ML20155K493

Analyses of Natural Circulation During a Surry Station Blackout Using SCDAP/RELAP5

ML20155K493
Person / Time
Site:	Surry
Issue date:	10/31/1988
From:	Bayless P EG&G IDAHO, INC.
To:	NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
References
CON-FIN-A-6360 EGG-2547, NUREG-CR-5214, NUDOCS 8811010254
Download: ML20155K493 (161)

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e NUREG/CR-5214 EGG-2547 n-.

Analyses of Natural Circulation

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During a Surry Station Blackout Using SCDAP/RELAP5 f

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Prepared by P.D Bayless EGGG Idaho, Inc.

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U.S. Nuclear Re0ulatory Conunission 8811010254 881031 ADOCK 050g0 DR

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i NOTICE

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NUREG/CR-5214 EGG-2547 R3 Analyses of Natural Circulation During a Surry Station Blackout Using SCDAP/RELAP5 s=She:ddar*-=

Bayless EG6G Idaho, Inc.

Idaho Falls, ID 83415 Prepared for Division of Systems Research Office of Nuclear Regulatory Research j

U.S. Nuclear Regulatory Commission Washington, DC 20555 NRC FIN A6360

l ABSTRACT l

The effects of reactor coolant system natural circulation on the response of the l

Sur.g nuclear power plant during a station blackout transient were investigated. A TMLB' sequence (loss of all ac power, immediate loss of auxiliary feedwater) was simulated from transient initiation until after fuel rod relocation had begun. Integral 3

analyses of the system thermal-hydraulics and the core damage behavior were per.

formed using t.5e SCDAP/RELAP3 computer code and several different models of the plant. Three scoping calculations were performed in which the complexity of the plant model was progressively increased to determine the overtil r"ects of in vessel and hot les naural circulation nows on the plant response. The natural circulation flows extended the transient, slowing the core heatup and deleying core damage by transferring energy from the core to structures in the upper plenum and coolant loops.

Increased temperatures in the ex-core structures indicated that they may fail, however.

I Nine sensitivity calculations were then performed to investigate the effects of model-I ing uncertainties on the multidimensional natural circulation nows and the system response. Creep rupture failure of the pressurizer surge line vta.s predicted to occur in eight of the calculations, with the hot leg failing in the ninth. The failure time was fairly itsensitive to the parameters varied. The failures occurred near the time that j

fuel rod relocation began, well before failure of the reactor vessel would be expected.

A calculation was also performed in which creep rupture failure of the surge line was modeled. The subsequent blowdown led to rapid accumulator injection and quench.

I ing of the entire core, l

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i EXECUTIVE

SUMMARY

i An increased awareness of severe accidents has one of the reference plants used in source term and been reflected in the U.S. Nuclear Regulatory risk analyses.

Commission's research after the accident at Three The analyses were performed using the SCDAP/

Mile Island. A parallel experimental and analytlerJ RELAPS computer code. This code performs inte-effort formed the basis of understanding of severe gral calculations of the system thermal hydraulics accidents. This research effort culminated in and the core behavior. The model for the hot leg l

NUREG-0956, the "Reassessment of the Technical countercurrent flow was developed using work per-i Bases for Estimating Source Terms," which formed at Argonne National Laboratory with the l

described the current ** v knowledge in the COMMIX computer code.

I severe accident e~a

..sified eight major In vessel natural circulation occurs between the areas of uncertaints 1,.- oraft "Reactor Risk Ref-uncovered part of the core and the upper plenum.

erence Document," NUREG ll50, used the source Vapor rises from the center of the core to the upper term methodology in NUREG-0956 to provide a plenum, w here it is cooled by the internal structures basis for new estimstes of reacior risk.

before returning to the core through the peripheral Natural circulation in the reactor coolant system fuel assemblics. Hot leg countercurrent flow trans-was one of the major severe ac:ident uncertainties fers energy to the hot leg piping and steam genera-identified in N11 REG-0956. Three types of natural tors. Hot vapor flows from the reactor vessel to the circulation that may occur during a high pressure steam generators along the top of the hot leg, while bo:loff transient in a pressurized water reactor were an opposing flow of cooler vapor proceeds from investigated for the Nuclear Regulatory Commis-the steam generators to the reactor vessel along the sion: in. vessel, hot les countercurrent Dow, and bottom of the hot leg loop natural circulation I

flow through the coolant loops. The objective of involves superheated vapor flowing through the i

the analyses was to investigate the effects of various coolant loops, and it only occurs if the loop seals types of natural circulation on the transient, severe are cleared of liquid. This did not occur in any of accident response of the plant. Sensitivity studies these calculations for the Surry plant.

were also performed to investigate the effects of The major influence of these flows is to transfer l

multidimensional natural circulation modeling energy from the core to other structures in the reac-l uncertainties on the plant transient response. Of tor coolant system, thereby reducing the heatup l

particular interest were changes in the events that rate of the core. Besides extending the transient, a j

occur, in event timing, and in the extent of core slower core heatup may alter the amount of clad.

i damage. A transient was also analyzed in w hich the ding that is oxidized while still in fuel rod geometry, surgeline failure and subsequent blowdown of the thereby changing the composition of the molten reactor coolant system were modeled.

core. The energy removed from the core will heat in meeting the objective of the analyses, insight other structures in the reactor coolant system, into the phenomena controlling the plant response Higher temperatures will tend to reduce fission and the natural circulation flows have been devel-product retention, and sufficiently high tempera-oped. The importance of mechanistic integral anal-tures may result in failure of the pressure boundary yses of severe accidents has been demonstrated.

prior to breach of the reactor vessel by the molten The results of the analyses have also helped to iden-core. Should this failure occur in the steam genera-tify particular areas of research that could provide for tubes, containment bypass through the steam l

a better understanding of the natural circulation lines may occur. If the failure occurs early enough, flows and their effects on the plant response, the system may depressurize sufficiently to avoid The translern selected for the analyses was the direct containment heating when the core debris is TMLB' sequence in the Surry nuclear power plant, ejected following vessel failure.

The TMLB' sequence involves the loss of all ac The natural circulation flows did not change the power and auxiliary feedwater, and was selected core dan age propession, only the times that because all three types of natural circulation may vmious stages of damage occurred up to fuel rod occur during the transient. Surry is a three loop relocation. Fuel rod relocation began at about 161 Westinghouse pressurized water reactor with U-min with no multidimensional natural circulation tube steam generators, it has a rated core thermal flows modeled. In vessel natural circulation l

power of 2441 MW. Surry was selected because it is delayed this event by */ min. The modeling of both v

in vessel and hot leg natural circulation resulted in These differences in timing and events demonstrate a further delay of 11 min. This difference in timing the importance of mechanistic modeling of severe was not as significant as the difference in ex-vessel accidents.

Structure temperatures.

Sensitivity calculations were performed to inves-With no natu il circulation Dows modeled, the tigate four areas of modeling uncertainty: axial vapor leaving the reactor vessel was near the satura-power profile, steam generator inlet plenum mix-tion temperature throughout the transient. With in-ing, heat loss from the piping, and radic! now vessel circulation modeled, the vapor leaving toe resistance in the upper plenum and core. In all of vessel was superheated. As a result, the hot leg and these calculations, both in-vessel and hot leg natu-surge line piping heated up, so much so that creep ra: circulation were modeled.

rupture failure of at least the surge line pipe would A base case was first performed that differed from be expected before the vessel i breached by the the scoping calculations primarily in that a lower decay molten core. Such a failure would alter the course power was used. Creep rupture failure of the surge line of the transient, compared to the calculation with-was predicted to occur at about 246 ndn after the out in vessel natural circulation. The extent of the beginning of the transient, shortly before the onset of change depends on the size of the failure and the fuci rod relocation, and about 50 min later than in the resultant depressurization. With hot leg counter.

comparable scoping calculation. Changing the axial current now also modeled, the steam generator power prof'de from a r-lati.ely Gat to a chopped cosine tubes also heated up, but they were several hundred shape had very little effect on the transient, with surge degrees below the surge line temperature. There-line failure occurring about I min earlier than in the fore, steam generator tube rupture and tne associ-base case, ated containuent bypass would not be expected to Reducing the amount of flow that mixed in the occur during the transient.

steam generator inlet plena increased the hot leg Ballooning occurred in the scoping calculations now and the heat transfer to the coolant loops. A with in vessel natural circulation, resulting in a 25% flow increase delayed the surge lirie failure by flow area reduction of 20-60% in the inner part of about 9 min. A bounding calculation that maxi.

the core. Although some dow was diverted around mized the hot leg now was also performed, in the ballooned region, the recirculating flow which there was no mixing in the steam generators.

between the core and the upper plenum was main-This led to a more uniform heatup of all the loop tained. Ballooning of the cladding affected the structures, and failure of the hot leg occurred amount of fuel dissolved by molten Zircatoy. Oxi-nearly 45 min later than the surge line failure in the dation of the inner cladding surface in the bal.

base case.

looned regions prevented the molten Zircatoy from Heat loss through the hot leg and surge line pipes coming into contact with the fuel pellets, prevent-was modeled with radiative and/or comective heat ing fuel dissolution in those regions.

transfer coefficients. While the heat loss increased The TMLB' transient was also analyzed as part the temperature difference across the pipes, the of the draft Reactor Risk Reference Document surge line was still pred!cted to fail. Delays in the (NUREG-il50) effort. A significant difference in failure of 6 to 13 min compared to the base case timing of events exists between that analysis and were realized, the scoping analyses performed here. In the draft Changing the radial Dow resistances in the core and NUREG-Il50 calculation, vessel failure was pre-upper plenum had little effect on the predicted surge dicted to occur 155 min into the transient. In the line failure. Decreasing the resistances in either location SCDAP/RELAP5 scoping calculation with no nat-led to failures less than 2 min earlier than in the base ural circulation, fuel rod relocation did not begin case increasing the resistances in both locations until 161 min. and vessel failure would not occur resulted in surge line failure about 12 min sooner than until some time well after 200 min. Even larger in the base case. The faster heatup of the core in this delays occurred when thein vessel and hot leg natu-last case was the result of ballooning in the core. The ral circulation now s were modeled. Fuel rod reloca-ballooning occurred when the cladding temperature tion did not begin until 248 min in the was about 1400 K, and the resultant surface area best-estimate sensitivity calculation. More signifi-increase and double-sided cladding oxidation acceler-cantly, the likelihood of cuessel structural failure ated the core heatup.

was illustrated in the calculations with multidimen.

The magnitude of the hot leg now was sensitise I

sional natural circulation, but was not explicitly only to the amount of mixing in the steam considered in the draft NUREG ll50 analyses.

generator inlet plena and the hot leg inlet vapor vi 1

l temperature. Changes in the reactor vessel upper Fragmentation of the fuel rods was predicted, plenum flow affected the hot leg only in that they although debris formation was not permitted in the altered the hot leg inlet vapor temperature. Con-calculation. At the end of the calculation, the sys-versely, the upper plenum flow was affected by the tem pressure was below 1.0 M Pa, the accumulators hot leg flow. The flow recirculating within the were empty, and the two-phase liquid level in the upper plenum increased as the loop hest transfer core was decreasing, although the core structures increased, because cooler vapor was being returned were still at the saturation temperature, a

to the plenum. This higher density fluid increased Further work is recommended in the natural cir-

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the buoyant driving force, and hence the flow.

culation area. The size of the reactor coolant sys-1 A 0.15-m diameter hole in the pressurizer surge tem failure needs to be determined, as does the line was modeled at the time of the surge line failure ensuing system behavior until the time of reactor in the final calculation. The system pressure rapidly vessel failure. The effects ol'the natural circulation l

decreased, allowing the secumulators to inject liq.

flows on the fission product transport and reten-uid into the reactor coolant system. This water tion need to be quantified. Interactions between the entered the reactor vessel, where it quenched the natural circulation flows and noncondensible gases i

entire core before any fuel rod relocation occurred.

should be investigated further.

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l CONTENTS iii ABSTRACT.........................................................................

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4 EX E CUT I V E SU M M A RY............................................................

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AC K N OW LE DO M E NTS.............................................................

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I NT R O D U CT I O N..............................................................

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N ATU R AL CI RCU LAT ION FLOWS..............................................

4 2.1 I n Vessel Nat u ral Circulation.................................................

5 2.2 H ot Leg Cou nt ercu r rent Flow................................................

7 2.3 LoopFlow.................................................................

A 8

3.

S CO P I N G AN A LY S ES..........................................................

8 3.1 O n ce-Th ro u gh M od el.......................................................

13 4

3.2 I n-Ves sel Circulatio n........................................................

20 3.3 H ot Leg Cou ntercu rrent Flow................................................

1 24 3.4 LoopFlow.................................................................

24 3.5 Re s ult Co m paris o ns.........................................................

24 3.5.1 Effects of Natural Circulation Flows...................................

28 3.5.2 Comparison with Draft NUREO.ll 50.................................

30 3.6 Uncertainties and Limitations................................................

32 4.

S E N S IT I V IT Y A N A LY S E S.......................................................

t 32 i

l 4.1 BaseCase.................................................................

44 4.2 Axlal Power Profile Sensitivity...............................................

48 4.3 I nlet Plenum Mixing Sensitivity...............................................

4.3.1 Reduced inlet Plenum Mixing.........................................

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3 4.3.2 No i nlet Plen um M ixing..............................................

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62 4.3.3 Summary..........................................................

I 65 4.4 Pi pi n g H eat loss Se n sitivit y..................................................

i 4.4.1 Convective Boundary Condition.......................................

65 4.4.2 Convection and Radiative Boundary Condition..........................

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4.4.3 Summary..........................................................

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4.5 Crossuow Resistance Sensitivity...............................................

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1 4.5.1 Decreased Upper Plenum Crossuow Resistance..........................

74 4.5.2 Decreased Core Crossaow Resistance...................................

74 4.5.3 Increased Core and Upper Plenum Crossflow Resistance..........

84 4.5.4 Summary..........................................................

93 4.6 Su rge Line Failure Calculation...............................................

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4.7 Summary of Sensitivity and Surge Line Failure Analyses..........................

98 4.8 Uncertaint ies and Limitations................................................

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CONCLUSIONS AND RECOhlh1ENDATIONS...................................

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R E F E R E N C ES.................................................................

109 APPENDIX A -PLANT AND hlODEL DESCRIPTIONS...............................

A.!

APPENDIX B-COh!PUTER CODE DESCRIPTION.................................... B-1 APPENDIX C-USE OF RELAPS THERh1AL. HYDRAULIC hf0DELS IN THE SEVERE CORE DAh! AGE ACCIDENT ANALYSIS PACKAGE...................

C1 FIGURES i

1.

Severe accident natural circulation nows............................................

2 1

2.

Hot leg natural circulation stream dows.............................................

6 4

3.

Pressurizer pressure for scoping Case 1.............................................

9 l

4.

Pressurizer collapsed liquid level for scoping Case 1 10

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Fuel rod (ladding surface temperatures at 0.18,1.28,2.01, and 3.47 m above the core j

bot to m fo r sc opi n g Ca se 1........................................................

11 4

6.

Total hydrogen generation rate for scoping Case !...................................

12 7.

hiass now rate through the core bypass for scoping Case 1.............................

12 8.

Reactor vessel collapsed liquid level for scoping Case 1................................

13 4

9.

Volume average temperature of the hottest upper plenum structure for scoping Case 1......

14

10. Volume-average temperatures of the hottest part of the hot leg pipe, surge line, and t

4 steam generator t ubes for scoping Case !............................................

14

11. Vapor velocity vectors in the core and upper plenum at 167 min in scoping Case 2.........

17 t

12. Center channel fuel rod cladding surface temperatures at 0.18,1.28,2.01, and 3.47 m i

above the bottom of the core for scoping Case 2......................................

18 2

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13. Fuel rod cladding surface temperatures at the tor (3.47 m above the core bottom) of 18 the three core channels for scoping Case 2...........................................

14. Outer channel fuel rod cladding surface temperatures at 0.18,1.2P.,2.01, and 3.47 m 19 above the bottom of the core for scoping Case 2......................................

15. Volume average temperatures of the hottest part of the hot leg pipe, surge line, and 21 steam generator tubes for scoping Case 2...........................................

21

16. Pressurizer collapsed liquid level for scoping Case 2...................................

17. Mass flow rate in the top of a non pressurizer loop hot leg for scoping Case 3.......

22 Ig. Center channel fuel rod cladding surface temperatures at 0.18,1.28,2.01, and 3.47 m 23 above the bottom of the core for scoping Case 3......................................

19. Fuel rod cladding surface temperatures from the three core channels 23 fo r sco p i n g Ca se 3..............................................................

20. Volume average temperatures of the hottest part of the hot leg pipe, surge line, and 25 steam generator t ubes for scoping Case 3............................................

28

21. Reactor vessel collapsed liquid level for the three scoping cases.........................

22. Pressurizer collapsed liquid level for the three scoping cases...........................

29

23. Reactor vessel collapsed liquid level for sensitivity Case 1.............................

35

24. Peak cladding temperature for sensitivity Case 1.....................................

35

25. Fuel rod cladding surface temperatures at the top of the three core channels for 37 se n sit ivit y Ca se 1...............................................

26. Center channel fuel rod cladding surface temperatures at 0.18,1.28,2. d. t.nd 3.47 m above the core bottom for sensitivity Case 1........................................

37

27. Total hydrogen generation rate for sensitivity Case 1..................................

38

28. Mass flow rates exiting the three core channels, the core bypass, and recirculating in 38 t he u ppe r plenum for sensitivit y Case 1.............................................

29. Upper plenum recirculating mass flow rate as a function of maximum upper plenum va por te m perahre for semitivity Case i.............................................

39

30. Volume aserage temperatures of the upper plenum structures at the outlet of the three core chan nels for sensitivit y Case 1...............................................

40

31. Fraction of the core heat remosed by the coolant for sensitivit, Case 1..................

40

32. Volume-average temperatures of the three hot leg pipes near the reactor vessel for 42 sensitisityCase1...............................................................

33. Voli'me average temperatures of the hottest loop C hot leg, surge line, and t'eam generat or t ubes for sensitis it y Case 1...............................................

42 xi

34. _ llot leg nonle hot and cold vapor temperatures in Loops A ano O for sensitivity Case 1.........................................................................

43

35. Upper hot leg mass now in Loops A and C for sensitivity Case I 43 1

36.

Ho: leg How as a function on the hot leg inlet vapor temperaturein Loo sen sitivit y Case i.............................................p A for i

45

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37. Pressurizer liquid volume for sensitivity Case I.......................................

45 I

38. Peak cladding temperatures for sensitivity Cases 2 and 1...............................

47

39. liighest volume-average pipe temperatures in the Loop C hot :eg, surge hne, and steam generator tubes for sensitivity Case 2, and the surge line for Case I.....................

47

40. Upper hof leg mass flow in Loop A for sensitivity Cases 3 and 1........................

51 l

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41. Ilot leg Dow as a function of hot leg inlet vapor temperature for sensitivity Case 3.........

$1

42. Fraction of the core heat removed by the coolant for sensitivity Cases 3 and I.............

52 1

43. Volume-average hot leg pipe temperatures near the reactor vessel for sensitivity Case 3.....

52 44, liighest volume-awrage pipe temperatures in the Loop C hot leg, surge line, and steam f

generator tubes for sensitivity Case 3................................................

53 i

45. Fuel rod claddii.g surface temperatures at the top of the three core channels for r

f sen siti vi t y Ca se 3................................................................

54 t

46. Peak cladding temperatures for sensitivity Cases 3 and 1...............................

54 l

47. Mass now rates exiting the three core channels and the core bypass for sensitivity Case?.........................................................................

55

43. Volume-average temperatures of the upper plenum structures at the outlet of the three j

core channels for sen sitivity Case 3.................................................

56

49. Decay and oxidation power for sensitivity Case 3.....................................

56 j

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50. Upper hot leg mass now in Loop A for sensitivity Cases 4 and 1........................

59 1

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$1. Ilot leg now as a function of hot leg inlet vapor temperature for sensitivity Case 4.........

59 F

1 1

52. Fraction of the core heat removed by the coolant for sensitivity Cases 4 and 1.............

60 l

53. Volume average hot leg pipe temperatures near the reactor vessel for sensitivity Case 4....

60

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J 54, liighest vo'ume average steam generator tube temperatures in the three coolant loops fo r se nsit ivit y Ca se 4............................................................

61 3

55. liighest volume average pipe temperatures in the loop C hot leg, surge line, and steam

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generator t ubes for sensitivity Case 4...............................................

61 L

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56. Fuel rod cladding surface temperatures at the top of the three core channels for 63 se n s it ivit y Case 4................................................................

57. Peak cladding temperatures for sensitivity Cases 4 and I...............................

63

$8. Mass now rates exiting the three core channels and the core bypass for sensitivity 64 Case 4.........................................................................

59. Volume. average temperatures of the upper plenum structures at the outlet of the three core channels for sensitivity Case 4.................................................

64

60. Decay and oxidation power for sensitivity Case 4.....................................

65

61. Fraction of the core heat removed by the coolant for sensitivity Cases 5 and 1.............

67

62. Fraction of the heat removed from the core that was lost through the piping to the containment for sensitivity Case 5.................................................

67

63. Fraction of the heat removed from the core that was transferred to the steam generator tubes and tube sheets for sensitivity Cases 5 and 1....................................

69

64. liighest volume average pipe tempentures in the loop C hot leg, surge line, and steam generator t ubes for sensitivity Case 5...............................................

69

65. liot leg flow as a function of hot leg inlet vapor temperature for sensitivity Case 5.........

70

66. Peak cladding temperatures for sensitivity Cases 5 and 1...............................

70 67, llot leg flow as a function of hot leg inlet vapor temperature for sensitivity Case 6.........

72 68, liighest volume. average pipe temperatures in the loop C hot leg, surge line, and steam generator t ubes for sensitivity Case 6..............................................

72 69.

Peak cladding temperatures for sensitivity Cases 6 and 1...............................

73

70. Upper plenum recirculating mass flow for sensitivity Cases 7 and 1......................

75

71. Upper plenum recirculating mass flow rate as a function of maximum upper plenum vapor temperature for sensitivity Case 7.............................................

75

72. Volume. average temperature of the upper plenum structure at the outlet of the outer core channel for sensitivity Cases 7 and 1...........................................

76

73. Peak cladding temperatures for sensitivity Cases 7 and 1...............................

76

74. Fraction of the core heat removed by the coolant for sensitivity Cases 7 and 1.............

77

75. Upper hot leg mass flow in loop A for sensitivity Cases 7 and 1........................

77 76, liighest volume. average pipe temperatures in the Loop C hot leg, surge line, and steam generator t ubes for sensitivit y Case 7...............................................

78 l

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77. Fuel rod cladding surface temperatures at the top of the three core channels for se n s i t i s i t y Ca se 8..............................................................

81 i

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78. Peak cladding temperatures for sensitivity Cases 8 and I...............................

82 t

79. Fraction of the core heat removed by the coolant for sensitivity Cases 8 and 1.............

82

80. Volume average temperatures of the upper plenum structures at the outlet of the center and outer core channels for sensitivity Cases 8 and 1..................................

83 t

81. Mass flow rates exiting the three core channels, the core bypass, and recirculating in the upper plenum for sensitivity Case 8.............................................

83

82. Highest volume-average pipe temperatures in the Loop C hot leg, surge line, and steam generator tubes for sensitivity Case 8...............................................

84

83. Mass flow rates exiting the three core channels, the core bypass, and recirculating in the upper plenum for sensitivity Case 9.............................................

86 l

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-84.

Upper plenum rreirculating mass, sow rate as a function of maximum upper plenum i

vapor temperature for sensitivity Case 9.............................................

86

85. Fuel rod cladding surface temperatures at 0.55 m above the core bottom in the three i

core cha nnels for sensitivity Case 9.................................................

87 t

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86. Center channel fuel rod cladding surface temperatures at 0.18,1.28,2.38, and 3.47 m above the core bottom for sensitivity Case 9........................................

87

87. Peak cladding temperatures for sesitivity Cases 9 and 1...............................

88

88. Fuel rod cladding surface temperatures at the top of the three core channels for sen si t ivi t y Ca se 9...............................................................

88

89. Total hydrogen generation rate for sensitivity Case 9..................................

89

90. Volume-average temperatures of the upper plenum structures at the outlet of the three core channels for sensitivity Case 9.................................................

90 1

91. Reactor vessel collapsed liquid level for sensitivity Cases 9 and I...,....................

90 l

l 92.

Hot leg noule hot and cold vapor temperatures in Loops A and C for sensitivit Case 9.............................................................y 91

93. Highest volume average pipe temperatures in the Loop C hot leg, surge line, and steam I

generator tubes for sensitivity Case 9 and the surge lin' for Case 1.....................

92 e

{

94. Fraction of the core heat removed by the coolant for sensitivity Cases 9 and 1.............

92

95. Pressurizer pressure for the surge line failure calculation...............................

95 l

96 Liquid volume in one accumulator for the surge line failure calculation..................

t

97. Peak cladding temperature for the surge line failure calculation.........................

95 i

98. Center channel fuel rod cladding surface temperatures at 0.18,1.28,2.38, and 3.47 m f

?

above the core bottom for t he surge line failure calculation.............................

97 i

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99. Collapsed liquid level in the center core channel for the surge line failure calculation.......97-100. Reactor vessel collapsed liquid level for the surge line failure calculation.................

99 101. Volume-average structure temperatures in the three volumes above above the center core channel for the surge line failure calculation.....................................

99 l

102. Void fraction in the bottom of the loop A loop seal for the surge line failure 100

' c al c ul a t i o n.....................................................................

103. Highest volume average steam generator tube temperatures in the three coolant loops 100 i

for the surge line failu re calculation................................................

i 104. Volume-average hot leg pipe temperatures near the reactor vessel for the surge line i

fail u re calcul a t i o n...............................................................

101 l

i 105. Mass flow out the break for the surge line failure calculation...........................

101 A l. Nodalization of the pressurizer coolant loop for the Surry SCDAP/RELAP5 J

calc ul a t i o n s.....................................................................

A-4 A-2. Nodalization of the single-channel reactor vessel for the once-through Surry SC DA P/ R E LA P5 calculatio n.....................................................

A5 A-3. Nodalization of the reactor vessel for the Surry SCDAP/RELAP5 calculations with j

in vessel n at u ral circulation.......................................................

A-6 i

i A-4. Cross section of the three-channel core region........................................

A-7 A-5. 'lipical arrangement of a Surry fuel assembly........................................

A-8 t.

A-6. Nodalization of the hot leg and steam generator for the Surry SCDAP/RELAP5 calculations with hot leg natural circulation..........................................

A 10 4

1 C-1. RELAP5 multi region core model with blockage, velocity vectors at volume centers.......

C-7 I

[

C 2. RELAP5 multi-region core model with heated center channel, velocity vectors at i

volumecenters..................................................................

C-9 I

C 3. Velocity vector plot for flow with blockage, from page 117 of Reference C.ll.............

C.11 i

C-4 Calculated viscous flow over a backstep, from page 337 of Reference C.13...............

C 12 TABLES i

I j

1.

Sequence o f event s for scoping Case 1..............................................

9 2.

Sequence o f event s for scoping Case 2..............................................

16 1

3.

Sequence of events for the scoping calculations.......................................

26 l

4.

Conditions w hen fuel rod relocation began in the three scoping calculations..............

27 i

XV 1

l l

l 5.

Sequence of events for the Surry TMLB' transient from five different calculations.........

29 6.

Matrix o f sensitivity calculations...................................................

33 7.

Sequence of events up to core heatup for the sensitivity and scoping calculations..........

34 4

I 8,

Sequence of events for t he base case................................................

34 I

9.

Sequence of events for the base case and axial power profile sensitivity calculations........

46 l

10. Conditions near the time of the surge line failure for the base case and axial power profile sensitivity calcula tions.....................................................

49 i

f

11. Sequence of events for the base case and steam generator inlet plenum mixing sensitivit y calculations..........................................................

50 i

t

12. Conditions near the time of the reactor coolant system failure for the base case and steam generator inlet plenum mixing sensitivity calculations............................

58 t

13. Sequence of events for the base case and piping heat loss sensitivity calculations...........

66 4'

14.. Conditions near the time of the surge line failure for the base case and piping heat loss

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sensitivit y calculat ions...........................................................

71 4

15. Sequence of events for the base case and crossflow resistance sensitivity calculations.......

79

+

16. Conditions near the time of the surge line failure for the base case and crossflow t

l resistance sensitivity calculations...................................................

80

{

1

17. Sequence of events for the surge line failure calculation................................

94 l

t L

j A.I. Comparison of computed and desired steady state parameters..........................

A.12 4

A.2. Initial inventory for five fission product elements.....................................

A.12 A.3. Decay power for t he scoping calculations...........................................

A.13 i

A-4. Decay power for the sensitivity calculations.......................................

A.15 t

+

A.5. Computer calculation statistics....................................................

A-17 C.l. Axial moment um equation component s.............................................

C.8 t

t C.2. Radial momentum equation components............................................

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I ACKNOWLEDGMENTS f

I would like to thank Dr. James Han of the U.S. Nuclear Regulatory Commission, i

the sponsor of this work, for his guidance and reviews of the models and analyses.

The staff at Argonne National Laboratory, particularly Drs. Hank Domanus and j

t j

William Sha, were most helpful in providing and discussing the results of the COMMIX calculations.

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ANALYSES OF NATURAL CIRCULATION DURING A SURRY STATION BLACKOUT USING SCDAP/RELAPS

1. INTRODUCTION An increased awareness of severe accidents was surized water reactor with three coolant loops and a reflected in the U.S. Nuclear Regulatory Commis-rated core thermal power of 2441 MW. Surry is one sion's research after the accident at Three Mile of the reference plaats used in source term and risk Island. A parallel experimental and analytical analyses. Therefore, calculations of the TMLB' effort formed the basis of understanding of severe sequence are available in w'ilch other computer accidents. This research effort culminated in the codes and modeling assumptions were used. Com-

"Reassessment of the Technical Bases for Estimat-paring the results of those calcu'ations with the cur-ing Source Terms," NUREG 0956,1 which rent analyses may provide insight into how the described the current state of knowledge in the codes and models used affect the calculated plant severe accident area and identified eight major

behavior, areas of uncertainty. The draft "Reactor Risk Ref-The SCDAP/RELAP5 computer code 4 was crence Document," NUREG il50,2 used the used to perform the calculations of the plant source term methodology in NUREG-0956 to pro-response. This code provides best-estimate integral vide a basis for new estimates of reactor risk.

calculations of the system thermal hydraulic and Natural circulation in the reactor coolant system, core damage response. Therefore, effects of the one of the areas of major uncertainty identified in various natural circulation flows on the core dam-NUREG-0956, was discussed in the asscciated age progression could be explicitly determined, uncertainty papers liowever, phenomenological There are three natural circulation flows in the 3

analyses in support of draft NUREG-Il50 did not RCS that are of interest in severe accidents. They account for the natural circulation flows. Analyses are inaessel circulation, hot leg countercurrent have been performed for the U.S. Nuclear Regula-Dow, and flow through the coolant loops. Figure I tory Commission to mechanistically investigate the illustrates these natural circulation Dows. In vessel effects of various types of naturalcirculation on the natural circulation is characterized by a hot plume transier.t. severe accident response of the Surry of vapor rising from the center of the core into the nuclear power plant. Of particular interest were upper plenum. Ileat is transferred to the upper i

changes in the events that occur, in event timings, plenum structures, thereby cooling the vapor. The and in the extent of core damage. Sensitivity stud-now turns radially outward in the upper plenum, les have also been performed to investigate areas of then returns to the core through the lower powered, major uncertainty in the modeling of multidimen-and hence cooler, peripheral fuel assemblies. The sional natural circulation Dows. Blowdown of the flow continues downward into the core, slowly plant following a postulated surge line failure has turning iow ard the center, where it again rises to the also been simulated.

upper plenum, liot leg natural circulation is char-The tramient used to imestigate the natural circula-acterized by single-phase, countercurrent flow in tion flows wm the TMLD' sequence, which is a loss of the hot leg between the upper plenum and U tebe both omite and offsite ac power, '.<ith early (immedi-steam generators, lit t sapor Dows out the top of ate) failure of the steam-dtben auxiliary feedwater the hot leg to the stem generator inlet plenum, pump. A high-pressure boiloff ensues. This tramient home of the now mhes we the fluid in the intet wm selected twause it presents the potential for natu-plenum, while some continues uO the steam gen-ral circulation flow in metal regions of the reactor crator tubes. The hot sapor cools as it.1ow s to the coolant system (RCS). It is also a useful trarnient in outlet plenum through about 35% of the steam that it has been analyzed by modeling different plants generator tubes.5 The cooler vapor returns through with various computer mdes, the tubes to the inlet plenum. Again, some of this The Surry nuclear power plant was used in these flow mhes in the inlet plenum, and some continues analyses. Surry is a Westinghouse-designed pres-directly into the bottom of the hot leg. The cooler i

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Steam Pressurizer Stea m generator A

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vapor then returi.s to the reactor vessel along the allow in-vessel natural circulation. Finally, the hot bottom of the hot leg. Superheated vapor flow leg model was changed to allow hot leg countercur-through the coolant loops will occur only if the rent flow. By comparing the results of each calcula-loop seals ase cleared of liquid. If this now occurs, tion with those of the previous one, the changes the in. vessel and hot leg natural circulation Dows effected by each natunt circulation flow could be will no longer exist, because the loop flow rate is determined. The coolant loops were modeled so high enough to preclude then4.

that loop natural circulation could occur if the loop The most important effect of the natural circulation seals cleared of liquid in any of the calculations.

Hows is to transfer energy from the core to other parts However, the loop seals did not clear in any of the of the RCS. 'Ihis energy tramfer reduces the heatup calculations.

rate of the core, and delays the various stages of core A series of sensitivity studies was then performed damage. The heatup of the other structures will also to investigate the effects of major phenomenologi-affect the fission product deposition and retention, cal and modeling uncertainties on the RCS behav-because higher structure temperatures will proside an ior until the time of RCS pressure boundary failure.

Impetus for drising the fission products further a ong Of particular interest were changes in the timing or i

the flow path. Higher temperatures in the ex vessel location of the RCS failure. Parameters that were stru<;tures also present the possibility of failure of the varied in the calculations included the axial power RCS before failure of the reactor wssel lower head.

prafile, the amount of mixing in the steam genera.

The possibility of such a failure is very important in the tor inlet plena, radial flow resistances in the core consideration of direet containment heating, because a and upper plenum, and heat loss through the hot large enough failure may allow the RCS to depressurize leg and pressurizer suige line piping. Finally, the sufficiently by the time the vessel fails to preclude early failure of the surge line and the ensuing blowdown containment failure by direct containment heating.

were analyzed.

Failure of the s' ram generator tubes may also provide a Chapter 2 discusses the three natural circulation path for containment bypass, in that Ossion products nows in g: cater detail. The results of the scoping and could be tmmported through the ruptured steam gen-sensitivity analyses are presented in Chapters 3 and 4, crator tubes into the secondary side of the steam gener-respectively, followed by conclusions drawn from the ators and from there, through the relief valves to the analyses in Chapter 5 and references in Chapter 6.

atmosphere.

Appendix A briefly describes the Surry plant and the A scoping study of the effects of the naturalcir-SCDAP/RELAP5 models of the plant. Appendix B culation flows was performed by systematically prosides a description of the SCDAP/RELAP5 com-increasing the complexity of the Surry plant model.

puter code. Appendix C contains a paper addressing First, the transient was calculated using a once-the applicability of the crossflow junction model in t

through model of the reactor vessel. Second, the RELAP5 to the Dow situations that might be encoun-core and upper plenum model were changed to tered during a severe accident.

