ML20126D054

RELAP5 THERMAL-HYDRAULIC Analyses of Pressurized Thermal Shock Sequences for H.B. Robinson Unit 2 Pressurized Water Reactor

ML20126D054
Person / Time
Site:	Robinson
Issue date:	04/30/1985
From:	Bolander M, Burtt J, Chon Davis, Fletcher C, Kullberg C, Ogden D, Stitt B, Waterman M EG&G, INC.
To:	NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
References
CON-FIN-A-6047, REF-GTECI-A-49, REF-GTECI-RV, TASK-A-49, TASK-OR EGG-2341, NUREG-CR-3977, NUDOCS 8506140624
Download: ML20126D054 (233)

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Idaho National Engineering Laboratory Operated by the U.S. Department of Energy RELAP5 Thermal-Hydraulic Analyses ,

of Pressurized Thermal Shock Sequences for the H. B. Robinson Unit 2 Pressurized Water Reactor N

4 C. Don Fletcher Benjamin D. Stitt Mark A. Bolander Craig M. Kullberg

-[ Michael E. Waterman Cliff B. Davis j John D. Burtt Donald M. Ogden April 1985 8506140624 e50430 PDR NUREG CR-3977 R PDR Prepared for the i

U.S. Nuclear Regulatory Commission Under DOE Contract No. DE-AC07 761DO1570 jl/ \s#

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e, NOTICE This report was prepared as an account of work sponsored by an agency of the United States Government 'is ther the United States Covernr.we.t nor any agency thereof,

- nor any of their employees, makes any warranty, expressed of Tiplied, or assumes any g

. legalliabihty or responsibehtv for any invi party's use, of the results of such use, of any

• information, apparatt s, potet or process disclosed in this report, or represents that its use by such third 5.vty would not infnnge privately owned rights.

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NUREGICR-3977 EGG-2341 Distribution Category: R4 e

RELAP5 THERMAL-HYDRAULIC ANALYSES OF PRESSURIZED THERMAL SHOCK SEQUENCES FOR THE H. B. ROBINSON UNIT 2 l 1

PRESSURIZED WATER REACTOR i i

e C. Don Fletcher Mark A. Bolander Michael E. Waterman John D. Burtt

,. Benjamin D. Stitt Craig M. Kullberg Cliff B. Davis Donald M. Ogden Published April 1985 EG&G Idaho, Inc.

Idaho Falls, Idaho 83415 l

e o Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C. 20565

- Under DOE Contract No. DE AC07 761D01570 FIN No. A6047 l

ABSTRACT Thermal-hydraulic analyses of fourteen hypothetical pressurized thermal shock

, (PTS) scenarios for the H. B. Robinson, Unit 2 pressurized water reactor were per-

. formed at the Idaho National Engineering Laboratory (INEL) using the RELAP5 computer _ code. The scenarios, which were developed at Oak Ridge National Laboratory (ORNL), contain significant conservatisms concerning equipment failures, operator actions, or both.

The results of the thermal-hydraulic analyses presented here, along with additional analyses of multidimensional and fracture mechanics effects, will be utilized by OR NL, integrator of the PTS study, to rssist the U. S. Nuclear Regulatory Commission in resolving the pressurized thermal shock unresolved safety issue.

i i

O $

a FIN No. A6047-Code Assessment and Applications (Transient Analysis).

il t

r EXECUTIVE

SUMMARY

Using the RELAPS computer code, thermal- Some scenarios start with the plant operating at full hydraulic analyses were performed at the Idaho power while others start from hot standby National Engineering Laboratory (INEL) to sup- conditions.

port the U. S. Nuclear Regulatory Commission's o investigation of the pressurized thermal shock (PTS) In general, steam line break scenarios were found unresolved safety issue. to be thermal-hydraulically severe for PTS (low reactor vessel downcomer fluid temperatures and The plant analyzed was the H. B. Robinson high pressures). Automatic clost.res of the main Unit 2 (HBR-2) pressurized water reactor (PWR); steam isolation valves (NISIVS) were not demanded, C

a Westinghouse three-loop design with c rated resulting in continuous plant cooldowns. hianually thermal power of 2300 N1W. An extensive computer closing the htSIVS and terminating auxiliary feed-model of the plant was developed specifically to water to affected steam generators are important perform the calculations needed for the analyses operator action techniques for these scenarios.

presented in this report. The model contains detailed thermal-hydraulic representations of Small break loss-of-coolant accident (LOCA) pertinent PWR primary and secondary systems, scenarios were found to be of limited thermal-including the feedwater train and steam lines. hydraulic severity. For very small breaks, the Detailed models are also included for the steam primary coolant system does not void sufficiently dump, steam generator level, pressurizer level, and to cause a loss of loop na: ural circulation flow and other plant control systems.

this allows the reactor vessel downcomer fluid tem-perature to remain elevated. For larger breaks, loop The model was quality-assured in four ways.

First, the development of each model component naturaWculanon hw is bst, mubg in very cd reacter vessel downcomer fluid temperatures. How-was documented on worksheets that include references to the plant documents supporting the ever, f r the larger breaks tha primary system depressurizes more than for the smaller breaks and development. Second, the worksheets were inde-

. this limits the thermal-hydrauhc severity. Counter-pendently reviewed by an analyst other than the one part calculations, performed with the same break who developed them. Third, utihty analysts, already size, indicate the hot leg break is more thermal-famihar with design and modehng of the plant, hydraulically severe than the cold leg break because reviewed both the model at various stages of com-pletion and the calculational results. Fourth, the the depressurization with the cold leg break is larger simulation of a plant transient was performed with during the first two hours of the event sequence.

the completed model, and results were compared The steam generator tube rupture scenarios were with measured plant data. The comparison appears not found to be thermal-hydraulically severe. Plant in is rep rt.

response for these scenarios was principally deter-Fo arteen scenarios of PTS interest are analyzed mined by the assumed operator actions.

in this report. The scenarios were defined at Oak Ridge National Laboratory (ORNL), the While results shown in this report cover analyses integrator of the PTS study. Computer simulations of 14 key scenarios, additional results are presented of the scenario event sequences were performed in NUREG/CR-3935 for 169 other scenarios of using best estimate conditions and assumptions. PTS interest. These additional scenarios are similar However, the scenarios themselves contain signifi. to those presented in this report with variations in o cant conservative assumptions concerning equip. operator actions and equipment failures.

ment failure 3, operator actions or omissions, or combinations of these. The results of the thermal-hydraulic analyses in this report represent part of the information The scenarios analyzed were initiated by main required by ORNL for the assessment of the PTS 0 steam line break, stuck-open steam line valve, steam issue. The results of this report are not to be used generator overfeed, small hot and cold leg break, di ectly as an indication of pressurized thermal stuck-open pressurizer valve, steam generator tube shock severity for the scenarios investigated. For rupture, and loss of secondary heat sink events. this purpose, comprehensive results of the analyses iii

m (far beyond the results shown here);have been : and publish a report that estimates the likelihood transmitted to ORNL. Following additional of reactor vessel failure and identifies important analyses of multidimensional and fracture event <equences, operator and control actions, and mechanics effects, ORNL will integrate all results uncertainties.

O 4

e iv

r CONTENTS ABSTRACT ......... ........ .................. ....................... ........... ii

SUMMARY

. .............. .... . ... .......... ............. ................... iii o NOMENCLATURE ....................... ......................... ................ xx

-1. INTRODUCTION ......... ................... ................................. I o 2. MODEL DESCRIPTION ............ .............................. ........ .... 3 2.1 Thermal-Hydraulic Model .... .. .......................... ...... ...... .. 3 2.1.1 Primary System ............ ........... ............................. 3 2.1.2 Secondary System ............... .................... ........... .... 3

' 2.2 Control System Model ............................ ......................... 11 1

2.2.1 Steam Dump Control System ............. ............ ................ 11 2.2.2 Steam Generator Level Control System ..... ................... ...... 12 2.2.3 Pressurizer Pressure Control System .... ............ .. ................ 13 2.2.4 Pressurizer Level Control System .................... ............. ... 13 2.2.5 Additional Control Systems . .................................... ..... 13 2.3 Steady State Conditions ... ......... ....................................... 13 2.3.1 ~ 2300 MW Steady State ................. .. ............ ............. 14 2.3.2 2200 MW Steady State ................................................ 14 2.3.3 Hot Standby Steady State .............. .................. ........... 14

3. SIMULATION OF H. B. ROBINSON PLANT TRIP FROM 2200 MW ............... 16 3.1 Description of Plant Trip Test .............. ........................... ..... 16 3.2 Comparison of Results ...................................................... 16 3.2.1 Steam Dump Control System Response ........... ...................... 16 3.2.2 Steam Generator Level Control System Response ......................... 18 3.2.3 Pressurizer Pressure Control System Resim.;e ..... .... ................. 20 3.2.4 Pressurizer Level Control System Response ............. . .............. 20 3.3 Conclusions ..... ........................... .............................. 20 4 4. SCENARIO 1,1.0-FT2 STEAM LINE BREAK AT HOT STANDBY ................. 22 4.1 Scenario Description .. ..................................................... 22 4.2 Model Changes .......... ........... ........ .... ..... .................. 22 0  %

4.3 ' Results ..................................................I............... 22 l

4.3.1 Calculation Results ...... ...... ... ... ............... ..... ........ 22 4.3.2 Extrapolations and Uncertainties ......... .............................. 31 v

4.'4 Conclusions ................. .... ....... .......... ........ ............. 31

5. SCENARIO 2, DOUBLE. ENDED STEAM LINE BREAK AT HOT STANDBY ........ 32 5.1 Scenario Description ... .......................................... ......... 32 5.2 Model Changes .. ........ ................................... ............. 32 5.3 Results ................................. .......... ......... ...... ...... 32 5.3.1 Calculation Results ... ...................................... ......... 32

,5.3.2 Extrapolations and Uncertainties ........................................ 38 a 5.4 Conclusions ........................ .................... ................. 41

6. SCENARIO 3, STUCK-OPEN STEAM LINE PORV AT HOT STANDBY ............ 44 6.1 Scenario Description ........................................................ 44 6.2 Model Changes ............................................................. ' 44 6.3 Results ................... ............ .... .............................. 44 6.3.1 Calculation Results .................................................... 44 6.3.2 Extrapolations and Uncertainties ........................................ 52 6.4 Conclusions ............................................... ................ 52

7. SCENARIO 4, THREE STEAM DUMP VALVES FAIL OPEN AT FULL POWER .... 56 7.1 Transient Scenario Description ............................................... 56 7.2 Model Changes ............................................................. 56 7.3 Results .................................................................... 56 7.3.1 Calculation Results ............ ............................... . ..... 56 7.3.2 Extrapolations and Uncertainties ........................................ 61 7.4 Conclusions ................................................................ 68

8. SCENARIO 5. OVERFEED WITH AUXILIARY FEEDWATER AT FULL POWER .................................................................. 69 8.1 Scenario Description ........................................................ 69 a

8.2 Model Changes ............................................................. 69 8.3 Results .................................................................... 69 8.3.1 Calculation Results .................................................. . 69 8.3.2 Extrapolations and Uncertainties .................. ..................... 75 8.4 Conclusions ................................................................ 75 vi

9. -SCENARIO 6,2-1/2-IN. HOT LEG BREAK AT FULL POWER ............ ... . 80 l 9.1 Scenario Description ........ . ..................... ................ . .. 80 L

9.2 Model Changes ............. ...................... .............. ... . ... 80 9.3 ' Results ......... ........................... ......... . . .. ........... 80 e

9.3.1 Calculation Results ................... ..... ....... ............... .. 80 9.3.2 Extrapolations and Uncertainties ............ ...... ......... ..... ... 89

o. 9.4 Conclusions ....................... ........................................ 92

10. SCENARIO 7, STUCK-OPEN PRESSURIZER PORV AT FULL POWER ............ 93 10.1 Scenario Description ........................................................ 93 10.2 Model Changes . ..... .................................. .................. 93 10.3 Results ...................................... ... ......................... 93 10.3.1 Calculation Results .................................................... 93 10.3.2 Extrapolations and Uncertainties ................... .................... 101 10.4 Conclusions ............ ...................................... ............ 101

11. SCENARIO 8,21/2.IN. HOT LEO BREAK AT HOT STANDBY ................... 104 11.1 Scenario Description ........................................................ 104 11.2 Model Changes .............................................. .............. 104 11.3 Results .................................................................... 104 11.3.1 Calculation Results .................................................... 104 11.3.2 Extrapolations and Uncertainties ........................................ 111

!!.4 Conclusions ................ ...................... ........................ 115

12. SCENARIO 9, STEAM GENERATOR TUBE RUPTURE AT HOT STANDBY ........ 116 12.1 Scenario Description ............................ .......................... 116 12.2 Model Changes ............................ ................................ 116 o 12.3 Results .................................................. ................. 116 12.3.1 Calculation Results ....................................... ............ 116 12.3.2 ExtrapolPtions and Uncertainties ........................................ 127 o

12.4 Conclusions ................................................................ 132

13. SCENARIO 10, STEAM GENERATOR TUBE RUPTURE AT FULL POWER ........ 133 vii

13.1 Transient Scenario Description .............................................. 133 13.2 Model Changes .............................. .............................. 133 13.3 Results .................................................................... 133 13.3.1 Calculation Results .................................................... 133 ,

13.3.2 Extrapolations and Uncertainties ........................................ 142 13.4 Conclusions ................................................................ 146

14. SCENARIO 11, LOSS OF SECONDARY HEAT SINK WITH PRIMARY SYSTEM FEED-AND. BLEED RECOVERY ................................................. 147 14.1 Scenario Description ........................................................ 147 14.2 Model Changes ............................................................. 147 14.3 Results .................................................................... 147 14.3.1 Calculation Results .................................................... 147 14.3.2 Extrapolations and Uncertainties ........................................ 154 14.4 Conclusions ................................................................ 158

15. SCENARIO 12,2.IN. COLD LEO BREAK AT FULL POWER ...................... 160 15.1 Scenario Description ........................................................ 160 15.2 Model Changes ............................................................. 160 15,3 Results .................................. ................................. 160 15.3.1 Calculation Resuhs .................................................... 160 15.3.2 Extrapolations and Uncertainties ........................................ 167 15.4 Conclusions ................................................................ 170

16. SCENARIO 13, STEAM OENERATOR TUBE RUPTURE AT HOT STANDBY WITH OPERATOR INTERVENTION ............................................. 171 16.1 Scenario Description ........................................................ 171 16.2 Model Changes ............................................................. 171 16.3 Results .................................................................... 171

• 16.3.1 Calculation Results .................................................... 171 16.3.2 Extrapolations and Uncertainties ........................................ 178 16.4 Conclusions ................................................................ 179

17. SCENARIO 14,2-IN. HOT LEO BREAK AT FULL POWER ....................... 183 17.1 Scenario Description ........................................................ 183 viii

t-17.2: Model Changes ................................. .... .......... .......... '183 17.3 Results .................. ....... ................. ..................... 183.

17.3.1 Calculation Results ................. ............... ......... .... ... 183 17.3.2 Extrapolations and Uncertainties ........ ................ .............. 190 1

17.4 Conclusions .. ...................... ........................ .......... 190

18. OVERVIEW AND CONCLUSIONS ... .................. .. . ....... ........ .. 195

, APPENDIX A-COMPUTER RUN TIME STATISTICS ....... ......................... A-1 FIGURES 2-1. Nodalization of primary coolant loops (Loop C shown) ..... ........ . ............ 7 2-2.' Nodalization of reactor vessel ............. .. .......... ........................ .8 2-3. Nodalization of steam generator (SGA shown) ...... ...... ...... .............. 9

.2-4. Nodalization of feedwater and steam systems ... .............................. .. 10 3 1. Plant trip test measured and calculated primary system highest average temperature responses ...................................................................... 17 3-2. Plant trip test measured and calculated steam generator level responses ............... 17 3-3. Plant trip test measured and calculated feedwater flow rate responses ................. 19 3-4. Plant trip test measured and calculated steam flow rate responses .................... 19 3-5. Plant trip test measured and calculated pressurizer pressure responses ........... . 21 3-6. ' Plant trip test measured and calculated pressurizer level responses ................. . 21 4-1. Scenario I primary and secondary pressures versus time ............................. 25 4-2. . Scenario i pressurizer narrow range indicated level versus time ............... ....... 25 4-3a. Scenario I break mass flow versus time for first 60 s of transient .................... 26 4-3b. Scenario I comparison of break .. ass flow and auxiliary feedwater flow versus time .... 26

' 4-4. Scenario I comparison of steam generator masses versus time ....................... 27 4-Sa. Scenario I comparison of primary loop cold leg mass flows versus time ............... 27 4-5b. Scenario I comparison of primary loop cold leg mass flows versus time (reduced ,

8 28 scale) - .............................. .... ................................ .... l l

4-6. Scenario I fluid temperatures versus time in primary Loop B ........................ 28 4-7. Scenario I fluid temperatures versus time in primary Loop C (W/PZR) .... .......... 29 ix

4-8. Securio 1 fluid temperatures versus time in primary Loop A (W/ASG) ............... 29 4-9. Scenario I comparison of primary loop cold leg temperatures and downcomer temperature versus time ....... . . .... . ............... .................... 30 1

410. Scenario 1 extrapolated downcomer inner wall surface heat transfer coefficient versus time .. .. ......... ..... ...... . . . . ............ .... . ...... ... 30 5-1 Scenario 2 primary and secondary pressures versus time ........... ...... ... .... 34 5-2a Scenario 2 break mass flow versus time for first 30 s of transient ................ ... 35 2b Scenario 2 comparison of break mass flow and auxiliary feedwater flow versus time .... 35 5-3. Scenario 2 comparison of steam generator masses versus time .... ..... ............. 36 5-4a. Scenario 2 comparison of primary loop cold leg flows versus time .. ............. . 36 5-4b. Scenario 2 comparison of primary loop cold leg flows versus time (reduced scale) ...... 37 5-5. Scenario 2 fluid temperatures versus time in primary Loop B ........ ....... ....... 37 5-6. Scenario 2 fluid temperatures versus time for primary Loop C (W/PZR) .. ........... 38 5-7. ' Scenario 2 fluid temperatures versus time for primary Loop A (W/ASG) ............. 39 5-8. Scenario 2 comparison of primary loop cold leg temperatures and downcomer temperature versus time .................................... ..... . .. ... .. 39 5-9. Scenario 2 extrapolated downcomer pressure versus time ............. .............. 40 5-10. Scenario 2 extrapolated downcomer fluid temperature versus time .................... 40 5-11. Scenario 2 extrapolated downcomer inner wall surface heat transfer coefficient versus time' ....................... ............................................ 42 512. Scenario 2 extrapolated primary loop cold leg flows versus time .......... ........... 42 5-13. Scenario 2 extrapolated primary loop cold leg temperatures versus time ... ...... . .. 43 6-1. Scenario 3 secondary system pressures ..................... . .................... 46 6-2. Scenario 3 steam generator heat removal rates .................... ...... . ....... 46 6-3. Scenario 3 reactor vessel downcomer pressure .................................... . 47 6-4. Scenario 3 affected steam generator auxiliary feedwater and break mass flow rates ..... 48

'6-5. Scenario 3 pressurizer normalized level indication ................................... 48 6-6. Scenario 3 high pressure injection mass flow rate (per loop) ... ...... .............. 49 6-7. Scenario 3 cold leg mass flow rates near the reactor vessel ........................... 50 6-8. Scenario 3 steam generator secondary masses .. ....................... . ........ 50 x

.s _ _...__ . . _ . . . . . . _ . . . _ . . . . _ . . . . . . _ . . _ . . _ _ _ _

6-9. Scenario 3 normalized steam generator narrow range indicated levels .... . .. 51 6-10. Scenar;o 3 cold leg and reactor vessel downcomer fluid temperatures .... . . .. . 51 6-11. Scenario 3 extrapolated reactor vessel downcomer fluid pressure ..... ... . .. . 53 6-12. Scenario 3 extrapolated reactor vessel downcomer fluid temperature . ... . . ... 53 6-13. Scenario 3 extrapolated reactor vessel wall inside surface heat transfer coefficient . 54 6-14. Scenario 3 extrapolated cold leg flow rates ..... ... . ......... . .. ... .. 54

~

6-15. Scenario 3 extrapolated cold leg fluid temperatures .. ........ . . 55 7-1. Scenario 4 primary and secondary system pressures .. ..... .... . .. . .. . 59 7-2. Scenario 4 decay power and total steam generator power . ..... . .. ..... ... .. 59 7-3. Scenario 4 motor driven AFW flows ... .. .... . .. . ... .. . . 60 7-4. Scenario 4 steam driven AFW flows ..... . .. .. . . ... ..... .... ..... .. 60 7-5. Scenario 4 steam generator mass inventory .. .... . . ...... . . . . .. ... . 62 7-6. Scenario 4 primary coolant mass flow rates . . .. ........ .. . .. ..... 62 7-7. Scenario 4 HPI mass flow rates .... . ...... .. ... . ... .... .... .... ... 63 7-8. Scenario 4 CVCS net injection flow rate ..... ... . .... .... . ... ....... 63 7-9. Scenario 4 reactor vessel downcomer fluid temperature ........ .. ... ...... . 64 7-10. Scenario 4 normalized pressurizer level indication ..... ...... .. ..... ... . ... 64 7-11. Scenario 4 extrapolated reactor vessel downcomer pressure . .. .... ....... 65 7-12. Scenario 4 extrapolated reactor vessel downcomer fluid temperature ........ ..... 66 7-13. Scenario 4 extrapolated cold leg fluid temperatures .. .. . . . . .. .. 66 7-14. Scenario 4 extrapolated reactor vessel downcomer inside surface heat transfer coefficient . .... .... .. .. . .. .. . . ... . .. ..... ... 67 7-15. Scenario 4 extrapolated cold leg mass flow rates . ... ....... . . .. ... 57 8 1. Scenario 5 primary system pressure ..... . .... . . . ........ .. ........ 72 8-2. Scenario 5 total dump valve flow rate . .. . ..... .. . . .. .. . 72

! 8 3. Scenario 5 cold leg flow rate ... .. .. .. ...... ...... ........ . . . 73 r o 8-4. Scenario 5 feedwater flow rates .. . . .. .. ..... . ..... . . ... 73 8-5. Scenario 5 feedwater fluid temperatures . .... . . . . . . ... .. .. .. 74 xi l

i

8-6. Scenario 5 total primary to secondary heat transfer ... ................ ............ 74 8-7.- Scenario 5 reactor vessel downcomer fluid temperature .............................. 76-8-8.- Scenario 5 cold leg fluid temperatures ........ ......... ......................... 76

~ 8-9. I 3cenario 5 pressurizer normalized liquid level ........... ................... ...... 77 8-10. Scenario 5 steam generator narrow range normalized liquid levels .......... . ...... 77 8-11. Scenario 5 extrapolated reactor vessel downcomer pressure .................. ....... 78 8-12. ~ Scenario 5 extrapolated reactor vessel downcomer fluid temperature .................. ~ 78 8-13. Scenario 5 extrapolated reactor vessel downcomer heat transfer coefficient ............ 79 9-1. Scenario 6 primary system pressure ........................................... ... 83 9-2. Scenario 6 normalized pressurizer liquid level ...................................... 83

3. Scenario 6 steam generator secondary pressure ... ................... .. .......... 84 9-4. Scenario 6 core power versus total primary to secondary heat transfer rate ............ 84 9-5. Scenario 6 total steam dump valve mass flow rate ............ ..................... 85

6. Scenario 6 total break mass flow rate versus total ECC/CVCS mass flow rate ........ 85 9-7. Scenario 6 motor. driven auxiliary feedwater mass flow rate .......................... 86 9-8. Scenario 6 steam'-driven auxiliary feedwater mass flow rate .......................... 86 9-9. Scenario 6 normalized steam generator narrow range liquid level .......... .......... 87 9-10. Scenario 6 primary cold leg mass flow rate ..................... .................. 87 9-11. Scenario 6 primary cold leg fluid temperatures and vessel downcomer fluid

. temperatures .................................................. . .............. 89 9-12. Scenario 6 extrapolated reactor vessel downcomer pressure at an elevation equal to the top of the core ........................................................... 90

13. _ Scenario 6 extrapolated reactor vessel downcomer fluid temperature at elevation equal to t he top o f t he core ' . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 90 9-14. Scenario 6 cxtrapolated reactor vessel downcomer wall surface heat transfer coefficient at elevation equal to the top of the core ........... .... ................ 91 9-15. Scenario 6 extrapolated cold leg mass flow rates ................................... 91 916. _ Scenario 6 extrapolated cold leg fluid temperature ........................ ......... 92 =

10-1. Scenario 7 primary system pressure ............... ...............................  %

10-2. Scenario 7 hot and cold leg temperatures .................. .......................  %

xii

10-3. Scenario 7 cold leg mass flow rates ... ..... .... .. . .. ... ..... . .. 97 10-4. Scenario 7 reactor vessel upper head and top of downcomer void fractions ... . .. 97 10-5. Scenario 7 normalized pressurizer level .. . .. ..... .. .. ... ... 98 10-6. Scenario 7 steam generator pressures . . .. .. . . ... . .. ... . 98 e

10-7. Scenario 7 total HPl flow and flow out PORV .... . . . . . .. . .. ... 99 10-8. Scenario 7 feedwater flow rates . ... .. ... ... . . .. . .. . .. 100 9

10-9. Scenario 7 steam generator normalized liquid levels . .. .. . ..... . .. 100 10-10. Scenario 7 extrapolated reactor vessel downcomer pressure .. ..... .. .. .. 101 10-11. Scenario 7 extrapolated reactor vessel downcomer temperature ........... .. . .. 102 10-12. Scenario 7 extrapolated cold leg mass flow rates .. . .. ... . . . . . 102 10-13. Scenario 7 extrapolated reactor vessel downcomer heat transfer coefficient . .... . 103 11-1. Scenario 8 pressurizer pressure ... .... .. ... ..... .. .. . .. .. . 107 11-2. Scenario 8 break and HPl plus makeup flows .. .. .. ......... .... ...... 107 11-3. Scenario 8 void fraction at the break . .. .. . . ..... ... ............ 108 11-4. Scenario 8 accumulator liquid volumes . .. .. . .. . ....... . . ....... 108 11-5. Scenario 8 upper head and core void fractions . ..... . ...... . ............ 109 11-6. Scenario 8 steam generator heat transfer rates . ....... .. . . . . .... .. .. .... 110 11-7. Scenario 8 cold leg mass flow rates . .. .. . ... . ... . ........ .... 110 11-8. Seenario 8 cold leg and downcomer fluid temperatures . .. . . . .. .... ..... 112 11-9. Scenario 8 extrapolated reactor vessel downcomer pressure ... .... ... ............ 112 11-10. Scenario 8 extrapolated reactor vessel downcomer temperature ..................... . 113 11-11. Scenario 8 extrapolated reactor vessel wall inside surface heat-transfer coefficient .. . 113 11 12. Scenario 8 extrapolated cold leg mass flow .. . ... . . . . .. . 114 a

11-13. Scenario 8 extrapolated cold leg fluid temperature .. . ... .. .. ... .. . .. 114 12-1. Scenario 9 nodalization for broken tube in steam generator A . ... ... ... ....... 118 O

! 12-2. Scenario 9 break mass flow rates . .. ... .. ... ....... .. ... ..... ..... 120 l

12-3. Scenario 9 reactor vessel downcomer fluid pressure . . . .. ... .... ...... 120 12-4. Scenario 9 pressurizer level indication .. .. . . . .. ... . . . .. .... 121 xiii

{ 12-5. Scenario 9 steam generator secondary mass inventories ............................... 121

! !2-6.' Scenario 9 main feedwater bypass regulating' valve flow rate .......... ........ ..... 122

7. . Scenario 9 comparison of injection and break' mass flow rate .............. ......... 123 12-8. Scenario 9 steam generator secondary pressures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 123 12-9. ' Scenario 9 steam generator heat removal rates ..................................... 124 L 12-10. Scenario 9 cold leg mass flow rate - . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 125 12-11. Scenario 9 cold leg fluid temperatures ........ .................................... 126

• t

'12-12. Scenario 9 reaurd vesse! downcomer fluid temperature .............................. 126 12-13. Scenario 9 caletss:ed and adjusted reactor vessel downcomer fluid pressure ............ 128--

t 112-14.' . Scenario 9 calculated and adjusted reactor vessel downcomer fluid temperature ........ 128 12-15. ~ Scenario 9 calculated and adjusted reactor vessel downcomer inside surface heat transfer coefficient ........ ..................................................... 129

~ 12-16.' Scenario 9 calculated and adjusted Loop A cold leg mass flow rate ................... 129 12-17. Scenario 9 calculated and adjusted Loop B cold leg mass flow rate ... . ............. 130 12-18. Scenario 9 calculated and adjusted Loop C cold leg mass flow rate . . . . . . . . . . . . . . . . . . . 130

19. Scenaric 9 calculated and adjusted Loop A cold leg fluid temperature ................ 131 12-20. Scenario 9 calculated and adjusted Loop B cold leg fluid temperature ................. 131

21. Scenario 9 calculated and adjusted Loop C cold leg fluid temperature ................ 132' 13-1. Scenario 10 primary system pressure .............................................. 134

~13-2. Scenario 10 normalized ' pressurizer liquid level .................... ................ 136 13-3. Scenario 10 total steam dump valve mass flow rate ................................. 136 13-4. Scenario 10 core power versus total primary to secondary heat transfer rate ........... 137 13 5. ' Scenario 10 steam generator secondary pressure .................................... 137 13-6. Scenario 10 total break mass flow rate versus total ECC CVCS mass flow rate ........ 139 13-7. - Scenario 10 normalized steam generator narrow range liquid levels ................... 139 13-8. Scenario 10 motor driven auxiliary feedwater mass flow rate ............. ........... 140 f Scenario 10 primary cold leg mass flow rate

. . . t j 13 9. ....................................... 140  ;

13-10. Scenario 10 primary cold leg fluid temperatures and vessel downcomer fluid temperature .................................................................... 141

. xiv

~

13-11. Scenario 10 extrapolated vessel downcomer pressure at elevation equal to the top of.the core .............................. ... .. ........ .................... 142.

J13-12. . Scenario 10 extrapolated vessel downcomer fluid temperatures at elevation equal to the top of the core ............... .. ... ..... ..... ............ .... ... 143

.. .13-13. -. Scenario 10 extrapolated vessel downcomer wall inside surface heat transfer coefficient at elevation equal to the top of the core . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 143 13-14. ' Scenario 10 extrapolation of cold leg mass flow rates . . . . . . . . . . . . . . . . . . . . . . . . . . . . 144

• 13-15. Scenario 10 extrapolation of cold leg fluid temperatures .... ........................ 144

- 13-16.- Scenario 10 vessel downcomer pressure with uncertainty ... ......................... 145 13-17. Scenario 10 vessel downcomer fluid temperature with uncertainty ................... 145 14-1. Scenario 11 reactor vessel downcomer temperature ... ... .............. ........ 150 14-2. _ Scenario 11 reactor vessel downcomer pressure ....... .............. . ......... 150 14-3. Scenario 11 primary mass inflows and outflows ....... .... . ................... 151 14-4. Scenario 11 cold leg flow ......................................................... 153 14-5. Scenario 11 steam dump flow .......................... ......................... 153 14-6. Scenario 11 steam generator wide range levels .................. ................... 155 14-7.' Scenario 11 steam generator heat transfer rates ..................................... 155 14-8. Scenario 11 reactor vessel downcomer pressure response, 4000-11000 s ................ 156 14-9. Scenario 11 reactor vessel downcomer temperature responses, 4000-11000 s ............ 156 14 10. Scenario 11 reactor vessel downcomer wall heat transfer coefficient 4000-11000 s ....... 157 14-11. Scenario 11 cold leg discharge mass flow rate responses, 4000-11000 s ................ 157 14 12. Scenario !! cold leg discharge volume temperature responses,4000-11000 s ............ 159 15-1. Scenario 12 primary system pressure .............................................. 163 15 2. Scenario 12 net CVCS injection flow .............................................. 163

1. 15-3. Scenario 12 cold leg break mass flow rates ........................................ 164 15-4. Scenario 12 primary loop mass flow rates .............................. .......... 164

' 15 5. Scenario 12 main flow auxiliary feedwater mass flow rates .......................... 165

'15-6. Scenario 12 break fluid and saturation temperatures ................................ 165 15 7. Scenario 12 primary system pressure .............................................. 166 xv

'15-8. Scenario 12 downcomer mass flow rate .......... ................................. 166

9. Scenario 12 extrapolated downcomer pressure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 168 15-10. Scenario 12 extrapolated downcomer fluid temperature ... .......................... 168 15-11. Scenario 12 extrapolated downcomer wall heat transfer coefficient .................... 169 .

15 12. Scenario 12 extrapolated cold leg flow rates ............. ......................... 169 15-13. Scenario 12 extrapolated cold leg fluid temperatures ............................. .. 170 e

'16-1. Scenario 13 break mass flow rate ........ ....... ................................ 174 16-2. Scenario 13 primary and secondary system pressures ................................ 174 16-3. . Scenario 13 normalized pressurizer level ................................ .......... 175 16-4.- Scenario 13 steam generator secondary liquid masses ................................ 175 16-5. Scenario 13 main feedwater mass flow rates ................. ..................... 176 16-6. Scenario 13 HPI and makeup mass flow rates ..................................... 177 16-7.- Scenario 13 reactor vessel downcomer fluid temperature ............................. 177 16-8. Scenario 13 cold leg mass flow rates ............................................. 178 16-9. Scenario 13 reactor vessel upper head void fraction ................................. 179 16-10. Scenario 13 extrapolated primary system pressure . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

• 180 16-11. Scenario 13 extrapolated reactor vessel downcomer pressure ......................... 180 16-12. Scenario 13 reactor vessel wall heat transfer coefficient ............................. 181 16-13. Scenario 13 extrapolated cold leg mass flow rates .................................. 181 16-14. Scenario 13 extrapolated cold leg temperatures ..................................... 182 17-1. Scenario 14 primary system pressure .............................................. 186 17 2. Scenario 14 normalized pressurizer pressure ........................................ 186 17-3. Scenario 14 secondary system pressures ........................................... 187 17 4. Scenario 14 core power and steam generator heat remo' val rates ...................... 187 17 5. Scenario 14 break and total ECC mass flow rates .................................. 188 P

17-6. Scenario 14 steam generator levels ................................................ 188 17 7. Scenario 14 cold leg mass flow rates .............................................. 189 17 8. Scenario 14 cold les and reactor vessel downcomer fluid temperatures ................ 189 xvi

e _

17-9. Scenario 14 extrapolated primary system pressure ......... . ... ... .. ............ 190 r 10. - Scenario 14 cxtrapolated reactor vessel downcomer .............. .................. 191 17-11. Scenario 14 extrapolated reactor vessel wall heat ...... ............................ 191

'17-12. Scenario 14 extrapolated loop A cold leg mass flow rate ............ ....... ....... 192 17-13. Scenario 14 cxtrapolated loop B cold leg mass flow rate .......................... . 192 17-14. Scenario 14 cxtrapolated loop C cold leg mass flow rate ................. .......... 193 e

17-15. . Scenario 14 extrapolated loop A cold leg temperature ~...................... ........ 193

'17-16. Scenario 14 extrapolated loop B cold leg temperature ' . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 194 t

17-17. Scenario 14 extrapolated loop C cold leg temperature ............................... 194 l A 1. Plant trip rate of CPU time usage ................................................A-5 A 2. - Scenario I rate of CPU time usage ...............................................A-5 i

! - A-3. Scenario 2 rate of' CPU time usage ...............................................A-6 I

A-4. Scenario 3 rate of CPU time usage ...............................................A-6 i A 5. Scenario 4 rate of CPU time usage ...............................................A-7 i.

i A-6. Scenario 5 rate of CPU time usage ...............................................A l A 7. Scenario 6 rate of CPU time usage ............................................... A.8 I A-8. Scenario 7 rate of CPU time usage ...............................................A-8 i

A-9. Scenario 8 rate of CPU time usage ...............................................A-9 i A-10. Scenario 9 rate of CPU time usage ............................................... A 9 i

A II. Scenario 10 rate of CPU time usage .............................................. A 10 l

l A 12. Scenario 11 rate of CPU time usage .............................................. A 10 A 13. Scenario 12 rate of CPU time usage .............................................. A ll i

A 14. Scenario 13 rate of CPU time usage .............................................. A Il i a j A 15. Scenario 14 rate of CPU time usage ..............................................A.12 i

TABLES 1 1. Summary of scenarios analyzed .................................................. 2 l

i

! 2 1. RELAP5 model nodalization numbering scheme .................................... 4 2 2. 2300 MW initial conditions ...................................................... 14 xvil

p..

2-3. 2200 MW initial conditions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 2-4. Hot standby initial conditions .............................................. ..... 15 4-1. Scenario description No.1 ....................................................... 23 4-2. Scenario I sequence of events .................................................... 24 .

5-1. Scenario description No.2 ......... ............................... ............. 33 5-2. Scenario 2 sequence of events .................................................. . 33 6-1. Scenario description No. 3 ...................................... ................ 45 6-2. Scenario 3 sequence of events .................................................... 45 7-1. Scenaria description No. 4 ....................................................... 57 7-2. Scenario 4 sequence of events ........................... ........................ 58 8-l. Scenario description No. 5 ............................................... ....... 70 8-2. Scenario 5 sequence of events ......... .......................................... 71 9-1. Scenario description No. 6 ................................................... .. 81 9-2. Scenario 6 sequence of events .................................................... 82 10-1. Scenario description No. 7 ..................... ................................. 94 10-2. Scenario 7 sequence of events .................................................... 95 111. Scenario description No. 8 ....................................................... 105 112. Scenario 8 sequence of events .................................................... 106 12.I. Scenario description No. 9 ....................................................... 117 12-2. Scenario 9 sequence of events .................................................... 119 131. Scenario description No.10 ...................................................... 134 13 2. Scenario 10 sequence of events ................................................... 135 14-1. Scenario description No.11 ...................................................... 148 14-2. Scenario 11 sequence of events ................................................... 149 14-3. Scenario 11 extrapolated values .................................................. 159 D

15-l. Scenario description No.12 ...................................................... 161 15 2. Scenario 12 sequence of events ................................................... 162 xylii

7,

~__

r 16-1. Scenario description No.13 .................... ................................. 172

,. 16-2. Scenario 13 sequence of events .. ... ............................................ 173 17-1. Scenario description No.14 .................. ................................... 184 17-2. -Scenario 14 sequence of events ...... ............................................ 185 e

1. Summary tabulation of HBR-2 PTS analytical results ....................... ....... 1%

A-1. Timing statistics ................................................................A-4

-p.

O O

9 4

xix

NOMENCLATURE FW- Auxiliary feedwater ASO Affected steam generator CL Cold leg CP&L Carolina Power and Light Company CPU Central processing unit CVCS Chemical and volume control system DC Downcomer DELTA-T or (AT) Differential temperature ECC Emergency core cooling ESFAS Emergency safeguards actuation signal HBR-2 H. B. Robinson, Unit 2 HL Hot les HPI High pressure injection INEL Idaho National Engineering Laboratory LPI Low pressure injection LRC Load rejections controller MFIV Main feedwater isolation valve MFW Main feedwater MSIV Main steam isolation valve NR Narrow range NRC U.S. Nuclear Regulatory Commission ORNL Oak Ridge National Laboratory PI Proportional integral PLCS Pressurizer level control system PORY Power-operated relief valve PPCS Pressurizer pressure control system PTC Plant trip controller xx

4

-PTS Pressurized thermal shock

~PWR  ;

' Pressurized water reactor PZR Pressurizer RC Reactor coolant a

RCP- Reactor coolant pump RCS Reactor coolant system 9

RV. Reactor vessel l.

SDCS Steam dump control system SO (or S/0) Steam generator SGLCS Steam generator level control system SI Safety injection SIAS Safety injection actuation signal SPC Steam pressure controller T-ave Average of hot and cold leg temperatures USO Unaffected steam generator WR Wide range o

l-i e

2 xxl

RELAP5 THERMAL-HYDRAULIC ANALYSES OF PRESSURIZED THERMAL SHOCK SEQUENCES FOR THE H. B. ROBINSON UNIT 2 PRESSURlZED WATER REACTOR

1. INTRODUCTION Rapid cooling of a reactor pressure vessel during bases for the screening criteria in the proposed PTS

• a transient or accident, accompanied by high rule (proposed 10 CFR 50.61) and to determine the coolant pressure, is referred to as pressurized content required for licensees' plant-specific safety thermal shock (PTS). In late 1981 the U. S. Nuclear analysis reports and the acceptance criteria for cor-Regulatory Commission (USNRC) designated PTS rective measures.

as an unresolved safety issue and developed a task action plan (TAP A-49) to resolve the issue. The computer simulations presented in this report were performed using best estimate modeling The safety issue exists because rapid cooling at assumptions for plant conditions and responses to the inner surface of the reactor vessel wall produces the events specified in the scenario descriptions. The thermal stresses within the wall. As long as the frac- reader is cautioned, however, that for bounding ture resistance of the reactor vessel is high, over- purposes the scenario descriptions were based on cooling transients will not cause vessel failure, extremely conservative assumptions concerning flowever, USNRC staff analy<.es (SECY-82-465) equipment malfunctions, operator actions and showed that certain older plants with copper and omissions, or combinations of these. Thus, while other impurities in the vessel weldments may the computer simulations represent best estimate become sensitive to PTS after several years as the plant responses to the scenarios as defined, they do nil-ductility transition temperature of the weld not represent the most probab/c plant responses to material gradually increases, the scenario initiating events.

The purpose of the the. mal-hydraulie analyses Analyses presented in this report were performed presented in this report is to better understand the for the 11. II. Robinson, Unit 2 (IlllR-2) pressur-behavior of a plant during various kinds of Ized water reactor operated at liartsville, South postulated severe overcooling transients with multi- Carolina by the Carolina Power and Light Com-ple failures of equipment and without correctise pany. The reactor is of Westinghouse three loop operator action. The understanding gained from design and, at the time of this study, was operated these detailed calculations will be used to interpolate at a reduced power because of steam generator tube coolant temperature and pressure responser in the plugging. The simulation of the plant transient downcomer for other postulated transients using a presented in Section 3, was performed at a thermal simplified man and-energy balance approach. For power of 2200 AlW to agree with the operating each of these postulated transients, Oak Ridge power at the time of the transient. The simulation National Laboratory (ORNL) will then calculate the , of Scenarios I through 14, presented in Sections 4 reactor vessel temperature distribution and stresses through 17 were performed anuming a full rated during the transient, and the conditional probability thermal power of 2300 N1W. This is the power level of sewel failure if the tramient should occur. ORNL at which the plant is expected to operate following will pubthh a report that integrates these results to replacement of steam generators. Other anticipated estimate the likelihood of PIS driving a crack plant changes were also incorporated into the through the reactor seuel wall and to identify analyses of the scenarios, important esent sequences, operator and control

• actions, and uncertainties. Table 1 1 briefly identifies Scenarios I through 14 as defined at ORNI.. Full scenario lhls series of analyses is intended to proside descriptions appur in Sections 4 through 17, information to help the USNRC staff confirm the Analpes were performed for a 2.h period starting I

l l

4 at the beginning of cooldown. For all except Section 2, a comparison of computer simulated and Scenario 11 the period started at the time of the in- . measured data for a plant transient is presented in itiating event. For Scenario !! the period started at Section 3, analyses of PTS Scenarios I through 14 the time of steam generator dryout, si h after the are presented in Sections 4 through 17 followed by initiating event. an overview and conclusions in Section 18. Appen-dix A presents a timing survey that discusses the

• This report is organized in the following format: computer run time required to perform the a description of the computer model is given in calculations.

Tatdo 11. Summary of soonerlos analyzed ,

Analysis initial Plant Section in Scenarios Condition Initiating Event This Reporth

! Hot standby 1.0 ft2 break in main steam line 4.0 -

2 Hot standby Double-ended main steam line break 5.0 3 Hot standby Stuck + pen steam line PORY 6.0 4 Full power Three steam dump valves fail open 7.0

$ Full power Overfeed with auxiliary feedwater 8.0 6 Full power 21/2 in, hot leg break 9.0 7 Full power Stuck open pressurizer PORY 10.0 8 Hot standby 21/2 in hot les break 11.0 9 Hot standby Steam generator tute rupture 12.0 10 Full power Steam generator tube rupture 13.0 11 Full power Loss of heat sink with primary system feed and bleed 14.0 recovery 12 Full power 2 in. cold les break 15.0 13 Hot standby Steam generator tube rupture with operator action 16.0 14 Full power 2 in. hot les break 17.0 ,

s. CAUTION-Scenario numbering scheme has been changed by ORNL in their final HRR 2195 report,

b. A summary of pressure and temperature results,for these scenarios is given In Table 18 l.

2

l

2. MODEL DESCRIPTION i

j- This section describes the RELAPS HBR-2 PWR represented in the RELAPS model. The loops are i

model used for three steady state initializations and designated as A, B, and C. Each modeled loop con-l subsequently for each of the PTS transient calcula- tained a hot leg, U-tube steam generator, pump suc-tions and a simulation of a HBR-2 plant trip tion leg, pump, and cold leg as shown in Figure 2-1.

transient. The subsections describe the thermal- The pressurizer was attached to the C loop and the l hydrauHe components of the model, the control pressurizer spray lines were attached to the B and C system model, the steady state conditions for each loop cold legs, as shown in Figure 2-1. Attached to steady state initialization calculation, and the cach cold leg was a low pressure injection (LPI) port documentation control used for the models and and an accumulator with its associated piping. Also o code for these calculations, attached to the cold leg was a high pressure injec-l tion (H PI) port. The LPI and H PI models were set The model was quality-assured in four ways. up to inject one third of the total HPI and LPI flow First, the development of each model component into each loop. Also attached to the Loop B cold was documented on worksheets that include leg was the chemical and volume control system r references to the plant documents supporting the (CVCS). Makeup and letdown were modeled with development. Second, the worksheets were in- a single junction. Heat structures were added to dependently reviewed by an analyst other than the each volume in the primary loops to represent the one who developed them. Third, utility analysts, metal mass of the piping and steam generator tubes.

already familiar with design and modeling of the Heat structures were also used to represent the plant, reviewed both the model at various stages of pressurizer proportional and back.up heaters, completion and the calculational results. Fourth, the simulation of a plant transient was performed Figure 2 2 shows the RELAP5 nodalization used with the completed model, and results were com- to represent the HBR 2 vessel. The downcomer, i pared with measured plant data. The comparison downcomer bypass, lower plenum, core, upper appears in Section 3 of this report. plenum, and upper head were represented in the RELAP5 vessel model. The following leakage paths Calculations were performed using the were represented in the vessel model: downcomer .

RELAP5/ MODI.6, Cycle 16 computer code, ex- to upper plenum, downcomer to downcomer cept for Scenarios I and 14 which were performed bypass, downcomer bypass to lower plenum, cold  ;

using RELAPS/ MOD 2, Cycle 16. leg inlet annulus to upper plenum, and upper i plenum to the upper head by way of the guide tube.  ;

"'*' '"""' represented both external and 2,1 Thermal Hydraulic Model internalmetalmassof thevesselas wellasthecore rods. Decay heat was assumed to be at the ANS The RELAP5 modelis a detailed representation standard rate. )

of the HHR 2 PWR power plant, describing all the  ;

major flow paths for both primary and secondary There were 130 volumes associated with the I systems, including the main steam and feed systems. primary loops and 33 volumes associated with the Also modeled are primary and secondary power vessel.

operated tellef valves (PORV), and safety valves.

The emergency core cooling system (ECCS) was 2.1.2 Secondery System. The RELAP$ llHR 2 i included in modeling the primary side, and the PWR secondary system model is shown in ausillary feedwater system was included in Ihe I:Igures 2 3 and 2 4. The steam generator secondary

• secondary side modeling. The model contained model, shown in Figure 2 3, represents the major 224 solumes,242 junctions and 218 heat structures. flow paths in the secondary and includes the  :

A description of the primary and the secondary downcomer, boiler region, separator and dryer i systems are prnented in the following sections. region, and the steam dome. Due to modeling con.

. Table 21 summarlies the correspondence between straints, the steam generator secondary separators ,

l reactor nyitem and model components, and dr> cts were modeled within a single hydro- l l

d> namle volume. Separation in the model thus takes  ;

l 2.1.1 Primary System.1he lillR 2 PWR plant place at a sing!c clesation rather than at two loca. j has three primary coolant loops and each loop is lions (separator and dryer), as in the prototype  ;

3 1

Tahde 2-1. RELAP5 model Wd= tion numbering scheme Component Number (s) Description 100-129 Reactor Vessel 100, 102, 104, 106 Downcomer 110, 112 Lower Plenum 114 Core 116 Core bypass 118, 120, 122 Upper plenum 126 Upper head 129 Guide tubes 200-220 Loop A Primary System 204 Hot les -

206 Inlet plenum of S.O.

208 S.O. tube primaries 210 Outlet plenum of S.O.

212 Cold leg (pump section) 214 Reactor coolant pump 216, 218, 220 Cold les (pump discharge) 300 320 Loop B Primary System (Numbering comparable to 200-220) 400 420 Loop C Primary System (Numbering comparable to 200-220) 335 343 Pressurizer 335, 337, 339 Spray lines 336, 338 Spray control valves 344, 34$ PORV and containment 346,347 Safety valve and containment 340,341 Pressurizer 5essel 343 Surge line 254 282 Loop A Secondary System 254, 258, 262 Downcomer 266 lloller 270, 274 lloller separator region 278 Separator .

