ML20113C438
| ML20113C438 | |
| Person / Time | |
|---|---|
| Site: | 05200003 |
| Issue date: | 01/26/1996 |
| From: | Boyack B, Lime J LOS ALAMOS NATIONAL LABORATORY |
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| LA-UR-95-4431, NUDOCS 9607010234 | |
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t j 2 Energy under contract W.7405-ENG h By acceptance of ttus arucle, the pubhsher recogruzes that the US Government retams a n ewacluswe, royalty-im license to pubbsh or reproduce the pubbshed form of tlus contnbuten, or to allow others to do so, for US Goverrunent purposes The Los Alamos National Laboratory requests that the pubhsher idennfy ttus artcle as work performed under the auspices of the US Department of Energy. Los Alamos Natanal Laboratory strongly supports academac imdorn and a researcher's nght to pubhah, there8 ore, the Laboratory as an msatuhon does not endorse the wwwpoint of a pubhcanon or guarantee ses techrucal correctness Form No 836 RS ST 262910/95
- 9607010234 960626 PDR ADOCK 05200003 Attachnent 2
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Title:
UPDATED TRAC ANALYSIS OF 80% DOUBLE-i ENDED COLD-LEG BREAK FOR THE AP600 i i w i s I t I l l Author (s): James F. Lime ! Brent E. Boyack l f i l I t i i> l l l ( ) i Submitted to: For USNRC Distribution I i i
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; Energy under contract W-7405-ENG 36. By acceptance of this article, the publisher recognizes that the U.S. Govemment retains a nonexclusive, royalty free
{ license to pubhsh or reproduce the published form of this contribution, or to allow others to do so, for U.S. Govemment purposes. The Los Alamos National Laboratory requests that the pubhsher identify this article as work pertormed under the auspices of the U.S. Department of Energy. Los Alamos National l Laboratory strongly supports RCademic freedom and a researcher's nght to publish; therefore, the Laboratory as an institution does not endorse the viewpoint
- of a publication or guarantee its technical correctness.
i Form No. s36 R5 j ST 26291o95 1_ ----
O LA-UR-95-4431 Los Memos Netonal Laborstwy is operated by the Urwerony of Carriorrua for the Unned States Depatiment of Energy inder contact W.7405-ENGG6 TITLE: UPDATED TRAC ANALYSIS OF AN IKP/. DOUBLE-ENDED COLD LEG BREAK FOR THE AP600 DESIGN i 9 AUTHOR (S): J. F. Lime and B. E. Boyak i SUBMITTED TO: J J i i Bj acceptance of this artcle, the pthksher recognues lhet the U.S. Government retems a nonexcluerve, royett ree f licanoe b pubhoh or reproduce the put2shed form of this contnbuben, or to allow cibers b do so, for U.S. Govemmera purposes. The Los Ahmos NeesW Laborsemy foemas te me metener Wenpy tus erede es nem peermed unew em suspues eene U.S. Depenmort or Enugy. ~ ~ _os A amos Los Alamos National Laboratory Los Alamos, New Mexico 87545 FORM NO. 82s R4 ST NO as2e WB1
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l 4 1 j -a 4 I UPDATED TRAC ANALYSIS OF AN 80% DOUBLE-ENDED COLD-LEG BREAK FOR THE AP600 DESIGN 1 l J. F. Lime and B. E. Boyack i
- Technolc,gy and Safety Assessment Division i Los Alamos National Laboratory i Los Alamos, New Mexico 87545 b
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- CONTENTS i
i j N O M EN CLATURE .................... ...... .. ........... . .. .. .. .............. ....... .. ... .. .. .... ... ..... .. .. ... . ..... . .... ... xii MCUTIVE S UMMY . ...... ........... . . ... .... . ........... .. ............ . . . .. ... . .. . .. . . . ......... . . . .... .... ........ 1 i ABSTRACT.........................................................................................................................4 1.0. I NTR OD UCTI O N ........ .. ... ......... ........... .... ..... ..... ... ... ....... ........ ... .. .. ..... . . . . . . . .. .... .... .. .. 4 j 2.0. PLANT DESCRIPTI ON ....................... ................................................ .. ........................ 5 l 2.1. AP600 Description ... ............. ............... .......................................... .................. 5 i 2.2. Key AP600 Features-LBLOCA Response ......... .. ........................................... 7 i 2.3. Comparison with Current Generation Westinghouse PWRs...................Z.:..... 7 1 3.0. TRAC M ODEL DESCRIPTION .......... .... ..... ...... . .............................................. 7 3.1. Plan t Model Database ........................... .. .............................................................. 7 ! 3.2. Reactor Vessel ................ ............ ................ . .. ... .. .... . ........... ................ ........ 8 a 3.3. Loop Components ......... .......... ........ .............. ................................................ .. 9 l l 3.4. Fassive Safety Systems ................. . . ............... ........ . ...... ...................... 10
.... .. ... .................... .............. ...... 10 . 3.5. Updated Model Changes ........ ...
i 4.0. TRAC CODE DESCRIPTION ..... ..... .... . ..... . ....... ....................... ....... ...... 12 l, 5.0. LBLOCA CALCULATION RESULTS .. . ... ...... ..... ................... ...... .............. 13 i 5.1. Blowdown Period (0 to 12.3 s) ... ..... .. ............................ .................... 14 i 5.2. Refill Period (12.3 to 43 s) ..................... . ............................... ......................... 22
- 5.3. Reflood Period (43 to 240 s) .............. .. ...... ....... ...... .................................. 26 l 5.4. Long-Term Cooling Period (240 s and Beyond) ................................................. 31 4
6.0. COMPARISON WITH ECOBRA/ TRAC .. . .... ................:........ ................ .... 33 l i + 7.0. CONCLUSIONS AND RECOMMENDATIONS ....... .................. ........................ 35 j REFER EN CES ........... .. . ..... . ... .. .... .. . .. .... .......... . .......... .. .... .. .............. 36 i APPENDIX A. CORE-INLET MASS FLOWS ... . ............ .................................. .... A-1 J l APPENDIX B. CORE-OUTLET MAS S FLOWS .........................................................B-1
- APPENDIX C. AVERAGE-ROD CLADDING TEMPERATURES ................. .........C-1
- APPENDIX D. HOT-ROD CLADDING TEMPERATURES........................................D-1 s
APPENDIX E. AVERAGE-ROD CLADDING TEMPERATURES VS ROD ELEVATION AT SELECTED TRANSIENT TIMES ................E-1 APPENDIX F. HOT-ROD CLADDING TEMPERATURES VS ROD ELEVATION AT SELECTED TRANSIENT TIMES ................F-1 APPENDIX G. LBLOCA TIMESTEP AND CPU TIME INFORMATION ................G-1 H
B a TABLES I. EXTENDED LBLOCA SEQUENCE OF EVENTS ............................... ...................... 37 H. COMPARISON TO.WCOBRA/ TRAC 80% DEGB LOCA ......................................... 38 FIGURES
- 1. AP600 plant isometric ........................ ...................................................... .................... 39
- 2. AP600 containment cutaway view .... .............. ................. ............... ...................... 40
- 3. Schematic of AP600 reactor coolant system and passive safety systems................. 41
- 4. Isometric view of reactor vessel model ............................................................... .. 42
- 5. Elevation view of seactor vessel model, including downcomer noding.................. 43
- 6. Plan views of reactor vessel model .. ........ .......... ............. ........ .......................... 44
- 7. Reactor vessel heat structures .......................... ..................................................... ..... 44
- 8. Reactor coolant loop-1 model overview ................... ... ................. ......................... 45
- 9. Reactor coolant loop-2 model overview .......... ..... .............................. .................. 45
- 10. Passive Safety Systems model overview .... ...... ........ .. . ......... ........................ 46
- 11. Vessel-collapsed liquid levels ......................... ...... ............................... .................... 47
- 12. System p n:aures .. - . ................ ......... . ...... ............ ...... ..... ....................... 47
- 13. Fuel-rod maximum cladding temperatums . ... ....... .......... .................................. 48
- 14. Pressurizer mass flows ........................ .......................................................................... 48
- 15. Accumula tor mass flows . .... .. . . .................. ........ ..... ............................... ..... 49
- 16. Accumulator liquid level ................... ................ ................. ........... .... ................... 49
- 17. Accumulator liquid volume fraction .............. ... .... .......... . ..... ...................... 50
- 18. Core makeup tank mass flows ... . . ...... ..... ................ ..................... .............. 50
- 19. Core makeup tank liquid levels ...... . .................. .............................. .... ................ 51
- 20. PRHRS mass flows ............ .............. . ....... .................... ...... ........................ ...... 51
- 21. Core makeup tank-pressure balance-line mass flows ................................................ 52
- 22. B rea k ma ss flo ws .... . ............. ...... ............ ............... .. .. ........ ...................................... 52
- 23. B reak exi t voiding ................................................. ......................................................... 53
- 24. Steam-generator primary-outlet mass flows ........................................ ...................... 53
- 25. Cold -leg mass flows .................. ......... .................. ...................................................... 54
- 26. In ta ct cold-leg voiding ......... .................................................................. .................... 54
. 27. H ot-leg 1 mass flows .............................. .................. ............................ .................... 55 All m
O 1 FIGURES (CONT) e
- 28. Hot-leg 2 mass flows ..... . ....................................................................................... .... ... 55
- 29. H o t-leg 1 v oiding ... .......... .. ...... .. ... ..................... .. ........ .. ..... .... . ..... .......... .. . . .. .. .. .... 56
- 30. Hot-leg 2 voidirn . .................... ...... .......... .... ...................... .. ...... ........ . ....... . . . . .. .. . . .... . ... 56 1
- 31. Total makeup flow ... .. ...... ................................. ............................ ........................ 57 3
- 32. Net-system mass-flow loss ...... ......................... ................................................ ...... 57
- 33. Integrated net-system mass-flow loss ...... ........... ............ .. .............. .................... 58 4
- 34. Heated core-average vapor fraction .................... ........................................................ 58 4
- 35. Core-inlet mass flows ... . ...................... .......... .................... .................................. 59
- 36. Core-outlet mass flows . ......... .. .......... ............ ............ .............. ........................ 59
- 37. Core-outlet liquid-mass flows ....... .......... ..... .................. ........................................ 60
- 38. Core-outlet vapor-mass flows ... . .. ..... ........... ................ ................ ............ ...... 60 f
- 39. Fuel-rod reactivities . ... ........................... ........ .... ... ..... . ............... ............ ...... 61
- 40. Stored-energy distribution in fuel rod with time . ... ... .. .................................... 61
- 41. Average-rod temperatures in cell 1 (r = 1, 8 = 1) .. . .. ....................... ..... . . 62
- 42. Average-rod temperatures in cell 4 (r = 1, 0 = 4) .... .
.. ...... .. .................... 62 4
- 43. Hot-rod temperatures in cell 1 (r = 1, 8 = 1) ... .... . ..... ..... ...... ............ .. 63
- 44. Hot-rod temperatures in cell 4 (r = 1, 0 = 4) ... .......... .... .... .......... .. .. 63
- 45. Average-rod temperatures in cell 9 (r = 2, 0 = 1) . ... .. ............. .................... . 64
- 46. Average-rod temperatures in cell 12 (r = 2, 0 = 4) .. ... -.... .................................. 64
- 47. Cell 4 average-rod temperatures vs rod elevation, O to 10 s ...... ........................ 65
- 48. Cell 4 average-rod temperatures vs rod elevation,20 to 140 s ............. .................. 65
- 49. Cell 4 hot-rod temperatures vs rod elevation, to 10 s ....... .................................. .... 66
; 50. Cell 4 hot-rod temperatures vs rod elevation,20 to 140 s ....................................... 66 4
- 51. Downcomer average vapor fraction ... ... . .. .. ... .. ...... .......................... 67 4
- 52. Total guide tube flow ....... .......... ... ..... .. .... .......... . .. ........... ....................... 67
- 53. Iower-plenum average vapor fraction ....... ...................... ................................. . 68
- 54. Lower-plenum coolant temperatures. ....... ... .......... .................. . ..................... 68
! 55. Upper-head average vapor fraction .................................. ................. .................. 69 56 Upper-plenum average vapor fraction . ..... ..... ................ ...................................... 69
- 57. Steam and feedwater mass flows ..... ................ ................... ......... .......................... 70
- 58. Total upper-head support-plate drain-hole mast t ..........................................70
- 59. Downcomer-to-upper-head mass flow ............ ............................ .......................... 71
- 60. Reactor-coolant pump-rotor speeds ................................ ................ .......................... 71 t
1 ( IV 1
! FIGURES (CONT)
- 61. Comparison of ECOBRA/ TRAC and TRAC upper-plenum pressure................... 72
- 62. Comparison of WCOBRA/ TRAC and TRAC accumulator mass flow.................... 72 l
j 63. Comparison of ECOBRA/ TRAC and TRAC lower-plenum liquid level .............. 73 l 64. Comparison of WCOBRA/ TRAC and TRAC downcomer liquid level .................. 73
- 65. Companson of WCOBRA/ TRAC and TRAC hot-rod peak cladding temperatures 74 l 66. Comparison of ECOBRA/ TRAC and TRAC
- hot-rod cladding temperatures at 6-ft rod elevation... .................... ...................... 74 j 67. Comparison of ECOBRA/ TRAC and' TRAC hot-rod cladding temperatures at 8.5-ft rod elevation.................. ............................ 75 l l l j 68. Comparison of WCOBRA/ TRAC and TRAC ! hot-rod cladding temperatures at 10-ft rod elevation....... . ..................................... 75 y j 69. Companson of ECOBRA/ TRAC and TRAC average-rod cladding temperatures at 6-ft rod elevation ....... ................................. 76 l ' 70. Comparison of ECOBRA/ TRAC and TRAC I average-rod cladding temperatures at 8.5-ft rod elevation . .... .... ................ .... 76 i 71. Comparison of ECCitsRA/ TRAC and TRAC j average-rod cladding temperatures at 10-ft rod elevation .. .... . .... ............ .... 77 j 72. Comparison of ECOBRA/ TRAC peripheral-rod cladding temperatures to TRAC outer-ring average-rod cladding temperatures at 6-ft rod elevation...:.. 77 l
- 73. Comparison of ECOBRA/ TRAC peripheral-rod cladding temperatures _
to TRAC outer-ring average-cladding temperatures at 8.5-ft rod elevation ... . ... 78
- 74. Comparison of ECOBRA/ TRAC peripheral-rod cladding temperatures (
f to TRAC outer-ring average-rod cladding temperatures at 10-ft rod elevation..... 78 A-1. Core-inlet mass flow, core sector cell 1 (r = 1, 0 = 1) .... ............................ ..... A-1 l 1
- A-2. Core-inlet mass flow, core sector cell 2 (r = 1, 0 = 2) ................... ................ ..... A-2 )
l A-3. Core-inlet mass flow, core sector cell 3 (r = 1, 0 = 3) ..... .......... . ..... ............... A-2 1
- A-4. Core-inlet mass flow, core sector cell 4 (r = 1, G = 4) ... . ........ ....................... A-3 A-5. Core-inlet mass flow, core sector cell 5 (r = 1, 0 = 5) ....... ............................ ..... A-3
- A-6. Core-inlet mass flow, core sector cell 6 (r = 1, 0 = 6) ......... ................................. A-4 i A-7. Core-inlet mass flow, core sector cell 7 (r = 1, 0 = 7) ..... ................................. .. A-4 r
A-8. Core-inlet mass flow, core sector cell 8 (r = 1, 0 = 8) ................... ....................... A-5 j i A-9. Core-inlet mass flow, core sector cell 9 (r = 2, 0 = 1) ............................................ A-5 . ' A-10. Core-inlet mass flow, core sector cell 10 (r = 2, 0 = 2) ................... ..... . .... ..... A-6 1
; A-11. Core-inlet mass flow, core sector cell 11 (r = 2, 0 = 3) . .............. .............. ..... A-6 a
i A-12. Core. inlet mass flow, core sector cell 12 (r = 2, 0 = 4) ... ...... ......... .................. A-7 i A-13. Core-inlet mass flow, core sector cell 13 (r = 2, 0 = 5) ... ............. .... ............... A-7 i = V i 1
4 i FIGURES (CONT)
- A-14. Core-inlet mass flow, core sector cell 14 (r = 2, 0 = 6) .......................................... A-8
~ ~
l A-15. Core-inlet mass flow, core sector cell 15 (r = 2, 0 = 7) ............................. ........... A-8 A-16. Core-inlet mass flow, core sector cell 16 (r = 2, 0 = 8) .......................................... A-9 l B-1. Core-outlet mass flow, core sector cell 1 (r = 1, 0 = 1)...........................................B-1 l B-2. Core outlet mass flow, core sector cell 2 (r = 1, 0 = 2)........................... ..............B-1 l* B-3. Core-outlet mass flow, core sector cell 3 (r = 1, 0 = 3)..... ....................................B-2 B-4. Core-outlet mass flow, core sector cell 4 (r = 1, 0 = 4).............................. ...........B-2 i B-5. Core-outlet mass flow, core sector cell 5 (r = 1, 0 = 5)...........................................B-3 j B4. Core-outlet mass flow, core sector cell 6 (r = 1, 0 = 6)...........................................B-3 l B-7. Core-outlet mass flow, core sector cell 7 (r = 1, 0 = 7)...........................................M . B-8. Core-outlet mass flow, core sector cell 8 (r = 1, 0 = 8) ......... ............ ................B-4 B-9. Core-outlet mass flow, core sector cell 9 (r = 2, G = 1)..... ....................................B-5 B-10. Core-outlet mass flow, core sector cell 10 (r = 2, G = 2)........................... ............B-5 B-11. Core-outlet mass flow, core sector cell 11 (.r = 2, 0 = 3)....... ................................B4 B-12. Core-outlet mass flow, core sector cell 12 (r = 2, 0 = 4).................. ..................B-6 B-13. Core-outlet mass flow, core sector cell 13 (r = 2, 0 = 5)... ....................................B-7 B-14. Core-outlet mass flow, core sector cell 14 (r = 2, G = 6) .... .............................=B-7 B-15. Core-outlet mass flow, core sector cell 15 (r = 2, G = 7).........................................B-8 _ B-16. Core-outlet mass flow, core sector cell 16 (r = 2, B = 8).........................................B-8 C-1. Average-rod cladding temperatures, core sector cell 1 (r = 1, 0 = 1) .................C-1 C-2. Average-rod cladding temperatures, core sector cell 2 (r = 1, 0 = 2) ...... .......C-1 C-3. Average-rod cladding temperatures, core sector cell 3 (r = 1, 0 = 3) .................C-2 C-4. Average-rod cladding temperatures, core sector cell 4 (r = 1, 0 = 4) ...... .........C-2 C-5. Average-rod cladding temperatures, core sector cell 5 (r = 1, 0 = 5) .................C-3 C-6. Average-rod cladding temperatures, core sector cell 6 (r = 1, 0 = 6) .................C-3 C-7. Average-rod cladding temperatures, core sector cell 7 (r = 1, 0 = 7) .................C-4 C-8. Average-rod cladding. temperatures, core sector cell 8 (r = 1, 0 = 8) .................C-4 C-9. Average-rod cladding temperatures, core sector cell 9 (r = 2, 0 = 1) .................C-5 C-10. Average-rod cladding temperatures, core sector cell 10 (r = 2, 0 = 2) ...............C-5 C-11. Average-rod cladding temperatures, core sector cell 11 (r = 2, 0 = 3) ...............C-6 C-12. Average-rod cladding temperatures, core sector cell 12 (r = 2, 0 = 4) .. .........C-6 C-13. Average-rod cladding temperatures, core sector cell 13 (r = 2,~0 = 5) ...............C-7 C-14. Average-rod cladding temperatures, core sector cell 14 (r = 2, 0 = 6.) ...............C-7 f 1
O l ! l l i i , FIGURES (CONT) 4 y C-15. Average-rod dadding temperatures, core sector cell 15 (r = 2, 0 = 7) ...............C-8 ,
~
j C-16. Average-rod cladding temperatures, core sector cell 16 (r = 2, 0 = 8) ...............C-8 1 D-1. Hot-rod cladding temperatures, core sector cell 1 (r = 1, 0 = 1) .........................D-1 !
~ ;
l D-2. Hot-rod ~ cladding temperatures, core sector cell 2 (r = 1,0 = 2) .........................D-1 ! D-3. Hot-rod cladding temperatums, core sector cell 3 (r = 1,0 = 3) ................ .......D-2 j D-4. Hot-rod cladding temperatures, core sector cell 4 (r = 1,0 = 4) .........................D-2 j D-5. Hot-rod cladding temperatures, core sector cell 5 (r = 1,0 =- 5) .........................D-3
- D-6. Hot-rod cladding temperatures, core sector cell 6 (r = 1, G = 6) ........... ............D-3
! D-7. Hot-rod cladding temperatures, core sector cell 7 (r = 1,0 = 7) ........ ...............D-4 l D-8. Hot-rod cladding temperatures, core sector cell 8 (r = 1,0 = 8) .................... ...D-4 i D-9. Hot-rod cladding temperatures, core sector cell 9 (r = 2,0 = 1) .........................D-5 . D-10. Hot-rod cladding temperatures, core sector cell 10 (r - 2,0 = 2) ................ .....D-5 D-11. Hot-rod cladding temperatures, core sector cell 11 (r = 2, G = 3) .. ...................D-6 D-12. Hot-rod cladding temperatures, core sector cell 12 (r = 2, G = 4) .. ................ .D-6 l l D-13. Hot-rod cladding temperatures, core sector cell 13 (r = 2, G = 5) .......................D-7 l D-14. Hot-rod cladding temperatures, core sector cell 14 (r = 2, G = 6) .. ................ .D-7 ! D-15. Hot-rod cladding temperatures, core sector cell 15 (r = 2,0 = 7) .............. .....D-8 [ D-16. Hot-rod cladding temperatures, core sector cell 16 (r = 2,0 = 8) ..... . .............D-8 ! l E-la. Average-rod cladding temperatures vs core elevation l at selected transient times from 0 to 10 s for core cell 1.................. ... ..... .... ..E-1 l E-lb. Average-rod cladding temperatures vs core elevation i at selected transient times from 20 to 140 s for core cell 1............................... ..E-2 l E-2a. Average-rod cladding temperatures vs core elevation ! at selected transient times from 0 to 10 s for core cell 2 .... ..... ............... ...... E-2 r
- E-2b. Average-rod dadding temperatures vs core elevation l at selected transient times from 20 to 140 s for core cell 2.............. .. . .............E-3 l E-3a. Average-rod cladding temperatures vs core elevation
!. at selected transient times from 0 to 10 s for core cell 3 .............. ....................E-3 I E-3b. Average-rod cladding temperatures vs core elevation l , at selected transient times from 20 to 140 s for core cell 3...... ............................E-4
- E-4a. Average-rod cladding temperatures vs core elevation
- at selected transient times from 0 to 10 s for core cell 4. ................... ............ ..E-4 i E-4b. Average-rod cladding temperatures vs core elevation at selected transient times from 20 to 140 s for core cell 4............................ ......E-5
) e l 1 i
FIGURES (CONT) E-5a. Average-rod cladding temperatures vs core elevation at selected transient times from 0 to 10 s for core cell 5........................................E-5 E-5b. Average-rod cladding temperatures vs core elevation at selected transient times from 20 to 140 s for core cell 5....................................E-6 E-6a. Average-rod cladding temperatures vs core elevation at selected transient times from 0 to 10 s for core cell 6........................................E-6 E-6b. Average-rod cladding temperatures vs core elevation at selected transient times from 20 to 140 s for core cell 6 ..................................E-7 , E-7a. Average-rod cladding temperatures vs core elevation at selected transient times from 0 to 10 s for core cell 7................................ ......E-7 E-7b. Average-rod cladding temperatures vs core elevation . at selected transient times from 20 to 140 s for core cell 7...................................E-8 E-8a. Average-rod cladding temperatures vs core elevation at selected transient times from 0 to 10 s for core cell 8... ................................E-8 E-8b. Average-rod cladding temperatures vs core elevation at selected transient times from 20 to 140 s for core cell 8................ ..................E-9 E-9a. Average-rod cladding temperatures vs core elevation at selected transient times from 0 to 10 s for core cell 9... .... ...... ........ ....E-9 E-96. Average-rod cladding temperatures vs core elevation at selected transient times from 20 to 140 s for core cell 9.. .............................E-10 E-10a. Average-rod cladding temperatures vs core elevation , at selected transient times from 0 to 10 s for core cell 10.... ..............................E-10 E-10b. Average-rod cladding temperatures vs core elevation .. at selected transient times from 20 to 140 s for core cell 10................................E-11 E-11a. Average-rod cladding temperatures v. core elevation at selected transient times from 0 to 10 s for core cell 11.. ................................E-11 E-11b. Average-rod cladding temperatures vs core elevation at selected transient times from 20 to 140 s for core cell 11........................ ......E-12 E-12a. Average-rod cladding temperatures vs core elevation at sc.lected transient times from 0 to 10 s for core cell 12 ................................E-12 E-12b. Average-rod cladding temperatures vs core elevation at selected transient times from 20 to 140 s for core cell 12................................E-13 E-13a. Average-rod cladding temperatures vs core elevation at selected transient times from 0 to 10 s for core cell 13.. ................................E-13 E-13b. Average-rod cladding temperatures vs core elevation i at selected transient times from 20 to 140 s for core cell 13................................E-14 i i
1e j l FIGURES (CONT) I E-14a. Average-rod cladding temperatures vs core elevation ; at selected transient times from 0 to 10 s for core cell 14....................................E-14 l E-14b. Average-rod cladding temperatures vs core elevation at selected transient times from 20 to 140 s for core cell 14................................E-15 l ] E-15a. Average-rod dadding temperatures vs core elevation at selected transient times from 0 to 10 s for core cell 15....................................E-15 i l } E-15b. Average-rod cladding temperatures vs core elevation l t at selected transient times from 20 to 140 s for core cell 15...... ........................E-16 l E-16a. Average-rod cladding temperatures vs core elevation __ j at selected transient times from 0 to 10 s for core cell 16............ ......................E-16 E-16b. Average-rod cladding temperatures vs core elevation at selected transient times from 20 to 140 s for core cell 16.... ..........................E-17 F-la. Hot-rod cladding temperatures vs core elevation i at selected transient times from 0 to 10 s for core cell 1........................................F-1 : F-Ib. Hot-rod cladding temperatures vs core elevation at selected transient times from 20 to 140 s for core cell 1....... .............. ..........F-2 F-2a. ' Hot-rod cladding temperatures vs core elevation at selected transient times from 0 to 10 s for core cell 2...................... .... .. ......F-2
.F-2b. Hot-rod cladding temperatures vs core elevation at selected transient times from 20 to 140 s for core cell 2...... ............................F-3 F-3a. Hot-rod cladding temperatures vs core elevation -
at selected transient times from 0 to 10 s for core cell 3.. ...... .................... ....F-3 F-3b. Hot-rod cladding temperatures vs core elevation at selected transient times from 20 to 140 s for core cell 3.... ............................F-4 I F-4a. Hot-rod cladding temperatures vs core elevation at selected transient times from 0 to 10 s for core cell 4.............. .......... .... ......F-4 F-4b. Hot-rod cladding temperatures vs core elevation at selected transient times from 20 to 140 s for core cell 4.. ........ ........... .........F-5 F-5a. Hot-rod cladding temperatures vs core elevation at selected transient times from 0 to 10 s for core cell 5.. ....... . .............. ......F 6 F-5b. Hot-rod cladding temperatures vs core elevation at selected transient times from 20 to 140 s for core cell 5....................................F-6 F-6a. Hot-rod cladding temperatures vs core elevation at selected transient times from 0 to 10 s for core cell 6........... ...........................F-7 F-6b. Hot-rod cladding temperatures vs core elevation at selected transient times from 20 to 140 s for core cell 6....................................F-7 F-7a. Hot-rod cladding temperatures vs core elevation - at selected transient times from 0 to 10 s for core cell 7........... ..........................F 8 4 m
4. l FIGURES (CONT) s a j F-7b. Hot-rod cladding temperatures vs core elevation
- at selected transient times from 20 to 140 s for core cell 7................ ..................F-8
{ F-8a. Hot-rod cladding temperatures vs core elevation at selected transient times from 0 to 10 s for core cell 8........................................F-9 i F-8b. Hot-rod cladding temperatures vs core elevation l at selected transient times from 20 to 140 s for core cell 8....................................F-9 F 9a. Hot-rod cladding temperatures vs core elevation j at selected transient times from 0 to 10 s for core cell 9......................................F-10 ' F-9b. Hot-rod cladding temperatures vs. core elevation at selected transient times from 20 to 140 s for core cell 9.. .. ........................F-10 F-10a. Hot-rod cladding temperatures vs core elevation at selected transient times from 0 to 10 s for core cell 10. .................................F-11 F-10b. Hot-rod cladding temperatures vs core elevation at selected transient times from 20 to 140 s for core cell 10.......... ....................F-11 F-11a. Hot-rod cladding temperatures vs core elevation at selected transient times from 0 to 10 s for core cell 11....................................F-12 F-11b. Hot-rod cladding temperatures vs core elevation at selected transient times from 20 to 140 s for core cell 11 ...... ...... ..............F-12 F-12. Hot-rod cladding temperatures vs core elevation at selected transient times from 0 to 10 s for core cell 12.... ........ ....................F-13 F-12b. Hot-rod cladding temperatures vs core elevation at selected transient times from 20 to 140 s for core cell 12.......... ....... ...........F-13 F-13a. Hot-rod cladding temperatures vs core elevation at selected transient times from 0 to 10 s for core cell 13................ ...... .. ......F-14 F-13b. Hot-rod cladding temperatures vs core elevation at selected transient times from 20 to 140 s for core cell 13............ .. ..............F-14 F-14a. Hot-rod cladding temperatures vs core elevation at selected transient times from 0 to 10 s for core cell 14 ........ .. .... ..............F-15 F-14b. Hot-rod cladding temperatures vs core elevation at selected transient times from 20 to 140 s for core cell 14...... ........................F-15 F-15a. Hot-rod cladding temperatures vs core elevation at selected transient times from 0 to 10 s for core cell 15.............. ....................F-16 F-15b. Hot-rod cladding temperatures vs core elevation at selected transient times from 20 to 140 s for core cell 15.. ............................F-16 F-16a. Hot-rod cladding temperatures vs core elevation at selected transient times from 0 to 10 s for core cell 16.......... .. ....................F-17 F-16b. Hot-rod cladding temperatures vs core elevation at selected transient times from 20 to 140 s for core cell 16 .. ...... ... ...........F-17 K
4 0 1 1 1 1 HGURES (CONT) ,'t is i G-1. Timestep size ... ....... . ...... .... ...... . ...... .... ..... ..... ... . ...... ......... ...... . . ... . . .... ........ . ... . G-3 G-2. Total CPU dme per calculation run ....................................................................... G-1 1 l i l 4 t l
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- o NOMENCLATURE i
! . ID One dimensional ! 3D Three dimensional i ACC Accumulator i ACC-A - Accumulator connected to broken direct vessel injection line A ] ACC-B Accumulator connected to intact direct vessel injection line B l ADS Automatic depressurization system CMT Core makeup tank i CMT-A Core makeup tank connected to direct vessel injection line A { CMT-B Core makeup tank connected to direct vessel injection line B l DEGB Double-ended guillotine break
- DVI Direct vesselinjection l DVI-A Direct vesselinjectionline A j DVI-B Dueet vesselinjectionline B 4
ECC Emergency core coolant ECCS Emergency core coolant system i IBLOCA Intermediate-break loss-of-coolant accident l IRWST In-containment refueling water storage tank
- JAERI Japan Atomic Energy Research Institute j LBLOCA Large-break loss-of-coolant accident 15fF large Scale Test Facility i NRC United States Nuclear Regulatory Commission l PBL-A Pressure balance line connected to cold-leg A -
j PBL-B Pressure balance line connected to cold-leg B
- PCCS Passive contamment cooling system j PRHRS Passive residual heat removal system j PSIS Passive safetyinjection system
- PWR Pressurized water reactor l RCP Reactor coolant pump RCS Reactor coolant system ROSA Rig of Safety Assessment, AP600 integral test facility, Japan Safeguards S
i SBLOCA Small-break loss-of-coolant accident 1 I
=
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] UPDATED TRAC ANALYSIS OF AN 80% DOUBLE-ENDED COLD-LEG BREAK
- FOR THE AP600 DESIGN *
! J. F. Lime and B. E. Boyack Teclinology and Safety Assessment Division l Los Alamos National Laboratory j Los Alamos, New Mexico 87545 i EXECUTIVE
SUMMARY
l The AP600 is an advanced passive 600-MWe reactor design being developed by j Westinghouse in conjunction with the US Department of Energy's Advanced Light i Water Reactor Technology Program. The AP600 has been submitted for Nuclear i Regulatory Commission (NRC) design certification. In accordance with 10CFR52.47 for i design certification, advanced reactor applicants are requirci to submit neutronic and
- thermal-hydraulic safety analyses over a sufficient range of normal operation, transient l conditions, and specified accident sequences. Review and confirmation of these i analyses for the AP600 design constitute an important activity in the NRC's review for
- design certification. In support of its design certification review, the NRC is using best-estimate thermal-hydraulic codes to perform audit calculations. The NRC is using TRAC-PF1/ MOD 2 as its primary tool for confirmatory safety analyses of large-break !