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2. NATURAL CIRCULATION FLOWS Three natural circulation flows that can be exist between fuel assemblies, between parts of fuel important during severe weidents are in-vessel cir-assemblies, and even between subchannels. Wher-culation, hot leg counterrurrent now, and flow ever a radial temperature gradient exists, a natural through the coolant loops. Each of these Dows may convection flow may be established. Rapid be present during a high pressure boiloff transient increases in local temperature associated with the such as the TMLH' sequence accelerated oxidation of the Zircaloy fuel rod clad-The primary effect of natural circulation flows is ding, at a temperature around 1850 K, will result in to redistribute the energy being generated in the the establishment of these relatively small natural core. This energy redistribution will slow the circulation cells. These smaller flow cells are heatup of the core, which in turn may affect the believed (assumed) to exist within larger overall damage progression or the extent of the core dam-natural convection cells involving the entire core.

age. Slowing the core damage would allow more These analyses investigate and discuss such core-time for systems to be recosered to mitigate or ter-wide patterns, minate the accident, flowev r, energy removed in vessel natural circulation now may affect the from the core will affect the structures to w hich it is cladding oxidation. Because steam is being recircu-transferred, in both the upper plenum and the cool-lated from the upper plenum back into the core, it ant loops. The discussions below address the basic is less likely that the oxidation reaction will become phenomena associated whh the three types of natu-steam starved. The slower cladding heatup caused ral circulation being considered, torther with how by the removal of some of the core energy to the the transient progression may be affected by the upper plenum structures, combined with a steam-coolant flow. Appendix A prosides information rich environment. may result in more extensise oxi-on how the muhidimusional Dows were modeled dation of the cladding at lower temperatures. More in these analyses.

extensive oxidation of the cladding, in turn, may 2.1 In Vessel Natural Circulation result in smaller amounts of unoxidized Zircatoy melting as the temper'ature increases, which could delay relocation of molten material and reduce dis-In vessel natural circulation begins w hen the core solution of the fuel pellets.

heatup begins. Because the center part of the core is The heating of the upper plenum structures at a higher power than the periphery, the super-could also result in their oxidation or melting, heated steam in the center is hotter and less dense, which could add to the hydrogen generated in the and a radial density gradient is estaolished. The RCS hfore vessel failure and add more material to denser vapor in the outer part of the core tends to the melt that nows from the vessel at the time of flow tow ard the center, replacing the hot vapor that lower head failure.

rises into the upper plenum. This vapor plume rises Both the smaller amount of fuel dissolved by liq.

to the top of the upper plenum, where it is turned uld Zircaloy and the higher temperature of the radially outward to the core barrel, and then back upper plenum structures affect fission product down toward the top of the core. }{ eat transfer to behavior. The smaller amount of dissolved fuel the structures in the upper plenum cools the vapor, may delay the release of many of the fission prod-reinforcing the density gradient between the center ucts until the fuel melts. At that later time, which of the vessel and the periphery. The cooler steam will be closer to the time of vessel failure, there will reenters the core through the top of the peripheral be less time for the fission products to be retained fuel assemblies. As core uncovering continues and on surfaces in the RCS, which may result in more the liquid level drops, the recirculating now estends fission products in the containment. The higher farther into the core. Depending on the axial power structure temperatures would tend to reduce the profile, the now may esentually extend to the bot-amount of fission products retained in the upper tom of the core. The density gradient in the upper plenum, causing them to be deposited elsew here or plenum also establishes a recirculatin2 now within released to the containment, the upper plenum, liigher vapor temperatures in the upper plenum in reality, many natural circulation cells will be will also make hotter vapor available to the hot established in the core, especiall, during the core legs. Flow through the hot leg and surge line to the damage portion of the transient. These cells will pressuriier power operated relief valses (PORVs) 4

will heat the piping. If the pipe temperatures are between the two fluid streams. The result is that Tn,i > T.2. As the Dow enters the steam generator high enough, creep rupture failure of these pipes h

may become a concern. Failure of the RCS piping inlet plenum, some of it mixes with the fluid in the before vessel failure could allow the system to plenum and with the cold now exiting from some of depressurize, initiating accumulator injection. If the steam generator tubes. The mixing reduces the the depressurizatian continues far and fast enough, temperature of the steam entering the steam gener-ator tubes, and Tn,3 < T.2. Heat is transferred to the RCS pressure at the time of sessel failure may be h

low enough to preclude direct containment the tubes as the steam nows through the steam gen-

heating, erators. When the flow returns to the inlet plenum, some of it mixes with the hot leg now. This mixing raises the temperature of the steam returning 2.2 Hot Leg Countercurrent Flow through the hot leg, so that T.2 > T,y As the flow c

proceeds along the bottom of the hat leg to th-Single-phase countercurrent Gow in the hot leg is reactor vessel, heat is being transferred from the the least well characterized of the three natural cir-hotter Guid above to this cooler steam, and from ct.lation Dows being imestigated. A general discus-this cooler steam to the hot leg pipe. Whether tnese sion of the basic considerations of the now is energy transfers tesult in a net heating or cooling of presented below.

the return now has et been quantified, but the Superheated vapor enters the top of the hot leg, vapor temperature will probably net change signifi-displacing saturated vapor, which then nows back cantly along the bottom of the hot leg. Also unac-to the reactor vessel along the bottom of the hot counted for in this discussion is the effect of leg. When the hotter vapor enters the steam genera-circumferential heat transfer in the hot leg piping, tor inlet plenum,it will rise toward the steam gener-in which heat would be conducted from the hot ator tubes. Vapor enters some of the tubes, upper part of the hot leg to the coollower part.

displacing the cooler steam that was in the tubes.

Assuming a steady Dow, the total energy transfer The displaced sapor enters the outlet plenum, then in the coolant loop is t'.te product of the hot leg reenters other steam generator tubes, forcing vapor mass now rate, the average heat capacity of the into the inlet plenum. A density gradient is thus flowing vapor, and the temperature difference established between tubes. This density gradient between the opposing nows at the hot leg nozzle.

then pulls more hor vapor into the tubes, displacing Analyses associated with the Westinghouse experi-cooler steam. The process continues until a steady ments showed that the hot leg mass flow rate is a flow is established, with hot vapor nowing from the function of geometric parameters, the Guld den-inlet plenum to the outlet plenum through some of sity, and the square root of the temperature differ-ence (T,i-T,,i). Thus, the heat transferred by the the steam Fenerator tubes, and cooler sapor return-n ing to the inlet ptenum through the remaining hot leg natuial circulation 00w depends on the tem-

tubes, perature difference at the nozzle, and any interac-Now conrider the flow streams shown in tions that tend to increase the cooler vapor Figure 2. The hot ('i ) and cold (T,) Ouid tempera-temperature will reduce the now rate and the heat n

tures at t hree locations will be examined: the hot leg transfer. Both the mixing in the steam generator norile (1), the ster.m generator end of the hot leg inlet plenum and heat transfer from the hottar (2), and the inlet to the steam generator tubes (3).

vapor abose act to increase the temperature of the At each of the locations, hotter fluid is Dowing returning sapor.

from the reactor sessel toward the steam generator Sin.ilarly, the heat transfer in the steam generator outlet plenum, and cooler Guid is Dowing toward tubes is the product of the mass now rate through the reactor sessel, the tubes, the aserage sapor heat capacity, and the The hot vapor entering the hot leg from the reac-temperature difference (T,3 T,,3). Ileat transfer in n

tor vessel Dows toward the steam generator along the tubes will be affected by interactions that alter the top of the pipe. As it nows, heat is tran ferred either of these temperatures. Again, the mixing in to both the hot leg piping and the returning cooler the inlet plenum tends to reduce T,3, thereby limit-n Ouid stream. This heat transfer occurs in both the ing the heat transfer in the steam generators.

5 Westinghouse natural circulation experiments and Now consider the case in w hich there is no mix-calculations performed with the COMMIX com-ing in the steam generator inlet plenum. The hot puter code in w hich no hot leg structures were mod-sapor temperatures in the hot leg will change little; eled.6 There may also be some mass transfer a lower temperature in the cooler vapor will 5

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increase the amount of heat transferred between interaction with the cooler fluid in the inlet plenum the opposing flow streams slightly, liowever, may result in the condensation of the vapors, either T,3 = T.2. The higher temperature Guid entering on aisting aerosols or as newly generated aerosols.

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the steam generator tubes will result in increased in liquid form, these fission products would be heat transfer to the tubes. The absence of mixing deposited more quickly, and probably in the inlet also means that T,,,

= T,,,, so that the now plenum rather than in the tubes. The countercur-returning through the hot leg is colder. Since the rent now in the hot leg itself may also affect the flow is driven by the temperature difference fission product transport. lf gravitationai settling is between the hot and cold Guid streams in the hot an important mechanism for fission product depo-leg, the mass now will increase. The higher mass sition in the hot leg, fission products falling from now rate willincrease the heat transfer in the loop, the flow heading toward the steam generators slowing the core heatup. liigher temperatures in would enter the return vapor stream, where they the steam generator tubes will also change the rela-would be carried back toward the reactor vessel tive energy deposition between the hot leg and the rather than away from it. This phenomenon is tubes, with more energy being transferred to the be>wd the capability of r.,rrent analytical meth-t ubes.

ods, since the now in the hot legs is considered to be Mixing in the steam generator plenum h a con-one-dimensional. liowever, the magnitude of the trolling parameter for the hot leg natural circula-effect should be calculable since the amount of tion Dow. It limits the mass now in the hot leg by deposition that is caused by gravitational settling increasing the temperature (and lowering the den-

<hould be known from the fission product trans-sity) of the vapor returning from the steam genera-port calculation, tor along the bottom of the hot isg. It limits the heat transfer in the steam generator by reducing the temperature of the hot vapor entering the tubes.

2.3 Loop Flow While accurate modeling of the mhing is impor.

tant in prosiding a realistic simulation of the hot leg now behasior, it is also clear that neglecting the Should the loop seals clear of liquid during the mixing in the steam generator inlet plenum in the transient, loop natural circulation would be rees-analyses will yield steam generator tube tempera-tablished. !n contrast to the natural circulatloa that tures and hot leg mass flow rates that are higher occurs following the initial reactor coolant pump l

thar would be expected in an actual transient, coastdown, the Guid Rowing through the coolant The primary structural consideration associated loops would be superheated vapor. Loop natural with the hot leg countercurrent now is the integrity circulation flow is a buoyancy driven one-of the steam gene:ator tubes. Steam generator dimensional now with heat addition in the core and tubes are very thin compared to the loop or surge heat rejection primarily in the steam generators.

line pipmg, and therefore, are quickly heated if Iloweser, in this situation, heat would be transfer-exposed to high temperature vapor. Should the red to the piping throughout the coolant loops, tubes fail, a direct path outside of cocainment Because of the resulting large vapor density differ-(through the steam line relief s alves) becomes avail-ences and the height of the steam generators, this able to any fission product

• carried in the coolant.

flow is generally large enough that it disrupts any Fission product behavior may also be affected by multidimensional natural circulation nows that the Oaw to the steam generators. An extremely might exist, large surface area is asailable on the steam genera.

The high now rate and large amount of metal ter tubes for deposition of fission products, if the structures available as h.at sinks result in a much tubes remain cool, deposited species may remain slower core heatup. The :!ower heatup rate could there and not be re eased to the containment, if the result in complete oxidation of the cladding before tubes continue to heat up so that revolatilliation any of the Zirenloy melts. Fuel rod relocation occurs, the Dow may simply carry the resuspended would be delayed for several ho,tts. Failure of the j

fission producis to cooler parts of the tubes, where piping anywhere in the RCS is possible, although they would again be deposited. The mixing in the the steam generator tubes would be particularly steam generator inlet plenum may also play a part susceptible because they are much thinner than the in the Ossion product behasior, if gaseous fission hot or cold legt lleating of all the piping will also prmlucts are carried with the hot sapor along the tend to reduce the extent of fission product reten-top of the pipe, the sudden cooling associated with tion in the RCS.

7

1

3. SCOPING ANAWSES s

Several scoping calculations were performed as The sequence of events for the transient is con-the first part of the natural circulation analyses, tained in Table 1. After the transient was initiated.

While these were best-estimate simulations of the decay heat was removed from the core to the steam plant response, they are referred to as reoping cal-generators by a natural circulation How through the cu!ations because the overall system effects of the loops. When sufficient water had been boiled in the in vessel and hot leg natural circulation Dows were secondary side of the steam generators so that the a

4 being investigated, and to distinguish them from remaining liquid was unable to remove the decay the sensitivity analyses presented in the next chap-heat, the RCS began to heat up and pressurize. The ter. The analysis began with a single-channel, once-pressurizer PORVs relieved the pressure by cycling j

through model of the core and upper plenum. This between their open and close setpoints of 16.2 and model provided a basis for determining the effects l$.7 MPa, respectively. When the saturation tem-of the natural circulation nows. Next, the core and perature was reached at 101 min, bolling began in upper plenum model was changed to investigate the the core. As the boiling continued, the liquid inven-effects of in vessel natural circulation on the tran-tory in the RCS decreased until the core began to

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sient response of the plant. For the hot leg counter-uncover and heat up at 130 min. With no source of current now analysis, a model was prepared which water, the heatup continued unmitigated until fuel allowed both in vessel and hot leg natural circula-rod relocation began at about 161 min. The calcu.

l tion Hows. Thus, the approach was to use increas-lation was terminated at 200 m!n.

3 ingly detailed models of the Surry plant to Fuure 3 presents the RCS pressure during the determine the incremental effects of the various transient. The pressure decreased from the steady natural circulation Dows. In each case, the pump state value shortly after the transient began, as the suction loop seal piping was modeled, so that clear.

steam generators were able to remove more energy ing of the loop sesjs could occur. All three of the from the reactor coolant than was being added in

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coolant loops were modeled separately, with one the core. After the reactor coolant pumps coasted r

l containing the pressurizer. Appendh A provides down, natural circulation through the coolant i

j information on the various models used.

loops transferred the decay heat from the core to i

q' The sections below describe results of the scop-the steam generators, and the RCS pressure ing analyses. The single-channel analysis (no multi-remained relatively constant. The small oscillations dimensional natural circulation) will be described in the pressure between 3 and 72 min renected the l

first, followed by the in vessel, in-vessel and hot cycling of the relief valves on the secondary side of leg, and loop natural circulation now analyses.

the steam generator. As the steam generator pres.

This latter section briefly discusses the results from sure decreased, so did the saturation temperature, a prior analysis of the Bellefonte nuclear power increasing the heat transfer from the primary sys-i plant, since t he loop seals did not clear in any of the tem and consequently lowering the RCS pressure.

i Surry calculations. Finally, the results of these After the steam generators dried out, the pressure Surry analyses are compared to one another and to increased until the PORVs opened. The pressure j

the Surry analyses used in draft NUREG il50, in then cycled between the opening and closing set-order to provide insight into the impact of the natu.

points of the PORVs for the remainder of the tran-y I

ral circulation nows on the transient behaslor, sient. For thvse analyses, it was assumed that the l

PORVs could cycle throughout the transient with-i 3,1 Once Thrr) ugh Model

"'fal""8-The collapsed liquid lesel in the pressurlier is l

shown in Figure 4. The level decreased as the tran-A once 4 gh model of the core and upser sient began because heat removalin the steam gen-i j

plenum

• i d in scoping Case I to provide a crators cooled the RCS liquid, causing it to l

basis for evaluting the effects of the various natu-contract, thereby reducing the les el. When the RCS i

tal circulation flows on the plant transient began to heat ca the steam generators dried out, the i

response. This has been the traditional modelins level increased. The les ci continued to increasc until j

approach Ior analyring plant behavior, and was all of the steam in the pressurizer had been relieved used in the source term analyses presented in draft through the PORVs, at about 90 min. The pressur-l NUREO il50.

lier remained liquid full until after boiling began in I

i s

t i

e l

l Table 1. Sequence of events for scoping Case 1 Time Event (min)

Transient initiation 0

l PORY cycling begins 71.8 Steam generators dry 75.4 77.2 Hot legs reach saturation 100.6 loop natural circulation flow ends 109.7 Core heatup begins 129.6 Cladding oxidation begins 144.1 Control rod relocation begins 157.2 Accelerated oxidation begins 157.9 Fuel rod relocation begins 160.5 Calculation ends 200.0 i

17 16 e

a g 15 i

14 1

0 50 10 0 150 200 1

Time (min)

Figure 3. Pressurlier pressure for scoping Case 1.

)

9

. =.

F L

15 i

l l

[

top of presserlaer-go e>

eere heetup begins

.2

/

o 3

D*

$ (

l

]

RCS heetup begine l

fuel red relocation begin 0

i 0

50 100 150 200 Time (min)

F Figure 4. Pressurlier collapsed liquid level for wping Case 1.

a I

I I

the core. Steam then entered the pressurizer, rees.

lower plenum by heat transfer from relocated molten tablishing the liquid level. The level decreased control red material then Cowed through the core i

through the rest of the transient, but some liquid bypass rather than through the core. The cladding tem.

[

remained in the pressurizer as the core damage peratures in the top half of the core dropped to nearly i

occurred.

1000 K after relocation began for socral reasons.

l Figure 5 shows fuel rod cladding surface tempera-Cladding oxidation uopped at eloations from which

].

tures at four of the ten axial nodes, including the top material relocated because all of the unoxidized Zirca-l and bottom nodes, from 127 to 200 min. The top of loy Howed to lower cloations. Besides the lack of heat i

l the core began to heat up shortly before 130 min, and a generation from oxidation, heat was being conweted l

top < lown dr>out of the core followed. The tempera-to the core baf!'.e plates and core b>rass region. Also, i

tures increawd more rapidly as the cladding began to because of tLt now blockage,little superheatej stenm oxidize, and continued to increase until the tempera-was rising from '.he lower part of the core to heat the ture reached 2500 K. When trat temperature was cladding in the upper part. Temperat urer in the bottom i

attained, the oxide shell on the cladding was anumed of the core continued to increase because oxidation was

[

]

to rupture, initiating the relocation of molten material still taking place. At about 192 min, the upper part of

{

that had been contained within the cladding. This the core began to heat up again, the result of higher l

l materia! Howed down the outside of the fuel rods until temperatures in the core bypass region, which reduced

]

it frore at lower cloutions, it should be noted that the heat tramfer from the upper part of the core so that

[

i when fuel rod relocation beg:.n in the top part of the the decay heat was not being removed. The changes in 7

i core at about 161 min, the bottom part of the core the transient behasior assodated with the coheshe I

j contained enough water so that the cladding was still at debris formation demonstrate the limitations of analy.

the saturation temperature. All of the material froze in ses that model the core as a single channel; had more l

Ihe two adal nodes in the middle of the core, fonning a than one channel been modeled, Aow could haw pro-coheshe debris at about 162 min. The coheshe debris ceeded through the core around a coheshe debris in 1

caumi the flow area to be reduecd to leu than l of one region of the core.

[

]

the original axlal now area nearly stopping steam now The coheshe debris was composed of Ziresloy j

through the core. Steam that was being generated in the and dissobed fuel from upper parts of the core,

[

i t

10 l

t

3000 i

i i

Height obove core bottom

)

a 0.is m

/,,,,,, n,,

O 1.28 m Q

4 2.01 m v

X 3.47 m 8

$* M*

M 2

3 0

e n

,000

=

O O-m m

i e

12 0 140 16 0 14 0 200 Time (min) h i

Figure 5. Fuel rod cladding surface temperatures at 0.18.1.28,2.01, and 3.47 m above the core bettom j

for scoping Case 1.

i

)

and contained 27.90s of the zirconium and 2.$re of Fission product release to the coolant began the fuelinitially in the core. At the end of the calcu-when the cladding oxide shell was breached at t

lation,90re of the control rods had relocated, as about 161 min At the end of the calculation, more had 35.70e of the cladding and 7.2re of the fuel, than 15fe of the xenon and krypton,13re of the i

The total hydrogen generation rate in the core cesium, !!re of the iodine, and 17?e of the tellu-during the transient is show n in Figure 6. The rate rium that was originally in the fuel had been of hydrogen production steadily increased as more released.

of the cladding was being oxidized. After fuel rod The mass now rate through the core bypass is relocation began, the coheshe debris formation presented in Figure 7. As the core uncovered, the i

and associated low core now reduced the cladding density difference between the Guld in the core and oxidation. The incieased generation rates at about that in the bypass caused a natural convecticn now 177 and 186 nin renected higher steam flows to be established. Steam flowed up through the core i

entering the core, which were caused by relocated to the upper plenum, and a sm.ill amount of steam core material b iiling water in the lower plenum. At returned to the lower plenum through the core the time of fuel rod relocation, about 97 kg of bypau. This now pattern was maintained until the hydrogen had been produced, corresponding to cohesive debris was formed in the core. After that, oxidation of about 14re of the Zircaloy in the core.

How through the bypass vias from the lower plenum i

The hydrogen generated after relocation began to the upper plenum. Increases in now occurred would probably be greater than that calculated by when molten material from the control rods and SCDAP/RUl.AP5 because the code does not cal-fuel rods relocated to the lower plenum. This mate-culate the oxidation of material while it is relocat.

rial boiled some of the liquid remaining fn the lower ing. There was sufficient steam in the core to allow plenum, and the resulting steam nowed through the oxidation throughout the transient. The maximum bypass to the upper plenum and hot leg.

i Sydrogen mole fraction in the vapor of 0.94 Figure 8 presents the collapsed liquid levella the l

occurred in the top of the core just before cladding itactor vessel, whlch decreased throughout the i

1 relocation began, transient. The ic5cl decrease slowed as less water l

11 l

1 i

e i

e O

L 0.8 C

O O

o,g EN

.a m3 V

0.4 C.*

On O

u l

v 0.2 x

a e

i 0

i 12 0 140 14 0 14 0 200 Time (min) r Figare 6. Total hydrogen generatbn rate ftv scoping Case 1.

60 s

relocallon to g

lower plenum I

j 40 v

i e

c

[

]

+=

0 20

\\

y 0

C:

o I n u.t i j $>*i_i1 AL t t_

v.

_1 a

2 I

I

[

-20 12 0 14 0 16 0 140 200 I

Time (min) l i

j Figure */. Mass flow rate through the core bypass for scoping Case 1.

I

?

i F

l f

1:

4

~

8 i

i i

L top of core j 4 v

e 4

releoetlen to i

V

/ lower plenum bottom of eere "

g l

4 J

2 bellom of core berrel -

I 8

i g

[

i 12 0 14 0 16 0 14 0 200 j

Time (min) j Figure 8. Reactor sessel collapsed liquid level for scoping Case 1.

1 1

i remained in the core, nearly stopping after the core the saturation temperature throughout the core l

dried out. The incremental level decreases after heatup. None of these structures would be expected

?

i 167 min occurred when molten material dropped to fall before the bottom head of the vessel is i

from the core to the lower plenum, causing some of breached. These results are typical of calculations the remaining water to boll. at the end of the calcu.

that ne, 'ect natural circulation dows.

l lation, the lower plenum was completely filled with

?

steam, and both the two-phase and the collapsed liquid lesels were in the lower head.

3.2 In Vessel Circulation The temperature of the hottest structure in the upper plenum is shown in Figure 9. This stainless i

steel structure is located just above the cose. The in vessel natural circulation is simulated with j

temperature steadily increased until the fuel rod SCDAP/RELAPS by modeling parallel now chan.

i

{i relocation began. The temperature then decreased nels in the core and upper plenum, connected radi.

from its maximum va!ue of about 1200 K because ally by crossflow junctions. The crossflow I

junctions neglect certain momentum Oux terms in i

i virtually no vapor was Dowinst through the core, the momentum equations. Neglect of these terms Cooler vapor Dowed to the upper plenum through has been shown to have a second order effect on l

the core bypass, reducing the upper plenum struc*

calculated results for applications such as natural

]

ture temperature. The cooling continued until near cin;ulation. A discussion of these terms is incicied the end of the calculathn, when the vapci Gowing in Appendix C. Three parallel channels were used through the bypass was hot enough to again cause to model the core and upper plenum. In the folicw.

the structure to heat up. The other structures in the li.,.tiscussions, the center core channel refers to a upper plenum remained at or near the saturation grouping of 25 fuel assemb!!cs in the center of the i

j temperature throughout the trans)ent.

core, the otter core channel refers to the 36 fuel j

Figure 10 shows the highesi structure average asser.Nies on the core periphery, the middle chan.

]

temperatures of the piping in the pressurlier loop nel refers to the 96 fuel assemblies located between i

hot leg, surge line, and steam generator tubes. The the center and outer channels, and the inner chan.

temperaturrs of all three structures remained near nets refers to the center and middle channels l

i

)

15

?

i i

. ~.

1200 i

i i

/

2 v

fuel red relocation e

6s 1000 6e

1. 803 k

6 i

e>

f 4

600 12 0 14 0 WO 14 0 200 Time (min)

Figure 9. Volume aserage temperature of the hottest upper plenum struture for scoping Case 1.

I 1

4 700 i

i O Hot leg mx O surge line V

6 Steam generaler tubes e

6 475 3

e O

l 6

e I

i O,

l E 850

~

i

(

(

O425 Mk i

A^^

s e

' p e

m f

I I

600 l

12 0 140 M0 540 200 Time (min)

Figues 10. Volume avinge temperatures of the hottest part of the hot les pipe, surge line, and steam j

generator tubes for scoping Case 1.

(

i 4

I

[

14 I

j i

l l

l l

together. More detail on the modeling of the chan.

that temperature, a change in the calculated oxida-nels is presented in Appendix A.

tion kinetics resulted in a much more rapid rate of The scoping Case 2 calculation was initiated at cladding oxidation and a more rapid heatup. The about 127 min, shortly before the core began to temperature then increased quickly to 2500 K, and uncover. The initial conditions for the calculation relocation of molten cladding began at about were taken from the Case I calculation at 127 min.

167 min at the top of the core. When the molten The sequence of esents, presented in Table 2, was Zircaloy relocated, no unoxidized material was left nearly the same with in vessel natural circulation in the upper part of the burJie. With no additional modeled as without, but the damage progressed at heat being generated from the oxidation reaction, a slower rate. The core heated up at a slower rate the cladding temperatures decreased, esentually because energy was being transferred by the natural increasing again when the sapor flowing past was convection flow from the core to the structures in too hot to remose the decay heat. The cladding the upper plenum. Higher vapor temperatures in temperatures at the top of the core in all three chan-the upper plenum also led to heating of the hot leg nets are show n in Figure 13. The rates at which the and surge line piping in the pressurizer loop, three regions of the core heated up renected the Figure 11 shows the Dow velocity vectors in the radial pow er distribution. The temperature inerease core and upper plenum at about 167 min, shortly from 1850 to 2500 K was slower in the outer chan-before fuel rod relocation began. There t.re four cir-nel than in the two inner channels because the oxide cul ting nows. There is a now from the center of shell was thicker. This was the result of the slow the core rising into the upper pler.um, where it heatup, w hich allowed osidation to occur for about turns toi trd the outer part of the plenum before 17 min longer in the outer channel before the oxi-descendl. ; back into the core through the periph-dation kineties changed. The center channel was cral fuel assemblies. A second flow exists la the the first to relocate at approsimately 167 min, fol-1 upper plenum, with vapor recirculating from the lowed by the middle channel at 171 min and the periphery of the plenura toward the center, above outer channel at about 173 min. The relocation in the top of the core. This flow is not obslous in the the middle and outer channels cfcurred at lower figure because the large radial flow areas in the elevations than those shown in Figure 13. Reloca-upper plenum result in sery low radial velccities. A tion at the top of the core in these channels did not small amount of vapor flows across the top of the occur because the cladding was completely oxi-core from the middle to the outer channel because dized before the temperature reached 2500 K. The 1

the fuel assembly upper end boses present a hrge decreasein temnerature at the bottom of the core at asial flow resis:ance at the core exit. The fonth 183 min (seen m Fpre 12) was caused by reloca-flow recirculates vapor from the upper plenutn to tion of molten material to the lower plenum Lig-thelower plenum through tis ecre b> pass regior.. A uid was boiled by the relocated material, and the masimum selocity of about 1.2 m/s occurs in the resulting steam ibw cooled the lower part of the upper plenum. Although the velocities are nearly core.

equal, the recirculating mass is higher in the upper The effects of the in-sessel natural circulation plenum than in the core because of the larger now flow on the temperature distribution are esident in area. About 15 kg/ sis leaving the top of the core in Figure 14, w hich presents the cladding surface tem-the center and middle channels, and about 10 kg/s peratures in the outer (low power) channel at the is returning to the co.e through the outu channel, same clewions as presented in Figure 12 for the The flow through the bypass is just oser 2 kg/s.

center channel. Unlike the two inner channels, in The flow recirculating in the upper plenum is owr which the temperature increased with increasing 32 kg/s, approximately three times greater than the elevation in the core, the outer channel temperature flow reciter'ating between the core and upper increased from both the bottom and top of the core

plenum, toward the middle. The tiow returning to the core l

Fuel rod cladding surface temperatures from from the upper plenum was heated as it nowed I

foar elevations in the center (high power) channel dow n, and the flow entering from the bottom of the i

j are presented in Figure 12. Similar trends to those core was heated as it flowed up, resulting in a maxi-I j

of scoping Case I are present. A top-down dryout mum cladding temperature just cbove the middle of the core led to a heatup of the entire core. The of the core. The cladding at this eles ation relocated L

core was completely uncoscred by about 157 min.

at 173 min, shortly after the cladding at the top of The temperatures increased at a sate of about the middle channel relocated. This was the only ele-0.6 K/s until the temperature reached 1850 K. At vation at w hich cladding relocation oecurred in the I

l f

i l

)

l

[

Table 2. Sequence of events for scoping Case 2 Time Event (min)

Transient initiation 0

Calculation begins 126.7 Core heatup begins 129,8 Cladding oxidation begins 146.3 Control rod relocation begins 166.2 Fuel rod cladding balloons 166.7 170.0 +

Fuel rod claddiag fails 167.0,170.0 +

Acceleraica oxidation begins 167.2 Fuel rod relocation begins 167.3 Calculation ends 200.0 i

l i

1 i

16 4

Top of upper plenum r

.....I...o.....................

l 4

i

\\

1 a

l

/

Hot leg

.......~...................-

1 i

\\

Top of

.....L......../..............

i i

Core J

....T.....T.....

m.. \\........

lL 1

j

.... Y...

" 0,5 m/s l

o

....T....

s i

s a

1,

...................../.....

t s

I

/

s Core j

bypOSS q

i r

s 1

l Bottom _...../...........'.......'..........

I Of Core i

.1 i

I Vessel C P394-LN87017-4 1

Fleuro 11. Vapor velocity vectors in the core and upper plenum at 167 min in scoping Case 2, s

1 1

j i

17 4

I.

_ _ _ _ -.,,.,,.. _7._-

,-7__.,,

_y g

l 1

fleight obove relocallon core bottom

//

\\

j O 0.18 m

/

\\

O t.2a m mx 6 2.01 m v

X 3.4 7 m l

E 2000

f*dV!n'*d 3

l

)

8 a

h 1000 H

O O

C-w relocotten te lower plenum 0

i i

12 0 14 o WO 14 0 200 Time (min)

Figure 12. Center channel fuel rod cladding surface temperatures at 0.18,1.28,2.01, and 3.47 m above the bottom of the core for scoping Case 2.

3000 i

i ggyg r oleeellen l

O cent er channoi O ulddle channe

/

f N

6 Outer channel l

m5

,j

, 2000 s

3 O

Le G.

WM H

l l

O i

i i

12 0 14 0 WO 14 0 200 Time (min)

Figure 13. Fuel rod claddin2 surface temperatures at the top (3.47 m abose the core bottom) of the three core channels for scoping Case 2.

18

l 3000 i

i i

Height obove Nt en core bottom r elocollon O o.ta m O t.28 m Q

6 2.01 m v

X 3.47 m k

2000

=

u 3

)

• =

O 6e Q.

~

M

~

F.

OM3

=

C O-m l

i l

0 12 0 14 0 16 0 14 0 200 Time (min)

Figure 14. Outer channel fuel rod cladding surface temperatures at 0.18,1.28,2.01, and 3.47 m above the bottom of the core for scopin? Case 2.

i J

i 1

outer channel; the cladding at the other elevations 20re. However, in this case, the ballooned cladding i

was completely oxidized before the temperature temperature ranged between 1560 and 1800 K, and

]

reached 2500 K. This was the result of the lower the rupture occurred at a temperature of 1864 K.

power in the outer channel, which led to a slower Balloonin did not occur in the outer channel heatup with more time to oxidize the cladding, because the fuel rods were at a lower average tem.

Some ballooning of the fuel rod cladding was perature than those in the inner channels, resulting predicted to occur. The fuel rods in the center chan-in a lower gap pressure. No changes in the flow nel ballooned and ruptured at about 167 min. The patterns were observed as a result of the ballooning l

4 l

balloon extended over the upper 70re of the fuel because of the relatively small flow blockages, rod length, and caused a flow area reduction of It is questionable whether cladding ballooning j

about 15%. The ballooning occurred with the clad-would occur at these elevated temperatures because l

l Ang temperature varying from 1475 to 1775 K oser of interactions between the Zircaloy cladding and the length of the affected cladding. The cladding the Inconel grid spacers. Zirconium forms eutectics rupture occurred at a temperature of 1481 K. The with both iron and nickel that, depending on the rupture occurred at this lower temperature because composition, melt at various temperatures well the higher temperature regions of the cladding had below 180G K.7 Experiments at KfK have shown been oxidized more. (The cladding at the top of the pinhole cladding failures at grid spacers, generally core had more than twice as much oxide as that in by the time the temperature reaches 1500 K.8.9 Ifa the middle of the core, where the rupture eutectic is formed, a hole in the cladding may occurred.) Although it was at a hlgher temperature, result, allowing the gap and coolant pressures to the more heavily oxidized cladding was stronger equalize. With no pressure difference across the and did not rupture. Similar behavior was observed cladding, ballooning would not occur. Formation in the middle channel, in which the fuel rod clad-of a cladding / grid spacer eutectic would also be ding ballooned and ruptured shortly after 170 min, influer.ced by other factors, such as the heatup rate i

Again, the balloon extended over the upper 70% of and the thickness of the oxide shell on the cladding, j

the cladding, with a flow area reduction of about Time is required for the eutectic composition to 19

form and fail the cladding. The presence of an some of the liquid and cooled the pipe. The liquid oxide shell inhibits the chemical interdiffusion did not reach the hot leg before the PORV' opened required to form the eutectic composition.

again, drawing the liquid back into the p essurizer.

Over 400 kg of hydrogen was geneated over the The liquid was slowly boiled from the pressurizer as course of the transient, corresponding to oxidation the PORVs cycled and as it interacted with the hot of57re of the Zircaloy in the core. The maximum vapor entering through the surge line. All of the hydrogen mole fraction in the core was about 0.80, liquid was gone at about 198 min.

so that the oxidation reaction was never steam-During the time that liquid remained in the pres-starved. The top 60% of the fuel rod cladding in the surizer,it may have helped to scrub fission products outer channel was completely oxidized at the end of from the effluent entering the containment through j

the calculation. 7ircatoy had melted and relocated the PORVs. Ilowever, few fission products were frori nodes 410 in the center channel and from released during this part of the transient. The nodes 4 8 in the middle channel, but only from the release of fission products began when the cladding sixth axial node in the outer channel, llowever, fuel first ruptured, at about 167 min in the center chan-liquefaction occurred only in the outer channel, nel. At the end of the calculation,less than 4We of w here the cladding did not balloon. Since the clad.

the xenon and krypton, 3re of the cesium and ding did not balloon, the molten Zircaloy on the iodine, and $re of the tellurium that was originally inside of the cladding was able to make contact in the core had been released from the fuel rods to with and dissohe some of the fuel pellets. This mix-the coolant.

ture (52re zirconium, 48re uranium dioxide by mass) then Gowed down the fuel rod when the zir-conium dioude shell on the cladding breached. In 3,3 Hot Leg Countercurrent Flow the inner channels, molten Zircaloy only existed in regions w here the cladding had ballooned. The fail.

A scoping calculation was performed in which ure of the cladding after ballooning introduced hot leg countercurrent now was modeled (Case 3).

steam to the gap, allowing the Zirealoy on the The results of CONIMIX analysesll were used to inside surface of the cladding to oxidize, in the cal-guide the modeling of the now. How this was done culation, this inner oxide layer presented dissolu-is discussed in Appendix A.

tion of the uranium dioxide by the molten Zircaloy.

As in the Case 2 calculation, conditions gleaned The highest piping temperatures in the pressur-from Case I at 127 min were used as initial condi-irer loop hot leg, surge line, and steam generator tions. The transient calculation began at 127 min tubes are shewn in Figure 15. The temperatures in as the core began to uncoser, and continued until the other loops were lower than those in the pres-180 min, shortly after fuel rod relocation had surizer loop because the cycling PORVs drew Dow begun. The pressure during this part of the tran-from the reactor sessel to the pressurizer. Based og sient was controlled by the pressuriier PORVs, creep rupture considerations, which indicate that cycling between 15.7 and 16.2 MPa. The core structural failure becomes a concern at tempera-began to heat up at 130 min, and was completely tures above about 1000 K,10 both the hot leg noi.

uncovered by 157 min.

zie and the surge line piping would be candidates Figure 17 presents the mass now rate in the top for failure. This failure would probably occur of the hot leg in one of the non-pressuriter loops. A before the reactor sessel bottom head fails, and fairly steady natural circulation now was estab.

would not be predicted by once through calcula.

lished by about 133 min, and the magnitude tions, llowes er, the steam generator tubes were still decreased slowly as the transient progressed. The near the saturation temperature, so that failure of now through the hot leg was interrepted by the the tubes would not be expected, opening of the PORVs. When the PORVs opened.

