ML20128M020
| ML20128M020 | |
| Person / Time | |
|---|---|
| Site: | Robinson  |
| Issue date: | 07/31/1985 |
| From: | Chon Davis, Fletcher C, Ogden D EG&G IDAHO, INC. |
| To: | NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| References | |
| CON-FIN-A-6047 EGG-2335, NUREG-CR-3935, NUDOCS 8507250171 | |
| Download: ML20128M020 (291) | |
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-L- A o Idaho National Engineering Laboratory Operated by the U.S. Department of Energy Thermal-Hydraulic Analyses of Overcooling Sequences for the H. B. Robinson Unit 2 Pressurized Thermal Shock Study D
C. Don Fletcher Cliff B. Davis Donald M. Ogden
,s May 1985 8507250 1 PDR P AnocK 0507aj 05000pgg t)DP Prepared for the U.S. Nuclear Regulatory Commission Under DOE Contract No. DE AC07 76lDO1570 ['# E s g s ,,,n,
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NOTICE
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- This report'was prepared as an account of work sponsored by an agency of .
the United States Government. Neither the United Sates Government nor any G. -: '* . agency thereof, nor any of their employees, makes any warranty, expressed t ," _ , or implied, or assumes any legal liability or responsibility for any third party's use, or the results'of such use, of any information, apparatus, product or proc-M;' > \ ess disclosed in this report, or represents that its use by such third party would . 'l; 'not infringe privately owned rights. t' l ,, - - t r
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NUREGICR-3935 EGG-2336 Distribution Category: R4 4 THERMAL-HYDRAULIC ANALYSES OF OVERCOOLING SEQUENCES FOR THE H. B. ROBINSON UNIT 2 PRESSURIZED THERMAL SHOCK STUDY C. Don Fletcher Cliff B. Davis Donald M. Ogden Published May 1985 i EG&G Idaho, Inc. Idaho Falls, Idaho 83415 i
- Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Under DOE Contract No. DE AC07761D01570 FIN No. A6047
ABSTRACT - Oak Ridge National L.aboratory(ORNL), as a part of the Nuclear Regulatory Com-mission's (NRC's) pressurized thermal shock (PTS) integration study for the resolu-tion of Unresolved Safety issue A49, identified overcooling sequences of interest to the 11. B. Robinson PTS study. For each sequence, reactor vessel downcomer fluid pressure and temperature histories were required for the two-hour period following the initiating event. Analyses previously performed at the Idaho National Engineering Laboratory (INEL) fully insestigated a limited number of the sequences using a de-tailed RELAPS model of the 11. B. Robinson, Unit 2 (llBR-2) plant. However, a full investigation of all sequences using the detailed model was not economically prac-tical. New methods were required to generate results for the remaining sequences. Pressure and temperature histories for these remaining sequences were generated at the INEL through a process combining: (a) partial-length calculations using the de-tailed RELAPS model (b) full-length calculations using a simplified RELAPS model, and (c) hand calculations. This report documents both the methods used in this process and the results. The sequences investigated contain significant conservatisms concerning equipment failures, operator actions, or both. Consequently, care should be taken in applying the results presented herein without an understanding of the conservatisms and , assumptions. The results of the thermal-hidraulic analyses presented here, along with additional analyses of multidimensional and fracture mechanics effects, will be utilized by ORNL to assist the NRC in resolving the PTS unresolved safety issue. 3 e FIN No. A6047-Code Assessment and Applications (Transient Analysis) ii l I
SUMMARY
Oak Ridge National Laboratory (ORNL)identi- developed specifically to address the controlling fied overcooling sequences of interest to the phenomena for the group. The simplified model
- 11. II. Robinson Unit 2 (IlllR-2) pressurized ther- was then benchmarked against the detailed mal shock study. Sequence-initiating events include RELAPS model to assure the validity of the isolatable and nonisolatable primary system breaks, simplified model. Pressure and temperature large and small secondary system breaks, steam histories were then generated by combining results generator tube ruptures, and control system failures of a partial-length, detailed model RELAPS calcula-following reactor trip. Some sequences involve com- tion over the initial portion of the transient binations of primary and secondary system breaks. (typically 10-15 minutes) and a simplified model RELAPS calculation over the remainder of the two-Reactor vessel downcomer fluid pressure and hour period. In general the detailed model was temperature histories were required for the two- required during the early portions of transients to hour period following the initiating event in simulate relatively complicated phenomena such as 183 sequences. The histories will be used by ORNL the effects of reactor trip, initiation of high pressure as boundary conditions for fracture mechanics injection, reactor coolant pump trip, and secondary analyses of the reactor vessel wall. steam relief. Ilowever, during later portions of the transients the phenomena are, in general, quasi-Analyses previously performed at the Idah steady or slowly varying; benchmarks between National Engineering Laboratory (INEL) fully detailed and simplified models show the simplified investigated a limited number of the sequences using models adequately represent the phenomena under a detailed, quality-assured RELAPS model of the these circumstances.
IlllR-2 plant.I Performing similar analyses for all 183 overcooling sequences was not economically In addition to pressure and temperature histories, feasible. Instead, pressure and temperature histories Mo i is provided on heat transfer coefficients
- were generated at INEL through a process combm-on the inside surface of the reactor vessel wall. As mg: (1) partial-length calculations using the detailed requested by ORNL, INEL identified for each RELAP5 model, (2) full-length calculations using sequence the previously reported scenario that has
. a simplified RELAPS modci, and (3) hand the most representative heat transfer coefficient.
calcul tmns. Estimates of uncertainty in the pressure and The sequences were grouped at ORNL by initiat- temperature histories are also provided. ing event (small steam line break, large steam line break, small primary break, etc.). These groupings The results of the thermal-hydraulic analyses in this were not convenient for thermal-hydraulic analyses. report represent part of the information required by Therefore, the sequences were regrouped according ORNL for the assessment of the PTS issue. The to the controlling thermal-hydraulic conditions or results of this report are not to be used directly as phenomena. For example, a group titled " Steam an indication of pressurized thermal shock severity Line llreaks With One Affected Steam Generator" for the sequences investigated. Following additional includes sequences with initiating events of large analyses of multidimensional and fracture mechan:cs and small steam line breaks at hot standby and full. effects, ORNL will integrate all results and publish power conditions. This group also includes a report that estimates the likelihood of reactor vessel sequences initiated by reactor trip, with subsequent failure and identifier important esent sequences, failing open of one steam line power-operated relief operator and control actions, and uncertainties. Com-valve. All sequences in the group thus share puter simulations of the sequences were performed
. thermal. hydraulic similarity because heat removal using best-estimate conditions and assumptions.
to the single affected steam generator is the Ilowever, the sequence definitions themselves contain controlling thermal-hydraulic mechanism, significant conservatise assumptions concerning I equipment failures, operator actions or omissions, or Within each new group, the pressure and combinations of the two. Consequently, care should temperature histories were generated for each be taken in applying the results presented herein sequence by consistently applying the following without an understanding of the conservatisms and general method. A simplified RELAPS model was assumptions. iii
Results of the analyses presented here indicate that Medium break lou of-coolant accidents the most thermal-hydraulically sescre sequences for (LOCAs) w cre found to produce not only scry low I'TS are those with secondary-side breaks. The larger temperatures but also low pressures which limited the break and the longer auxiliary feedwater contin- sescrity. Small break 1.OCAs were less sescre than ucs, the more severe the results. A failure of the charg' medium break LOCAs. ing to throttle makes these sequences even more secte.
' Results for reactor trip and steam generator tube Combination primary- and secondary-side breaks rupture sequences were relatively less severe, with were found to be less severe than the corresponding results for the latter controlled mainly by the as- secondary-side-only breaks, due to the lower pri.
~
sumptions of operator action. mary system pressures. 0 0 e av
l I i NOMENCLATURE i AFW Auxiliary feedwater ASO Affected steam generator i DC Downcomer . i
- ECCS Emergency core cooling system 2
HBR 2 - H. B. Robinson, Unit 2
, HPI High pressure injection
= INEL Idaho National Engineering j Laboratory j i
q LOCA Loss-of-coolant accident LPI Low-pressure injection i MBLOCA Medium break loss-of-coolant , accident i MFW Main feedwater . - MSIV Main steam isolation valve NR Narrow range NRC U. S. Nuclear Regulatory [
, Commission !
ORNL Oak Ridge National Laboratory P1 Proportional integral l PORY Power operated relief valve , PTS Pressurized thermal shock i PWR Pressurized water reactor RCP Reactor coolant pump RCS Reactor coolant system
, = RV Reactor vessel j SBLOCA Small break loss-of-coolant accident
, SDV Steam dump valve (turbine bypass)
! SO Steam generator i SOTR Steam generator tube rupture ! USO Unaffected steam generator i l I r I 4 1 Y i r
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CONTENTS AllSIRACT .
. ... . . .. ... . .. ......... .... ........... . . . ... ii SUNINIARY . .. . . ....... . . . .... ........... . ...... .......... . ... iii NONIENCl.ATURE .. .. . .. ... .. .. .. ..... ...... . ...... . . . .. . .. v
- 1. IN TRODUCTION .. .. . ............... ..... .......... ... .......... ....... I ,
- 2. SEQUENCE DElclNITIONS ..... . ........ ...... ..... ............. .......... 2
- 3. NIETilODS ..... ..... . ..... .. ... ...... .... .................... .. . .. 18 3.1 Regrouping of Sequences by Controlling Conditions or l'henomena ... .... . ... 18 3.2 Identification of Equisalent Sequences .................... .. ................ IN 3.3 Application of Stodels ........... ......... .... .......... ...... ....... 18 3.3.1 Description of Detailed Ntodel .. .. ................................... 21 3.3.2 Description of liase Simplified hfodel ...... ............................ 28 3.3.3 Simplified hiodel Variations ................... ........................ 30 3.4 Simplified hfodel llenchmarks ............ ....................... ........... 33 3.4.1 Small Steam fireak Affecting One Steam Generator At ilot Standby ........ 33 ,
3.4.2 Small Steam Ilreak Affecting One Steam Generator At Full Power ...... . 33 3.4.3 I arge Steam tireak Affecting One Steam Generator At flot Standby ........ 33 3.4.4 Steam Ilreak Symmetrically Affecting Three Steam Generators At Full l'ower .................. ..................................... 33 3.4.5 SilLOCA at flot Standby ..... ......................... .... ......... 37 3.4.6 Sill OCA Combined with Steam lireak, Symmetrically Affecting Three Steam Generators at Full l'ower ........................................ 40 3.4.7 Afill.OCA Combined with Small Steam lireak, Affecting Two Steam Generators at l'ull Power ................ ............................. 40
- 4. GROUI' A RiiSUI TS: SECONDARY. SIDE IIREAK AFFECTINO ONE STEANI GENiiRA10R .... ................................................ 44 4.1 Group A Definition ........................... ............. ............... 44 4.2 Small Steam lireak at ilot Standby ........................................... 44 4.3 Small Steam lireak at Full l'ower ..... ........... .......... ................ 46 4.4 1.arge Steam lireak at flot Standby ........................................ . 47 .
4.5 1.arge Steam lireak at I:ull l'ower ....................... ................. ... 49 4.6 Conclusions ................................................................ 49 l $. GROUI'11 Rl! SUI.TS: SECONDARY. SIDE IIRl!AK WITil TilRiill l SYhth1ETRICAl.I.Y AlITCTED SIEANI GENiiRATORS .......................... 51 i 5h
5.1 Group 11 Definition ......................................................... 51 5.2 One Stuck Open Steam PORY and Steamline Check Val,e Failure at ilot Standby ............................................................. 51
$.3 One or More Steam Dump Valses, or Three Steam PORVs Stuck.Open Following Reactor Trip From Full Power ............. ........................ 52 5.4 Conclusions ................................................................ $5
- 6. GROUP C RESULTS: SECONDARY SIDE IIREAKS AFFECTING TWO STEAM GENERATORS. OR TilRFE ASYMMETRICALLY AFFECTED STEAM GENERATORS ......................................................... $6 6.1 Group C Definition ......................................................... $6 6.2 Combinations of Stuck Open Steam PORVs and Steam Dump Valves ............. $6 6.3 Two Stuck.Open Steam PORVs .............................................. 58 6.4 Conclusions ................................................................ $8
- 7. GROUP D RESULTS: REACTOR TRIP FROM IULL POWER WITil NO PRIMARY. OR SECONDARY. SIDE IIREAKS ..................................... 60 7.1 Group D Definition ......................................................... 60 7.2 Results and Conclusions ..................................................... 60
- 8. GROUP E RESULTS: MAIN FEEDWATER OVERFILL ............................ 62 8.1 Oroup E Definition ......................................................... 62 8.2 Results and Conclusions ..................................................... 62
- 9. GROUP F RESULTS: LOCA ..................................................... 63 9.1 Uroup F Definition ......................................................... 63 9.2 SilLOCAs ................................................................. 63 9.3 Mill.OCAs ................................................................. 64 9.4 Conclusions ................................................................ 6$
- 10. GROUP O RESULTS: LOCA COMillNED WITil SYMMl!TRIC SECONDARY.
SIDl! IIREAKS .................................................................. 66 10.1 Uroup G Definition ......................................................... 66 10.2 Results ............................... .................................... 66 10.3 Conclusions ................................................................ 67
- vil
7
- 11. GROUP 11 RESUI.TS: 1.OCA CON 1HINED WITil ASYhlNIETRIC SECONDARY- L SIDE HRI!AKS ........................................ ......................... 68 11.1 Group 11 Definition ......................................................... 48.
fl.2 Results ............................................ ....................... 68 . 11.3 Conclusions ................................................................ 69
- 12. GROUP I RESUI.TS: ISOI.ATAllt.E PRINIARY SIDE IIRl!AKS ..................... 70 12.1 Group I Definition .. ....................................................... 70 12.2 Results ................................................... ................ 70 e
12.3 Conclusions ................................................................ 72 11 GROUP 3 RESUI.TS: STEANI GENERATOR tulle RUPIURES .................... 73 13.1 Group J Definition .......................................................... 73 13.2 Results and Conclusions ................................................. ... 73
- 14. Ill!AT TRANSII!R COElllCIENTS ............................. .........-...... 75
- 15. UNCERTAINill!S .............................................................. 79
- 16. SENSITIVITil!S ................................................................. 82 16.1 Secondary. Side lircaks ....................................................... 82 e
16.2 Reactor Trip With hiinor l' allures ............................................ 85 16.3 Primary. Side lireaks ........................................................ 85 16.4 Combination Primary and Secondary. Side lireaks .............................. M 16.5 Steam Ocnerator Tube Ruptures .............................................. 90 )
- 17. CONCl.USIONS ................................................................ 92 17.1 Secondary. Side lircaks ....................................................... 92 17.2 Reactor Trip with hilnor l' allures ............................................. 93 17.3 Primary. Side litcaks ........................................................ 93 17.4 Combination Primary. and Secondary. Side lireaks .............................. 94 17.5 Steam Generator Tube Ruptures .............................................. 94
- 18. Rlil1!Rl!NCl!S .................................................................. 94 APPENDIX A-Pl Ol'il!D RESUI.TS Olt Rl! ACTOR VI!SSEl. DOWNCONil!R 1:1.U10 PRl!SSURl!S AND 'll!N1Pl!RA'IURi!S ..........................................A.I Slil
f 1 ! FIGURES
- 1. Detailed model nodalization of primary coolant loops (Loop C shown) ........... ..... 22 T 2. Detailed model nodalization of reactor vessel ............................... ....... 23 I .
- 3. Detailed model nodalleation steam generator (SGA shown) ....................... .. 24
;, 4. Detailed model nodalization of feedwater and steam systems ....... .................. 26 I
- 5. Ilase simplified model nodalization ................................................ 29
- 6. Sequence 7-4 simplified model pressure benchmark . ................................ 34
- 7. Sequence 7-4 simplified model temperature benchmark ... ........................... 34
- 8. Sequence 9-25 simplified model pressure benchmark ................................. 35
- 9. Sequence 9 25 simplified model temperature benchmark .. .... ...................... 35 i
- 10. Sequence 8-4 simplified model pressure benchmark ................. ................ 36 I II. Sequence 8-4 simplified model temperature benchmark ... ........................... 36
- 12. Sequence 915 simplified model pressure benchmark ................................. 37
)
. 13. Sequenec 915 simplified model temperature benchmark .............................. 38
- 14. Sequence 3-1 simplified model downcomer pressure benchmark ...................... 38
- 15. Sequence 31 simplified model average RCS temperature benchmark ............... ... 39
- 16. Sequence 31 simplified model downcomer temperature benchmark .................... 39
- 17. Sequence 18 simplified model downcomer pressure benchmark ....................... 40
- 18. Sequence 1-8 simplified model average RCS temperature benchmark ... .......... .... 41
- 19. Sequence 18 simplified model downcomer temperature benchmark .................... 41
- 20. Sequenec 2.ll simplified model downcomer pressure benchmark ...................... 42
- 21. Sequence 211 simplified model average RCS temperature benchmark .................. 42
- 22. Sequence 211 simplified model downeomer temperature benchmark ................... 43 e 23. I!!fect of secondary. side break slic on reactor senci downcomer temperature ........... 83
- 24. I!ffect of secondary side break location on reactor venci downcomer temperature ....... 83
- 25. I!ffect on reactor venci downcomer temperature of initial core power level for secondary. side break sequences ................................ . ................. 84 lx
- 26. Effect of charging throttling failure on reactor vessel downcomer pressure for secondary-side break sequences . .. . . . . . . . . . 84
- 27. Effect of auxiliary feedwater isolation on reactor vessel downcomer temperature for secondary-side break sequences .. . . .... . . . . . . 85
- 28. Effect of auxiliary feedwater isolation on reactor vessel downcomer pressure for secondary-side break sequences . ... . . . . . 86
- 29. Effect of primary system break size on reactor vessel downcomer pressure ... . 86
- 30. Effect of primary system break size on reactor vessel downcomer fluid temperature ... 87
- 31. Effect of initial core power level on reactor vessel downcomer pressure for primary-side breaks . . . .. . . . . . ... . . 88
- 32. Effect of initial core power level on reactor vessel downcomer fluid temperature for primary-side breaks ... .. . .. . . ...... . . 88
- 33. Effect of break location on primary system pressure for primary-side breaks 89
- 34. Effect of primary-side break isolation on reactor vessel downcomer pressure . 89 35 Effect of primary-side break isolation on reactor vessel downcomer fluid temperature 90
- 36. Effect on reactor vessel downcomer fluid temperature of steam dump valve failure to close on demand during normal recovery from SGTR . .. .. . .. . . ... . 91 ,,
TABLES .
- 1. Sequence descriptions-small break LOCA at power .. . 3
- 2. Sequence descriptions-medium break LOCA at power . . .. . . 4
- 3. Sequence descriptions-small break LOCA at hot standby . . . 4
- 4. Sequence descriptions-mrdium break LOCA at hot standby . . ... .... . 5
- 5. Sequence descriptions-small steamline break at power . . . . 5
- 6. Sequence descriptions-large steamline break at power .. . 6
- 7. Sequence descriptions-small steamline break at hot standby . . . . .. 7
- 8. Sequence descriptions-large steamline break at hot standby . . .. .. . . 7
- 9. Sequence descriptions-reactor trip at full power . . .. . . . . 8
- 10. Sequence descriptions-steam generator tube rupture at hot standby . . . 12 .
- 11. Sequence descriptions-isolatable SBLOCA at power .. ..... . . .. . 13
- 12. Sequence descriptions-isolatable MBLOCA at power . . 13 X
- 13. Steamline break event descriptions ....... . ............... .... ................ 14
- 14. Reactor trip and LOCA event descriptions .... .. .... .. . ...... ......... .... 16
- 15. Regrouping of sequences by controlling conditions or phenomena ... .. . .. ..... . 19
- 16. Thermal-hydaulically equivalent sequences .. ............ ........................ 20
- 17. Results of small steam breaks affecting i SG at hot standby .............. ......... . 45
- 18. Results of small steam breaks affecting 1 SG at full power .. . .... . . ........ 47
- 19. Results of large steam breaks affecting i SG at hot standby . ................ ....... 48
- 20. Results of large steam breaks affecting 1 SG at full power . . . . ... .. ............ 50
- 21. Results of one stuck-open steam PORV and steamline check valve failure at hot standby ........ . ... . ......... . . ... . ... ...... . .. ... . . 52
- 22. Results of one stuck-open steam dump valve at full power .......... ........ ... ... 53
- 23. Results of two stuck-open steam dump valves at full power ... .... . .. ... ... 53
- 24. Results of three stuck-open steam dump valves at full power ..... ... .. ............ 54
- 25. Results of five stuck-open steam dump valves at full power ..... . ....... ... ... 54
- 26. Results of three stuck-open steamline PORVs at full power ............. ........... 55
- 27. Results of combination SDV and steam PORV breaks ... ................. ........ 57
- 28. Results of two stuck-open steam PORVs ......................... .. . ... ........ 59
- 29. Results of reactor trip sequences with minor failures . . .. ....... .... .. ........ 61
- 30. Results of main feedwater overfill . ...... .. . . . . .... . .............. ...... 62
- 31. Results: SBLOCAs ... .. . ........... .. ..... ............. ...... ....... 64
- 32. Results: MBLOCAs . .... . .. . ....... ... ........ .. ... . . 65
- 33. Results of LOCAs combined with symmetric secondary-side breaks .... . . . . 67
- 34. Results of LOCAs combined with asymmetric secondary. side breaks .. .. .. ... 69
- 35. Results of isolatable primary-side breaks ...... . .. .. . . ... . . .. .. 71
- 36. Results of steam generator tube ruptures .......... .. . . . . . .. .. ... . 74
- 37. Representative heat transfer coefficients ... ....... ... . ... .. . .. ... .. . 75
- 38. Uncertainties in reactor vessel downcvwr fluid temperature and pressure . .. .. .. 79
- 39. Summary of severe sequences ... .......... ..... ................ .... ... .. 93 t
Xi _ _ _ _ _ _ _ _ _ _ _ _ - - . - - - _ l
THERMAL-HYDRAULIC ANALYSES OF OVERCOOLING SEO.UENCES FOR THE H. B. ROBINSON UNIT 2 PRESSURlZED THERMAL SHOCK STUDY
- 1. INTRODUCTION Rapid cooling of a reactor pressure vessel, accom- descriptions. The reader is cautioned, however, that panied by high coolant pressure, during a transient the sequence descriptions were based on extremely or accident is referred to as pressurized thermal conservative assumptions regarding equipment mal-shock (PTS). In late 1981 the U. S. Nuclear functions, operator actions and omissions, or Regulatory Commission (NRC) designated PTS as combinations of these. Thus, while the analyses an unresolved safety issue and developed a task results represent best-estimate plant responses to the action plan (TAP A-49) to resolve the issue, sequences as defined, they do not represent the most probable plant responses to the sequence-initiating The safety issue exists because rapid cooling at events.
the inner surface of the reactor vessel wall produces thermal stresses within the wall. As long as the Analyses presented in this report were performed resistance te fracture of the reactor vesselis high, for the H. B. Robinson, Unit 2 (HBR-2) pressur-overcooling transients will not cause vessel failure, ized water reactor operated at Hartsville, South However, NRC staff analyses (SECY-82-465) Carolina by Carolina Power and Light Company. showed that certain older plants with copper and The reactor is of Westinghouse three-loop design other impurities in sessel weldments may become and is currently undergoing steam generator sensitive to PTS after several years as the replacement. Prior to steam generator replacement, nil-ductility transition temperature of the weld the unit was operated at reduced power, partially material gradually increases. The purpose of the because of steam generator tube plugging. Analyses I thermal-hydraulic analyses presented in this report presented here assumed a full rated thermal power
. is to quantify the behavior of a plant during various of 2300 MW. This is the power level at which the kinds of postulated severe overcooling transients, plant is expected to operate following a replacement with multiple failures of equipment and with only of steam generators. Other anticipated plant minimal operator corrective action. For each of changes during the steam generator replacement these postulated transients, Oak Ridge National outage were also incorporated into the analyses to Laboratory (ORNL) will calculate both the reactor more realistically model the plant.
vessel temperature distribution and the stresses during the transient as well as the conditional This is the second plant for which PTS thermal-probability of vessel failure if the transient should hydraulic analyses have been performed at INEL. occur. ORNL will publish a report that integrates Similar work, described in Reference 2, was these results to estimate the likelihood of PTS performed for the Oconee-1 plant, a PWR of driving a crack through the reactor vessel wall, and Babcock and Wilcox design. to identify important event sequences, operator and control actions, and uncertainties. This report is organized as follows: the sequences of interest to PTS, as defined at ORNL, appear in This series of analyses is intended to provide Section 2; the methods by which pressure and information to help the NRC staff confirm the temperature histories were determined appear in bases for the screening eriteria in the proposed PTS Section 3; results of the analyses appear in rule (proposed 10CFR 50.61) and to determine the Sections 4 through 13; estimates for heat transfer content required for licensees' plant-specific safety coefficients of the reactor vessel wall appear in
- analysis reports and the acceptance criteria for Section 14; discussions of uncertainties appgar in corrective measures. Section 15 and sensitivities in Section 16; con-clusions are stated in Section 17; and references The analyses presented in this report were appear in Section 18. Plotted results, showing performed using best-estimate modeling assump- reactor vessel downcomer pressure and temperature tions for both plant conditions and operator histories for each sequence, are presented in responses to the events specified in the sequence Appendix A.
1
2, SEQUENCE DEFINITIONS Oak Ridge National Laboratory (ORNL) defined line break is a double-ended rupture of one steam for detailed study sequences of interest to the line downstream of the flow restrictor and upstream HBR-2 pressurized thermal shock study. The of the main steam isolation valve. The term "at , sequence list includes thos. sequences expected to power" means during 2300 N1W full-power opera-have the highest probability of vessel fracture. The tion; "at hot standby" means 100 hours after reac-sequences were primarily grouped by ORNL tor shutd awn, assuming infinite operating time. The according to sequence-initiating event and operat- term "isolatable LOCA" refers to a loss-of-coolant ing power level. Tables I through 12 present the accident in ivhich the operator, after a delay, takes sequence definitions for Groups 1 through 12, action to terminate the loss of primary system in-respectively. Corresponding initiating events for ventory. In fly case of an "isolatable small break each group are as follows: LOCA," for example, it is assumed that the operator isolates the break (in this case closes the Group 1 Small break LOCA at power PORY block valve), thus terminating the primary-Group 2 hiedium break LOCA at power side break. Group 3 Small break LOCA at hot standby Group 4 hiedium break LOCA at hot standby Each sequence is defined by the initial power Group 5 Small steam h,ne break at power level, the initiating event, and subsequent hardware Group 6 Large steam i,me break at power and operator actions as defined in Tables 1 Group 7 Small steam line break at hot standby through 12. Column headings in these tables are Group 8 Large steam line break at hot standby abbreviations for events; Table 13 (for steam line Group 9 Reactor trip at full power break sequences) and Table 14 (for all other Group 10 Steam generator tube rupture at hot sequences) provide more detailed descriptions of the standby events. To simplify the identification of sequences, Group 11 Isolatable small break LOCA at the following numbering scheme was adopted. power , Sequences are identified by two numbers: first, the Group 12 1solatable medium break LOCA at group number, and second, the line number within P * *f the group. Sequence 1-4, for example,is the fourth The reader is cautioned that, after the analysis sequence listed in Table 1. Sequence 1-4 represents presented here had been completed, ORNL made a small break LOCA during full-power operation, significant changes in the number and type of with a hardware failure of the auxiliary feedwater sequences in each of the above groups. Therefore, (AFW) flow controller and the failure of the opera-there is not a one-to-one correspondence between tors to manually throttle AFW based on steam the sequences reported here and in the ORNL final generator levels. report. For these additional sequences, ORNL used the results shown in this report as bounding cases. For each sequence, ORNL required for its frac-ture mechanics analyses a reactor vessel downcomer in the terminology used here: (1) a small break fluid pressure and temperature history for the two-LOCA (Groups 1 and 3) is initiated by a single, hour period following the initiating event, an stuck-open pressurizer power-operated relief valve indication of a representative heat transfer coeffi-(PORV);(2) a medium break LOCA (Groups 2 and cient on the reactor vessel wall, and an estimate of
- 4) is initiated by a 0.0635-m (2-1/2 in.) diameter hot uncertainties in the pressure and temperature leg break; (3) a small steam line break is a single, histories. This report documents both the methods stuck-open steam line PORV; and (4) a large steam used to provide this information and the results.
e 2
P dddddddddddddddddd dddddddddddddddddd T nnnnnnnnnnnnnnnnnn nnnnnnnnnnnnnnnnnn aaaaaaaaaaaaaaaaaa aaaaaaaaaaaaaaaaaa P mmmmmmmmmmmmmmmmmm eeeeeeeeeeeeeeeeee T mmmmmmmmmmmmmmmmmm e eeeeeeeeeeeeeeeee M dddddddddddddddddd C d'ddddddddddddddddd P nnnnnannnnnnnnnnnn A ooooocoooooooooooo nnnnnnnnnnnnnnnnnn W R I oooooooooooooooooo F iiiiiiiiiiiiiiiiiipppppppppppppppppp I s r s r s r s r s r s r ss r rsrsrsrsrsrsrsrsss r r r r r r r r r r r r r r r r r r r r r R uuuuuuuuuuuuuuuuuu M TTTTTTTTTTTTTTTTTT cccccccccccccccccc . cccccccccccccccccc OOOOOOOOOOOOOOOOOO dddddddddddddddddd E nnnnnnnnnnnnnnnnnn aaaaaaaaaaaaaaaaaa S O mmmmmmmmmmmmmmmmmm e L C eeeeeeeeeeeeee eeee dddddddddddddddddd m r m mmmmm r r r r r r mmm r r r mmmmr rr r nnnnnnnnnnnnnnnnnn a a laaaaa aaa aaaa V oooooooooooooooooo l a l a l ll l aaaaa laaa l l laaaa ll l I F eeeeeeeeeeeeeeeeee l l eeeeeeeeeeeeeeeeee l l ll l l l l l lll s s s s s s s s s s ss s s s oooooooooooooooooo s s s W vl vl v v vvv lt vv vt vvvv l M F let t ol et t oll eeeeet eeet ol leeee CCCCClCCCCCCCCCCCCG A l ll l l l l l lll l l l lll r l ollll r h hrh rhhhhh h h iiigggt hhh h hhhh R gt igt iiiiigggggt iiiigggg 1 ih dddddddddddddddddd nnnnnnnnnnnnnnnnnn aaaaaaaaaaaaaaaaaa T 1 o tohohhhhh t et r o ooooo ooo tohhhohhhh et t t r t t t et t t et r oooo t t t t r mmmmmmmmmmmmmmmmmm ii r ol ur u r r r r r ur r r ur rr r iol roooool J N eeeeeeeeeeeeeeeeee dddddddddddddddddd r ariar PFPFPPPPPFPPPFPPPP iiiiii r r r ariiioool r r iariiiioooo rr r I nnnnnnnnnnnnnnnnnn P l oooooooooooooooooo l s ss s s s s s s ssss s s s s s 9 r r r r r r r r r r rr r r r r r r uuuuuuuuuuuuuuuuuu cccccccccccccccccc cccccccccccccccccc dddddddddddddddddd OOOOOOOOOOOOOOOOOO nnnnnnnnnnnnnnnnnn aaaaaaaaaaaaaaaaaa J mmmmmmmmmmmmmmmmmm ee eeeeeeeeeeeeeeee dddddddddddddddddd N I dddddddddddddddddd nnnnnnnnnnnnnnnnnn nnnnnnnnnnnnnnnnnn aaaaaaaaaaaaaaaaaa C oooooooooooooooooo N mmmmmmmmmmmmmmmmmm eeeeeeeeeeeeeeeeee C s s s s s s s s s s s s s s s ss s E dddddddddddddddddd A r r r r r r r r r r r r r r r r r r uuuuuuuuuuuuuuuuuu G nnnnnnnnnnnnnnnnnn cccccccccccccccccc cccccccccccccccccc L oooooooooooooooooo OOOOOOOOOOOOOOOOOO G dddddddddddddddddd eeeeeeeeeeeeeeeeee S t t t t t t t t t t t t t t t t t t 1 I aaaaaaaaaaaaaaaaaa r r r r r r r r r r r r r r r r r r S eeeeeeeeeeeeeeeeee nnnnnnnnnnnnnnnnnn eeeeeeeeeeeeeeeeee O GGGGGGGGGGGGGGGGGG S I e G r dddd ddddddd dd S e nnnn nnnnnnn nn W w E S aaaae mmmms eeermmmmmms aaaaaaa eaa mm F AAAAAAAAAAAAAAAAAA O os s s r o A NNNNNNNNNNNNNNNNNN o p L eeeel oooeeeeeeel ee dddd cl l l ccc ddddddd cdd . C nnnno nnnnnnnonn t V oooot tooo t t ooooooot oo a D eeeel s s s s l l l siiii aaaa eeeeeeel ee s s s s s s si s s s S a dd dddddd ddd ddd A ooooffff ooooooof lool ll l lCCCCI 235GCCCCCCICC O l ll l l l lee l leeeeee l ll l l leeel l leee l l C T loo l l l ll l l oooooo r r r r r r looo l l r r looo r r l l O U r r t t t t t t t t r t t t t tr L dddd A nnddnnnnnnd ooeeooooooeoooeooo t nnndnnn E nnnn s W cceecccccceccceccc k S aaaa eeee oor eet ff r oooooor ooor ooo f f a O L mmmm eeee innnn i ii F A t t t t t t t et t t uuvvuuuuuuv uuuv uuu et t t e r C dddd l l ll I I I2 AAOOAAAAAAOAAAOAAA b nnnn nnnn Y oooo oooo l R eeee l O s s s s ls ls lss a P l ooooAAAAiiiiAAAAAA ll l aaaa l dddddddddddddddddd m CCCCNNNNFFFFNNNNNN nnnnnnnnnnnnnnnnnn s aaaaaaaaaaaaaaaaaa
- T mmmmmmmmmmmmmmmmmm eeeeeeeeeeeeee - eee s dddddddddddd nnnnnnnnnnnn C dddddddddddddddddd n V aaaaaaaaaaaa A nnnnnnnnnnnnnnnnnn s s oooooooooooooooooo i
t o L V mmmmmmmmmmmmeeeeee WF eeeeeee ee eeennnnnn ddddddddddddiiiiiil l ll ll s s s s s s ss s s s ss s s s ss eeeeeeeeeeeeeeeeee p G nnnnnnnnnnnnI I I I 23 A t t t t t t t t t t t t t t t aaaaaaaaaaaaaaaaaa t t t i E oooooooooooo uuuuuuuuuuuuuuuuuu r R nnnnnn t t t t t t cccccccccccccccccc t t t t t t t t t t t t c s r sr rsr s r r rs r rsr s r r s s s s s oooooo AAAAAAAAAAAAAAAAAA s W uuuuuuuuuuuus s d e F cccccccccccc l l ls cccccccccccciiiiii OOOOOOOOOOOOFFFFFF aaaaaa lsls ls e
. c d n
n dddddddddddddddddd nnnnnnnnnnnnnnnnnn E S a e P aaaaaaaaaaaaaaaaaa O m u I R mmmmmmmmmmmmmmmmmm eeeeeeeeeeeeeeeeee L d e q dddddddddddddddddd C e T n nnnnnnnnnnnnnnnnnn V o
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oooooooooooooooooo I S e s s ss s s s s s s s ss s s s s s s U r rr r r r r r r r r r r r r r r r uuuuuuuuuuuuuuuuuu M AAAAAAAoAAAAAAAAAA l 1 T cccccccccccccccccc NNNNNNNCNNNNNNNNNN cccccccccccccccccc OOOOOOOOOOOOOOOOOO le b a O 1 23456789012345678 O I234567890I 2345678 T N 1 11 1 1 1 1 1 1 N 1 i 1 1 1 1 11 1
Table 2. Sequence descriptions-medium break LOCA at power N_O TURB TRIP FW REG VLV PORV CLOSE SDV CLOSE SI SGL GEN HP INJ MFIV CLOSE MFW PMP TP I Occurs on demand Occurs on demand Cose on demand Cose on demand Generated on demand Occurs on demand Close on demand Trip on demand 2 Occurs on demand Occurs on demand Gose on demand Close on demand Generated on demand Occurs on demand Gose on demand Trip on demand 3 Occurs on demand Occurs on demand Cose on demand Close on demand Generated on demand Occurs on demand Gose on demand Trip on demand 4 Occurs on demand Occurs on demand Cose on demand Close on demand Generated on demand Occurs on demand Gose on demand Trip on demand 5 Occurs on demand Occurs on demand NA 1 fails to close Generated on demand Occurs on demand Gose on demand Trip on demand 6 Occurs on demand Occurs on demand NA 2 fail to close Generated on demand Occurs on demand Close on demand Trip on demand 7 Occurs on demand Occurs on demand NA 3 fail to close Generated on demand Occurs on demand Close on demand Trip on demand 8 Occurs on demand Occurs on demand NA 5 fail to close Generated on demand Occurs on demand Close on demand Trip on demand 9 Occurs on demand Occurs on demand Fails on I line Close on demand Generated on demand Occurs on demand Cose on demand Trip on demand 10 Occurs on demand Occurs on demand Fails on I line Close on demand Generated on demand Occurs on demand Close on demand Trip on demand 11 Occurs on demand Occurs on demand rails on 2 lines Close on demand Generated on demand Occurs on demand Close on demand Trip on demand N_O MSIV CLOSE AFW ACT Alv AUTO AFW SG ISO ACC INJ TIIR AFW Ri!R ACT
-4 NA Actuates on demand Auto controlled NA Occurs on demand Pr or to high level alarm Occurs on demand 2 NA Actuates on demand Auto controlled NA Occurs on demand Failure to throttle Occurs on demand 3 NA Actuates on demand Overfeed NA Occurs on demand Prior to high level alarm Occurs on demand 4 NA Actuates on demand Overfeed NA Occurs on demand Failure to throttle Occurs on demand 5 NA Actuates on demand Auto controlled NA Occurs on demand Prior to high level alarm Occurs on demand y 6 NA Actuates on demand Auto controlled NA Occurs on demand Prior to high level alarm Occurs on demand 7 NA Actuates on demand Auto controlled NA Occurs on demand Prior to high level alarm Occurs on demand 8 Close on demand Actuates on demand Auto controlled NA Occurs on demand Prior to high level alarm Occurs on demand 9 NA Actuates on demand Auto controlled NA Occurs on demand Prior to high level alarm Occurs on demand 10 NA Actuates on demand Auto controlled NA Occurs on demand Failure to throttle Occurs on demand il NA Actuates on demand Auto controlled NA Occurs on demand Prior to high levet alarm Occurs on demand Table 3. Sequence descriptions-small break LOCA at hot standby N_O SDV CLOSE SI SGL GEN IIP INJ MFIV CLOSE MFW PMP TP MSIV CLOSE Alv ACT I Close on demand Generated on demand Occurs on demand Close on demand Trip on demand NA Actuates on demand 2 Close on demand Generated on demand Occurs on demand Close on demand Trip on demand NA Actuates on demand 3 I fails to close Generated on demand Occurs on demand Close on demand Trip on demand NA Actuates on demand N_O AFW AUTO AFW SG ISO ACC INJ TilR AFW RiiR ACT I Auto controlled NA Occurs on demand Prior to high level alarm Occurs on demand 2 Auto controlled NA Occurs on demand Failure to throttle Occurs on demand 3 Auto controlled NA Occurs on demand Prior to high level alarm Occurs on demand
T n E e e C o S l lt l t A s ted O C i t o r t o n L r W a a C I dtdtd h h F u P e e c m t V r ore A Ad e I S AAAAAAAAAAAAAAAAAAAAAAAA R l l itiu u qreqreq toriu t h NNNNNNNNNNNNNNNNNNNNNNNN 1 1 e ueue rl rl r . T s iasias AFAFA E S dddddddddddddddddddddddd O P nnnnnnnnnnnnnnnnnnnnnnnn L T aaaaaaaaaaaaaaaaaaaaaaaa C P mmmmmmmmmmmmmmmmmmmmmmmm eeeeeeeeeeeeeeeeeeeeeeee V M dddddddddddddddddddddddd ddddd P nnnnn I S A nnnnnnnnnnnnnnnnnnannnnn oooooooooooooooooooooooo aaaaa h t N W pppppppppppppppppppppppp J mmmmm eeeee F f iiiiiiiiiiiiiiiiiiiiiiii r r r r r r rrr r r r r r r r r rr r rr rr N ddddd d A TTTTTTTTTTTTTTTTTTTTTTTT I nnnnn n P ooooo a I I ss s s s T m r r r r r d C e uuuuu ccccc P n d ccccc T a A n OOOOO dddddddddddddddddddddddd P m R o E nnnnnnnnnnnnnnnnnnnnnnnn 1 e l s S aaaaaaaaaaaaaaaaaaaaaaaa A P d n i R r u O mmmmmmmmmmmmmmmmmmmmmmmm eeeeeeeeeeeeeeeeeeeeeeee o c c L dddddddddddddddddddddddd W O C nnnnnnnnnnnnnnnnnnnnnnnn F ip V oooooooooooooooooooooooo 1 r N T I F eeeeeeeeeeeeeeeeeeeeeeee s s s s s s ss s s s s s s s s s s s s s s s s mm r r m r f h oooooooooooooooooooooooo aa a CCCCCCCCCGCGCCCCCCGCCGCC l l ll ll l l l l l l laa la W l eeeee l vvll v l m r F leet ool t tt e A l d la E n a hhggt thh r rh S a dddddddddddddddddddddddd R ig O m e W l e v nnnnnnnnnnnnnnnnnnnnnnnn 1 1 iihhooh L d F aaaaaaaaaaaaaaaaaaaaaaaa T oo t t o y C n A l e N mmmmmmmmmmmmmmmmmmmmmmmm eeeeeeeeeeeeeeeeeeeeeeee t t eet r r V o h E dddddddddddddddddddddddd r r uu r o b I e R g G nnnnnnnnnnnnnnnnnnnnnnnn iiool r r iaar l ii d F s I I ih oooooooooooooooooooooooo PPFFP n N 1 o T L a C to r G dddddddddddddddddddddddd eeeeeeeeeeeeeeeeeeeeeeee r S e t s o e I t t t t t t t t t t t t t t t t t t t t t t t t aaaaaaaaaaaaaaaaaaaaaaaa r r r r r r rrr r r r r r r rr r r r r r r r i r S eeeeeeeeeeeeeeeeeeeeeeee t P w nnnnnnnnnnnnnnnnnnnnnnnn o o p eeeeeeeeeeeeeeeeeeeeeeee GGGGGGGGGGGGGGGGGGGGGGGG h d dddd eeee n t r r r r t a O iiii . a a m S I uuuu qqqq k eeeed A J N d e d n a dddddddddddddddddddddddd G S r rr r s s s s te C I n a e V nnnnnnnnnnnnnnnnnnnnnnnn aaaala O P I o s N J m e b r L V aaaaaaaaaaaaaaaaaaaaaaaa mmmmmmmmmmmmmmmmmmmmmmmm eeeeeeeeeeeeeeeeeeeeeeee WF s rsss r uuuui r r so L I r u d dddddddddddddddddddddddd A cccct e I c n G cccco k c C o n E nnnnnnnnnnnnnnnnnnnnnnnn OOOON a O C s i R oooooooooooooooooooooooo e r A r u l s s s s s s sss ss s s s s s sss s s ss s r r r r r r rr r r r r r r r r r r r r r rr r c m W uuuuuuuuuuuuuuuuuuuuuuuu cccccccccccccccccccccccc b c a F cccccccccccccccccccccccc O e OOOOOOOOOOOOOOOOOOOOOOOO m t s u d n ddddd i a l O leeeee l l l l d l a T l ll l looooo e N E m e U r tr r r t t t r m d O m ddddddddddddd nnnnnnnnnnnnn dddd nnnn A tnnnnn ooooo G n S s E aaaaaaaaaaaaaeeeee aaaa ccccc s
- L G d o I S
O mmmmmmmmmmmmms eeeeeeeeeeeeel l lll ooooos s s s s eemmmm s ooeeee WF ooooo G s ddddddddddddd cccccl l dddd A t t t t t uuuuu n S te a S n L C nnnnnnnnnnnnnooooo cc nnnn AAAAA i o I S r e W o V ooooooooooooot t t t t t t s s oo oooo t n F A i D eeeeeeeeeeeeel s l lsllsl l eeee p e A N t s s s s s s s s s s s s siiiiiii aaaaaaa s s s s i r G ip S ooooooooooooofffffff loooo l l l l l l l l l c r c lCCClCCCCCClCCCCI I III 22CCCC s s e e ddddd d nnnnn d aaaaa e d n d e dddddddddddddddddddd T mmmmm eeeee
> c E a O le c L nnnnnnnnnnnnnnnnnnnn C ddddd n S m T l
o n C aaaaaaaaaaaaaaaaaaaa mmmmmmmmmmmmmmmmmmmm A nnnnn e O e l tr eeeeeeeeeeeeeeeeeeee eeee ooooo u L d l A n e P s s s s W q C n o c u R ddddddddddddddddddddoooo nnnnnnnnnnnnnnnnnnnncccc l lll F ss s s s teeeee o W q A t t t t e V e F o e L oooooooooooooooooooo oooo aaaaa uuuuu a S D s t V s ss s s sss ss s ss ss s ss s st t t t t t t t t S S o A u K eeeeeeeeeeeeeeeeeeee s s s s s s s s s s s s s s s s s s s s sl lsl s s ccccc C A C ooooooooooooooooooooiiiil AAAAA _ aaaa CCCCCCCCCCCCCCCCCCCCFFFF l lll l l lll l l l l l llll l _ 4 5 l e e b l b a l O I a O O T O_ N N T N I 234567R90l 2345678901234 1 I 1 1 1 1 11 1 1 22222 N 12345 u _
Table 5. (Continued) NO AFW ACT AFW AUTO AFW SG ISO THR AFW IIP INJ TliR llPI CH 6 Actuates on demand Auto controlled Not isolated Prior to high level alarm Occurs on demand Failure to throttle 7 Actuates on demand Auto controlled Not isolated Failure to thrott!c Occurs on demand As required 8 Actuates on demand Auto controlled Not isclated Failure to throttle Occurs on demand Failure to throttle 9 Actuates on demand Overfeed Occurs as required Prior to high level alarm Occurs on demand As required 10 Actuates on demand Overfeed Occurs as required Prior to high level alarm Occurs on demand Failure to throttle II Actuates on demand Overfeed Occurs as required Failure to throttle Occurs on demand As required 12 Actuates on demand Overfeed Not isolated Prior to high level alarm Occurs on demand As required 13 Actuates on demand Overfeed Not isolated Failure to throttle Occurs on demand As required 14 Actuates on demand Auto controlled Occurs as required Prior to high level alarm Occurs on demand As required 15 Actuates on demand Auto controlled Occurs as required Prior to high level alarm Occurs on demand Failure to throttle 16 Actuates on demand Auto controlled Occurs as required Failure to throttle Occurs on demand As required 17 Actuates on demand Auto controlled Not isolated Prior to high level alarm Occurs on demand As required 18 Actuates on demand Overfeed Occurs on demand Prior to high level alarm Occurs on demand As required 19 Actuates on demand Auto controlled Occurs as required Prior to high level alarm Occurs on demand As required 20 Actuates on demand Auto controlled Not isolated Prior to high Icvel alarm Occurs on demand As required 21 Actuates on demand Auto controlled Occurs as required Prior to high level alarm Occurs on demand As required 22 Actuates on demand Auto controlled Occurs as required Prior to high level alarm Occurs on demand Failure to throttle 23 Actuates on demand Auto controlled Occurs as required Failure to throttle Occurs on demand As required 24 Actuates on demand Overfeed Occurs as required Prior to high level alarm Occurs on demand As required a Table 6. Sequence descriptions-large steamline break at power NO CKVL RP CL MSIV RP CL RM MSIV CL SDV CLOSE SI SGL GEN FW REG VLV MFIV CLOSE MFW PMP TP I Closes on demand NA Close on demand NA Generated on demand Occurs on demand Clo e on demand Trip on demand 2 Closes on demand NA Close on demand NA Generated on demand Occurs on demand Close on demand Trip on demand 3 Closes on demand NA Close on demand NA Generated on demand Occurs on demand Close on demand Trip on demand 4 Closes on demand NA Close on demand NA Generated on demand Occurs on demand Close on demand Trip on demand 5 Closes on demand NA Close on demand NA Generated on demand Occurs on demand Cose on demand Trip on demand 6 Closes on demand NA Close on demand NA Generated on demand Occurs on demand Close on demand Trip on demand 7 Closes on demand NA Close o a demand NA Generated on demand Occurs on demand Close on demand Trip on demand 8 Closes on demand NA Close on demand NA Generated on demand Occurs on demand Close on demand Trip on demand 9 Closes on demand NA Close on demand NA Generated on demand Occurs on demand Close on demand Trip on demand NO AFW ACT AFW AUTO AFW SG ISO THR AFW HP INJ THR HPI Cll i Actuates on demand Auto controlled Occurs as required Prior to high level alarm Occurs on demand As required 2 Actuates on demand Auto controlled Occurs as required Prior to high level alarm Occurs on demand Failure to throttle 3 Actuates on demand Auto controlled Occurs as required Failure to throttle Occurs on demand As required 4 Actuates on demand Auto controlled Occurs as required Failure to throttle Occurs on demand Failure to throttle 5 Actuates on demand Auto controlled Not isolated Prior to high level alarm Occurs on demand As required 6 Actuates on demand Suto contro!!cd Not isolated Prior to high level alarm Occurs on demand Failure to throttic 7 Actuates on demand Auto controlled Not isolated Failure to throttle Occurs on demand As required 8 Actuates on demand Overfeed Occurs as required Prior to high level alarm Occurs on demand As required 9 Actuates on demand Overfeed Not isolated Prior to high level alarm Occurs on demand As required