loss-of-coolant accidents (LBLOCA). This reports documents the results and analyses of l 4 a multidimensional TRAC-PF1/ MOD 2 calculation of an updated AP600 LBLOCA, an
- 80% double-ended guillotine cold-leg break next to the reactor-coolant pump. This '
i analysis was performed with an AP600 plant model updated for model corrections and
- plant design changes, and supersedes an earlier TRAC AP600 LBLOCA analysis.1 1 \
The TRAC model used for the LBLOCA calculation is finely noded. The reactor vessel l is modeled with two multidimensional vessel components in order to preserve the !. AP600 vessel geometry. The first vessel component models the lower plenum, core, l core bypass, reflector, upper plenum, and upper head. The second models the j downcomer. One-dimensional (ID) components are used to model the remainder l of the reactor coolant system. The passive safety systems are also modeled. The j accumulators and core makeup tanks (CMT) are modeled with ID components. J Because this analysis focused on the first few minutes of a LBLOCA, models of the
- aptomatic depressurization system (ADS), passive residual heat removal system l (PRHRS), and in-reactor refueling water storage tank (IRWST) were not included j because they are not activated during the interval of the calculation. The containment is
! modeled as a constant back pressure of one atmosphere. l 1 )
- This work was funded by the US Nuclear Regulatory Commission's Offu:e of Nuclear Regulatory Research 1 1 a
i The TRAC AP600 plant model was updated for plant design changes and for model corrections identified from a peer review of the TRAC plant model at INEL on May 26-
- 27,1994. The plant design changes included deletion of the pressurizer pressure i
balance lines to the core makeup tanks, adding a venturi to the direct vessel injection
- (DVI) nozzle, and piping geometry changes in the CMT and accumulator injection lines.
1 The modeling corrections included the modeling of counter-current flow limitation l (CCFL) in the upper core plate, reactor power deposition to the reactor coolant flow, l core azimuthal and radial power distibution, and other changes. In addition, the break i location was changed from next to the vessel downcomer to between the reactor coolant
- pump and the CMT pressure balance line connection, to match the ECOBRA/ TRAC l break location.
i i The TRAC-PF1/ MOD 2 code version used was interim version 5.4.04. There were major code error fixes in this code version. Major code corrections include a blowdown heat t transfer error fix, tee-component momentum source fix, and mass loss error fix, and a downcomer interfacial drag. The transient thermal-hydraulic behavior of an LBLOCA can be characterized in major phases:
- 1. Blowdown Period,0 to 12.3 s. During this phase, there is a rapid system depressurization and a high rate of flow out the break. By 6 s the core is over 90% voided. The rapid loss of coolant causes fuel cladding temperatures to increase rapidly; they reach a blowdown maximum at 7 s and then decrease slightly from an increase in cooling. The maximum-power fuel-rod peak clad temperature during this period reached 1021 K (1378'F). Increased cooling is the result of a decrease in net mass-flow loss out of the reactor vessel when two-phase flow occurs at the bmak exits.
- 2. Refill Period,12.3 to 43 s. The system pressure decreases below 700 psia and the accumulators begin to discharge. The vessel net mass-flow loss decreases and by 20 s more flow begins to enter the vessel than is leaving. Fuel cladding temperatures, however, increase gradually as the core, upper plenum, and upper head regions continue to void and core cooling becomes less effective. In the latter phase of this period, the primary has stopped depressurizmg and the break flow is essentially all vapor. The lower plenum starts to refill from accumulator flow injection. Fuel temperatures continue to increase.
- 3. Reflood,43 s to 240 s The lower plenum has refilled and core reflood and quenching begins. The maximum-power fuel rod peak clad temperatures reach a maximum of 1197 K (1695'F) at 58 s and then decreased gradually thereafter. The average-power fuel rods are quenched by 165 s, and the maximum-power fuel rods are quenched by 240 s. The accumulators are completely empty of liquid at the same time that the maximum-power fuel rods are quenched. .
2
l t j l
- 4. Long-term cooling,240 s and beyond. There is a short period,240 to 265 s, during which there is no emergency core-coolant flow. The accumulators 3
have completely emptied; however, the core makeup tanks and the IRWST l have not yet started to inject because the pressure in the DVI line is still too
- high to allow the check valves to open. The pressure finally is low enough to
! all the core makeup tanks that they can drain at 265 s. The IRWST drain line j check valves finally open at ~275 s. The core, still in a state of' saturated j nucleate boiling at the beginning of this period, eventually will revert to
- single-phase natural-convective cooling. There is sufficient water in the IRWST to cover the hot and cold legs completely and form a natural-
{' circulation path between the containment and the reactor core. _.__ ! The TRAC calculation results are compared with the Westinghouse ECOBRA/ TRAC l' results presented in the AP600 Standard Safety Analysis Report (SSAR).10 The l comparison shows reasonable agreement between the two calculations. The system depressurization was about the same in both calculations. The TRAC accumulator
- flows were higher than the ECOBRA/ TRAC flows, so refill and reflood event times are earlier than ECOBRA/ TRAC event times. We believe that the accumulators in the ECOBRA/ TRAC analysis were calibrated to a lower flow than the accumulators in the
! TRAC. analysis. ) A significant similarity between the two calculations is in the calculation of the hot-rod j peak cladding temperature (PCT)..Both the TRAC and ECOBRA/ TRAC calculations
- show the reflood PCT to be higher than the blowdown PCT. The TRAC and j ECOBRA/ TRAC calculated reflood PCTs compare within 50 K, although the TRAC-l calculated peak occurred much earlier in the transient. The reason for the earlier PCT is
! being investigated. The TRAC-calculated hot-rod PCT is lower than the App. K limit of 2200*F, but near the cladding oxidation temperature of 1700*F. Should the calculated ! PCT exceed 1700*F, the TRAC-calculated PCT would no longer be accurate because TRAC does not account for oxidation effects. l l A major difference between the two calculations is in the cooling of average-powered
- rods during the late blowdown period. The ECOBRA/ TRAC calculation showed top-i down rewetting and quenching of average-powered rods during the late-blowdown j phase, with cladding temperatures being cooled down to saturation-temperature levels.
l The TRAC calculation showed considerable radial and azimuthal variation in fuel l cladding temperatures. Fuel rods near the break experienced less top-down quenching l and higher cladding temperatures than the fuel rods opposite and away from the break.
- In addition, the fuel rods in radial ring i experienced less top-down quenching than the i
fuel rods in the outer radial ring 2, a consequence of CCFL. In general, the degree of i top-down quenching and rewetting in the TRAC calculation was not as high as in the
- ECOBRA/ TRAC calculation. This may be because the TRAC reflood model does not -
j account for quenching resulting from a falling liquid film. j 3 i t
. - . . - . - - . ~ - - . _ _ _ . - . .- . . . . - .
4 i UPDATED TRAC ANALYSIS OF AN 80% DOUBLE-ENDED COLD-LEG BREAK j FOR THE AP600 DESIGN' i J. F. Lime and B. E. Boyack i 2 ABSTRACT An updated TRAC 80% large-break loss-of-coolant accident (LBLOCA) l! has been calculated for the Westinghouse AP600 advanced reactor design, , The updated calculation incorporates major code error corrections, model j corrections, and plant design changes. The 80% break size was calculated i by Westinghouse to be the most severe large-break size for the AP600 i design. The LBLOCA transient was calculated to 280 s, which was the j time of IWRST injection. All fuel rods were quenched completely by 240 s. i Peak cladding temperatures (PCTs) were well below the App. K limit of l 1478 K (2200*F) but very near the cladding oxidation temperature of 1200 i K (1700*F). Transient event times and PCT for the TRAC calculation were j in reasonable agreement with those calculated by Westinghouse using theirWCOBRA/ TRAC code. However, there were significant differences in the detailed phenomena calculated by the two codes, particularly
- during the blowdown phase. The reasons for these differences are still l being investigated. Additional break sizes and break locations need to be j analyzed to confirm the most severe break postulated by Westinghouse.
t i l i l 1.0. INTRODUCTION - The AP600 is an advanced passive 600-MWe reactor design being developed by Westinghouse in conjunction with the US Department of Energy Advanced Light Water Reactor Technology Program. The AP600 has been submitted for Nuclear Regulatory i Commission (NRC) design certification. In accordance with 10CFR52.47 for design i certification, advanced reactor applicants are required to submit neutronic and thermal-l hydraulic safety analyses over a sufficient range of normal operation, transient j conditions, and specified accident sequences. Review and confirmation of these ! analyses for the AP600 design constitute an important activity in the NRC's review for j design certification. In the process of design certification, the NRC will use best-l estimate, thermal-hydraulic codes to perform audit calculations. The best-estimate code
- selected by the NRC for analyzing large-break loss-of-coolant accident (LBLOCA) transients is TRAC-PF1/ MOD 2,2 developed by Los Alamos National Laboratory. Los j Alamos was requested by the NRC to perform LBLOCA analyses with TRAC in support
! of the design certification review of the AP600.
i
- This work was funded by the US Nuclear Regulatory Commission's Office of Nuclear Regulatory Research.
4 ), l
Q 2.0. PLANT DESCRIPTION A brief description of the AP600 system is provided with a review of those systems and components having the greatest impact on the course of the LBLOCA during the blowdown, refill, and reflood periods of the transient. Similarities and differences between the AP600 and the Westinghouse four-loop PWR safety systems and their impact on the LBLOCA are discussed. 2.1. AP600 Description The AP600 is an advanced passive 600-MWe reactor design being developed by Westinghouse in conjunction with the US Department of Energy Advanced Light Water Reactor Technology Program. The AP600 is a two-loop design. Each loop contains one l hot leg (HL), one steam generator (SG), two reactor coolant pumps (RCP), and two cold legs (CL). A pressurizer is attached to one of the hot legs. The reactor coolant pumps are a canned-motor design and are attached directly to the steam generator. The loop seal is eliminated an added safety feature in that core uncovery caused by the existence of water-filled loop seals is eliminated during a postulated small-break (SB) LOCA. The core is designed for a low-power-density and consists of 145 fuel assemblies with an active fuel length of 12 ft. The fuel assembly is a 17 x 17 array of fuel and control rods. The AP600 incorporates passive safety systems that rely only on redundant / fail-safe valving, gravity, natural circulation, and compressed gas. There are no pumps, diesels, or other active machinery in these safety systems. During plant shutdown, all the passive safety features will be tested to demonstrate system readiness, flow, and_ heat removal performance. These systems are shown in an isometric cutaway view of the AP600 reactor design in Fig.1, containment cutaway in Fig. 2, and in a schematic _ diagram in Fig. 3. Two Passive Safety injection System (PSIS) trains, each with an accumulator (ACC), a Core Makeup Tank (CMT), and an injection line from the In-containment Refueling Water Storage Tank (IRWST) and sump are connected directly to the reactor-vessel downcomer via a direct vessel injection (DVI) line. Depressurization of the primary system is an essential process that is required to ensure long-term cooling of the AP600. For example, the accumulators inject coolant into the reactor coolant system (RCS) only after the primary pressure has dropped to 700 psia. Coolant injection from large, safety-class water pools, specifically the IRWST and sump, can occur only after the reactor coolant system pressure decreases below the gravitational head of each pool. An Automatic Depressurization System (ADS) permits a controlled pressure reduction of the RCS. The ADS has four stages. Each of the first three stages consists of two trains providing redundant flow paths between the top of the pressurizer and the IRWST. The coolant discharged to the IRWST is condensed and accumulated for later injection into the RCS. The actuation signal for first stage ADS is a reduction in the coolant inventory of one CMT to 67% of its initial value. ADS stage 2 is actuated 70 s after stage 1 actuation. ADS stage 3 is actuated 120 s after stage 2 actuation. The actuation signal for fourth-stage ADS is a reduction in the inventory of one CMT to 20% of its initial value and an interval of 120 s following third-stage ADS 3 actuation. The fourth stage ADS consists of two trains, one train connecting the top of , 5
ia i 4 the pressurizer hot leg (loop 1) and the containment, and the other train connecting the i loop-2 hot leg and the containment. A direct discharge path to the containment is i needed to ensure that th PCS pressure will equilibrate with the containment pressure l . so that the head-driven IRWST injection can proceed. The fractions of the total ADS j discharge area for ADS stages 1--4 are 0.038,0.171,0.171, and 0.62, respectively. After the accumulators and CMTs are depleted and the primary system has depressurized , and approached the containment pressure, water injection is provided from the IRWST. l This tank empties after several days. Provisions are also made for recirculating coolant ! from a sump. IRWST and sump recirculation may occur at the same time for some 1 < transients. ! The AP600 containment plays an essential role in the long-term cooling of the primary l via the Passive Containment Cooling System (PCCS). Steam entering the containment, ! either through a break in the primary or through operation of the ADS, condenses on i the inside of the steel containment shell. The condensate drains downward and a large i fraction is delivered via gutters to either the IRWST or the sump. Heat transfer on the outside of the containment steel shell is by evaporation of liquid sprayed near the top of l the steel reactor containment dome by the PCCS, and by convection to an air stream ) induced by buoyancy-driven flow (unforced). This air steam enters a high-elevation i inlet, flows downward to an elevation near the bottom of the cylindrical portion of the j steel reactor containment structure, passes upward through the annular gap between ; i the steel reactor containment structure and the concrete shield building, and is l ! exhausted to the atmosphere near the top of the concrete shield building. The PCCS spray inventory is eventually depleted. However, by the time the PCCS water supply is j depleted, the decay heat has decreased sufficiently so that the buoyancy induced air - j flow through the air gap between the steel containment structure and the concrete l shield building can remove the core decay heat. For non-LOCA accidents, long-term heat removal is provided by a Passive Residual Heat Removal System (PRHRS) that removes core heat thmugh natural circulation. The
- PRHRS receives water from the top of the het leg to which the pressurizer is connected.
The single PRHRS line connected to the hat leg divides into two lines, each feeding one of the two PRHRS heat exchangers residing in the IRWST. The two PRHRS discharge
- lir.es then rejoin and connect to the outlet plenum of the steam generator outlet in the i same loop. When functioning as the heat sink for the PRHRS heat exchangers, the
- IRWST has sufficient water volume to remove decay heat for two hours before the l inventory reaches saturation temperature. Isolation valves on the PRHRS lines open l upon receipt of the safeguards (S) signal, and a buoyancy induced flow transports
! primary coolant through the PRHRS. The PRHRS may also contribute to the removal of i energy from the primary system during LOCA events. However, operation of the i PRHRS is interrupted when the PRHRS inlet void fraction becomes very high,
- degrading the energy transport to the PRHRS heat exchangers. Thus, the PRHRS is
! ineffective for LBLOCAs, has a limited interval of effectiveness for intermediate-break l LOCAs, and has an extended period of effectiveness for SBLOCAs. The coolant in the i PRHRS lines and heat exchangers will, however, contribute to the total makeup flow during a LOCA . i 6 4
e Westinghouse has changed several features of the AP600 design since it was submitted for design certification. Although only the final design is of interest when assessing safety,it is important to assure that the design information used in performing various safety assessments reflects either the current or final design. In performing the assessments presented in this document, we reflect the design features known to us as of November 15,1994. 2.2. Key AP600 Features - LBLOCA Response The selected scenario is an 80% break in a single cold leg between the RCP and the nozzle to which the cold-leg pressure balance line (PBL) is connected. For this PWR LBLOCA scenario, a large amount of primary coolant is discharged through the two-sided break, the core completely fills with vapor, the fuel rod cladding temperatures rapidly increase, emergeacy core coolant (ECC) is injected, and the core is quenched. This sequence of events occurs within the first five min following the break initiation. From break initiation through the time the entire core is quenched, the accumulators and core makeup tanks am the key and dominant components ensuring core cooling. After the accumulators have emptied and depressunzed completely, the core makeup tanks start to inject along with IRWST injection. Although the ADS will be initiated on j CMT level, they serve no purpose in depressurizing the system. 2.3. Comparison with Current Generation Westinghouse PWRs Several design differences do enter into the early phases of the LBLOCA scenarios for . i the two plants. For example, in the AP600, ECC is injected directly into the downcomer through the two DVI lines of the PSIS. Thus a cold-leg break does not directly result in the loss of ECC from either accumulator. In the earlier Westinghouse designs, each of l the four accumulator injection lines are connected to a different cold leg. Thus, the I entire inventory of one accumulator is lost through the break and is unavailable for core cooling. In the AP600, a turning vane directs the flow from each DVIline downward into the downcomer, thereby affecting the amount of ECC bypass. The average linear power in the AP600 core is ~70% of that in the four-loop Westinghouse PWR. One ! manifestation of the lower power density core is that the internal diameter of the AP600 is fully 90% of that in the four-loop Westinghouse PWR and the internal vessel cross- i sectional area is -82%. The two accumulators contain about the came volume of borated water as three accumulators in the Westinghouse four-loop plant. 3.0. TRAC MODEL DESCRIPTION The TRAC model of the AP600 is a finely noded, multidimensional model with 161 hydrodynamic components (752 3-D and 8681-D computational fluid cells) and 45 heat-structure components in the model. The LBLOCA plant model has undergone an independent quality-assurance check. 3.L Plant Model Database The TRAC AP600 LBLOCA plant model reflected as accurately as possible the AP600 design based on the design information we had at the time the analysis was performed. Where design information was lacking, modeling assumptions had to be made or the 7
- i
- a 1
- \
l l Component TRAC Model Database Reactor Vessel i Vesselinternal geometry Westinghouse AP600 drawings.
- Fuel-bundle geometry Westinghouse AP600 drawings.
i Fuel-rod reactivity coefficients Assumption. Used USPWR plant model 4 { Vessel flow and internal resistances Westinghouse. Reactor Coolant System ! -Hot- and cold-leg piping Westinghouse AP600 drawings. s Reactor coolant pumps Westinghouse rated vaines and l COBRA / TRAC homologous pump l curves. __ j Steam generator Intemal geometry and thermal-i hydraulic data-Westinghouse ! . computer output. l ~i'ressunzer and pressunzer Westinghouse AP600 drawings. l surgeline j Passive Safety injection System i j CMT and accumulator tank Westinghouse dimensions and geun e f volumes. l
- Injectionline piping Westinghouse reduced-size AP600 drawings. Dimensions approximated for non-legible values.
l CMT pressure balance 1mes Westinghouse AP600 drawings. l Check valves 0.4 psi to open, based on i Westinghouse in-situ check valve tests,
-0.4 psi to close, assumed l Flow resistances Westmghouse, design-basis flows and l resistances
- Control systems j Trips, set points, and delay times Information provided by Westinghouse for some trips but not Pressurizer pressure and level all.
- l. control These and other plant control systems adapted from a TRAC modelof a Steam generator level control Westinghouse three-loop plant.
modeling from a RELAP5 AP600 plant mode 13 had to be used. A summary of the database sources for the TRAC plant model is given below: 3.2. Reactor Vessel The reactor vessel is modeled in three-dimensional (3D) cylindrical coordinates, with 4 radial rings,8 azimuthal sectors, and 17 axial levels. An isometric view of the reactor vessel model is shown in Fig. 4. Elevation and plan views of the reactor vessel model , 8 i _ . _ . ~ . _ _
iO l i l are also shown in Figs. 5 and 6, respectively. Two TRAC vessel components are needed j to model the reactor vessel in order to preserve the elevations of the hot-leg, cold-leg, ! and PSIS connections to the vessel. Otherwise, there would have to be a compromise on i modeling the vessel true geometry. The elevation differences of these piping ! connections to the reactor vessel are clearly seen in the elevation view of Fig. 5. I The first vessel component, component 10, models the lower plenum, core, upper j plenum, and upper head. The core region is modeled with the first two radial rings. 1 The third radial ring models the reflector region. The fourth radial ring is used in the l
- lower plenum and in the upper head but in the axial levels modeling the core and upper l l plenum it is not used. The second vessel component, component 20, models the j downcomer annulus and is noded into two radial rings, eight azimuthal sectors, and 13
- axial levels. Radial ring 1 is a noncomputational ring and is used only to mo-del the
! annular geometry of the downcomer. Where possible, the same axial-level noding ;
- heights are used in both vessel components. The two vessel components are connected I
- together by short one-dimensional (1D) pipe components. Other ID pipe components
! are used to model the fuel-assembly guide tubes, upper-head cooling spray flow, core i bypass flow, and reflector-block cavity fluid regions. A total of 33 TRAC heat structures components are used to model the core and reactor { vessel structure, as shown in Fig. 7. There is heat conduction in the core barrel between j the downcomer annulus and the core region. Outer heat-structure surfaces are treated ! adiabatically. The fuel rods are modeled with one powered heat structure. The fuel i j rods are combined and modeled as 16 lumped assemblies each with the same number l ! of fuel rods, one for each of the active core sectors shown in Fig. 6. Supplemental hot-i rod components are also modeled that represent the maximum-power fuel rods. The , j core decay power is calculated using reactivity feedback coefficients. Control-rod and ! l reactivity feedback coefficients are from a 15 x 15-fuel large-break US/ Japanese PWR
- model4 because we did not have sufficient reactivity information for the AP600 design j at the time the TRAC model was developed. The reflector-block heat structum models
! 0.25% of the total core power during steady state. Power deposition of 2.35% to the com l coolant flow was modeled with a very thin artificial heat structure that was coupled to { each of the core hydro cells. When the reactor is scrammed, the reflector-block and j coolant-deposited power are reduced to zero. 3.3. Loop Components Figures 8 and 9 show a modeling overview of reactor coolant loops. Loop 1 is modsled with 33 hydro components and 181 17 computational cells. The pressurizer is connected to the hot leg of loop 1. The PRHRS is also connected to loop 1, with the PRHRS inlet line being connected to the hot leg of loop 1 and the PRHRS return line being connected to the steam generator outlet plenum (see Fig.1). Loop 2 has 32 components and 1621D cells, and is similar in noding to loop 1 except that there is no pressurizer, and instead of pressurizer spray sources, core makeup tank pressure balance lines connect to the cold legs. Loop 2 also models the broken cold leg,_which is modeled with a series of TEE and VALVE components that allow for a more mechanistic modeling of a large break. 9
ie i j The reactor-coolant pump homologous performance curves are from the Westinghouse
- Cobra / TRAC input. The steam generator model reflects the AP600 A75 design, which is similar to the current Westinghouse Model F design, but has a taller tube bundle and j redesigned secondary-side risers and separators. Where specific component design data were not available ibr the AP600, components from a Westinghouse three-loop plant were used. All external piping wall structures .are modeled with an outer
! adiabatic boundary assumed. i j 3.4. Passive Safety Systems
- Figure 10 shows a modeling overview of the AP600 passive safety systems, the PSIS, l ADS, PRHRS, and IRWST. There are two separate trains in each of the systems and i each of the trains are modeled separately. The TRAC LBLOCA plant model includes
. the components of the PSIS, the accumulators, core makeup tanks, pressure balance i lines, and injection lines, as these are the only safety components activated in the first , few minutes of a LBLOCA transient. Also included in the LBLOCA plant model are the ! PRHRS primary-side components, to account for its liquid inventory. In the latest ! AP600 plant design, the PRHRS isolation valve will open on an "S" signal, which will
- allow the PRHRS fluid to drain into the primary coolant system. The other passive
! safety systems, the ADS and IRWST, will be modeled for longer-term transients such as ! for intermediate- and small-break LOCAs, where the transient is calculated out to when i the IRWST drains into the reactor coolant system (RCS). i j Both PSIS trains are orificed to delivered the same amount of flow even though the j piping geometry of the two trains are different. Separate stand-alone cases of both PSIS q trains were used to calibrate the accumulator and CMT flows to Westinghouse specified
- values.
t j 3.5. Updated Plant Model Changes l The plant model was upd.ated for a number of design changes and model corrections. I An AP600 Working Group Meeting was held at the Idaho National Engineering l Laboratory to review the TRAC plant model and TRAC LOCA calculations with the i NRC and Idaho National Engineering laboratory staff.5 The plant model peer review identified a number of modeling changes that were incorporated into the plant model. ! These design and modeling changes are noted as follows: t i 1. DVI Injection Nozzle Venturi. A venturi was modeled in the DVI injection ! nozzle based on an AP600 design change. l 2. Pressure Balance Line from Pressurizer. The pressure balance line from the ! pressurizer to the core makeup tank was removed based on an AP600 design ! change
- 3. Feedwater and Steam Lines. The feedwater and steam lines were modeled
- from AP600 drawings received at the INEL peer review meeting. _
i 1 10 1
'O
- 4. DVI Line Geometry Changes. Westinghouse changed the DVIline A to avoid piping stress problems. The previous piping arrangement had core makeup tank flow and accumulator flow coming together from opposite ends of the main pipe of a tee and then flowing through the side section of the tee to another tee before flowing into the main DVI line. This was changed so that the accumulator flow discharged through the main pipe and the CMT flow was from the side pipe.
- 5. Orificing in the PSIS Lines. The CMT, ACC, and IRWST lines are each orificed to deliver a specified flow at a given net driving head. The orifice locations are shown in the noding diagrams. We assumed the orifice to be a sharp-edge orifice, modeling the actual orifice area and applying a theoretical irreversible pressure loss coefficient for a sharp-edge orifice. Stand-alone flow calibration cases wem used to adjust the orifice area and corresponding loss coefficient. The orifice area ratio used for each flow are noted in the modeling diagrams
- 6. Power deposition to reactor coolant. The Westinghouse database specified that 97.4% of the total reactor power is deposited in the fuel rods. The remaining power was assumed to be deposited into the reflector block and core coolant . The non-fuel power deposition was corrected so that 0.25%
power was deposited in the reflector block and the remaining 2.35% was deposited into the moderator (coolant). The moderator deposited power was modeled with a very thin slab powered heat structure that coupled to_each com sector fluid cell. The very thin slab allows a very rapid power deposition and transient response to the core coolant. -
- 7. Core azimuthal power distribution. Westinghouse provided a fuel-bundle power distribution for a 1/8 core sector. This sector is offset 22.5* from the hydro sector. In modeling the average power distribution and accounting for the offset, alternate sectors will have slightly more power. The core sectors corresponding to the cold leg sectors will have slightly less power than the core sectors to the hot legs and DVI injection line.
- 8. Counter-Current Limiting Flow (CCFL). The TRAC CCFL modeling option was added to the AP600 plant model at several places, in the reactor vessel upper core plate, in the hot leg inclination, in the pressurizer surgeline at the 90' elbow bend, and in the pressurizer heater upper support plate.
- 9. Check Valve Pressure Differential. All plant check valves were modeled to open on a 0.40 psi pressure differential across the valve, based on in-situ check valve testing by Westinghouse. We assumed that the valves would close on a -0.40 psi pressure differential.
- 10. Cold Leg Break Location. The break location was changed from next to the downcomer to between the reactor coolant pump and the pressure balance 11 l
D line connection to conform with the break location assumed by
- Westinghouse in their ECOBRA/ TRAC LBLOCA analyses.
- 11. Reflector Block Cavity Fluid. Four pipe components were added to the plant
{ model to account for the reflector block cavity fluid. These pipe components j were connected to the core by assuming the very thin gaps between reflector blocks formed a flow path. A flow resistance based on parallel flat plates ! gaps was assumed for the leakage paths between the cavity region and the l mm. i 4 4.0. TRAC CODE DESCRIPTION i
- Analysis was performed with an interim version TRAC-PF1/ MOD 2, version 5.4.04.