Figure 16 shows the collapsed liquid leselin the sapor was drawn toward the pressuriier surge line pressurizer, which illustrates that the liquid did not from throughout the RCS. This temporarily drain quickly into the hot leg. When the PORVs reduced the Dow proceeding from the reactor vessel were open, the vapor s elocity through the surge line toward the steam generators, and increased the was high enough to prevent the liquid in the pres-flow from the steam generators to the vesselin the surizer from draining into the surge line. The hot loops without the pressuriter. In the pressurizer du;. also heated the piping. When the PORVs were loop, the now was drawn toward the surge line closed, liquid did begin to drain into the surge line, from both the reacter sessel and the steam genera-licat transfer from the hot surge line piping beiled tor. The now through the steam generators was also 20

I i

1500 i

i i

O Hot leg

)

mw O Surge line i

V A Stoom generef or tubes i

e i

6 3

3 O

6 fuel red relecellen begine e

f DM

~

e

.e.

I Le 4

i l

12 0 14 0 WO 14 0 200 j

Time (mln)

Figure 15. Volume average temperatures of the hottest part of the hot leg pipe, surge line, and steam generator tubes for scoping Case 2.

J l

4 1

to j

l i

l n

i E

i v

j

-e>e S

D releeellen begine

/

l J

l

\\

l 0

12 0 14 0 40 14 0 200 Time (min)

Figure 16, Pressurizer collapsed liquid lesel for scoping Case 2.

21 l

10 i

n(

PORVe eleoed On rel M

l Ife* F f

' E h II*fdNN dyrLp 0

l l

l ll

'i If 5-0 o

}

PORVs apen 502

.m 12 5 14 5 Time (min) 5 16 18 5 Figure 17. Mass Dow rate in the top of a non pressurizer loop hot leg for scoping Case 3.

affected by the PORY opening, although there was peratures from the highest temperature regions of always some now proceeding from the inlet plenum each of the three core channels. The temperature to the outlet plenum. When the PORVs closed, the differences renected power differences between the natural circulation Dows in the hot leg and steam three core regions. The maximum temperature in generator tubes were quickly reestablished.

the outer channel was near the center of the core Figure 18 shows the fuel rod cladding surface because of the in vessel natural circulation flow, temperatures at four of the ten asial nodes in the The vapor was heated by fuel rods as it flowed center channel. The temperat ures steadily increastti down the outer core channel, from saturation to about 1500 K. when the heatup Oxidatien of the Zirealoy cladding began at rate increased as the rate of cladding oxidation 148 min. The oxidation rate gradually increased as increased. Wh(n the temperature reached 1850 K, the ten'perature increased. When cladding reloca-the osidation kinetics changed, and the tempera-tion began, nearly 34 kg of hydrogen had been gen-ture increased rapidly to 2500 K. At that point, the crated. At the end of the calculation, 52 kg had cladding oside shell was assumed to be breached, been produced. The oxidation reaction was at no allowing molten unoxidized Zircaloy to now from time steam starsed, inside the cladding down the otside of the fuel During the nrst part of the heatup, the cladding i

rods; no fuel liquefaction occurred because the was collapsed onto the fuel pellets. Howeser, the cladding had ballooned. The ballooning and rup-continued hestup of the fuel rods allowed the gap ture allowed the inside surface of the cladding to pressure to inercase above the coolant pressure, at d l

develop an otide shell, which presented molten the cladding ballooned and ruptured. Ballooning Zirealoy from coming into contact with and dis-began in the center channel at about 176 min. The solving the fuel. With no further material to oxi-balloon extended oser the top 70re of the fuel rod dire, the heat generation at these elevations length, and caused a now area reduction of about j

decreased, as did the temperature. Relocation 60r. The temperature of the cladding in the e

began at 178 min in the center channel; no reloca-affected region ranged from 1230 to 1510 K. The j

tion had occurred in the other channels by the end cladding ruptured about I min after the ballooning i

of the calculation. Figure 19 presents cladding tem-occurred, at a temperature of 1567 K. No changes 22

l 5000 i

i l

Helght obove l

core bottom O 0.18 m r elecoll on--

O 1.25 m Q

6 2.01 m i

v X 3.47 m 2*o

.e 4 Lei.d i

oe L

3 r

O L

h

• 000 om 0

12 5 14 5 16 5 18 5 Figure 18. Center channel fuel rod cladding surface temperatures at 0.18,1.28,2.01, and 3.47 m above i

the bottom of the core for scoping Case 3.

l l

j i

3000 i

i Height above core bottom O center chornet 3.47 m releco vion

_ a O widdle chonnet, 3.47 m Q

A Ou t e r chonne l. 2.01 m v

r

,,,, g,,,,,4 _.

2WO e

oxidollon u

3 i

O L

e et w

j E.

3000 E

i i

n 12 5 14 5 14 5 18 5 Time (min)

Figure 19. Fuel rod cladding surface temperatures from the three core channels for scoping Case 3.

l 23

in the flow patterns in the core were caused by the now carried hot vapor to the steam generator ballooning, but some of the vapor Dowing through tubes, causing them to heat up as well. However, the center channel was diserted around the bal-the temperature of the tubes was still low enough looned region, that they would not be expected to fail. The largest The cladding rupture also initiated the release of temperature difference between the inside and out-fission products from the fuel rods to the coolant.

side surfaces of the pipe at 180 min was 97 K in the Decause the calculation ended shortly after fuei rod hot leg,25 K in the surge line, and 0.2 K in the relocation began in the center channel, most of the steam generator tubes.

core was at lower temperatures. The low tempera-tures resulted in the release of less than 0.3% of the noble gases, cesium, iodine, and tellurium by the 3,4 Loop Flow end of the calculation.

Approximately 210 of the energy that was Natural circulation now through the coolant removed from the core during the heatup was loops was not calculated in the Surry analyses.

deposited in the coolant loops nlong the path of the However, previous analyses of sesere accidents in hot leg natural circulation now Of this,18% was the Bellefonte plantl2 showed that natural circula-retained in the hot leg piping,6re w as deposited in tion through the loops reduced the heatup rate of the steam generator plena,19re was transferred to the core by an order of magnitude, to about the steam generator tube sheets, and 24re was 0.06 K/s, because all of the piping in the RCS was stored in the steam generator tubes. The remaining available as a heat sink. The pipe temperatures 33re was transferred through the steam generator lagged behind the maximum fuel rod cladding tem-tubes to the steam and structures on the secondary perature by about 100 K. Such high temperatures in side of the steam generators. The pressuriier surge the loop structures would cause the RCS pressure line abso: bed tre of the heat remosed from the boundary to fait long before the fuel rods began to core-relocate, probably in the steam generator tubes The effect of this energy transfer on the structure s nee they are the thinnest structures. Although temperatures in the loop is seen in l'igure 20, w hich Bellefonte is a Babcock and Wilcox plant with presents the highest hot leg, surge line, and steam once-through steam generators, similar results generator tube solume average temperatures in the would be expected in a U tube steam generator pressuriter loop. The hot leg temperature decreased from the reactor vessel to the steam gen-plant, such as Surry, should the loop seals clear of I

liquid, erator. The steam generator tube temperature also steadily decreased from the inlet plenum to the out-let plenum along the hot now tubes, and then from 3.5 Result Comparisons the outlet plenum to the inlet plenum along the cold i

Dow tubes. The surge line temperature increased more rapidly w hen the PC R\\. wcre open because Results of the Surry natural circulation calcula-there was more now, and hence more heat transfer, t ons presented in the presious sections will be com-through the surge line. The oscillations in the steam pared with one another. This will provide an generator tube temperature were caused by the indication of the effects and importance of model-POR\\ cycling. When the POR\\ was open, the now ing the various natural convection Dows that may through the steam generator end of the hot leg exist during a severe accident. The results of these l

res ersed. W ith no source of hotter vapor, the vapor analyses will then be compared with the Surry sta-in the steam generator mlet plenum cooled because tion blackout analyses performed for draft heat was transferred to the structures in the plenum NUREG-1150. This comparison will proside some and because cooler sapor was entering the p;enum insight into differences in the timing and extent of from the tubes. Thus, the sapor entering the hot core damage that result from the models used to How tubes was cooler, a id the tube temperature analyze the plant behasior.

decreased. As in scoping Case 2, the average temperature of the surge line was higher than that 3 5.1 Effects of Natural Circulation Flows. The of the other structures in the loop, and was high inclusion of the sarious natural cir;ulation Dows enough at the end of the calculation that creep rup-into the model used to calculate the ThlLB' tran-ture failure of the surge line could occur before sient did not affect the esents that occurred up to reactor sessel failure. The presence of the hot leg the early putt of core relocat on. Rather, thenc i

24 1

1200 i

i O Hot les m

l x

0 Surge Itne l

v 6 Stoom generator tubes i

e 6

3

~

• • 1000 we f

P0ftVe open q

O

~

~

o 6e>

l Poltve slooed 0

i i

400 12 5 14 5 16 5 18 5 Time (min)

Figure 20. Volume-average tempcatures of the hottest part of the hot leg pipe, surge line, and steam i

generator tubes for scoping Case 3.

f

)

i flows served to extend the transient by removing ingful to examine the state of the plant at the time a l

energy from the core and transferring it to other specific event occurs. The latest common event in

[

parts of the reactor coolant system. The ultimate the three analyses was the beginning of fuel rod i

l effect of these Dows on the plant behavior cannot relocation.

be completely o aluated until the transient proceeds Table 4 presents several major parameters that I

to failure of the reactor vessel bottom head. At that describe the state of the Surry plant at the time fuel I

time, differences in RCS pressure, melt compost-rod relocation began in the three scoping calcula-I tion, and fission product distribution will affect the tions. Fuel rod relocation began in the scoping cal-containment response and ultimately the source culations when the cladding in one of the ten axial term resulting from the transient, nodes reached a temperature of 2500 K.

The most significant effect of the natural circula-Although they had been exposed to superheated tion !L. was ine heating of ex vessel structures, vapor for a longer time, the hot leg and surge line which was not seen in the once-through calcula-temperatures were only slightly higher when fuel j

tion. The heating was sufficient so that failure of rod relocation began in the hot leg natural circula-l the RCS boundary would occur before the sesselis tion case than with only in vec:<l circulation mod.

l breached, probably in the surge line. Such a failure eled. Although more energy was transferred out of j

1 may allow the RCS to depressurire before the s essel the core, the participation of both the steam gener.

l fails, perhaps sufficiently to significantly change ators and ali three coolant loops (rather than just i

the containment load at the time of melt ejection, the pressurlier loop) as heat sinks resulted in a 1

The higher piping temperatures will also affect the more even distribution of the energy throughout j

j fission product behavior in the RCS.

the RCS.

Table 31ists the sequence of major events for the The temperature difference between the center three scoping calculations. Since the addition of and middle channels caused by the different aver-

)

natural circulation flow s extended the core heatup, age power was accentuated by the natural comec-I a comparison of the three analyses at a specific tion flows. The cooling provided by the coolant I

time is not very meaningful. Ilowever, it is mean-loops in Case 3 reduced the temperatures of both I

j 23

I Table 3. Sequence of events for the scoping calculations Time (min)

Event Case 1 Case 2 Case 3 i

j Transient initiation 0

0 0

i Calculation begins 0

126.7 126.7 PORY cycling begins 71.8 Steam generators dry 75.4 77.2 r

Hot legs reach saturation 100.6 L

i loop natural circulation 11ow ends 109.7 5

i Core heatup begins 129.6 129.8 129.6 i

i F

Cladding oxidation begins 144.1 146.3 148.0 Control rod relocation begins 157.2 166.2 178.2 Fuel rod cladding balloons 166.7 170.0 +

176.4

)

Fuel rod cladding fails 167.0,170.0 +

177.3 f

Accelerated oxidation begins 157.9 167.2 178.0 I

l Fuel rod relocation begins 160.5 167.3 178.3

}

Calculation ends 200.0 200.0 180.0 i

1' r

I i

i I

i I

i i

26 I

l I

I

Table 4. Conditions when fuel rod relocation began in the three scoping calculations f

Parameter Case 1 Case 2 Case 3 Time (min) 160.5 167.3 178.3 liydrogen generated (kg) 96.9 47.2 33.7 hiasimum middle channel fuel 1747 1546 cladding temperature (K) hlaximum upper plenum structure temperature (K) 1100 I?48 1153 llot leg nonle temperature (K) 633 789 829 hiaximum surge line temperature (K) 637 973 1001 51asimum steam generator tube temperature (K) 624 629 731 Reactor vess(Iliquid level (m) 3.26 2.59 2.1i Pressurizer liquid level (m) 4.03 3.07 1.99 Core outlet flow (kg/s) 1.2 10 11 Core return flow (kg/s) 8 8

Upper plenum recirculating flow (kg/s) 38 49 the cose and upper plenum structures compared to occurred, and the molten Zircatoy was able to dis-Case 2. When fuel rod relocation began in the cen-solve some of the fuel pellets, llallooning did not ter channel, the maximum cladding temperature in occur because the axial aserage temperature of the the middle channel was 1747 K in Case 2, and fuel rods was low enough that the gap pressure 1546 K in Case 3. Similarly, the hottest upper never exceeded the coolant pressure (the bottom of plenum structure was 100 K cooler with the hot leg the core was still submerged in liquid wben the top countercurrent flow considered. The lower clad.

of the core was relocating). With in-vessel natural ding temperatures also resulted in less hydrogen circulation modeled (Cases 2 and 3), sausage-type j

hasing been generated when relocation began, ballooning occurred in the higher powered regions The amount of hydrogen generated when fuel of the core. With an oxide shell on the inner surface l

rod relocation began was about 97 kg in Case 1, of the cladding, no contact was made between the 47 kg in Case 2, and 34 kg in Case 3. This compar.

fuel pellets and the molten Zircaloy in the cladding.

Ison must be viewed with the understanding that Thus, no dissolution of the fuel occurred and only different amounts of material are at high tempera-Zircaloy relocated in the inner core channels.

I ture when relocation begins. In Case 1, the once-Reactor sessel collapsed liquid levels from the through calculation, all of the fuel rods in the core three cases are presented in Figure 21. The liquid j

are af fected. in the other two cases, fuel rod reloca-level decreased more rapidly with inaessel natural tion begins in the 25 fuel assemblies in the center circulation modeled because sapor was being recir-core channel, with an accordingly smaller amount culated to the lower parts of the core and lower of material being relocated, and most or the core plenum, accelerating the boiling of the liquid that I

still undergoing oxidation.

was there. The liquid lesel in the once-through cal-The core damage was affected by the behas tor of culation esentually decreased to the lesel of the l

the fuel rod cladding. In Case 1. no ballooning other two calculations as molten core material l

27 l

8 i

O cose t O cose 2 top of core A cose 3 T4 v

~

In 4

?

m g

bottom of sert **

3 2

bottom of eere berrE i

i i

12 0 14 0 10 0 ISO 200 Time (min)

Figure 21. Reactor vessel collapsed liquid level for the three scoping cases.

relocated to the lower plenum, where it boiled the Table 5 presents the sequence of events for the l

remaining liquid.

draft NUREO-l150 analysis (obtained from Refer.

The pressurlier collapsed hquid lesel from the ence 13), the three scoping calculations presented j

three calculations is shown in Figure 22. More lig-earlier, and a MELPROO calculation.14 The tab!e i

uld was retained in the pressurizer in Case I shows that the transient wa extended both by the because the vapor nowing out the surge line was inclusion of the naturalcirculation ' lows and by the near the saturation temperature throughout the more detailed modeling provided by the SCDAP/

transient. In contrast, the amount of liquid RELAP5 and MELPROO codes. Scoping Case i decreased more rapidly with the natural circulation was similar to the MARCH calculation used in Dows modeled because superheated vapor was draft NUREG ll50 in the hydraulic treatment of flowing into the pressuriser, w here it could transfer the core. Yet the SCDAP/RELAPS calculation heat to the remaining liquid. Also, the heating of reached the temperature at which fuel rod reloca-the surge line pipe by the superheated vapor tion began over 5 mir. after the bottom head failed allowed it to transfer heat to the liquid that drained in the MARCH calculation. While part of this dif-from the pressurizer when the PORVs were closed, ference can be attributed to a difference in decay again increasing the amount of liquid that was power, which led to an earlier steam generator

boiled, dryout, some is also attributable to the manner in which the codes treat the system behavior. The l

3.5.2 Comparisons with Draft NUREG.

SCDAP/RELAP5 and MELPROO event timing 1150. Surry was one of the reference plant',inves-were similar.

tigated in the "Reactor Risk Reference This lengthening of the transient could lead to i

Document," NUREG il50 (Draft). Natural circu-sery different conditions at the time of wssel fail-i lation in the RCS was not explicitly considered in ure. By heating the pipes in the coolant loops, fall-the thermal hydraulle analyses performed for ure of the RCS is likely to occur before the core NUREG-1150. Therefore, it is of interest to cum-melts through the reactor vessel. The resulting 2

ine differences in esent timing and core damage depressurization may be suffleient to reduce the l

caused by the natural eirculation Dows.

impact of direct containment heating on the i

i 28 i

4 2

i i

O cose 1 O Case 2 A Case 3 m

E v

,o l

~e>

.e n

v r

i j

&3

., b_ m,.

)

w o

O O i

t20 14 0 16 0 14 0 200 Time (min) 1 1

Figure 22. Pressurlier collapsed liquid level for the three scoping cases, i

l Table 5. Sequence of events for the Surry TMLB' transient from five different calculations

[

l j

Ti.me l

(min)

MARCil SCDAP/RELAP5 MELPROG l

I Draft

Once, llot Les and NUREG/CR Event NUREG 1150 Through In. Vessel in. Vessel

-4742 t

i Steam generators 69.0 77.2 77.2 77.2 69.5 dry l

Core heatup began 97.2 129.6 129.8 129.6 117.8 Fuel rod relocation i18.5 160.5 167.3 178.3 169.7 i

a began b

i43.5

> 200.0 S 200.0

> 180.0 248.0 i

Core slumped i

Bottom head failed 155.3

> 200.0

> 200.0c

> 180.0C 265.5

(

I

a. Fuel rod rekwation correstonocJ to the fuel rod melting tercretsture of 2$$o K 6e the draft NO RI G il50 cakulatia. to the claJdtag oude shell bnach temperature of 2KO K in the SCIMPs RL LAPS cakulatiorts, and to 22(O K in the Mt L PR'X) cakulation.

b. Core slumpics occurnd w hen the bottom node in a region rekwated or w hen the debris crun fatted, i

c. Iailure outade the wuct is likely to cwur before the bottom head iadure.

l l

29 1

containment response. The longer time before ses.

With the hot leg now s physically separated in the set failure may also allow more oxidation of the model, the piping temperature difference between materialin the reactor vessel. Additional oxidation the top and bottom halves is too large. Circumfer-in the sessel could affect the containment behasior ential conduction would transfer heat from the top through the core / concrete interaction. If less unox-of the pipe to the bottom. By increasing the ternper.

IEied material is imolved in the interaction, less ature of the pipe wall along the bottom of the pipe, hydrogen may be generated in containment after the vapor Dowing past may also increase in temper-vessel failure and fewer fission products may be ature, thus reducing the temperature difference released.

between the hot and cold now streams and reduc.

Another difference in the calculations is the ing the natural circulation now rate. This tempera-amount of hydrogen produced. At the time of fuel ture variation around the pipe may also introduce rod relocation, the h!ARCil calculation predicted stresses in the pipe that would accelerate its failure.

that 5% of the cladding would have been osidised.

The phpical separation of the opposing Dows also When fuel rod relocation began in the center chan*

nelin the SCDAP/RELAP5 calculations,14% of precludes accounting for any interactions that may the cladding was oxidized in the once through case.

occur between the Dows, such as heat and mass transfer.

7% was oxidited with !n-vessel natural circulation modeled, and 5% was oxidired with both hot leg The pipe temperatures are also affected by the and in. vessel natural cirealation modeled*

boundary conditions imposed on the outer surfem In these calculations, the surfaces exposed to the containment atmosphere were assumed to be adia-3.6 Uncertainties and Limitations bitic. In reality, there would be some heat loss through the pipes to the containment atmosphere, i

This section describes some of the uncertainties The heat loss would increase as the pipe tempera-associated with the analyses presented in previous ture increased because the insulation may begin to sections, and some of thelimitations of the models break down, allowing a higher heat transfer rate, and codes used in the analyses. Areat for future and because the temperature difference between the analysis are also discussed. Some of the uncertain-pipe surface and the containment is increasing.

ties are addressed by Ihe sensitisity studies While the temperatures of the surge line and hot leg described in the next chapter, would be lower than those calculated, it is expceted Decause the now in the coolant Imps is modeled that the relative temperatures would remain the in a one-dimensional manner, the countercurrent same. That is, the surge line would be hotter than hot leg now is not strictly mechanistically treated.

the hot leg, which would be hotter than the steam This introduces an uncertainty into the results, generator tubes. The lower pipe temperatures because the model was deseloped to redect the would also allow more heat to be removed in the behavior during a similar transient calculated by coolant loops by the hot les countercurrent now, the CON 15flX computer code. Differences in the estending the time to s arious stages of core damage nows and heat transfer rates predicted by the two esen further. Ileat loss from the hot legs and surge codes dictate that the results will not be identical; line is addressed in the sensitisity analy ses, one can only match the results of the CONIN!!N The amount of hydrogen generated is also an calculations as best as possible. The resulting area of uncertainty. The SCDAP/RELAPS code heatup li beliesed to be close to that which would does not calculate the oxidation of Zircaloy that is actually oe6ur, but the distribution of the energy in relocating. Nor does the code calculate the osida-the coolant loops may be somewhat different. The tion of structures outside the core, such as the steel results under these conditions are also dependent structures in the upper plenum that are certainly on the applicability of the modeling used to simu.

hot enough to osidire in the presence of steam, late low pressure, low temperature esperiment data The ballooning that occurred in Case 2 may not to the high pressure, high temperature conditions occur in the plant because of interactions between of the transient being imestigated.1he sensitisity the Zirealoy cladding and the Ineonel grid spacers, analyses imestigate effects of hot leg now rates on These materials form eutecties such that the clad-the plant tran lent response, ding may be breached at temperatures below

)

i 30 l

1500 K. Since the ballo ming in the calculation did amount of fission products released during the cal-not occur until some of the cladding had reached culations was generally small enough that there nearly 1800 K, it is likely that the cladding would would be little effeet on the structure temperatures.

have breached be' ore then. That would allow the The physical separation of the opposing nows in l

coolant and rap pressures to equalize, and thereby the hot legs may affect the fission product trans-preclude ballooning. This may also be true for port. If a significant portion of the deposition in Case 3, in which the cladding ternperatures were the hot leg is the result of gravitational settling, l

I abcve 1500 K when failure occurred.

then the fission products would drop into the now Another area of the code in which improsed stream returning toward the reactor vessel, where l

modeling is needed is in the cladding breach model, they may be retained on the cooler piping in contact l

Currently, the user inputs a temperature at which with the cooler now stream.

the zirconium dioxide shell on the claddlag The "bottom line" of the natural circulation breaches, initiating the now of molten material analyses is haw these Dows affect the condition of from inside the fuel rod. A more realistic model the core and RCS at the time of reactor vessellower would account for failure of the oxide shell by dis-head failure. To that end, the location, site, and l

solution of the oxide phase by the molten Zirealoy timing of any RCS failure before the vessel failure j

on the inside of the cladding. This would allow a needs to be determined. The surge line or hot leg more accurate prediction of the timing and physical appears to be the lik.-iy location, but the actual time l

l J

content of the relocation, at which failure wcurs and the site of the hole are The formation of a cohesive debris in the once-still unknow n. The slie is important in that a large through calculation had a significant impact on the enough break may permanently disrupt the natural i

transient progression, With sery little stesm How.

circulation nows. The site of the failure may also ing through the core, the heatup of the top of the be dynamic, in that continued heating of other core stopped, and oxidation tbroughout the core parts of the RCS may result in additional failures or j

nearly ceased. Flow tbrough the core bypass cooled the hot vapor nowing through the original failure the upper part of the core. While a cohesive debris location may ablate the pipe and increase the slie of

]

thai blocked the entire core would not be expected the hole. By diverting now from other g+.ris of the q

in an actual transient, it does demonstrate the RCS, those locations may cool, increasing the

[

l drawbacks of using a single channel to model the retention of fission products. Finally, the depres.

~

entire core.

suritation resulting from the failure needs to be cal-i The transport of fission products was not culated to determine the pressure of the RCS when included in these analyses. Consequently, the heat-the sessel fails, in order to determine the impor-j ing of the structures in the RCS by deposited fission tance of direct containment heating on the contain-t products was not accounted for. Iloweser, the ment response.

I i

?

l I

I l

I l

Il l

4, SENSITIVITY ANALYSES A series of sensitisity calculations was per-superheated steam, with the pressures varying formed to further investigate the effects of multidi-between the relief valve setpoints of 7.24 and mensional natural circulation now s on the response 6.89 MPa.

of the plant. In each of these calculations, both in-vessel and hot leg natural circulation Dows were 4,1 B aSe C aSO modeled. These sensitisity calculations addressed uncertainties in the modeling of the Surry reactor The first calculation used best-estimate values coolant system and within the SCDAP/RELAP5 for the steam generator inlet plenum mising, and code itself, including some uncertaintir.s identified the core and upper plenum crosinow resistances.

In the scoping analyses.

These values were the stme es those used in the Ses crat changes in the code and input model were scoping calculation with both in. vessel and hot leg made from the version used for the scoping analy-natural circulatix nows modeled (Case 3) An ses. The decey power was reduced by about 9%

end-of-life power profile was modeled, and the The asial power profile used was Datter. Interae-heat structures in contact wi h the containment t

t ons between the Inonel grid spacers and Zircaloy stmosphare were sdiabatic. This calcuation is cladding were simu!ated, so that the cladding fail-referred to as tht base case.

ute was based on citNr strain or temperatute. The Table 8 contains the sequence of events for the surge line model wa: also changed, so thet it con.

base case. The calentation began at 160 min, w hen nected to both the ton and bottom halses of the hot the core heatup was just begie: ting. As the heatup leg piping. These changes are discuned in mors continued, the cladding reached 1000 K, the tem.

detailin Appendices /.and D. Theimpact of these perature at which SCDAP/RELAP5 begins to ca:

changesis that the results of the sensitisity analyses culate oxidation of the Zirealoy cladding. This cannot be compared directly with the results of the occurred about 5 min before the core completely scoping calculations.

dried out. Eutectie formation between the Zirc.aJoy The nine sensitisit) calculations presented in the cladding and inconel grid spacers caused failure of following sections are shown in Table 6. The case the fuel rod cladding in pH three channals when the numbers will be used in most of the discussions for temperature eseceded 1410 K; no l>allooning was simplicity and clarity. Following these sensitisity calculated to occur. Creep rupture of the pressur-calculations, a final case will be pmsented in w hich lier surge line occurred at 246.3 min, shortly Case I was continu-d with failurr of the surge line before the onset of fuel rod relocation. The calcula-piping modeled, tion was terminated at 250 min, since the RCS fail.

As was the case with the sect >ing calculations, ure time and location had been established for each of the sensitisity calculations began as the comparison with the sensitisity calculations to core began to uncoser and heat up (at 16J min),

follow.

and continued until shortly after the RCS pressure The collapsed liquid loel in the reactor sessel is boundary failed. The conditions at this time wese presented in Figure 23. The inel decreased below taken from a calculation of the plant response from the bottom of thecore at about 186 min, and below transient initiation into the core heatup using a the bottom of the core barrel at about 228 min. At model without inoessel or hot leg natural circul+

the end of the calculation, the inct was within the l

tion Dows. Table 7 presents the sequence of cents lower head.1he only other liquid in the RCS was in I

for this early part of the transient. For compa,ison, the three loop seats, w hich did not clear during the the nent timings from the scoping calculation are transient, also presented. The difference in timing was caused Figure 24 presents the peak cladding tempera-i by the use of a lower decay power in the sensitisity ture in the core. The initial heatup rate of about I

calculationi. The two calculations were consistent 0.27 K/s deercased to 0.20 Kes at about 187 min.

l ir. that the core uneoscring began in both cases At that time, the core bypass cleared of liquid, l

when the same amount of energ) had been gener-allowing steam La now from the upper plenum to ated in the core. T he RCS pressure throughout the the lower plennm. This increased the core inlet transients cy cled between the PORY open and close Dow, reducing ihe core heatup rate. The peak tem-setpoints of 16.2 and 15.7 MPa, respectisely. The perature inercased steadily until the osidation secondary sides of the steam generators contained Linetia changed at about 245 min, when the peak 1

32 1

u

Table 6. Matrix of sensitivity calculations Sensitivity Calculation Case Number Parameter 1

2 3

4 5

6 7

8 9

Asial power profile l

End of life X

X X

X X

X X

X Beginning-of life X

Inlet plenum mising flest-estimate X

X X

X X

X X

Reduced X

None X

Piping heat loss None X

X X

X X

X X

Convection X

X Radiation X

UJfer plenum crossflow resistance Best ettimate X

X X

X X

X X

{

Deerease-1 X

Increased X

Core crosiflow resistance 11est< stimate X

X X

X X

X X

Deercased X

Increased X

33

4 Table 7. Sequence of events up to coes heetup for the senettivity and scoping calculat6ons Time (min)

Event Sentitivity Scoping _

i Transient initiation 0

0 PORY cycling begins 82.9 71.8 1

Steam generators dry d5.4 90.7 75.4 77.2 j

Hot legs reach saturation 120.9 100.6

<J s

loop natural circulation flow ends 129.9 109.7 i

i Core heatup begins 159.8 129.6 1

j Table 8. Sequence of events for the base case i

1 j

Time f

Event (min)

Calculation begins 160.0 Center channel osidation begins 185.3 l

hiiddle channel oxidation begins 186.1

[

i

(

1 Two-phase liquid level below core 190.2 j

Outer channel osidation begins 192.6 I

I Center channel fuel rod cladding falls 223.4 i

I Pressurlier empties ofliquid 224.8 I

l l

Middle channel fuel rod cladding falls 225.3 l

Outer channel fuel rod cladding fails 241.3

[

Pressurlier surge line fails 246.3 Center channel fuel rod relocation begins 248.0 l

hiiddle channel fuel rod retxation begins 248.8 1

l Outer channel fuel rod relocation begins i

Cakulation ends 2!0.0 r

t

}

l 34 i

e 4

f

[

4 i

i i

i 4

n E

v bottom of eere

$a wiio.,e,ee,o w, A 2

0 16 0 160 200 220 240 240 Time (min)

Fleure 23. Reactor vessel collapsed liquid level for sensitivity Case 1, 4000 7 3000 v

e i

6

?

O 2000

)

H eseeleteled esl4 ellen.

1000 0

16 0 100 200 220 240 See Time (min)

Figure 24. INak claddins temperature for sensitivity Case 1, 35

cladding temperature increased rapidly. The short ated to keep the temperature near the melting point hmperature escursions just prior to this time of zirconium dioxide (2973 K) or uranium dioxide redected the calculated behavior of the control rod (3013 K).

guide tubes. The later temperature excursions Because the calculation was terminated shortly showed the behavior of individual nodes in the after temperatures exceeded 2000 K, the core damage core. The temperature increased as the cladding was primarily in the form of Zirealoy oxidation, with rapidly oxidlied. The temperature then decreased some material relocation. Figure 27 shows the total when the cladding in a node was completely oti.

hydrogen generation rate during the tramient. The d!ied, because the decay heat alone was unable to osidation increased steadily until parts of the clad-sustain the higher temperature, ding became completely oxidited and the rate l

Fuel rod cladding surface temperatures from the decreased. At its peak, the energy relened by the osi-top node in each of the three core channels are pre.

dation reaction was about 6 times the decay power. At sented in Figure 25. The center and middle channel the end of the calculation (250 min),222 kg of hydro-temperatures were sery close together throughout sen had been generated, corresponding to osidation the transient. The outer channel temperature was of about 32re of the Zirealoy in the core. Material lower because it had a lower power and because relocation was essentially condned to the control steam reentering the outer channel from the upper rods, as onh about 0.03 kg of fuel melted in the core plenum helped to cool Ihe fuel rods, liowner, after and there was no relocation of Zircaloy from the fuel accelerated oxidation began in the inner two chan-rods. The upper 6Dre of the control rods in the center nels, the hot vapor from the upper plenum helped and middle channels relocated, with all of the Zirca-to aceclerate the temperature increase in the outer loy and stainless steel and 25re of the relocated Ag-channel. That is w hy there w as a rapid heatup in the In-Cd control material refreezing in the core. The i

)

outer channel prior to the change in otidation remaining 75re of the relocated absorber dropped kinetics, beginning at a temperature of about into the lower plenum. In the outer channel, the top i

)

1500 K. The heatup rate increased further as the 50re of the control rods relocated, with 70re of the temperature increased and accelerated oxidation relocated absorber being the only material to drop

occurred, below the core. The calculation w as terminated before l

Figure 26 show s the center channel fuel rod clad-the outer channel fuel rod temperatures were high ding surface temperatures at socral einations, enough to initiate relocation. Fiuion product release Changes in the core inlet conditiom generally had a during the tramient us also relatisely minor, with greater impact on the lower half of the fuel rod. As 4re of the noble gases,3ro of the cesium and tellu-I discussed prniously, the core b> pau cleared oflig-rium, and 2re of the iodine originally in the core J

uid at about 187 min, and the increased now into being released by the end of the calculation.

the ccre cooled the cladding in the lower part of the Figure 28 shows the now entering the upper core while slowing the heatup rate in the top part, plenum frorn each of the three core channels and i

An increase in the center channelinlet now shortly from the core bypsu, and the upper plenum recir-i after 200 min resulted in more cooling of the bot,

culating now. The reeirculating flow moses down tom part of the core. At about 218 min, the down-the outer channel of the upper plenum and turns comer liquid Inet dropped below the core barrel, inward just abose the core esit rather than entering i

allowing cooler sapor to now into the core, again the core. The now in the upper plenum was about L

cooling the lower regions of the core. With the core four times greater than the now returning to the barrel cleared, a now path w as established from the core through most of the tramient 't he increase in upper head to the downcomer into the core. Relo-the now from the upper plenum to the core bypass cation of controt rod absorber material to the lower at about 187 min occurred when the liquid loci plenum began at about 240 min. This material cleared the core bypan, allowing sapor to now l

boiled liquid in the lower plenum, w hich resulted in freely from the upper plenum to the lower plenum.

j some cooling in the lower cloations of the core.

A slight decrease in the now reentering the outer i

1 After the accelerated osidation, the temperature of channel from the upper plenum can also be the cladding in the top node peaked and then obsersed at that time. The relation betweeti the deetcased. The decrease occurred because the clad-upper plenum reeirculating now and the upper ding was completely osidised, so that the only heat plenum sapor temperature just abose the center generation in that node was the decay heat in the core channel is shown in figure 29. The now i

fuel. Without the esothermie Zircaloy osidation decreased with increasing temperature because the j

reaction, there was not enough heat being gener-density difference drising the flow was decreasing.

i j

36 l

4000 i

O C.nl.r chonn.1 O Widdl. chonn.1 6 Out.r chann.1 7 3000 l

v s

S-f 3

O 2000 g

e...l. ret.d.uidell.e H

goo 1

0 40 40 200 220 240 260 Time (min)

Figure 26. Fuel rod cladding surface temperatures at the top of the three core channels for sensitivity Case 1.

4000 s.i.h.

..,. 6. ii m O 0.18 m O t.2a m 7 3000 A 2.38 m v

X 3.4 7 m x

s.

3 0 2000 6.

";i.#f"'

l

~#

c a-N O -e 0

0 H

I i

..r. 6.re.i ir. i r.d

.iyr.

c.ipen 40 20 200 220 240 260 Time (min)

Figure 26. Center channel fuel rod cladding surface temperatures at 0.18.1.28,3.38, and 3.47 m above the core bottom for sensithity Case 1.

37

I s

i

.e O'

O.8 5=

C 0.6 m<

en

.w 0.4 u

V 0.2 AI 0

WO WO 200 220 240 260 I

Time (min)

Figure 27. Totai hydrogen generation rate for sensithity Case 1, T

l f

no i

i f

O Cent er chamet n

O Widdle charmet m

A Outer chamel

\\

X Core bypois I

j j 10 0 0 Uppet pl enum v

t 1

N a

c D

o f

C M a a Q.. e,,4 (p j ^^ w-O-Q,n i s 'gsg m

m g

0, Q,,s-;;"y ; Ed p l

3 l

)

.g j

l WO 40 200 220 240 260 l

Time (min)

Figues 28. Mass flow rates esiting the three core channels, the core bypass, and recirculating in the upper f

]

plenum for sensithity Case 1, t

i.

l 38 t

I i

tso m

n N

y too L

Ch.!!,,,

v A',QN(w..