. . . , e
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ooooooocoooo sssss s st s s s s W oooooo s s s s ss F eeeeeeeeeeee F eeeeee t t t t t t A t t t t aaaaaaaaaaaa t t t t t t t t A aaaaaa uuuuuu uuuuuuuuuuuu t t t t t t t t t t t t t t t t
. ccccccccctccc cccccc AAAAAAAAAAAA AAAAAA P dddddd T nnnnnn . E aaaaaa S P mmmmmm eeeeee O M dddddd L P nnnnnn C oooooo V e e W F uiiiiippppp le I l lt r r r r r r l t AAAAAAAAAAAA M l t S i t o to TTTTTT i C
t o M NNNNNNNNNNNN C h r h r h r I I dtddt dtdddd P e eeddddddd eeeeeee P e eeee l f r itiuu u orirt or ii iiii r r r r r r iuuuuuuu E dddddd nnnnnn I I iu toriii r iuuuu r r r R qreqqreqqqqqqq S aaaaaa R qeqqqq e mmmmmml e ru eeee P dddddddddddd i l rlu ree rlu r r eeeeeee r r rr r O eeeeee i r l r r r r T nnnnnnnnnnnn aaaaaaaaaaaa T sia s sias s s s sss L dddddd T siassss P mmmmmmmmmmmm AFAAFAAAAAAA C nnnnnn AFAAAA M eeeeeeeeeeee dddddddddddd V oooooo P I F eeeeee nnnnnnnnnnnn s s s s s oooooo s W oooooooooooo M l l l l ll F iiiiiiiiiiiipppppppppppp CCCCCC r r r r r r r r r r r r M TTTTTTTTTTTT dddddddddddd nnnnnnnnnnnn dddddd nnnnnn aaaaaaaaaaaa aaaaaa J mmmmmmmmmmmm eeeeeeeeeeee dddddd nnnnnn J mmmmmm eeeeee aaaaaa dddddd NI dddddddddddd nnnnnnnnnnnn N mmmmmmNI eeeeee nnnnnn dddddddddddd oooooooooooo E oooooo E nnnnnnnnnnnn P G dddddd P I ss s s ss y S aaaaaaaaaaaa l l ss s s s s s s s sss r r r r r r r r r r r r nnnnnn I r r r r r r mmmmmmmmmmmm uuuuuuuuuuuu y L oooooo uuuuuu b O eeeeeeeeeeee cccccccccccc G dddddd cccccc L dddddddddddd cccccccccccc b eeeeee cccccc d C OOOOOOOOOOOO d S t t t t t OOOOOO n V nnnnnnnnnnnn oooooooooooo n I taaaaaa r r r r r r a t I eeeeeeeeeeee a S eeeeee nnnnnn F s s s s s s s s s s s s t eeeeee s M oooooooooooo s GGGGGG a CCCCCCCCCCClC l l l l l ll l l t o t o mm mmm h mm r r mm r r mmmmmm r r r r r r h dddddd r r r r r t aa aa aaaaaa nnnnnn aa l aaa a laa l laa l l ll lll aaaaaa t E aaaaaa laa laaa l l l l eeeeeeeeeeee ll l lllll a S mmmmmm l l eeeeee l ll
. k dddddddddddd nnnnnnnnnnnn W vs l t vvl v v vvvv k O
L eeeeee dddddd F W vvt vvv l a t aaaaaaaaaaaa F leetoleet eeeeee leetoleee e mmmmmmmmmmmmA l l ol l l ll l a C nnnnnn l l l r N E eeeeeeeeeeee hhh r hh rhhhhhh h e V oooooo A R hh ggtiggg r h hhh dddddddddddd R ggtiggt iiiiiigggggg r eeeeee b i i D i ihh ohhh ii G ihh tohh tohhhhhh b l l s s s s s s nnnnnnnnnnnn i S ooooooi e ooooooooooooT oo oo oooooo T oo t ooo L e CCClCCC l l ll et et t t et t t n G dddddddddddd t t r rr t r ur r ur r r r r r t t t t t r r u r r r r i S eeeeeeeeeeee ool iool in iiool iooo irr iariiariiiiiioooooo t t t t t t t t t t t t
'l aaaaaaaaaaaa i l r r iarii r r I
r r r r r r r r r r r r r r r r r r m S eeeeeeeeeeee PPFPPfPPPPPP dddddd PPFPPP a nnnnnnnnnnnn eeeee eeeeeee m L nnnnnn e GGGGGGGGGGGG a C aaaaaa t e mmmmmm eeeeee s t V s I S dddddd l e nnnnnn l a ddd d d dd g M oooooo ddd nnn d e eee e e ee f eeeeee m dddddddd dn OS r r r r r r r r s s s s s s O aaa r nnnnnnnn iii uuu i i u u iuu i a O loooooo S nmm i i qu s E aaaaaaaa ee a qqq q q qq l ll l l l CCCCCCI eee mmmmmmmms I S O eeeeeeeel l s oos oe Gem eeeddd rr r eee ededee r e r e r r - G dddd ede r e s L ddddddddccl d S c ss ss ss taaa t t s ala ala t st s s a s S annt st cooa al a n C nnnnnnnnoo o n aaall l ooos o s os n o V oooooooot t t o W r r r s s s r s r s r rs o L W r r r slos s ss r so i eeeeeeeel s sl l e F uuuiii uiui uu i C F uuuiui t p D s s s s s s s siii aaa s A ccct t t ct ct cc cccooocococc t P A c c ct ct cccoco S loooooooofff o p i r CClCCCCCCI I2Cl l l l ll OOONNNONONOO i r R OOONON c c V s s I S AAAAAA e e M NNNNNN d d e ddddddddddd nnnnnnnnnnn dddddd eeeeee dddd e dddd c L aaaaaaaaaaa O l l l l l l leeee ll l c dddddd O leeeel l l n C mmmmmmmmmmm eeeeee eeeee e T l l l l oooooo l l ll l loooo n L nnnnnn T loooo r l l l r r e P s U r tr t t rt r r r t r r r t t r e C aaaaaa U mmmmmmA tr t t tnnnndd u q R ddddddddddd nnnnnnnnnnnc o l A tnnnnnndd ooooooeeoooo tnnnnt u P eeeeee ooooee L ooooooooooo cccccceecccc q dddddd W ccccee e V oW ff e R nnnnnn F oooor rff S s s s s s s ss ss s t eeeeeeeeeee F toooooor r toooo eet t L tuuuuv ee t t t v
, K s s s s s s s s s s s ls A t t t uuuuuuv v uuuu t t t S V ooooooA s s s s s s AAAAOO C oooooooooooi a AAAAAAOOAAAA K eeeeee ClCCCCCCCCCCF l ll l ll l l l s s s s s s
. . C loooooo 7 8 l l l l l CCCCCC l
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Table 9. Sequence descriptions-reactor trip at full power = NO TURB TRIPS FW REG VLV PORV CLOSE SDV CLOSE SI SGL GEN MFIV CLOSE MFW PMP TP I Occurs on demand Occurs on demand Close on demand Close on demand NA NA NA 2 Occurs on demand Occurs on demand NA 1 fails to close Occurs on demand Close on demand Trip on demand ' 3 Occurs on demand Occurs on demand ' NA I fails to close Occurs on demand Close on demand Trip on demand 4 Occurs on demand Occurs on demand NA 1 fails to close Occurs on demand Cose on demand Trip on demand 5 Occurs on demand Occurs on demand NA 1 fails to close Occurs on demand Close on demand Trip on demand 6 Occurs on demand Occurs on demand NA 1 fails to close Occurs on demand Close on demand Trip on demand 7 Occurs on demand Occurs on demand NA 1 fails to close Occurs on demand Close on demand Trip on demand 8 Occurs on demand Occurs on demand NA I fails to close Occurs on demand Close on demand Trip on demand 9 Occurs on demand Occurs on demand NA 2 fail to close Occurs on demand . Close on demand Trip on demand 10 Occurs on demand Occurs on demand NA 2 fail to close Occurs on demand Close on demand Trip on demand II Occurs on demand Occurs on demand NA 2 fail to close Occurs on demand Close on demand Trip on demand 12 Occurs on demand Occurs on demand NA 2 fail to close Occurs on demand Close on demand Trip on demand 13 Occurs on demand Occurs on demand NA 2 fail to close Occurs on demand Close on demand Trip on demand 14 Occurs on demand Occurs on demand NA 3 fail to close Occurs on demand Close on demand Trip on demand 15 Occurs on demand Occurs on demand NA 3 fail to close Occurs on demand Close on demand Trip on demand 16 Occurs on demand Occurs on demand NA 3 fail to close Occurs on demand Close on demand Trip on demand 17 Occurs on demand Occurs on demand NA 3 fail to close Occurs on demand Close on demand Trip on demand 18 Occurs on demand Occurs on demand NA 3 fail to close Occurs on demand Close on demand Trip on demand 19 Occurs on demand Occurs on demand NA 5 fail to close Occurs on demand Close on demand Trip on demand 20 Occurs on demand Occurs on demand NA 5 fail to close Occurs on demand Close on demand Trip on demand 21 Occurs on demand Occurs on demand NA 5 fail to close Occurs on demand Close on demand Trip on demand 22 Occurs on demand Occurs on demand NA 5 fail to close Occurs on demand Close on demand Trip on demand 23 Oecurs on demand Occurs on demand NA 5 fail to close Occurs on demand Close on demand Trip on demand 24 Occurs on demand Occurs on demand NA 5 fail to close Occurs on demand Close on demand Trip on demand 25 Occurs on demand Occurs on demand Fails on I line Close on demand Occurs on demand Close on demand Trip on demand NO MSIV CLOSE AFW ACT AFW AUTO AFW SG ISO HPIN3 THR HPI CH THR AFW I NA NA NA NA NA NA NA 2 NA Actuates on demand Auto controlled NA Occurs on demand As required Prior to high level alarm 3 NA Actuates on demand Auto controlled NA Occurs on demand As required Failure to throttle 4 NA Actuates on demand Auto controlled NA Occurs on demand Failure to throttle Prior to high level alarm 5 NA Actuates on demand Auto controlled NA Occurs on demand Failure to throttle Failure to throttle 6 NA Actuates on demand Overfeed NA Occurs on demand As required Prior to high level alarm 7 NA Actuates on demand Overfeed NA Occurs on demand As required Failure to t7rottle 8 NA Actuates on demand Overfeed NA Occurs on demand Failure to throttle Prior to high level alarm 9 NA Actuates on demand Auto controlled NA Occurs on demand As required Prior to high level alarm 10 NA Actuates on demand Auto controlled NA Occurs on demand As required Failure to throttle 11 NA Actuates on demand Auto controlled NA Occurs on demand Failure to throttle Prior to high level alarm 12 NA Actuates on demand Auto controlled NA Occurs on demand Failure to throttle Failure to throttle 13 NA Actuates on demand Overfeed NA Occurs on demand As required Prior to high level alarm 14 NA Actuates on demand Auto controlled NA Occurs on demand As required Prior to high Icvel alarm 15 NA Actuates on demand Auto controlled NA Occurs on demand As required Failure to throttle 16 NA Actuates on demand Auto controlled NA Occurs on demand Failure to throttle Prior to high level alarm 17 NA Actuates on demand Auto controlled NA Occurs on demand Failure to throttle Failure to throttle
mm r r mrmmmP r r r T ddddddddddddddddddddddddd nnnnnnnnnnnnnnnnnnnnnnnnn m m mr r r aa a aaa aaaaaaaaaaaaaaaaaaaaaaaaa a la a laa la laaa P l l l mmmmmmmmmmmmmmmmmmmmmmmmm eeeeeeeeeeeeeeeeeeeeeeeee l a a la W l eeeeeeee l l l l l M ddddddddddddddddddddddddd W l leeeeeee l l F vvl vlt vv v P t nnnnnnnnnnnnnnnnnnnnnnnnn F t s l s l ett et et t vl t A leet l ol ol l let eee ooooooooooooooooooooooooo tol ol rol o hhrhh r hhh W A r r r R h F iiiiiiiiiiiiiiiiiiiiiiiiippppppppppppppppppppppppp h h h h h h h iiggt igt iiiggg r r r r r r r r rr r r r r r r r r r r r r r r r R tigt igt igt l i hhohohhh M TTTTTTTTTTTTTTTTTTTTTTTTT l oh o toh o toho i T toot etr eto trooo t t t t T et et ret e o t
. r r r r r u r urooo r r u r u r u r u iool r r arol iiii ariiir r l iol ioliiiol iar iar ar a PPFPFPPP ddddddddddddddddddddddddd FPFPFPF E nnnnnnnnnnnnnnnnnnnnnnnnn S aaaaaaaaaaaaaaaaaaaaaaaaa . O mmmmmmmmmmmmmmmmmmmmmmmmm eeeeeeeeeeeeeeeeeeeeeeeee L ddddddddddddddddddddddddd lee t ltt C ee lee l
l nnnnnnnnnnnnnnnnnnnnnnnnn lt l t lt C too r r V ooooooooooooooooooooooooo I I too t t too t I hh I F eeeeeeeeeeeeeeeeeeeeeeeee s s s s s s s s s s s s s s s s s ss s s s s s s C r r r r P dddt ddd hh hh t ooooooooooooooooooooooooo I dt t ddt t l l r r r ooreee eee iiit t iii r rM CCCCCCCCCCCCCCCCCCCCCCCCC l l l l l ll l l l l l l ll l ll ll l P l e r oor eer oo R uuu qqqreeqqq uuu l it t ii t t u uu l eee r R qeeqqee r r r r i T r rr luu i r eee l r r l i e r uu ree r lluu s s s iaas s s l l AAAFFAAA T s iiaas s iaa i ddddddddddddddddddddddddd AFFAAFF nnnr nnnnnnnnnnnnnnnnnnnnn N aaaaaaaaaaaaaaaaaaaaaaaaa E mmmmmmmmmmmmmmmmmmmmmmmmm eeeeeeeeeeeeeeeeeeeeee eee dddddddd nnnnnnnn G ddddddddddddddddddddddddd ddddddd 3 aaaaaaaa L mmmmmmmmG eeeeeeee oooonoooooooooooooooooooo nnnn nnnnnnnnnnnnnnnnnnnn s s s s s s < s sss s ss s s s s s s s s s s s nnnnnnn aaaaaaa mmmmmmm N dddddddd S r r r r r r r r r r r r r r r r r r r r r r r r r 3 eeeeeee I nnnnnnnn S I uuuuuuuuuuuuuuuuuuuuuuuuu ccccccc eccccccccccccccccc N ddddddd P oooooooo cccccc: ccccccccccccccccc I nnnnnnn I I s s s s s s s s r r r r r r r r OOOOOO0LOOOOOOOOOOOOOOOOO P I ooooooo s s s s s s s uuuuuuuu cccccccc I r r r r r r r cccccccc uuuuuuu ccccccc OOOOOOOO ccccccc OOOOOOO ddddddddddddddddddddddddd E nnnnnnnnnnnnnnnnnnnnnnnnn aaaaaaaaaaaaaaaaaaaaaaaaa d S O mmmmmmmmmmmmmmmmmmmmmmmmm eeeeeeeeeeeeee eeeeeeeeeee e ddddddddddddddddddddddddd O r L O S iu C nnnnnnnnnnnnnnnnnnnnnnnnn S I q e V ooooooooooooooooooooooooo I G r D eeeeeeeeeeeeeeeeeeeeeeeee s s s s s s s s ss s s s s s s s s s s s s s s s G S s S
. a S oooooooooooooooooooooooool CCCCCCCCCCCCCCCCCCCCCCGlCC WF l l l ll l l l lll l l l l l l l l l l W
F s r AAAAAAA A u c A NNNNNNN AAAAAAcA NNNNNNON E s sss s s s s s s s s s s s ddd S eeeeeeeeeeeeeeeeeeeeee dddd dd O nnnnnnnnnnnnnnnnnnnnnn O leee l l l lee L l li ll i i i li ll i il ill ii ll i il ili li ll i illlii ii i ll T loool O leece l !! ll l U r r r T loor C I I I I i I I222222223333333 l U loooo r r r nnnnnnnnnnnnnnnnnnnnnn A t t t nnndddd A r t t t tnnnnd tnn R r t V oooooooooooooooooooooo oooeeee ccceeee d eooooeoo lslsl s ls l s lsls ls lss ls ls sl sl sl lsls s ll s s s l s W ffff W ecccceccf O iiiiiiiiiiiiiiiiiiidl aaaaaaaaaaaaaaaaaaaaaaAAA i l i F tooor eeee t t r r r F f r oooor oo P A uuuv v v v et t t t et fFFFFFFFFFFFFFFFFFFFFFNNN AAAOOOO A v uuuuv uut OAAAAOAA dddddddddddddddddddddd ddddddd nnnnnnnnnnnnnnnnnnnnnn nnnnnnn dddddddd nnnnnnnn V aaaaaaaaaaaaaaaaaaaaaa aaaaaaa aaaaaaaa L mmmmmmmmmmmmmmmmmmmmmmeee T mmmmmmm eeeeeee T mmmmmmmmV eeeeee ee eeeeeeeeeeeeeeeeeeeeeennn ddddddddddddddddddddddl l l i i i C ddddddd C dddddddd G nnnnnnnnnnnnnnnnnnnnnnI I I A nnnnnnn A E oooooooooooooooooooooo ooooooo snnn ooo W nnnnnnnn W oooooooo R s r rsr s r rs r ss r r sr r s rss r rsr s r rs r sr rs rsr sr sss s F s s s s s s s eeeeeee s s s s s s s s uuuuuuuuuuuuuuuuuuuuuus s t t t t t F eeeeeeee W A taaaaaaat A t t t t aaaaaaaa t t t t F ccccccccccc ccccccccc cc l aaa cccccccccccccccccccccciii lsl uuuuuuu t uuuuuuuu cccct t t tcccc t t t OOOOOOOOOOOOOOOOOOOOOOFFF t t t t t t t ccccccc AAAAAAA AAAAAAAA dddddddoddddddddddddddddd
)
ddddd nnnnnnnnnnnnnnnnnnnnnnnnn d nnnnn S aaaaaaaaaaaaaaaaaaaaaaaaa E e E aaaaa P mmmmmmmmmmmmmmmmmmmmmmmmm S u S O mmmmme eeeee s I R eeeeeeeeeeeeeeeeeeeeeeeee ddddddddddddddddddddddddd O o T L in t L C ddddd l nnnnn c l i nnnnnnnnnnnnnnnnnnnnnnnnn C ooooooooooooooooooooooooo n V oooooo R sruuuuuuuuuuuuuuuuuuuuuuuuu rsr s r rsr s r rsr s r rsr sr s s s s r r r r r r r r r r r r s s s s s s s s s s s s V o I S ss s s s r r r r r s t U ccccccccccccccccccccccccc I S
. C uuuuul M AcccccacccccfA i T ccccccccccccccccccccccccc M AAAAAAA
( OOOOOOOOOOOOOOOOOOOOOOOOO NNNNNNN NOOOOOiN 9. le b O O a 8901 2345 O 678901 2345678901 234567890 678901 2 N 1 1222222 N 2222333333333344444444 %45 N 2222333 T . C
Table 9. (Continued) N_O MSIV CLOSE AFW ACT AFW AUTO AFW SG ISO IIP IN3 TilR IIPI CH TilR AFW 33 NA Actuates on demand Auto controlled NA Occurs on demand As required Prior to high level alarm 34 NA Actuates on demand Auto controlled NA Occurs on demand As required Failure to throttle 35 NA Actuates on demand Auto controlled NA Occurs on demand Failure to throttle Prior to high Icvel alarm 36 NA Actuates on demand Auto controlled NA Occurs on demand Failure to throttle Failure to throttle 37 NA Actuates on demand Overfeed NA Occurs on demand As required Prior to high level alarm 38 NA Actuates on demand Overfeed NA Occurs on demand As required Failure to throttle 39 NA Actuates on demand Overfeed NA Occurs on demand Failure to throttle Prior to high level alarm 40 NA Actuates on demand Overfeed NA Occurs on demand Failure to throttle Failure to throttle 41 NA Actuates on demand Auto controlled NA Occurs on demand As required Prior to high level alarm 42 NA Actuates on demand Auto controlled NA Occurs on demand As required Failure to throttle 43 NA Actuates on demand Auto controlled NA Occurs on demand Failure to throttle Prior to high level alarm 44 NA Actuates on demand Auto controlled NA Occurs on demand Failure to throttle Failure to throttic 45 NA Actuates on demand Overfeed NA Occurs on demand As required Prior to high level alarm 46 NA Actuates on demand Overfeed NA Occurs on demand As required Failure to throttle 47 NA Actuates on demand Overfeed NA Occurs on demand Failure to throttle Prior to high level alarm 48 NA Actuates on demand Auto controlled NA Occurs on demand As required Prior to high level alarm 49 NA Actuates on demand Auto controlled NA Occurs on demand As required Failure to throttle 50 NA Actuates on demand Auto controlled NA Occurs on demand Failure to throttle Prior to high level alarm NO TURIl TRIPS FW REG VLV PORV CLOSE SDV CLOSE SI SGL GEN MFIV CLOSE MFW PMP TP 8 51 Occurs on demand Fails on I line NA Close on demand Occurs on demand Close on demand Trip on demand 52 Occurs on demand Fails on I line NA Close on demand Occurs on demand Close on demand Trip on demand 53 Occurs on demand Fails on I line NA Close on demand Occurs on demand Close on demand Trip on demand 54 Occurs on demand Fails on I line NA Close on demand Occurs on demand Close on demand Trip on demand 55 Occurs on demand Fails on I line NA Close on demand Occurs on demand Close on demand Trip on demand 56 Occurs on demand Fails on I line NA Close on demand Occurs on demand I fails to close Fail to trip 57 Occurs on demand Fails on I line NA I fails to close Occurs on demand Close on demand Trip on demand 58 Occurs on demand Fails on I line NA I fails to close Occurs on demand Close on demand Trip on demand 59 Occurs on demand Fails on I line NA I fails to close Occurs on demand Close on demand Trip on demand 60 Occurs on demand Fails on I line NA I fails to close Occurs on demand Close on demand Trip en demand 61 Occurs on demand Fails on I line NA I fails to close Occurs on demand Close on demand Trip on demand 62 Occurs on demand Fails on I line NA 2 fail to close Occurs on demand Close on demand Trip on demand 63 Occurs on demand Fails on I line NA 2 fail to close Occurs on demand Close on demand Trip on demand 64 Occurs on demand Fails on 1 line NA 2 fail to close Occurs on demand Close on demand Trip on demand 65 Occurs on demand Fails on I line NA 2 fail to close Occurs on demand Close on demand Trip on demand 66 Occurs on demand Fails on 2 lines NA Close on demand Occurs on demand Close on demand Trip on demand 67 Occurs on demand Fails on 2 lines NA Close on demand Occurs on demand Close on demand Trip on demand 68 Occurs on demand Fails on 2 lines NA Close on demand Occurs on demand Close on demand Trip on demand 69 Occurs on demand Fails on 2 lines NA Close on demand Occurs on demand Close on demand Trip on demand 70 Occurs on demand Fails on 2 lines NA Close on demand Occurs on demand Close on demand Trip on demand 71 Occurs on demand Fails on 2 lines NA Close on demand Occurs on demand Close on demand Trip on demand 72 Occurs on demand Fails on 2 lines NA Close on demand Occurs on demand Close on demand Trip on demand 73 Occurs on demand Fails on 2 lines NA Close on demand Occurs on demand Close on demand Trip on demand 74 Occurs on demand Fails on 2 lines NA Close on demand Occurs on demand I fails to close Fail to trip 75 Occurs on demand Fails on 2 lines NA 1 fails to close Occurs on demand Close on demand Trip on demand s . . . . .
Table 9. (Continued) NO MSIV CLOSE AFW ACT AFW AUTO AFW SG ISO HPIN3 TilR 11Pt Cll TilR AFW SI NA Actuates on demand Auto controlled NA Occurs on demand Fadure to throttle Failure to throttle 52 NA Actuates on demand Overfeed NA Occurs on demand As required Prior to high lesel alarm 53 NA Actuates on demand Overfeed NA Occurs on demand As required Fadure to throttle
$4 NA Actuates on demand Oserfeed NA Occurs on demand Failure to throttle Prior to high loci alarm 55 NA Actuates on demand Overfeed NA Occurs on demand Failure to throttle failure to throttle 56 NA Actuates on demand Auto controlled NA Occurs on demand As required Prior to high toel alarm 57 NA Actuates on demand Auto control!cd NA . Occurs on demand As required Prior to high lesel alarm 58 NA Actuates on demand Auto controlled NA Occurs on demand As required Failure to throttle 59 NA Actuates on demand Auto controlled NA Occurs on demand failure to throttle Prior to high level alarm 60 NA Actuates on demand Auto controlled NA Occurs on demand Failure to throttle Failure to throttle 61 NA Actuates on demand Overfeed NA Occurs on demand As required Prior to high level alarm n2 NA Actuates on demand Auto controlled NA Occurs on demand As required Prior to high lesel alarm 63 NA Actuates on demand Auto controlled NA Occurs on demand As required Fadure to throttle 64 NA Actuates on demand Auto controlled NA Occurs on demand Failure to throttle Prior to high lesci alarm 65 NA Actuates on demand Overfeed NA Occurs on demand As required Prior to high le el alarm 66 NA Actuates on demand Auto cor. trolled NA Occurs on demand As required Prior to high loc! alarm 67 NA Actuates on demand Auto controlled NA Occurs on demand As required Iadure to throttle 68 NA Actuates on demand Auto controlled NA Occurs on demand I ailure to throttle Prior to high lesel alarm 69 NA Actuates on demand Auto controlled NA Occurs on demand Failure to throttle Failure to throttle 70 NA Actuates on demand Overfeed NA Occurs on demand As required Prior to high level alarm 71 NA Actuates on demand Oserfeed NA ' Occurs on demand As required Failure to throttle 72 NA Actuates on demand Overfeed NA Occurs on demand Failure to throttle Prior to high IcVel alarm 73 NA Actuates ou demand Overfeed NA Occurs on demand Failure to throttle I ailure to throttle 74 NA Actuates on demand Auto controlled NA Occurs on demand As required Prior to high lesel alarm 75 NA Actuates on demand Auto controlled NA Occurs on demand As required Prior to high lesel alarm NO TURB TRIPS FW REG VLV PORV CLOSE SDV CLOSE St SGL GEN MiIV CLOSE MFW PMP TP 5 Occurs on demand Fads on 2 lines NA I fails to close Occurs on demand Close on demand Trip on den +nd 77 Occurs on demand I ails on 2 lines NA 1 fails to close Occurs on demand Close on demand Trip on demand 78 Occurs on demand f ads on 2 lines NA I fails to close Occurs on demand Close on demand Trip on demand 79 Occurs on demand f ails on 2 lines NA 2 fail to close Occurs on demand Close on demand Trip on demand 80 Occurs on demand Fails on 2 hnes NA 2 fail to close Occurs on demand Close on demand Trip on demand 81 Occurs on demand Iails on 3 hnes NA Close on demand Occurs on demand Close on demand Trip on demand 82 Occurs on demand fails on 3 h*s NA Close on demand Occurs on demand Close on demand Trip on demand 83 Occurs on demand f ails on 3 hnes NA Close on demand occurs on demand Close on demand Trip on demand 84 Occurs on demand fails on 3 lines NA Close on demand Occurs on demand Close on demand Trip on demand 85 Occurs on demand Iads on 3 hnes NA Close on demand Occurs on demand Close on demand Trip on demand 86 Decurs on demand fails on 3 hnes NA i fails to close Occurs on demand Close on demand Trip on demand 87 Occurs on demand Fails on 3 knes NA 2 fail to close Occurs on demand Close on demand Trip on demand NO MSIV ClOSE Al W ACT Al W AUTO AFW SG ISO IIP INJ T11R llPI Cil TilR Al W 76 NA Actuates on demand Auto controlled NA Occurs on demand As required Failure to throttle 77 NA Actuates on demand Auto controlled NA Occurs on demand Failure to throttle Prior to high level alarm 78 NA Actuates on demand Overfeed NA Occurs on demand As required Prior to high lesel alarm 79 NA Actuates on demand Auto controlled NA Occurs on demand As required Prior to high level alarm 80 NA Actuates on demand Overfeed NA Occurs on demand As required Prior to high level alarm 81 NA Actuate, on demand Auto contro!!cd NA Occurs on demand As required Prior to high lesel alarm 82 NA Acturies on demand Auto controlled NA Occurs on demand As required Failure to throttle 83 NA Actuates on demand Auto controlled NA Occurs on demand Failure to throttle Prior to high leve! alarm 84 NA Actuates on demand Auto controlled NA Occurs on demand Failure to throttle Failure to throttle 85 NA Actuates on demand Oserfeed NA Occurs on demand As required Prior to high kves alarm 86 NA Actuates on demand Auto controlled NA Occurs on demand As required Prior to high level alarm 87 NA Actuates on demand Auto controlled NA Occurs on demand As required Prior to high Icvel alarm
Table 10. Sequence descriptions-steam generator tube rupture at hot standby All sequences below are initiated by the rupture of a single tube in Steam Generator A. The rupture occurs at the tubesheet on the cold-leg side with the reactor at hot standby conditions. Sequence 1 (Base case, Scenario 13 in Reference 1) The operator responds as follows:
- 1. If SIAS signalis 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% SG level.
- 3. At 500 s the operator closes the affected steam generator MSIV.
- 4. At 10 minutes the operator fully opens 3 steam dumps and cools down the primary system until 45'F subcooling is attained. Subcooling is measured using the core outlet temperature and saturation temperature in the affected SG 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.
- 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 HPI.
Sequence 2 Same as Sequence 1, except the steam dumps fail to close for 10 min after the subcooling require-ment is met. Sequence 3 Same as Sequence 1, except the pressurizer PORV sticks open for 10 min on first opening. Sequence 4 Same as Sequence 1, except the second pressurizer PORV opening does not occur, and HPI and charging are throttled when pressurizer setpoint level is attained.
~
Sequence 5 Same as Sequence 4, except the HPI and charging are not thre!ed. 12
Table 11. Sequence descriptions-isolatable SBLOCA at power Sequence I below is identical to Scenario 7 as reported in Reference 1. This scenario involves a stuck-open pressurizer PORV at full-power conditions which the operator isolates at 10 minutes into , the transient.* Sequences 2 through 4 below are defined as changes to Sequence 1.
- 1. Scenario 7 as previously defined by ORNL.I
- 2. Isolate at 10 minutes, charging fails to throttle.
- 3. Isolate at 20 minutes.
- 4. Isolate at 20 minutes, charging fails to throttle.
As a result of the analysis presented here, ORNL assumed a later isolation time (1-1/2 hours) than is shown in this table for the fracture mechanics analysis. Table 12. Sequence descriptions-isolatable MBLOCA at power . The sequences below are based on Scenario 6 as presented in Reference 1. ~~. sis scenario involved a 0.0635-m (2.5-in.) diameter hot leg break at full power which was not isolated. Sequences I through 4 below are defined as changes to Scenario 6.*
- 1. Isolate at 10 minutes, charging fails to throttle.
- 2. Isolate at 10 minutes, charging throttled 3 minutes after pressurizer level setpoint reached.
- 3. Isolate at 20 minutes, charging fails to throttle.
- 4. Isolate at 20 minutes, charging throttled 3 minutes after pressurizer level setpoint reached.
As a result of the analysis presented here, ORNL assumed a later isolation time (1-1/2 hours) than is shown in this table for the fracture mechanics analysis. 9 13
Table 13. Steamline break event descriptions Heading Description Entries Description CKVL RP CL Check valve on ruptured steam line Coses on demand Check valve on ruptured steam line closes on demand. closes to prevent backflow from . remaining steam lines. Fails to close Check valve on ruptured steam line fails to prevent backflow from , remaining steam lines. MSIV RP CL. MSIV on ruptured steam line closes on NA Closure not required if check valve demand (automatic closure not demand operates properly. for small break). Coses on demand MSIV on ruptured steam line closes on demand; required if check valve fails to close.- RM MSIV CL MSIVs on unaffected steam lines close Cose on demand Remaining 2 MSIVs close on demand. on demand (automatic closure not demanded for small break). I fails to close MSIV on I line fails to close. 2 fail to close MSIVs on 2 lines fail to close. SDV CLOSE Steam dump valves (SDVs) open on tur- Cose on demand All 5 SDVs close when required. bine trip and close when required. Tur-bine trip assumed to occur as part of I fails to close i SDV fails to close when required, the initiating event. 2 fail to close 2 SDVs fail to close when required. 3 fail to close 3 SDVs fail to close when required. Sequences involving 3 or more SDV failures coupled with I or more FW
- reg valve failures were not considered.
5 fail to close 5 SDVs fail to close when required. Since the failure probabuility for
- failure of 4 SDVs is the same, it was not considered as a separate sequence.
MFW PMP TP MFW pumps trip on demand. Trip on demand Both MFW pumps trip on demand. Fails to trip MFW pumps fail to trip on demand. Failure must be coupled with at least i MFIV failing to close. AFW ACT AFW actuates on demand. Actuates on AFW always actuates on demand and demand delivers flow based on 2 motor-driven AFW pumps and turbine-driven AFW pump where 2/3 low SO level indica-tions are reached. AFW AUTO AFW auto control. Auto controlled AFW auto control operates as required. Overfeed AFW auto control fails resulting in steam generator overfeed at maximum rate. AFW SO ISO ' AFW isolation to affected steam Occurs as required Operator isolates affected steam generator. generator as required. Not isolated Operator fails to isolate affected , steam generator as required. THR AFW AFW throttled. Prior to high level Operator throttles AFW prior to alarm steam generator high level alarm. Failure to throttle Operator fails to throttle AFW. 14
Table 13. (continued) IIcading Description Entries Description . IIP INJ llP! occurs on demand. Occurs on llPI occurs as required, given that 51 demand signal has been generated. TIIR llPI Cil }{PI and charging pumps throttled. As required Operator throttles llP! and charging pumps to prevent repressurization. Failure to throttle Operator fails to throttle liPI and charging pumps. FW REG VLV FW reg. valves runback on turbine trip. Occurs on FW reg. valves runback and modulate Not included for sequences initiated at demand as required on 3 lines. hot standby. Turbine trip assumed to occur as part of the initiating event. Fails on I line FW reg. valve remains open on I line. Sequences involving 1 or more FW reg. valve failures coupled with 3 or more SDV failures were not considered. Fails on 2 lines FW reg. valve remains open on 2 lines. Fails on 3 lines FW reg. valves remains open on 3 lines. 51 SGL GEN SI signal generated when required. Generated on SI signal always generated when demand required, based on system response to initiator. , MFIV CLOSE MFIVs close on demand. Close on demand All close on demand, given that SI signal has been generated. I fails to close 1 MFIV fails to close on demand. Failure must be coupled with failure of MFW pumps to trip. Failure of MFIVs to close not considered when FW reg. valve runback occurs successfully. RM PORY CL PORVs open on turbine trip (>70% Close on demand Remaining 2 PORVs close on demand power) and those on the two unaffected if opened. steam lines close when required. Not included for sequences initiated at hot NA Not required. If I or more FW reg. standby. Turbine trip assumed to occur val es fail to runback, PORVs are not as part of the initiating event. demanded. PORV failures are not significant if one or more SDVs fail to close. Fails on I line PORV on I line fails to slose. Fails on 2 lines PORVs on 2 lines fail to close. G 15
Table 14. Reactor trip and LOCA event descriptions IIcading Description Entries Description TURB TRIPS Turbine trips on demand. Not included Occurs en Turbine always trips on demand, for sequences initiated at hot standby, demand Steam side PORVs open (if trip is . from >70% power) and SDVs open. FW REG VLV FW reg. valves runback on turbine trip. Occurs on FW reg. valves runback and modulate Not included for sequences initiated at demand as required on 3 lines. , hot standby. Fails on I line FW reg. valve remains open on I line. Sequences involving i or more FW ' reg. valve failures coupled with 3 or more SDV failures were not considered. Fails on 2 lines FW reg. valve remains open on 2 lines. Fails on 3 lines FW reg. valve remains open on 3 lines. SDV CLOSE Steam dump valves (SDVs) open on tur. Close on demand All $ SDVs close when required. bine trip and close when required. I fails to close i SDV fails to close when required. 2 fall to close 2 SDVs fail to close when required. 3 fail to close 3 SDVs fail to close when required. Sequences involving 3 or more SDV failures coupled with I or more FW reg. valve failures were not considered. , 5 fail to close 5 SDVs fail to close when required. Since the failure probability for failure of 4 SDVs is the same,it was not considered as a separate sequence. SI SGL GEN Si sigrud generated when required. NA SI signal not expected to be generated. Generated on SI signal always generated when demand required, based on system response to initiator. MSIV CLOSE MSIVs close on demand. NA MSIV closure signal not expected. If signal generated, MSIV closure occurs. Failed open steam PORVs will not lead to demand for MSIVs to close. Occurs on All MSIVs close on demand demand I fails to close 1 MSIV fails to close on demand. If l evaluation shows that MSIV closure is not demanded, this sequence can be eliminated. SDVs will be manually isolated if MSIVs fail to close. AFW ACT AFW actuates on demand. NA AFW not demanded. . Actuates on AFW always actuates on demand and demand delivers minimum required flow. AFW AFW au:0 control. NA AFW auto control not required. Auto controlled AFW auto control operates as required. Overfeed AFW auto control fails, resulting in steam generator overfeed at maximum rate. 16
Table 14. (continued) Heading Description Entries Description THR AFW . AFW throttled. NA Not required. Prior to high level Operator throttles AFW prior to alarm steam generator high level alarm. Failure to throttle Operator fails to throttle AFW. ~* ACCINJ Accumulators discharge when required. Occurs on Accumulators always discharge when Not included for sequences initiated by demand required. reactor trip. RHR ACT RHR injection occurs on demand. Not Occurs on RHR injection occurs as required, included for sequences initiated by reac- demand given that SI signal has been tor trip. generated. PORV CLOSE PORVs open on turbine trip (>70% NA Not required. If I or more FW reg. power) and close when required. Not valves fail to runback, PORVs are not included for sequences initiated at hot demanded. PORV failures are not standby. significant if I or more TBVs fail to close. Close on demand All 3 PORVs close on demand. Fails on I line PORY on I line fails to close. Fails on 2 lines PORVs on 2 lines fail to close. PORVs on 3 lines failing to close is equivalent to 1 SDV failing to close. MFIV CLOSE MFIVs close on demand. NA Remain open; closure not demanded
- Close on demand All close on demand, given that Si signal has been generated.
I fails to close ! MFIV fails to close on demand.
. Failure must be coupled with failure of MFW pumps to trip. Failure of MFIVs to close not considered when FW reg. valve runback occurs successfully.
MFW PMP TP MFW pump trips on demand. NA MFW pumps remain running; trip not demanded. Trip on demand Both MFW pumps trip on demand. Fail to trip MFW pumps fail to trip on demand. Failure must be coupled with at least 1 MFIV failing to cicse. AFW SG ISO AFW isolation to affected steam NA AFW isolation not required unless generator. MSIVs fail to close on demand. Occurs as required Operator isolates affected steam generator as required. HP INJ HPI occurs on demand. NA HPI not demanded. Occurs on HPI occurs as required, given that SI demand signal has been generated. THR HPI CH HPI and charging pumps throttled. NA Not required. Not included for LOCA initiators. As required Operator throttles HPI and charging pumps to prevent repressurization. Failure to throttle Operator fails to throttle HPI and charging pumps. 17
- 3. METHODS This section describes the general methods used a group, an analyst has thus developed a method to determine the reactor vessel pressure and generally applicable to all sequences within that temperature histories for the sequences shown in group. Furthermore, this approach assures that ,
Section 2. Where application of the general sequences with the same controlling phenomena are methods was not feasible or inappropriate, details analyzed in a consistent manner, of alternate methods used appear in Sections 4 through 13. , 3.2 identification of u alent Sequences 3.1 Regrouping of Sequences by Controlling Conditions or The reader may have noticed that not all Phenomena sequences defined by ORNL (Tables I through 12) appear in the regrouped list (Table 15). The reason As described in Section 2, the sequences were for this is that, during the process of regrouping, transmitted by ORNL in 12 groups, with sequences many sequences were identified as being thermal-in each group sharing a common initiating event. hydraulically equivalent to other sequences. In cases To determine the thermal-hydraulic sequence where this occurred, only one of the sequences in responses, it was convenient to regroup the an equivalent set is shown in Table 15. For com-sequences according to controlling thermal- pleteness, however, plotted results in Appendix A hydraulic phenomena rather than by initiating are given for every sequence in Tables I through 12. event. For example, all sequences involving only a Plotted results for all sequences in an equivalent set secondary-side break affecting one steam generator were identical. Table 16 lists the equivalent were regrouped together into Group A (note: sequences and the reasons for equivalency. ORNL groupings are designated by a number while . the INEL regroupings are designated by a letter). 3.3 Application of Models All Group A sequences share the controlling phenomena of heat removal to a single, affected , steam generator. Sequences within Group A differ Sequences begin from the steady plant conditions in break size, power level, and minor complicating associated with either full-power or hot standby failures which do not change the controlling operation. When the initiating event occurs, the phenomena. Group A thus contains all sequences plant experiences a transient defined by: (1) the in Groups 6 and 8 and part of the sequences in initiating event, (2) operator or hardware failures Groups 5, 7, and 9. specified in the sequence description, ind (3) auto-matic and operator plant actions encorntered as a Table 15 presents the regrouping of sequences by result of changes in conditions due to (1) and (2). controlling thermal-hydraulic phenomena. Such transients generally include an early phase, Groups A through C include sequences with only during which a complicated scenario of operator secondary-side breaks; Groups D and E include and automatic actions occur, and a late phase, sequences with no primary- or secondary-side during which such actions have ceased and the breaks, Groups F and I sequences with only thermal-hydraulic plant conditions are determined primary-side breaks, Groups G and H sequences by relatively simple thermal-hydraulic processes. with combinations of primary- and secondary-side breaks; and Group J includes steam generator tube During an event with one or more steam dump rupture sequences. valves , ailing open, for example, the initiating event - would be expected to cause a rapid succession of The purpose of regrouping as just described is events such as: reactor and turbine trips, safety to organize sequences in such a way that all injection and auxiliary feedwater initiation, main . sequences within a group share common control- feedtrain isolation, letdown isolation, termination ling phenomena. By deseloping and qualifying a of pressurizer heater power, an increase in makeup specific method to determine the pressure and flow, and reactor coolant pump trip. After these temperature histories for any single sequence within events, how ever, the plant conditions are controlled 18
I i I Table 15. Regrouping of sequences by controlling conditions or phenomena Group Controlling Conditions or Phenomena Sequences A Secondary-side break, I affected steam 5-1, generator 6-1 through 6-9, 7-1 through 7-8, 8-1 through 8-6, 9-25 through 9-32. B Secondary-side break,3 symmetrically 7-12, affected steam generators 9-2 through 9-23, 9-41 through 9-47. C Secondary-side breaks with 2 affected 5-14, 5-15, steam generators or with 5-17 through 5-20, 3 symmetrically affected steam 7-9 through 7-11, generators 9-33 through 9-40. D Reactor trip from full power, no 9-1, primary- or secondary-side breaks 9-49 through 9-55. E Main feedwater overfill 9-56. F Primary-side breaks 1-1 through 1-4, 2-1 through 2-4, 3-1, 3-2, and 4-1. G Primary-side breaks combined with 1-5 through 1-8, symmetric secondary-side breaks 2-5 through 2-8, 3-3. H Primary-side breaks combined with I-9 through 1-12, asymmetric secondary-side breaks 2-9 through 2-11. I Isolatable primary-side breaks 11-1 through 11-4, 12-1 through 12-4. J Steam generator tube ruptures 10-1 through 10-5. t 1 l e 19
Table 16. Thermal-hydraulically equivalent sequences Sequence (s) Equivalent Sequence Reason (s)a 1-13, 1-17, 1 18 1-1 B . 1-14 1-2 B l-15 1-3 B 1-16 1-5 B - 5-2 through 5-13 5-1 D 5-16 5-14 F 5-21 through 5-24 5-1 D 9-24 9-19 E 9-48, 9-66, and 9-81 9-1 B 9-57, 9-75, and 9-86 9-2 A,B 9-5R and 9-76 9-3 A,B 9-Sy and 9-77 9-4 A,B 9-60 9-5 A,B 9-61 and 9-78 9-6 A,B 9-62, 9-79, and 9-87 9-9 A,B 9-63 9-10 A,B 9-64 9-11 A,B 9-65 and 9-80 9-13 A,B 9-67 and 9-82 9-49 B 9-68 and 9-83 9-50 B 9-69 and 9-84 9-51 B 9-70 and 9-85 9-52 B
.9-71 9-53 B 9-72 9-54 B 9-73 9-55 B 9-74 9-56 C
- a. A-Same break size and location.
B-Feedwater regulating valve failure is inconsequential following a reactor trip, unless accom-panied by a feedwater isolation valve failure on the same feedline. C-Since only one feedwater isolation valve fails open, the failure of I or 2 feedwater regulating valves is equivalent. D-Sequences initiated by a stuck-open steam line PORV at full power do not result in reactor or turbine trips. For further discussion see sequence 5-1 results in Section 4.1. , E-Main steam isolation valve (MSIV) closure signal is not generated because coincident high steam line flow (in 2 out of 3 lines) and low average temperature were not encountered. "MSIV Fails To Close On Demand" is therefore inconsequential. Since the MSIV did not fail to close, , the "SG ISO" is changed from " Occurs As Required" to "NA." For further discussion see sequence 9-19 results in Section 5.3. F " Failure to Throttle" auxiliary feedwater (AFW) is inconsequential because steam generator narrow range levels had not recovered to 40% before AFW was isolated at 10 minutes. 20 a
by the relatively stable processes of core heat Calculations with the simplified model were per-addition, natural circulation loop flow, and heat formed using an updated version of the removal to the generators. Depending on the sever- RELAPS/ MODI.6, Cycle 18 computer code. ity of the initiating event and subsequent failures, Detailed model calculations were performed using this later stable phase typically begins at from 5 to this version, as well as RELAPS/ MOD 2.
- 20 minutes after the initiating event.
3.3.1 Description of Detailed Model. This sec-Note that the term " stable phase" does not mean tion describes the base detailed RELAPS HBR-2 i
. " steady phase." In the above example, factors con- PWR model. Subsections following describe the tributing to nonsteady behavior during the stable thermal-hydraulic and control system components phase include: continually decreasing core decay of the model.
heat, secondary system pressure, and stored energy in metal components. The detailed model was quality-assured in four ways. First, the development of each model com-A limited number of sequences were previously ponent was documented on worksheets which investigated, using a detailed RELAPS thermal- include references to the plant documents sup-hydraulic and control system model of the HBR-2 porting the development. Second, the worksheets pressurized water reactor (PWR).I Investigating all were independently reviewed by an analyst other sequences in Table 15 over the two-hour periods than the one who developed them. Third, utility following initiating events was not economically analysts, already familiar with design and model-feasible using this method. The method generally ing of the plant, reviewed both the model at various used for this purpose was to apply the detailed stages of completion and the calculational results. RELAPS model during the early stage and a Fourth, the simulation of an actual plant transient sim'plified RELAPS model during the later stage of was performed, and the completed model and a sequence. Using this approach, the detailed model results were compared with measured plant data. calculation (1) defined plant response during the The comparison appears in Section 3 of
. complicated early phase of a transient,(2) allowed Reference 1.
confirmation that a stable stage had been reached, (3) defined the starting conditions for a simplified Thermer.nydraulic Model. The detailed RELAPS model calculation over the stable stage, (4) provided model is a representation of the HBR-2 plant, an understanding of the stable stage controlling describing all the major flow paths for both primary phenomena that was required to assemble a valid and secondary systems, including the main steam simplified model, and (5) provided results against and feedwater systems. Also modeled are primary which the results from the simplified model could and secondary power-operated relief valves be compared and qualified. Once qualified, a (PORVs), and safety valves. The emergency core simplified model calculation was used to determine cooling system (ECCS) is included in modeling the the thermal-hydraulic response over the later stage primary-side, and the auxiliary feedwater system is i of the sequence. For analyses presented in this included in the secondary-side modeling. The model report, the simplified models typically ran faster contains 224 volumes,242 junctiom and 218 heat than the detailed models by a factor of about 200. structures. Descriptions of the primary and the Comparisons between detailed and simplified model secondary systems are presented in the following. calculation results show agreement adequate to paragraphs. justify applying the simplified model over the later stages of the sequences. As will be shown in Each loop of the HBR-2 plant's three primary Section 4, agreement ranged from fair for calcula- coolant loops is represented explicitly in the detailed tions of primary-side breaks to excellent for RELAPS model. The loops are designated as
, secorJary-side breaks. The stable portions of the A, B, and C. Each modeled loop contains a hot leg, sequences were therefore controlled by quasi-steady U-tube steam generator, pump suction leg, pump, simple mass and energy balances, which could be and cold leg as shown in Figure 1. The pressurizer well-represented with the simplified model. is attached to the C loop, and the pressurizer spray lines are attached to the B and C loop cold legs. A The following subsections document the basic low-pressure injection (LPI) port, an accumulator detailed and simplified models and variations of the with its associated piping and a high-pressure simplified models used in these analyses. injection (HPI) port are attached to each cold leg.