282 Stsam dome 354 382 Loop H Fecondary System (Numbering comparable to 234 282) 4

r; Tetde 21. (continued)

Component Number o (s) Description 454-482 Loop C Secondary System

, (Numbering comparabic to 254-282) 500-565 Loop A Feed and Steam Systems 505 Main feedwater control valve

$10,520 Main feedwater line

$15 Check valve

$25 Check valve

$30 Steam-driven auxiliary feedwater

$40 Motor-driven auxiliary feedwater

$50,560 Main steam line 555 Main steam isolation valve l 565 Check valve 570,575 Safety valve 580,585 PORY 600-665 Loop 11 Feed and Steam Systems (Numbering comparable to 500 565) 700 765 Loop C Feed and Steam Systems 1

l (Numbering comparable to 500-565) 800 810 Common Steam Systems 800 Steam header 802 Main steam line 8M, 806 Turbine stop valve 808, 810 Steam dump valves 822 878 Common Feedwater Systems l

l 822 Condenser hotwsil L , 824 Condensate pump 828, 834 Main feed line 830 1.ow pressure heaters I and 2 840 f.ow pressure heaters 3,4, and 5 844 1.ow pressure heater bypass line

1. $0, 852 lleater drain system l 834 Main feed header 860 Main feed pump suction 861,864 Main feed pumps 862, 865, 867 Main feed pump discharge 86),866 Check valves

]

5

Ta M e 2-1. (continued Component Number (s) Description 870 Main feed pump bypass line High pressure heater =

874 878 Main feed header 900 971 Additional Systems 911, 912, 913 Loop A, B, and C accumulators 921, 922, 923 Loop A, B, and C ECC lines 931, 932, 933 Loop A, B, and C low pressure injection 951, 952, 953 Loop A, B, and C high pressure injection 971 Makeup injection 6

344 PORV r c 345 346 Safety

_ _ j347l i

Steam 340 339 generator JT -

y-- ) 341 5 4 n p /- y--

~

Pressurizer 338 336 j

/

_(] g j_

' ~~

l / 7 l3 408 L_/

2 /

337 333 f'