2 r

=

g t;-

t l

o f

1 i

.g 1

soo 1000 isoo 2000 asos i

i Vopor temperature (K) 1 I

J F;gure 29. Upper plenum recirculating mass now rate as a function of masimum upper pienum vapor l

temperature for sensitivity Case 1.

t i

i l'

s Upper plenum struuure temperatures in each reactor vessel head bolts was 621 K at the end of the l

]

channel just above the core outlet are show n in Fig-calculation. The high structure temperatures in the l

4 ute 30. The center channel temperature was highest upper plenum and bafne plates indicate 1, hat more

{

becau'e it was abose the hottest part of the core, hydrogen would have bern producsd had the oxida.

l The outer channel temperature was,he lowest tion of these steel structuies been accounted for in i

because the sapor in that channel had been cooled the calculatiore. There may also have been melting l

us it nowed dow n; the upper plenum outer channel and relocation of sonne of these structures.

i temperatures increased with increasing elesation.

Figure 31 presents d,4, rraction of the core heat I

The middle channel temperature was bc; ween the that had been removed by the cou:c.:., sny time.

}

centet and outer channel temperatures, and might The integral core best removal remained near 73%

I hase been closer to the center channel structure for most of the calculation. Of the energy temowd temperature escept that the recirculating now from the core at the time of the surge line failure within the upper plenum caused cooler sapor from (246.3 min). 9.Or had been transferred to Hruc.

e Ihe outer channelin mis w ith that above the 'niddle turts in each of the non-prenurlier loops and channel, cooling the structures thete. The hottest li.tr to toop C structures. The hot les piping la e

structures within the upper elenum wcre the control all three loops had absorbed 4.8e% of the energy

)j rod housings. These very thin structures haJ a tem-remosed from the core, the surge line piping 0.8%,

perators of 180) K at the end of the calculation.

and the steam generator tubes end tube sheets This temperature is high enough to meh these stain-23.'.%. Sotae of the energy transferred to the steam l

leu steel housings, e4though that hvhastor was not generator tubes was in turn transferred to the steam modeled, Other solume-average structure tempera-ca the second.uy side of the steam gewrators, so tures of note at the end of the calculation were a that the net euergy storage in these structures was masimum of 11% K for the core barrel,1477 K for less than 23.1%.

j the core baffle plates, and 733 K for the uppce htore energy was deposited in the pressuriser 1

head, w hich w a: the highest reactor sessel wall tem-loop (Loop C) than in the other loops because the j

perature. The local temperature in the region of the PORVs drew now into that loop when they were 1

39

[

l

ll i

1 2000 i

i i

j' O Center chonnel nw O Widdle chcanel V

A outer channel 9

6 3

j 1500 U

6 i

6 e

l 1

m ISO 200 330 240 See i

Time (min) i J

l i

Figure 30. Volume average temperatures of the upper plenum structures at the outlet of the three core i

channels for sensitivity Case 1.

}

1 L

i I

r I

1 t

1.5 i

i i

I k

t l

i i

t 1

l 5

{

g

-w 1

w t

L.

j l

0.s 1

l i

?)

o t

2 16 0 too 206 220 240 240

(

j Time (min) j l

I Fleute 31. I'raction of the core heat remosed by the coolant for sensithity Case 1.

p i

?

'l I

J d

i 40 i

i f

j

j open. The increased Oow in the hot leg, combined top of the pipe to the bottom, flowner, this cannot be g

with the decreased now into the other tw o hot legs, modeled with the one-clirt 1sional heat structures Ir, l

resulted in about Me more energy being absorbed SCIMP/REl.AP5.

bythe structuresin Loop Cthaninloop Aor B. A The temperature of the upor entering and edting result of this difference is illustrated in Figure 32, the reactor wssel from the loop A and C hot legs is w hich presents the volume-average temperatures of presented in l'igure 34. The upor temperatures in the he three hot less near the reactor sessel. The top halws of the hot kgs were nearly identical through-loop A and il bot leg temperatures are identical, out t!w tramient. The temperatures in the bottom part and slightly cooler than the leop C hot leg. 51milar of tne hot kgs were fairly close together, escept when I

temperature differences were observed in the rest of the PORVs were open. % hen the PORVs were open, l

the hot legs and :n the steam generators.

now rewrsed in the bottom of the mop C hot leg, figure 33 shows the highot temperatures in the drawing vapor from the reactor wssel tcward the surge l

loop C hot kg, surge bne, and steam generator tubes, line. This causal the large increases in temperature.

Aho included in the figure are the corrnponding struc-When the PORVs closed, the flow wm qukkly rmtab-ture temperatures on the bottom of the hot leg and lished toward the reactor msel, and the temperature surge hne. All of the structures began to heat up abow decremed. The open PORVs caused slight decreases in the saturation temperature inunediately. The surge line the temperature in loop A because the now rate heated up the fastest became it had nearly the same through the bottom of the hot kg increased, resulting upor temperatures as the hot kg, but it is ortly one-in higher heat transfcr rates to the pipins and corte-third as thick. The steam generator tutes are the thin-spondingly lower Hund temperatures. The hot leg bot-nest structures, but they had the coolest upor (of the tom temperatures did diverge slightly as the tramient structures shoan). Aho, the steam generator tubes are p.ogreswd. This was the result of higher pipe tempera-somewhat protected by the tube sheet, which har a tures in loop C, which were dewloped during the large thermal capacity and helps to cool the upor PORY cycles when hot vapor from the upper plenum before it ernounters the tubes. Creep ruptua of the wm drawn into loop C.

st.rge line was predived to occur at 246.3 min, at a The mass flow into the top of two of the hot leg, temperature of 1219 K. At that time, the hot leg was is prtsented in Figure 35. The now rate increased as near 1(00 K, and th I steam generator tubes were near the transient began, reaching a masimum at about 800 K. The surge li te temperature was affected most UO min. Displacement of the saturated vapor in by the eyeling of thi PORVs. When the PORVs were the loops and heat traasfer to the structures com-open, hot upor wm drawn from the reactor wuel to binet to keep the return now at the saturation tem-the preuuricer, hea ing the surge line. When the perature during this time. The now then gradually PORY: were closed, wry little now paswd through the decreased as the return now temperature increased surre line. Until the presunier emptied at :24.8 rain, abose saturation, because the density difference liquid draining into the surge line when the PORVs between the opposing hot les flow s was decreasing were closed coold the pire. After the pressuruer emp-esen hough the temperature difference was tied, the temperature inerceed skuly w hen the valm increasing. When the PORVs were closed, the now were ckwd.11y contrast, the hot les temperature did in Loop C was higher than that in loop A.. The not folkm the PORY cgling. The greater thkkness, loop !! Oow was the same as that in loop A.

and heru.e longer tune cwtstant, of the hot leg pipng.

About 5% less of the hot leg Dow mised in loop C

ether with the fact that there wu s!*ms fkw from than in loops A and II. This resulted in a higher s r reactor wuci into the hot leg. resuhed in a steady now rate in loop C, as will be discuued in the h 4 tup of the hot Leg. Sewral short dxtra5o in the steam generator inlet plenum mising sensitivity see-upper hot leg temperature betwven 170 and 2t0 min tion. When the PORVs were open, the now in the were the roult of bquid draining from the preuurizer; preuurizer loop increased because Gow was bein; the sat urated bquid coolni the pipe. The large tempera-drawn to the surge line, while the loop A Gow ture differerwrs between the top and bottom halves of deereased for the same reason. When the PORVs the surge hne (up to 4M K) and hot leg (up to 250 K) closed, the two Cows quickly returned to their woukt indwe circumferential streues in the pipn.

quasi steady salues. Unlike the Westinghouse There woukt alo be conduction heat tramfer from the e perimerts5 and CONtN1IX cateulations* 3, the t

41

l 1200 i

i i

O Loop A ng a Loop B v

O Loop C e

6 3

noo o-O L

p600 1

6e 4

i 400' too 14 0 200 220 240 260 Time (min) j Figure 32. Volume average temperatures of the three hot leg pipes near the reactor seasel for sensitivity 1

Case 1.

l 1

i l

}

1500 t

O Hetles top l

n x

O Het leg bottom V

6 Steam generef or twbes

(

X surge une top i

C Su*ge li ne bo t t om eroep i

L 1250 3

rupture

[

,1 feuvre a

\\

u l

o f moo preewl er metr J

ea O

l 750 l

6 L

e j

t j

i i

goo t

14 0 160 200

+.0 240 250 Time (min) 1 I

f

)

Figure 33. Volume-asciage temperatitres of the hottest 1oop C hot leg, surge line, and steam generator j

i tubes for ser sitisity Case 1.

42 l

I

2000 i

O Loop A Iop n

O Loop A bottom M

A Loop C top L

l X Loop C bottom 3

e I

L

/

3 1500 s[J h

c 1000 o

j jr UPNd 500 16 0 180 200 220 240 260 Time (min)

Figure 34. Hot leg noz21e hot and cold vapor temperatures in loops A and C for sensitivity Case 1.

40 O Loop A O Loop c nHN$

20 e

0 O

O O

l O

g D

,,swwy b

70

'C' o

0 14 1

-20 16 0 18 0 200 220 240 260 Time (min)

Figure 35. Upper hot leg mass flow in loops A and C for sensitivity Case 1.

4)

flow in the non pressurizer loop did not reverse investigated with sensitivity Case 2. In this calcula-completely when the PORVs were open; a small tion, a chopped cosine axial power profile with a positive flow toward the steam generator persisted.

peak to average power ratio of 1.200 was used in all While this behavior does not reproduce the desired three core channels. In the base case (Case 1), a behavior,it was also present in the calculations that flatter profile with a peak to-average oower ratio of were performed to test the hot leg countercurrent 1.155 was used. All other parameters were the same flow nodalization. Hence the mixing in the steam as in the base case.

{

generator inlet plenum was adjusted to provide the Thble 9 presents the sequence of events for the correct total heat transfer with this flow present.

base case and the axial power profile sensitivity The flow into the Imop A hot leg is correlated with case. There was very little difference between the the vapor temperature in the upper plenum at the two calculations. The onset of various stages of hot leg nozzle in Figure 36. The lower How values core damage occurred slightly earlier in Case 2, as at each temperature, which occurred less frequently did the surge line failure. However, a difference in i

than the higher flows, reflect times when the failure time of just over 1 min for a nearly 250 min PORVs were open.

transient is insignificant.

He now through the steam generator tubes was I;igure 38 shows peak cladding temperatures about 70% 'arger than the flow in the hot legs in from Cases I and 2. As with most comparisons loops A and B, and about 120% higher in Loop C.

between the two cases, there was little difference in About 88te of the hot vapor flow in the hot leg mixed the temperatures. The temperature in Case 2 in the steam generator inlet plenum, with the other increased slightly faster. The short increases before 12% proceedmg directly into the steam generator tubes the temperature reached 1850 K were again the in loops A and B; about 84% of the hot leg flow result of oxidation excursions in the control rod mixed in loop C. He same fraction of the flow guide tubes.

returning from the cold flow steam genemtor tubes There were small differences in the flows in the mixed in the inlet plenum, with the rest proceeding reactor vessel between Cases 2 and I. In Case 1 directly into the bottom of the hot leg. In the the return flow from the upper plenum to the outer Westinghouse naarl circulation experiments, it was core channel was slightly higher than in Case 2.

estimated that abo. 30% of the hot leg flow partici-The flow also penetrated further down the outer pated in the inlet pSr.am mimg.5 channelin the base case. Both of these changes in Figure 37 show, the pressurizer liquid volume.

flow resulted from the faster heatup of the down-The liquid volume decreased w hen the PORVs were ward flow in Case 2. This downward flow was open for two reasons. First, the pressure was heated faster because of the steeper power gradient, decreasing, causing some of the liquid to flash to which transferred more energy to the fluid in the steam. Second, the vapor entering from the surge top half of the cc.e, reducing the density gradient line was superheated, and transferred heat to the between the outer and inner channels. Similar liquid, causing it to boil. Virtually no liquid was decreases in the outer channel flow and flow pene-entrained with the flow through the PORVs. The tration were observed in the base case when the amount of liquid in the pressurizer increased cladding in the top part of the outer channel began slightly when the PORVs were closed, because the to oxidize heavily. This oxidation had the effect of increasing pressure resulted in the condensation of changing the avial power profile to be more like some of the steam, which compensated for the loss that of Case 2.

l of a small amount of liquid that drained back The energy removal from the core and redistribu-I through the surge line to the hot leg. A brief calcu-tion throughout the system were nearly identical in lation was also performed in which the Wallis the two cases. The hot leg mass flow rate at a given flooding correlationl5 was applied at the junction hot les inlet temperature was identical to the base between the surge line and the pressurizer. Os er the case. Figure 39 presents the highest hot leg, surge 3

40 min that the calculation covered, there was little line, and steam generator tube temperatures from change in the pressurizer liquid volume.

the pressurizer loop, as well as the surge line tem-perature from the base case. The temperatures increased throughout the transient, with the surge

/. 2 Axial Power Profile line failing at 245.0 min at a temperature of r

Sensitivity 1238 K. There was little difference in the surge line temperature until close to tM time of failure, when The effect of a different asial power profile was the vapor temperatures leavir. the core were higher i

I 44 t'

-,_,,..c-

___,____,-__,,,,._-..--.__m_

m

e 7N cn5 5

i NW'%.%.s%.

1

-T 1

_l g,,.;;e,.;v:M W :rn'. -

+n

/

M to O3 1

-s l

300 1000 1500 2000 I

Vapor temperature (K) l Figure 36. Ilot leg flow as a function of the hot leg inlet vapor temperature in Imop A for sensitivity Case 1.

30 p

W

\\

E v

, 20 E

s o>

t 5

D 0-

.J 0

14 0 18 0 200 220 240 260 Time (min)

Figure 37. Pressurlier liquid volume for sensitivity Case 1.

45

- -_1

(

d Table 9. Sequence of events for the base case and axial power profile sensitivity i

calculations Time (min)

Event Case 1 Case 2 Fuel rod cladding oxidation begins Center channel 185.3 185.3 Middle channel 186.1 185.7 Outer channel 192.6 191.7 Liquid lesel drops below core 190.2 190.2 Pressurizer empties of liquid 224.8 224.8 Fuel rod cladding fails Center channel 223.4 222.0 Middle channel 225.3 223.5 1

Outer channel 2J1.3 239.3 Fuel rod relocation begins Center channel 248.0 245.2 Middle channel 248.8 246.2 L

l Outer channel RCS pressure boundary fails 246.3 245.0 j

i RCS failure location Surge line Surge line Calculation terminated 250.0 248.2

[

l l

I i

E i

I J

l i

s 1

1 46 r

4000 O case 2 O cose 1 2 3000 v

e i

Ls

+

c 2000 d

nf' %d i

to.

E e

M0 C

0 16 0 18 0 200 220 240 260 Time (min) i Figure 38. Peak cladding temperatures for sensitivity Cases 2 and 1.

l 1500 O Hot leg Case 2 nM O 'leam generator tubes, cose 2 V

6 Surge line, Case 2 creep rupture X surge nne, case i f ailure oL 1250 case 2 =2:

3 Case 1 i

0 L

ec.

E 1000 e

+

e tn O 750 e>4 i

i 500 16 0 18 0 200 220 240 260 Time (min)

Figure 39. liighest volume average pipe temperatures in the loop C hot leg, surge line, and steam generator tubes for sensitivity Case 2, and the surge line for Case 1.

47

in Case 2 than in Case I because of the higher clad-steam generator tubes, and therefore, no heatup of dinh :em xratures, the tubes. The hot leg countercurrent now would be Some of the plant conditions shortly before the driven only by heat transfer to the hot leg piping surge line failure for Cases I and 2 are contained in itself, and the behavior would be similar to that Table 10. As in the sequence of events, there was with no hot leg natural circulation modeled.

very little difference between the two calculations at about 242 min. The peak cladding temperatures in 4.3.1 Reduced inlet Plenum Mixing. As in the each channel were slightly higher in Case 2, as was base case, the calculation with reduced inlet the hydrogen generation. The higher middle chan-plenum mixing proceeded until shortly after the nel cladding temperatures in Case 2 resulted in RCS pressure boundary failed. Table 11 presents more control rod relocation than in Case I at that the sequence of events for the base case (Case 1) time. The fission product release was slightly and the two inlet plenum mixing sensitivity cases.

higher in Case I because the fuel rod temperatures By increasing the now in the coolant loops, the in the bottom half of the core were higher in Case I transient was extended. Ballooning in the center i

than in Case 2.

and middle core channels, which resulted from the Changing the axial power profile had little effect slower heatup, was the major difference between on the plant transient response. A slight decreasein this and the base case.

the outer channel core now resulted in slightly With reduced mixing in the steam generator inlet higher cladding temperatures, which led to surge plena, the hot leg now rates increased. Figure 40 4

line failure 1.3 min earlier than in the base case, shows the now entering the top of the loop A hot The hot leg now characteristics and energy removal leg from the reactor vessel for Cases 1 and 3. The were nearly identical to the base case.

trends in the Dows were the same, but the magni-tudes were higher in Case 3. The hot leg now as a functi n Iinlet temperature is shown in Figure 41 4.3 Inlet Plenum Mixing for Case 3. Compared to the base case, the now at Sensitivity any given temperature was awut 25% higher. With cooler vapor returning to the upper plenum, the The natural circulation flow in the hot leg and recirculating now in the upper plenum was also steam generators is the most uncertain of the flows higher than in the base case, by about 10%.

modeled because of the method used to develop the The higher loop Dow rates also changed the model. A model for full scale commercial plant energy distribution compared to the base case.

severe accident conditions was developed from First, more energy was removed from the core as computer code calculations and low pressure, low the hot leg flow increased. Figure 42 shows the temperature experiments in a scaled facility using a fraction of the core heat removed during Cases I steam simulant. Accordingly, the hot leg model was and 3. Second, more of the energy removed from changed significantly in these sensitivity the core was transferred to the loops in general, and calculations.

the steam generator tubes in particular. Near the As discussed in Chapter 2, the amount of mixing time of the RCS failure in the base case, heat trans-in the steam generator inlet plenum determines the fer to the hot legs accounted for 4.8% of the energy density difference between the opposing hot leg removed from the core, and heat transfer to the Dows, and hence the hot leg now rate. The amount steam generator tubes and tube sheets accounted of energy deposited in the loops is therefore directly for 23.1% of the energy removed from the core.

Influenced by the inlet plenum mixing 'iko calcu.

These values were 3.6% for the hot legs and 30.1%

!ations were performed with decreased mixing, and for the tubes and tube sheets in Case 3 near the hence higher hot leg dows. In sensitivity Case 3, time of the RCS failure, the mixing was decreased arbitrarily. In sensitivity The effect of the higher energy deposition on the Case 4, there was no mixing at all. Case 4 prosided loop structure temperatures is shown in Figures 43 an upper bound for the steam generator tube tem-and 44. Figure 43 presents the hot leg temperatures peratures and for the heat transfer to the loops with nearest the reactor vessel for each of the thrce hot leg countercurrent now because the steam gen-loops. The Loop C hot leg temperature was slightly erator tubes were exposed to vapor at the same tem-higher than that of the other two loops because of perature as that entering the steam generators from the PORV cycling, which periodically drew more the hot leg. In the limiting case of complete mixing now into that hot leg. The temperatures increased in the inlet plenum, there would be no now in the steadily until shortly after accelerated oxidation 48 a

i

Table 10. Conditions near the tirne of the surge line failure for the base case and axial l

power profile sensitivity calculations 1

(

Value I

Parameter Case I _

Case 2 Time (min) 241.7 241.7 Center channel peak clad temperature (K) 1752 1815 hiiddle channel peak clad temperature (K) 1714 1756 Outer channel peak clad temperature (K) 1513 1543 hiaximum upper plenum structure temp. (K) 1355 1381

^

htaximum hot leg temperature (K) 973 985 j

hiaximum steam generator tube temp. (K) 783 791 htaximum surge line temperature (K) 1864 1191 Core outlet flow (kg/s) 12.9 12.2 Core return flow (kg/s) 8.6 8.9 Upper plenum recirculating flow (kg/s) 35.7 36.8 Reactor vessel collapsed liquid level (m) 1.48 1.43 Core heat removal (%)a 76.1 76.2 Core energy removed and deposited in:a loop A structures (%)

8.9 8.9 Loop B structures (%)

8.9 8.9 Loop C structures (%)

11.6 11.6 Ilot leg piping (%)

4.6 4.7 Steam generator tubes, tube sheets (%)

22.6 22.8 Surge line piping (%)

0.8 0.9 Center channel oxidation (%)

14 15 htiddle channel oxidation (%)

12 14 Outer channel oxidation (%)

7 8

flydrogen generated (kg) 80 89 Fuel relocation (%)

0.0 0.0 Fuel rod cladding relocation (%)

0.0 0.0 Control rod relocation (%)

3 45 Xenon / krypton release (%)

0.8 0.7 Cesium release (%)

0.5 0.4 l

lodine release (%)

0.0 0.0 Tellurium release (%)

1.1 1.0 Surge line failure time (min) 246.3 245.0

s. Intesral quantities tiom the Mart of the calculation.

i 49

l 1

l i

Table 11. Sequence of events for the base case and steam generator inlet plenum mixing sensitivity calculations Time (min)

Event Case 1 Case 3 Case 4 Fuel rod cladding oxidation begins Center channel 185.3 187.8 193.5 hliddle channel 186.1 188.3 194.7 Outer channel 192.6 197.4 206.4 Liquid level drops below core 190.2 191.3 191.1 Pressurizer empties of liquid 224.8 231.6 224.0 Fuel rod cladding balloons f

Center channel 233.3 +

250.0 +

hliddle channel 233.3 +

Outer channel 4

Fuel rod cladding fails Center channel 223.4 242.3 252.2 hiiddle channel 225.3 244.2 257.3 Outer channel 241.3 255.3 268.3 Fuel rod relocation begins Center channel 248.0 hliddle channel 248.8 275.3 Outer channel 254.3 279.8 ll RCS pressure boundary fails 246.3 254.8 290.5 RCS failure location Surge line Surge line 11otleg i

Calculation terminated 250.0 266.7 300.0

'l r

J 50 u

10 i

O cose 3 O Case 1 m

M Nm 3v Dt/ P 5

?

UE Nkh a..

h

~

E O

o o

O C

0 M

M O2

-5 16 0 200 240 280 Time (min)

Figure 40. Upper hot leg mass flow in loop A for sensitivity Cases 3 and 1.

10 7Nm a

v

@h s

1

'O.O.Q e

5

,f jpi!pi.'% i.i,&... ' ""$yy...,:.g

..l 2

'~..

l..-

~

0 N

t O2

-s s00 1000 1500 2000 Vapor temperature (K)

Figure 41. Ilot leg flow as a function of hot leg ir.let vapor temperature for sensitivity Case 3.

51

15 i

i O Case 3 O Case I l

C O

".g C

6 la.

0.5 g

Oc 14 0 200 240 280 Time (min)

Figure 42. Fraction of the core heat removed by the coolant for sensitivity Cases 3 and I.

1200 O Loop A mM A Loop B v

O Loop C G

L

3 1000 6

0.

E

.e

$ 800 gfgjgtedcladding.

O Le><

600 16 0 200 240 280 Time (min)

Figure 43. Volume average hot leg pipe temperatures near the reactor vessel for sensithity Case 3.

i

$2

1500 i

i O Hot leg nM O surge Itne V

A Stoom generolor tubes rupture 3

foNure e.

O 6.e Q.

EN e

1 F

i t

oueieroi.d olodding 4

G ex t dellen 500 14 0 200 240 280 Time (min)

Figure 44. liighest volume average pipe temperatures in the loop C hot leg, surge line, and steam generator tubes for sensitivity Case 3.

began in the core, at about 250 min. The enhanced oxidize. The increased oxidation accelerated the oxidation in the core increased the vapor tempera-heatup. The outer channel heatup rate did not ture in the upper plenum and the heatup rate of the increase until a few minutes later, when accelerated hot leg piping. There was little difference in the oxidation began in the center channel. The acceler.

steam generator tube temperatures between the ated oxidation led to higher vapor temperatures in loops in Case 3. The highest volome average tem-the upper plenum. This vapor then entered the peratures in the hot leg, surge line, and steam gen-outer channel, causing the cladding heatup rate to erator tubes for Case 3 are presented in Figure 44, increase, liowever, the faster inner channel heatup As with the hot leg, the surge li.~e temperature associated with ballooning was not reDected in the increased much more rapidly after the core heatup outer channel, indicating that the additional oxida-l accelerated. Creep rupture failure of the surge line tion energy was retained in the cladding rather than was calculated to occur at 254.8 min, at a tempera-transferred to the coolant, ture of 1256 K.

Peak cladding temperatures from Cases I and 3 l

Figure 45 shows the fuel rod cladding surface are shown in Figure 46. The curves exhibited the temperatures from the top nodein each of the three same behavior, but with the time extended. The l

core channels for Case 3. The delay in the onset of higher hot leg flow rates led to increased heat accelerated oxidation between the center and mid.

remor a in the loops, which in turn led to increased I

die channel temperatures increased compared to heat removal from the core and a slower heatup, the base case. This was because cooler vapor was The slower core heatup allowed the cladding to bal.

being returned from the upper plenum to the core, loon in both the center and rniddle channels at I

where it prosided cooler vapor first to the outer about 240 min. The ballooning in both channels channel, then to the middle channel as it flowed extended over the top 60% of the cladding and inward. The heatup rate increased in the center and reduced the axial now area by just oser 60%.

middle channels at about 240 min because the clad-The ballooning caused an overall decrease in the ding ballooned. When the ballooned cladding rup.

core nows in Case 3. Figure 47 shows the mass tured, the inner surface of the cladding began to now rata entering the upper plenum from each of 1

53

(

3000 i

O Center chonnel O Widdle chonnel

}

A Outer channel I

2000 6

occelerated oxidation i

J y

j o.l M0 balooning 0

16 0 200 40 280 Figure 45. Fuel rod cladding surface temperatures at the top of the three core channels for sensitivity Case 3, 4000 i

i O case 3 O case 1 2 3000 5

I 6

0 2000 L

O Q.

E A

C-1000 boiconing C

0 16 0 200 240 280 Time (min)

Figure 46. Peak cladding te.uperatures for sensitivity Cases 3 and 1.

(

54

40 i

i O Center channel O ulddle channel 7

A Outer channel N

X core bypass llhL i 1 j m"

p

- w u-m.

n dk W vi g p e g g d]

o C

h-20 2

bolooning

-40 16 0 200 240 280 Time (min)

Figure 47. Mass flow rates exiting the three core channels and the core bypass for sensitivity Case 3.

the core channels and the core bypass. When bal-set temperature,752 K, while the temperature in looning occurred at about 240 min, the flows leav-the region of the head bolts was 630 K.

l ing the center and middle channels and entering the The slower heatups led to more extensive oxida-outer channel decreased. The flow penetration in tion of the cladding at the time of the RCS pressure the outer channel was also reduced, although the boundary failure. In Case 3,382 kg of hydr gen basic now pattern in the core was maintained. The had been generated shortly after the surge line fail-flow pattern did change briefly at about 250 min.

ure, compared to 109 kg at the time of surge line When the cladding in the middle channel began to failure in the base case. Figure 49 shows the heat oxidize rapidly, causing a rapid heatup in that chan-generated by the oxidation reaction together with nel, the flow reversed in the center channel. Flow the decay power. The peak oxidation power was was drawn toward the middle channel from both about 7 times greater than the decay power, the center and outer channels until the oxidation The core damage was also affected by the bal-rate decreased when most of the cladding had been looning. With the cladding ballooned, oxidation oxidized comp etely.

occurred on both the inner and outer cladding sur-Temperatures of the upper plenum structures faces. By the time the temperature reached levels at just above each of the core channels in Case 3 are which relocation might occur, the cladding was presented in Figure 48. The temperatures increased completely oxidized, so that no relocation of fuel steadily throughout the transient, with the heatup rods occurred in the ballooned components. Relo-rate increasing at about 246 min as accelerated oxi-cation did occur in the control rods and the unbal.

dation began in the core. By the end of the calcula-looned fuel rods,in Case 3,31% of the Zircaloy in tion, the structures above the center and middle the outer channel fuel rods had melted and relo-channels were hot enough that they could begin to cated by the end of the calculation. The molten melt. The highest temperatures of other internal Zircaloy dissolved 1.0% of the fuelin that channel, structures at the end of the calculations were allowing it to relocate with the Zircaloy to lower 1734 K for the control rod guide housings,1670 K portions of the fuei rods; none of the fuel rod mate-for the core baffle plates, and 1315 K for the core rial dropped into the lower plenum. The top 60%

barrel. The upper head had the highest reactor ves-of the control rods in each of the channels

2000 i

i O center chonnel nM O Widdle chonnet v

A Outer channel e

6

3

• • 1500 6o Q.

Ee

.e.

8 1000 O

occelerated

>4 elodding onldelien 1

500 14 0 200 240 280 Time (min)

Figure 48. Volume average temperatures of the upper plenum structures at the outlet of the three core channels for sensitivity Case 3.

200 i

O Decay power O 0xidotten power 15 0 n

]

32 v

6 10 0 e

No I

CL 50 l

C O

O O

O O

O O

O C

0 0

1 0C O

O O

O'O O

-0

-0

'O

^

14 0 200 240 280 Time (min)

Figure 49. Decay and oxidation power for sensitisity Case 3.

l 56

relocated, with 75% of the relocated absorber the energy removed from the core, compared to val-material and none of the stainless steel and Zircaloy ues of 4.8% and 23.1%, respectively, in the base flowing into the lower plenum, case.

The longer transient also resulted in higher fis-The effect of the higher energy deposition on the sion product releases than the base case, in Case 3, loop structure temperatures is show n in Figures 53 6.1% of the xenon and krypton, 4.7% of the through 55. Figure 53 presents the hot leg tempera-cesium, ?.0% of the iodine, and 6.6% of the tellu-tures nearest the reactor vessel for each of the three rium in the fuel rods at the beginning of the tran-loops. As in previous cases, the Loop C hot leg sient was released by the time the surge line failed, temperature was slightly higher than that of the Table 12 presents the plant conditions for other two loops because of the PORY cycling. The Cases 1,3, and 4 near the time of the RCS failures.

temperatures increased steadily until shortly after in general, the longer the transient lasted, the more accelerated oxidation began in the core, at about severe the damage was at the time of the RCS fail-256 min in the center channel and at about 275 min ure. The integral core heat removal for Case 3 is in the middle channel. The enhanced oxidation in l

somewhat misleading,in that Figure 42 shows that the core increased the vapor temperature in the the fre.ction increased to a value above that of upper plenum and the heatup rate of the hot leg Case 1 after 260 min.

piping. Figure 54 shows the highest steam genera-tor tube temperatures in each of the loops. The 4.3.2 No inlet Plenum Mixing. For Case 4, the pressurizer loop temperature was higher than the flow from the top of the hot legs proceeded directly other loops, again because of the efrect of the into 11 - steam generator tubes, and the flow return

• PORY cycling. The highest solume-awragt tem-ing from the tubes went directly into ?he lower part peratures in the hot leg, surge line, and steam gen-of the hot legs. No mteractions b.< ween the flows crator tubes for Case 4 are presented in Figure 55, occurred in the steam gene.ator inlet plena.

The steam generator tubes were the hottest struc-Table 11 presents the sequuce of events for this ture until 250 min because they were the thinnest transient. Two major differences between this and structure and there was a smaller difference the base case resulted from the slower heatup. Bal-between the hot leg and, team generator tube vapor loomng occurred in the center core channel and the temperatures than in the other calculations. The RCS failure location was the piessurizer loop hot hot leg near the reactor sessel was the hottest struc-(Vith no mixing in the steam generator inlet ture after 250 min. Its high temperature led to creep rupture failure at 290.5 min, at a temperature of plenum, the hot leg flow rate increased more than 1233 K. The hot legs in the other two loops failed at in Case 3. Figure 50 shows the flow entering the 291.9 min, and the surge line at 296.9 min. None top of the loop A hot leg from the reactor vessel f the steam generator tubes had failed by the end for Cases 1 and 4. The trend of the now was the of the calculation. The surge line temperature was same in both cases, with higher flow rates in Case 4. The hot leg flow as a function of inlet tem-strongly affected by the PORY cycling. When the valves were open, the surge line temperature perature is shown in Figure $1. Compared to the increased rapidly. When the valves were closed, the base case, the now at any gisen temperature was 50-60% higher in Case 4. Cooler vapor returning temperature slowly decreased. The calculated surge to the upper plenum inereased the density gradient, line temperature was probably too low. The tem-so the recirculating flow in the upper plenum was perature remained near saturation until the pressun also 10-20% higher in Case 4 than in the base case.

izer was nearly empty of liquid. Liquid draining As in Case 3, more energy was removed from the from the pressurizer kept the upper part of the core and deposited in the loops as a result of the hot surge line cool, while the lower part heated up. In leg flow increase. Figure $2 shows the fraction of reality, this behavior should be reversed, with the the core heat removed during Cases 1 and 4. Near liquid draining through the lower half of the surge the time of the RCS failure, heat transfer to the hot line.The effect of the change may be that the surge legs accounted for 3.9% of the energy removed line would fail before the hot legs, given the more from the core, and heat transfer to the steam gener-rapid temperature increase of the surge line once it ator tubes and tube sheets accounted for 37.0% of did begin to heat up.

i i

57

l l

l l

Table 12. Conditions near the time of the reactor coolant system failure for the base case and steam geneistor inlet plenum mixing sensitivity calculations Value Paramral Case I Case 3 Case 4 Time (min) 246.3 255.3 285.8 Center channel peak clad temperature (K) 2066 2763 2562 htiddle channel peak clad temperature (K) 1926 2601 2600 Outer channel peak clad temperature (K) 1618 3017 2633 hiaximum upper plenum structure temp. (K) 1430 1488 1718 hiaximum hot leg temperature (K) 1012 1012 1175 hiaximum steam generator tube temp. (K) 805 905 1110 h1aximum surge line temperature (K) 1219 1262 1176 Core outlet flow (kg/s) 11.5 5.2 6.2 Core return flow (kg/s) 9.0 2.6 3.5 Upper plenum recirculating flow (kg/s) 37.1 36.5 32.2 Reactor vessel collapsed liquid level (m) 1.38 1.44 0.75 Core heat removal (re)a 75.4 74.1 79.4 Core energy removed and deposited in:a Loop A structures (re) 9.0 11.6 13.8 loop B structures (re) 9.0 11.6 13.8 Loop C structures (re) 11.8 12.8 16.4 flot leg piping (re) 4.8 3.6 3.9 Steam generator tubes, tube sheets (re) 23.1 30.1 37.0 Surge line piping (re) 0.8 0.7 0.5 Center channel oxidation (re) 20 61 63 Allddle channel oxidation (re) 17 62 63 Outer channel oxidation (re) 10 28 62 Hydrogen generated (kg) 109 382 440 Fuel relocation (re) 0.0 0.3 0.0 Fuel rod cladding relocation (re) 0.0 4.1 0.0 Control rod relocation (re) 33 60 70 Xenon / krypton release (re) 1.0 6.1 19.4 Cesium release (re) 0.6 4.7 14.9 lodine release (r )

0.0 3.0 12.4 e

Tellurium release (re) 1.3 6.6 13.3 Reactor coolant system failure time (min) 246.3 254.3 290.5

a. Intesral quantnies frorn the start of the calculation.

58 v----

i l

l i

10 O case 4 O cas 1 n

M

\\

J CD dd 5

\\

5 -

t1e WP m nQ, e

j b" iu,,o -[' B -y g

~ q,

[

o S w I

l g

()

o D

C 0

l M

M O2

-5 14 0 180 200 220 240 280 280 300 Time (min)

Figure 50. Upper hot leg mass now in imop A for sensitivity Cases 4 and 1.