1 21
l l 344 PORV
= = 345 346 Safety
= = l347l Steam 340 339 generator l ,
# ' ) 341 5 4 u n Pressurizer
,-m /
n 338 33S
-= ;
/. - -
/-- 408 337 335
/ 7
~
l2
/ From Loop B f
/ 8 1 a
'///// // /
/
1 T fff1 vsss
.I2 3 2 343 410 406
, 404 405 3 2 1 _
ZZ r / / s s / / / / / s / Reactor vessel _ D ///i 414 + 7
/ 1
[ 416
~~~
418 1 2
,-- ~ RC pump.
420
/ - 2 /
913 /'
/-a / / // ///////////////
4 l 953 l HPl Accumulator 412 923 I 933 LPI I I I I L i Makeup L_971____ (Loop B only) INEL 4 5114 Figure 1. Detailed model nodalization of primary coolant loops (Loop C shown). The LPI and HPI models are set up to inject one- piping and steam generator tubes. Heat structures third of the total HPI and LPI flow into each loop. are also used to represent the pressurizer propor-Also attached to the Loop B cold leg is the chemical tional and backup heaters.
- and volume control system (CVCS). Makeup and letdown are modeled with a single junction. Heat Figure 2 shows the RELAPS nodalization used structures are connected to each volume in the to represent the HBR-2 vessel. Represented in the primary loops to represent the metal masses of the model are the downcomer, downcomer bypass, 22
////// ead 126
, h, 1 Y
/ 3 / 1
/ Upper
, 100 / / $ 2 / plenum 2
/ / "
/ 122
/ 1 / 3 /
From cold legs i . [ g r To hot U l I I legs
// 102
/ 120
/ 1 It Il
/,
104
/
118 / I 'll i f I
/ / /
1 / 1 6
/ l /
, 2 l 2
/ 5 3
/ Bypass 3 / 4 Downcomer 106 / 116 p / 114 l 5 / S 2 l l /
/
6 / 6 / 1
/
y U H
/ 7 / 112 ,
3
, l i Lower pienem 110 INEL 4 5111 Figure 2. Detailed model nodalization of reactor vessel.
lower plenum, core, upper plenum and upper head. There are 130 volumes associated with the The following leakage paths are represented in the primary loops and 33 volumes associated with the vessel model: downcomer to upper head, down- vessel. l comer to downcomer bypass, downcomer bypass
- to lower plenum, cold leg inict annulus to upper The detailed RELAPS IIBR-2 PWR secondary
( plenum, and upper plenum to the upper head via system model is shown in Figures 3 and 4. The the guide tube. Ileat structures represent the core steam generator secondary model, shown in rods as well as the external and internal metal mass Figure 3, represents the major flow paths in the j of the vessel. Decay heat is assumed to be at the secondary and includes the downcomer, boiler region, separator and dryer region, and the steam 1979 ANS standard rate. 23
To main steam line
//
Steam , dome 282
/ \
e I
/ 278
/
7 7 254 / / 274 From feedwater PIPO 270 258 v 1/ 1
/ 1 / 4
/ % 4
/
Downcomer / / N Boller 262
/- /
/ I
-J
\ , 266
/ l---( /
! /
/ 2 3 / l I 3 ,
_ / l__ /
/_f l__
/ / ! l 3 / 2, l /
7 l l 2 .
/- -
l - --- 4 / 1 / l l l 1
' / '
/ i Ii d 2 . .J L_. 2
~
Outlet l l l inlet plenum plenum l / i I / I
. \ l l /
To cold leg \ ' From hot leg l
\ s L-Loop Component Numbers A 2xx B 3xx C 4xx INEL 4 5112 -
l l Figure 3. Detailed model nodalization steam generator (SGA shown). 24 I
dome. Due to modeling constraints, the steam A total of 14 volumes represent the secondary for generator secondary separators and dryers were each steam generator. The steam line consists of modeled within a single calculational volume. 16 volumes, and the feedwater system contains , Separation in the model thus takes place at a single 31 volumes. elevation rather than at two discrete elevations, as , in the prototype steam generator. The effect of this contm/ system Models. The purpose of this section difference is a perturbation of the flow field at the is to provide the reader with a general overview of upper steam generator level tap, which affects the the functions of the major control systems used in indicated level in a minor way. A further discus- the detailed mcdel. Information regarding the set-sion of this effect appears in Section 3 of points of the Westinghouse control systems will not Reference 1. be provided due to the proprietary nature of the various control system specifications. In general, the The major flow paths of the steam line out to the control systems were modeled as clasely as possi-turbine governor valves were modeled, and are ble and are co isidered to be good representations shown in Figure 4. Each line from the steam of the actual systems. generator secondary out to the common steam header was modeled individually and included a The steam dump control system will be described, main steam isolation valve (MSIV), PORVs, a followed by descriptions of the steam generator check valve, and safety valves. The flow restrictor level control system, the pressurizer pressure con-was modeled in combination with the flow nozzle trol system, the pressurizer level control system, and at the top of the steam dome. From the header to additional systems. the turbine governor valves, the two steam lines in the plant were represented as one line in the model. st m cump contmisystem. The purpose of the The steam dump valve banks were modeled as one steam dump control system (SDCS) is to: valve, with appropriate controllogic to simulate the opening of each valve in the banks. 1. Permit the nuclear plant to accept sudden losses of load without tripping the reactor . The major flow paths of the feedwater system were modeled and are shown in Figure 4. The feed- 2. Remove stored energy and residual heat water system consists of the condensate system, following a reactor trip and bring the plant main feedwater system, and the auxiliary feedwater to equilibrium no-load conditions without system. The components included in modeling the actuation of the steam generator safety condensate system were both condensate pumps, valves low-pressure feedwater heaters, low pressure heater bypass, heater drain system, and the main feedwater 3. Permit control of the steam generator pump suction header. The condensers were modeled pressure at no-load conditions and permit using a constant-pressure boundary condition. The a manually controlled cooldown of the components included in modeling the main feed- plant. water system were both main feedwater pumps, main feedwater pump recirculation, high pressure The above tasks are accomplished by three modes feedwater heaters, main feedwater header tank, of steam dump control. Requirements I and 2 are main feedwater/ bypass valves, and piping to the met isy control of the primary system averate fluid steam generators, including the feedwater header temperature, whereas requirement 3 is met by con-ring. The auxiliary feedwater system modeling trolling the secondary system steam pressure. The includes the motor-driven and turbine-driven SDCS is divided into three separate systems: Load systems, with a common header for each system and Rejections Controller, Plant Trip Controller, and valves from the header to each feed line. Steam Pressure Controller which will be described next. Ileat structures for the secondary system include the internal and external metal for each of the steam The load rejections controller (LRC)is designed generator secondaries, and the piping for both the to control the primary system average temperature steam and feedwater systems. during periods of load rejection. Control of the 25
655 755 MSIV 555 Steam MSIVx \MSIV 565 dump
' 550 N Nq \ 9 560 W 808 Condenser l 585 l -4 650 \ ; T 660 h h ader 665 Steam l 602
+: '
' 810 l y 685 l C750 760 Tu'rbine f 1 64 PORV 680 stop TuMne l 575 l ([' Safety (8670 l 785 1760% 770' >d 775 l 804
( Steam Steam Steam
. Condensate Condenser generator generator l 675 l generator PORV pump 824 A and B I A 0 C Safety 525 2 1 2 1 P1 822 l LP. heaters Heater drain system
~~
720 __ 1 and 2 3 844 _
! ! Motor driven LP. heater bypass 852 aux feed 834 520 620 line C I
y Q Q Q f740M t 715 1 LP. heaters 3,4 and 5 540 ,- 640 2 l 3 l 4 H { l h 5 ,
. 615 1 Main Motor 4 riven Motor 4 riven aux feed line B 710 feedwater aux feed line A 863Y, 862 header MFW pump 854 f
( 705 ; 878 f A 861 510 *** *" " *** 610 aux feed line C l 874 1- MM pmp B Steam -driven 864 aux feed line B _ I $ MFW 605 866y' -- h 530 630 -*< pump l870 l 865 Steam-driven 505 bypass aux feed line A MFW/ bypass valve T INEL 4 5113 Figure 4. Detailed model nodalization of feedwater and steam systems.
l 1 primary system average temperature is performed steam Generator Level control system. The steam by modulating the steam dump valves and, if the generator level control system (SGLCS) is designed load rejection is greater than 70%, the steam line to regulate the liquid level in the steam generator power-operated relief valves (PORVs). The turbine (SG) downcomer. This control system uses three impulse stage pressure signal is linearly converted input signals to regulate the feedwater How rate into into the primary system average temperatur set- each of the three steam generators. These three point. The filtered derivative of the turbine impulse signals are: (1) the steam generator liquid le,el, stage pressure signal is used to determine whether (2) the steam flow rate, measured in the steam line or not a load rejection has occurred, and the size at the SG outlet, and (3) the feedwater flow rate, of the rejection when one does occur. htodulation measured downstream of the feedwater regulating of the steam dump valves is blocked if the condenser valve. The SGLCS is used only when the plant load does not have sufficient vacuum, or if the primary is above 15%. system average temperature decreases below the minimum temperature setpoint. Other bistables The steam generator liquid level is determined by exist in the real plant but were not modeled because measuring the differential pressure between pressure they are not used in the various calculations taps in the SG downcomer. The liquid level is presented here. inferred from this differential pressure, and can be perturbed by events that influence these two taps The function of the plant trip controller (PTC) in a nonsynchronous manner; for example, a main is to bring the primry system average temperature steam line break or turbine stop valve closure. The down to the equilibrium no-load setpoint,559 K steam generator liquid level signal is compared to (547*F) for the 2300 htW full-power case, after the the setpoint level, which is a function of the turbine turbine has been tripped. The PTC performs this impulse stage pressure. The resulting error is then function by modulating the steam dump valves. used as an input signal to a proportional-integral Unlike the LRC, the PTC does not have any con- (P-1) controller. trol over the steam line relief valves. hiodulation of the steam dump valves is blocked if the condenser To determine the feed-steam mismatch signal the does not have sufficient vacuum, or if the primary feedwater and steam now rate signals are compared. system average temperature decreases below the This signal is added to the level error signal, and minimum temperature setpoint. Steam dump con- the result is used as the input signal for another P-1 trol system operation using the PTC is replaced with controller. The output of this P-1 controller is used the steam pressure controller when the primary to modulate the appropriate feedwater valve. system average temperature is decreased to the no-load setpoint, with the additional constraint that When the plant load is less than 15%, instead of 60 s must have expired since the plant was tripped. using the SGLCS, the main feedwater valves are The 60 s delay is used to simulate the reactor closed and feedw ater control is performed operator's response time. manually, using the feedwater bypass valves to maintain tae desired steam generator level. The steam pressure controller (SPC) is used to Additionally, one main feedwater pump and one regulate the secondary system steam header condensate pump are used,instead of two of each, pressure. This system is used when the plant is at asi- > f ull-power case. no-load conditions, or to replace the PTC. The steam header pressure is controlled by modeling the - onditions that can result in main feedwater steam dump valves. The setpoint pressure i 1 in the plant have been incorporated into 7.03 N1Pa (1020 psia). No modulation of the steam the wLCS model. These conditions include plant line PORVs is performed by this system. hfodula- trip, main feedwater pump trip, and initiation of tion of the steam dump valves is blocked if the the engineered safety features actuation signal condenser does not have sufficient vacuum, or if (ESFAS). the priraary system average temperature decreases below the minimum temperature setpoint; howeser, Pressuriier Pressure controlsystem. The purpose of unlike the LRC and PTC systems, the minimum the pressurizer pressure control system (PPCS) is temperature condition may be overridden by the to maintain the desired primary system pressure. plant operator to enable plant cooldown to cold This function is performed using spray valves, relief shutdown conditions. valves, proportional heaters, and backup heaters. 27
The pressurizer pressure is compared to its set- ing pump speed to effect the desired change in the point to determine the error. This error signal is primary system coolant inventory. used as the input signal to a P-I controller. The output of the P-I signal is used to control the The level error signal is also used to actuate the function of both spray valves, the proportional and backup heaters when the pressurizer level error backup heater source demands, and the valve area exceeds the setpoint level by 5%. Presserizer heater . of one of the two pressurizer PORVs. The other demand is blocked when the pressurizer level PORV area is a function of the uncompensated becomes less than the low-level limit of 14%. pressurizer pressure signal. The PLCS is modeled to include both the reac-The PPCS is modeled as accurately as possible tor coolant pump seal injection contribution and and includes all the trips and setpoints in the actual the charging flow demanded by the compensated ( plant, with two exceptions. First, the spray valves pressurizer level error signal. ' do not maintain a minimum flow as in the plant because of difficulties incurred due to thermal- Additione/contro/ systems. Included in the control hydraulic considerations. The minimum flow system package are miscellaneous controllers and requirement is imposed in the plant to maintain the trips that are modeled to represent various system spray line temperature at the temperature of the functions that cannot be classified in any of the primary system cold legs. This is required to avoid aforementioned systems. These controllers perform the possibility of thermally stressing the spray lines functions such as: (1) feedwater recirculation to the when pressurizer spray is demanded. To compen- condenser during periods of low feedwater demand, sate, the model spray lines were initialized at cold (2) low pressure feedwater heater bypass due to low leg temperatures, and no heat losses from the lines main feedwater pump suction pressure, (3) specifi-to containment were considered. The second model- cation of turbine impulse stage pressure as a func-ing exception is in the amount of power supplied tion of steam flow rate and turbine governor valve to the proportional heaters during steady state area, and (4) control of the auxiliary feedwater operation. The heaters normally operate at systems.
~
2000 kW to make up for plant heat losses and pressure decay due to the continuous minimum 3.3.2 Description of Base Simplified Model. This spray operation. Since the pressurizer tank walls section describes the base simplified RELAPS HBR-2 were modeled as perfectly insulated heat structures, PWR model. He simplified model was developed - and the spray valves were completely isolated dur- primarily by combining calculational cells of the ! ing the steady state initialization phase, this detailed model. As described in the last section, the j 2000 kW heater source was subtracted from the detailed model was quality-assured in many ways. total possible proportional heater source of his philosophy was extended to the simplified model 4000 kW. development; calculations supporting the combining of cells were independently checked by an analyst Pressuriierleve/Contro/ System. The purpose of the other than the one performing the calculations. pressurizer level control system (PLCS) is to main-tain the desired amount of liquid inventory in the The simplified model was developed to address primary coolant system. The amount of water thermal-hydraulic plant phenomena during the later inventory in the primary coolant system may be stages of sequences. The base model described here inferred from the liquid level in the pressurizer, was specifically designed to address plant conditions which varies as a function of the primary system where: (1) reactor and turbine trips have occurred, average coolant temperature. (2) safety injection and auxihary feedwater flows have been initiated, and (3) reactor coolant pumps have The pressurizer setpoint levelis a function of the been tripped. Furthermore, the model assumes that
~
primary system average coolant temperature. The significant transient effects of the above actions have setpoint level is subtracted from the actual level ceased and symmetric loop natural circulation which is determined from a set of differential continues. pressure taps located in the pressurizer. The - pressurizer level error signal is used as the input The base model described here was generally signalin a P-I controller. The output of the P-I con- applied, as is, to the later stages of sequences with troller specifies the amount of change in the charg- stuck-open steam dump valves (most sequences in 28
Group B) or small primary breaks (part of Group F). For sequences invoMng only a small primary break Variations on the base model, tequired to properly (defined in Section 2 as a single, stuck-open address sequences controlled by different phenomena, pressurizer PORV), the controlling phenomena dur-are discussed in the next subsection. ing the later portions of the sequences are core heat addition, virtually symmetric loop natural circulation,
, For sequences involving stuck-open steam dump heat removal to steam generators, and energy and valves, the later portions of the sequences are con- mass removal at the PORV. Minor loop asymmetries trolled by decay heat addition in the core, symmetric due to the break location are ignored.
loop natural circulation, and symmetric heat removal through all steam generators. The symmetry exists A nodalization diagram, showTiin Figure 5, of the because the steam dump valves are located on the por- base simplified model was developed to address the tion of the steam line common to all three steam primary and secondary system mass and energy generators, balances that control the phenomena described in the 344 1 345 . PORV l 340
-__________ {
126 l Steam 341 Upper ! break 810 Pressurizer
~---------
head heaters _____,___," iL 808 TDJ O g V V s Odecay 266 100 heat SG secondaries RCS d n i l n ji 7 TDJ TDJ U-tubes [ TDJ TDJ Core and passive 534 538 961 972 heat structures 536 540 951 971 911
- MDAFW TDAFW HPl/LPI Charging Accum INEL 4 5110 Figure 5. 11ase simplified model nodalization.
29 ~ - -. - - - .
previous two paragraphs. Component 100 represents cumvent these problems, constant heat transfer coef-all fluid in the reactor coolant system (RCS), except ficients were specified at both locations based on for that in the pressurizer (Components 340 and 341) representative coefficients calculated with the detailed and that above the hot leg centerline within the reac- model for similar conditions. In the same manner, tor vessel (Component 126). Component 266 a constant heat transfer coefficient was specified for represents the fluid volume of the three steam the heat structure representing the core and other . generator secondaries. Auxiliary to the primary metat system are time-dependent junction Components %1, representing HPI and LPI flow as a function of As was discussed earlier, initial conditions for the , primary system pressure, and 972, representing charg- simplified model were derived from conditions ing flow as a function of pressurizer level. The three calculated using the detailed model. As a result of accumulators are modeled by Component 911, and qualifying the simplified model against the detailed the pressurizer PORV by valve Component 344. Aux. model (results of these qualifications appear in iliary to the secondary system are time dependent Section 3.4), it was found that the best agreement was junction Components $34 and 538 representing obtained if the RCS cell in the simplified model motor- and turbine-driven auxiliary feedwater injec- (Component 100) was initialized at a temperature tion and time dependent junction Component 808 consistent with the third U-tube cell of the detailed representing a secondary system steam break. The model (Cell 3 of Component 408 in Figure 1). The three heat structures shown represent the pressurizer temperature of this U-tube cell generally differs from heaters, the U-tubes of the three steam generators, the reactor vessel dowrcomer fluid temperature only and the core and passive heat structures. This latter slightly [ typically 2 K (3.6*F)]. Thus the simplified heat structure represents the metal of the core, reac- model RCS temperature generally was used directly tor vessel, loop piping, and the steam generator shells. as an indication of downcomer temperature. As will Heat input is based on the ANS standard for decay be discussed in Section 3.3.3 for the primary break heat and the time since reactor trip. sequences, natural loop circulation is slower and HPI flow higher as compared with the base model assump-Auxiliary feedwater logic requires the calculation tions. In these cases, an alternate method (also , of steam generator narrow-range levels. Due to the discussed in Section 3.3.3) was used to calculate the simplicity of the model, this was not possible. Instead, downcomer temperature. results of detailed model calculations were resiewed during periods of similar plant behavior to determine 3.3.3 Simplified Model Variations. Variations to the secondary masses corresponding to key, narrow- the base simplified model(from Section 3.3.2) were range level setpoints. The auxiliary feedwater control required to address various sequence-controlling was then based on the mass setpoints rather than on phenomena; this section documents those variations. level setpoints. Variation 1. Small Steam Break Affecting One Steam The steam break (Component 808) was represented Generator, nasetor cooient Ptimps Tripped. For sequences with a time-dependent junction that calculated break involving a small break that affects only one steam steam flow based on the secondary system pressure, generator, the base simplified model from the break area, and homogeneous equilibrium model Section 3.3.2 was first modified by reducing the size l critical flow tables.3The steam generator secondaries of the U-tube heat structure and the secondary l were assumed to be at saturation conditions with in- volume to represent a single, affected steam generator, flow of subcooled auxiliary feedwater and outflow Next, the unaffected steam generator (USG) metal of saturated steam. masses (but not an equivalent for the liquid masses)
- were added to the primary sptem heat structure. This Due to the simplicity of the model,it was necessary was a modeling compromise to account for a to specify heat transfer coefficients on both sides of slowdown in unaffected loop cliculation flow over .
the U-tube heat structure. On the inside surface, the the period when the simplified model is applied. As fluid in Component 100 of the simplified model is the unaffected loop flow slows, the USGs become not flowing, while the fluid inside the U-tubes is flow- more loosely coupled to the primary system. This
- l ing due to natural loop circulation. On the outside simplified model was benchmarked, for Sequence 7-4, surface, the void fraction in Component 266 does not against the detailed model; results of the benchmark well-represent that in the lower boiler section. To cir- comparison appear in Section 3.4.1.
30
Variation 2 Small Steem Broek Affecting One Steem stagnant unaffected loop liquid and metal masses. Generator, Reactor Coodent Ptimps Operating. For se- Finally, the steam break critical flow model is not quences involving a small steam line break affecting applicable because of the very low ASG secondary a single steam generator where the reactor coolant pressure; it was replaced by a friction-dominated pumps remain operating, the base simplified model steam break model. Friction flow was implemented from Section 3.3.2 was first modified by reducing the by determining an effective loss coefficient from sizes of the U-tube heat structure and secondary detailed model calculation results under similar flow volume to represent a single affected steam generator. conditions. This simplified model was bench-Unlike Variation 1, in which the RCPs are tripped, marked, for Sequence 8-4, against the detailed the unaffected steam generators (USGs) remain close- model. Results of the benchmark comparison ap-ly coupled to the primary system. To account for this pear in Section 3.4.3. phenomenon, both the USG metal and liquid mass equivalents were added to the primary system heat vanerion su Smea Stenm une sisek Symmetnce#y Affbet-structure. In a .d tion, core decay heat was increased ing Thine Steam Generators, Reactor Coodent Ptimps to compensate for the pump power due to continuous operating. For sequences involving a small steam line RCP operation, and U-tube and primary system heat break with three symmetrically affected steam structure inside surface heat transfer coefficients were generators, the base simplified model from increased (based on results using the detailed model) Section 3.3.2 was modified by: (1) increasing the core to account for forced primary system flow. This decay heat to simulate pump heat addition, and simplified model was benchmarked, for (2) increasing the inside surface heat transfer coeffi-sequence 9-25, against the detailed model; results of cients for the U-tube and primary system heat struc-the benchmark comparison appear in Section 3.4.2. tures, to account for forced RCS fluid circulation. Vanietion 2 Steam Break Affecting One Steem Generator; Variation 6, Steam Breaks Affecting Two Steam Primary System Host Removal Controlled by Unoffected Generators. For sequences invohing steam breaks af-Steem Generefors. For sequences invohing a steam fecting only two steam generators, the base simplified break affecting a single steam generator but in which model was modified as follows. The steam generator the unaffected steam generator heat removal is domi. secondary volume and the heat structure representing nant, the following variation was made to the base the U-tubes were reduced in size to correspond to two simplified model described in Section 3.3.2. The steam generators. Based on experience from the secondary volume and U-tube heat structure were benchmarking of simplified model Variation I reduced in size to represent the two unaffected steam (presented in Section 3.4.1), the unaffected loop generators. This variation was typically applied over steam generator liquid and metal masses were periods of a transient when, due to ASG AFW ter_ removed from the model. mination, the heat removal to the ASG was negligi-ble and heat removal to the USGs was controlling. V8. son 7. SsL0cA Sequences. The base simplified This could involve either: (1) unaffected loop natural model was modified to represent the local effect of circulation (due to very low USG temperatures), or ECC on dovncomer temperature for all the SBLOCA (2) continued RCP operation, in which case the sequences (see Tables 1, 3, 4, and .11). The average changes regarding RCP heat input and U-tube heat RCS temperaturt was modified to compensate for transfer coefficients, discussed under Variation 2, both the mixing of ECC and loop flocs and for the apply as well. fluid transit time between the steam generator and the downcomer. Specifically, the downcomer Variation 4. Large Steem une Break Affecting One Steam # " D, was & d as Generator. For a large steam line break affecting only one steam generator, the unaffected loops stagnate L p Tm T ECC ECC,y dT T =
, completely early in the sequence. Due to the size of D m +m ECC the break, the affected steam generator secondary pressure is very near atmospheric pressure. T where T and TECC were the average RCS and ECC account for the above distinctions, the base simplified , temperatures, respectively; mtoo model from Section 3.3.2 was first modified by reduc- due to natural circulation in all t$was the to ree loops; mECC ing the sizes of the U-tube heat structure and was the sum of the HPI, LPI, accumulator secondary volume to represent a single affected steam and charging flows; d. Twas the time derivative of the generator. Next the RCS volume and its heat struc- dt ture were reduced in size to remove the effects of the average RCS temperature; and Y was the time 31
required for the Guid to travel from the steam the U-tube heat structure and secondary volume to generator outlet to the downccmer. The loop flow, represent two affected steam generators. Second, mLoop, was estimated for each sequence, based on Component 100 was modified to represent both the the results of a representative detailed model reactor vessel below the hot leg centerline and two calculation. The transit time, Y, was treated as a affected loops. Third, the primary system heat constant (80s) based on results frcm detailed model structure of the base model was modified by sub- - calculations. The local effect of ECC on down- tracting the metal mass of the unaffected loop. The comer temperature had to be estimated in the metal in the unaffected loop would not cool as simplified model because results from the detailed rapialy as the metal in the affected loops because . model showed that the downcomer temperature the Dow in the unaffected loop stagnates. An could be more than $6 K (100'F) colder than the additional control volume was added to the model average RCS temperature. The LOCA sequences to represent the unaffected primary loop. The resulted in low RCS pressures and high ECC flow additional control volume was necessary because the rates that were significant when compared to the unaffected loop responds differently than the rest loop Dow due to natural circulation, and thus it was of the RCS during sequences with asymmetric steam necessary to represent the effect of ECC mixing on breaks. The additional control volume accounted downcomer temperature. for Dow stagnation in the unaffected loop, which results in flashing that can affect the pressure Variation 8, MBLOCA Sequences. The Variation 7 response of the RCS. model was modified to represent MBLOCA se-quences. A single junction, representing the break, "# ' " ##' #8100## ** U"' 8'"**~0 #*" 8'*** and a time-dependent volume, reprecenting the con-POnv."'The Variation 8 model was modified to repre-tainment, were added to the Variation 7 model. The sent the asynunetric loop response resulting from one single junction connected the RCS (Component 100, stuck-open steam PORV. First, the model was mod-Figure 5) to the time-dependent volume. ified by reducing the size of both the U-tube heat structure and the secondary solume, to represent a single affected steam generator. Second, ~ Varkrtion 9, SBLOCA Sequences With One Stuck-Open Component 100 was modified to represent both the Srmm POnV. The Variation 7 model was modified t reactor vessel below the hot leg centerline and the represent the asymmetric loop response resulting from single affected loop. The unaffected loops were one stuck-open steam PORV First, the Variation 7
- modeled separately from the affected portion of the model was modified by reducing the size of the U- RCS because of the asymmetric steam break. Third, tube heat structure and secondary volume to repre- the primary system heat structure of the base model sent a single affected steam generator. Second, was modified by subtracting the metal mass of the Component 100 was modified to represent both the unaffected loops. Finally, the primary-side break was reactor vessel below the hot leg centerline and the moved so that it was attached to the unaffected loops single affected loop. An additional control volume rather than the affected loop. This change more ac-was added to the model to represent the two unaf- curately represeMed the location of the break by fected primary loops. The additional volume was allowing saturated water, rather than the highly sub-necessary because the unaffected loops respond dif- cooled water from the affected loop, to Dow through ferently than the rest of the RCS during sequences the break.
with asymmetric steam breaks. The additional con-trol volume accounted for Dow stagnation in the Variation 12. MBLOCAs With Two Stuck-Open Steam unaffected loops, which results in flashing that can Ponvs. The Variation 8 model was modified to affect the pressure response of the RCS. Third, the represent the asymmetric loop response resulting from primary system heat structure of the base model was two stuck-open steam PORVs. First, the model was modified by subtracting the metal mass of the unaf- modified by reducing the size of the U-tube heat struc- ~ fected loops. The metal in the unaffected loops would ture and secondary volume to represent two affected not cool as rapidly as the metal in the affected loop steam generators. Second, Component 100 was because the Dow in the unaffected loops stagnates. modified to represent both the reactor vessel below the hot leg centerline and two affected loops. The Variation 10. SBLOCA Sequences With Two Stuck Open unaffected loop was modeled separately from the af-Steam POnVs. The Variation 7 model was modified fected portion of the RCS because of the as>Tnmetric to represent the asymmetric loop response resulting steam break. Finally, the primary system heat struc-from two stuck-open steam PORVs. First, the Varia- ture of the base model was modified by subtracting tion 7 model was modified by reducing the size of the metal mass of the unaffected loop. 32
The primary-side break was left attached to system pressure drops until the HPI volumetric Component 100, which represents the affected addition rate equals the thermal contraction rate. portion of the RCS. This was different from the break i location used in the Variation 11 model, but was 3.4.2 Small Steam Break Affecting One Steam justified by the assumption that the C loop contains Generator at Full Power. Sequence 9-25 (see
, both the primary-side break and one of the stuck- Table 9) involves a reactor trip from full power open PORVs. followed by a single, stuck-open steam PORV. For this sequence primary system pressure does not decline venterion 1.1 sr m Generator rube Rupture. For steam sufficiently to cause RCP trip. The detailed model generator tube rupture sequences, the base simplified (Section 3.3.1) and Variation 2 of the simplified model was modified as follows. The existing steam model (Section 3.3.3) were run for this sequence over generator secondary volume and tube heat structures a common period from 75 to 800 s. Figures 8 and 9 were reduced in size to represent the two unaffected show the results using the two models. The com-steam generators. A new volume, representing the af- parisons indicate that both the simplified model fected steam generator secondary was connected by pressure and the temperature responses agree well a break junction to the primary system volume. The with those of the detailed model. The minor dif-break junction area was the total for both break flow ferences in the temperature responses shown in paths. A loss coefficient which was consistent with Figure 9 are caused by modeling the reactor vessel up-the detailed model was chosen. per head as a dead-end flow path in the simplified model. In the plant, and in the detailed model (see Figure 2), minor flow paths into and out of the up-3.4 Simplified Model Benchmarks per head tend to circulate upper head fluid when the RCPs are operating. With the simplified model, this This section shows the results of benchmark com- circulation is not accounted for and the upper head parisons between simplified model and detailed model fluid is not cooled, resulting in the colder primary calculations of the same sequences. These com- system temperatures calculated with the simplified parisons lend confidence to the use of a method rely- model. While the temperature differences are small ing on calculations performed with a greatly simplified [ typically 4 K (7"F)], the extra fluid shrinkage due model. to these differences, caused the moderate discrepancy in primary system pressures shown in Figure 8.
3.4.1 Small Steam Break Affecting One Steam Generator at Hot Standby. Sequence 7-4 (see 3.4.3 Large Steam Break Affecting One Steam Table 7), starting from hot standby conditions, Generator at Hot Standby. Sequence 8-4 (see involves a single, stuck-open PORV. The detailed Table 8) involves a double-ended break of one steam model (Section 3.3.1) and Variation 1 of the line, starting from hot standby conditions. The de-simplified model(Section 3.3.3) were run for this se- tailed model (Section 3.3.1) and Variation 4 of the quence over a common period from 1390 to 2490 s. simplified model (Section 3.3.3) were run for this se-Figures 6 and 7 show the results using the two quence over a common period from 200 to 1198 s. models. The comparisons indicate that both the Figures 10 and 11 show the results using the two simplified model pressure and the temperature models. The comparisons indicate that both the responses agree well with those of the detailed model. simplified model pressure and the temperature re-Differences of up to about 11 K (20 F) in the sponses agree well with those of the detailed model. temperature comparison are believed to be caused by Figure 11 indicates very good agreement for the reac-the inclusion of the USG metal mass in the primary tor vessel downcomer temperature; the two culves dif-system heat structure (see Section 3.3.1, Variation 1). fer by at most 8 K (14*F). The downcomer pressures, Had the USG metal mass been deleted, as was the shown in Figure 10, are in adequate agreement with USG liquid mass, the comparisons would have been a maximum difference of 0.73 MPa (106 psi). better. The drops in primary system pressure at about 1900 s are caused by termination of charging flow 3.4.4 Steaf- Break Symmetrically Affecting Three when the pressurizer setpoint level is attained. The Steam Generators at Full Power. Sequence 9-15 drop in pressure when using the simplified model is (see Table 9) involves the sticking open of three steam smaller than when using the detailed model because dump valves (SDVs) following a reactor trip from the primary fluid thermal contraction rate is slightly full-power conditions. It is further assumed that the smaller with the simplified model, and the primary operator does not close the main steam isolation 33
2000 , , , , DETAILIED MODEL
~
-2200
--- SIMPLIFIED MODEL 14000 _- --2000 m n ,
O U L 5 5
- a
-1800 ~
I. 3 12000 L e 3 e -1600
- t. .....
10000 - i-------- -- --
-1400 8000 O 500 1000 1500 2000 2500 Time (s)
Figure 6. Sequence 7-4 simplified model pressure benchmark 600 , , , , DETAILIED MODEL -600 -
--- SIMPLIFIED MODEL ^ n M b " 550 -
V
- e
$ -500 L-U O L.
'e Q-500 - - e O.
E E
* -400
- T)
~~'
, , ' - n
*3 450 ..,,,-
L~ [
-300 400
< 0 500 1000 1500 2000 2500 Time (s) . Figure 7. Sequence 7-4 simplified model temperature benchmark. 34
16000 i , , DETAILIED MODEL
--- SIMPLIFIED MODEL -2200 14000 -
m ',
--2000 m o ',-
' .2 0-6 ,'.,' "
m c. . - -1800 8 12000 -
',, - 0 3
v ~~,
' 3 m - m e ',, -1600 g
** ' e 10000 -
'."-he -1400 0-8000 O 200 400 600 800 Time (s)
Figure 8. Sequence 9-25 simplified model pressure benchmark. 580 i , , DETAILIED MODEL
--- SIMPLIFIED MODEL m ^
M l'- V v 560 --550 0 , o L '.,' L 3 , 3 o ' o u ,' ,, u 0
- Q.540 c.
h .,, -630 h D '. o
'5 520
*3 L. L.
. -450 500 O 200 400 600 800
, Time (s)
Figure 9. Sequence 9-25 simplified model temperature benchmark. 35
16000 , , , , , DETAILIED MODEL
--- SIMPLIFIED MODCL 14000
--2000 m n .
O- 0 1 6 12000 - - v E.
- 8. #
3 e -1500 3 e 10000 -
.. - - M e
u G. .*** ..... - { 8000 -
-1000 6000 O 200 400 600 800 1000 1200 Time (s)
Figure 10. Sequence 8-4 simplified model pressure benchmark. 600 , , , , , DETAILIED MODEL
--- StMPLIFIED MODEL ^ ^
v M 550 - F v e
-500 ,
' u 3 s
** 500 - - *-
U U
' u e -400 ,
k '
, G.
E ' E e 450 - - e 3 -300 V-2.- - 2' 400 - - - k __ tu
-200 350 O 200 400 600 800 1000 1200
- Time (s)
Figure 11. Sequence 8-4 simplified model temperature benchmark. 36
valves. The detailed model(Section 3.3.1) and the RCS fluid temperatures and downcomer fluid tem-base simplified model (Section 3.3.2) were run for peratures, respectively. The temperature oscillations this sequence over a common period from 390 to calculated with the detailed model were related to the 1790 s. Figures 12 and 13 show the results using the low loop flows for this sequence. Similar oscillations two models. The comparisons indicate that both the were reported in Reference 1. Analysis indicated that simplified model pressure and the temperature those oscillations would not occur in the plant but, responses are in excellent agreement with those of , g the detailed model. The pressure responses differ by at most 0.23 MPa (34 psi), and the temperature were reasonable. Figure 15 shows that the average
. responses differ by at most 4 K (8'F). During most temperature from the two models agreed closely. The of the period of comparison, the agreement was calculated downcomer temperature with the simplified much better than that indicated by these numbers. model (shown in Figure 16) was within the oscilla-tions calculated with the detailed model, although the 3.4.5 S8LOCA at Hot Standby. Sequence 3-1 (see average cooldown rate was slightly smaller. The Table 3) involved a single, stuck-open pressurizer difference between the models was due to the assump-
! PORV at hot standby. The detailed model (Sec- tion of a constant loop flow for the HPI mixing , -tion 3.3.1) and Variation 7 of the simplified model calculation in the simplified model. Although the (Section 3.3.3) were run for this sequence over a companson was thought to be reasonable, it illustrates common period from 400 to 1725 s. Downcomer the difficulty in calculating the effect of HPI mixing > pressures calculated with the two models are shown on downcomer temperature when using a simplified in Figure 14. The pressure was relatively constant, model. The results shown in Figures 14,15, and 16 between 400 and 1500 s. A gradual depressurization show generally good agreement between the detailed began near 1500 s in both calculations when the and simplified models and indkate that the simplified pressurizer filled with liquid and the flow out the model can represent most of the important phenom-PORV increased. Figures 15 and 16 show average ena calculated with the detailed model. i 16000 , I ' DETAILIED MODEL ! --- SIMPLIFIED MODEL i 14000 -
-2000 m o
- 8. 'E I
6 12000 -
~
O e e L L 3 -1500 10000 -
..- e u L-
- n. ,
1 8000 - 4
~ -1000 6000 O 1000 2000 l ,
Time (s) Figure 12. Segoence 9-15 simplified model pressure benchmark. t I 37 i'
600 , i i DETAILlED MODEL -600
--- SIMPLIFIED MODEL m ^
V M l'- 550 - - W e . 500 3
+
0 0 L L * {500 - - { E E e - ', 400 e
~
D '
~, D
's 450 - -
*g Ls Ln.
.. -300 400 O 500 1000 1500 2000 Time (s)
Figure 13. Sequence 9-15 simplified model temperature benchmark.
.a00 , , , ,
DETAILIED MODEL
--- SIMPLIFIED MODEL 14000
--2000 m ^
O g .O 5 12000 - - v
' e L
E -1500 3 g 10000 g u e 1 I 8000 -~- -
... _..........._ ............. *' 1000 e i 6000 0 500 1000 1500 2000 l Time (s) .
Figure 14. Sequence 3-1 simplified model downcomer pressure benchmark, 38
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3.4.6 SBLOCA Combined with Steam Break, downomer temperature was also calculated well, Symmetrically Affecting Three Steam even after the start of accumulator flow. Generators at Full Power. Sequence 1-8 (see Table 1) involved a single, stuck-open pressurizer 3.4.7 MBLOCA Combined with Small Steam PORV and five stuck-open steam dump valves at Break, Affecting Two Steam Generators at Full full power. The detailed model (Section 3.3.1) and Power. Sequence 2-11 (see Table 2) involved a . Variation 7 of the simplified model(Section 3.3.3) 0.0635-m (2.5-in.) diameter hot leg break and two ! were run for this sequence over a common period stuck-open steam PORVs at full-power conditions. from 200 to 385 s. Downcomer pressures calculated The detailed model (Section 3.3.1) and Variation 9 , with the detailed and simplified models are shown of the simplified model (Sectbn 3.3.3) were run for in Figure 17. The pressures show remarkably good this sequence over a commou period from 400 to agreement, even after accumulator injection was 580 s. Downcomer pressures calculated with the initiated near 330 s. The rapid depressurizations detailed and simplified moads are shown in calculated with the detailed model at 265 and 375 s Figure 20. The average depressurhation rate was were caused by condensation spikes that were not shnitar with the two medels. Figur.s 21 and 22 show reasonable but did not have a significant long-term average RCS fluid temperatures and downcomer fluid effect on the results. Figures 18 and 19 show temperature, respectively. The average RCS temper-average RCS fluid temperatures and downcomer atures were within a few degrees shortly after the start fluid temperatures, respectively. The comparison of of the benchmark calculation. The downcomer tem-the average RCS temperatures indicates that the peratures also showed fairly good agreement. The simplified model calculated the overall cooldown simplified model was also able to calculate the trend rate well since the two temperatures were nearly of the rapid cooldown following the start of ac-parallel after 230 s. The effect of ECC mixing on cumulator injection near 450 s. 20000 , , , DETMLIED MODEL
--- SiWPLIFlED MODEL -2500
~
m 9 0
$ 15000 - -
E v
-2000 O *
- e L
S '3 e 10000 '- - 1500 E L o.
- 1000 e E E 5 5000 - -
E o o 3----500 > 0 ' ' ' O O 10 0 200 300 400 Time (s) Figure 17. Sequence 18 simplified model downcomer pressure benchmark.
[3 D eEQe8 3'e
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4 5 5 6 4 5 5 6 5 0 5 0 5 0 5 0 0 0 0 0 0 0 0 0 O O - F
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) r ) - -
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1 20000 , , DETA! LIED MODEL
--- SIMPLIFIED MODEL 2500 15000 -
9o. -2000 - 0 8 10000 - - --1500 e s M 3 m m-e
- u -
-1000 e 1
N 5000 -
-500 0 O O 200 400 600 Time (s)
Figure 20. Sequence 2-11 simplified model downcomer pressure benchmark. r 580 , , , , , DETAILIED MODEL SIMPLIFIED MODEL
^ 560 -- --550 m v
M l'- v e e
' 540 - - L 3
+- -
-500 -.3 0 0 L L
{520 - - { E E
". 450 *
, 500 - ...",
' ,, o T3 ,
3 N 3
.,N La- 480,- -
400 460 O 10 0 200 300 400 500 600 Time (s) l Figure 21. Sequence 2-11 simplified model average RCS temperature benchmark. f 42
- - . _ . ~ . . . _ , _. ._ , . - . - _ . . _ . , _ - - . _ _ _
600 , , DETAILIED MODEL 600
--- SIMPLIFIED MODEL n ^
V M- M-550 - c e L
- 3. 500 '
3-o o L L {500 - { E E
= ...,
-400 e.
.t ', ' -
t 3 450 - '
- .. - *s
. k 6
-300
. 400 O 200 400 600 Time (s) Figure 22. Sequence 2-11 simplified model downcomer temperature benchmark. 4 43
- 4. GROUP A RESULTS: SECONDARY-SIDE BREAK AFFECTING ONE STEAM GENERATOR The following sections present the results or 4.2 Small Steam Break at thermal-hydraulic analyses for sequences in Hot Standby .
Group A of Table 15, using the methods presented in Section 3 to determine reactor vessel downcomer fluid pressure and temperature histories. Section 4.1 Sequences involving small steam breaks at hot defines the group and describes the controlling standby are numbered 7-1 through 7-8 (see phenomena. Within Group A, sequences were Table 7). The sequences differ based on: organized by initiating event and power level as (1) isolation of auxiliary feedwater (AFW) at follows: (1) small break at hot standby, (2) small 10 minutes, (2) AFW flow rate (normal or high) to break at full power, (3) large break at 1.ot standby, the affected steam generator,(3) throttling of AFW and (4)large break at full power. Sections 4.2 to the unaffected steam generator, and (4) throttling through 4.5 present the results for these subgroups. of charging flow when pressurizer setpoint level is Conclusions covering Group A sequences appear in surpassed. Section 4.6. To facilitate referencing of data, plot-ted results showing the pressure and temperature For these sequences it is possible to overfill the - histories are organized in numerical sequence in USGs only if AFW is isolated or throttled to the Appendix A rather than appearing within this ASG. Due to very low ASG pressures, all AFW section. The figures shown in Appendix A were flows from the headers to the ASG, unless AFW computer generated using a limited number of valves to it are closed, in which case AFW can be points, thus discontinuities in slope are not delivered to the USGs. necessarily indicative of specific events in the sequences. Due to the large number of sequences Table 17 presents the methods used, and gives an investigated in this report, detailed discussions of indication of results for, sequences 7-1 through 7-8. thermal-hydraulic processes for each sequence are Plotted results for all sequences appear in not practical. Such discussions are documented for Appendix A. representative sequences in Reference 1. For Sequences 7-1 through 7-8, turbine-driven AFW is not initiated because it requires two steam 4.'l Group A Definition generators to have low level indications, and this condition was present only in the affected steam Group A includes all sequences controlled by a generator. steam line break which affect only one steam generator. Within this section a "small" steam line Calculations using the detailed model were per-break refers to a single, stuck-open steam line formed for initial portions of Sequences 7-4 power-operated relief valve (PORV) and a "large" (Scenario 3 in Reference 1) and 7-7. Simplified steam line break refers to a double-ended rupture models used were Variations 1, 2, and 3, as of a single steam line. Both breaks are located described in Section 3.3.3. downstream of the steam line flow restrictor and upstream of the main steam isolation valve (htSIV) For Sequences 71, 7 2, and 7-7, AFW is ter-and steam line check valve. So located, the break minated at 10 minutes; the affected steam generator cannot be isolated by closure of the htSIV or affeet subsequently drys out. Following dryout, the the other two steam generators, due to the action primary system heats up due to core decay heat, of the check valve. Thus, all sequences in Group A RCP power, and a loss of secondary heat sink. For , are controlled by primary system heat removal to convenience, hand calculations were performed a single, affected steam generator (ASG) and heat over the heatup phases of these sequences. Using addition from the two unaffected steam generators the simplified model results at the time of dryout, (USGs). the primary system temperature was extended to 44
! Table 17. Results of small steam breaks affecting 1 SG at hot standby Detailed Simplified Model Model ( Sequence Initiating Power Calculation Variation Used l Number Event Level Failures Used (from 3.3.3) 71 Small SLB HSB None 7-4 2 7-2 Small SLB HSB Charging not throttled 7-4 2
, 7-3 Small SLB HSB AFW overfill 7-4 3 7-4 Small SLB HSB AFW not isolated 7-4 I 7-5 Small SLB HSB AFW not isolated, charging not throttled 7-4 1 7-6 Small SLB HSB AFW not isolated. AFW overfill 7-4 1 7-7 Small SLB HSB AFW overfeed 7-7 2 7-8 Small SLB HSB AFW not isolated. AFW overfeed 7-7 1 Maximum Methods Used, Periods in Seconds Minimum Subsequent Figures RV DC RV DC Showing Sequence Detailed Simplified lland Temperature Pressure Plotted Number Model Model Calculations (*F) (psia) Results Notes a 7-1 0 - 600 600 - 2216 2216 - 7200 412 2371 A-131, A-132 A,B,C 7-2 0 600 600 2268 2268 - 7200 409 2371 A-133, A-134 A,B,C 73 0 600 600 - 7200 - 369 2371 A-135, A 136 A,B,D 7-4 0 - 1290 1290 - 7200 -
283 1565 A-137, A-138 A 7-5 0 - 1290 129') 7200 - 270 2371 A-139, A 140 A 7-6 0 - 1290 1290 7200 - 283 1565 A-141. A-142 A 7-7 0 - 600 600 - 2400 2400 - 7200 393 2371 A 143, A 144 A,B,C 7-8 0 - 1000 1000 - 7200 - 288 1528 A 145 A-146 A
- a. A-Turbine-driven AFW not initiated.
- B-Reactor coolant pumps do not trip.
C-Hand calculation performed over the heatup phase following dryout of ASO.
- D-Simplified model results were adjusted for ASG blowdown effects from 600 - 2216 s based on results in 7-1.