TRAC-PF1/ MOD 2, Ref.2,is the latest TRAC version to be developed by Los Alamos
. National Laboratory. TRAC was developed to provide advanced, best-estimate
! predictions for postulated accidents in pressurized-water reactors. The fluid-dynamics solution is formulated in the six equation, two-fluid, nonequilibrium model with a i staggered difference scheme. Multidimensional flow capability is available in the vessel ! component, either in cylindrical or Cartesian coordinates. One-dimensional flow ] capability is available in the other hydrodynamic modeling components. l TRAC-PF1/ MOD 2 features a 3D stability-enhancing two-step method, which remove.4 the Courant time-step limit within the vessel solution. There are also internal code-i structure vectorization changes to improve computational times. MOD 2 also includes improvements to the interfacial drag and heat transfer constitutive relationships for stratified flow and air-water mixtures, a more efficient matrix solution method, and - ! conservation of momentum flux. Other major features of TRAC include: (1) a two-
- phase, two-fluid, nonequilibrium hydrodynamics model with a noncondensable gas i
field; (2) flow-regime-dependent constitutive equations; (3) either a 1D or 3D reactor-vessel model;(4) complete control-systems modeling capability; and (5) a generalized j heat-structure component. i ! A major modeling option in MOD 2, exercised in the LBLOCA analysis, is an improved ! mechanistic reflood model based on the inverted annular flow regime mapping of Dejarlais and Ishii.6 This reflood model and its assessment against experimental data are reported in Ref. 2 and also in Refs. 7--9. Additional reflood model development { efforts have been identified, such as the modeling of top-down quenching from a falling j liquid film and applicability of the reflood model to high-power applications. l Another modeling option exercised in the LBLOCA is the counter-current flow limit j option. The Bankoff CCFL correlation was applied at the core upper plate and at the pressurizer heater upper support plate. CCFL based on a Wallis diameter-dependence scaling was applied at the hot-leg inclinations and at the vertical elbow in the i pressurizer surge line. 1 - i i 12 j
1 0 l I There were major code errors and problems corrected in version 5.4.04. These included i
- correctmg for asymmetric flow problems in TEE components, a post-CHF heat transfer error that significantly affected the blowdown heat transfer, and total mass-loss error, i l --
- 5.0. LBLOCA CALCULATION RESULTS i
- The limiting LBLOCA has been identified as an 80% DEGB in a single cold-leg pipe l between the primary coolant pump and the connecting point for the CMT PBL to the j cold leg.10 Westinghouse has concluded that this hypothesized break will produce the j maximum fuel rod cladding temperature. Table I gives the sequence of events that j occur for the TRAC calculation of this limiting LBLOCA.
l To facilitate analysis, the LBLOCA calculation is subdivided into four time periods that l characterize events during the sequence. The format and period definitions of the
- LBLOCA in AP600 closely follow those of Ref.11. These time periods, termed l blowdown, refill, reflood, and long-term cooling are defined by the start of accumulator i injection and by the core and lower plenum liquid mass fraction behaviors (Fig.11).
- The first two periods and most of the third period have been calculated. Therefore, only j the first three periods will be discussed in detail in this report.
I The blowdown period is the result of a break in the coolant system through which the l pnmary coolant is expelled. Blowdown physical processes / phenomena include critical j flow at the break, fluid flashing and depressunzation, redistribution of fuel-rod stored
- energy, and heating of the fuel rods due to degraded heat transfer. During blowdown, i some components are affected more than others. In particular, the heat removal from
! the core results from the changing flow and heat transfer regimes in the core. The j performance of the primary coolant pumps degrade as the coolant flashes. The steam ! generator heat transfer degrades after the steam-generator secondaries are isolated. The i blowdown period ends when the intact-loop accumulator injection is initiated.11 ; \ i During the refill period, the reactor system starts to recover as the PSIS components i (CMTs and accumulators) start to inject coolant into the primary system. The important j refill components and processes / phenomena concern the introduction of water into the j reactor vessel downcomer and its subsequent distribution. Refill physical
- processes / phenomena are the operation of the PSIS, including interactions between the accumulators and CMTs, bypassing injected water through the downcomer to the l broken cold leg, and penetration of safety injection water into the lower plenum. The j refill period ends when the mixture level in the lower plenum approaches the core inlet, j and conditions are established for reflooding the core with coolant.
l When sufficient water from the PSIS has entered the lower plenum and the core liquid j inventory enters a period of sustained recovery, the reflood period is in progress and the j hot fuel rods are being recovered with water. The reflood process is highly oscillatory i but the overall trend with increasing time is increasing core coolant inventory, i.e., a
! sustained recovery. The reflood processes may be quite slow because much of the ! water is boiled and transported as steam and entrained droplets into the upper plenum 1
i 13 l.
- -- - - r - - -- ,- , , - . . - - - . - - , - ---
o ,
\
J { and hot-leg piping. The reflood period ends when the entire core is quenched, that is, l all fuel rod cladding temperatures am at or below the coolant saturation temperature. i ! The long-term cooling period continues after total core quench. At the time the fuel rod 3 cladding is completely quenched, the core is only partially full. The accumulators also j have just emptied of liquid. However, the noncondensable gas in the accumulator continues to discharge into the DVI lines, which keeps the DVI line pressure up and .
- delays CMT draining for ~30 s. Finally, the CMTs resume draining their inventory into l i the primary. A few seconds later, IRWST injection is initiated when the primary ,
l pressure decreases to a levelless than the static head in the IRWST. CMT and IRWST l
- draining may occur simultaneously because both are gravity-head driven. Draining of l the IRWST is expected to take several days, after which water from the sump is 4
recirculated indefinitely. For many accident scenarios, the depressurization-process
- must be assisted by operation of the ADS. The ADS is actuated when the liquid volume i j of one CMT is reduced to 67% and the ADS valve-opening sequence is initiated.
- However, the LBLOCA produces sufficient area to depressurize the primary, even in the absence of ADS actuation.
l For the TRAC calculation, the containment back pressure was specified to be a constant one atmosphere. The scenario description is supported by Figs.11 through 60 that display significant calculational results. In some cases, the figure legends identify , components with alphabetic or numeric descriptors, for example, DVI line A or CMT B. l Figure 3, a schematic of the AP600 and its passive safety systems, shows the i relationships between the plant components and the TRAC input model labels. With i reference to Fig. 3, loop I contains hot leg 1 (HL 1) and cold legs 1A and IB (CL 1A and i IB). The pressunzer is connected to HL 1. The PRHRS is connected between HL 1 and l the outlet plenum of steam generator 1 (SG 1). One train of the fourth stage ADS l attaches to the PRHRS inlet line which is in turn connected to HL 1. Loop 2 contains ! HL 2 and CL 2A and 2B. The second train of the fourth-stage ADS connects directly to l' . HL 2. The PSIS has two trains which feed coolant directly to the downcomer through 3 DVIlines A and B. Connected to DVIline A are CMT A, ACC A, and a IRWST drain line. Connected to.DVI line B are CMT B, ACC B and a second IRWST drain line. j DVI A is connected to CL 2A by PBL A. DVI B is connected to CL 2B by PBL B. The postulated 80% DEGB occurs in CL 2B between the PBL B connection to CL 2B and i RCP 2B.
- 5.1. Blowdown Period (0 to 12.3 s)
- Overview j The AP600 LBLOCA blowdown starts at the time the break opens, and ends when intact
- loop accumulator injection is initiated, a period of ~12.3 s. The initial break mass flows j are high, reflectmg subcooled critical flows at the break planes. The mass flow from the j vessel side of the break is much larger than that from the pump side because of the high j hydraulic resistances of the pump and the steam generator piping. Because coolant j flows from the vessel to the break through both the hot and cold legs, core flow rapidly
- stagnates; the flow at the core inlet reverses shortly after the break occurs. Early in the i
i 14 l - 4
o 4 transient, flow from the intact loops is generally bypassed around the reactor vessel downcomer to the broken cold leg and out the break. As the primary coolant system rapidly depressurizes (Fig.12), flashing occurs first in the highest temperature parts of the system, starting in the pressurizer and rapidly progressing through the upper plenum, hot legs, and then proceeding through the core and the steam generator and finally the lower plenum, downcomer, and cold legs. The pressurizer starts flashing earliest, but because of the resistance to flow out of the surge line, it takes almost 10 s to empty. The highest pressure in the primary is in the pressurizer until 7 s (Fig.12). The flow from the pressurizer also contributes to the flow of upper plenum liquid into the top of the core. The steam volume addition rate from flashing reduces the primary coolant depressurization rate in the higher temperature
- portions of the primary system. Flashing and nucleate boiling begin in thrcore.
Because of the core radial power profile, voiding in the core is nonuniform. The core approaches a vapor-filled condition within 2 s and the accompanying reactor kinetic effects produce a core power reduction. The reactor is tripped at -1 s on either low
- primary pressure (pressurizer) or containment high pressure. The fuel rod cladding i temperatures increase rapidly because of the degrading rod-to-fluid heat transfer
- (Fig.13)a.
The pressure in the broken cold leg decreases rapidly and to a lower level than in the hotter parts of the primary system where the rate of pressure decrease is reduced by i flashing (Fig.12). The break flow regime changes from subcooled to saturated critical e flow, voiding occurs in the break and increases the resistance to flow, and the break l mass flow rate decreases rapidly. Processes / phenomena in the primary coolant system are tightly coupled during the
- carly part of the blowdown period. The decreasing pump-side break flow rapidly affects core cooling processes / phenomena. RCP 2A and RCP 2B share a common plenum in SG 2. Immediately following accident initiation, the flow through RCP 2B to the break is so high that flow delivered to RCP 2A is small. However, as the CL 2B break flow decreases, the flow delivered to RCP 2B recovers to a near-normallevel. The j increased flow through RCP 2A is delivered through CL 2A to the downcomer inlet annulus where it joins with the flows from CL 1A and CL 1B. With the additional flow 5
from CL 2A, the coolant supplied through the intact cold legs (IA, IB, and 2A) exceeds the bypass flow to the vessel side of the break through CL 2B for a brief interval. The l excess coolant enters the still liquid-full downcomer and displaces liquid into the lower
, plenum, partially refilling the lower plenum, introducing some liquid into the core
- region, and restoring upward flow through the core for a brief interval. Concurrently, core power decreases via void reactivity insertion and insertion of the shutdown rods.
The release of stored heat from the core diminishes. The cladding temperature peaks ', (first peak) and begins to decrease. aCladding temperatures for each computational rod are searched and the maximum cladding temperature s located each time plotting data is stored. During the transient, different computational rods will have the maximum temperature. Figure 5 is a composite repre*entation of the maximum cladding temperature.
- 15 i
. As the primary system pressure continues to decrease, flashing begins in all four cold I . legs and voiding appears in the pump-suction inlets. Pump performance degrades, and the intact cold-leg mass flows rapidly diminish and approach zero. The interval of i lower plenum refilling that began with recovery of the flow in CL 2A is termmated and 1 the lower plenum resumes emptying. Flow enters the top of the core through the guide ! tubes and holes in the upper core plate (countercurrent flow). Sources of the liquid in the upper plenum include coolant existing from the time of accident initiation, liquid entering the upper plenum from the pressurizer and HL 1, and liquid entering the
- upper plenum from the upper head. The core liquid volume fraction remains about the j same for the remainder of the blowdown period, and the core temperature decrease i continues. Top-down cooling is provided by the liquid flow passing from the upper l plenum to the top of the core. The fuel assemblies receive different amounts of water l l'
, depending upon their location. Some of the design factors producing the nonsymmetrical top-down flows are location relative to the control rod guide tubes, ,
location relative to the pressurizer hot leg (HL 1), and the core radial power profile.
- Thus, some portions of the core experience a top-down quench and others are cooled
- only a little from the top-down flow of coolant. The TRAC modeling of the guide tubes
! and core is such that the guide tube flow is distributed to all fuel rods. Early in the blowdown period, the valves isolating the CMTs from the DVI lines are ! opened. A small flow is induced by the liquid head in the CMT and flows into the
- downcomer. However, all the flow from the CMT bypasses the core as it is entrained in
! the flow rising upward through the downcomer and carried into the broken cold leg and out the break. -
- i j System Response j Automatic Depressurization System
i
- The ADS is not activated. Operation is based upon CMT level. The time at which I the CMT level decreases to the trip level occurs well beyond the blowdown j period.
i Passive Containment Cooling System: i i Vapor and two-phase fluid pass through the break and exhaust into the l containment. A constant containment back pressure of one atmosphere was specified for the TRAC calculation, In reality, the containment atmosphere
- pressure and temperature increase and a reactor trip signal is generated when a containment overpressure of 5 psiis reached. The increasing containment back
- pressure will have no effect on the primary because the break remains choked j throughout this interval.
Passive Heat RemovalSystem:
- The S signal is generated on one of its initiators (e.g., high containment pressure, i
radiation leakage, or low pressurizer pressure). The S signal initiates CMT 4 actuation (opens the isolation valve). CMT actuation initiates PRHRS actuation 16
.. . . . - - - . . _ - - . - - - - - - - _ _ - - - _ . . - ~ . _ - _ - -
l ) (opens the isolation valve). Flashing and flow reversal occur in the PRHRS inlet
, piping immediately following break initiation, and two-phase flow passes into l l HL 1 and renders the PRHRS ineffective. Draining of the PRHRS inlet piping j supplies an indeterminate amount of the liquid to HL 1 and thence the top of the l
- core where it can participate in the top down quench of the core that occurs
- during this phase. The draining of the PRHRS heat exchanger (PRHRHX)
- volume and outlet piping inventory into the SG 1 exit plenum provides added liquid inventory to the RCS. The mass flows back into the RCS from the PRHRS
- at the hot leg and SG 1 exit plenum are shown in Fig. 20.
- Passive Safety Injection System
- '
The primary system pressure remains above the 4.83 MPa (700 psi) accumulator set point (Fig.12) so there is no accumulator injection during this period (Fig.15).
- CMT recirculation begins during this period once the S signal induces CMT l actuation (opens the isolation valve). The openmg of the CMT isolation valve is l followed by a small recirculation flow as shown in Fig.18. However, the CMT l remains filled with liquid (Fig.19) because the CMT is replenished by liquid
! delivered through the PBLs (Fig. 21). The IRWST and sump do not drain during i this period.
- Primary Coolant System:
4
- Break flow. Liquid flows through the break and into the containment during the
] first few seconds following the break. Figures 22 and 23 show the break inass j flows and exit vapor fractions, respectively. For the remainder of the blowdown _ { ~ period, two-phase critical flow with increasing vapor content passes through the break, and by the end of the blowdown period the liquid content in the break flow is small. The break flow peaks early, during the interval when liquid passes out of the break. The vessel-side break flow is significantly larger than the pump-side break flow. Flashing in the core maintains a higher pressure on the vessel side. Also the operating pump actually is a resistance to flow out of the pump side of the break. The vessel-side break flow retains more liquid content longer than the pump side. RCPs. Figure 24 shows the total mass flow in each loop at the steam generator exit plenum. The break causes an increase in mass flow in each loop but more so in loop 2, the loop with the broken cold leg. The RCPs operate in single-phase mode until void appears in the pump suction shortly after 5 s. However, pump performance degrades rapidly thereafter, and the loop flows also rapidly decrease. The RCPs continue to operate throughout this period. Cold legs. Because the RCPs continue to operate, the intact-loop cold leg flows (CL 1A and 1B) remain constant until voiding occurs in the pump suctions at
-5 s. The flows then rapidly decrease. Figures 25 and 26 show the cold-leg mass flows and voiding, respectively. The flow in CL 2A is reduced immediately following accident initiation as the break in CL 2B consumes most of the flow 17 i
i j arriving in the SG 2 plenum that is common to both CL 2A and 2B. The flow in l CL 2B is simply the break flow. However, as the pressure in the broken cold leg i decreases, the flow through the break flow decreases. Voiding at the break then ! . begins, which increases flow resistance resulting in a further reduction in mass l flow. The incoming flow from the steam generator plenum is diverted to the i intact pump RCP 2A. Voiding begins in the inlets of all RCPs shortly thereafter j and the flows through the intact cold legs (IA,2A, and 2A) rapidly decrease. j Processes / phenomena in the break, cold legs, and vessel are tightly coupled as ! discussed in the blowdown period overview. ! Hot lees. HL 1 (intact loop) and HL 2 (broken loop) display significantly
- different behavior immediately following accident initiation. Flow in HL 2 j continues in the normal flow direction (Fig. 28). However, a brief interval of high
! flow results from the increased pressure difference induced by the break. The HL 1 flow bifurc.ites. Part of the flow continues in the normal flow direction i through SG 1 (Fir. 27). The remainder of the flow reverses and flows backward into the upper plenum. Pressurizer draining contributes to this flow (Fig.14). i Liquid entering the upper plenum from HL 1 participates in top-down cooling of I the core. Voiding begins almost immediately in the hot legs because the , temperatures are higher in this portion of the primary coolant system. Figures 29 e and 30 show the hot-leg voiding. i ! Pressurizer. The pressunzer empties a large fraction of its inventory through the
- surge line and into HL 1 within the first 8 s following the break (Fig.14). -The i pressurizer blowdown occurs while its pressure is higher than in the remainder -
of the primary system (Fig.12). Once the pressurizer pressure equalizes with the . remainder of the primary system,its discharge rate slows. During the blowdown i period, the pressurizer discharge is the primary source of the total makeup flow l (Fig. 31). 4
- Steam eenerators. Voiding occurs in the hottest and highest portions of the
- steam generator primary much as that described for the hot legs. Similarly, the i outlet flows track those of the corresponding cold legs.
Reactor System: l Control rods. Upon receipt of the S signal, the control rods insert over a period i of ~1 s and are fully inserted by ~1 s following the S signal (i.e., essentially 2 s j after the break). The initial shutdown of the reactor results from negative l reactivity inserted by core voiding (Fig. 39). However, the inserted control rods maintain the reactor in a shutdown condition once the core is refilled. The i control rods will not be discussed further for the remaining periods. i j Lower elenum. The coolant inventory in the lower plenum decreases (Fig. 53)
- for the first 3 s after the accident. During this interval, the break is supplied by
- both the flows through the intact cold legs (1A, IB, and 2A) and the vessel l inventory. The lower plenum inventory recovers briefly once the intact cold legs 18
} 4 I . _ _ _ . , .
i 4 j can supply the break (see the cold-leg discussion for this period), but this partial j refilling of the lower plenum lasts only until the RCPs void at 5 s and the flows ! through the intact cold legs decrease and approach zero. At the end of this period,50% of the liquid inventory of the lower plenum is lost. Figure 54 shows j the lower plenum liquid and saturation temperatures. The lower plenum coolant i reaches saturation at 5 s; it remains saturated until ~40 s, when the lower plenum ! is refilled with subcooled accumulator flow. \ i Core recion. Flashing begins shortly after the break occurs as a result of primary i system depressurization (Fig. 33). The core liquid flow reverses and flows I downward to the lower plenum (Fig. 34) and upward to the broken loop through
- the downcomer. Vapor flow continues in the normal direction throughout much l of the core, resulting in countercurrent flow within the core and at the core exit.
j During this period, flow passes downward from the upper plenum into the upper portions of the core (Fig. 35). In addition, flow passes downward from the
- upper head and is delivered to the top of the core through the guide tubes. l However, the deliver of coolant from the upper plenum to the core is interrupted !
for ~2 s beginning at 2 s. This interrupts the top-down quench of the fuel rods l j (see fuel-rod discussion for this period). More top-down quenching occurs in the outer core region because of more liquid down flow in the outer core region. , I { Figure 37 shows the core-outlet liquid mass flow for the inner ring and outer ring ! of the modeled core region. Figure 38 shows the corresponding vapor mass j flows. The flow in the core has a multidimensional character and this is reflected , l in different rates of cladding heating and cooling throughout the core. Voiding ! in the core inserts negative reactivity and is the initial mechanism for core power l j decrease before the control rods are inserted (Fig. 39). A two-phase mixture ' ! continues in the core throughout this period, but the amount of liquid remaining j j in the core at the end of this period is small (Fig.11). Core-inlet and core-outlet '
- flows for each core sector cell are given in Apps. A and B, respectively.
j Upper plenum. The vessel upper plenum begins to void immediately following
- break initiation and is almost fully voided by the end of the period (Fig. 56).
l Fluid exits the upper plenum through the broken-loop hot leg and via top-down j draining of liquid through the upper core support plate into the top of the core
- (Fig. 37). Countercurrent flow exists in the upper plenum. Vapor enters the upper plenum from the core region (Fig. 38) and early in the period, the intact-l loop hot leg delivers water to the upper plenum.
Upper head. The vessel upper head communicates with the upper plenum
- through the guide tubes and drain holes in the upper head support plate. It also j communicates with the top of the downcomer annulus. The guidetubes can also i deliver coolant directly to the top of the core. After a delay of ~5 s, the upper
! head begins to void and is almost fully voided by the end of this period (Fig. 55). j The largest flow from the upper head is to the upper plenum through the drain ] holes (Fig. 58), although this flow does not begin for ~3 s following the accident. j Some of this upper-head flow is swept across the upper plenum into the broken-i ! 19 i 4
, _ _ _ _ __m . _
b. l cold-leg hot leg and thus does not reach the upper core plate. Upper head ! inventory is also delivered directly to the top of the core through the guide j tubesb (Fig. 52). The guide tube flow begins immediately following accident initiation. The upper head to downcomer flow is small (Fig. 59). Downcomer. The downcomer annulus remains either full or nearly full until j ~7 s (Fig. 51). Then the downcomer begins to rapidly empty and -65% of the
- liquid inventory is lost by the end of the period. "Ihe voiding proceeds from the j top of the downcomer annulus downward under the influence of the cold-leg i break. When the voiding reaches the level of the broken cold-leg nozzle, voiding
! contributes to a continuation of the break flow decrease, which to this time has i been dominated by the decrease of the primary system pressure. Early in the j transient, flow from the intact cold legs passes around the downcomer annulus j and out the broken cold leg. In addition, inventory from the other areas of the { vessel supply the break until the flow has diminished to a level that can be supplied by flow through the intact cold legs. Fuel rods. Heat transfer from the fuel rods to the coolant degrades as portions of the core void and the fuel cladding temperatures begin to rapidly increase. Essentially all the stored energy in the fuel rods is redistributed from the center i of the fuel rods to the outer periphery of the fuel rods and the cladding (Fig. 40). The cladding temperature increase slows markedly beginning at ~5 s, and the blowdown or first peak cladding temperature is reached at ~7 s (Fig.13). The blowdown period fuel-rod temperature increase is terminated by a combination of processes / phenomena occurring in the primary system. These processes / phenomena alter the core power to flow ratio. The decreasing core power and removal of core stored energy affect the power component. The flow ! component is more complex. Immediately following break initiation, the vessel side flow is larger than can be supplied by coolant entering the downcomer annulus from the intact cold legs. Thus, the required additional flow is supplied from vessel inventory. When core inventory depletion by flow processes is combined with coolant flashing in the core, a large fraction of the core becomes vapor filled. At ~3 s, the vessel-side break flow can be completely supplied by the flows entering the downcomer annulus from the intact cold legs. This is caused by the combination of flow recovery in CL 2A (Fig. 24) and the decrease in break flow (see blowdown period overview and cold-leg discussions for this period). The coolant delivered to the downcomer annulus in excess of that needed to supply the break is delivered to the lower plenum and core. The liquid inventories of both the lower plenum and the core increase (Fig.11). The coolant entering the core boils and the resultant b In the 'IRAC model, the guidetube flow is delivered directly to the top of the core and is distributed uniformly over all assemblies modeled in the cell. In the actual AP600 reactor design, the guidetube flow is delivered to a select number of assemblies in the core and not to all the usemblies. Furthermore, the guidetubes have slots that communicate between the inner portion of the guidetube and the upper pienum. These guidetube details are not included in the TRAC model. 20
i I steam flow cools the core. The cladding temperatures are decreasing at the end i o.'the blowdown period. l The TRAC calculation showed considerable variation in top-down quenching i and rewetting with radial and azimuthal location and relative location to the break. To illustrate, Figs. 41 and 42 show average rod cladding temperatures at l ! two vessel cell locations, cells 1 and 4. Cell 1 is in the inner core ring of azimuthal l sector 1, the same azimuthal sector of hot-leg 1; it is also the cell that the j maximum hot-rod peak clad temperature occurs in. Cell 4 is in the inner core ! ring of cold leg 2A azimuthal sector, two sectors away from the broken cold leg i sector. Figures 43 and 44 show the hot rod cladding temperatures at the same i two cells. Figures 45 and 46 show the average rod cladding temperatures for j cells 9 and 13, which correspond to the outer core ring at the same azimuthal i sectors. As can be seen, cell 1 receive more top-down quenching and cooling } than cell 4. In general, those cells away from the break receive more top-down ! flow than those near the break. Also, the outer ring cells receive more top-down i quenching and cooling than the inner ring cells, a consequence of CCFL that j limits the liquid downflow in the inner core ring because of a higher vapor upflow. Average-rod and hot-rod temperatures for each core sector cell are , presented in Apps. C and D, respectively. Fuel-rod cladding temperatures as a function of core elevation at selected
- transient times are in Figs. 47 through 50. A detailed view of the progression of l rod heatup, cooling, and reflood can be obtained through careful study of such data. Figures 47 and 48 are for an average-power rod and Figs. 49 and 50 are for i a maximum-power fuel rod, both rods in core sector cell 4. For example, much of
- tiie upper portion of the core (approximately the upper 1-1/2 m,5 ft) is quenched j during the blowdown period (Figs. 47 and 49). This quenching arises from liquid l flows entering the top of the core through the upper core support plate (Fig. 36).
j Fuel-rod cladding temperatures as a function of core elevation at selected j transient times are presented in Apps. E and F for each core sector cell, App. E for average-power fuel rods and App. F for maximum-power fuel rods. l Guide tubes. The vessel guide tubes provide a flow path whereby fluid from the j upper head proceeds directly to the top of the core (Fig. 52). Although some of ! the holes in the upper core plate are covered by the guide tubes, others are not. j This distinction is lost in the TRAC input model which divides the vessel into 16 i sectors, each of which represent in some average way fuel elements that are both directly under guide tubes and fuel elements that are removed from guide tube locations. Significant guide tube flow delivery to the top of the core occurs only during the blowdown period. This flow begins immediately as the pressure in the upper plenum and core decrease under the influence of the break. The i steady-state fluid temperature in the upper head is between the hot- and cold-leg
! temperatures, but closer to the cold-leg temperature. Thus, flashing in the upper
! head is delayed for several seconds after the start of the accident but sustains j guide tube flows as the primary pressure decreases. The guide ' tubes play no i 1 21
~
4 a j
l l significant role during either blowdown or the remainder of the transient and . willnot be discussed further. l t . I Vessel structures. The vessel structures include the core support plates, core i barrel, downcomer walls, etc. These structures also contain stored heat that i interacts with the coolant. Vessel structures appear to play an insignificant role
- in this or the remaining periods of the LBLOCA and they will not be discussed I j further. I l Bvoass flows. There are several small flows that bypass the core and are
- unavailable for core cooling. These include rod thunble flow, core bypass flow, i reflector cooling flow, and reflector cavity flow. During steady state, these flows l are ~7.5% of the total loop flow. The fluid in these regions flash during the
; depressurization transient. The bypass flows play no significant role during
! either blowdown or the remainder of the transient and will not be discussed
- further.
Steam Generator System (Secondary Side): The S signal initiates feedwater valve closure, and closure is completed during this
- period. In addition, the containment overpressure signal leads to closure of the main i steam isolation valve (MSIV) during this period. Thus, the steam generator is isolated
! during this period and both feedwater and steam flows decrease to zero (Fig. 57). The 4 SG-pressure is slightly higher than.the primary pressure at the end of this period but l secondary-to-primary heat transfer is insignificant because the tubes on the primary ; j side of the SG are highly voided. l 4 l 5.2. Refill Period (12.3 to 43 s) { Overview ! i The refill period begins at the time accumulator injection flow is initiated, and ends !' when sufficient water from the PSIS has entered the lower plenum to nearly or fully refill the lower plenum. The end of the refill period also corresponds to the start of the reflood period. Identification of the transition from the refill period to the reflood period is somewhat uncertain, as the lower plenum is not completely refilled and the collapsed liquid level is fluctuating when the sustained core inventory increase begins. Refill begins at ~12.3 s and ends at ~43 s (Fig.11). Early in the refill period, the CMTs supply water to the DVI line. However, as the primary system pressure continues to decrease, ECC flows from the accumulators increase and eventually terminate the CMT flows. Accumulator injection is initiated at the start of the refill period when the primary pressure decreases to 4.83 MPa (700 psi) (Fig.12). The coolant flows to the reactor vessel downcomer through the DVI lines.
- Early in the period the injected liquid does not enter the lower plenum. Rather, ECC entering the downcomer through the DVI lines is entrained by core-generated steam passing upward through the downcomer, carried into the broken cold leg, and exhausted into the containment through the break. Thus, the liquid inventories of the
i j core, lower plenum, and downcomer continue to decrease during the early part of the refill period. In this steam-filled environment, the core begins to reheat (Fig.13). l ECC bypass continues only as long as core-generated steam passes through the I j downcomer and exhausts to the containment through the break. Steam can be , i generated only so long as coolant is entering the core. No coolant enters the core from j 3 the lower plenum during the early part of the refill period (Fig. 34). More important is l j the cessation of coolant flow from the upper plenum into the top of the core at ~20 s 1 j (Fig. 35). Shortly thereafter, all remaining liquid in the core turns to steam and by 23 s
; the core is steam filled (Fig.11).
