core heatup and the maximum claddin temperature reached in the period when the core stayed uncovered (Fiure ), was predicted very well in both simulations. Still the calculation results show, reardless o the interacial riction correlation choice, that enouh liquid was kept at the top o the core in the later phase o the transient to prevent core overheatin. Only when the vapor quality exceeded.95 in the top volumes o the core, the heat transer coeicient dropped drastically and the overheatin was started. This could not be observed rom the experimental data. Proceedins o the International Conerence Nuclear Enery in Central Europe, Portorož, Slovenia, Sept. -,

6 (m) reactor vessel collapsed level (.9 / 9. m ) -Bestion core top core bottom level dierence about m Fiure 5: IAEA-SPE- on PMK- reactor vessel collapsed level claddin temperature at the top o the core (. m ) Bestion (K) Fiure : IAEA-SPE- on PMK- rod claddin temperature. ISP- on PACTEL In the ISP- experiment there were several interestin phenomena observed beore the inal core dryout and heatup. Ater the nd drainin irst voids appeared in the hot les, which caused the blockae o sinle-phase natural circulation low. Shortly ater that the periodic hot le loop seal clearin periods observed. Pressure spikes (Fiure 7) occurred due to the vapor eneration, beore the rowin vapor bubble could break throuh one o the hot le loop seals. This also caused periodic liquid push-up into the pressurizer vessel (Fiure ). All these phenomena were captured very well by RELAP5/MOD.. code pressurizer pressure (MPa) drainin. drainin periodic loop seal clearin. drainin 5 7 Fiure 7: ISP- on PACTEL pressurizer pressure. drainin 5. drainin. drainin 7. drainin Proceedins o the International Conerence Nuclear Enery in Central Europe, Portorož, Slovenia, Sept. -,

7 .7 pressurizer level (. /.9 m) (m) 5. drainin periodic loop seal clearin. drainin. drainin 5 7. drainin 5. drainin. drainin 7. drainin Fiure : ISP- on PACTEL pressurizer level It can also be observed that the reactor vessel and core liquid level was predicted very well in all staes o the transient (Fiure 9), which also resulted in correct prediction o the parameters in the core heatup period (Fiure ). The timin o the core heatup onset, the heatup rate and the maximum claddin temperature reached were very close to the experimental data. (m) drainin. drainin reactor vessel level. drainin 5 7. drainin 5. drainin. drainin 7. drainin Fiure 9: ISP- on PACTEL reactor vessel level (K) claddin temperature, core channel A, rod no.7 (.7 m) periodic loop seal clearin 5 7 heatup Fiure : ISP- on PACTEL rod claddin temperature 5 CONCLUSIONS For the NEPTUN boil-o experiments it was established that Bestion correlation perormed better at medium and hih power level in lower and middle parts o the core while the EPRI correlation enerally served better at low power level and or all power levels at the top o the core. Larer discrepancies in liquid level were observed or low power levels. Proceedins o the International Conerence Nuclear Enery in Central Europe, Portorož, Slovenia, Sept. -,

8 . In both simulation o interal eect tests (IAEA-SPE- and ISP-) MOD.. prediction matched most o the experimental data very well. No important dierences observed usin EPRI or Bestion correlation. In IAEA-SPE- experiment simulation the primary coolant remained trapped in the hot le loop seal ater inal loop seal clearin. Core heatup timin and maximum PCT predicted very well, while discrepancies were observed comparin experimental and calculated core liquid level at the heatup occurrence. The core was almost empty at that time in the simulation with MOD.., while experiment data show considerable amount o primary coolant in the core reion. In ISP- experiment simulation the loop seal clearin was predicted very well and no liquid remained trapped in any o the loop seals. Core heatup timin and maximum PCT predicted very well, at the correct core liquid level. Some phenomena occurred in wron loops, but since the experiment was symmetric, the MOD.. prediction could still be ound very satisactory. Simulation showed more liquid bein held in horizontal SG tubes than in the experiment. SG tubes were deliberately modeled slihtly inclined to capture the third loop seal eect and to describe the boiler-condenser heat transer mode in horizontal SG better. REFERENCES [] RELAP5/MOD Code Manual, Vol.,,,, 5, 7,, NUREG/CR-555, Scientech, Inc., Rockville, Maryland and Idaho Falls, Idaho, USA, June 999. [] I. Parzer, Reactor Core Heatup Model Durin the Loss-o-Coolant Accident, dissertation (in Slovene), University o Ljubljana, Ljubljana,, pp.5. [] B. Mavko, I. Parzer, S.Petelin, "A Modelin Study o the PMK-NVH Facility", Nucl. Technol., 5, 99, pp.-5. [] I. Parzer, S.Petelin, B. Mavko, "Vertical Stratiication Model in RELAP5 Computer Code", Z. anew. Math. Mech., 7, 99, pp.7-7. [5] B. Mavko, A. Prošek, Peak claddin temperature response surace eneration based on simulations o a small-break loss-o-coolant accident scenario, Proc. Int. Con. Nuclear Enery in Central Europe '97, Bled, Slovenia, September 7-, Nuclear Society o Slovenia, 997, pp. 5-. [] I. Parzer, S. Petelin, A % cold le SB LOCA test simulation on the PMK- acility, Proc. Int. Con. Nuclear Enery in Central Europe '9, Terme Čatež, Slovenia, September 7-, Nuclear Society o Slovenia, 99, pp. 7-. [7] B. Mavko A. Prošek, F. D'Auria, Determination o code accuracy in predictin smallbreak LOCA experiment, Nucl. Technol., 997, pp. -9. [] S. N. Aksan, F. Sterli, G. Th. Analytis, Boil-o Experiments with the EIR-NEPTUN Facility: Analysis and Code Assessment Overview Report, Paul Scherrer Institute (PSI), Villien, Switzerland, EIR-Bericht Nr. 9, NUREG/IA-, US NRC, Washinton, DC, USA, March 99. Proceedins o the International Conerence Nuclear Enery in Central Europe, Portorož, Slovenia, Sept. -,