/

~~~ ~-  !

Fr m Loop B p

'/////// /J l / 8 1 h 1 l t T

e ef  : ,s,s 1a 3 2 343 410 M i

404 405 - 3 2 1 _

zZ i// / / // / / / / 7- Reactor vessel

~

7~r r sssa c

! 41 410 2

• 7 418 1 l

l~I ~

RC pump '

l 420

/

/ 953 HPl 9I3

/ 4 Inu/u/uu/ Accumulator l 412 923 933 LPl t

=

l

! o

.30. (Makeup Loop B only) ligure 21, Notfalunlion of primary toolant loops (I oop C ilioant i

o l l t

I i 7

Upper

///// head 126 f

f 1 I

/ 3 / j '

Upper ,

100 / / g / plenum 2 2

/ / -

-/ 122

/, / ,

3 /

From cold O

( }' n

To hot

[ y g

( 102 ( 120 /

U u 104 118 h

I i l I l f

/ 1 / 1

6 /

l /

2 l 2 / s /

/ / Bypass

/

/ 3 3 j Core 3y Downcomer 106 p / 116 7

/ l f 0 / 0 j 2 /

j

/ / /

L-6 --

6 1 u u

_ {

/ 7 / 112 /

f h Lower t plenem

,ff// INCL 4 6111 ,

sisui, a munmi ,n a ,,.a ., ,,,,,i, e

i s

f ll

i To main steam line

//

Steam domo 282

/ t I

o ,

/ 278

/

7 7 254 / / 274 From feedwater - I Ed -  !

E'E* 270 258 y 1r 1/ 1 t

/ 1 / 4

% A

/ /

Downcomor

/ / N Dollor 202 y - \ ', 266

/ f /

/2 3 / l ._ yl , _ _ _ I/ 3

/ ,_ .., /  ! /

/

/

/

/ b 1 )l l

/

/

/3 2 , l l 2

/- ', ,. /

-_.k L-.

4 / 1 / l l j

' l

/ / l l z2_ . .; _ _ . <

s Outlet l l ] Inlot plonum l / plonum l l /

Tocoldlog \ /From hot 100 o \  % L!

Loop Componnnl Numbors

~~E 2nn o n 3sa C 4** INil 4 Bitt l

Ilptiff I l N0il4li/Allett of 4fretti grfwf atof (M)A blumft).

l 9

f R l -

OSS l 755 MSIV 555 . Steam MSNx kMSN 565 dump

-, s8o x gm So g- , 808 q1 =

l 545 l H 650 - T480% header l 802 PORw t 765 M 806 Ge5 l 750 TNh TurtWne Safeer }673-stop TurtHne O I 57s i Sasam Steam POW 5600 S*er. ,m ,

Steam l785 MM a n5 l So.

Condensate Condenser generaint generator M 9'"*'MC' PORV pump 824 A and B

^ 8 Sa,etr 52s y _725 , 2i , 8221 7  !

7 g-LP. heaters 1 and 2 v 844 Heater drain system Motorasrwen f LP. heaterbypass 852 aux feed 834 520 $20 line C I

Q Q Q f *5 71 I 1 LP. heaters 3,4 and 5 4

=

3.

L 2 l 3

=

l 1

f Moeorm uca*=en"3

,gwg

>615 710 7 Main feedwater aa 9eed Ene A 863y 862 header MFW pump 854

-- -- 705 i 861

878 h 510 ' #

610 air. ree:8 kne C 874 }- MFW pum B Saeam N 864 U 867 T' MFW aa feedkne B { 6o5 866 -

I Se 4*e.

anca Seed bne A

_ . ses

-~

no  :

MFW;typass watve e [-]

pumo 8m

]

,,3 mEL 4 5H3 rw 24 .u+ s or ree == =s se= .me==-

e e e e

r steam generator. The effect of this difference is a 2.2 Control System Model perturbation of the flow field at the upper sicam generator lesel tap that affects the indicated lesci in a minor way. A further dkeunion of this effect The purpose of this section is to proside the appears in Section 3. reader with a general merview of the functions of the major control systenu used in the 12 cateula.

The major flow paths of the steam line out to the lions that wcre described in subsequent sections, turblne gmcrnor vahes wcre modeled and are Detailed information regardir.g Ihe setpoints of the o

shown in l'igure 2-4. liach line from the steam Westinghome control sprems wili not be prosided generator secondary out to the common steam due to the proprietary nature of the various con-header was modeled individually, and included a trol sprem specificatiom. in general, the control main steam holation vahe (MSIV), a check sahe, systems were modded as dosely as poulble and are sar cty, and PORV sahes. Ihc flow restrictor was comidered to be good representatiom of the actual modeled in combination with Ihe flow noitle at the sptems, top of the steam dome. I' rom th? header to the tur-bine gmcrnor vahes, the north and south lines wcre lhe steam dump control sptem will be desetibed modeled as one line. 'I he steam dump sabe banks in Subsection 2.2.1, followed by descriptiom of the were modeled as one sabe, with appropriate con. steam generator Icsci control sprem in Subsce-trollogic to simulate the opening of each sahe in tion 2.2.2, the preuuriter preuure control system the banks, in Section 2.2.3, preuuriter lesel comrol sptem in Subsection 2.2.4, and identification of additional

'the major flow paths of the feedwater spicm sptems in Subsection 2.2.5.

were modded and are shown in ligure 2 4. 't he Icedwater sprem comhted of the condemale 2.2.1 Steam Dump Control 8ystem Ihe purpme spicm, main ferdwater splem, md the amillary of the steam dump control splem (SI)CS) h to:

feedwater splem, t he components induded in modeling the condemate intem were the A and ll 1. Permit the nudcar plant to accept sudden (ondetoate pumps, low preunte ferdwater heaters, toucs of load withou tupping the nator low preuute he tcr b)pau, heater dialn setem, and the inain ferdeater pump suction header. Ihe son' denwrt wcre modeled mmg a omstant temperature gg , g, g gg g g boundary cornhilon. Ihe (omponentiintluded in modding the main fredwatcr spicm were both m dbrh w kWwedom m m lilaIrl ferdwater purnpt, snant Iceda alcr purnp init a M uad e of W e m m u W culation, high preuute icedwater heatrri, main Icedwater header tank, main ferdwater/bypau s ab es, ami piping to the steam generatori, L l'ctmit control of the sfcam generator inthiding the feedwatcr header ting. Ihe amiliasy pnuun at no Mad coniMons and pgtmk fccdwater inicm enodeling in,luded the inotor a nianuah (onuolled woldown m the drisen and stram drhen optemi, with a wmmon Old"I-header for cAh sulcm and 5 ahes itom the head-r lo cAh feed hn[. IbfA N ldd %AN ak W M Ib'bfd bIlbkf WOdC%

of steam dump wntrol, itcquitements I and 2 are ifrat inusturn for the inomfary n orm induded met b sonttolof the prunary spiern merage fluid the internal and esternalinctalin.oilor cas h of the temperatuff, last, whettat requirement 3 h met steam generator scumila to amt ihc metal man of by 4ontf olkng the inondary sptem itc.on prcourc.

1. I hr 'dKN h disided into the / sepainte spirms as the piping for both the stram kne intem and the ferdwater intem, docribed hi the following s nontioni.

Iherv wetc l 4 golumco #cytocnling ihe sesond. isf7 d ettr.yn#/on ceatm N the load rejco

' af y for cash itcam venciator, t he sicam linc IIom sontfollcr (l 14(164 doignol to sontful the somitted of 16 splumn atal the feed *ater opfem primary optem merage temprinf urf duihtg petloth soritained 11 5 010111 0 . of load f ejhttoll, forif f oi of lbf prittiary s)stert) 11

average temperature h performed by modulating decreases below the minimum temperature setpalat; the steam dump vahes and,if the load rejection h howeser, unlike the LRC and PTC Sptems, the

>70r ,e the steam line power operated relief sabes minimum temperature condition may be oserrid-(PORVs). The turbine impuhe stage preuure signal den by the plant operator to enable plant cooldow n h linearly cometted into the primary sptem aserage to cold shutdown conditions, temperature setroint; the littered derisative of the turbine impuhe stage preuure signal is med to 2.2.2 Steam Generator Level Control determine whether or not a load rejection has System. The steam generator lesci control system a occurred, ar.d tbc site of the rejection when one (SGl.CS) is designed to regulate the liquid lesel in does occur. the steam generator (SG) dow ncomer.1 his control sptem mes tbree input signah to regulate the feed.

Moduladon of the steam dump sahes k bh>cked water flow rate into each of the three steam .

if the condemer does not base suflicient vacuum, generators. T hese three signah are: (a) the steam or if the primary sptem aserage temperature generator liquid lesel, Ib) the steam flow rate decreases below the minimum temperature setpoint. measured in the steam line at the 50 outlet, and 0:hcr control functiom eskt in the real plant but (c) the feedwater flow rate measured dowmtream were not modeled became they would not base been of the feedwater regulating sahe. The SGl.CS h used in the sariom calculatiom presented here, med only when the plant load h > l$re.

32 f 3 near rieg stum Dump control Ihe function t he steam generator liquid lesel h determined by of the plant trip controller f riC) h to bring the meMuring the dif ferential preuure betw een preuure primary sptem average temperature down to the laps in the SG downeomer, the liquid level b equihbrium em load setpoint, $$9 K ($47'l') for the inferted from thle dilferential preuure, and can be 2100 MW full power case, after the turbine he been perim bcd by esents that do not influence these Iwo tripped. ^lhe PIC performs this funstion by taps in the moc manner, for esample, a main steam modulating the steam dump sahet Unlike the une breal, or surbine stop sahe clmure. 'I he steam 1,RC, the PIC don not hoc any control mer the bencrator liquid lewt dynal h compared to the sel-stram kne rehrf whet point lesci, whkh h a funstion of the turbine impuhe stage preuure, and the resuking error h Modulation of the uram dump uhn h bhis ked then med n an input signal to a proportional.

if the sondemer don not her suf tkicnt usuum, integral IP4) (ontroller, or if the primary setem ecrage temperature destran twlow the minknum tempriature setpoint. The fctdwater and steam flow rate signah are Stram dump omitol intem operation udng the compared to dctermine the ferd eteam mhmats h PiC h replaced with the steam preuure controller dynal. I hh signal h added to the leici crror dgnal when the primary optem merage tempcialme h and the roull k med c ibc Input dgnal for another desreced to the no load wtpoint, wkh the addb P.I controllcr. T he output of thh P l controller h tional tonernint that 60 s nmst hoc nphnl Un(c med to modulate the ap,nopriate feedwater sahe.

the plant wn tripped. the to s delay h mni to simulate the trastor operator inporoe time, the SGI CS h not mal when the plant loci h leu than I?Co . Imteml, the main icedwater uhn JJ f Jsrum n. ewe c.mtres lhe stcam picoute ate shned anil feedwater control h performni sontroller (SpC) h mni lo regulate the snondary inanually ming tha lculanter b> pau sahn to main-spicm steam hemier picoute, Ihh spirm h mol ,ain the dotted sicam generator loci Additionall),

I whrn lhe plant h at no load sondillom, or 10 one main ferdwater pump and one omdensate

. Icplass the PIC t he steam headct preuure is son, pump are med, instead of two of cash, e in the .

! Irellnl by modchny the sicam dump sahrt the full power use, setpoint preuure h 7.01 Mi a ll010 pd4) No modulation of theitcam kne Pul4% h performal I he omdlliom that un roull in main Icot* ater by this splem. holation ht the plant het been imorporated into ,

the Sul ( % inodel, I bcw tondillom ilulude plant Modulation of the itcain dmnp who h l>los knl tilp, inain ferd*aler pump tilp, and Inlilation of if the tomicmcr don not hee sullhlent u.uum, the enginccted safcty featurrs mtuation signal or if the primary sotem ecrage tempiratute llTI A%h 12 l

t _ _ _ _ _ _

I

! 2.2.3 Pressuriser Pressure Control System. The The pressurizer setpoint level is specified as a ,

purpose of the pressurizer pressure control system function of the primary system average coolant (PPCS)is to maintain the desired primary system temperature. The setpoint level is subtracted from pressure. This function is performed using spray the actuallevel determined from a set of differen-valves, relief valves, proportional heaters, and tial pressure taps located in the pressurizer. The back up heaters. pressurizer level error signal is used as the input ,

signalin a P I controller. The output of the P l con- l

• troller specifies the amount of change in the charg-The pressurlier pressure is compared to the set.

point pressure to determine the pressurlier pressure ing pump speed to effect the desired change in the error. This error signal is used as the input signal primary system coolant inventory, to a P-I controller. The output of the P 1 signal is  !

e used to control the function of both spray valves. The level error signal is also used to actuate the the proportional and back up heater source back up heaters when the pressurizer level error demands, and the valve area of one of the two exceeds the setpoint level by $%. Pressuriier heater pressurlier PORVs. The other PORV area is a fune- demand is blocked when the pressurlier level j tion of the uncompensated pressurlier preuure becomes leu than the low level limit of 14%.

signal. ,

The Pl.CS is modeled to include the reactor The PpCS is modeled as accurately as poulble coolant pump seat injection contribution in addi- l and includes all the trips and setpointiin the actual tion to the charging flow demanded by the compen-plant, with two exceptions. l'irst, the spray valves sated pressurlier level error signal. All the pump i do not maintain a minimum now as in the plant seal injection How is added to only one loop, but becau e of difficulties incurred due to thermal- is quantitathcly correct. l hydraulic considerations. The minimum now requirement it imposed in the plant to maintain the 2.2.5 Additional Control Systems, included in the spray line temperature at the temperature of the control system package are mheellaneous con- .

primary system cold legs. Ihis h required Io eold trollers and tript that are modeled to reprnent  !

the ptmibillty of thermally streulns the spray lines various system functions that cannot be claulfled j when preuurlier spray is demanded. To compen- In any of the aforementioned systems. These con-sale, the model spray lines were initiallied at cold troller perform functiom such as: (a) feedwater l leg temperatures, and no heat loues from the lines reeltculation to the condemer during periods of low to containment were com6dered.1 hc other model, feedwater demand, (b) low preuure feedwater i ing esception is in the amount of power supplied heater bypan due to low main feedwater pump suc- i to the proportional heaters dming steady state lion preuure, (c) specirleation of turbine impuhe l operation. The heaters normally operate at stage prenure as a function of steam How rate and l 2(WWikW to make up for plant heat loues and turbine gmernor vahe arca, and(d) controlof the l preuure decay due to the minimum continuous anillary feedwater systemt. j spray operation. Since the preuurlier tank walk i

• *" ""* d "' 3* d"" > I"'"'d' #d h'd' '" " "'"' 2.3 Steady State Conditions i and the spray 5 ah n were completely hotated during j the itcady state initialliation phme, thh 2(No kW heater somce wai subtracted from the total ptml. 1hree steady itate Initialliation calculatlom were !

ble proportional heater source of 4(WM) kW. performed with the Rl; TAPS lillR.2 model.1he calculatiom were performed with the core power 2.2.4 Preeeurteet Level Content system, t he pur, at 2W,22(NI, and A.29 MW thot standby), I!ach l

• pme of the preuurlier loci control splem (pl CM of the subsequent f ramlent calculatiom med either h to maintain the dotted amount of Ilquld the 2hMI MW power steady state or the hot stand.  !

i (mentory in the primary coolant tptem, the by steady state, t he 22(M) MW power steady state amount of waler imentory in the primary coolant was used for a plant tilp tramient performed for  !

, tptem may be inferred from the liquid inel in ihe model ched out purpan. The following subsec. t preuurlier, whish taries as a function of the llom present the ticady state condillom for cash l primary splem merage coolant temperature, of the three power Inch. j l)  :

l  !

t 2.3.1 2300 MW Steady State. The initial condi- represent the behavior of the primary pressure con- l tions for the 2300 MW steady state were repre. trol and loop flow, and the chemical and volume  :

sentative of IIBR-2 operating conditions at the control system. RELAPS steam generator l proposed 100% rated core power. Table 2 2 com- secondary initial conditions were calculated from  ;

pares initial values of the selected parameters f rom the use of control systems that represented manual  !

the RELAPS model with desired initial conditions control of the feedwater bypass valve to maintain  ;

that were obtained from a revised precautions, levelin the generators and action of the steam dump limitations, setpoints document, and from other vahes in plant trip control mode. The table shows -

documents describing the conditions for proposed that the actual initial conditions were generally in  ;

2300 MW power operation for ilBR 2. The excellent agreement with the desired initial RELAPS initial conditions were obtained from a conditions.

steady state run that used control systems to ,

represent Ihe behastor of the primary pressure con. 2.3.3 Hot Standby Steady State. The initial com-trol and loop flow, the chemical and volume con- puter model conditions for hot standby steady state i trol system, the secondary liquid level, and main were representative of the llBR.2 operating condi- l feedwater and steam control valves. The table shows tions at hot standby. Table 2-4 compares initial values that the actual initial conditions were generally in of selected parameters from the RELAPS model with excellent agreement with the desired initial desired initial conditions obtained from documents conditions, describing hot standby conditions for the ilBR 2 i plant. The RELAPS initial conditions were obtained j 2.3.2 2200 MW Steady State. The initial condi. from a steady state run that used control systems to j tions for the 2200 MW steady state were repre- represent the behavior of the primary pressure con.

• sentative of IIHR 2 operating conditions at the time trol and loop flow, and the chemleal and volume con-of the plant trip transient. Table 2 3 compares trol system. REl.AP$ secondary initial conditions initial values of selected parameters from the were obtained from the use of control systems that RELAPS model with desired initial conditions reprewnted manual control of the feedwater bypass which were obtained from start.up data and other sal e to maintain lesel in the generators and action documents describing the operating conditions of of the steam dump valves in steam pressure control j the llHR 2 Plant at 2200 MW core power. The mode. The table shows that the actual lnitial condi.  ;

REl.APS initial conditions were obtained from a lions were generally in esectlent agreement with the  !

steady state run that used control systems to desired initial conditions.

Table 2 2. 2300 MW initial conditions j i

Parameter REl.APS Desired Core power, MW 2300 2300 i Pressurlier pressure, MPa (psla) l$.$(2250) 15.$(2250) llot les temperature, K ('ll $91.4 (NH ll) $91.4 (604.$)

Cold leg temperature, K (*l') $$8.9 ($46.3) $$8.9 ($46.3) i Pressurlier level, % $4.$ $ 3.3 -

Reactor coolant flow, kg/s (lbmh) 12726 (280$$.$) 12726 (2NO$6)

Reactor coolant pump speed, RPM 1247.1 1190 Net makeup flow, spm DAM)I O  ;

Steam pressure MPs (psla) 3.3(804) $.7 (M2M) .

Sicam generator level,8's $1.7 $2 = }

Steam flow (each), ksh (lhmh) 42$(937) 424.2 (9)$.2) i hicam generator mass (each), Kg (Ibm) 44233 (97$N)) 42302 (912N4 j Condenwr tempetiture, K Pl') 312 (102) 312 (102) [

1cedwater temocrature, K PI') $00.4 (441) $00.6 (441.$) .

a  ;

lleater drain Ifow, kg/s (Ibm /s) 333.3 (783.2) l'ecdwater recirculation, spm 0.0 a j

a. No dtia asalfable. 1 j

14 f

1 Table 2-3. 2200 MW initial cond;tions Parameter RELAPS Desired Core power, htW 2192 2192

, Pressurizer pressure, blPa (psia) 15.5 (2250) 15.62 (2265) llot leg temperature, K (*F) 585.6 (594.4) 583.2 (590)

Cold leg temperature, K (*F) 555.9 (541.0) 552.6 ($35.0)

Pressurlier level, % 43 40 o Reactor coolant now, kg/s (Ibm /s) 13275.3 (29267) 12458.3 (27466)

Reactor coolant pump speed, RPhi 1244 119040 Net makeup Dow, gpm 34.8 a Steam pressure, hlPa (psla) 5.16 (748) 5.20 (754.7)

Steam generator level, % %52 s52 Steam now (each), kg/s (lbm/s) 401.7 (885.7) 426.3 (939.8)

Steam generator mass (each), Kg (ibm) 43650 (96232) 42302 (93260)

Condenser temperature, K ('F) 312 (102) 312 (102)

Feedwater temperature, K ('F) 498 (437) 498 (437) lleater drain now, kg/s (ibm /s) a 353 (783)

Feedwater recirculation, spm 0.0 a

n. No data available.

l Table 24. Hot standby Initial conditions l'arameter ,

RI!!.APS I)csited Core power, htW H.29 N.29 l'reuurlier preuure, hil'a (psla) 15.5 (2250) 15.5(2250) 110 leg temperature, K ('l) 560.3 (34N.9) 559 (347)

Cold leg temperature, K ('I') 560A4 (54N.4) 559 (347) l l'rcuurlier itsel, % 23.6 24.4  ;

Reactor coolant How, kg/s (Ibm /s) 12629 (27841.9) 12626.8(27817) 1 Reactor coolant pump speed, RI'Al 1226 livo Net makeup How, ppm 0 0 Sicain preuure, hil'a (pala) 7,0) (1020) 7.03 (1020)

Steam generator lescl. % JM.4 l9 Strain flow (ca6h), kg/s (lbfn/4) 2.27 ($ 0) =a o Nitam generalor mats (cash), Kg (lbm) 577MM (127400) $44)2 616N9 (1200tM) 1)MNN))

l

('ondenser lettiperature K ('l') 299.N (MO) A l'cedwater lettiperature, K ('l ) 299.N (MO) A l

lltater dralti flow, kg/e (lbm/s) 0.0 A l' red *nf tt retirculation, ppm 1300 A

m. No d.tta avalfabic.

- _ _ _ _ _ _ . _ ~ , _ _ _ _ _ _ = _ _ _ _ - _ _ _

l$

3. SIMULATION OF HBR PLANT TRIP FROM 2200 MW The results of the plant trip calculation are 3.2 Comparison of Results presented in this section. This calculation was per-formed to assess the ability of the model to simulate the responses of the corresponding plant thermal- The results of the plant trip calculation are

• hydraulic and centrol systems to a trip from the full presented in this section. Similarities and discrepan-power state. Although the plant trip simulation did cies between the calculated results and the measured not exercise the thermal-hydraulic models as fully plant data are addressed for each of the major plant as did the PTS sequences, the control system models control systems described in Section 2.2. .

were exercised fully.

In comparing measured plant with code-cal- [

3 A description of the plant trip test that was culated data, the reader should be aware that the I simulated by this calculation is presented herein. former is generally not in a form that allows for The subsections present the results of the calcula. easy comparison. Data was taken from strip charts tion as compared to the plant data, along with the that covered a 2 h period during the test. As such, conclusions regarding the performance of the con. the width of the lines representing the measured trol system models, data were 40 s. Therefore the occurrences of events in time were subject to some interpretation.

~ 3.1. Description of Plant Trip Test 3.2.1 steam cump controi system Response.

The functioning of the SDCS may be inferred from A description of the plant trip calculation is the response of the primary system highest average presented in this section. Information concerning temperature, T-ave. The calculated and measured the plant trip test sequence of events was obtained responses of this parameter are shown in Figure 3-1.

from the HBR-2 plant startup data. Additional The difference in steady state initial temperatures details may be obtained from Section 43 of the is only 2.5 K (4.5*F), which is insignificant due to Plant Startup Test Report (WCAP-7844). the nature of the temperature response during the test.

The plant trip test was initiated from the 2200 MW full power steady state. The proposed The response of the calculated T-ave is in close 100% power level is 2300 MW and will be used after agreement with the measured T-ave through the the plant is upgraded with new steam generators, initial period of temperature decrease following the however 2200 MW was the power level at which the plant trip. The rate of increase of the calculated plant was operating at the time of the test. Approx- temperature was somewhat faster than the imate!y one hour of plant data was reported in the measured increase. This discrepancy in primary startup test report. The first 900 s of this data was system heatup rates is due to the difference in the considered the most significant since this was the total feedwater flow rates.

period when the various control systems were required to operate in a transient recovery mode. Steam generator levels are compared in Fig-ure 3-2. The measured steam generator levels The plant trip test was performed to demonstrate decreased to 0% at the start of the test, while the the ability of the plant control systems to bring the calculated levels decreased to 20%. Consequently, plant to equilibrium no-load conditions following steam-driven auxiliary feedwater was initiated in the -

a plant trip from full power. test but not in the calculation (a requirement for initializing steam-driven auxiliary feedwater is

1. The test was initiated by tripping the turbine. All 2/3 steam generators with 15% NR level). The control systems operated as designed, with the par- additional feedwater in the test resulted in higher ,

tial exception of the main feedwater control valves, primary to secondary heat transfer rates than those which failed to immediately close when the turbine calculated. These higher rates resulted in the slower was tripped. The main feedwater system was primary system temperature responses indicated by isolated 50 s after the start of the test. the data.

16 x

580 , , , ,

~

WEASURED -580

---

CALCULATED

- m S70 - -

m M F u

~

-) -560 v I u 3  % o o ~0 560 - - +

g

' I L

• e

c. -

1 -540 o, E I E U 550 -

/~~'~~~~~' ~-- -

\ s' -520 g/

540

-250 0 250 500 750 1000 Time (s)

Figure 3-1. Plant trip test measured and calculated primary system highest average temperature responses.

0.6 , , , ,

WEASURED CALCULATED

- -% s 7 's

~

/  %

! / 's N e 0.4 -

/ __

/

i /

D I to e

N ll

= if E I,I

' O.2 - I -

o Z

0.0

-250 0 250 500 750 1000 Time (s) o Figure 3-2. Plant trip test measured and calculated steam generator level responses.

17

The discrepancy between the measured and uncertainties in the plant data relative to the calculated steam generator NR levels will be modulation of the main feedwater and steam dump addressed in the discussion of the SGLCS. valves. The effect of modeling the separators and dryers at a single elevation (see Subsection 2.1.2)

It is important to note that while the differences was found to be only a minor perturbation in in primary system temperature responses between indicated level due to flow effects at the upper measured data and calculated results appear to be reference pressure tap. The continued increase in ,,

significant, the maximum difference was only 2.6 K calculated level following termination of AFW was (4.7'F), which is probably within the plant's found to be caused by a redistribution of liquid temperature measurement uncertainty. The general from the boiler region to the downcomer. Further trend of the calculated response is the same as the investigations are being conducted to understand measured response indicating that the SDCS model these discrepancies.

adequately simulated the response of the actual system. The steam generator total feedwater flow rate response is shown in Figure 3-3. The main feed-3.2.2 Steam Generator Level Control System water control valves were reported to have not Response. The functioning of the SGLCS may be closed during the first 50 s of the test due to a delay determined using the response of the steam in the feedwater controller response. No other generator narrow range (NR) levels, and the feed- information was available regarding the actual water and steam flow rates. The three steam response; therefore, the modeled valve areas were generator calculated responses were essentially held constant for the first 50 s. After the first 50 s identical; therefore, the discussion of SGLCS the calculated and measured feedwater flow rate response will be centered on the responses of steam responses were approximately the same, except for generator B (SGB). The measured and calculated the period from 140 to 660 s. During this period the SGB NR level responses are shown in Figure 3-2. modeled auxiliary feedwater system ceased flow due The trends were qualitatively the same with two to the calculated recovery of steam generator level important exceptions. (the model regulated steam generator levels between 39 and 41%). Feedwater flow was resumed at 660 s The first discrepancy between the calculated and when the NR level decreased below 39%. The measured level responses was in the size of the measured feedwater flow rate did not decrease to downward spike at the start of the test. The zero because some feedwater was required to main-measured data indicated that the NR level tain SG level at the setpoint. This feedwater demand temporarily decreased to 0% before recovering; was required due to continuous steam flow out of whereas, the calculated level only decreased to 20%. the SG during this period.

This discrepancy is significant because the SG NR level determines the demand from the auxiliary The measured and calculated steam flow rate feedwater systems (and the main feedwater as well). responses are shown in Figure 3-4. The calculated These differences appear to be associated with the steam flow rate was less than the measured flow rate uncertainty in behavior of the feedwater control throughout the calculation. The higher measured valve. Sensitivity calculations indicate that modula- flow rate was due to steam-driven auxiliary feed-tion of the valves will reproduce the response seen water operating in the test but not in the calcula-in the plant data. tion. The extra feedwater caused higher primary to secondary heat transfer, which in turn resulted in The second discrepancy was in the response of greater steam generation.

the level taps after the feedwater to the steam generators had been terminated in the calculation, Overall, the largest discrepancy in SGLCS -

which happened when the NR level reached 39%. response was in the calculation of the NR levels.

The measured data indicated that once the level This discrepancy can be important when the steam-reached 39% no further increase in level occurred; driven auxiliary feedwater system is required in whereas, the calculated level continued to increase addition to the motor-driven system due to low SG .

for an additional :.00 s. NR level. With the exception of the auxiliary feed-water discrepancy, the calculated feedwater and A complete explanation for these two discrepan- steam flow rate responses are in reasonable agree-cies has not been determined. Important factors are ment with the measured plant data. Since these three 18

500 , , , ,

. WEASURED -1000

'---- CALCULATED 400 -

^

I "

- 7 -

I,.,

-800 N

N m E 6 300 - -

v 0

-600 3

< o 3 C 0 200 - - C n -

-400 m n 0 m 2 0 2

10 0 - -

200 0 . O

-250 0 250 500 750 1000 Tirle (s)

Figure 3-3. Plant trip test measured and calculated feedwater flow rate responses.

500 , , , ,

MEASURED - 1000

---

CALCULATED 400 -

- 1 -

- 800 q l

300 - -

m i E x -

- %0 a v .

v 3 200 - -

1

.O. - 400

~ l .$_

M g 100 -

I

-- 200 $

f 2 O h 2 0-- l-------------------- --0

' ' ' ' - -200

-10 0 -

l -250 0 250 500 750 1000 I

Time (s)

Figure 34. Plant trip test measured and calculated steam flow rate responses.

l 19 I

m-. -,,,m.m-,. a,-, ,r -

m- n -

,,._.,-p---,.-

,. gm. aw 3 --- ,w- -m-

parameters (NR level, feedwater and steam mass Overall, the PPCS model performed adequately, flow rates) are the input signals for the SGLCS it relative to the actual plant system.

was concluded that the SGLCS is adequately modeled with the exception of the minor discrepan- 3.2.4 Pressurizer Level Control System cies in upper reference tap transient response char- Response. The measured and calculated responses acteristics. The reference tap problem is only of the PLCS are inferred from the pressurizer nor-important in those situations similar to the plant malized level responses shown in Figure 3-6. The ,,

trip test. In situations where the NR level decreases measured pressurizer setpoint level is also included below 15% for a prolonged period of time the to provide a characterization of the desired modeled level tap response behaves in the desired response, manner; hence, there are few cases where the modeled SGLCS would not respond like the actual The measured and calculated level responses were '

system. This response was evaluated for each of the nearly identical with the exception of the minimum 14 PTS calculations presented in this report and level achieved at N100 s. This discrepancy was due discussed in the appropriate analysis section (see to a lag element included in the level tap model to analyses for Scenarios 4, 6, 7, 8,12, and 14). simulate instrument response time. The overall response of the calculated data indicates that the 3.2.3 Pressurizer Pressure Control System PPCS is adequately modeled.

Response. The PPCS measured and calculated responses may be inferred from the information 3.3 Conclusions presented in Figure 3-5, which illustrates the calculated and measured pressurizer pressure A comparison of the calculated and measured responses, data indicated generally good agreement, thus providing a limited, but useful, qualification of the The agreement between the measured and cal- plant model beyond the detailed quality assurance culated data is very good through the first 400 s. measures described in Section 2.

The measure'd data indicate the pressure increased to sl5.65 MPa (2270 psia), whereas the calculated The modeled control systems appear to be suffi-pressure did not increase beyond 15.41 MPa ciently capable of simulating the plant control system (2235 psia). This discrepancy was due to unresolved responses that were expected to occur in the various anomalies in the actual plant response (the setpoint PTS transients. The modeled thermal-hydraulic pressure was 15.51 MPa, 2250 psia) and the systems adequately simulated the plant trip test apparent inability of the model to maintain the set- however the PTS transient responses were generally point pressure at the lower pressurizer setpoint level. of larger magnitude than the plant trip test.

s e

20

14 i , , , ,

WEASURED -2300

^

---

CALCULATED 15.5 -- 1 --2250 1 /~ ~ ~ ~ ' ~ ~ _ _._ -

[ -

-2200

' b 15 -

t j

' - e W l / 3 l f -2150 g

" l / e d

a l / 5-A

/

a. 14.5 .- 1 f --210 0 1 /

) /

\ /

s

-2050 14

-250 0 250 500 750- 1000 Time (s)

Figure 3-5. Plant trip test measured and calculated pressurizer pressure responses.

0.5 , , , ,

O MEASURED SETPOINT LEVEL O MEASURED LEVEL e A CALCULATED LEVEL

- 0.4 -

%3 -

e c

"O I N 0.3 - -

= , .- a. -

u I-O -

Z 0.2 -

a 0.1

-250 0 250 500 750 1000 Time (s)

Figure 3-6. Plant trip test measured and calculated pressurizer level responses.

21

4. SCENARIO 1,1.0-FT2 STEAM LINE BREAK AT HOT STANDBY The following section details the analysis of main and auxiliary feedwater flow to the steam Scenario 1, a 0.093 m 2 (1.0 ft 2) break in Steam generators had ceased. The deletions decreased the Line A, downstream of the flow restrictor and computer memory requirements for the problem upstream of the main steam isolation valve (MSIV). and thereby decreased run time. '

The subsections contain a description of the scenario, model changes effected to perform the calculation, analysis of the results, extrapolation of 4.3 Results the key PTS parameters, and the conclusions. ,

The following section contains the results of the Scenarios investigated in this report generally Scenario I calculation. The first subsection include conservative assumptions concerning equip- discusses the results of the calculation. The second ment failures, operator actions, or combinations of subsection discusses the uncertainties involved in the these. Conclusions relative to pressurized thermal calculation.

shock severity are not to be drawn directly from the results preser.ted in this report (see Section 18). 4.3.1 Calculation Results. The sequence of events that occurred during the Scenario I calculation is shown in Table 4-2. The transient was initiated by 4.1 Scenario Description theinsertion of a 0.093 m2(1.0 ft ) break j, unction to atmosphere in Steam Line A (Component 550).

The description of Scenario I, as provided by Approximately 100 ms after break initiation a SIAS Oak Ridge National Laboratory, is shown in signal was generated by a high AP between the Table 4-1. system header and Steam Line A. The SIAS signal shut down the operating main feedwater pump and The transient was initiated by the occurrence of activated motor driven auxiliary feedwater (AFW).

a 0.093 m 2 (1.0 ft2) break in Steam Line A. All Turbine driven AFW initiation requires two-out-of-automatic plant functions were assumed to respond three low steam generator level indications and thus normally. Three operator actions were assumed as was not initiated because only one steam generator well. The first was to trip the reactor coolant pumps is affected in this scenario. High pressure injection (RCPs) if the primary system pressure fell below (HPI) flow was initiated at 68.5 s when the primary 9.07 MPa (1315 psia), provided a SIAS signal had system pressure dropped below the pump's shutoff been generated. The second operator action was to head of 10.13 MPa (1470 psia). At 72.2 s, when the stop auxiliary feedwater to the unaffected steam pressure dropped below 9.07 MPa (1315 psia), the generators when liquid carryover to the steam lines RCPs were tripped. At 600 s auxiliary feedwater is observed. The third operator action was to stop flow was terminated as prescribed in the scenario auxiliary feedwater flow to the affected steam description. By about 1300 s heat transfer to the generator (ASG) 600 s after the transient initiation. ASG had degraded to the point where core decay power exceeded the ASG's heat removal capability and the primary coolant temperatures started to 4.2 Model Changes increase. At 4889 s, primary pressure had mereased to the power operated relief valve (PORV) setpoint The basic RELAPS model used to perform the and the valve began cycling.

Scenario I calculation is described in Section 2.

• Figure 4-1 presents the primary and secondary The break in Steam Line A was simulated by the pressure responses. Pressure in the ASG decreased insertion of a break flow path at the downstream continuously reaching near atmospheric conditions end of Component $50. After 600 s of the transient (0.14 MPa,20 psia) by 900 s. Both unaffected ,

calculation, all components upstream of the steam steam generators (USGs) experienced a slight generators, including auxiliary feedwater headers decrease in pressure early in the transient when the and valves, and the main feedwater train, were primary cooldown caused them to become primary deleted. These deletions were performed after all system heat sources. Once the RCPs were tripped, 22 1

Table 4-1. Scenario description No.1 Plant Initial State - Just prior to transient initiator General

Description:

Hot 0% Power,0% Power after 100h of shutdown System Status a- Turbine: Not latched, TSVs closed Secondary PORY: Automatic control

. Steam Dump Valves: Automatic control Charging System: Automatic control Pressurizer: Automatic control

• Engineering Safety Features: Automatic control Power Operated Relief Valves (PORVs): Automatic control Reactor control: Manual Main Feedwater: In bypass mode, manual control to provide 39% level in S/Gs; I condensate pump,1 MFWP operating.

Aux Feedwater: Automatic control Main Steam Isolation Valves (MSIVs): Open, Automatic control Main Feedwater Isolation Valves (MFIVs): Closed, Automatic control Transient Initiator A 1.0-ft2 hole appears in steam line A outside containment upstream of the MSIV and downstream of the flow restrictor.

Equipment Failures That Occur During the Transient if the Equipment is Demanded.

None Operator Reactions to Reported Information

1. If safety injection actuation signal (SIAS). is generated, the operator will trip the reactor coolant pumps when RCS pressure reaches 1300 psig.

2. Stop AFW flow to the unaffected S/G when liquid carryover is observed in the main steamline.

3. Stop AFW flow to the affected S/G at 10 min or when carryover occurs, the USGs were effectively isolated due to primary Figure 4-3a shows break flow over the first 60 s loop stagnation and their pressures stabilized. of the transient. Break flow, choked at the break junction, initially peaked, then fell as the ASG In the primary system, the pressure decreased as depressurized. Figure 4-3b compares break flow the break induced heat transfer to the ASG peaked. with auxiliary feedwater flow. Motor driven As the ASG emptied and the tubes became sur- auxiliary feedwater (AFW) flow was initiated with rounded by high quality fluid, heat transfer to the the SI AS signal at 0.15 s and directed by the header o ASG degraded and, with the RCPs tripped, primary to the ASG. Since the pressures of the ASG and pressure stabilized around 6.55 MPa(950 psia). At AFW header were significantly lower than those of 587 s the pressurizer began refilling, as shown in the USG, all AFW flow was delivered to the ASG.

Figure 4-2, and the primary pressure began to By 390 s auxiliary feed flow exceeded break flow increase. Repressurization continued until, at and the ASG began to refill, as shown in Figure 4-4.

4889 s, the power operated relief valve (PORV) set- This refill terminated at 600 s when auxiliary feed-point of 16.2 MPa (2350 psia) was reached. water ceased as defined in the scenario description.

Primary system pressure continued to oscillate around the setpoint for the remainder of the Figure 4-Sa presents the mass flow rates in the transient calculation. cold legs of all three primary loops. Prior to the 23 )

. . . . -- - .=. - . - - - . . . - - . - ~ _ . - - - - _ . . -

I Table 4-2. Soonerlo 1 sequence of events Figures 4-6 and 4-7 show the temperature response in the hot leg, steam generator inlet, steam  !

generator outlet, and cold leg in Loops B and C, Time respectively. Both figures show similar responses.

, (s) Event - There is an initial decrease in all temperatures in '

response to the cooldown of Loop A. The steam O Steam line ruptures - generator outlet temperature became higher than ,

the inlet temperature, signalling that the USGs were 0.1 High Steam Line A AP supplying heat to the primary. The stagnation in SIAS signal loop flow v;as reflected in the divergence between

' Main feedwater pump tripped off steam genuator outlet and cold leg temperatures.

i

• Motor auxiliary feedwater tripped on The cold les temperatures reflected the insurge of cold HPI fluid into the stagnant cold les piping.

20.3 Pressurizer low level alarm Differences between the Loop B and C temperature reponses are caused by pressurizer effects (Loop C 4 43 Pressurizer indicated empty only) and makeup effects (Loop B only).

[ 68.5 High pressure injection (HPI) Figure 4-8 shows the hot leg, steam generator initiated inlet, steam generator outlet, and cold leg temper-atures in Loop A. Both the cold leg and the steam 72.2 Reactor coolant pumps tripped generator outlet temperatures were always lower than the hot leg / steam generator inlet temperatures, 4

587 Pressurizer level indication returned reflecting the fact that the ASG acted as a heat sink throughout the transient. The close coupling of the 600 Auxiliary feedwater tripped off temperature pairs was an indication of the natural circulation flow through the loop. Afler 1000 s, 1079 HPI shut off with the ASG empty, the effects of the heat transfer degradation caused hot and cold leg temperatures 1300 Downcomer temperature staned to to converse.

. increase Figure 4-9 presents a comparison between all 4889 Power operated relief valves began three cold leg temperatures and the downcomer gyg ;,,

temperature. The downcomer temperature was between the USG and the ASG loop temperatures

! 5031 - Pressurizer indicated full during the time the RCPs were operational and c asting down. Once the USG loops stagnated, the l 7200 End of calculation downcomer temperature conversed on Loop A's

cold les temperature and they remained closely coupled throughout the transient.

RCP trip, the mass flow in all loops increased due

' to the increasing coolant density as the fluid cooled.

Flow in Loop A was higher than in the two USG Figure 410 presents the heat transfer coefficient loops because the heat transfer into the ASG caused for the inner wall surface of the reactor vessel the Loop A fluid to be cooler and, thus, more downconwr. The correlation used for the subcooled ,

dense. At 72.2 s the primary system pressure single-phase heat transfer occurring on the wall sur- i dropped to 9.07 MPa (1315 psia) and the RCPs face is the Dittus-Boelter correlation, which is =

were tripped. Flow in all three loops coasted down dependent on flow and fluid conditions.

and established an asymmetric natural circulation condition. Figure 4-5b shows this flow condition in The transient calculation was terminated at more detail. Loop A established a substantial flow, 7200 s. At this time, the ASG was completely

• l driven by the continued heat transfer to the ASG. empty, the primary system pressure was cycling ,

This flow decreased as the heat transfer to the around the PORV setpoint of 16.2 MPa (2350 psia) emptying ASG degraded. Loops B and C stagnated and the reactor vessel downcomer temperature was t and this allowed the ASG loop to dominate the rising, having reached a minimum of 3% K (253*F) ,

thermal downcomer response during the transient.

at 1071 s.

0 24  ;

CAUTION: THE SCENARIOS SIMULATED

. CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, E0VIPMENT FAILURES, OR BOTH.

20 , i i 0 PRIMARY

, . O SG A -2500 a so a T 15 J -

k -

-2000 $

. =- .

s

• io - _-1500  :*

s PORV setpoint reached o- c.

2

$ r=: : : : : : : : :  :  :  :  : -1000 3

[

g 5 - -

g

-500 0

~

0 0 'O 2 0 0 C t 2 0 0' C 0 0 0 2000 4000 6000 8000 Time (s)

Figure 4-1. Scenario I Primary and secondary pressures versus time.

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. ,

1.5 , , i c 1

.2 00 t

a 0.5 - -

e-HPI off 0

O 2000 4000 6000 8000 Time (s)

Figure 4 2. Scenario I Pressurizer narrow range indicated level versus time.

25

r CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

1500 , ,

- 3000 1000 - -

g -

- 2000 N

E N b v

$ 500_ -_ 1000 *

. Separator filled j i =

$ 8 O-- --0

-500 ~ ' r - -1000 0 20 40 60 Time (s)

Figure 4-3A. Scenario 1 Break mass slow versus time for first 60 s of transient.

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EDUlPMENT FAILURES, OR BOTH.

200 , , ,

_ a BREAK FLOW -400 0 AUXILI ARY FEEDWATER 7

D

-300 k E

.x A v

$ 100 - -

-200 j 2

i )

-100 g yn\ t, -,,- - - t u- .., u - Lu- a g 0 2000 4000 6000 8000 Time (s)

Figure 4 3B. Scenario I Comparison of break mass flow and auxiliary feedwater flow versus time.

26 i

(.

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

60000 _ ____ , , ,, ,,, , t , , ,, , ,

- 1.2 0*10' o a SG A

- 1.0 0*10'

. 40000 -

(( -

m 9 - E a d$, - 8.00*10*v d e

0 -

E 3 - 6.00*10' $

20000

- -- 4.00*10,

- 2.0 0*10' r 2 b2 r : c h c "- C' O 2 0 0.00 0 2000 4000 6000 8000 Time (s)

Figure 4-4. Scenario I Comparison of steam generator masses versus time.

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

5000 , , ,

-j +---RCP trip a LOOP A (W/ASG) - 10000 ll 0 LOOP B 4000 -

A LOOP C (W/PZR) -

- 8000 m 7

( 3000 - N y -

- 6000 ,$

c j 2000_ -- 4000 $

N

• g 1000 c -- 2000 8 o l 2

0--

-H HEHHHE E E 3 E 8 --0

_jogo - . i i - -2000 0 2000 4000 6000 8000 Time (s)

Figure 4-5A. Scenario I Comparison of primary loop cold leg mass flows versus time.

27

CAUil0N: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES. OR BOTH.

400 i ; , , ,

~

u LOOP A (W/ASG) - 800 0 LOOP B a LOOP C (W/PZR)

- 600 =

n 7

{o200 - -

- 400 N

E

d. :9 v .

m e c-o-c W - 200 g 2

0-- I

'l b: : : : : 0 -

2 - re- --0 E j

- -200

~

-200 ' ' , - - 00 0 2000 4000 6000 8000 Time (s)

Figure 4-58. Scenario I Comparison of primary loop cold les mass flows versus time (reduced scale).

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

600 , , ,

-600 7 i E

• 550 -

• g c c c c_ c- c :  :  :  : e o 500

• g f500 I SS5$S-  ! O E 2 -

}

[ a HOT LEG

• 400 E o SG INLET I 3 450 - A SG OUTLET _ j x COLD LEC 3 .

300 3 8- 8-

= 400 -

_x

=

g -

-200 g 350 -

~

300 O 2000 4000 6000 8000 Time (s)

Figure 4-6. Scenario I Fluid temperatures versus time in primary Loop B. ,

28

l CAUT ION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

600 , , ,

-600 E

v i - $

v e550 -

a 5 -

er _c c g--e c-c c- -500 $

E500 E

- a : ;  ; a . -4 -

k

• E

-400

• v v 5 450 - -

g 5 . . , ' W $

ex -300 ,

E E s400 -

0 HOT LEG -

s o o SG INLET o

> A SG OUTLET >

~

x COLD LEG -200

' ' ~

350 0 2000 4000 6000 8000 Time (s)

Figure 4-7. Scenario 1 fluid temperatures versus time in primary loop C (W/PZR).

Figure 4-7. Scenario I Fluid temperatures versus time in primary Loop C (W/PZR),

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

600 , , ,

~

O HOT LEG 7

• I o SG INLET p I

A SG OUTLET v e 550 -

x COLD LEG

~

e 3 -

-500 3 0 0

' ~

L E500 -

E # I 2 -

W -400 *

$450 -

,p' -

}

E - ^

E

-300 ,

e b400 o

o

-200 350

, 0 2000 4000 6000 8000 l Time (s)

Figure 4-8. Scenario 1 fluid temperatures versus time in primary loop A (W/ASG).

Figure 4-8. Seenario l Fluid temperatures sersus time in primary Loop A (W/ASG).

29 L

CAUTION: THE SCENARIOS SlWULATED CONTAIN SIGNIFICANT CONSERVATISWS IN OPERATOR ACTIONS, EQUIPMENT FAILURES. OR BOTH.

600 , , ,

-600 0 LOOP B 7

" 550 i

o a

LOOP C LOOP A -

7 V

E x DOWNCOWER 500 2 E 2

' - o

&.500 -400 g g ,

2 450 - -

2 3 -._c c c c _- 0 300 3 I

E S

= 400 -

-a -a e

l

- 350

-200

~

300 O 2000 4000 6000 8000 Time (s)

Figure 4 9. Scenario I Comparison of primary loop cold les temperatures and downcomer temperature versus time.  !

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

40 , , ,

7i "E -' -6000 Y E 3 30 r ip

-4000  !"

O 20 -

sl

$ Sk w >

' E *3

. m

-2000 .

= :I i

I 0

L ' __

O l

l 0 2000 4000 6000 8000 Time (s)

! Figure 410. Scenario I Extrapolated downcomer inner wall surface heat transfer coefficient versus time. .-

l i

30

4.3.2 Extrapolations and Uncertainties. Since operation. Therefore the uncertainty in reactor the calculation was carried out to 2 h, extrapola- vessel downcomer pressure and temperature tion of results to that time was not necessary. responses due to separator op.tation is considered to be small.

The principal uncertainty is a consequence of approximating the reactor vessel as a one-dimensional system. The model by definition passes 4'4 CONCLUSIONS equal temperature fluids to the hot legs. This is a limitation where significant loop asymmetric The calculation and extrapolation of Scenario I behavior is to be expected and where sufficient Huid led to the following conclusions:

mixing does not occur within the reactor vessel. For o the sequence being investigated here there is signifi- The peak primary pressure was calculated to cant asymmetry between affected Loop A and be the PORV setpoint 16.2 MPa unaffected Loops B and C. It is not known, (2350 psia), which was reached at 4889 s however, to what extent the cold leg Guids are mixed and continued to the end of the 2 h as they pass through the reactor vessel. Therefore transient it is not possible to quantify the effect of this uncertainty on the parameters of interest. The minimum downcomer temperature Qualitatively, little effect on primary system reached was 396 K (253*F) at 1071 s into pressure is expected. The unaffected hot legs would the transient likely be warmer and the affected hot leg cooler than calculated. As a result the unaffected loops would The minimum cold leg temperatures reached flow somewhat faster and the affected loop slower were than calculated.

Loop A 396 K (253*F) at 778 s There is an uncertainty concerning operation of Loop B 327 K (129'F) at i141 s the separator under the extreme conditions Loop C 392 K (246'F) at 1047 s presented by a large steam line break. The RELAPS separator simulates a perfeet separator at void frac- The scenario developed into a continuing primary tions > 0.5 and a partially effective separator at system cooldown until the ASG was emptied; then void fractions below 0.5. It is unknown how close- the decay heat source in the core overwhelmed the ly this modeling represents actual separator heat removal capacity of the system, which had behavior during the transient conditions. Previous developed an asymmetric natural circulation Dow analysis of main steam line break events has, condition. This heat imbalance turned the however, indicated that calculated primary system temperatures around and a primary system heatup behavior is not significantly sensitive to separator commenced.

O O

31

l

5. SCENARIO 2, DOUBLE-ENDED STEAM LINE BREAK AT HOT STANDBY The following section details the analysis or 5.3 Results Scenario 2, a double ended guillotine break in Steam Line A, downstream of the flow restrictor and upstream of the htSIV. The subsections con. The following section details the results of the tain a description of the scenario, model changes Scenario 2 calculation. The first subsection effected to perform the calculation, analysis of the discusses the results of the calculation. The second results, extrapolation of the key PTS parameters, subsection discusses the extrapolation to 7200 s of and the conclusions drawn, the key PTS parameters and the uncertainties =

involved in the calculation.

Scenarios investigated in this report generally include conservative assumptions concerning equip' 5.3.1. Calculation Results. The sequence of ment failures, operator actions, or combinations of events that occurred during Scenario 2 is shown in these. Conclusions relative to pressurized thermal Table 5 2. The transient was initiated by the inser-shock severity are not to be drawn directly from the tion of a 0.287 m2 (3,094 rg2) junction to results presented in this report (see Section 18). atmosphere in Steam Line A (component $50).

Approximately 50 ms after break initiation, an 5.1 Scenario Description SIAS signal was generated by high AP in Steam Line A. The SIAS signal shut down the one operating main feedwater pump and activated the The description of Scenario 2, as provided by motor driven auxiliary feedwater system. llPI flow Oak Ridge National Laboratory, is shown in into all 3 loops began at 50.5 s when primary Table 51. pressure dropped below 10.13 h!Pa (1470 psia).

IIPI flow continued until 1464 s when the primary The transient was initiated by the occurrence of system had repressurized above the pump's shutoff a full double ended break in Steam Line A with the head. Reactor coolant pumps (RCPs) were tripped reactor at hot standby conditions. All automatic at 53 s when the primary system pressure dropped plant functions are assumed to respond normally, below 9.07 h1Pa (1315 psla).

Operators were assumed to trip the RCPs when the primary pressure fell below 9.07 h1Pa (1315 psia),

provided a SIAS signal had been generated Figure 51 presents the primary and secondary previously, in addition, operators were assumed to system pressure responses during the transient, fail to isolate auxiliary feedwater to the ASG. Pressure in the ASG fell to near atmosphene pressure (0.14 h!Pa,20 psia) by 140 s. The two unaffected steam generators (USGs) began to act 5.2 Model Changes as heat sources initially as the primary system cooled down and the secondary pressures dropped. Once The basic RELAP5 model used to perform the the RCPs tripped at 53 s pressure in the USGs Scenario 2 calculation is described in Section 2. The stabilized at 6.89 h1Pa (1000 psia) as the USGs were transient was initiated from the hot standby steady efrectively isolated from the primary system by state presented in Subsection 2.3.3. stagnant primary loop flows.

The break in Steam Line A was simulated by the in the primary system, pressure decreased initially -

insertion of a break flow path in the downstream as the heat transfer to the ASG decreased average end of Component 550, and the deletion of steam primary system temperature. The depressurization line Components $55, 560, and 565. Trips were lasted until the ASG had empt!cd sufficiently to altered to insure continuous auxiliary feedwater to degrade its heat transfer and t'ie RCPs had been ,

Steam Generator A throughout the transient after tripped. Primary system pres.ure then began to the time it is automatically tripped on. increase as llPI flow refilled the primary system.

32

i

! l i Table 51. Scenario description No. 2 [

l~

Plant initial State . Just prior to transient initiator

!~ General

Description:

Hot 0% Power,0% Power after 100-h of shutdown

! System Status l Turbine: Not latched, TSVs closed i o Secondary PORY: Automatic control l Steam Dump Valves: Automatic control Charging System: Automatic control

• Pressurizer: Automatic control Engineering Safety Features: Automatic control o PORVs: Automatic control Reactor Control: Manual Main Feedwater: In bypass mode, manual control to provide 39% level in S/Gs; I condensate pump,1 MFWP operatius.

i Aux Feedwater: Automatic control

! MSIVs: Open, Automatic control MFIVs: Closed, Automatic control

'i ransient Initiator Full double-ended guillotine pipe break in Line A upstream of the MSIV and downstream of the flow restrictor.

Fquipment Failures That Occur During the Transient if the Equipment is Demanded.

The operator fails to isolate AFW to the affected S/O.

! l Operator Reactions to Reported Information:

l. If SIAS signal is generated, the operator will trip the reactor coolant pumps when RCS pressure reaches ,

j 1300 psig.

2. Stop Alv flow to the unaffected S/G when liquid carryover is observed in the main steam line.

Table 52. Scenario 2 sequence of events i

Time (s) Event ,

l 0 Steam line ruptures

[l 0.0$ High steam line AP  !

SIAS signal  ;

Main feedwater pump tripped Motor auxiliary feedwater initiated 17 Pressurizer low level alarm i

35 Pressurizer indicated empty e

50.$ HPI initiated

$3.05 Reactor coolant pumps tripped

• 310 Pressurizer level returned 1170 HPI flow stopped at shutoff head  ;

1300 Calculation terminated 7200 End of extrapolation

! 33 '

I r

__.,o _ , , . . . . . , , . . _ _ _ . _ _ _ _ _ , _ _ . , , m._ -_ . . . _ _ . . . _ _ _ _ . , _ . _ ___ , , , , , , . . . . . . _ _ . - - - - , . , . .. __, ~ _ _ _ . . - _ . . . . , .

CAUil0N: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

20 i i D PRIMARY

• o SG A -2500 a SG B n ( x SG C g 2

15 7

?, -

-2000 3 ,

E U 3 3 h 10 -- ,t --1500 y 1 i E ' -a-tale -a-a-a-a-a-a-a-a-a -1000 [

3 2 g 5 -

g

-500 0 #"""* - "": - - -- k 2 2 ^

0 0 500 1000 1500 Time (s)

Figure 5-l. Scenario 2 Primary and secondary pressures versus time.

Figure 5-2a shows break flow for the first 30 s Figure 5-4a presents the mass flow rate in the of the transient. Initially, the break flow peaked, three primary loop cold legs. All three loops then became choked downstream of the break at experienced an increase in mass flow prior to the the 0.13 m2 (1.4 ft 2) flow restrictor in the steam RCP trip due to the cooldown of the primary system generator outlet nonle. The flow rate fell until 1.5 s that increased the density of the reactor coolant.

when the separator in the ASG filled with liquid and After the pumps were tripped at 53 s, an asym-moisture began being carried over to the break, thus metric natural circulation condition was established, decreasing the void fraction of the exit fluid. This with flow in Loop A much higher than either of the decreased void fraction increased the mass lost and other loops. This higher flow was due to the greater break mass flow increased until 4.5 s, when the density head in the loop caused by the continued separators again became effective, cutting off the heat removal through the ASG. Figure 5-4b shows liquid flow to the break. Once the break quality this asymmetric flow in more detail. As shown, flow began to increase, break flow became a function in Loops B and C essentially stagnated and occa-only of pressure, increasing or decreasing as the sionally seversed. The unaffected loop flows were ASG pressure responded to the transient. nearly two orders of magnitude less than the cir-Figure 5-2b shows a comparison of break flow and culation in Loop A. This stagnant condition t.uxiliary feedwater flow during the transient. From isolated the USGs from the primary system and ,

120 s, except for five spikes caused by momentary allowed Loop A to dominate the downcomer ther-pressure surges because of slug flow of liquid to the mal response during the transient, break, auxiliary feedwater flow exceeded break flow and began refilling the ASO. Whether these spikes Figures 5 5 and 5-6 present the temperatures in are reasonable is not known; voiding may actually the hot leg, steam generator inlet and outlet, and occur in a more stable manner. The overall results cold leg in Loops B and C, respectively. In the of the calculation, however, will be shown initial part of the transient, prior to the RCP trip, insensitive to the spikes. Steam generator refill is the cooldown of the primary caused the unaffected shown graphically by the mass inventory in steam generators to become heat sources. This is Figure 5-3. reficcted in the figures by the steam generator outlet 34

fi

' AUTION: THE SCENARIOS SIMULATED C

CONTAIN SIGNIFICANT CONSERVATISMS IN ,

OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

2000 , ,

-4000 i l.

l o ,

1500 - -

a 7 a -

-3000 N

• " \ Separator O 5

c

! 1000 filled -

-2000 ..o a

, a l 8 3 2 g

( 500_ -

-1000 l

0 0 O 10 20 30 Time (s)

Figure 5 2A. Scenario 2 Break mass flow versus time for first 30 s of transient.

CAUT ION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

200 , i o BREAK FLOW -400 o AUXILI ARY FEEDWATER 150 - -

m 7 -

-300 $

b a a E

v

$ 100 - -

a ,

c -

-200 .o i 8 I

• 3 j ni 50 - ' -

-10 0 m j .

= = --c = o 9

0 ;. O O 500 1000 1500 Time (s)

Figure 5 28. Scenario 2 Comparison of break mass flow and ausiliary feedwater flow versus time.

35 e

- . _ _ ., . _ . . , _ . - , _ , , _ , _ . _ _ . _ - , . . , , _ , _ _ . , . . . _ _ _ . _ . . , . - , , . . . = _ , ~ - . - _ . _ . , . - . _ , _ _ , . , . _ . , , _ , _ _

CAUTION: THE. SCENARIOS SIMULATED CONTAIN SIGNiflCANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

60000 ..z ;. , :. .. . . :, :. -;. .a.a ,

.22 .....:.. - o

- 1.2 0 *10'

- 1.0 0 + 10' 40000 -

^ 4

- 8.0 0*10 j e

d. 0 SG A -

. a 6.0 0*10' 20000 - _

-- 4.0 0*10'

- 2.00*10' 0 ' '

O.00 0 500 1000 1500 Time (s)

Figure $.3. Scenario 2 Comparison of steam senerator nusses versus time, j CAUT ION: THE SCENARIOS SIMULATED CONTAlH SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

5000 , ,

4000

[ -

--- RCP trip $ A (W/ASG)

A LOOP C (W/PZR) -

- 10000

- 8000 m 7 (e3000 - - N

- 6000 j c

j 2000,

+

n -- 4000 $

c 3

• g 1000. -

~

- 2000 j ,

1. - -M

-9 91 -O g 9 0- - . ' - - - - -- -- - - - --O e

t 1

- -2000

-1000 -

0 500 1000 1500 Time (s)

Figure $4A. Scenario 2 Comparison of primary loop cold leg flows versus time.

36

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

400 i , ,

~

o LOOP A (W/ASG) - 800 o LOOP B A LOOP C (W/PZR) i - 600 m  ?

"a 200,- -

- 400 N

E

.x 1 .o t> O f .

- 200 g

M E O-- - WE --0 E S

- -200

'~

-200' 0 500 1000 1500 Time (s)

Figure 5 4B. Scenario 2 Comparison of primary loop cold les flows versus time (reduced scale).

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

600 , ,

_ ~

o HOT LEG Qv i o SG INLET 7 a SG OUTLET

• e550 x COLD LEG

)o -

A p-

-500 j a

- ~

h* e h 2 -

-400

• B450 8-4

-300 E - ~ E 3 400 3

-200 350 O 500 1000 1500 Time (s)

Figure 5 5. Scenario 2 Fluid temperatures versus time in primary Loop B.

37 L__

CAUT ION: THE SCENARIOS SIM'JLATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

700 i , 800 0 HOT LEG Q

V o SG INLET p .

A SG OUILET v

• x COLD LEG

• j 600_- --600 ) '

E i E

E .. E 500 - -

m -

-400 m h '

=

}

=

400 -

q_ _

x x

, -200 -

300 O 500 1000 1500 Time (s)

Figure 5-6. Scenario 2 Fluid temperatures versus time for primary Loop C (W/PZR).

temperatures being greater than the steam generator lished. At the time of termination the primarv inlet temperatures. Once the RCPs had tripped, the pressure was steadily increasing, the downcomer cold leg temperature diverged from that in the steam temperature was decreasing, and the ASG was still generator outlet as a result of the Dow stagnation filling.

discussed previously and the injection of cold ilPI liquid into the cold legs. 5.3.2 Extrapolations and Uncertainties. The following section presents and discusses the Figure 5-7 presents the Duid temperatures in the extrapolation of key PTS parameters out to 7200 s hot leg, steam generator inlet and outlets, and cold and the uncertainties involved in these parameters.

leg of Loop A, the affected loop. Over the entire The parameters discussed are downcomer pressure transient, the steam generator outlet temperature and fluid temperature, downcomer wall inner sur-was lower than the inlet temperature, reflecting the face heat transfer coefficient, primary loop cold leg continued heat removal through the ASO. The close mass flows, and fluid temperatures.

coupling of the steam generator outlet and cold leg temperatures, as well as the hot leg and steam Figure 5-9 presents the extrapolated downcomer genes ator inlet temperatures, resulted from the high pressure. At the time the transient was terminated, mass flow in the affected loop. pressure was still rising due to the continued influx ,

of raakeup flow. It was extrapolated that this rise Figure 5-8 shows a comparison of the three loop would continue until the makeup flow is terminated, cold leg temperatures with the downcomer fluid when pressurizer level attains its setpoint (see temperature. Once the RCPs were tripped, the Scenario 3 for further discussion). Following ter- ,

dominate loop in the system was Loop A, as seen mination the pressure will stabilize just below the by the close coupling of the affected loop's cold leg HPI pump shutoff head, and the downcomer temperature.

Figure 510 shows the extrapolated downcomer The calculation was terminated at 1300 s when temperature. The key to extrapolating this trends suitable for extrapolation had been estab- parameter was its close coupling to the affected 38

C AUT ION: THE SCENARIOS SIMULATED CONTAIN SIGNiflCANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

600 i ,

' ~

a HOT LEG 7

O SG INLET f A SG OUILET v e550 x COLD LEG e

-500 $

~

5 -

5w  %-

E 500 E

kE '

E E

2 -

-400

• y ,

450 -

%y%g - -

O -300 s400 - -

b 5  % -n :r -.. 0 3

-200 350 O 500 1000 1500 Time (s)

Figure 5 7. Seerario 2 Iluid temperatures versus time for primary Loop A (W/ASG).

CAUll0N: THE SCENARIOS SIMULATED CONTAIN SIGNIflCANT CONSERVATISMS lH OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH, 600 i ,

~

D LOOP A (W/ASG)

Q V I o LOOP R p v-A LOOP C (W/PZR) e550 x 00wnc0uta e

? -500 3 E E E 500 - -

E E o E

-400 2 m n

'3450 - - -

tr E-0

( -

^'

/ -300

$400 s % . rr%% - *=6 5 2 2 e

350

' ' \ -200 O 500 1000 1500 Time (s)

Figure 5 8. Scenario 2 Comparison of primary loop cold leg temperatures and downcomer temperature versus time.

39

C AUT ION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

16 , , , , , , ,

CALCULATED

-- EXTRAPOL ATED I4 ~

2000 9 g -

I

- a" E 12 - -

R a a a w ,q, .

j ___________________________ - 500 i 0

e 10

!e -

2 E

i

.i/ l a

o o 8 -

-1000 6

0 1000 2000 3000 4000 5000 6000 7000 8000 Time (s)

Figure 5-9. Scenario 2 Extrapolated downcomer pressure versus time.

CAUT ION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACil0NS, EQUIPMENT FAILURES. OR BOTH.

600 , , , , , , ,

~ '

CALCULATED 7

v EXTRAPOLATED p v

e 550 - -

5 -

-500 $

56 o 6

E 500 - -

E I

-400

• u u 450 _ _

E E

) -

-300 ( ,

f400 -

~- -- ._

fy f _'-

- ___ .,___ ,, -200 ,

350 O 1000 2000 3000 4000 5000 6000 7000 8000 Time (s)

Figure $.10. Scenario 2 Extrapolated downcomer fluid temperature sersus time.

40

I loop's cold leg temperature. Once the IIPI flow because a sery small reverse flow was present. This stops, the Loop A cold leg temperature rises to the reverse flow drew warmer fluid out of the inlet ASG outlet temperature. With auxiliary feedwater annulus and into the cold legs. Projecting this flow continuing to fill the ASG with cold water, the ASG to continue, cold leg temperatures in Loops 11 will become subcooled and continue to bring down and C would rise asymptotically to the downeomer the primary temperatures. This conclusion was temperature. Temperature uncertainties in the based on the results of Scenario 3 (Subsec. affected loop were minimal due to the constant flow

• tion 6.3.1). Using the downcomer temperature at through the loop, in the unaffected loops, flow the end of the calculation, coupled with the rate of behavior poses a substantial uncertainty For temperature decrease in the ASG outlet tempera- example, Loop Il experienced a flow reversal ture, the extrapolated temperature at 7200 s is 359 K around 900 s causing its cold leg temperature to rise.

o (187'F). If that flow reversal had not occurred the Loop 11 cold leg temperature could have gone dow n to very Figure 511 presents the extrapolated heat near the llPI injection temperature,305 K (90'F).

transfer coefficient at the downcomer wall inner surface. The heat transfer coefficient in the low An additional source of uncertainty concerning flow, subcooled regime in the downcomer is found Scenario 2 is discussed in Subsection 4.3.2, the using the Dittus Iloclier correlation, which possible asymmetric flow and temperature behavior decreases as flow decreases. Using extrapolated in the vessel, water properties and the flow extrapolation shown in Figure 512, the heat transfer coefficient was calculated at 7200 s. 5.4 Conclusions Figure 512 shows the three cold les flows The Scenario 2 calculation and extrapolation led extrapolated to 7200 s. Iloth Loop 11 and Loop C to the following conclusions:

remain stagnant, given the assumption of no operator action on the USO secondary side. Cold Peak pressure was extrapolated to be leg flow in the affected loop is very sporadic but 11.0 hlPa (15% psia),

generally drifting down as the cold leg temperature drifts down. Using the slope of the curve at the end hiinimum downeomer fluid temperature was of the transient, the flow was extrapolated to extrapolated to be 357 K (186'F).

7200 s.

The Scenario 2 calculation resulted in a con.

Figure 513 extrapolates the three cold leg tinuous cooldown of the primary system due to the temperatures o 7200 s. For Loop A, the AT continued flow of auxiliary feedwater into the ASG.

calculated between the end of the calculation and The primary system descloped an asymmetric 7200 s was identical to the AT used to extrapolate natural circulation condition with flow in the the downcomer temperature in Figure 510. Cold affected loop and stagnation in the remaining two leg temperatures increased in the stagnant loops loops, o

9 41

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

MM i i i i i i i 2 '

CALCULATED

-200 n' --

EXTRAPOLATFD E %% - -

-800 $

j 4000 hE 8i y -

-600 ua g j 3000 - -

g[.

• 2:

o N

6 -

-400 3

• 2000 - -

AE ~

I --------------------.-__________

~

f 1000 - . -200 1 , , , , , , ,

o , ,

0 1000 2000 3000 4000 5000 6000 7000 8000 Time (s)

Fisure 511. Scenario 2 Entropolated downcomer inner well surface twat transfer coefficient versus time.

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISWS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

400 ii , i . . . i i 0 LOOP A - 800 0 LOOP R a LOOP C i ;

- - EXTRAPOLATED - 600 m

7

( 200,- - 400 E

$ . - 200 g

- =

0- - 'llb-------------------------------- -

-0 ,

- -200

~~

-200' 0 1000 2000 3000 4000 5000 6000 7000 8000 Time (s)

Figure 512. Scenetto 2 Entropolated primary loop cold les flows versus time.

42

a CAUil0N: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

• 600 , , , , , ,

~

O LOOP A Q

V o LOOP B p I

A LOOP C e 550 l

-- EXTRAPOLATED

~

) q -500 j 2 2 f 500 2 -

-400

• 3 450 - - 3 g.

-300

$400 - ' -

. ,,,5 " ' % --

200 350 ' ' ' '

0 200 2000 3000 4000 5000 6000 7000 8000 Time (s)

Figure 513. Scenario 2 Extrapolated primary loop cold les temperatures versus time, o

o 43

6. SCENARIO 3, STUCK OPEN STEAM LINE PORV AT HOT STANDBY The following section describes the investigation 6.3.1 Calculation Results. A sequence of events of Scenario 3. A calculation was performed to for the calculation appears i? Table 6-2. This evaluate the consequences of a postulated transient sequence assumes the PORV in Steam Line A fails initiated by the failure-open of a single steam line open at time zero and the operator fails to isolate power operated relief valve (PORV) with the reac- auxiliary feedwater (AFW) to the affected steam -

tor at hot standby conditions. generator. The break size used was 0.009035 m2 (0.0978952 ft2 ). This was the size of the PORV A description of the scenario is provided, fol- required for the RELAPS model to pass 73.08 kg/s lowed by a discussion of the model changes required (161 lbm/s) of saturated steam to atmosphere at a ,

to perform the calculation. The results of the pressure of 5.45 MPa (790 psia). At the time of calculation, the extrapolated results, the uncertain- sequence initiation the reactor was at hot standby ties associated with the calculation, and the conclu- conditions. Initial reactor power was 8.29 MW sions are also discussed. representing decay heat at 100h after reactor shutdown.

Scenarios investigated in this report generally include consenative assumptions concerning equip- Upon opening the break, the Steam Generator A ment failures, operator actions, or combinations of (SGA) secondary pressure fell rapidly as shown in these. Conclusions relative to pressurized thermal Figure 6-1. Steam Generators B and C (SGB and shock severity are not to be drawn directly from the SGC) were isolated from the effects of the break results presented in this report (see Section 18). because of the steam line check valves in each steam line upstream of the common steam line header. As 6.1 Scenario Description a result SGB and SGC secondary pressures remained elevated except for the minor downward drift associated with the changing heat balance A description of the scenario as developed at among the steam generators shown in Figure 6-2.