7200 s by calculating the effect of decay heat and not become colder than the primary system. As a RCP power on the temperature of the primary and result, the USGs are not recoupled by natural cir-secondary metal and liquid masses over this period. culation to the primary system, and the overfilling This temperature response was then used to cal- of the USGs has no effect on the primary system culate the pressure response through the assump- pressure and temperature. Therefore, results for tions of isentropic pressurizer behavior. Sequence 7-6 are identical to those for Sequence 7-4. For Sequence 7 3, in which only the USG over-fill was accounted for in the simplified model varia. For Sequences 7-1,7-2,7-4, and 7-7, the AFW tion, the effects of the ASG blowdown from 600 to is terminated to the ASO at 10 minutes; by about 2216 s (as calculated for Sequence 71) were 45 minutes, the ASG secondary was dry. As a superimposed on the simplified model results over result, in these sequences the reactor coolant pumps this period. remained operating, and primary system temper-atures increased after about 45 minutes. The The description for Sequence 7-6 is identical to resulting thermal expansion of the primary fluid that for Sequence 7-4 except that in Sequence 7-6 drove the primary system pressures up to the PORV the USGs are overfilled with AFW, following the setpoint. throttling of AFW to the ASG. In Sequence 7-4 this throttling occurred at 3318 s, a time when the For Sequences 7-4, 7-5, 7-6, and 7-8 the AFW primary system and the ASG were significantly was continued until the time of spillover of liquid colder than the USGs. As the USGs are overfilled to the steam line from the ASG. After spillover, the with AFW they become colder; however, they do AFW was throttled to maintain ASG secondary 45
mass. Primary system temperatures declined con- Table 18 presents the methods used, and gives an tinuously in these sequences, with the slope reduced indication of results for, Sequences 5-1, and following AFW throttling. Due to continuous 9-25 through 9-32. Plotted results for all sequences thermal contraction, the primary system pressures appear in Appendix A. were maintained below 11 MPa (1600 psia) except for Sequence 7-5 in which the failing of charging For Sequences 9-25 through 9-32, both the , throttling drove the pressure to the PORV setpoint. motor- and turbine-driven AFW operate. Unlike the This caused Sequence 7-5 to be the most severe small steam break at hot standby, at full power the sequence of the subgroup, with a minimum temper- USGs relieve steam for a period following reactor ature of 405 K (270*F) and a maximum pressure trip. As a result, the USG levels decrease sufficiently of 16.35 MPa (2371 psia), both of which occurred to cause initiation of turbine-driven AFW. at the end of the two-hour period. In addition to Sequence 5-1, a detailed model calculation was performed for the initial portion of 4.3 Small Steam Break at Sequence 9-25. A major finding of this calculation Full Power was that primary system pressure does not decline sufficiently to cause tripping of the reactor coolar.t pumps (RCPs). Simplified model calculations were Sequences involvm.g small steam breaks at full performed using Variation 2, as described in power are 5-1 and 9-25 through 9-32 (see Tables 5 Section 3.3.3. and 9). Sequence 5-1 is initiated by the steam break, while Sequences 9-25 through 9-32 are initiated by For Sequence 5-1, the detailed model was run to a reactor trip followed by a steam PORV failing 337 s, at which time a new steady state condition open. This distinction is important because a had been attained. This steady condition is expected detailed model calculation of Sequence 5-1 indi- to continue to 7200 s, so pressures and temperatures cated that, for an initiating event of a single, stuck- do not change between 377 and 7200 s. For open steam PORV, an automatic reactor trip is not Sequence 5-1 the absence of a reactor trip controls encountered. Instead, the steam and feed systems
- the response. Reactor vessel downcomer pressure respond to the effects of the break and, after and temperature remain near their full-power, approximately 5 minutes, the plant reaches a new steady state values.
steady operating point, with the cold leg temper- . ature 1 K (1.8'F) below the starting point. This For Sequences 9-25 through 9-32, the temper-drop in temperature is not sufficient to cause an atures calculated using the simplified model were automatic overpower reactor trip due to moderator adjusted after the time the ASG was water-filled temperature effects. (1877 s with normal AFW delivery rate, and 1047 s at the overfeed rate). After that time, AFW is There is a possibility that the operator would throttled to the ASG, and AFW flows to the USGs. respond to conditions present following a stuck- The simplified model does not account for a cool-open steam PORY by manually tripping the reac- down of the USGs during this period. To adjust for tor. For purposes of this study, it was assumed for this, the effect of the USG AFW demanded on the Sequence 5-1 that this would not occur. If manual overall system temperature was determined by an reactor trip did occur, Sequence 5-1 could be energy balance hand calculation and this effect was expected to be virtually identical to Sequence 9-25 superimposed on the simplified model calculation. in which the reactor trip was the initiating event. Primary system pressure during this period was then Note in their final analysis, ORNL chose to assume adjusted accordingly, based on the primary system manual reactor trip would occur and thus used the fluid shrinkage caused by the temperature results from Sequence 9-25 for their analysis, adjustment. Sequences 9-25 through 9-32 all assume no AFW For Sequences 9-25 through 9-32, the primary isolation. Sequences 9-25 through 9-32 differ based system temperatures continuously decline. The rate on: (1) AFW flow rates (normal or high), of decrease is significantly slowed when AFW is - (2) throttling of AFW to the USGs at 40% level, throttled to the ASG, at the time the ASG begins and (3) throttling of charging flow when pressurizer to spill liquid to the steam line. Primary system setpoint level is attained. pressures are stabilized below 11.55 MPa 46
Table 18. Results of small steam breaks affecting 1 SG at full power Detailed Simplified hiodel Model Sequence initiating Power Calculation Variation Used
. Number Event Level Failures Used (from 3.3.3) 5-1 Small SLB Full None 5-1 -
9-25 Reactor trip Full Stuck-open steam PORV 9-25 2
, 9-26 Reactor trip Full Stuck-open steam PORV, AFW overfill 9-25 2 9-27 Reactor trip Full Stuck-open steam PORV, charging not throttled 9-25 '
2 9-28 Reactor trip Full Stuck-open steam PORV, AFW overfill, charging 9-25 2 not throttled 9-29 Reactor trip Full Stuck-open steam PORV, AFW overfeed 9-25 2 9-30 Reactor trip Full Stuck-open steam PORV, AFW overfeed, AFW 9-25 2 overfill 9-31 Reactor trip Full Stuck-open steam PORV, AFW overfeed, charg- 9-25 2 ing not throttled 9-32 Reactor trip Full Stuck-open steam PORV, AFW overfeed, AFW 9-25 2 overfill, charging not throttled Maximum Methods Used, Periods in Seconds Minimum Subsequent Figures RV DC RV DC Showing Sequence Detailed Simplified fland Temperature Pressure Plotted Number Model Model Calculations (*F) (psia) Results Notes a 51 0 - 337 - 337 - 7200 544 22 % A-65, A-66 A,B,C 9-25 0 75 75 - 7200 1877 - 7200 343 1675 A-215, A-216 A,D 9-26 0- 75 75-7200 1877 - 7200 292 1647 A-217, A-218 A,D 9-27 0 - 75 75 - 7200 1877 - 7200 333 2371 A-219, A-220 A,D 9-28 0 - 75 75 - 7200 iR77 - 7200 282 2371 A-221, A-222 A,D 9-29 0 - 75 75 - 7200 1047 - 7200 344 1623 A-223, A-224 A,D , 9 0 - 75 75 - 7200 1047 - 7200 293 1602 A-225 A-226 A,D 9-31 0 - 75 75 - 7200 1(M7 - 7200 334 2371 A-227. A-228 A,D 9-32 0 - 75 75 - 7200 1047-7200 283 2371 A-229, A-230 A,D
- a. A-Reactor coolant pumps do not trip.
B-Reactor does not trip. C-Steady conditions at 337 s expected to continue through 7200 s. D---Iland calculations performed to adjust simplified model results for AFW delivery to the USO. (1675 psia), except in cases where the charging fails fer based en: (1) AFW isolation to the ASG at to throttle and the pressures are driven to the 10 minutes, (2) AFW delivery rate (normal or pressurizer PORV opening setpoint. overfeed), (3) AFW throttling to the USG at 40% level, and (4) throttling of charging when Sequence 9-28, involving failure to throttle AFW pressurizer setpoint level is attained. to the USGs and failure to throttle charging, was the most severe sequence of the subgroup, with a Table 19 presents the methods used, and gives an minimum temperature of 412 K (282*F) and a max-indication of results for, Sequences 8-1 through 8-6. imum pressure of 16.35 MPa (2371 psia), both of Plotted results for all sequences appear in
, which occurred at the end of the two-hour period.
Appendix A. 4,4 Large Steam Break at .. . For these sequences it is possible to overfill the Hot Standby USGs only ir AFW is isolated or throttled to the ASG. Due to very low ASG pressure, all AFW The sequences involving large steam breaks at hot flows from the headers to the ASG, unless AFW standby are 81 through 8-6. These sequences dif- valves to it are closed. 47
r Table 19. Results of large steam breaks affecting 1 SG at hot standby Detailed Simplified Model Model Sequence Initiating Power Calculation Variation Used Number Event Level Failures Used (from 3.3.3)
- 8-1 Large SLB HSB None 8-4 4 8-2 Large SLB HSB Charging not throttled 8-4 4
' 8-3 Large SLB HSB AFW overfill 8-4 4 .
8-4 Large SLB HSB AFW not isolated 8-4 4 8-5 Large SLB HSB AFW overfeed 8-4 4 8-6 Large SLB HSB AFW not isolated. AFW overfeed 8-4 4 Maximum Methods Used. Periods in Seconds Minimum Subsequent Figures RV DC RV DC Showing Sequence Detailed Simplified Hand Temperature Pressure Plotted Number Model Model Calculations (*F) (psia) Results Notes" 8-1 0 600 600 - 7200 - 284 2371 A-155. A-156 A 8-2 0 600 600 7200 - 283 2371 A-157, A-158 A 8-3 0 600 600 2850 2850 - 7200 284 1792 A-159. A-160 A,B 8-4 0 200 200 - 7200 - 2M 1947 A-161, A-162 A 8-5 0 - 200 200 - 7200 - 274 2371 A 163. A 164 A 8-6 0 - 200 200 - 7200 - 201 1995 A-165. A 166 A
- a. A-Turbine-driven AFW does not initiate.
B-Hand calculation used to introduce effects of overfilling the USG with AFW. For these sequences, turbine-driven AFW is not system at 2850 s; decreases again until the USGs are initiated, because it requires two steam generators filled at 3200 s; and then increases again when AFW to have low level indications, and this condition was is throttled to all steam generators. - present only in the affected steam generator. Sequences 8-1 through 8-6 start with a rapid A detailed model calculation was performed for cooldown and depressurization of the primary the initial portion of Sequence 8-4 Simplified system. In Sequences 8-1,8-2,8-3, and 8-5, AFW models used Variation 4, as described in is isolated to the ASG at 10 minutes and the primary Section 3.3.3. system heats up. The resulting primary system liquid thermal expansion drives the primary system In Sequence 8-3 there is an AFW overfill of the pressure up to the pressurizer PORV opening set-unaffected steam generators that begins at 600 s, point (except for 8-3, where cooling to the USGs when the AFW is isolated to the ASG. At 2850 s, prevents it). In Sequences 8-4 and 8-6, AFW isola- , the USG temperatures fall below the primary system tion does not occur until the ASG begins to spill temperature, and natural circulation of the liquid to the steam line (3700 s in 8-4 and 2452 s unaffected loops commence. Variation 4 of the in 8-6). As a result,in Sequences 8-4 and 8-6, ASG simplified model does not include USG cffects, and cooling continues much longer and the primary the following hand calculation was performed to system temperatures fall well below those for the address the time period from 2850 s to 7200 s. The other sequences. Due to additional primary system primary system temperature response was deter- fluid shrinkage, however, the primary system
- mined by taking a mass-weighted average of the pressures were maintained below 13.8 MPa primary system temperature response from (2000 psia) in Sequences 8-4 and 8-6.
Sequence 81 and the USG temperature response as . it is overfilled. As a result, the primary system All of Sequences 8-1 through 8-6 are severe. temperature for Sequence 8-3: decreases rapidly Sequences 8-1, 8-2, and 8-5 result in minimum !. until AFW is isolated to the ASG at 600 s; increases temperatures of about 407 K (274*F) and maximum until the USGs become colder than the primary pressures of 16.35 MPa (2371 psia). Sequences *-4 48
and 8-6 result in minimum temperatures of about steam dump control system. Primary system 367 K (20l'F) and maximum pressures of temperature excursicns would be limited by it. 13.8 h1Pa (2000 psia). Sequence 8-3 is slightly less severe, with a minimum temperature of 413 K Sequence 6 " was found to be equivalent to (284*F) and a maximum pressure of 12.35 h1Pa Sequence 6-5. The two sequences differ only by
. (1792 psia). USG AFW overfill criteria. The overfill of the USGs in Sequence 6-7 does not cause the USG 4.5 Large Steam Break at temperature to fall below the primary system temperature calculated for Sequence 6-5.
Full Power Therefore, unaffected loop natural circulation does not commence; the USGs remain decoupled from Sequences involving large breaks at full power are the primary system; and the USG overfill does not 6-1 through 6-9. These sequences differ, based on: affect the primary system pressure and temperature (1) AFW isolation to the ASG at 10 minutes, responses. (2) AFW delivery rate (normal of overfeed),
; (3) throttling of AFW to the ASG at 40% level, and Because AFW is not isolated to the ASG at (4) throttling of charging flow when pressurizer 10 minutes, primary system temperatures fall con-setpoint level is attained. siderably lower in Sequences 6-5,6-6,6-7, and 6-9 than in the other sequences where AFW is isolated For these sequences it is possible to overfill the at 10 minutes. The extra primary fluid shrinkage USGs only if AFW is isolated or throttled to the caused by cooling to the lower temperatures, ASG. Due to very low ASG pressures, all AFW however, prevented the primary system pressure flows from the headers to the ASG, unless AFW from reaching the pressurizer PORV opening set-valves to it are closed. Both the motor- and turbine- point pressure (except in Sequence 6-6 that included driven AFW operate in these sequences because the a failure to throttle charging).
USGs relieve steam for a period following reactor
. trip. As a result, the USG levels decrease sufficiently Because the AFW was isolated at 10 minutes in to cause initiation of turbine-driven AFW. Sequences 6-1, 6-2, 6-3, 6-4, and 6-8, primary system temperatures fell to only about 422 K 2
. Table 20 both presents the methods used, and (300*F). However, in all these sequences, the gives an indication of results for, Sequences 6-1 primary system pressure was later driven to the through 6-9. Plotted results for all sequences appear pressurizer PORV opening setpoint pressure by in Appendix A. thermal expansion of the primary system fluid.
A detailed model calculation was performed for The most severe sequence of this subgroup was the initial portion of Sequence 6-1. Simplified thus 6-6. The minimum temperature was 381 K models used Variations 3 and 4, as described in (227 F), and maximum subsequent pressure was Section 3.3.3. 16.35 hlPa (2371 psia). In Sequences 6-5,6-6,6-7, and 6-9, AFW is not 4.6 Conclusions isolated to the ASG at 10 minutes, and heat removal to the ASG dominated the sequence. In Sequences 6-1 through 6-4 and 6-8, however, AFW Group A includes 32 sequences involving steam to the ASG was isolated at 10 minutes, and heat breaks affecting one steam generator. These removal to the USGs became a dominant mecha- sequences were investigated in subgroups according nism. For Sequences 6-1 through 6-4 and 6-8, to break size (small or large) and power level (hot
+ therefore, detailed model results were used to 135 s; standby or full-power initial conditions). General then Variation 4 simplified model results were used findings for the subgroups are summarized below.
j to 600 s, followed by Variation 3 simplified model
, results to 7200 s. In Sequences 6-1 and 6-8, the For the small steam break (stuck-open steam simplified model results were adjusted so that PORV) at hot standby, Sequence 7 5 (no AFW primary system temperatures did not exceed $54 K isolation and a failure of charging throttling) was (547'F). This is the saturation temperature cor- found to be the most severe sequence. This sequence responding to the opening setpoint pressure for the had a minimum temperature of 405 K (270*F) and 49
Table 20. Results of large steam breaks affecting 1 SG at full power Detailed Simplified Model Model Sequence Initiating Power Calculation Variation Used Number Event Level Failures Used (from 3.3.3) 61 Large SLB Full None 6-1 4,3 6-2 Large SLB Full Charging not throttled 6-1 4,3 6-3 Large SLB Full AFW overfill 6-1 4,3 - 6-4 Large SLB Full AFW overfill, charging not throttled 61 4,3 6-5 Large SLB Full AFW not isolated 6-1 4 6-6 Large SLB Full AFW not isolated, charging not throttled 6-1 4 6-7 Large SLB Full At W not isolated, AFW overfill 6-1 4 6-8 Large SLB Full AFW overfeed 6-1 4,3 6-9 1.arge SLB Full AFW not isolated, AFW overfeed 6-1 4 Maximum Methods Used. Periods in Seconds Minimum Subsequent Figures RV DC RV DC Showing Sequence Detailed Simplified Hand Temperature Pressure Plotted Number Model Model Calculations (*F) (psia) Results Notesa 6-1 0 - 135 135 6532 6532 - 7200 302 2371 A-Il3, A II4 B,C 6-2 0 - 135 135 - 7200 - 302 2371 A 115. A 116 B 6-3 0 135 135 - 7200 - 302 2371 A-117, A I18 8 6-4 0 - 135 135 - 7200 - 302 2371 A-119, A 120 B 6-5 0 - 135 135 - 7200 - 229 1772 A-121, A 122 0 135 6-6 135 - 7200 - 227 2371 A-123. 5-124 6-7 0 135 135 - 7200 - 229 1772 A-125, A-126 A 6-8 0 135 135 6600 6600 - 7200 299 2371 A 127, A-128 B,C 6-9 0 135 135 - 7200 - 229 1781 A-129, A 130 e
- a. A-Sequence 6-7 was determined to be equivalent to 6-5 because USGs do not become colder than the primary system.
B-Variation 4 simplified model used to 600 s, Variation 3 simplified model used for remainder of period shown. C-Simplified model results were adjusted by limiting primary system temperature increases such that 547'F is not exceeded. 547'F is saturation temperature corresponding to opening setpoint pressure for the steam dump control system. maximum pressure of l6.35 h!Pa (2371 psia). Other For the large steam break (double-ended main sequences in the subgroup were significantly less steam line break) at hot standby, all 6 sequences severe in temperature, pressure, or both. investigated were severe Two classes of severity were found. The first class, with AFW isolated, has For the small steam break (stuck-open steam 3 sequences near 407 K (274*F), 16.35 h1Pa PORV) at full power, Sequence 9-28 (failure to (2371 psia). The second class, with AFW not isolated, throttle AFW and charging) was the most severe has 2 sequences near 367 K (201*F),13.8 h1PA sequence [412 K (282*F),16.35 h1Pa (2371 psia)], (2000 psia). and Sequence 9-32 was nearly as severe. An impor-tant finding for this group was that, if the initiating For the large steam break at full power, the most event is the stuck-open PORY (Sequence 5 l), a severe sequence was 6-6 (no AFW isolation and a reactor trip does not occur and an overcooling event failure to throttle charging). Other sequences in the is avoided, unless the operator manually trips the group were significantly less severe because either: , plant in response to the transient observed. If the (1) AFW is isolated at 10 minutes and the cooldown initiating event is a reactor trip and the stuck-open is stopped, or (2) AFW and the cooldown continue, PORV is a subsequent failure, the primary system but the resulting primary system fluid shrinkage pre-pressure does not decline sufficiently to cause trip- vents the primary system pressure from increasing to ping of the reactor coolant pumps, the pressurizer PORV opening setpoint pressure. 50
- 5. GROUP B RESULTS: SECONDARY-SIDE BREAK WITH THREE SYMMETRICALLY AFFECTED STEAM GENERATORS The following sections present the results of decrease sufficiently, due to loss of secondary in-thermal-hydraulic analyses for sequences in ventory. Turbine-driven AFW is assumed to be Group B of Table 15 using the methods presented unavailable when the steam generator secondary -
in Section 3 to determine reactor vessel downcomer pressures are below 0.69 N1Pa (100 psia). Below this fluid pressure and temperature histories. Section 5.1 pressure, insufficient steam is available to run the defines the group and describes the controlling turbine-driven AFW pumps. phenomena. Section 5.2 discusses the results for a single, stuck-open steam line PORV (and steam line check valve failure) at hot standby conditions. 5.2 One Stuck-Open Steam PORV Results for sequences initiated from full-power con. and Steaml.ine Check Valve ditions are shown in Section 5.3. These sequences Failure at Hot Standby involve one or more stuck-open steam dump valves or three stuck-open steam PORVs. Conclusions Sequence 7-12 involves a stuck-open steam covering all Group B sequences appear in Sec- PORV on one steam line and a failure of the steam tion 5.4. To facilitate referencing of data, plotted line check valve on the same line. With both of these results showing the pressure and temperature failures, all three steam generators blow down histories are organized in numerical sequence order through the steam PORV. As specified in the in Appendix A rather than appearing within this sequence description, all AFW is isolated at 10 min. section. Due to the large number of sequences As a result, the primary system pressure does not investigated in this report, detailed discussions of decline sufficiently to cause tripping of the reactor thermal-hydraulic processes for each sequence are coolant pumps. Restoration of AFW will eventually not practical. Such discussions are documented for be required (after the two hour period) to avoid dry-representative sequences in Reference 1. ng out the steam generator secondaries. 5.1 Group B Definition Table 21 presents the methods used, and gives an indication of results for, Sequence 712. Plotted results for this sequence appear in Appendix A. Group B includes all sequences controlled by a steam break that symmetrically affects all three A detailed model calculation for the initial por-generators and no primary system break. Sequences tion of Sequence 7-4 was used to predict the of this type fall into three categories: sequences behavior for Sequence 7-12. Sequence 7-4 involves involving (1) stuck-open steam line PORVs with the same steam break withour
- check valve coincident failure of the check valve on the same faib ire. The primary system heat remo ral was thus line, (2) three stuck-open steam line PORVs, and nearly the same for Sequences 7-4 and 7-12.
(3) one or more stuck-open steam dump valves Variation 5 of the simplified model(as described (SDVs). In all of these cases, the ste.:m break (or in Section 3.3.3) was used oser the later stage of breaks) are located so that the steam flows exiting the sequ(uce. A mass-weighted average of the steam the steam generators are equal. As a result, the generator tecondary conditions (of the 1 ASG and responses of the three steam generators are 2 USGs) from the detailed model calculation for identical. Sequence 7-4 was used to initialize the simplified model calculation for Sequence 7-12. Sequences in Group 11 are controlled by equal primary system heat removal to each steam The minimum temperature of the fluid in the generator and heat addition from core decay heat reactor sessel downcomer for Sequence 7-12 was and stored energy in metal heat structures. 440 K (332'F), and the maximum subsequent l prenure was 10.4 NIPa (1505 psia). The ses crity of i For all Group 11 sequences, turbine.drisen AFW the sequence was limited by the isolation of AFW i is initiated because all three steam generator lesels at 10 min. i i l 51
Table 21. Results of one stuck-open steam PORV and steamline check valve failure at hot standby Detailed Simplified hfodel hfodel Sequence Initiating Power Calculation Variation Used
- Number Event Level Failures Used (from 3.3.3) 7 12 Small SLB 115B None 7-4 5 Afaximum Afethods Used, Periods in Seconds hfinimum Subsequent Figures i RV DC RV DC Showing Sequence Detailad Simplified fland Temperature Pressure Plotted Number Afodel N1odel Calculations (* F) (psia) Results 7 12 0 - 600 600 7200 -
332 l$0$ A 153, A 154 5.3 One or More Steam Dump Sequences 9-2 through 9-8 involve one stuck-Valves, or Three Steam open SDV; 9-9 through 9-13 two SDVs; 9-14 through 9-18 three SDVs; and 9-19 through 9-23 PORVs, Stuck Open Following five SDVs. Sequences 9-41 through 9-47 mvolve Reactor Tn,p From Full Power three stuck-open steam line PORVs. Detailed model calculations were performed for the initial portions Sequences 9-2 through 9-23 involve one or more of one sequence with each break size: 9-3,9-9,9-15, stuck-open steam dump valves (SDVs) at full 9-19, and 9-41. Simplified model calculations used power. Sequences 9-41 through 9-47 involve three the base simplified model described in stuck-open steam PORVs at full power. For all Section 3.3.2. these sequences, isolation of AFW is not required unless a main steam isolation valve (htSIV) fails to Sequences 9-6, 9-7, 9-8, 9-13, 9-18, 9-23, 9-45, close on demand. The steam flow was not sufficient 9 46, and 9-47 include an overfeed of AFW at a rate to demand the NISIVs with one, two, or three SDVs 48% above normal. The detailed model calculation
- or three steam PORVs stuck open. With five SDVs used for these sequences did not include the effects stuck open, the steam flow was sufficient over the of the overfeed. To account for this, the initial con-first few seconds of transient, but the additional ditions for the simplified model were adjusted by requirement for low average temperature was not increasing the steam generator secondary mass for met until later in the transient. To demand the the extra AFW which would have been injected over htSIVs, these two conditions must be satisfied at the periods when the detailed model was applied.
the same time; therefore, the htSIVs were not demanded for five SDVs stuck open as well. Since Results of these sequences show a continuous the htSIVs are not demanded, AFW is not isolated cooldown of the primary system. The cooldown rate in these sequences. The sequences differ by: slows considerably when the turbine-driven AFW (1) AFW feed rates (normal or overfeed),(2) AFW, is terminated due to low secondary pressure. The throttling at 40% level or when hquid spills to the cooldown rate slows even more when all AFW is steam line, and (3) throttling of charging flow when throttled because of liquid spilling to the steam the pressurizer setpomt level is attamed. Note it is 11 or .50% level attainment, depending on the assumed that the operator does not manually close sequence description, the htSIVs throughout the 2 hour period and this is a conservative assumption. Table 22 presents the analysis methods used, and caused by the cooldown, limits the repressurization gives an indication of results for, one SDV stuck to about 11.7 hlPa (1700 psia) except in sequences open, Table 23 for two SDVs, Table 24 for three involving a failure to throttle chargind. For these SDVs, Table 25 for fise SDVs, and Tabic 26 for sequences, the primary system pressure increases to three steam PORVs. Plotted results for all the opening setpoint pressure of the pressurizer sequences appear in Appendix A. PORV, thus increasing the severity of the sequence. 52
Table 22. Results of one stuck-open steam dump valve at full power Detailed Simplified hfodel hfode! Sequence Initiating Power Calculation Variation Used . Number Event Level Failures Used (from 3.3.3) 92 Reactor trip Full I stuck-open SDV 9-3 Base 93 Reactor trip Full I stuck-open SDV, AFW overfill 9-3 Base 9-4 Reactor trip Full I stuck-open SDV, charging not throttled 9-3 Base 9-5 Reactoe trip Full I stuck-open SDV, AFW ove fill, charging not 9-3 Base throttled 94 Reactor trip Full I stuck open SDV, AFW overfeed 9-3 Base 9-7 Reactor trip Full I stuck-open SDV, AFW overfeed, AFW overfill 9-3 Base 9-8 Reactor trip Full I stuck open SDV, AFW overfeed, charging not 9-3 Base throttled hfaximum Afethods Used Periods in Seconds hfinimum Subsequer.t Figures RV DC RV DC Showing Sequence Detailed Simplified lland Temperature Pressure Plotted Number hiodel hfodel Calculations l'F) (psia) Results Notess 9-2 0 900 900 7200 - 312 1544 A 169, A 170 9-3 0 900 900 7200 - 294 1553 A 171, A-172 9-4 0 900 900 7200 - 305 2371 A 173, A 174 9-5 0 900 900 7200 - 282 2371 A 175, A 176 94 0 900 900 7200 - 306 1527 A 177, A 178 A 9-7 0 900 900 7200 - 297 1694 A 179, A 180 A 9-8 0 900 900 7200 - 308 2371 A 181. A-182 A
- a. A-Adjustment made in simplified model initial condition to account fx AFW overfeed from 0 to 900 s-Table 23. Results of two stuck open steam dump valves at full power Detailed Simplified hfodel hfodel Sequence initiating Power Calculation Varlation Used Number Fsent 1.evel Failures Used Ifrom 3.3.3) 9-9 Reactor trip Full 2 stuck open SDVs 9-9 Dase 9 10 Reaetor trip Full 2 stuck cren SUVS. AFW oserfill 99 Hase 9 11 Reactor trip Full 2 stuck < pen SDVs, charging not throttled 99 Base 9-12 Reactor trip f ull 2 stuck-open SDVs, AFW overfill, charging not 9-9 Base throttled 9-13 Reactor trip Full 2 stuck open SDVs, AFW oveifeed 99 Dase hfatimum hiethods Used Periods in Seconds hl nimum Subsequent Figures RV DC RV DC Showing Sequence Detailed Simphried lland Temperature Pressure Plotted Number Nfodel Afodel Cateulations ('ll (psW Results Notesa e 9-9 0 400 400 7200 -
264 1611 A 183 A 184 9 10 0 400 400 7200 - 248 1318 A 185 A 186 9 11 0 400 400 7200 - 257 2371 A 187, A 188 9-12 0 400 400 7200 - 241 2371 A 189 A 190
. 9 13 0 400 400 7200 - 267 1545 A 191, A 192 A
- a. A-Adjustment made in simphfied modelinitial condition to account for AFW overfeed from 0 to 400 s.
53
Table 24. Results of three stuck open steam dump valves at full power Detailed Simplified hiodel hiodel Sequence Initiating Power Calculation Variation Used Nuruber Event Level Failures Used (from 3 3 3) , 9-14 Reactor trip Full 3 stuck-open SDVs 9-15 Base 9 15 Reactor trip Full 3 stuck-open SDVs, AFW overfill 9-15 Base 9-16 Reactor trip Full 3 stuck-open SDVs, charging not throttled 9-15 Base 9-17 Reactor trip Full 3 stuck-open SDVs, AFW overfill, charging not 9 15 Base throttled 9-18 Reactor trip Full 3 stuck-open SDVs. AFW overfeed 9-15 Base Staximum hiethods Used, Periods in Seconds blinimum Subsequent Figures RV DC RV DC Showing Sequence Detailed Simplified lland Temperature Pressure Plotted Number htodel htodel Calculations (*F) (psia) Results Notesa 9-14 0 390 390 7200 - 234 1610 A 193, A 194 9-15 0 390 390 7200 - 226 1613 A 195, A-l% 9 16 0 390 390 7200 - 225 2371 A 197, A-198 9-17 0 390 390 7200 - 220 237i A 199, A-200 9 18 0 390 390 7200 - 258 1754 A-201, A-202 A
- a. A-Adjustment made in simplified model initial condition to account for AFW overfeed from 0 to 390 s-Table 26. Results of five stuck open steam dump valves at full power Detailed Simplified klodel htodel Sequence initiating Power Calculation Variation Used Number Event Level Failures Used (from 3.3.3) 9-19 Reactor trip Full 3 stuck-open SDVs 9-19 Base 9 20 Reactor trip Full 3 stuck-open SDVs, AI:W oserfill 9-19 Base 9-21 Reactor trip rull 5 stuck open JJVs. charging not throttled 9 19 Base 9-22 Reactor trip Full 5 stuck-open SDVs, AFW overfill, charging not 9-19 Base throttled 9-23 Reactor trip Full $ stuck-open SDVs, AFW overfeed 9 19 Base hlatimum klethods Used Periods in Seconds Minimum Subsequent Figures RV DC RV DC Showing Sequence Detailed Simplified lland Temperature Pressure Plotted Number Model Stodel Calculations ('F) (psla) Results Notesa 9-19 0 400 400 7200 - 203 1594 A 203 A 204 9 20 0 400 400 7200 - 201 1573 A 205, A-206 9 21 0 400 400 7200 - 199 237i A 207, A 208 9-22 0 400 400 7200 - 1% 2371 A 209, A-210 9-23 0 400 400 7200 - 213 1669 A 211, A-212 A
- a. A-Adjustment made in simplified model initial condition to account for AFW overfeed from 0 to 400 s.
54
Table 26. Results of three stuck-open steamline PORVS at full power Detailed Simplified Model Model Sequence Initiating Power Calculation Variation IJsed
. Number Event Level Failures tJsed (from 3.3.3) 9-41 Reactor trip /ull 3 stuck-open SPORVs 9-41 Base 942 Reactor trip Full 3 stuck open SPORVs, AFW overfill 9-41 Base
, 9-43 Reactor trip Full 3 stuck open SPORVs, charging not throttled 9-41 Base 9-44 Reactor trip Full 3 stuck-open SPORVs, AFW overfill, charging 9-41 Base not throttled 9-45 Reactor trip Full 3 stuck-open SPORVs, AFW overfeed 9-41 Base 9-46 Reactor trip Full 3 stuck-open SPORVs, AFW overfeed, AFW 9-41 Base overfill 9-47 Reactor trip Fun 3 stuck-open SPORVs, AFW oserfeed, charging 941 Base not throttled Maxiraum Methods Used. Periods in Seconds Minimum Subsequent Figures RV DC RV DC Showing Sequence Detailed Simplified lland Temperature Pressure Plotted Number Model Model Calculations ('F) (psial Results Notesa 9-41 0 - 400 400 - 7200 -
268 1615 A-247, A-248 9-42 0 - 400 400 - 7200 . 251 1517 A-249, A-250 943 0 400 400 7200 - 262 237I A 251. A 252 9-44 0 - 400 400 7200 - 244 2371 A-253, A-254 9-45 0 - 400 400 - 7200 - 272 1490 A-255, A 256 A 9-46 0 400 400 7200 - 259 1632 A-257, A 258 A 947 0 - 400 400 7200 - 265 2371 A-259, A-260 A
- s. A-Adjustment made in simplified model initial condition to account for AFW overfeed from 0 to 400 s.
The severity of the results usually varied directly 5.4 Conclusions with the size of the steam break. Minimum temper-atures of the reactor vessel downcomer fluid within Group B includes 30 sequences involving steam each break size category were calculated for the se. b eaks that symmetrically affect all three steam quence involving failure to throttle AFW and charg. generators. Sequences in this group were in general ing. The smallest break size (one SDV) resuhed in un to k sese,re because auxiliary feedwater con-tinues for a sigmficant tune. The severity of results a minimum temperature of 412 K (282*F). As the was found 'o be sensitive to break size, with the break size increased, the minimum temperature larger break size producing the most severe results. decreased: 391 K (244*F) for three steam PORVs, Furthermore, sequences in which charging fails to 389 K (241*F) for two SDVs, 377 K (220*F) for throttle are more severe than those in w hich charg-three SDVs, and 364 K (196 F) for five SDVs. ing throttles, because primary system pressure is driven up to the pressurizer PORV opening setpoint pressure. The most severe Group B sequence was 9-22, involving five stuck-open SDVs with failures to The most severe Group B sequence was 9-22, throttle AFW and charging. The minimum temper, inv Iving five stuck-open SDVs with failures to throttle AFW and charging. The mmimum temper-
-. ature of the reactor vessel downcomer for th,s i ature of the reactor vessel downcomer for this sequence was 364 K (196*F); the maximum subsc- sequence was 364 K (196 F); the maximum subse-quent pressure was 16.35 MPa (2371 psia). quent pressure was 16.35 MPa (2371 psia).
e 55
- 6. GROUP C RESULTS: SECONDARY-SIDE BREAKS AFFECTING TWO STEAM GENERATORS, OR THREE ASYMMETRICALLY AFFECTED STEAM GENERATORS The following sections present the results of secondary pressures are below 0.69 MPa (100 psia). ,
thermal-hydraulic analyses for sequences in Below this pressure insufficient steam is available Group C of Table 15 using the methods presented to run the turbine-driven AFW pumps. in Section 3 to determine reactor vessel downcomer ~ fluid pressure and temperature histories. Section 6.1 defines the group and describes the controlling 6.2 Combinations of phenomena. Section 6.2 discusses the results for Stuck-Open Steam PORVs sequences containing combinations of stuck-open and Steam Dump Valves steam PORV and steam dump valves. Section 6.3 discusses results for sequences containing two stuck- Sequences 5-14,5-15,5-17 through 5-20, and 7-9 open steam PORVs. Conclusions covering all through 7-11 involve combinations of steam PORV group C sequences appear in Section 6.4. T and steam dump valve (SDV) breaks (see Tables 5 facilitate referencing of data, plotted results, show- and 7). Sequences from Table 5 begin from full-ing the pressure and temperature histories, are power conditions, and sequences from Table 7 organized in numerical sequence order in begin from hot standby conditions. The sequences Appendix A rather than appearing within this sec- further differ by: (1) AFW isolation at 10 min, tion. Due to the large number of sequences in- (2) AFW feed rate (normal or overfeed), and vestigated in this report, detailed discussions of (3) throttling of charging flow when the pressurizer thermal-hydraulic processes for each sequence are setpoint level is attained. Each of the sequences not practical. Such discussions are documented for involves one stuck-open SDV and either one or two representative sequences in Reference 1. stuck-open steam PORVs. 6.1 Group C Definitions Table 27 presents the analysis methods used, and gives an indication of results for, sequences involv-ing combination of SDV and steam PORV breaks. ~ Group C includes all sequences controlled by Plotted results for all sequences appear in multiple steam line breaks that asymmetrically Appendix A. affect the steam generators. This includes two categories: (1) sequences involving one or more A detailed model calculation was performed for stuck-open steam PORVs and one stuck-open steam one sequence at each power level: Sequences 5-14, dump valve (SDV), and (2) sequences involving two 5-19,7-9 and 7-11. Simplified model calculations stuck-open steam PORVs. In the first category, the were performed, using the base simplified model open SDV affects all steam generators the same, but described in Section 3.3.2. an open PORV affects only the steam generator on which it is located. In the second category, the steam The base simplified model contains a single generator without the open PORV is not affected, secondary volume that represents three sym-due to the action of the steam line check valves. metrically affected steam generators. By using this model, it is assumed that the asymmetries in these Sequences in Group C are controlled by primary sequences may be ignored. This approach was system heat removal to each affected steam found to adequately model the phenomena present generator and heat addition from steam generators in these sequences because the total primary system in flowing unaffected loops, core decay heat, and heat removal rates calculated using the detailed and , stored energy in metal heat structures. simplified models agreed to within 5%. The simplified model was initialized based on conditions For all Group C sequences, turbine-driven AFW present at the end of the detailed model calculation: is initiated because the levels in at least two steam the secondary system mass was conserved, and the generators decrease sufficiently, due to loss of secondary pressure wa* the average of the three secondary inventory. Turbine-driven AFW is steam generator pressures which had been assumed unavailable when all steam generator calculated with the detailed model. 56
I Table 27. Results of combination SDV and steam PORV breaks Detailed Simplified Afodel hiodel Sequence Initiating Power Calculation Variation Used
- Number Event Level Failures Used (from 3.3.3) 5 14 Small St.B Full I stuck-open SDV 5-14 Base 5-15 Small SLB Full I stuck-open SDV, charging not throttled 5 14 Base . 5-17 Small SLB Full I stuck-open SDV, AFW not isolated 5-14 Ease 5-18 Small SLB Full I stuck open SDV, AFW overfeed 5-14 Base 5-19 Small SLB Full 2 stuck-open SDVs 5 19 Base 5-20 Small SLB Full 2 stuck-open SDVs, AIM not isolated 5-19 Base 7-9 Small SLB 115B i stuck-open SDV 7-9 Base 7-10 Small SLB IISB 1 stuck-open SDV, AFW not isolated 7-9 Base 7 11 Small SLB flSB 2 stuck-open SDVs 7 11 Base blaximum hfethods Used, Periods in Seconds hiinimum Subsequent Figures RV DC RV DC Showing Sequence Detailed Simplified lland Temperature Pressure Plotted ,
Number hfodel Afodel Calculations ('F) (psia) Results Notesa 5 14 0 - 600 600 4400 4400 - 7200 311 2371 A-91. A-92 A 5 15 0 - 600 600 4000 4000 - 7200 309 2371 A-93, A-94 A 5-17 0 - 600 600 7200 - 277 1553 A-97. A-98 5-18 o 600 600 - 5400 5400 - 7200 303 2371 A-99, A-100 A 5-19 - 600 600 3400 3400 - 7200 286 2371 A-101. A-102 A 5-20 0-6W 600 7200 - 244 1618 A-103, A-104 7-9 0-6N 600 - 7200 - 239 1508 A-147 A 148 7 10 0 - 600 600 7200 - 229 1470 A 149, A 150 7-11 0-60 600 7200 - 209 1537 A 151 A-152 e a- A-Iland calculations performed over heatup phase following steam generator dryout-System responses fell into two categories, Due to the low decay heat and large break size, depending on AFW isolation. The first category, Sequence 7-11 (two steam PORVs and one SDV sequences that include AFW isolation at 10 min, open at hot standby) reached the lowest reactor proceeded through a cooldown phase, followed by vessel downcomer temperature for the group,371 K a steam generator dryout phase, and finally a (209'F). The primary system repressurization for heatup phase. For convenience, 'the responses this (and other hot standby) sequences was limited during the heatup phases were hand calculated because decay heat was insufficient to overcome the based on the elfeet of decay heat on the primary steam generator heat removal. Thus, the primary system liquid and metal masses. For sequences in Guld was not heated; thermal expansion of the fluid this category, AFW restoration will eventually be was avoided; and only a slight repressurization was required (after the two hour period of interest) for encountered, long-term plant cooling. For first-category sequences initiated from full power (5-14, 5-15, Sequences 5-14,5-15,5-18, and 5-19 encountered 5-18, and 5-19), steam generator dryout was heatup phases, and the resulting primary fluid calculated before two hours. For first-category thermal expansion was sufficient to force the sequences initiated from hot standby (7-9 and 7-11), primary system pressure to the pressurizer PORV steam generator dryout did not occur before two opening setpoint pressure. Of these sequences,5-19 hours, so the heatup phase was not entered. (two steam PORVs and one SDV open at full i power) had the lowest reactor vessel downcomer l
- The second category, sequences where AFW is fluid temperature,414 K (286*F).
j not isolated at 10 min (5-17, 5-20, and 7-10), proceeded through a continuous primary system The severity of sequences in this group is limited cooldown. The cooldown rate slows considerably because significant primary:ystem repressurization I as AFW throttling begins w hen the steam generator requires AFW termination, and this action narrow-range levels have recovered to 40%. terminates the cooldown The sequences would be l 57
made much more severe if AFW isolation were assuming no interaction between it and the primary delayed until 20 min, for example, thus allowing an system. Next the primary system temperature extra 10 min of cooldown before the severe response was determined by taking a mass-weighted repressurization. average of the USG temperature response and the primary system temperature response, assuming no 6.3 Two Stuck-Open USG efrects (note for Sequence 9-34 the latter , response is available from the Sequence 9-33 Steam PORVs simplified model calculation, etc.). The primary system pressure response was then calculated, using Sequences 9-33 through 9-40 involve two stuck- the primary system fluid thermal contraction rate. open steam PORVs. All of these sequences are HPI and charging volume addition rates, and an initiated from full-power conditions; AFW is not adiabatic compression / expansion model of the terminated by the operator at 10 min. The pressurizer. sequences differ by:(1) AFW delivery rate (normal or overfeed), (2) throttling of AFW to the Results for all af these sequences indicate a con-unaffected steam generator (USG) at 40% narrow- tinuous cooldown of the primary system until AFW range level or when liquid spills to the steam lines, throttling to the ASG occurs. After throttling, the and (3) throttling of charging flow when the temperature responses are virtually flat. The effect pressurizer setpoint level is attained. For these of overfilling the USG was found to be minor in sequences, AFW can be fed to the unaffected steam all cases. generators only when AFW is throttled to the ASGs; in all these sequences this throttling occurs Minimum temperatures for all sequences were when liquid spills from the ASG to the steam line. within 5 K (9'F) of 416 K (289*F). The more severe Due to the low ASG pressures, all AFW flows from sequences were 9-35,9-36,9-39, and 9-40 in which the headers to the ASGs, unless the AFW valves to the failure to throttle charging caused the rimary the ASGs are closed, in which case AFW can be system pressure to rise to the pressu. 'e PORV delivered to the USG. opening setpoint pressure. Table 28 presents the analysis methods used, and gives an indication of results for, sequences involv-6.4 Conclusions ing two stuck-open steam PORVs. Plotted results for all sequences appear in Appendix A. Group C includes 17 sequences involving only steam breaks asymmetrically affecting the steam A detailed model calculation was performed for generators. the initial portion of Sequence 9-33. Simplified model calculations were performed using model For sequences involving three asymmetrically Variation 6, as described in Section 3.3.3. affected steam generators, representing the phenomena with a single, simplified model For sequences not involving AFW overfilling of secondary cell, usicg average conditions, was found the USG (9-33,9-35,9-37, and 9-39), it was found to be a satisfactory method. that the USG was not sufficiently cooled to cause heat removal from the primary system to the USG. For sequences that specify AFW isolation at in these sequences, the USG response does not af- 10 min, AFW restoration will eventually be required feet the primary system response. The simplified to maintain long-term cooling. The restoration model results were used to two hours. would occur after the two-hour period of interest for PTS. For these sequences, thermal-hydraulic For sequences involving AFW overfilling of the severity would be increased significantly if AFW , USG (9-34, 9-36, 9-38, and 9-40), the USG was isolation were delayed, for example, to 20 min. found to affect the primary system response. For these sequences, hand calculations w ere performed, The sequences initiated from hot standby condi-from the time of AFW throttling to the ASG to two tions were found to lead to colder reactor coolant hours after the initiating event, to predict the temperatures than those initiated from full-power primary system response. The hand cale' hns conditions. For these sequences, however, repres-first determined the USG temperature rnww, surization of the primary system is limited because 58
decay heat is insufficient to overcome steam to throttle charging flow, the PTS severity of these generator heat removal. Thus the primary system sequences would be increased significantly. fluid is not heated, does not expand, and does not compress the pressurizer steam bubble which would The effects on the results of failure to throttle increase the pressure. Had one or more of the hot AFW to USGs and AFW overfeed (at a higher rate) , standby sequences in this group specified a failure were found to be minor. Table 28. Results of two stuck-open steam PORVs Detailed Simplified hiodel h1odel Sequence Initiating Power Calculation Variation Used Number Event Level Failures Used (from 3.3.3) 9-33 Reactor trip Full 2 stuck-open SPORVs 9-33 6 9-34 Reactor trip Full 2 stuck-open SPORVs, AFW overfill 9-33 6 9-35 Reactor trip Full 2 stuck-open SPORVs, charging not throttled 9-33 6 9-36 Reactor trip Full 2 stuck-open SPORVs, AFW oserfill, charging 9-33 6 not throttled 9-37 Reactor trip Full 2 stuck-open SPORVs, AFW oserfeed 9-33 6 9-38 Reactor trip Full 2 stuck-open SPORVs, AFW overfeed, AFW 9-33 6 overfill 9-39 Reactor trip Full 2 stuck-open SPORVs, AFW overfeed, charging 9-33 6 not throttled 9-40 Reactor trip Full 2 stuck-open SPORVs, AFW overfeed, AFW overfill, charging not throttled 9-33 6 hlaximum hiethods Used, Periods in Seconds Minimum Subsequent Figures RV DC RV DC Showing Sequence Detailed Simplified Iland Temperature Pressure Plotted Number Model Model Calculations (*F) (psia) Results Notes a 9-33 0 - 665 665 7200 - 295 1833 A-231, A-232 9-34 0 - 665 665 - 3748 3748 - 7200 287 1782 A 233, A-234 A 9-35 0 - 665 665 - 7200 - 285 2371 A-235, A-236 9-36 0 - 665 665 - 3G49 3049 - 7200 280 2371 A-237, A-238 A 9-37 0 - 665 665 - 7200 - 298 1823 A-239, A-240 9-38 0 - 665 665 - 2337 2337 - 7200 290 1816 A-241, A-242 A 9-39 0 - 665 665 -7200 - 289 2371 A-243, A-244 9-40 0 - 665 665 - 2340 2340 - 7200 283 2371 A 245, A-246 A
- a. A-Iland calculation performed for period following AFW throttling to the ASGs.
