International Nuclear Safety and Cooperation International Nuclear Safety and Cooperation RBMK SP-2 Validation Results RBMK SP-2 Validation Results (KS PH Rupture Simulation) (KS PH Rupture Simulation) Bruce Schmitt Bruce Schmitt Battelle, PNNL Battelle, PNNL April, 2002 April, 2002 IG0101008.1
International Nuclear Safety and Cooperation International Nuclear Safety and Cooperation RINSC/INSC Code Validation Support � RINSC/INSC Code Validation Support � Joint Project 6 established to assist with code validation, primarily with the application of western codes like RELAP5 � Standard problems were defined to investigate important phenomena for VVER and RBMK designs � PNNL provided analysis support to ANL for RBMK problems � PNNL and KI jointly analyzed SP-2 � RBMK Standard Problem 2 (SP-2) � RBMK SP-2 was defined from a series of stop flow experiments that were performed with the KS facility � SP-2 was to simulate a pressure header rupture – although it was performed at system pressure IG0101008.2
International Nuclear Safety and Cooperation International Nuclear Safety and Cooperation RINSC/INSC Code Validation Support � RBMK SP-2 phenomena investigated � water release (ejection) from the fuel channel (FC) model and fuel simulator surface drying, � dryout under sharp flow deceleration at the inlet of the RBMK-1000 and RBMK-1500 fuel assembly (FA) models, � post dryout heat transfer and fuel simulator temperature conditions in the FA model under channel drying, � steam and water counter current flows in the steam-water piping, and in the FC with the FA model, � propagation of the reflood and quench front in the FA model under flow resumption at the channel inlet. IG0101008.3
International Nuclear Safety and Cooperation International Nuclear Safety and Cooperation RINSC/INSC Code Validation Support 60 7 6 08 U p pe r he a d er 3 c 1 c S e p er a tor s 2c 3 0 X X X V II 53 ~ X I I ~ 10 ~ 9 8 ~ X X X V II I К О H e a t e x c ha ng e rs V I II 5 3' ~ X X V I I V II 2 0 IX 1 8 ~ 1 9 ~ ~ XX X I X X X I X 1 3 ~ 12 ~ ~ 1 4 2 9 ~ 21 2 2 X II I ~ ~ 35 1к 2 к 6 3к ~ X I 40 7 т 8т V 4т 5 т 1 8т 6т V I п 27 2 X X I V X X I II II 2 7 38 X V X X X II In ta ke h e a de r п 22 7 3 9 п 2 24 п 22 6 X X X I X X V T e st ~ 9т 11 4 5 ~ 4 6 s ec tio n 1 Ф 3 3 3 7 47 32 Э П Д Д Д 3 Д X 4 Ф Д Д 5 6 Э П Д Д I T e st 24 ~ ~ 41 23 5 1 1 28 С оле ме р se ct ion 2 12 9 Ф Д C 12 1 7 A B X X X 1 0т T e st se c tion 3 X X V II I X I V _ V = 4 м 3 1 37 V = 2 м 3 п 1 17 п 2 09 п 10 8 P um ps X X V I п 2 80 1 38 2 э ~ 1 19 1 э 3э ~ Д 7 ~ 5 0 ~ 7Н X X X V IV I I I 5 Н ( Х Т Р 4/32 0) 1 б 1 р 2р 3 p ~ ~ 2б 3б (1 ,5 X -6 Л ) 2 Н 1Н X V I I 25 2 8 42 Д 1 12 0 X V I 3Н ( Т4 А ) 54 X I X Д 2 X V I II X X 16 L ow e r he a d er 4 6 4 5 X X II 1 39 44 1 22 26 1 7 X X 1 15 36 3 6A 61 IG0101008.4
International Nuclear Safety and Cooperation International Nuclear Safety and Cooperation RINSC/INSC Code Validation Support To separators Выход теплоносителя Upper Header R267 89х8 To hear echangers 756 Конец зоны 311 тепловыделения 475 3х120=360 Отбор 4 SWC model Обозначения: - решетка-интенсификатор с полным набором ячеек - решетка-интенсификатор с неполным набором ячеек Bypass - штатная решетка ТВС реактора РБМК-1000 19х360=6840 5250 Sampling 4 TW1 8468 TW2 7000 (обогреваемая длина) 7000 TW3 6894 DP16-4 6854 6836 Sampling 15 3814 TW4 PIN Sampling 16 1586 Отбор 15 1044 186 FC model TIN 1400 G LWC model Отбор 16 2r 475 Начало зоны 289 тепловыделения 490 From the circulation pumps Вход теплоносителя 222 Lower header 75х5 IG0101008.5
International Nuclear Safety and Cooperation International Nuclear Safety and Cooperation RINSC/INSC Code Validation Support 12 13 11 14 Припой ПСР-72 Термопара из провода КТМС ∅ 1.5 3 4 10 15 1 2 5 9 16 7 6 8 17 19 18 Имитатор твэла 32 61.8 80 109 121 IG0101008.6