Oak Ridge National Laboratory appears in Following the break the SGA heat removal rate Table 6-1. The scenario is initiated with a failed increased dramatically, while SGB and SGC became open PORV on Steam Line A with the reactor at heat sources to the primary coolant system. The hot standby conditions. Operator action is assumed peak in the SGA heat removal rate corresponds to to trip off reactor coolant pump (RCP) power when the time at which the SGA downcomer fluid flash-primary system pressure falls below 9.07 MPa ed and removed the subcooling effect on the out-(1315 psia). It is assumed the operator fails to side of steam generator tubes in the lower boiler isolate auxiliary feedwater (AFW) to the affected section. The effect of losing forced circulation heat steam generator, transfer on the inside of SGA tubes is also evident in Figure 6-2.

6.2 Model Changes The primary system pressure response is shown With the cyception of trip changes necessary to in Figure 6-3. The pressure initially fell as primary simulate the stuck open valve, the model used to liquid volume shrank due to the high SGA heat perform . this calculation is described in removal rate. The falling pressure caused the Subsections 2.1 and 2.2. The transient was initiated pressurizer heaters to be fully powered at 13 s. As from the hot standby conditions presented in shown on Figure 6-3, when the pressurizer emptied Subsection 2.3.3. at about 400 s the depressurization rate increased.

At 27 s a safay injedion activation signal (SI AS) 6.3 Results was generated due to high differential pressure be-tween the common steam line header and Steam The. following sections describe the analysis Line A. Immediate actions caused by the genera-results for a calculation of Scenario 3 and tion of the SIAS were: closure of the feedwater extrapolation and uncertainty of those results. (FW) bypass valves, tripping of power to the 44

T able 6-1. Scenario description No. 3 Plant Initial State - Just prior to transient initiator General

Description:

Hot 0% Power,0% Power after 100h of shutdown System Status Turbine: Not latched Secondary PORY: Automatic control

a. Steam Dump Valves: Automatic control Charging System: Automatic control Pressurizer: Automatic control Engineering Safety Features: Automatic control PORVs: Automatic control Reactor Control: Manual O Main Feedwater: ' In bypass mode, manual control to provide 39% level in S/Gs; I condensate pump,1 MFWP operating.

Aux Feedwater: Automatic control MSIVs: Open, Automatic control MFIVs: Open, Automatic control Transient Initiator A hole appears in Steam Line A outside containment upstream of the MSIV and downstream of the flow restrictor. The hole size corresponds to that of the steam line PORV.

Equipment Failures That Occur During the Transient if the Equipment is Demanded.

The operator fails to isolate AFW to S/G A.

Operator Reactions to Reported Information If SIAS signal has been generated the operator will trip the reactor coolant pumps when RCS pressure reaches 1300 psig.

Table 6-2. Scenario 3 sequence of events Time (s) Event 0 Break opens, stuck open PORV in Steam Line A, reactor at hot standby 2 Pressurizer proportional heaters powered 13 Pressurizer back-up heaters powered 27 Safety injection actuation signal (SIAS) on high steam line differential pressure Immediate actions caused by SIAS: FW bypass valves closed, main FW pump tripped, MFW pump recirculation flow terminated, motor.

driven AFW initiated, HPI and LPI pumps started (shutoff heads HPI: 1470 psia. LPI: 145 psia), letdown isolated .

30 Steam generator B becomes primary system heat source 33 Steam generator C becomes primary system heat source 81 Low pressurizer level indication, pressurizer heaters tripped off and makeup rate increased C 205 Pressurizer level indication zero (< 1%)

441 HPI flow starts (1470 psia primary system pressure) 999 RC pump trip (see text) I 9

1861 RC pump rotors stopped 1913 Makeup reduced to 15 spm RC pump sealinjection only, pressurizer level > level setroint 2455 - End of RELAP5 calculation 7200 End of extrapolation 45

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN

,0PERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

8 , , , ,

~

• = = = -1000

===:::== = ,

e ^

• 6 -

2

.k -

-800 $

e a LOOP A . *

$ o LOOP B g 2 A LOOP C -600 E e 4. _

' L o- n.

E -

RCP trip -400 E 3 W 3 o

>o2

-200

., 0 ' ' ' ' O O 500. 1000 1500 2000 2500 Time (s)

Figure 6-1. Scenario 3 Secondary system pressures.

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

150 , i , ,

4 , o SG A SG A downcomer flashes o so e a so c 10 0 RCP trip o

g 50 - -

I *===O , %

i O .... - - -

; - ,_.-.r -----

o

-50 O 500 1000 1500 2000 2500 Time (s)

Figure 6-2. Scenario 3 Steam generator heat removal rates.

46

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

16 , , , ,

-2200 m ~

~ ~

. _ y 14 - -2000 $

w 4 &

E 4 Pressurizer empty E R~ -1800 g

.g 12 MAKEUP flow throttled -[a.

a. ..

. -1600 l

E -

E 2 o - -

o

> 10 - -1400 >

, , , , -1200 0 500 1000 1500 2000 2500 Time (s)

Figure 6 3. Scenario 3 Reactor vessel downcomer pressure.

operating main feedwater (MFW) pump, termina- Figure 6-5 presents the pressurizer level response.

tion of MFW pump recirculation flow, initiation At 81 s the low level indication caused the termina-of motor-driven auxiliary feedwater (AFW) and tion of pressurizer heater power and an increase in isolation of letdown flow. The SI AS also caused the the makeup rate to full makeup capacity. The level high pressure injection (HPI) and low pressure indication was essentially zero at 205 s.

injection (LPI) pumps to be started. However no _ _

1 flow could be delivered from these pumps unless The primary system pressure continued to fall as primary system pressure fell below their respective shown in Figure 6-3 and, at 441 s, the pressure fell shutoff heads. Turbine-driven AFW flow was not below the HPI pump shutoff head. The HPI mass initiated during this sequence because 2-of 3 steam flow rate response, per loop, is shown in Figure 6-6.

generators are required to have low level indications HPI flow is inversely proportional to the primary and this condition was met only in SGA. system pressure. The volume addition rate to the primary system due to HPI flow more than offset Figure 6-4 presents a comparison of break and the continuing shrinkage rate due to cooling and auxiliary feedwater mass flow rates for SGA. The the primary system depressurization was reversed break flow initially peaked, then declined along with shortly after initiation of HPl. As HPI injection the SGA secondary pressure shown in Figure 6-1. _ continued, however, the pressurizer began to refill The AFW flow was essentially constant after AFW - with highly subcooled liquid. The mixing of the sub-initiation at 27 s at a rate that represented total cooled liquid and saturated or superheated vapor motor-driven AFW capacity. All AFW was directed within the pressurizer could result in a further to SG A because in HBR-2 a common AFW header depressurization of the primary system and this is used to feed all steam generators. Due to the would cause a reactor coolant pump trip if the break in' the SGA secondary system and the pressure declines to 9.08 MPa (1315 psia). It is-previously mentioned effective isolation of SGB and uncertain how well the RELAP5 cede can calculate SGC, the pressure in the AFW header was below the phenomena during the beginning of a pressur-that of SGB and SGC. As a result all AFW flow izer insurge. Furthermore, the scenario outcome is -

was directed to SGA. sensitive to the reactor coolant pump operation. It 47

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

10 0 , , , ,

O BREAK FLOW o AFW FLOW -200

~{cn -

-150 kE 6 e v

=

$ 50 -

3

. -100 .o w

w w

w w v w w w w w v v w w w w w

w a 2

3

-50 0a O O 500 1000 1500 2000 2500 Time (s)

Figure 6-4. Scenario 3 Affected steam generator auxiliary feedwater and break mass flow rates.

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN

- OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

0.4 , , , ,

0.3 - -

i

.e.

m 2 0.2 -

=

0 E

o x ,

0.1 - -

d 0.0 O 500- 1000 1500 2000 2500 Time (s)

Figure 6-5. Scenario 3 Pressurizer normalized level indication.

48

t L

CAUTION: THE SCENARIOS SIMULATED r i

l CONTAIN SIGNIFICANT CONSERVATISMS IN

- OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

15 i i , ,

-30

-25 ,

n . <

(_10 - -

N

!. -20

[ -

l

-15 f.

j 5,- -

-10 E" MAKEUR >

throttled l

-5 l, .

i

.0 0 O 500 1000 1500 2000 2500 Time (s)

Figurc 6-6. ' Scenario 3 Hish pressure injection mass flow rate (per loop).

ll was decided to assume that the primary system exceeded the volume addition rate and the primary depressurized further during the beginning of the system pressure dropped as shown in Figure 6-3. As pressurizer insurge and that the reactor coolant a result of the falling pressure the HPI flow pump trip occurred at 999 s. This time was calcu . Increased until the primary system shrinkage rate lated with RELAPS using alternate pressurizer was again exceeded by the volume addition rate, modeling.

The affected steam generator initially lost mass k,

Following the trip of the reactor coolant pump due to the effect of the break. However, after about power at 999 s, the pumps coasted down with the 800 s the AFW flow to SGA exceeded the break rotors stopping at 1161 s. Figure 6-7 shows that flow as shown in Figure 6-4 and the secondary following coastdown, Loop A natural circulation began to refill. Figure 6-8 shows the steam continued while Loop B and C flows approached generator secondary masses and Figure 6-9 shows stagnation. This difference was caused by the dif- - the corresponding narrow-range level indications.

ferent steam generator heat removal rate character. The dip and recovery in the SGA level at about istics shown in Figure 6-2. By about 2400 s flow 120 s was caused by flashing of the SGA rates in both Loops B and C were virtually zero, downcomer fluid during the blowdown.

Stagnation of Loops B and C occurred because SGB and SGC were heat sources to the primary Figure 610 shows the fluid temperature system and the heat added caused a sufficient responses in each of the cold legs near the reactor density difference to balance the pressure head that vessel and in the reactor vessel downcomer at the -

o- was driving fluid through Loop A. Stagnation was elevation corresponding to the top of the core. The not a result of voiding within the SGB and SGC Loop A cold leg was generally the coldest as a result tubes. of the continued SGA heat removal as indicated in Figure 2. Following RC pump trip the Loop B a- At 1913 s the pressurizer level had advanced to and C cold legs were essentially stagnant, however, the setpoint level and makeup flow was reduced to the minor Loop B and C flows shown in Figure 6-7 a constant 0.95 liters /s (15 spm) for the remainder were sufficient to prevent the Loop B and C cold of the sequence. This rate represents the net reac- leg temperatures from plunging, in general the

i. tor coolant pump seal Injection rate expected dur- Loop B cold leg fluid temperature is below that of ing this period. When the makeup flow throttled, Loop C because the makeup is injected only into the primary system shrinkage rate momentarily Loop B.

l 49 gy--wm w e ,,aw-,,v., --ew ..-c-y- .+r-' , , .- . , , . - +,.m-w , y,m,_- __..%e -,%.m=,.-_r,,....wy..wp, ew.,-,-%.v.r--~%--c..r..y

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISWS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

400 , , ;, , ,

~

O LOOP A - 800 0 LOOP B 300 -

a Loop c -

f e - 600

• m

• RCP coastdown O 200 - ~ N ccc - 400 E O =

- l .100 _- 1

-- 200 g w

0-- 2  ; ^--0 E

-100 -- -- -200

~

' ' ' ' - -400

-200 0 500 1000 1500 2000 2500 Time (s)

Figure 6-7. Scenario 3 Cold les mass flow rates near the reactor vessel.

CAUTION: THE SCENARIOS SIMULATED

. CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

90000 , , , ,

0 SG A -200000 o SG B 80000_- --180000 7,70000- -

--160000 E

.x .o C

N *

$60000 ~ - - -140000 5

= = -

======== = = = =

-120000 50000 - -

-100000 .

40000 '

O 500 1000 1500 2000 2500 Time (s)

Figure 6-8. Scenario 3 Steam generator secondary masses.

50

CAUT ION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, E0VIPMENT FAILURES, OR BOTH.

1.5 , , , ,

a SG A i o SG B

,o A SG C

.e S

8 i - -

6

• 5 E~

B c

o 0.5 - -

^

% i N  ::: ---. . . _

E 5

z

- caca_

0 O 500 1000 1500 2000 2500 Time (s)

Figure 6-9, Scenario 3 Normalized steam generator narrow range indicated levels.

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

600 , , , ,

~

a LOOP A -600 7

o LOOP B p v

A LOOP C e ' e

' x DOWNCOWER s 550 -

g E

N -500 o E 8.

E E E 500 -

y z z z :. z a :. -

2 3 -

-400 g 3

g.

=

l450 g

2 2

-300 $

s 400 O 500 1000 1500 2000 2500 i

Time (s) s Figure 6-10. Scenario 3 Cold les and reactor vessel downcomer flu!d temperatures.

51 w___-__-_- _

m

)

1 1

I The calculation was terminated at 2455 s because the 15 gpm minimum makeup volume addition a suitable indication of trends exiaed to allow an rate. The extrapolated pressure at 7200 s is extrapolation of results to 7200 s. At the ena of the 10.13 hlPa (1470 psia).

calculation the primary system pressure was rising slowly and the reactor vessel downcomer fluid Figure 6-12 shows the extrapolated and adjusted ,

temperature was falling slowly. The affected steam reactor vessel downcomer fluid temperature at an l generator separator was almost flooded. Loop A elevation equivalent to the top of the core. The  ;

continued to circulate. extrapolation from the end of the calculation to l 7200 s assumes that the existing cooldown continues 6.3.2 Extrapolations and Uncertainties. This sec- until 3800 s. After 3800 s the affected steam tion describes the extrapolation of results to from generator blowdown is complete and the cooldown 2455 s to 7200 s and adjustment of results to is based on an affected steam generator heat

• account for uncertainties in the computer removal rate of 14.4 AlW and a core decay heat of .

calculation. 8.2 hlW. The temperature extrapolated at 7200 s is 397 K (256*F).

Two uncertainties in the calculation will be addressed: pressurizer effects, and asymmetric hot Figure 6-13 shows the extrapolated heat transfer leg temperatures. coefficient on the inside surface of the reactor vessel downcomer wall. The extrapolation was performed As discussed in Subsection 6.3.1 there is uncer- simply by extending the trend present at the end of tainty about phenomena expected at the beginning the calculation. This is justified because trends in of the pressurizer insurge. It was assumed that flow conditions and temperatures are well estab-mixing of subcooled liquid and saturated or super- lished at the end of the calculation, heated vapor within the pressurizer would cause a depressurization of the primary system and cause Figure 6-14 shows the extrapolated cold leg mass a reactor coolant pump trip. flow rates between the llPI sites and reactor vessel.

Conditions present at the end of the calculation are Next, is the uncertainty due to asymmetric loop expected to continue through 7200 s. Loop A is behavior discussed in Subsection 4.3.2. For t:.: expected to continue in natural circulation while sequence being investigated here there is significant Loops B and C remain virtually stagnant.

asymmetry between affected Loop A and unaffected Loops B and C. It is not known, Figure 6-15 shows the cold leg fluid temperatures however, to what extent the cold leg fluids are mixed between the IIPI sites and the reactor vessel. The as they pass through the reactor vessel. Therefore Loop A extrapolation used the results of the it is not possible to quantify the effect of this uncer- downcomer temperature extrapolation from tainty on the parameters of interest. Qualitatively, Figure 6-12 and the relationship between the little effect on primary system pressure is expected. downcomer and Loop A cold leg temperature from The unaffected hot legs would likely be warmer and Figure 6-10. The minimum extrapolated Loop A the affected hot leg cooler than calculated. As a cold leg temperature is 395 K (252'F). As llPI flow result, the unaffected loops would flow somewhat decreases, Loop B and C cold leg temperatures are faster and the affected loop slower than calculated. expected to be the unaffected steam generator secondary saturation temperatures.

Extrapolations to 7200 s are shown on Figures 6-11 through 6-15 as dashed lines. These 6.4 conclusions dashed lines represent the best estimate responses for this scenario. The minimum reactor vessel downcomer fluid

• temperature occurred at the end of the two-hout Figure 6-11 shows the extrapolated and adjusted period. The extrapolated minimum temperature is primary system pressure. The horizontal dashed line 397 K (256'F). The extrapolated primary system indicates the primary system shrinkage rate exceeds pressure at that time is 10.13 AlPa (1470 psia). ,

52

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

16 . , , i i . .

• CALCULATED

-- EXTRAPOLATION -2200 m ^

14 - -

-2000

.. - S' E

• 3 -

-1800 g 5 12 - -

3

3. h.

-1600 E

3 I 3

>0 10

- - - - - - - ' - - - - ' " - - - ~ ' ' - - - - - - - - -

o

-1400 >

8 ' ' ' ' ' ' '

0- 1000 2000 3000 4000 5000 6000 7000 8000 Time (s)

Figure 6-11. Scenario 3 Extrapolated reactor vessel downcomer fluid pressure.

CAUTION: THE SCEANARIO SIMULATED CONTAIN SIGFNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

600 i , , i i i i CALCULATED n EXTRAPOLATION g g 550 -

g

-500 2 E' E f500 - -

h

{

E

-400

=

2.450

'?

')

. , " 300 8-

= . . , , , -

400 _ - - _ _ , -

e o

' b o 200

> 350 O 1000 2000 3000 4000 5000 6000 7000 8000 TIME (s)

Figure 612. Scenario 3 Extrapolated reactor vessel downcomer fluid temperature.

53

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

6000 , , , , , , ,

2 '

CALCULATED -1000 d -- EXTRAPOLATloN .

E k -

-800 j.4000 hp

• SJ

-600 g u:

;L 8

u -

-400 oE J 2000 -

AS E -_______ o A .

- - -200 $ '

O 0 ' ' ' ' ' ' '

O O 1000 2000 3000 4000 5000 6000 7000 8000 Time (s)

Figure 6-13. Scenario 3 Extrapolated reactor vessel wall inside surface heat transfer coefficient.

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

400 i, , , , , , .

o Loop A - 800 o Loop B 300 -

A LOOP C -

4 RCP coastdown --

EXTRAPOLATED - 600 7

$ \

S -

.__________________________ - 400

}200_-

U **

• 10 0 _ -

-- 200 E I

f, =

) 0-- --------------- ----------- --O s

_joo -

-- -200

' ' ' ' ' ' ' ~~

-200 ~

0 1000 2000 3000 4000 5000 6000 7000 8000 Time (s)

Figure 6-14. Scenario 3 Extrapolated cold les flow rates.

54

o CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

600 i , , , , , i

~

a LOOP A Q

V I o LOOP B @

v A LOOP C e550 -- EXTRAPOLATED

~

3 -

-500 3 E E f* 500

'i

/

r400 *  !

v <

v 450 - - -

E E

, -300 l 400 - ' ' " ' ' - - . -

2 2

-200 350 O 1000 2000 3000 4000 5000 6000 7000 8000 Time (s)

Figure 6-15. Scenario 3 Extrapolated cold leg fluid temperatures.

o I

e I

e

7. SCENARIO 4, THREE STEAM DUMP VALVES Fall OPEN AT FULL POWER The following section describes the investigation For this transient both the motor an! steam-of Scenario 4. The simulation was performed to driven auxiliary feedwater systems were activated.

evaluate the consequences of three steam dump Once activated, the motor-driven auxiliary system valves failing open with the reactor at full power, was left on. The steam-driven feedwater auxiliary .

This section contains the scenario description, was assumed to degrade to zero flow after the steam modeling changes ef fected to perform the transient generator steam dome pressures reached 0.69 MPa calculation, analysis of the transient results, (100 psia). The transient calculation was initiated extrapolation of the results and conclusions drawn from the full power steady state described in ,

from the analysis. Subsection 2.3.1.

Scenarios investigated in this report generally include conservative assumptions concerning equip-7.3 Results ment failures, operator actions, or combinations of these. Conclusions relative to pressurised thermal The following section details the results of the shock severity are not to be drawn directly from the Scenario 4 calculation. The first subsection results presented in this report (see Section 18). discusses the results of the calculation. The second subsection discusses extrapolation of results to 7200 s and uncertainties involved in the calculation.

7.1 Transient Scenario Description 7.3.1 c.icuistion riesults. The calculation of Scenario 4 was performed to evaluate the conse-A description of Scenario 4 appears in Table 71. quence of three steam dump valves failing open with This sequence definition was developed at Oak the reactor at full power. A sequence of events for Ridge National Laboratory (ORNL). The transient this simulation is summarized in Table 7 2.

was initiated from full power steady state conditions by locking open 3 out of 5 main steam line dump At zero time,3 out of 5 steam dump valves failed valves. All automatic plant functions with the open in the steam line. This hypothetical equipment malfunction is equivalent to a 0.043 m 2 (0.465 ft2 )

exception of the main steam isolation valve (MSIV) for Loop A are assumed to be operative. It was steam line break that affects all three steam assumed that the Loop A MSIV falls to close if a generators. After the break was initiated the MSIV trip signal is generated, however this condi. secondary pressure began to drop, resulting in a tion was not reached during the transient. The subsequent primary system cooldown.The reactor reactor was tripped at 41.1 s into the transient and was tripped at 41.1 s after transient initiation. This operator action is assumed to trip the reactor trip was imposed to account for reactor overpower coolant pumps on a low pressurizer pressure signal. conditions caused by moderator cooldown effects.

The reactor trip is based on a separate effects calculati n used t estimate the reduced average 7.2 Model Chan9es moderator temperature needed to mduce overpower conditions. The results of this separate effects The basic RELAP5 model used to perform the calculation were consistent with the time frame three dump steam valves failure open transient given by ORNL for the plant trip to occur.

calculation is described in Section 2. The follow. ,

ing changes were made to the control system model The primary and secondary system pressure to simulate the transient. The reactor was tripped responses are shown in Figure 71. Afler reactor trip at 41.1 s to simulate an overpower trip caused by the primary depressurization rate increased moderator cooldown effects. The steam dump con. significantly. The increased depressurization rate trol system was modified such that at the beginning was a consequence of enhanced primary system of the transient the total steam dump valve area was cooldown caused by a large mismatch between the set at a constant value representing the 3 stuck open core decay power and the much larger total power dump valve condition, removal through the steam generators (Figure 7 2).

56

Table 71. Scenario description No. 4 Plant initial State - Just prior to transient initiator General

Description:

100% Power steady state System Status o Turbine: Automatic control Secondary PORV: Automatic control Steam Dump Valves: Automatic control Charging System: Automatic control Pressurizer: Automatic control Engineering Safety Features: Automatic control PORVs: Automatic control Reactor Control: Automatic hiain Feedwater: Automatic control Aux Feedwater: Automatic control htSIVs: Open, Automatic control h1FIVs: Open, Automatic control Transient Initiator Three (3) steam dump valves fail open.

Equipment Failures That Occur During the Transient if the Equipment is Demanded.

1. The affected steam dump valves will not close.

2. htSIV A fails to close.

Operator Reactions to Reported Information

1. If SIAS signal is generated, the operator will trip the reactor coolant pumps when RCS pressure reaches 1300 psig.

2. Stop AFW Dow to the unaffected S/O when liquid carryover is observed in the main steamline.

3. Stop AFW Dow to S/G A 10 min after attempted automatic htSIV closure or when car-ryover occurs.

4. Operator fails to manually close htSIVs for the entire 2-h period.

The primary pressure reached a minimum of Throughout the simulation, the auxiliary feed-6.9 h1Pa (1000 psia) at 300 s. Although the primary water played a key role in maintaining enough cooldown continued after 300 s, repressurization steam generator mass inventory to sustain primary was induced by liquid volumetric addition from side cooldown. After the main feedwater train was both the llPI and CVCS injection. Although the isolated at $2.9 s the motor and then steam driven

, secondary system pressure initially dropped, a auxiliary feedwater systems where activated at substantial pressure increase occurred after the $2.9 and 98.2 s respectively (Figures 7 3 and 7-4).

steam turbine valves shut [ peak value of 6.1 N1Pa The trip signal that caused the feedwater valves to (885.0 psla) at 50 sl. This phenomena was caused close was triggered by a low primary average tem-when the secondary steam production rate tem- perature [<563 K (554*F)]. Final feedwater valve porarily exceeded the steam mass now rate out the closure occurred over a 4.0 s period. For each loop stuck open dump valves. As the reactor power partial feedwater valve closure had occurred prior dropped however, the total secondary side steam to the reactor trip. This was a consequence of the generation rate decreased below the mass now rate feedwater control system responding to conJitions out the dump valves and the secondary sides began induced by the stuck open dump valves. The to depressurize again. reduced feedwater now induced a trip signal to both

$7

l i

l Table 7-2. Scenerlo 4 sequence of events Time j (s) Event j 0.0 Three steam dump valves fail open 1.0 Proportional heaters on 5.0 Proportional heaters reach maximum power 17.35 Back-up heaters on .

41.10 Trip reactor 41.10 Trip turbine 42.0 Turbine stop valve closed 49.85 hiain feedwater valves start to close ,

l 52.9 Trip feedwater pumps 52.9 hiain feedwater valves closed 52.9 hiotor-driven auxiliary feed on 89.30 Back up and proportional heaters latched off 98.95 Steam-driven auxiliary feed on 132.30 SI signal on low pressurizer pressure (1730 psia) 132.30 CVCS flow reaches maximum due to letdown isolation 132.30 llPI pumps activated 175.0 llPI shutoff head overcome 180.0 Pressurizer emptied  ;

181.0 RCP pumps tripped 300.0 Transition to natural circulation loop flow-IIPI reaches maximum flow l l

895.0 Steam driven auxiliary feed off l l

1100.0 Primary side pressure stabilizes at 10 AIPa (1450 psia) l 1635.0 CVCS drops to minimum flow rate of 15 gpm 1700.0 End of calculation 7200 End of extrapolat;on i

l l

58

-r

CAUT ION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

20 , , , , , , , , ,

a PRlWARY 0 SG A -2500 h SG B T j$

X SG C _

Q I -

"a.

w -2000 v a

• il B g to -.- c- w _-1500 a.-

o.

-1000 l 5'-

-500 f

0

' ' ~*I*~ ==5''*~"r-* '

O O 200 400 600 800 1000 1200 1400 1600 1800 2000 Time (s) ,

Fisure 71. Scenario 4 Primary and secondary system pressures.

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES. OR BOTH.

500 , , , , , , , , ,

a TOTAL SG POWER 0 0 CORE DECAY POWER o

400 - -

RC pump trip y 300 -

( -

o n 8,

l 200 - Steam driven .

AFW on o

10 0 -

0 0 200 400 600 800 1000 1200 1400 1600 1800 2000 Time (s)

Figure 7 2. Scenario 4 Decay power and total steam generator power.

~

59

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

15 i , , , i i i i i

_ =-= .  ; -*;-- _. , -30 M

~~

-25 k 10 - -

I -

-20 .o

• O

- a SG A -15 $

o 50 8 *-

E 5 - - U 2 -

-10 j

-5 0n O O 200 400 600 800 1000 1200 1400 1600 1800 2000 Time (s)

Fisure 7 3. Scenario 4 Motor driven AFW flows.

CAUT ION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACil0NS, E00lPMENT FAILURES, OR BOTH.

20 , , , , , , . . .

0 SG A - 40 0 SG B RCP trip a so c -

15 -

1-a - 30 7

4

$; 10 , - -

)

- 20 $

$ 8 2

,, 5 e . to ,,

n -

2 3 .

O n- m-#+#+n +n-a--n-n -

-0 i i i i i i i .i i - -10

-5 O 200 400 600 800 1000 1200 1400 1600 1800 2000 Time (s)

Figure 7 4. Scenario 4 Steam driven AFW flows.

60

~ _ _ _ _ _ - _ _ _ _ _ - _ _ _ _ _ _ _ - _

i l

feedwater pumps that in turn activated the motor- Dows (Figures 7 7 and 7-8). Although, llPI con-

driven auxiliary feedwater pumps. As each steam tributed to cooling the primary fluid after 300 s,

, generator continued to lose mass inventory, a two- heat removal to the steam generators was the domi-out-of-three narrow range low level signal (narrow nant mechanism by which the primary coolant range <l5% of full span) activated the steam- temperature was reduced. Variations in CVCS driven auxiliary feedwater system. injection did not significantly perturb the o downcomer temperatures or produce significant Presented in Figure 7 5 are the steam generator temperature differences in the cold legs (only mass inventories for the three loops. Each steam Loop B had CVCS injection).

generator is characterized by approximately the

, same response with the mass inventories being Both the llPI and CVCS systems were responsi-somewhat larger in Steam Generators B and C since ble for the repressurization of the primary coolant the feedwater lines to these two steam generators system. Initially, the llPI was significantly larger than are shorter than that connected to Steam the CVCS mass flow rate. Prior to 1100 s the llPI Generator A. Thus Steam Generators B and C flow was the principal cause for repressurization of received more auxiliary feedwater. During the initial the primary coolant. This condition changed around 600 s of the transient each steam generator lost 1100 s when the primary pressure approached the N40% of its total mass inventory, llP! shutoff head (Figure 7-9). The increased primary system mass inventory eventually caused a After s600 s the steam generator pressures were repressurization of the primary system cl)ng with a reduced to 1.03 h1Pa (150 psia) such that the total refilling of the pressurizer (Figure 710)?

auxiliary feedwater mass flow rate execeded the mass flow rate through the steam dump valves.

Thus, each steam generator began to refill. At 895 s At i100 s the pressurizer pressure reached a the steam-driven auxiliary steam feedwater system plateau of N10 h!Pa (1450 psia). At Nils time the was no longer available due to insufficient steam primary coolant shrinkage rate due to cooldown generator pressure to drive this system (Figure 7-4). effects was balanced by approximately the same Thereafter, each steam generator continued to refill am unt ofliquid volume injected from the !!PI and as a consequence of the motor-driven feedwater CVCS systems. During the remainder of the simula-system flow. tion the primary system pressure was maintained at sl0 h1Pa (1450 psia) as the primary system For a brief period of time the auxiliary feedwater c Idown continued.

flow contributed to increasing the energy removal between the primary and secondary sides. This 7.3.2 Extrapolations and Uncertainties.

enhanced removal rate occurred at approximately Figure 711 presents the extrapolated pressure the time the steam-driven auxiliary feedwater system response in the reactor vessel downcomer at an was activated (Figure 7 2). This additional flow elevation adjacent to the top of the core. At 4830 s approximately doubled the amount of liquid going primary side repressurization was predicted to occur to each steam generator, llowever, this enhanced at the time the steam generator secondary sides cooldown continued only until the primary reactor reached atmospheric pressure conditions and the coolant pumps were tripped at the pressurizer secondary depressurization stopped. It was con.

pressure setpoint of 9.07 h!Pa (1315 psla). The cluded that the core decay power matched the total subsequent primary coolant flow reduction resulted steam generator heat removal rate. As a conse-in an immediate reduction in primary to secondary quence, when the secondary depressurization and heat transfer, cooldown stopped, shrinkage of the primary coolant ceased. Continued CVCS injection then fly 300 s the primary reactor coolant flow in all caused a gradual primary side repressurizaticn. By three loops had transitioned to natural circulation 7200 s the system pressurized to it.H h1Pa (Figure 7-6). During the remainder of the simula. (1711 psla) with an estimated uncertainty of tion the loop flows remained relatively stable at 1.4 h1Pa (203 psia). The principal source of uncer-sl50 kg/s (330 lbs/s). Generally, the natural cir- tainty in the final primary pressure was the culation loop mass flow rates wcre at least an order estimated time at which the steam generators would of magnitude larger than either the llPI or CVCS blow down to atmospheric conditlom.

61

CAUT ION: THE SCENARIOS SIMULATED CONTAIN SIGNlflCANT CONSERVAllSMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

45000 , , , , , , , , ,

a so A -100000 0 SG B .

A SG C 40000 - -

-90000 9 E df. B 35000 -- -

-80000 Steam driven J AFW off j 30000 ~ -

^^

--70000 s' ' ' '

-60000 25000 0 200 ~400 600 800 1000 1200 1400 1600 1800 2000 Time (s)

Fisurs 7.$. Scenario 4 Steam senerator mass inventory.

CAUil0N: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISWS lil CPERATOR ACTIONS EQUIPMENT FAILURES, OR BOTH.

5000 , , , , , , i i i a LOOP A -10000 o LOOD R I A LOOP C 4000 -

m

-8000 g E

I 3000 -

- a

" - -6000 0 8

= e 2000 g , -4000 a u

.o 1000 . - -2000

, ;00,000, : :i : a ; : g-a -- :, .- , _,

0 200 400 600 800 1000 1200 1400 1600 1800 2000 Time (s)

Figure 7-6. Scenario 4 Primary coolant mass flow rates.

62

i

, CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

15 , , , , , , , , ,  !

30 t

-25 ,

n 7

• (e M -

- - N E

g -20 g a

-15 f

~ ~

-10 5

0 0 0 200 400 600 800 1000 1200 1400 1600 1800 2000 Time (s)

Fisure 7 7. Scenario 4 HPt mass flow rates. ,

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISWS IN  ;

OPERATOR ACTIONS, E0VIPMENT FAILURES, OR BOTH.

200 , , , , , , , , ,

' ' SIAS j 150 - -

3 2 100 -

r E

$ 50 - -

G ,

3 0 C -

I '

) -50 - -

t

-100 i 0 200 400 600 800 1000 1200 1400 1600 1800 2000 "

Time (s)

Figure 7 8. Scenario 4 CVCS net injection flow rate.

63

?

CAUil0N: THE SCENARIOS SIMULATED CONTAIN SIGNiflCANT CONSERVATISMS IN OPERATOR ACTIONS, E0VIPMENT FAILURES, OR BOTH.

600 , , , , i i , , ,

-600 7

v p

v e e y550 }

2

-500 o ,

& 1 E E E $00 - -

2 e u j -

-400 =}

=

g

-g450 2 S -

-300 $

400 0 200 400 600 800 1000 1200 1400 1600 1800 2000 Time (s)

Figure 7 9. Scenario 4 Reactor vessel downcomer fluid temperature.

CAUT ION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

0.6 , , , , , , , , ,

\

0.4 - -

2 k

E j 0.2 - -

l l l I I l 1 l 1 0 200 400 600 800 1000 1200 1400 1600 1800 2000 Time (s) l Figure 710. Scenario 4 Normalized pressurizer level indication.

l 64

CAUil0N: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

16 , , , , , ,

CALCULATED

, -- EXTRAPOLATED I4 - ~

9 -

2000 7 I 7 s

8 12 -

?,,

m a ~

, ,, i

---

------~~~'

• 10 e

E E

E 3

$e - _ $

-1000 6 ' ' ' ' ' ' '

0 1000 2000 3000 4000 5000 6000 7000 8000 Time (s)

Figure 711. Scenario 4 Extrapolated reactor vessel downcomer pressure.

Coincident with the stabilization of the secondary Loop B were not significant. This is because the side conditions was a stabilization of the primary loop natural circulation mass flow rate is much side downcomer and cold leg temperatures greater than the CVCS mass flow rate. The (Figures 712 and 713). After natural circulation minimum downcomer/ cold leg temperatures were had been established in the primary coolant loops extrapolated to 373 K (212'F) with an estimated -

the temperatures of the fluid exiting the primary uncertainty of + 5 K (9.8'F). The principal uncer-side steam generator outlets were approximately tainty in estimating the extrapolated downcomer equal to the secondary side steam generator boiler and cold leg temperatures responses was a function saturation temperatures. It was concluded that this of when the steam generator secondaries reach trend would persist out to 7200 s. This is because atmospheric conditions, of relatively low primary loop mars flows and because the core decay power out to 7200 s is ade- The extrapolated values for the cold leg loop quate to maintain the liquid in the steam generator flows and downcomer heat transfer coefficient were boiler sections at saturation conditions. At 4830 s maintained at the values calculated at 1700 s when the steam generators reached atmo3pheric (Figures 714 and 715). These estimates were made conditions both the downcomer and cold leg using the assumption that the primary system temperatures were estimated to stabilize at 373 K . density gradients that drive the natural circulation o (212'F). This temperature corresponds to at- loop flows would not significantly change in the mospheric saturation conditions without accounting 1700 to 7200 s period. The extrapolated cold leg for the steam generator boiler hydrostatic head mass flow rates and downcomer heat transfer coef-effects that would slightly increase the boiler ficient had uncertainties as a consequence of the e saturation temperature, assumption that the primary system density gra-dients did not significantly change from 1700 to The extrapolated primary cold leg temperatures 7200 s. These extrapolations are biased on the high were symmetric as a consequence of nearly sym- side since both the loop flows and downcomer heat metric secondary side conditions. The temperature transfer coefficients decrease as a consequence of mixing effects caused by CVCS injection into core power decay and secondary side stabilization.

65

CAUT ION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, E0VIPMENT FAILURES, OR BOTH.

600 , , , , , , ,

_ ~

CALCULATED 7

v EXTRAPOLATED p v

. 550 -

5 -

-500 $

%6 %s =

500 - -

2 -

-400

• v v 450 - - -

E E

:

. -300

- 400

%,,~ _

f

~..'

......... . -200 350 O 1000 2000 3000 4000 5000 6000 7000 8000 Time (s)

Figure 712. Scenario 4 Extrapolated reactor vessel downcomer fluid temperature.

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

600 i , , , , . .

-600 o LOOP A Q

v o LOOP 0 7 w

i a LOOP C e 550 -- EXTRAPOLATED e

3 -

500 3 E E 500 - -

400 .

t v v

~ 450 - - -

E E

~ "

., -300 .

~

$400 -

'.s......

' ~

-200 .

350 0 M00 2000 3000 4000 5000 6000 7000 8000 Time (s)

Figure 7.I.1. Scenario 4 Extrapolated cold les fluid temperatures.

66

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISWS IN l OPERATOR ACTIONS EQUlPMENT FAILURES, OR BOTH.

50000 , , , , , , ,

CALCULATED 0 {

n

-- EXTRAPOL ATED

-8000 E iA N 40000 5

7 v .I

, q -

-6000 hp i l

$ 30000 - -

8i c i I

-4000 OI 20000 - -

l}

x5" g

-2000 j

• 10000 "- -

0 O 0 1000 2000 3000 4000 5000 6000 7000 8000

_ Time (s)

Figure 714. Scenario 4 Extrapolated reactor vessel downcomer inside surface heat transfer coefficient.

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

1000 ij i , , , , , ,

. O LOOP A -2000 o LOOP B 4 LOOP C 800 -

EXTRAPOLATED k -

-1500 k

$ 600 - -

-1000 $

. 400 - -

8 0 l o 2 j 200'-'I --500 9

0 O O 1000 2000 3000 4000 5000 6000 7000 8000 Time (s)

Figure 713. Scenario 4 Extrapolated cold les mass flow rates. L 67

i

! - 7.4 Concluelone were extrapolated to 373 K (212'F) and 11.8 MPs l (1711 psia) at 7200 s respectively, l

The temperature responses in all three loops were The simulation of Scenario 4 indicated that the characterized as beins relatively symmetric primary coolant system initially experienced a rapid throushout the simulation. The principal cooldown and depressurisation until about 300 s, mechanism for the cooldown of the primary coolant followed by a stadual cooldown as natural circula- was natural circulation. ,

tion cominued and heat was transferred to Ihe steam By enerapolating ihe simulation results to 7200 s generators. As a consequence of CVCS and HPI it was found that the primary system was main.

operation, partial recovery of the primary system tained in a natural circulation cooldown mode with pressure occurred. The minimum downcomer fluid ihe downcomer and cold les fluid temperatures ,

temperature and subsequent maximum pressure stabilizing at W)0 s.

9 e

0 0

Y

8. SCENARIO 5, OVERFEED WITH AUXIUARY FEEDWATER AT FULL POWER The following subsections describe the Insestiga. 8.3 Results tion of Scenario S. This calculation was performed to evaluate the consequences of a postulated steam o generator overfeed with ausiliary feedwater with the This section is divided into two subsections. The I

reactor at full power conditions. first presents the calculated results, and the second extrapolates the calculated results to 2 h and A description of the scenario ls provided followed discuocs uncertainties associated with the calcula-

, by a discussion of the model changes required to tion and extrapolation.

perform the calculcion. The results of Ihe calcula.

tion, the extrapolated results, the uncertainties with 8.3.1 Calculation Results. The sequence of events the calculation an,i the conclusions regarding the for this transient is presented in Table 8 2. The calculation are also presented, transient was initiated by tripping both MI:W pumps. This trip caused an immediate turblne trip, Scenarios investigated .n this report generally which then tripped reactor power. T he steam con-include conservative rnumptions concerning equip. trol vahe closeu by I s, simulating the clos ure of ment failures, operator actions, or combinations of the turbine isolation valve. Also at 1 s, the steam these. Conclusions relative to preuurited thermal dump 5ah es opened on a load rejection signal, llot shock severity are not to be draw n direstly from the leg temperatures dropped immediately as core results presented in this report (see Section 18). power dropped. This cooldown shrank the primary system fluid volume and caused an outsurge from the penurlier, w hish in turn caused primary system 8.1 Scenario Description preuure to fall, as shown in l'igure 8.l. At 3 s, the preuurlier heaters came on as a result of the drop Table 81 givcs the description of the desired In preu ure, scenario as supplied by ORNL. The transient was initiated from full power steady state conditions by At 4 s, the steam dump valves were closed as tripping both main feedwater (MI W) pumps. T he shown in l'igure 8 2 and at 29 s were modulated auxillary feedwater (Al W) pumps failed to start open to control the primary system average when first demanded but, at il min into the tran. temperature to $$9 K ($47'IT). 't he primary preuure i

tient, both motor and turbine driven ausillary feed. continued Io fall until approsimately 90 s, at w hlch water pumps were manually started and set to time the preuurlicr heaters overcame the ef fect of provide maximum flow. The ausiliary feedwater the cooldown, and primary system preuure began was terminated when the vold fractions in the to rise, it rose steadily until it reached the heater volumes representing the steam domes of all the control point of l$.6 Mpa (2270 psia) at (J4 s. The steam generators dropped below $0% This cutoff pressurlier heaters then cyc!cd on and off to hold point simulated the time when liquid carryover the preuure at that point. The primary splem would be observed in the steam line. preuure never dropped low enough to activate the lipl, or trip off the reactor coolant pumps.

All other control systems operated in their 11gure 8 3 presents the three cold les flow rates, automatic modes.

Energy remosal out the steam dump valses liept a 8.2 Model Changes ' h' P ' '" ' Y '""'" '"" P" *' " '" "' 8 6" k l ($48'I') from 29 to 4NO s. At 4NO s. Al:W flow was initiated (see i Igure 8 4). The initial AlV llow did No changes in the h>drodynamic model were not cause a significant cooldown in the primary neccuary.1 he only changes in the model werc the sptem. This was because the fluid inillally injected trips that controlled the main feed *ater pumps, tur. to the steam generators was the relattsely warm bine trip, and the ausiliary feed

• ater setem.1he 11uld whkh had stagnated in the feedwater lines.

changes to the main feedwater pump and the nusil. l'igure 8 3 show5 the feedwater temperatures for lary feedwater control systems imple*nent the each of the steam generators. As more cold AI:W scenarlo described abose, fluid was injected into the feedwater lines, the i

l 69

Table 81. Scenario desceiption No. 5 Plant initi J State Just prior to tramient initiator General

Description:

100Ve Power steady state System Status

• Turbine: Automatic control Sceandary PORV: Automatic control Steam Dump Valves: Automatic control Charging System: Automatic control e Pressurlier: Automatic control Engineering Safety I'entures: Automatic control PORVs: Automatic control Reactor Control: Automatic Main l'cedwater: Automatic control -

• Aus Itcedwster: Automatic control MSIVs: Open, Automatic control MIlVs: Open Automatic control Transient initiator lloth Main 17eedwater pumps trip simultaneously.

Equipment l'adures That Occur During the Tramient if the Equipment is Demanded.

Ausillary feedwater pumps fall to start.

Operator Reactions to Reported information I. Operator inillates actiom to correct Aus now problem and restarts all Al W pumps to pro-vide mas now at 8 min. Aus feedwater How at manimum la all 3 steam generators, and all Al:W pumps started Weedwater source: ConJentate tank).

2. Stop all AI W How when liquid carryover is observed in the main steam lines, temperatures in the lines dropped, and colder water How then diverted to % team Generator 11. At 1300 a was injected into the sicam generators, the feedwater injection volume filled in Steam 00 erator Il and the AI W flow shifted to Steam During thh itage of the tramient, the liquid leseli Omtrator A. At 1340 e the same volume filled in in the secam sencrators were below the elevation of Steam Generator A and the Al:W How spill the feedwater ring. As a result, the feedwater was between the three steam generators, injetted into sicam filled volumes. At 784 s, the temperature of the feedwater entering Steam l'igure 8.$ shows the total energy tramictred Generator C had dropped to the point where 11 from the primary system to all three steam o induced an increate in condemation in that generators. As the now shifted between generators, generator, and dropped the preuure slightly. Thh different temperature w ater was delbered to one of drop in preuure dherted all the Al:W How from the generators, therefore total heat tramfer the Iwo other steam generstors and delbered it all changed. Also, there was a time lag between when e to Steam Ocnerator C.1hh estra now deopped the cold feedwater was injected into the top of the steam feedwater temperature even further, and kept the generator downcomer, and when it arrived at the generator preiture depreued. At 970 s, the volume tube bundle.1he result of these shifts in feedwater that recched the feedwater flow in Steam Now and temperatures were the oncillations seen in Ocnerator C filled with liquid, and the condema- 11gure 84,1he$e meillations in energy remmal rate tion induced depreuurliation ended. The AliW camed the oscillatlom neen in the downcomer and 10

l Table 62. Soeneele 5 seguenee of events t Time '

(i) Event 0 Tripped MFW pumps; turbine tripped; reactor tripped O

I Turbine isolation valve closed; steam dump valves opened on load rejection; propor-tional preuurizer heaters turned on 3 Backup preuurlier heaters turned on o

4 Steam dump valves closed 5 Main feedwater valves closed 29 Steam dump valves opened 440 AFW, motor sad turbine, turned on manually 674 Backup preuurizer heaters turned off (cycle on and off until %950 s) 784 AFW shifted to SGC 854 Steam dump valves closed 880 High Tave dropped <$47'F, steam dump control shifted from plant trip control (PTC) to steam pressure control (SPC) 970 AFW Dow shifted to SOR; preuurizer proportional heaters turned off 991 Preuvelier sprays turned on (cycle on and off until sil00 s) 1300 AFW shifted to SOA 1540 AFW now splits between all 3 Sus 1924 AFW now shifts to SOA 1976 Pressurlier heaters turned off on low preuurlier level 20l3 SIAS signal on high differential prenure between Steam Line A and header; letdown flow isolated 2352 SOA filled and llolated; AFW flow shifted to SOR 269) SUN filled and isolated; AFW How shifted to SGC 3027 SOC filled and loolated; AFW llow turned off 3131 Prenurlier spray turned on o 3201 Makeup flow at minimum value l 3600 I!nd of cakulation 7200 I!nd of entrapolation l

t 71 j

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

16 i , ,

-2300 t Spray m a on m O ~ 15.5 --  % --2250 .

.h

, \ n v Pressurizer Primary lleaters off Liquid -2200 E AFW Filled R 15 -

shifted

.! E a - among SG -215 0 5

,E Letdown E 3 Isolated 2 y 14.5 - --210 0 g

-2050 14 O 900 1800 2700 3600 Time (s)

Figure 8-1. Scenario 5 Primary system pressure.

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

100 , , , ,

-200 80 - -

-150 I ' 60 m

O

[

! -100 S

. 40 - -

$ E 2 i .

-50 20 ~ - -

4 .

0 O O 200 400 600 800 1000 Time (s)

Figure 8-2. Scenario 5 Total dump valve flow rate.

72

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

4600 i , i u LOOP A

- o LOOP B -10000 0 4500 - -

-9800 k

, [4400 - ' b g -I -9600 3 2 43 -

- A ,

/ -

5

-9400 j

-k#"

I 8

2 4200 - -

-9200 410 0 O 900 1800 2700 3600 Time (s)

Figure 8-3. Scenario 5 Cold leg flow rate.

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

100 i i i g a STEAM GENERATOR A -200 g o STEAM GENERATOR B a STEAM GENERATOR C 150 $

. 1 O

l

-100 $

+

Q- m 8

~

,p#5 '@ ( ) ~5 g

4 4

1. ' '~

0 - ^^ " ' -^ ^^

^^^-^^"O O 900 1800 2700 3600 Time (s)

Figure 8-4. Scenario 5 Feedwater flow rates.

73

i CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUlPMENT FAILURES, OR BOTH.

550 , , ,

O STEAM GENERATOR A -500 7 0 STEAM GENERATOR B p

• 500I a STEAM GENERATOR C - V

• E 3

-400 I 3

2450 - - o 3 400

-300 f I 3 o -

-200 o E350 - -

I

• . ._ e b ~

~ .:. - -

g. -100 $

g 300 - ~

mM _ . . . .

250- i i i -0 0 900 1800 2700 3600 Time (s)

Figure 8-5. Scenario 5 Feedwater fluid temperatures.

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

200 , i i 3 -

6 10 0 E

e j [

A

\

q -

0 O 900 1800 2700 3600 Time (s)

Figure 8-6. Scenario 5 Total primary to secondary heat transfer.

74

. . . . ._ - .-. ~ .. - - . . .-

t cold leg temperatures between 900 and 1800 s, as pressurizer. At 3131 s, pressurizer sprays came on and shown in Figures 8-7 and 8-8. The oscillations in held the primary system pressure at sl6.0 MPa primary system temperature caused the oscillations (2320 psia). At the end of the calculation, 3600 s, in primary pressure over the same time interval by primary system pressure was stable at 16.0 MPa alternating swelling and shrinking the primary fluid. (2320 psia) and the reactor vessel downcomer fluid The oscillations are also seen in the pressurizer temperature was at 546 K (523*F) and rising steadily.

, liquid level, shown in Figure 8-9. The changes in primary system pressure were limited by the pres- 8.3.2 Extrapolations and Uncertainties.

. surizer heaters and sprays, which turned on and off Figures 8-11 and 8-12 show the downcomer

throughout this period as determined by their pressure and temperature extrapolated to 7200 s.

respective controllers. Pressurizer spray capacity was sufficient to control primary system pressure during the heatup By 1800 s, the effects of the shifting AFW flows experienced at the end of this transient. Therefore, damped out as all the warm water in the feedwater the primary system pressure stayed at sl6.0 MPa lines was flushed out. The AFW flows shifted (2320 psia) through 7200 s. Primary and secondary around again between 1924 and 3027 s due to con- system temperatures were closely coupled and densation induced pressure drops, but since the increased until the secondary pressure reached feedwater temperatures to all the steam generators 7.03 MPa (1020 psia) at about 4600 s, at which time were the same, it made no difference which the steam dump valves opened to hold steam gener-generator received the feed flow. Between l800 and ator pressure at that point. This pressure cor-3027 s, the primary system experienced a gradual responds to a saturation temperature of 559 K cooldown as energy was removed from the primary (547'F). The primary temperature would therefore system by the cold AFW flow. The pressurizer rise until it reached 559 K (547'F) and then stay heaters attempted to recover the drop in primary there through 7200 s. The cold leg temperatures system pressure associated with this cooldown until would be essentially the same as the downcomer 1976 s, when the heaters were shut off on low temperature through 7200 s because the primary 4 pressurizer level. coolant pumps would stay on Cold leg flow rates would also stay nearly constant through 7200 s At 2015 s, the A steam generator was receiving because of primary coolant pump operation.

! all the AFW flow, and that flow caused a sufficient Figure 8-13 shows the heat transfer coefficient at depressurization in the A steam line to cause an the outer wall of the downcomer, extrapolated to SIAS signal on high AP between the A steam line 7200 s. There were no major uncertainties I and the steam header. The only significant result associated with the calculation.

, of the SIAS signal was that the letdown flow was 2

isolated. This resulted in a net increase of 8.4 Conclusion 8 3.3 x 10-3 m3/s (54 gpm)in the makeup flow. This increase in flow was sufficient to overcome the This transient was mild primarily for two reasons.

volumetric shrink as the primary system cooled, and The primary system never depressurized enough to therefore, pressurizer level began to increase and initiate HPI or to trip the RCPs. Because no HPI primary system pressure began to rise, was injected, the primary system did not cool as quickly, and more importantly, because the RCPs Figure 8-10 shows the narrow range liquid levels did not trip, the primary fluid was kept well mixed of the three steam generators. At 2352 s Steam throughout the transient. Secondly, the AFW flow Generator A filled with water, and feedwater to that rates were not sufficiently large to cause a serious generator was isolated. Steam Generator B filled and cooldown. The minimum reactor vessel downcomer

• isolated at 2693 s, and Steam Generator C filled and temperature reached was $35 K (503*F), and occur-isolated at 3027 s. After Steam Generator C isolated, red at 3027 s. The maximum reactor vessel all AFW flow was stopped, and the energy transfer- downcomer pressure after this time was 16.0 MPa red out of the primary system dropped immediately. (2320 psia). After extrapolating the results out to

+ As a result, primary system temperatures started to 7200 s, the downcomer pressure and temperature rise. Primary system pressure also started to rise more would be 16.0 MPa (2320 psia) and $59 K (547'F) i rapidly as the heatup caused an insurge into the respectively.

75 d

- , . , , -- , - . , _ . , . - , _ . , _ . , , - . , , . . - . , , _ , . r.-, , . . , , -- , n

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

570 , , ,

n.

-560 n v AFW Shifted P-

$ m Among SG m 5 560 k g

5  %

-540 g a a .

E E 2 550 - -

2 3 3 E -

-520 =E

=

g 540 - -

g 2 2 AFW off -500 530 O 900 1800 2700 3600 Time (s)

Figure 8-7. Scenario 5 Reactor vessel downcomer fluid temperature.

CAUTION: THE SCENARIOS SIMULATED CONTAlH SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

570 i , ,

_ a LOOP A -560 7

V o LOOP B [

v a LOOP C l

560, [= =- -

h 3 3 5 -

-540 g a a E E 2 $50 - -

2 2 , 2

=

E -

-520 =E g540 - -

g 3 3

-500 ,

~530 O 900 1800 2700 3600 Time (s)

Figure 8-8. Scenario 5 Cold leg fluid temperatures.

76

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES,' OR BOTH.

0.6 , , ,

0.5 -

7

~

0.4 -

o 2 ir' 0.3 -

I e

-g 0.2 -

5 z

0.1 -

0.0 O 900 1800 2700 3600 Time (s)

Figure 8-9. Scenario 5 Pressurizer normalized liquid level.

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

1- , , c c q -

=

0 -STEAM GENERATOR A o STEAM GENERATOR B A STEAM GENERATOR C e

2

• 0 0.5 -

3 5  :

D' il o

z 4

0 ' ' '

O 900 1800 2700 3600 Time (s)

Figure 8-10. Scenario 5 Steam generator narrow range normalized lig.iid levels.

77

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

16 y , , _,_,i_____t____;_____1_____

- 1O CALCULATED -2300. 4

-- EXTRAPOLATED 15.5 --

h --2250 $

E 1 f

-2200 2 3 15 - -

g E -

- 215 0 5

• e

! E f 14.5 - --210 0

-2050 14 O 1000 2000 3000 4000 5000 6000 7000 8000 Time (s)

Figure 8-11. Scenario 5 Extrapolated reactor vessel downcomer pressure.

CAUTION: THE SCENARIOS SIMULATED o CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

570 , , , , , , .,

CALCULATED -560 p

7 v

-- EXTRAPOLATED v

e .

.g 560 ,

________________.:- g

% /  %

g -

/ -540 g

& / &

2 550 -

j p- a g

'k -

-520 .E

.= =

3 540 - -

g

- 2

-500 ,

530 ' ' ' ' ' ' '

O 1000 2000 3000 4000- 5000 6000 7000 8000 Time (s)

Figure 8-12.- Scenario 5 Extrapolated reactor vessel downcomer fluid temperature.

78-

o.

O CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, E0VIPMENT FAILURES, OR BOTH.

m 48 , , , , , , ,

CALCULATED

)E -- EXTRAPOLATED N

E j 46 -

!c

\ s

-8000 *$ p Ed

.o

44 -

's s _________________ o=

u :-

=

31 E) y- 0+

-7500 1:3

' E 42 - -

$ :E O

e I ' ' ' ' ' ' '

40 O 1000- 2000 3000 4000 5000. 6000 7000 8000 Time (s)

. Figure 8-13. Scenario 5 Extrapolated reactor vessel downcomer heat transfer coefficient.

'e p

79

9. SCENARIO 6,2-1/2-IN. HOT LEG BREAK AT FULL POWER The following subsections present the transient 9.3.1 Calculation Results. A sequence of events scenario description, modeling changes effected to of the transient is provided in Table 9-2. The small perform this calculation, detailed analysis of the break was assumed to occur at time zero in the bot-transient results, extrapolations and uncertainty tom of the Loop C hot leg. The primary system analyses, and conclusions drawn from the analysis rapidly depressurized as shown in Figure 9-1. Both .

for Scenario 6; a small hot leg break at full power. the proportional and backup heaters were turned on to recover the pressure. Also, as a result of the Scenarios investigated in this report generally in- break, the pressurizer level rapidly decreased, as clude conservative assumptions concerning equip- shown in Figure 9-2, and the makeup flow increased ,

ment failures, operator actions, or combinations of to compensate for the lost liquid inventory.

these. Conclusions relative to pressurized thermal shock severity are not to be drawn directly from the At sl6 s, the reactor tripped on a 2/3 reactor results presented in this report (see Section 18).

overtemperature deita T signal. The turbine stop valves closed and secondary pressures began to 9.1 Scenario Description incr-ase as shown in Figure 9-3. The primary system depressurization rate increased due to the rapid te u n Ne pown and a lag in th response The transient was initiated from full power steady th pdmary to secoMary hat removal rate as state (nominal temperature and pressure), and all ,

. shown in Figure 9-4. The steam dump control control systems were in automatic control. The tran-system changed from load rejection mode to plant sient was imtiated by a 0.0635 m (2.5 in.) dnmeter

, trip control mode at the time of reactor trip to bring break at the bottom of the horizontal section of the the plant average temperature down to 559 K C loop ho:. leg, just upstream of the pressurizer (547*F). The steam dump valves opened and closed surge line connection. It was assumed that all as shown in Figure 9-5, due to a large mismatch systems opeiate automatically as designed. The only between the plant average temperature and the operator actions assumed to take place were: (a) trip temperature setpoint in the plant trip control the reactor coolant pumps when the primary pres-mode. Also as a result of reactor trip, the break sure reached 9.1 MPa (1315 psia)if a SIAS signal mass flow rate increased as shown in Figure 9-6.

was generated, and (b) throttle auxiliary feedwater flow to maintain a 40% narrow range levelin each of the steam generators. A transient scenario is At s27 s the primary system pressure had detailed in Table 9-1. dropped to 11.9 MPa (1730 psia) actuating the SIAS signal. As a consequence the main feedwater valves ci sed, isolating main feedwater from the 9.2 Model Changes steam generator secondaries. The main feedwater pumps tripped on low flow and the heater drain Changes made to the steady state model to initiate flow was terminated. At the termination of the main the small hot leg break included the addition of a feedwater pump power the motor-driven auxiliary break valve in the C loop hot leg connected to a feedwater system was activated and auxiliary feed-time dependant volume set at atmospheric condi- water began flowing into the A and B steam gener-tions. The break components were set to represent ators as shown in Figure 9-7. Due to the use of a a break at the bottom of the hot leg pipe. The valve common header for both the motor and steam-was set to open at the initiation of the transient. driven auxiliary feedwater systems, the differential 7

pressures between the header and the steam 8'"" ' 'S d*'"*i"*d *hich steam generators 9.3 Results received the flow. At the imtiat,on i of the motor-driven auxiliary feedwater flow the C steam This section presents the results, extrapolations, generator pressure was slightly higher than the other and uncertainties of the small hot leg break tran- two, therefore, only the A and B steam generators sient at full power. received auxiliary feed flow.

i 80

Table 9-1. Scenario description No. 6 Plant Initial State - Just prior to transient initiator General

Description:

100% Power steady state 4 System Status

, Turbine: Automatic control Secondary PORV: _ Automatic control Steam Dump Valves: Operative / Automatic control

+ Charging System: Automatic control Pressurizer: Automatic control Engineering Safety Features: Automatic control PORVs: Automatic control Reactor Control: Automatic Main Feedwater: Automatic control Aux Feedwater: Automatic control MSIVs: Open, Automatic control MFIVs: Open, Automatic control Transient Initiator A 2.5-in. hole appears in the hot leg.

2 Equipment Failures That Occur During the Transient if the Equipment is Demanded.

.Nonc j

Operator Reactions to Reported Information

-1. If SlAS signal is generated, the operator will trip the reactor coolant pumps when RCS pressure reaches 1300 psig.

2. The operator will throttle AFW flow to maintain 40% S/G level.

= Primary system pressure dropped below the HPI shown in Figure 9-5. Also by this time, the pres-shutoff head by N40 s and flow from this system surizer was completely empty. At s55 s the primary was established to each cold leg. Also at this time, system pressure had dropped below 9.1 MPa-two out of three steam generators had reached the (1315 psia) and the reactor coolant pumps were low-low level signal setpoint and the steam-driven tripped. As the pumps coasted down, the loop flow -

auxiliary feedwater flow ~ w'as initiated to the transitioned from full flow to natural circulation generators. Only the A and B steam generators as shown in Figure 9-10. Primary system heat

- received the steam-driven auxiliary feed flow at this removal was nearly equal to decay power (see '

o time as shown in Figure 9-8. No auxiliary feed flow Figure 9-4) and the secondary system pressures was delivered to Steam Generator C because of its began to decrease due to the influence of the cold slightly higher pressure. With both the motor and auxiliary feedwater and heat removal through the steam-driven auxiliary feedwater systems operating, steam dump valves. Auxiliary feed flow to the y the level in Steam Generators A and B began to C steam generator was established at this time.

increase as shown in Figure 9-9. Primary system depressurization was slowed as a result of the equalization between decay heat Between 50 and 55 s several events occurred. At generation and removal, and vapor generation in

. s50 s the previously mentioned temperature mis- the reactor vessel upper head as a result of the e match in the plant trip control system had corrected pressure there reaching the saturation pressure of itself and the steam -dump valves opened as the fluid.

81

a-Table 9 2. Scenario _6 sequence of events -

Time (s)~ Event 0.0 - 0.0635 m (0.2083 ft) diameter break appeared in bottom of C loop hot leg. Primary system depressurizes and pressuriz:r heaters energized.

16 Reactor tripped on 2/3 reactor overtemperature AT signal. Turbine stop valves close.

Steam dump system switches to plant trip control mode.

25 . Pressurizer heaters are deenergized due to pressurizer level dropping below 14.4% of measured level.

27 SI signal received on low pressurizer pressure, main feedwater pumps trip, main feed valves close, motor. driven auxiliary feedwater initiated.

40 HPI flow initiated, steam-driven auxiliary feedwater initiated.

50 Pressurizer empty, steam dump valves open.

55 - Reactor coolant pumps trip, vessel upper head begins to void.

'295 Tave dropped below 559 K (547'F), steam dump system switched to steam pressure con-trol mode, steam dump valves close.

400 Hot legs and steam generator tubes void, natural circulation to Loops A and B stopped.

1607- Auxiliary feed flow to Steam Generator C stopped due to 40% narrow range level criteria.

849 Auxiliary feed flow to Steam Generator B stopped due to 40% narrow range level criteria.

~

989 Auxiliary feed flow to Steam Generator A stopped due to 40% narrow range level

. criteria.

1000 Natural circulation to Loop C stopped.

2440 Accumulator flow initiated.

v 2800 Calculation terminated. Reactor vessel downcomer pressure and temperature are

' 3.6 MPa (522 psia), 362 K (193*F) respectively.

e' 7200. End of extrapolation 4

82

. _ _ _ . _ - _ _ _ _ - _ ~

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

20 , , , , ,

e

-2500 m ^

U '

15 - -

y -

-2000 $

5 Pressurizer Empty 5 1500

! 10 - g Upper Head Voided 1 f.

-1000 l 3

~ ~

$ h

-500 0 O O 500 1000 1500 2000 2500 3000 Time (s)

Figure 9-1. Scenario 6 Primary system pressure.

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

0.6 , , , , , , , , ,

7

.~

0.4 - -

3 1

5 4 g 0.2 - -

o z

,9-0.0 O 20 40 60 80 10 0 12 0 14 0 16 0 18 0 200 Time (s)

Figure 9-2. Scenario 6 Normalized pressurizer liquid level.

83

r CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR _ ACTIONS, EQUlPMENT FAILURES, OR BOTH.

7 , ,

0 STEAM GENERATOR A -1000 .

I -

0 STEAM GENERATOR B A STEAM GENERATOR C m

o R ^

)3 .?

b 6 .5 ~ -950 3 ,

! kg% = = = = = = = = = = =! -

-900

{' {

E 6 - -

E 3 s s -

cc c c em -850 j All AFW flow terminated 5.5 I ' ' -800 0 1000 2000 3000 Figure 9-3. Scenario 6 Steam generator secondary pressure.

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQulPMENT FAILURES, OR BOTH.

200 , , ,

O CORE POWER i

o SECONDARY Q 15 0 -

m

• 10 0 -

5

$ 50 - -

=

'==+===c = c-c o m 0 -

+c , c -- c c co - .

-50 ' '

O 1000 2000 3000

• Time (s)

Figure 9-4. Scenario 6 core power versus total primary to secondary heat transf er rate.

Figure 9-4. Scenario 6 Core power versus total primary to secondary heat transfer rate.

84

r-CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

500 i i

-1000 400 - -

m

-800 g

)

[ 300 -

-600 m_

v b $

c c 200 - -

N -

-400 g 1

10 0 _- -

200

' ^

0 0 0 10 0 200 300 Time (s)

Figure 9-5. Scenario 6 Total steam dump valve mass flow rate.

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

300 i i i O BREAK Flow -600 0 INFLOW

-500

?

)

n

{* 200 -

-400 o 5 High Quality Fluid -

• Reaches Break ,

j -

-300 g

100,-

1. Lg u -

-o j  % -200 h

^

I i -'- cg b c c =

On O O 1000 2000 3000 Time (s)

Figure 9-6. Scenario 6 Total break mass flow rate versus total ECC/CVCS mass flow rate.

85

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

60 , , , , , , , , ,

, o STEAW GENERATOR A 12 0

• O STEAM GENERATOR B A STEAM GENERATOR C

-100

^  ?

( -40 - - N .

y-

- ^ C D .go f O

-60 2 m j 20_- * -40 E s

' ' g ^.

f. g

-20

. E ^ E '  :

0: :0 0 10 0 200 300 400 500 600 700 800 900 1000 Time (s)

Figure 9-7. Scenario 6 Motor-driven auxiliary feedwater mass flow rate.

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

60 , , , , , , , i i o STEAM GENERATOR A -12 0 0 STEAW GENERATOR 8 a STEAM GENERATOR C

-10 0

{ 40 y .

o r"

~

g -

-80

)

.o v O 4

-60 y g 20,- l

, , N -40 3 0 .

'2

-20 0- ' ' ' ' ' ^ ' ^ '^ ^ *^ + 1 :0 0 10 0 200 300 400 500 600 700 800 900 1000 Time (s)

Figure 9-8. Scenario 6 Steam-driven auxiliary feedwater mass flow rate.

86

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUlPMENT FAILURES, OR BOTH.

1 i i

-e D STEAM GENERATOR A o STEAM GENERATOR B A STEAM GENERATOR C 0.8 - ,

0.6 -

_ 282 E E 3

-N 8

5 E 0.4 -

o z

0.2 0

O 1000 2000 3000 Time (s)

Figure 9-9. Scenario 6 Normalized steam generator narrow range liquid level.

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

5000 i i 0 LOOP A - 10000 i[ 0 LOOP B 4000 -

A LOOP C -

- 8000 n 2 4 3000 - - N p

v

- C000 j O

$ 2000,- -- 4000 $

=

E a g 1000.- -- 2000 E

.. :s 0--

m-

[ -';"r 72 I kr2 ^, 2 W- --0 i i - -2000

-1000 -

0 1000 2000 3000 Time (s)

Figure 9-10. Scenario 6 Primary cold leg mass flow rate, 87

1 At sl50 s the reactor vessel upper head had com- The cold leg and reactor vessel downcomer fluid pletely voided and voiding in the upper plenum, temperature responses are shown in Figure 9-11.

downcomer above the inlet annulus, and hot legs Between 400 and 1500 s the liquid temperature commenced. Two-phase fluid reached the break and decreased at a faster ratein the A and B loops than the break mass flow drastically dropped as shown in the C loop because those loops has stagnated and in Figure 9-6. Vapor generation in the vessel upper mixing of the warmer loop flow with the incoming plenum and the reduction in the break flow rate cold HPI liquid was minimized. When the C loop resulted in a stabilization of the primary system stagnated at N1000 s, the cold leg temperature in

• pressure for sl50 s as shown in Figure 9-1. that loop decreased at a rate similar to the other loops. The oscillations in the fluid temperature dur-At N300 s, the highest average loop temperature ing this period were due to oscillations in the loop dropped below $59 K (547*F) and the steam dump flow shown in Figure 9-10. The reactor vessel e system control changed from plant trip control downcomer fluid temperature was higher than in mode to steam pressure control mode. The steam the cold legs due to heat addition to the cold leg dump valves closed, and secondary system depres- fluid as it entered the steam filled environment in surization was continued solely due to cooldown the downcomer, downcomer wall heat transfer, and effects from the addition of cold auxiliary feed warm fluid entering the downcomer through the flow. Between 300 and 1000 s vapor generation and vessel internal leakage paths.

condensation effects resulted in perturbations in the primary system pressure response, shown in When the C loop hot leg had drained at 1000 s Figure 9-1, as the loops and upper plenum drain- high quality fluid was at the break, resulting in the ed. Two-phase flow periodically reaching the break sharp decrease in the break mass flow observed in also contributed to the pressure oscillations. Also Figure 9-6. The total primary inflow from the HPI contributing to the primary system pressure and makeup after this time equaled or exceeded the response during this time, was the primary-to- break mass flow as shown in Figure 9-6. The secondary heat removal rate shown in Figure 9-4. primary system depressurization was stopped and The heat removal rate was affected by the amount a slow repressurization began due mainly to ter-of auxiliary feedwater that the steam generators mination of auxiliary feedwater. The repressuriza-were receiving. The common header for each aux- tion continued until the core subcooled, reducing iliary feedwater that (steam and motor-driven) was the vapor generation rate. From 1800 s to the ter-modeled and the flow to each generator was deter- mination of the calculation, primary system mined by the differential pressure between the pressure decreased as shown in Figure 9-1 until the header and the generator. At times one generator energy removal at the break balanced with the receised all of the flow as shown in Figures 9-7 energy addition, and break flow equaled total ECC and 9-8. As specified in the scenario, the operator flow.

controlled the secondary level around the 40% nar-row range level. At s600 s the levelin the C steam Between 1800 and 2200 s flow oscillations in the generator reached the 40% level and auxiliary feed A loop allowed more mixing of the warm loop fluid flow to that generator was terminated. At N849 and and the cold ECC fluid, resulting in an increase in 989 s respectively, auxiliary feed flow to the the cold leg fluid temperature as shown in Fig-A and B steam generators was terminated. The ter- ure 9-11. At s2160 s a flow reversal in all three mination of auxiliary feedwater to each steam loops occurred and the temperature in all three cold generator stopped the cooldown and depressuriza- legs increased due to warmer liquid from the down-tion of that secondary as shown in Figure 9-3. comer entering the cold legs. After 2200 s the oscillations in the A loop subsided somewhat and By s400 s enough voiding in the A and B loops the temperature in that loop decreased.