4 59 w
- 7. GROUP D RESULTS: REACTOR TRIP FROM FULL POWER WITH NO PRIMARY- OR SECONDARY-SIDE BREAKS The following sections present the results of Table 29 presents the analysis methods used, and thermal-hydraulic analyses for sequences in gives an indication of results for, the Group D -
Group D of Table 15, using the methods presented sequences. Plotted results for all sequences appear in Section 3 to determine reactor vessel downcomer in Appendix A. fluid pressure and temperature histories. Section 7.1 . defines the group; Section 7.2 presents the results A detailed model calculation was performed for and conclusions for sequences in Group D. To the initial portion of Sequence 9-1, a reactor trip facilitate referencing of data, plotted results show- from full power. By 900 s this sequence reaches a ing the pressure and temperature histories are steady, post-trip condition expected to continue organized in numerical sequence in Appendix A through 7200 s. rather than appearing within this section. Due to the large number of sequences investigated in this The hand calculations for Sequences 9-49 report, detailed discussions of thermal-hydraulic through 9-55 were used to modify the results processes for each sequence are not practical. Such calculated for Sequence 9-1 for effects of AFW discussions are documented for representative overfill, AFW overfeed, and failure to throttle sequences in Reference 1. charging. 7.1 Group D Definition For AFW overfill, a new steam generator secondary temperature response was calculated, assuming no heat addition from the primary system. Group D includes sequences not controlled by Next, a new primary system temperature response primary- or secondary-side breaks. These sequences was calculated by taking a mass-weighted average are initiated by reactor trip from full-power condi- of the primacy system temperatures from , tions, with only minor failures within the auxiliary Sequence 9-1 and the new secondary temperatures. feedwater (AFW) or pressurizer level control Since the charging system flow capacity is sufficient systems. These sequences are controlled by near- to make up for the primary system fluid shrinkage , normal decay heat removal to the steam generators rate caused by the cooldown, the primary system including continued reactor coolant pump opera- pressure response from Sequence 9-1 remains tion. Sequences 9-1, and 9-49 through 9-55 are applicable. included in Group D. For AFW overfeed, the results of Sequence 9-1 7.2 Results and Conclusions " " " di"'d ' "" ""' I r the faster AFW delivery rate. Th.is modification affected only the i timing of events. Sequences 9-1 and 9-49 through 9-55 are initiated r by a reactor trip from full-power conditions. The For the failure to throttle charging, the primary j sequences differ by: (1) the AFW delivery rate (nor- system pressure response was modified, based on i mal or overfeed), (2) AFW throttling at 40% steam an adiabatic compression of the pressurizer steam generator level, or when liquid spillover to the steam bubble. The pressurizer inflow was assumed to be lines occurs, and (3) throttling of charging flow the net makeup flow, i.e., charging Dow minus let-when the pressurizer setpoint level is attained. down flow. As indicated in Table 9, Sequences 9-49 through All Group D sequences are oflimited severity for . 9-55 include one main feedwater regulating valve PTS. Only sequences involving an AFW overfill (NIFWRV) failing open. This failure is inconsequen- (9-49,9-51,9-53, and 9-55) result in a cooldown. tial since the main feedwater isolation valves The minimum primary system temperature for these , (N1FWlVs) close automatically following a turbine sequences is 503 K (446'F). Primary system trip. The NIFWIVs are in series with the NIFWRVs, pressures rise to the pressurizer PORV opening set-so that the closure of either valve effectively isolates point pressure, except in Sequences 9-1 and 9-52, main feedwater to a steam generator. where they remain at the normal operating point. 60
1 Table' 29. Results of reactor trip sequences with minor failures Detailed Simplified
- Model Model Sequence initiating ' Power Calcu% tion Variation Used
. Number Event Level Failures Used (from 3 3.3) 9-1 Reactor trip Full ~ None 9-1 -
9-49 ' Reactor trip Full AFW overfill 9-1 -
., 9-50 Reactor trip Full Charging not throttled . 9-1 -
9-51 Reactor trip - Full AFW overfill, charging not throttled 9-1 9-52 Reactor trip Full AFW overfeed 9-1 - 9-53 Reactor trip Full AFW overfeed, AFW overfill 9-1 -- 9-54 Reactor trip Full . AFW overfeed, charging not throttled 9-1 - 9-55 Reactor trip Full AFW overfeed, AFW overfill, charging not 9-1 - throttled Maximum Methods Used. Periods in Seconds Minimum Subsequent Figures RV DC RV DC Showing Sequence Detailed Simplified Hand Temperature Pressure Plotted Number Model Model Calculations (* F) (psia) Resuhs Notes a 9-1 0 - 900 - 900 - 7200 54 2296 A-167. A-168 A 9-49 0 - 900 - 900 - 7200 44 2371 A-263, A-264 B 9-50 0 900- - 900 - 7200 546 2371 A-265, A-266 B 9-51 0 - 900 - 900 - 7200 446 2371 A-267 A-268 B 9-52 0 - 900 - 900 - 7200 54 2296 A-269, A-270 B 9-53 0 - 900 - 900 - 7200 446 2371 A-271. A-272 B , 9-54 0 900 - 900 - 7200 54 2371 A-273 A-274 B l 9-55 0-900 - 900 - 7200 446 2371 A-275 A-276 B l j - a. A-Steady detailed model results at 900 s extrapolated to 7200 s-B-Pressure and temperature responses calculated by hand based on detailed model results and effects of the AFW or charging failures. Specifics of these hand calculations appear in the text. i e 9. l 61
- 8. GROUP E RESULTS: MAIN FEEDWATER OVERFILL The following sections present the results of 8.2 Results and Conclusions thermal-hydraulic analyses for the sequence in Group E of Table 15 using the methods presented in Section 3 to determine reactor vessel downcomer Table 30 presents the analysis methods used, and .
fluid pressure and temperature history. Section 8.1 gives an indication of results for, Sequence 9-56. defines the group, and Section 8.2 presents the Plotted results for this sequence appear as results and conclusions for the sequence in Figures A-277 and A-278 in Appendix A. , Group E. To facilitate referencing of data, plotted results showing the pressure and temperature A detailed model calculation was run for the histories are organized in numerical sequence in initial 440 s of Sequence 9-56. The affected steam Appendix A, rather than appearing within this generator level reached its high level setpoint at 60 s, section, at which time main feedwater pumps were tripped. After 440 s, the primary system pressure response assune charging How continues untH the nonnal 8.1 Group E Definition pressurizer level is attained. Afterwards, the pressurizer spray and heaters control the pressure Group E includes only Sequence 9-56 (see normally. The primary system temperature response Table 9). This sequence is initiated by a reactor trip assumes that the minor cooldown, underway at from full power, followed by failing open of the 440 s, continues until 559 K (547"F) is reached. main feedwater regulating and isolation valves on Thereafter, the steam dump system is assumed to one feed line, and a failure to trip main feedwater maintain that temperature. pumps until high steam generator level is reached. Sequence 9-56 is of very low PTS scverity. Sequence 9-56 was considered by itself as a Significant cooling capability is not present during separate group because it is the only sequence from a main feedwater overfill transient because main . Tables 1-12 that involves the overfilling of the steam feedwater is warm, relative to AFW, and because generators using the main feedwater system. the main feedwater pump trip is reached quickly. Table 30. Results of main feedwater overfill Detailed Simplified N!odel h1odel Sequence Initia:ing Power Calculation Variation Used Number Event Lesel Failures Used (from 3.3.3) 9-56 Reactor trip Full NtFWRV and 51FWIV fail open on same line, 9-56 - N1FW pumps do not trip after turbine trip blatimum hiethods Used. Periods in Seconds hiinimum Subsequent Figures RV DC RV DC Showing Sequence Detailed Simplified lland Temperature Pressure Plotted Number htodel htodel Calculations ('F) (psia) Results 9-56 0 - 440 - 440 7200 544 22 % A-277, A-278 62
- 9. GROUP F RESULTS: LOCA
^
The following sections present the results of Detailed model calculations were performed for thermal-hydraulic analyses for sequences in initial portions of Sequences 1-1 (see Scenario 7 in Group F of Table 15, using the methods presented Reference 1) and 3-1. The two detailed model - in Section 3 to determine reactor vessel downcomer calculations provided a starting point for the fluid pressure and temperature histories. Section 9.1 simplified model, which was then used to calculate defines the group and describes the controlling phe- the remaining portions of the sequences. For each - nomena. Within Group F, sequences werc organiz- sequence, Table 31 shows the time periods ed by break size into Small Break Loss-of-Coolant calculated with the detailed model and with the Accident (SBLOCA) and Medium Break Loss-of- simplified model. Variation 7 of the simplified Coolant Accident (MBLOCA) subgroups. model, which is described in Section 3.3.3, was used Sections 9.2 and 9.3 present the results for these for all the SBLOCA sequences. subgroups. Conclusions covering all Group F sequences appear in Section 9.4. To facilitate Both motor-driven and turbine-driven AFW were referencing of data, plotted results showing the initiated in the SBLOCA sequences at full power pressure and temperature histories are organized in (Sequences 1-1 through 1-4). Only motor-driven numerical sequence order in Appendix A, rather AFW was initiated in the sequences at hot standby than appearing within this section. Due to the large (Sequences 3-1 and 3-2). The turbine-driven AFW number of sequences investigated in this report, was not initiated at hot standby because the large detailed discussions of thermal-hydraulic processes initial secondary mass and the small steam produc-for each sequence are not practical. Such discus- tion prevented the steam generator levels from sions are documented for representative sequences decreasing enough to actuate the turbine-driven in Reference 1. system. e 9.1 Group F Definition . B '.h the simplified and detailed model calcula-tions mdicated that the ECC How was generally as large as, or exceeded, the How through the PORV. Group F includes all sequences controlled by a Consequently, the loops did not drain, the U-tubes . primary-side break that is not isolated. Within this were not voided, and natural circulation flow was section, a "small" break refers to a single stuck-open maintained. Eventually the pressurizer filled with pressurizer PORV. A " medium" break refers to a subcooled liquid, causing the flow out the PORV 0.0635-m (2.5-in.) diameter hole located at the bot- to increase and the RCS to depressurize. The energy tom of the C loop hot leg. The controlling phenom- removed by the combination of ECC and flow ena are the mass and energy flow due to the break through the PORV eventually exceeded the decreas-and the ECC, possible draining of the U-tubes which ing core decay power. Consequently, the steam stops natural circulation, and the heat transfer to, or generators Secame heat sources rather than heat from, the steam generators. sinks. The steam generators became heat sources at 2000 s in Sequence 1-1 (SBLOCA at full power) and at 3 s in Sequence 3-1 (SKOCA at hot 9.2 SBLOCAS standby). Even after the steam generators became heat sources, natural circulation continued in the Sequences 1-1 through I-4 involve SBLOCAs at liquid-filled loops because of the effects of core full power (see Table 1). Sequences 3-1 and 3-2 decay power, ECC, and loop transient times. At hot involve SBLOCAs at hot standby (see Table 3). The standby (Sequence 3-1), the natural circulation How sequences differ based on: (1) initial power level, was about twice the HPI How. (2) AFW Howing at normal or high (overfeed) rates, and (3) AFW throttled or not throttled (overfill) at The dow neomer pressure and temperature were 40% narrow range level. significantly affected by the initial core power level, e as shown in Table 31. As expected, the calculated Table 31 presents the analysis methods used, and downcomer pressures and temperatures at het gives an indication of the results of, the SBLOCA standby were lower than at full power. The sequences. Downcomer pressure and temperature minimum downcomer temperatures were from results for all sequences are presented in 76 to 108 K (136 to 195*F) lower at hot standby; Appendix A. the final pressures were about 1.4 MPa (200 psi) 63
. __ __ _ . _ _ . . . _ _ _ . ~ . . _
Table 31. Results: SBLOCA s Detailed Simplified Afodel h!odel Sequence Initiating Power Calculation Variation Used Number Event Level Failures Used (from 3.3.3) = l-1 SBLOCA Full None 11-1 7 1-2 SBLOCA Full AFW overfill 11 1 7 I.3 SBLOCA Full AFW overfeed Il-1 7 - I-4 SBLOCA Fut! AFW overfeed, AFW overfill 11-1 7 3-1 SBLOCA HSB None 3-1 7 3-2 SBLOCA HSB AFW overfill 3-1 7 htaximum hiethods Used. Periods in Seconds hiinimum Subsequent Figures RV DC RV DC Showing Sequence Detailed Simplified Hand Temperature Pressure Plotted Number hiodel h1odel Calculations ('F) (psia) Results Notesa 1-1 0 - 600 600 - 7200 - 474 924 A-I, A-2 12 0 - 600 600 - 7200 - 112 578 A-3, A-4 l3 0 - 150 150 - 7200 - 4 73 919 A-5, A-6 1-4 0 150 150-7200 - 381 576 A-7, A-8 31 0 - 790 790 - 7200 - 279 679 A-59, A-60 A 3-2 0 - 400 400 7200 - 246 393 A-61, A-62 A
- a. A-Turbine-driven AFW not initiated, lower at hot standby. AFW overfill also signif- sequences. Downcomer pressure and temperature .
icantly affected the downcomer pressure and results for all sequences are presented in temperatures. The final downcomer temperatures Appendix A. and pressures were from 18 to 51 K (33 to 92*F), . and about 2 MPa (300 psi) lower when AFW was Detailed model calculations were performed for not throttled at 40% level. However, the AFW the initial portions of the MBLOCA at full power overfeed did not significantly affect either the with no other failures (Sequence 2-1), the downcomer temperature or pressure at two hours. MBLOCA at full power with AFW overfill (Sequence 2-2), and the MBLOCA at hot standby The SBLOCA sequences resulted in relatively low (Sequence 4-1). These sequences were extrapolated downcomer pressures and high downcomer from the end of the detailed model calculation to temperatures. Consequently, the SBLOCA two hours using hand calculations. The thermal-sequences are probably not severe PTS transients. hydraulic conditions at two hours were based on steady state mass and energy balances. At steady state, the loops were assumed to be stagnated. The 9*3 MBLOCAS ECC was assumed to flow into the downcomer, past the core, into the C loop hot leg, and out through Sequences 2-1 through 2-4 (see Table 2) involve the break. The break thus removed the heated ECC MBLOCAs ' at full power. Sequence 4-1 (see and consequently the core decay power. The results Table 4) involves a MBLOCA at hot standby. The of Sequences 2-3 and 2-4 were based primarily on sequences differ based on: (1) initial power level, engineering judgment relative to the effects of AFW . (2) AFW flowing at normal or high (overfeed) rates, overfeed. The effect of AFW overfeed was to cool and (3) AFW throttled or not throttled (overfill) at and depressurize the primary system more rapidly 40% narrow range level. during the overfeed period, but the effect was , assumed to diminish after AFW termination. Thus, Table 32 presents the analysis methods used, and the results of Sequence 2-3 (AFW overfeed) even-gives an indication of the results for, the MBLOCA tually converged with those of Sequence 2-1 64
Table 32. Results: MBLOCA s Detailed Simplified Model Model Sequence Initiating Power Calculation Variation Used , Number Event Level Failures Used (from 3.3.3) 2-1 MBLOCA Full None 2-1 - 2-2 MBLOCA Full AFW overfill 2-2 - 2-3 MBLOCA Full AFW overfeed 2-1 - 2-4 MBLOCA Full AFW overfeed, AFW overfill 2-1 - 4-1 MBLOCA IISB None 4-1 - Maximum Methods Used, Periods in Seconds Minimum Subsequent Figures RV DC RV DC Showing Sequence Detailed Simplified lland Temperature Pressure Plotted Number Model Model Calculations (*F) (psia) Results Notes a 2-1 0 - 2800 - 280J - 7200 100 144 A 37, A-38 A 2-2 0 - 1730 - 1730 - 7200 100 144 A-39, A-40 A 2-3 0 - 200 - 200 - 7200 100 144 A-41, A 42 B 2-4 0 - 200 - 200 - 7200 100 144 A-43, A-44 B 4-1 0 - 1740 - 1740 - 7200 100 144 A43, A44 A,C
- a. A-Iland calculations based on extrapolations of detailed model results and steady state mass and energy balances.
B-Iland calculations based on engineering judgment and steady state mass and energy balances. C-Turbine-driven AFW not initiated. (normal AFW dow). Similarly, the results of power were judged to be more severe relative to PTS Sequence 2-4 (AFW overfeed and overfill) even- than at hot standby because the RCS depressurized tually converged with those of Sequence 2-2 (AFW more slowly at full power.
- overfill). The simplified model was not used to calculate any of the MBLOCA sequences shown in Table 32 because it was not thought to be adequate
9.4 CONCLUSION
S for these sequences. The break was big enough to cause voiding in the loops, but the simplified model The SBLOCA sequences are probably not severe could not calculate a mixture level within the RCS relative to PTS. The MBLOCA sequences are more and hence could not reliably calculate the void severe f r PTS because the downcomer temper-fraction or flow at the break. atures dropped nearly to the ECC temperature. The MBLOCA was large enough to cause voiding However, the downcomer pressures were relatively in the loops, depressurize the RCS, and eventually low for the MBLOCA sequences. remove the core decay power. Loop flows stagnated
- due to voiding in the U-tubes. The final conditions Both the small and medium breaks were large for all the sequences were the same. The final enough to remove more energy than was generated downcomer temperature was 311 K (100*F), slightly by core decay heat within two hours. Consequently, higher than the ECC temperature, due to bypass the RCS could be cooled and depressurized with flows within the vessel. The final downcomer either break size. The thermal-hydraulic results were pressure was 1.0 MPa (144 psia), slightly below the sensitive to AFW overfill, but not to AFW over-LPI shutoff head. The effect of AFW was to feed. Loop flow stagnation was calculated to occur
-depressurize and cool the RCS and to promote only when the U-tubes were voided. The medium
- natural circulation flow. Consequently, AFW over- break was large enough to void the U-tubes but the fill significantly affected the pressure and small break was not. Loop flow stagnation did not temperature histories. The effect of AFW overfeed occur w hen the loops were liquid full, even with heat was judged to be insignificant. The sequences at full transfer from the steam generators to the RCS.
65
- 10. GROUP G RESULTS: LOCA COMBINED WITH SYMMETRIC SECONDARY-SIDE BREAKS The following sections present the results of Table 33 summarizes the analysis methods used thermal-hydraulic analyses for sequences in and gives the results of the LOCA sequences com-Group G of Table 15, using the methods presented bined with stuck-open SDVs. Downcomer pressure in Secti<m 3 to determine reactor vessel downcomer and temperature results for all sequences are fluid pressure and temperature histories. Sec- presented in Appendix A.
tion 10.1 defines the group and describes the con- - trolling phenomena. Section 10.2 presents the Detailed model calculations were performed for results for all the sequences in this group. Conclu- initial portions of Sequences 1-5,1-8, 2-5, and 2-8. sions related to Group G sequences appear in These sequences represented SBLOCAs and Section 10.3. To facilitate referencing of data, h1BLOCAs with either one or five stuck-open SDVs plotted results showing the pressure and temper- which bounded the possible secondary break sizes, ature histories are organized in numerical sequence The detailed model calculations provided a starting order in Appendix A, rather than appearing within point for the simplified model, which was used to this section. Due to the large number of sequences calculate the bounding sequences to two hours. investigated in this report, detailed discussions of Variation 7 of the simplified model was used for thermal-hydraulic processes for each sequence are the SBLOCA sequences, and Variation 8 was used not practical. Such discussions are documented for for the A1BLOCA sequences. These models were representative sequences in Reference 1. Note that described in Section 3.3.3. Results for automatic closure of the main steam isolation valves Sequences 1-6,1-7,2-6, and 2-7, which represented was not demanded for sequences in this group. LOCAs with either two or three stuck-open SDVs, Reasons for this are discussed in Section 5.3. were obtained by interpolating the results of the appropriate boundmg sequences. For example,
""hs r r Sequence 1-7 were obtained by inter-10.1 Group G Definition .
polatmg Sequences 1-5 and 1-8, and Sequence 2-7 . was interpolated from Sequences 2-5 and 2-8. Group G includes all sequences controlled by a Sequence 3-3, a SBLOCA at hot standby with one primary-side break, combined with a secondary-side stuck-open SDV was not analyzed because of . break that symmetrically affected all three steam probabilistic considerations. SDVs are only occa-generators. A SBLOCA refers to a LOCA initiated sionally demanded to open at hot standby. Once by a single, stuck-open pressurizer PORV. A the PORV is opened, it can remove core decay h1BLOCA refers to a LOCA initiated by a 0.0635-m power, and the SDVs would not be demanded to (2.5-in.) diameter hole located at the bottom of the open. The probability that both the pressurizer C loop hot leg. The secondary-side break was PORV and a SDV would stick open at hot standby initiated when a given number of steam dump valves was thought to be relatively low. Consequently, (SDVs) stuck open as they were demanded to open Sequence 3-3 was not analyzed, and results for this by the plant control system following reactor trip. sequence do not appear in either Table 33 or The controlling phenomena were the mass and Appendix A. energy flows due to the primary-side break and ECC, and heat transfer to the ste'am generators. The calculations showed that significant voiding of the RCS loops did not occur for the combina-ti ns f primary- and secondary-side breaks. Even 10.2 ResultS m the h1BLOCA sequences, only minor voiding occurred in the loops early in the transients. The Sequences involving SBLOCAs combined with loops filled with subcooled liquid due to the com- - stuck-open SDVs are numbered 1-5 through 1-8 bined effects of ECC and the steam line break. With (see Table 1) for full power and 3-3 (see Table 3) the loops liquid-full, natural circulation was main-for hot standby. Sequences involving N1BLOCAs tained throughout the transient. In the SBLOCA . combined with stuck-open SDVs are numbered 2-5 sequences, a pressure plateau was reached near through 2-8 (see Table 2). The sequences differ 5.5 51Pa (800 psia) as the pressurizer filled. The based on primary-side break size, number of stuck- RCS then depressurized relatively rapidly until open SDVs, and initial core power level. accumulator flow was initiated, at which time the 66
Table 33. Results of LOCAs combined with symmetric secondary-side breaks 4 Detailed Simplified Model Model Sequence Initiating Power Calculation Variation Used , Number Event Level Failures Used (from 3.3.3) 1-5 SBLOCA Full I stuck-open SDV l-5 7 14 SBLOCA Full 2 stuck-open SDVs - - 17 SBLOCA Full 3 stuck-open SDVs - - l-8 SBLOCA Full 5 stuck-open SDVs 1-8 7 2-5 MBLOCA Full I stuck-open SDV 2-5 8 24 MBLOCA Full 2 stuck-open SDVs - - 2-7 MBLOCA Full 3 stuck-open SDVs - - 2-8 MBLOCA Full 5 stuck-open SDVs 2-8 8 3-3 SBLOCA IISB 1 stuck-open SDV - - Maximum Methods Used Periods in Seconds Minimum Subsequent Figures RV DC RV DC Showing Sequence Detailed Simplified Hand Temperature Pressure Plotted Number Model Model Calculations (* F) (psia) Results Notes a I-5 0 - 750 750 - 7200 - 233 338 A-9, A 10
-14 - -
0 - 7200 218 328 A-II, A-12 A l-7 - - 0 - 7200 203 317 A-13, A-14 A l-8 0 350 350 - 7200 - 171 300 A-15. A-16 2-5 0 - 550 550 - 7200 - 210 144 A-45, A-46 24 - - 0 - 7200 198 144 A-47, A-48 A 2-7 - - 0 - 7200 185 I44 A-49, A-50 A 2-8 0 - 540 540 - 7200 - 160 144 A-51, A-52 3-3 - - - - - - B
- a. A-Results interpolated from the sequences with I stuck-open SDV and 5 stuck-open SDVs.
B-Seque. ice deleted because of probability considerations. rate of depressurization slowed. In the MBLOCA downcomer pressures in the combination LOCA sequences, the pressure dropped relatively quickly and steam line break sequences were more than to the LPI shutoff head and then remained con- 8 MPa (1200 psi) lower than the steam line break stant. The downcomer temperatures were controlled of the same size. Because the pressures were much by the heat transfer to the steam generators and by less, the combination LOCA and steam line break the ECC mixing. The cooldown rate always sequences tre not as potentially severe PTS tran-decreased when AFW was throttled. Table 33 shows sients as th: steam-line-break-only sequences. that the minimum downcomer temperatures decreased as the break size on either the primary-or secondary-side rnereased. All the sequences were 10.3 Conclusions at or near steady state at two hours. The sequences with combined LOCAs and steam The sequences with combinations of LOCAs and line breaks were probably not as severe relative to steam line breaks had temperature histories that PTS as the corresponding steam line break tran-were similar to the corresponding steam line break, sients described in Section 5. 1 - while the pressure histories were more similar to the corresponding LOCA. Because of the additional The sequences with combined LOCAs and steam i cooling due to the LOCA, the final downcomer line breaks had temperature histories that were , temperatures in the combination LOCA and steam similar to a steam line break, while the pressure line break sequences were colder by 17 to 56 K histories were more similar to those of a LOCA. (30 to 100*F) than in the steam line break of the same size described in Section 5. Because of the Natural circulation was maintained in these depressurization due to the LOCA, the final sequences because the RCS loops did not void. l l l 67 l
- 11. GROUP H RESULTS: LOCA COMBINED WITH ASYMMETRIC SECONDARY-SIDE BREAKS The following sections present the results of Table 34 summarizes the analysis methods used, thermal-hydraulic analyses for sequences in and gives an indication of the results for, the LOCA Group H of Table 15, using the methods presented sequences combined with stuck-open steam PORVs. -
in Section 3 to determine reactor sessel downcomer Downcomer pressure and temperature results for fluid pressure and temperature histories. Sec- all sequences are presented in Appendix A. tion 11.1 defines the group and describes the con- . trolling phenomena. Section i1.2 presents the Detailed model calculations were performed for results for all the sequences in this group. Con- initial portions of Sequences 1-9, 1-12, 2-9, clusions related to Group H sequences appear in and 2-11. These sequences represented SBLOCAs Section i1.3. To facilitate referencing of data, and h1BLOCAs with either one or two stuck-open plotted results showing the pressure and steam PORVs. The detailed model calculations temperature histories are organized in numerical provided a starting point for the simplified model, sequence order in Appendix A rather than appear- which was then used to calculate all the Group H ing within this section. Due to the large number of sequences to two hours. Four different simplified sequences investigated in this report, detailed models were used, one for each combination of discussions of thermal-hydraulic processes for each primary-side and secondary-side break size. These sequence are not practical. Such discussions are simplified models are described in Section 3.3.3. documented for representative sequences in Reference 1. The calculations showed that the affected RCS loop did not significantly void, even in the h1BL CA sequences. The affected loop subcooled, 11.1 Group H Definition due to the combmed effects of ECC and the steam line break. Natural circulation flow was maintained Group H includes all sequences controlled by a through tne affected loop, but less flow passed , primary-side break combined with an asymmetric through the unaffected loops, as described in secondary-side break. A small primary-side break Scenario 3 of Reference 1. The asymmetric steam refers to a single, stuck-open pressurizer PORV. A line break caused a less rapid cooldown of the , medium primary-side break refers to a 0.0635-m unaffected loops, which, when coupled with the low (2.5-in.) diameter hole located at the bottom of the RCS pressure due to the LOCA, caused flashing in C loop hot leg. The secondary-side break was the unaffected loops. The flashing caused voiding initiated by either one stuck-open steam PORV on in the U-tubes and stopped natural circulation in the A loop, or two stuck-open steam PORVs, one the unaffected loops, on the A loop and one on the C loop. The steam PORVs stuck open when the reactor tripped. The in the SBLOCA sequences, a pressure plateau t controlling phenomena are the mass and energy was reached near 5.5 h1Pa (800 psia) as the flows, due to the primary-side break and ECC, and pressurizer filled. The RCS then depressurized heat transfer to the steam generators. relatively rapidly until the accumulator flow was initiated and the unaffected loops began flashing, si wing the rate of depressurization. In the 11.2 Results N1BLOCA sequences, the RCS pressure dropped quickly to the LPI shutoff head and then remained r Sequences 1-9 through I-12 (see Table 1) involve constant. The downeomer temperatures were con-SBLOCAs combined with stuck-open steam trolled by heat transfer to the steam generators and PORVs at full power. Sequences 2-9 through 2-11 by ECC mixing. The downcomer temperatures were . (see Table 2) involve h1BLOCAs combined with significantly affected by ECC mixing, since the stuck-open steam PORVs at full power. The downcomer temperatures could be more than 56 K sequences differ based on primary-side break size, (100 F) lower than the average RCS temperatures. , number of stuck-open steam PORVs, AFW flow- Table 34 shows that the minimum downcomer ing at normal or high (overfeed) rates, and AFW temperature decreased as the primary-side break throttled or not throttled (overfill) at 40% narrow- size increased. For the SBLOCA, as expected, the range level. minimum downcomer temperature was lower with 66
Table 34. Results of LOCAs combined with asymmetric secondary-side breaks Detailed Simplified Model Model Sequence Initiating Power Calculation Variation Used . Number Event Level Failures Used (from 3.3.3) 19 SBLOCA Full I stuck-open steam PORY l-9 9 l-10 SBLOCA Full I stuck-open steam PORV, AFW overfill 1-9 9 , 1 11 SBLOCA Full i stuck-open steam PORV, AFW overfeed I-9 9 l-12 SBLOCA Full 2 stuck-open steam PORVs 1-12 10 2-9 MBLOCA Full I stuck-open steam PORY 2-9 !! 2-10 MBLOCA Full I stuck-open steam PORV, AFW overfill 2-9 11 2-11 MBLOCA Full 2 stuck-open steam PORVs 2-11 12 Maximum Methods Used, Periods in Seconds Minimum Subsequent Figures RV DC RV DC Showing Sequence Detailed Simplified lland Temperature Pr ssure Plotted Number Model Model Calculations ('F) (psia) Results I-9 0 - 480 480 - 7200 - 224 368 A-17. A-18 1-10 0 - 480 480 - 7200 - 224 356 A-19, A 20 1 11 0 - 150 150 - 7200 - 224 368 A-21, A-22 1-12 0 - 475 475 - 7200 - 217 358 A-23, A-24 2-9 0 - 715 715 - 7200 - 190 144 A-53, A-54 2-10 0 - 625 625 - 7200 - 190 144 A-55, A 56 2-1I O - 580 580 - 7200 - 203 144 A-57, A-58 two stuck-open steam PORVs than with one, than 56 K (100*F) lower in the combination se-However, for the MBLOCA, the minimum quences than in the corresponding steam hne break downcomer temperature was 7 K (13*F) lower with sequences. However, because of the depressuriza-one stuck-open steam PORV than with two. The tion due to the LOCA, the final downcomer 7 K (13*F) temperature difference was less than the pressures were more than 9 MPa (1300 psia) lower uncertainty in the results. The average RCS in the combination sequences than in the temperature was colder with two stuck-open steam corresponding steam line break sequences. PORVs, but the calculated ECC mixing effect (see Variation 7, Section 3.3.3) was larger with one stuck-open steam PORV, because the total loop 11.3 Conclusions flow was assumed to be smaller with two stagnated unaffected loops than with one unaffected loop. All The sequences with combined LOCAs and stuck-the sequences were at or near steady state at two open steam PORVs were probably not as severe - hours. relativ'e to PTS as were the corresponding sequence with only stuck-open steam PORVs. The combination LOCA and stuck-open steam PORV sequences were probably not as severe PTS The combination of a LOCA and stuck-open transients as were the corresponding steam line PORVs resulted in sustained natural circulation in break sequences described in Sections 4.3 and 6.3. the affected loops. However, natural circulation in Because of the additional cooling due to the LOCA, the unaffected loops stopped, due to voiding of the the minimum downcomer temperatures were more U-tubes. O I 69 I
- 12. GROUP 1 RESULTS: ISOLATABLE PRIMARY-SIDE BREAKS The following sections present the results of Detailed model calculations were performed for thermal-hydraulic analyses for sequences in Group i initial portions of Sequences 11-1 (see Section 7 in of Table !5, using the methods presented in Reference l) and l2-3. These sequences represented ,
Section 3 to determine reactor vessel downcomer a SBLOCA isolated at 10 min and a h1BLOCA iso-fluid pressure and temperature histories. lated at 20 min. The detailed model calculations Section 12.1 def'mes the group and describes the provided starting points for the simplified models, ~ controlling phenomena. Section 12.2 presents the which were used to calculate the Group I sequence results for all the sequences in this group. Conclu- to two hours. Variation 7 of the simplified model sions related to Group 1 sequences appear in Sec- was used to calculate the SBLOCA sequences, and tion 12.3. To facilitate referencing of data, plotted Variation 8 was used for the A1BLOCA sequences, results showing the pressure and temperature These models were described in Section 3.3.3. The histories are organized in numerical sequence order results of simplified model calculations were in Appendix A rather than appearing within this augmented with hand calculations. The last 5000 s section. Due to the large number of sequences of Sequence Il-1 vas based on the end point of the investigated in this report, de: ailed discussions of calculation, at which time the downcomer pressure thermal-hydraulic processes for each sequence are and temperature were held constant by the pres-not practical. Such discussions are documented for surizer code safety valve and steam dump valves, representative sequences in Reference 1. respectively. The effect of maximum charging on SBLOCA Sequences Il-2 and Il-4 prior to 600 s (the end of the appropriate detailed calculation) was 12.1 Group i Definition calculated by hand and added to the results of the detailed calculation. A corresponding hand calcula-Group I consists of sequences initiated by a tion was not required for the N1BLOCA sequence primary-side break which is isolated at a given time because maximum charging was calculated with the
~
during the two-hour period. A "small" break refers detailed model. to a single stuck-open pressurizer PORV. A
" medium" break refers to a 0.0635-m (2.5-in.) Table 35 shows that the isolatable LOCA diameter hole located at the bottom of the C loop '
sequences were not potentially severe PTS tran-hot leg. The controlling phenomena are the mass sients. The final downcomer pressures were rela-and energy flows due to the primary-side break and tively high, but the downcomer remained warm. ECC, and heat transfer to the steam generators. The NIBLOCAs that were isolated at 20 min were the most severe transients of this group, but reached 12.2 Results a minimum downcomer temperature of only 453 K (356*F). Because the downcomer temperature was Sequences 11-1 through 11-4 (see Table ll) in-decreasing at the time of break isolation, the h1BLOCA sequences would have been more severe ! volve isolatable SBLOCAs. Sequences involving if the break had been isolated later. For example, isolatable h1BLOCAs are numbered 12-1 through if the break had been isolated at 40 m,m rather than 12-4 (see Table 12). All the isolatable LOCA 20 min, the minimum downcomer temperature sequences were initiated at full core power. The sequences differ based on primary-side break size, w uld have been more than 56 K (100'F) lower. time of LOCA isolation, and throttling of the charging flow. The charging flow was either throt- In the SBLOCAs and h1BLOCAs that were tied automatically by the pressurizer level control isolated at 10 min, the downcomer remained warm l system, throttled manually 3 minutes after the level because natural circulation was maintained - control should have throttled charging, or was not throughout the transients. Natural circulation was I throttled. maintained because the RCS loops were not highly I voided and the steam generators acted as heat sinks , Table 35 summarizes the analysis methods used throughout the transient, in the h1BLOCAs that and gives an indication of the results of the were isolated at 20 min, the steam generators were isolatable LOCA sequences. Dou.womer pressure heat sources, the U-tubes were draining, and the and temperature results for all sequences are loop flow s were stagnating at the time of isolation. presented in Appendix A. After isolation, HPl filled the loops, the RCS 70
Table 35. Results of isolatable primary-side breaks Detailed Simplified Model Alodel Sequence Initiating Power Calculation Variation Used
. Number Event Level Failures Used (from 3.3.3) 11 1 SBLOCA Full Isolated at 10 min. 11 1 7 11-2 SDLOCA Full Isolated at 10 min., charging not throttled 11-1 7 11-3 SBLOCA Full isolated at 20 min. 11-1 7 11-4 SBLOCA Full Isolated at 20 min., charging not throttled 11-1 7 12-1 MBLOCA Full isolated at 10 min., charging not throttled 12-1 8 12-2 MBLOCA Full isolated at 10 min., charging throttled manually 12-1 8 12-3 MBLOCA Full Isolated at 20 min., charging not throttled 12 3 8 12-4 MBLOCA Full Isolated at 20 min., charging throttled manually 12-3 8 Maximum Methods Used, Periods in Seconds Minimum Subsequent Figures RV DC RV DC Showing Sequence Detailed Simplified lland Temperature Pressure Plotted Number Model Model Calculations ('F) (psia) Results Notes a 11-1 0 - 2200 -
2200 - 7200 517 2554 A-351, A-352 A,B 11-2 0- 600 600 - 7200 150 - 600 501 2554 A-353, A-354 A,C 11-3 0- 600 600 - 7200 - $16 2554 A-355, A-356 A 11-4 0 - 600 600 - 7200 150 - 600 506 2554 A-357, A-358 A,C 12-1 0- 600 600 - 7200 - 463 2371 A-359, A-360 D 12-2 0- 600 600 - 7200 - 463 1942 A-361, A-362 12-3 0 - 1930 1930 - 7200 - 356 2371 A 363, A-364 D 12-4 0 - 1930 1933 - 7200 - 356 1914 A-365, A-366
- a. A-Maximum pressure determined by code safety valve opening setpoint.
B-fland calculations based on extrapolations of detailed model results. C-The effect of not throttling charging was calculated by hand and superimposed on results tiom the detailed model. D-Maximum pressure determined by PORY opening setpoint. heated to above the steam generator temperatures, reached in all the SBLOCA sequences. This dif-a large natural circulation flow was established, and ference between the SBLOCA and MBLOCA the downcomer was warmed. sequences was probably not real but was due to different calculation techniques. In the NIBLOCA Prior to the isolation of the break, the calculated sequences, the pressure increased due to the com-results were based on the LOCA sequences that piession of the steam bubbles in the pressurizer and were described in Section 9. All the sequences upper head. In the SBLOCA sequences, the steam responded similarly after the isolation of the break. bubbles were condensed rather than compressed, After break isolation, the cooling mechanism, due and most of the pressure increase occurred after the to energy flow through the break, no longer existed. RCS became liquid-full. The complete condensa-The downcomer temperature increased after AFW tion of the steam bubbles in the SBLOCAs was was throttled, and the RCS pressure exceeded the probably not reasonable, and the uncertainty of IIPI shutoff head. The temperature increased until these sequences, described in Section 15, was the SDVs opened, and then held the temperature increased to compensate for this. The uncertainty constant. The isolation of the break caused the in the calculated pressure of the isolatable downcomer pressure to increase as HPI and charg. SBLOCAs was not significant relative to PTS. ing filled the RCS. The downcomer pressure was ultimately limited by the setpoint of either the Throttling the charging significantly affected the pressurizer PORV or code safety valves. In the downcomer pressure, but had only a minor effect NIBLOCA sequences with charging throttled, the on the downcomer temperature response. The RCS pressure did not reach the PORV setpoint primary effect of break size was on the amount of within two hours. However, the PORV setpoint was void in the RCS at the time of break isolation. Of 71 L
. course, the RCS was more highly voided in the transients because the loops were voiding, causing MBLOCAs than in the SBLOCAs. loop flow stagnation and colder downcomer temperatures at the time of break isolation. The
'12.3 Conclusions p1BLOCA sequences would have been more severe ifisolation of the break had been delayed. Failure to throttle charging significantly affected .
The isolatable LOCA sequences were not poten. downcomer pressure response, but did not
- tially severe PTS transients. The MBLOCAs that significantly affect the downcomer temperature wuc isolated at 20 minutes were the most severe response. .
e i l l t 72
~ ,- -
,. - . . - - . ,--n _ - , . . - - . , - . . , _ . . . - . . . . - . , . . . , - - . - . . - _ _ _ _ . _ _ - - - , . _ - _
- 13. GROUP J RESULTS: STEAM GENERATOR TUBE RUPTURES The following sections present the results of A detailed model calculation was performed over thermal-hydraulic analyses for sequences in Group J the initial portions of Sequences 10-1 and 10-2. The
. of Table 15, using the methods presented in Section 3 detailed model was modified by representing the to determine reactor vessel downcomer Huid pressure broken tube with four control volumes and two and temperature histories. Section 13.1 defines the break paths. A discussion of this modification . group, and Section 13.2 presents the results and con- appears in Section 12.2 of Reference 1. To address clusions for sequences m Group J. To facilitate g ; gg g r:ferencing of data, plotted results showing the .
rupture event, Van.ation 13 of the s.implified model pressure and temperature histories are organized in was developed (see Section 3.3.3). numerical sequence order in Appendix A rather than appeanng within this section. Due to the large number of sequences investigated in this report, detailed Steady temperature and pressure conditions were discussions of thermal-hydraulic processes for each reached for all sequences, using either detailed or sequence are not practical. Such discussions are simplified model calculations. These steady condi-documented for representative sequences in tions were extrapolated to continue to 7200 s. Reference 1. For all tube rupture sequences, the effect of the 13.1 Group J Definition break flow limits the potential for repressurization f the primary system. Even with a failure to throt-Group J includes Sequences 10-1 through 10-5 tie HPI and charging (Sequence 10-5), the repres-(see Table 10). All sequences involve the double-surization was limited to 9.66 MPa (1401 psia). At ended rupture of a single steam generator tube with the reactor operating at hot standby conditions. that pressure, the break flow equaled the HPI and
. Sequence 101 (Scenario 13 in Reference 1)is a base charging injection flow, and the pressure stabilized.
sequence. Sequences 10-2 through 10-5 are modifi- For Sequences 10-1 through 10-4, the repressuriza-cations of Sequence 10-1. The sequences vary due tion was limited by the action of the affected steam
, to different operator responses to the events and generator PORV and safety valves.
to hardware failures. The cooldown rates in Sequences 10-1, and 10-3 The base sequence (10-1) requires the opcrator to isolate steam and feedwater to the affected steam through 10-5 were similar, with minor differences generator (ASG) and cool the primary system to a resulting from the different pressurizer PORV con-specified temperature using the steam dump valves. trol criteria. The cooldown in Sequence 10-2 was The operator then equalizes the primary and more severe as a result of the steam dump valve secondary system pressures by opening the pressurizer sticking open for 10 min. The minimum reactor PORV in two cycles. Sequences 10-2 through 10-5 vessel downcomer fluid temperature for assume failures such as stuck-open steam dump valves Sequence 10-2 was 409 K (277'F). or a pressurizer PORV, or a failure to throttle HPl and charging. Results for the steam generator tube rupture sequences were, in general, strongly arrected by the 13.2 Results and Conclusions assumptions for operator action. For example, if it is assumed the operator opens the steam dump valves Table 36 presents the analysis methods used and and cools the primary system to a prespecified tem-
- gives an indication of results for all Group j sequences. Plotted results for all sequences appear perature, then the temperature (but not its rate of in Appendix A. change) is already denned by the assumption.
L f l 73
Table 36. Results of steam generator tube ruptures Detailed Simplified Model Model Sequence initiating Power Calculation Variation Used Number Event Level Failures Used (from 3.3.3) . 10 1 SGTR HSB None 10-1 - 10-2 SGTR HSB SDV fails to close for 10 minutes 10-2 - 10 3 SGTR HSB POR,V sticks open for 10 minutes on first 10-1 13 , opemng 10-4 SGTR HSB Second PORV opening does not occur - 10-1 - 10-5 SGTR HSB Second PORY opening does not occur and HP!/ 10-1 13 charging not throttled Maximum i Methods Used Periods in Seconds Minimum Subsequent Figures RV DC RV DC Showing Sequence Detailed Simplified Hand Temperature Pressure Plotted Notes a Number Model Model Calculations ('F) (psia) Results 10-1 0 - 2500 - 2500 - 7200 386 1080 A-341, A-342 A 10 2 0 - 2500 - 2500-7200 277 1080 A-343, A-344 A 10-3 0 - 990 990 - 1590 1590 - 7200 394 950 A-345, A-346 B 10-4 0 - 1000 - 1000 - 7200 406 1020 A-347, A-348 A 10-5 0 - 990 990 - 2635 2635 7200 366 1401 A-349, A-350 B i.
- a. A-Steady conditions at end of detailed model calculation were extrapolated to 7200 s.
B-Steady conditions at end of simplified model calculation were extrapolated to 7200 s. O O 4 + 1 74
~ en- - , + - - - - -
- 14. HEAT TRANSFER COEFFICIENTS Another contractor has been enlisted by ORNL through 13 of this report. Representative heat transfer to determine, for the sequences presented in this coefficients for the remaining sequences correspond report, the heat transfer coefficients on the inside to the sequence equivalency shown in Table 16. The
- surface of the reactor vessel wall. ORNL indicated, assignments shown in Table 37 were made primarily hoaever, that it would facilitate their fracture by comparing the reactor vessel downcomer flow con-mechanics analyses if a preliminary estimate of the ditions in the detailed model calculations from this - heat transfer coefficient was made for each se- report with those from Reference 1, and selecting the quence. Reference i documents the reactor vessel Reference I scenario that most closely matched the wall heat fransfer coefficient histories for twelve conditions. Typically, this process resulted in the selec-PTS sequences. INEL agreed to indicate, for each tion of the scenario having the most representative sequence, which of the Reference I scenarios has break size or initiating event, secondary heat sink a heat transfer coefficient that is most behavior (source or sink), and reactor coolant pump representative. trip control (trip and timing of trip). For sequences involving significant changes in the heat transfer coef-Table 37 lists the scenario in Reference I that has ficient as a transient progresses, more weighting was the most representative heat transfer coefficient for given to the end-state conditions in selecting the each of the sequences presented in Sections 4 representative scenario. Table 37. Representative heat transfer coefficients , Is Best IIcat Transfer Represented By . Coefficient Reference I for Sequence Scenario Number Comments - l-1 11 Subtract I hr from the time in Scenario 11 1-2 11 Subtract I hr from the time in Scenario 11 1-3 11 Subtract I hr from the time in Scenario 11 1-4 11 Subtract I hr from the time in Scenario 11 1-5 3 Subtract 1000s from the time in Scenario 3 1-6 3 Subtract 1000s from the time in Scenario 3 1-7 3 Subtract 1000s from the time in Scenario 3 1-8 3 Subtract 1000s from the time in Scenario 3 1-9 3 Subtract 1000s from the time in Scenario 3 1-10 3 Subtract 1000s from the time in Scenario 3 1-11 3 Subtract 1000s from the time in Scencrio 3 1-12 3 Subtract 1000s from the time in Scenario 3 2-1 6 2-2 6 2-3 6 2-4 6 2-5 6 2-6 6 2-7 6 2-8 6
- 2-9 6 2-10 6 2-11 6 75
Table 37. (continued) Is Best - Heat Transfer Represented By Coefficient Reference I for Sequence Scenario Number Coraments - 31 11 Subtract I hr from the time in Scenario 11 3-2 11 Subtract I hr from the time in Scenario 11 - 4-1 6 5-1 5 RCPs not tripped 5 14 11 Subtract I hr from the time in Scenario 11 5-15 11 Subtract I hr from the time in Scenario 11 5 17 4 5-18 11 Subtract I hr from the time in Scenario 11 5-19 11 Subtract I hr from the time in Scenario 11 5-20 4 6-1 11 Subtract I hr from the time in Scenario 11 6-2 11 Subtract I hr from the time in Scenario 11 6-3 11 Subtract I hr from the time in Scenario 11 6-4 11 Subtract I hr from the time in Scenario 11 6-5 2 6-6 2 6-7 2 6-8 11 Subtract I hr from the time in Scenario 11 6-9 2 7-1 5 RCPs not tripped + 72 5 RCPs not tripped 73 5 RCPs not tripped 7-4 3 - 7-5 3 7-6 3 7-7 5 RCPs not tripped
- 7-8 3 7-9 11 Subtract I hr from time in Scenario 11 7-10 3-7-11 11 Subtract I hr from time in Scenario 11 7-12 3 8-1 1 AFW '.solated 8-2 1 AFW isolated 8-3 1 AFW isolated 8-4 2 8-5 i AFW isolated 8-6 2 9-1 5 RCPs not tripped 9-2 4 9-3 4 9-4 4 9-5 4 9-6 4 -
9-7 4-9-8 4
, 9 4 9-10 4 76
Table 37. (continued) Is Best Heat Transfer Represented By Coefficient Reference I for Sequence Scenario Number Comments 9-11 4 9-12 4 9-13 4 9-14 4 9-15 4 9-16 4 9-17 4 9-18 4 9-19 4 9-20 4 9-21 4 9-22 4 9-23 4 9-24 4 9-25 5 RCPs not tripped 9-26 5 RCPs not tripped 9-27 5 RCPs not tripped 9-28 5 RCPs not tripped 9-29 5 RCPs not tripped 9-30 5 RCPs not tripped 9-31 5 RCPs not tripped 9-32 5 RCPs not tripped 9-33 4 9-34 4 9-35 4 9-36 4 9-37 4 9-38 4 9-39 4 9-40 4 9-41 4 9-42 4 9-43 4 9-44 4 9-45 4 9-46 4 9-47 4 9-49 5 RCPs not tripped 9-50 5 RCPs not tripped
~'
9-51 5 RCPs not tripped 9-52 5 RCPs not tripped 9-53 5 RCPs not tripped 9-54 5 RCPs not tripped 9-55 5 RCPs not tripped 9-56 5 RCPs not tripped 10-1 9 77
i 1 Table 37. (continued) . Is Best Heat Transfer Represented By Coefficient Reference 1 , for Sequence Scenario Number Comments 10-1 9 ~ 10-2 9 10-3 9 10-4 9 9 10-5 11-1 7 11 2 7 11-3 7 11-4 7 12-1 6 (t s 1000s),
, 7 (t > 1000s) 12-2 6 (t s 1000s) 7 (t > 1000s) 12-3 6 (t s 2500s),
7 (t > 2500s) 12-4 6 (t s 2500s) 7 (t > 2500s) i - i . .f a i e 1 1 d 78
- 15. UNCERTAINTIES A comprehensive study of applicable uncertain- tion of temperature. This was found to be a useful ties for all 183 sequences is beyond the scope of this approach since the components of uncertainty fre-work. Es'. mates of uncertainty are required, quently varied in magnitude as a sequence
. however, to facilitate the ORNL fracture mechanics progressed. As an example, during a large steam analyses. This section presents estimates of uncer- line break sequence the uncertainty in reactor vessel tainty for the reactor vessel downcomer fluid downcomer temperature is initially small and . temperatures and pressures documented in this related to measurement error or minor uncertain-report. ties in initial conditions. As the transient begins, a relatively large uncertainty in the break flow Estimates of uncertainties were developed dominates and the overall uncertainty rises as the separately for each category of sequence. First, for temperature falls. After some time, the affected each category the controlling thermal-hydraulic steam generator reaches a stable low-pressure phenomena were identified, and an estimate of operating point, with heat removal controlled by uncertainty from the effects of each phenomenon a stable steam generator secondary saturation was made. Second, an estimate of the uncertainty temperature, and the overall uncertainty diminishes. due to the use of the simplified model was deter-mined primarily from the benchmark comparisons In situations where comparisons between experi-shown in Section 3.4. Third, a representative mental and code-calculated data were available, the uncertainty due to measurements (including initial comparisons were used directly as the indication of conditions) was determined. Next, the components uncertainty due to controlling phenomena. of uncertainty (phenomena, simplified model, and measurement) were combined using the root-sum- Table 38 shows the uncertainties resulting from square method. Finally, the uncertainty was limited the above evaluation. The maximum temperature based on physical constraints,if any. For example, uncertainty is 139 K (170'F) and the maximum . an uncertainty range was not allowed to project a pressure uncertainty is t 1.38 MPa (1200 psi). reactor vessel downcomer temperature below 305 K Uncertainty ranges shown in the table should be (90'F), the temperature of the HPl. considered as 95% confidence ranges; that is, the , actual value will usually fall within the stated range, The uncertainties, in general, were stated as a but not always. For temperatures and pressures function of the value of the variable itself. For between values shown in the table, linear interpola-example, the uncertainty in temperature is a func- tion may be used. Table 38. Uncertainties in reactor vessel downcomer fluid temperature and pressure Sequences involving Small Steam Line Breaks Only (All of Tables 5 and 7, plus Sequences 9-2 through 9-47, 9-57 through 9-65, 9-75 through 9-80, 9-86 and 9-87. Temperature Error ('F) (* F) 550 15 400 1 25 200 1 25 100 15 Pressure Error (psia) (psi) 2400 1 20 1800 t125 1500 t160 1400 t165 1000 t t80 800 t180 79
Table 38. (Continued). Sequences involving Lerge Steam Line Brooks Only (All of Tables 6 and 8) Temperature Error . (*F) ('F) 550 15
- 400 + 25, -40 200 + 25 100 t5 Pressure Error (psia) (psi) 2400 t 20 1800 t125 1500 t160 1400 t165 1000 t 180 800 + 180, -220 Sequences Not involving Primary or Secondary Brooks (Sequences 9.1,9-48 through 9-56,9-66 through 9-14, and 9-81 through 9-85.)