! Concurrent with the termination of top-down liquid flow at 20 s, core-generated steam ! production decreases. The downcomer begins to refill at 20 s and a second later the
- lower plenum begins to refill. At 43 s, the lower plenum first fills to the lower core
! support plate (Fig.11). There is a small accumulation of water in the core beginning at j 37 s but the core is still heating at the end of the period (Fig.13). l System Response
- Automatic Depressurization System:
) The ADS is not activated during this period (See Blowdown Period System l Description). i 1
- Passive Containment Cooling System
- )
l l The containment atmosphere pressure and temperature continue to increase - j during this period. The pressure peaks at ~50 psia (Ref. 8) and remains between l 45 and 50 psia for an extended period. The system pressure appears to remain
- high enough to keep the break choked throughout period. We again note that a
!' constant containment back pressure of one atmosphere was specified for the TRAC calculation. Precise statements regarding the time at which the break
- planes unchoke are not possible because the calculation was not performed with j a coupled model of the primary system and the containment.
i j Passive Heat RemovalSystem: i The flow entering the PRHRS piping from the hot leg is highly voided at the start i of the refill period and fully voided by the end of the period. Vapor entering the ! PRHRSHX condenses, but the energy transferred to the IRWST heat sink is j minor. The flow at the PRHRS outlet is liquid and contributes to the liquid i inventory in SG 1. Figure 20 shows the PRHRS inlet and outlet mass flows. The . i PRHRS is not designed to operate during accidents in which highly-voided j conditions exist in the primary system. j Passive SafetyInjection System: i l 23 i
1 Accumulator flow injection begins at -12.3 s (Fig.15) when the system pressure drops below the 4.83 MPa (700 psi) accumulator set point (Fig.12). The accumulator flow rapidly terminates the CMT flow that began during the blowdown period. Choking in the DVI nozzle venturi occurred at 22 s, limiting the accumulator flow thereafter. Neither the IRWST nor the sump drain during this period. , u . l Primary Coolant System: Break flow. The pump-side break flow is essentially all vapor, and critical flow l conditions persist during this period (Fig. 23). The vessel-side break flow l entrains some liquid flow from the downcomer but it too is highly voided. The { decrease of the primary pres 3ure (Fig.12) and the highly voided nature of the i flows from each side'of the break contribute to a diminishing break mass flow l (Fig.22). At the start of the refill period, the total break flow is ~5700 kg/s. At the end of the refill period, the break flow is near zero. ( RCPs. The reactor coolant pumps trip at -16 s. This has very little effect on the j mass flow because the flow through the pumps is highly voided at the start of i the refill period, and pump performance is highly degraded. ! Cold legs. The cold legs progress from a highly voided to a near fully voided condition durmg this period (Fig. 25). The connection to the cold-leg PBL voids, j and a vapor path is opened to the top of the CMT, which initiates CMT draining ! as previously discussed. 4 _._ i Hot lees. The broken-loop HL 2 is essentially fully voided from the beginning of i the refill period (Fig. 29). The intact-loop HL 1 is highly but not fully voided at
- the begmning of the period ({ig. 28). There is a flow from this hot leg back into i the upper plenum between 10 and 20 s. This flow is from the pessurizer and j backflow from the loop steam generator. Both hot legs are fully voided by the j end of the refill period.
j Pressurim. After the pressurizer pressure equalizes with the primary pressure ! near the end of the blowdown period, the rate at which the pressurizer inventory i is discharged into the loop-1 hot leg slows. The pressurizer discharges the
- remainder of its inventory into the loop-1 hot leg by the end of the refill period.
i l Steam generators. The steam generator tube bundles void at -16 s and remain voided. As the primary system pressure decreases, secondary-to-primary
- reverse heat is possible. However, because the primary-side tube bundles are j vapor filled, the secondary-to-primary heat transfer is insignificant. Some mass j flows back into HL 1 from the steam generator between 10 and 20 s.
1 t. 1 i e 24 l
Reactor System: . 1 _. Lower Plenum. The key process during this period is refilling of the lower plenum to set the stage for core reflood. At the start of the refill period, the lower plenum inventory decreases (Fig.11), and it continues to decrease until ~22 s.- During the interval between 12.3 and 22 s, the break flow rapidly decreases as discussed previously. Steam generation in the core has diminished because the . core approaches a fully voided condition and their is little flow of either liquid or i vapor at the bottom of the core, which can pass into the downcomer and retard the flows (Fig. 34). Consequently, the ECC flows injected through the DVI lines begin to refill the downcomer and lower plenum at -22 s. De liquid level in the lower plenum first reaches the bottom of the core at ~43 s and this time is declared as the end of the refill period and the beginning of the reflood period. Flow oscillations occur during the lower plenum refilling process. Core Recinn. The core is at decay power levels and is maintained in a shutdown condition by the control rod reactivity (Fig. 39). The core is ~90% voided at the start of the refill period and fully voided by 20 s. Between 12.3 and 20 s, the downward flow of liquid into the top of the core from the upper plenum is decreasing and approaches zero (Fig. 35). After 20 s, the vapor in the core is heated and moves upward but the vapor mass flow is very small and produces little cooling of the core. Upoer Plenum. The upper plenum is -90% voided at the start of this period, becomes fully voided by 25 s, and remains fully voided throughout the remainder of the refill period (Fig. 56). The upper plenum supplies essentially all the liquid flow into the core between 12.3 and 20 s. The coolant flow from the upper plenum into the core supplies the liquid that sustains steam generation in the core and ECC bypass. Upoer Head. The upper head is -90% voided at the start of this period, becomes fully voided shortly thereafter, and remains fully voided throughout the , remainder of the refill period (Fig. 55). The liquid in the upper head at the begmning of the period enters the upper plenum through the upper support plate drain holes (Fig. 58) and the downcomer through the downcomer-upper ; head flow path (Fig. 59). There is no flow into the upper plenum through the guide tubes after the upper head liquid level drops below the top of the guide tubes near the end of the blowdown period. Downcomer. See the description of lower-plenum processes and phenomena for the refill period. Fuel Rods. At the start of the refill period, the fuel rods cool as a result of liquid introduced into the core late in the blowdown period. The temporary restoration of liquid flow into the core during the blowdown period occurred when the intact-loop cold leg flows were sufficient to supply the rapidly decreasing vessel-25
i ! side break flow However, subsequent voiding in the cold legs at the end of the
- blowdown period terminates the core flow recovery and the core becomes vapor l filled during the refill period as previously discussed. In this vapor filled environment, the cladding-txoolant heat transfer degrades and the fuel rods
- begin to reheat (Fig.13). As previously discussed, the flows refilling the lower i plenum are oscillatory. Thus, some liquid begins to appear at the core inlet
] before the end of the refill period (Fig. 34). The core heats up (Fig.13) and
- continues to heat to the end of the period. The flow of liquid downward from l the upper plenum into the core between 12.3 and 20 s removes some of the core j decay heat, especially at the top of the core as shown in Fig. 48.
1 Steam Generator System (Secondary Side): i l Secondary-to-primary heat transfer is initiated during this period, but the steam-l generator primary is vapor filled and the reverse heat transfer is limited. ! 5.3. Reflood Period (43 to 240 s) j Overview ! Reflood begins after the lower plenum refills and the core begins to refill (Fig.11). j Reflood is completed when the entire core is quenched. Initially, core reflood is quite
- rapid because (1) the downcomer head is initially resisted by only a steam-filled core, j and (2) com-generated steam flows through the core, coolant loops, and break are small i so the pressure resisting the downcomer head is small. For this analysis, the reflood
! period begms -43 s after accident initiation. .. 1 i Because of the high fuel rod temperatures at the beginning of reflood, the entire - l spectrum of thermal regimes, starting with single-phase liquid and progressing upward j through the core with nucleate boiling; transition boiling; film boiling; churn two-phase i flow; dispersed droplet flow; and single-phase steam flow, are encountered. However, i by the end of this period, fuel temperatures have peaked and the entire core is i quenched (Fig.13). i ~ Because of droplet carryover from the core and subsequent deentrainment at the upper core plate and grid spacers, top quenching and local quenching occur in addition to i
- bottom quenching. Higher vapor velocities and liquid entrainment occur in the central j region of the core where higher-powemd fuel rods are located. The entrained liquid has j a cooling effect on the fuel rod regions above. The upper portions of the core remain
( cooler because significant cooling occurred during the previous period as liquid from
- the upper plenum entered and cooled the upper portions of the com (Figs. 47 and 48).
l Some of the entrained lie aid is deentrained at the upper core plate. The remainder is ! carried into the upper plenum, where it is deentrained, forming a two-phase pool. l Liquid from the pool can reenter the low-power regions of the core through the upper i support plate because of the lower vapor velocities in those regions. A three-l dimensional flow pattern results: in the core, flow is from the low- to high-power i regions, while in the upper plenum the flow passes from the high 'to low-power j regions. Liquid from the upper plenum two-phase pool may be further entrained and ! 26 l. 4
io i i l carried over into the hot legs. In traversing the upper plenum, this liquid may be j further deentrained on upper plenum internal structures. ! j . As the bottom quench progresses upward through the core, more liquid is carried over ~ to the upper plenum pool. Conditions exist (steam passing through the upper plenum pool) by which liquid can be entrained and carried into the hot legs. If the liquid j carried through the hot legs reaches the steam generators, it will be boiled by reverse j steam generator heat transfer, causing a pressure increase in the steam' generator i bundle. In the current generation Westinghouse PWRs, the loop seals are liquid full so j any vapor generated in the steam generators increases the hot-leg pressure and can j retard the core reflood process. This phenomenon is called " steam binding." The AP600 ) does not have loop seals, and a vapor path is open to the vessel from both sides of the steam generator, making it unlikely that steam binding will be significant in the7tP600. i There is some liquid accumulation in the steam-generator inlet late in this period, which ! will alter the loop pressure losses somewhat. I ! For much of the reflood period, manometer oscillations between liquid in the downcomer and the core lead to an oscillatory reflood rate. The downcomer liquid level (head) . .e driving potential for liquid to enter the core. Liquid passing through the downcomer and entering the core from the lower plenum boils; the increased pressure in the core alters the core-downcomer force balance and reduces the flow from the downcomer. With reduced core flow, there is less steam generation, the core-side pressure decreases and again alters the core-downcomer force balance causing an ) increased downcomer to core flow in this part of the cycle. Associated with the oscillations is increased liquid entrainment and carryover that accelerate core ' quenching. Quenching of the maximum-power fuel rods occurred at the same time that the accumulator emptied. System Response Automatic Depressurization System: The ADS is not activated during this period (See Blowdown Period System Description). Passive Containment Cooling System The cooling processes of the PCCS become more important as the transient proceeds. The primary functions of the containment are several. First, the containment is to provide a barrier to fission product release. To do so, the integrity of the containment structures must be preserved. Thus, the containment pressure increase must be limited. Second, the decay heat of the core must be transferred to the ultimate heat sink, the atmosphere. If systems were not provided to cool the containment, the pressure would rise and possibly fail the containment which is an essential system during the long-term cooling period. - 27
* ~
,- pe-, m
- The break planes unchoke during the refill period. Precise statements regarding
- the time at which the break planes unchoke are not possible because the I calculation was not performed with a coupled model of the primary system and
{ the containment. The containment and primary coolant system are coupled l during this period but the coupling does not appear to have a strong impact on i core cooling processes. The PCCS is not modeled in the TRAC calculation. The l following qualitative description of PCCS operation is provided for completeness. 1
- Early in the transient, energy transfer to the atmosphere through the steel j containment structure is enhanced by evaporation of liquid deposited on the i outside of the steel containment structure near the top of the structure.
- Buoyancy induced air flow through the air gap between the steel concrete
! structure and the concrete shield building also cools the structure. By the time i the PCCS water supply is depleted during the long-term cooling period, the
- decay heat had decreased sufficiently so that the buoyancy induced air flow
- through the air gap between the steel containment structure and the concrete shield building can remove the decay heat of the core.
i j Passive Heat Removal System: j The PRHRS mass inventory is depleted either from flashing or from draining, and this system does not contribute to energy removal from the primary. j Passive SafetyInjection System: l The sole source of emergency coolant injection during this period is via accumulator injection (Fig.15); this circumstan'ce continues until the accumulator
- empties, which occurred in this transient at about the same time that the last of
- the maximum-power fuel rods were quenched. Figures 16 and 17 respectively
! show the accumulator liquid levels and the fraction of initial liquid volume. i Noncondensable gas from the accumulators is discharged into the DVI lines at
-190 s, which rapidly decreases the mass flow of the remaining accumulator li liquid. The discharge of gases into the DVI lines increases the pressure in the l DVI lines such that CMT flow injection is delayed for a short time after the l accumulators have discharged their liquid inventory. Typically, analyses of the
- LBLOCA transient are terminated before CMT injection resumes because the key j safety acceptance parameter, the peak cladding temperature, has been reached j and the core is cooling.
) Primary Coolant System:
. Break flow. The impact of the break on primary system behavioi, major during ; the blowdown and early refill periods, becomes less important throughout the . refill period and is of little importance during the reflood period. Neither the
- pump-side nor vessel-side break flows are critical, and flows are near zero
; (Fig. 22). The pump-side flow is vapor until the refill process is well advanced -
i 1 28 1
D 1 l
- (Fig. 23). The vessel-side break flow is nearly so, but significant amounts of
! liquid appear at the vessel-side break plane after 60 s, as the oscillating j downcomer liquid level approaches the cold-leg nozzle connections to the vessel. ! The liquid entrainment out the break increases the net system mass flow loss l l (Fig. 32) and slows the vessel refill rate as can be seen in the integrated net mass j flow loss, Fig. 33. i E2L The reactor coolant pumps still are rotating at a fairly high rate (Fig. 60), i even though they were tripped at 16 s. The average mass flow through the loop-1 pumps during the reflood period is 17 kg/s. The average mass flow , . through the loop-2 intact cold leg pump is 10 kg/s, whereas the mass flow l l through the broken cold-leg pump is 24 kg/s. The flow consists mostly of vapor 1 j during this period; therefore, even though the mass flows are low, the volumetric l
- flows are relatively high. The volumetric flows of the loop-1 pumps are i
! ~13.7 m3/s. The volumetric flow of the loop-2 intact cold-leg pump is 12 m3/s, j and the volumetric flow of the broken cold-leg pump is 26 m3/s. For ! l comparison, the pump volumetric flow at steady-state full power is 3.2 m3/s. ! 1 ! Cold lees. The intact cold legs are essentially voided, and the flow consists l i mostly of vapor (Figs. 26 and 25, respectively). After 60 s, the oscillating j downcomer liquid level reaches the cold-leg-nozzle connections to the vessel and 1
'significant amounts of liquid begin to move through the broken-loop cold leg to j the break.
i l j Hot lees. The hot legs are vapor filled during the early part of the reflood period ! (Figs. 28 and 29). As the upper plenum begins to accumulate liquid later in the i transient (Fig.11), small amounts of liquid are entrained and carried into the hot
- legs. If carried into the steam generators, rew.rse heat transfer from the ,
- secondary side of .the steam generators vaporizes the liquid. In current l Westinghouse commercial plants, the vapor generation in- the steam generators l participates in a process / phenomena called " steam binding." Steam binding occurs when liquid vaporizes in the steam generator and the pressure increases propagate toward the core. The increased core pressure offsets some of the j downcomer liquid head and the core reflood rate decreases. In plants with a loop seal, such pressure increases are possible because the liquid in the loop seal l blocks pressure equalization. AP600 does not have loop seals and we believe j that steam binding will not occur because vapor generated in the steam j generators can flow unobstructed to both sides of the core.
i j Pressurizer. The pressurizer is voided. i [ ' Stamm generators. See the previous discussion of reflood phenomena in the hot legs. During the latter phases of the reflood period, liquid is carried to the steam
;' generator inlets (Figs. 28 and 29). Some of the liquid is deentrained there, and that which is carried into the steam generator tubes is vaporized.
{ 29 )
P l i Reactor System: i Lower Plenum. The reflood period begins when lower plenum refilling is , ! completed. Because of the oscillatory filling process, the actual time assigned to l 1 the start of this period is somewhat uncertain. However, by 45 s, the lower j plenum is essentially full (Fig. 53) and lower pienum liquid is subcooled (Fig. 54). ; Core Recion. The core-outlet flow is highly oscillatory from a time shortly before { the end of the refill period and continuing until the core is completely reflooded l (Fig. 34). The flow oscillations are considerably damped in the confined flow l passages of the core [ compare the core-outlet mass flows (Fig. 35) to the core- , outlet mass flow (Fig. 34)]. The liquid fraction, although oscillatory, steadily l
- increases (Fig.11) and then decreases at 200 s because of the decrease in l accumulator flow (Fig.15). For much of the reflood period, manometer j oscillations between liquid in the downcomer and the core lead to an oscillatory reflood rate. Liquid entering the core from the lower plenum boils and the i incre
- sed pressure in the core reduces the flow from the downcomer. The
, downcomer flow is set by balance of the liquid level (head) in the downcomer
- and the pressure in the core.
, Upoer Plenum. The upper plenum begins the reflood period in a fully voided i condition (Fig. 56). However, as the reflood period continues, the core liquid level increases, accompanied by significant steam generation (Fig. 35) that carries j liquid into the upper plenum (Fig. 36). A small amount of liquid accumulates in i the upper plenum (Fig. 56), but there are several competing processes that i prevent a large accumulation of liquid. These processes include entrainment and - l transport of liquid into the hot legs, as previously discussed, and the flow of l l liquid to regions of the upper core support plate above the lower powered l i- regions of the core. As the steam generation in these regions is small, some of the ! j upper plenum liquid flows downward into the top of the core. l Upper Head. The upper head remains fully voided during this period (Fig. 55). { Downcomer. The downcomer continues to refill from acaulator flow injection j through the DVI line (Fig.11). Large flow oscillations continue to occur. The i downcomer does not fully refill because once the liquid reaches the level of the
- cold-leg nozzle connections to the vessel, downcomer liquid flows into the t broken cold leg and out the break.
! Fuel Rods. At the start of the reflood period, fuel rod temperatures in much of 3 the core are increasing (Figs. 41-46). However, as the core begins to refill, the j heat transfer environment of the fuel rods change. The primary direction of the i quench is from the bottom up, as shown in Figs. 48 and 50, for the average-power i and maximum-power rods respectively. The upper 1-1/2 m (~5 ft) of the core were quenched during the refill period. There is a minor reheat of the upper core
! during the early part of the reflood period, but the cladding temperatures remain ] well below the peak cladding temperatures lower in the core. Coolant advancing l
30 l
~
upward from the bottom of the core is the primary mechanism for core cooling during reflood. The lower sections of the fuel rods are cooled by liquid convection and nucleate boiling, higher elevations of the fuel rods by film boiling, and even higher elevations of the fuel rods by nucleate boiling. The quench front advances upward in the core and eventually the entire core is quenched as the core refills. , Steam Generator System (Secondary Side): See the discussion of hot-leg phenomena during the reflood period. 5.4. Long-Term Cooling Period (240 s and Beyond) Overview Long-term cooling starts when the fuel rods have been quenched completely. At that point, system pressures have decreased enough that the gravity-driven core makeup tanks and IRWST can start draining into the reactor vessel. There is sufficient water from the IRWST alone to flood the containment to a level at which the broken cold-leg sections are completely submerged, thus forming a liquid natural-circulation cooling path between the core and containment. This mode of cooling is designed to remove the reactor core heat indefinitely. System Response Automatic Depressunzation System: The ADS will be activated on a core makeup tank liquid level trip signal but will not have any effect because the primary system pressure already has decreased to containment pressure levels. Passive Containment Cooling System: The cooling processes of the PCCS is the dominant mode of cooling for the long-term cooling period. The heat removed by the natural-circulation cooling loop eventually will be transferred to the containment wall and to the external atmosphere. The PCCS is not modeled in the TRAC calculation. By the time the PCCS water supply is depleted during the long-term cooling period, the decay heat has decreased sufficiently so that the buoyancy-induced air flow through the air gap between the steel containment structure and the concrete shield building can remove the decay heat of the core. Passive Heat RemovalSystem: The PRHRS mass inventory is depleted either from flashing or from draining; this system does not contribute to energy removal from the primary system. 31 y=-
i- ; I
- Passive Safety Injection System
The core makeup tanks will start to drain early in the long-term cooling period. i They are orificed so that at their designed flow, the CMTs will drain for several hours. l l Primary Coolant System: 1 1 Break flow. The break eventually will be submerged with water from the ! IRWST. 1 EQs. The pumps will cease rotating as they become submerged with water from the IRWST. Cold legs. The cold legs eventually will be submerged with water from the IRWST. Hot lees. The horizontal sections of the hot legs eventually will be submerged l with water from the IRWST. l Pressurim. The pressurizeris voided. ' Steam generators. The voided primary side of the steam generator will have no effect. Reactor System: Lower Plenum. The lower plenum will be completely full. Core Recinn. The core eventually will be filled with water from the IRWST. Upoer Plenum. The upper plenum eventually will be filled to above the hot-leg i Connections. I Upoer Head. The upper head will remain voided during this period. l Downcomer. The downcomer eventually will be filled to above the cold-leg Connections. Fuel Rods. The fuel-rod heat transfer eventually will change from saturated nucleate boiling to single phase liquid convective heat transfer. Steam Generator System (Secondary Side): The secondary side will have no effect because there is_ no primary flow circulation. 1 32
l 1 6.0. COMPARISON WITH ECOBRA/ TRAC I l
~
The TRAC calculation results were compared with the Westinghouse ECOBRA/ TRAC i results presented in the AP600 Standard Safety Analysis Report.10 Table Il presents a ; summary comparison of the TRAC and ECOBRA/ TRAC results for transient event t times and selected calculated parameters. ECOBRA/ TRAC transient calculation plots l were also scanned, digitized, and overlaid on TRAC calculation results for a graphical ' i comparison. It should be noted that the ECOBRA/ TRAC results do not reflect the latest AP600 design. We have requested the latest Westinghouse 80% DEGB LBLOCA l calculation but have not yet received it. A 100% DEGB LBLOCA ECOBRA/ TRAC , calculation, recently received from Westinghouse, also provided some insight in the ! general differences between TRAC and ECOBRA/ TRAC. Figure 60 compares the upper plenum pressures for the first 40 s of the transient, which l shows very good agreement between the two calculations. Figure 61 compares the l accumulator flow rates. The TRAC accumulator flow is ~60 kg/s higher than the ECOBRA/ TRAC flow. As indicated in the previous TRAC LBLOCA report, we believe the ECOBRA/ TRAC accumulators to be calibrated to a different design flow rate. j Figure 62 compares the lower plenum liquid levels. TRAC shows a momentary refill i during blowdown whereas ECOBRA/ TRAC did not. TRAC shows the lower plenum , i being etoptied more than ECOBRA/ TRAC but then refilling at a faster rate. Figure 63 l compares the downcomer liquid level. TRAC shows the downcomer level not J emptying as much, a slightly faster refill, and considerably stronger flow oscillations l between the lower plenum and downcomer. The next set of figures compare fuel rod cladding temperatures. The method of ! modeling the core in ECOBRA/ TRAC is different from TRAC, which makes the ; comparison somewhat difhcult. In ECOBRA/ TRAC, the core is modeled with vertical ! one-dimensional fluid channels that are subdivided into a number of cells. Crossflow , paths can be specified between these fluid channels. Five types of fuel assembly channels (referred to as rods by Westinghouse) are modeled: Rod 1 models a hot rod at the maximum allowed linear heat ratec; Rod 2 models an average rod in the hot assembly containing the hot rod; Rod 3 models an average rod in an assembly covered by an open hole or support column; Rod 4 models an average rod in an assembly covered by a guide tube; and Rod 5 models the low power peripheral assembly next to the reflector blocks. The fluid channels are connected together with crossflow connections. Aside from the peripheral assembly representation, there appears to be no radial or azimuthal dimensionality to the ECOBRA/ TRAC core model. By contrast, the TRAC core model is divided into two radial rings and eight azimuthal sectors for a total of 16 fluid e The number of actual fuel rods n=W in each type of fuel assembly channels is Westinghouse propnetary infonnanon. 33
channels, each modeling the same number of average-power fuel rods. Each TRAC fluid channel also models a supplemental maximum-power fuel rod. Figure 64 compares the hot rod peak cladding temperatures. For the ECOBRA/ TRAC 1 pet, this represents the maximum cladding temperature at all rod elevations for Rod 1. For TRAC, the hot rod modeled in core sector cell 4 is the fuel rod that has the l maximum peak cladding temperature throughout the transient. The comparison shows j that TRAC has a lower blowdown PCT and a higher reflood PCT than
- ECOBRA/ TRAC (the temperature values are given in Table II). Also, TRAC reflood j PCT occurs much earlier than ECOBRA/ TRAC, and after the reflood PCT, TRAC j shows a slightly more rapid cooldown rate than WCOBRA/ TRAC.
i
- Next, rod cladding temperatures are compared at specific rod elevations,6 ft,8.5 ft, and j 10 ft. For each comparison, two TRAC rod temperatures are shown to the one j ECOBRA/ TRAC rod temperature. The two TRAC temperatures on each plot provide
- an indication of the radial and azimuthal variation in rod temperatures calculated by l TRAC. Figures 65 through 67 compare hot rod cladding temperatures at 6 ft,8.5 ft, and
- 10 ft, respectively. Figures 68 through 70 compare average rod temperatures at the !
l same respective rod elevations. Finally, Figures 71 through 73 compare the TRAC l average rod temperatures in radial ring 2 to the ECOBRA/ TRAC peripheral assembly ! Rod 5. ! At the 6 ft. rod elevation, TRAC blowdown PCTs are lower than ECOBRA/ TRAC, but i TRAC reflood PCTs are significantly higher than ECOBRA/ TRAC. ECOBRA/ TRAC j show higher cooling and quenching after blowdown PCT, and quenching after 80 s. At the 8.5 ft. rod elevation, TRAC blowdown PCTs are lower than ECOBRA/ TRAC.
- ECOBRA/ TRAC rod heatup period after blowdown is much longer than TRAC and
! cooling turnaround does not occur until after 100 s. At the 10 ft. rod elevations, TRAC
- blowdown and reflood PCTs are significantly lower than H. COBRA / TRAC.
- ECOBRA/ TRAC quenches longer after blowdown peak but heatup period after I j quench is longer than TRAC.
f j Core vapor and liquid mass flow plots were also presented in the ECOBRA/ TRAC i analysis, but these are not compared with TRAC results because the number of fuel l assemblies represented in the ECOBRA/ TRAC mass flow plots are not known. l In general, there is reasonable agreement between the two calculations. A significant } similarity is in the calculation of the hot-rod PCT. Both the TRAC and j ECOBRA/ TRAC calculations show the reflood PCT to be higher than the blowdown l PCT. A major difference in calculation results is in the cooling of average-powered rods i during the late-blowdown phase. The ECOBRA/ TRAC calculation showed a complete i rewetting and quenching of average-powered rods during the late-blowdown phase,
- . with average-rod cladding temperatures being cooled down to saturation-temperature j levels. The TRAC calculation shows a considerable variation , with those rods furthest
- away from the break showing more quenching and rewetting than those rods nearer the
! break. i i )
Q
- Westinghouse calculations predict more extensive core cooling following the first peak than predicted in the TRAC-PF1/ MOD 2 calculation. As described in the previous paragraph, TRAC predicts that the increased delivery of coolant to the downcomer and core following recovery of flow through CL 2A. This coolant entering the core boils and the steam that is generated passes upward through the upper core support plate. The steam flow inhibits the downward flow of coolant from the upper plenum and guidetubes. As a consequence, top-down cooling of the core is limited. The Westinghouse calculations show an increase in the lower plenum liquid content at about the same time predicted by TRAC but this liquid does not enter the core. Thus, there is no steam generation and related retardation of downward flow of liquid from
, the upper plenum into the core. It is not possible to declare which representation of blowdown phenomena is more correct at the present time. l However, we do note that changes made to core heat transfer models as TRAC $v~olved from MODI to MOD 2 result in diminished heat transfer following the first peak in several LBLOCA calculations. Thus, changes to the TRAC heat transfer models may
- play a substantial role in creating the phenomenological differences observed in the l TRAC and Westinghouse calculations.
The two-phase flow representation in the vessel component of ECOBRA/ TRAC also may be the reason for the difference in quenching behavior. The TRAC reflood model j does not account for quenching due to a falling liquid film. ECOBRA/ TRAC uses a
, two-fluid, three-field representation of two-phase flow in the vessel component. The
- three fields are a continuous vapor field, a continuous liquid field, and an entrained liquid droplet field. TRAC has just a two-field representation.
7.0. CONCLUSIONS AND RECOMMENDATIONS An updated 80% DEGB cold-leg LBLOCA has been calculated for the Westinghouse AP600 advanced reactor design. The calculation reflects AP600. design changes and corrections to the TRAC code and plant model. Hot-rod PCT of 1021 K (1378'F) and 1197 K (1695'F) were calculated for the blowdown and reflood phases, respectively. 4 These temperatures are below the App. K limit of 2200*F but near the cladding l oxidation temperature of 1700*F. The TRAC calculation results were compared to the ! Westinghouse SSAR E. COBRA / TRAC LBLOCA results for the same size break. l Transient event times and peak clad temperatures were in reasonable agreement but ! there were significant differences in the detailed phenomena calculated by the two codes. The reasons for these differences are still being investigated. Further study of
- the differences in plant modeling, transient calculation assumptions, code computational method, and code models and correlations are recommended. The
{ calculation was performed with a plant model that had undergone an independent j quality assurance check. Additional break sizes and break locations need to be analyzed to confirm the most severe break postulated by Westinghouse. k - i 35 i
i-i REFERENCES u 4 i
- 1. J. F. Lime and B. E. Boyack, " TRAC Analysis of an 80% Double-Ended Cold-Leg j' Break for the AP600 Design," Los Alamos National Laboratory document LA-UR-
- 94-2503 Ouly 1994).
l
- 2. J. C. Lin, et. al., " TRAC-PF1/ MOD 2 Code Manual," Volumes 1-4, Los Alamos National laboratory aeport LA-12031-M, NUREG/CR-5673 (in press).
l l 3. R. J. Beelman, S. M. Sloan, and J. E. Fisher, "AP600 Qualig Assured RELAP5 Input j Model Description," Idaho National Engineering Laboratory report EGG-NRE-10824 (rough draft and proprietary) (last revised: June 28,1993). l 4. P. R. Shire and J. W. Spore," TRAC-PF1/ MODI Analysis of a Mmimum-Safeguards Large-Break LOCA in a US/ Japanese PWR with Four Loops and 15x15 Fuel," Los Alamos National laboratory report LA-2D/3D.TN-86-18 (December 1986). l 5. B. E. Boyack to F. Odar, " Minutes of May 26-27,1994, AP600 Working Group l Meeting," los Alamos National Laboratory letter 'ISA-12-94-220 Gune 13,1994). j 6. G. Dejarlais and M. Ishii, " Inverted Annular Flow Experimental Study," Argonne
- National Laboratory report ANL-85-31 (NUREG/CR.4277) (1985).