had occurred that natural circulation had stopped and these loops were very nearly stagnant through- At s2440 s the primary system pressure dropped out the remainder of the calculation as shown in below 4.6 MPa (673 psia) and accumulator injec-Figure 9-10. The C loop, however continued to cir- tion began. With accumulator flow, the total mass ,

culate due to the effects of the break, until s1000 s. inflow to the primary system and the break mass At this time the liquid in the loop C steam generator flow were nearly equal as sho a n in Figure 9-6. The tubes and hot leg had drained and natural circula- calculation was terminated at 2800 s when the tion was lost, primary system was slowly refilling. The prin ary 88

7_ .

CAUTION: THE SCENARIOS SlMULATED CONTAIN SIGNiflCANT CONSERVATISMS IN OPERATOR ACTIONS, EQUlPMENT FAILURES, OR BOTH.

600 , .

-600 0 LOOP A 7

v i o LOOP B a LOOP C Q

v

• x DOWNCOMER

-500 e I

E 500 - -

2 l

E -

-400 1

! n . -

5 2 -

-300 2 400 - -

E - Mf -200 E 2 3 S $

8 -10 0 300 O 1000 2000 3000 Time (s)

Figure 9-11. Scenario 6 Primary cold leg fluid temperatures and vessel downcomer fluid temperatures.

system pressure at the termination of the calcula- through the break or steam generators was balanced tion was 3.6 MPa (522 psia) and decreasing, and with the core decay power. It was estimated this the reactor vessel downcomer fluid temperature was steady state would occur around a pressure of 363 K (193*F) and decreasing. 0.9 MPa (142 psia), which is just below the LPI shutoff head, and would occur at s3200 s. The addition of LPI fluid appeared to be sufficient to 9.3.2 Extrapolations and Uncertainties. This sec-hold the pressure at .9 MPa (142 psia) through the tion presents the extrapolations of the vessel 2-h pedod of m conmn.

downcomer pressure, fluid temperature and wall inside surface heat transfer coefficient. Also, Figure 9-13 shows the extrapolation of the reac-extrapolations of the cold leg flow rates and fluid tor vessel downcomer temperature. Loop mass temperatures are shown. Any known uncertainties flows at the end of the calculation were nearly equal in the calculation are also addressed.

to the ECC flow rate (essentially stagnant), and the downcomer temperature was approaching the An extrapolation to 2-h of the pressure, temperature of the ECC fluid temperature of 305 K temperature and heat transfer coefficient curves in (90*F). It is estimated the reactor vessel downcomer the reactor vessel downcomer and the mass flow temperature will be dominated by the ECC fluid

. rate and temperature curves in the cold legs are temperature because of loop flow stagnation, and shown in Figures 9-12 through 9-16. Parameters in will decrease until it approaches the temperature of

o. the downcomer are shown for the elevation adja- the ECC fluid. Mixing of the cold leg fluid coming cent to the top of the core. At the termination of from the upper head and plenum leakage paths will the calculation, heat removal from the primary to keep the reactor vessel downcomer temperature the secondary was nearly zero and the primary slightly above the cold leg temperature. The L , system was gaining very little mass due to the total estimated reactor vessel downcomer temperature at i ECC mass inflow (including accumulator flow) 2-h is 310 K (100*F).

equaling the break mass flow rate. It is estimated the primary system would continue to depressurize Figure 9-14 shows the extrapolated reactor vessel as shown in Figure 9-12, until the break flow and downcomer inner surface wall heat transfer coeffi-total ECC inflow were balanced and energy removal cient. After reactor coolant pump trip occurred, the 89

' CAUT ION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

20 , , , i i i i CALCULATED

-- EXTRAPOLATED -2500 n ^

g 15 - -

3 -

-2000 3 *

' b 3 3 l jo~ - _ _-1500 g i 1

-E 1

-1000 2

l o 5 - -

o

-500

\

\

0 O O 1000 2000 3000 4000 5000 6000 7000 8000 Time (s)

Figure 9-12 Scenario 6 Extrapolated reactor vessel downcomer pressure a :un elevation equal to the top of the core.

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

600 i i i i , i i

-600 CALCULATED Q

v

-- EXTRAPOLATED 7 v

-500

• 3 2 E 500 - -

E E. -

g -400 E.

I bl' E 2

2 .

300 3 Pr Er

= 400 - -

= .

e e g -

-200 E 2- \ 2

> \' > ,

, , , " - - - - - - T - - - - r - - -- 1_ -10 0 0 1000 2000 3000 4000 5000 6000 7000 8000 Time (s)

Figure 9-13. Scenario 6 Extrapolated reactor vessel downcomer fluid temperature at elevation equal to the top of the core.

90 1.

C AUTI ON: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

10 , , , , , , ,

0 -- CALCULATED Y --

EXTRAPOLATED "E -

-1500

) 8 -

T*

6 .

c hE 1 6 -

- *~

.o

-1000 0C 0 ul

" 2.R 4 -

E3

% ES

} -

-500 $

? 2 -

?! -

?

o ._-----______________ 1 0 ' ' ' ' ' ' ' O O 1000 2000 3000 4000 5000 6000 7000 8000 Time (s) Figure 9-14. Scenario 6 Extrapolated reactor vessel downcomer wall surface heat transfer coefficient at elevation equal to the top of the core. CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNiflCANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 600 g i , , , i i , O LOOP A o LOOP B _ a LOOP C -1000

                                                                                            -- EXTRAPOLATED j

400 - - m 7 d"

                                                                                                                                <E D

x J '

                                                                                                                                .o
                      ~
                               '         I 200      -l
          =                                                                                                                     e C

h } E 2 _ l gl g k------------------------ _, S 0 o

            -200 O      1000                    2000      ~3000     4000      5000        6000     7000     8000 Time (s)

Figure 915. Scenario 6 Extrapolated cold leg mass flow rates. 91

P CAUTION: THE SCENARIOS SlWULATED CONTAIN SIGNIflCANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 600, , i i i i ' '

                                                                                                                    -600 0 LOOP A Q

l O LOOP 8 A LOOP C 7 v

• I -- EXTRAPOLATED e h 500 - -

s

                                                                                                                    -400 g i                 i     .

t

a. a.

• E E I 400 -

i 2 2 - 1

                                                                                                                    -200 2 e
              =-

g = g300 - - - - - - - - - - - - - - - - - - - - - - - g 3 3 j -

                                                                                                                    -0   y 200 O    1000          2000        3000   4000          5000           6000           7000  8000 Time (s)

Figure 9-16. Scenario 6 Extrapolated cold leg fluid temperature. wall to fluid heat transfer regime changed to a natural tures will decrease and approach the ECC fluid circulation / pool boiling mode as the loop flow tran- temperature of 305 K (90'F). sitioned from forced flow, to natural circulation, to . stagnation. After 2800 s it is estimated the wall heat There were no significant uncertainties that af-transfer coefficient will not change significantly. fected the overall results of the calculation. Figures 9-15 and 9-16 show the mass flow rate 9.4 Conclusions and fluid temperature extrapolations for each of the three cold legs. At the termination of the calcula- As a consequence of the size of the hot leg break tion (2800 s), loop flow was stagnant because the it was shown.that the break was capable of remov-steam generator tubes and hot legs were voided. At ing core decay power. Therefore there was no

 - 2800 s the steam generators were heat sources                       mechanism that would allow the primary system to because the fluid temperatures were slightly higher                  repressurize. It was estimated that primary system than the primary, and the break flow essentially                    -pressure would stabilize at @.9 MPa (142 psia).

equaled the total ECC inflow, thus no primary side Also, because of primary mass inventory loss, natural fluid volume gain. It was estimated these conditions circulation was lost and the minimum reactor vessel will exist throughout the 2-h period, therefore, loop downcomer temperature was estimated to be 310 K flow will remain stagnant. The cold leg tempera- (100*F), slightly above the ECC fluid temperature. b 92

l

10. SCENARIO 7, STUCK-OPEN PRESSURIZER PORV AT FULL POWER The following subsections describe the investiga- 10.3.1 Calculation Results. The sequence of tion of Scenario 7. This calculation was performed events for this transient is presented in Table 10-2.

to evaluate the consequences of a stuck open The transient was initiated by opening one primary

• primary PORV with the reactor at full power

  .                                                             PORV. Primary system pressure, as shown in Fig-conditions.                                                ure 10-1, began to fall as mass was lost out of the PORV. However, almost all other parameters in the A description of the scenario, the results of the       plant remained essentially unchanged until 33.4 s, e    calculation, tl.e extrapolated results, and the uncer-    at which time the reactor was tripped. The reactor tainties associated with the calculation are described    tripped on a two-out-of-three loop high reactor herein, along with the conclusions regarding the          delta temperature. The reference delta temperature calculation,                                               used to determine the overtemperature was based on the primary system pressure, so as the primary Scenarios investigated in this report generally        system pressure dropped, the reference delta include conservative assumptions concerning equip-        temperature dropped. Therefore, the over delta ment failures, operator actions, or combinations of       temperature was reached without the primary these. Conclusions rela;ive to pressurized thermal        system temperatures changing appreciably.

shock severity are not to be drawn directly from the results presented in this report (see Section 18). Following reactor trip, the primary system pressure dropped more rapidly as the primary 10.1 Scenario Description SY5* C l'd. H t and c Id les fluid temperatures are shown m. Figure 10-2. By 45 s, the turbme isolation valve closed, and the steam dump valves Table 10-1 gives the description of the desired opened on a load rejection signal, and stayed open scenario as supplied by ORNL. The transient was for 3 s. The feedwater valves also began to close initiated from full power steady state conditions by following reactor trip and by 52 s were completely opening one primary power operated relief valve closed. The main feedwater pumps tripped off and (PORV). The PORV was closed 10 min into the both motor and turbine-driven auxiliary feedwater transient to simulate an operator closing a block pumps began delivering feedwater at 53 s. At 72 s valve. It was assumed this block valve would pre- the steam dump valves modulated open and brought vent any further flow through either PORV. All the primary system average temperature down to other control systems operated in their automatic 560 K (547 F). modes. At 113 s, primary system pressure dropped below 0.1 MPa (1470 psia), and HPI flow was initiated. 10.2 Model Changes At 139 s, primary system pressure dropped below No changes in the thermal-hydraulic model were 9.07 MPa (1315 psia), and the primary coolant necessary. The only change in the control system pumps were tripped. Hot leg temperatures then model was in the trips that controlled the primary increased and cold Icg temperatures decreased as the PORVs. The transient was initiated from the full- loop flow rates dropped. By N300 s, a stable natural power steady state conditions presented in circulation flow was established in all three loops o Subsection 2.3.1. as energy was added to the primary fluid from decay heat and removed through the steam generators. The cold leg flow rates for the three loops are shown 10.3 ResultS m Figure 10-3. By 180 s, the primary system pres-

.                                                              sure had dropped to the saturation pressure of the This section is divided into two subsections. The       fluid in the reactor vessel upper head and the fluid first presents the calculated results, and the second      there began to flash. The void fraction within the extrapolates the calculated results to 2-h, and            reactor vessel upper head and in the volume at the discusses uncertainties associated with the calcula-       top of the reactor vessel downcomer are shown in tion and extrapolation.                                    Figure 10-4. As liquid was forced out of the upper 93

r Table 10-1. Scenario description No. 7 Plant Initial State - Just prior to transient initiator General

Description:

100% Power steady state System Status Turbine: Automatic control - Secondary PORV: Automatic control Steam Dump Valves: Operating / automatic control Charging System: Automatic control Pressurizer: Automatic control . Engineering Safety Features: Automatic control PORVs: Automatic control Reactor Control: Automatic Main Feedwater: Automatic control Aux Feedwater: Automatic control MSIVs: Open, Automatic control MFIVs: Opening, Automatic control Transient Initiator Pressurizer PORY transfers full open. Equipment Failures That Occur During the Transient if the Equipment is Demanded. PORY blocking valve will not close until 10 min into transient. Operator Reactions to Reported Information

1. The operator shuts PORV blocking valve at 10 min into transient.

2. If SIAS signal is generated, the operator will trip the reactor coolant pumps when RCS pressure reaches 1300 psig.

head when the bubble there grew, liquid surged into after llPI initiation. At N300 s, the volumetric the pressurizer rapidly increasing the pressurizer injection of the HPI system exceeded the volume level, as shown in Figure 10-5. The pressurizer level lost out the PORV, and the primary system began exceeded its indicated range at 278 s, although the to repressurize. Repressurization was very slow at pressurizer was not completely full. Some voids also first. At 600 s, the PORV was closed simulating the formed in the volume at the top of the reactor vessel closure of a block valve by an operator. The rate downcomer, as seen in Figurc 10-4. of repressurization increased at that time, but was still fairly slow as the voids in the reactor vessel At 272 s, the steam dump valve control shifted upper head and downcomer collapsed. to the steam pressure control (SPC) mode, and the , dump valves closed. Secondary pressures, shown in During the early stages of this transient, the liquid Figure 10-6, began to rise but there was no signifi- levels in the steam generators were below the eleva-cant effect on primary conditions because the cold tions of the feedwater rings. As a result, the feed-auxiliary feedwater removed sufficient energy from water was injected into steam filled volumes. At , the primary fluid without having to draw steam off 620 s, the temperature of the feedwater entering the steam generators. Steam Generator C had decreased to the point where it induced an increase in the condensation in The total IIPI flow rates, and the flow rate out that generator, and dropped the pressure slightly. the PORV are shown in Figure 10-7. The total HPI This drop in pressure diverted all the AFW flow flow rate exceeded the flow out the PORV very soon from the other two steam generators and delivered 94

Table 10-2. Scenario 7 sequence of events

Time (s) Event 0.0 PORV opened o.

33.4 Reactor tripped on 2/3 reactor over delta temperature; turbine stop valve closed; feed-water valves began to close. 34 Steam dump valve opened 37 Steam dump valve c!osed 52 Feedwater valves closed 53 MFW pumps tripped

           .72      Steam dump valve opened 113      HPI flow initiated on low primary system pressure 139      Reactor coolant pumps tripped on low primary system pressure 180      Reactor vessel upper head fluid saturated and began to flash 272       Steam dump valve control shifted to steam pressure control (SPC) mode, dump valve closed 278       Normalized pressurizer level went off scale high 600       PORV block valve closed 620       AFW flow started to SGC 674       AFW flow isolated from SGC, diverted to SGB 706      AFW flow isolated from SGB, diverted to SGA 848       AFW flow isolated from SGA, all AFW flow stopped 947      Minimum reactor vessel downcomer temperature,538 K (509'F), reached 1490-      Reactor vessel upper head refilled with liquid 1506       Primary system pressure rose above HPI shutoff head, HPI flow stopped 1840      Steam dump valves opened
   'O 2088       Pressurizer safety relief valve opened, then immediately closed 2200      End of calculation 7200      End of extrapolation 95

CAUT10N: THE' SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN , OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH, 20 i i i. 5 i i i i i i

                                                                                                                            -2500   ^

m 0 2 v I - v E 15 h-E

                                 +-Reactor trip                                                                             -2000

{ N

               ?
             'a-                                                                                                                     o.

E go -, _ _-1500 E s a 4 Primary

                                                                                                                             -1000 PORV                                   system
                                      ,      ,        c,l osed,         ,       ,      ,     ]iquidfil}ed O'.200 400 '600 800 1000 1200.1400 1600 1800 2000 2200 Time (s)

Figure 10-1. Scenario 7 Primary system pressure. CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 620 ; .i i i i i i i i i i _650 0 LOOP A HOT LEG ,, g v o LOOP A COLD LEG p A LOCP B COLD' LEG v 600 - x LOOP C COLD LEG

                                                                                                                           ~

u 3 3 g  ? -600 Eu

             .u.

E 580 - ~ E e 2 E 2 E 560 id 550 3 f~ $ [ h540 -

                                                  - ~

j -

                                                                                                                              -500 y 520 0 200 400 600 800 1000 1200 1400 1600 1800 2000 2200 Time '(s)

Figure 10-2. Scenario 7 Hot and cold leg temperatures.

e CAUT ION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS .lN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 300 ip , , , , , , , , , O LOOP A

   '..,                              _                                                                                                                        o LOOP R                                -600 a LOOP C 250     -                                                                                                                                                            -

j {m -

                                                                                                                                                                                                      -500 E N

6 2 v

                         $200          -                                                                                                                                                            -

C .2

                                                                                                                                                                                                      -400 E                                                                                     ,                                                                                                      E 150    -                                                                                                                                                            -
                                     -                                                                                                                                                       I M! 300 a

l 10 0  ! 0 200 -400 600 800 1000 1200 1400 1600 1800 2000 2200 Time (s) f Figure 10-3. Scenario 7 Cold leg mass flow rates. l CAUT ION: THE SCENARIOS SIMULATED CONTAIN SIGNIFlCANT CONSERVATISMS IN L OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 1 i A10 C iC i i i i i i O UPPER HEAD 0 00WNCOWER , c

                            .3                                    h O                                                                                                                                                                                                      r M

! 3 0.5 - - ! E g a 8- .

    -0                       >

l 0 O nec 'O ' ' ' 3 0 'O c'c d - 'M)--oL-o-o 0 200 400 600 800 1000 1200 1400 1600 1800 2000 2200  ! Time (s) ? Figure 10-4. Scenario 7 Reactor vessel upper head and top of downcomer void fractions. 97

CAUT ION: THE. SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 1 i r i e i i i i i i i

                   ~

• g 0.8 - -

3

• 0.6 - -

I 6 a E o 0.4 - - z 0.2 O 200 400 600 800 1000 1200 1400 1600 1800 2000 2200 Time (s) Figure 10-5. Scenario 7 Normalized pressurizer level. CAUil0N: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 7.5 , i i i i , i , i i n 7 - - n o -

                                                                                                                                    ,                                     -1000 2 I

v v E y E 6.5 - 3 3

                ,I     -
                                                                                                                                                                          -900              ,E e    6   -                                                                                                                                          -

E O STEAM GENERATOR A E 2 o STEAM CENERATOR B 3

               $ 5.5 iF; ; ;                    ~2     0 0 ;                                                                          ^ ""                           --800                 $

5 0 200 400 600 800 1000 1200 1400 1600 1800 2000 2200 Time (s) Figure 10-6. Scenario 7 Steam generator pressures. 98

C AUT ION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 40 i i i i i i i i i i _ 0 TOTAL HPl -80 o PORV 30 - b N - o km -

                                                                                                         -60 E

6 :9 v

             $ 20                                                                                      -

s

             =     -
                       ,                                                                                 -40     g E                   ]                                                                                $

2 -NQ ( s 10 r \ --20 0 0 ue 'o-d - h - n La__ul-o - u 0 0 200 400 600 800 1000 1200 1400 1600 1800 2000 2200 Time (s) Figure 10-7 Scenario 7 Total Hpl flow and flow out pORV. it all to Steam Generator C, as shown in Fig- After the isolation of the AFW system, the heat ure 10-8. The steam generator narrow range levels sink for the primary system was lost, and the are presented in Figure 10-9. At 674 s, the liquid primary system began heating up. The primary level in Steam Generator C reached the 40% con- system was also being filled by the liPI flow. At trol point, and feed flow to that generator was 934 s, the pressurizer went liquid full, and at 1490 s, isolated. The AFW then diverted entirely to Steam the upper head of the reactor vessel completely filled Generator II. The feed flow to that generator was with liquid. Primary system pressure rose rapidly isolated at 706 s as the liquid level exceeded 40%, after this time. Primary system pressure rose above Steam Generator A then received all the feedwater the llPI shutoff head of 10.1 h1Pa (1470 psia) at until 848 s w hen its liquid level rose to 40% and all 1506 s, and ilPI flow stopped. The pressure con-AFW ficw was isolated. tinued to increase because the primary fluid was swelling as it heated up, and because the CVCS was The elfects of these shifts in AFW flow are the still injecting water at its minimum rate, tempora:y drops in the cold leg temperatures bet- 9.2 x 10-4 m3/s (15 gpm). ween 60f and 1200 s, as seen in Figure 10-2. As the AFW 11ew shifted to a specific steam generator, the The primary and secondary systems were closely energy removed by that generator increased, while coupled throughout this transient. At 1840 s, e the energy removed by the other two generators secondary pressures had risen to 7.03 h1Pa decreased. As a result, the cold leg temperature in (1020 psia), and the steam dump valves opened to the loop receiving all the feedwater dropped below hold the pressure at that point. This pressure cor-the temperature in the other two loops. After the responds to a saturation temperature of 559 K o AFW was isolated to a given steam generator, the (547'F). Therefore, the steam dump valves operated cold leg temperature in tnat loop began to rise, to control the primary and secondary fluid After all AFW flow was isolated, all three cold legs temperatures close to $59 K (547'F) through the temperatures converged. The shifts in AFW flow end of the calculation. Primary system pressure rose did not significantly affect the primary system high enough to open the primary safety relief valve pressnre, once, very briefly, at 2088 s. The primary system 99

p l-i CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 100 , i. , ,_ i , , , , i

- 90 -- a STEAM GENERATOR A --200.

O STEAM GENERATOR B 80 4; a STEAM GENERATOR C _ a o .. y x 70_- -_15 0 7 60 - - E

                  .a                                                                                                                                                                                         o .

50 - - G

                  .{    40
                                                                                                                                                                                                    -10 0    y o
                                                                                                                                                                                                             =

l 30 -

                                                        ,g
                                                                                                                                                                                                    -50 g

j 20 ~ - -

                         .10   -                                                                                                                                                                  -

0--  :.

: : : : : : = = = = ::0

                       -10 0 200 400 600 800 1000 1200 1400 1600 1800 2000 2200 Time (s)                                                                                                                     ;

1 Figure 10-8. Scenario 7 Feedwater flow rates. CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 1 i i i i i i i i i i a STEAM GENERATOR A o- STEAM GENERATOR B a STEAM GENERATOR C _ _e_ 3 v

                       '3      0.6      -

cr . 3 n

                       =                                                                       m                      . .

E l

                       .5      0.4      -

6 - o l E O

                      .g.                                                                                                                                                                                        *
                                                                           ~

Z 0.2 - - l 0 O 200 400 600 800 1000 1200 1400 1600 1800 2000 2200

                                                                                              - Time (s)

Figure 10-9. Scenario 7 Steam generator normalized liquid levels. 100

pressure then stabilized at just under the safety relief through 7200 s. The heat transfer coefficient in the valve setpoint,17.5 h1Pa (2535 psia), through the downcomer, Figure 10-13, is a strong function of end of the calculation. At the end of the calcula- the mass flow rate, and would also remain essen-tion, 2200 s, both downcomer pressure and tially constant through 7200 s. temperature were stable at 17.5 h1Pa (2535 psia), and 560 K ($48'F), respectively. There were no major uncertainties associated with this calculation. e 10.3.2 Extrapolations and Uncertainties. Figures 10-10 and 10-11 show the downcomer pressure and temperature extrapolated to 7200 s. 10.4 Conclusions Steam dump valve capacity was more than suffi- _, cient to remove decay heat, so the primary system This transient was mild primarily for two reasons, temperatures would stay very close to 560 K (548*F) First, the open PORV was not large enough to void out through 7200 s. Because mass is still being the primary system and stagnate the loop flows. added to the primary system by the CVCS, the Secondly, the energy removed by the PORV and safety relief valve would occasionally open to relieve AFW flow was not sufficient to cause a serious pressure, and the primary system would remain at cooldown. The minimum reactor vessel downcomer the safety relief valve setpoint, 17.5 h1Pa temperature reached was 538 K (509'F), and occur-(2535 psia), out through 7200 s. Figure 10-12 red at 947 s. The maximum primary system pressure presents the cold leg mass flow rate extrapolated after this time was 17.5 h1Pa (2535 psia). After to 7200 s. All three loops would act syrnmetrically, extrapolating the results out to 7200 s, the reactor and since a stable natural circulation flow had been vessel downcomer pressure and temperature would established, the flow rates would stay essentially be 17.5 h1Pa (2538 psia), and 560 K (548'F), constant from the end of the calculation out respectively. CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 20 i , i i i i i

                                                                                                                           -2500 o                                                                                                                      E G-                                                                                                                      %

3 15 - E -

                                                                                                                           -2000 0 3                                                                                                                       3 U                                                                                                                       E u                                                                                                                       u
             "                                                                                       CALCULATED                      Q-o                                                                                 -- EXTRAPOLATED                       *
                                                                                                                           -1500 E 10 _                                                                                                      _

E 3 ' 3

                                                                                                                           -1000 5
   .                0      1000     2000       3000               4000               5000          6000   7000      8000 Time (s)

Figure 10-10. Scenario 7 Extrapolated reactor vessel downcomer pressure. I 101

i .:. - CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR' ACTIONS, EQUIPMENT ~ FAILURES, OR BOTH. 570 . . ., i i i i

                             .                                                                       CALCULATED          -560 7

v-

                                                                                               -- EXTRAPOLATED                 7 y

o' , s_560 ----_---_---_----_-----. - g 3 3

              ~5-             -
                                                                                                                          -540 g    *

o. a.

E E I550L t 3 3 5 -

                                                                                                                          -520 ?r
                =                                                                                                               =
              'g540
                                                                                                                               ,g 3-                                                                                                              3
                                                                                                                          -500 530 0         '1000      2000       3000          4000        5000        6000    7000    8000
                 ,.                                                        Time (s)

Figure 10-11. Scenario 7 Extrapolated reactor vessel downcomer temperature.

                                                   ' CAUT ION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

l3001 .. i , i i ., i CALCULATED

                                                                                                                         -600
                                                                                               -- EXTRAPOLATED
                    '250        -                                                                                     -
                                                                                                                         -500
                  - 20'0l       -

7 -

                                                                                                                         -400  .o v                                                                                                                    O
          .3.         150       -
                                                                                                                         -300 J

_ j . ~ 100 _- r

                                                                                                                      --200'f

• 50 _ -- --10 0 O

                        'O 0            1000      2000       3000          4000        5000       6000     7000    8000 Time (s)

Figure 10-12. Scenario 7 Extrapolated cold leg mass flow rates. 102

4-

   *-                                  CAUT ION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN

_ OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

_ 10 - , , , , , , ,

        'u  [l        _

CALCULATED

                                                                                                -- EXTRAPOLATED
                                                                                                                     -1500
           -N     '8     -                                                                                         -

e [ w-U

              ~
                                                                                                                           ' 5Y E  _s
                         -                                                                                         -           8,1 g       -
                                                                                                                     -1000     u:

5 L

 ~

4 - _

               ;                              _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .                                As
                                                                                                                     -500 C                                                                                                               D 2

E -

               .o E                  '         '               '                 '                '        '      '

O O 0= -1000 2000 3000 4000 5000 .6000 7000 8000 Time (s) Figure 10-13. Scenario 7 Extrapolated reactor vessel downcomer heat transfer coefficient. O: a 103

11. SCENARIO 8,2-1/2-IN. HOT LEG BREAK AT HOT STANDBY The following subsections describe the investiga- extrapolates the calculated results to two hours, and tion of Scenario 8. This calculation was performed discusses uncertainties associated with the calcula-to evaluate the consequences of a postulated small tion and extrapolation.

hot leg break with the reactor at hot standby conditions. 11.3.1 Calculation Results. The calculated sequence of events for this scenario is presented in A description of the scenario, followed by a Table 11-2. The transient started when the break discussion of the model changes required to perform opened at 0.0 s. The backup and proportional the calculation are presented herein. The results of heaters were energized within I s and were tripped the calculation, the extrapolated results, the uncer- at 6 s because oflow pressurizer level. The makeup a tainties with the calculation and the conclusions pumps delivered maximum flow after 7 s. A safety regarding the calculation are also presented. injection actuation signal (SIAS) was generated at 21 s when the pressurizer pressure dropped to Scenarios investigated in this report generally in- 11.93 h1Pa (1730 psia). The SIAS caused the fol-clude conservative assumptions concerning equip- lowing actions: the HPI pumps started, the ment failures, operator actions, or combinations of operating main feedwater pump tripped, the main these. Conclusions relative to pressurized thermal feedwater bypass valves closed, motor-driven AFW shock severity are not to be drawn directly from the flow started, and the letdown isolation valve closed. results presented in this report (see Section 18). The reactor coolant pumps were tripped at 32 s to simulate an operator action taken when pressurizer pressure dropped to 9.07 51Pa (1315 psia). 11.1 Scenario Description Calculated pressurizer pressure is shown in The description for this scenario, as defined by Figure Il-1. Flow out the break caused the pressure ORNL, appears in Table 11-1. The transient was to decrease rapidly. HPI was delivered to the cold assumed to be initiated by a small hot leg break legs after 29 s, when the pressure dropped below while the reactor was at hot standby. The break was 10.1 h1Pa (1470 psia), the shutoff head of the HPI located at the bottom of the C hot leg, near the con- pumps. The depressurization rate slowed at 53 s nection to the pressurizer surge line, and had a w hen flashing began in the steam generator U-tubes. diameter of 0.0635 m (2.5 in.). The operators were The steam generator secondaries began acting as assumed to trip the reactor coolant pumps when the heat sources at 57 s. The pressure remained nearly primary coolant pressure dropped below 9.07 h1Pa constant between 57 s and 370 s because of heat (1315 psia). The operators were also assumed to transfer, first from the steam generators and then throttle auxiliary feedwater (AFW) flow to main- later from the core. The heat transfer produced tain steam generator narrow range levels at 40%. enough steam to compensate for the volumetric flow out the break. The steam production in the "e dr pped near 370 s because of an increase in 11.2 Model Changes flow mto the reactor vessel that caused the core to subcool. Without the steam production in the core The model used to perform this calculation was to maintain the pressure, the break was able to essentially the same as that described in Sub- depressurize the reactor coolant system. The reac-sections 2.1, 2.2, and 2.3.3 except for nodalization tor coolant pressure dropped to the accumulator changes required to represent a small hot leg break. pressure of 4.65 h1Pa (675 psia) at INO s, initiating A junction (597) and time-dependent volume (599) accumulator flow. The pressure remained nearly a were added to the model to simulate the break and constant after 1650 s when the reactor coolant containment, respectively. The break was attached to system was refilled to the point where liquid the bottom of Volume 4m03 (refer to Figure 2-1). droplets entered the U-tubes. The subsequent heat transfer from the hot steam generator secondaries

• St pped the depressurization of the reactor coolant 11.3 ResultS system.

This section is divided into two subsections. The Figure Il-2 shows the break mass flow rate and first presents the calculated results, and the second the sum of the HPl and makeup flow rates. The 104

Table 111. Scenario description No.8 Plant Initial State - Just prior to transient initiator General

Description:

Hot OTo Power, OTo Power after 100 h of shutdown System Status

• Turbine: Not latched, TSVs closed Secondary PORV: Automatic control Steam Dump Valves: Automatic control Charging System: Automatic control

 -o            Pressurizer: Automatic control
             - Engineering Safety Features: Automatic control PORVs: Automatic control Reactor Control: Manual Main Feedwater: In bypass mode, manual control to provide zero power level ia Gs; I condensate pump,1 MFWP operating Aux Feedwater: Automatic control MSIVs: Open, Automatic control MFIVs: Closed, Automatic control Transient Initiator A 2.5 in. hole appears in the hot leg.

Equipment Failures That Occur During the Transient if the Equipment is Demanded. None Operator Reactions to Reported Information

1. If SIAS signal is generated, the operator will trip the reactor coolant pumps when RCS pressure reaches 1300 psig.

2. The operator will throttle AFW flow to maintain 400/o S/G level.

break flow rate decreased during the first 53 s of pressure decreased. By the end of the calculation, the transient and then reached a plateau, similar to the sum of the HPI and makeup flows approxi-the reactor coolant pressure curve that was mately equaled the break flow. described previously. The break flow then remained 4 relatively constant until 900 s when the break Calculated accumulator liquid volumes are shown partially uncovered. The break was generally in Figure Il-4. Flow from the accumulators began covered with liquid before 900 s, as illustrated by at 1040 s when the reactor coolant pressure dropped

     - Figure 11-3 which shows the void fraction of the          below the accumulator pressure of 4.65 MPa e   fluid exiting the break. A two-phase mixture flowed        (675 psia). The accumulators retained about 2087o through the break between 900 s and 1300 s. The            of their initial liquid volume at the end of the break was usually covered with liquid again after         calculation. The average accumulator flow after 1300 s. The partial uncovering of the break at 900 s      1040 s slightly exceeded the sum of the HPI and
  ,-  caused a sharp reduction in break mass flow rate,          makeup flows. Before accumulator injection began, The mass flow rate tended to increase after 1300 s         the break flow generally exceeded the sum of the when the break recovered. Although the calculated          HPI and makeup flows, causing the void in the flow rate' and void fraction at the break were             reactor coolant system to increase. After accumu-relatively noisy after 900 s, the trends were              lator injection began, the total injection flow reasonable. HPI flow was initiated at 29 s and            generally exceeded the break flow, refilling the increased during the transient as the reactor coolant      reactor coolant system.

105

Table 11-2. Scenario 8 sequence of events Time -

   -(s)                                                   Event 0.0        Break opened I          Heaters on -

6' Heaters off 7 ' Maximum makeup flow 17 Pressurizer emptied 21 SIAS; MFW pump tripped; bypass valves closed; AFW initiated, letdown isolated; HPI pumps started 29- HPI initiated 32 RCPs tripped 53 Flashing in U-tubes 57 Steam generator secondaries acting as heat source 100 Voiding begins in upper head 125 Voiding begins in core 300 Upper head completely voided 400 Core subcooled 470 AFW off 800 U-tubes voided 1040 Accumulator flow initiated 1300 Break recovered 1650 Vessel refilled; U-tubes refilling 1740 Calculation terminated 7200 End of extrapolation e The voiding and refilling of the reactor coolant the core at 125 s due to a reduction in flow follow-system are illustrated by Figure 11-5, which shows ing the reactor coolant pump trip at 32 s. An void fraction in the upper head and near the center increase in core inlet flow at 370 s brought sub-of the core. Voids first appeared in the reactor cooled liquid into the core, causing the voids in the . coolant system in the steam generator U-tubes at core to disappear by 400 s. The core was subcooled - 53 s. The upper head began draining at 100 s and for the remainder of the calculation. The steam was completely voided by 300 s. Voids formed in generator U-tubes were voided completely by 800 s, 106

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 20 , , .,

                                                                                                    -2500 15                                                                                    -

g -

                                                                                                    -2000 g gg                                                                                                    ;

v m o. 10 -- Flashing _-1500 . R 5

                                                                                                    -1000 Heat transfer 5   -

from steam

                                                                                                 ~
              .             Core Subcooling                                  generators             -500 h

0 O O 500 1000 1500 2000 Time (s)

                                                                                                               ~

Figure 11 1. Scenario 8 Pressurizer pressure. CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 400 , , , O HPI + MAKEUP -800 0 BREAK 300 - ( m

                                                                                                      -600  $

en E i 45 2 v t

       ! 200        -                                                                               -
                                                                                                      -400 y

i . .o Break uncovered o s $'

          .100__                                                      qra             ll            --200 i .

b occ c l t r HPI begips , , 0 500 1000 1500 2000 Time (s) Figure 112. Scenario 8 Break and flPI plus makeup flows. 107

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 1 i i i 1.0 0.8 -- -- 0.8 E i E

• 5 0.6 -- I

                                                                                      --0.65 E                                               i                                              E L                                                                                             L o                                                                                             D g   0.4 --                                                                          -- 0.4 2>

Break  ! 0.2 -- uncovered j ,

                                                                                      -- 0.2 0

I I "' kJ l' "" 0.0 0 500 1000 1500 2000 Time (s) Figure 113. Scenario 8 Void fraction at the break. CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 25 , , i M###e#--e-e-e -o-e--e=9 O A LOOP -800 O B LOOP 20 - Accumulator / a C LOOP _ flow initiated

                                                                                     -600 m                                                                                           n E  15    -                                                                      -

v v e e

                                                                                     -400      !

o 10 - - 2 I 5~ - -

                                                                                     -200 0                                                                                0 0              500              1000                  1500             2000 Time (s)

Figure 11-4. Scenario 8 Accumulator liquid volumes. 108 m

r-CAUT I ON: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. I q c c c c- c, o e o o c 1.0 Upper head voided 0.8 -- a CORE -- 0.8 o UPPER HEAD o 8 8 5 0.6 - -- 0.6 5 E E u u. v v i 0.4 -- -- 0.4 5 0.2 -- Core subcooled Upper head -- 0.2 refilled 0 uos = dc c c c c' c c c c c' c -: 0.0 0 500 1000 1500 2000 Time (s) Figure 11-5. Scenario 8 Upper head and core void fractions. and by 900 s the loops had drained down to the ing the U-tubes at 1650 s, resulting in a large heat elevation of the hot leg nozzles. Accumulator in- transfer rate from the hot steam generators to the jection began at 1040 s, starting the refill of the reactor coolant, reactor coolant system. The reactor vessel had com-pletely refilled at 1650 s as evidenced by the void Calculated cold leg mass flow rates are shown in fraction in the upper head going to zero. By 1650 s, Figure Il-7. The vertical scale has been limited to the hot legs had nearly refilled and liquid drops were present an expanded view of the flow rates after entering the U-tubes. completion of pump coastdown. The mass flow rates decreased rapidly following the trip of the The rate of heat transfer to each steam generator reactor coolant pumps at 32 s. The loop flows secondary is shown in Figure Il-6. The heat transfer nearly stagnated at 125 s when the volume expan-into the C steam generator increased at the start of sion due to vapor production in the core approxi-the transient because of an increase in the C hot leg mately balanced the volumetric flow out the break. fluid temperature that was caused by an outsurge In fact, between 125 and 330 s the flow in the A of flow from the pressurizer. The heat transfer into and B loops was from the vessel backwards through the C steam generator decreased shortly after the the pump towards the steam generator. Between 125 pressurizer emptied at 17 s. The steam generator and 300 s, the density difference between the core 2 secondaries began acting as a heat source at 57 s. and downcomer was balanced by a difference in The steam generators were generally heat sources liquid levelin the uphill and downhill sides of the for the remainder of the calculation, although they steam generator U-tubes. Because of the reverse were heat sinks between 370 s and 480 s because of flow in the A and B cold legs, the HPI refilled the cooling effects related to AFW. The steam genera- U-tubes at 350 s and the density difference between tors were nearly thermally isolated after the U-tubes core and downcomer could no longer be balanced voided at 800 s. Consequently, the steam generators within the steam generator, resulting in a positive remained hot even though the primary coolant natural circulation flow in the A and B loops. The system was cooling. Liquid droplets began enter- increase in flow caused the voids in the core to 109

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.  ; 100 i i , O A LOOP

• o B LOOP A C LOOP y Due to pressurizer

                                                                                                          ^

y Due to AFW $ 3 3 u 0 % Lam _ w. _ u _ _ _ _ _ _ . _

                                                                                    -      --0       '

g

n. SG refill

                                                                         /                           !

n-began

      -                                                                               -- -50 2L
     -10 0 O              500               1000                 1500                   2000 Time (s)

Figure 11-6. Scenario 8 Steam generator heat transfer rates. CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 1000_ 4' ' ' '

                                                                                             - 2000 Natural circulation flow 500 _-                                                 ,                             -- 1000 Loops drained n

N i n

                                                                                                      ^

q o __ . we;. p  ; pern{.g,. __0 g i , n v

     -500--                                                                                -- -1000 ,
$                                                                                                       o g                      O A LOOP O B LOOP                           Due to SG
! -1000 ~    -

A LO P heat transfe _- -2000 g 2 S .

                                                                                             - -3000
    -1500    -
                                                                                              - -4000     .
   -2000 0              500               1000                 1500                   2000 Time (s)

Figure 11-7. Scenario 8 Cold leg mass flow rates. I10

disappear as the core subcooled at 400 s. The driv- generator secondaries would have cooled to the ing potential for natural circulation decreased as the primary saturation temperature, based on the voids in the core disappeared, and the loop mass calculated heat transfer rate at 1740 s. The pressure flow rates nearly stagnated again. The low mass then decreased to below the shutoff head of the LPI flow rates in the cold leg between 450 and 900 s were pumps and remained constant thereafter. The final primarily due to draining of the downhill side of pressure value was determined from steady state the U-tubes and loop seal piping down to the eleva- mass and energy balances. At steady state, the break o tion of the hot leg nozzle. After the loops stopped flow equaled the sum of the HPI and LPI flows and draining at 900 s, the mass flow in the cold leg was removed core decay power. The steady state mostly due to HPI and accumulators. The calcu- downcomer pressure was 0.97 MPa (140 psia). The lated mass flow rates became very noisy after 1650 s steady state fluid temperature at the break was o when the reactor coolant system refilled sufficient- 320 K (116*F), and the break was covered with ly for water droplets to reach the U-tubes. A subcooled liquid. simultaneous reflooding of three hot steam generators would probably be an oscillatory Downcomer fluid temperature is shown in process, but the magnitudes of the calculated flow Figure Il-10. The temperature was calculated by the oscillations are probably too large. code before 1740 s and extrapolated by hand after 1740 s. The extrapolated temperature was constant Calculated cold leg and downcomer fluid temper- between 1740 s, when the calculation terminated, and atures are shown in Figure 11-8. Significant dif- 2700 s, when the steam generator secondaries cooled ferences between cold leg temperatures were to primary saturation temperature. The downcomer calculated between 125 and 900 s. After the loop temperature was then assumed to cool to 310 K mass flow stagnated at 125 s, slight differences in (100*F). The final downcomer temperature was 6 K the direction and magnitude of cold leg flows caused (10 F) above the HPI/LPI temperature because of relatively large differences between the cold leg bypass and leakage flows within the reactor vessel. temperatures in different loops. The cold leg temperatures decreased rapidly at 900 s when the The heat-transfer coefficient for the reactor vessel loops drained down to the elevation of the hot leg wall is shown in Figure 11-11. The heat-transfer coef-nozzles. After the loops finished draining, the cold ficient was calculated by the code before 1740 s and leg flows were due primarily to the ECC flows, and extrapolated by hand after 1740 s. The extrapolation consequently the cold leg temperatures decreased assumed a constant coefficient, equal to the average nearly to the ECC temperature. The minimum cold calculated value during the reflooding of the U-tubes, leg temperature was N310 K (100 F), just 5.6 K until the steam generator secondaries were cooled to (10* F) above the HPI temperature. The downcomer primary saturation temperature at 2700 s. The heat-temperature, although generally higher than the transfer coefficient then decreased to 320 W/m 2 .g cold leg temperatures, al30 decreased after 900 s. 2 (58 Btu /h-ft *F) based on a forced convection heat-The downcomer was warmer than the cold legs transfer correlation (Dittus-Boelter) and the combined primarily because ofleakage of warmer fluid from HPI and LPI flow at 7200 s. the upper plenum, upper head, and downcomer bypass regions into the downcomer. The cold leg The mass flow rate in the A cold leg is shown in and downcomer temperatures increased after 1650 s Figure 11-12. The mass flow rate was calculated by due to mixing caused by the flow oscillations that the code before 1740 s and extrapolated by hand after accompanied the refill of the steam generator 1740 s. The B and C loop flow rates were not shown U-tubes. because the flow was similar in all three loops after 500 s. The extrapolation for the A loop flow was o 11.3.2 Extrapolations and Uncertainties. No applicable for the other loops. The extrapolated flow significant offsets or biases due to uncertainties were equaled the sum of the HPI and LPI flows, apparent in the calculation. The fluid temperature in the A cold leg is shown

  . Downcomer fluid pressure is shown in Fig-                in Figure 11-13. The temperature was calculated by ure Il-9. The pressure was calculated by the code           the code before 1740 s and extrapolated by hand after prior to 1740 s and extrapolated by hand after               1740 s. The B and C loop temperatures were not 1740 s. The extrapolated pressure was constant at           shown because the temperatures in all three loops 1.3 MPa (190 psia) until 2700 s when the steam              were similar after the loops finished draining at 900 s.

I1I i

f CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS,- EQUIPMENT FAILURES, OR BOTH. 600 i i i

                                                                                                      -600 0 A LOOP
               ,                                                                  o B LOOP                      .

550 - A C LOOP - x DOWNCOMER -500 2 v 500 ' - E v E o

                                                                                                       -400 Eo
    }450 E

[ E

                                                                                                      -300 E E 3 400          -                                                                                -

3

                                                                                                      -200 350         -

3 Loops finished ' draining

                                          '                     '                    '                -100 300 O                     500                    1000                  1500            2000 Time (s)

Figure 11-8. Scenario 8 Cold leg and downcomer fluid temperatures. CAUT ION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 20 , i i CALCULATED _ -- EXTRAPOLATED -2500 15 - g -

                                                                                                    -2000 9
 -v
                                                                                                            %u-v E    jo -  _

_-1500

• R R 5

E  :

                                                                                                    -1000   1 5   -                                                                                    -              *
                                                                                                    -500 0                                                                                            O O                    2000                   4000                   6000           8000 Time (s)

Figure 119. Scenario 8 Extrapolated reactor vessel downcomer pressure, 112 L

+ E ' CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 600 i i i

                                                                                                                                                        -600 CALCULATED
                                                                                                                            -- EXTRAPOLATED 550                                                                                                                        -
                                                                                                                                                        -500 E

v 500 - - E v E -

                                                                                                                                                        -400 L-                    2                                                                                                                                             2                      :

o450 - - o  : i

s. -

C -

                                                                                                                                                        -300       a.

E E 4 400 - -

• I_____,i -200 350 -

t - t t i 1

                                 -                                                                          '-----------------------                    -100 300 O                                    2000                                      4000           6000            8000 Time (s)

Figure 11-10. Scenario 8 Extrapolated reactor vessel downcomer temperature. CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 5000 i i i 2 . CALCULATED

                                                                                                                                                       -800 n'                                                                                                         -- EXTRAPOLATED E                                                                                                                                                ~

g'4000 - - g v o 7

                                                                                                                                                       -600                             !

E.C . , g 3000 - - 3,1

- s.

                 .                                                                                                                                                 .        I o             -
                                                                                                                                                       -400

,- , 2000 -

                                                                                                                                                                  %).5 8

. _____, 3 ;;

_v

                 ,,                                                                               1 c                          .)                                                     i                                                               o
    *-           o                                                                                  '                                                  -200        -

2: 1000 - :E

                                                                                                     \

l

                                 -(

i o i

                ;                                                                                       i_______________________.

O O 0 2000 4000 6000 8000  ! Time (s) Figure 1111. Scenario 8 Extrapolated reactor vessel wall inside surface heat transfer coefficient. i l ! 113 i

A CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUlFMENT FAILURES, OR BOTH. 1000 , i ,

                                                                                                           - 2000 CALCULATED
                                                                                -- EXTRAPOLATED

• 500. -- 1000 7

s  ; , 0-d i - Jv ---- -------~~--~~~-~~-------

                                                                                                        --0 5                                                                                                                  5
   $    -500'-                                                                                          -- -1000 C 2

E

                                                                                                          - -2000
       -1000          -                                                                                 -
                                                                                                          - -3000
       -1500 O               2000                         4000                 6000             8000 Time (s)

Figure 1112. Scenario 8 Extrapolated cold leg mass flow. CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 600 , , .

                                                                                                          -600 CALCULATED EXTRAPOLATED m
                 -\     *
                                                                                                          -500 m

5 500. - - D E -

                                                                                                          -400 E 2                                                                                                           ?

E E g 400

                                                                                                          -300 y

f p____,\ -200

                                                   \
                                                    \

n

                                                      \
                                       ,               i_____,__________,_______                          -100 0                 2000                         4000                 6000             8000 Time (s) t Figure 1113. Scenario 8 Extrapolated cold leg fluid temperature.

114

g The same extrapolation was applicable for all three 1616 s. The calculated downcomer pressure at loops. The extrapolated temperature was constant 1616 s was 1.77 MPa (257 psia). The downcomer between 1740 s, when the calculation terminated,. fluid temperature and pressure extrapolated to and 2700 s, when the steam generator secondaries 7200 s were 310 K (100*F) and 0.97 MPa cooled to primary temperature. The extrapolated (140 psia), respectively. Three-dimensional effects temperature then decreased the 306 K (90*F), the within the downcomer were probably not important temperature of the HPI and LPl. during this transient because all three Icops o responded similarly after 900 s, which was long 11.4 Conclusions before the minimum downcomer temperature occurred. The calculation was thought to be The minimum calculated downcomer fluid reasonable. No significant offsets or biases in the s ' temperature was 329 K (f 33*F) and occurred .it calculation were apparent. o-6 115 m.

12. SCENARIO 9, STEAM GENERATOR TUBE RUPTURE AT HOT STANDBY The following subsections describe the investiga- tube, the model used to perform this calculation is tion of Scenario 9. This calculation was performed described in Subsections 2.1 and 2.2. The transient to evaluate the consequences of a postulated rup- was initiated from the hot standby conditions ture of a single steam generator tube with the presented in Subsection 2.3.3.

reactor at hot standby conditions. The broken steam generator tube was simulated A description of the scenario is provided, fol- by the nodalization shown in Figure 12-1. The tube lowed by a discussion of the model changes required primary side was represented by four cells. Heat to perform the calculation. The results of the transfer was represented between each cell and its # calculation, the extrapolated results, the uncertain- adjacent steam generator secondary cell. In addi-ties associated with the calculation, and the conclu- tion to wall friction, lumped flow losses were used sions regarding the calculation are also presented. at junctions 289,291, and 292 representing contrac-tion from plena to tube and expansions from tube Scenarios investigated in this report generally in- to the secondary region. clude conservative assumptions concerning equip-ment failures, operator actions, or combinations of these. Conclusions relative to pressurized thermal 12.3 Results shock severity are not to be drawn directly from the results presented in this report (see Section 18). The following sections describe the analysis results for a calculation of Scenario 9 and extrap-lati n and uncertainty f those results. 12.1 Scenario Description 12.3.1 Calculation Results. A sequence of events A description of the scenario as developed at for this calculation is presented in Table 12-2. At Oak Ridge National Laboratory, appears in zero time a single tube in steam generator A was Table 12-1. assumed to have ruptured at the tubesheet on the cold leg end. Figure 12-2 shows the break mass flow The scenario was initiated with the double-ended rates through both ends of the ruptured tube. The rupture of a single tube in steam generator A with flow through the hot leg end was significantly less the reactor at hot standby conditions. The break than that through the cold leg end due to the large was located at the tubesheet on the cold leg end of wall friction pressure drop imposed on the fluid the tube. Operator actions were assumed to trip off exiting through the full length of the tube. Flow reactor coolant pump (RCP) power when primary through both break paths was friction dominated, system pressure fell below 9.07 MPa (1315 psia) that is, the flow rates were determined by the flow and restart RCPs if subcooling, pressure, and losses of the paths and not by choking phenomena. pressurizer level criteria were later satisfied. The Both paths passed single-phase liquid throughout operator was also assumed to throttle auxiliary the calculation. feedwater to maintain 40% steam generator narrow-range levels. As a result of the break the primary system pressure fell, as shown in Figure 12-3, causing the The assumed operator actions represent those pressurizer proportional heaters to be powered at expected in response to a small break loss of coolant I s and backup heaters to be powered at 8 s. The transient that the operator fails to recognize as a pressurizer level indication, shown in Figure 12-4, a steam genera'or tube rupture event. That is, opera- fell due to the fluid volume lost through the break, ting procedures to recover from a steam generator and this caused the makeup rate to increase and also tube rupture event are not followed. tripped off pressurizer heater power. Because the pressurizer level indication did not recover, the

• 12,2 Model Changes ' " 3 "' P"
• P '*5' " ii^ (5" P"' '

actions in Table 12-1) was never ach.ieved. With the exception of nodalization changes The affected steam generator mass inventory necessary to simulate the broken steam generator increased dramatically as shown in Figure 12-5 and i16

Table 12-1. Scenario description No. 9 Plant Initial State - Just prior to transient initiator General

Description:

Hot 0% Power,0% Power after 100 h of shutdown System Status

 ,_          Turbine: Not latched, TSVs closed Secondary PORY: Automatic control Steam Dump Valves: Automatic control Charging System: Automatic control Pressurizer: Automatic control Engineering Safety Features: Automatic control PORVs: Automatic control Reactor Control: Manual Main Feedwater: In bypass mode, manual control to provide zero power level in S/Gs; I condensate pump,1 MFWP operating.

Aux Feedwater: Automatic control MSIVs: .Open, Automatic control MFIVs: Closed, Automatic control

    ' Transient Initiator A steam generator tube rupture on the cold leg side of tube sheet of S/G A.

Equipment Failures That Occur During the Transient if the Equipment is Demanded. None Operator Rerctions to Reported Information

1. If SIAS signal is generated, the operator will trip the reactor coolant pumps when RCS pressure reaches 1300 psig.

2. The operator will restart reactor coolant pumps 10 min after all the following criteria are met.

A. >40*F subcooled B. Pressurizer level 220% or increasing C. R.C. pressure >325 psig

3. The operator will throttle AFW flow to maintain 40% S/G level.