10 i

7N Cn

'!N U jii*:'elld(. ". % '/k % ' +*~' %,: e .i 0 ,i H Gr.v,0.g. ,s -,, W ,, ',, Mr Cr 8 y o ., 1 C s 1 0 l M i: 9. g O2 i -5 500 1000 1500 2000 i Vapor temperat ure (K) Figure 51. Ilot leg now as a function of hot leg inlet vapor temperature for sensitivity Case 4. 59

1.5 O cose 4 O case 1 ~ C ~ O 2 ;, "En 5 C C 3 u Y w 7 0 w L La. 0.5 - l i i i e i oc 14 0 180 200 220 240 260 280 300 Time (min) Figure 52. Fraction of the core heat removed by the coolant for sensitivity Cases 4 and 1. J 1500 8 i i i i O Loop A mM O Loop a b Loop C h e6 1250 3 O i 6 occelerated e 'I'ddInf 1000 oxidelion f ~ e r e cn 8 750 f e [ r O i 8 i e i f 500 14 0 18 0 200 220 240 280 300 Time (min) 260 j { j Figure 53. Volume average hot leg pipe ternperatures near the reactor vessel for sensitivity Case 4. l t ] r 60

1200 ) O Loop A mM A Loop B v O Loop C B '~ e 6 3 1000 6e Q. E l e 8 800 O Le>< t 600 14 0 18 0 200 220 240 260 280 300 Time (min) Figure 54. liighest volume average steam generator tube temperatures in the three coolant loops for sensitivity Case 4. 1500 O Hol leg nM O Surge Itne V A Sleom generolor tubes ) eL 1250 creep rupture 3 f ailur e ) k O 6e Q. 1 E 1000 e I em 6 750 e> 4 ^ % n -Q ^^^ y 500 a 16 0 18 0 200 220 240 260 280 300 Time (min) 1 Figure 56. Ilighest solume average pipe temperatures in the loop C hot leg, surge line, and steam generator tubes for sensitivity Case 4. 61

i ~ Figure 56 shows the fuel rod cladding surface Case 3 and 109 kg in Case 1. Figure 60 shows the temperatures from the top node in each of the three heat generated by the oxidation reaction from core channels. The difference between the center Case 4 together with the decay power. The peak and middle channel temperatures increased even oxidation power was about eight times greater than more than in Case 3 because the heat transfer in the the decay power. loops had increased more. The center channel The core damage was again affected by the bal. heatup rate increased at about 250 min, when the looning,in that double-sided oxidation of the clad-cladding ballooned and ruptured. Oxidation of the ding led to no relocation of fuel rods in the center inner claddu.g surface again caused the faster channel. The slow heatup also allowed the cladding i heatup. In the other channels to oxidize completely before Peak cladding temperatures from Cases 1 and 4 Zircaloy relocation could occur. Relocation did i are shown in Figure $7. The curves exhibited the occur in the control rods and the unballooned fuel same behavior, but with the time extended in rods. In Case 4,0.3 kg of fuel melted and relocated Case 4. The higher heat removal in the loops led to in the middle and outer channels. The upper 70% increased heat removal from the core and a slower of the control rods in each of the channels relo. heatup. The slower core heatup allowed the clad-cated, with none of the structural material and ding to balloon in the center channel in Case 4, 79% of the absorber relocating to the lower reducing the axial now area by 62% over the upper plenum. 60% of the fuel rod at about 250 min. More assion products were released than in the The effects of the ballooning on the core nows base case because of the longer heatup and time at were more localized in Case 4 than in Case 3. Mass high temperature prior to the RCS pressure bound-How rates entering the upper plenum from the three ary failure, in Case 4,19% of the noble gases,15% core channels and the core bypass are presented in of the cesium,12% of thelodine, and 13% of the Figure 58. The center channel core exit now tellurium had been released before the hot leg decreased when the ballooning occurred at about failed. ( 250 min. Flow was diverted around the blocked Table 12 presents the plant conditions for [ region, increasing the middle channel now. The Cases I,3, and 4 near the time of the RCS failures. [ now into the outer channel, as well as the distance in general, the darr. age was at the time of the RCS j it penetrated down the channel, were essentially failure was more severe in Case 4 because the tran-unchanged, sient lasted longer and was at higher temperatures Figure 59 presents the core exit upper plenum for a longer period of time than the other two cases. structure temperatures for Case 4. Shortly after The loop heat structure temperatures in Case 4 ~ 250 min, rapid heatup of the center core channel showed that the temperatures were relatively close caused a more rapid heatup of the upper plenum together throughout the loops. structures. The heatup rate slowed as much of the cladding in the center channel had been oxidized, 4.3.3 Summary. Decreasing the amount of mix-j 4 q then increased again at about 268 min when the ing in the steam generator inlet plena resulted in t rapid heatup of the middle core channel began. The slower core heatups and more energy deposition in i temperatures were high enough at the end of the the loop structures. Despite the increased energy [ j calculation that significant oxidation and melting deposition in the loops,'. ie steam generator tubes of the stainten steel structures in the upper plenum were not predicted to fallla either sensitivity calcu. might occur. At the end of the calculation, the lation. The surge line failed 8.5 min later than in highest temperatures in other structures in the reac. the base case with a 25% increase in the hot leg l tot vessel were 1507 K for the core barrel,1840 K countercurrent flow (Case 2). With a 50% increase l j for the control rod guide housings,832 K for the in the hot les now (Case 4), the RCS failure loca. l reactor sessel at the norile elevation, and 654 K for tion was the hot leg,44.2 min later than the failure i the head bolt region. The top 60% of the core baf-in the base case. The slower heatup also resulted in l i ne was above 1760 K, so that these plates would be ballooning of the fuel rod cladding. Extensive ) expected to melt. sausage type ballooning in both Cases 3 and 4 l The slower heatup led to more extensive oxida, resulted in flow redistributions, but not in a change t ] tion of the cladding at the time of the RCS pressure of the basic now pattern in the core. Temperatures i I boundary failure. In Case 4,440 kg of hydrogen in the core and upper plenum were high enough l had been generated before the hot leg failure, com. that melting of some internd structures could j 4 pared to 382 kg near the time of surgeline failure in occur near the time the RCS piping falls. l g 62 I l I l

4000 i i i i i O center chonnel O utddie chonnoi 6 Outer chonnel 73@ v belooning g O 2000 E ^ 1000 I i 0 18 0 180 200 220 240 240 240 300 Time (min) Figure 56. Fuel rod cladding surface temperatures at the top of the three core channels for sensitivity Case 4. 1 4000 O CJse 4 O Case 1 1 l 7 3000 F,< 1 3 b L v \\ Y (2 L N O 2000 L ) 4 E. WM C i 0 14 0 18 0 200 220 240 260 240 300 Time (min) Figure 67. Peak cladding temperatures for sensitivity Cases 4 and I, 63

50 i O center chonnel O Widdle chonnel 7 A Outer chonnel N X core bypos O 25 l ? '" Nauua C =Cz -- 3z rt_. 3,_ y 0f id'

• C-E 52

_v m a g i-y ,,__m x ._w 2 i y -25 bolooning 02 -50 16 0 18 0 200 220 240 260 280 300 Time (min) Figure 68. Mass flow rates exiting the three core channels and the core bypass for sensitivity Case 4. 2000 O center channel m 3 x O widdle chonnel A Outer channel ) U e L i 3 1500 wv O. E .o & 1000 o 6e>< gp,, - m 500 16 0 16 0 200 220 40 260 300 min) 260 Time Figure 59. Volume average temperatures of the upper plenum structures at the Outlet of the three core channels for sensithity Case 4. I 64 l

300 O Decay power O oxidofion power $ 200 3 se l Do Q-C ,oo I I A l i ) Ann /k l L n n n 3 m m m m m m j my m ' 'O OC O 'O O' O O O O 14 0 14 0 200 220 240 240 280 300 Time (min) Figure 60. Decay and oxidation power for sensitivity Case 4. i J 1 4.4 Piping Heat Loss Sensitivity of the heat loss were localized in the piping. The overall energy removal from the core wts nearly the same as in the base case, as illustrated in Figure 61, flow the imulation on the plant will perform under mere accident conditions is not well known. Two sen-which shows the integral energy removal from the core for Cases 5 and 1. Heat transfer to the coolant i sitivity calculations were performed to imestigate the effects of heat loss from the hot leg and surge line loops was slightly higher than in the base case. piping. The first applied a convective heat transfer lleat loss from the piping kept the hot legs and i coef0cient to the outer surface of the pipes. The sec. surge line cooler than in the base case. i I ond applied convective and radiative heat transfer l'igure 62 shows the fraction of ti;e energy coef0cients to the outer surface of the pipe, removed from the core that was lost thtough the piping to the containment. The fraction remained fairly e nstant near 0.G4 through most of the tran-4.4.1 Convective Boundary Condition. For Case 5, a constant comective heat transfer coeffi-sient, indicating that 4te of the energy was being cient of 28.4 W/m K was applied to the outside lost at any point in time as well. Since cbout 75 fe of 2 I surface of the hot leg and surge line piping. This is a the core power was being removed by the coolant, 1 high heat transfer coefficient for natural convec. about 3fe of the core power (roughly 0.65 MW) ] tion vapor flow.16 and, since it is applied directly was belng transferred through the uninsulated to the pipe outer wall, no credit is taken for any pipes. The decrease in the fraction near the end of i insulation. The heat sink (containment) tempera, the calculation occurred when the core was heating i ture was anumed to be a constant 311 K. up rapidly, with more of the energy removed from The sequence of esents for the base case and the core being stored in structures within the reac-both heat loss semitivity cases are contained in tor vessel. This was a temporary redistribution of 1 Table 13. Convective heat loss through the piping the energy, in that the heat loss would be expected I in Case 5 delayed the omet of various stages of core to return to about 4'~e if the calculation were i damage, but only by 12 min. The surge line failure continued. was delayed by 6.6 min. The larger difference in the The fraction of the energy removed from the core surge line failure time occurred because the effects that was deposited in the steam generator tubes and 1 I 65

I i Table 13. Sequence of events for the base case and piping heat lo4s sensitivity calculations 1 Time E (min) 't Event Case 1 Case 5 Case 6 i r Fuel rod cladding oxidation begins I Center channel 185.3 185.5 185.7 Middle channel 186.1 186.0 186.0 2 Outer channel 192.6 193.1 193.7 1.iquid level drops below core 190.2 192.2 In2.4 Pressurizer empties ofliquid 224.8 225.8 225.0 Fuel rod cladding fails a 1 Center channel 223.4 224.1 227.5 Middle channel 225.3 226.2 228.9 Outer channel 241.3 241.8 243.0 Fuel rod relocation begint Center channel 248.0 249.8 251.8 ? Middle channel 248.8 250.7 252.5 4 Outer channel 253.2 255.0 RCS pressure boundary falls 246.3 252.9 259.2 l RCS failure location Surge line Surge line Surge line l Calculation terrahated 250.0 262.9 266.7 l j i l i i 1 i l l r 1 i 1 m b t

i I 1.3 O cose 5 O cose t c O = N 0^M J0 L ) L 1 l 0.5 OC 16 0 200 240 280 Time (min) Figure 61. Fraction of the core heat removed by the coolant for sensitivity Cases 5 and 1. 0.06 i 0'04 ~ -- C o o Q L L 0.02 0.00 16 0 200 40 280 Figure 62. Fraction of the heat remosed from the core that was lost through the piping to the containment for sensitisity Case 5. 67

tube sheets is presented in Figure 63 for Cases 5 wall in Case 6. The pipe was assumed to be a dif-and 1. Slightly les, energy was -leposited in Case 5 fuse gray emitter, and the containment was because slightly cooler vapor was entering the assumed to be a black body absorber at a constant steam generators. There was more heat transfer temperature of 311 K. The emissisity varied lin-from the vapor to the hot leg in Case 5 because of early from 0.24 to 0.31 as the pipe outer surface the heat loss to the containment. temperature increased from 480 to 1310 K.17 The The volume average temperatrres of the hottest resulting radiation heat transfer coefficient was parts of the hot leg, steam generator tubes, and about 20% of the convection heat transfer coeffi-surge line are presented in Figure 64. As in previous cient at a wall surface temperature of 600 K, and cases, th, pressurizer loop structure temperatures about 130% at 1150 K. were slightly higher than similar structures in the The sequence of r: vents for Case 6 is presented in other two loops. The hot leg and surge line temper-Table 13. The effects of the additional heat loss atures were lower than in the base case because of became more evident as the transient progressed the heat loss through those pipes, while the steam and the piping temperatures increased. Fuel rod generator tubes were also cooler than in the base relocation was delayed by nearly 4 min, and the case because of the heat loss through the hot leg surge line failure was delayed nearly 13 min, com-piping, which reduced the temperature of the vapor pared to the base case, entering the steam generator tubes. The surge line The heat loss through the piping represented was the hottest structure throughout the transient about 5% of the energy removed from the core, or and the steam generator tubes the coldest for most about 3.8% of the core heat generation. However, of the time. The temperatures increased at a faster as in Case 5, this heat loss from the piping did not rate at about 249 min, after the rapid heatup in the alfeet the hot leg flow rate. Figure 67 shows the hot core began. Failure of the surge line occurred at leg inlet flow as a function of the inlet vapor tem. 252.9 min, at a temperature of 1256 K. The heat perature. The flows at any temperature were the loss through the piping increased the temperature same as those in Case 5 and the base case. The difference across the hot legs from 97 K in the base additional heat loss in the hot leg did lead to a fur-case to 193 K, and across the surge line from 8 to ther reduction in the heat transfer in the steam gen. 45 K near the time of the surge line failure. erators, and hence'in the steam generator tube Figure 65 plots the hot leg inlet vapor flow temperatures, against the inlet vapor temperature. The flow at As in Case 5, the effects of the heat loss were any gisen temperature was the same as that in the primarily noticeable in the loop piping tempera-base case (Figure 36). This means that heat loss tures. Figure 68 shows the highest hot leg, steam through the hot leg piping was compensated for by generator tube, and surge line temperatures in the less heat transfer in the steam generators. pressurizer loop. The temperatures in the other two i Peak cladding temperatures from Cases 5 and I loops were lower than those shown in the figure. are presented in Figure 66. A slight delay in the Again, the loep structure heatup rates increased I onset of the rapid heatup near a temperature of shortly after t he core heatup rate inereased at about 1850 K was the only real difference between the two 250 min. The surge line failed at 259.2 min, when calculations. The delay resulted from the slightly its temperature was 1250 K. The heat loss through t highet energy remosalin theloops in Case 5. Other the pipes increased the temperature difference I temperatures throughout the reactor vessel were between the inner and outer walls, Near the time of also very similar in these two cases, as were the the surge line rallure, the hot leg temperature differ. 0*** ence had increased from 97 K in the base case to The conditions of Ihe plant near the time of 254 K in Case 6, and the surge line iemperature dif. j surge line failure are presented in Table 14 for ference had increased from 8 to 125 K. Cases 1,5, and 6. The differences between Case 5 Figure 69 shows the peak cladding temperatures and Case I were the result of the later time in from Case 6 and the base case. The heat loss Case 5, which allowed the core to heat up further' through the pipes increased the heat remosalin the oxidize more, and release more fission products. loops, resulting in about a 4 min delay in the onset J 4.4.2 Convective and Radiative Boundary Con-of the rapid core heatup.1.ike the cladding temper. [ dition, in addition to the comectise boundary ature, structure temperatures throughout the condition applied in Case 5, a radiatise heat trans. reactor sessel were sery clo:e to the salues in the i fer coefficient was also modeled on the outer pipe base case. l 68 i l 1

i 0.8 O cose s O case 1 0.4 = l 0 O I L L 0.2 l 4 0.0 C Time (minf Figure 63. Fraction of the heat remosed from the core that was transferred to the steam generator tubes and tube sheets for sensitivity Cases 5 and 1. l. i 1500 O Hot leg l mM O surge nne A Stoom generator tubes orsep rupture 6 3 folure Q b e Q. i E 1000 .e d i e m { O 'e i J 4: gg hd 1 t e i 500 14 0 200 240 280 Time (min) i Figure 64, liighest solume aserage pipe temperatures in the loop C hot leg, surge line, and steam j q generator tubes for sensitivity Case $. 1 l 69 i d

m i nMN cm i 6 s e

k "NAhh 1

';;iin,,t t ',.8;t? It e,.i i(" 'J. yM". r e . s..... M M l. l 8 i -s 500 1000 1500 2000 Vapor temperature (K) Figure M. Hot leg flow as a function of hot leg inlet upor temperature for sensitivity Case $. l i ' ',00 l lO0 case 5 cose 1 e Q 3000 L Cl i e ri L. 3 O 2000 } 6 ottelera t ed onldellen i e H 1000 e C [ O '40 200 240 280 ) Time (min) ) F4.se M. Peak cladding temperatures for sensithity cases 5 and 1. I i 9 70 i

~% Table 14. Conditions near the time of the surge line failure for the base case and piping heat loss sensitivity calculations i 1 Value [ , Parameter Case I Case 5 C.sse 6 Time (min) 246.3 250.0 258.3 Center channel peak clad temperature (K) 2066 2577 2809 htiddle channel peak clad temperature (K) 1926 2300 2783 Outer channel peak clad temperature (K) 1618 1679 3120 L h1aximum uppn plenum structure temp. (K) 1430 1500 1767 hiaximum hot leg temperature (K) 1012 953 1014 hiasimum steam generator tube temp. (K) 805 816 848 i hiaxirnum surge line temperature (K) 1219 1125 1202 5 Core outlet flow (kg/s) 11.5 9.4 6.9 Core return flow (kg/s) 9.0 8.0 5.2 Upper plenum recirculatin. ",w e 37.1 36.4 30.3 Reactor sessel collapsed liquit o 1.38 1.33 1,11 ] Core heat removal (r,)a 75.4 74.6 72.3 Core energy removed and deposited in:a Loop A structures (r ) 9.0 8.3 8.1 e Loop B structures (*.'e) 9.0 8.3 8.1 Loop C structures (r,) 3i,g i1,t 30,7 Ilot les piping (re) 4.8 3.3 3.2 i Steam generator tubes, tube sheets (r ) 23.1 22.5 21.5 l e ] Surge line piping (re) 0.6 0.7 0.6 4,o 4,7 l Core energy removed through piping (r )a e Center channel osidation (r ) 20 30 55 e hliddle channel oxidation (re) 17 23 55 l Outer channel oxidation (r,) 30 g 34 l liydrogen generated (kg) 109 152 384 Fuel relocation (%) 0.0 0.0 0.1

i

) Fuel rod cladding relocation (re) 0.0 0.0 0.0 i i Control rod relocation (r ) 33 50 60 e I Xenon / krypton release (re) 1,0 1.6 18.8 l l Cesium release (re) 0.6 1.0 14.8 j 1 lodine release (r,) o,0 0,3 32,3 Tcliurium release (re) 1.3 1.9 12.5 i Surge line failure time (min) 246.3 252.9 259.2 l i l l, a, Integrat quantities fiom the Mart of the cakulation. 1 l 1 l i 71 1 1

m nn N cn5 s e & AS4%*4 N fi 6 g, O

P"'

j ~ ~ > ,;,,'j;<! g ! t t $,A D D '... <' ~. 'l h ,i i c ~ n a i. 3 v -s soo woo isoo 2000 Vapor temperature (K) Figure 87. Hot les flow as a function of hot leg inlet vapor temperature for sensitivity case 6. 2 isoo i O Ho t leg n 1 x O stoom generator tuben 1 V 6 Surge line j steep rupture f ailure 1250 1 3 J gg, O i M ~ t 1-g 7M t ,>4 I [ l 500 - 16 o 20o 240 28o Time (min) j Figure St. Highest volume-aserage pipe temperatures in the loop C hot leg, surge line, and steam i generator tubes for sensitivity Case 6. i i .i l 1 72 i 1 s

4000 i i O cass 6 O cose t 1ll 0 7 3000 v s E f a 2v00 l s. I e J m C 0 10 200 240 280 Time (min) Figure 69. Peak cladding temperateres for sensitivity Cases 6 and 1. Some of the plant conditions near the time of the with a convective boundary condition, and surge line failure are presented in Table 14. Differ. 12.9 min with a convectise and radiative boundary ences in the temperatures and the extent of core condition. The heat loss resulted in smaller delays damage were attributable to the lawr time at whkh in the onset of rapii core heatup and relocation the conditions were evaluated in Case 6, since the because most of the effects of the heat loss were core heatups were similar. The effects of the later localized in the hot legs and surge !!ne. Increased time, and longer time at high temperatures, are heat transfer from the vapor flowing in the hot legs seen particularly in the outer channel cladding tem-led to lower temperatures in the steam generators, perato. and the extent of the core damage. The and hence lower steam generator tube temperatures cladding in the outer channelin Case 6 had already than in the base case. The piping heat loss did not gone through the heatt.p associated with acceler-affect the hot leg flow rate, in that the flow rate at ated oxidation, m its peak temperature was much any given inlet vapor temperature was unchanged higher than in the other two cases. The longer time from the base case in both Case $ and Case 6. at high temperatures throughout the core also resulted in more oxidation and much greater fission 4.5 Crossflow Resistance product release in Case 6 than in Cases I knd $. Sens.t. ity i iv Because of the increased heat transfer from the hot leg and surge line piples, those structures retairied less of the energy removed from the core in Case 6 The natural circulation flow within the reactor than in Case 1. As discussed sbove, the steam gen-vessel has both axial and isdial components. Many cratcr tubes and tube sheets also removed a smaller structures are present in the upper plenum and core fraction of the core energy because of the heat loss to pr vidt resistance to flow la the radial direction, along the hot legs. such as fuel rods, control rod drise shafts, control rod guide housings, and support columns. The sen-sitivity of the calculated in-vessel (lows was investi-4.4.3 Summary. Modeling of heat loss from the gated by cht.nging the erossflow loss coefficients, surge tine and hot legs dela>rd the surge line failure These loss coefficients were increased and compared to the base case. The delay was 6.6 min decreased by factors of 10, 73

4.5.1 Decreased Upper Plenum Crossflow rupture of the surgeline occurred at 244.9 min, at a Resistance. The Case 7 crossflow loss coeffi-tempera'ure of 1258 K. cients in the upper plenum were a factor of 10 lower Table 15 presents the sequence of events for this than the base case model values. The upper plenum casr, as well as for the other crossuow resistance recirculating f'ow for Cases 7 and I are shown in sensitivity cases and the base case. The damage Figure 70. The decreased crossuow resistance in progression was slightly accelerated compared to Case 7 resulted in a higher recirculating flow in the the base case, with fuel rod relocation beginning upper plenum than in the base case. Figure 71 3.3 min earlier and the surge line failing 1.4 min shows the upper plenum recirculating now as a earlier. function of the vapor temperature at toe outlet of Table 16 lists some of the plant conditions near the center core channel. The Dow at any given tem-the time of the surge line failure for the base case perature was 30-40% higher than in the base case, and the crossuow resistance sensitivity cases. In The increased now in the upper plenum changed comparing Cases 7 and I, it is seen that the tem-the energy distribution in the upper plenum peratures in case 7 werc higher throughout the sys-slightly, increasing the temperatures of the vapor tem, because of the earlier rapid core heatup. The and structures in the outer channel. The higher core damage was slightly more extensive for the vapor temperatures resultut in slightly less Dow same reason. The energy removal in the loops was returning to the core, w hich in turn caused the core nearly identicalin the two cases, indicating that the to heat up slightly faster. Figure 72 shows the vapor hot leg flow was unaffected by the changes in the i temperature in the upper plenum just above the upper plenum model, as would be expected. outer core channel for Cases 7 and 1. Slightly 4.52. Decreased Core Crossflow Resistance. In higher temperatures were also present in the strue-sensitivity Case 8, the crossnew loss coefficients tures above the outer core channel and in the core beem the &ree core channels were decreased by a barrel temperatures in the upper plenum. facwe v. m.naking it easier for How to move radi-Figure 73 shows the peak cladding temperatures ally between the channels. from Cases 7 and 1. The temperatures began t The sequence of esents for Case 8 is presented in desiate after 215 min, the result of the lower return Table 15. The core heatup and the onset of various flow frem the upper plenum. The more rapid damage stages occurred slightly earlier than in the l heatup ofIhe core anociated with the change in the base case. An increased recirculating Dow within Zircaloy oxidation kinetics began at about the core was responsible for the faster core heatup. 240 min At the top of the core, the upper end boxes present j The amount of core energy removed by the cool

• a reasonably large axial Dow resistance that tends ant is shown in Figure 74 for Cases 7 and 1. The to drive the now radially, in the base case, the energy remosal was nearly identical in both cases radial (crossuow) resistances were also large, so until late in the calculations, w hen the difference in that the radial now component between the upper-the timing of the rapid core heatup was renected in most core nodes was small. With the decreased an earlier drop in the remosal fraction in Case 7.

cross 00w resistance, this now increased, prosiding ( Figure 75 presents the now entering the top of more vapor from the top of the middle channel to the loop A het les for Cases 5 and I. The now the top of the outer channel. This vapor then mixed rates when the i ORVs were closed were nectly with the sapor entering the core from the upper I q identical, as was the flow at any given temperature, plenum, increasing its temperature. The increased The temperatures of some of the loop structures vapor temperature, and hence lower vapor density, [ are shown in Figure 76. The temperature differ-in the outer chaunct resulted in a lower buoyant enees between the loops were small, with the pres-drising head, reducing both the recirculating How I suriser loop temperatures being higher. The between the core and upper plenum, and the dis-j temperature increase during the transient was tance the outer channel now was able to penetrate steady, with the surge line heatup seeclersting w hen down the cote. With less now, less cooling of the l the pressurizer emptied of liquid at about 225 min core occurred, and the core heated up faster. l (because liquid was no longer draining through the The ineressed outer channel tempnature is illus-l surge line, helping to cool it), and both the surge tiated in Figure 77, which presents the fuel rod line and hot leg heatup rates increasing when the cladding surface temperatures at the top of the core j core heatup acerlerated at about 240 min (because for each of the three core channels. The tempera-hotter sapor was leasing the reactor sessel). Creep tures were closer together than in the base case 74

15 0 i O case 7 O case 1 m M l !~AD%, ~ o gj 50 1W )o C h0 0 2 -50 40 20 200 220 240 260 Time (min) Figure 70. Upper plenum recirculating mass flow for sensitivity Cases 7 and 1. 15 0 7 bi ii lt N '$ Ul lie ym N q.y%M.,,, i. v D {f.' g%. e 'N [k5 t v v.p. ',- - o ,e 4.i 5-i.,.; 3o N oc 0 2 n .n 500 1000 1500 2000 Vapor temperature (K) Figure 71. Upper plenum recirculating mass flow rate as a function of muimum upper plenum sapor temperature for sensitisity Case 7. 75

2000 O Case 7 mx O case 1 v e 6 3 1500 6e Q. E. e. g 1000 0 l 'e-C 500 14 0 18 0 200 220 240 260 Time (min) Figure 72. Volume average temperature of the upper plenum structure at the outlet of the outer core channel for sensitivity Cases 7 and 1. i i I 4000 O Case 7 O Case 1 l Q 3000 } e I s 3 O 2000 eecelereted J t .. i wi. Q. Ee 1000 [ 0 14 0 18 0 200 220 240 260 Time (min) Figure 73. Peak cladding temperatures for sensitivity Cases 7 and I, a 1 76

1.5 2 O Case 7 O Case 1 1 o 0 C +-C O ^ 6 ts. 0.5 l l s CC 16 0 18 0 200 220 240 260 Time (min) l Figure 74. Fraction of the core heat remosed by the coolant for sensitivity Cases 7 and 1. to O cose 7 O case i nn N cnU 5 e p(4 L al g p 1 ~ A; l. t t m g E ) g; O l [t U g O N O2 -5 16 0 18 0 200 220 240 260 Time (min) Figure 71,. Upper hot leg mass now in imp A for sensitivity Cases 7 and 1. 77

i s 1500 i i i i O Hol leg f nx O steem generator tubes V a Surge line 7 t eroep rupture feture 3 O 6e j fg. preeeurlier erwir - I I L e> j f i i e i g 1 wo 18 0 200 220 240 240 Time (min) a i Figure 76. liighest volume-average pipe temperatures in the Loop C hot leg, surge line, and steam i ll generator tubes for sensitivity Case 7. I ) f I I r i I 78 l 1 1 l

Table 15. Sequence of events for the base case and crossflow resistance sensitivity calculations Time (min) Event Case ! Case 7 Case 8 Case 9 l Fuel rod cladding oxidation beings j Center channel I85.3 184.8 I85.0 185.6 2 Middle channel !86.1 185.3 185.3 186.9 ] Outer channel 192.6 191.5 191.5 194.3 Liquid level drops below core 190.2 189.5 189.4 192.2 Pressurizer emp:les of!! quid 224.8 224.7 226.0 224.5 l t j Fuel rod cladding balloons i l J Center channel 216.7 + Middle channel 227 Outer channel t ] Fuel rod cladding fails I ) Center channel 223.4 220.9 221.3 223.7 l Middle channel 225.3 222.9 222.9 226.8 Outer channel 241.3 237.5 + 236.3 236.4 I ] Fuel rod relocation begins Center channel 248.0 244.7 243.5 236.5 )' Middle channel 248.8 245.3 244.0 233.2 t Outer channel 247.8 246.0 242.2 [ I t j RCS pressure boundary falls 246.3 244.9 244.6 233.9 I t ] RCS failure location Surge line Surge line Surge line Surge line l Calculation terminated 250.0 250.0 250.0 246.3 l l l l l l l t l 1 l 79 i

.able 18. Conditions near the time of the surge line failure for the base case end crossflow resistence sensitivity calculations Value parameter Case 1 Case 7 Case 8 Case 9 Time (min) 241.7 241.7 241.7 233.3 Center channel peak clad temperature (K) 1752 1935 2058 1698 bliddle channel peak clad temperature (K) 1714 1807 1972 2398 l l Outer channel peak clad temperature (K) 1513 1579 1636 1400 hlastmum upper plenum structure temperature (K) 1355 1383 1396 1339 i hlasimum hot leg temperature (K) 973 982 980 949 blatimum steam generator tube temperature (K) 783 793 794 778 a blatimum surge line temperature (K) I164 1179 1159 1245 Core outlet flow (kg/s) 12.9 11.8 10.8 12.1 Core return flow (Lg/s) 8.6 8.7 8.0 5.7 Upper plenum recirculating now (kg/s) 35.7 50.7 39.2 16.8 Reactor sessel collapsed liquid lesel (m) 1.48 1.43 1.40 1,81 l Core heat removal (r )a 76.1 75.4 74.4 72.6 e Core energy removed and deposited in:a loop A structures (r,) g,9 8.9 8.8 8.6 loop Il structures (r ) 8.9 8.9 8.8 8.6 e loop C structures (r ) 11.6 11.6 11.5 11.0 e llot leg piping (re) 4.6 4.7 4.7 4.5 Steam generator tubes. 22.8 22.7 22.6 21.9 tube sheets (r ) e Surge line piping (r ) 0.8 0.8 0.8 1.0 e j Center channel oxidation (re) 14 17 19 1I hiiddle channel oxidation (r ) 12 15 16 31 e Outer channel osidation (r.) 7 9

o 4

l 1 Hydrogen generated (kg) 80 96 108 150 ~ l Fuel relocation (r ) 0.0 0.0 0.0 0.5 e l Fuel rod cladding relocation (r ) 0.0 0.0 0.0 7.2 e 1 Control rod relocation (r ) 3 27 35 33 l e Xenon / krypton release (r ) 0,g 0,9 0,9 g,g 1 4 Cesium release (r ) 0.5 0.5 0.5 1.3 e I lodine release (r ) 0.0 0.0 0.0 0.6 e Tellurium release (r ) 1.1 1.2 1.3 2.1 e 1 Surge line failure time (min) 246.3 244.9 244.6 233.9 [ ] ) .. inte,r i quaniitie, trom ihe,ian or the c.ieui.iion. r t 1 L i i j 80 I I

4000 O Center channel O Widdle chonnel 6 Ou t e r chonnel 7 3000 k + e j 6 I O 2000 L eseelerated ontdellen H 1000 1 0 14 0 14 0 200 220 240 240 Time (min) Figure 77. Fuct rod cladding surface temperatures at the top of the three core channels for sensitivity Case 8. J r 4 (Figure 25), particularly in the center and middle case. Other structures in the reactor vessel exhibited channels, which were nearly identical. similar behavior compared to the base case. 1 Peak cladding temperatures from Cases 8 and i Flows leaving the top of the core and core bypass are presented in Figure 78. The temperatures began are show n in Figure 81, together with the now leav. to deviate at about 217 min, with the higher tem-ng the outer channel in the base case. The now peratures in Case 8 leading to the rapid core heatup behavior was the same as in the base case, but the about 4 min sooner than in the base case

• magnit'ades of the now s were lower, similar to what The increased recirculation within the core coup-is show n for the outer channel. The upper plenum led with the decreased now between the core and recirculating now was the same as in the base case upper plenum combined to reduce the amount of at any gh en center core channel outlet temperature, energy removed from the core. Figure 79 shaws the fraction of the core energy removed by the coolant Howeser, as discussed earlier, the initial upper plenum heatup was slower than in the base case, so i

durir.g the transient for Cases 8 and 1. As did the the now was lower, temperatures, the energy removal began to deviate at about 217 min.The lower heat removalin Case 8 led Volume average metal temperatures in the pres. to the faster core heatup. The faster core heatup led surizer loop hot leg, surge line, and steam genera-l to a faster heatup of structures in the upper plenum tor tubes are shown in Figure 82. As in the other and the coolant loops, calculations, the non-pressurizer loop structure Figure 80 shows the upper plenum metal temper. temperatures were lower than those in the pressur. J atures at the outlet of the center and outer core izer loop. The surge line heatup rate increased at j channels for Cases 8 and I. The temperatures in about 226 min, when the pressurizer dried out, Case 8 were lower until about 216 min because because liquid draining from the pressuriier was no more energy was being retained in the core rather longer available to help cool the pipe. The hot leg than transferred to the upper plenum. The Case 8 and surge line heatup rates both increased when the temperatures were higher later in the transient corc heatup rate inereased at about 235 min. Creep because the rapid core heatup provided hotter rupture of the surge line occurred at 244.6 min, vapor to the upper plenum earlier than in the base when its temperature was 1252 K. j l 81 1 1

4000 O case a O case i 2 3000 v eu 3 O 2000 L Y e O. E e 1000 [ i 0 14 0 18 0 200 220 240 240 Time (min) Figure 78. Peak cladding temperatures for sensitivity cases 8 and 1. i J i L j L 4 LS O Case 8 O case 1 ) c I 1 0 e =o 0 O - g_ m,, Q = -d u L 0.5 4 l I i oc [ i i i 14 0 18 0 200 220 240 240 Time (min) [ Figure 79. Fraction of the core heat remosed by the coolant for sensitivity Cases 8 ani 1. I I I 82 [ j

2000 i i i i O center chonnet, case a nx O outer chonnel, cose 8 V 6 Center chonnel, case 1 X outer chonnel, case 1 e 6

s

• • 1500 6e CL E

.e & 1000 O Le> 1 4: 1 ( SM 16 0 18 0 200 220 240 260 Time (min) Figure 80. Volume-average temperatures of the upper plenum structures at the outlet of the center and outer core channels for sensitivity Cases 8 and 1. 40 i i i O center channet, cose a O Widdle chonnet, cose a n 6 Outer chonnel. Case 8 H t \\ X Core bypose, cose a I j 20 - 0 outer chonnet. cose 1 ( b 1 yy 1 g_ - o -w-0,, N ' NW" %: m f _- r C -20 2 i i i i 4n 16 0 18 0 200 220 240 260 Time (min) Figure 81. Mass flow rates exiting the three core channels, the core bypass, and recirculating in the upper plenum for sensitivity Case 8. 83