Temperature Error + 5'F Pressure Error i 20 psi Sequences Involving Non Isoletable LOCAs or Steem Generator Tube Rupture (Tables 1,2,3,4 and 10) Temperature Error - ('F) (* F) . 550 15 400 1 40 300 t 70 160 1 70 100 + 40, -10 l Pressure Error ( (psla) (psi) i 2500 1 50 l 1000 t 50 800 i 160 400 i160 200 i 75 140 1 25 4 80
r s t Table 38. (Continued). Sequences involving isolatable Small Break LOCAs (Table 11) Temperature Error (*F) (*F)
. 550 15 400 1 40 300 i 70 160 1 70 100 + 40, -10 Pressure Error (psia) (psi) 2500 + 50, -50 2000 + 100, -250 1200 + 250, -100 1000 + 250, 50 800 t160 400 i 160 200 1 75 140 1 25 Sequences involving Isolatable Medium Break LOCAs (Table 12) - Temperature Error
(* F) (*F)
- 550 15 400 i 40 300 1 70 160 ,
1 70 100 + 40, -10 Pressure s Error (psia) (psi) 2500 1 25 2250 t 25 1800 1200 800 1 160 400 t160 200 1 75 140 1 25 81
- 16. SENSITIVITIES The purpose of this section is to discuss the major The effect of secondary-side break location on sensitivities of the thermal-hydraulic parameters of the reactor vessel downcomer fluid temperature is interest (reactor vessel downcomer pressure and shown in Figure 24. The temperature responses temperature). The discussions are segregated as compared are for a single stuck-open steam !ine
~
follows by sequence type: secondary-side breaks PORV beginning from hot standby conditions v.ith (Section 16.1), reactor trip with minor failures and without a steam line check valve failure. With (Section 16.2), primary-side breaks (Section 16.3), the check valve failure, all three steam generators - combination primary- and secondary-side breaks are affected while without the check valve failure (Section 16.4), and steam generator tube ruptures only one steam generator is affected. Figure 24 (Section 16.5). These discussions are based on shows that the cooldown is more severe for the case experiences and understandings gained through the with only one steam generator affected. This analyses presented in this report and in Reference 1. indicates the importance of the secondary side depressurization characteristics for secondary-side 16.1 Secondary-Side Breaks break sequences. The results show that, for a given break size, the cooldown will be more severe for a break located on a single steam line than on the Sequences with secondary-side breaks involve common steam line because a single affected steam energy removal from, and resulting depressuriza- generator will depressurize more rapidly than three tion of, one or more steam generator secondary affected steam generators. Furthermore, for a break systems. Primary system heat removal across the on only one steam line, loop stagnation in the unaf-steam generator tubes is enhanced resulting in a fected loops requires further analysis of multi-cooldown and shrinkage of the primary fluid. dimensional effects in the cold leg and reactor vessel llecause of shrinkage, the primary system downcomer regions. depressurizes, causing reactor, turbine, and reactor coolant pump trips. Large natural circulation loop The effects of initial core power level on the reac- ' flows continue in loops affected by the break while tor vessel downcomer fluid temperature are shown unaffected loop flows are virtually stagnant. in Figure 25. The temperature responses shown are Primary system cooldown continues asymptotically for a single stuck open steam line PORV starting
- toward the saturation temperature associated with from full power and hot standby conditions. As tne final secondary-side pressure in the affected would be expected, lower temperatures are steam generator. This final pressure is determined experienced in sequences starting from hot standby primarily by break size and core power level. The than from full power conditions.
primary system pressure decline is reversed as high pressure injection and charging flow replace Failure of the charging system to throttle when primary system coolant volume lost because of the setpoint pressurizer level is reattained can l shrinkage. dramatically effect the reactor vessel downcomer pressure response. This effect is shown in Figure 26 Major sensitivities for secondary-side breaks are: for sequences with one stuck-open steam dump (1) the break size, (2) the break location, (3) the valve starting from full power. Failure to throttle j initial core power,(4) performance of the charging charging causes the pressurizer level to continue system, and (5) auxiliary feedwater isolation. These increasing and results in an increasing primary are discussed further in the following paragraphs. system pressure. The pressure excursion is arrested only by opening of the pressurizer PORV at its The primary system temperature response is par- setpoint pressure.
- ticularly sensitive to break size as shown in '
Figure 23 for sequences starting from full power Isolation of auxiliary feedwater to the affected with 1,2,3, and 5 open steam dump valves (SDVs). steam generator (s) has a dramatic effect on the reac-Two effects are noted. First, the initial rate of tor vessel downcomer temperature response. This - t temperature decrease becomes larger as the break effect is shown in Figure 27 for a sequence involving size increases. Second, the asymptotic temperature one stuck-open steam line PORV beginning from discussed above decreases as the break size hot standby conditions. As shown, when the increases. auxiliary feedwater is isolated at 10 min (600 s) the 82
650- '- ' ' ' ' ' i
-700 0 1 SDV O 2 SDV 600 -
A 3 SDV - i X- 5 SDV -600.
- m ^
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v 550 x - v l -n -
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400 - O O O OO -
- l. -
u o aA
. X X X X -200 350 ' ' ' ' ' ' '
O 1000 2000 3000. 4000 5000 6000 7000 8000 Time (s) Figure 23. Effect of secondary-side break size on reactor vessel downcomer temperar2re.
.. \
l
- r. l
-1 L- 600 , , , .i , , , -~ l
.O 1 steam PORV and check volve failures . -600 0 1 steam PORV failure r
[ m 550 - m x v - o'
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O 1000 2000 3000 4000 5000 6000 7000 8000 Time (s) Figure 24 Effect of secondary-side break location on reactor vessel downcomer temperature. 83
600 i , , , , , ,
- O Full power -600 1
O Ho t s t codby C
,550 -
n v M b
-500 v 0 o u u .
3 3
$ 500 g
u
-400 $
E E v e 450 - - H
-300 400 0 1000 2000 3000 4000 5000 6000 7000 8000 Time (s)
Figure 25. Effect on reactor vessel downcomer temperature of initial core power level for secondary-side break sequences. 20000 i , , , , , ,
, O Normal charging O Failure to throttle charging
~
-2500 ct 16000CI -
5 b 14000
--2000 $
m 3 m M O U
' 12000 - -
E 1 Q. n m m m n 10000 - - v v v v v v --1500 8000 ' ' ' ' ' ' ' O 1000 2000 3000 4000 SOCO 6000 7000 8000 ~ Time (s) Figure 26. Effect of charging throttling failure on reactor vessel downcomer pressure for secondary-side break sequences. 84
600 , , , , , , , O No AFW isolation -600 0 AFW isolation at 600 s m 550 - v x - m.
-500 v 0
' e
. 3 L 3
$ 500 g
u
-400 $
E E U o 450 - H
-300 400 O 1000 2000 3000 4000 5000 6000 7000 8000 Time (s)
Figure 27. Effect of auxiliary feedwater isolation on reactor vessel downcomer temperature for secondary-side break sequences. rate of cooldown slows considerably and when the result, the primary system depressurizes and starts affected steam generator is dry (2200 s) the to void. Emergency core cooling system injection , cooldown is stopped and a heatup of the primary increases as the primary system depressurizes with coolant begins. This is compared against the con- significant increases in injection flow occurring at tinuous cooldown for the case where auxiliary feed- the accumulator and low pressure injection setpoint , water is not isolated. While auxiliary feedwater pressures. If the break is large enough, coolant loop isolation limits the cooldown it causes a more severe natural circulation flow may be lost as large voids repressurization of, the primary system due to form in the U-tubes of the steam generators. A thermal expansion of primary coolant during the larger break size, however, results in generally lower heatup phase. This effect is shown in Figure 28 for primary system pressures. The only mechanism that the same sequence. can repressurize the primary system is isolation of the break (which may be accomplished for breaks 16.2 Reactor Trip With in certain locations by the operator closing a block Minor Failures *^l")- Major sensitivities for primary-side breaks are: Sequences m.volvm.g only minor failures follow-
, (1) break size, (2) initial core power level, (3) break mg reactor trip were found not to be thermal- location, and (4) break isolation. These are dis-hydraulically severe. Minor failures are defined as cussed further in the following paragraphs.
main'or auxih,ary feedwater overfill, auxihary feed-water overfeed, and failure to throttle charging. Both the reactor vessel downcomer pressure and Smce these failures do not mvolve breaks of either fluid temperature responses are sensitive to the , the primary or secondary systems, a mechanism t break size, as shown in Figures 29 and 30, severely cool the primary s); stem fluid does not ; g g g ggg exist. No important sensitivities were observed for breaks in the hot leg with diameters of 3.55 cm sequences in this group. , (1.4 in.), 5.08 cm (2.0 in.), and 6.35 cm (2.5 in.). As expected, as the break size increases so do the 16.3 Pr.imary-Side Breaks depressurization and cooldown rates of the reactor vessel downcomer. For the smallest break, high Sequences with primary side breaks involve direct pressure injection (HPI) flow exceeded the break primary system energy and mass removal. As a flow, the steam generator U-tubes were not voided, 85
20000 , , , , , , , C No AFW isair tion l 4 O AFW isolation at 600 s 18000 - -
-2500 o -
O. 16000[ g 0 0
- 0 14000 g- --2000 I m -
m e
' 12000 - -
m 1 Q. 10000 '- . G O O O' --1500 8000 O 1000 2000 3000 4000 5000 6000 7000 8000 Time (s) Figure 28. Effect of auxiliary feedwater isolation on reactor vessel downcomer pressure for secondary-side break sequences. 20000 , , , , , , , , O 3.55 cm O 5.08 cm -2500 A 6.35 cm O 15000 a - 3 a
-2000 9.
a . m f, O O JM u 10000 _-1500 0 a ' m 'n n ^ m m m m 5
-1000 f.
C. 5000 -
-500 1
m m m m n .
' ' ' ' ' ' ' I 0 o 0 1000 2000 3000 4000 5000 6000 7000 8000 Time (s) .
Figure 29. Effect of primary system break size on reactor sessel downcomer pressure. 86
700 , , , , , , , 800 0 3.55 cm O 5.08 cm A 6.35 cm
, n 600,- -
^
M -600 v k'- v Y, ' n v m v m v n v v m v v O j SCO - s- -
-400
%t e o C1.
O. E E e e 400 -
-200 300 O 1000 2000 3000 4000 5000 6000 7000 8000 Tirne (s)
Figure 30. Effect of primary system break size on reactor vessel downcomer fluid temperature. and loop natural circulation flow continued. For Figure 33. The depressurization for the hot leg the larger two breaks this was not the case and, as break is not as large as for the cold leg break. The a result, the reactor vessel downcomer temperatures primary reasons for this difference are that the dropped dramatically due to increased influence of density of fluid at the break, and hence the break the cold HPI fluid. mass flow rate, are significantly larger with the cold Ieg break during the first 5000 s. After that 'ime the The sensitivities of reactor vessel downcomer hot leg break begins to pass steam and, as a rault, pressure and fluid temperature to the initial core the hot leg break depressurization accelerates. The power level are shown in Figures 31 and 32 respec- significantly higher primary system pressure with tively. As expected, pressures and temperatures are the hot leg break makes it a more severe sequence generally lower for primary-side break sequences for pressurized thermal shock than a cold leg break initiated from hot standby conditions than for those sequence with the same size break. As a conse-initiated from full power conditions. The curves quence of this analysis, ORNL' used thermal-shown are for a 3.55 cm (1.4 in.) diameter hot leg hydraulic results for hot leg primary-side breaks to break (stuck-open pressurizer PORV). bound the resu'ts for cold leg breaks as well. Most of the primary-side break analysis per- The effect of break isolation is shown in formed for this study has been for breaks in a hot Figures 34 and 35 for a medium break, 6.35-cm
- leg or for stuck-open pressurizer power operated (2.5 in.) diameter, in the hot leg. The cases shown i
relief valves (essentially equivalent to a hot leg are for no isolation, isolation at 10 min, and isola-break). To study the effects of break location, tion at 20 min. With no isolation the primary equivalent RELAPS calculations for 5.08 cm system pressure generally declines continuously. As (2.0 in.) hot and cold leg breaks were performed. shown, if the break is isolated, a dramatic increase Results indicate loop natural circulation flow was in primary system pressure occurs and this increases l- stopped for breaks at both locations and this caused the potential for pressurized thermal shock. The
- similar dramatic decreases in reactor vessel reactor vessel downcomer fluid temperature downcomer fluid temperatures (similar to that decrease is arrested by isolation of the break and shown in Figure 30). The primary pressure this decreases the potential for pressurized thermal responses differed markedly, however, for the hot shock. Since the time at which an operator is leg and cold leg break locations as shown in assumed to isolate a break is arbitrary, ORNL 87 L.
20000 , i i i i i i O Fu ll power O Hot stondby 2500 C1 15000 - g I
-2000 9 G. :
m s s . 10000 ~-' - 1500 e 3 y m a n n m n m , m
-1000 5000 -
-500 0 O O 1000 2000 3000 4000 5000 6000 7000 8000 Time (s)
Figure 31. Effect of initial core power level on reactor vessel downcomer pressure for primary-side breaks. 600 i i i i i , i
~
O Full power -600 0 Hot stondby n 550 , , m m y v v v v v v . P-
-500 v 0
6 e L 3 3
"O500 -
L O 0 L ct -
-400 o
a. E U E H o 450 -
-300 .
400 ' ' ' ' ' ' ' O 1000 2000 3000 4000 5000 6000 7000 8000 Time (s) ' Figu*e 32. Effect of initial core pow er level on reactor vessel downcomer fluid temperature for primary side breaks. 88
20000 , , , , i i i O Hot leg breck O Cold leg breck -2500
, 15000 0 -
9 Q
-2000 3
'E 6 -
S 8 10000 ~- - 1500 q; a ' M m 0 Q + u -
-1000 0 1
a. 5000 - - -
-500 3 0 0 O O 1000 2000 3000 4000 5000 6000 7000 8000 Time (s)
Figure 33. Effect of b cak location on primary system pressure for primary-side breaks. B
=
20000 , , , , , , , O No isolotion O Isolation at 10 minutes -2500 A isolation at 20 minutes E9 15000 - 9 a.
-2000 o
.x a v
v a 8 10000 '- - 1500 e 3 M 3 W m m 8 -
-1000 0 Q-a.
5000 -
-500 m v v v v v O O ' ' ' '
O O 1000 2000 3000 4000 5000 6000 7000 8000 Time (s) Figure 34. Effect of primary-side break isolation on reactor vessel downcomer pressure. 89
700 ' ' , i ' ' , 7800 g so isoicticn C isolution ct 10 miiutes A .wlet.on et 20 minutes
, ,600 -
v x p- - v w -- g L L 3 . 3 . E $00 g
-400 '
S " E E o e 400 - -
-200 l
I ' ' ' 9 ' ' 300 ' O 1000 2000 3000 4000 5000 6000 7000 8000 Time (s) Figure 35 Effect of primary-side break isolation on reactor vessel downcomer fluid temperature. analyses assumed a very late break isolation time 16.5 Steam Generator (1-1/2 hours). This bounded the results for all isola- Tube Ruptures tion times since the reactor vessel downcomer fluid + temperature will be very low as compared to earlier isolation times. A further sensitivity in the cases with break isolation is the failure to throttle charg- Steam generator tube rupture (SGTR) sequences . ing flow as discussed in Section 16.1. involve a break between the primary and secondary systems. Sequences investigated involved the rup-16.4 Combination Primary- and ture or a single tube. This break size was selected Secondary Side Breaks primarily because the rupture of one tube is the most probable failure and bec.mse the primary Sequences involving combinations of primary- system pressure remains more elevated than it would and secondary-side breaks share many common with a larger number of broken tubes. The SGTR l characteristics with the secondary-side or primary- secuences investigated represent various combina-
- side breaks already discussed. Generally, the tiv,s of operator actia
- md complicating combination break pressure responses were similar harifware failures. In fact the thermal-hydraulic to those of the primary-side break alone. General- severity of SGTR sequences is principally dependent ly the combination break temperature responses on assumption of operator action and any hardware i
were similar to those of the secondary-side break failures unique to such action. While SGTR alone. Important sensitivities for the combination sequences with no operator intervention (other than breaks are expected to be similar to those already that required by normal small primary-system break described in Section 16.3 for the primary system procedures) were analyzed, it is unlikely an SGTR pressure response and in Section 16.1 for the reac. event would be undiagnosed by the operators for
- tor vessel downcomer fluid temperature response, a 2-hour period. Thus SGTR analyses concentrated 90
on sequences with realistic operator intervention plicating hardware failures would be the failure of and subsequent hardware failures. the SDVs or a PORV to close when demanded.
, Realistic operator intervention for SGTR -
sequences first requires a rapid cooldown of the Figure 36 shows the sensitivity of the reactor primary system by the opening of the steam dump vessel downcomer fluid temperature to failure to valves (SDVs). Upon attaining proper primary close the SDVs for 10 min (after their closure was system subcooling, the SDVs are closed and the demanded). The results shown are for the rupture primary system pressure is controlled near the steam of a single tube beginning from hot standby reac-generator safety valve opening setpoint pressure by tor conditions with normal operator intervention manual control of a power operated relief valve and with normal and delayed SDV closure. As (PORV) on the pressurizer. These actions are shown, the temperature response and, as a result intended to minimize the loss of primary system the thermal-hydraulic severity of the sequence, are coolant through the broken tube. During such very sensitive to both the assumed operator actions operator intervention, the most significant com- and a hardware failure pertinent to those actions. 600 , , , , i i i
-600
~
O Nor not recevery 0 10 minute delayed SDV closure m 550 - n M 12-v
-500 v
0
' o L
3 3
- 500 -
. E O
^
c o o o o o o E
- c. -
-400 o
ct E E O o
- H 450 -
-300 400 -
O 1000 2000 3000 4000 5000 6000 7000 8000 Time (s) Figure 36. Effect on reactor vessel downcomer fluid temperature of steam damp valve failure to close on demand during normal recovery from SGTR. e 91
i I ! 17. CONCLUSIONS Reactor vessel downcomer fluid pressure and Significant findings of the analyses in this report temperature responses were determined for are presented below, arranged by sequence type. i 183 sequences of interest to the pressurized thermal . I shock study for the liBR-2 Pressurinxi Water Reactor. 17.1 Secondary-Side Breaks A unique economical method was developed to . accomplish this effort. The method used may be generally applicable to the thermal-hydraulic Seventy-nine sequences involving only secondary-analysis of other large groups of sequences sharing side breaks were investigated. These sequences con-
- common characteristics, such as the output of fault tained one or more stuck-open steam PORVs, one i tree analyses. The method involved the combina- or more stuck-open steam dump valves, double-tion of partial-length calculations, using a detailed ended main steam line breaks, and combinations RELAPS model of the plant; full-length calcula- of these. General findings were
- (1) the larger break
; tions using a simplified RELAPS model; and hand size produced colder reactor vessel downcomer calculations. The simplified model was constructed temperatures; (2) a failure to throttle charging specifically to address the controlling phenomena results in high primary system pressures; and
- of a group of sequences, as determined from (3) sequences involving isolation of all auxiliary analyses of detailed model calculations. The feedwater result in high primary system pressures.
simplified model was then benchmarked against the The most severe sequence investigated in this report detailed model to assure the validity of the was Sequence 9-22. This sequence involves five simplification process. Sequence responses were stuck-open steam dump valves and failures to
, typically determined by joining results using the throttle auxiliary feedwater and charging following
! detailed.model over the initial portion of the a reactor trip from full power. For Sequence 9-22 i sequence (when relatively complicated transient the minimum reactor vessel downcomer fluid . phenomena predominate) with results using the temperature was 364 K (196*F), and the maximum l simplified model over the later portion of the subsequent pressure was 16.35 MPa (2371 psia). sequence (w hen steady or quasi-steady phenomena predominate). Once a simplified model has been For sequences initiated by a stuck-open steam line developed and qualified for a particular type of PORV at full power, an automatic reactor trip does i sequence (for example, stuck-open steam dump not occur, thus precluding PTS concern. In such I valves at full power), the model was then applied an instance, the operator may manually trip the
- to other sequences of that type. Thus, the thermal- reactor. Thermal-hydraulic responses have been i hydraulic responses of many sequences may be provided for these sequences as well.
determined at a considerable cost savings, compared with performing a detailed model calculation for For sequences initiated by a reactor trip and each sequence. followed by a stuck-open steam line PORV, the I primary system pressure does not decline suffi-As a convenience for the reader, a preliminary in- ciently to cause tripping of the reactor coolant dication of the most thermal-hydraulically severe se- pumps, thus minimizing the PTS concern. quences for PTS are summarized in Table 39. Judgments concerning PTS severity require considera- Sequences involving large (double-ended) main
- j. tion of probability and fracture mechanics aspects not steam line breaks are, in general, severe. liowever, considered in this report. These judgments will be severity is limited because, if the cooldown con-made later at ORNL. Table 39 lists all sequences tinues (auxiliary feedwater not isolated), then the ,
which have both reactor vessel downcomer resulting primary fluid thermal contraction prevents temperatures below 394 K (250'F) and pressures a severe primary system repressurization. Thus only above 3.45 MPa (500 psia). These ranges were ar- sequences involving a failure to throttle charging , bitrarily chosen only to limit the size of the table and result in both a very low temperature and a pressure are not to be considered as limits of PTS severity, at the pressurizer PORY opening setpoint. l i 92 i 4
-.n -
--m.n., ---,n,--~---,-+.,,,,e-,__,. . , , - - ,-,--,-,---,,,,g, , _ . . , . , , , - , , - , - , , , . , , , - - ,,n-n_, .- - . , _ _ - - ,,
i i Table 39. Summary of severe sequences f Maximum ' Minimum Subsequent RV DC RV DC Sequence Initiating Power Temperature Pressure Number Event Level Failures or Conditions (psia) (* F)
, 84 Large steam break 115B AFW not isolated 204 1947 8-6 Large steam break flSB AFW not isolated, AFW overfeed 201 1995 6-5 Large steam break Full AFW not imlated 229 1772 6-6 Large steam break Full AFW not isolated, charging not throttled 227 2371 6-7 Large steam break Full AFW not isolated, AFW overfill 229 1772 6-9 Large steam break Full AFW not isolated, AFW overfeed 229 1781 9-10 Reactor trip Full 2 stuck-open SDVs, AFW overfill 248 1518 9 12 Reactor trip Full 2 stuck-open SDVs, AFW overfill, charging 241 2371 not throttled 9-14 Reactor trip Full 3 stuck open SDVs 234 1610 9-15 Reactor trip Full 3 stuck-open SDVs, AFW overfill 226 1613 9-16 Reactor trip Full 3 stuck-open SDvs, charging not throttled 225 2371 9-17 Reactor trip Full 3 stuck-open SDVs, AFW overfill, charging 220 2371 not throttled 9-19 Reactor trip Full 5 stuck-open SDVs 203 1594 9-20 Reactor trip Fult 5 stuck-open SDVs, AFW overfill 201 1575 9-21 Reactor trip Fult 5 stuck-open SDVs, charging not throttled 199 2371 9-22 Rector trip Full 5 stuck-open SDVs. AFW overfill, charging 1% 2371 not throttled 9-23 Reactor trip Full 5 stuck-open SDVs, AFW overfeed 213 1669 9-44 Reactor trip Full 3 stuck-open steam PORVs. AFW overfill, 244 2371 charging not throttled , 5-20 Small steam break Full 2 stuck-open SDVs AFW not isolated 244 1618 7-9 Small steam break ilSB 1 stuck-open SDY 239 1508 7-10 Small steam break IISB I stuck-open SDV, AFW not isolated 229 I470 7 11 Small steam break 115B 2 stuck-open SDVs 209 1537 17.2 Reactor Trip with higher temperatures and higher pressures than cor-Minor Failures responding sequences with a medium break LOCA.
Nine sequences initiated by reactor trip, follow- Both the medium and small break sizes were large g g ed by minor control system failures, and involving no primary- or secondary-stde breaks were by decay heat within two hours. Consequently, the mvestigated. All nine of the sequences were found reactor coolant system could be cooled and not to be severe. The minimum temperatures [503 K depressurized with either break size. Loop flow (446*F)] were calculated for sequences mvolvmg stagnation was calculated to occur only when the steam generator overfill using auxiliary feedwater. U-tubes were voided. The medium break was capable of voiding the U-tubes, but the small break was not. Loop flow stagnation did not occur when 17.3 Pn, mary-Side Breaks the loops were liquid rull, even with heat transfer from the steam generators to the primary system. Nineteen sequences involving only primary-side breaks were investigated. Sequences with a medium Sequences involving isolation of a primary-side break LOCA resulted in reactor vessel downcomer break resulted in high primary system pressures. As temperatures a few degrees above the HPI a result of isolation during the early stage of the temperature; however, the resulting primary system sequence, however, severe overcooling of the pressures were very low. thus limiting PTS concern. primary system did not occur before the time of Sequences with a small break LOCA resulted in isolation, therefore minimizing PTS concern. 93
17.4 Combination Primary- and For all sequences, resuhing temperatures are very Secondary-Side Breaks low [344 K (160 F) for Sequence 2-8 is the lowest]. liowever, resulting pressures are also low [2.54 MPa (368 psia) for Sequences 1-9 and 1-11 is Sixteen sequences involving combination of the highest}, thus limiting PTS concern for com-primary- and secondary side breaks w ere bined breaks. investigated. In general, these sequences are not as severe as the corresponding sequences involving 17,5 Steam Generator only the secondary-side break. The sequences with Tube Ruptures ~ combined breaks have temperature responses similar to a corresponding secondary-side-break- Five sequences involving the rupture of a single only sequence, and pressure responses similar to a steam generator tube with the reactor at hot stand-corresponding primary-side-break-only sequence. by conditions were investigated. Repressurization of the primary system is limited by the broken tube. For sequences with a symmetric, secondary-side m n mum temperatures are enentiaHy deten break (stuck-open steam dump valves), natural cir. rmned by the operator actions and failures specified culation continued in all loops. For sequences with in tk sequence descr4 tion. For exan@, U tk an asymmetric, secondary-side break (one or two g rat r anume to open tk steam dump vahe stuck open steam line PORVs), natural circulation until a subcooling criterion is met, the magnitude continued only in the affected loop or loops. c wn sgc y th anumption. The steam generator tube rupture sequences were, Sequences involving a medium, Primary-side in general, not severe, except for Sequence 10-2 break resulted in lower reactor vessel downcomer which involved a 10-min period with a stuck-open temperatures and pressures than did the cor- steam dump valve. This sequence resulted in a responding sequences with small, primary-side minimum temperature of 40. K (277 l') and a max- , breaks. imum subsequent pressure of 7.45 MPa (1080 psia).
- 18. REFERENCES
- 1. C. D. Fletcher, et al., "RELAP5 Thermal-liydraulie Analyses of Pressurized Thermal Shock Sequences for the II. B. Robinson Unit 2 Pressurized Water Reactor," EG&G idaho, Inc., NUREG/CR-3977, April 1985.
- 2. C. D. Fletcher, et al., "RELAP5 Thermal-liydraulie Analyses of Pressurized Thermal Shock Sequences for the Oconee-1 Pressurized Water Reactor," EG&G Idaho, Inc., NUREG/CR 3761, EGG-2310,
- June 1984.
- 3. D. G.11all and L. S. Czapary, " Tables of flomogeneous Equilibrium Critical Flow Parameters for Water in SI Units," EG&G Idaho, Inc., EGG-2056, September 1980.
l O l
~
t t 94
m 9 APPENDIX A PLOTTED RESULTS OF REACTOR VESSEL DOWNCOMER FLUID PRESSURES AND TEMPERATURES e e e A1
APPENDIX A PLOTTED RESULTS OF REACTOR VESSEL DOWNCOMER FLUID PRESSURES AND TEMPERATURES This appendix contains plotted results showing The analyses presented in this report were per-the reactor vessel downcomer fluid pressure and formed using best-estimate modeling assumptions temperature histories for each sequence analyzed. for plant conditions and responses to the events Results are shown for the 2-hr period immediately specified in the sequence descriptions. The reader following the initiating event, is cautioned, however, that the sequence descrip-tions were based on extremely conservative assump-Figures in this section are arranged by numerical se- tions concerning equipment malfunctions, operator quence, first by table number and second by the line actions and omissions, or combinations of these, number within each table (see Section 2). For exam- Thus, while the analyses results represent the best-ple, Sequence 7-6 is the sixth sequence shown in estimate plant responses to the sequences as defined, Table 7, and, in this Appendix, results for Sequence 74 they do not necessarily represent the most probable are found following those for Sequence 7-5. plant responses to the sequence initiating events. 9 e i e A-3
16000 r ' ' 14000 , - ~ O O
- n. "
6 12000 * - O - L 3 '
-1500 @
E 10000 - - m e 0 L Q O-8000 - _
-1000
, t I 6000 0 2000 4000 6000 8000
, Time (s) Figure A 1. Sequence 11 reactor vessel downcomer fluid pressure. 360 i 8 '
-540 550 -
p n
-520 b
- 540 -
_ e 3 ' g *- u
-500 O g
e a.530 _ E
- E p e
-480 520 -
510 1 I , -460 0 2000 4000 6000 B000 Time (s) . Figure A-2. Sequence 1 1 reactor vessel downcomer fluid temperature. A-4
20000 , , ,
-2500 15000 - -
Q -
-2000 E a m U v 0-
[ 10000 ~ --1500 e 3 3
- m M m 8
8 -
-1000 Q- n.
5000 - -
-500 0 O O 2000 4000 6000 8000 Time (s)
Figure A 3. Sequence 12 reactor vessel downcomer fluid pressure.
. 560 , , i 540 - -
n -
-500 ^
M 12-v v 8 520 - -
- 3 3 8 -
-450 8 (500
[ E E o e
& H 480,- --400 460 O 2000 4000 6000 8000 Time (s)
Figure A-4. Sequence 12 reactor vessel downcomer fluid temperature. A-5
16000 , , , 14000 ~
~-2000 . ^ m O g Q- -
m 6 12000 O
~ ' C 3 L $ 10000 - -1500 @
c m
' c Q. L Q.
8000 - _
~
-1000 6000 ' i i 0 2000 4000 6000 8000 Time (s)
I~igure A 5. Sequence 13 reactor vessel downcomer fluid pressure. 560 , , ,
-540 550 - _
^ ^
~
-520 b 8 540 -
C 3 O 0 L
-500 ES30
. e E E e *
~'
520 - 5t0 ~ i i i -460 0 2000 4000 6000 8000 ' Time (s) l Figure A-6. Sequence 13 reactor vessel downeomer fluid temperature. A-6
'20000 , , ,
-2500
. 15000 -
9 a.
-2000 -
0 w M U
=
h u 10000 -- --1500 e 3 '- W 3 W M. e M w -
-1000 e 5000 - -
I
-500 0 ' ' '
O O 2000 4000 6000 8000 Time (s) Figure A-7. Sequence 1-4 reactor vessel downcomer fluid pressure. MO , , , 540 - -
^ -
500 ^ M &^- v v 8 520 - -
- 3 3-D 0 g -
450 g [500 - - [ E c e 3 H >= 480,- -400 460 O 2000 4000 6000 8000 Time (s) Figure A 8. Sequence 14 reactor vessel downcomer fluid temperature. A7
20000 i i i
-2500 15000 - -
n . o -
-2000 0
- a. ,;
6 8 ~
' --1500 e u 10000 '-.
u 5 3 a m
- e C u -
-1000
"- I 5000 - -
-500 0 O O 2000 4000 6000 8000 Time (s)
Figure A-9. Sequence I 5 reactor vessel downcorrer fluid pressure. 600 , , ,
-600 550 -
^ -
-500 ^
M l'- v v 8 500 -
- e
= 2 0
-400 +
g
' u E450 g
E E
- 0 p -
-300 400 - -
-200 350 O 2000 4000 6000 8000 Time (s)
Figure A 10. Sequence 15 reactor sessel downcomer fluid temperature. A-8
I-20000 , , ,
-2500 15000 -
. m y -
-2000 a G- m O
8 10000 ~ - 1500 e 3
- 3 M m m
8 -
-1000 C a- n.
5000 - -
-500 0 O O 2000 4000 6000 8000 Time (s)
Figure A-II. Sequence 1-6 reactor vessel downcomer fluid pressure. 600 , , i
-600 550 -
n -
-500 ^
M v l'- v 8 500 - - 8 3 3
-400 -
0 0 L L [450 - - [ E E 8 -
-300
- 400 - -
-200 350 O 2000 4000 6000 8000 Time (s) l'igure A 12. Sequence 16 reactor vessel downcomer fluid temperature.
A9
r 20000 , , ,
-2500 15000 -
- Q Q-
-2000 ,o m
6 v C- - 8 10000 ~ --1500 e 3 3
- m M
m 8 -
-1000 S Q- n.
5000 - -
-500 0 ' ' '
O O 2000 4000 6000 8000 Time (s) Figure A 13. Sequence 17 reactor sessel downcomer fluid pressure. 600 i i i
-600 550 -
m .
-500 ^
M v l'- v 8 $00 - - [ 3 3
- -400 +-
0 0 6 L [450 - - E E E o -
-300 8 g
400 - -
-200 . 350 O 2000 4000 6000 8000 ,
Time (s) Figure A 14. Sequence 17 reactor vessel downcomer fluid temperature. A.10
20000 , , ,
-2500 15000 _
m m O
-2000 o
- n. -
+
U v 8 10000 - _-1500 e 3 u M 3 M M
@ M
~
-1000 8 n'. '
L 5000 - _
-500 0 O O 2000 4000 6000 8000 Time (s)
Figure A 15. Sequence I-8 reactor vessel downcomer Guid pressure. 600 , , ,
-600 550 -
^ -
-500 ^
b b v 8 500 -- - e 3 $ 0
-400 +
' 0 L
$450 g
E E
-300 8 400 - -
-200 350 -
O 2000 4000 6000 8000 Time (s) l l ligure A 16. Sequence I 8 reactor vessel downcomer Guld temperature. A Il
20000 i i i
-2500 15000 -
g -2000 0
~
- a. -I s 6 S ,
8 10000 -- - 1500 e 3
- 3 m m m
8 -
-1000 8 0-CL 5000 -
-500 0 O O 2000 4000 6000 8000 Time (s)
Figure A 17. Sequence I 9 reactor vessel downcomer Guid pressure. 600 e i i
- 550 -
n M -500 ^ v b v
,S 500 -
3 3 0 400 + L 0 6 [450 - E E E o 8 g - 300 400 -
-200 350 ' ' '
O 2000 4000 6000 8000 Time (s) - Figure A 18. Sequence l 9 reactor sessel downcomer duid temperature. A 12
I i i 20000 i , i l 2500 15000 - n 9 Q-
-2000 ,o m . O I 10000 -- -
1500 e a M 2
- m m
8 -
-1000 8 E- n.
5000 - - 500 o ._ 0 O O 2000 4000 6000 8000 Time (s) Figure A 19. Sequence 1 10 reactor vessel downcomer Huid pressure. 600 , i i
-600 550 - -
^ -500 ^
M v l'- v
,8 500 - - 8 3 3
-400 +
0 0 L L E450 1 E E g -300 8 400 - -
-200 350 O 2000 4000 6000 8000 Time (s) figure A 20. Sequence 1 10 reactor sessel downcomer Huid temperature.
A 13
r 20000 , i i 2500 15000 -
^ -
m
-2000 0 S l
'm dS e
S - u 10000 __ _-1500 e 3 u m ' 3 M m M u' 1000 8 0. o 5000 -
-500 0 t '
O O 2000 4000 6000 8000 Time (s) rigure A 21. Sequence 1 11 reactor vessel downcomer fluid pressure. 600 i i i 600
$50 -
m -500 ^ M b v v 8 500 -
~
- u
? -400 5 0 0 L '
E450
- ~
E E e -300 8
& H 400 - ~
-200 350 ' ' '
O 2000 4000 6000 8000 , Time (s) l'igure A 22. Sequence 1 11 reactor senel downeomer fluid temperature. A 14
20000 , , ,
-2500
, 15000 - m o -2000 C
- a. ,
6 - Q-8 10000 -- _ -1500 e a =
=
a 8 -1000 0 1 5000 - c.
-500 0 O O 2000 4000 6000 8000 Time (s) l'igure A 23. Sequence 112 reactor vessel downcomer fluid pressure.
600 , i i .,
-600 550 -
^ 500 ^
M l'- v v 8 500 -
- 3 3
+
0 400 + L 0 L {450 - [ E E g -300 8 400 - 200 350 ' ' ' O 2000 4000 6000 8000 Time (s) l'igure A 24. Sequence 1 12 reactor senel downeomer fluid temperature. A 15
16000 , , , 14000 - _-2000 m n O
- n. .O 6 12000 _
{ v . L e 3 ' L N 10000 - 1500 @ e
,n 6 e G. L Q.
8000 - .
~
1000 6000 i ' i 0 2000 4000 6000 8000 Time (s) Figure A 25. Sequence 113 reactor vessel downcomer Guid pressure. 560 , , ,
-540 550 -
m b" ~ 520 b 8 540 - _ e 3
?
6
~
500 0
$530 -
e E E e
- 480 520 - -
i i i 460 510 - 0 2000 4000 6000 8000 Time (s) - Figure A 26. Sequence 113 reactor vessel downcomer Guld temperature. A 16
20000 , , ,
-2500 15000 -
m 7 a.
-2000 0 V" v a.
8 10000 - --1500 e 3 ' w 3 m M e " u -
-1000
- Q-5000 -
I
-500 0 ' ' '
O O 2000 4000 6000 8000 Time (s) ; Figure A 27. Sequence 1 14 reactor vessel downcomer fluid pressure.
$60 T i i 540 -
^ -
-500 ^
M l'- v v l
- 520 - -
- i 3 3 0 0 w -450 g (SCO -
E E E e e H H 480_- --400 460 ' ' ' O 2000 4000 6000 8000 Time (s) Figure A 28. Sequence 1-f 4 reactor sessel downcomer fluid temperature. A 17
16000 i i i 4 14000 - --2000 n g ' E 7 6 12000 - S , e e
' u j .
-1500 m 10000 - - @
e e L e L Q- Q. 8000 - -
-1000 6000 O 2000 4000 6000 8000 Time (s)
Figure A-29. Sequence 1-15 reactor vessel downcomer fluid pressure. 560 i i i
-540 550 - -
a ^ 6 -
-520 b 8 540 - - '
3 3 0 -
-500 O g
(530 - - E E E o e
-480 520 - -
510 - i e i -460 0 2000 4000 6000 8000
~
Time (s) Figure A-30. - Sequence 1-15 reactor vessel downcomer fluid temperature. A-18 _
i 20000 , ,, ,
-2500 15000 -
o -
-2000 o
- a. -
i- . I ' 10000 -- -1500 e s .
- 3 m M e m
-1000 e CL '
5000 -
-500 0 o 0 2000 4000 6000 8000 Time (s)
Figure A-31. Sequence I-16 reactor vessel downcomer fluid pressure. 600 , , , j
-600 i l
)
550 _ Q
-500 p v
8 500 - _ e 3 b
-400
]u +-
o 6 [450 - e o. E E
-300 g 400 -
_ i l
-200 350 ' ' '
O 2000 4000 6000 8000 Time (s) Figure A-32. Sequence 1-16 reactor vessel downcomer fluid temperature. A-19
16000 , , , 14000 - _-2000 7 a.- 7 6 12000 . v E e . L e 3 L E ' 10000 -
-1500 y e v, L. e
- n. L Q.
8000 - _
-1000
~
6000 ' ' ' O 2000 4000 6000 8000 Time (s) Figure A-33. Sequence 1-17 reactor vessel downcomer fluid pressure. 560 , , ,
~
-540 550 -
^ ^
~
-520 D 8 540 -
e 3 0 0 6
-500 (530 - _ e E E e
- r 520 -
510- ' ' ' -460 0 2000 4000 6000 8000 Time (s) . Figure A 34. Sequence 1-17 reactor vessel downcomer fluid temperature. A-20
16000 , , , 14000 _ -
--2000 . m m o o n- -
6 12000 - v E e L 8 3 ' e -
-1500- 3 m 10000 - - M a e m ,
L 8 Q. ' , C-8000 - -- -
-1000 I
6000 < 0 2000 4000 6000 8000 , Time (s) Figure A-35. Sequence 1-18 reactor vessel downcomer fluid pressure. 4 J-e A*
- 560 , , ,
-540 550 -
a m b -
-520 b 8 540 - - e 3 3
.e-0 -
-500 0
[530 - g E E 8 e
& W
-480 520 -
510 - ' ' ' -460 0 2000 4000 6000 8000 Time (s) Figure A-36. Sequence 1-18 reactor vessel downcomer fluid temperature. A-21 w r---r ,,4.-. -- w - - - ,.m , . . - - - - , - - - - - , -.--a ., - ,- --e- - -,s
r --- 20000 i i ,
-2500 15000 -
9a.
-2000 -
0
~d CL .
[ 10000 ~ - _-1500 e 3 ' e 3 m: M e - 6 -1000 8 1 5000 - I
-500 0 ' ' '
O O 2000 4000 6000 8000 Time (s) Figure A-37. ' Sequence 2-1 reactor vessel downcomer fluid pressure. 600 i , _ .i 550 -
-500 m ^
M b v 500 -
- V f -
-400
- 3 3
]L 450 - -
g e u Q.
-300 e o.
E 400 - E e e F- >-
-200 350 -
300 O 2000 4000 6000 8000 ^ Time (s) a Figure A-38. Sequence 2-1 reactor vessel downcomer fluid temperature. A-22
-20000 , , ,
-2500 15000 -
'~
7 a.
-2000 0 6 c.
I 10000 - -
--1500 e 3 '
M U M M e
- u -
-1000 8 1
n. 5000 -
-500 v
0 O O 2000 4000 6000 8000 Time (s) Figure A-39. Sequence 2-2 reactor vessel downcomer fluid pressure.
. 600 , , i ;
550 -
-500
! n ^ M l'- l v 500 - - v u -
-400 3 3
*O 450 -
l 6 U L l 8 -
-300 e i Q. n.
! E 400 - - E l t o e I H H l
-200 I 350 - -
i 300 0 2000 4000 6000 8000 Time (s) Figure A-40. Sequence 2-2 reactor vessel downcomer fluid temperature. t A-23 L
if 20000 , , r'
~
-2500 15000 - -
9 n.
-2000 o 5 a 8 10000 - - -
1500 cu 3 " m m M m 8 -
-1000 8 1 Q.
5000 - -
-500 e
0 O O -2000 4000 6000 8000-Time (s) Figure A-41. Sequence 2-3 reactor vessel downcomer fluid pressure. 600 , , , 550 - -
-500 m ^
M 18-v 500 - - v E. -
-400 S u
3 3
$ 450 - -
fL U e o.-
-300 c.
E v 400 E H e W
-200 350 - -
' ' ' -10 0 300 O 2000 4000 6000 8000 Time (s) -
Figure A-42. Sequence 2-3 reactor vessel downcomer fluid temperature. A-24
20000 i . .
-2500 15000 -
^ -
-2000 0 2 m s 3 o --1500 o e 10000 - -
a u m a to m
- m
-1000
- O Q.
5000 -
-500 s
0 O O 2000 4000 6000 8000 Time (s) Figure A-43. Sequence 2-4 reactor vessel downcomer fluid pressure. 600
-600 550 -
-500 m ^
y . "l'- v 500 - 0 S'
-400 u 3 3
$ 450 -
Y
.u u o .
-300 0
. o. o.-
(400 F. h H- -
-200 350 -
' ' ' -10 0 300 O 2000 4000 6000 8000 Time (s) .
Figure A-44. Sequence 2-4 reactor vessel downcomer fluid temperature. A-25
. _ _- ._ , . . . _ - . - - ~ - . . .
20000 i , .
-2500 15000 -
9 a.
-2000 O 6 & .
8 -10000 - -
..-1500 o 3
M 3 M M M I -
-1000
- Q-O_
5000 - 4
-500 I
O ' ' ' O O 2000 4000 6000 8000 Time (s) 4 Figure A-45. Sequence 2-5 reactor vessel downcomer fluid pressure. l 600 , , . 600 550 -
^ -
500 ^ M b v v C L 500
- - S u
3 3 , O
-400
- L O
L E450 - - [ E E g' -
-300
- 400 - -
-200
-350 ' ' '
~ 0 2000 4000 6000 8000 .
- Time (s) 1 i Figure A-46. Sequence 2-5 reactor vessel downcomer fluid temperature.
1 2 A-26
20000 i i i
-2500
- 15000 -
m g -2000 ,0 O_ m 6 0 e 10000 _-1500 e u - 3 L us 3 tn m O m
-1000
- Q, O.
5000 - ~
-500 e ' '
0 O O 2000 4000 6000 8000 Time (s) Figure A-47. Sequence 2-6 reactor vessel downcomer fluid pressure. 600 i i 5
-600 550 ~
n -
-500 ^
6 V-0 500 - - o g 3 .
-400 g o
' u o
E450 - ~ g
-300
- F-g 400 - -
-200 t I I 350 0 2000 4000 6000 8000 Time (s)
Figure A 48. Sequence 2-6 reactor vessel downcomer fluid temperature. A-27
20000 , , i
-2500 15000 -
9 a.-
-2000 9 6
8 . 8 10000~ -
--1500 e a '
-1000
- O-Q.
5000 -
-500 0 '-
0 0 2000 4000 6000 8000 Time (s) Figure A-49. Sequence 2-7 reactor vessel downcomer fluid pressure. 600 i i i
~
-600 550 -
n Y -500 ^ v 12-v 0 500 - 0 3 3
+
O
-400 s L O L
$450 -
E E O
- g -
-300 400 -
-200 350 O 2000 4000 6000 8000 Time (s)
- Figure A-50. Sequence 2-7 reactor vessel downcomer fluid temperature.
A-28
20000 - - - i r- i
-2500 15000 - -
n-
-2000 [
m 0 8 10000 ' - - 1500 e 3 a m m M m 8 -
-1000
- Q- Q-5000 - -
-500 0 O O 2000 4000 6000 8000 Time (s)
Figure A-51. Sequence 2-8 reactor vessel downcomer fluid pressure. . 600 , i i 550 -
-500 m ^
Y v 500 - - b v
- 0
-400 3 3
$ 450 u
y u { -
-300
{ E o 400 E a> F-- F--
-200 350 - -
300 O 2000 4000 6000 8000 . Time (s) Figure A-52. Sequence 2-8 reactor vessel downcomer fluid temperature. A-29
20000 -
-2500 15000 -
O -2000 0 a . . _ d . I 10000 - - _ -1500 e a ' W 3 m m Q m 0 Q.
-1000 '
Q. 5000 -
-500 0 ' ' '
o 0 2000 4000 6000 8000 Time (s) Figure A-53. Sequence 2-9 reactor vessel downcomer fluid pressure. 600 , , ,
-600 550 - ^ -500 ^
M v 12-v 0
' 500 -
- o u
3 3
~
0
-400 s ' 0 L $450 -
{ E E [ -
-300 [
400 -
-200 350 ' ' '
O 2000 4000 6000 8000 . Time (s) Figure A-54. Sequence 2-9 reactor vessel downcomer fluid temperature. A-30
20000 , , ,
-2500 i
15000 - 9 a
-2000 .o--
.2 M O
8 10000 - - - 1500 e a " m 3 , m M
@ M
-1000
- 1 5000 -
I - 500 1 t 0 O. 0 2000 4000 6000 8000 [. Time (S) - ! Figure A-55. Sequence 2-10 reactor vessel downcomer fluid pressure. i e - . 600 , , ,
-600 550 -
^ -
-500 ^
M 12-v v
'0 500 - -
- u
~O 3 1
-400 s
" 0 0 L L
$450 - -
1 E E [ -
-300 ['
400 - -
-200 350 i .