the associated increase in steam generator level the increasing level in steam generator; option (b) caused a slight throttling back of the main feedwater assumes he observes it and controls feedwater as to the affected steam generator as shown in Fig- he has been; option (c) assumes he observes it, C ure 12-6. The transient was assumed to start with recognizes a tube rupture has occurred and acts main feedwater under manual control with the reac. accordingly. The calculation was performed using tor at hot standby conditions. Accordingly, there option (b). The gradual initial decline in affected was an uncertainty as to what control, if any, should steam generator main feedwater bypass valve flow A be placed on the affected steam generators main rate was an indication that the operator was slowly l feedwater bypass valve. Options included (a) freez. throttling back feedwater but did not recognize the i ing the valve at its pretransient area, (b) allowing transient as a tube rupture which would require the valve to modulate at a rate consistent with that isolation of feed and steam functions of the affected used prior to the transient, and (c) closing the valve, steam generator. This choice was consistent with the Option (a) assumes the operator does not observe operator actions in the scenario description 117 o

Secondary *

     -----_---------_______--_---cfh-______

f I

                                           /
                                                                            /

290 290 02 / N 03 /

                                                                            /
                                           /                                /
                                           /                                /
                                           >                                /                  26603
     -----____i                    r----____l                      i__-___
                                           /                                /

290 / 290 / 26602 04 / 01 /

     ----___.                                    . - -  .                        --_             ____ ~ _

p j

                                           /                                /
                                           /                                /                  26601                                                                    -

i J291

                   =

J292 J289 l l Outlet inlet INEL 4 5451 plenum plenum

                            -210                           206' Figure 12-1. Scenario 9 Nodalization for broken tube in steam generator A. ..

I l l 118

r: 4 f f Table 12-2. Scenario 9 sequence of events E 6 Time f (s) Event l 4 0 Tube rupture occurs 1 Proportional pressurizer heaters powered, makeup rate starts increasing i 8- - Backup pressurizer heaters powered l 64 Pressurizer heaters depowered on low level 68 Makeup rate reaches maximum capacity l 276 SIAS signal gene:ated by low pressurizer pressure (1730 psia) F ' - Actions due to SIAS are: feedwater bypass valves closed, main feedwater pump tripped; main feed-water pump recirculation flow terminated; motor-driven auxiliary feedwater initiated and steam generator levels controlled to 40% NR; HPl and LPI pumps started (shutoff head HPI: 1470 psia, LPI: 145 psia), letdown isolated. 366 HPI flow starts (1470 psia) L 625 RC pump trip (1315 psla) 794 RC pump rotors stopped 1332 Affected steam generator separator flooded (void = 0.) i-l, 1530 Liquid flow to affected steam generator steam line begins 2500 Affected steam generator liquid solid 7200 . Calculation terminated C

   ,e

{ l 119 l t I t

J CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ~ ACTIONS, EQUIPMENT FAILURES, OR BOTH. i, 40~ , i i i i , i _ o HOT LEG SIDE _80 Initial value cold LEG SIDE 30 m

                '7                                                    -
                                                                                                                                                                                                                                                  -60   $E h

6 f * { 20 a

                                                                                                                                                                                                                                                  -40   j i                                                                                                                                     c c a -c c c - c- =                           c-      c&- =               a              w
                  . :E E

2 U -~'

    .                                                                                                                                       LInitial value                                                                                      --20
                                                                                                                                       -ccccccc c c c-a                                                                    c        c     c O ..                                                                                                                                                                                      O O                               1000                                             2000                              3000       4000      5000        6000       7000     8000 Time (s)

Figure 12-2. Scenario 9 Break mass flow rates. CAUT ION: THE SCENARIOS SIMULATED CONTAIN SIGN!FICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 16 i , i i i i i

                                                                                                                                                                                                                                                  -2200
                               ,                                                                                                                                                                                                                           n h I# -                                                                                                                                                                                                      ,
                                                                                                                                                                                                                                                  -2000 v                                                                                                                                                                                                                            v E                                                                                                                                Pressurizer                                                                             E R                                 ~

empty -1800 g E 12 - _ g 1 'k HPI initiated h

  .                            [ 10 _

g -RC pump trip _

                                                                                                                                                                                                                                                  -1400    $

8 O 1000 2000 3000 4000 5000 6000 7000 8000 Time (s) Figure 12-3. Scenario 9 Reactor vessel downcomer fluid pressure. 120

m. -

m Y CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT _CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

0.3 , , , ,

       .o 0.2     -                                                                             -

E

                     .=

0 E E 0.1 - - 0.0 0 200 400 600 800 1000 Time (s) Figure 12-4. Scenario 9 Pressurizer level indication. s CAUT ION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 120000- , , , , ., , ,

                                                                                                                       -250000 100000          -                            -                                                -

o SG A ^ n' f - o SG B -200000 j

                  "                                                                      A     SG.C                            -

m 80000 - - N N 2 s 0 j..cc c c c-c c c c c a c. c.  : : :  : c c c c c 1 a

                                                                                                                       -150000 60000,                                                                                       -

40000 0 1000 2000 3000 4000 5000 6000 7000 8000 Time (s)

                                            ~
                                                                                                                                 \

Figure 12 5. Scenario 9 Steam generator secondary mass inventories. 121

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 3 i i i i

                               .                                                                                                                                       O SG A      -6

• o SG B a SG C

_5 n  ? (2 - -

                                                                                                                                                                                      }

• J -

                                                                                                                                                                                   -4 .a C

D D 2 .

                                                                                                                                                                                   -5

• i -

C j I. -

                                                                                                                                                                                 --2 j
                               -                         I                                                                                                                         -t O                                 : 6: : : i : : : 6 : : :                                                                                             0 0                200             400                    600                                                                          800       1000 Time (s)

Figure 12 6. Scenario 9 Main feedwater bypass regulating valve flow rate. i , (Table 121) that indicate the operator was aware as a result was not started for this transient. a small break transieM was in progress but did not Therefore, only motor-driven AFW was initiated recognize it to be a steam generator tube rupture. during this transient. The results of the calculation were insensitive, however, to the selection of options (a), (b), or (c) Figure 12-7 shows a comparison of total injec-because the combined break mass flow rates (Fig- tion and break mass flow rates. Injection flow com-ute 12-2) exceeded the affected steam generator bines the HPI rate for the three loops with the main feedwater Dow rate (Fig,ure 12-6) by more makeup rate. Break flow is a combination of the than an order of magnitude. Thus, results were two break mass flow rates shown in Figure 12-2. relatively insensitive to the delivery of main Following initiation of HPI at 366 s the injection feedwater. flow exceeded the break flow by a small amount throughout the remainder of the transient. During At 276 s the safety injection activation signal this period the primary system pressure (Fig-(SIAS) was generated due to low pressurizer ute 12-3) was stabilized slightly below the HPI pressure. As a result of the SIAS the following shutoff head with injection volumetric flow just actions occurred: Main feedwater (MFW) bypass balanced by break volumetric flow, the pressurizer valves were closed, the MFW pump was tripped, virtually empty, and the pressurizer surge line flow MFW pump recirculation flow was terminated, virtually zero. The injection mass flow rate during motor-driven auxiliary feedwater (AFW) was ini- this period was slightly higher than that of the break tiated and controlled to each steam generator to in Figure 12 7 because the injection flow density maintain 40% narrow-range level indication, high was higher than that at the break, and low pressure injection (HPI and LPI) pumps were started, and letdown was isolated. Figure 12-8 shows the steam generator secondary pressure responses and Figure 12-9 shows the steam Turbine-driven AFW initiation requires two-out- generator heat removal rates. The heat removal rate of three low steam generator level indications and shown for Steam Generator A does not include the 122

        ~.       _                .                  .           _              _ _ _                 _ _ .-              ..

W n CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN , OPERATOR ACT10NS, EQUIPMENT FAILURES, OR BOTH. 60- , , , , , , ,

                     .                                                        O TOTAL INJECTION FLOW                         -120 i

o TOTAL RREAK FLOW

                                                                                                                             -10 0
           ^                                                                                                                       7

{a 40 - N E 3

                                                                                                                             -80   g v

o

            ,.                                                                                                               -60   ::=

20 - S-S E E i-E E E ,- -

                                                                                                                             .gg U
                                                                                                                             -20 0--

O O 1000 2000 3000 4000 5000 6000 7000 8000  ;

                           . Figure 12 7. Scenario 9 Comparison of injection and break mass flow rate.

CAUTION: THE SCENARIOS SIMULATED CON,TAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 7.1 i i , i i i i

                      ,          cce_,mm                       ce                     C       0    c=       0-     u
                                                                                                                             -1020 7    -                                                                                               -

T 9

n. %a.

2

• 6.9-- --1000
• I E 3 3 2 6.8 -

g 1 -

                                                                                                                             -980      I E 6.7      -                                                     '                                         -

E o 3 'O SG A 3 0 0 SG B o a SG c -960 6.6 - - 6.5 l 0 1000 2000 3000 4000 5000 6000 7000 8000 L Time (s) (' Figure 12 8. Scenario 9 Steam generator secondary pressures. > 123 l r

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. , 20 , , , , , , , 0 SG A o SG B a SG C

        , 10        -                                                                             -

h j ' p e c - weuremnym - i l l 1 I I I 1 0 1000 2000 3000 4000 5000 6000 7000 8000 Time (s) Figure 12-9. Scenario 9 Steam generator heat removal rates. break flow energy removal rate, only that energy Mass flow rates between the HPI sites and reac-removed through the tube walls. Steam Gener- tor vessel in the three cold legs are shown in ator A (SGA) remained pressurized to the steam Figure 12-10. Following reactor coolant pump dump setpoint as a result of the broken tube break coastdown smooth natural circulation was estab-flows and became a primary system heat source at lished at around 1300 s. Due to the effects of the 511 s. Because the SGA level was elevated, no AFW broken steam generator tube, the Loop A cold leg was delivered to SGA. Steam Generators B and C natural circulation flow rate was about one-third (SGB and SGC) secondary pressures remained that in Loops B and C as shown in Figure 12-10 elevated until 593 and 703 s, respectively, when from about 1300 to 2200 s. After this time, flow these generators became primary system heat oscillations in all three loops were observed The sources. AFW was delivered to SGB from 276 to oscillations were caused by phenomena occurring 543 s and SGC from 276 to 617 s. AFW to these in the Loop A cold leg. Loops B and C flow oscilla. steam generators was terminated when their narrow tions were driven by the oscillations emanating from range levels exceeded 40%. The initial high SGC Loop A. The oscillation of Loop A flow was heat removal rate was caused as hot pressurizer fluid developed by a time-varying driving force for loop entered the Loop C hot leg during the initial natural circulation flow caused by a time varying , pressurizer outsurge. The peak in SGC heat removal loop A cold leg fluid temperature during periods rate occurred from 543 to 617 s when SGC was oflow flow or stagnant Loop A conditions. As loop receiving all the AFW flow, flow approached stagnation, llPI fluid made up an increasing percentage of the fluid reaching the reac- , At 625 s the primary system pressure declined to tor vessel. The llPI fluid was colder than the fluid 9.07 MPa (1315 psia) and power to all reactor coming from the reactor coolant pump discharge coolant pumps was tripped off. The ensuing pump so as stagnation was approached the reactor vessel coastdown ended at 794 s when the pump rotors downcomer fluid temperature decreased. As a had stopped, result, the reactor vessel downcomer density 124

CAUT ION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 200 ni i i i i i i

                                                                                                                                                                      - 400 e                                        +               3 RC pump coastdown                                                                                                    i 10 0 ,-      k                                                            .
                                                                                                                                                                      - 200 ^

0-- - O g f i t I o o a 2

                              -10 0
                                     ~

0 LOOP A -- -200 S ' 0 LOOP B A LOOP C ,

                                    ~
                                                                                                                                                                        "4
                             -200 0    1000      2000       3000        4000       5000           6000                                                7000     8000 Time (s)

Figure 12-10 Scenario 9 Cold leg mass flow rate. Increased and this caused the natural circulation complete stagnation as shown in Figure 1210. The flow rate to increase. One would suspect a steady discussion above is intended to explain the causes equilibrium point might have been reached where of the oscillations observed in the RELAPS calcula-the loop flow slowed until the natural circulation tion. A discussion of whether or not these oscilla-driving potential required to sustain that steady flow tions are physical is presented in Subsection 12.3.2. was achieved. This steady equilibrium is attainable, i however, only if the natural circulation driving Figure 1211 shows the cold leg fluid tempera-potential rapidly responds to changes in loop flow, tures between the llPI sites and the reactor vessel in the calculation presented here this was not the downcomer. All temperatures shown generally case; loop velocities were less than @.05 m/s decreased as a result of continued primary system (0.2 ft/s). Thus a change in loop flow did not affect energy removal due to liPI and break flow. The the natural circulation driving potential (by chang- temperature oscillations were large but are , ing the liquid density in the downcomer) until explained fully by the loop flow oscillations just significantly later in time. At a loop velocity of discussed. The minimum calculated downcomer 0.05 m/s the transit time from the liPI site to the fluid temperature was 451 K (352'F) as shown in reactor vessel was 80 s. This delay was the cause of Figure 1212.

   ,                the loop oscillations observed in Figure 1210. To                                                                                                             i further the investigation, a sensitivity RELAPS                   At 1332 s the affected steam generator separator                                            ;

calculation was performed in which the llPI injec- was 11ooded and at 1530 s liquid flow to the affected i tion sites were moved from the cold legs to the steam generator steam line commenced. At 2500 s the

   .,,.             reactor vessel downcomer inlet at $300 s. The                  affected steam generator secondary was liquid solid.

oscillations damped out shortly after the injection sites were moved confirming that the delay effeet The calculation was terminated at 7200 s, the end discussed above was responsible for the flow oscilla- of the pressurized thermal shock investigation tions. The change in the nature of the oscillations period as defined at Oak Ridge National at about 5000 s was caused by Loop 11 also reaching Laboratory. 125 e______________________-.----_.--_____

CAUil0N: THE SCENARIOS SIMULATED $ CONTAIN SIGNIFICANT CONSERVAllSMS IN  ;

                                    ' OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.                                                                                      l 600                     ,          ,         ,        i               i             ,            ,

600 0 LOOP A Q i o L OOP R p

                                                                                                                                                                  ~

l

            " 550           -                                                                           A Loop c              -                                           f i

40

            - 450
                                                                               '                 "   d ip hh j f 3

3 3 4

            = 400 l

f -

                                                                                                                                -300 3
                                                                                                                                                                  =
                                                                                                   ,        ,  1     ,

I l

           - g 350           -
                                                                                                                                -200 g

l

                                         '          '         '         '              '             '            '             ~100 300' 0             1000       2000      3000       4000           5000         6000        7000            8000                                           :

Time (s) t Figure 1211. Scenario 9 Cold les nuld temperatures.  ; i CAUil0N: THE SCENARIOS SIMULATED CONTAIN Sl'GNIFICANT CONSERVAllSMS IN i OPERATOR ACil0NS, EQUlPMENT FAILURES, OR BOTH. 600 i , , , , , i

                      -                                                                                                         -600                              ,

v F i E - 550 I 3 , 3

                                                                                                                              -                                    0 E550             -

g. .

E

              ;                                                                                                                  -500

( t v . h a 500

                              -                                  l U r r,

[ ~ 450ul ffft b -

                                                                                                                                  -400 _.g t

450 t 0 1000 2000 3000 4000 5000 6000 7000 8000 Time (s)  ; l'isure 1212. Scenario 9 Reactor venet downcomer Guld temperature. , 126

12.3.2 Extrapolations and Uncertainties. Since discussed in Subscstion 12.3.1 this change suc-the calculation was carried out to a full 2 h after ceeded in damping out the flow oscillations. The the initiating esent, no extrapolation of data was sensitivity calculation was run for 1200 s, starting required. at $300 s, and the results were used to generate the dashed lines that appear on the remaining figures The major uncertainty in the calculated results in this section. The dashed lines represent the best centers on the flow oscillations discussed in the estimate of plant responses for this sequence, o previous section. The oscillations were determined to be caused by a delayed response of natural cir- Figure 1213 shows the reactor vesseldowneomer culation drising potential to changes in loop flow, fluid pressure. The dashed line indicates the best The delay is caused by long transit times of fluid estimate pressure response. Very little change in

       ,,                                        from the IIPI site to the reactor vessel. The uncer-          pressure was observed as a result of damping out tainty is whether or not the RELAPS-calculated                the loop flow oscillations. At 7200 s the b.st delay phenomena in the affected loop cold leg                 estimate pressure is 9.624 MPa (1396 psla),

represents reality. Figure 1214 shows the reactor vessel downcomer RELAPS is a water and steam, two fluid com- fluid temperature and Figure 1215 shows the puter code. If a cell is liquid fdled, as the cold leg downcomer wallinside surface heat transfer coef-was during the oscillations, the liquid is considered ficient. The dashed lines indicate the best estimate to be homogeneous and represented by a single Guld responses as determined through the use of the temperature. The model contains three cells between sensitivity calculation. At 7200 s the best estimate the llPI slie and reactor sessel. As long as loop flow downeomer Muid temperature is 465 K (378'F). 15 suf(1clently large, a high degree of mixing between llPI and loop flow is to be expected and the model Figures 1216 through 1218 show the mass flow will faithfully represent reality. If the loop flow rates between the llPi location and reactor vessel stagnates, as does Loop A after about 2400 s, a in each of the cold legs. The dashed line for each homogeneous model does not represent well the loop was based on the results of the sensitivity now conditions cf peeted within the cold leg. Speci- calculation with adjustments in loop flow expected fically, under stagnant conditions a high degree of if IIPI and reactor coolant makeup injection are vertical thermal stratification of liquid within the accounted for. The adjusted cold leg flow rates at cold legs is to be expected. Cold ilPI liquid is the end of calculation are 3.6 kg/s (8 lb/s) in expected to flow along the bottom of the cold leg I oop A,33.6 kg/s(74 lb/s)in Loop 11and 60 kg/s cross section underneath warmer liquid flowing in (132 lb/s)in Loop C. Factors affecting these dif-the top. Under such circumstances, the average ferences in loop 00ws are: (a) the effects of the density of Guld reaching the reactor sessel cold leg break on Loop A flow, (b) all makeup injection nonle would be smooth in time. Thus the flow occurring into Loop II, and (c) Loop C steam oscillations observed during near stagnant loop con- generator secondary fluid temperatures below those ditions were caused by a limitation of the computer of 1.oop 11. This latter factor caused less degrada-model and were not physical. Ilecause thermal tion of natural circulation driving head in Loop C stratification within the cold leg is to be expected, than in I oop II, hence the higher I.oop C flow, however, this sequence should be investigated fur-ther using a multidimensional computer model Figures 1219 through 12 21 show the cold leg capable of simulating thermal stratifiention within Guld temperatures between the llPI injection loca. the cold leg and reactor venel dow neomer regions. tions and the reactor scuel. Adjustments, indicated by the dashed lines, have been made to account for o Since the loop flow oscillations were not physleal the climination of oscillatory behavior. 'ihe and have a significant cffcct on mised cold leg and adjustments wete accomplished by using the sensi-reactor scuci dow neomer Muld temperatures (1 ig. tivity calculation as an indication of loop now rates urcs 12.ll and 1212)it was considered neceuary and RCP discharge fluid temperatures. Man. flow-

       ,                                        to predict the respomes to be espected with the                weighted 0uld temperature mising calculations were oscillations removed. This prediction was a com.               then performed ming the loop and llPl/ makeup plished by performing a RELAP5 semitivity cateu.               flow rates and tcmperatures. The Loop 11 and C lation in which the llPI and makeup were reh>eated             adjmted temperatures were higher than those from the cold Ica to the reactor scucl inlet. As               calculated due to higher than calculated adjmted 127                                                          i i

m____._

CAUTION: THE SCENARIOS SIMULATED. CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS. EQUIPMENT FAILURES, OR BOTH. 16 i _i , , , , , CALCULATED

                                                                                                   -- ADJUSTED                       -2200 14                                                                                                                      -
                                                                                                                                     -2000 %a v

t. .

3 -

                                                                                                                                     -1800    5

• 5 12 - - E

-c . a

                                                                                                                                     -1600    ,
    ~
     !                                                                                                                                       E E
    >w _                               .______________________.--a                               _
                                                                                                                                     -1400   $
                         '          '          '        '                                    '           ,             ,             -1  0
        -8 0        1000      2000 '3000           4000            5000                            6000           7000       8000 T!me (s)

Figure 12-13. Scenario 9 Calculated and adjusted reactor vessel downcomer fluid pressure. CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN

                    ' OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

600 , , , , , , , CALCULATED -600 ^

  $                                                                                                 -- ADJUSTED t                                                                                                                                         .

3

                                                                                                                                      -550 $
   ?.550          -                                                                                                                .       %
  .[
                                                                                                                                      -500 E k

2 m

              ~
                                                                                                                                      -450

_ 500 - _ f

                                                                                                                                           ~

[I[f(ffffr x'

   .                                                                 I                                                                           -

g s

                                                                                                                                      -400 2 j

450 ' ' ' ' ' ' ' O- 1000- 2000 .3000 4000 -5000 - 6000 7000 8000 Time (s) Figure 12-14. ' Scenario 9 Calculated and adjusted reactor vessel downcomer fluid temperature. 128

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 6 , , , , , , ,

                                        ~

CALCULATED -1000 eg -- ADJUSTED i i e' h -

                                                                                                                                            -800
                                                                                                                                                    'E 3                                                                                                                   .

g4 - h f- . 5

                                                                                                                                            -600        d mi
                                $                                                                                                                   II u

as

                                                                                                                                            ~400      52 u

s2 n r

                                                                                                         ;l      I                   -

am {ff

                                                                                                                                            -200 y

0 ' ' ' ' ' ' r nw ' O O 1000 2000 3000 4000 5000 6000 7000 8000 Time (s) Figure 1215. Scenario 9 Calculated and adjusted reactor vessel downcomer inside surface heat transfer coefficient. CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 200 i , , , , i i CALCULATED - 400

                                                                                                          -- ADJUSTED
                                                                                                                                            - 300 100 ,-                                                                                                 -
                                                                                                                                            - 200
                            .x                                                                                                                         o v
                                                                                                                                            - 100      3 S

to O-- l Wh /

                                                                                                                            "        ^
                                                                                                                                          --0 6

I - -10 0

    *                                   ~
                                                      '                    '      '           '       '            '             '          '~
                                -100' 0       1000      209L               3000       4000    5000       6000           7000         8000 Tirrn(s)

Figure 12-16. Scenario 9 Calculated and adjusted Loop A cold le8 mass flow rate. 129

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGHlflCANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 200 ., , , , , , ,

                                                                                                                                                                                     - 400 CALCULATED
                                                                                                                                              -- ADJUSTED 100      -

m -- 200 ^ < L . 4 i fl

                                                                                                                                             '~    ...                                        .5 To                                                                                                                         j%         J      ) )                Tl f ,--0                      *       '

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=                                                                                                                                                                                               .,
$                                                                                                                                                                                               8
   -100
          ~
               -                                                                                                                                                                  -  - -200 2
   -200 O                   1000                               2000                                  3000      4000            5000           6000             7000           8000 Time (s)

Figure 1217. Scenario 9 Calculated and adjusted Loop B cold leg mass flow rate. CAUTION: THE SCENARIOS SIMULATE 0 CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 200 , , , , , , ,

                                                                                                                                                                                     - 400
                                                                                                                                                                                      - 300 n

7 $, 100 ,- - 200 E _ __ .. _. b o - si l i I  ! - 100 g O )YJ J

                                                                                                                                                                                    --0          E E     0' --

2 $ , CALCULATED

                                                   -- ADJUSTED                                                                                                                        - -10 0
                                                                                                                                                                                      ~ -
    -10 0 '

0 1000 2000 3000 4000 5000 6000 7000 8000 Time (s) Figure 12-18. Scenario 9 Calculated and adjusted Loop C cold leg mass flow rate. 130

CAUTION: THE . SCENARIOS SIMULATED , CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 600 i i , , , , ,

                                                                                                                                                    -600 CALCULATED 7

v

                                                                                                      -- ADJUSTED                                               p v

_Soo o 500 - - E. E

                                                                         'i                     L
                                                                                                                                                     -400        E.

E Y '1 T *

                                                                         -l[

1 ') p

                                  -                                             I f                   -300       }

I400 -

                                                                                                ' lljjf E

1 e ss i e g -

                                                                                                         's s                                        -200 E O                                                                                            's,'                                        o      !

300 O. 1000 2000 3000 4000 5000 6000 7000 8000 Time (s) Figute 12-19. Scenario 9 Calculated and adjusted Loop A cold leg fluid temperature. CAUT ION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT - FAILURES,' OR BOTH. 600 , i i i i i i CALCULATED

7. -- ADJUSTED '@

                 .v                                                                                                                                             v
                                                                                                                                                     -500

• 500 - 1 1

43 -400 i  ; n'

• 2 1 J i- s .,

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                    '}          _ _

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                       .g          -
                                                                                                                                                     -200 E f                       'o                                                                                      1 1    b                                  o
                                                                                                                                                     -100 0      1000    2000     3000         4000            5000           6000           7000           8000 Time (s)

Figure 12-20. Scenario 9 Calculated and adjusted Loop B cold leg fluid temperature. , 131 l t

~ 4 4

                                      - CAUT ION: ITHE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS,' EQUIPMENT FAILURES, OR BOTH.

600 i . . , , , . CALCULATEC- -600

   ?!       :7:

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            ]        :-
                                                                                                         -550 $
              ~O550-    -                                                                              -           o
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                                                                             ~.                                   3 t              -450 =&
              = 500     --
                                                                                           )     $.    -

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                                                                                                         -400 2 5'                                                                                                  $

450 O 1000 2000 3000. 4000 5000' 6000 '7000 8000 Time . (s) Figure 12-21. - Scenario 9 Calculated and adjusted Loop C cold leg fluid temperature. loop flow rates as shown in Figures 12-17 and and were caused by a limitation of the computer 12-18. Conversely, the adjusted Loop A tempera- model during near-stagnant flow conditions. The ture was lower-than-calculated as a result of the computer model precludes calculation of thermal lower-than-calculated adjusted Loop A mass flow stratification effects within a liquid-filled cold leg shown in Figure 12-16. The adjusted cold leg fluid that would remove the driving potential for the

      . temperatures .at 7200 s are 330 K (135'F) in               oscillations.

Loop A, 433 K (320*F) in Loop B,' and '502 K (444*F) in Loop C. The minimum reactor vessel downcomer fluid temperature occurred at the end of the 2-h period.

      -12.4 Conclusions                                            After adjustment for removal of the nonphysical flow oscillations,' the minimum downcomer The calculated loop flow oscillations that were          temperature is 465 K (378'F). The associated -

observed starting at about 3000 s were not physical downcomer fluid pressure is 9.624 MPa (13% psia). e a 132

13. SCENARIO 10, STEAM GENERATOR TUBE RUPTURE AT FULL POWER The following subsections contain the transient 13.3 Transient Results scenario description, modeling changes effected to perform this calculation, detailed analysis cf the This section presents the results and extrapola-transient results, extrapolations and uncertainty tions and uncertainties of the steam generator tube analysis, and conclusions drawn from the analysis rupture transient at full power.

for Scenario 10: a single steam generator tube rup-ture at full power. 13.3.1 Calculation Results. A calculated sequence of events for the transient is provided in Table 13-2. o Scenarios investigated in this report generally The double-ended rupture in the cold leg side of a include conservative assumptions concerning equip- single tube in Steam Generator A was assumed to ment failures, operator actions, or combinations of occur at time zero. Upon rupture of the tube, the these. Conclusions relative to pressurized thermal primary system began to depressurize as shown in shock severity are not to be drawn directly from the Figure 13-1. The pressurizer proportional and back-results presented in this report (see Section 18). up heaters were automatically initiated as a result of the depressurization in an effort to recover primary system pressure. Also, as a result of the break, the 13.1 Transient Scenario pressurizer liquid level shown in Figure 13-2 began Description t decrease, resulting in an increase in the make-up flow rate to recover the level. At M9 s a high tur-bine overtemperature AT signalin the reactor protec-tion system was generated and the turbine governor The transient was initiated from full power steady state (nominal temperature and pressure), with all an to mn bad heen M and M s load rejection setpoints of 15, 35, and 55% were reached system controls in automatic control. The transient and steam dump valve banks 1,2, and 3 were began by a double-ended break occurring in a single tube in the cold leg side of the Loop A steam respedely as m, &ap h tk tutal mam dump mass flow rate shown m Figure 13-3. generator. During the transient it was assumed that all systems operate automatically as designed. The At 114 s into the transient, a reactor over-only opcator actions were: (a) trip the reactor temperature AT signal tripped the reactor. Primary coolant pumps when the primary pressure reached system pressure shown in Figure 13-1 rapidly 9.1 MPa (1315 psia) and a SIAS signal was gener- decreased as a result of the decrease in core power ated, and (b) throttle auxiliary feedwater flow to at reactor trip and a delay in the reduction of maintain a 40% narrow range level in each of the primary to secondary heat removal rate as shown steam generators. it should be noted, as in in Figure 13-4. The turbine stop valves closed as a Section 12, the above actions are not the normal result of reactor trip, and secondary pressure procedures the operator would take in this type of rapidly increased to 4.9 MPa (1000 psia) as shown transient. A transient scenario is provided in in Figure 13-5. The steam dump system switched Table 13-1. from load rejection control to plant trip control mode due to reactor trip. The steam dump valves closed as a result of an overswing of a lead-lag con-l 13.2 Model Changes troller in the plant trip control system logic, but opened thereafter, as shown in Figure 13-3, and modulated to bring the primary system a.erage G Changes made to the steady state model to temperature down to S59 K (547'F). After reactor incorporate the tube rupture include the addition of a trip, the primary system pressure continued to pipe component describing a single tube, and are the decrease at a slower rate as shown in Figure 13-1, same as for the steam generator tube rupture at hot due to the reduction in primary system heat removal standby, described in detail in Subsection 12.2. rate shown in Figure 13-4. 133

m Table 13-1. Scenario description No.10 ,

 %nt Initial State - Just prior to transient initiator General

Description:

1009e Power steady state System Status Turbine: Automatic control Secondary PORV: Automatk control Steam Dump Valves: Automatic control Charging System:. Automatic control Pressurizer: Automatic control Engineering Safety Features: Automatic control PORVs: Automatic control R: actor Control: Automatic Main Feedwater: Automatic control Aux Feedwater: Automatic control MSIVs: Open, Automatic control MFIVs: Open, Automatic control Transient initiator A steam generator tube rupture on the cold leg side of the tube sheet on S/O A. Equipment Failures That Occur During the Transient if the Equipment is Demanded. None Operator Reactions to Reported Information

1. If SIAS signal is generatM. the operator will trip the reactor coolant pumps when RCS pressure reaches 1300 psig.

2. The operator will throttle AFW flow to maintain 40Ve S/O level.

CAUTION: THE SCENARIOS SIMULATED CONTAlH SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 16 , . , , , , . . .

                 -                                                                                            -2200 1                  Reaptor trip                                                                 9 g 14 -      -
                                                                                                              -2000 %o-g v                                                                                                              v e                                                                                                              .
         '              4                  Secondary pressure                                                           s.

3 -1800 g increase

                  ~

E 12 - g

         -                                                                                                              s.

o. a.

N ressurizer empty. -1600 g g 3 3 0 - o

        > jQ .

1400 > C - t . Primary SyStea inflow

                             ,        , ,and outflow e,qualized
                                                                                      ,      ,        ,       -1200 0     250 500            750 1000 1250 1500 1750 2000 2250 2500 Time (s)

Figure 13-1. Scenario 10 Primary system pressure. 134

Table 13-2. Scenario 10 sequence of events Transient Time (s) Event O Single tube rupture in Steam Generator A occurs Primary system begins to depressurize.

10. Pressurizer heaters turn on.

o 89 2/3 turbine high overtemperature AT signal, turbine valve began to run back. 95 15% load rejection setpoint reached and steam' dump valve bank i opened. 101 35% load rejection actuation setpoint reached and steam dump valve bank 2 opened. 108 55% load rejection actuation setpoint reached and steam dump valve bank 3 opened. 114 Reactor tripped on 2/3 reactor overtemperature AT. Turbine stop valve,s close. Steam

               . dump system switches to plant trip control mode and steam dump valves modulate to bring the plant Tave to 559 K (547*F).

130 Pressurizer level is <14.4% and heaters are tripped off. Main feedwater regulating valves close on reactor trip and low Tave. 137 Main feedwater pumps trip; motor-driven auxiliary feed water system turned on'. 199 SIAS signal received due to low pressurizer pressure. 275 Pressurizer and pressurizer surge line empty. 335 Pressure dropped below HPI pump head and HPI flow is established. Auxiliary feed to Steam Generator A terminated. 370- Primary depressurization stopped. 560 Total aux feed flow goes to Steam Generator C for about 50 s. 600 Primary pressure increase began. 1130: Plant Tave dropped below 559 K (547*F) and steam dump system switched from plant trip control to steam pressure control. Auxiliary feed flow to Steam Generator C increased. 1160 Level in Steam Generator C reached 40% narrow range level and auxiliary feed flow to that generator is terminated. -o. 1200 Level in Steam Generator B reached 40% narrow range level and auxiliary feed flow was terminated.

 ,,   2400      Transient terminated. Primary system pressure stable at 9.6 MPa (1400 psia) and vessel downcomer temperature stable at 560 K (549'F). Secondary heat removal was equal to core power, and total primary inlet flow equaled the break flow.

7200 End of extrapolation 135

1 CAUTION: THE SCENARIOS SIMULATED

                                 ' CONTAIN SIGNIFICANT CONSERVATISMS IN --

OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

0.6 . . .i_ i , , , , ,

         - 0.5                                                                                                                                -

e

g

    ;     0.4     -

E o,3 - - i s. gl0.2 - o z-0.1 - 0.0 ' ' ' ' ' ' ' ' ' O :250 500 750 1000 1250 1500 '1750 2000 2250 2500 Time (s) Figure 13-2. . Scenario 10 Normalized pressurizer liquid level. CAUTION:- THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 1500 , , , , , i i i i

                                                                                                                                                -1000 400      -
                                                                                                                                                -800 g

(

    $ 300         -
                                                                                                                                                      )

n

                                                                                                                                                -600  v
   .c e

l g . 200 , -

                                                                                                                                                -400   a O.

2 Plant trip control $

• 2

                         /
         .100_-                                                                                                                               -

200 y- Steam pressure control

            -O                                                                                                                                    O O'-250            500    750 1000 1250 1500 1750 2000 2250 2500 Time (s)

Figure 13-3. Scenario 10 Total steam dump valve mass flow rate. 136 s

CAUTION: THE SCENARIOS SIMULATED

                                              - CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

200 i i i , i i , , , , i i CORE POWER l -- SECONDARY Q

 .o-i 150    -

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 'o.          3:                    s.

~ d 's u - 10 0 -

                                     -\

I \ l \l, 50 - ii ,i

                                                    ' J 'i f '

lij i s_ ,' % ,I ., I i 8g

                                                                               -\       r g-,,-*------.

s_ 0 O 250 500 750 1000 1250 1500 1750 2000 2250 2500 Time (s)- Figure 13-4. Scenario 10 Core power versus total primary to secondary heat transfer rate. CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 7.5 i i i , i , i i i

                                                                                            - - = = = = = ==-

F-y n

                        '7 -       >                        : :

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                 ' 6.5 R                                                                                                                       R E'         -
                                                                                                                                 -900    E r-          &                                                                                                                       g
                 ,-       6   -                                                                                             -

E o STEAM GENERATOR A E

 .#             3                                                                                                                       2 o STEAM GENERATOR 8
               $            q                                                         a STEAM GENERATOR C
                                                                                                                            --800       $

Turbine stop valves close 5 O 250 500 750 1000 1250 1500 1750 2000 2250 2500 Time (s) Figure 13-5. Scenario 10 Steam generator secondary pressure. F 137

  ,   ,       .                        ~=      __ . -.                 - - - - . -.                                              - . - - .                        .- .-

F _ Break flow, shown in Figure 13-6, rapidly shown in Figure 13-1. The primary system pressure

          ~ decreased upon reactor trip due to the reduction in                                             remained relatively stable until s500 s, at which primary - system pressure and the increase in                                                  time, the press'urizer surge line began to fill with

secondary system pressure. Choking, which liquid. The. incoming liquid interacted with the dominated the break flow from the initiation of the steam in the first volume of the surge line and transient, was terminated when the conditions at the resulted in condensation that depressurized the break changed from saturated liquid to subcooled primary system. The volumetric condensation rate ,,

liquid due to the consequences of the reactor trip. briefly exceeded the volumetric inflow to the The break mass flow rate increased as a result of primary system and the primary pressure decreased subcooling at the break until the steam dump valves as shown in Figure 13-1. Also affected by the con-opened, as discussed above, at which time the break densation was HPI flow, which increased, as shown flow again decreased. in Figure 13-6, due to the decrease in primary pressure. At s525 s the condensation in the surge . At 130 s the pressurizer liquid level decreased line had stopped and primary pressure began to below 14.4% of the measured level range and the increase as shown in Figure 13-1. The increase was pressurizer proportional and back-up heater power due to more heat being generated in the core than was terminated. Also, at approximately this time was removed by the secondary as shown in the main feedwater regulating valves were closed Figure 13-4. due to the reactor trip and a low average

!            temperature signal. No effort by the operator was                                                 Figure 13-8 shows the auxiliary feedwater flow assumed to open the bypass valves to maintain main into each of the steam generator secondaries. Dur-
          - feed flow to the generators. Approximately 7 s                                                 ing the transient, the operator controlled the narrow

. later, the main feedwater pumps tripped and motor- range level to 40%. At N335 s the 40% narrow driven auxiliary feedwater was initiated. Turbine-range level in the affected steam generator was , ' driven auxiliary feedwater was not initiated as it reached and auxiliary feedwater to that generator requires two-out-of-three low steam generator level was terminated. Because of the common auxiliary indications, feedwater header, the flow was split between the two unaffected steam generators. The amount of The secondary narrow range liquid level is shown flow each generator received was determined by the

in Figure 13-7 for all three generators. Initially, the pressure difference between the auxiliary feedwater

level in the affected steam generator (Steam header pressure and the steam generator pressure.

_ Generator A) increased due to the break flow into As shown in Figure 13-8 at s400 and 550 s

          . the secondary. Upon reactor trip, the turbine stop                                             auxiliary feed flow preferred the C steam generator valves closed and the level in the generators col-                                             for short periods of time. The effect of this behavior lapsed due to the elimination of flow effects on the (notably at 550 s) enhanced the overall primary 4            differential pressure that was used in the level                                               system heat removal (see Figure 13-4) such that calculation. When the motor-driven auxiliary feed-                                             primary repressurization at 550 s was arrested

+ water system was initiated at sl37 s the levels began temporarily as shown in Figure 13-1. Condensation a to merease. effects in the C steam generator were the major l cause of the preferential auxiliary feedwater flow. At s200 s the pressure in the pressurizer dropped below I1.93 MPa (1730 psia) generating a SIAS signal and energizing the HPI pumps. By s275 s Between 620 and i130 s primary system pressure

         . the pressurizer and pressurizer surge line had                                                  had increased to s9.6 MPa (1400 psia) and was emptied; thus the primary depressurization rate                                                stable due to the total primary inflow equating the increased as shown in Figure 13-1, and the mass                                                break flow, and the total energy be,ms removed               .

flow rate out the break decreased as shown in from the primary system (including energy removed Figure 13-6. at the break) equaling the core decay power. Also, the loop mass flow rates, shown m Figure 13-9, By N335 s primary pressure had dropped below were very nearly equal during this time, whereas .

' the HPI pump head and HPI flow was initiated. before, the Loop C mass flow rate was higher as Upon initiation of HPI, the total mass flow into a result of emptymg the pressurizer.

the primary system equaled the break mass flow rate as shown in Figure 13-6. The primary depressuriza- At s1130 s the auctioneered high average tion was terminated at s9.4 MPa (1363 psia) as temperature dropped below 559 K (547'F) and the 1 138

CAUTION: 1THE SCENARIOS. SIMULATED ~ CONTAIN SIGNIFICANT CONSERVATISMS IN OPERAT0ft ACTIONS, EQUIPMENT FAILURES, OR BOTH. 50 i , i i , i i , , O BREAK FLOW 0 INFLOW -100 o 40 - A

                                                                                                                    -80  g
    $                                                                                                                    N
    $ 30          -

E

                                                                                                                         .o
       ,                                                                                                           -60   v
                      ~
                                                                                                                 ~

y ~

                                             .--       C : e                      a c      - - -

a -40 10 -- - Primary sytem - 20 pressure stablized Od ' ' ' ' ' ' ' ' O O 250 500 750 1000 1250 1500 1750 2000 2250 2500 Time (s) Figure 13-6. Scenario 10 Total break mass flow rate versus total ECC CVCS mass flow rate. CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR .BOTH. 1 i i i i i i i i i c c ca e

  .e_ .
  "O            l         .)

2 0.5 - -

  %                     o E

o. a STEAM GENERATOR A Z

1. ' o STEAM CENERATOR B

t. STEAM GENERATOR C 0

0 250. 500 750 1000 1250 1500 1750 2000 2250 2500 Time (s) Figure 13-7. Scenario 10 Normalized steam generator naricw range liquid levels. 139

CAUT I ON: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN CPERATOR ACTIONS, EQUIPMENT FAILURES OR BOTH. 40 , ,  ; , , , , i i O STEAM CENERATOR A _gg o STEAM GENERATOR B A STEAM GENERATOR C 30 - 7

 \

m

                                                                                                              -60  $

E 6 0 e a "

   $ 20       -

e c 3 - y o e c : a -40 g s 4 10 _-

                                                                  /                                        --20 0;      -

c d a d c da de c' = e' = c' c' e c -0 0 250 500 750 1000 1250 1500 1750 2000 2250 2500 Time (s) Figure 13-8. Scenario 10 Motor driven auxiliary feedwater mass flow rate. CAUT ION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES. OR BOTH. 4450 - i i i i i i i i i -9800 0 LOOP A o LOOP B 4400 -- A LOOP C --9700 A 2 L k 4350 - a

                                                                                                         --9600 $

6 {W ; 2 E

                                                                                                             -9500

$4300 - - 3 C g -

                                                                                                             -9400 g ; ; gy gv; ; ; ; gg-g4250 g                                                                                  -

g ,

                                                                                                             -9300 4200 -M %                      '
              &             "   ^

1

                            -t - f                                                                           -9200        .

4150 i ' ' ' ' ' ' ' ' O 250 500 750 1000 1250 1500 1750 2000 2250 2500 Time (s) Figure 13 9. Scenario 10 Primary cold leg mass flow rate. 140

steam dump system switched from plant trip con- being removed from the primary system than was trol to steam pressure control mode, and the steam being generated by decay heat. dump valves closed. In steam pressure control, the steam dump valves modulate open if the secondary At s1200 s the level in the B steam generator ' pressure increases to 7.0 MPa (1020 psia). As a reached 40% of the narrow range level and auxiliary result of the steam dump valve closure the feed flow to that steam generator was terminated, secondary pressure in Steam Generators A and B thus all feed flow to the steam generators was

 .. increased, ~ however the pressure in Steam                       terminated. At that time the primary heat removal Generator C decreased as shown in Figure 13-5 due                 rate decreased and the primary pressure increased to condensation effects from auxiliary feedwater                  as shown in Figure 13-1. At sl350 s condensation entering the secondary. Auxiliary feedwater prefer-               in the steam generator downcomer above the feed
 ,  red the C steam generator because of the lower                    ring occurred in Steam Generators B and C. This pressure, and therefore the level in that generator               condensation resulted in a temporary flow reversal increased faster, reaching the 40% cutoff level at                in the secondary side that enhanced the primary to sil60 s. Auxiliary feed flow was then terminated                  secondary heat transfer, resulting in a small from the C steam generator and the B steam gener-                 decrease in pressure and a small decrease in the cold ator received all of the auxiliary feed flow. The                 leg loop iluid temperatures (see Figure 13-10).

effect of preferential auxiliary feed flow to one or the'other steam generator was to increase the overall Primary system pressure continued to increase primary heat removal rate, as shown in Figure 13-4, until sl700 s. At this time, a balance between the which in turn resulted in a depressurization of the break flow rate and the total primary inflow (see primary side (see Figure 13-1) and an increase in Figure 13-6) existed. Also, a balance between the HPl flow (see Figure 13-6). The cold leg and reac- total primary system heat removal (heat transfer tor . vessel downcomer temperatures, shown in across the steam generator tubes and energy Figure 13-10, during this time also decreased due removal at the break) and core decay power existed. to the increased cold HPl flow, and more energy Because of a balance in these thermal-hydraulic CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 566 , , , , , , , , , a DOWNCOMER 2 O LOOP A p '

           *e564 -      -

x A LOOP B

                                                                                                              -555 x LOOP C                         ,

a a

            -             n                                                                                          .,3 562     -
                                                                                                         -             o E

7 -- g g g3 -550 g

• 560 -

oo - .

           .]-       l                  ' i$ $ 2                                                                     .}

E 558 -- " --545 3 , .c-o e . E E E 556 - - 2 S -

                                                                                                              -540 S l               554 O     250 -500      750 1000 1250 1500 1750 2000 2250 2500 Time (s) l Figure 13-10. Scenario 10 Primary cold leg fluid temperatures and vessel downcomer fluid temperature.

141

mechanisms, the primary system pressure increase to remain stable throughout the 2-h period, with was terminated and the pressure remained stable to minor perturbations as the system adjusts to the the end of the calculation at 2400 s. The primary decreasing core decay power. system pressure at the end of the calculation was 9.6 MPa (1400 psia) and the reactor vessel down- Figure 13-16 shows the vessel downcomer pres-comer temperature was stable at 560 K (549*F). sure response. As discussed in Subsection 13.3.1 the slight depressurization at sl350 s was a result of 13.3.2 Extrapolations and Uncertainties. This section presents the extrapolations of the reactor

                                                                      "    **         . "I e steam generator secondary due to condensation in the upper portion of the vessel downcomer pressure, fluid temperature, wall generator downcomer. The flow reversal enhanced inside surface heat transfer coefficient, cold leg flow rates, and fluid temperatures. Also, any known                         the primary system heat removal rate, resultmg m uncertainties in the calculation are addressed.                        ae Idown of the primary and thu., a slight depres-surization. The flow reversal may have been over-An extrapolation to 2-h of the pressure, temper-                    stated and had it not occurred, the pressure would sture and heat transfer coefficient curves in the                      not have decreased, but increased as shown by the reactor vessel downcomer and the mass flow rate                        dashed line in Figure 13-16, until a balance in the and temperature curves in the cold legs are shown                      system existed as described in Subsection 13.3.1.

in Figures 13-11 through 13-15. The reactor vessel The major effect of this uncertainty was a decrease downcomer parameters are shown at an elevation in reactor vessel downcomer temperature as shown equal to the top of the core. At the termination of in Figure 13-17. When the primary system pressure the calculation the system was in a quasi-steady decreased, the HPI flow increased introducing more state, i.e., ECC flow equaled the break flow and cold liquid into the primary system. Again, had this the total energy removed from the primary system depressurization not occur' red, the temperature equaled the core decay heat, which resulted in a would have responded in the manner of the dash-stable system behavior. This behavior was expected ed line shown in Figure 13-17. CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, E0VIPMENT FAILURES, OR BOTH. 16 i i i i i i i CALCULATED

              ~
                                                                              -- EXTRAPOLATED              -2200
     ^

9 E 14 J --200'0 7 L & R -

                                                                                                           -1800     g E

u 12 - - E u

o. ct

                                                                                                           -1600     .

E E o a j10 _ - [o 8 O 1000 2000 3000 4000 5000 6000 7000 8000 Time (s) Figure 13-11. Scenario 10 Extrapolated vessel downcomer pressure at elevation equal to the top of the core. 142

CAUTION: THE' SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 564 i i .i i i i i

                                                                                                                           -555 CALCULATED 7

v-

                                                                                              -- EXTRAPOLATED                    p v

to e . s 562 - g E E g -

                                                                                                                           -550 g

g. __________________________

                                                 ,                                                                                g I 560            -

2 3

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             =                                                                                                                    ?
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e 558'- - 545 . E E 2 2 o o