400 O Hot les mM O Surge line V 6 Stoom generator tubes 8 eroep ;wplure feNure 6, 3 O b. 9 M = I 1 ) ~ i i e i i i g N0 no 200 220 240 260 Time (min) Figure 82. Highest volume average pipe temperatures in the loop C hot leg, surge line, and steam generator tubes for sensitivity Case 8, i Table 16 lists some of the conditions in the RCS The sequence of events for Case 91s contained in near the time of the surge line failure. The reactor Thble 15. Several important differences from the sessel temperatures were higher in Case 8 than in base case are indicated. Ballooning occurred in the base case because the rapid core heatup had both the center and middle core channels. Core already begun. The loop structure temperatures relocation began sooner, and started in the middle were only slightly higher, or in the case of the surge channel, not in the center channel as had all the line lower, than those in the base case, indicating other sensitisity calculations. Finally, the surge line that the temperature increase in the reactor vessel failure occurred more than 12 min earlier than in 1 had not yet propagated fully into the loops. The the base case. 1 lower surge line temperature may also lndicate that The core now pattern was affected by the bal. i the PORVs had not cycled since the core heatup looning in the center and middle channels. Prior to ) rate increased, since PORY cycling is needed to the ballooning, the now through the core was from draw hotter sspor into the surse line. The heat the inlet to the outlet in ihe center and rniddle chet-3 transfer to the loops was nearly the same as in the nels, and from the outlet to the inlet in the outer ] base case. The higher hydrogen generation and fis. channel. When a locallied balloon formed after sion product release in Case 8 resulted from the 217 min in the fifth node of the center channel, i higher core temperatures, reducing the now area by 57%, the flow pattern 4 changed. Flow in the center channel below the bal. loon was ownward to the lower plenum, while l 4.5.3 Increased Core and Upper Plenum vapor Gowed upward from the lower plenum into l Croseflow Rosletance. The crossnow loss coef. the middle and outer channels. The nows in the j ficients in both the core and the upper plenum in upper part of the core were as before, with the outer Case 9 were increased by a factor of 10 oser the channel downnow penetrating to the third node values used in the base case. This change tended to from the bottom of the core. A sausage type bal. i make the flow more one-dimensional in the axial loon occurred in the middle channel at about l direction. 227 min, resulting in a 60% now area reduction 84

from nodes 4 through 8. The return now in the looning occurred in the middle channel, the outer channel penetrated as far as before, but the increased surface area coupled with the oxidation Dow magnitude was reduced. The outer channel of the cladding inner surface caused an increase in flow penetrated only to the seventh node from the ths oxidation rate in the ballooned region, bottom after the rapid heatup of the niiddle chan-accelerating the heatup in nodes 4 through 8 so nel began.That rapid heatup also caused the center that they were hotter than the top two nodes. The channel now to roerse, Howing from the upper temperatures in the upper two nodes soon fol-plenum into the core for a few minutes. Iowed, as the temperature of the vapor available to Mass now rates leaving the three core channels cool the fuel rods increased. The heatup rate and the core bypass, and recirculating within the increased further when the temperature reached I upper plenum are shown in Figure 83. The center 1850 K and the oxidation Linetics changed. At i I channel outlet flow is seen to reverse during the about 232 min, the liquid inel in the downcomer l rapid heatup of the middle channel around dropped below the core barrel, allowing cooler 233 min. The asial nows generally exhibited simi-vapor from the downcomer to enter the lower lar behavior to the other sensitivity calculations, plenum and core. This cooler vapor caused a short but with higher magnitudes. The upper plenum decrease in the temperatures in the bottom half of recirculating now, how n er, w as much low er than in the channel, but did not affect the temperatures of any of the other calculations. Figure E4 show the the top four nodes because those elevations were upper plenum recirculating now as a function of above 1850 K and oxidizing rapidly. the vapor temperature above the center core chan-The peak cladding temperature in the core dur-net. The Dow at any ghen temperature was nearly ing the transient is shown in Figure 87, together 60% lower than in the base case, with the temperature from the base case. The more The now in the outer channel was heated as it uniform asial temperatures in Case 9 resulted in a nowed downward. Because of the increased radial lower peak cladding temperature than in the base now resistance, which reduced the amount of flow case until ballooning occurred in the middle chan-I turning in to the middle channel, the temperatures net. After that time, the Case 9 temperature in the lower part of the core increased with increas-increased much more rapidly, ing radius; that is, the outer channel had the highest Figure 88 presents the fuel rod cladding surface temperatures and the center channel the lowest, temperatures at the top of the core for the three core Figure 85 illustrates this by showing the fuel rod channels. As in the other sensitivity calculations, cladding surface temperatures from node 3 in each the center channel was the hottest for the first part of the core channels. The temperature distribution of the transient. Ilowner, the middle channel bal-was such that the lower part of the core was hotter tooning and associated heatup allowed the middle than in the base case, while the upper part was channel temperature to esceed that of the center j cooler. The cfIcet of the increased cross 0ow resist-channel, leading to fuel rod relocation 3 min earlier ance was to make the temperatures more un!!orm than in the center channel. The outer channel tem-i asially. perature was lower than in the base case because Fus! cladding surface temperat ures at sneral dif-there was less mixing in the upper plenum. Cool ferent elevations in the middle core channel are Dow returning from the bottom of the hot legs to shown in Figure 86. The temperatures in the top the upper plenum was not mised with much hotter half of the shannel were close together throughout vapor, so tt.at the sapor entering the outer channel the tnansient. When the core bypass dried out at was cooler than in the base case. The lower temper-about 185 min, the core inlet now inercased, cool-ature caused the outer channel relocation to be f ing the fuel rods near the bottom of the core. The delayed, compared to the inner channel relocation entire core then heated up until ballooning and thc other crossnor semitisity calculations, occurred in the center channel at about 217 min. The total core hydrogen generation rate is show n The change in the now pattern in the t'ottom part in Figure 89. When balloomng occurred in the of the core resulted in cooler sapor enteeing the middle thannel at about 227 min, a step increase to middle channel from the lower pienum, with no 3 more ranic y inercasing hydrogen generation rate i recirculation of hotter sapor from the o> cr chan-oecarred. T ne rate then increased ar.d decreased as nel; the result was a slowly decreasing temperature jift-ent regior of the-ore went through acceler-in the bottom two nodes of the middle c mel, stM aidddini saidation at terr.peratures above a with a slower increase in nodes 3 and 4. % nen bal-1; 3 *C u ..c

40 i O Center chonnel O Widdle chonnel 9 6 Outer chonnel N X Core bypass 03 0 Upper plenum g (g^^ O -O "E O 9 % __ _ of / ' N34, au .v -r prg-ti'X M' ~ ir,__.v, , n, -,m __u_ g O2 i i i i _3, 16 0 18 0 200 220 240 260 Time (min) Figure 83. Mass now rates exiting the three core channels, the core bypass, and recirculating in the upper plenum for sensitivity Case 9. 1 60 ? N f. y 40 f'. i'i v o D}).pu..i o .gg.g. 3.,.. 20 t @ W,. o E o 02 -20 500 1000 1500 2000 2500 Vapor temperature (K) Figure 84. Upper plenum recirculating mass now rate as a function of muimum upper plenum vapor temperature for sensitisity Case 9. t l 86 1

1200 i i i O Center chonnel I O utddie chonnel 6 Ou t e r channe l m

w:

v , 1000 6 3 e l u a e ~ 800 H i. i ~ y~ 3.s - w g g i i 14 0 14 0 200 220 240 240 l Time (min) Figure 86. Fuel rod cladding surface temperatures at 0.55 m above the core bottom in the three core channels for sensitisity Case 9. l l 3000 i Height obove l core bottom O 0.ta m O t.2 8 m Q 6 2.38 m v X 3.47 m mlMle X 2000 e ehennel u beIeeniog 1 3 ] eere bypees cl oere g l 9 I Q. 1000 s=M w ~ m m m u eester l eheme l sore barral beteening eloore 0 14 0 18 0 200 220 240 240 Time (min) Figure 86. Ce.tter channel fuel rod cladding surface temperatures at 0.18,1.28,2.38, and 3.47 m above the core bottom for sensitivity Case 9. f 1 87

4* i i i O case s O case 1 7 3M T l S p l\\ 6 a f3 l 0 2000 L i e 1 WM C e i e i o 14 0 14 0 200 220 240' 260 Time (min) Figure 87. Peak cladding temperatures for sensitivity Cases 9 and I. j l 3000 O Center chonnel O utddle chonnel i 1 6 Outer chonnel m b f l 2%0 l L= i i ~ O r L l e 1 c. h 1000 l W I t l l i e i i o MO 20 200 220 240 260 i Time (min) ] Figure 98. Fuel rod cladding surface temperatures at the top of the three corc chan:.cls for sensitisity i Case 9. J ,{ 88 i l [

l L 2 i i i 6 8= 7 4 s, tzd k o' middle i 1 shermel r begeening f iI i 0 i 16 0 ISO 200 220 240 260 Time (min) i 1 Figure St. Tbtal hydrogen generation rate for sensitivity Case 9. [ l l i The structure temperatures throughout the upper between the vapor and the liquid. When the rapid ] plenum were higher in Case 9 than in the hse case, core heatup occurred at about 230 min, hotter l with the exception of the outer channel just above vapor Dowed through the core bypass, and the level the core outlet. The higher core exit flows seen in approached that of the base case, i Figure 83 transferred more of the core energy to the Vapor temperatures at the top and bottom of the upper plenum early in the transient. Iower temper. hot leg for two of the coolant loops are shown in atures were seen in the Guld volume into which the Figure 92. The rapid heatup in the core led to about i cooler hot les now returns. With a lower recirculat-a 400 K temperature increase over 2 min in the top ing now in the upper plenum, there was not as of the hot legs. The vapor temperatures in the pres. } much miting with higher temperature vapor in this surizer loop were slightly higher after the rapid volume, keeping the structure temperature cooler, increase, because the PORY cycling hr.d increased l Figure 90 shows the upper plenum structure tem-the heat transfer to the hot les piping, and not as peratures just above each of the three core much heat could be transferred from the vapor to

channels, the piping. The cooler (hot les bottom) s apor tem-Collapsed liquid levels in the reactor sessel for peratures were the same in the two loops through l

Cases 9 and I are shown h Figure 91. The leselin most of the' transient except when the PORVs were Case 9 decreased more slowly after the core dried open. When they were open, hotter vapor was out because of the lower temperature abose the drawn from the reactor vessel to the surge line outer core channel. The Guld in this volume also through the bottom of the pipe. The temperatures Howed down the core bypass to the lower plenum, also diverged bric0y during the rapid heatup in the where it transferred heat to the liquid. The heat top of the pipe. The PORVs cycled often during transfer to the liquid caused it to boil, reducing the this period, so that the hotter vapor in the piping reactor vessel liquid lesel. Since the now entering did not have time to completely clear the loops the top of the core bypass was cooler than that in between cycles, and the pressurizer loop tempera-the base case, the now exiting the bottom was also ture was higher. The temperatures were closer cooler. Less energy was transferred to the liquid together after the cycling frequency decreased because of the smaller temperature difference again. 89

2000 i O Center chonnel nw O Widdle channel V 4 Outer thennel e 6 3 1500 6 M l i e i i m 14 0 10 0 200 220 240 240 Time (min) Figure 90. Volume average temperatures of the upper plenum structures at the cutlet of the three core channels for sensitivity Case 9. O i i e i E O cose 9 O cose 1 m E 4 e e bottom of sore Es tr 2 be H om o f eer e berr el - 0 10 0 10 0 200 220 240 260 Time (min) Figure 91. Reactor vessel collapsed liquid level for sensitivity Cases 9 and 1. 90 ~

2000 i i i O Loop A top O Loop A bottom n I M A Loop C lop X Loop c bottom i e 6 3 1500 i 0 L e O. Ee 3C l ~ W% ~ l u ( L Lhl;' ' Y a g( ~ i i I t g 14 0 14 0 200 220 240 260 Time (min) i Figure 92. Ilot les nonle hot and cold vapor temperatures in Loops A and C for sensitisity Case 9. I Figure 93 shows the pressurizer loop hot leg, was caused by the rapid core heatup, w hich heated surge line, and steam generator tube maximum the structures closest to the core first. The calcula-temperatures, as well as the surge line temperature tion was terminated before that eneigy could be i from Case 1. The surge line heated up faster in redistributed to the loops, and one would expect Case 9 than in the base case, as did the hot leg. The the ratio to increase had the calculation continued ] temperature increase associated with the rapid core further. ( heatup was more pronounced in Case 9 because the Conditions of the RCS near the time of the surge l middle core channel heated up first rather than the line failure are presented in Table 16. The tempera- ~ center core channel. The middle core channet repre-tures in the system were generally lower than in the [ sents % fuel assemblio while the center channel base case, except for the middle channel cladding l repre,ent s 25. Since the t tiddle channel was heating and surge line. The more extensise total oxidation i up in Case 9, there wa s a larger amount of hot and fission product release were primarily caused [ 2 vapor entering the uppet plenum, which resulted in by the heatup of the middle channel occurring higher average vapor temperatures in the upper before that of the center channel, since the middle plenum and hot legs The surgeline failed at a tem-channel represents nearly 4 times as many fuel perature of 1252 K at 233.9 min.The heatup of the assemblies as does the center channel. 'the lower i surse line occurred rapidly because the PORVs energy deposition in the loops is misleading, in that I cycled fise times in 2.5 min during the time of the a rapid core heatup was occurring at 233.3 min. l rapid core heatup, drawing hot sapor into the rela-and the energy remosed from the core had not yet tively thin surge line often, been redistributed to the loop structures. Similarly, Figure 94 shows the fraction of the heat remused the core heat removal fraction is probably a bit low. I from the core that was transferred to the steam gen-The higher reactor sessel liquid lesel is the only l erator tubes and tube sheets in Cases 9 and I. The other major difference between Case 9 and the higher upper plenum temperatures in Case 9 at the other calculations. hot leg inlet led to increased flow in the hot legs and Resiewing the figures,it can be seen that the core more ens gy transfer to the steam generators. The and surge line heatups in Cases 9 and I were quite [ deerease in Case 9 near the rad of the calculation similar before the fuel rod cladding ballooned. The ] 91

2000 O Hot leg Case 9 mx O surge une, case t V 4 Stoom generator tubes, Case t e X surge une, case 1 6 3 I$ 0 eroep rupture g feRure f h8 = e Ceee t e @ en ./ = _ e i e i goo 14 0 10 0 300 220 240 280 Time (min) Figure 93. liighest volume aserage pipe temperatures in the loop C hot leg, surge line, and steam i generator tubes for sensitivity Care 9, and the surge line for Case 1. C.4 1 O cose s O case t l 0.4 C O t o D t.'a. t O.3 j 0.00 i 14 0 14 0 200 220 240 240 Time (min) Figure M. Fraction of the core heat remosed by the coolant for sensitivity Cases 9 and I. I 1 2 92 I

ballooning then changed the now and heatup, ballooning, the surge line failure time probably accelerating the core damage and surge line failure, would have been much closer to the base case than llad ballooning not occurred, the dif ference in tim-it was. ing of the surge line failure would probably have The changes in the reactor vessel modeling did been much less than 12.4 min. not alter the hot leg countercurrent now character-istics. The hot leg now rate was affected by the changes in the upper plenum only in that the hot leg 4.5.4 Summary. Decreasing the crossflow resist-inlet temperature was changed; the now at any ances in either the core or the upper plenum led to given inlet temperature did not change from the surge line failure less than 2 min earlier than in the base case. Fase case. Increasing the resistances in both loca-tions resulted in lower crosinows, which reduced the mising in both the core and upper plenum. The 4.6 Surge Line Fallure Calculation more one-dimensional now led to ballooning in the center and middle core channels, which accelerated the core heatup. Surge line failure occurred A final analpis was performed in which failure 12.4 min earlier than in the base case. of the surge line was modeled. The base case calcu-With a reduced crossflow resistance, the upper lation was changed at 246.3 min to include a plenum recirculating now increased by more than 0.15sn diameter break in the side of the surge line 30%. The increased mixing resulted in higher tem-near the hot leg. The calculation was then contin-peratures abose the outer core channel, which ued to esamine the effects of the blowdown on the reduced the flow returning to the core. This led to system behasior. Since the transient up to the time slightly higher core temperatures and earlier core of surge line failure is the base case, which has

damage, already been presented, the discussion that follow s With reduced cross 0ow resistances in the core, will concentrate on the behasior during the blow-more flow recirculated within the core, turning down. Similarly, the time scale on most of the fig-radially outward at the top of the core because of ures will focus on the later part of the transient.

the relatisely large asial Cow resistance at the upper The sequence of esents for the transient is pre-end botes. This now heated the sapor at the top of sented in Table 17. After the surge line failed, the the outer channel, reduced its density, and thus, RCS pressure decreased rapidly, initiating accumu-reduced the buoyant driving force. The higher latcr liquid injection. The accumulator water, outer channel temperatures and reduced core flow together with water from the loop seals.cr tered the rates accelerated the core heatup and the onset of reactor sessel and core, esentually quencahig the core damage, but only by a few minutes. entire core. The pressure continued to decrease, increased crossflow resistances in both the core and the accumulators emptied completely, At the and upper plenum led to a more one-dimensional end of the calculation, the two phase liquid leselin now. A more uniform asial temperature distribu-the reactor senel was just abcne the top of the core, j tion in the core resulted, as did a higher core flow indicating that core uncovering and heatup were rate. The higher 00* rate remosed more of the core about to begin age.m. I energy early in the transient, and allowed the fuel The pressuriier pres *ure after 245 min is shown rod cladding to balloon in both the center and mid-in Figure 95. The pressure cycled between the die channels. The localized balloon in the center PORY opening and closing setpoints of 16.2 and channel altered the flow pattern in the bottom of 15.7 MPa, respectisely, until the surge line failed, the core, changing the asial temperature distribu-A rapid pressure decrease ensued, with the pressure tion. T he sausage type balloon in the middle chan-decreasing from 16.2 ts 1.7 MPa in 1.25 min. This nel caused an increased heatup rate by increasing rapid depressurization caused the log seals to the cladding surface area, and allowing osidation clear of liquid, with the water flowing to the reactor i of the inner surface, both of which inercased the sessel. The pressure then increased as water from /irealo) osidation rate. A rapid heatup of the core the accumulators and loop seals boiled in the core. l ensued, followed closely by a rapid heating of the A pressure decrease allowed more water to be surge hne to failure. The rapid core heatup started injected from the aceumulators, resulting in more w hen the cladding temperat ure was around 1400 K, boiling and another prenure increase. Subsequent well below the tempe,".c at which the Zirealoy accumulator injection cycles caused only slight osidation Lineties enange (1850 K). Without pressure increases, with the pressure slowly 93 4

I l Tintdo 17. Sequence of events for the surge line failure omloulation 4 Time Event (min) 3 Calculation begins 160.0 Center channel osidation begins 185.J I I F Middle channel oxidatica begins 186.1 ( i i TWo phase liquid level below core 190.2 ( i t 1 Outer channel osidation begins 192.6 ) 4 Center channel fuel rod cladding falls 223.4 l l j Pressurizer empties ofliquid 224.8 l Middle channel fuel rod cladding fails 225.3 { i Outer channel fuel rod cladding falls 241.3 s' i i Pressuriter surge line fails 246.3 Accumulator injection begins 247.2 ? Core quench begins 247.4 4 d Core quench complete 250.3 I i Accumulators empty ofliquid 255.5 I t Calculation ends 266.7 l l j i I I i [ i i l l 94 i I t

20 T n. l 2 v e i u n I M E t w n. f 0-245 250 255 240 245 270 Time (min) Figure 95. Pressurlier pressure for the surge ime failure calculation, c r i decreasing until the end of the calculation. At that Figure 98, which presents the fuel rod cladding time, the pressurizer pressure was 0.91 Mpa. surface temperatures r,t several elevations in the center channel, better illustrates the cooling of the Figure % presents the liquid volume in one of e re. Tempratures om th endrdengs of the fuel the three accumulators, which was nearly identical r d decreased as soon as the surge line failed at to that of the other two. Each accumulator con-i m n. co w n then slowed, aM N top tained 29.4 m of 322 K water at the beginning of f the core began to heat up again. The tempera-the translent. Accumulator injection began at tures decreased again at 247.4 min, when water 247.2 min, and the accumulators were dry by from the accumulators began entering the bottom 255.$ min. The injection occurred in fise stages. of the core. The bottom core node quenched Liquid flow from the accumulators was interrupted (reached the saturation temperature) at 247.6 min. four times by repressuritation of the RCS. This is At 250.3 min, the top of the core was also show n by the rapid loci decreases, followed by per-quenched. Quenching of the oxidized, embrittled lods where the loel was not cht aging. The final cladding was ptedicted to shatter it,inhiating for-injection stage was a slow injection as the RCS mation of a debris bed within the core. Ilowoct, a pressure gradually decreased

• rod like geometry was assumed for the remainder The peak cladding tempcrature is shown in of the calculstion.

Figure 97. The temperature had reached about The center core channel collapsed liquid loc! is 2l00 K w hen the surge line failed. Depressurization presented in Figure 99. The loc! increased rapidly of the RCS allowed some of the liquid remaining in when the accumulators injected. The brief the lower head to Oash to steam, and increased decreases in level occurred when the RCS pressure steam now through the core cooled the core struc-was temporarily above the accumulator pressure, tures, When the accumulator and loop seat water so that no liquid was being added to the system. reached the core, a more rapid cooldown began, The collapsed lewt nearly reached the top of the followed by a quench of the core. The entire core core before decreasing at the end of the transient. wus at th; saturation temperature at the end of the The two-phase toel at the end of the calculation 1 calculation. was in the top core node. 95

1 m i i i i i C no E v i i 1 o e I I L i 4 r l o 24s tso ass neo nos 270 l l Time (min) i i i Fleure 96. L.lquid 5olume in one accumulator for the surge line failure calculation. F l I r

• D i

i t j l ] m 5, 1 4 i i tooo j i i j O l .1 6 t \\ ) 1 i-l l i o i 16 0 200 240 240 l l Time (min) r 'l i j Figure 97 Peak cladding temperatye for the surge line failure calculation, f i I l i ? f l t l l

3000 I i i i i h',9* Height above y re bellom felure g,,

== ';en o ue m Q 6 2.38 m v X 3.4 7 m 1000 r b i __ b O b c g + h 1000 a x, b. b c 0 245 244 247 248 249 250 2 51 Time (min) Figure 98. C.nter channel fuel rod cladding wiface temperatures at 0.18,1.28,2.38, and 3.47 m abose the core t.ottom for the surge line failure calculation. 4 i i i i top of core e i E P v 2 .3 Er l M eesumulater Inloetten 2 0 245 250 255 280 265 270 Time (min) Figure 99. Collapsed liquid leselin the center core channel for the surge line failure calculation. 97

Hgure 100 shows the reactor sessel collapsed lig-the secondary side of the ste&m generators was aid loel during the transient. A sharp drop in the cooler or hotter than the RCS sapor temperature, lesel oecurred w hen t he surge line failed, as some of The hot leg pipe temperatures near the reactor the water that remained ;i the vessel lower head sessel are presented in Figure IN. When the surge Cashed as the RCS depreuurized. Water from the line failed, the temperature in loop C increased loop seals and accumulators then pal,lally refilled b(cause hot vapor was being drawn from th: reae-the vnsel. As in the core, the loci decreased brie 0y tot vessel toward the surge line, and the increased w hen the accumulators w ere not injecting, arid con. velocity increased the heat transfer rate. In the tinuously after they emptied. other t wo loops, the temperature decreased because Upper plenum structure temperatures m the the now reversed in the top of the hot legs, and three solumes abose the center core channel are cooler sapor was draw n from the steam generators shown in Figure 101. The temperatures were close towani the reactor so.cl. The increased now, csm together through the heatup and initial stage of the bined with lower sapor temperatures emanating blow dow n. Ilowner, the reactor senel liquid loc! from the se<sel as the core cooled and quenched, increased sufficiently so that the structures in the then led to a more rapid cooling of the loop C hot upper plenum just abose the core querehed, while leg. Cooling of all the piping, together with the those higher L the plenum did not. Another fea. RCS depressurization, indicates that creep ruptcre ture of interest is that the struc;ure at the core outlet failure of another RCS structure subsequent to the w as cooler than the stru, t in the nest higher sol-surge line failure is unlikely for this failure site, ume, e en though the sapor temperature was Figure 105 presents the mau now through the hi her. The asceleration of the now in the upper break in the surge line. The now rate deer ased as t plenum >iclJed a higher now selocity in the second the pressure dropped, then increased as the RCS solume and hence, a higher heat transfer coeffi-repressurized The now rate then generally fol-cient, leading to the higher structure temperature. Iowed th" system pressure behavior, although there f igure 102 presents the void fraction at the bot-were increases in the flow caused by localincreases tom of the loop A loop seal. The pipe was full of in sapor density as cooler sapor entered the surge i water throughout the transient until the surge line

line, failed. The rapid depressuritation of the RCS in the base case, fuel rod relocation did not begin caued all of the loop seals to clear. The water from until after the surge line was predicted to fail.

loops A and il Gov.ed into the reactor seuel. Flow Ilecause the cooling and quench of the core began from loop C entered both the reactor senei and as soon as the surge line failed, there was no fuel the steam generator tubes; the latter oceurred rod relocation in the blowdown cal.ulation. The because now was drawn toward the surge line, calculated core damage wailimited ro osidation ol i l There was about 12.7 m'ofliquid in theloop seals the cladding, which roulted in the geneca; ion of w hen the surge hne failed, not enough to refill the 114 kg of hydrogen, and to relocation of 27r, of senel from its Inct at that time to the bottom of the the control rods in the core from the center and I core.11 252 min, w hen the pressu c had stabilized, middle channels. Ihe code predicted that the i 3 water from the accumulators refilled the loop quench of the ccre would result in the fragmenta. seali. tion of the fuel rods and the formation of a debris The hottest steam generator tube temperatures bed at about 24S min, but a rod like geometry was [ from each loop are shown in f igure 10L The tem-anumed to be maintained until the end of the I t peraturn in each deercased when the surge hne calculation. l failed and the loops seals cleared tts steam Dowed i toward the surge hne on the loop C hot leg. Some 4,7 Summary of Sensitivity and hauid entered the I cop C steam generator because of its prosimity to the break, roulting in a more Surge Line Failure Analyses rapid cooling of the tubes, although they did not reac h the saturation temperat ure. The temperaturn This section presides an oseniew of the roults i began to inercaw again w hen the loop seals tefilled of the sarious sensitisity calculations and the surge with hquid, pre enting 00w from entering the line failure cateulation. General patterns of behas-steam generators from the cold legs. The tempaa-ior are Jneribed, and the roults of different analp turn at the end of the transient w ere stoal) deercas-ses are compareJ to better understand how the ing (loops A and tu er increasing (loop C), natural circulation Dow> affect the plant snere dependmg on whether the superheated steam on accident inponse, i 98 I

I 7.5 i i top of core ) m s !? ] _ bo tt om o f co r e O' 2.5 ~ d beitom of eore barre 1 surge line / f ailur e e i 16 0 200 240 280 Time (min) Figura 100. Reactor sessel collapsed liquid level for the surge line failure calculation. i 1500 i O Core exit nx O Second volume V 6 Third volume e L 3 1000 '/ u \\ c Q. w Ee + t $00 Q 0 Le>< 0 16 0 200 240 280 Time (mln) Figure 101. Volume average structure temperatuns in the three volumes abose above the center core channel for the surge line failure calculation. i N

1 e i i l 8 l = 8 M o.5 = o> l2 i wo 200 240 280 Time (min) Figure 102. Void fraction in the bottom of the Loop A loop seal for the surge line failu:e ca:eulation. 900 i i sure* Ene O Loop A ,g f ailure O Loop B v g A Loop C eg800 q ~ a-m we O. E 700 O Q O O n n n n n .e m m m m m m m w e en E 600 e.

4 f

f., 4 A ,A A A 24b 250 255 260 245 270 Time (min) Figure 103. }{ighest volume average steam generator tube temperatures in the three coolant loops for the wrge line failure calculation. 100

i l 110 0 i i i i O Loop A ^ O Loop B 6 Loop C e L

s u

^ ^ m n n g } _a u a u a o , ea g .or e line f ailure g 900 0 Le> 4: g t i I 8 45 250 255 260 265 270 Time (min) Figure 104. Volume-average hot leg pipe temperatures near the reactor vessel for 'he surge line failure calculation. 400 i i i i 7N Osb e + 0' 200 ~ .2 + E O f 245 250 255 260 265 270 Time (min) Figure 106. Mass flow out the break for the surge line failure calculation. { 101

The hot leg now was affected orJy by the inlet overall now pattern did not. There was some reduc-vapor temperature and the steam generator inlet tion in the now penetration in the outer channel, plenum mixing. With decreased inlet plenum mix-but the recirculating flow between the core and ing, the flow increased, increasing the heat transfer upper plenum was maintained in all the calcula-in the loops and moving more of the energy to the tions. Double-sided oxidation of the ballooned steam generator tubes. In all of the sensitivity cal-cladding accelerated the heatup and allowed the culations except Cases 3 and 4, the hot leg now as a cladding to be completely oxidized at relatively low function of temperature did not change. Any temperatures. With no unoxidized Zirealoy, disso-changes in the upper plenum conditions affected lution of the fuel by molten cladding could not the hot leg flow only by altering the hot leg inlet occur, and fuel liquefaction and relocation could vapor temperature. only occur w hen the temperatures reached the melt-Heat loss through the hot leg pipe also affected ing point of uranium dioxide. This reduced the i only the loop energy distribution, and not the hot amount of fuel relocation that occurred by the time i leg now. The heat loss through the pipe reduced the of the RCS pressure boundary failure. temperature of the vapor entering the steam genera-The pressure boundary failure in all the cases tors. With lower temperatures, the steam genera-occurred very close to the time that fuel rod reloca-tors removed less energy, and t he t ube temperatures tion began. The largest differences in the timing of j were reduced. these two events were 7.4 min in Case 6 and l The upper plenum flow was affected by the local 15.2 min in Case 4; the initial fuel rod relocation modeling and by the hot leg behavior. Increasing and surge line failure occurred within 3.1 min of the upper plenum crossuow resistance reduced the each other in all the other calculations. The signifi-How recirculating within the upper plenum, while cance of that is its relation to the timing of the reac-l decreasing the resistance increased the now. tor vessellower head melt through. The results of a 4 increasing the heat transfer in the coolant loops MELPROO calculation presented in Table 5 in the i also increased the upper plenum liow by returning scoping analyses chapter showed that the melt-cooler vapor from the hot legs to the upper plenum, through occurred nearly 100 min after the initial with the same hot leg inlet temperature. The cooler fuel rod relocation. While relocation occurred at vapor had a higher density, increasing the buoyunt lower temperatures than in the SCDAP/RELAP5 driving force in the upper plenum and conse-calculations, the core damage occurred at a higher i i quently the now. decay power, so that the delay indicated by the l With a simple modelin SCDAP/RELAP5 simu-MELPROG calculation is probably not too long. lating the effects of cladding / grid spacer material Since the RCS failure occurred no more than interactions, ballooning did not occur as frequently 15.2 min later than the initial fuel rod relocation in as in past calculations. In most of the sensitivity the sensitivity calculations, it should occur at least calculations, the fuel rod cladding failed when the an hour bef ore the sessel failure. This allows a long calculated temperature reached 1470 K, a value time for the RCS to depressurize, input to simulate the effects of eutectic formation With a 0.lS-m diameter failure, this would be between the Zircaloy and iron or nicke!, rather than more than enough time to depressurize the RCS to failing because of excessne strain following bal. nearly the containment pressure. In the surge line looning. By contrast, ballooning occurred in both failure calculation, the pressure decreased from of the scoping calculations in w hich inoessel natu-16.2 to less than 1.0 MPa within 16 min. ral circulation was modeled. Liquid remained in the loop seals for all of the The ballooning that did occur affected the core sensitivity calculations, sd

• Teets ofloop natu.

damage more than it affected the core nows and ral circulation How on the p,.ucture temperatures j early heatup behavior. Only in Case 9 did the bal-were not calculated, flowever, sensiti"ity Case 4, in looning have a significant effect on the core now which there was no mixing in the steam generator i pattern and heatup rate, causing the now in the inlet plena gave an indication of what those effects bottom of the center channel to reverse while would be. With no mixing, the dow traveled from J changing the temperature distribution in the core, the reactor vessel, through the top of the hot leg Ballooning of the middle channelled to increased and the hot now steam generator tubes to the outlet oxidation of the ballooned cladding, and the mid-plenum, where the now turned back into the cold die channel healed up faster than the center :han. How steam generator tubes, proceeding to the reac-4 nel, in the other cases in which ballooning tur vessel through the bottom of the hot leg. With occurred, the Cow magnitude changed while the loop natural circulation flow, the now in all of the l l 102 i I i i

tubes would be from the inlet plenum to the outlet pressure and temperature were modeled with plenum. The Dow rate would also be much higher COMMIX, w hichwas then used to calculate a tran-than in the sensitivity calculation because the eleva-sient in a full-scale plant. The SCDAP/RELAP5 tion over which the density difference was driving model was adjusted to provide the same heat trans-the now would be much larger (from the top of the fer for the large plant calculation, then was applied steam generators to tha bottom of the core, rather to the Surry station blackout transient. The analy-than half the height of the hot leg). Even with the ses that have been performed have bounded the lower now, the steam generator tube temperatures now rate, in that Case 4 provided an upper nov/ were very close to the hot leg and surge line piping limit and that the lower now limit is no countercur-temperatures. The higher How rate in loop natural rem now, which was addressed in scoping Case 2. circulation would probably increase the tube tem. The behavior of the liquid in the loop seals is also peratures even further, so that they may fail before uncertain. Although the loop seals did not clear in Ihe other RCS structures, any of these calculations, the efrect on the transient The RCS failure time and location were fairly would be significant if they did clear. Other stud-insensitive to the parameters varied in the sensitiv-iesl8 have indicated that the loop seal clearing is a ity calcelations. The surge line was predicted to fail random process, so its potential should be consid-in all of the calculations except Case 4, where the cred w hen analyzing the plant respor,se. not leg failed shortly before the surge line. How. Effects of insulation degradation on the piping cur, even in that case, the calculated surge line temperatures were examined in sensitivity Cases 5 behavior may have resulted in lower surge line tem. and 6. Heat loss from the piping, which is an peratures than one might expect. Failure of the uncertainty in the plant modeling, was shown to surge line occurred within 13 min of the base case de'ay the surge line failure by 6-13 min, time of 246.3 min in all of the cases, again with the The size of the RCS failure is not known. The exception of Case 4, where the hot leg failed rate and extent ol the depressurization depend on 44.2 min later. It mast be remembered that Case 4 the failure size. A large enough failure, such as that was a counding calculation that maximized the hot in the blowdown calculatian, may lead to core leg sountercurrent now, so a longer delay was quenching and low reactor vessel pressures at the expected. time t he core melts through the lower head. Smaller The largest effect on the failure time was the failures may icaA to different core damage sce-decay power. Although it was net specifically part carios, with slow intermittent injection of accumu-of the sensitivity calculations, the main difference lator liquid that cools only part of the core while between scoping Case 3 and sensitivity Case I was the rest continues to heat up. a decrease in the decay power. The result was that How much now blockage actually occurs during surge line failure occurred about 50 min later in the ballooning is another area of uncertainty. With all sensitivity calculation than in the scoping of the fuel rods in a region of the core behaving calculation, identically, blockages caused by localized baaloon-ing may be too high. In the plant, while all cf the 4,8 Uncertainties and Limitations fuel r ds in a region may indeed balloon, the block-age would probably not be coplanar, as is the case with a code calculation. This section identifies some of the uncertainties The transient timing could be greatly affected by in the plant modeling and calculations. Also changes in the initial conditions. A lower steam addreued are some of the limitations of the analy-generator liquid Icvel would accelerate the time of ses. Some of these uncertainties were 'he subject of steam generator dryout, leading to an earlier core the sensitivity calculations. heatup. With the cot e heating up earlier in the tran. Uncertainties exist in the decay power during the sient, the decay power would be higher, causing a transient. The difference between the scoping and faster core heatup and possibly changing the nature sensitivity calculations showed the importance of of the core damage (less oxidation and more reloca-the power in determining the timing of the RCS tix., for example). A lower initial reactor power failure. would reduce the fission product decay heat, The hot leg counteicurrent now rate is very lengthening the transient. A lower burnup would uncertain, but was addressed in sensitisity Cases 3 reduce the actinide decay power, which would and 4. The uncertainty stems from the way the affect the core damage portion of the transient medel was desetoped. Scaled experiments at low more than the initial heatup because the actinide 103

power is a larger fraction of the total power later in dire, adding to the amount if hydrogen g:nerated. the transient. Melting tempera'.ures were also attained in some Failure of components other than those that cases, but the relocation of those structures was not define the accident sequence were not considered. calculated. Core debris fc mation caused by frag-For example, the los1 of cooling water to the reactor menting fuel rods also wa, not accounted for. The coolant pumps may lead to failure of the shaft core quench in the wc; !!ne failure calculation seals, initiating a small break loss of coolant acci-caused the embrittled fu 1 rods to shatter, but an dent, or there may be insufficient battery power or intact geometry was nai.tained for the duration of air pressure to keep the relief valves operatirig the calculation. Chen9ng this model would not throughout the transient, affect the surge line folure, since it occurred earlier, The model of the Surry plant included only the but it would affec' the subsequent reheating of the major components. Small pipes and instrument c e. 4 penetrations of the system were not modeled, so The calculations were not continued to vessel I that their behavior was not calculated. Features failure. Accordingly, while the pressure at that time such as these on the hot leg may be subject to condi-can be estimated,it has not yet been determined, so i tions such that they v>ould fall before the surge that the extent of any direct containment heating is

line, unknown. In particular, the effect on the pressure OxidrAlon and melting of structures outside of of molten core relocation into a water pool in the the core were net considered The temperatures in lower head is unknown, as are the effects of smaller

+ the upper plenum were high enough during many creep rupture failures with different accumulator of the calculations that the stainless steel would oxi-injection rates, i i l f 1 l l l \\ ? ) L ] IN 4 1 .,m,, -. - --,., - ---.--_-.---,_---_--,y

' i. CONCLUSIONS AND RECOMMENDATIONS A comprehensive analyJs on.he response of the hot leg, steam generator, and surge line struc-Surry plant to a station blackout transient has been tures were high enough that caidation, melt-performed. The analysis has provided insight into ing, or creep rupture failure may occur. With the phenomena that control the plant response and no natural circulation flows modeled, the ex-natural circulation nows, and to the level of detail vessel structures remained at or near the satu-needed to model the plant. Conclusions drawn ration temperature. The hot les from the natural circulation analyses performed for countercurrent flow transferred about 30% of the Strry plant are presented below. Based on the the energy remond from the core to the cool-results of the analyses, recommendations are also ant loops. i made for further investigation to help resolve the l phenomenological uncertainties resated to RCS 4. Hot les countercurrent flow led to the likely natural circulation during severe accidents. creep rupture failure of the pressurizer surge line about 4 hr into the transient.