1 ! 7. R. A. Nelson and C. Unal, "A Phenomenological Model of the Thermal Hydraulics j of Convective Boiling During the Q.nenching of Hot Rod Bundles. Part I: Therinal j Hydraulic Model," Nuclear Engineering and Design 136,277-298 (1992). _ 4 i 8. C. Unal and R. A. Nelson, "A Phenomenological Model of the Thermal Hydraulics j of Convective Boiling During the Quenching of Hot Rod Bundles. Part II:
- Assessment of the Model with Steady-State and Transient Post-CHF Data," Nuciar l Engineering and Design 136,299-318 (1992).
i ! 9. C. Unal, E. Haytcher, and R. A. Nelson,"A Phenomenological Model of the Thermal Hydraulics of Convective Boiling During the Quenching of Hot Rod Bundles. Part III: Model Assessment Using Winfrith Steady-State, Post-CHF, Void-Fraction and i Heat-Transfer Measurements and Berkeley Transient-Reflood Test Deta," Nuclar l Engineering and Design 140,211-227 (1993). i ' 10. Westinghouse Electric Corporation, "AP600 Standard Safety Analysis Report," prepared for the US Department of Energy, Document No. DE-AC03-90SF18495, l Revision 0 Gune 26,1992). j 11. Boyack, B., Duffey, R., Giffith, P., Lellouche, G., Levy, S., Rohatgi, U., Wilson, G., { Wulff, W. and Zuber, N., " Quantifying Reactor Safety Margins: Application of j Code Scaling, Applicability, and Uncertainty Evaluation Methodology to a Large-j Break Loss-of-Coolant Accident," EG&G Idaho, Inc. report NUREG/CR-5249, also
- i. EGG-2552 (October 1989). .
- 36 i
i* 4 i i TABLEI l EXTENDED LBLOCA SEQUENCE OF EVENTS
- Time (s) Event i
! O Break occurs. i
- 0.35 "S" signal occurs from contamment 5-psi overpressure signal. l 0.86 Reactor trip from "S" signal, steam generators isolated.
l
- 2.1 CMT and PRHRS isolation valves start to open. A very small amount of CMT mass flow occurs but ceases once accumulator flow starts.
7 Maximum hot-r7xi peak cladding temperature during blowdown: 1021 K , (1378*F). l 8 Pressunzer emptie's. l 12 Accumulators stah to inject. 17.1 RC pumps tripped. ' l 22 Peak accumulator flow of 408 kg/s (900 lb/s) occurs; flow limited by j choking at the DVI nozzle venturi. I i 30 lower plenum starts to refill. l l l 43 Core reflood begms. I
- 58 Maxunum hot-rod peak cladding temperature during reflood
- 1197 K l (1695'F).
165 All average-power rods are quenched to saturation temperature levels. 190 Accumulator noncondensable gas enter DVI hnes. I a i 240 All maxunum hot rods are quenched to saturation temperature levels. j Accumulators completely empty of liquid. j 265 Core makeup tank A start to drain. l 275 DVI line pressure sufficient low enough that IRWST drain line check valve i can open allowingIWRSTinjection. i j 280 Calculation termmated. Decay power remaining to be removed: 52.5 MW t. I 1 t i I b 4 l 4 37
?- ) ; TABLE II l l j COMPARISON TO ECOBRATTRAC 80% DEGB LOCA !' TRAC HCOBRAfrRAC l l Reactor trip 0.86 s <1 s ) , CMT isolation valves start to open 2.1s 2.2s j Pressunzer empties 8s 7s i Accumulators start toinject 12 s 12 s RCP trip 17.1 s 17.2 s Maxunum accumulator flow 408 kg/s (900 331 kg/s (730
- lb/s) Ib/s) j Lower plenum starts to refill 30 s 34 s
- Average-rod PCT during blowdown 808 K (995 F) 847 K (1064*F)
Hot-rod PCT during blowdown 1021 K(1378 F) Jtp73 K (1472 F) { Core reflood begms 43 s ,j 56 s j Average-rod PCT during reflood 864 K (1095'F) 738 K (868 F)
- Hot-rod PCT during ref tood 1197 K (1695*F) 1125 K (1565'F)*
Time that hot-rod PCT occurs 58 s 102 s i *With code uncertamty included, the.ECOBRA/ TRAC PCT was cited to be 1254 K (1*/98'F) I l 38
la 1 1 j ADS Valves j (1/2 ADS Trains Shown)
- l. .
i 1
% Steam
- Generator IRWST PRHR HX O Proesurimer [ N ( sparser CD o/s 9 4 v
/b -
v 0 1 (O . v Une 8 -- Y
**"'" M -
l W Pumps vesses umA
/
e i Fig.1. AP600 plant isometric. 1 i i 4 5 l 39
! A - Internal condensation and natural circulation transfers heat from the core to the steel containment. B -The containrnent is continuously . ~~~. cooled by naturalcirculation of air - between the containment vessel and surrounding shield building. i G g C - Initially, containment cooling is j enhanced by gravity-fed water
- from tanks above the containment. ,
i 3< 3 C - N l / r i . O * ^;
/> w 1 i
t "
; M i
.1 ce s h+ .
/y 3 4
, t 3
==
EE = EE y- i A g m _
~
g - 3 V Fig. 2. AP600 contaiment cutaway view.
. . _ . _ _ . _ . _ . . . _ m . _ _ . _ _ _- _ _ . - ._ - _ . _ _ _ . !=
i j Breaklocation for ACCs Cuis j ADS 8 DVI Lee 8 y i "
- ~
_ r ('~ [,
- Cold Leg Pressure
- 8"C' L"' 8 i - CL28 ADS A RCP1 RCP28 N a
L ? Reactor ! SG1 l HL1 Vessel i ML2 g l SG 2 __ i . ADS 1 tage d L ADS Stage 4
- pp RCP 2A Coed Log Pressure i e
' ~ Balance Lee A CLig CL 2A l 2 .. .
)(
~
Dvi on. A ( eRwAS ! mws CMTA j 'l' ACC A l
)(
Speger A i IRWST Sump IRWST Drom unee I f l l Fig. 3. Schematic of AP600 reactor coolant system and passive safety systems. i 4 4 1 .. 4 h i l 1 I } 4 I J i i I 4 4 i 41 i - J A
a
= = .
Upper Head > f(//. % y; \. . N
= = .
_ - s Reg 2 Guide A L) R Tubes (8) Ring 1 Ge A U Upper Head Coohng U Tubes (8) Flow Paths (B) MW/ 6 0 y 'j.b , ( 'Upper Plenum/ j/ _ _ - . N Downcomer Reflector Region (Ring 3) y b
%d
/ A)
Core (Rings 1 and 2) ot-Le /
=
=
=
= W / J
= =
% W+ /
, J j
core Bypass and - Thimble Flow (8) ( / j J
,J)
Active Core k * * )
/
(6 levels) ( (' Core Region'/
/ /
% d
- /- - \w s J w /
~
Inst m
- 6
= 9
_ x _ F, .10 _e
= - 5 CC, 5 R. _ ,
Fluid (4) s b . VESSEL i Lower Plenum
~
Component 20 l
- (r=2, 0=8, z=13)
~
VESSEL Component 10 (r=4, 0=8, zz17) Fig. 4. Isometric view of reactor vessel model.
)
42
.= b ~
/ n N
/ _
N GddeTubes k c % N U 16 Upper Head
$3 N N N-I l Space br
% mmectro pipe comporants Doln Hdes l ! _R '3 U
_ s! 'GS 14 M3 !h h i 12 +- Cold Leg Hot L% = 13 l 11
. PSIS Flow Upper Plenum 12' ,
11 9 p , n, nmen merrn_r -- Unheated Core Region q 10 8 j l Darncane r l-- j 9 7 / Axial Level Noding l Relectr l 6 , Blods % ,! 8 I _ i ca. syne-.re 7 , S.,# TNmbleflou Heated Core Rep on - 6 4 l I Refleda d ' CartyFlud 3 _l-- - 4 2 U l _ l Unreatad Coe Rogon and I I Lower Cae Supp)rt Ph to
' os y '.
.t.
; j errporurts g
Lower Plenum \ /
\% ' /
Y J 1 2 34 Fig. 5. Elevation view of reactor vessel model, including downcomer noding. 43
J 4 s i . 1 Core Vessel Plan View Downcomer Vessel Plan View Modeled with 4 radial rings and 8 azimuthal sectors Modeled with 1 radial nng and 8 azirnuthat sectors using separate VESSEL cornponent t DVI B l Cold Leg 1 A , Cold Leg 2B Core Region 3, l
~
, A. i. M'5' e s.- . . ro mo 7 ) MM- ?'U: .i. :M. :( - 8 6 m .
/
~
_. 3 :7 /: . ' N!:M "
-. .): J: Nif .': - 9 .-
ege M+1 CEC * ' ciift:g.2ii i Hot Log 1 : s Hot Leg 2
~
- l:Q:i 1' ;
^
- i. : ..I '
l "- .C7: 2: / :_: \ :' JC "'
. 2 a
. 3
~ l o n '
l as ilf*f i _. Cold Log 2A Reflector Region Ring 1 not used DVI A s l Ring 4 not used in axial levels of core and upper pienurn : i Fig. 6. Plan views of reactor vessel model. l! I l Upper Head @ j Core Support Guide Tubes (16) @ - Q Columns 1 g Upper Head Mount 6ng Fiange Thick Vessel Wall Upper Core Support Flange l Upper Support Plate h @ l i Upper Core Plate ! Core h 0*"*' Downcomer Annulus h j i Reflector Block Fuel Roos Coolant Power l Deposition , ' Control
' RMs q
I Lower Core Support Plate- ! ' Vessel Wall Lower Head Level 2
- j
- Secondary Core Support @ g
- and Energy Absorber i 33 heat structures total
; @ Lower Head Level 1 Fig. 7. Reactor vessel heat Structures. } ]
44
RC Loop 1 Components g 8.c
@ - @ 8 '@ =
L . e b: O
~
g _ _,_ =f = -- O - e, - d @
@g A @ -1~
fCal g gg e . _
~ J \~p 1 --
- iw0 g Q/ "l.*.* ,,
( 6 - 8 - u_ e e. _ .. , e l@ 1""*'
- OO
- -rh - e ::::=
80 i 8 ca = '. i 1. - g-%
- ~~"hg @@ i 9 d'=
@ O eves ,,c ,,,,, ,, @ *" *
- Fig. 8. Reactor coolant loop-1 model overview.
nc Loop 2 Components @O g = @ 8 ig - e--
= ~ ~ * = ~ '
L .
@ @ D ==
e_f
~~ b@m@O
. -- A
- a @
1 k=l g gg
\-p i - imO
= 4 i
=- . e -
l @m 6 _.g ,t {B y , =,,,,,,- ,
- 1. - g y-
! WM ,, , 8 00 *** @ 8 8 @ e- se : 2 - 0 3 A=- etp e o1-80 n = ' ' ' ' = "* *' g g i is g
'g" ,,
@O O@
Fig. 9. Reactor coolant loop-2 model overview. . 45
. - - ~, -. . ~ . . . . _ _ . _ . . . . ,.
,o I d 4
.......................................................... Note: Eacn system hee two separete trees
- Automatic DepteOSuftEstion system *
- !O @ .w. . . . . . . .One..tresi shown for seen system
. *- " ~ @ @ @ @ ;g g *. *'*
Paselve Residual Heat ; a R.m.v.i sy.i.m i i . I @ r p@ +L f-' Le-o... Y
! : @- @l i e,,_e!
j@
.-e. n- :
any ( g ;! en nm e _P f-a-ne@......... s .y m @
. .-- - o .,
s : 2,, . !
- e... ........;@ :
@- D @o: -
@ A.. .... sos e Water Seerage Tent ia i C,, ass 2.
sum.s temasse en aus es ) g
"' pawst) : ,,,,,,, ,,, 6., .
...........mm.ww.......................
( ,......................................... .................................................., b
- i A @ !
- r T j
- Psecoureer neur.T Dem.Lt. 1 e r e. e-- !
l V- ! om
~'"""*
o., ! . ! @[P" @ FC g gg " @""*m s, !
**""'" h @ n j 1."L ,,,
n.,4~,, l ~ lj
- @ t
, Og :
, i
@ @ i.
q :
- P 6 sesm..yo.n : ;
............. .. . e.......sy.w m Fig.10. Passive Safety Systems model overview.
' a. 1 { t u a 2 - 3o I,a, i Top of wper pierum.,, w a- ' f e
'..'p *
- g. , . ' 0% cornertne ---
. f - - ti - HoHog corecere -
/p;jjj r-3 r-
% t
- h DVloereertrw-- E
'y:::;;.,y f; mas'94hj::=
4:'. 20 I e . : .+o or or m.m w a 8 l1 cor. woi .
- !_u.. T. or w enr==
- o. --0 use-mw
- , flD1LL , ftFLWJD , C00uMG
! o 25 50 75 10 0 125 150 D5 200 225 250 275 300 l M 4) 4 Fig.11. Vessel-collapsedliquidlevels. 1 - - 2450 l LPPDt PLDAM
====. ..........- . ,,,,
- BROKEN COLD LEG 2B
- STEAM CDERATOR SECOMWtY 4 12500000, ACCUWULATOR -,-
- 3
===. -
1050
; \
25 0=.- N .-350 i 1 s'-- ___ _
~'% -
j O t.ONC-TERW d ,8tE7u , ItFLD00 , C00UNC 2500000 l 0 25 50 75 100 125 50 T75 200 225 250 275 300 M h) Fig.12. System pressures. 47
- . _ ~ . , . - ... ... = _ _ . - . _ . .- . _ _.
j
~
HOT R00 PCT j 200 - ** ******* AVERACE ROD PCT 2 CORE AVG. Liou3 TEWP. _ goo i b tm - k w 1400 3 1000 _ 1 1 4 I
. goo O '
900. Sco* r** ,.
*.s.,
- g. .
300
. ,: d so0 ' :. , .
oo ;
. l i S00, 400 3 l 400
', ,,_-,...~-.~-
, *..6.... -
, REFu , flEFLOCO C00UNG m 1
! o a 60 3 a =$ .o = 200 m no m = 4 TNE (s)
. Fig.13. Fuel-rod maximum cladding temperatures.
i
, , i i i i 3000 1
20 , mo l 2o00 4000 l 4 6 noo -
- 1 I
3000 g i 1 d d h ". -
~
2o0o 0 s 3 m.- .
.0 0 , --o I unc-ww C00LM El, flETu , IIEft000 ,
-m 0 2S SC 4 2 US 1ho W 200 22$ 250 25 J00 THE (s)
Fig.14. Pressurizer mass flow.
i a b e 1 '
. ACCUMULATOR A . goo
- .......... AccuuutAtoR e
, 350 - so
- m -
soo m - ! l 2 o.. l h ".- . m h j = - M 50 - .i 1 ! 0 - --O ! , , , , , i i i o a .0 3 mo as no ra zoo m m := m TE (s)
- Fig.15. Accumulator mass flows.
4 4 l l ACCUMULATOR A l
.. ...... Am na e ATOR B
)
, 3 _ i J a 1a 2 - e W s 4 . . s ~ e W E
, i -
g i ! o.- --0 e i i I LDNG-TERW j s gru , artooo com
- .t 0 25 50 3 10 0 12b 20 15 20o 22b 260 25 300 TE (s)
- Fig.16. Accumulatorliquid level.
i 4 i s 49 4 4
i- ' ' ' u ' ACCUMULAT08t A
, .......... ACCuuutAToa e -
- o. -
g V
- j. E 0.6 -
i c.4 a u . 0 - aCRDoo , CooLMG ;
-n lM; fitTLL_ ,
o n so n no m no e oo m m m m WE (s) Fig.17. Accumulatorliquid volume fraction. l l CORE MAKEUP TAM A 2co
.......... CoRc waxtup tam e
.o 6o
- 12 o 6 so . 6
- - So 20 .
4a
@ l @
s s
- o. u. q.. ..o 2o
. anu acnooo , ucoau o-=c
-4o o a so n eo e no = 2o0 m no m soo m:(s)
Fig.18. Core makeup tank mass flows. 50
. _ _ _ . _ . . . - ....._m__.= ~ , _ _ _ _ ___ . . . . _ _
m a T 7 CORE MAKEUP TAN < A
., ..-------- corg. MAKEUP IAM 5 -
20 4
. 3 .
~
a [ -
?
v e - b! i l 3 .. 10 5 E g . - 3 j 1
- unc- =
, =ru , etooo , ame
, 0 0
- O 26 ho 4 10 0 12b 50 Ub 200 22b 2bo 25 300 l Tat (s) 4 1
- Fig.19. Core makeup tank liquid level.
I t , , , , , . . . 9 3, s . . no a : PRHRs rLOW NTO RCS AT Sc PLINUM ' Joe -
.......... mes rtow mo RCs At Not tic s00
- 230 ; -
'1 . l 200-
*o 1
a. I
^
1
, so
~
- @ ! ') h
, a : - m , 2 ' [ I 's
. Q. . 1I ,
0
. . ..... . .. . (. ._l. . ._0 .
. 30 - unc- =
,eu, wrto0o , ame
. i i
,g 0 2$ SO M 10 0 123 50 15 200 22b 260 25 3o0 Tac (s)
Fig.20. PRHRS mass flows. - i i 51 1
- C 5
i
~
Cur PBL A
.0 ,
.......... Cur pet e 3 . -
no i . - 3 e0 0 . 3 6 l
, T E
50 20 , I;! i_ I E i 0- ! --O m w j' i : 1
-20 -
-30 1
-40 - -
.0 u c-=
,cu et0m coaac
-0 0 26 50 3 10 0 125 to 175 200 22b 250 27b 300 tee (s)
Fig. 21. Core makeup tank-pressure balance-line mass flows. L E I Torats o<rww . 0w0 l
.......... nup sot een rww vcssa soc seu now i 1
e=0 - 30000 I d 2000 - ~ 20000 h h e 3 **. j .
=000 a 2
0-- 'Y ' d '- ^ T ' ^- --0 _. _ . u _nc- =
-s=0 0 a 30 m m0 m 30 m 2m m 250 m 300 TEE (s)
Fig. 22. Break mass flows. 52
O I J
, i g i i
- PUWA SCE i
I b 5 N i$ ' : :- :- : i ::ig! ::i -
- - n: :.
0 e .::
- . ! s ::::!,.:: e
- g.
g: ! j
~#:: : ::: :, .:...:: : . ,
.t ~ .,
- C' ::: : *I: :
I U *f
- j:, ::
h 0.6 l
- :: : ?
- 1 n E ' %;
- : ::= .
0: ::" :.
!: j 0.4 !! j:ji 4
~. .
02 n . 1 : O t . LDNG-TDtw RtrL000 C00UNO 1 -02 0 2S SO 2 10 0 12 5 15 0 f5 200 226 250 24 300 i TNE (s) 4 Fig. 23. Break exit voiding. STEAM EpstATOR 1 s000.- . 0500
, .......... STEAM EPOtATOR 2 i :
i 7000- j 15000
- l 6000
- i. .f .
12500 5000 - wStenay-Sanes Plour - 10000 4000 - - 7500 1 3000 - - 2 1 5000 ' 2000 - l l
', -.2500 g- '
i . 2 0 -a .-0 ' LDMG-TDIW E , REru , RtrLoco , C00LNC l 4
.,000 4 0 2S So b 10 0 12h 16 0 T4 200 22b 250 24 300 '
l WE (s) 1 Fig. 24. Steam-generator primary-outlet mass flows. 53 a l
- .. - ~. . .- --_ .- . .- _.
i 4
, I & I 20ooo 750o . HTACT COLD LEC 1A s
.......... NTACT COU) LEG 1B
, WTACT COLD LEG 2A 5000 - BROKEN COLD LEG 28. PUMP SOE .~cooo i BROKEN COLD LEG 28. VESSEL SOE noo em - h -noo
"~
,. ig e .qmq, 1 - - --
'v - "*
f i 0 f
.sooo-.l
-= @s j _ . _ _ -
, -noo -
-2oooo
-mooo l
-" un.c unu
~
,atTLL , RtrLDoo CooLNC
-mooo i o 25 So 75 too 12 5 eo 17 5 2co 225 25o 275 3o0 i 4 l TWE (s)
Fig. 25. Cold-leg mass flows. u
- l 3
NTACT COLD LEC 1A
.......... NTACT COLD LEG E o.a NTACT COLD LEG 2A 5 h -
l g o. - E f o4 0.2 - o h E ana_nnu s , =cru , atrtooo , c= Lac o 25 to n 100 125 eo to 200 225 2so m 300 TME (s) Fig. 26. Intact cold-leg voiding. t =.
- . . . -. .-. . ~ . -- .~ - -- - . .-- .- _ . _ - - _
1
- l C i
i
, , , , , , , , , i 1200 5000 -Senney-temas Row FLOW AT VESSEL
.......... FLOW AT STEAM GENERATOR
* )
l
. 10000 3 4000 -
1
. - 7500 l 1 3000 - -
l j 4 -1 - S000 2000 4 g -.! - 2500 l
\ l t 2 I 2 1
. _. _ _ __ w ,:
. = ,-- nl.
3, 0- ,- -,,,; 3;.
--0
-iOOO. .
_2300 4 i
-2000 -
tDNC-TDeW
- .wru , art 000 , c00t=
3000 - 0 25 50 75 100 125 15 0 f7S 200 22S 250 275 300 1 TNE (s) i Fig. 27. Hot-leg 1 mass flows. l 4 J l 4 ,000, _ 1 ,I, j FLOW AT VESSEL 7000- .......... FLOW AT STEAM EPERATOR
~15000 8000 -
a500 5000 - *---Senney Sense Piour - e . . 10000
}
=
4000 n00
}
6 9 3000 - g . . 5000 g 2 2 1000*- "2500
" " ~ ~ ~
0- --O _,000 - to.e-n=w s , wru , art 000 , c00tm 1 _2000 0 25 50 75 10 0 12 5 15 0 175 200 225 250 275 300 TNE (s) Fig. 28. Hot-leg 2 mass flows. 55
e
, i i V0ONG AT VESSEL
, - - ---- VDONG AT STEW TNERATOR j n: j ' o! jj F i :i' pm si::Aj t pi f
1 . : :
- =, a: .
- . ! :s :::l .
04 J : :: i : 5:5w s- .
..!: ! f $ !:f !
'd 5
05 i ji ' H ili-i i: DA l 1 02 - 0 - LONO-TERW
, RETu RETLOOD , C00UNC j
-0J I O a 50 m 30 0 in 30 rm 200 2a a0 za 300 TE (s) !
I Fig. 29. Hot-leg 1 voiding. l l l u ' ' _, j l VDONG AT VESSEL 1
, , ---- -- VODNG AT STEAM WPERATOR ~
gg
'g' gj'- .
s' : ; t ?a* E I 0.8
- j jj ,-
- l:
h
- 'h 0.6 '1 -
l l 04 02 - 0 - o LONG-TERW
. d , RETu , REFLOOD , C00UNC
-02 0 25 50 4 10 0 12 S 20 im 200 22b 250 24 J00 T*AE (s)
Fig. 30. Hot-leg 2 voiding. - 56
i I i e TOTAL
. .......... m ss m a .
.ooo )
...- .. _ . ACC AND CMT N <
2Soo .p - Sooo ) mo -
.ooo 500 - l 1
l l l ~- ,. 1
- m. - c, ,
mo j Y- ..o i
..l -... .:v.w w l
, =ru artooo , "ER7.E"
-Se o 25 So M too 123 no 175 2oo 225 2So 25 Joo TWE (s)
Fig. 31. Total makeup flow.
- ooooo _
l mooo
.moa
'ma -
[ mm 1
! h h wa.- . ioooo l o.- IJ.b.g)alf b " .-o
, atTu atrLooo co0 LNG
_Sooo o 25 So M too 12 5 no 15 2oo 22S 2So 25 Joo TME (s) Fig. 32. Net-system mass-flow loss. 57
c. i
,anoo , ,
1 4 -.
,oooo . -_3eooo goooo-- -
27oooo
-. -.22Sooo b ..._
h g _.. , 5
.oooo- -.issooo g
_ ... _ g 3 2oooo- --45000 ', o-d -a W , arru , arrtooo , "coau"c"
~'
o a so a no e no e zoo m 2so m aoo j E (s)
- Fig. 33. Integrated net-system mass-flow loss.
u - - - - - - - - A . o.8 . I 5 f t l i " ~ I fa . I i . o2 - o -
._ =
o a so n eo m no a zoo 2n no m soo TNE (s) Fig. 34. Heated core-average vapor fraction. 58
e=0
==0 TDrAL
. ... ..... two -
.0 .
. . . VAPOR n000 3000 2000 - ~
~ 3 l
]
{ sm Ll i, -~* f!(V
*' V I >
r ._. y 8000
-4000. -
l 5 j e , eu s to0e " cot?
)
I
-o=0 0 2s so n m0 as no e 2m 223 230 m 3m TEAC (S)
Fig. 35. Core-inlet mass flows.
.000 TDTAL
. . .. ... ... LWD -
2000
... .. . vApoa n000 4000 9000 l 3
- E= .
_ } m . - 3000 o- - = hk --* 3000
-2w0
-40=
._ __ me .
0 25 50 75 =0 12 5 60 T75 200 225 250 275 300 WE (s) Fig. 36. Core-outlet mass flows. 1 59
go, 3000 h>LR CoRc RNc
........ a m acoatnue o 00 2000 I
J m 500, j, l
. goo
, kR jl !
! Il'l .:
3 < . .. ,q ,
- p
"~ t 3 t::.:i]i ;g. :. !!n::
I i :,
-~
- E . bi
,il!.'MfI i Edij., $
4
' 4 : p:. e ii ;
- h
_S00-yi ll' '
- r:a:::liig!"
i j j : t; j j
- :is : j;i!!!!.tl
-o* @ !I' i i i!
2000 _ o 00 LONC-itnu E , RETILL RETLOOD , C00 LNG
-t500
., 0 25 50 M 10 0 125 50 15 200 225 250 23 300 l TME (s) ) l Fig. 37. Core-outlet liquid-mass flows. l 1
, 1 I i I I I f 4
ru nconcnuo
- concnua -
30 300 - k 800 200-eo jf - ~ f a g w - ii - 300 8 g m 2 e0 - W 2
. . go 0--
h_ --O C LONG-TERW E RETILL 1 RETLOOD COOLNG
-u 0 25 50 M 10 0 125 50 175 200 225 250 25 300 TIME (s)
Fig. 38. Core-outlet vapor-mass flows. t
Q o.o75 NET REA7., M*Y
.......... CONTROL ROO REACTMTY
. ... .. .. . FUEL TEMPERATURE RE A7., MTY o.o5o - - COOLAN* TE99ATURE REA7., MTY
-- v00 FRA"., TION REACTMTY o.o,3 .
e .h, u .. - g . . . . _ . _ .
~
h . __ ~ -A%wAAwA^ VM\%*^# y -o o25 W
-0.oso #
-am5 . . . . .................................................................
-o.too 4 -
1 e , eu e tooo ucoot o.nc.
- o.125 o 25 So 75 10 o 12 5 So 175 2oo TME (s)
Fig.39. Fuel-rod reactivities. isso OT=Os "00-^ ... isoo
. *T=1s iooo-. *3 +T=2s -.uso E .
. p
=T=3s
-N'N'N N-1200 4
- A , ., .y.4,
~
m w
- g. .,
, ,7,3, 1050 b:' soo ~
= - 4.:.:.:. - w
- ... : ~
,oo
-. m,__ g
.-._ . .,.os .
73o
. *T=Os soo.- 'M *T=us
. goo
, , i ,
g o.o to 2.o .to 4.0 s.o s.o i ROD RADUS (rr$
- Fig. 40. Stored-energy distibution in fuel rod with time.
61 I.
e R ROD Z= o.ooo
" ~
~r z. zow
(. , l Z= 4.000
$ m -
2- 6.000 0 .
, K _
z - s ooo s 9 g J z . nooc
$ 30o _ -
N}% . 'z - 12.000 x L ,b L
'~
.Az - -
=
Roo g , 0 -900 e 25 So M too 12 5 15o *75 2ao TWE (s) Fig. 41. Average-rod temperatures in cell 1 (r = 1, 0 = 1).
' ' ' ' ' l soo R ROD z- o.000 j Mo -
f I - 2.000
' 'Z = 4.00o g ,
f Z= 6.000
. ) _
"2 - 8.000 h.
N W z - 10.000 g Soo . s . Z = 12.000 N w ,
~
1 j ^_ _;
==
, , , , , e-900 o 2s so a 100 12 s no n zoo TWE (s)
Fig. 42. Average-rod temperatures in cell 4 (r = 1, 0 = 4). 62
O I
- R ROD 1
o 8 f7 2- 0.000
~
MO ~ l D gg _ _ Z = 2.000 l 'Z = 4.000 g _
' Z = 6.000 j
< 800 m l 2= 8.000
~
W 2 - 10.000
~
h % _ 'z - 2000
.00 ll 5= - N - e__ _ _ _--
400 ,
=
w . = - - . "- 2 %:b.b h_ t
, , , o .,m ;
O 2S 50 75 X)0 12 3 BC *7b 200 TWE (s) l l Fig. 43. Hot-rod temperatures in cell 1 (r = 1,0 = 1).
- R ROD _
8
=0 -
"z .9.0m O
x300 - . , z- 2.0m g , ,
' 'z- 4.0m k '2= 6.000 m .= .
e z - e.000 W
- 1 Z = 10.000 W
$ .00 l), : 'z- 20m Sm - o N w ___ _
<= -
,p b-A 54 . ;- _
, , , , o .,m 0 25 50 75 10 0 12 5 15 0 ~75 200 TWE (s)
Fig. 44. Hot-rod temperatures in cell 4 (r = 1,0 = 4). _ 63
C l
! l l
"O R ROD 2 = 0.000 em -
o 2= 2.000 2 = 4.000 E m -
^
g 2= 6.000 h '2= 8.000
~
- T g j. 2 = 10.000
- 2 = E000 S00 -
, s l
1 400 -
~ $ f. ~. . . . - . _ _ _ . __ - I R00 l
, , , , , 0 =900 g
0 25 so n 100 125 15 0 *3 200 : i TWE (s) I Fig. 45. Average-rod temperatures in cell 9 (r = 2, 0 = 1). o00 R ROD 8U "2= O.0x 8@ - o 2 = 2.000 J
'2 = 4.000 )
E = - y 2 = 6.000
=
d - 2 ..w0 g-s 3( s[
- 2 = nom l Soo -
N
- 2. u.000 igNb - - - : . _ . _ ;; _
R00
, , . o-*O O 25 50 5 00 12 5 150 "M 200 TWE (s)
Fig. 46. Average-rod temperatures in cell 12 (r = 2, 0 = 4).
o a aso
- o T = 0.0 5 300 -
- T = 2.0 S 7
z ,,o . g ..-
.~ .
g '
/ .