0 2000 4000 . 6000 B000 Time (s) 7 Figure A-56. Sequence 2-10 reactor vessel downcomer fluid temperature. P I k A-31
,U-, ,,v-o . , . , ..+-M.,,ev- t--+w-+-,-.- -v -,.t- --e- , , ., - . - . - r
20000 - -- - r- - - i i
-2500 15000 -
o -
-2000 a .o 6
5 . 8 10000 - - 1500 o a ' m 3 us D Q U L -
-1000
- 1 Q.
5000 -
-500 0 ' ' '
O O 2000 4000 6000 8000 Time (s) Figure A-57. Sequence 2-11 reactor vessel downcomer fluid pressure. 600 , , ,
-600 1
550 - n -
-500 ^
M b v v
- 500 -
0 3 3 0
-400 s > ' o L $450 -
1 E E p -
-300 g 400 -
-200 350 O 2000 4000 6000 8000 Time (s) -
Figure A-58. Sequence 2-1I reactor vessel downcomer fluid temperature. A-32
. . . . . . - . - . . - . _ . . - . _ . . - . _ . - ~ . _ -
d 4
'20000 , , ,
-2500
.15000 -
* ^
n o -
-2000 0
- a. -
-[3 i
10000 'i. --1500 e 4 p 3
- m M 1
.e m u -
-1000 8 Q-5000 -
k l
-500 0 O O 2000 4000 6000 8000 Time (s)
Figure A-59. Sequence 3-1 reactor vessel downcomer fluid pressure. ' r J e I
' -- 600 , , ,
-600 m 550 -
n M l'-
-500 v a o e
' L 3 3
,- *O 500 -
- +-
1 i L n L 2. e e O. -
-400 a.
E E
= e 450 -
- H
-300 400 ' ' '
1- . 0 2000 4000 6000 8000 Time (s) Figure A-60. Sequence 3-1 reactor vessel downcomer fluid temperature. f l A-33
1 20000 , , ,
-2500 15000 - -
- 9 a-
-2000 o 5 3 .
8 10000 ~ --1500 e i u S 3 a I -
-1000 e A c.
5000 - -
-500 0 O O 2000 4000 6000 8000 Time (s)
Figure A-61. Sequence 3-2 reactor vessel downcomer fluid pressure. 600 , , ,
-600 550 -
^ -
-500 ^
M ld-v v 8 500 - 3 3
~
0
-400 *--
6 0 y l, [450 - - [ l E E , g' -
-300 .e 400 - -
-200 350 ' ' '
! 0 2000 4000 6000 8000 - Time (s) Figure A-62. Sequence 3-2 reactor vessel downcomer fluid temperature. A-34
i 20000 , , ,
, -2500 , . 15000 -
n m o -
-2000 .o a
.x m v
v a. E 10000 - -
.-1500 u
e 3 vs a en en Q M
-1000
- i CL '
O 5000 -
-500 i
0 ' ' ' O 0 2000 4000 6000 8000 Time (s) , Figure A-63. Sequence 4-1 reactor vessel downcomer fluid pressure. 4 600 , , , 600 550 - 500 m n M 12-V 500 - v u
-400 3 m j 450 - -
g
'v -
-300 L
o C- ct. E e 400 E i o W
-200 3
350 - 1 ; 300 ' ' ' i , 0 2000 4000 6000 8000 Time (s) Figure A-64. Sequence 4-1 reactor vessel downcomer fluid temperature. 4 f f 4 4 ) A-35 4
, --. - - _,., - ,.-_- - .--- , - . . . - , , - . . - . _ , . , - , _ . , _ , , . - , .,, - - ~ . . _ - ,
. - - - ,.- , .~ . -
. _ _ _ . _ - = _ _ _ . _ -
+ i l 15840 , , , 15820 -
--2295 m n -
g 15800 - 0 a m
-2290 O ,
[ 15780 -
- e 3
M 3 m m e m
' 15760 -
-2285 I
- i. i 15740 -
15720 2280 0 2000 4000 6000 8000 Time (s) Figure A-65. Sequence 5-1 reactor vessel downcomer fluid pressure. 3 559- i i i - 546.5
- 546.0 m ^
M
" 558.5 - - v l'-
e~ -
- 545.5 e L L 3 3
*- e-0 0 g -
- 545.0 g
- o. a.
E 55a - - E e e
- 544.5 i, .
- 544.0 557.5 O 2000 4000 6000 8000 Time (s) -
Figure A-66. Sequence 5-1 reactor vessel downcomer fluid temperature. l, A-36
'i
. - . = ... . . . . .. .. . . - - - . - - - . . . . . . - . . . . -
15840 , , ,
-2295 15820 ~ - -
m n g 15800 - -
.U a m
-2290 O
' e u 15780 - -
3 3 M M M e " 8
' 15760 -
g 1~ -
-2285 n.
15740 - 15720 2280 0 2000 4000 6000 8000 Time (s) [ Figure A-67. Sequence 5-2 reactor vessel downcomer fluid pressure. 559 - i i i - 546.5
- 546.0 m n M
" b v
558.5 - - e -
- 545.5 e
' L 3
+- 3.
O o 5
- 545.0 g
- a. a E 558 - -
E
- e
- 544.5 *
- 544.0 557.5 O 2000 4000 6000 8000 Time (s) l-Figure A-68. Sequence 5-2 reactor vessel downcomer fluid temperature.
l A-37
- ,r - , - - - > - ~ , - -
,n, , , - - -,-n- - - - - ----,,,--n . - . - ~ - - - ,
15840 , , , 15820 '- - n ^ U g 15800 -
-2290 0 -
8 15780 -
- 8 a o
=
- n
' 15760 -
Q- u
-2285 c.
15740 - 15720 2280 0 2000 4000 6000 8000 Time (s) Figure A-69. Sequence 5-3 reactor vessel downcomer fluid pressure. 559 - i i i 546.5 546.0 m ^ M b v 558.5 -
- v e -
- 545.5 e 3 3
+-
o o g -
- 545.0 '
- o. a.
E 558 - - E e e
- 544.5 *
- 544.0 557.5 O 2000 4000 6000 8000 ,
Time (s) . Figure A-70. Sequence 5-3 reactor vessel downcomer fluid temperature. A-38
~ e-----
- , , - - - - . , - , - a .e , - - - --&
15840 , , , 15820 '- --2295 m n 0 [ 15800 - -
-2290 8 15780 - - e 3 '
M 3 m M e M
' 15760 -
e
-2285 I 15740 -
15720 2280 0 2000 4000 6000 8000 Time (s) Figure A-71. Sequence 5-4 reactor vessel downcomer fluid pressure, e 559- , i i - 546.5
- 546.0 m a M l'-
" 558.5 -
e -
- 545.5 e
' L 3 3
*~ .e.
O o
- 545.0 g 0- c.
E 558 - - E
- 544.5
- 544.0
' 557.5 O 2000 4000 6000 8000 Time (s) l Figure A-72. Sequence 54 reactor vessel downcomer fluid temperature. I A-39
? 15840 , , , i 15820 ' -
--2295 m n .
g 15800 -
- ,0
.x M
-2290 O
- e u 15780 -
- e 3 '
- m. 3 M M e M 8
' 15760 -
-2285 I i
15740 - l 15720 2280 0 2000 4000 6000 8000 Time (s) Figure A 73. Sequence 5-5 reactor vessel downcomer fluid pressure. ? 4 559. i i i - 546.5 546.0 m a M l'-
" 558.5 - - v e -
- 545.5 e L- g
. 3 3
*~ .e-l 0 o
- 545.0 '
O- O. E 55a - - E
- e
- 544.5 *
- 544.0 557.5 i 0 2000 4000 6000 8000 Time (s) - -
Figure A 74. Sequence 5 5 reactor vessel downcomer fluid temperature. J 4 A-40
15840 , , 15820 ~ - -
-2295 .. m n 0
[ 15800 - - M
.x
-2290 O
[ 15780 - - e 3 ' e 3 m M e M 8 L 15760 - 1 -
-2285 I 15740 -
15720 2280 0 2000 4000 6000 8000 Time (s) Figure A-75. Sequence 5-6 reactor vessel downcomer fluid pressure. , e.
- 559. i i i - 546.5
- 546.0 m m M
v l'-
" 558.5 - -
e -
- 545.5 e
' u 3 3 O o .
5
- 545.0 g 0- CL
-E 558 - -
E
- e
- 544.5
- f
- 544.0 557.5 O 2000 4000 6000 8000 Time (s) 7 Figure A-76. Sequence 5-6 reactor vessel downcomer fluid temperature.
L A-41 , i
. . _ _ _ . _ _ . _ _ . . -.- m _ __ . _ _ _ _ _ . . _ . _ _ _ .
15840 , , , 15820 '- --2295 n n [ 15800
- ,0
.x "
v - U *
-2290 v" I 15780 -
- e 3 '
n 3 m M e "
' 15760 -
4 -
-2285 I 4-15740 -
15720 ' ' ' 2280 l 0 2000 4000 6000 8000 Time (s)
, Figure A-77. ' Sequence 5-7 reactor vessel downcomer fluid pressure.
i 1-i i' ' 559 - a i - 546.5
- 546.0
' a m i M la-i " 558.5 -
- v e -
- 545.5 e L u 3 3 O' c
- i. ' O
- 545.0 e O. O.
, E 558 - - E 1 e e
- 544.5
- 2 -
- 544.0 557.5 ' ' '
O 2000 4000 6000 8000 ' Time (s). Figure A-78. Sequence 5-7 rIactor vessel downcomer fluid temperature. 1-A-42 i
15840 , , , 15820 ~ -
-2295 m n 0
g 15800 -
.x M
-2290 O 15780 -
- e
,u . u e 3 M M e M
' 15760 -
-2285 I 15740 -
15720 2280 0 2000 4000 6000 8000 Time (s) Figure A-79. Sequence 5-8 reactor vessel downcomer fluid pressure. 559- i i - 546.5
- 546.0 m ^
M l'- V
" 558.5 - -
e -
- 545.5 e L. L 3 ~3 e- .e .
O O
- 545.0 '
e e Q. Q. E 558 - - E e e
- 544.5
- o-
- 544.0 557.5 O 2000 4000 6000 8000 Time (s)
Figure A-80. Sequence 5-8 reactor vessel downcomer fluid temperature. A-43
15840 i , i 15820 ~ - - n n 0 -
- a. 15800 .
b -
-2290 O ,
[ 15780 -
.o ,
- m
- M 6 15760 -
e a- u
-2285 n.
15740 -
}
15720 2280 , 0 2000 4000 6000 8000 Time _ (s) Figure A-81. Sequence 5-9 reactor vessel downcomer fluid pressure. 559 i i i - 546.5
- q 546.0 m ^
M " l'- 558.5 - - v e -
- 545.5 e L L 3 3 e- -
D , o g -
- 545.0 g
- o. c.
E' 558 - - E e e
- 544.5
- 544.0 557.5 O 2000 4000 6000 8000 Time (s) -
Figure A-82. Sequence 5-9 reactor vessel downcomer fluid temperature. A44
d 15840 i i i
-2295 15820 - - -
0 Q- 15800 - - M
. b -
-2290 O I 15780 - - 8 3 o
# M
- m
- e
' 15760 -
u Q- -
-2285 cL 15740 -
15720 2280 0 2000 4000 6000 8000 Time (s) Figure A-83. Sequence 5-10 reactor vessel downcomer fluid pressure. t l (.* 559 - 6 i i - 546.5 l
- 546.0 m a M l'-
" 558.5 -
- v e --
- 545.5 e
' u 3 3
+ -
0 0
- 545.0 Q- n.
E 558 - - E
- e
- 544.5 *
~
- 544.0 557.5 O 2000 4000 6000 8000 Tirne (s)
Figure A-84. Sequence 5-10 reactor vessel downcomer fluid temperature. A-45
15840 , , , 15820 '- _-2295 m n [ 15800 _ ,0 a m
-2290 O
- 8 15780 -
_ e 3 ' n 3 m n e M 15760 C n'.
~
-2285 c'.
15740 _ 15720 ' ' ' 2280 0 2000 4000 6000 8000 Time (s) Figure A-85. Sequence 5-11 reactor vessel downcomer fluid pressure. 559 - i i i - 546.5
- 546.0 ^ m M l'-
558.5 - _ v e -
- 545.5 e ' L 3 *- 3_
O O l
- 545.0 g
- o. a.
E 558 - - E
- e
- 544.5 *
\ -
- 544.0 557.5 C 2000 4003 6000 8000 ~
Time (s) Figure A-86. Sequence 5-11 reactor vessel downcomer fluid temperature. A-46
15840 i i i 15820 - -
-2295 m
[ 15800 Q 3
-2290 vc.
v - [ 15780 -
- 0 3
M 3 M m S m L 15760 - e Q- t n.
-2285 15740 -
k ' ' ' l 15720 2280 0 2000 4000 6000 8000 Time (s) Figure A-87. Sequence 5-12 reactor vessel downcomer fluid pressure. l
. 559 - i i i - 546.5
- 546.0 m
y O v P-558.5 -
-- U e -
- 545.5 e u
L 3 3
~
0 0 O
- 545.0 O Q. O.
E 558 - - E
. e
- 544.5
- 544.0 557.5 O 2000 4000 6000 8000 Time (s)
Figure A.88. Sequence 5-12 reactor vessel downcomer Huid temperature. A-47
15840 . . .
~ '
15820 - -
=
m g 15800 - - x a
' V -
-2290 v .
u 15780 - - 3 3
'M g
# M
- e L .15760 -
u 1 -
-2285 a_
15740 - 15720 2280 0 2000 4000 6000 8000
, Time (s)
Figure A-89. Sequence 5-13 reactor vessel downcomer fluid pressure. 559 - i i i - 546.5
- 546.0 m ^
M l'- V U 558.5 - - e -
- 545.5 e L L 3 3
+-
U U g -
- 545.0 g o- o.
E 558 - - E e e
- 544.5
- 544.0 557.5 O 2000 4000 6000 8000 Time (s)
Figure A-90. Sequence 5-13 reactor vessel downcomer fluid temperature. A-48 r f
18000 , i i
-2500 16000 - - . m Q E %
6 14000 _ --2000 O e e L L 3 3
$ 12000 - -
E e e L L Q- Q_
-1500 10000 - -
8000 O 2000 4000 6000 8000 Time (s) Figure A-91. Sequence 5-14 reactor vessel downcomer fluid pressure.
.' 600 i i i
-600 m 550 -
^
v M h
-500 v e e L L 3 3
]L 500 - -
3L
-400 $
E E e e H 450 - -
-300 400 O 2000 4000 6000 8000 Time (s)
Figure A-92. Sequence 5-14 reactor vessel downcomer fluid temperature. A-49
~ 18000 i i
-2500 16000 - -
~
.m Q f 7 6 14000 _ -_2000 O . e e L L 3 3 E 12000 - - E e e L- 6 Q. G.
-1500 '
10000 - - 8000 O 2000 4000 6000 8000 Time (s) Figure A-93. Sequence 5-15 reactor vessel downcom' fluid pressure. 600 i i i
-600 550
~
ps v -500 e e L ' 3 3
}u S00 p
'{, . -400 $
E E
, e H ~
450 - _ -300 . 400 O 2000 4000 6000 8000 Time (s) Figure A-94. Sequence 5-15 reactor vessel downcomer fluid temperature. A-50
b.. - -' t 18000 , , ,
~
-2500 16000 -
m n 0 0 o., i.
+
_ d 14000 _-2000 av
.e
- L. e I- 3 L m 3 4 m .12000 - - m e m i L e
- n. '
O.
-15 %
10000 - - ! 8000 O 2000 4000 6000 .8000 Time (S) I~ Figure A-95. Sequence 5-16 reactor vessel downcomer fluid pressure. J 600 , , ,
-600 2
m 550 - m M
-500 v G e
' L 3 3
]'500 - -
g L t e ,
- c. -
-400 c.
E E W- e H H 450 - - y_ .
-300 400 O 2000 4000 6000 8000 Time (s)
Figure A-96. Sequence 5-16 reactor vessel downcomer fluid temperature. 1 l 4 h' A-51
3 16000 , , ,
-2200 14000 _ --2000 m U O
cL % 3 a.
-1800 8 12000 - - e it s
-1600
- e. h I i 10000 - -
-1400 8000 O 2000 4000 6000 8000
. Tim.e (s)
Figure A-97. Sequence 5-17 reactor vessel downcomer fluid pressure. 600 , i i
-600 m 550 -
m M
" l'-
-500 v
.e e L L 3 3
}6 500 - -
g 6
-400 'E E E G e H
- 450 - -
-300 ,
400 O 2000 4000 6000 8000 Time (s) - Figure A-98. Sequence 5-17 reactor vessel downcomer fluid temperature. A-52
7 18000 i i i
-2500 16000 -
m f S
'a 6 14000_ --2000 O L e 3 L 3
E 12000 - E L e Q. L Q.
-1500 10000 -
8000 ' ' ' O 2000 4000 6000 8000 Time (s) Figure A-99. Sequence 5-18 reactor vessel downcomer fluid pressure.
. 600 i i i
-600 550 -
v _
-500 v e
- L '
3 3
]L 500 - -
3' [ -400 $ E E o
- H 450 - -
-300 400 O 2000 4000 6000 8000 Time (s)
Figure A-100. Sequence 5-18 reactor vessel downcomer fluid temperature. A-53
r 18000 , , ,
-2500 16000 - - , m y 14000_7 --2000 O .
8 12000 - - 0 3 ' E m 8 10000 '- - a- a. 8000 - -
-1000 6000 O 2000 4000 6000 8000 Time (s)
Figure A-101. Sequence 5-19 reactor vessel downcomer fluid pressure. 600 i , i
-600 m 550 -
n M 12-v -
-500 v 0 @
L L 3 3
]6 500 - -
g 6
-400 $
E E o e H H 450 - -
-300 -
400 0 2000 4000 6000 8000 Time (s) - j Figure A 102. Sequence 5-19 reactor vessel downcomer fluid temperature. A-54 4
f 16000 i i i ll 14000_ -
--2000 '
- . m Q e 2 6 12000 -,
3
-1500 m .10000 - -
m
# e L L Q- O.
8000 - -
-1000 6000 O 2000 4000 6000 8000 Time (s)
Figure A-103. Sequence 5-20 reactor vessel downcomer fluid pressure.
. 600 i i i
-600
.550 -
n -
-500 ^
M v b v 8 500 - u 3 3
-400 -
0 U L L
- 450 - -
- g a E E o -
-300
- W W 400 -
- 2! )
350 0- 2000 4000 6000 8000
.- Time (s)
Figure A 104. Sequence 5-20 reactor vessel downcomer fluid temperature. l-l- l A-55 - l
- v. . . . . . . , . . . - . . . ~. . . _
~
=15840 .. , ,
A 15820'- _-2295 m n 0 [ 15800 - - m a
-2290 O ,
u 15780 -
- e o u j M 3 m m ;
.e m 8
4-L e' 15760 - -2285 at 15740 - i i 15720 2280-O. 2000 4000 6000 8000 Time (s)
) Figure A-105. Sequence 5-21 reactor vessel downcomer fluid pressure.
J 559. 5 -i i - 546.5 *
- 546.0 m- m M k'~
v 558.5 -
- v e -
- 545.5 e
' L
-0
?g
.e-
- 545.0 g o- c.
. E 558 - - E 1 e .
- 544.5 *
- 544.0 557.5 O 2000 4000 6000 8000 Time (s) -
Figure A.106. Sequence 5-21 reactor vessel downecmer fluid temperature. A-56
, . _ . , _ , . . _ . . _ _ - . - - - . . . _ . - - _ - - , . . . . .. .. , - _ . - - _ . - _ . _ _ _ _ , . _ _ _ - , . - .~. .-- ~ - - - .
15840 i i i
~
15820 '- ~
^
.n
[ -15800 -
.x - a v -
-2290 v
[ 15780 - - e
=
R. *
' 15760 -
u 1- -
-2285 n.
15740 - t 15720 2280 0 2000 4000 6000 8000 Time (s) Figure A-107. Sequence 5-22 reactor vessel downcomer fluid pressure. e
- 559_ , i i - 546.5
- 546.0 m ^
M b v 558.5 - e -
- 545.5 e L L 3 3
+-
O O g -
- 545.0 g 1- c.
E 558 - - E e e
- 544.5 *
- 544.0 557.5 O 2000 4000 6000 8000 Time (s)
Figure A-108. Sequence 5-22 reactor vessel downcomer fluid temperature. A-57
15840 , , , 15820 '- _-2295 m m y 15800 - . ,0 4 -
-2290 5 *
[ 15780 - _ e 3 ' M ~3 e E 15760 e l o' _
-2285 n'.
15740 _ 15720 ' ' ' 2280 0 2000 '4000 6000 8000 Time (s) Figure A-109. Sequence 5-23 reactor vessel downcomer fluid pressure. 559. _, , , _ 546.5 546.0 ^ m M ld-558.5 - _ v e -
- 545.5 e 2 5 0 0 e
- 545.0 '
O'
- a E 558 -
E
- 544.5
- 544.0 557.5 ' ' '
O 2000 4000 6000 8000 Time (s) l l Figure A.llo. Sequence 5 23 reactor vessel downcomer fluid temperature. A-58
15840 , i i
~
15820 '- -
- n m
[ 15800 3 a. V
-2290 v 8 15780 - - 8 a a
= Q
' 15760 -
u Q- -
-2285 c.
15740 - 15720 2280 0 2000 4000 6000 8000 Time (s) Figure A-Ill. Sequence 5-24 reactor vessel downcomer fluid pressure. 559- , i i - 546.5
- 546.0 m ^
M l'- U U 558.5 - -
- 545.5 e L L 3 3
+- -e-0 0 g -
- 545.0 g O. a.
E 558 - -
. .E
- 544.5 *
- $44.0 557.5 O 2000 4000 6000 8000 Time (s)
Figure A.ll2. Sequence 5-24 reactor vessel downcomer fluid temperature. A-59
18000 , , ,
-2500-16000 - _
m- m O .O Q. 6 14000 _- .2000 "a v , e e
< L i L 3
m - 3 e 12000 - _ m , e m L e. L Q. Q
-1500 10000 _
8000 O 2000 4000 6000 8000 Time (s) Figure A-113. Sequence 6-1 reactor vessel downcomer fluid pressure. 600 , , ,
-600 m 550 - -
m 5 l'~
~
-500 v
- e-i 6 '
- 3
-.3-5g 500 - -
0 e ! Q. -400 1 E' E 450 - e
-300 .
400 O 2000 4000 6000 8000 Time (s) i Figure A Il4. Sequence 61 reactor vessel downcomer fluid temperature. 1 i A-60
I -- 18000 , , ,
-2500 16000 - .. m ^
O O g .- 6 14000 _-
--2000 L
3 e M 3 g 12000 y u e 1
-1500 10000 -
8000 0 2000 4000 6000 8000 Time (s) Figure A-II5. Sequence 6-2 reactor vessel downcomer fluid pressure.
. 600 i , ,
-600 m S50 -
n M l'- v . 500 v e e
' L 3 3
]' 500 -
g L
-400 $
E E e e H
- 450 -
-300 400 O 2000 4000 6000 8000 Time (s)
Figure A Il6. Sequence 6-2 reactor vessel downcomer fluid temperature. A-61
18000 i i i
-2500 16000 - -
n y Q - 6 14000 _- --2000 0 -
@ G)
L L 3 3
$ 12000 -
C e
' L Q- c_
-1500 10000 -
8000 O 2000 4000 6000 8000 Time (s) Figure A-117. Sequence 6-3 reactor vessel downcomer fluid pressure. 600 i i i
-600 m 550 - -
^
M v b { -500 v e \ @ L L 3 3 3L
] SCO L
[ -400 $ E E o @ H 450 - -
-300 ,
400 O 2000 4000 6000 8000 Time (s) Figure A-Il8. Sequence 6-3 reactor vessel downcomer fluid temperature. A-62
18000 , , ,
-2500
'16000 -
- n n 0 g
- o. -
M 6 14000 _- .-2000 gv e e 6 3 L M 3
# 12000 -
m e a e G. L o.
-1500 10000 8000 ' ' '
O 2000 4000 6000 8000 Time (s) Figure A-Il9. Sequence 6-4 reactor vessel downcomer fluid pressure. l '. 600 , , ,
-600 m 550 - .
m
-500 e e u u y 3
]u500 - .
o e Q- -
-400 $
E E
# 450 - - H i
i
-300 I
400 ! 0 2000 4000 6000 8000
.. Time (s)
Figure A-120. Sequence 6-4 reactor vessel downcomer fluid temperature, l i A-63 1
16000 i i 1
-2200 14000 _- --2000 m *
^ O .O CL M J O. 1300 8 12000 - S 3 3 M w
-1600 g
' u Q- CL 10000 -
-1400 8000 O 2000 4000 6000 8000 Time (s)
Figure A 121. Sequence 6-5 reactor vessel downcomer fluid pressure. 600 i i i
-600 550 -
^ -
-500 ^
M v l'- v S 500 - - S 3 3
-400 -
0 0 L L
$450 - -
{ E E j -
-300 g 400 - -
-200 350 O 2000 4000 6000 8000 '
Time (s) Figure A 122. Sequence 6-5 reactor vessel downcomer fluid temperature. A-64
18000 , , ,
-2500 16000 - -
. n ^
' O O
- n. 7 6 14000 _- -_2000 O e ,
' L U 3 in 12000 - -
- e
' L A Q.
-1500 10000 -
8000 O 2000 4000 6000 8000 Time (s) Figure A 123. Sequence 6-6 reactor vessel downcomer fluid pressure. . 600 , . .
-600 550 -
^ -
-500 ^
v M l'- v 8 500 - - ' u 3 3
-400 4-0 0 L L
[450 - - E E E
-300 e 400 - -
-200 350 O 2000 4000 6000 8000 Time (s)
Figure A 124. Sequence 6-6 reactor vessel downcomer fluid temperature. A-65
16000 , , ,
-2200 14000 _- --2000 n n
0 0 1 3a. 6 * *
-1800
- 8 12000 -
3 ' (n 3 e -
-1600 6
1 I 10000 -
-1400 8000 O 2000 4000 6000 8000 Time (s)
Figure A-125. Sequence 6-7 reactor vessel downcomer fluid pressure. 600 i i i
-600 550 - -
^ -
-500 ^
M 12-v v 8 500 - - 8 3 3
-400 +
0 0 6 L [450 - - 1 E E 8 -
-300 8 g
400 - -
-200 350 O 2000 4000 6000 3000 '
Time (s) Figure A-126. Sequence 6-7 reactor vessel downcomer fluid temperature. A-66
18000 , , ,
-2500 16000 - -
0 0
~
n
, 5 14000 _- --2000 e #
L g L. m 3 m 12000 - - M e m 6 #
' ~
10000 - 8000 O 2000 4000 6000 8000 Time (s) r Figure A 127. Sequence 6-8 reactor vessel downcomer fluid pressure. 600 , , ,
-600 m S50 - -
m M l'-
-500 v e e L L 3 3
]'500 - -
g L
- e
- c. -400 a.,
E E
= e H H 450 - -
-300 400 O 2000 4000 6000 8000 Time (s)
Figure A 128. Sequence 6-8 reactor vessel downcomer fluid temperature. i A-67 _ . ~ . . - - . _ , . _ . - . _ . - - - , , - - _ _ . _ - _ _ _ . . . . . . _ _ _ - - - .
16000 i i ,
-2200 14000 - ~
~
-2000 m .
7a.
- o
.x $
-1800 -
I 12000 - - 3 3 E '
-1600 >
8 u u- Q-10000 ,
-1400 ,
~
-1200 8000 0 2000- 4000 6000 8000
/
IIme (S) Figure A-129. Sequence 6-9 reactor vessel downcomer fluid pressure. 600 i i
-r i
-600 550 -
- u. .
v l'- v 8 500 - _ , u 3
.e- .
-400 .3-0 U u
I
'E450 - ,
a E E e .
-300 e
- H >--
400 - , i
-200 350 0 2000 4000 6000 8000 Time (s) i Figure A.130. Sequence 6-9 reactor vessel downcomer fluid temperature.
i I ' A-68 i t.
. . . - - - - - . - - --,-..,,-._,..n., - - , . , ~ .
t
~
[ 18000 . . i
-2500 16000 - -
n 0 q A M 5 14000 -
--2000'O e -e L u 3' 3
$ 12000 - -
E
. e L L L 1
-1500 10000 - -
8000 O 2000 4000 6000 8000 Time (s) Figure A 131. Sequence 7-1 reactor vessel downcomer fluid pressure. l 1 L 580 , i i 564 - - --550 m ^ M v l'- v e w 540
- - e u
3 . 500 3 j -
+
! O O I L L , k520 - R E E e e
-450 W 500 - -
l. l 1. i F 480 O 2000 4000 6000 8000 Time (s) . Figure A 132. Sequence 7-1 reactor vessel downcomer fluid temperature. A-69
18000 , , ,
-2500 16000 - -
n m O U
- n. -
6 14000,-
--2000 O .
- e 3
n 3 m 12000 - - M
- l8 L
O- Q.
-1500 10000 - -
8000 O 2000 4000 6000 8000 Time (r) Figure A-133. Sequence 7-2 reactor vessel downcomer fluid pressure. 580 i i i 560~- --550 m ^ M v l'- v 8 540 - -
- 3 -
-500 3 0
L
/ 0 L
E520 [ E - E v e H . 450 H 500 - - 480 O 2000 4000 6000 8000 Time (s) Figure A 134. Sequence 7-2 reactor vessel downcomer fluid temperature. ' A-70
18000 , , .
-2500 16000 - -
n n O O E- Y 6 14000 .- --2000 O
- e L
S - 3 g 12000 - 6 *
' I
-1500 10000 - -
8000 O 2000 4000 6000 80C0 Time (s) Figure A-135. Sequence 7 3 reactor vessel downcomer fluid pressure. 580 , i i 560 -- - -550 m ^ M b v 540
- - V e -
-500 e w L 3 3
}6 520 - -
g L e o Q. -
-450 c.
E 500 - - E e e 480, - -
-400 460 O 2000 4000 6000 8000 Time (s)
Figure A.136. Sequence 7 3 reactor vessel downcomer fluid temperature. A 71 1 i i
16000 i , i
-2200 14000 ,- '
^ -2000 e ' O 2 ;;
-1800
[ 12000 -
- S 3 u m 3
" o
-1600 m 10000 - -
-1400 8000 O 2000 4000 6000 8000 Time (s)
Figure A-137. Sequence 7-4 reactor vessel downcomer fluid pressure. 600 i i i
-600 m 550 -
^
v M h
-500 v v
L e L 3 3
]L 500 - -
3L
$ -400 $
E E v e H 450 - -
-300 -
400 O 2000 4000 6000 8000 , Time (s) Figure A 138. Sequence 7-4 reactor vessel downcomer fluid temperature. A-72
18000 , , ,
-2500 16000 -
1 , .. , n 0 0 n - 6 14000 _- i . . -2000 E v l e 6 e 3 ' \ m 3 I e 12000 - m I e . m L e
- n. '
O.
~
10000 - -15 M [ -
. e 8000 ' ' '
0 2000 4000 6000 8000 Time (s) Figure A 139. Sequence 7-5 reactor vessel downcomer fluid pressure. l t 400 , , , 600 m 550 - n M l'- v - 500 v 8 e
' L 4 3 3
] SCO g
L 8 e
-400 O. -
a E E
= e H
- 450 - -
-300 400 ' ' '
O 2000 4000 6000 8000 Time (s) Figure A 140. Sequence 7-5 reactor vessel downcomer fluid temperature. A-73
I 16000 , , ,
-2200 14000 ,- ,
-s --2000 -s O .O.
- g. "
.x v
-1800 S .
[ 12000 - - # 3 3 M e -
-1600 E I
10000 - - [
-1400
~
8000 I O 2000 4000 6000 8000 l Time (s) Figure A 141. Sequence 7-6 reactor vessel downcomer fluid pressure. f 600 , , ,
-600 m 550 -
m M l'-
-500 v e e L L 3 3
]6 SM - -
g L
-400 $
E E e e
'H 450 - - *
-300 .
400 O 2000 4000 6000 8000 Time (s) Figure A 142. Sequence 7-6 reactor vessel downcomer fluid temperature. A-74
18000 i i i
-2500 16000 -
g 9 a. 6 14000,- .
--2000 0 3
E 12000 -
- E v .*
- n. l I /
/ [
10000 - -1500 8000 ' ' ' O 2000 4000 6000 8000 Time (s) Figure A-143. Sequence 7-7 reactor vessel downcomer fluid pressure. 580 i i i 560 -- --550 m ^ M E v 540 -
~
- e .
-500 e u 6 3 3
]L 520 -
5L e e
- c. .
-450 c.
E 500 - E e e H & 480 -
--400 460 ' ' '
O 2000 4000 6000 8000 Time (s) Figure A-144. Sequence 7-7 reactor vessel downcomer fluid temperature. A-75
16000 , , ,
-2200 t
.14000 ,-
n --2000 m O O ~ f E n y - v
-1800 3
[ 12000 - - e 3 '
.m 3 e -
-1600 E L e 1 I 10000 - -
-1400
~8000 .
O 2000 4000 6000 8000 j, Time (s) Figure A 145. Sequence 7-8 reactor vessel downcomer fluid pressure. d 600 , , ,
-600 m 550 -
m i-x
.v b
-500 v ,
e e L L 3 3
]L 500 - -
i
-}
L E -
-400 E.
E E e e H
& iso - -
-300
- 400 O 2000 4000 6300 8000
- Time (s) 4 Figure A 146. Sequence 7-8 reactor vessel downcomer fluid temperature.
A-76
a 16000 , , ,
-2200 14000 -
m
-2000 m O .O g
b v O.
.i,00
[ 12000 - - e 3 " in 3 e -
-1600 in L
L - p 10000 - -
-1400
~ '
8000 O 2000 4000 6000 8000 Time (s) Figure A-147. Sequence 7-9 reactor vessel'downcomer fluid pressure. 600 , , ,
-600 550 -
^ -
-500 ^
v M l'- v 8 500 - - e 3 3
-400 -
0 0 L
- u l[450 - -
{ E E g -
-300 e 400. - -
-200 350 O 2000 4000 6000 8000 Time (s)
Figure A-148. Sequence 7-9 reactor vessel downcomer fluid temperature. A-77
16000 i i 5
-2200 14000 ~
-2000 m g .o e
- b a
- -1800 L 12000 - ~
s ' n = e M e -1600 m 6
- A '
U-10000 - ~
-1400
' ' ' -1200 8000 O 2000 4000 6000 8000 Time (s)
Figure A-149. Sequence 7-10 reactor vessel downcomer fluid pressure. 600 i i '
-600 550 ~
n -
-500 ^
E V-0
" 500
- _ e 5-3 3
-400 g 0
L L
* - _ e Q-h- 450 ,
-300 E
8 400 - ~
-200 350 ' '
O 2000 4000 6000 8000 " Time (s) Figure A-150. Sequence 7-10 reactor sessel downcomer fluid temperature. A-78
16000 , , , 14000 -
--2000 n ^
0 0
- n. *-
, 6 12000 -
v c. L E -
-1500 3 g 10000 y
u e A n. .: 8000 - -
-1000 6000 O 2000 4000 G000 8000 Time (s)
Figure A-151. Sequence 7-11 reactor vessel downcomer fluid pressure. 600 , , ,
-600 550 -
^ -
-500 ^
M k'- v v 8 500 - - e 3
-400 -.3 0 o
' L
$450 - -
{ E E g -
-300 e 400 - -
~
-200 350
, 0 2000 4000 6000 8000 Time (s)
Figure A-152. Sequence 7-11 reactor vessel downcomer fluid temperature. A-79
16000 , , ,
-2200 s
N000 -
' n-
--2000 m -
O O
- a. M
.x Q.
v
-1800
[ 12000 - - e 3 3 m M o m e -
-1600 6
t Q- -
- I 10000 - -
-1400 8000 O 2000 4000 6000 8000 Time (s)
Figure A-153. Sequence 7-12 reactor vessel downcomer Guld pressure. 600 , , ,
-600 m 550 n
- M
~"
-500 v e
L 3- 3
]'500 - -
g L
- e G. -
-400 a E E
- e H
450 - -
-300 .
i
-400 O- 2000 4000 6000 8000 Time (s)
Figure A-154. Sequence 7-12 reactor vessel downcomer Huid temperature. I I' A-80 ? - r
18000 , , ,
-2500 16000 - -
- m n
y 14000_- - 2000 5 5 8 12000 - 3 m 3 m M b 10000 ' - 1 (1. 8000 -
-1000 6000 O 2000 4000 6000 8000 Time (s)
Figure A-155. Sequence 8-1 reactor vessel downcomer fluid pressure. 600 , , ,
-600 t
n 550 - - m M v l'-
-/ -500 v e e L L 3 3
]6 500 -
g L
-400 $
E E e e H H 450 - -
-300 400 0 2000 4000 6000 8000 Time (s)
Figure A-156. Sequence 8-1 reactor vessel downcomer fluid temperature. A-81
. . .. .. _ . , - . . . - . . . ~ .
. - . .~ - . - . _
18000 i i i
-2500 16000 .
d O g 14000_- - 2000 *g
.x - a v v .
- 8 L 12000 - -
3 3
# M
< M gfg e -
-1500 e L 10000 -
6
-1 a, 8000 -
-1000 6000 O 2000 4000 6000 8000 Time (s)
., Figure A-157. Sequence 8-2 reactor vessel downcomer fluid pressure. t -i 600 i i i
-600 m 550 - -
^
M v h
-500 v
.e e L L 3 3
*O 500 - -
U L L e e a -
-400 a.
E E e e M 4g _ - F
-300
- 400 O 2000 4000 6000- 8000 , l Time - (s)
Figure A-158. Sequence 8-2 reactor vessel downcomer fluid temperature. A-82
16000 , , , 14000 , - _-2000 e
^ n o -
o D. 6 12000 - _ { 6 ".L 3 '
-1500 g E 10000 _
. m.
6 L Q. Q. 8000 .
-1000 6000 O 2000 4000 6000 8000 Time (s)
Figure A-159. Sequence 8-3 reactor vessel downcomer fluid pressure. 600 , , ,
-600 m 550 - .
U b 500 v
= si
+
a 500 o 6 ' e Q- -
-400 1
! E 450 - .
-300 400 0 2000 4000 6000 8000 Time (s)
Figure A-160. Sequence 8-3 reactor vessel downcomer fluid temperature. A-83
<~
16000 , , , 14000 _- --2000 n ^ O C
- a. ;;
6 12000 - v
- c. -
L S -
-1500 3 en 10000 y
' e L
E- n. 8000 -
-1000
+ 6000 ' ' ' O 2000 4000 6000 8000 Time (s) Figure A-161. Sequence 8-4 reactor vessel downcomer fluid pressure. 'i . s
-600 , , ,
-600 550 r -
m 500 ^ M -f b v v 8 500 6 L i 3 3
-400 +
0 0
-6 y
{450 - - { E E
-300 e 400 - -
-200 350 ' ' '
O 2000 .4000 6000 8000 . Time (s) Figure A-162. Sequence 8-4 reactor vessel downcomer fluid temperature. A-84
18000 , , ,
-2500 16000 - - . , m
[ 14000_- - 2000 6 o [ 12000 - - e a 3 m e -
-1500 e
' 10000 -
u 1 CL 8000 -
-1000 6000 O 2000 4000 6000 8000 Time (s)
Figure A-163. Sequence 8-5 reactor vessel downcomer fluid pressure, e 600 i , ,
-600-n 550 - -
n v M l'-
-500 v e e
' L 3 3 5 500 6
g L
. -400 $
E E e e H F 450 - -
-300 400 O 2000 4000 6000 8000 ,- Time (s)
Figure A-164. Sequence 8-5 reactor vessel downcomer fluid temperature. A-85
16000 , , , 14000 _-
--2000 ,
6 12000 - v E , L 8 3 " to -
-1500 3 g 10000 -
F, L 8 8000 -
-1000 6000 O 2000 4000 6000 8000 Time (s)
Figure A-165. Sequence 8-6 reactor vessel downcomer fluid pressure. 600 i , ,
-600 550 - -
^ -
-500 ^
M ld-v v 8 500 - - e 3 3
+ .
-400 +
0 0
't 6 1450 - -
[ E E g -
-300 g 400 - -
-200 350 O 2000 4000 6000 8000 .
Time (s) Figure A.166. Sequence 8-6 reactor vessel downcomer fluid temperature. A-86
16000 __ -2300 15500 -- --2250
, m O O E-E d -
-2200 3 i
8-15000 -
- 8
= 3 E -
-2150 g e e Q- L n.
14500_ --2100
-2050 14000 O 2000 4000 6000 8000 Time (s)
- Figure A-167. Sequence 9-1 reactor vessel downcomer fluid pressure.
e. 564 i i i
-554 m ^
M
--552 Y
v v 562 -- e e L L 3 3 3L
-550 o L
e e Q. Q. E 560 - E e -
-548 e H ! l--
-546 558 O 2000 4000 6000 8000 Time (s)
I Figure A-168. Sequence 9-1 reactor vessel downcomer Guid temperature. A-87 m, - -- .
. -- . . ~ . - -_ _ .- _ . -
16000 , , , o .
-2200 i 14000 ,
^ _-2000 m .
O - o Q-E e
-1800 .
u 12000 _ 3 u 3 y
-1600 10000 -
X _ e
-1400 4
' ' i -1200 8000 0- 2000 4000 6000 8000 Time (s) 4 Figure A-169. Sequence 9-2 reactor vessel downcomer fluid pressure.
f ! 600 , , , i
-600 m 550 - .
b b
-500 v e e u '
3
] SCO
- _ 2 U
L u 4 e < Q. - 400 { E h 450 - _ i
-300 ,
l 400 O 2000 4000 6000 8000 i-t Time (s) Figure A-170. Sequence 9-2 reactor vessel downcomer fluid temperature. 1-A-88
_=_ . _ .. _ _. _ .
~
l 16000 , , ,
-2200 14000 -
n
-2000 m O O g
.x ".
Q v
- s -1800 u 12000 - -
3 3 M M M e -
-1600 m 10000
-1400 i ' ' '
8000 O. 2000 4000 6000 8000 Time (s) Figure A-171. Sequence 9-3 reactor vessel downcomer fluid pressure.
.a
'l 600 , , , j
-600 t
n 550 - - m M v l'- 4 - 500 v 8 e b
]L S00 - -
g
, g
# e Q. -
-400 a.
E E H 8 e 450 - - H
-300 400 O 2000 4000 6000 8000 Time (s) , Figure A-172. Sequence 9-3 reactor vessel downcomer fluid temperature.
A-89
18000 , , ,
-2500
'6000 m m
.O O
- a. -
6 14000 , ". a 7 --2000 v .
.e e u '
3 M 3 g 12000 y u e I
-1500 10000 - -
8000 O 2000 4000 6000 8000 Time (s) Figure A-173. Sequence 9-4 reactor vessel downcomer fluid pressure. 600 , , ,
-600 m 550 - -
m M v- l'-
-500 v e e L u 3 3 ]6-500 - -
g u e e
- c. -
-400 a.
E E-e e H
- 450 - -
-300 .
400 O 2000 4000 6000 8000 Time (s)
- Figure A-174. Sequence 9-4 reactor vessel downcomer fluid temperature.
i A-90
i 18000 , , , t
-2500 16000 -
- n. ^
0 C
- a. -
. 6 14000 _
--2000 O
' e
- L l 3 so 12000 - e- E e Q- L Q.
-1500 10000 -
8000 O 2000 4000 6000 8000 Time (s) Figure A-175. Sequence 9-5 reactor vessel downcomer fluid pressure. e 4 600 . i . ,
-600 J
m 550 - -
^
v M b
-500 v e e L L 3 3 l
] SCO L
3L
-400 $
E E e e 450 - -
. 1
-300 l 400
. 0 2000 4000 6000 8000
., Time (s) . Figure A-176. Sequence 9-5 reactor vessel downcomer fluid temperature. A-91
16000 , , ,
-2200 14000_ --2000 m O O
n.. %o-5 v .
-1800 8 12000 - - #
3 3
" m E -
-1600 g E / t 10000 -
l
-1400 t
4 8000 O 2000 4000 6000 8000 Time (s) Figure A-177. Sequence 9-6 reactor vessel downcomer fluid pressure. 600 , , i
-600 m 550 - -
m M v
}'-
-500 v e e L L 3 3
]6 500 - -
g 6
$ -400 a.
E E o . H 450 - - N
-300 -
400 O 2000 4000 6000 8000 , Time (s) Figure A-178. Sequence 9-6 reactor vessel downcomer fluid temperature. A-92
. ~ . . _ _ -. . . _ - - ..
'16000
-2200
- 14000 , --2000 o
g G-a 5a V 1300 [ 12000 - - 3 s
" n
-1600 g
' L 1
- 0. a 10000 - -
-1400 8000 O 2000 4000 6000 8000 Time (s)
Figure A-179. Sequence 9-7 reactor vessel downcomer fluid pressure. i s. 4 i. 600 , , ,
-600 i
n 550 - - n M l'-
-500 v 8 e
' L 3 3
]'500 - -
g L e e C. -
-400 a E E
~
j' 450 - -
-300 400 ' ' '
0 2000 4000 6000 8000 Time (s) Figure A-180. Sequence 9-7 reactor vessel downcomer fluid temperature. i' i 4 A-93 1
, . . . ~ -. - _ _ . . - . - .. . . - - . . .-
a 18000 , , , ,
-2500 _
16000 m ^ 0 0
. n. -
6 14000 _ --2000 v E , e L 8 3 "
- . n 3
- m 12000 -
- M e *
- u. e Q. '
O.
-1500 10000 -
8000 O 2000 4000 6000 8000 Time (s) Figure A-181. Sequence 9-8 reactor vessel downcomer fluid pressure. t ! 600 , , ,
-600 m 550 - -
a M v b 500 v e e L L 3 3
]L S00 - -
g L
-400 1 -
E E e e
~450 - -
-300 *
, 400 O 2000 4000 6000 B000 ,
. Time (s)
? Figure A.182. Sequence 9-8 reactor vessel downcomer fluid temperature. A-94
. - -- . . - . . -- - . - . - .- . . . . -- _. - -- - - . - - - - - - - . ---- - . - . _ _ .~. . - . - _ - - -
16000 , , ,
-2200 14000 -
n
-2000 m O .O
~
Q. d V 1300 [ .12000 - - e 3 " to 3 e -
-1600 L
Q- I 10000 - -
-1400 8000 O 2000 4000 6000 8000 Time (s)
Figure A-183. Sequence 9-9 reactor vessel downcomer fluid pressure. ~ 600 , , .
-600 m 550 -
a M v l'-
-500 v e2 e L L 3 3
] SCO L
5L
-400 $
E E o e H 450 -
-300 400 O 2000 4000 6000 8000 Time (s)
. Figure A-184. Sequence 9.9 reactor vessel downcomer fluid temperature.
A-95
16000 , , ,
-2200 14000 _
--2000 m -
O . U Q-3 6
- a-
-1800 '
12000 - - # a 3 e -
-1600
_ L 10000 - -
-1400 1
8000 O 2000 4000 6000 8000 , Time (s) Figure A-185. Sequence 9-10 reactor vessel downcomer fluid pressure. 600 , , . 550 -
^ -500 ^
M ld-v v 8 500 - -
- 3 3
-400 +
0 0 L L [450 - - 1 E E 8 -
-300 8 400 - -
-200 350 O 2000 4000 6000 8000 Time (s) -
i Figure A-186. Sequence 9-10 reactor vessel downcomer fluid temperature. A-%
18000 i i i
-2500' 16000 - -
~
=-
~g. y CL .
3 6 14000 -
--2000 O 3 L 3
$ 12000 - -
E e e
' L CL ct
-1500 10000 - -
8000 O 2000. 4000 6000 8000 Time (s) Figure A-187. Sequence 9-11 reactor vessel downcomer fluid pressure.
'- 600 , i i
-600 550 -
^ -
-500 ^
M v l'- v
- 500 - - '
u 3 3 0
-400 +
L 0 L [450 - - [ E E C -
-300 e 400 -
-200 350 O 2000 4000 6000 8000 Time (s)
Figure A-188. Sequence 9-11 reactor vessel downcomer fluid temperature. 4 A-97 =
18000 , , ,
-2500 16000 - -
m n D 0
- n. ~
6 14000 _ r --2000 , 3 " M 3 g 12000 y u o
"" I
-1500 10000 - -
8000 O 2000 4000 6000 8000 Time (s) Figure A-189. Sequence 9-12 reactor vessel downcomer fluid pressure. 600 i i i 550 -
^ -500 ^
M v l'- v 8 500 - -
- 3 3
-400 +
0 0 L L v (450 - - 1 E E W -
-300 #
400 - - ^
-200 350 O 2000 4000 6000 8000 Time (s) -
Figure A 190. Sequence 9-12 reactor vessel downcomer fluid temperature. A-98
16000 . i i i i i i
-2200
- 14000 -
m --2000 n o t U
- o. ;
6 -
- a
-1800 8 12000 -
- 0 3
en 3 en
- e -
-1600 y 10000 -
5
+ -
-1400 8000
'O 1000 2000 3000 4000 5000 6000 7000 8000 Time (s)
Figure A-191. Sequence 9-13 reactor vessel downcomer fluid pressure. 600 i i i i i i
-600 550 -
^
2 v .