International Nuclear Safety and Cooperation International Nuclear Safety and Cooperation RINSC/INSC Code Validation Support Figure 6) KS Model Node Diagram 600-02 600-01 600-07 N2 Receiver -01 -09 335-01 Seperator 605-00 852- Condenser Upper SWC 853-01 610 (01-06) 320-07 331-01 320-02 331-01 332-01 320-01 Upper Drum 315-00 330-01 310- 165-00 615-00 610-01 -01 -02 -02 -16 -17 -18 500-01 160-09 620-02 400-00 305-00 410-01 500-02 Accumulator 850-10 620-03 160-08 300-41 850-09 500-03 62-04 SWC (0.72 o slope) -01 250 850-08 850-07 300-40 620-05 300-02 300-01 850-06 -07 850-05 410-05 500-04 850-04 500-05 620-06 850-03 240 850-02 410-06 500-06 850-01 160-02 500-07 -01 Lift Path 410-07 820-06 Bypass 160-01 -03 After 505-00 Cooler 410-08 415-01 430-01 -02 234 510 (01-05) 620-07 820-05 230-01 155-00 1 0 420-01 775- 430-02 - Pressure Tube 220-01 -02 150-04 210-58 -01 -03 620-08 Pump Bypass 210-57 820-05 515-00 -01 -02 770-01 -01 -02 420-02 520- 780-00 770- 785- -03 01 -03 435-00 150-03 Bypass 620-09 Hx 440 (01-06) -01 -02 -03 -04 420-03 820-04 210-03 765- 210-02 -04 720- -01 150-02 210-01 725-01 -04 -03 -02 420-04 3 0 630-02 200-01 5 - 5 820-03 110-03 7 150-01 110-07 725-02 710-02 110-03 445-00 -01 -02 -03 -04 110-06 Circulation Pumps 630-01 700-03 110-02 100-01 745-01 735-01 450- -05 750- 740- Lower Header 110-05 820-02 -02 -01 -01 700-02 -06 100-01 110-01 700-01 105-00 Intake Header -02 -04 -03 730-02 710-01 700-01 810-00 820-01 -03 755-01 -06 -05 -04 710-01 -01 730-01 760-01 700-04 -04 -02 705- -03 760-01 IG0101008.7
International Nuclear Safety and Cooperation International Nuclear Safety and Cooperation RINSC/INSC Code Validation Support Table 2) KS Facility Initial Conditions No. Experiment Electrical power Pressure at inlet of Water temperature at Water flow of fuel assembly fuel assembly, P16, inlet of fuel channel, rate at inlet model, MWth MPa TF1, ºK GL, kg/s 1 Test-4 1.691 7.68 516.1 3.90 2 Test-5 2.486 8.40 527.4 4.70 3 Test-5? 2.532 7.95 533.1 4.17 4 Test-6 2.926 8.23 527.3 4.28 5 Test-7 3.488 8.23 529.3 4.13 6 Test-8 4.566 8.74 531.1 6.27 IG0101008.8
International Nuclear Safety and Cooperation International Nuclear Safety and Cooperation RINSC/INSC Code Validation Support Table 1) RELAP5 Model Equivalent Data Locations Data RELAP5 Model Thermocouples HS Volume TW-1 2102-59 mesh 8 210-58 TW-2 2102-58 mesh 8 210-57 TW-3 2101-33 mesh 8 210-32 TW-4 2102-10 mesh 8 210-09 Pressure Taps Volume P4 210-57 P16 210-02 IG0101008.9
International Nuclear Safety and Cooperation International Nuclear Safety and Cooperation RINSC/INSC Code Validation Support Table 3) Calculation Matrix Test ID Case Options e s m t h vp b 4 X O X O O O 5 X X O 5? X O X O O O 6 X X O 7 X X O 8 X O X O O X O Case Option Definitions (x –results presented here, o –results not presented) e - EPRI bundle friction correlation (this is the ‘basecase’ model setup) s - the SWC piping junction diameter is reduced by 1/8, based on SP-1 results [5] for liquid drainback m - improved CHF multiplier coefficient, based on SP-1 results [5] for dryout prediction v - valve leakage allowed p - early power shutdown t - time step size reduction h - heat structure radial mesh reduction b - Bestion bundle friction correlation IG0101008.10
International Nuclear Safety and Cooperation International Nuclear Safety and Cooperation Conclusions Conclusions � Overall Performance considered minimal or poor � The predicted steady state pressure drop in the heater bundle is not well correlated by RELAP5, and is considered to be minimally acceptable. � This suggests that the RBMK bundle requires a more specific correlation than the Lockart-Martinelli correlation used in RELAP5 or that the mass-flux dependent coefficients be defined specific to the RBMK bundle. � Time to dryout is reasonably predicted for each case. However, this would be expected for even significant errors in the predicted CHF for this evaluation. IG0101008.11
International Nuclear Safety and Cooperation International Nuclear Safety and Cooperation Conclusions Conclusions � RELAP5 consistently under-predicts rewet for the cases where power is maintained constant and the overall prediction is considered poor. � In general this is in the conservative direction. However, for the case of power reduction, Test 5 ′ , early rewet is predicted. � Sensitivity studies performed do indicate that an improved CHF correlation (specific to the RBMK fuel assemblies) would likely provide significant improvement. � It should also be noted that for RELAP5/MOD3.2, the reflood model is disabled because of incompatibilities. Updated versions of RELAP5 with a reflood model may provide additional improvement as well. IG0101008.12
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