= T = 6.0 S q ..'
e ~
~
- T - 8.0 5
- - ' T = 10.0 5 ,
.so A*',,...-
."G 4 y
soo -,,,,,....'" '*..q '"6....., t t t e a 263 27 27 3 26 28.S 29 29 S 2S 26.5 26 ELEVATON (m) Fig. 47. Cell 4 average-rod temperatures vs rod elevation, O to 10 s. i ,
, e t
g
#-4. ..', , .. . . . .
300 - ;1'
- g *
' ' *** ,. - ' o T = 20.0 5
/ N a T = 40.0 S
/ ;
E
/ . T = 60.0 $
Too - j ,. g q
/4 I . - ..
c.- . af=50.0S 2 i 's
--* ., b.,
~
f * .'
- T = 100.0 5
~
w , ! , j h, ' T = 120.0 5
/ I ',$
- [,/
500 -
= T = wo.0 S j ,
jf i a 1 = A .. .. . * ' a as a as a vs 2e as a as ELEVATON (m) Fig. 48. Cell 4 average-rod temperatures vs rod elevation,'20 to 140 s. 65 1
O J 4 1050 i
* . . W * ~
gg . _ o T = 0.0 5 1
- 900 - - . -
' *20I gg , ,
4 .. * ,
- T = 3.0 S
* = = .0 S 800 - -
\A *T=8.0S no - ! ,
o .'.
- T = 10.0 5 g 700
.30 / 6.,
s...,'
,.a- _
soo . . SSO g , a m a ns v vs n ns a as ELEVATON (m) Fig. 49. Cell 4 hot-rod temperatures vs rod elevation,0 to 10 s. 1200 - -
.- . ..- . T . 20.0 s "8
~ ~
((___g- . , , , , , , i=0 g - g n
,o , .
f.*
~
~
= T - 80.0 s a : :
(e 8" -
/ ,
. . . T . iOO.0 s g * 's, -
' T = 120.0 S f l ,
a . .
.00 300
,,/ // :
l
=.
'w
. T . u.0 s 00 _ . : g .. .. I .
, i 1 i a ns a ns v vs n ns a as ELD /ATON M Fig. 50. Cell 4 hot-rod temperatures vs rod elevation,20 to 140 s.
66
c 1 o.9 o.8 - c.7 Y o.6 8 oh - od - ll u - 1 a.2 - o.1
~
3 eu ewoo "dL" c
~'o 23 so a eo a no rm zoo m a m m TNE (s)
Fig. 51. Downcomer average vapor fraction. I i i t f I [ l 2ooo
". . oo, ,
W. ~ Coo 25o. . Soo f o--w yi - kg ' l l
.-o e
.2so-1 j- -= 8 g
2
.,oo -_ -moo
-75o -
l.
-ooo 1 y
e , eu emoa , "c6L"
~" o 23 so a wo e no e 2o0 m 23o m u TME (s)
Fig. 52. Total guide tube flow. _ 67
1 O.8 i l 0.6 - 0 2 b l j 04 l n f
.. . . . -. -_- _. _ - -_.s _
E um-m come e , mu etocc ,
-u 4
0 2S 50 M 10 0 12S 20 14 200 22b 250 24 300 mE6) Fig. 53. Lower-plenum average vapor fraction. UQUO TEMPERATURE cSO.- -- - - - SATJRATON TEWAE4ATURE . a ~ 600 - 600 i. E m - 300 E m 5 -
.00 g .
- w .0 .
g 300
- '00 -
~...~.....
- 200 b LONG-f(RM e , cu miooe , comm 200 0 25 50 M 10 0 12 5 50 f5 200 225 250 275 300 NE (s)
Fig. 54. Lower-plenum coolant temperatures. 68
O u - - 1 0.8 5 l O.s - f 0.. 0.2 -
]
1 0 d
!,=ru = 00e , "com'E -02 0 25 50 M E 12 S 15 0 175 200 22S 250 27b 300 E (s)
Fig. 55. Upper-head average vapor fraction. I I I I 1 9 t i - 0.a - M - I 5 0.s h 4 E ! 0.4 C2 t 0 e,=ru = 00e "cd'#
-02 0 25 50 n a nS e0 in 200 m 250 m 300 M (S)
Fig. 56. Upper-plenum average vapor fraction. __ 4
1 o { l i 1 y FEEDWATR FLOW 1 WA~ER FLOW 2
- e AM ROW 1 -500 700- - { AM RDW 2 r.: -- * .
eso 300 000 4 ^1 s *o u a h - 30 h y m - m s . 200 - 250 go- . 0 L_ --0
.- to c-rta.
E , _REF1LL_ RETLOOO , C00LNc
-.0 0 25 50 M 00 12 S 20 T5 200 223 250 2M 300 TNE (s)
Fig. 57. Steam and feedwater mass flows.
, I ! I l 3 i 1 i 0- --O
-250
-.- -. -iom s -no -
O a -
-20w g g -e00 r
< W s 2
-oso -
. 0
- 300 -
_r50 y -
. 45 Looec-ftnu a , acru acn.000 , coa c
-mo 0 25 50 3 10 0 125 50 f5 200 225 250 275 300 TWE (s)
Fig. 58. Total upper-head support-plate drain-hole mass flow. 70
O g 200 SD,- .~ go
--C 0- _
8 8 d c
-_ - = @
s 3- s __ lit
-200
_m LONO-TERW
-so -
0 25 50 M M 12S 50 15 200 22S 250 25 300 NE (s) Fig. 59. Downcomer-to-upper-head mass flow. I 1 l 1 l I W 1.A 400 . . . . . _ . . - PUMP 2A
', - w 2s 350 -
T. m -
. \.
m . ,
\'
m
\4 .- sx
, /-
- ~ ~
.. _ _ =
0 0 2S So n a a$ .0 rm m m a m 300 mt (s) Fig. 60. Reactor-coolant pump-rotor speeds.~ 71 t
rm oooo 245o WOOBRA/'RA0
. . ... .. . . . me 2c.
12Sooooo, . f75.
~
uoo ka nooooo.- . oso soooooo.- .
. E
_.. .. .,,o l _nooooo o 5 m 16 2o 2S .), 3S 4o l TNE (s) l l l Fig. 61. Comparsion of ECOBRA/ TRAC and TRAC upper-plenum pressure. 1 f. 9ao 4o0- .
. . . ........ nm 35o,-
l
. g 3oo .
m 3 l . . - s #c -
; - . 1 s . . . . . - . . . . .'". . . ,
9 2ao.. : -.450 o N "m 3o0 $ co - so So - o --a t I I o 2o do 6o So to 12 0 wo 16 . TWE (s) Fig. 62. Comparsion of WCOBRA/ TRAC and TRAC accumulator mass flow. 72
es 22
. y W^@ " -
' *""**"* M I p ?i.ll.,O.::.?f.iN\lhlVt*MW .
f*/ I ' 6 175 -{- ,, ! 'd
* ;i:' l)' :: *,.' s' %
- :([ t
- ' ?
-_S uo-f3 i
$ s.
E 'i. O o Y L25 {l\
<?
- 4 d ,. .
a 1 ! ,! $ i :: 3 E
, :- 5 0.75 - ,'. ,3 lI -
2 0.50 -
- 5. . 1 025 gii Mt 0 O O 20 40 60 80 10 0 12 0 wo 16 0 mt(6)
Fig. 63. Comparsion of WCOBRA/ TRAC and TRAC lower-plenum liquid level. g- -_30 h,. .......... m
- .", s 8 *: ; -
i f-1 :" i
) : :; :.: : I M
7 -
- :. : :: r.
- . : : : :: ,t :: .
- :: ., . ' ' ::,,3..:d
- :::.1
- :: :!':i 1 5-
- 7. : :
- i::! i::
- .i -
20 o o g -k:i. :: .. : ' '4; .:w.::!:::n::::
- 1!!. ::::1..!:-
m :. : :n ; W w p::s
.. :: e
- :"r 3. :- .
J S - l - s a . :. ..- - e e< 4 : ; -l i a =. m
- e. .
3-- : l* ._ to
* *J * '
2 -
\b:::4,*;lal:,N' f t v
. - 5 1
0 O O 20 40 60 80 10 0 12 0 WO 16 0 l NE (s) l l l Fig. 64. Comparsion of WCOBRA/ TRAC and TRAC downcomer liquid level. j I I I 1 l 73
O , l l 1300
~
wCoeRA/ TRAC. e 1 l 1200 ,. -- -- TRAO. HOT RO3 4 1600 p g . . g ,
- a
- 1. ' .
u00 Q l E 1000 , ' .
\pl. .
1200 N 900 I
"N k 800" I
4 2 700- --800 i Q i l 2 500 I . ,oo 500 O 20 40 60 80 00 12 0 MO 16 0 TWE (s) l Fig. 65. Comparsion of ECOBRA/ TRAC and TRAC hot-rod peak cladding temperatmes.
~
1200 WCOBRA/ TRAC. R001 - 1600 330 0 . --- -- TRAO. CEM.1 HOT ROC
- .* r.- TRAO. CELL 4 HOT ROD wm we E ,x' .
.u..
12 " p
..-..,, ,oo, g
; ,oo. 'i. ,
~
a . y g 7m- .- j .. 00 8
= .=. ,
i .
. [
wo.- l . 00 400 L - soo 0 20 40 60 80 10 0 12 0 MO 160 TWE (s) Fig. 66. Comparsion of EC OBRA/ TRAC and TRAC hot-rod cladding , temperatures at 6-ft rod elevation. 74
m0 500 WCOBRA/TRC. ROD 1
" ........... mC. CEL 1 @ ROC TRC. CELL 4 HOT R00 200
'a _.
um - M
- um $
.00. m a \
=. -_1000 h m
f
........ .. .., y
.. 1 sw. ., .
sw m .
, , i g
0 20 40 80 80 10 0 00 wo 20 NE (s) Fig. 67. Comparsion of ECOBRA/ TRAC and TRAC hot-rod cladding temperatures at 8.5-ft rod elevation. WCOBRA/ TRAC. ROD 1 200 l
,, _ .......... me. Cou.1 @ ROC -
TRM. CQ.L 4 HOT R00 Woo l g)og .
~
E
~
.= -
E
- -_1000
< 800 ,
4 E h ,,,. ., I
..e00 f
~
e00 $ m.- : ,oo em i ' g i O 20 40 60 so 10 0 120 WO 20 NE (s) Fig. 68. Comparsion of ECOBRA/ TRAC and TRAC hot-rod cladding temperatures at 10-ft rod elevation. 75
= soo WCOBRA/ TRAC, R00 3
. .... ..... TRAC, CELL 1 AVC. R00 - 105c e y:- - TRAC, CELL 4 AVC. ROC soo , , ,
r !
.oo n
E = b 75o 5 '. Y q , . . . . . . . .. .,,.. -
- ,. g
.co.1 ,
..- .s,
. ,oo m
- ..- : 5
- ! .- i w
@ ,oo..
/ { ._eo
[ s 1 i.. . . - 300
,oo . ............ -
l l o 2o do ao So too 12 0 Mo Iso ) NE (s) Fig. 69. Comparsion of W. COBRA / TRAC and TRAC average-rod cladding temperatures at 6-ft rod elevation. WCOBRA/ TRAC ROD 3
.......... TRAC.CEu.i m ROD .
ioso
. ....... . . TRAC, CELL 4 AVG. ROO j
eno . 1
^
. goo b 700 v. -
E w g . .
=
q . x . g soo.r..,
. . ,oo i y ..,
$ soo-. ,
,- l ,,, -_a soo soo a 2o ao Go So too 12 0 Mo teo TNE (s)
Fig. 70. Comparsion of .W. COBRA / TRAC and TRAC average-rod cladding temperatures at 8.5-ft rod elevation. . 76
1 o 900 WCORRA/ TRAC, ROD 3
- .......... TRAC. CEU.1 AVC. ROO -
10 5:'
- TRAC. CEu. 4 AC ROC _
900
$ M $
m g g
*C E. .,
f , . . . . .~600
.... w g -*.. . 4,0 g
,00 .
- \
M0 400
..... j y
0 20 40 60 80 10 0 12 0 wo 16 0 mt (s) Fig. 71. Comparsion of ECOBRA/ TRAC and TRAC average-rod cladding temperatures at 10-ft rod elevation. WCOBRVTRAC. ROD 5
. .......... mAC. CEtt , Am Roo .
1030 TRAC. CELL 12 AVG. ROD _
- .- 900
$ M -
Q
. ps0
~
a s. ~
=
E H 600-g. 5 . . ' ' . 500 y W
,00-. ./
450
'." . .- : \ - x0 400 -
i i 0 20 40 60 80 10 0 12 0 14 0 16 0 E (s) _ Fig. 72. Comparsion of ECOBRA/ TRAC peripheral-rod cladding temperatures to TRAC outer-ring average-rod cladding temperatures at 6-ft rod elevation. 77
e 900 WCOBRA/ TRAC. ROD 5
. ..-....... TRAC. CEu. 9 Am R00 -
105c
- TRAC. CEU. U AVG. ROD -
800 - 900 E m E 750 lw j - em W .
,l ,
l
$ ee f
300- . .;
. j
- 1
- j
- h. .
= ,
I 300 O 20 40 60 80 10 0 12 0 WO 16 0 4 TWE (s) Fig. 73. Comparsion of ECOBRA/ TRAC peripheral-rod cladding temperatures to TRAC outer-ring average-rod cladding temperatures at 8.5-ft rod elevation. WC00RA/ TRAC. ROD 5 ]
. .......... mAC. CEU. 9 Am ROD -
1050
. _.- .. . . mAC, CEU. U AVG. ROO 900 E m -
F g . m a e g om.. .
.00 g
y .. ... .., w n . .... , g ,00. ,
...o g
. \; -
m a - ... . . . . . . . . . . . . . . . . . . .
, R I O 20 40 60 80 10 0 12 0 #0 10 0 TME (s)
Fig. 74. Comparsion of ECOBRA/ TRAC peripheral-rod cladding temperasres to TRAC outer-ring average-rod cladding temperatures at 10.ft rod elevation. 78
APPENDIX A CORE-INLET MASS FLOWS Figures A-1 through A-16 depict core-inlet mass flows at each core sector cell. A noding i diagram is provided in Fig. 6 of the main text. UOtJD WASS FLOW
.......... v m m . tow
- 300 -
' . goo 1
- 450 200 _
4 6 . m } l m - l 9 1 m s 9 o- 4 l, l --" I j l j 4
-20
- 10 0
.i . -
-m 0 25 50 M 10 0 125 50 U5 200 225 250 275 300 i TWE (s)
Fig. A-1. Core-inlet mass flow, core sector cell 1 (r = 1, 0 = 1). A-1
400 UCX.O WASS FLOW ,
. ........ vApon uAss tow 300 -
l 600 4 450 200-
$ T 300 too
? -
q l i , l l1 ' ee 5 o-
-j h .
i j I l l
'l
--O
_ e0 3
.c0 L l l
~
- 300 l l
o 2s so n e0 m no em roo 223 2so m soo l 2 mt (s)
Fig. A-2. Core-inlet mass flow, core sector cell 2 (r = 1, 0 = 2).
,og UQUO MASS Flow
... .. ... vApm uASs ww 300 -
600 200* 4s0
$ 3 f . .
3ao 7 g e0 l - p u y
\
y a s k
\lll l '
\ l o g 0-k ,.
. '. tj I q
i
- i. !fl[q . -o
-'.l I i I I' ; i .
_eo
-%0 300
.,oo o 2$ 60 b VC t2h e0 In 200 22S 230 25 300 3 IlME (5)
Fig. A-3. Core-inlet mass flow, core sector cell 3 (r = 1, 0_= 3). A-2
, i t UOVO WASS R.0W ,
.......... vApop uAss : tow 300 600 45C 200 3
0 C
- . 30c k
- 4 8 1.
d ' 'l l ! , I e0
- i h il <
en g " 2 b, Il h !i I h $ d' l gfik""yy
" ~
-w0
{ ) ; l r
]
q l
\
(1 -
-=0
-300 0 23 w 3 wo 12s 30 r5 200 m 230 m 300
. Tut (s)
Fig. A-4. Core-inlet mass flow, core sector cell 4 (r = 1, 0 = 4).
,l, UQUO MASS W , ng
.......... VAPOR MASS :lDW 300 -
600 200-4s0 I - - 300 g m
=0 1 l g
~ '
- Q l
, I I l b g 2 i l> 2 j
[ l --O 0-t I f lRn i l if , a t f l
-so
-a -
-300 1 1 f I t t f ,
- O 2b so 5 w0 tb 15 0 15 200 22b 250 25 J00 s TME (s)
Fig. A-5. Core-inlet mass flow, core sector cell 5 (r = 1, 0 = 5). 4 A-3
400 UQJD WASS Flow ,
....... . VAPOR uAss low 300 600 200- - *450 f - -
300 k, g - g g - a m
'l s l I I l d l ag i I g 0- '
L ."!j l J {!.l-
'p
-0 i, f Il '4'j l- g l
i; n l ' fll! ! l j
-to l
-oo
.1 .
300 l I I l I o 25 50 75 10 0 t25 15 0 15 200 225 250 25 300
. Tut (s) l Fig. A-6. Com-inlet mass flow, core sector cell 6 (r = 1, 0 = 6).
1
@0 uouc urss rtow , , l
......... VAPOR WASs Row soo - .
. . goo 200- -.60 A - -
= k, e- ~
,1 .
I
~
e l l
@ l i
i l M s I
; [ p :
1 l 0-g d --0
,, p 'p i
-j l
!h. h
.\
h 1
-so
; 1
- 10 0
.m
.,oo 0 25 50 5 10 0 125 15 0 im 200 225 250 25 300 Tut (s)
Fig. A-7. Core-inlet mass flow, core sector cell 7 (r = 1, 0 = 7). A-4
- - .~ _ - .
400 LIQUO WASS FLOW 0
...... ... VAPOR MASS Row 300 -
600 450 200-x0 k, wo o w i I g m - 1 so w W !!.g;l , j' l l i, i W 0- .y fi' I
--O H .!qd j-
, II 1 .
l 1 i - -mo
-l l i
, d i
-50 i i l
- - -300
-MO 0 25 50 75 00 12 5 15 0 f5 200 22S 250 25 300 1
. Tiut (s) l Fig. A-6. Cere-inlet mass flow, core sector cell 8 (r = 1, 0 = 8).
i t t t t i e i i r Lwo urss rtow _ , i
.......... vAPon uAss 10w 300 . _ _ ,
600 460 200-- s - 300 h g - { g =0 - p _ , . ,50 s . I s . i ' l -0 0-lj ]h- i S ; ;
,j -
'(
I il l l _ _so
-mo l
300
.xo I O 23 SC 2 00 12b 15 0 th 200 22S 250 24 300 o TIME (s)
Fig. A-9. Core-inlet mass flow, core sector cell 9 (r = 2, 0 = 1). _ i i i A-5
40c, U(X#D WASS FLOW ,
. ........ VAPOR WASS LOW 300 600 450
% . 200-e 6 .
300
' j -
no ($,4i f I o lij g , l l d l! f. _0 5 2 O. . l
-i .
p ,! ,.
.O.
1 l j t h n 9; . l f il l i I ; . _eo i _a - 300 0 25 50 75 10 0 125 e0 f5 200 225 250 275 300 Fig. A-10. Core-inlet mass flow, core sector cell 10 (r = 2, 0 = 2). 00
~
U000 MASS FLOW ,
. VAPOR MASS : LOW 300 _
600 200-450 E y M b
) --@e e2 l
i l
!i
, 1-l s O- , , , -
j >1 , --0 [ fl j Y i . _eo _iOO _300
.g i i , , i . 0 26 50 4 10 0 123 b0 15 200 225 250 25 300 v TNE (s)
Fig. A-11. Core-inlet mass flow, core sector cell 11 (r = 2, 0 = 3). G A-6
=
@ i UOuD WASS Flow ,
......... VAPOR uASS 210W 300 -
600 200"- - 450 f - - 300 { 9 L m i 9 j u m - l ' i ; i l i 50 m Q ! i l M
*' h- I l i 1 1
lI k )l l
,:i
, i I
-~*
5
] f 1
, p t \
l
\ p I i r
,_ _w
_ico L j l - 300 l _,,o 0 25 50 75 10 0 12$ 50 ff5 200 225 250 275 300 o Tut (s) Fig. A-12. Core-inlet mass flow, core sector cell 12 (r = 2, 0 = 4). j UQuo uASS Flow ,
......... yApon uAss : Low 3oo _ _
600 i 200- e0 z l f j l
- i I
E
- i l t i
( 1
= n !
'I s
o-1 .gf { , V .
--o i
' ~
I 1 l, il -
-50
-10 0 l' -
J
)
_, . . . , . . . . . j 0 23 50 75 10 0 12b 50 f5 200 22b 250 2?b 300 ! u TWE (s) I i Fig. A-13. Core-ir let mass flow, core sector cell 13 (r = 2, 0 = 5). ! A-7
t i UQUO MASS FLOW 33
......... . VAPOR MASS -10%
300 - 600 200-- *-4S0 O . - 300 v 2 1 c i . 1 7!
. ,SO I 0 9
. i h h ! , t l I. hl p;Iq i 0
E {[ l I 'l i qN}y ;} . l l g i _e0 - [
-m l -
-300 0 25 $0 75 10 0 ut 50 17 5 200 225 250 275 300 m TIME (s)
Fig. A-14. Core-inlet mass flow, core sector cell 14 (r = 2, 0 = 6). 400
'~
UQUO WASS FLOW ,
.......... vApon uAss Low 300 - -
600 200- - eSO
. -. 300 g
g iOO 1 no l m l l , s l ! I $ 0 . W l '
.-0
' I I
l j- - 30
-100 -
300 O 25 $0 5 10 0 Sb 30 17b 200 225 250 25 J00 e TIME (s) Fig. A-15. Core-inlet mass flow, core sector cell 15 (r = 2, 8 = 7). _ A-8
m i i i
, g i e e i UQUD WASS FLOW 30
...... .. . VAPOR WASS 10W
! 300
*~ - -
600 450 200-- } 9 4 - - 300 1 e 100 i O w I I c e0 g m ! 5 l t ll ; lll I )
, l.l hI s
..a m 0-L j. ,,
I 0 i l
,] 1 I
' i tl I 4
4
-3 > ; , i .
.e0
- 10 0 300 3
t
,g i 1 0 25 50 M 10 0 12 S e0 15 200 225 250 25 300
= E (s) i l Fig. A-16. Core-inlet mass flow, core sector cell 16 (r = 2, 0 = 8).
1 1 i s O A-9 e
APPENDIX B CORE-OUTLET MASS FLOWS Figures B-1 through B-16 depict core-outlet mass flows at each core sector cell. A noding diagram is provided in Fig. 6 of the main text. M
.ae ms w
.......m,
- m. ee h= .
}
g e
>
- s .
h h1 pewnnm% 4je
% s
-a
-300 0 D N n MM M M M M N M M ime w Fig. B-1. Core-outlet mass flow, core sector cell 1 (r = 1,0 = 1).
3oo g g. 600 I
...... ... VAPOR WAss now 200-- -
40 300
}=-
1, e g
. I i q
sNn--o E 0-
&hh0N ---- - -
l
. I -
.e0 1
-co -
300 0 2$ *4 5 c0 125 50 175 200 225 250 24 300 mt (s) Fig. B-2. Core-outlet mass flow, core sector cell 2 (r = 1,0 = 2). B1 i
i l e. l l
~
LoJ3 uASS Flow
.......... v4 pop urss gow 40 200--
300
? -
e t l ' l 0-f b - -- --- -- - - '
--0 _
.e0
.c0 -
300 0 25 50 5 10 0 13 e0 175 200 225 250 25 J00 mE (s) Fig. B-3. Core-outlet mass flow, com sector cell 3 (r = 1,0 = 3). 3 t 1 i I I 1 1
~
UCMD WASS Flow
. . . . . . . . . . v4 pop m now 200-
#0 m
h 30 = 0- -n }
--O F -
.e0 l 10 0 -
i . . 300 t 200 0 25 50 75 10 0 12 5 150 15 200 225 250 775 300 i mt (s) Fig. B-4. Core-outlet mass flow, com sector cell 4 (r = 1,0 = 4). B-2
.o l 9a j ,. UOVO WASS FLOW
... ...... v4PoR uASS stow too-- --@
300 g s t 2 o-J I ddL
-yrpnpp{
j< ( i N
--o 5
{
%.y ]'
.so
-Co 300
-20o o a w n eo m no = , 2oo m ao m soo TWE (s)
Fig. B-5. Core-outlet mass flow, core sector cell 5 (r n 1,0 = 5). 3oo
~ ~
UQUID WASS FLOW
.......... VAPOR MASS : LOW 200-- -
eo Joo l 5 I h l no g illd .11 N
, o.
(w r wynn mmh . l
--o G
l
- f ..o
-wa -
-3oo
-Ma J, o a so a eo m so m 20o m no m soo 4
Tut (s) Fig. B-6. Core-outlet mass flow, core sector cell 6 (r = 1,0 = 6). B3
D I l LOUDWASSFLOW
... ....., VAPOR MASS :10W
-_eo 1
m-- . 1 300
- i k so a g - l g o- s. .
7 u i.c.u4 qppi npi $ #- Nf - - l
@[k- i
--o g I
. l .
..o
.mo 300
,g o 25 So 7b Vo 12$ 150 f1b 200 22S 2So 27b 30o WE (s)
Fig. B-7. Core-outlet mass flow, com sector cell 7 (r = 1,0 = 7). l l t g , LOUO WASS FLDW
.......... VAPOR WASS
- LOW eo 200--
l m to' h
= .
$ } I !' y M @
0 } --o h o-b}hhi h
. 1
.a
..o
. . 300 1
~*
a 25 so E eo e no fa a m 2so m soo l wt (s) l Fig. B-8. Core-outlet mass flow, core sector cell 8 (r = 1,0 = 8). i B-4 l l
O 30o
~ '
LOUO MASS FLOW
.......... vApog uAss gow 200*- *-450 l
300 h*- h 8 - - l d 8 l c l o. i- - c, l fl i 4 .J. i p
....f..!f,p l
..a @
s -
)
ji ' ! ',l 'I l l
-eo
-too -
300 l l
, , , , I ! ,
,,oo , ,
l o a 5o 5 eo us no ra roo m ao re soo TNE (s) ; 1 Fig. B-9. Core-outi.at mass flow, core sector cell 9 (r = 2,0 = 1). l Mo
~ ~
LCuD MASS FLOW
.......... v490n ues now
~^
200- -.so l t - - soo 1 5 h l m W
, o-r d
la it l
,fsvpe..nn u d O <
-7 l
-- l --= $
l 3
- i _
..o
.co .
300 _,oo , , , _ o 25 50 75 eo 125 no f5 20o 225 25o 25 300 TNE (s) l Fig. B-10. Core-outlet mass flow, core sector cell 10 (r = 2,0 = 2). B.5
i g i i
~~
LOUO WASS FLOW
.......... VAPOR MASS W 450 200-- _
300 fp
~
k l0 ,y 4,4)k$..\.. N.W 0 l
..o 300 300 0 25 50 5 2 125 50 175 200 223 250 25 300 TWE (s)
Fig. B-11. Core-outlet mass flow, com sector cell 11 (r = 2,0 = 3). 300
' ~
- LIQuD WASS FLOW
......... . v4Pm uASS : tow 200-- -.e0 300 l
8 -
.( n -
e0 g h 0-l';f ,:. + 4\)ilNYpf1 --
--0 l
300 0 25 50 75 10 0 125 150 15 200 225 250 275 300 TNE (s) Fig. B-12. Core-outlet mass flow, core sector cell 12 (r = 2,0 = 4). B.6
O l l 300
~
UouD WAss Flow
.......... veon uAss now ec 200-l no l
i ,f a -
}
g 3 d , e
~ \ --o h
~
h o
..o
.M -
.m
, , , , J j!
o a 50 m a 125 eo rm zoo 225 ao 2m 300 l u (s) l Fig. B-13. Core-outlet mass flow, core sector cell 13 (r = 2,0 = 5). 1 3ao
~
uouc uAss rtow
.......... veon uAss now
- g. -.eo I
3oo i 3 So g w , I
; d
- J nd lll d J \ ,
g jegq m. I "r p pu [ lj --o l i A - - o-l 8 y'i . . l .
.w l
.m -
3ao
.g ,
o 25 So 75 W 125 So 175 2co 225 25o 275 300 TNE (s)
~
Fig. B-14. Core-outlet mass flow, core sector cell 14 (r = 2,0 = 6). B-7
O 500 '
" ~
Loud MASS FLOW
.......... VAPOR uASS n Ow 850 200-300
! k
.o ,
N lJ! l,l l llINi !
--O 0- t -- 9 n..yv nyyprugpp- - --
3 g __ f
-60
_eo . 300
, , i , , , , , , ,
,g 0 25 50 7a 2 12 5 50 17b 200 225 250 27b 300 TWE (s)
Fig. B-15. Core-outlet mass flow, core sector cell 15 (r = 2,0 = 7). I f I I 1 l I f j touo uAss rtow
.......... vAP0n uAss tow ,
m- a 300 g W
=
go
}g h,0- \l nl ' u.lhaI 1 N" !g p,, y/ i nig
- t
* --O 7
f /
-t( .s0
-m 300 0 25 50 75 m 125 50 175 200 225 250 275 300 NE (s)
Fig. B-16. Core-outlet mass flow, core sector cell 16 (r = 2,0 = 8). B-8 1
l' b APPENDIX C AVERAGE-ROD CLADDING TEMPERATURES Figures C-1 through C-16 depict average-rod cladding temperatures at selected rod elevations for each core sector cell. A noding diagram is provided in Fig. 6 of the main text. - l l R RO) l *z B 'on
- == .