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556 O- 1000 2000 3000 4000 5000 6000 7000 8000 Time (s) - Figure 13-12. Scenario 10 Extrapolated vessel downcomer fluid temperatures at elevation equal to the top of the core. CAUT10N: THE SCENARIOS SIMULATE 0 CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS,- EQUIPMENT FAILURES, OR BOTH. 44 i i i i i i i m n y CALCULATED

                                                                                             -- EXTRAPOLATED E
                                                                                                                           -7500   -

e

             .*2) 42 -                                                                                                 -            *
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               .E                                                                                                                  :E
                                                                                                                           -6500
                    -36 O       1000        2000       3000       4000         5000                6000     7000    8000 Time (s)

Figure 13-13. Scenario 10 Extrapolated vessel downcomer wall inside surface heat transfer coefficient at elevation equal to the top of the core. 143 I

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN . OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 4450. , i i i i i i -9800 o LOOP A o LOOP 8 4400- l A LOOP c --9700 . iii -- EXTRAPOLATED {oih350 - --9600 kE 45 :m e -

  $4300 ~                                                                                                   -

c 2

                                                                                                              -9400      ,,

a 4250 = -reann-------- ------------------

                                                                                                            -            a g

2 1 1

- j

                                                                                                              -9300 I

4200 g - [

                                                                                                              -9200 4150 O           1000 '2000         3000        4000            5000              6000   7000   8000 Time (s)

Figure 13-14. Scenario 10 Extrapolation of cold leg mass flow rates. CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 556 , , , , , , , I O LOOP A 7 0 LOOP B p

  **564 ' n     -

A LOOP C

                                                                                                           --555 V
                                                                                      -- EXTRAPOLATED                 ,

3 U 3 E 562

                -                                                                                          -          o E
            ~
                                             --------------------- ----.                                      -550 g
                                             --~~'--------- -----------

• 3 560 - -

3. 3 d 3 3

$ 558 -- d

                                                                                                           --545 g  .

E E 2 556 -

                                                                                                              -540 $

554

          .0           1000         2000    3000       4000            5000               6000   7000   8000 Time (s)

Figure 13-15. Scenario 10 Extrapolation of cold leg fluid temperatures. 144

                                                                                                                               )

                                                                                          ~        . , _

E L CAUT ION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN

                           . OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

564 , ,_ , , , , , , ,

                                                                                                            -555 CALCULATED
      $                                                                        -- UNCERTAINTY                     p v

a e

• s 562 -

i 3 L 0

                                                                                                            -550
 , ,                                                                        j, --- _             _

E 560 - l - 2 l z i 1 / } j558- - _-545 .

     ->                                                                                                           j 556 O         250- 500 750 1000 1250 1500 1750 2000 2250 2500 Time (s)
                            ' Figure 13-16. Scenario 10 Vessel downcomer pressure with uncertainty.

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 16 i , , , , , , , , CALCULATED

                                                                               -- UNCERTAINTY               -2200
      'T     14     -                                                                                    _

9 g -

                                                                                                            -2000    *J
       -                                                                                                             e e                                                                                                            .

R -

                                                                                                            -1800     5 0u.

12 - - l

o. a e -

                                                                                                            -1600     .

E E

       .3                                                                                                             a o
                                                                                                            -1400 i       ,      ,       ,        ,          ,         ,         ,      ,      -1200 g

0 250 500 750 1000 1250 1500 1750 2000 2250 2500 Time (s) Figure 13-17. Scenario 10 Vessel downcomer fluid temperature with uncertainty. 145

13.4 Conclusions more than 5.5 K (10*F) from the initial steady state value. Also, because the pumps did not trip, multi-As a consequence of the primary system pressure dimensional effects in the loops and reactor vessel not decreasing below 9.1 MPa (1315 psia) during the were negligible, therefore, multidimensional mixing transient, the reactor coolant pumps were never trip- calculations for this transient are not needed. The ped. This resuked.in good thermal nuxmg of the minimum reactor vessel dowTwomer temperature was pnmary coolant fluid with the incoming ECC fluid 556.9 K (542.7'F), and the maximum subsequent and the vessel downcomer temperature did not vary pressure was 9.6 MPa (1400 psia). = e O

                                       +

146

14. SCENARIO 11, LOSS OF SECONDARY HEAT SINK WITH PRIMARY SYSTEM FEED-AND-BLEED RECOVERY

        'Ihe following subsections describe the results of                (SGA) inlet after RCP trip, and the dif-the Scenario 11 calculation. This calculation was                    ference between the minimum temperature performed to evaluate the consequences of a                          and the current temperature. This tempera-
 ,   postulated loss of secondary heat sink accompanied                   ture difference was used to start HPI and by an uncontrolled primary feed and bleed type of                    open the pressurizer PORVs.

recovery. The feed and bleed was performed by opening both pressurizer power-operated relief The transient was initiated from the 2300 MW valves (PORVs) and initiating high pressure injec- full power steady state conditions described in tion (HPI). Subsection 2.3.1. A description of the scenario is provided, fol- 14.3 Results lowed by a discussion of the model changes required to conduct the calculation. The calculated data, the The results of the Scenario 11 calculation are extrapolated results, the uncertainties in the calcula- presented in this section. Included in the discussion tion, and the conclusions regarding the calculation of the results are the assumptions made in extrap-are also presented, olating the calculated results beyond the end of the RELAPS calculation. Scenarios investigated in this report generally include conservative assumptions concerning equip- The first subsection describes the RELAPS ment failures, operator actions, or combinations of results. A discussion of the method used to extrap-these. Conclusions relative to pressurized thermal olate the calculated data and the uncertainties shock severity are not to be drawn directly from the involved in'the calculation are presented in the results presented in this report (see Section 18). second subsection. 14.1 Scenario Description MM CaWadon Results. The results of the RELAP5 calculation are described :n this section. A synopsis of the events that occurred in the calcula-A description of Scenario 11 is provided in tion are presented in Table 14-2. Table 14-1. As with the other calculations, this scenario was developed by Oak Ridge National The reactor vessel downcomer fluid temperature Laboratory (ORNL). The sequence of the calcula- response is shown in Figure 14-1. The initial tion followed the scenario description through the increase in downcomer temperature at the start of required 7200.s. The calculation was extended for the transient was due to a combination of secondary another 900 s to more adequately address the pressure increase and decreasing feedwater flow. pressurized thermal shock (PTS) question. The increase in secondary pressure was due to the closing of the turbine stop valve [which was initiated by the tripping of the main feedwater (MFW) 14.2 Model Changes pumps). The constart downcomer temperature dur-ing the first 3600 s was the result of the plant trip The model described in Section 2 was changed control system (PTC) response, which was designed as described below to satisfy the operator action to control the steam dump valves to bring the requirements given in Table 14-1. primary average temperature down from the full o power value (575 K,575.4*F) to the no-load set-

1. The RCP trip was changed from a low point (559.3 K, 547'F).

nressure trip (9.07 MPa,1315 psia) to an evaluation of any SG WR level below 5%. The reactor vessel downcomer pressure response is shown in Figure 14-2. The initial decrease in pressure

2. Additional control variables were added to was due to a reactor trip caused by the turbine trip.

the model to monitor the Loop A hot leg The reduction in core power resulted in a coincident temperature, determine the minimum reduction in the hot leg temperatures which caused

temperature at the Steam Generator A a shrinkage of the pressurizer liquid volume. The 147 l

1. Tatdo '14-1. Scenario' description No.' .11

                                                      ~

P'lant Initial State - Just prior to transient initiator General

Description:

100% power steady state 1 System Status Turbine: Not latched, TSVs closed . . Secondary PORY: - Automatic control Steam Dump Valves: - Automatic control. Charging Systems: Automatic control Engineering Safety Features: - Automatic control Pressurizer PORVs: : Automatic control . Reactor Control: Automatic control-

               -Main Feedwater: . Automatic control
              ; Auxiliary Feedwater: Automatic control Main Steam Isolation Valves (MSIVs): Automatic control Main Feedwater Control Valves: . Automatic control Transient Initiator -

Both Main Feedwater pumps trip simultaneously. Equipment Failures That Occur During the Transient if the Equipment is' Demanded Auxiliary Feedwater pumps fail to start. Operator Reactions to Reported Information J1. -Operator trips reactor coolant pumps (RCPs) when 1/3 Steam Generator (S/G) wide range

                 ~( WR) levels decrease below 5%.
         - 2. - Operator initiates safety injection (HPI) and opens the pressurizer' PORVs after RCP trip and when the A Loop hot leg temperature has increased 5'F.

Jreduction in pressure ended as the hot leg . sides of the steam generator tube bundles at 3818 s,-

     ~ temperatures stabilized at 560 K (548'F) and the             which ended by 3878 s due to voiding in the top of pressurizer heaters began recovering the pressure.          .the tube bundles and reactor vessel upper plenum The primary system pressure - was restored to              - and. head. The 0.70 MPa (102 psia) pressure 15.51 MPa (2250 psia) by 650 s and remained there           increase between 3878 s and 3%2 s was due to the until the RCPs were tripped.                                 refilling of the primary coolant system by the HPI and CVCS, which compressed the voids in the reac-The RCPs were tripped at 3626 s when the WR               tor vessel and steam generators. The subsequent level in SGA had decreased to 5% of full range. The          pressure decrease, starting at 3%2 s, was caused by reduction in core flow caused the hot leg tempera-           the resumption of primary-to-secondary heat ntures to increase. This temperature response con-             transfer as natural circulation was reestablished.

tinued until 3718 s when the pressurizer PORVs

     .- were opened and HPI initiated. These actions were

• taken, per the scenario description, .when the The mass flow rates of the pressurizer PORVs, Loop A hot leg temperature had increased by HPI and CVCS are shown in Figure 14-3. HPI was initiated at 3781 s, but no coolant was injected into 2.78 K (5'F). Initiation of HPI was assumed to require a manual SIAS which in turn caused _ the cold legs until the primary system pressure had ,

letdown to be isolated. decreased to the HPI shutoff head (10.14 MPa, 1470 psia) at 3767 s. The pressurizer indicated level The opening of the PORVs at 3718 s caused the began increasing when the PORVs were opened at

     . primary ' system' to depressurize to 6.75 MPa                3718 s. The increase in level was due to the insurge -

(979 psia) by 3878 s. The opening of the PORVs of primary system coolant as the pressurizer upper also resulted in reversed flows in the hot leg head depressurized. The indicated level was 100% by 148 l

Table 14-2. - Scenario 11 sequence of events Time (s) Event 0 Transient initiated by manually tripping MFW pumps Reactor and turbine tripped - PTC initiated 5 MFW valves closed (feed train isolated from steam generators) 100 Primary pressure decreased to 14.27 MPa (2070 psia) 650 Primary pressure restored to 15.51 MPa (2250 psia) 1500 SGC FW header blowdown began 1700 SGB FW header blowdown began 1900 SGC feedwater header blowdown completed 2100 SGB feedwater header blowdown completed

    '3626     SGA WR level reached 5% - RCPs were manually tripped.

3718 Loop A T-hot increased by 2.78 K ($*F) Pressurizer PORVs opened and SIAS tripped on 3767 Primary system pressure reached HPI shutoff head 3817 PTC switched to SPC 3830 Vessel upper head began voiding. 3857 Natural circulation in loops degraded due to combination of voiding in hot leg and PORV flow. 3878 Pressurizer indicated level reached 100%. 3937 Natural circulation resumed 3960 Reactor vessel upper head completely voided. 4093 SGB secondary became primary system heat source 4096 SGA secondary became primary system heat source 4100 SGC secondary became primary system heat source 4444 Natural circulation ended in Loop C 5500 SGA feedwater header blew down.

• 6000~ HPI and CVCS inflow exceeded PORV outflow 6026 Condensation depressurization in the Loop C pump suction.

,-   7159     Condensation depressurization in reactor vessel upper downcomer 7160     Accumulator injection initiated 8100     End of calculation 11000     End of extrapolation 149

CAUTION: THE SCENARIOS SIMULATED

                        .CONTAIN SIGNIFICANT CONSERVATISMS IN
                  - OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

580' i i i i 560 - .

                                                                                           --550              .,

3 m f ^ 5 540 - - D

                                                                                              -500    ,       ,

5 5 .E 520 - - - - E t a f-g 500

                                                                                              -450 [

y 480_- h 400 k

  -460 O              2000            -4000            6000           8000           10000 Time (s)

Figure 14-1. Scenario 11 Reactor vessel downcomer temperature. CAUTION: THE SCENARIOS SIMULATED

                         ~ CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

20 i i i i

                                                                                              -2500 15     [

m- -

                                                                                              -2000 g E                                                                                                       %

E- a e - 1500 e 10 ~- - a a g -

                                                                                              -1000 g 5      -                                                                              -
                                                                                              -500 0                                                                                         O O             2000             4000            6000           8000           10000 Tirae (s)                                                -

Figure 14-2. Scenario 11 Reactor vessel downcomer pressure. 150

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 100 , , , , O PORV MASS FLOW -200 0 HPl MASS FLOW 80 - l A cycS MASS FLOW -

                                                                                                         -15 0 m
                                                                                     ,,g                         g 3        (03 60    -

a t N E

                                                                                                         -10 0

{ 40 - Ilj 3

          =                                          !

e 20 ~ - 1 _

                                                                                                         -50     g s

O MC  : Uo^t^ ^ooo'ocoo^ --O c -20 O 2000 4000 6000 8000 10000 Time (s) Figure 14-3. Scenario 11 Primary mass inflows and outflows. 3878 s, but the pressurizer upper head remained The constant downcomer temperature response partially voided for the rest of the transient because between 4090 s and 4444 s was the result of nearly the PORV flow rate was greater than the pressurizer constant flow conditions and heat additions to the insurge caused by the refilling of the primary system primary coolant from the core and steam generator by the HPIS and CVCS. secondaries that were offset by heat removal due to the open PORV and the HPI fluid. The heat Isolation of the CVCS letdown valve at 3718 s transfer from the secondary to the primary system resulted in a net CVCS injection flow rate of was positive for the remainder of the calculation. 3.41 1/s (54 gpm). The increase in net injection flow rate was offset by the pressmizer level control The minor temperature oscillations between system level error, and the net injection flow 4500 and 4900 s were due to flow effects in the decreased to 0.951/s (15 gpm) by 3787 s. C loop. During this period of time, SGC began voiding in the top of the tube bundle. This voiding Temperatures in the vessel downcomer (as shown induced a flow oscillation in the C toop, which in Figure 14-1) decreased between 3768 and 3817 s affected the amount of HPI fluid that flowed into due to the HPI fluid from the three cold legs. The the vessel downcomer. These oscillations were even-increase in temperature between 3817 s and 3937 s tually damped out when the tube bundle was suffi-o was the result of opening the PORVs, which caused ciently voided to prevent further natural circulation reverse circulation in the three loops and resulted in that loop. Positive cold leg flow in the three loops in core fluid circulating back through the for the next 900 s resulted in a steadily decreasing downcomer. The resumption of positive circulation downcomer temperature. at 3937 s resulted in a rapid temperature decrease due to the HPI fluid entering the downcomer. This The relatively large decrease in downcomer decrease ended after the fluid in the cold legs was temperature at 6026 s was caused by condensation flushed through the downcomer, followed by the depressurization in the Loop C pump suction and warmer fluid from the hot legs. subsequently in the vessel. The decrease in pressure 151

caused an increase in cold leg and HPI flow from Another condensauon-induced depressurization the A and B loops into the vessel downcomer. This occurred in the upper downcomer of the reactor temperature oscillation lasted N15 s and ended vessel at 7159 s. The pressure in the cold legs was when the loop seal was refilled. s4.76 h1Pa (690 psia), and the depressurization of the primary system was sufficient to reduce the ne Dow rate responses of the three cold leg pressure to the accumulator injection pressure discharges are shown in Figure 14-4. He flow rates re- (4.62 h1Pa,670 psia), mained essentially constant until the RCPs were trip- . ped at 3626 s. The flow rates then-decreased to The downcomer temperature and pressure at the 136 kg/s (300 lbm/s) by the time the HPI shutoff head end of the calculation were 468.6 K (383.8"F) and was reached (3767 s). He injection of HPI fluid 4.04 N1Pa (585 psia). The response during the last resulted in a brief period of flow rate increase before 100 s indicated a temperature decrease of 90.9 K/h the initial voiding of the tube bundles resulted in flow (163.6

• F/ h), and a pressure decrease of stagnation. Natural circulation was reestablished at 3.72 h1Pa/h (540 psia /h). It should be noted that 3937 s in the three loops as the refill of the primary the downcomer temperature is a combination of the system collapsed the voids in the tube bundles. The mass three loop temperatures. The A and B loop cold leg flow rates through the three loops then increased until temperatures downstream of the HPI and accumu-3980 s as the reactor vessel upper plena and head drain- lator injection point were 479.5 and 478.7 K ed into the hot legs. Flow rates then decreased to the (403.4 and 402.0 F), respectively; while the C toop natural circulation rate of 70 kg/s (154 lbm/s). The A cold leg temperature was 315.8 K (108.8'F).

and B loops maintained natural circulation for the remainder of the calculation. However, the Loop C . steam dump mass Dow rate gponse h

• f"in Figure 14-5. The large flow spike at the tube bundle voided at 4400 s, as a result of the PORV flow, and the Imop C flow began to decrease. He flow begmm.ng of the transient was due to the response rate in Loop C mntinued to decrease until 4444 s, and f the plant trip controller (PTC). The function of was essentially stagnant for the remainder of the tran- the PTC is to reduce the primary average temper-sient, except for the oscillations calculated between a ure the nohad value, M3 K WM h difference between the actual average temperature 4500 s and 4900 s as discussed below.

and the setpoint temperature determines the steam dump valve arca. At the beginning of the transient The flow oscillations calculated for Loop C the primary average temperature was 575 K between 4500 s and 4900 s occurred when the SGC (575.4*F), which represented a PTC error of suffi-tube bundle refilled during the initial period of cient magnitude to cause the steam dump valve to reverse flow (4444 to 4482 s) and then began voiding from the top down. The resultant bubble ".ip almost completely open. This large error rapidly m the two upper volumes of the SGC tubes tended dimmished due to the effect of the reactor trip. The rapid decrease in mass flow rate following the initial pres ur n he empt ng er o s Thi proc s o 8[g**d compression-decompression contmued until 4860 s, - g (lead / lag controllers tend to overrespond to sudden when the void fractions m the two upper volumes changes in the input variable). The initial of the tube bundle were sufficiently large to damp temoerature variations were gone within the first out further oscillations. The tube bundle continued 50 s of the transient and were followed by a more to void until it was completely dramed (4900 s). stable period of temperature behavior. A condensation depressurization of the Loop C The minor oscillations in steam dump flow rate at pump suction at 6026 s resulted in an increase in 650,1450,2300 and 3150 s occurred as a result of the A and B cold leg loop flow rates and a reverse changes in the heat transfer regimes in the SG boiler , of the C loop cold leg flow rate. The pump suction region as the secondary mass decreased. The chang-depressurization stopped when the volume refilled, ng heat transfer regimes caused minor fluctuations but the reverse flow induced a condensation depres- in the primary average temperature, which induced surization of the reactor vessel upper downcomer changes in the steam dump valve area demand. These . by drawing in cold water from the A and B loop oscillations were relatively short-lived and did not cold legs, which reestablished positive flow in affect the results of the calculation. Loop C cold leg. These loop flow oscillations con-tinued (but with lower magnitudes) for the The increase in steam dump mass flow rate at remainder of the transient. 3626 s was due to the effect of tripping the RCPs. 152

                                            ' CAUT I ON: THE SCENARIOS SIMULATED
                                                   ~

CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 1000 , n , , ,

                                                                                                              - 2000
                       '800     -                                                                           -
         ~-             600     -                                                                           -
                                                                                                              ~
                 ^

a 400 ~ - - 7 N .N Q E

         .. J 200,          -                                                                           -

g { O-- L ,

                                                                                    ;=

rr r 3y -

                                                                                                            --O         g
                  .. -200       -                                                                          -

m i O I

                '2 -400,-                       O LOOP. A MASS FLOW                                        ~- -1000 o LOOP B MASS FLOW
                      -600      -

A LOOP C MASS FLOW

                                                                                                           ~
                      -800      -                                                                          -
                                                                                                              - -2000
                     -1000 0           2000               4000           6000          8000           10000 Time (s)

Figure 144. Scenario 11 Cold leg flow. CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS. IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 600- i i i i

                                                                                                              -1200
                                                                                                              -1000 k400 f           -      -
                                                                                                              -800   .o 3

C 3

                                                                                                              -600   j j 200_-

_4ag j

                                                                                                              -200
       '4                                                -

y m_ L .-

                                                                               '             '           l 0                                    '.                                               0
                       -1000               0               1000'          2000          3000            4000
                                                 ,                Time (s)

Figure 14-5. Scenario 11 Steam dump flow. 153

The RCP trip reduced the flow rate in the primary downcomer pressure, temperature and heat transfer coolant loops and consequently increased the coefficient and the cold leg discharge volume mass primary system average temperature by increasing flow rates anu temperatures. the temperature rise across the core. The increase in the primary average temperature resulted in an The reactor vessel downcomer pressure response increased demand on the steam dump valve area for the last 4100 s of the calculation (solid line) is from the PTC. The decrease in steam dump mass shown in Figure 14-8. The primary system pressure flow rate at 3718 s occurred when the pressurizer is expected to continue decreasing steadily to . PORVs were opened, causing a decrease in the 2.29 MPa (333 psia) at 11000 s. primary system average temperature, and a cor-responding overresponse in the PTC signal. The reactor vessel downcomer fluid temperature response for the last 4100 s of the calculation (solid , The flow reversal in the primary system due to line)is shown in Figure 14 9. As can be seen in this the PORVs opening caused a temporary increase figure, the downcomer temperature was essentially in the primary system average temperature with the independent of the condensation depressurization expected steam dump valve response. The steam events at 6000 and 7200 s. The temperature dump valves closed at 3817 s when the primary response during this period of time was dominated system average temperature finally decreased below by the A and B loops. These loops were only the no-load setpoint of 559.3 K (547'F) and the slightly affected by the accumulator injection which PTC was switched over to the steam pressure con- began at 7159 s. trol mode (SPC). The SPC regulates the steam dump valve area based on secondary header pres- The reactor vessel downcomer fluid temperature sure. This pressure never became great enough to is expected to continue decreasing at a constant rate open the steam dump valves during the remainder of 57.8 K/h (104*F/h) through 11000 s (dashed of the transient. line). The final temperature based on the estimated The SG wide range (WR) level responses are ra e tempaatum decease b W K @% shown in Figure 14-6. The initial oscillations in SG levels were due to changes in the secondary read r vessel wnc me waH cat tranda me cent mponw during tM last 4100 s of the pressures in response to the closure of the turbine stop valve and the opening of the steam dump valve.

                                                                   "" "                                 UnQ *wnm, gum M W relatively large sp.kes                     i   m the calculated data were the The relatively smcoth decrease in the SG levels from result of oscillations m the downcomer pressure, s50 to 3817 s was due to the removal of the secondary mass through the open steam dump                   tempuaye d man hw ram h dasns valve. The change in the slope in SGB and SGC                w"e udn part to 6e mponw oprimary system  ,, ,

t acmniu ato mjecuon events. Dunng pen,ods of levels between 1500 and 2l00 s was due to the accumulator m, {jection colder fluid was circulated into respective feedwater headers blowing down as the ma ene wnmma m legs. M SG pressures decreased. The SG A feedwater header c Ider fluid resulted in the higher heat transfer coeffi-did not blow down until 5500 s, and as a result, the

                                                               #"IS'                            #*

• 8" * ## #"" "

SGA WR level was the level that tripped the RCPs c e ic ent am quama6ely masonaW, k may & at 3626 s' overstated. The heat transfer coefficient is expected to The increase in SGA mass starting at $500 s did remain nearly constant through 11000 s (dashed line). not affect the results of the calculation because the The final value is estimated to be between 0.7 kW/m2 temperature of the feedwater was approximately the K (0.034 BTU /s 2ft aF) and 1.3 kW/m2g same as the existing secondary temperature and the (0.0M BTU /s ft2 .p), heat transfer during this period of time was essen-

• tially zero. This is shown in Figure 14-7, whwh The calculated cold leg mass flow rate responses dur-presents the primary-to-secondary heat transfer ing the last 4100 s (solid line) are shov n in response for the three SGs. Figure 14-11. The Loop A and B discharge flo.v rate responses were essentially the same. As can be seen 14.3.2 Extrapolations and Uncertainties. Uncer- in Figure 14-11, the A and B loops behaved rather tainties regarding the results of the Scenario iI predictably and indicated no unusual response. The calculation and predicted conditions at the end of the mass flow rates should remam nearly constant through 2-hour period of interest are presented in this section. the last 2900 s (dashed line) due to the effect of the The parameters to be addressed are the vessel driving head produced by the core decay heat coupled 154

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 0.8 , , , , , a SG A WR LVL o SG B WR LVL

      ,                                                                          A SG C WR LVL 0.6  -

7 _3 p= , E 0.4 - t  := 4 n E 5 z 0.2 -

                                                                                         ~ f i

0.0 e T - - MM s g gg

                   -2000             0        2000         4000        6000           8000      10000 Time (s)                 ,

Figure 14-6. Scenario 11 Steam generator wide range levels. CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 800 , , , i i 0 CG A WR LVL O SG B WR LVL A SG C WR LVL 600 - - m

• 400 - -

d t [200 - -

• C 0 -

                                                          * %:::::::::::E:                          -
       ,o
                  -200
                  -2000              0        2000         4000        6000           8000      10000 Time (s)

Figure 14-7. Scenario 11 Steam generator heat transfer rates. 155

                            .i Qt:

                                            ' CAUT I ON: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

20- , , , , , , CALCULATED

                      .                                                                                        -- EXTRAPOLATED                -2500
             '15        -

q, -

                                                                                                                                              -2000 g

u. -

5 $ '

       's ' 10 --                                                                                                                           --1500        e

. 5 g -

                                                                                                                                              -1000      g
                         -~
              '5                                                                                                                            -
                                                                                                                           ~~~_ ___,          -500
              .0
                                   '                                                  '       '           i           '          '

O 4000 -5000 6000 7000 8000 -9000 10000 11000 Time (s) Figure 14-8. Scenaric 11 Reactor vessel downcomer pressure response,4000-11000 s. - CAUTION: THE SCENARIOS SIMULATED - CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES - OR BOTH. 600 , , , , , ,

                                                                                                                                              -600 CALCULATED
                                                                                                               -- EXTRAPOLATED 550          -'                                                                                                                  -

m m 6 -

                                                                                                                                              -500 t 500          -

5 t -400

o-td D h- ,. s 5

 .   +                                                                                                              -                               -

450 - ss s _300

          '400-                                                                                                    '           '

4000- 5000- 6000 7000 8000 9000 10000 110C0 ' Time (s). Figure 14-9. Scenario 11 Reactor vessel downcomer temperature responses,4000-11000 s. 156

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EOUIPMENT FAILURES, OR BOTH. E5 i , , , , , c) . CALCULATED E -- EXTRAPOLATED

                                                                                                                                                                  -800
    -                                                             4      -

v E. 3

                                                                                                                                                                  -600      E f3                       -
                                                                                                                                                                  -400 g2                     -

g3

                                                 -                       4, 9                                         .

4' mm E p )i; l  ; L '"

                                               .b                            1 l' ' '    -.

l l ll# ----------------- 1

                                                                                                                                                                -_200     g

_ 2 E I O 0 4000 5000 6000 7000 8000 9000 10000 11000 Time (s) Figure 14-10. Scenario 11 Reactor vessel downcomer wall heat transfer coefficient 4000 11000 s. CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 500 i i i i i i 0 LOOP A MASS FLOW

                                                                                                                                                                  - 1000 400          -

o LOOP B MASS FLOW a LOOP C MASS FLOW 300 -

                                                                                                                                   -- EXTRAPOLATED S
                                                                                                                                                                  ~
                         ^

a 200'y - N N , y 100 h - " N4

• M--~--------------- g E

g o._ w Mpp ll M h----------------- 0 , I

                               . -100                                   -

I I e o N c 2 -200_- - _ -500 $

                                               -300                       -
     ,                                         -400                     -
                                                                                                                                                                  - -1000
                                                -500 4000             5000        6000          7000             8000         9000        10000   11000 Time (s)

Figure 14-11. Scenario 11 Cold leg discharge mass flow rate responses, 4000-11000 s. 157

with the effect of the gradually increasing HPI flow. function of the HPl and accumulator injection flow The flow rates of these two loops are expected to rates and the direction of flow in the Loop C cold be N100 kg/s (220 lbm/s) at 11000 s. leg discharge piping. The C loop cold leg discharge trass flow rate

14.4 CONCLUSION

S response was significantly affected by accumulator injection and condensation depressurization events. Conclusions regarding the results of the This was due to the nearly stagnant flow conditions Scenario 11 calculation are presented in this section. existing in this loop. The large flow oscillations A compilation of the values of the parameters between 7200 and 7400 s were due to the effects of described in Subsection 14.3 will be presented. accumulator injection. These oscillations could be qualitatively correct, but in all likelihood, are The extrapolated values from the Scenario 11 < significantly overpredicted. These oscillations are calculation corresponding to 11000 s are summar-expected to continue for the duration of the desired ized in Table 14-3. period of the calculation, but at a reduced magnitude as the primary system continues to Extrapolation of the primary system pressure depressurize. response past 11000 s indicates that the system pressure will continue to decrease until the low The Loop C cold leg discharge mass flow rate will pressure injection (LPI) shutoff head is reached continue to equal the HPI plus accumulator injec. (0.97 mpa, 140 psia). Initiation of LPI will tion flow rate for the last 2900 s (dashed line) since thereafter maintain the primary system pressure at this loop would be stagnant without these contribu. the LPI shutoff head as long as the PORV is open. tions. The flow rate at 11000 s is expected to be approximately 10 kg/s (22 lbm/s). . The reactor vessel downcomer fluid temperature is expected to continue decreasing due to the The cold leg temperature responses of the three influence of the HPI, accumulators and LPI. The cold leg discharge volumes for the last 4100 s of the minimum fluid temperatures will continue to be calculation (solid line) are shown in Figure 14-12. f und in the Loop C cold leg discharge volumes. The Loop A and B responses were predictably The reactor vessel downcomer wall heat transfer smooth since the flow rates through these two loops coefficient should remain essentially constant except were nearly constant and nonoscillatory in during periods of accumulator or LPI which will behavior. The Loop C discharge volume tempera-cause brief periods of increased flow and decreased ture was significantly more erratic. The relatively large temperature oscillations at 7200 s and 7700 s M Th % nbId fluid properties will tend to increase the heat were the result of reverse flow during periods of ac-transfer coefficient; however, these periods are cumulator mjection. The reversed flow events were expected to be transient in nature and should not caused when the Loop C cold leg suct,on i depressur-significantly affect the overall heat transfer ized, thereby inducing the warmer downcomer fluid coefficient response. to flow mto the cold leg discharge volume. The subsequent decreases in the Loop C discharge The cold leg discharge mass flow rates are expected temperature were due to the resumption of positive to remain essentially constant due to the driving head flow back into the downcomer. These oscillations provided by the core decay heat. The SGC bundle are the result of condensation depressurization is not expected to refill due to the open PORV. If events. the PORV were to be closed, it is expected that the SGC tube bundle would refill and the primary system The Loop A and B cold leg discharge tempera- loop mass flow rates would eventually become about - tures are expected to continue deceasing at a con' 70 kg/s (154 lbm/s) in all three loops. stant rate of s60 K/s (108*F/s) for the last 2900 s (dashed line). The final temperature of the fluid in The cold leg discharge fluid temperatures in the these two loops will be s420 K (296*F). A and B loops are expected to continue decreasing . due to the effect of the HPI, accumulator, and LPI The Loop C cold leg discharge temperature is contributions. The Loop C cold leg fluid tempera-expected to vary between 300 K (80*F) and 305 K ture was essentially at the HPl fluid temperature (90*F) at i1000 s. The final temperature will be a prior to 11000 s. 158

p CAUTION. THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

 -        600              ,          ,            ,         ,            ,          ,
                                                                                                 -600 0 LOOP A o LOOP B a LOOP C o             {[-    ? :                                         -- EXTRAPOLATED                -500
                                                                   ~
                                                                                                 -400 3                                                            N 0

L l h

                                                                        ~,,r,'   s, n

g E -

                                                                                       's        -300 i b400 5
                                                                                                 -200
                                                                  ~ - - - - - - - - - - - - - -  -300   .

4000 5000 6000 7000 8000 9000 10000 11000 Time (s) Figure 14-12. Scenario 11 Cold leg discharge volume temperature responses, 4000-11000 s. Table 14-3. Scenario 11 extrapolated values Parameter Value Downcomer pressure 2.29 MPa (333 psia) Downcomer temperature 422 K (300'F) Downcomer wall heat transfer 0.7-1.3 kW/m2 .K coefficient (0.034-0.064 BTU /s ft 2 . p) Cold leg mass flow rates:

 ,            Loops A and B                                       100 kg/s (45 lbm/s)

Loop C 10 kg/s (4.5 lbm/s) 4 Cold leg discharge temperatures: Loops A and B 420 K (296*F) Loop C 303 K (85*F) 159

15. SCENARIO 12,2-IN. COLD LEG BREAK AT FULL POWER The following section presents the transient discusses the results of the calculation. The second scenario description, modeling changes effected to subsection discusses the extrapolation of the results perform this calculation, detailed analysis of the to 7200 s and related uncertainties.

transient results, extrapolations and uncertainty analyses, and conclusions drawn from the analysis 15.3.1 Calculation Results. The sequence of , for Scenario 12: a small cold leg break at full power. events for this simulation is summarized in Table 15-2. After break initiation the primary Scenarios investigated in this report generally system rapidly depressurized (Figure 15-1 shows the include conservative assumptions concerning equip- first 200 s of the calculation) activating the ment failures, operator actions, or combinations of pressurizer proportional and backup heaters at these. Conclusions relative to pressurized thermal I and 3 s, respectively. The loss of pressurizer level shock severity are not to be drawn directly from the caused the CVCS flow to increase as shown in results presented in this report (see Section 18). Figure 15-2. By 16 s the reactor had tripped on a 2/3 over-temperature delta T signal. The subse-quent C re power reduction enhanced the 15.1 Scenario Description depressurization rate. The reactor trip caused the closure of the turbine stop valves, which induced The transient was initiated from full power steady a temporary increase in the secondary pressures. state (nominal temperature and pressure), and all The reduction of core power, the opening of the control systems were in automatic control. The steam dump valves, and AFW activation reversed transient was initiated by a 0.0508-m (2.0-in.) the initial steam generator pressure increase, diameter hole appearing in the bottom of the horizontal section of the A loop cold leg between At 18 s the steam generator level control system the RCP discharge and HPI injection location. It caused the partial closure of the feedwater regula-was assumed that all systems operate automatically tion valves as a consequence of a steam /feedwater as designed. The only operator action assumed to flow mismatch signal. By 23 s the main feedwater take place was tripping of the reactor coolant pumps valves had closed and the motor driven AFW when the primary pressure reached 9.1 MPa pumps were activated. At 37 s, a two-out-of-three (1315 psia) if a SIAS signal had been generated. steam generator low-low signal initiated the steam Instead of controlling AFW to 40% steam generator driven AFW system. The initiation of the motor and level, AFW was continued untilliquid carryover to steam driven AFW systems did not have a signifi-the steam lines occurred. A transient scenario is cant effect on the primary side pressure response. provided in Tatle 15-1. At 24 s the pressurizer heaters were deenergized when the pressurizer level dropped below 14.4% of 15.2 Mcdel Changes full se le. This event caused a slight increase in the primary system depressurizat,on i rate. Changes made to the steady state model to initiate During the initial 200 s of the transient, the break the small ce!d leg break included the addition of flow trend (Figure 15-3) followed the primary system a break valve in the A loop cold leg connected to pressure response. After this period the primary loop a time dencadant volume set at atmospheric con- mass flows transitioned to a natural circulation regime ditiom. The break components were set to repre- (Figure 15-4) and the AFW switched to Loop B sent a beak at the bottom of the cold leg pipe. The (Figure 15-5). The mechanism that caused the AFW valve was set to open at the initiation of the to cycle to different steam generators is the same as

• transient. discussed in Subsection 8.3.1.

Changes in the primary to secondary heat transfer 15.3 ResultS , and primary mass flow conditions eventually induced large changes in the break flow response The following section details the results of the beginning at 485 s. The sudden increase in the break Scenario 12 calculation. The first subsection mass flow rate was caused by the onset of loop flow 160

Table 15-1. . Scenario description Nr.12 Plant Initial State - Just prior to transient initiator General

Description:

100% Power steady state System Status Turbine: Automatic control Secondary PORV: Automatic control Steam Dump Valves: Operative / automatic control Charging System: Automatic control Pressurizer: Automatic control Engineering Safety Features: Automatic control PORVs: Automatic control

• ~ _ Reactor Control: Automatic Main Feedwater: Automatic control MSIVs: Open, Automatic control MFIVs: Open, Automatic control Transient Initiator A 2.0-in. hole appears in loop A cold leg.

Equipment Failures That Occur During the Transient if the Equipment is Demanded. None l Operator Reactions to Reported Information

1. If Safety Injection Actuation Signal (SIAS) is generated, the operator will trip the reactor coolant pumps when RCS pressure reaches 1300 psig.

2. Stop AFW flow to each S/G when liquid carryover is observed in the main steam line.

stagnationin Loops A(breaklocation)and C.The At 915 s the AFW switched from Loop B to C. stagnation of the loops was a direct consequence There was a transition period during which natural of degradation of primary to secondary heat circulation terminated in Loop B but had not transfer in Steam Generators A and C due to the developed in Loop C. This is observed in the preferential flow of AFW to Steam Generator B. reduction in downcomer flow (Figure 15-8). The The loop stagnation enabled HPI liquid to convect reduced core mass flow induced a temporary to the break. Presented in Figure 15-6 are the fluid repressurization of the primary system. At 1230 s and saturation temperatures upstream of the break. repressurization was terminated when a temporary Coincident with the increase in fluid subcooling was steam path to the break increased the volumetric the increase in break flow at 485 s. break flow rate. At 1350 s Loop C natural circula-tion had developed, and as a consequence the core At 750 s the primary depressurization rate vapor generation rate was reduced so that subcooled increased (Figure 15-7) as a consequence of sub- conditions at the break (Figure 15-6) were o cooled ECC liquid condensing steam in the vessel reestablished. An anomaly concerning break mass inlet annulus, enhanced heat transfer between the flow is evident at this time in Figure 15-3. This is primary and secondary systems, and reduced core further discussed in Section 15.3.2. vapor generation. During this period, the reactor

• vessel downcomer mass flow (Figure 15-8) increased Similar behavior in the primary pressure response as a consequence of enhanced natural circulation occurred when the AFW switched from Steam in Loop B. The enhanced flow was caused by an Generator C to A at 1707 s. The low primary mass increased static pressure head difference between inventory at that time prevented the development the vessel core and downcomer, of steady natural circulation flow conditions in 161

m

 ^

J

                                                    ~

Table 15 2.5 Scenario 12 sequence 'of events Time , Event

              ~

(s) 0.~  : Loop A~,0.051 m (0.167 ft) cold leg break initiated -

                                    - 1.0 -       Pressurizer heaters activated
                                  ~16.0'          Reactor tripped, turbine tripped, and turbine stop valves close 23.0 -       - Main feedwater valves closed main feedwater pumps tripped, and motor driven AFW activated
  =

24.0 Pressurizer heaters off due to pressurizer level dropping below 14.4% 7 30.0 ~ SI signal generated on low pressurizer pressure 3'7.0 = Steam driven AFW activated 50.0- HPI flow initiated, pressurizer emptied 69.0 RCP tripped, upper head began to void 200.0 Primary loops transitioned to natural circulation 216.0 ~ Steam dump control valves closed 230.0 . AFW switched to Steam Generator B

     ~

g 870.0 Accumulator flow initiated 915.0 AFW switched to Steam Generator C 1707.0~ AFW switched to Steam Generator A

                               -2315.0           Calculation terminated
                               -7200.0           End of extrapolation l

Loop A before the calculation was stopped. After previously mentioned periods of primary side a period.of repressurization from about 1750 s to repressurization. As a consequence, there was not L 1900 s there was sufficient oscillatory loop flow to enough ECC available to induce an increase in t'te induce primary system depressurization. primary system mass inventory before the calcula- . tion was stopped. At the termination of the calcula-At 470 s the primary system pressure dropped - tion, the reactor vessel upper head, most of the below 4.6 MPa (673 psia), which inithted accumu- upper plenum,' and steam generator tubes were. lator _ injection. During periods of accumulator steam filled. . injection the total ECC flow generally exceeded the break mass flow rate. However, the accumulator .During the period from 2100 to 2315 s, code flows were not continuous. The principal cause of problems developed that were characterized by large this lack of continuity in accumulator flow was the CPU /real time ratios. It was suspected that code -

                                                                           ~ 162 L

r:- 1 CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 20 i i i a SG A o SG B -2500 a SG C X PRIMARY ^ n ~ 15 - y RCP trip $ a -2000 3

     **     y                                                                                             y
                                                                                                 --1500   g
           ' ! .10 --

5 't 5 E -

                                    - 2 = = =
                                    =
                                                                          - - - - - = = --1000
                                                                                          -               l
                       -                                                                           -500
                                            '                   '                    '              0 0

0 50 10 0 15 0 200

                                                         -Time (s)

Figure 15-1. Scenario 12 Primary system pressure. CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 200 i i ,

               .n E
                -a 3 150            -                                                                     -

2 E

                  !        10 0  -                                                                     -

o

                 'E

, e c 5' 50 - - 2 o ) 0 O 50 10 0 15 0 200 Time (s) Figure 15-2. Scenario 12 Net CVCS injection flow. 163

CAUTION: THE-SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 250 , , , ,

                   -!                                                                                -500 200         -
      .n
                                                                                                     -400 q
         "                                                                                                    N
        ". 15 0
                                                                                                     -300 0        #

10 0 - a; -

                                                                                                     -200 g 2                                                       ;                     %                       $
              -50      -

100 i l  ! O O O' 500 1000 1500 2000 2500 Time (s) Figure 15-3. Scenario 12 Cold leg break mass flow rates. CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 5000 , , i O LOOP A ~~

           '4500 --                  2--                  RCP trip               0 LOOP B B

4000 b b 5 " -

      ^                                                                                           --8000        g a    3500 ~     -

N N' g I 3000 e. v g

                                                                                                       -6000

[ 2500 - g c -

=

g 2000 ,.

                                                                                                       -4000     a O

2 1500 - 3" ,

                                                                                                  ~

1000 -- - 2000 500 - _-c 5 a.

                                        '                   '                  '                         0 0

0 50 10 0 15 0 200 Time (s) Figure 154. Scenario 12 Primary loop mass flow rates. 164

7 CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

            .500                      i                i                 i              i
                   -                                                                    a SG A             - 1000 0                                                                   0 SG B
           '400 I-                                                                       a SG C        -
 ~
                   -                                                                                       - 800
                                                                                                                  .n N

(o 300 -

                                                                                                           - 600    j
 <3   v j 200,
                                                                                                           - 400    -

E a o 10 0 _ -

                                                                                                        -- 200        8 2                           e c c       O-   o       a; ; o            - na c c c o                          2 0-            - c a a a             :Ecc c c c c-c c c c c c                            --0 i                i                 i              i                 - -200
            -100 -

0 500 1000 1500 2000 2500 Time (s) Figure 15-5. Scenario 12 Main flow auxiliary feedwater mass flow rates. l CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 700 , , , i 800 0 SATURATION 2 v o FLUID f v e ( . h 600, --600 5 5 '

          %                                                                                                        t

o. _ o.

E E 2 500 -

                                                                  }-                                    -

v -

                                                                                                           -400   m h                                                                            I  J                        h
                                                                                                                  =

o-

         =

4 V

                                                                                               .1[9 9

g400 f g

          "                                                                                                        2
                                                                                                           -200
         $                                                         I                                              $

300 O 500 1000 1500 2000 2500 Time (s) Figure 15-6. Scenario 12 Break fluid and saturation temperatures. 165

      ~'

j CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 20 .. , , .

                                                                                                   -2500 n                                                                                                       n     .

15 p v

                                                                                                   -2000     n.

E E 3- 3 y

     !      10 -                                                                               -

1500 g i i

                                                                                                    -1000    $

3 2 g 5 - - g

                                                                                                    -500 0                                                                                        O O         500             1000               1500         2000              2500 Time (s)

Figure 15-7. Scenario 12 Primary system pressure. CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 2000 , i i i 1800 -- --4000 1600 - - g 1400_- - 3000 $

 \                                                                                                             E

• a

 .x v

1200 - -

  $       1000     -                                                                              -
 ,                                                                                                    -2000   j
  ,,       800     -                                                                              -

E E 2 600 - - j 4ng - _ l _-1000 1 l y i 200 - q l o

                                 ,              i   i L M    l      i   u lMI    d   W I ' '

o a 0 500 1000 1500 2000 2500 Time (s) Figure 15-8. Scenario 12 Downcomer mass flow rate. 166 m

problems related to interphase mass transfer were at 2 h was extrapolated to be 1.38 h1Pa (200 psia) the principal cause of difficulty. However, these as shown in Figure 15-9. This estimate was made problems were not judged to significantly affect the by assuming the vapor generation from the core is conclusions related to PTS concerns. At 2315 s the balanced by approximately the same volumetric calculation was terminated and at that time primary flow out the break. It was further assumed that system pressure and reactor vessel downcomer most of the HPI exited through the break junction temperatures were 1.76 h1Pa (255 psia) and 442 K with the remainder used to replace liquid boiled off - (338"F), respectively. in the core. 15.3.2 Extrapolations and Uncertainties. This Figure 15-10 presents the extrapolated reactor section presents the extrapolations of the vessel vessel downcomer temperature. Loop mass flows o downcomer pressure, fluid temperature, and wall were nearly stagnant by 2499 s with the cold leg to mside surface heat transfer coefficient. Also, vessel mass flows in Loops B and C equal to the extrapolations of the cold leg mass flow rates and ECC mass flow rates. Because of the break loca-fluid temperatures are shown. Uncertainties in the tion in Loop A, the flow was negative at the cold calculation are also addressed. leg /vesselinterface to Loop A. ECC flow resulted in the upper downcomer region cooling down to Extrapolations to 2 h of the pressure, temper- 310 K (100 F) by %%300 s. The low flow conditions ature, and heat transfer coefficient curves m the also resulted in a rapid stabilization of the reactor reactor vessel downcomer and mass flow rate and vessel downcomer inner surface wall heat transfer temperature curves in the cold legs are shown in coefficient (Figure 15-11). Figures 15-9 through 15-13. Parameters in the downcomer are shown for the elevation adjacent to the top of the core. Figures 15-12 and 15-13 show the cold leg mass flow rates and fluid temperatures for Loops A At the termination of the calculation, Loops B through C. It was estimated that within 2 h natural and C were stagnant and the flow in Loop A was circulation conditions would not be reestablished oscillatory. In was concluded that the oscillatory because of the stable low primary system mass flow in Loop A was caused by condensation of inventory present when the simulation was ter-steam in the Loop A steam generator tubes. Addi- minated. The stagnant loop flow conditions allowed tionally, suspected code problems tended to enhance each cold leg to reach 305 K (90'F) by 3000 s. the size of these flow oscillations. It was judged that L p A was c led with ECC h, quid from Loops B these oscillations would not affect the final time- and C that entered the vessel inlet annulus and averaged extrapolated responses of pertinent passed back from the vessel to the break in Loop A. parameters. This assumed mode of transfer between the loops does not consider the inherent three dimensional Based on the refill times for the secondary sides character of the vessel downcomer. These additional of Steam Generators B and C, it was estimated that c nsiderations would possibly allow less ECC liquid Steam Generator A was isolated at 2499 s due to to recirculate to the break in Loop A. steam line liquid carryover. The isolation of Steam Generator A resulted in core flow stagnation. A Between 1200 and 1300 s an anomaly in break subsequent hand calculation indicated that the mass flow is noticed in Figure 15-3. As discussed, primary system is repressurized to 2.1 h1Pa the break Huid becomes alternately saturated vapor (305 psia)(Figure 15-9) at which time the Loop A and subcooled liquid during this period (see pump seal liquid inventory was depleted enough to Figure 15-6). It does not appear physically correct allow a steam flow path between the vessel and that break ilow momentarily be driven to zero dur-break. By 2892 s the break flow was comprised of ing this period. Because of the momentary nature both vapor from the vessel and liquid from the ECC of the anomaly and the minimal disturbances in system. The two-phase mixture exiting through the reactor vessel downcomer pressure and fluid a break resulted in a slow depressurization of the temperature responses (Figures 15-7 and 15-10), the primary system as a consequence of an increase in effects on overall calculation results were considered the volumetric break flow rate. The final pressure insignificant. 167

CAUT ION: THE SCENARIOS SIMULATED

                                 . CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ~ ACTIONS, EQUIPMENT. FAILURES, OR BOTH.'

20 , , , i i i i CALCULATED EXTRAPOLATED -2500 o -

    - [ 15 _                                                                                            -2000 $

e 3 3 ~ 3 _-1500 10 - --

      .[2        .
                                                                                                        -1000    E 3

g p 5 -

                 .                                                                                      -500 t            '        '          '         '           '         '

O 10 O' 1000 2000 3000 4000 5000 6000 7000 8000 Time (s) Figure 15-9. Scenario 12 Extrapolated downcomer pressure. CAUTION:-THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 600 , , , , , , .

                                                                                                        -600 CALCULATED 7

v

                                                                            -- EXTRAPOLATED                   p-v
                                                                                                        -500~

• 2 2 2500 -

Q g E E-

                                                                                                        -400 E

• L E T

2 hti

                                              \                                                         -300 2 E                                        \                                                              E-
      = 400        -

g -

                                                                                                              =
       *                                             \                                                         e     .

g - N -200 E n 2

    -$.                                                    'N                                                 $
                             ,            ,         ,         ,\--3---             r- - - -- 7
                                                                                                        -100

• O 1000'- ~ 2000 3000 4000 5000 6000 7000 8000 Time (s)

Figure 15-10. Scenario 12 Extrapolated downcomer fluid temperature. 168 L

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 50000 i i i i i i i 2 _ CALCULATED _8000 J -- EXTRAPOLATED , k 40000 - E b t

                                                                                                                      .3 g            -
                                                                                                            -6000        p

( 30000 - 8d 5 mi E -

                                                                                                            -4000     25 20000                                                                                             -

3 28 E -

 .5    10000 -    -
                                                                                                            -2000     j a l . l l yi _rd ,. _ _ ,_ _ _ _ _ .r - - - - r - - - ,-- - - r -

m 0 0 1000 2000 3000 4000 5000 6000 7000 8000 Time (s) Figure 15-11. Scenario 12 Extrapolated downcomer wall heat transfer coefficient. CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 800 g i , , , i i i O LOOP A 700_- o toop g 1500 a LOOP C 600 -- EXTRAPOLATED

    ^ 500           -
                                                                                                          -          7
    $           -             {                                                                              -1000   \

I 400 - h - I H c { 300 - 3 200 ~ --500 $ f

     $                                                                                                                N 2     100      -                                                                                      -

j o =========================== 0-- l ___________________________ --O

        -10 0      -

l -

        -200 O            1000       2000        3000        4000        5000       151 )   7000    8000 Time (s)

Figure 1512. Scenario 12 Extrapolated cold leg flow rates. 169

CAUT ION: THE SCENARIOS SIMULATED

 ~

CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

              .800              ,        ,        ,         ,         ,          ,        ,

o LOOP A 2 o LOOP 8 p v A LOOP C .

            . 700       -
                                                                        --- EXTRAPOLATED           --800        .

3 2 E E 600 f -_600 ' h T 9h  : n 37 500 - - g

                                                                                                     -400       e
          -e                                                                                                    e f400 y       -

N '

                                                                                                     -200 fy 300              '        '        '         '\---'--             ' ---*-

0 1000 2000 3000 4000 5000 6000 7000 8000 Time (s) Figure 1513. Scenario 12 Extrapolated cold leg fluid temperatures. The principal source of uncertainty in the above 15.4 Conclusions extrapolations was the assumption of how ECC was partitioned between the break and vessel. It was As a consequence of the break location and size, concluded that this assumption introduced some the Scenario 12 simulation was characterized by uncertainty relative to the extrapolated pressure but periods of primary system repressurization. It was did not significantly effect the other extrapolated estimated that the primary system would not drop parameters. It was further concluded that the uncer- below 1.38 MPa (200 psia) before 7200 s. Moreover, tainty in the extrapolated pressure was biased on primary system mass inventory depletion eventually the high side since additional ECC flow into the suppressed natural circulation so that loop stagnation vessel would tend to reduce the extrapolated system developed. Loop stagnation enabled the downcomer pressure. The maximum estimated uncertainty in temperature to reach 310 K (100*F) which was slightly the extrapolated pressure was 0.69 MPa (100 psia). above the ECC fluid temperature. O 170