1. The modeling of each additional naNral circulation flow slowed the core heatup, Creep rupture failure of the surge line was extending the transient.

predicted in all of the sensitivity calcula-tions within 13 min of the base case time The in vessel and hot leg natural circula-of 246 min, except when there was no mix-tion Hows slowed the core heatup by trans. Ing in the steam generator inlet plena ferring energy from the core to the upper (Case 4). In that calculation, the lot legs plenum and coolant loops, respectively. failed at about 290 min, shortly before the The alower heatup allows more time to surge line, although the calculated surge recover systems to mitigate core damage. line temperature may have been too low. Fuel rod relocation L gan at about Wo important points should be terr.cm-161 min in the once.through calculation, bered m regard to this conclusion. Should 7 min later with inoessel circulation mod-the loop seals clear, tacre will be no hot leg e!cd, and 18 min later with both in vessel countercurrent now, and the steam gener-and hot leg natural circulation modeled, ator tubes may fail first. Second, no pene-trations of the hot legs were modeled 2. The decay power had the greatest effect on besides the surge line (instrument taps, the event timings. small pipes, etc.). The predominant difference between scop-5. Steam generator tube failure will probably ing Case 3 and sensitivity Case I was not occur. about a 9% decrease in decay power. This delayed steam generator dryout by 10 min, The temperature of the steam generator i core heatup by 30 min, and RCS failure by tubes remained near the saturation tem-about 50 min. By comparison, the longest perature without hot leg natural circula-delay of the RCS failure in the sensitivity tion modeled. With the hot leg flow, the calculations was aoout 45 min for the tube temperatures were generally several bounding analysis in which no steam gen-hundred K lower than the surge line tem-erator intet plenum mixing occwrtd. perature. In the limiting hot leg flow case of no inlet plenum mixing, the tube tem-

3. Natural circulation nows resulted in high peratures were close to the hot leg and ex. vessel st ructure temperatures.

surge line temperatures, yet those struc-tures failed by the end of the calculations. Modeling of the multidimemional flows in while the tubes did not. Should the loop the wssel and hot leg resulted in hot wpor seals clear of liquid, the tubes may be i flowing through the upper plenum and hot heated fast enough to fail first. Without leg. We temperatures of the upper plenum, failure of the tubes, containment bypass 105 I i

through the secondary system relief valves the loop heat transfer increased, for the would not occur. sarne hot leg inlet temperature, cooler vapor was returned to the upper plenum, 6. The RCS will have time to depressurize increasing the buoyant driving force and before the reactor vessel melt through. the recirculating flow in the plenum, in seven of the nine sensitivity calcula-

10. Fuel rod cladd,ng ballooning generally i

tions, the RCS failure occurred within had little effect on the core nows. 3.1 min of the onset of fuel rod relocation; I relocation began 7.4 and 15.2 min before The ballooning that occurred in most cases the RCS failure in the other two. Since the changed the core Dow rates without chang-MELPROG calculation showed that vessel melt-through occurred nearly 100 min ing the flow pattern. However, in sensitiv-after fuel rod relocation began, the RCS ity Case 9, ballooning changed the Dow failure should occur at least 85-100 min pattern and heatup rate by accelerating the before the core leaves the reactor vessel, oxidation of the ballooned cladding in the middle channel. 7. The RCS depressurization may quench the I1.13allooning did affect the core damage. core. With a 0.15-m diameter surge line failure, Cladding that ballooned and failed oxi-the rapid injection of accumulator water dized on both the inner and outer surfaces, quenched the entire core. Smaller failures so that the cladding was completely oxi. may result in slow accumulator injections dized by the time Zircaloy would begin to that may quench only part. or none, of the melt. With no molten Zircaloy, fuel disso-core. The quench may also result in debris lution and relocation at temperatures bed formation as the oxidized cladding below the fuel melting point could not fragments. occur. Oxidation of the cladding inner sur. face increased the hydrogen generation 8. The hot leg countercurrent now was deter-rate and ac:eterated the core heatup. mined by the mlet vapor tempermure and the steam generator inlet e,enum mixing.

12. Increasing the hot les How extended the time to RCS failure.

The hot leg How as a function of the inlet vapor temperature only changed when the A 25% increase in the hot leg flow (sensi-amount of mixing in tht iniet plenum tivity Case 3) delayed the surge line failure changed. Changes in the upper plenum by 3.5 min compared to the base case. behavior affected the hot leg flow only in With a bounding $0-60% flow increase, that they altered the hot leg inlet vapor the RCS failure occurred in the pressurizer temperature. Ileat loss from the piping did loop hot leg 44.2 min later than the surge not affect the hot leg flow rate because line failure in the base case. cooler vapor then entered the steam gener-ator tubes, reducing the heat transfer in the

13. l{ cat loss from the hot leg and surge line steam generators commensurately.

piping did not significantly alter the transient. 9. The upper plenum now was affected by the hot leg behasior. With a convective be ndary condition on the pipe outer v it (Case 5), heat loss to As expected, changing the upper plenum she co ".al%ent accounted for 4% of the crossuow resistances changed the now cir-heat remosed from the core, delaying the culating in the upper plenum. The hot leg surge line failure by 6.6 min compared to flow also affected the upper plenum now the base case. Adding a radiative heat by changing the temperature of the vapor transfer coefficient as well (Case 6) returning from the steam generators. As increased the heat loss by 25% and delayed 106

the surge line failure an additional analysis. Addition of the natural circula-6.3 min. tion flews enlarged the difference in tim-ing, and added the structural failure

14. Changing the crossflow resistances had a concerns that were not addressed in the small effect on the transient.

draft NUREG-IISO calculations. These differences demonstrate the importance of Reducing the upper plenum (Case 7) or performing deterministic analyses of core (Case 8) crossuow loss coefficients by severe accidents. a factor of 10 resulted in surge line failure less than 2 min earlier than in the base

18. Detailed calculations of the effects of the case. Increasing the loss coefficients in natural citeulation Dows should be per-l both locations by a factor of 10 (Case 9) formed to the time of vessel failure.

l had a similarly small effect on the temper-atures until ballooning occurred, acceler. The effects of the natural circulation dows ating the heatup and leading to surge line and slower heatup on the core debris for-failure 12.4 min earlier than the base mation and composition need to be inves-case. tigated. The composition of the melt will affect the core / concrete interaction in con-

15. Changing the axial power profilc had little tainment, and subsequently the contain.

effect on the transient, ment loading, fission product behavior, and ultimately the source term for the in the base case, a bottom peaked, rela-accident. Jvely nat axial power profile was used. In sensitivity Case 2, a center pcaked

19. The RCS failure size should be deter-chopped cosine profile was used, resulting mined, and the resultant depressu'ization in slightly different core Cows that acceler-calculated.

ated the surge line failure by 1.3 min. The relation between natural circulation

16. The slow draining ofliquid from the pres-and direct containment heating dictates surizer will have little effect on the fission that the transient be cor.h.aed until vessel product behavior.

failure to determine the RCS pressure that will drive the melt ejection. A 0.15 m if fission products must pass through the diameter surge line break caused the RCS liquid in the pressurizer before entering the to depressurize fast enough to quench the containment through the PORY discharge entire core before f" sd relocation could piping, many of them mey be scrubbed by h* gin, indicating IMt the vessel melt-the liquid so that few are released to the throu l..vould probably occur at low pres-a containment. Although liquid remains in sure, reducing the impact of direct the pressurizer for a long time, the scrub-containment heating. A larger failure size bing benefits will be minimal because of would affect the plant similarly, while the the small release (less than 3%) of fission effects of a smaller break need to be products before all of the liquid is boiled. determined.

17. Mechanistic analyses are needed to accu-

20. The effects of natural circulation nows on rately describe the plant severe accident fission product transport and deposition

response, should be investigated.

A comparison of the once through calcu-With the natural circulation flows provid-lation (scoping Case 17 with the draft ing a path for moving released fissioi NUREG ll50 resultt.nich should have products through most of the RCS, sigt.lfi similar detail in the c,re modeling, showed cant retention of fissic,n producta may that vessel failurr.>ccurred in the draft occur. Fission products deposited on the NUREG Il50 an.syWs before fuel rod pipes may also heat the structures, leading relocation began in the SCDAP/REl AP5 to their reesolution or to stru. ural failure. 107 l i

The disposition of these deposited fission The computer codes being used to model products after the vessel fails, or after an the natural circulation flows can only sepa-ex vessel failure, needs to be determined rate and concentrate noncondensibles by since it will directly affect the fission prod-condensing the steam. Additional investi-uct inventory in the containment, and thus sation would be useful in dete..ining the source term. whether significant separation of the steam and noncondensibles may occur in

21. Experiment data from more prototypic these low velocity flows, and whether any conditions would aid in the modeling of such separation would affect the flows or l

the hot leg countercurrent flow. the heat transfer. l The experiments that have been performed to date have been at relatively low pres-

23. The fuel rod model in SCDAP/RELAP5 sures and temperatures with cooling water g

provided in the steam generator secondary. ding interactions with the grid spacers. Computer simulations of these experi-ments have been used to build models for full scale plant applications at high pres-sure and high temperature. Data at higher In the scoping calculations, ballooning temperat ares, in which heat transfer to the was being calculated under conditions in hot leg pi,dng may be significant, and with which failure of the cladding may have no cooling water flow in the steam genera-already occurred. In the sensitivity calcu-tors would be very usefulin evaluating the lations, cladding / grid spacer interactions applicability of the current modeling were simulated by imposing a temperature

approach, failure criterion on the cladding in addi-tion to the strain failure criterion usoci-

22. The effects of noncondensibles on the nat-ated with ballooning. The cladding failed ural circulation flows 4h.uld te before ballooning in nearly all of the sensi-investigated.

tivity calculations. i l l I l l I 108 l

REFERENCES 1. M. Silberberg, J. A. Mitchell, R. O. Meyer, and C. P. Ryder, Reassenment ofthe 7?chnicalBasesfor Estimating Source 7?rms, NUREG-0956, July 1986. 3. U.S. Nuclear Regulatory Commission, Reactor Risk Reference Document (Drgft for Comment), NUREG 1150, February 1987. 3. U.S. Nuclear Regulatory Commission, Uncertainty Ptipers on Severe Accident Source Terms, NUREG-1265, May 1987. 4. T. C. Cheng et al.,"RELAP5/SCDAP An Integrated Code for Severe Accident Analysis," Proceed-ings of the Thirteenth i+' ter Reactor Sqfety Research htformation Afeeling, Gaithersburg, AfD, a October 22 25,1985, NUREG/CP-0072, pp. 347-355. 5. W. A. Stewart, A. T. Pieczynski, and V. Srinivas, "Experiments on Natural Circulation Flows in Steam Generators During Severe Accidents," Proceedings of the International ANS/ ENS Tbpical Afeeling on ThermalReactor Sqfety, San Diego, CA, February 2-6,1986. 6. H. M. Domanus, R. C. Schmitt, W. T. Sha, and V. L. Shah, Analysis o/ Natural Convection Phe-nomenon During Postulated Station Blackout 1hmsient Accident, ATH RP-30, January 1987. 7. M. llansen and K. Anderko, Constitution of Binary Alloys, Second Edition, Schenectady, NY: Genium Publishing Corporation, August 1986, pp. 742,1060. 8. S. Hagen and P. I10ffmann, Physicaland ChemicalBehavior ofLi4'R FuelElements up to le'ry High Temperatures, KfK4104, June 1987. 9. S. flagen, L. Sepold, P. Hoffmann, and G. Schanz,"Out of Pile Experimenu en Severe Fuel Dam-age D. havior of LWR Fuel Elements (CORA Program)," lA EA Internationa/ Symposium on Seven Accidents in Nuclear iMwer Plants, Sorrento, Italy, Afaith 21-23,1988. 10. B. L. Ilartis, V. N. Shah, and C. E. Korth, Creep Rupture of Three Components of the Reactor Primary Coolant System During the ThfLB' Accident, EGG EA 7431, Noveraber i986. I1. H. M. Domanus and W. T. Sha, Analysis ofNatural-Convection Phenomenain a 3-Loop Pll'R dur-ing a TAfLB' 7hmsient Using the COhlhilX Code, NUREG/CR 5010, ANL 87 54, January 1988. I2. P. D. Bsyless, C. A. Dobbe, and R. Chumbers, Feedwater Thmsient and SmallBreak Loss of Cool-ant Actident Analysesfor the Bellefonte Nuclear Plant, NUREG/CR-4741 EGO 247l, March 1987. 13. R. S. C enning et al., Radionuclide Release CalculationsforSelectedStvere Accident Scenarios (Pl4'R, Subatmmpheric Containment Design), NUREG/CR 4624, BMI.2139, Vol. 3, July 1986. 14. J. E. Kehy, R. J. llenninger, and J. F

Dearing,

AfELPROG Pil'R/AIODI Analysis of a TAILB' Accident Sequence, NUREG/CR-4742, SAND 86-2175, January 1987, 15. G. B. Wallis, One Dimensional 7ko-Phase Flow, New York: McGraw liill,1%9. 16. W. M. Rohsenow and H. Chol, Heat, Afan, and Atomentum 7hrntfer, Englewood Cliffs, NJ: Prentice liall, Inc.,1%1, pg.102. 109 1

__ ~ _ r. 17. R. Siegel and J. R. Howell, 77:ermal Radiation Heat 7?ansfer, Second Edition, Washington: Hemisphere Publishing Corporation,1981, pg. 834, 18. N. Lee, "Discussion on Imp Seal Behavior During Cold Leg Small Break LOCAs of a PWR," Nuclear Engineering and Design 99, pp. 453-458,1987. I k 1 i h i l s a j l l \\ i l l l l I I10 l I l 4 i L

l l APPENDlX A PLANT AND MODEl. DESCRIPTIONS A1 \\

APPENDlX A PLANT AND MODEL DESCRIPTIONS M. ' - The SCr1AP/RELAP5 model of the plant upper head. The core bypass represented the il

k...

included all of the major co nponents necessary to volume between the core barrel and the core bafile, [,'}' perform the station blackout analyses. The reactor and connected the lower plenum and the upper 6:;( vessel, three coolant loops, three steam generators, plenum. Flow paths were modeled between the inlet ' q' and the pressurizer were modeled. Infore wtion was annulus and the upper head and between the inlet obtained from References A-1 and A-2. annulus and the upper plenum (leakage around the g Each of the three coolant loops was modeled, as hot leg nonles). Flow between the upper plent.m were the associated steam generators. The loop and upper head occurred only through the guide model included both the Guid volume and the tube assemblies. Heat structures modeled include metal structui 3. Figure A-1 shows the nodaliia. the reactor vessel walls, the core t cel and bafne, tion of one of the coolant loops. The piping was the thermal shield, the upper and lower core sup-6 assumed to be adiabatic on the outer surface. The port plates, and structures in the lower and upper pressuriier and surge line were attached to one of plena. The three-channel model nodalization is the coolant loops (loop C). The surge line piping shown in Figure A 3. The core and upper plenum and pressuriier shell were modeled as heat struc-volumes were divided into three radial regions. The { tures, again with adiabatic outer surfaces. The core regions were selected so that similarly powetcJ pressurizer heaters were not modeled, as power was fuel assemblies were grouped togaher. The upper not asailable to them during the transient. The two plenum regions were exansions of the core region pressurizer power operated relief vahes (PORVs) boundaries. The three channels were connected at were modeled as a single sabe connected to the top each elevation (escept the top of the upper plenum) of the pressurizer. The vahe was sized to proside a by crossflow junctions, so that a nearly two-saturated steam flow rate of 45.1 kg/s at dimensional model of the core and upper plenum 16.2 N1Pa, which is twice tht rated capacity of one was asailade for the inacssel natural circulation PORV. Similarly, the three safety relief salves calculations. The deep beam weldments were (SRVs) on the pressuriier were modeled as a single assumed to proside barriers to radia10ow in the top saln. It was assumed that there was sufficient bat. of the upper plenum, so tim the uppermost sol-tery power and plant air to operate the PORVs as umes in the urner plenum were not connected to long as the transient continued. Piping down-ech other.1igurt A-4 shows a cross section of the "; n of the PORV3 and SRVs was not modeled. three channel core model. The number of fuel accumulators were the only asailable part of assemblies in each region, along with their relative the emergency core cooling system for these tran-powers, is also prosided. sients. Each of the three accumulators injected into The 3.66 m active fuel length was divided into a different cold leg and contained 29100 kg of ten equal asial nodes, numbered I to 10 from bot-322 K borated water with a nitrogen coser gas pres-tom to top, corresponding to the nodalization of sure of 4.24 hlPa. the Guid volumes in the core. SCDAP structures The steam generator model included the tubes, were used to model the fuel rods, control rods, and downeomer, riser, separator, steam line, main and instrument tubes and empty control rod guide ausiliary feedwater sptems, main steam isolation tuba in each channel in the core. Thus, threc sahe, PORVs and SRVs. The metal maues anoei. SCDAP components were used in the once through ated with the steam generator walls and internals core model, and nine SCDAP components were were modeled, with the outer surface of the steam used in the three channel model. Figure A 5 shows generators assumed to be adiabatie. a ty pical fuel assembly. Grid spacers were assumed Two reactor sessel models were deseloped for the to be located at the bottom of nodes 1,3,5,6,7, analpes. The single-channel model, w hose nodali. and 9. There are 48 fulllength and 5 partiallength ration is shown in ligure A 2, was used for the control anemblies in the Surry core: 8 full and once through calculation. The Guid solumes mod-1 partial length in the center channel,32 full and cled include the downcomer, lower head, lower 4 partial length in the middle channel, and 8 full plenum, core, core b> pan, upper plenum, and and 0 partial length in the outer channel. The A3

444 , O.F. k 482 r) 484 449 450 487

• h485.

f_- s l [$lAux. F / 489 445 '.< 1 -*.d 4.;gyn/ th usi safe safelles l, lies p ~, 476 8 t S t / s. A A ,f ) so e o pk / K I [440 ~ *" dp d, "[0 9 j,........................ .../ ;39'D 2 j / i g3 6 3 l Stoom 3 ! -c f 441 Generator 474 4 s 4 / Pressur-0 3 2 7 2 Izer 8 p 5 / $ 4 1 8 1 d I_ d 6 7 l o I 475 41 0 406 s/s/s wsn [ 443 1 3!2 l d p ///ss/////n b I 400 404 402 2 l1 e ////. ////////n> m rn//-,, 777) 7777 7777 7777 7 422 -+ 1 414 416 41 8 420 -+ / Accumulofor Reactor 2 Ve s s el 540 412 542 HPl 3 5 4 i 544 LPI l l t 1 P425 5T-0245-01 Figure A 1. Nodalitation of the pressuriier coolant loop for the Surry SCDAP/RELAPS calculations. A-4

i /////////////// P[ 190 o 1P 175' 100 a ypp,7 plenum 1 170 a n To l from cold 7M 102[(d { 160 [/jj._*. hof A legs legs ~ n 3 u VM 1 M l 150 K/) m a a o 10 2 5 .......9........ 8 ) 3 4 7 i Oowncomer Byposs ,,,,,,,,6 104 4 1 3 d

• • • • • • • ~

O 5 2 ~~-- ---- - .......3........ i 2 t 6 1 1 a n u3 l 108 K/2 pfe*num Lower 106 head ll$lllllllll i ) ) P428 ST-0245-03 Figura A 2. Nodalliation of the single channel reactor vessel for the once through Surry 1 SCDAl'/REl.Al'5 calculation. 1 i 4 J i l A5 i 1 i - ~ -, - ,,,n

//////////////// l 19 0 i 1 1 1 3 3 3 2 l 174 l 1 16 4 l 1 15 4 1

8

8 i

i l 8 i 10 0 173 l-!'2-.163 l-!^1 1 3 l 15 3 -+ L_ L_ L_ n_ n n n 322,!22+l102l 3 17 2 l 162 l 152 l 200,300 : y 400 18 l 171 l = 161 l 151 l = ] at t a I I j i 39-- g 39. g 19--- 2 5 .. 9... _1.3_ _q 12 8, .. 9... g_.. 9... g 3 4 ...j... g...j... g 335 12 5 13 7 --@-- W 104 4 3 .5 T37.. 5... g.. 5.. 4 T32 - $...-- M -$--. 5 2 jyf' 121 -g-- -k-- Q -k--M y-- 6 1 i " 11 8 6 113 T 11 2 " 111 f78 h f 108 u a 106 ////////////////// ~ P394-LN87017-1 Figure A 3 Nodalization of the reactor vessel for the Surry SCDAP/RELAP5 calculations with in vessel natural circulation. ] l i A.6 P

///// '$1 z t, '-l ~ ': Core barrel l !.-l Core bypass ,3 l--h Core baffle 7--; i i i H Outer channel 36 fuel assemblies / l j i.__. .__.i l relative power = 0.76 l d = l i j {~' Yjp Middle channel -l i 96 fuel assemblies -i rH relative power = 1.05 3 r ~;

J Center channel Lj-i L_l 25 fuel assemblies L,,,,

relative power = 1.17 ' ///// P394-LN87017-13 Figure A4. Cross section of the three-channel core region.

O O O O O O O Control rod O O = guide tube l O O -~ ,O O g instrument tube O O _ Fuel rod ' ' t' " O O O O O P431-LN87031-1 Figure A-5. Typical arrangement of a Surry fuel assembly.

_ _.. ~ _. _ _. control material is Ag In-Cd. The fuel rods were steam generator plena to provide the desired assumed to be prepressurized to 2.4 htPa. The amount of heat transfer to the steam generator i same axial power profile was used in each of the tubes and hot leg piping for a given hot leg inlet core channels. The scoping calculations and sensi-temperature, based on the COhlhtlX calculation. tivity Case 2 used an axisymmetric chopped cosine TWo sets of tubes co:mected the inlet and outlet 4 axial power profile with a peak to average power plena. The hot now tubes represented 35% of the ratio of 1.200 located in nodes 5 and 6. The rest of now and heat transfer areat the cold return flow the calculatfore used a flatter axial power profile tubes represented 65% of the total now and heat 3 i having a peak to average power ratio of 1.15.( transfer area. This division of the steam generator located in node 3. The core was assumed to be at tubes was based on results of the Westinghouse nat-the end of an equilibrium cycle, which corre-ural circulation experiments.A-4 The reactor cool. sponded to an average burnup of ant pump suction piping was still connected to the j 21,000 htWd/t.A 1 Both radial and axial conduc-steam generator outlet plenum, so that clearing of i tion were accounted for in the cladding heat trans-theliquid from the loop seals could occut if appro-i fer calculations. A control system was used to priate conditions existed. Since the pressurizer approximate the behavior of molten material that surge line connected to the side of the hot leg, the l relocated below the core. A heat structure was horizontal portion of the surge line (the two vol-located in the lower plenum that received the energy umes at the hot leg end) was also split into top and of the relocated material. The structure then dissi-bottom halves, connected to the top and bottom pated this energy, allowing water in the lower halves of the hot leg, respectively, in the sensitivit.: l plenum to be boiled. calculations. For the scoping calculations, the pres i in order to simulate the hot leg countercurrent surizer surge line was not split. Because of diffictl. flow, several modeling changes were made. ties with reverse now through the pressurizer lo)p 1 l Figure A-6 shows the nodalization of one loop for (encountered during testing of the model), a va ve t the hot leg countercurrent now calculations. All was inserted at the connection of the bottom of t te [ ] three loops were modeled this way, except that the hot leg to the reactor vesset for the scoping calcula-j surge line was connected only to loop C. The hot tions. This valve was closec. when the PORVs were i leg was divided into top and bottom halves. The open, preventing vapor in the vessel from entering i return now from the bottom half of the hot leg the bottom of the hot 14 until the PORVs closed. ) entered the reactor vessel at the top of the volume This valve was nst needed with the split surge line t below the hot leg nocle connection. This approxi-model. Creep damage calculations were performed j mated the eLivation difference between the top and for each hot leg near the reactor verci, for the hot- ) bottom h,:hes of the hot leg, so that the drising test part of the steam generator tubes in each loop, i 2 head for the natural circulation flow should be and for the hot leg end of the surge line, i close to that of the actual situation. It also pre-This hot leg model evolved from low pressure, J vented the hot and cold steam Dows from mixing law temperature experiments performed by l immediately in Ihe upper plenum. The steam gener. Westinghouse using Sly as the working fluid.A i ator inlet plenum was divided into three volumes. A These experiments were then simulated using the mixing volume in the middle connected to the flows COhlh11X computer code. COhlhilX was then [ entering and leasing both the hot leg and the steam used to simulate a high pressure, high temperature ] generator tubes. The volumes on either side of the transient in a commercial plant. The plant model mixing volume passed the hot and cold vapor that included the core, upper pienum, hot legs, steam i did not mix with the other Guld in the inlet plenum generators, surge line, and PORV. Pipes and steam l between the hot leg and steam generator tubes. The generator tubes were modeled with adiabatic outer 1 amount of flow entering and leaving the mixing boundary conditions. The transient was a 2000 s volume was adjusted by changing the loss coeffi. heatup from saturated vapor conditions through-cients and flow areas. Guidance in adjusting the out the model. This transient was then calculated j nows was obtained from calculations of a similar using SCDAP/RELAP5. The mixing volume, j transient performed at Argonne National Labora-junction areas in the inlet plena, and lost coeffi-I tory using the three-dimensional CONINilX com-cients were adjusted to reproduce the COhlh11X. I puter code.A'3 The unmixed Guld now areas were calculated integral heat transfer in the loops as a reduced to 5% of the now area leaing the hot leg function of the hot leg inlet vapor temperature. The and steam generator tubes. Large loss coefficients model developed, which did not have a split surge were added to the junctions in the hot legs and line, was within 7% of the COhthilX valuci. The i A9 I a

4 5 408 5 4 3 6 3 6 l 2 7 2 7 Reactor 1 8 1 8 6 y vessel AA T T f f 405 406 407 410 455 g 400 T 1 l 2 + 402 + 404 412 172 -+ '//HHHHf HHH/HHJ f/H/H/HH a 430 l 3 l I 2 l 4 (HHHHH/HHH/HHH//HH///HHHH//H/Hf 171 P431-LN87031-2 Figure A4. Nodalization of the hot leg and steam generator for the Surry SCDAP/RELAP5 calculations with hot leg natural circulation.

i hot leg flow rates were lower than those calculated for the end of an equilibrium cycle by COhihilX, but that was not a large concern (21000 h1Wd/h1T) was assumed. To achieve this l because the heat transfer to the loops was the most burnup, the fuel assemblies in each channel were important parameter, and the one that was being assumed to have been in that region of the core dur-l matched. The flow characteristics were also some-ing a 22.5 month period of full power reactor oper. what different than in the COh1N11X calculation ation. The prompt power was assumed to decrease and the Westinghouse experiments, in that the linearly from full power to zero in the first 2 s. flows in the tops of the non pressurizer loop hot Table A-3 lists the decay power for these legs did not reverse when the PORVs were open, calculations. although their magnitudes were reduced. This hot For the sensitivity calculations, ORIGEN2 cal-leg model was ther used for the scoping calcula-culations were performed to generate input power tions, and the split surge line was added fcr the tables for SCDAP/RELAP5. It was assumed that sensitivity calculations, fuelis loaded such that fresh fuelis added to the l For the calculation in which inlet plenum mixing core periphery, from w here it moves radially inward was not considered, the mixing volume was on subsequent refuelings until it is eventually removed. The two through flow solumes were removed from the center of the core. It was increased in size, so that each modeled one-half of assumed that 52 assemblies were added at the the inlet plenum. The steam generator tubes were begiring of the current cycle,52 assemblies were not changed from the model discussed above. in their second cycle, and 53 assemblies were in Full power steady state calculations were per-their third and final cycle. Accordingly, the center j formed for both the scoping and sensitivity analy-channel contained 25 assemblies with three cycle I ses, since different sersions of the code were used, burnup, the middle channel contained 28 assem-Results of the steady state calculations are pre. blies with three-cycle burnup,52 assemblies with sented in Table A 1, together with the desired val-two-cyc!c burnup, and 16 assemblies with one-ues for sescral key parameters. There was good cycle burnup, and the outer channel contained agreement between the calculated and desired 36 assemblies with one-cycle burnup. ORIGEN2 steady state conditions, calculations were performed to determine the fis-Table A 2 presents the initial imentory of five sion product and actinide decay after the reactor significant finion product elements for the various scram. Calculations were performed for the irradi-calculations. It can be seen that there is a signifi-ation of ammblies in each of the three core chan-l cant difference in initialimentory for four of the nels. The last irradiation cycle was assumed to be at elements for the sensitivity analyses, for which cal-the relative radial power described earlier (see Fig-culations were performed using both SCDAP/ ute A4). The two and three-cycle burnup asrem-RELAp5 and ORIGEN2 5, with the ORIGEN2 blies were assumed to have been irradiated at A values b?irg 840% lower. SCDAP/RELAP5 uses aserage core power prior to the current cycle. The A l PARAGRASS 6 to calculate t! c fission product resulting fission product decay power values were initial inventory and subsequent release. The values multiplied by a factor to account for neutron calculated by ORIGEN2 are believed to be more absorption by the fission products [G(t) in ANS accurate because of the more extensise physics and 5.l], which had a very small effect on the power cross scetion data used, af d became PAR AGRASS untJ after 10000 s. The prompt power decay was was primarily deseloped for trace irradiated fuel dese:oped from a RELAP5 kinetics calculation, it rather than the high.burnup fuel associated with was assumed that the prompt power began to decay the end of an equilibrium fuel cycle. ORIGEN2 when the control rods started to enter the core, at calculations were not performed for the scoping 0.7 s after the loss of power. The decay of the fis-i

studies, sion proJuets and actinides was assumed to begin The power used in SCDAP/RELAP5 is disided at 1.0 s after the start of the tramient. Table A4 l

into threc components: prompt (fission) power. fis-lists the decay power input for the sensitisity sion product decay power, and actinide decay calculations. power. Different decay power salues were used in Table A 5 prosides some details on the sizes of the scoping and semitisity calculations. the various models used and the computer time for the scoping calculations, the decay heat was needed to perform the natural circulation calcula-calculated internally by the code. This calculation tions. The scoping calculations were performed on l uses the ANS 5.l^4 standard and assumes only a CRAY IS computer, w hile the semitivity calcula-I one finile isotope (UW). A core aserage burnup tions were performed on a CRAY X-N1P/24 i l 2 A Il

= Table A 1. Comparison of computed and desired steady state parameters Parameter Scoping Study Sensitivity Study Desired Core thermal power, MW 2442 2443-2441 1 Pressurizer pressure, MPa 15.5 15.5 15.5 i Pressurizer liquid level, m 6.05 6.62 6.62 liot leg temperature, K $92.0 591.7 591.9 Cold leg temperature, K $57.3 $$7.0 557.0 l l Coolant flow per loop, kg/s 4230 4230 4230 Steam generator pressure, MPa 5.49 5.71 5.41 l t J Steam generator liquid mass, 44000 44000 44000 l kg(each) Steam generator feedwater 443.9 444.6 441.8 flow, kg/s 4 i i I Table A 2. Initialinventory for five fission product elements f I Finion Product Mass i (kg) C Scopint alculations Sensitivity Calculations [ l I Element One channel Three-channel SCDAP/RELAP5 ORIGEN2 j { f Xenon 257.4 253.9 258.3 '238.4 i I i Krypton 28.99 29.16 B 10 17.44 i Cesium 186.1 187.2 116.7 125.5 i l lodine 10.39 10.45 10.42 10.79 t i Thllurium 23.91 24.05 23.98 21.12 l f i i t I I A 12 f i i I

Table A 3. Decay power for the scoping calculations Center Channel Middle Channel (MW) (MW) Time Fission Fission (s) Prompt Product Actinide Prompt Product Actinide 0.0 419.8 31.00 1.465 1446.8 106.8 5.050 2.0 0.0 27.03 1.465 0.0 93.12 5.045 i i 10.0 0.0 21.90 1.461 0.0 75.43 5.035 I 100.0 0.0 14.18 1.428 0.0 48.86 4.919 1000.0 0.0 8.574 1.163 0.0 29.53 4.007 5000.0 0.0 5.238 0.747 0.0 18.04 2.574 7600.0 0.0 4.572 0.693 0.0 15.74 2.389 8000.0 0.0 4.496 0.689 0.0 15.48 2.374 8500.0 0.0 4.408 0.684 0.0 15.I8 2.359 9000.0 0.0 4.328 0.681 0.0 14.90 2.346 f 9500.0 0.0 4.252 0.678 0.0 14.64 2.335 I 10000.0 0.0 4.182 0.675 0.0 14.40 2.325 l 10500.0 0.0 4.338 0.672 0.0 14.95 2.317 l1000.0 0.0 4.275 0.670 0.0 14.73 2.310 11500.0 0.0 4.216 0.668 0.0 14.53 2.303 12000,0 0.0 4.160 0.667 0.0 14.34 2.297 i 12500.0 0.0 4.109 0.665 0.0 14.16 2.292 1 13000.0 0.0 4.060 0.664 0.0 13.99 2.287 14000.0 0.0 3.968 0.661 0.0 13.68 2.277 20000.0 0.0 3.559 0.647 0.0 12.26 2.229 A.13

Table A 3. (continued) Outer Channel Single Channel (MW) (MW) Time Fission Fission (s) Prompt Product Actinide Prompt Product Actinide 0.0 392.7 28.99 1.371 2267.8 166.4 7.866 2.0 0.0 25.26 1.370 0.0 145.0 7.862 10.0 0.0 20.44 1.367 0.0 117.5 7.845 100.0 0.0 13.25 1.336 0.0 76.I2 7.666 1000.0 0.0 8.008 1.088 0.0 46.00 6.244 i j 5000.0 0.0 4.887 0.699 0.0 28.09 4.010 7600.0 0.0 4.264 0.648 0.0 24.51 3.721 3000.0 0.0 4.193 0.644 0.0 24.11 3.699 8500.0 0.0 4.110 0.640 0.0 23.63 3.674 1 9000,0 0.0 4.034 0.637 0.0 23.20 3.654 9500.0 0.0 3.963 0.634 0.0 22.80 3.637 10000.0 0.0 3.898 0.631 0.0 22.42 3.623 10500.0 0.0 4.057 0.629 0.0 23 29 3.610 11000.0 0.0 3.998 0.627 0.0 22.75 3.598 11500.0 0.0 3.943 0.625 0.0 22.63 3.589 12000.0 0.0 3.892 0.624 0.0 22.34 3.579 12500.0 0.0 3.844 0.622 0.0 22.05 3.570 13000.0 0.0 3.796 0.621 0.0 21.79 3.563 l l 14000.0 0.0 3.711 0.6I8 0.0 21.30 3.548 ] 20000.0 0.0 3.329 0.605 0.0 19.11 3.472 i 1 l e i l A 14

l t l Table A4. Decay power for the sensitivity calculations l i i Center Channel Middle Channel j (MW) (MW) I Time Fission Fission [ (s) Prompt Product Actinide Prompt Product Actinide 0.0 427.5 25.85 1.412 1472.8 90.13 4.324 l I 0.7 427.5 25.85 1.412 1472.8 90.13 4.324 j t 1.0 382.8 25.85 1.412 1318.7 90.13 4.324 I 1.5 324.3 24.68 1.412 1117.2 85.% 4.323 2.0 276.3 23.92 1.412 951.9 83.28 4.323 f l 3.0 61.29 22.86 1.411 211.1 79.58 4.321 6.0 8.887 21.03 1.410 30.62 73.15 4.318 11.0 5.550 19.36 !.409 19.12 67.33 4.313 I6.0 4.I15 18.30 1.407 14.I8 53.64 4.308 j 21.0 3.29I i7.52 1.405 11.34 60.92 4.302 31.0 2.307 16.41 1.402 7.948 57.05 4.290 $ 1.0 1.299 14.98 1.395 4.476 52.!! 4.269 j 101.0 0.397 13.06 1.378 1.368 45.45 4.216 l 201.0 0.068 11.29 1.345 0.234 39.32 4.115 a 501.0 0.001 9.325 1.255 0.005 32.46 3.838 t 1001.0 0.0 7.918 1.132 0.0 27.58 3.458 1 j 2501.0 0.0 6.030 0.900 0.0 21.02 2.741 l 1 5001.0 0.0 4.755 0.744 0.0 16.55 2.263 1 e i 10001.0 0.0 3.760 0.675 0.0 12.99 2.052 20001.0 0.0 3.215 0.646 0.0 11.11 1.%5 i l t ,I i J 1 I l 4 i i l J A 15 l 1 i

i I Tatde A4. (continued) Outer Channel i l (MW) f Time Fission i (s) Prompt Product Actinide i a i 3 j 0.0 399.1 25.24 1.013 l j 0.7 399.1 25.24 1.013 l.0 357.4 25.24 1.013 1.5 302.8 24.03 1.013 1 1 2.0 258.0 23.26 1.01J i t j 3.0 57.22 22.20 1.013 j 6.0 8.297 20.37 1.012 { I 11.0 $182 18.7! 1.011 r 1 16.0 3.842 17.67 1.009 l 21.0 3.072 16.91 't008 I i ) 31.0 2.154 15.83 1.005 i l 51.0 1.213 14.45 1.000 t a t j 101.0 0.371 12.61 0.988 [ 201.0 0.061 10.91 0.964 j { 501.0 0.001 9.000 0.898 i 1 l 1001.0 0.0 7.654 0.807 6 2501.0 0.0 5.839 0.637 r i 5001.0 0.0 4.583 0.523 [ ) IPJ01.0 0.0 3.567 0.474 l 20001.0 0.0 3.027 0.454 l 1 L l l 2 } g h A.16 i i 1 i ., - -, -,, -,. ~ - -,,,,..,