! f z - z.ooo , g z - 4.ooo l z- sooo l E n . 1 G l e .m-\ q w z . om l 9 sm _-_ 'z - oum l %
. LLk
,w...--
b - ! l ! m c-" l l c 25 so n m us no to 200 1 W~ N 1 l Fig. C-1. Average-rod cladding temperatures, core sector cell 1 (r = 1, 0 = 1). l R ROD i "z8 . $m I soo , z - 2.000 g ( z - 4.000 j ( z - 6.000 le } q '2- a= g( 9 Q z - io.ooo
@ . N Y 'z - u.ooo N
/ *
= ,
p .- __ e Ro) j' , c .900 a 25 so n a as no n 2a TSE (s) Fig. C-2. Average-rod cladding temperatures, core sector cell 2 (r = 1, 0 = 2). C1
I c l l
- R ROD l
o83 z - o.ooo 800 - - o 2 - 2.000 s 'z - 4.000 I E
= -
g 'z - seco
~
l= } e NK % - 2-z nooo
)
l h 3ao 'z - 12.000 h i N w l
= -
, p=t,w. ==h ; bd ___ _ ;;
noo
, , , O -900 0 25 SO 3 10 0 125 20 *4 200 TWE (s) l Fig. C-3. Average-rod cladding temperatures, core sector cell 3 (r = 1, 0 = 3).
m - - - - g ggg
.o84 z - o.000 h z - 2.000 i 'z = 4.000 E = -
g z- e.oco 3 ,,,, ] g "z- e. coo N 'z nooo l $ 3ao ,. z - t2.000 N i N w a -
, V N_ _ _ ;
noo
, O -900 0 25 SO 75 10 0 12 5 50 ~5 200 TWE (s)
~
Fig. C-4. Average-rod cladding temperatures, core sector cell 4 (r = 1, 0 = 4). C2
o R ROD e85 2= 0.000 em - - o
, Z = 2.000 1
'2 = 4.000 8 m -
g Z= 6.000 h i "2= Q 8.000 g ** ~
- ? .
g 2 = 10.000 h, _
\- _ 'z - u.000 g
+= , p .- bkU ___.__;--
ROD
, , , , O WO o 25 50 75 10 0 12 5 150 5 200 TWE (s)
Fig. C-5. Average-rod cladding temperatures, core sector cell 5 (r = 1, 0 = 5). soo R R00 o86 2= 0.000 goo . - z = 2.000
'I = 4.000
% m -
g Z= 6.000
= n f,
A N Z- 8000 g ,oo , w k
~
h, % h) - Z = 10.000 z = u.000 N .
<m ,p . -k ~b_ ~.d. . . . _ _ ;
ROD
, , , O -900 0 25 50 75 10 0 12 5 150 ~75 200 TWE (s)
Fig. C-6. Average-rod cladding temperatures, core sector cell 6 (r = 1, 0 = 6). C3
- =
R ROD
~
z= o.000
~
800 - o z - 2.000 i . _. z= 4.000 X .m ,\ - 1 g ' z = 6.000 E z. em
,oo , g _
z - nooo N o 8 z - 200o
\
%'1 400 -
p kc [ _cb;.__; : noo
, , 0 -900 0 25 60 75 10 0 12S 16 0 *M 200 TWE (s)
Fig. C-7. Average-rod cladding temperatures, core s(rtor cell 7 (r = 1, 0 = 7). 300 R RCD o88 z = o.ooo so - 1 , z - 2.000 z = 4.000 g _ y z 6.000 3
"' g "z s.ooo NNg
~
N % z = nooo e 1
% N '
g . . I , _ z- 200o w 400 . _ s v.. . ; - Roo
, , , , 0-900 0 25 50 75 10 0 125 15 0 75 200 TWE (s) _
Fig. C-8. Average-rod cladding temperatures, core sector cell 8 (r = 1, 0 = 8). C-4
O R ROD o89 2= 0.000 800 - o Z= 2.000
'I = 4.000 8 m - -
Z= 6.000 3 5 , .
"Z= s.000
- Z = 10.000 s2 s .
500 - - Z= 12.000 l
\
a -
, b .c , : = - . - ; _ . =- =
ROD g , , , . . O -900 0 2S 50 M 10 0 12 S 50 ~75 200 TWE (s) l l Fig. C-9. Average-rod cladding temperatures, core sector cell 9 (r = 2, 0 = 1). 1 l 300 R ROD o 8 10 2= 0.000 800 - - 2 = 2.000
- 'Z = 4.000 6 MO - -
y Z= 6.000
= .
g, W g Z= s.000 Z = 10.000 m 8 m . s1 _ Z = 2000 N
# ~ ' 1 ROD 300 O 25 50 75 10 0 12 5 50 % 200 TWE (s)
Fig. C-10. Average-rod cladding temperatures, core sector cell 10 (r = 2, 0 = 2). G5
i O R ROD j e 811 z = 0.000
~
800 - o l z = 2.000 z <.000 z n 2* 6.00C 3 i
'z- e.0m m;
g w z - u000 I. %* E, _ g " _ 'z - ew0 , s,f__ -- a . h N
,= ;,,..---- -- - . - ._
ROO
, , 0 =900 0 25 50 m c0 12 5 50 *M 200 TWE (s)
Fig. C-11. Average-rod cladding temperatures, core sector cell 11 (r = 2, 0 = 3). R RO9 o 8 12 z = 0.000 800 -
- a z . 2.m0 g 2 = 4.000 g 'z = 6.000
& : z - e.0x g *11
~
e 4 %s s
- 2. uwo e
8 30, . w w s z - com
@h . ___: :
ROO
, , , O-900 0 2d 60 5 00 12 S 50 '5 200 l TWE (s) l
[ Fig. C-12. Average-rod cladding temperatures, core sector cell 12 (r = 2, 0 = 4). l C-6
0
)
I i l l l R R}) I 2 = 0.000 800 - a 2 = 2.000 Z= 4.000
$ m -
$ 2= 6.0C0 g _
2 = 8.000 9 k g 2 = nOOO g # s , 0: Soo _ % _ Z = 12.000
\
_ . aa -- -e- -, RCD l l l g , , , O -900 0 25 50 M 10 0 12S B0 *B 200 Ts/E (s) Fig. C-13. Average-rod cladding temperatures, core sector cell 13 (r = 2, 0 = 5). 90') R RT 2= 0.000 am - 2 = 2.000 2= 4.000 g _ f ' 2 = 6.000
= th .
600 2 = n000 W g $ ' h soo 3 _ 2 - 12.000
, ._ x ;?
ROO g , , O WO O 25 50 75 10 0 12 5 150 ~75 200 T*'e (s) Fig. C-14. Average-rod cladding temperatures, core sector cell 14- (r - 2, 0 = 6). - C-7
c
- R ROD
- o 8 15 i= 0.000 800 - e z - 2.000 l g _
'z = 4.000 y z = 6.000
= .
Z= 8.000 W Z = 10.000 soo _
\ 2 = 12.000 N %
i soo . . A -. ,_ _;;- " ROD
' ' ' O 300
< o a 60 n no a u n 200
- TuE (s)
Fig. C-15. Average-rod cladding temperatures, core sector cell 15 (r = 2, 0 = 7). l l 900 R ROD
,o 8 16 2- 0.000 800 -
Z= 2.000 1
'Z = 4.000 8 700 - -
f Z= 6.000
= .
=BM 600 -
W 2 = 10.000 .,
' % , i soo . % 7 . 2 = 12.000 l
' i N
# ~
ROD i g i r , , , 0-900 l 0 2b SO B 10 0 12 5 15 0 *B 200 i TWE (s) i l Fig. C-16. Average-rod cladding temperatures, core sector cell 16 (r = 2, 0 = 8). l l C-8 l
O APPENDIX D HOT-ROD CLADDING TEMPERATURES Figures D-1 through D-16 depict hot-rod cladding temperatures at selected rod elevations for each core sector cell. A noding diagram is provided in Fig. 6 of the main text. R ROO ;
- am 2- 0.000 eaa z - 2.000
~
4 z . .f 'z. 4.000 , g 2 .-
- l. ~_ q; "
z . .. l j $ o, 'z - c.ooo m W % __
.0,
_ __L m - -- Lai ____ o
, , , , O =900 0 25 so n m0 as no em m --
T u 4) Fig. D-1. Hot-rod cladding temperatures, core sector cell 1 (r = 1,0 = 1). i
- R ROD
"# z o.ooo
- i .
moo . z - 2.000 g ,
'z = 4.000 f
~
p 2= 6.000 i ! - 1 z . ..
!- i . h % %
NV z.- 4 E. ~6 'z , u.ooo a N s- _a j a00 ,
- w. _ - -
-2 b"--
_b- _ a , g . . 0-900
; O 2s 30 5 10 0 uS 50 75 200 , ru s)
I j Fig. D-2. Hot-rod cladding temperatures, core sector cell 2 (r = 1,0 = 2). i D-1
UM R ROD
-' 8
- imo -
"z . 8c. coo
- z- 2.ox z - (mo z = 6.oCC "z. s.oco
, o M 2. oox y~ -% : 'z . o.wo
- V ~- :
400 ,
. r- "? M _'-- b_ 3 i rod
, , , , 0 -900 0 25 50 a 10 0 125 tho 5 200 TWE (s)
Fig. D-3. Hot-rod cladding temperatures, core sector cell 3 (r = 1,0 = 3). o00 R ROD too - z 0.000 _ "z - 2.mo
'z 4.000 E - -
g z - s.aco
" ~
"z - s.mo g =
( N 2 z- nom
'z . u.ow N % _ ;
400 - 2 ; .- R@
, , O =900 0 25 50 M 10 0 125 N 3 M TWE (s)
Fig. D-4. Hot-rod cladding temperatures, core sector cell 4 (r = 1,0 = 4). D-2
O SO R RCD tc0 - 2=' O.000
, _ Z= 2.000
'2 = 4.000 E m -
g 2= 6.000 h '2 = 8.000 700 2 = 10.000
- ~ .
. 2 = 12.000 e00 l ,
300 -
% ___o 400 p& - N _ v% : ,o
, , , , , o =900 0 25 30 3 10 0 125 50 ~7S 200 TWE (s)
Fig. D-5. Hot-rod cladding temperatures, core sector cell 5 (r = 1,0 = 5). voo g go9 100 - - 2= .000 goo . _ 2 = 2.000 n 2= 4.000 F, 300 - y l ' '2 = 6.000
% '2 = 8.000 ' N -
W 2 = 10.000 6 400 -
= ;_. _ aP. Yf -[7-300 O 25 50 75 100 125 150 ~75 200 TWE (s)
Fig. D-6. Hot-rod cladding temperatures, core sector cell 6 (r = 1,0 = 6). D3 i '
O
" R RCD 100 2= .000
. _ z = 2.000
'z = 4.000 g _
{ " Z= 6.000 3 'z= e. coa 5 = - z = 10 00o e . u
$ < _' 'z = t2. coo
= .Y %- .
.00 -
y =.=-;.=- & b.% & ,
, . o =900 0 as so n 80 tzs e0 n m TWE (s)
Fig. D-7. Hot-rod cladding temperatures, core sector cell 7 (r = 1,0 = 7).
" R ROD 8
nm - 2=%.00o coa . 2 = 2.000 2 = 4.000 y Z = 6.000
~
'2 = 8.000 M -
Z = 1o.000
$ .00 g ..z= 12.000 500 -
Aw - _ _ j_ ; 400 - i .r ; _ 1 t Sf . ..: f)he;- 300 -- O 25 50 M 20 125 50 '75 200 TWE (s) Fig. D-8. Hot-rod cladding temperatures, core sector cell 8 (r = 1,0 = 8). - D-4
m 12 R ROD
~
100 - - 2= .000 e coe . . 2= 2.000 2= 4.000 8 000 - - { ' 2= 6.000 "2= 8.000 MO - 2 = 10.000 W 600 ll - 2 " 12.000 500 - -- - 400 % : _^ - b _ _ ! _M _ * " a r . Roo gn , , O -900 0 2$ $0 M 10 0 125 15 0 ' 'M 200 TWE (s) Fig. D-9. Hot-rod cladding temperatures, com sector cell 9 (r = 2,0 = 1). t200 R RCD
'50 o 8 26 2= 0.000 o
1000 - - 2= 2.000 g _ Z= 4.000 k 2= 6.000 Q 800 - a T 2= 8.000 W M0 - W , 2 = 10.000 600j . 2 = 12.000 500 - * - 400 i =w_ a __z_- 2 '.'_-- b_@ _ k-go , i O =900 0 25 50 75 10 0 125 150 ~75 200 TWE (s) Fig. D-10. Hot-rod cladding temperatures, core sector cell 10 (r = 2,0 = 2). D-5
{
"O R RCD too Z= .000
,, . 2 = 2.000
'Z = 4.000
$ wo .
Z= 6.000 l E 800 = ct 2= 8.000 700 % - Z = 10.000
~ .
Z = u.000 600 l _
=
N % % f I l L- m
.oo L_ _ .M. . --- _
,00
, , , , , o -9m 0 25 50 5 to 12 5 50 *75 200 TWE (s)
Fig. D-11. Hot-rod cladding temperatures, com sector cell 11 (r = 2,0 = 3). l 1200 R ROD
'50 . 2=h.000 mon _ ' "z = 2.000 g .,
,oo .
g l 'z= s.000 3 'z= e.000 700 g z = 10.000 g ,oo ll . z = u.000 m - % - _ 400 - ": . 5 ? '_ d a- Roo
, , o -900 0 25 50 75 20 125 150 % 200 TWE (s)
Fig. D-12. Hot-rod cladding temperatures, core sector cell 12 (r = 2,0 = 4). - D6
O R ROD 1 o 8 29 2= 0.000
*10 0 g 2= 2.000 k 'Z = 4.000 E ** " < -
y 2= 6.000 2= 8.000 g 700 'f % - 2 - 10.000
@ W 'z = u.000 hN <
,oo .
,00 . .
.00 k_. _,. _k__
-- 6 ._ _
,oo
, , , , i O =900 0 25 50 75 10 0 12 5 50 *M 200 TWE (s)
Fig. D-13. Hot-rod cladding temperatures, core sector cell 13 (r = 2,0 = 5). R RCD mo - - 2= k000
,ooo . . 'z'= 2.000 g _ _
'Z = 4.000
{ E a00 -
=
Z = 6.000 l N
- z= 8.000 s 20 g -
z = 10.000
$ .00 ;I . 'z = u.000
. 500 -
~c -
"w -
400 -
=. , _ -N ; b-
___ bi-g , , , , 0 -900 0 25 50 75 100 125 50 '75 200 TWE (s) Fig. D-14. Hot-rod cladding temperatures, core sector cell 14 (r = 2,0 = 6). - D-7
O 1
- R ROD o 8 31 150 2= 0.000 o
_ Z= 2.000 Z = 4.000 g _
^
{ Q 800 a Z= 6.000 2 2= 8.000 700 Z = 10.000
% i
.00 'l l
S- - Z = 12.000 4 i ~% % 500 -
' I L
400 _=...'...
^-*.?-- .. y
' ' ' ' O 300 o 25 So 75 to 12 S 15 0 '7b 20o
; TWE (s)
Fig. D-15. Hot-rod cladding temperatures, core sector cell 15 (r = 2,0 = 7). I l 1 1200 R RT l l i sco - 4 "z0. , coo ! o coo . . Z = 2.000 Z= 4.000 g , g 2= s.c00 Q 800 - - u g Z = 8.000 700 - ~ Z = 10.000 W % -- -
$ .oo I _ Z = u.000 Soo . - _
soo - w% h% %,1. 2 d- , g , , O -900 i o 25 so 4* co 12 5 tso 75 200 TWE (s) _ Fig. D-16. Hot-rod cladding temperatures, core sector cell 16 (r = 2,0 = 8). - D-8 I
i
= ,
l APPENDIX E I l 4 AVERAGE-ROD CLADDING TEMPERATURES VS l
..-- ROD ELEVATION AT SELECTED TRANSIENT TIMES Figures E-1 through E-16 depict average-rod cladding temperatures vs rod elevation at I selected transient times for each core sector cell. Two plots am presented for each core ]
sector cell. The first plot is for transients times of 0,2.0,3.0,6.0,8.0, and 10 s. The ! second plot is for transient times of 20,40,60,80,100,120, and 140 s. A noding diagram l l 1s provided in Fig. 6 of the main text. j M l
# ~
[ ~ o T = 0.0 S l
#[ . T = 2.0 S
- j. , _ / .. .
+ T - 3.0 5 mo -
= T = 6.0 5
- T = 8.0 S i
- w m . d '
I L i . T = no S ]
+ ,..'., '
.oo . \ ,.f . '~.. ,-
p' -4 a N = ='; . i a g i i i
- a as a as v vs a as a as ELEVATION (m)
- Fig. E-la. Average-rod cladding temperatures vs core elevation at selected transient 2
times from 0 to 10 s for core cell 1. E-1
C a 750 9 m - e T = 204 S eso -
,h 3
- 1 = 40.0 S b +ca.
&- + Y = 6C.0 $
*' ,Y........a m T = M.0 S f **w.. .
Isoo * *
~ ~
# f + T = 100.0 $
+
- gan - , __
,hg
- T = 120.0 $
f \1 I . e T = 140.0 S I 40 - l 400 - ,
-) , ....................
e a as a as v v.5 a a3 = as ELEVATON M Fig. E-1b. Average-rod cladding temperatures vs com elevation at selected transient times from 20 to 140 s for core cell 1. i m ' ' 3' - 7$Q -
.eT=0.05
,.i ..
{ , l
*J -
iaT=2.0S
,oo -
[%\/ y- [, +T=3.0S
~
mT=6.0S a ~ /&s \
~ .
- T = 8.0 $
h e /f E25 - ; , T = 10.0 S
$ * . . . . .~.. * . . . ( .
soo
~. -
550 -
. , , , i g
25 253 26 205 27 273 28 283 29 29 3 ELEVATD4 (m) Fig. E-2a. Average-rod cladding temperatures vs core elevation at selected transient times from 0 to 10 s for core cell 2. E-2
750
.. . 1 o T = 20.0 $
/
- T = 40.0 5
, ,, g 8 .: ) .
- T = 60.0 $
1
.' * . . ,,8 l
- , ,l ,
1
?. .. c.
, .: ' * . , .g . T = 80.0 s
- g
- T = 100.0 5 f
W \\* gm . , 5'
- ;* _ = T - 120.0 s i a
' j// ,I '*.N '
. T . wo.O s l
M . . .; . . . ; ..........*- 1 g i i 25 255 26 263 27 273 28 283 29 293 l ELEVATION (m) Fig. E-2b. Average-rod cladding temperatures vs core elevation at selected transient , times from 20 to 140 s for core cell 2. l
.30 . .
800 - ~ s[ oT=0.0S A=
- T = 2.0 s E
,30 .
/
g* 1\.} 't
+ I = 3.0 5 g
- T = 6.0 5
$ 9. A T = s.O s e .30 ...,, \\ .
' T = 10.0 S k'
.., f
= . I ... ,... .. ,
.; ... c.q -
C; m - . m 25 253 26 263 27 273 28 28.5 29 29 3 ELEVATION (m) l Fig. E-3a. Average-rod cladding temperatures vs core elevation at selected transient i times from 0 to 10 s for core cell 3. t l E-3
eso I'* *
*M i ' .s ... .
/ O, M's ., . e T = 20.0 5 no - .: 'A . * -
,f I U
\ Q a T = d.0.0 S 7eo -
p,
/
,. [ \ l -
- T = 60.0 $
g ,. i l
.s
= T = 80.0 S O' ,'.
l k'a\,\*, y g + T = 100.0 s g l Soo -
- T = 120.0 s
- jl , l soo -
e T = wo.0 s l l , no - l : . - a -
- d. .. .. . . . - * * -
350 a ns n us v as n as a as ELEVATION (m) Fig. E-3b. Average-rod cladding temperatures vs core elevation at selected transient times from 20 to 140 s for core cell 3. 000 - eT=0.05
/ ( ~ . T = 2.0 5 2 no ['
. T = 3.0 S l
f ' . = T - 6.0 S
. T = 8.0 5 h i W '
. s
- T = 10.0 5 20
. . . . . . <., -l
.' y
, ......,.....- _ .*i ..,/\- .m
, j I
20 n n.s a ns o as n us n as ELEVATION (m) l i 1 I Fig. E-4a. Average-rod cladding temperatures vs core elevation at selected transient times from 0 to 10 s for core cell 4. __ E-4
.=
1 900
.* ,, fr[-d ...) , .s..
800 * ' ' '.. - o T = 20.0 S
/ ;I * [
- A'*'V.,
U A a T = 4.0 S b 700
o*. 4,
. T = 60.0 S
\ f,*,.* ,-
= T = 80.0 S g T s m. . , Q,l-., ,
e f . 3 y ** I l
.I. -
. T = o0.0 S W , :
j 8,
- T = t20.0 S p
E 300 . l(j #/ l. ',\. _
., f ,
.,., = T = w0.0 S
- l[ !
. . [ s ... . . ' .
J00 a m a as a v3 m as a as ELEVATION (d Fig. E-4b. Average-rod cladding temperatures vs core elevation at selected transient times from 20 to 140 s for core cell 4. 850 800 g, o T = 0.0 S g a T = 2.0 S
$ 750
, + * ...+ - ( i
. T = 3.0 S
$ d g*'
N = T = 6.0 5 q \. x
- T = 8.0 S h -
W ,
- T = e.0 S
,..... ~
8 . . e 830 _
,... .,,, f . \ .
. -\ .
.'.t 800 ,,.......-
2s as a na a v3 m as n as ELEVATION (m) Fig. E-5a. Average-rod cladding temperatures vs core elevation at selected transient times from 0 to 10 s for core cell 5. E-5
\
i y goo . ..., *.., NT*' . ;- , e T - 20.0 s a T = 40 5 m . ! ' \, .
- T = 60.0 $
.I i
= /' ,\ -
q ! ;,
* = T = 80.0 S f'
" ~
I e., .
. . T .100.0 S o-*
m -
,e
! a.
. T . uco s
't soo -
,./ ~ j i \, -
= T = =0.0 s e .
- l :
l 'e . eco . /_ Q, . g . ' * **
.no as n as a as v v3 m as a ELEVATION @n)
Fig. E-5b. Average-rod cladding temperatures vs core elevation at selected transient times from 20 to 140 s for core cell 5. 350
.C soo o T = 0.0 S
- T = 2D S
+
g ,....
+ 7 = 3.0 S W M m T = 6.0 $
Q a
}, . T - s.0 S W 82 -
e......... ,n 1
. T - i0.0 S p**,,... b w
= w .
,oo a a3 x m3 v v3 m m3 a a3 ELEVATION (m)
Fig. E-6a. Average-rod cladding iemperatures vs core elevation at selected transient times from 0 to 10 s for core cell 6. E-6
mi Soo 3..4..... Soo - .
"' ~ ' * " '
o T = 20.0 $
. -- l l'~~~ , ' '"*
, ..i
[~ y a7-*eS 7ao - -
- T = 60.0 $
/ l q
, /-
* !lI .
. = T = 60.0 S
# ~
, f.,, , (.* .
- T = 100.0 S e '
I ; ' . N ; ...
.
- T = 120.0 S soo -
,i : .. -
. 's. * \ e T = M0.0 $
j i i ! a - 3 L :l__ d .....- - soo a as a as v vs a as a as ELEVATON (m) Fig. E-6b. Average-rod cladding temperatures vs core elevation at selected transient times from 20 to 140 s for core cell 6. soo ' 75o -
....+- ' - ! I -
o T = 0.0 5
'
- T = 2.0 S g s.
* / ,
+T=3.0S
'/
% 7 = T = 6.0 5 g
g 56o !/ ., , . . -* ' ' ' ' . ,
- , . 7 . ,3 3 w
4 ,. ,
. a i *.,'
- T = 10.0 S
- . a'...*"
soo ,...- n -. m
"'*w..,
t . h ..r_ , a as a ns v vs a as a ns ELEVATON (m) i Fig. E-7a. Average-rod cladding tempemtures vs core elevation at selected transient times from 0 to 10 s for core cdl 7. i E-7 l l
a 80c
+ . , ~ . . . . ~ . ..
no -
.,g.V . 4. ... .. -
.e a D's - o T = 20.0 S 700 / /ebA a T = @.0 $
/ \.
E -
- T = 60.0 S l .
" T = 80.0 S j
6,, I.00 = *
/
e I I
.. * \;~,.
2
- T = 100.0 S
- T = 120.0 $
500 - f l 'n '
\ sT=20.0$
l
<00
- _ .- Q. ... . . . . * - .
350 a as a n.s v v.s a as a a3 ELEVATON (m) Fig. E-7b. Average-rod cladding temperatures vs core elevation at selected tra:uient times from 20 to 140 s for core cell 7. m - - - - - 750 -
, i o T = 0.0 S m - .
*~ .f 'aT=MS g 700 -
/; y,
+ T = 3.0 $
f q m - A
/
/ *Y^ -
= T = 6.0 $
=( " -
./f \} N
- T = s.0 5 e an
// i -
. T = m.0 S g - *//
.00 . j..'........--,.........-
575 -
$50 -
g , a as a x3 v v.s a as a n3 ELEVATON (m) Fig. E-8a. Average-rod cladding te .tperaturr.s . vs core elevation at selected transient times from 0 to 10 s for core cell 8. E-8
>a l
e , hI . .\a, . e T = 20.0 s
.: \ ' . ... .. e 65o -
[ M ~. . . . a T - 40.0 S g ,' *' . ..I e D 4'g% .* + T = 60.0 S
/* f
- g soo ,g g .. .
I q j
; ag., ' ._ =T=80.0s 8 i k.'
. + T = 100.0 S w f W
- T - t20.0 s 3ao sy _
e - e f f.s . 1 = =0 s i , , . \ . .. .. : l l e gn 2. .M....... - 35o 25 25.5 26 263 27 273 26 2E3 29 29 3 1 ELEVATON (m) Fig. E-8b. Average-rod cladding temperate ~. s vs core elevation at selected transient times from 20 to 140 s for core cell 8. 4 , 700 1 oT=0.0s m - / '1
,7
. r . 2.O s u\ ,l g 11 + T,.= 3.0 s eso -
j[ 4 -. 2 jl r - s.O s c
= l\
g
.; 4 .
- 1. a0 s e ..
' . .m.. ... . .. . ...
J
~
' ' T - 10.O s g saa
. ,; f- }
- /,
~ '
y , 2s 2s.s a a3 v v3 a 263 a as ELEVATON (q) Fig. E-9a. Average-rod cladding temperatures vs core elevation at selected transient times from 0 to 10 s for core cell 9. E-9
=
225 600 - 55 - s.**s.'+
- s, o T = 20.0 5
. T = 4.0 5 550 -
S :
+ T = 63.0 5 525
, v. .
_ _ - __a
'., m i = 80.0 S
; i
- T = 00.0 S 45 -
w - I
- T = 120.0 5
.o -
/.
l 8 se - a T = 140.0 S 425 - 4 - gae/')ee.4.. l\
....../.e...
g....e+
= - -
g i 25 2Ss n as a vs u as n as ELEVATON (m) Fig. E-9b. Average-rod cladding temperatures vs core elevation at selected transient times from 20 to 140 s for core cell 9.
* ~
= T = 0.0 S
..,*- . . T = 2.0 S soo .
...s..4..-
g . ~ .' . *.. ....* + T = 3.0 5
...+
q Sao -
, - : = T = 6.0 S g ....
, I
- T = 8.0 5 W $7o -
'N -
' T = 10.0 S Soo
'g,w.%= -
, - f ~~* '~ ~ - e . ,. . SSo - _ - - - _ _% - S.o 25 2Ss 26 26.5 27 27.5 28 28s N Ms ELEVATON (m) Fig. E-10a. Average-rod cladding temperatures vs core elevation at selected transient times from 0 to 10 s for core cell 10. _._ E-10
e I 625 600 - o T = 20.0 5 m - s/, . .
,,,.................*.' . T = e.0 S sso s' bW'\
p' f \ i. * .,
+ T = 60.0 $
$ = / ; i \ J.. -
q / I *
= T = 80.0 S cr s' ! l 's W 500 g
_ t .. - - - -_ \. ' . _
- T = 100.0 S 3
w -
/ \-
g a r f
\ -
T = uco S
. , j .
. T = =o S
. l .
I soo
- p/ aa-..
....~...........
m 25 253 a as v zu a 23 a a3 ELEVATION (m) Fig. E-10b. Average-rod cladding temperatures vs core elevation at selected transient times from 20 to 140 s for core cell 10. m , , , , , 725 - a T = 0.0 5
,f-iW*g 700 .
- T = 24 5 I A ~
gg
+ T = 3D S s 1,
" T = 6.0 5 l Q 650 -
['[ . t .
- T = 8D S w
.*...as. . . . ' . .
625 C .- ! ' ,./ ,
. T = 104 S 8 Goa .. .
*/..c ~' f . . . ,:m
{
~...
/
~
.=- . _ ,,.]
~ &... .
560 - W a - 25 253 26 263 27 273 28 26 3 29 29 3 ELEVATION (fri) Fig. E-11a. Average-rod cladding temperatures vs core elevation at selected transient times from 0 to 10 s for core cell 11. E-11
d I 1 g 1
~
+, , . . . . _ . . . o T = 20.0 $
$...A...*A,',
! f '*****"**s.,**
/ , W ,3 a 7 = 40.0 $
i 400 : O ; s' 9-e ';
' *"
- 8
/ .*
- T = 60.0 5 580 f ! k$ '
a T = 60.0 5 1 g
- T = 00.0 S q
W 500 -
. I -
) o ! . t = t20.0 $ g :
- l Y.
g .
,! f .
eT=#0.0S I l b I b
*o
& :.J: /... s.. ....... ...
i i n as a n.5 o as a ns a ns !- arv e @) - \ Fig. E-11b. Average-rod cladding temperatures vs core elevation at selected transient i I ! times from 20 to 140 s for core cell 11. i ! l l 750 i i M - - - l 3 #A o T = 0.0 5 I 700 f'\ $ . .
'j J\ *T=1.05
! E m -
. T . 3.0 5 .
1 Q 350 . = T = 6.0 5
* +
i .
- T = 8.0 %;
4 W \ O *
' T = 0.0 S
& .0a . . .
....-................~..1.. a -q i
gg ,* \ ., - ' s i . . n as u as v vs a as a as
- arv e @)
Fig. E-12a. Average-rod cladding temperatures vs core elevation.at-selected transient
! times from 0 to 10 s for core cell 12.
1 i E-12 i 4 J
l O 700 1 65o - f .* * * * **h- ....... o T = 20.0 S
.~~ ,'
/
f* *T=40.0S
,4 9% 5.