16. SCENARIO 13, STEAM GENERATOR TUBE RUPTURE AT HOT STANDBY WITH OPERATOR INTERVENTION The following subsections describe the investiga- and 10 as shown in Figure 12.1. The tube primary tion of Scenario 13. This calculation was performed side was represented by four cells. Heat transfer was to evaluate the consequences of a postulated rup- represented between each cell and its adjacent steam ture of a single steam generator tube with the reac- generator secondary cell. In addition to wall fric-tor at hot standby conditions with limited operator tion, lumped flow losses were used at junctions 289, intervention. 291, and 292 representing contraction from plena to tube and expansions from tube to the secondary 5 A description of the scenario is provided, fol- region.

Iowed by a discussion of the model changes required to perform the calculation. The results of the calculation, the extrapolated results, the uncer- 16.3 Results tainties associated with the calculation, and the con-clusions regarding the calculation are also The following sections describe the analysis presented. results for a calculation of Scenario 13 and extrap-olation and uncertainty of those results. Scenarios investigated in this report generally in-clude conservative assumptions concerning equip- 16.3.1 Calculation Results. A sequence of events ment failures, operator actions, or combinations of for this calculation is presented in Table 16-2. The these. Conclusions relative to pressurized thermal events of Scenario 9 and Scenario 13 were the same shock severity are not to be drawn directly from the for the first 500 s, therefore, Scenario 13 was results presented in this report (see Section 18). restarted from Scenario 9, at 500 s. 16.1 Scenario Description At zero time a single tube in Steam Generator A was assumed to have ruptured at the tube sheet on A description of the scenario, as developed at eg en gm ws tweak man w ra mug en f th mpted tuk Oak Ridge National Laboratory, appears in The flow through the hot leg end was significantly Table 16-1. The scenario was initiated with the less than that through the cold leg end due to the double-ended rupture of a single tube in Steam large wal1 friction pressure drop ,mposed i on the Generator A with the reactor at hot standby con- , flu d exitmg through the fulllength of tube. Flow ditions. The break was located at the tubesheet on through both break paths was friction dommated, the cold leg end of the tube. that is, the flow rates were determmed by the flow I sses f the paths and not by choking phenomena. The assumed operator actions described in Both paths passed single-phase hquid throughout Table 16-1 represent the general procedures to th cakulation, recover from a steam generator :ube rupture event, except the actions leading to a second opening of g, g g7 g g ; the pressurizer PORV are not a part of the general depressurized, as shown in Figure 16-2, causing the procedure. pressurizer proportional and backup heaters to be powered at I and 8 s, respectively. The pressurizer 16.2 Model Changes level indication, shown in Figure 16-3, decreased 3 due to the fluid volume lost through the break, and With the exception of nodalization changes caused the makeup rate to increase and tripped off necessary to simulate the broken steam generator the pressurizer heater power. The affected steam tube, the model used to perform this calculation is generator mass inventory increased dramatically as . described in Subsections 2.1 and 2.2. The transient shown in Figure 16-4 and the associated increase in was initiated from the hot standby conditions steam generator level caused a slight throttling back presented in Subsection 2.3. of main feedwater to the affected steam generator as shown in Figure 16-5. The transient was assumed The broken steam generator tube was simulated to start with the main feedwater under manual con-with the same nodalization used for Scenarios 9 trol with the reactor at hot standby conditions. 171

i ! Table 16-1. Scenario description No.13 h [ Plant Initial State - Just prior to transient initiator t General

Description:

Het 0% Power,0% Power after 100 h of shutdown System Status Turbine: Not latched, TSVs closed - Secondary PORV: Automatic control Steam Dump Valves: Automatic control Charging System: Automatic control Pressurizer: Automatic control # Engineering Safety Features: Automatic control PORVs: Automatic control Reactor Control: Automatic hiain Feedwater: In bypass mode, manual control to provide zero power Icvel in S/Gs; I condensate pump,1 AIFWP operating. Aux Feedwater: Automatic control MSIVs: Open, Automatic control MFIVs: Closed, Automatic control Transient initiator A steam generator tube rupture on the cold leg side of tube sheet of S/G A. Equipment Failures That Occur During the Transient if the Equipment is Demanded. None Operator Reactions to Reported Information

1. If SIAS signal is generated, the operator will trip the reactor coolant pumps when RCS pressure reaches 1300 psig. (see text)

2. At 500 s the operator closes the affected steam generator MSIV.

3. The operator will control AFW to maintain 40% S/O narrow-range levels.

4. At 10 min the operator fully opens 3 steam dumps and cools down the primary until 45'F subcooling is attained. Subcooling is measured between core outlet temperature and satura-tion temperature in the affected S.G. secondary.

5. When subcooling is attained close the steam dumps.

6. Wait 260 s after Event 5 then open one pressurizer PORV and depressurize the primary system.

7. When the pressurizer and affected steam generator steam dome pressures have equalized close the PORV.

a

8. Wait 500 s after Event 7 then open one pressurizer PORV and depressurize the primary system to 1000 psia. Close the PORV.

9. Wait 100 s after Event 8 then secure llPl. .

10. After securing flPI the operator will use the steam dump valves to maintain secondary pressure below the opening setpoint pressure of the steam line PORVs.

172

I. e p [: Table 16-2. Scenario 13 sequence of events Time J (s) Event 0 Tube rupture occurred 1 Proportional pressurizer heaters powered, make up rate started increasing 8 Backup pressurizer heaters powered 64 Pressurizer heaters depowered on low level

   =

68 Makeup rate reached maximum capacity 276 SIAS signal generated by low pressurizer pressure (11.9 MPa; 1730 psia) Actions due to SIAS are: feedwater bypass valves closed; main feedwater pump tripped; main feedwater pump recirculation flow terminated; motor-driven auxiliary feedwater in. itiated and steam generator levels controlled to 40% NR; HPl and LPI pumps started [ shutoff head HPI: 10.1 MPa (1470 psia); LPI: 0.99 MPa (145 psia)], letdown isolated. 366 HPI flow started (10.1 MPa; 1470 psia) 500 Operator isolated affected steam generator by closing MSIV 546 40% NR level reached in Steam Generator B motor-driven aux feed switched to Steam Generator C 600 Operator opened 3 steam dump valves to obtain 25 K (45'F) subcooling through secon-dary depressurization. 40% NR level reached in Steam Generator C, motor-driven aux-iliary feed terminated 605 RCS pumps trip (9.06 MPa; 1315 psia) 624 Upper head began to void

      '707      Motor driven auxiliary feed to steam generators established 730      25 K (45'F) subcooling obtained, steam dump valves closed 753      2/3 low steam generator level signal, steam-driven turbine auxiliary feed established 990     Operator opened primary PORV to depressurize primary system pressure to affected steam generator secondary pressure 1003     Operator closed pressurizer PORV 1250     40% NR level criteria met in steam generators, all auxiliary feed terminated 1277     Affected steam generator secondary pressure reached PORV setpoint (7.2 MPa; 1050 psia) o   1500     Operator opened pressurizer PORV and depressurizes primary system to 6.89 MPa (1000 psia) 1525     Operator closed pressurizer PORY 1625     Operator terminated ilPI 2400     Calculation terminated 7200     End of extrapolation 173

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN

                ' OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

40 , , , , 0 TUBESHEET SIDE - 80 o TUBE SIDE 30 -

                                                                                               - 60     .

7 4 N

     - 0   ,
                                                                                               - 40
  $                                                                                                  i 10 --                                                                                 -- 20 OiF                                                                               o--0 i

i i e t - -20

     -10~ -

0 500 1000 1500 2000 2500 Time (s) Figure 16-1. Scenario 13 Break mass flow rate. CAUTION: THE SCENARIOS SIMULATE 0 CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 20 , , , , u PRlWARY o SG A -2500 A SG B e o q x SG C - g 15 - k -

                                                                                               -2000 $

{ SDV opened { 10 -- --1500 g h p PORV opened {3 .  :  :- 2', 4 _ #mccq:;.NOu-2-d% 1 to00 g 3 g 5 g

                                                                                               -500      .
                                          . - m .. ..               . , - .       ..       .

0 O O 500 1000 1500 2000 2500 - Time (s) Figure 16-2. Scenario 13 Primary and secondary system pressures. 174

p - CAUTION::THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 1 i i i i

 " '                     O.8- -

N P-I BL 1 0.6 -

                ?N O

0.4 E u-

E 0.2 PORV opened 0

                            .0             500             1000                  1500              2000                 2500 Time (s) r Figure 16-3. Scenario 13 Normalized pressurizer level.

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, E0VIPMENT FAILURES, OR BOTH. 90000- , , , , D SG A -200000 o SG B 80000 -

                                                                                                                     --180000 9 70000 --                                                 .......                        .       .       .--160000    E

_g -c-c's e c : ec:  : o :o

v M

j 60000~- --140000 j

  *                  -                                                      s                                          -120000 50000     -

i

                     -                                                                                                 -100000 4

40000 O 500 1000 1500 2000 2500 Time (s) Figure 164. Scenario 13 Steam generator secondary liquid masses. 175

F CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 3 i i i i _ o SG A -6 o SG B A SG C be,-% @ ^ j { -5 7 N N 2 - - y v

                                                                                                   -4 E
                                                                                                      .o_

v = 3 3 0

            -                                                                                                                                                                                               o U                                                                                     ~

en S I_~ -2 E 2

                  .                                                                                _j 0                  '               '             ='       :      c'=      = =     =0 0            100             200              300            400             500 Time (s)

Figure 16-5. Scenario 13 hiain feedwater mass flow rates. Accordingly, there was an uncertainty as to what valves were closed, the h1FW pump was tripped, control, if any, should be placed on the affected h1FW pump recirculation flow was terminated, steam generator's main fetdwater bypass valve. motor-driven auxiliary feedwater (AFW) was Options included (a) freezing the valve at its initiated and controlled to each steam generator to pretransient areas, which assumes the operator does maintain 40% narrow-range level indication, high not observe the increasing level in the steam and low pressure injection (liPI and LPI) pumps generator; (b) allowing the valve to modulate at a were started, and letdown was isolated. Figure 16-6 rate consistent with that used prior to the transient, shows the increase in net makeup flow when let-which assumes the operator observes an increasing down was isolated and the initiation of IIPI flow steam generator level and controls the feedwater as when the primary system pressure fell below the he has been; (c) closing the valve, which assumes IIPI pump shutoff head. the operator observes the increasing level, recognizes it as a tube rupture and acts accordingly. At 500 s the operator closed the affected loop The calculation was performed using option (b). h1SIV but this action did not have a significant The gradual initial decline in affected steam effect on plant parameters. At 600 s the operator generator main feedwater bypass valve flow rate opened the steam dump valves which depressurized was an indication that the operator was slowly the unaffected steam generators and primary system throttling back feedwater, but did not yet recognize as shown in Figure 16-2 and caused a rapid the transient as a tube rupture which would require cooldown of the primary system as shown in

• isolation of feed and steam functions of the affected Figure 16-7.

steam generator. The primary system depressurization caused a At 276 s the safety injection activation signal reactor coo! ant pump trip at 605 s and the loop , (SIAS) was generated due to low pressurizer flows in all loops coasted down as shown in pressure. As a result of the SIAS the following Figure 16-8. The tripping of the reactor coolant actions occurred: main feedwater (h1FW) bypass pumps occurred after the time operator mitigation 176 L

r P CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 15 , , , , _ O HPl (per loop) -30 o MAKEUP g f"

                                                                                                         -25 g                 Letdown                                                                                7 t                                                                                                  N ho       -

isolated - 3 0 0 0 0 E

  ., d.                                                                                                  -20 g
  ~

v- " v

       $                                                                                                       t
      .                                                                                                  -15  j 5       -

c - E 2 -

                                                                                                         -10  j
                                                                                                         -5 ccc       ccc          c  c    o oc          a                                '               '

acc 'c c C 0 0 500 1000 1500 2000 2500 Time (s) Figure 16-6. Scenario 13 HPI and makeup mass flow rates. CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 580 , , , , 7 v 560 - 550 E v

                                ^>                          SOV opened E

3 2 o E540 - - E

                                                                                                         -500 i E

2 520 - - 2 3

g. -

                                                                                                         -450 g.
     = 500         -                                                                                  -
                                                                                                                =

e e O E E

     $ 480_-                                                                                          -_400 i
 ~

Q I terminated 460 0 500 1000 1500 2000 2500 Time (s) Figure 16-7. Scenario 13 Reactor vessel downcomer fluid temperature. 177

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 400 , r , , , o LOOP A - 800 o LOOP B .

                                     /                                           A LOOP C
                                                                                                     - 600 n                                                                                                       7 g200_-                                                                                       -
                                                                                                     - 400     E        -

s g 5 -

                                                                                                     - 200     g
                                                                                                              =

E O-- c - o--0

                                                                                                     - -200
         -200'
                                                                                                     ~

0 500 1000 1500 2000 2500 Time (s) Figure 16-8. Scenario 13 Cold leg mass flow rates. began at 600 s. The operator may elect not to trip Per the scenario description, at 990 s (260 s after the pumps because mitigation is la progress. This the closing of the steam dumps), the operator represents an uncertainty in the description of the opened the pressurizer PORV to depressurize the scenario. Due to heat removal through the unaf- primary system to the affected steam generator fected steam generators flow continued in Loops B secondary pressure. At 1003 this equalization was and C. Since the Loop A MSIV was closed, no heat accomplished and the PORV was closed. This cycle was removed through Steam Generator A and was repeated with the same results, between 1500 Loop A flow stagnated. and 1525 s. At 1625 s,100 s after the second PORY closure, the operator terminated the HPI as directed At 624 s the primary system pressure had fallen in the scenario description. The minimum reactor to the saturation temperature of the fluid in the vessel downcomer temperature of 470 K (387'F) reactor vesset upper head. As a result, a steam bub. was reached at the time of HPI termination as ble formed in the head, au shown in Figure 16-9. shown in Figure 16-7. The maximum subsequent This flashing ha!ted the primary system depres- downcomer pressure was defined by the steam surization, as 'hown in Figure 16-2. dump valve opening setpoint,7.03 MPa (1020 psia) as shown in Figure 16-2. At 730 s the subcooling requirement was met and After 1625 s the plant rapidly reached a long-term the operator closed the steam dump valves, slow- steady operating condition with the primary and ing the cooldown (Figure 16-7) and causing the affected steam generator system pressures equaliz-primary system to repressurize (Figure 16-2). As a ed, thus limiting the break flow. It was assumed that result of the repressurization the bubble in the upper the operator continued to keep the pressures _ head was compressed. The cooldown, however, was equalized by periodic opening of the steam dump not stopped at this time because the unaffected valves. steam generators, although severely depleted of inventory by the steam dump flow, continued to 16.3.2 Extrapolations and Uncertainties. The receive auxiliary feedwater, calculation was tetminated at 2400 s with the plant 178 L

e - - j f; *

CAUT ION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR ' BOTH.

0.4 -- i i i i I

  .o e     0.3   -                                                                                    -

z o

                            ' 3 - 0.2 -    -                                                                                   -

E o

                              >      0.1   -                                                                                    -
                                  . 0.0 0             500               1000              1500           2000             2500 Time (s)

Figure 16-9. Scenario 13 Reactor vessel upper head void fraction.

                 < in a stable long term cooling mode that was                       ing of the pressurizer PORY and assumed late ter-expected to continue indefinitely. Thus the                       mination of high pressure injection.

extrapolated parameters shown in Figures 16-10 through 16-14 assume the trends in conditions present at 2400 s continued through 7200 s. There 16.4 conclusions L are no major uncertainties in the calculation and

                 , extrapolations for Scenario 13. As previously                       The minimum reactor vessel downcomer fluid mentioned, the tripping of the RCPs, after operator               temperature,470 K (387'F) occurred at the time of intervention has oegun, represents an uncertainty                 HPI termination. The maximum subsequent down.

In the specification of the scenario. Furthermore, comer pressure was 7.03 MPa (1020 psia). The l the scenario specification is conservative (relative thermal hydraulic results for this scenario are strong to existing procedures) because of the second open. functions of the assumed operator actions. F

  -: o m

i 179  ;

7. -.- ,

i CAUTION: .THE . SCENARIOS SIMULATED-CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 16 . i i i i i i CALCULATED

                                                                              -- EXTRAPOLATED 14

• g -2000 a 3  %

            -                                                                                                        s E 12     -                                                                                 -

E R R

           'l                                                                                               -1500     i
              , 10 E
                     '--(                                                                                -

E s a

             $g        _
                                                                        ~                                             $
                                                                                                            -1000 6

O 1000 2000 3000 4000 5000 6000 7000 8000 Time (s) Figure 16-10. Scenario 13 Extrapolated primary system pressure. CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 580 . . . . . i i CALCULATED Q' -- EXTRAPOLATED p v 560 % --550 E E 2 - 1 o E540 - [ -

                                                                                                             -500 i E                                                                                                       E I 520        -

2 2 3

g. ' ,

                                                                  , ',                                       -450 g.

i.= 500 - -

                                                                                                                      =

i g 480 -

                                                                                                             -400 i

o

                                                                                                    '                   ~

460 0 1000 2000- 3000 4000 5000 6000 7000 8000 Time (s) Figure 1611. Scenario 13 Extrapolated reactor vessel downcomer pressure. t 180

m-- , , CAUT ION: THE SCENARIOS SIMULATED

                                              .CONTAIN SIGNIFICANT CONSERVATISMS IN 3,.

OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 10000 , , , , , , , 2 CALCULATED d ' -

                                                                                      ~~

EXTRAPOLATED

                                                                                                                  -1500 E
     .o.
              -}

y z cm -

               ?

y. -UOO W

u Oy

               }o       5000     -                                                                             -

Q gg t 85 hv

                                                                                                                  -500 5                                                                                                         o
               =                                                                                                         i
               ;                                          m_________________________.

I , 0 O O 1000 2000 3000 4000 5000 6000 7000 8000 Time (s) Figure 16-12. Scenario 13 Reactor vessel wall heat transfer coefficient. CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 400 i;i , , , , , , o LOOP A - 800 0 LOOP B A LOOP C

                                                                                     -- EXTRAPOLATED             - 600 n                                                                                                      ?
            ,            200,-       I n
                                                                                                                 - 400    E 4

6 :f! I v c f . ,

                                                             ==a==a==========::::::::                            . gog    y t                                                                                   f j_                                                                                            _o o._               94
                  ^                                                                                                      &

• I ,

                                                                                                                 - -200
      ~                      ~
                                                                                                                 ~~
                       -200 0     1000        2000         3000    4000     5000        6000    7000     8000 Time (s)

Figure 16-13 Scenario 13 Extrapolated cold leg mass flow rates. 181

  ~ ,

i. n ,.. 4 CAUT ION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. . 700 , , , , , , , 800 0 LOOP A p 7 0 LOOPS B AND C

      *                                                                  -- EXTRAPOLATION                 v 8                                                                                                 e
       .h. 600 ,
                                                                                                  --600    3
      -E          t       -

F . . p

       % 500,
                                                            - y ~~~~~~....                        -

2

                                                                                                     -400 y          ~
                            %                        /                                                    m re*1
                                                                                                          =

T 400 - { -200 - 300 , O 1000 2000 3000 4000 5000 6000 7000 8000 Time (s) Figure 16-14. Seenario 13 Extrapolated cold leg temperatures. t

                                                                                                 =

a 9 am 182

17. SCENARIO 14,2 IN. HOT LEG BREAK AT FULL POWER The following subsections present the transient 17.3.1 Calculation Results. A sequence of events scenario description, modeling changes effected to for the transient is provided in Table 17-2. The perform this calculation, detailed analysis of the break was assumed to occur at time zero in the transient results, extrapolations and uncertainty bottom of the Loop C hot leg. The primary system analyses, and conclusions drawn from the analysis rapidly depressurized as shown in Figure 17-l. Both for Scenario 14: a small hot leg break at full power. the proportional and backup heaters were turned on to recover the pressure. Also, as a result of the Scenarios investigated in this report generally break, the pressurizer level rapidly decreased, as include conservative assumptions concerning equip- shown in Figure 17-2, and the makeup flow ment failures, operator actions, or combinations of increased to compensate for the lost liquid these. Conclusions relative to pressurized thermal Inventory.

shock severity are not to be drawn directly from the results presented in this report (see Section 18). At s24 s, the reactor tripped on a 2/3 reactor over-temperature AT signal. The turbine stop valves 17.1 Scenario Description closed and secondary pressures began to increase as shown in Figure 17 3. The primary system depressurization rate increased due to the rapid The transient was initiated from full power steady reduction in core pow er. The primary-to-secondary state (nommal temperature and pressure), and all heat removal rate is shown in Figure 17-4, The control systems were in automatic control. The steam dump control system changed from load transient was initiated by a 0.0$08 m (2-in.) rejection mode to plant trip control mode at the diameter hole appearing in the bottom of the time of reactor trip and the steam dump valves horizontal section of the C loop hot Icg, just opened to bring the plant average temperature down upstream of the pressurizer surge im, e connection. to $$9 K ($47'F). It was assumed that all systems operate automat-ically as designed. The only operator actions assum- Also as a result of reactor trip, the break mass ed to take place were: (a) trip the reactor coolant flow rate increased as shown in Figure 17 5. This pumps when the primary pressure reached 9.1 MPa increase was due primarily to an increase in the (1315 psla)if a SIAS signal had been generated, and pressurizer surge line outflow rate. At the time of (b) isolate auxiliary feedwater flow 10 min into the reactor trip the primary system depressurized faster transient. A transient scenario is provided in than the pressurizer due to the energized pressurizer Table 171. heaters trying to maintain a pressure of 15.5 MPa (2250 psia). The increased driving potential on the 17.2 Model Changes fluid in the pressurizer surge line resulted in the increased mass flow. At s34 s the pressurizer level Changes made to the steady state model to initiate dropped below 14.4%, the heaters were turned off, the small hot leg break included the addition of a the driving potential decreased, and the break mass break valve in the C toop hot leg connected to a fl w rate decreased. time dependant volume set at atmospheric condi-

                                                                          ,g                   P gg tions. The break components were set to represent a break at the bottom of the hot leg pipe. The valve       t 11.9 MPa (1730 psia), actuating the SIAS s.ignal.

was set to open at the initiation of the transient. As a consequence the main feedwater valves closed, isolating main feedwater from the steam generator o secondaries. The main feedwater pumps tripped on 17.3 Results low now and the heater drain flow was terminated. At the termination of main feedwater pump power. This section presents the results and extrapola- the motor driven auxiliary feedwater system was ac-tions and uncertainties of the 2 in. hot leg break tivated and ausiliary feedwater began flowing into the transient at full power. A and 11 steam generators. 183

3 [~ p Table 171. Soonerlo description No.14 Plant initial State - Just prior to transient initiator

              , General

Description:

100% Power steady state System Status . Turbine: . Automatic control - Secondary PORY: Automatic control .

                       . Steam Dump Valves: Operative / Automatic control Charging System: Automatic control Pressurizer: Automatic control Engineering Safety Features: Automatic control PORVs: ' Automatic control Reactor Control: Automatic Main Feedwater: Automatic control Aux Feedwater: Automatic control.

MSIVs: Open, Automatic control MFIVs: Open, Automatic control ' Transient Initiator A 2-in, hole appears in the hot leg. Equipment Failures That Occur During the Transient if the Equipment is Demanded. None Operator Reactions'to Reported Information _l.; If SIAS signal is generated, the operator will trip the reactor coolant pumps when RCS ' pressure reaches 1300 psig.

                   '2.    ~ The operator will terminate AFW flow at 10 min.

At.45 s, two out of three steam generators had steam dump valves. Primary system depressuriza-

              - reached the' low-low level signal setpoint and the          tion was slowed as a result of the equalization steam-driven auxiliary feedwater flow was initiated .       between decay heat generation and removal, and to the generators.' With both the motor and steam-          vapor generation in the reactor vessel upper head

driven auxiliary feedwater systems operating, the as a result of the pressure there reaching the -

level in Steam Generators A and B began to - saturation pressure of the fluid.-

               -increase as shown in Figure 17-6. Primary system .

pressure had dropped below the HPI shutoff head At N248 s the reactor vessel upper head had com-by N60 s and flow from this system was established  : pletely voided and voiding in the upper plenum,

               . to each cold leg.                                        : downcomer above the inlet annulus, and hot legs cornmenced. Two-phase fluid reached the break and At N78 s the primary system pressure - had            . the break mass flow drastically dropped as shown
              .. dropped below 9.1 MPa (1315 psia) and the reac-            in Figure.17 5.
               . tor coolant pumps were tripped. As the pum;n coasted down, the loop flow transitioned from full            At s300 s, the high average temperature dropped   _

flow to natural circulation as shown in Figure 17-7. below $59 K (547'F) and the steam dump system

• Primary system heat removal was nearly equal to control changed from plant trip control mode to

                ~ decay power and the secondary system pressures -           steam pressure control mode. The steam dump began to decrease due to the influence of the cold -        valves closed and secondary system depressurization auxiliary feedwater and heat removal through the            continued solely due to cooldown effects from the gg4

f b f. t p

                  . TatWe 17-2. Scenario 14 sequence of events E

Time [ (s) Event 0.0 0.0508-m. (2-in.) diameter break appeared in bottom of C loop hot leg. C 24 . Reactor tripped, turbine stop valves closed. ( 34- ' Pressurizer heaters deenergized due to low pressurizer pressure. 39 SI signal received on low pressurizer level,' main feedwater pumps tripped, main feed. water valves closed, motor-driven auxiliary feedwater initiated. 45 . Steam-driven auxiliary feedwater initiated.

                           . 60                                                        HP! flow initiated.

78 ~ RC pumps trip on low primary system pressure, reactor vessel upper head started to . void. l -600 All auxiliary feedwater flow terminated (per scenario description). l:

                        -2000-                                                         Loop flows stagnated.

4056- Accumulator flow initiated. 5983 Calculation terminated. 7200 End of extrapolation. Laddition of cold auxiliary feed flow. Between By s2000 s enough voiding in all loops had 300 and 1000 s vapor generation and condensation occurred that natural circulation had stopped and effects resulted in perturbations in the primary the loops were very nearly stagnant throughout the system pressure response, shown in Figure 17-1, as remainder of the calculation as shown in

                   -- the loops and upper plenum drained. Two-phase                                                             Figure 17-7.

i flow periodically reaching the break ~ also con-tributed to the pressure oscillations. 'Also con- ~ The cold leg and reactor vessel downcomer fluid tributing to the primary system pressure response temperature responses are shown in Figure 17-8.

                  ' during this time, was the primary-to-secondary heat                                                         The cold leg temperatures dropped significantly removal rate (Figure 17-4) which was affected by :                                                        during periods when the loop flow stagnated as the amount of, auxiliary feedwater. the steam                                                             shown in Figure 17-7. The downcomer temperature generators were receiving. The common header for                                                          declined at a slower rate than the cold legs due to o     .   . each auxiliary feedwater system (steam and motor-                                                           heat addition from the vessel walls and warm fluid
                ~    ' driven) was modeled and the flow to each generator -                                                     entering the downcomer from the reactor vessel was determined by the differential pressure between                                                       internal leak and bypass paths.
                   ' the header and the generator. At times one gener-ator received all of the flow in a manner similar to                                                          At N4050 s the primary system pressure dropped
                  - that discussed for Scenario 6 in Subsection 9.3.1.                                                          below 4.6 MPa (673 psia) and accumulator injec-As specified: in .the scenario description, the                                                          - tion began. The calculation was terminated at operator terminated auxiliary feedwater to each                                                           5983 s with the primary system slowly refilling. The steam generator at 10 min.                                                                                primary system pressure at the termination of the I85 h

           .i CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

20 i , , , ,

                                                                                                                                                                                                                                          -2500 m                                                                                                                                                                                                                             ^   .
                 'o                   15         ..                                                                                                                                                                                     .        ,0
                                                                                                                                                                                                                                          -2000 $

E E s s . l

 ~

go - _-1500 g i L-E -

                                                                                                                                                                                                                                          -1000   [
                                                                                                                                                                                                                                          -500 O                                                                                                                                                                                              O O           1000       2000           3000 .                                                   4000                                                                         5000      6000 Time (s)

Figure 171. Scenario 14 Primary system pressure. CAUTION: THE SCENARIOS SIMULATED

                                                                   ~ CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

0.6 . , i 0.4 - - 7 a 0.2 - - 0.0 0 50 10 0 15 0 200 - Time (s) Figure 17 2. Scenario 14 Normalized pressurizer pressure. I86

CAUTION: THE SCENARIOS SIMULATED ' CONTAIN SIGNIFICANT CONSERVATISMS IN

j. OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. .

7 , , , , , a 3g 3 -1000 o SG B A SG C [ 6.5 ' -l cc o -c c c c c c c c c c : [950 y d

                                                                                                                            -900

• a a 3 6 --

h  :  :  :  :  : g i: _ gc v c c c ^ c c c c c _850 1

• e E E

            .'                                                                                                                       a
            >o - 5.5 l
                                                                                                                            -800     o
                                                                                                                            -750 5     -

O. 1000 2000 3000 4000- 5000 600v Time (s) Figure 17-3. Scenario 14 Secondary system pressures. CAUTION: THE-SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN I ' OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 100 a 0 , , , , , O SG FOWER O CORE POWER a 50 - - y 3 a

                                                                                                                      .       _m-e R                           i o-0       -
                                                                                     -c       c       c      c     c    c      c ti o

i '

                       -50 O            1000           2000          3000              4000            5000            6000 Time (s)

Figure 17-4 Scenario 14 Core power and steam generator heat removal rates. 187 t-

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 200 , , , , ,

                              .                                                                         O BREAK                -400 0 ECC I                                                                                                             .
                   ^                                                                                                                  7 q          -
                                                                                                                               -300 s     ,

a i E 2, I M e v f 8 10 0

                              ~

sm' b,, / , 200 j

                    =                                                                   1 1 o

1 l Q* 2

                                                     @@#Jg p#l
                              ~

lM h

                                                                                               %l ..

Li

                                                                                                                               "'00 O                                                                                                   O O             1000          2000         3000               4000        5000           6000 Time (s)

Figure 17-5. Scenario 14 Break and total ECC mass flow rates. CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 0.8 , , , , , o SG A o SG B

                                                                                                             ,a      SG C 0.6    -                                                                                          -

g-c c c = c = c c i l _4 a A a e o a { ' g,4 .. C _ n 0.2 - oc c c c c c c c c r c c e 40 _ 0 1000 2000 3000 4000 5000 6000-Time (s) Figure 17-6. Scenario 14 Steam generator levels. 188

l. CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR.. ACTIONS, EQUIPMENT FAILURES, OR BOTH. 2000 , , , , , 0 LOOP A - 4000 o LOOP B oc a LOOP C --

                                                                                                                     - 3000 m                                                                                                                    7

{ 1000 ,- -

                                                                                                                     - 2000 E N

m. 6 2 v c 1000

• o 0--  ;

Bft;;Mi7 E ; 7 ; ; T--  ; -f 2- O h. 2

                                                                                                                     - -1000
                      ~                                                                                              ~~
             -1000 0         1000               2000                3000         4000         5000       0000 Time (s)

Figure 17-7. Scenario 14 Cold leg mass flow rates. CAUTION: THE SCENARIOS SIMJLATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

              -600                  ,                   ,                  ,            ,             ,
                                                                                                                    -600 0 LOOP A          I 7

i o LOOP B p v A LOOP C

• x DOWNCOMER

                                                                                                                    -500 .

D 3 > l 3 0500 -

                                                   ;,                                                           -          o E-      -

_400 8. E E J l

)

S [ 2 -

                                                                                                                    -300 3
                                                        'T 400      -                                                                                       -

f

            .                           t                                                                                  .

o E -

                                                                                                                    -200 E 2                                                                                                               2 S                                         .

300 ' ' ' ' ' O 1000 2000 3000 4000 5000 6000 Time (s) Figure 17-8. Scenario 14 Cold leg and reactor vessel downcomer fluid temperatures. 189

calculation was 2.8 MPa (406 psia) and decreasing, The reactor vessel downcomer wall heat transfer and the reactor vessel downcomer fluid temperature coefficient is extrapolated in Figure 17-11. The coef-was 333 K (141*F) and decreasing. ficient was declining smoothly at the end of the calculation due to the decreasing fluid temperature 17.3.2 Extrapolations and Uncertainties. This in the downcomer and changing fluid properties that section presents the extrapolations of the vessel affect the Dittus-Boelter heat transfer correlation, downcomer pressure, flu?d temperature and wall inside surface heat transfer coefficient. Also, Figures 17-12 through !?-14 show the loop mass , extrapolations of the cold leg flow rates and fluid flow rates present at the end of the calculation are temperatures are shown. Any known uncertainties expected to continue through 7200 s. All cold legs in the calculation are also addressed. are expected to reach the HPI temperature,305 K (90*F) before 7200 s as shown in Figures 17-15 Extrapolations to 2 h of the pressure, temper- through 17-17. ature, and heat transfer coefficient in she reactor sessel downcomer and the mass flow rates, and There were no significant uncertainties that temperatures in the cold legs are shown in affected the overall results of the calculation and Figures 17-9 through 17-17. At the termination of extrapolations. the calculation, heat removal to the steam generators was nearly zero and the primary system

17.4 CONCLUSION

S was gaining mass due to total ECC flow exceeding break flow. It is estimated the primary system depressurization trend at the end of the calculation The 2-in. diameter hot leg break was showTt capable I would continue as shown in Figure 17-9. The f removing decay power. Therefore, no mechanism resulting extrapolated pressure at 7200 s is 2.0 MPa existed to repressurize the primary system; at 7200 s (290 psia). the pressure was estimated to be 2.0 MPa (290 psia). Because of loss of primary system mass, natural loop The reactor vessel downcomer temperature circulation was lost and the minimum reactor vessel extrapolation is shown in Figure 17-10. Again, the downcomer temperature was estimated to be 310 K trend is expected to continue and the estimated (100 F), slightly above the ECC fluid temperature, at temperature at 7200 s is 310 K (100*F). 7200 s. I CAUT ION: THE SCENARIOS SIMULATED CONTAlN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 20 , , , , , , i CALCULATED

                                                                 -- EXTRAPOLATED                _2500 m                                                                                                    ^

f 15 - - 3 -

                                                                                                -2000 $
     =                                                                                                     .

s. 3 h 10 ~ - - 1500 g 5-6. 0- c. E 3

                                                                                                -1000      [a

• o 5 - -

o

                                                                                                -500 N   s s                                                  -

0 ' ' ' ' ' ' ' O O 1000 2000 3000 4000 5000 6000 7000 8000 Tirne (s) Figure 17 9. Scenario 14 Extrapolated primary system pressure. 190 _ __.______-__.___s-

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS -IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.- 600 ,

                                                                                                                          -600-CALCULATED 7                                                                               -- EXTRAPOLATED                      p

_o " 550 -

                                                                                                                       -        v E                              )
                                                                                                                          -500

• 3 rd 3

    =
          'O 500 E.

h -

                                                                                                                                }

400 8. E E E 450 -'

• x -

                                                        )'                                                             -
                                                                                                                                .a

_3 -300 g

         .$400            -
                                                                                                                      ~

f E - e

                                                                                                                         -200 E f350 300                   '              '             '         '         '             '         '

0- 1000 2000 3000 4000 5000 6000 7000 8000 Time (s) Figure 1710. Scenario 14 Extrapolated reactor vessel downcomer.

                                              - CAUT I ON: THE SCENARIOS _ SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

6000 , , i i i , , 2l - CALCULATED -1000 u -- EXTRAPOLATED E .

      }             -
                                                                                                                         -800    $o
      -7 '4000           -

59 2 $~' u -

                                                                                                                         -600     o ::-

t- u I E-o *I e% u

                                                                                                                         -400     E2 2000-                                                                                                               A8 8

o u 3z: l -200 1

   .w            0                                       _

O O 1000 2000 3000 4000 5000 6000 7000 8000 Time (s) Figure 17-11. Scenario 14 Extrapolated reactor vessel wall heat. 191 L

i l CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 50 , , , , ,, , ip

          ~
                                                                                                                ~
                                                                        ~ FXTRAPOLA ED 40       -

n -

                                                                                                                -80   g m                                                                               i                                     E g 30         -
                                                                                                                      .o l
                                                                                                                -60   v
 ,  20       -                                                                     ~~--

n -

                                            !l                                                                 -40     m Y!

• 1 10 -

20 0 0 0 1000 2000 3000 4000 5000 6000 7000 8000 Time (s) Figure 17-12. Scenario 14 Extrapolated loop A cold leg mass flow rate. CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 50 i , , i , ,

                                                                                                             ~
                                                                      -- XTRAPOL ED 40     -

m

                                                                                                             -80     g E 30 1
                                                                                   - - - ' ~ ~
                                                                                                                     )

o

                                                                                                             -60     0
          ~                                                                                             '

N - i i -40 a 4 i 10 -

                                                                                                        -_29 0                          '

O - 0 1000 2000 3000 4000 5000 6000 7000 8000 Time (s) Figure 17-13. Scenario 14 Extrapolated loop B cold leg mass flow rate. 192

                                    - CAUT I ON: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH.

50 i , , , , ,i, i

                                                         !                       CALCULATED
                                                                                                            -100
                                                                           -- EXTRAPOLATED 40      -                                                                                       -
                                               ~

n

                                                                                                             -80 q
     .E 30
                                                                                      ~ ~'~~~

ai I l -60 C 20 - -

                                                                                                            -40

• s 10 _- -

20 i I i O I ' ' ' 0 0 1000 2000 3000 4000 5000 6000 7000 8000 Time (s) Figure 17-14. Scenario 14 Extrapolated loop C cold leg mass flow rate. CAUIlON: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 600 , i i i i i i i CALCULATED Q -- EXTRAPOLATED f h-

    ?
                                                                                                           -500
                                                                                                                 ?

0500 - - E E -

                                                                                                           -400 E 5                                                                                                            5 3       _             ,
                                                                                                           -300 3 a                          0                                                                                 a
    = 400     -                                                                                        -
                                                                                                                 =
                                                                                                           -200 f

s s

                        ,              ,            ,        ,        ,                     - ,.           -100
 ,     300 0       1000          2000          3000     4000     5000        6000         7000      8000 Time (s)

Figure 17-15. Scenario 14 Extrapolated loop A cold leg temperature. 193

p J CAUTION: .THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 600 , , , , , , ,

                                                                                                       -600 CALCULATED 7
       ~
                                                                          -- EXTRAPOLATED                     p v   .

E

                                                                                                        -500

• 2 lggf 2 E500 -

                                                                                                    -          o
                                                                                               ~

E -

                                                                                                        -400 E    e E                                                                                                     E
         ?                                                                                                     *
                                                                                                        -300  y
        $400       -                                                                                -

f

• e Q

                                                                                                       -200 E 2                                                                                                     3
                              ,           ,            ,       ,                  ,---- ,--            -100 0      1000          2000         3000     4000    5000       6000     7000      8000 Time (s)
                          ' Figure 17-16. Scenario 14 Extrapolated loop B cold leg temperature.

CAUTION: THE SCENARIOS SIMULATED CONTAIN SIGNIFICANT CONSERVATISMS IN OPERATOR ACTIONS, EQUIPMENT FAILURES, OR BOTH. 600 ,

                                ,           ,            i       i       i           ,       i CALCULATED Q                                                                  -- EXTRAPOLATED                     f
         .E'      -

l ,

                                                                                                         -500

• a

         ~ E 500 VD                                                               -

2 T 2 E -

                                                                                                         -400 E E-                                  '

E i 3 .

                                                           \                                             -300 3 400     -                                                                                -
          .                                                                                                     e g       -
                                                                                                         - 2*   g  ,

o o

                                '           '.           '       '       '           '     _"            -100 300                                                                                                  '

O 1000 2000 3000 4000 5000 6000 7000 8000 Time (s)

                        ' Figure 17-17. Scenario 14 Extrapolated loop C cold leg temperature.

194

y i 18.-. OVERVIEW AND CONCLUSIONS Described herein are the analyses of fifteen - Table 18-1 shows a summary tabulation of RELAPS computer code calculations pertinent to - minimum fluid temperatures and maximum subse-

             . the study of pressurized thermal shock (PTS)in the        quent fluid pressures in the reactor vessel H. B. Robinson,' Unit 2 (HBR-2) pressurized water        'downcomer for each of the eleven scenarios. The Q         reactor (PWR).                                            pressures and temperatures shown are located at an elevation adjacent to the top of the core, One of the calculations ~ simulated a plant trip transient that occurred in the PWR. A comparison -

of code-calculated and measured data indicated NOTE: The pressures . and temperatures 4 generally good agreement thus providing an infor- shown are generally not coincident. The mal and limited, but useful, qualification of the temperatures shown represent 'the lowest computer model beyond the detailed quality calculated or, in the event of calculations assurance measures described in Section 2. terminated before the end of the PTS 2-h The remaining fourteen calculations simulated period of interest, the lowest extrapolated temperatum wii,n the M penM. W hypothetical cooldown scenarios with varying pressures and temperatures shown m, the table

             . potentials for primary system repressurization.

These scenarios were defined at Oak Ridge National have been adjusted, if required, for any uncertainty or bias identified in the calcula-Laboratory (ORNL), integrator of the multi- , ti ns as dismssed in the " Extrapolations and laboratory PTS study. Detailed descriptions of the Uncertainties , subsections for each seenario. scenarios are presented in the " Scenario Descrip- , tion" subsections of Sections 4 through 17. scussi n f uncen nties in the, calculations presented here appears in Sec-The computer calculations were performed using tion 15 of NUREG/CR-3935. Table 18-1 is best estimate modeling assumptions for plant con- presented as a convenience to the reader and ditions and responses to the events specified in the is not intended to be used as an indicator of scenario descriptions. The reader is cautioned, pressurized thermal shock severity for each however, that for bounding purposes the scenario sequence. Determination of severity first descriptions were based on extremely conservative requires an evaluation of the plotted informa-assumptions concerning equipment malfunctions, tion presented in the scenario results sections. operator actions and omissions, or combinations of Second, additional analyses of multidimen-these. Thus, while the computer calculations repre- - sional effects and fracture mechanics, will be

              - sent best esiimate plant responses to the scenarios         performed by other PTS study participants.
              . as defined, they do not represent the mostprobable          Following these additional studies, judgments plant responses to the scenario initiating events.           of severity will be made at ORNL.

_O

        .4 195 L:

J3 s > l Table';18-1. Summary tabulation of HBR-2 PTS analytical results i Minimum RV Maximum Downcomer - Subsequent Fluid RV Downcomer Temperature Pressure

                                                ' Plant

• Scenario ' ' Description Condition (K) ('F) (MPa) (psia)

            .I'-         I ft2 steam       Hot standby     3%           253    16.2              2350 line break r-2         Double-ended . Hot standby     359           187-  11.0              15 %

steam line break 3 Stuck-open steam . Hot standby 397 256 10.1 1470 line PORY

            .4~         3 steam dumps      Full power      373          212    11.8              1711 fail open 5         SG overfill with   Full power      535          503    16.0             2320 AFW i

6 2-1/2 in. hot les Full power . 310 100 0.98 142 break 7 Stuck-open Full power 538 509 17.5 2538 pressurizer PORY 8 21/2 in. hot les Hot standby 310 100 0.97 140 break' 9.- SO tube rupture Hot standby 465 378 9.62 13 %

         ' 10           SO tube rupture    Full power      $57          543    9.65              1400 l-         -11           Loss of heat sink. Full power      422          300    2.29               333 primary feed and
                      ' bleed recovery 12 -     - 2 in. cold les      Full pwer       310           100    1.38              200 break 13          50 tube rupture    Hot standby     470          387    7.03              1020 with operator
                     - action

[ 14 2 in. hot les - Full power 310 100 2.00 290 4 b 1%

   'O APPENDIX A COMPUTER RUN TIME STATISTICS d

A-1 t__

APPENDIX A COMPUTER RUN TIME STATISTICS Table A-1 presents a timing survey of the calcula-. Laboratory. The calculations were performed us-tions ~ presented in this report. Figures A-1 ing the RELAP5/ MODI.6 (Cycle 16) computer

  ~~
    ,     through'A 15 show the continuous rate of CPU              code (except for Scenarios I and 14 which were time usage during each of the calculations.               performed using RELAPS/ MOD 2 Cycle 16).

The computer used to perform the calculations was the CDC 176 at the Idaho National Engineering l l l [ 6 w A-3

g. . . _ - - _ , . . . , . - . - - . . - - - ,

Tatdo A-1. Timing statistics Scenario Plant Trip 1 2 3 4 5 6 7 8 9 10 11 12 13 14_ Total number 213 193 198 224- 223 223 224 223 224 222 -222 223 224 222 224 volumes (#C) Total number 218 200 203 218 218 218 218 218 218 .226 226 218 218 '226 218 heat structures Transient time 900 7200 1300 2455 1700 3600 2800 2200 1737 7200 2400 8100 2315 2400 5983 s (RT)

         >                        Total CPU         4944   16501 6200  14759 11761 '21116 9697   5599      11915 20556   13704 31389 24094      17635 207 %

A s used Total number of 18004 54436 27150 47059 35992 73465 30959 18844 34384, 68972 48052 99127 51114 32150 48068 time steps (#DT) CPU /real time 5.49 2.29 4.77 6.01 6.92 5.87 3.46 2.55 6.86 2.86 5.71 3.88 10.41 7.35 3.48 - x 10 2.58 1.19 2.41 2.68. 3.10 2.63 1.54 1.14 3.06 1.29 2.57 1.74 4.65 3.31 1.55 RT x #C CPU x 10 1.43 0.22 0.89 0.57 0.86 0.36 0.50 0.60 0.89 0,19 0.53 0.18 0.91 1.03 0.32 RT x #C x #DT x 0 1.29 1.57 I.15 1.40 1.47 1.29 I.40 1.33 1.55 1.34 1.28 1.42 2.10 2.47 1.93

                                  #C x #DT
                                                                                                                                     .?

Y $' ) 49

10 e i i i 6

                   'N
                   -N O

g 8 S

c. w-

_ 2

                     ;- 6 Q-                              {                [ {,

g , , 1, _; 2 U

     .- M o 4      ..                                                                      -

G r z W2 - ~

    .>               E S

O. 0 0 '200 400 600 800 1000 TIME (s) Figure A-1. Plant trip rate of CPU time usage. 10 i i i 6 E N O c w

                    ?              I
                    ~            l 2     5     -                                                                       ~

U

                   'S
                    ~

C l 5 44 d Wm . . JJ 5 e  ! b t i t

      ./                  0 0                 2000                 4000               6000         8000 TIME (s)

_q Figure A-2. Scenario 1 Rate of CPU time usage. 6 A-5

10 i i. - I  ! l 8  ! $ l , 8 i g - t }, I a p- i i 3 ~S&

• a 5 - -

O r 8

                !                                             i         .

a d ( .6 l i 5 5 e 1 5 0 O 500 1000 1500 TIME (s) Figure A-3. Scenario 2 Rate of CPU time usage. 10 , l , o U$ o m I c 1 y  ! - I h.Ll>_[ lop MN .I l.I Jail <...,; {5 - - 8 2 C 5 5 ' b E O O 500 1000 1500 2000 2500

• TIME (s)

Figure A-4 Scenario 3 Rate of CPU time usage. A-6

10 , i G l$ N 8 6 4 w 7.5 - l -

 -s     3
    .3 o         4 tu w a-e                  L       .l4 jj                     Q 8

Q- 5 - _ c 5 5 b 5 2.5 ' ' ' O 500 1000 1500 2000 TIME (s) Figure A-5. Scenario 4 Rate of CPU time usage. 10 , , G U N S 8 - - e w M AJ hil l Mi i -J t_t g r i , t_ [ t 3 u 8 4 - - 2 W 2 - - E 5 Q- , g 0 i 0 1000 2000 3000 ! .' TIME (s) O Figure A-6.- Scenario 5 Rate of CPU time usage. A-7

10 i i i 6 N N' g .8 - S , > l

s 5 ,

l l p 6 h i g l  ! j o "- l ' S 4  : !i .

n 3 il ,I M L hs l 1,

l._ II. ll td j ;j liIIb ( _ dg l $ ) i e ' ' l l M 2 - 12 5 o O O 500 1000 1500 2000 TIME (s) Figure A-7. Scenario 6 Rate of CPU time usage. 10 , i i i E d N 3 8 - U I W p 6 -k - ie o 8 4 - a 1. t t h ) : I a i a a t_i. t. g 2 - W 2 - g .J ,a . i As La2 1l h o

                                '                       '                       '                  '               I O

O 500 1000 1500 2000 2500 TIME (s) s i Figure A-8, S,enario 7 Rate of CPU time usage. A-8

7:- 60 i i , , G m" . N O e

    <s w 4o.       _                                                              l i      _

3 w 2 0 3; _ B E 20 - 1 p 5 5 w l I l a i . _ t

                                                              . . dM O

O 500- .1000 1500 2000

                                                    ... TIME (s)

Figure A-9. Scenario 8 Rate of CPU time usage. 10 i i , N N O g 8 - - e Y . p 6 - 2 o a 4 _ _ N M z in. .. M 2 - - 0 5 0 4 ' ' '

                'O 0           2000                4000               6000                8000 TIME (s) c.

Figure A 10. Scenario 9 Rate of CPU time uuge. A-9

10 i , , E-d N-8 8 - - E-W 3 M y'Ltia,i_A111111 t I t i 1110 - h 3 n_

• t O 4- - _

w E 2 - - O b O O O 500 1000 1500 2000 2500 TIME (s) Figure A-II. Scenario 10 Rate of CPU time usage. 10 , , ,

        'G d

N ~ g8 - S w l- p 6 43 4 2 o 8 s 4 - f e d2 -

                                                        "  '~              

b al 4 O

                                   '               '                '          '              4 O

0 2000 -4000 6000 8000 10000 TIME (s) Figure A-12. Scenario 11 Rate of CPU time usage. A 10

20 i i i i 8 en - N O o 5 w 3 i l , i H , 9 2 10 -

                                                                                                        ~

o i 1 O i i y h JMh J.1l 1 1 L}J l

                       .                         .k.

s , il 1.,LLuh l 1 I i 0 0 500 1000 1500 2000 2500 TIME (s) Figure A-13. Scenario 12 Rate of CPU time usage. 10 i i i 6 d N 8 8 I l w

                                            ~

h 2 5 - O S 3 1]JlLil.I1. ' ' I. u

          $                                                                                  i t a a 5-o

( l i I O 500 1000 1500 2000 2500 TIME (s) 4 Figure A-14. Scenario 13 Rate of CPU time usage. A-I l

)

20 , i o d' N O 8 W 2 u jo _ _ 8 dvl W' , E ' i _-__ k

• ghtfif 4/Mt O

O 2000 4000 6000 TIME (s) Figure A-15. Scenario 14 Rate of CPU time usage.

                                                                                 )

EGG 2341 A-12

                                                   .oc6.........._              ,..,o..         ...        . ,..c                  ,
   ~l.< == =

T,.',","*/ sisWOGRAPHIC DATA SHEET NUREG/CR-3977

      ....oc,,o..........                                                         EG&G-2341
    '#El'AM"*iRERMAL-HYDRAULIC ANALYSES OF PRESSURIZED THERMAL SHOCK SEQUENCES FOR THE H. B. ROBINSON UNIT 2                                                                                    a oa'"'* ' c- a PRESSURIZED WATER REACTOR                                                                               l
   'C"tTdh Fletcher, Mark A. Bolander, Michael E.                            M                  ...,,.,,o.,,,,,o I985 Waterman, John D. Burtt, Benjamin D. Stitt,                                             o ..

j .... 9 Craig M. Kullberg, Cliff B. Davis, Donald M. Ogden April 1985

    ,...,o...o....,........,<,.oo.....,.c,                                      . - a.cra....o.. .. . - ...

EG&G Idaho, Inc. P. O. Box 1625 Idaho Falls, ID 83415 A6047

    ...,0..................oo....,.,.c,                                         ..........=>

Division of Accident Evaluation Technical Office of Nuclear Regulatory Research ,,,,,,,co,,,,,,,,,,,,,,,,,,,,,, U.S. Nuclear Regulatory Commission Washington, DC 20555 83 L. .t . . .. . . . .

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Thermal hydraulic analyses of fourteen hypothetical pressurized thermal shock (PTS) scenarios for the H. B. Robinson, Unit 2 pressurized water reactor were per-formed at the Idaho National Engineering Laboratory (INEL) using the RELAP5 computer code. The scenarios, which were developed at Oak Ridge National Laboratory (ORNL), contain significant conservatisms concerning equipment failures, operator actions, or both. The results of the thermal-hydraulic analyses presented here, along with additional analyses of multidimensional and fracture mechanics effects, will be utilized by ORNL, integrator of the PTS study, to assist the U. S. Nuclear Regulatory Commission in resolving the pressurized thermal shock unresolved safety issue.

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