L b l i i i Table A 5. Computer calculation statistics Nu.nber of l Numb's E RELAPS Heat Number of Transient CPU i Volage/ Structures / SCDAP Time Time Calculation Jdre.ons Mesh Points Componcats (s) (s) l Scoping l Analyses l ] Case 1 168/177 195/909 3 12000 7269 Case 2 198/235 211/955 9 4400 11676 l 4 e Case 3 240/289 259/1201 9 3200 8008 l i Sensithity f Analyses f 4 t j Case 1 247/297 262/1210 9 5400 7152 Case 2 247/297 262/1210 9 $289 7094 Case 3 247/297 262/1210 9 6400 10010 l Case 4 244/285 259/1192 9 8400 13357 J Case 5 248/297 262/1210 9 6174 8998 i Case 6 248/297 262/1210 9 6400 9527 Case 7 247/297 262/1210 9 $400 7290 Case 8 247/297 262/1210 9 $400 7884 i I Case 9 247/297 262/1210 9 $176 7802 4 0 } Surge line 247/297 262/1210 9 6400 19540 l l failure j i i l t A 17

REFERENCES A l. Virginla Power Company, Surry 1%wer Station Updated (1m Analysis Report, Docket 0$000280,0$000281, July 16,1982. A 2. J. D. Burtt,.tudit Calculations for a Main Steam Line Break in North Anna Unit 2 Using the RELAPJ Computer Code, EGO NTAP-6082, November 1982. A 3. H. M. Domanns and W. T. Sha,.1nalysis of Natural. Convection Phenomena in a 3 inop PWR during a TMLB' TVansient Using the COMMIX Code, NUREO/CR 5010, ANL 8154, January 1988. A-4. W. A. Stewart, A. T. Pieczynski, and V. Srinivas, Experiments on Natural Circulation Flows in Steam Generators During Severe Accidents," Proceedings of the International ANS/ ENS 7bpical Meeting on Thermal Reactor Sqfety, San Diego, CA, February 2-6,1986. A 5. A. O. Ctoff, A Users ManualforIhe ORIGEN2 Computer Code ORNtrTM 117$,3uly I985. A-6. Atgonne National Laboratory, Light Hkter Reactor Sqfety Research Program: Quarterly Progress Report, July-Septernber 1981, NUREO/CR 2437 Vol. Ill, ANir8177 Vol. Ill, February 1982. A 7. American NationalStandardfor Decay fleet twerin Light Hkter Reactors, ANSI /ANS $.I l979. l 4 l A-18

APPENDlX B COMPUTER CODE DESCRIPTION l ) B1

APPENDIX B COMPUTER CODE DESCRIPTION The computer code used in the transient analyses SCDAP components simulate core disruption by I wasSCDAP/RELAP5.d l A brief general descrip-modeling heatup, geometry changes, and mater.'ai tion of the code, together with information on the relocation. Detailed modeling of cylindrical and specific versions used, is presented below. slab heat structure geometries is allowed. Thus, The SCDAP/RELAP5 computer code is a light fuel rods, control rods and blades, instrument water reactor (LWR) system transient analysis code tubes, and now shrouds can be represented. All that is currently being dneloped. It can be used for structures of the same type, geometry, and power in simulation of a wide variety of system transients of a coolant channel are grouped together and one se: interest in LWh safety but is designed especially to ofinput parameters is used for each of these group-calculate the behavior of 'he reactor coolant system ings or components. Code input identifies the during snere accidents. The core, primary system, number of rods or tubes in each component and secondary system, feedwater train, and system con-their relative positions for the purpose of radiation troh can be simulated. The code models hase been heat transfer calculations. hiodels in SCDAP cal-designed to permit simulation of postulated acci-culate fuel and cladding temperatures, Zircaloy and dents ranging from small break loss of coolant acci-stainless steel oxidation, hydrogen generation, dents to snere accidents. Transient conditions can cladding ballooning and rupture, fuel and cladding be modeled up to the poini of sessel failure.

iquefaction, flow and freeting of the liquefied SCDAP/RELAP5 was produced by incorporat-materials, and release of fission products.

ing models from the SCDApil 2 and TR AP. Fragmentation of fuel rods during reflood is calcu-h1ELT 3.4 codes into the RELAP5/htOD2 5 lated. Oxidation of the inside surface of the fuel Il B code. The SCDAP components model the struc-tod cladding is calculated for ballooned and rup-tures in the reactor core. The TR AP-NtELT models tured cladding. were used as a basis for the fiaion product trans-The fission product behasior includes aerosol port and deposition modds. RELAP5 models the agglomeration, aerosol deposi,on, evaporation Guid behas ior throughout the system, as well as the and condensatioa, and chemisorption of sapors by thermal behasior of structures outside the core. stainless steel. qion products are anumed to be The feedbacks between the sarious parts of the released equally oser the entire length of the fuel code were doeloped to preside an integral analysis rodi. The released fission products enter the cool-capability. For example, the changes in coolant ant as aerosols, being put into the smallest size bin now area anociated with fuel cladding ballooning and allowed to agglomerate or naporate as condi-or relocation are taken into consideration in the tion: Gta;e. The number of aerosol site bins used, hydrody namics. as well as the fluion product species tracked, is SCDAP/RELAPS uses a one-dimemional, two selected by the user. The chemical form of the fis-Guid, nonequilibrium, sis equation hydrodynainie sion products is fiteo. All of the iodine is assumed model with a simplified capability to treat multidi-to be in the form of Csi, with the remaining cesium mensional Dows. This model prosides continuity, being transported as Csoll. Fission products do momentum, and energy equations for both the lig-not interact with the surfaces of SC0AP compo-uid and the sapor phases within a control solume, nents (fuel rods, control rods, control blades, and The energy equation contains source terms which shrouds). couple the hydrodynamic model to the heat struc-The sersions of the code used for the scoping ture conduction model by a comeetisc heat transfer calculations were SCDAP/RELAP5/htOOO formulation. The code contains special process Cycles 48 and $1, with updates. The updates models for critical flow, abrupt area changes, included error corrections that hase been added to branching, crounow junctions, pumps, accumula-subsequent sersions of the code, using steam prop-tors, vahes, core neutronies, and control systems. erties in a control solume w hen the noncondensible A Gooding model can also be applied at sertical quality was less than 0.001, and a generalized creep junctions. Appendit C contains more information rupture model for RELAP5 heat structures. The on the crounow junction, which is used to treat sensitisity calculations used an updated sersion of l multidimensional flow. SCDAP/RELAP5/NtODI Cycle 5. Updates to the B-3

code included several error corrections, and a the molten material. Modelt of the debris forma. change in the cladding failure subroutine that tion and behavior in the reactor vessel lower head caused the cladding to fail when its temperature have been deseloped and incorporated in more exceeded 1470 K,if it had not already failed follow-recent sersions of SCDAP/RELAP5. The oxida-ing ballooning. This change simulated the interac. tion of control rod stainless stael occurs only if the tion between the Zircaloy cladding and the inconel surrounding Zircaloy guide tube is completely oxi-grid spacers, which would be expected to form dited. The control material in the control rods is holes in the cladding by the time this temperature assumed to be Ag in Cd. Breaching of the oxidized 4 was attained.il 6,7 fuel rod cladding by the molten material within the The core damage part of the SCDAP/RELAPS fuel rod is controlled by the user. In the scoping code does not consider the metallurgical interae-studies, the oxide shell failure depended only on tion of Zircaloy cladding and grid spacers. As men-temperature; a tempcrature of 2500 K was selected. I tioned above, a simple rnodel of the effect of such in the sensi tvity calculations, the breach was interactions on the intact fuel rod cladding was dependent on both temperature and oxide layer i used in 'he sensithity calculations. Interactions thickness. The cladding would breach when the between molten material and thr; fluid below the temperature reached 2500 K only if the cladding core were not explicitly modeled in these sersions of were less than 60% oxidized The code does not the code, although enough information was availa-allow oxidation of materia' aaloy) while it is l ble to use a control system to dissipate the..,ergy in relocating. I i i 1 I f t 1 t .i n 1' [ l i i ) l i 4 1 l 1 0-t f i

1 REFERENCES B 1. T. C, Cheng et al., "RELAP5/SCDAP - An Integrated Code for Severe Accident Analysis " Pro-credings of the TNrteenth K'ater Reactor Sqfet) Research Irtformation Meeting, Gaithersburg, MD, October 22 23,1985, NUREO/CP-0072, pp. 347 355. f i B 2. G. A. Berna, C. M. Allison, and L. J. Slefken, SCDAP/ MOD //iO! A Computer Code for the Analysis of LH'R l'ssel Behavior Durirts Ssvere Accident Dunstents, IS SAAM 83-002, Rev.1, I e l July 1984. B 3,

11. Jordan, J. A. Olesche, and P. Baybutt, TRA AMELT User's Manual, NUREO/CR-0632, BMI-2017. February 1979.

l B-4. H. Jordan and M. R. Kuhlman, TRA AMELT2 User's Manual NUREO/CR-4205, BMI 2124, May 1985. I l B 5. V. H. Ransom et al., RELAP5/ MOD 2 Code Manual, iblumes / and 2. NUREO/CR-4312 EGO. 23%, August 1985. l 1 it 6. S. Iiagen and P. I10ffmann, Physicaland ChemicalBehavior qf LH'R EbeiElements up to l'ery High j j Nmperatures, KfK4104, June 1987. l B 7. S. Hagen, L. Sepold, P. Hoffmann, and O. Schanz, "Out of Pile Experiments on Severe Fuel Dam-age Behavior of LWR Fuel Elements (CORA Program)," lA EA Internationa/ Symposium on Severe Axidents in Nuclear IUwer Plants, Sorrento, Italy, March 2123,1988. i I i i i l r l [ I I l i i f 2 i I i i ) t 3 r 4 l t j l i ) B5 i )

-a w s --J .i m l l l ( APPENDIX C USE OF RELAPS THERMAL HYDRAULIC MODELS IN THE SEVERE CORE DAMAGE ACCIDENT ANALYSIS PACKAGE 1 \\ e L t t l i i { i b' l l. I C1 I

i APPENDIX C USE OF RELAPS THERMAL HYDRAULIC MODELS IN THE SEVERE CORE DAMAGE ACCIDENT ANALYSIS PACKAGEa A Position Paper V.11. Ransom J. C. Lin C. hl. Allison P. E. blacDonald Introduction and Summary w hen the couplings and interactions between these processes are modeled simultaneously. The mission of the Soere Core Damage Analysis Considerable thought and plaening has been Package (SCDAP)C l is to predict,in best-estimate addrened to the issue of which Jiermal hydrau'ic fashion, the consequences of a soere accident in a model should be incorporated into SCDAP. The TRAC-PWR -2 TRAC-IlWR -3, and REIAP5C4 C C light water reactor up to the point of accident ter. mination or major core relocation. The accumula-models were comidered along with the kjea of dewky-tion within and the release from the primary ing a new wssel multidimensional thermal-hydraulic coolant system of hydrogen and radiologically sig-model. A link with RELAPS was chosen for the fol-nificant fission products along with the overall lowing reasons. Iirst, the combined RELAP5/ geometry and coolability of the damiged core must SCIMP/IRAP MELT code will be able to treat all of be described. The thermal hydraulic behastor the important procesws including loss of geometry, within the reactor sessel and the entire primary sys-release and transport of hydrogen and the important tem will both influence and be influenced by the radionuclides, two-phase primary coolant system core damage progression and the fission product thermal-hydraulics, and muhidimensional flow in the release. For example, natural circulation within the wssd driwn by increawd axial pressure drops across primary system may lead to early structural failures blockage regions or buoyancy forces. Second, in either or both the vessel and primary system. RELAPS is fast running, and is based on a modular changes in core heating rates, and changes in fis* and user friendly architceture that can casily accom-sion product retention dusing certain risk domi-modate linking with SCDAP and TRAI4tELIG5, nant transients. Natural circulation flow s can move nnally, the cost to dotlop and maintain the linked significant thermal energy from the core to periph* code is signineantly less than would be required to crat structures, and the evaporation, condensation dorlop new capability since RELAPS prosides a and re-evaporation of the solatile fission products prmen capability to model pressurized and boiling can also relocate a significant fraction of the decay water reactors, as well as the light water reactor (LWR) heat source in addition, the runaway oxidation safety esperimental facilities, and core heatup is influenced by the steam supply llowner, questions hase been raiwd regarding the which is, in turn, mfluenced by core geometry ability or RELAP$ to adequately model multidimen-changes. Therefore, a best estimate calculation of sional flow phenomena within a damagni core. Specif-core damage progression, fission product transport leally the modeling of recirculating flow downstream and coolant flow patterns can only be achiewd of a blockage has been mentioned.C4 The recircula-tion region formed downstream of a bkxkage is the result of flow separation and reattachment with a semi-

e. Tha poution paper su entten in 1984. before the SCIMP stagnant flow within the separated region. The sepa-and R t L.APS codes mere integrated. Ahhough some of the da-rated flow region results from laminar / turbulent cumon in that regard n dated,it is reprc4wed here because the asumon or the crounom knetions and iiicous nom erfceti a boundary layer separatten that sceempanies an l

suu apptwable. adwrse pressure gradient in the mean flow. The C3

modeling of such viscous / turbulent effects requires a Severe Accident Modeling boundary layer model for the now at the wall coupled Requirements with a turbulence model for the mean flow. None of the LWR system codes have such models. Only the C COllRA 7 code contains a viscous mean Oow model, SCDAP is a significant step forward in the mech-but nen this rnodelis a simple isotropie eddy viscosity anistic modeling of the fuel and coolant behasior model that is limited to mean flow shear; specifically, during se ere accidents. Howner, uncertaintles still shear effeds near a boundary cannot be modeled and exist in the modeling of processes such as clad thus flow separation / reattachment cannot be mod-metal water reaction, fuel liquefaction and eled. TRAC and REl.AP5 do not have siscous/ fragmentation, and core material relocation. The turbulent rAcar models and separatal flow regions spatial variation of these processes is modelled in with recirculation cannot be calculated, let alone accu-SCDAP by disiding the core into regions or nodes. rately modeled, llowner, we beline that the uncer. The numbers of nodes which can be used is mostly a maner f c mputer economics in terms of core tainty resulting from neglect of these effects is smallin comparison to the uncertainties associated with model-storage and esecution time. The modeling work at ing mechanical behavior of snerely damaged fuel. the Idaho National Engineering Laboratory n CDAP and at Sandia National Labo-Recirculating Dows that result from body forces, ratory on TRAC /MIMAS mdicates that nodes such as gravity, play an important role m. the sesere smaller than about 30 cm characteristic dimensions accident process and must be modeled. These are impractical. This node size is about the same as effects can be modeled by all the system codes and nodalizations which are used for core thermal-do not require uscouty/ turbulence models. The hydraulic modeling (even for accidents not imotv-TRAC sessel component uses a full three-ing fuel damage) and consistent with the length dimemional nomiscous flow model, COllRA con-scale of two-phase mistures. tains user specified options either/or a Viscous / turbulent now effects can be modeled three-dimensional siscous flow model or a nomis-by either a microscopic approach using nodes cous crossnow model for parallel channel flow. smaller than the smallest turbulent eddies of inter-and REl.AP5 uses a simplified multi region est or by a bulk shear modelin which the details of crossuow modelin which some of the momentum the Dow near a boundary are approsimated by a Dus cross-product terms are neglected. All buoy-boundary layer model to proside boundary condi- ) ancy and wall shear forces are included in RE LAP 5 tions. Only bulk siscous/ turbulent effects could be i for both the asial as well as crossflow models. modeled using a node slie of 30 cm or larger in a Thus, REl.AP5 can predict recirculation roulting reactor core and thus special boundary layer from natural convection. models would be required in order to model the j The redistribution of now in a damaged core details of now separation and reattachment with i resulting from local flow blockage can also hase a formation of recirculation regions. Ilowner, the significant effect on the coolability of the core and anisotropic effects of the fuel geometry could not the circulation within the primary coolant system, be modeled by a bulk shear model and must be This effect can also be accurately modeled without approsimated by an empirical Darcy law formula-l consideration of bulk viscous / turbulent shear tion anyway. The primary reason for this is that the models and ell the system codes hase this capabil, length scale of the siscous/ turbulent effects is less ity. The flow redistribution is dominated by aniso. than the pore size (i.e., approsimately I cm or tropic wall shear effects (that result from the smaller)and cannot be represented mechanistically embedded mattis of fael rods) and the change in at the characteristic length of most numerical now resistance due to local blockage. These effects models (i.e., greater than 30 cm). are modeled in all the codes by empirical Darcy The primary motisations for core wide and sys-Ope friction factors. tem thermal-hydraulic models for use in snere The status of two phase system modeling in rela-accident umulatton are to properly represent the tion to the phenomena ofimportanee to se ere r.cel availability of coolant to the core, tue transport of dent modeling is discuued in greater detail in the E#"' I

1. "#'8Y a

e kan su n p s 8 b ut following section of this paper. In particular, the

• 1. 'I '

RELAP5 socre accident core wide fluid modeling Inelusion of fluid siscous/ turbulent shear models capabilities are discussed, would mainly affect the coolant selocity field C-4

i l l dowmtream of.:brupt changes in now area or ble code can model viscous / turbulent now effects in a i direction where now separation and formation of reactor core nor is it neceuary that such effects be i regions with recirculation would esist.

included, llowner, the hear tramfer correlatiom for natural The RELAP5 multi region modelis a simplified comection steam cooling or boiling (the comective formulation obtained by deleting the cross product cooling mechanisms of most importance in a damaged momentum Dus terms from the asial momenturn i

core) are not stmng functions of the Duid velocity and equation and by deletion of all momentum aus thus accurate descriptions of such flow regiora are not terms from the crounow momentum equation. neceuary. The tramport of energy and entrained th-The salidity of this approach for moderate or low sion produets throughout the sptem by the bulk now flow selocities can be verified by an order of magni-aho is not significantly innuenced by siscous/ tude analpis using the equivalent two nuid Nasier-turbulent shear effects, in summary, a detailed Stokes equations. The vector formulation for the sisemity/ turbulence modelis not feasible with present ensemble and time averaged momentum equations mathematical modeling and computational technol-has been derised by Ishii,C-8 and appears as fol-ogy and h not required for best estimate modeling of lows in noncomenatise form for the Lth,peeje. the soere accident progreuion, s s

• V VL * ~ VP PL (0Y /01) kVL k

Thermal Hydraulic Models for + pt g + r - ( r( - p\\,O,L) - V PL (1) s L Severo Accidents The terms of the equation can be interpreted as fol-llest estimate modeh of the severe accident proc-low 5: the time rate of change of momentum, force ess require that the macroscopic thermal and man gradient due to momentum comection, force gra-tramport procenes throughout the entire nuclear dient due to pressure, grasitational force gradient, steam supply sptem be included. The rate at w hich and siscous/ turbulent force gradient. Equation (1) j energy is transported away from the core to other in nondimemional form 9 becomes, C components significantly affects the course and soerity of core damage and anociated fiuion product release. Thus it k clear that a sptem (1/Ngg)(OV /Ot) + (VL V*) VL, L thermal hydraulic modeling capabihty is required 3 for coupling with the SCDAP code. Additional - (1/NRu) V'P' + (A%/V9 8 thermal hydraulic requirements include the ability to model the Guid procenes in the core and unel + (1/Nge) Ve (rt - pk VCVC) such as now redistribution resulting from local blockage and natural circulation within the core - (As/pk o) V ilk (2) Y and plenum regions. Iloth of these effects are sig-nificant mechanisms for core cooling and fluion product transport within the senel and must, there-fore, be included in the thermal hydraulic model, where The candidate codes that hast been comidered for (t V /t ), Strouhal amber integration with SCDAP to proside the system and NSt = oo y core wide thermal hydraulie modeling include TRAC-Y 2 (P o /P ), Ruark number PWR, TRAC LlWR, COllRA/ TRAC, and RELAP5. NRu " o The TRAC and CollRA/ TRAC codes include three-(l Y Prp), Reynolds number. limensional senel thernud-hydraulic nnteh, while NRe

• oo RELAP5 uses a muti-region crounow model to appmsimate multidimensional effects. The RELAP5 aportwh is the simplest and results in a highly sersatile The siscous stress tensor rk s assumed to have the i

'ast running code. The COllRA/ TRAC model same dependence on parameters of the now as the ) includes a siscouvturbulent now nniel for the mean siscous stress tensor of a Newtonian fluid. l now shear effcvts. Ilowoer, the model is based on an in the low Dow limit the momentum Gus terms isotropie eddy siseosity fomiulation that is inappro-of Equation (2) are small in comparison to the priate for now with embedded structures such as the other terms because the coefficients 1/NSt.1/NRu. seuel core and upper plenum. lhus none of the asalla-1/NRe, and the coefficients of g and P, all become L j i i C5

large for small V. Thus the low flow limit of Equa-much greater than velocities that could exist during o tion (1)is the core damage phase of an accident. The reason that the momentum nux terms are not significant s ~ k'IOV /at) = - VP + Pkg factors for flow in a light water reactor core is that k the now is dominated by the gravitational forces + V * ( 3 - pk V ' V ') - Y I'k (3) and the anisotropic wall friction (the ratio of k k k hydraulic resistances in the radial to axial directions The flow is dominated by the forces due to pressure is approximately 50). The simplified momentum gradient, grasity, viscosity / turbulence, and mass Dux formulation used in RELAPS is clearly ade-transfer. The forces due to momentum flux may be quate for predicting flow redistribution effects neglected.C 9 resulting from local blockage due to fuel damage. A more severe test of the simplified RELAPS Since the modeling of natural comection currents momentum nux formulation is provided by exam-within a reactor sessel is neceuary for best estimate ining the various terms of Equation (1) for calcu-modeling of a soere accident, a further numencal lated high velocity flows in a reactor cue with experiment was conducted. ne RELAP5 multi-region blockage. The flow geometry and the f ELAPS approximation was used to simulate the natural circu-nodalization for the case examined is illustrated in lation resulting from asymrnetric heating. The configu. Figure C l. The core is modeled by Ihree concen-ration modeled is the same as the proious case, i.e., an tric regions each divided into five axial nodes. Flow adsymmetric core hasing three concentric sertical pas-is totally blocked between the third and fourth sages with each divided into the equal length nodes. nodes in the Iwo innermost regions. An inflow Crossflow junctions link the axial nodes at each of the boundary condition at the bottom of the core sessel fhe loets. The core has no through.now and heating is specified such that the flow through the occurs in the center channel. Once the calculation unblocked core would be 1.0 m/s. reached steady state the flow pattern shown in The magnitudes of each of the terms appearing Figure C-2 resulted. The maximum selocity in the in Equation (1) are tabulated in Table C-l (using heated channel is approximately 0.1 m/s and the cylindrical coordinates) for the selected points indi. downward flow selocities in the two outer channels are cated on Figure C 1 (points I through 5). The time approximately 0.02 m/s. The momentum flux terms at rate of change of momentum is zero in all cases these low selocities are negligible as pre iously shown since the calculations were carried out to steady. by the order or magnitude analysis. ncse results con-state. For the axial nodes, only the radial momen. firm the ability of the RELAPS multi-region modeling tum flux term is neglected in the RELAP5 approach to approximate recirculation resulting from calculation. This term is less than 10% of the total buo>ancy forces, pressure gradient except at Node 3 where a strong radial gradient in the asial velocity exists in front of the blockage. Here the neglected term is 25 % of the Therm. I Hydraulic Model pressure gradient and is a result of the Dat plate Expectations configuration of the idealized blockage. This value is still within the uncertainty of the flow resistance Misconceptions edst with respect to what can be definition and the numerical approximation of the expected from the thermal-h>xtraube models employed derivative terms. in the light water reactor safety codes. First, as dis-The components of the radial momentum equa-cussed abose, none of the system codes such as TRAC. tion are tabulated in Table C 2 for selected points PWR, TRAC BWR, COBRA / TRAC, or RELAPS in the flow (points 6 through 10). In this case both can mechanistically model siscous-turbulence effects momentum flus terms are omitted from the in an anisotropic porous medium, such as a reactor crossflow momentum equation. The neglected core. The reason for this is that the length scale of the terms are again less than 10% of the total pressure turbulence is small compared to the length scale of the gradient except at Node 10 where the velocities are ensemble average two-phase nuid model and the spa-l Very small However, this is the region where recir-tial nodahzation used is too coarse. The COBRA / culation would exist in real flow, but cannot be TRAC code has a bulk siscosity/ turbulence model modeled by any of the system codes. suitable for free shear flows, but this model ls inappro. These calculations were made for flow velocities priate for LWR cores where the flow is dominated by approximately equal to the flow velocity at full anisotropic frictional effects due to embedded struc-power operatic n (1.0 m/s through the core) and are tures. Phenomena such as now separation and l C-6

Outlet Outlet l I I I I ~ Nodalization j Flow i schematic -l-schematic ] g h' 1 \\ l _t i sl l, lf e IC-NNNNNhAN Blockage Blockage } N i r-- % 'i 8 8! I k \\' 9 2 O 4 3 4 __1_ 1 l f 0 7 6 j r-i i t i i i i i 1 Normal Junction I k I ( -X-Crossflow junction inlet t inlet EC000199 l Figure C-1. RELAP5 multi-region core model with blockage, velocity vectors at volume centers.

Tbble C 1. Axlel momentum equation components Wall Friction aV Y 0Y 3Y z r Z V z 18P and Node T ai F p Tz~ Form In_ss 1 0.0 1.09a 3.32 27.61 14.48 9.81 2 0.0 0.00a 1.05 9.82 1.04 9.8) i 3 0.0 2.49a ,1,i7 .]0,33 3,74 9,33 4 0.0 2.25a 4,79 -32.25 17.65 9.8 i 5 0.0 0.00a 0.00 9.82 0.01 9.81 1

e. Neslected terms in RELAPS formulation.

i Table C 2 Radial momentum equation components l i Wall Friction DV V OV V OV r r f r 1, 8 P and j Node at ar az p or Form loss g 6 0.0 0.1Ia 0.06a 7.61 7.61 0.00 l 7 0.0 0.68a 3,7ga I8.63 18.63 0.00 l i 8 0.0 0.12a 0.02a ,g,03 g,03 o,00 l 9 0.0 0.85a g,7ga 29.62 29.62 0.00 l l 10 0.0 0.00a 0.08a 0.I4 0.I4 0.00 l i i

a. Neslected terms in RELAPS formulation.

[ i r I i C-8 l

i Outlet Outlet I I s l l t l kl Nodalization

pgo, schematic schematic y

h n n I 9 N sl [ x x l @> t k i I I k i 8 o l t, s E E i ) i w X X l h I 1 t I I i Normal junction ( ( -X-Crossflow junction i EC0H2M Rgwe C-2. RELAP5 multi-respon core model with heated center channel, velocity vectors at volume centers. i

reattachment with recirculation in the separated a detailed or a bulk Dow model. This is due to lack now region are the result of siscous/ turbulent of a detailed two phase model with siscous/ cffects near the wall or boundaries of the Dow, turbulent effects and in addition the large number Simulation of such cffects uould require a shear of nodes that would be required for solution due to layer or boundary layer model coupled with a the need to include detailed boundary layer consid-model for the bulk now turbulence effect such as erations for use with a bulk shear model. the La model.C 10 Thus, the recirculation that might occur behind a step or blockage in a core CONCLUSIONS cannot be modeled by any of the LWR system codes Second, the modeling of separated now regions None of the I.WR system codes can model with recirculation depends upon whether the boundary goserned siscous/ turbulent effects on siscouuturbulent effects are included in the model the bulk now in an 13VR core or plenum region, and not on whether all momentum Dus terms are Thus, none of the LW R system codes can predict the recirculat,on that occurs downstream of an i included. This fact is sisidly illustrated by a calcu-C a pt ste p r age Howeser, N uncertainnes lated result obtained using the llEACON Il code associated with modeling the mecham, cal, thermal, w hich is a containment sersion of the and chemical behasior of the core componenti dur-KACillNA code, lloth of these codes hase a ing a sesere accident are judged to be far greater full multidimensional model including all the than any uncertainties resulting from omission of momentum cross product terms, but do not con-siscous/ turbulent effects. Thus, such effects are sider siscous shear or turbulence. The results for not significant factors in the best estimtte predic-Simulation of a now hasing seseral blockages are tion of 13VR system behasior for sescre core dam-shown in f igure C 3. Note that no Dow recircula-age accidems. tion is predicted to esist downstream of any of the The now in an 1WR senel does not depend strongly E abrupt steps in the boundary. These results are sim-on the momentum Dus crou-product terms and is ilar to those obtained using the RELAP5 multi-dominated by grasity and kinetic loss effects. Thus the region model for flow in a blocked core, see RELAP5 multi-region model is completely adequate Iigure C-l. Recirculatior. regions would form for modeling the multidimemional core-wide flow dowmtream of the rearward facing steps in a real effects that are awriated with sesere core damage acci-now. Itoundary layer separations would result dents and large now blockages. from the adserse pressure gradient in the mean The linking of RELAP5 with SCI 1AP and TRAP-flow, l'o r comparison, a calculation by MEIT to obtain a system and core-wide sesere fuel Roache -13 for a single phase now oser a back-damage accident analysis capability is strongly recom-C step using a siscous flow code is illustrated in Fig-mended. This approach is completely adequate for ute C-4. Note the boundary layer separation at the modeling of the thermal hydraulie system respome rearward facing step followed by reattachment and takes advantage of the superior code architecture, downstream. A siscous shear-drisen recirculation efficient esecution, and user friendly features of region is formed within the separated now region. RELAP5. The results of this resicw reinforce the cor-This calculation was made using a sery detailed redness of the conclusions of an earlier INEL study nodalisation in order to resche the Dow process concerning the selection of a thermal. hydraulic model near the wall. A comparable calculation for a reac-to link with SCDAP. RELAPS was alw m;ommended tor core now is not technically feasible using either in that study. C-10

4 4 4 4 k 4 4 Y b T A A 4 g 4 Y e Y Y Y WW h A \\ 4_ 4 b l / 5 l' i 4-a e I / 4 F f p 9 u ] 2 Y Y T e d Y Y Y .;\\ 8 3 f f se %A %A A 4 s V \\ \\ A 1 a f k k k i C Il

-.----*-l-.--.---.--*-j-*.--*. - * - - - -- - -.---.---l-.----.--+I*.-.--.--- - -*. -- 3 - * - - - - - -- +a #_-.-l +__-e __._- -e +_l_+a #a + - _ _ -w -w -n -.- - l ->.- -.- -. 5 // / / -.--..---.---.-q---.-- j.--+.---.m-+-----..l.--.--.-----.--.--.------.--.--I+-.-_*-- y, ,----l+-.----.--*l-*--.--*--*. l v t w w *-1 -*- -- -- g s; -*. -- -* -* 2 - 4 / t 4 e p; ss I i L sww

• - H e g g s 3 -*- -.- -

-.-l / A 4 % w w ---j - - -*- -.- -- -- / % w -- m --- + - - --- ml -- *- -.- -.- j - - - - - - ~ g 2 r m (c) flow direction for compressible flow in a base region, Mo = 2.24. Re = 300, y = 1.4. Rgure C4. Calculated viscous flow over a backstep. from page 337 of Reference C-13.

REFERENCES C l. G, A. Ilerna, C. St. Allison, a.:d L. J. Sieflen, SCDAP/Af00//lD: A Computer Code for the Analysis of Lil'R l'essel Behavior During Severe Accident Dansients, IS-SA Aht 83-002, Rev.1, July 1984. C-2. D. R. Liles et al., TRAC PD2, An Advanced Best Estimate computer Progmmfor Pressuri:ed il' ter a Reactor inss of-Coolant Accident Analysis, NUREG/CR-20$4,1981. C 3. D. D. Taylor et al., TRAC-BD)/Af0DI: An Advanced Best Estimate Computer Progmmfor Boiling if ater Reactor hansient Analysis, iblume I: Afodel Description NUREGICR-3633, EGG 2294, April 1984. C-4. V. I1. Ransom et al., RELAP3/A10D2 Code Afanual, Iblume 1: Code Structure, System Afodels, and Solution Afethmh, NUREG/CR-4312. EGG 2396, August 1985. C-$.

11. Jordan, J. A. Gieseke, and P. Haybutt, TRAD3 FELT User's Afanual, NUREG/CR-0632, 11N112017, l ebruary 1979.

C-6. hiemorandum for O. E. Ilasset from C. N. Kelber on subject, "Trip Report, SNL and LANL, April 912,1984," dated April 19,1984. C 1. h1. J. Thurgood et al., COBRA / TRAC-A Thermal flydmulics Code for Dansient Analysis of Nuclear Reactor Ihsels and Primary Cooling Systems, Iblume 1, Equations and Constitutive A1odels, Pacific Northwest Laboratory, NUREG/CR 3N6, Starch 1983. C-8.

81. Ishii, Thermo-Fluid Dynamic Theory of Two-Phase Flow, Eyrolles (l9'$).

C 9. J. C. Slattery, Atomentum, Energy, and Stats Dan.sfer in Continua, SicGraw-liill (1972). C.10. W. Rodi, Titrbulent AIodeh and Their Application in flydmusics, IiiR (l980). C 1l. C. R. llroadus et al., BEACON /A10D3: A Computer Program for Thermal Hydraulic Analysis of Nuclear Reactor Containments, Users Afanual, NUREG/CR 1148, Artil 1980 C 12. D. D. Amsden and V i1. I1atlow, KACillNA: An Eulerian Computer Prognmfor AfultyieldFluid Hows, LA 56so,(1974). C-13. P. J. Roache, ComputationalHuid Dynamics, l{ermosa (1972). C 13

e. ...... m.. ....o.. ...- rx ~ 'l*f,*,';'/ SISt.lOGRAPHIC CATA SHEET NUREG/CR-5214 EGG-2547 es,... x, Analyses of Natural Circulation During a Surry Station Blackout Using SCDAP/RELAPS 5eptember 1988 P.D. Bayless l October 149A ,...,o....<,o.2..............a........i.c, s moes.s= =o.. v.a.v Idaho National Engineering Laboratory EG&G Idaho Inc. P.O. Box 1625 A6360 Idaho Falls, Idaho 83415 .. mo...<, <,...o i.o.......... u.

i. e.

Division of Systems Research Technical Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Comission Washington, DC 20555 l ,,....< u.. I The effects of reactor coolant system natural circulation on the response of the Surry nuclear power plant during a station blackout transient wtre imestigated. A TMLB' sequence (loss of all ac power, immediate loss of auxiliary feedwater) was simulated from transient initlation until after fuel rod relocation had begun. Integral analyses of the system thermal hydraulics and the core damage behavior were per, formed using the SCDAP/RELAPS computer code and several different models of the plant. Three scoping calculations were performed in which the complexity of the plant model was progressiwly increased to determine the owrall effects of in vessel and hot les natural circulation flows on the plant response. The natural circulation flows extended the transient, slowing the core heatup and delaying core damage by transferring energy from the core to siructures in the upper plenum and coolant loops. Increased temperatures in ihe ex core structures indicated that they mty fall, howtwr. Nine sensitivity calculations were then performed to lawstigate the effects of model. Ing uncertainties on the multidimensional natural circulation flows and the system response. Creep rupture failure of the pressuriser tturge line was predicted to occur in right of the calculations, with the hot les faillas in the ninth. The failure time was fairly insensitive to the parameters varied. The failures occurred near the time that fuel rod relocation began, will before failure of the reactor sessel would be expected. A calculation was also performed in which creep rupture failure of the surge line was modeled. The subsequeN blowdown led to rapid accumulator injection and quench. Ing of the entire core. ....,.,.;.,. g. ..m.......... ..........u..,. natural circulation SCDAP/RELAP5 Unlimited station blackout j Surry 1 acwa'" <6 * "c.'** I r a.. ........ o..i.e a.."' Unc1assified a...- Unclassified o... o.... n i .c. . u. s. u....... i.no or,ia,,s.. n i..s,.n,s

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