E ~
,i % ),
. T = 60.0 5 1 g
55o i \. ' ., , - T = so.0 5 I g ; ;.; l l ls m - , i l N
. T = m0 S t l ' T - 120.0 $
l . T = wo.0 S I J 4oo
- gj ,: & _ . ... . . . ... . .. . . . -
l 25 25.5 26 26.5 27 273 28 283 29 29 3 l ELEVATON (m) l l ! Fig. E-12b. Average-rod cladding temperatures vs core elevation at selected transient times from 20 to 140 s for core cell 12. g i i
,9 oT=0.0S N'
-
- T = 2.c 5 75o 8 *
. T = 3.0 S
.- :i f T = s.0 5 g
voo - g -
#
- T = 8.0 5 l e . _
1t . . T = o.0 S g fag ,
,/ , . . . * * ' . . . . . . t. .
" ~
hlk h;-. :';*-k . Vy L ~ ~-- ~- -=--., , 33, . W . 1 Soo . 25 255 26 2E5 27 27.5 28 283 29 29 5 ELEVATON (m) l i l Fig. E-13a. Average-rod cladding temperatures vs core elevation at selected transient
- times from 0 to 10 s for core cell 13.
1 E-13
l 1 l
. i 33 M ,
s' '. o T = 20.0 $
..} t gso -
"?*s,[ ' I t' -
a3,gg$ i g i i soo
- T = 60.0 5 i l ,
; *. m T = 80.0 5 u,o l , TM; N.=-
. T = mo.O s g +. ,
l y '
\
3,o
\ ._s . t . co.O s g - , .
ao j
=
l { = T = x0.0 s l , I
/
*m - 4' _; ; J _ _--............4 aso ,
a ns n us v vs a ns a as i ELEVATON (m) l Fig. E-13b. Average-rod cladding temperatures vs core elevation at selected transient times from 20 to 140 s for core cell 13. I j
,3, l
725 - o T = 0.0 5 f 700 - A, . l
' l
*T=2,0S
/ \
675 d. . -
+ T = 3.0 S Y J p\ ~
Q 650 N - " T = 6.0 S cr W =
- ._i// , .
, 1 = .o s e 2
. :.T=ubs a . 4-
- soo * -- -
sn - -
$50 - -
33 a m n as v os a as a ns ELEVATON (m) Fig. E-14a. Average-rod cladding temperatures vs core elevation at selected transient times from 0 to 10 s for core cell 14. E-14
. . . . .. _ . - . - _ . _ . _ . . . ~ - . - _ __ ._
I o n 75o l no ..
~* ...* e T = 20.0 S 4 ,
aso - , N' a T = 40.0 5 g r "V e !
./ \
t y.1 . . 7 = .O S l g
* = T = 80.0 $
Q s I p ,,o .' ,l Q*. ., 3 e
\s .
- T = 00.0 S
^
I
- T = 120.0 5 Soo g
, j s
= T = w0.0 5 l lN. .,
eo - 1 I
<m M. . . ... ..;.y........ ...
... ---l -
i 1 , a as a ns a os a ns a as . ELEVATON (m) Fig. E-14b. Average-rod cladding temperatums vs core elevation at selected transient times from 20 to 140 s for core cell 14. 1 4
' 1 I t f
) 7so - a 8 o T = 0.0 5 5
- T = 2.0 5 g Too -
M Q.. . 1
. . T = 3.0 5 7~
E m
;/
/\ = T = 6.0 5 m
) M 66o - I
+ T - tO S , 5
- i l
W h g s2$ - 7 '- g -
- T = e.0 5 a ..-I -
Goo - g i. _ % _
;f.. -f, ....
m .
,f
~
- =_
L-"...--: S$o - W - - g , a ns a ns v ns a ns a ns ELEVATON (m) Fig. E-15a. Average-rod cladding temperatures vs core elevation at selected transient times from 0 to 10 s for core cell 15. E-15 l l I
a g 650
* " * ~ *
- T = 2o.o s
? ,, , a , ','*, , , . .
- D*
- T = 'o.o s 6o0 .! .
k g .. - T - so.o s l l l
< ao -
l l g,{.\', -
- T . so.a s
- , \,
I T - ca.o s e m - ~ l I l
- T - tm.o s g , \
. i l . T . wo.o s i ,
j a - lL _J. . y_n................... ao a m a a.s v v3 a a3 a a3 ELEVATION (m) Fig. E-15b. Average-rod cladding temperatures vs core elevation at selected transient times from 20 to 140 s for core cell 15.
' ' c 6D 600
,..,s.......,, oT=o.os
* .. f " . . +
'
- T - 24 s gg ,.*
*..,,j,,'s..***
8 . .s'
**** .,*- + T = 3.0 s
{g 550 -
,' - ' = T = 6.o s o'
w sm - [' , N .
. . T = e.o s g '%. ,
- T - c.o s
** ~
,w----.-.. ,
~-.. ,
.f
~ ^ -~
550 - [ %
?
i MO a m a a3 v v3 a m3 a a3 ELEVATON @) Fig. E-16a. Averave rod cladding temperatures vs core elevation at selected transient timesi am 0 to 10 s for core cell 16. _._ I 1 E-16 1
G J 4 625 i soo *
. *'~. ..,*
o T = 20.0 S m . 4 . _
. . . .. ....... . . 94 g. . a T = 40 5 550 -
. 8' %' 's -
f .
- T = 60.0 S
,/ f g
g , l f \ a.' . . ,
= ,oo _
- l. _ \\t .
' - su s 5
~
g , T TV ,
- T = 100.0 5
\g Q , j t * ' T = 120.0 5 i
e l eT M0.0 S 425 - l - 400 . ,[ _ __! m.n..,................. m - - 4 F 1 1 25 255 26 263 27 273 28 263 29 29 5 ELEVATION (m) Fig. E-16b. Average-rod cladding temperatures vs core elevation at selected transient times from 20 to 140 s for core cell 16. l 1 1 l l l l 1 9 E-17
APPENDIX F HOT-ROD CLADDING TEMPERATURES VS ROD ELEVATION AT SELECTED TRANSIENT TIMES ! Figures F-1 through F-16 depict hot-rod cladding temperatures vs rod elevation at , selected transient times are presented for each core sector cell. Two plots are presented for each core sector cell. The first plot is for transients times of 0,2.0,3.0,6.0,8.0, and
. 10 s. The second plot is for transient times of 20,40,60,80,100,120, and 140 s. A
- noding diagram is provided in Fig. 6 of the main text.
d 1 ' ' ' ' ' toSo 1o0a - . 350 . _ o T = 0.0 S } - ,oo . _
.t-2.cS g ' '
- T - 3.0 5 eSo - .
< =T=6.05 800 gm N .
. T - a.0 5
. T - no s l g o . ,
**^ '. :-
,,o . ,...< .
.~_ _
z 2.::::$ F"3-5
- 35o - .
3oo a as a as v v3 m as a a3 ELEVATON (m) Fig. F-la. Hot-rod cladding temperatures vs core elevation at selected transient times from 0 to 10 s for core cell 1. l d k . F-1
O t t t
- s .
V .*\' ~ o T = 20.0 s
~
- Y -' . T = 40.0 5 800 -
,' ,, .
- T = 60.0 5 q Mo '
, ,'s . . . a . , ' -
=T=80.0S cr J/ ,
W wo - f f ' " ' * ..
+ T = 100.0 S
- T = t20.O s 500 -
'- - eT=WO.05
_ a y J/
.oo . .__;_ : W. ..... E i
25 25.5 26 26 3 27 273 26 263 29 293 ELEVATION (en) Fig. F-Ib. Hot-rod cladding temperatures vs core elevation at selected transient times from 20 to 140 s for core cell 1. f f f W [i
, A Q' o T = 04 S E
aso
! /. \/
,l
. T = 2.0 s g .. L,k, .
. T = 3.O s g
Q(, ' T = e.O s y -
\. -
.T=s.Os y ,
. T = no s g .
eso - -
,.....*~.;,. .. _ .\
2._ _ 2s 2s3 a x3 v v3 2e as a as ELEVATON (m) Fig. F-2a. Hot-rod cladding temperatures vs core elevation at selected transient times from 0 to 10 s for core cell 2. F-2
. . . - _ . - . . ~ - . - - - - -- -
O a 4 l l
, , )
g i i ,
- 1 1000
+' , ,
o T = 20.0 5 900 -
~~'...-
M - a T = 4.0 S 1 J g , ;..' w.**...,,,* + T = 60.0 S v . Boo -
- l. l' ? ..'., y s.,
1 Q ,/ , 's .. . a . = T = 80.0 $ cr / , . I W 100 - a I .
g '
- T = o0.0 S l 5 l W i .
- ! v ' 7 = 120.0 S eco -
} '.
' *. , , a T = #0.0 S
[ l sw - l ,[1 ; 1 4
.oo - __ -
- 4... - _
3 ,o a a3 a as v v3 m as a a3 i ELEVATON (rr$ ! Fig. F-2b. Hot-rod cladding temperatures vs com elevation at selected transiert times , i from 20 to 140 s for com cell 2. ; i (
- e , , , , , ,
c
\ dk\ ,
o T = 0.0 S
*] i .
- T = 2.0 S E [, + T = 3.0 5 m -
f ,
! ' = T = &0 5 Q
c: W 750 - -
+ T = B.0 S 5 +
w '" -
- ! . T = e.0 S g , ... ' "' d. ]. .,
eso ,.- .
./ '... _. -
~ .,' [ ..
600 - . sso
.: l
' ' ' ' ' i 500
- 2b 253 26 26 b 21 27h 28 28h 29 293 ELEVATON (m)
Fig. F-3a. Hot-rod cladding temperatures vs core elevation at selected transient times from 0 to 10 s for core cell 3. F-3
1 I f TW -
*~
+
o T = 20.0 S y , Qs. ,%*. *
*T=60.05
^<$s'
~
s .00 . // ,
. T . 0, $
g .- - ; !
,' /
, * s a T = 80.0 $
g 3o0 g . ; - y . T = w0n S e =
- i ;
X' ,
- T = 120.0 S
~
600 SCO - l I
'a"
= T = 140.0 S I i
a ..... ........ 3,o 25 255 26 26.5 27 27S 28 283 29 293 ELEVATON (m) Fig. F-3b. Hot-rod cladding temperatures vs core elevation at selected transient times from 20 to 140 s for core cell 3. 1050 1000 - - 950 - - oT=0.0$
~
goo . . . . , _ a T E 2.0 S
. *^ *' + T = 3.0 S ;
650 , q + = T = 6.0 $ g , , x
%
- T = 8.0 5 g, .
a,,,,....... .
- . T iO.0 $
.30 . n. .
,J _
**e+..,
goo _ .
$50 -
t f i 1 2S 25.S 26 26.S 27 27S 28 28 3 29 29 5 ELEVATON (m) Fig. F-4a. Hot-rod cladding temperatures vs core elevation at selected. transient times from 0 to 10 s for core cell 4. F-4
0 00 1200 ,..
. e T = 20.0 5 tioo -
. . ,g%.V., . -
'.*"*'/
- a 7 = 40.0 S 1000 - z. -
ar - *,
,i l 's ,,, e ',, .
n 1
- S0.0 $
800 - - h
- T = 100.0 5 j , ,.
g i : '. m 700 ! 'a -
, T = 120.0 S g l
.ao
//j :
=.,
. T = wo.O s lp soo - i -
23 25 S 26 263 27 273 28 284 29 293 D EVATON (d Fig. F-4b. Hot-rod cladding temperatures vs core elevation at selected transient times from 20 to 140 s for core cell 4. 6 F-5
1050 1000 s 950 e [-fn \ m i o T = 0.0 S 900 -
\ -
b ... s
- T = 3.0 S
'e -
850 -
* = T = 6.0 S q
E > i
#
- T = 8.0 $
750 'g -
- T = 10.0 S g *
,..., \
,a0 . ,.....- \,.\
v.
.30
-= = =
_ _ , , , , , . . . . . ' ' =* _
. i
,,o 25 253 26 263 27 273 26 28 3 29 293 ELEVATON (rn)
Fig. F-5a. Hot-rod cladding temperatures vs core elevation at selected transient times from 0 to 10 s for core cell 5. i
, , t , ,
l l
@ =
. -. . . . . . r . 20o , }
gjoo . .- #\ ,
+
i a T = 40.0 S 900 ,( . .,
,. - .
- T = 60.0 S h ,oo /'Il .
* \\.. . . . T = eco S f ,! j . .a . \0} l
(" y M l
/
'm,
- 3
- T = 100.0 S g '/ l ., , 3
- T = t20.0 5 2 300 : 's ,.
. ., .T=#0.05 300 / l 'm -
=a......
a , _ i i r g i e 25 253 26 26.5 27 273 28 263 29 29 3 ELEVATON (m) Fig. F-5b. Hot-rod cladding temperatures vs core elevation at selected transient times from 20 to 140 s for core cell 5. F-6
as a J 1050 1000 - d e T = 0.0 S aT=2.0S i o c
%j ,
- T = 3.0 S 830 -
J
= T = 6.0 5
< 800 -
j
+ T = 8.0 S m m -
- T = 10.0 S
= - , .. . - {
- no -
l '.... 600 - 6 _
- , , , , n
' 23 ass a as v vs a as a as J ELEVATON (m) ,
- 1 Fig. F-6a. Hot-rod cladding temperatures vs core elevation at selected transient times ; t ! from 0 to 10 s for core cell 6. 1200 go _... .._..-.-. . p o T = 20.0 5 ,
\
/;/ ry/% -
- T = 40.0 5 l
.* % .s *
- h. '
n E 900 -
,-,' .,
- T = 60.0 S -
g ,,, e.' - l ..
., ,j l ' . . .
\ 3 = T = 80.0 S l 5 l
,l ,:
- g i :
-
- T = 100.0 $
g m -
; '.='.
; 1
- T = 120.0 5
. Soo - i : . i / : ' s T = 140.0 $
. A. y=
SCO -
.' l 'a J !
. 400 -
......e-j g , , , ,
, 2S 253 26 26.5 27 27h 28 283 29 293 ELEVATON (m) _- 4
- Fig. F-6b. Hot-rod cladding temperatures vs core elevation at selected transient times from 20 to 140 s for core cell 6.
F-7 3
10 0 o [A
\' 'l o T = 0.0 S 900 ,\ v ..
\.l% l \' . b\
. T . 2.c 5 6
m -
'4 \l
- T = 3.0 5
'}
'!h.
800 - y = T = 6.0 S DO '
+ T +T=&OS
~
,r,......... ", )
=
n no -
, , , ,o.o , ;
g . aso ,- w ,l., u ns a n.s v vs a aos a na ELEVATON (m) j
; Fig. F-7a. Hot-rod cladding temperatures vs core elevation at selected transient times from 0 to 10 s for core cell 7.
l t t n g i 1000 .
=. b ; o T = 20.0 S
... L.D.4. %. ...
4 9ao '.
.T=40.0$
y \"? + b ,, . 4 T "' - .,
- T = 60.0 S
*00 * .
l . flf f q , i i=T=80.0S e / l \
- s
- T = 100.0 S f
w : ai
- l i '
- T = 120.0 S Q 600 -
3 l '.
+ -
, W fj . x*
, . e T = MO.0 S soo -
l[/. I . .oo ............. . 3,, 25 2Ss 26 26s 27 27s 28 252 29 29 s
< ELEVATION (m)
Fig. F-7b. Hot-rod cladding temperatures vs core elevation at selected transient times from 20 to 140 s for core cell 7. F-8 1
l l l l 95o - o T = 0.0 5 [ Y,/
~
. 1 - 2.c 5 E eso - '
f/ j /\ i -
. T .10 5 f.oo T - e.o 5 g
e no - I
\\(N"]
. T = e.0 5
. T - c.0 S g
E m - eo -
, ,,.....a~~........L. I i ....j. D. ......
, i t I t n as a ns v vs a ns a as ELEVATON (m)
Fig. F-8a. Hot-rod cladding temperatures vs com elevation at selected transient times from 0 to 10 s for com cell 8. cm - s ..
.. N. - - e T = 20.0 S
_9 .. .,. - . . . - Soo -
,h - -
a T = 40.0 S b l' $ 'N -
. T = 60.0 S g aa \ -
g cr j
/ /- // , ;
.... r..t...
i aT=80.0S
,I
- T = 100.0 S h ,f
\.
w o 'I I
! '.\. \\
\
- T = 120.0 S 600 - -
l f
'-.
- T = 40.0 S
.)
m -
; c' l
.ao -
a m n us a vs a ns a ns ELEVATON (m) Fig. F-8b. Hot-rod cladding temperatures vs core elevation at selected transient times from 20 to 140 s for core cell 8. F-9
. l I l i g i 900 oT=0.05 850 *A - a T = 2.C S g
}
) .
. T .1, 5
=T=6.05 4 750 /
i k h *T=8.05 g M - -
, . , :
- T = 10.0 5
.3, .
p .
,I. \ _ . . . .. .... . , . . .A. .
~
s00 -
./ '
.-...'..........p.,. . . . .
W- -, + _-- :
,,o a as a as v m a 23 a .53 ELEVATION (m)
Fig. F-9a. Hot-rod cladding temperatures vs com elevation at selected transient times from 0 to 10 s for core cell 9. l l 900 l s. 800 - [**'. j., o T = 20.0 S
! /' . .a . T = 40.0 5
. .....,... g M ~
~
+ T = 60.0 S g .
Q j }. . ... . .. . . . , " jl., a T = 80.0 S
& : : 's.
~ l '
! .
- T = 100.0 5 e , I ;
! ' T = 120.0 S 500 .
,' s T = WO.0 S 1
. . ....< - . ;4.... . 2 .
, t I ] f 25 23.S 26 26.5 27 274 28 285 29 29.S ELEVATION (m)
Fig. F-9b. Hot-rod cladding temperatures vs core elevation at selected transient times from 20 to 140 s for com cell 9. F- 10
= 630 620 oT=0.0S
* *~'"
6C
s....,
aT=2.CS
. ,,, .,s .
600 -
. T = 3.0 S
-.,,,..~ +
'g,***...d., .
a 6."'
~* *
" T - 6.0 S Q S00 ,
e ._. ( g . m -
~.. -
. T . so S g . . T = io.0 5 e sm N.% % -
p- ---. . Sea - 7
. _ _ _ x_ %
m . n m n us a vs a as a as D.EVATION (m) Fig. F-10a. Hot-rod cladding temperatures vs core elevation at selected transient times from 0 to 10 s for com cell 10. 900 850 - - r\ - . o T = 20.0 5 800 - w 5 m .
*' Y"(w ,, ,
a T = 404 5 [ * * * *.. , . , . + T = 60.0 5 g ,, , g ** y , , j
, . = T - 80.0 S
% 4- , 'e, ,
/ f '.
, *
- T = 100.0 5
* . ; +
/* j ,
- T = 120.0 5 pt _ _ _ _
'. e T = 40.0 5 Soo . - - -
_4 . 460 d $ - um . 33, 25 m n us v vs n as n as ELEVATION (m) Fig. F-10b. Hot-rod cladding temperatures vs core elevation at selected transient times from 20 to 140 s for core cell 10. . F-11
950 900 .
. eT=0.05 850 ~ ' *
- T = 2.0 5 E em -
y i
. T .10 5
$ 3 g m ,j/ T = 6.0 5 g . .t=e05
" * ~
~
- T = 10.0 5 a0 . . . . '.s,. - -
)
s@ ,. I ':h-wa; g 1 1 r , g i e a a 233 a a3 v v3 2e 2s.3 a as ELEVATION (m) Fig. F-11a. Hot-rod cladding temperatures vs core elevation at selected transient times ( from 0 to 10 s for core cell 11. g , , , , l l
~
* + v , ~ < o T = 20.0 5 al' .s..-* '
- T = 40.0 5 ang . ; ,
. T = 60.0 $
g i 'i . n't.g . g 700 -
- k. '. -
= T = 50.0 S tr
' l ', .
- T = 1004 S h ;
, f W 600 : j ' k. - ,
o ! l l k\,. '*T=u0.0S M -
'I I - a T = M0.0 5 500 -
! \ i a - . . . . r. - a l ; . ... .. A .
23 253 26 263 27 27 3 28 28 3 29 29 3 ELEVATION (m) Fig. F-11b. Hot-rod cladding temperatures vs core elevation at selected transient times from 20 to 140 s for core cell 11. . F-12
a 950
~
900 g N \ o T = 0.0 5
\ 7
- T = 2.C S b 800
. T = 3.0 S
\
Y . - * = T = 6.0 S g e
.
- T = 8.0 $
w * - '
~
, T = 10.0 5 650 600 .....
.. ..,i ..... m, .......
9 560 l i f i 1 I 2$ 253 26 264 27 273 26 26 3 29 294 ELEVATON (m) 1 1 Fig. F-12a. Hot-rod cladding temperatures vs core elevation at selected transient times j from 0 to 10 s for com cell 12. 110 0 1000
. . . . o T = 20.0 S 900 -
ls' l .
- f' -.. A'
..s.
. M
'%4 sy - _ s T = LO.0 S
.l E / ,fl / '... \ . T = co.O S
'n 600 .. ..
'4 *'
s T = 80.0 S
~ '
- T = 100.0 S w
600 l
,f '
i '..s
' T = 120.0 5 aT=M0.0S Soo - ! o -
<00 -
.!__m.......- -
25 215 26 263 27 273 26 263 29 29 4 ELEVATON (m) Fig. F-12b. Hot-rod cladding temperatures vs core elevation at selected transient times from 20 to 140 s for core cell 12. F-13
. . _ _ . . . - - . . - .- - - ~_.. . - -...- -- -
o C00 t
# ~
fT ,F , o T = 0.0 5 900 -
* *
- T = 2.C S g k ..
z
. /
- T = 3.0 5
{= f
= T = s.0 5 c: 4, no - o
- T = 8.0 $
w yon .* "
- T = 10.0 5 650 -
p .. - - - b I
=
550
- 7~ ~. . .2 fg Soo I ges gg 2g3 O 23 0 26 26h 27 17h 2e DD b)
- . O C dg a reS VS Core elevatj0n at Selected transient times 000 - -
-*
- T = 20.0 5 900 - .. ..; . .1' , -
*T=40.0$
sf./ v ,
- T = 60.0 S 800 g l
# f'. s;. *
*T=30.0$
700 - l l '
- i = 00.0 5 l ' '
' 'g'\-
W , _ . ' T = 120.0 S sa: . . .
, I f
I m T = wo.o S Soo . , j
. - ..r_ J_l_ .. ... .! _
300
#3A 26 26h 27 27h 28 ges gg 2,3 El.fVATION ((
' e elevation at Selected transient times to1@ o F-14
+
a 950 i 900 -
-er l a o T = 0.0 S 850 -
\ \ Mi a T = 2.0 S 2 \
. T = 3.0 5 v 800 - \
g , A.1 m T + 6.0 5 2 750 x
- y . \ . T = 8.0 5 g mo -
-
- T = 10.0 5 g *
% go - , -
i j... ; g _ 23 2h3 26 263 27 273 28 28 3 29 29 3 , D.IVATON (m) Fig. F-14a. Hot-rod cladding temperatures vs core elevation at selected transient times from 0 o 10 s for core cell 14. . l l 1000 ,.* . -
.-m.. e T = 20.0 S p
. .. ..g(
- 900 -
.q -
a T = 40 S m 1, 6 '*~~. * '
, ... '.,
- T = 60.0 S
{q 800 e../
, 'n, j
/ ,
= T = 80.0 5 l l
1 ,
w
~ '
j
- T = 100.0 5
" 'h ' ,'
j ', - R 600 s f ', - .
- ' T = 120.0 5 2E n :
J a T = MO.0 5
= . l .
a - f/_ ; = 4. . 4
,( -
1 f f I
, 2S 25.S 26 264 27 273 26 28 5 29 29 5 ELEVATON (m)
Fig. F-14b. Hot-rod cladding temperatures vs core elevation at selected transient times from 20 to 140 s for core cell 14. F-15
l e 500 1 950
< o T = 0.0 5 goo .
,f. *T=2.CS 850 # f p -
. T = 3.0 S v I 800 -
\ I q .
= T = 6.0 5 l W 750 -
- T = 8.0 S g
e 700 - P -
. T . o.O S l
a
@ .=- 1
.50
/ 's .. . ... _
l l 600
**..s- . .L '* .*. . % .. -
;,...~........
~n . :~~
,oo 26 253 26 263 27 273 26 283 29 29 3 j ELEVATION (my
- l 1 Fig. F-15a. Hot-rod cladding temperatures vs core elevation at selected transient times from 0 to 10 s for core cell 15. 000 9
~
.,,,.. ; o T = 2C.0 S
! ' ...-.
- T = 40.0 $
800 -
' f :' -
k . + T = 60.0 $
.' s ~ ?**st.,
f -
,,* = T = 80.0 S q
700 -
,t, m . .
l (e
, l . T = u0.0 S a .
l . . -
- ., = i = c0.0 $
l , 3oo .
, l ',, _
= T = wo.O $
4 / :
<m :- a bs..... -
soo 2S 233 26 263 27 273 28 25 3 29 29 3 ELEVATON (m) Fig. F-15b. Hot-rod cladding temperatures vs core elevation at selected transient times from 20 to 140 s for core cell 15. F-16
as 630 620 - -
., o T = 0.0 5 sto -
,* y ' * ~ . ,_ -
e ',
. 'j b 600 .- ' -
- T = 3.0 S
..'a...'., .
g g . .
. T = 6.0 $
tr 4 s
- T = 8.0 S 580 -
*
- T = 10.0 5 sm %- % N . -
m /---*--..-.-.'s - sso pg , i , 23 2S.S 26 264 27 273 28 283 29 29b ELEVATON (m) Fig. F-16a. Hot-rod cladding temperatures vs core elevation at selected transient times from 0 to 10 s for core cell 16. 900 850 - 800 / \ - o T = 20.0 S
-n. s ..
'8 3 _
- T = 4.0 5 z 700 - '
I
- T = 60.0 5 g 4............I . y
,.... ..... . T = a0.0 5 g ,,, _
l 8" ' g f f .
'
- T = 100.0 $
n j l . sso - ./
- j
- T = 120.0 S
_! ; _ m T = 140.0 5 t I i 430 - [ ! _ 400 a . l.: . . g..... . A 2b 2bb 26 26h 21 27h 28 28.6 29 29h ElfVATON (m) Fig. F-16b. Hot-rod cladding temperatures vs core elevation at selected tra 1si5 times from 20 to 140 s for core cell 16. F-17
i APPENDIX G LBLOCA TIMESTEP AND CPU TIME INFORMATION 1
- The entire extended LBLOCA calculation was performed on the HP9000 workstation and required 48 separate calculation runs. Figures G-1 and G-2 show the timestep size
- and total CPU time per calulation run. The tabulation below shows more detailed 3
information for each run. There were at least eight times when the calculation died because of code computation difficulties and had to be restarted using a smaller maxima timestep size. The code difficulties occurred primarily during the latter stages of reflood, between 140 and 230 s.
- The total calculation required 6,375,074 CPU s, which is equivalent to 73.8 days of ,
continuous computation on the HP workstation. Calculations 1 through 29 were 4 performed with an unoptunized code. The rest of the calculation runs were performed ! with the same code version optimized at an optimization level I but with selected j subroutines optunized at an optimization level 2. This optimized code was seven times faster than the unoptimized code for the AP600 LBLOCA calculation. l t l Calculation Run Calculation End Number of Total CPU Avg. Timestep { Number Time Time Steps Time (s) Size (s) 1 5 1448 62,296 0.003453 2 10 1418 54,647 0.003526
- 3 15 1247 60,558 0.004010
. 4 20 1091 58,508 0.004583 l 5 25 1240 73,313 0.004032 6 30 1324 62,654 0.003776 i 7 35- 1598 98,248 0.003129
- 8 40 1781 91,522 0.002807 l 9 45 4805 240,399 0.001041 i 10 50 3033 167,006 0.001649 11 55 2024 97,155 0.002470 _
12 60 1301 64,940 0.003843 i 13 70 2338 126,163 0.004277 14 80 4371 223,366 0.002288 l 15 90 4542 229,528 0.002202
- 16 100 4920 242,200 0.002033 J 17 110 8638 392,661 0.001158 j 18 120 7436 366,047 0.001345 19 130 8023 375,773 0.001246 1 20 140 7054 353,680 0.001418 21 144 3244 158,565 0.00123 j 22 147 3377 151,877 0.000888 i
4 G-1 4
\
lc Calculation Run Calculation End Number of Total CPU Avg.Timestep Number Time Time Steps Size (s) j _ Time (s) 23 150 2977 133,028 0.001008 i ! 24 152 2080 95,180 0.000962 25 153 1574 60,845 0.000635 l
- 26 156 1786 84310 0.001680 27 160 4983 212,729 0.000803 l
28 165 6310 186,838 _ 0.000792 29 168 3960 124,084 0.000758 I 30 174 21,320 223,172 0.000281 i 31 180 17,808 174,103 0.000337 32 185 10,435 97,466 0.000479 l ! 33 189 6760 64,111 0.000592 l 34 195 13,366 120,028 0.000449 _ j 35 200 13,228 130,089 0.000378 36 205 10,438 85,025 0.000479 l i 37 206 2000 16,723 0.000500 i I 38 210 16,234 137,886 0.000246 1 39 212 8840 75.964 0.000226 40 215 15,302 130,688 0.0001 % l ! 41 219 20,160 159,468 0.000198 42 225 15,026 132,682 0.000399 43 230 5056 39,997 0.000989 ' ~ , ! 44 240 6562 53,675 0.001524 .> 45 250 2819 24,212 0.003547 i I 46 260 3072 27,488 0.003255 ! 47 270 2016 16,745 0.004960 ! 48 280 2282 19,452 0.004382 Total 292,647 63,75,074 0.000957 i l Summary: l Totalnumberof timesteps 292,647 TotalCPU time (s) 6,375,074 l 106,251 j TotalCPU time (min) i TotalCPU time (h) 1770.85 TotalCPU time (days) 73.8 i Average timestep size for totalcalculation 0.000957 s l i t i . l i
- G.2 i
i
0 1 0.012 1 On l
- l 0.006 1 3 I i
h 0.00s 1 l o. 0.004
=
t Om2 l I 0.0=
. . _ . _ . me 0 25 50 5 10 0 15 150 15 200 225 250 24 300 mE (s) 5 Fig. G-1. Timestep size. l 4
1 t i 1 1 ! l L I 1 1 g 400000 - 350000'- j 300000 - . 80 g y 250000
~
8 2==0' i W f
- E '$a"-- -
.0 l
1 , . f - 20 I q l t O. AAA/ ..o m=
- 2. _ _ .
0 25 50 75 10 0 125 15 0 15 200 225 250 275 300 C TWE (s) Fig. G-2. Total CPU time per calculation run. G3
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