-500 v A^-
D o L L 3 3
$ 500 L
$L
$ -400 $
E E o o 450 - -
-300 400 -- '
O 1000 2000 3000 4000 5000 6000 7000 8000 Time (s) Figure A-192. Sequence 9-13 reactor vessel downcorner fluid temperature. A-99
16000 , , . 14000 _
--2000 m
Q E % 6 12000 - O. .
-1500 m 10000 - -
m O g b 6 Q-c 8000 - -
-1000 6000 O 2000 4000 6000 8000 Time (s) ,
Figure A-193. Sequence 9-14 reactor vessel downcomer fluid pressure.
- i. .
600 i i i
-600 s
550 -
^ -
-500 ^
v M b v
- 500 - -
- 3 \ 3
~
. -400 -
0 0 L L [450 - - 1 E E o .
-300
- W W 400 - -
-200 350 O 2000 4000 6000 8000 ,
Time (s) Figure A 194. Sequence 9-14 reactor vessel downcomer fluid temperature. A-100
(. 16000 , , , f. 14000 -
--2000
, l m O O Q- 3 6 12000 -
v c. c
-1500 g 10000 y
u e 1 8000 - -
-1000 6000 O 2000 4000 6000 8000 Time (s)
Figure A-195. Sequence 9-15 reactor vessel downcomer fluid pressure, l i 600 i , ,
-600 l' 550 -
m .
-500 ^
v M l'- v 8 500 - - S 3 3 0
-400 +
' 0 L
E450 l { E E S [ -
-300 400 - -
-200 350 ' ' '
O 2000 4000 6000 8000 Time (s) Figure A-l%. Sequence 9-15 reactor vessel downcomer fluid temperature. A-101
18000 i i i
-2500 16000 - -
m Q g 14000_ - - 2000 g 6 k O . [ 12000 - [ 3 3 m m E -
-1500 E L 10000 - -
6 a- c. 8000 - -
-1000 6000 O 2000 4000 6000 8000 Time (s)
Figure A-197. Sec.uence 9-16 reactor vessel downcomer fluid pressure. 600 , i i
-600 550 -
n -
-500 ^
M v l'- v [ 500 - - [ 3 3
-400 +
0 0 L L {450 [ E E y -
-300 g 400 - -
-200 350 O 2000 4000 6000 8000 '
Time (s) l Figure A-198. Sequence 9-16 reactor vessel downcomer fluid temperature.
I 18000 i i i
-2500 16000 - -
^
g g 14000 - - 2000 *g
. 6 8 8 12000 - '
l 3 u M 3 y E -
' 10000 - --1500 L Q.
o. l l 8000 - -
-1000 6000 O 2000 4000 6000 8000 l' Time (s)
Figure A-199. Sequence 9-17 reactor vessel downcomer fluid pressure. 600 i i i
-600 f
550 - n y
-500 ^
v k'- v j 8 500 - - [ 2 .
-400 2 c U l L. S- l g 450 E E E e e l p-
-300 >-
400 - 1
-200 350 O 2000 4000 6000 8000 Time (s)
Figure A-200. Seqt ence 9-17 reactor vessel downcomer fluid temperature. A-103 l l
l f 3000 , , , i~ 14000 -
--2000 m n O O Q-3c.
6 12000 - v . ! e e L I E -
-1500 3 g 10000 - -
y l u = Q-l t I I' 8000 - - l 1
-1000 6000 l 0 2000 4000 6000 8000 Time (s)
Figure A-201. Sequence 9-18 reac*ar vessel downcomer fluid pressure. 600 . . .
-600 550 -
^ -
-500 ^
M v l'- v e 500 - - C u u 3 3
-400 .-
0 0 L L [450 - - [ E E
-300 e 400 - -
-200 350 O 2000 4000 6000 8000
~
Time (s) ,. Figure A-202. Sequence 9-18 rea: tor vessel downcomer fluid temperature. A-104
n l 16000 , , , 14000 _- --2000 n m O O g -- 6 12000 - v E e
-1500 g 10000 y
u e
' I 8000 - -
-1000 6000 O 2000 4000 6000 8000 Time (s)
Figure A 203. Sequence 9-19 reactor vessel downcomer fluid pressure. 4 600 , , ,
-600 550 -
^ -500 ^
M v b v 8 500 - -
- 3 3
-400 +.
O O L L E450 1 E E 8 -
-300 8 g
400 - -
-200 350
, 0 2000 4000 6000 8000 Time (s)
Figure A-204. Sequence 9-19 reactor vessel downcomer fluid temperature. A-105
~
16000 i . . 14000 _- ~-2000 m a O O Q- E 6 12000 - v
- c. -
- e
-1500 en 10000 -
E 6 A a. 8000 - -
-1000 6000 O 2000 4000 6000 8000 Time (s)
Figure A-205. Sequence 9-20 reactor vessel downcomer fluid pressure. 600 , , i 550 - n -
-500 ^
M v l'- v 8 500 - ~ ' u 3.- -
-400 .'3-0 0 g 6-
- 450 - -
- g a.
E- E c .
-300 e
'400 - -
-200 350 O 2000 4000 6000 8000 Time (s) -
Figure A-206. Sequence 9-20 reactor vessel downcomer fluid temperature. A-106
m-18000 i i i
-2500 16000 - -
- n. q y .14000_- -
2000 *- 6 & 8 12000 - 8 3 3 M M e -
-1500 e
' 10000 -
L A Q. 8000 - -
-1000 6000 O 2000 4000 6000 8000 Time (s)
Figure A-207. Sequence 9-21 reactor vessel downcomer fluid pressure. e 600 i i i
-600 550 -
n -500 ^ M v b v 8 500 - f 3 3
-400 +
0 0 L L
- 450 a
1 E E y _
-300 e 400 -
-200 350 O 2000 4000 6000 8000 Time (s)
Figure A-208. Sequence 9-21 reactor vessel downcomer fluid temperature. A-107 l
18000 i i i
-2500 16000 - -
Q
~
n f
- [ 14000 - -
2000 g U O . [ 12000 - [ 3 3 E
-is00 E.
10000 - n'. L a. 8000 - -
-1000 6000 O 2000 4000 6000 8000 Time (s)
Figure A-209. Sequence 9-22 reactor vessel downcomer fluid pressure. 600 . i
-600 550 -
n -
-500 ^
v M l'- v 8 500 - -
- 3 3
-400 +-
0 0 L L [450 - - E E E 4 -
-300 8 400 - -
-200 350 O 2000 4000 6000 8000 ,
Time (s) Figure A-210. Sequence 9-22 reactor vessel downcomer fluid temperature. A--108
16000 i i , 14000_- --2000
-q o
. Q. E 6 12000 -
3
' L 3 -
-1500 3 e 10000 -
E
- e
' L Q- G.
8000 - -
-1000 6000 O 2000 4000 6000 8000~
Time (s) Figure A-211. Sequence 9-23 reactor vessel downcomer fluid pressure. e 600 , i i
-600 550 n -
-500 ^
M 1^- v v S ' u 500 u ~ 3 3
~ -
-400 +
0 0 L L {450 E E E y -
-300 8 400 -
-200 350 O 2000 4000 6000 8000 Time (s)
Figure A-212. Sequence 9-23 reactor vessel downcomer fluid temperature. A-109
l l 16000 , , , 14000_- --2000 n ^ 0 0
~
0- n 6 12000 - 3 , 3 m '
-1500 $
m 10000 - E
- e L
0-n. 8000 - -
-1000 6000 O 2000 4000 6000 8000 Time (s)
Figure A-213. Sequence 9-24 reactor vessel downcomer fluid pressure. 600 , , ,
-600 550 r -
^ <- -500 ^
M v b v 8 500 - - ' u 3 3- -
-400 +-
0 0 L L [450 - - E E E e .
-300 e 400 - -
-200 350 O 2000 4000 6000 8000 ,
Time (s) Figure A-214. Sequence 9-24 reactor vessel downcomer fluid temperature. A-Il0 y y--- r 4-- -
16000 i i i
-2200 14000 ~
-2000 m q 2 n-b E.
-1800 u 12000 - _ e 6
3 -- E l
-1600 "
L e CL n'- 10000 - ~ _1400
, , -1200 8000 0 2000 4000 6000 8000 Time (s)
Figure A-215. Sequence 9-25 reactor vessel downcomer fluid pressure.
. 600 i i '
-600 n 550 - -
m
$ }i.
-500 v o e G L 3 3
] 500 -
O e o- -400 E. E E o e H i-450 -
- -300
[ l 400 0 2000 4000 6000 8000 Time (s) Figure A-216. Sequence 9 25 reactor vessel downcomer fluid temperature. - 4 A-111
16000 i i i
-2200 14000, ~-2000 m -
O
- a. =
.x '
" Q-
-1800 I 12000 - ~
8 3 3 M M m m e -
-1600 L
e O. 0-10000 - ~
-1400
' ' ' -1200 8000 O 2000 4000 6000 8000 Time (s)
Figure A-217. Sequence 9-26 reactor vessel downcomer fluid pressure. 600 i i i
-600 m 550 - ~
^
M v b
-500 v e
- 6 '
3 3
]6 500 - ~
3' { _
-400 $.
E E o
- H 450 - ~
-300 .
400 O 2000 4000 6000 8000 Time (s) Figure A-218. Sequence 9-26 reactor vessel downcomer fluid temperature. A-Il2
18000 , , ,
-2500 16000 - -
m n 0- 0
~
Q. 6 14000 , --2000 3 3 to g 12000 y u e G. k
-1500 10000 - -
8000 O 2000 4000 6000 8000 Time (s) Figure A-219. Sequence 9-27 reactor vessel downcomer fluid pressure.
- 600 i i i
-600 m 550 - -
n M v l'-
-500 v e e L L 3 3
] SCO g
6 6
$ -400 $
E E e e H 450 - - l I
-300 400 O 2000 4000 6000 8000 Time (s)
Figure A-220. Sequence 9-27 reactor vessel downcomer fluid temperature. I i j A-ll3 l i
h 18000 i i i
-2500 3000 . -
~
. m q^
f 7 6 14000 --2000. O . e e L L 3 3
.E L
12000 .E L
- c. c.
-1500 10000 - -
8000 O 2000 4000 6000 8000 Time (s) Figure A-221. Sequence 9-28 reactor vessel downcomer fluid pressure. 600 i i
-600 m
6 550 - ' p 500 v e
- L '
3 3
]L 500 M'
{' _
-400 $
E E e
- n 450 - - H
-300 -
' t '
400 O 2000 4000 6000 8000 Time (s) Figure A-222. Sequence 9-28 reactor vessel downcomer fluid temperature. A Il4 S, %
16000 e i i
-2200
.14000 -
-2000 m q 2 a-6- E.
-1800
, e u 12000 - _
L 3 E ~ S
.e
-1600 88 L
e O. 0-10000 - '
-1400
, , , -1200 8000 0 2000 4000 6000 8060 Time (s)
Figure A-223. Sequence 9-29 reactor vessel downcomer fluid pressure. i i i
-600 m 550 -
~
-500 V
[ e e
' L 3 3 e-
*0 500 -
0
' L e -400 e
a. E- E e e F-F 450 -
~
- -300 g i I 00 0 2000 4000 6000 8000 Time (s)
Figure A-224. Sequence 9-29 reactor vessel downcomer fluid 'emperature. A-l l5
16000 , , ,
-2200 14000 ,
m --2000 n - 0 0-d * *
-1800 I 12000 - - e a 3 e -1600 L
E I 10000 - -
-1400
~
8000 O 2000 4000 6000 8000 Time (s) Figure A-225. Sequence 9-30 reactor vessel downcomer fluid pressure. 600 , , ,
-600 m 550 - -
n M 12-v -
-500 v e e L L 3 3
]6 500 - -
g L
-400 $_
E E e e H 450 - -
-300 .
400 O 2000 4000 6000 8000 Time (s) Figure A-226. Sequence 9 30 reactor vessel downcomer fluid temperature. A-l l6
18000 i i i
-2500 16000 - -
n Q d' g 5 14000 --2000 O L e L 3 3 E 12000 - - E
- e L L 0- 0
-1500 10000 - -
8000' O- 2000 4000- 6000 8000 Time (s) Figure A-227. Seqttence 9-31 reactor vessel downcomer fluid pressure. 600 , i i
- - -600 m 550 -
^
M h v - -500 v c e L L 3 3
]L 500 -
3L [ _ -400 $ E E e e H - 450 - , - -300 400 O 2000 4000 6000 8000 Time (s) Figure A-228. Sequence 9-31 reactor vessel downcomer fluid temperature. A-l l7
18000 , i
-2500 16000 -
m Q f % 6 14000 _ --2000 O .
'e 6 e g
3 3 E 12000 - - E
- e L
L Q-n.
. -1500 10000 -- -
8000 0 2000 4000 6000 8000 Time (s) Figure A-229. Sequence 9-32 reactor vessel downcomer fluid pressure. P 600 i i
-600 m 550 - -
^
M h v _
-500 v e e L L 3 3
}L 500 - -
3L [ _
-400 $
E E e e F 4g - - *
-300 -
400 O 2000 4000 6000 8000 , Time (s) Figure A-230. Sequence 9-32 reactor vessel downcomer fluid temperature. , 6 A-118
. . = . = . _ .. . . _ . - -- . ._
b f-16000 , , ,
-2200
, 14000 -
n-
-2000 n 0 0
, L m
.x l
. v
-1800 3
L 12000 - - e 3 ' w 3 m . in e -
-1600 m e
L a ' O. 10000 - - s
-1400 i
' ~
8000 0 2000 4000 6000 8000 Time (s) Figure A-231. Sequence 9-33 reactor vessel downcomer fluid pressure. ; i 600 , , ,
-600 m 550 -
n. M v h
-500 v e e L- L 3 3
~O 500 - -
O L L e e r a -400 a. E E e e H ' 450 - - i
-300 400 i 0 2000 4000 6000 8000 Time (s)
Figure A-232. Sequence 9-33 reactor vessel downcomer fluid temperature. 4 A 119 _ . - - , - . . - . - _ . _ _ - _ ~ . . . .. - _ . . _ _ . . _ . . . - . , , . , . . . - , , . . , . - -- .~ .,,. _ _._
16000 , ,
-2200 14000 n
-2000 m
- O U a .
d
- O.
-1800 L 12000 -
- o 3 "
M 3 M M e -
-1600 M L W Q. '
10000 -
-1400 8000 O 2000 4000 6000 8000 Time (s) ure A-233. Sequence 9-34 reactor vessel downcomer fluid pressure.
600 , , ,
-600 m 550 -
n M l'-
-500 v 0
L e L 3 3
]L 500 - -
g L 0 W
- o. -
-400 Q.
E E O @ H
- 450 - -
-300 .
400 O 2000 4000 6000 8000 Time (s) , Figure A-234. Sequence 9-34 reactor vessel downcomer fluid temperature. A-120
18000 , , ,
-2500 16000 - -
n y cf - % 6 14000 _ --2000 O e e
' L 3 3 E 12000 - -
E e
- L L Q- cL
-1500 10000 - -
8000 O 2000 4000 6000 8000 Time (s) Figure A 235. Sequence 9 35 reactor vessel downcomer fluid pressure. 600 i i i
-600 m 550 -
^
M v b
-500 v e
- 6 6 3 3
} SCO L L
[ -
-400 1 E E ,
e e H
- 450 _ -
-300 400 O 2000 4000 6000 8000 Time (s) -
Figure A-236. Sequence 9-35 reactor vessel downco.ner fluid temperature. A 121
. _ . . _ _ _ _ . - . _ _ , _ ,_ _ , . _ - . ~ _ _ _ _ . _ . _ . - _, _.
18000 i i i
-2500 16000 -
m O G_ Q._ vs 6 14000 --2000 O - O L e 3 L 3
$ 12000 -
E e e L Q- L Q_
-1500 10000 -
8000 O 2000 4000 6000 8000 Time (s) Figure A-237. Sequence 9 36 reactor vessel downcomer fluid pressure. 600 i i i
-600 m 550 -
^
M v b
-500 v o
- L '
3 3
]6 500 - -
o { -400 $ E E e
- H 450 - -
-300 .
400 O 2000 4000 6000 8000 Time (s) ' Figure A 238. Sequence 9-36 reactor vessel downcomer fluid temperature. A-122
l I-16000 , , ,
-2200 14000 -
m
-2000 m O
0-
?en
# Q.
-1800
- 6. 12000 - - 8 3 3
" m
-1600 g S- t G- CL.
10000 - -
-14C0 L
' ' ' ~'
8000 O 2000 4000 6000 8000 Time (s) Figure A-239. Sequence 9-37 reactor vessel downcomer fluid pressure. ! e ( l 1 l 600 i i i ! - -600 m 550 m
'M P-v - 300 v e e
' L 3 3
]'S00 -
g 6. e -400 e
- c. Q.
E E o e H -
- 450 -
t 9 300 4 400 O 2000 4000 6000 8000 Time (s) t Figure A-240. Sequence 9-37 reactor vessel downcomer fluid temperature. A-123 9
- , - , , . , - - , , , - , - - . - , - - - , - - - -.---c. %- --._ =,w, m.-.-- er - _ , , , - . , . - - - - ---m.,,.-r-. ,,--e-,,.,
16000 -
-2200 m
14000 _ --2000 m O O
- a. -
6 *
- c. *
-1800 I '12000 - - #
a 3 e -
-1600
' L A Q. 10000 - -
-1400 8000 O 2000 4000 6000 8000 Time (s)
Figure A-241. Sequence 9-38 reactor vessel downcomer fluid pressure.
.i SCO , _, ,
600 m 550 - m v M b 500 v e . 6 L 3 3
}6 500 - -
g 6
-400 1 E E = e-H
- 450 - -
300 400 O 2000 4000 6000 8000 , Time (s) Figure A-242. Sequence 9-38 retctor vessel downcomer fluid temperature. A 124
18000 . , .
-2500 16000 - - * 'n o Q.-
A M 5 14000 --2000 O L . L 3
$ 12000 - -
E
' e L
O- n.
-1500 10000 - -
8000 O 2000 4000 6000 8000 Time (s) Figure A-243. Sequence 9-39 reactor vessel downcomer fluid pressure. 600 i , ,
-600 m 550 -
m M }^- v -500 v e e L L 3 3
*O 500 - -
O 6 L e e
- a. -
-400 a.
E - E
= e H
450 - -
- i 300 400 O 2000 4000 6000 8000 Time (s) l Figure A-244. Sequence 9-39 reactor vessel downcomer fluid temperature.
I l
~
l ! A-125
F-18000 , i .
-2500 16000 - -
n O g
- a. ;;;
6 14000 --2000 O ,
- e
' L S 3 m 12000 - -
- e
' L 4
A A
-1500 10000 - - -
4
~' ' '
8000 r 0 2000 4000 6000 8000 ., Time (s) i Figure A-245. Sequence 9-40 reactor vessel downcomer fluid pressure. I J i 400 , , ,
-600 i
i m 550 - m M A'- v -
-500 v e e L L 3 s O 500 6
o 6 E -
-400 E.
E E
= e 450 -
i - 300 . 400 O 2000 4000 6000 8000 Time (s) I Figure A-246. Sequence 9-40 reactor vessel downcomer fluid temperature. 1 Y , A 126
16000 , , , C
-2200-14000 , -
m --2000 D^ 0 f
- a. ;
5
- o.
-1800
- e L 12000 -
3 ' M 3 e -
-1600 -
6 A I 10000 - -
-1400 8000 O 2000 4000 6000 8000 Time (s)
- Figure A-247. Sequence 9-41 reactor vessel downcomer fluid pressure, e
600 i , i
-600 m 550 -
n M v l'-
-500 v e e L . L 3 3
*OS00 *
, O L L
-400 $
, E E e e H
450 - - *
-300 N
400 O 2000 4000 6000 8000 Time (s) l i Figure A 248. Sequence 9-41 reactor vessel downcomer fluid temperature. l
)
A-127
i 16000 , , ,
-2200 i
14000 - n
-{
--2000 m
- O a .O b "
Q.
-1800
, --12000 u -
- e i
.s L M 3 M M e -
-1600 e
m i u A g 10000 -
-1400
, , , -1 00 8000 0 2000 4000 6000 8000 Time (s)
Figure A 249. Sequence 9-42 reactor vessel downcomer fluid pressure. 400 , , ,
-600 550 -
^ -
-500 ^
M v l'- v - 8 500 - - e 1 3 400 3. i O
' n L
[450 - - g E E
-300 g o
400 - - 1-l
-200 350 O 2000 4000 6000 8000 ~
Time (s) Figure A 250. Sequence 9-42 reactor vessel downcomer fluid temperature. A 128
18000 , , ,
-2500 16000 -
m n o .2 a-m 6 14000 -2000 O
' e L
S U m 12000 - E
' e 1 ' L a
1500 10000 - j l l BOCO ' ' ' O 2000 4000 6000 8000 l Time (s) Figure A-251. Sequence 9-43 reactor vessel downcomer fluid pressure. 600 , , ,
-600 m 550 -
m M l'-
-500 v 0
' e 3 L 3
'g 500 -
'o u c -
-400 e a
E C E H e 450 - l
-300 ;
1 400 ' ' O 2000 4000 6000 8000 ' Time (s) Figure A-252. Sequence 9-43 reactor sessel downcomer fluid temperature. A-129
.18000 , , ,
2500 3000 .
, m O O
. cL -
-2000 O
- 6. 14000 -k e
.L g. '@: 3 in 12000 -
E
' e A- L Q.
-1500 10000 -
80C0 O 2000 4000 6000 8000 Time (s) Figure A-253. Sequence 9-44 reactor vessel downcomer fluid pressure. 500 ,. , ,
,.600 550 - ^ -
x -500 ^ v F. v. 8
,500 -
- o 3
- 3.
O 400 .*
' g L
[450 -
.g_
E E 300 8 400 -
-200 350
-0 2000 '4000 6000 8000 Tim. (s) -
. Figure A-254. Sequence 9-44 reactor veuel downcomer fluid temperature.
A 130
16000 , ., ,
-2200.
14000 _~- --2000 n n O ,
.O.
- g. .
. 5 -1 1800 I 12000 -
u 3
.M 3 e -
-1600 N m #
E 10000 - - I
-1400 8000 O 2000 4000 6000 8000
- Tim. (s) i I
g Figure A-255. Sequence 945 reactor vessel downcomer' fluid pressure. ,.1 600 , , ,
~
-600 m 550 .
-500 e
u #
.3 b-
~g 500 -
2 u 0 c L-1 1
~
-400
.! -E 450 -- -
* ~
-300-400 ' '
~
0- 2000 4000 .6000, 8000 Time' (s)' l.
. Figure A-256. Sequence 9-45 reactor vessel downcomer fluid temperature.
~A 131
. . . _ ~ . . ., . . . - - . . . . . . . - . . ---.
W 4 T WMO , , ,
-2200' L
14000 -
.A -
i --2000 m
- I .O_ -
,O m
a. i, . Y. -
-1800 O .
' e u - 12000 - -
u-1
- o. 3 M a *
. m T e- 1600 g ,
.u
- 0. ..
jogoo . _ [
-1400
-1200
, 8000 O 2000 4000 . 6000 '8000-T ime . (S) Figure A-257. Sequence 9-46 reactor vessel downcomer fluid pressure. k A 600 , , ,
-600 1
{ 550 -
^ -
-500 ^
s M ks-v v .
' MO - - #
u 3 400 3 g O u g- [450 [
, .E E g -
-300 *
- 400 - -
.. 1
-200' ;
- 350
, -0 2000 4000 - 6000- 8000
- Time (s)
Figure A-258.. Sequence 9-46 reactor vessel downcomer fluid temperature. 'i !
! I r
. A 132-
i: 18000 ' , , ,
-2500 16000 - -
~_ .
o 0
- n. ;;; .
,_ d .14000- -2000 O L ,
g f 3
' E 12000 - -
E
' L Q- n. .
-1500 10000 - -
i 8000
'0 2000 4000' 6000 8000 Time (s) i Figure A-259, Sequence 947 reactor vessel downcomer fluid pressure.
. 4
# 600 , , ,
-600 m S50 -
.n J6 h
-500 v m- ,
6 g 3 3
~
500 - - +
-6 O o
.g
- e e C. ~400 a E E
- v. e 450 - - H i-
-300~
400 ' ' ' - l t 0 '2000 4000 6000 8000
- Time (s)
Figure A-260. Sequence 9-47 reactor vessel downcomer fluid temperature.
- IA133
.,-as.- w es---
e- y y + --m, -- n s y- 4 m
16000 , , , (, 2300 15500 -- --2250 , n -n
'O- 0 A m-3 v -
-2200 E~ v .
- 15000 - - e 3 '
> n 3
-m 2150 m.
'.I e 6
, M E
l (- A. -
-I 14500 . -_2M
-2050 14000 O 2000 4000 6000 8000 Time (s) ,
Figure A-261. Sequence 9-48 reactor vessel downcomer fluid pressure. ; s' f: i I i i 554 m a M
--552 b
v v 562-- e [ 6 r 3 l - .3 j c - 550 - o t 6 l 'e . e O. -a E 560 - E-
= -
54a e W >- 546 .. I l~ 558 : ! O- 2000' 4000 6000 8000 ' l Time (s)~ ' Figure A-262, Sequence 9-48 reactor vessel downcomer Duid temperature. f l A-134
-16500 ,. ,
16000 -
-2300 9
s 7._
, . 6 15500 - -
v E e e c. 3
-2200 '
so 3 g - 15000 - g
.u e g
A
'14500 - -
2100 14000 ' ' O 2000 4000 6000' 8000
-Time (s)
Figure A-263. Sequence 9-49 reactor vessel downcomer fluid pressure. 4 580 , ., , n M 560~ [ --550 m v Ad-v c -e
' L 3 3
]'540 - -
g L e' -
-500. e O- c.
E .E o e 520 - - H
.p.
-450 500 0- 2000 4000 '6000- 8000
..T i me . (s) .
Figure A-264. Sequence 9-49 reactor vessel downcomer fluid temperature.- A 135 - m
. . - . . . . . - ..~ .- . .. .
16500' . ..i . ' 16000 .j
-2300 m .
. ^ '
- O .O '
...1- ' O. m 0-6 15500 -
- v. ,
i' :e e 2200 ' 3 ' 3 , n .m 4 . m 15000 -
.m .,
t e- e L L
- n. n.
i- 14500 . -- 2100 ,1 : i I 14000 .' ' ' 8000 O' 2000 4000 6000 .I . Time '(s) , Figure A-265. Sequence 9-50 reactor vessel downcomer Guld pressure. i i a
- 564 , , ,
l - 554 i1 m ^ M
--552 v l'-
v 562-- s e L L 3 .3 O. - 550 o L L e- e O. a.
.E~560 -
E s 545 e H H
-546 . .
SM
-0 2000 ' 4000 6000 8000 '
Time (s) i
- - Figure A-266. Sequence 9 50 reactor vessel downcomer fluid temperature. .t 1
A.136 ' + 4s- t t wr r- i e- t - e, -,yr
- r e-4-, -
, = - ).
t 117000- , , ,
-2400 16000 -
m
-n .O L -2200 E 6 15000 - - -
c.
- e 6- g 3
. $ 14000 - -
E-
-2000 e
' u E- Q.
13000 -
-1800 12000 O 2000 .4000 6000 8000 Time (s)
Figure A-267. Sequence 9-51 reactor vessel downcomer fluid pressure. 580 i , , m 560- --550 m M }*-
'v v e e w g 3 3
]'540 - -
g L e -500 e
- c. . o.
E- E-c e F 520 -~ -
- 450-500 O 2000- 4000 -6000 8000
- Time- (s).
Figure A-268. Sequence 9-51 reactor vessel downcomer fluid temperature. o A-137 : N _-- -- - - . - - _ - - - '
-,-.,----,--a a ---_1 - - - - - - -_--s- - - . - - - - - - - - - - - - - - - - - - - - - - - - - . ~ - - . - - - - - u- - - - . - - - - - - . - , - - . - - . . -__s ----- -
i m 4 y a; i o #. _ x . ,.m~ . . 4 - _. .. .. l 16000 ,. , , L -2300 I j 15500 -- --2250 . . m n !- 0 0 .; L l^
- a. -
- .:g M
* ~
C
-2200 v- -
e u 15000 -
- e
- a. u n -
3 e -2150 m-e 1 m u_ l e n. i u I n.. 14500 . - l -.-2100 2050 11000 ' ' ' ' 0 2000 4000 . 6000 8000 Time (s) t Fisure A-269. Sequence 9-52 reactor vessel downcomer fluid pressure.~ . l 564 , , , i 563 '- -
.n m x i v 562 -- - -552 ' v}a- ~
. e w u .
3 3
]'561. -
550 -o u e o-
' O.' . O.
- E- :
E. Sao
--54a e
>- -H 559 - -
546 , 558
'O 2000 4000 6000' 8000 Time (s)
Figure A-270. Sequence 9-52 reactor vessel downcomer fluid temperature. I l (. . , . . A-138 1
.:_--_________-._--_-_=.-___-__=_. _ _ _ _ ..
, ,.. , , , ... - . - . . . . . .. ~ . .- _ . . _ .
c
~
+
16500 , , ,. ,
-16000 -
I m
-2300 n U. J .O
.g. .
. 6 .15500 -
v E i e e u.
. 3'
~
-2200 '
m: 3
- e 15000 J -
- e
, e. l e e u- .i. m- , o'. j j 14500 - -
-22 1
i f 14000
- O 2000- 4000 6000 8000 I L Time (s) !
] Figure A-271. Sequence 9-53 reactor vessel downcomer fluid pressure. 1 t . i i i
- 580 i i i f n 560- . -550 m 26, E v
l
- e w- g 1 3 2 i *U 540 - - +
.i L. O g r e - 500 e ! Q- o. 1 E E t
= e ;
, 520 - - M j i, 4 450 l 500- , ,' ~ 0 2000 4000 - 6000 8000 4 Time- (s) . i . Figure A-272. Sequence 9-53 reactor vessel downcomer fluid temperature. e o I. l i l' A-139 i
16500 i. i i 16000 -
-2300 - ^
O e ;; 6 15500 - v
- o. .
O
' o 3
-2200 g a
m 15000 -
- m # n L e L
G-O. 14500 - --2100 14000 O 2000 4000 6000 8000 l Time (s) Figure A-273. Sequence 9-54 reactor vessel downcomer fluid pressure. 564 i i i l l
-554 I
n m i M l'- ! " 562 -- --552 v v e L L 3 3 o -
-550 3L 6
v o Q. Q. E 5s0 - E o - 548 e H , H I 1 1 l
-546 558 O 2000 4000 6000 8009 -
l Time (s) Figure A-274. Sequence 9 54 reactor vessel downcomer fluid temperature. A.140
s 17000 , , ,
-2400
-16000 -
n 0 Q- .3
-2200 m 6 15000 - -
3
.'$ 14000 3
L 3 o E w.
-2000 e 0- g Q.
13000 --
-1800 12000 O 2000 4000 6000 8000 Time (s)
Figure A-275. Sequence 9-55 reactor vessel downcomer fluid pressure. s
'580 , i i m 560 - --550 . m M b v
v c e L L 3 3
}L 540 -
}w e -500 e
- c. . o. . .
E E.
~s e 520 - - *
-450 500 ' ' '
- O 2000 4000 6000' 8000
. Time (s)
Figure A.276. Sequence 9-55 reactor vessel downcomer fluid temperature.
..A 141
_-~- 4 4
-16000' . , .
-2500 i
15500-- --2250 . s n O Q
, .0.. m
} 6 -
-2200 O .-
- 15000 u - - 8 ;
3 3 i 6- * -
-2150 m a m
-
- e L. g G- c.
i 14500 - -
--2100
-2050 l 14000
-0 2000 4000 6000 8000 Time (s) i 4
) Figure A-277. Sequence 9-56 reactor vessel downcomer fluid pressure. 1 i 1 i . j M2 . , , i' . j 561_- - 550 a f m M &*- : 4 v v
- 'e 560 - .
548 i 3
- t. 3. .
.i O O 6 L [
* '- # i G. 559 O. .c' E -
54s c
. .3 W
) - S58 - 4 i 544 .- ! I ' ' '
'S57 * ! O '
2000 4000 6000 8000
- i- Time. (s)
.: Figure A 278. Sequence 9-56 reactor venet downcomer fluid temperature.
4 A 142-i
i 15000 , , ., 2200 14000 ,
. m --2000 O^
O
. ct
- d_ _
v
. ,,00
[ 12000 - -
.c d
- a 3 g M e -
1600 g I 10000 - - 1400 l 2
~
' ' ' 1200. .
. -8000 O .2000 4000 6000 8000 ,
-; Time (s) 1 x
- Figure A-279. Sequence 9-57 reactor vessel downcomer fluid pressure.
~ ,
. 600 , , _,
-600 m 550 - -
m M b 500 v s e 4 6 L
+ 3 3 ; } $00 - -
g , , .c u o 'e , i c. -400 o, l E s
.F '
5 450 - -
- s f --
300 ~
.400
.l.
- 0 2000 4000 6000 8000
; Time (s)
Figure A-280.' Sequence 9-57 reactor vessel de encomer fluid temperature. A 143.
,p . . ,.,..c_.
. . .. . . .~. . ..- . . . . - = . - , - . .. . - . .
~
l l- 16000 , , , I
-2200 l-l
-14000 , .,
, n -2000 m 0 i 0 e - i .x -
.v Q. , t
-1800 L [ 12000 - - e i a s-l e 3
~m
- e -
-1600 g A I l m .
.~.
-1400 P
, , , -1200 0 2000 4000 6000- 8000 Time (s)
Figure A 281. Sequence 9-58 reactor vessel downcomer fluid pressure. i l 400 , , , . i 600 ( i I ; I
- m 550 -
m i M V. v -
-500 v e e
' L
, 3' s 0 500 t
' o l 6 400 1 -
( e- ! j F 4g _- _ H ' 300 . I l ] l 0 2000 4000 4000 8000 " Time (s) *I l Figure A-282. Sequence 9-$8 reactor veneel downcomer fluid tanperature.
- A.144
E a i .
-2500 18000 - _
- n. m O O L " i 6 14000 , ,2000 g f
v
[ e e i 3 L E '12000 - . S. e e l ' L i A A-f 1500 10000 -- . I 8MO i
~
0 2000 4000 6000 .8000 Time (s) Figure A-283. Sequence 9-59 reactor vessel downcomer fluid pressure. , I I I 600 550 - . v b
-500 v e e u '
3 f6 5% ~ . ? O I. o ' 400. E f f l l
" 4M - .
l F ,
~
300 400 0 2000 4000 6000 8000 Time (s) . r Figure A.284. Sequence 9-$9 reactor vessel downcomer fluid temperature. l
.A.145 l- ,
l . I
l 18000 i , ,
-2500 16000 - -
n O 7 Q- e6 6 14000
-2000 O
- e e L
L 3 3 E 12000 - - E e . b L Q- G. 1500 10000 - - 8000 O 2000 4000 6000 8000 Time (s) l Figure A-28$. Sequence 9-60 reactor vessel downcomer fluid pressure. l 600 i i i 600 m 550 - -
^
v M h 500 V e e L L 3 3
}L 500 -
3L E 400 1 E
.E e H
450 - - 300 400 O 2000 4000 6000 8000 , Time (s) Figure A 286. Sequence 940 reactor vessel downcomer fluid temperature. A.146
, - .. , . ~ . - . - . . . , , . . . . . _ . - - . .
I 16000 ., , ,
-2200 . 14000 -
l n
-2000. m
'O
- O-
~
A. -m y
.. v 4 l
-1800 ;
u 12000 .
. .e s s' e *
- n m i
t e - 1600 e u - L g-A 10000 -- -- -
-1400
' ' ' ~'
8000 0 2000 4000 6000 8000 Time (s) . Figure A-287. Sequence 941 reactor vessel downcomer fluid pressure. L t 600 , , , 600 m 550 - m v M P-300 - v e L 3 3
*0 500 - -. * - -
0 e
- L, l a -400 g i E i = .
l " 4% - - H
]
-300-400 ,
- i. 0 2000 4000 6000 8000 '
? Time (s) . i Figure A.288. Sequence 9-61 reactor vessel downcomer fluid temperature. , 1 l 1 I A 147
16000 r- 1 ,
-2200
^ 14000 _,r --2000 m O O . G- m 3 O.
-1300
[ 12000 - - #
= a
- n
-1600 g ' u G- c.
10000 - - 1400 2000 O 2000 4000 6000 8000 Time (s) Figure A-239. Sequence 9-62 reactor sessel downcomer fluid preuure. 600 i , , 600 m 550 - a M v b 500 v e e L L 3 3
} SCO g
L w E -400 1 E E e e H 450 - - " 300 . 400 O 2000 4000 6000 8000 ~ Time (s) Figure A 290. Sequence 9-62 reactor wnel downcomer fluid temperature. A 14A
16000 , , ,
- * -2200 . 14000 _ -2000 ^
m O 0 t -
- o. M
.:c
. U
}
I -1800 h u 12000 - - 3 3 m M m M e -
-1600 u
1 _ I 10000 - -
-1400 8000 I
O 2000 4000 6000 8000 Time (s) Figure A-291. Sequence 9-63 reactor vessel downcomer fluid pressure. l i 600 , , ,
-600 550 -
^ -500 ^
b b v
# 500 - - e 3 3
~ ~
400 + 0 6 w e [450 - - n. E E
$ -300 #
l 400 - -
~
1
-200 350 O 2000 4000 6000 8000 Time (s) rigure A 292. Sequence 9-63 reactor venel downcomer fluid temperature.
A 149
c'
~18000 i i i
-2500 16000 -
n O
. q A e 6 14000 -
-,.2000 0 .
e e
' L 3
E 12000 - N
= =
' L Q- Q.
- 1500 10000 -
8000 0 2000 4000 6000 8000 Time (s) Figure A-29). Sequence 9-64 reaetor vessel downcomer fluid pressure. 600 , , , 600 550 -
^ 500 ^
M &^- v v 8
,500 - -
- 3 3 400 -
O o
' L
[450 { E E
- 300 e 400 - -
200 350 O 2000 4000 6000 8000 Time (s) Figure A 294. Sequence 9-64 reactor venci downcomer fluid temperature. A 150
16000 , , ,
-2200 14000_ -2000 0^
m 0 G- A
~
6
- O.
-1800 8
[ 12000 - - a a
-1600 e -
h I 10000 - X - e 1400
~
BOCO O 2000 4000 6000 8000 Time (s) Figure A-295. Sequence 9-65 reactor vessel downcomer fluid pressure. 600 , , ,
-600 m 550 -
m M v l'- 500 v v e 6 6 3 3
} 500 - -
g u
'o 400 e
- c. O.
E E v e H 450 - -
- 300 400
. 0 2000 4000 6000 8000 Time (s) Figure A 2%. Sequence 9-65 reactor seswl downcomer fluid temperature. A 151
16000 , , ,
-2300 15500 -- - 2250 ,
m ^ O- O
- n. -
.x v -
-2200 4 .
f v 8 15000 - - e 3 " E -2150 S e = L 8
-14500 . ~-2100 2050 14000 O 2000 4000 6000 8000 Time (s)
Figure A-297. Sequence 946 reactor sessel downcomer fluid pressure. 564 , , , 554 m a M
--552 l'-
v v 562 -- 8 b e 6 3 3 o - 550 g b L
- e O- O.
E 550 - E W- - 548 e H H 546 - 558 O 2000 4000 6000 8000
- Time (s)
Figure A 298. Sequence 9-66 reactor vessel downcomer fluid temperature. A 152
r. 16500 i i i 16000 - -
-2300
^
0 h o Q. to 6 15500 - - v 0-e ' e 3
-2200 3 h
~
E
.E L
15000 e L Q. Q. 14500 - --2100 14000 0 2000 4000 6000 8000 Time (s) l l Figure A.299. Sequence 9-67 reactor vessel downcomer fluid pressure. a 580 i i i m 560 - --550 m M v l'- v c e L L i 3 3 g 540 g 6 L c -500 e
- c. a.
E E v e H
- 520 - -
l
-450 500 . 0 2000 4000 6000 8000 Time (s)
Figure A.300. Sequence 9-67 reactor veswl downcomer fluid temperature. A-153
16500 i i i
~
16000 - -
-2300 .
m . n 3 6 15500 - a ,
# e 2200 g e 15000 -
E e e b b Q- O. 14500 - -210 0 1 14000 O 2000 4000 6000 8000 l Time -(s) 1 Figure A-301. Sequence 948 reactor sessel downcomer Guid pressure. l l l 564 i i i l 554 e ^ M - 552 v b v $62 -. l c e 1 s L 3 3 3u - 550 o 6
- e e a a
- E Ss0 -
E l c - 548 e \ L H H
. 546 ,
558 l 0 2000 4000 6000 8000 l 1 Time (s) Fisure A 302. Sequence 948 reactor vessel downcomer Guid temperature. l l l A 154 l
17000 , , ,
-2400 16000 -
n .
^
O O Q-2200 *E
, 6 15000 - -
v c. e 3 '
. 3 m 14000 -
8
-2000 e Q-I 13000 -
-1800 12000 O 2000 4000 6000 8000 Time (s)
Figure A-303. Sequence 9-69 reactor vessel downcomer fluid pressure,
~
580 , , , I m 550 - --550 m M v l'- v 8
' e 3 L 3
}' 540 -
g C L C- 500 e c. E S E e 520 -
- H 450 500 ' ' '
l 0 2000 4000 6000 8000 Time (s) Figure A-3Gl. Sequence 9-69 reactor vessel downcomer fluid temperature. A 155
.l, 16000 ._ 1 i
-2300 j
15500 '- --2250 ,
^
o - c : A vi 6- -
-2200 O .
.150C0 - -
- 3 u 3
-g -
-2150 .y e '
- 6 ' g Q. L 14500 . -
-- 2100 2050 14000 O 2000 4000 . 6000 .8000 T ime - (s)
Figure A-305. Sequence 9-70 reactor vessel domncomer fluid pressure. 564 i i i
~
563'- 4 - n m M b
" 562 -- - 552 v e e 6 6 3 3 56 561- -
-550 3L e e C. G.
E 560 - e -54e E. H H
'559 -
546 . S58 0 2000 4000 6000 8000
- Time (s)
Fisure A.306. Sequence 9 70 reactor veneel downcomer fluid temperature. A 1$6 I
16500 , , . f 16000 - - 2300
^
O _O O- m
, 6 15500 - -
0-O e
-2200 g E 15000 -
E O e b L G- , O. I 14500 5 --2100 14000 O 2000 4000 6000 8000 Iime (s) Figure A-307. Sequence 9-71 reactor vessel downcomer fluid pressure. 580 , , , m 560' --550 M 12-v v 0 e
' L 3 3
}' 540 -
g i L i 0 -
-500 e
- c. a.
E E v e l 520 - l
-450 500 l . O 2000 4000 6000 8000 1 Time (s)
Figure A-308. Sequence 9-71 reactor sessel downcomer fluid temperature. A-157
16500 . i i 16000 - {
-2300 ^ m n
5 S 6 .15500 , e e
$ -2200 g E 15000 - -
E e e L L Q-Q. 14500 - --2100 14000 ' ' ' O 2000 4000 6000 8000 Time (s) Figure A-309. Sequence 9-72 reactor veswl domncemer fluid pressure. 564 . . i 554 m '
^
.E i b
v 562 -- -- 552 v e e 6 L 3 3 o -
-550 o u u e e
- G. Q.
E Sec - E
= 54a e
>- -H
-546 .
558 -- O ";i 2000 . 4000 - 6000 .8000 '
-,t Time - (s)
- l'igure A-310.. Sequence 9-72 reactor vessel downcomer fluid temperature, f
i- A 158
..s. ,
17000 i i i
-2400 16000 -
m O
"n Q.
-2200 m 6 15000 - -
S e L e 3 L 3 E 14000 - e E e L -2000 Q. L Q. 13000 -
-1500 l 12000 O 2000 4000 6000 8000 Time (s)
Figure A-311. Sequence 9-73 reactor vessel downcomer fluid pressure. l 580 i i i m 560 - --550 m v x h v c e L L 3 3
]w540 -
5L e -500 e C. 0-E E o e H 520 - ~ 450 500 O 2000 4000 6000 8000 Time (s) Figure A-312. Sequ'ncee 9-73 reactor vessel downcomer fluid temperature. A-159
t 16000 i , .
-2300
., 15500 -- - 2250 , cr n
-o Q c
f Q. m ! d' -
-2200 O ,
i
- 15000 - - 8 3 3
- i * -
-2150 m a m
-e 1
L e L , n_ o.
- 14500 . - -
2l00 l 20$0 j 14000 i ~ 0 2000 4000 6000 8000
- Time (s)
}. Figure A-313. Sequence 9-74 reactor vessel downcomer fluid pressure. J 1 l i , i i . j M2 , , , . 4 3 561_ --
--550 m ^
x v- l'- v j I e go
- _ e
-548 '
t 3
. 3.
D 0 L L
-eO.559 - *~
o. 4 E-e
-s4s Ee '
t-- H
- $58 -
-544 .
i i i 337 0 2000 4000 6000 8000 ' Time (s)
~
i .- Figure A-314. Sequence 9-74 reactor venet downcomer fluid temperature.
. A 160-
. ._ _ . - , , , . . , . . , _ - . , . , _ , . s.- . . _ .
16000 , ., , e
-2200
. 14000 , --2000 m
0 g
. n. m 3 O.
-1800
. 8 12000 - - '
3 3
- m
-1600 g l E
10000 -
/ -
L
-1400
~
8000
.0 .2000 4000 6000 8000
, . Time (s) 1 Figure A-315. Sequence 9-75 reactor vessel downcomer fluid pressure. I
- 600 , , ,
-600
.i . I ' m 550 - - m M v b
-500 - v 8 e L u ,
, 3 3
' +-
~D 500 - -
o 6 g e e c.. -400 o, . 1 E E 1 e o H
"1450 >
300 - 400 1 . 0 2000- 4000 6000 8000 ,' Time (s) , Figure'A-316. Sequence 9-75 reactor vessel downcomer fluid temperature.
- a A-161
. , . .- ~ . . . , . . . . ,
.16000 i , .
-2200 14000 - -
. m.
~
-2000 n O O G- A 6
- a .
4
-1800
[ 12000 - - # 6. 3 e -1600
.t / t 10000 -
~
-1400 5000 1
-0 2000 4000 , 6000 8000 '
l - Time (S). l Figure A-317. Sequence 9-76 reactor vessel downcomer fluid pressure. 4 600 . , ,
-600 f-m 550 - -
m- . M- }*- v- _
-500 V
- 6. _L-3 3 l
+
0 30o _ . t 0 L. L e- e
- n. -400- a E. E :
e e F 450 - -- *
-300-400
-0 2000 4000. :
6000 - 8000 , _. - Ilm. D L - Figure A-318. Sequence 9-76 reactor wssel downcomer fluid temperature.
. A-162 -__..________._____________.__L____-,_______-_.__..'.____m_._..________
WHO , , ,
-2500 16000 -
m
- n O' O Q. +- .
--2000 E
,. 6 14000 v I e e L
.3- 6
- 3 m 12000 - - m .
,e en i L e Q. '
(L 10000 - . 8000 O 2000- 4000 6000 8000 Time (s) Figure A-319. Sequence 9-77 reactor vessel downcomer fluid pressure. 600 , , ,
-600 m 550 -
'5 -
-500 v k'-
L ' a
.3-s -.
.]g 500 -
O
- l. g 6.
C- ~
-400 E
, E E
'N 450 -
300 l 400 ' '
' 8 0- 2000 4000 6000 8000
. Time (s)-
Figure A-320. Sequence 9-77 reactor vessel downcomer fluid temperature. A-163 - ,
- . ~ . - - . . . - . .1 , . . . . . , . .
~
. i k
16000 , , ,
-2200 14'00 0 cm.
-2000 m O- .O 4
. g..
.x M 1 v -- 0- .
a , 1800 3 u ,12000 - - e 3- ' 3 -4 M M M e - 1600 M , u e
- 'L A. - .A r
10000 - -
-1400 I
' ' ' ~'
5000 ,
; O 2000 4000 6000 '8000 j Time (s) .
- - Figure A-321. Sequence 9 78 reactor vessel downcomer fluid pressure.
} 4 1 . 1 600 , , , i
-600 l
1 i j m 550 - - m M v
- p.
.500 - v
- e L . g 3
D 500 4 - - - +- g . b 6 ! b -
-400.I
- .E. . E' H ~H 450 - -
v .
.g 400 0 2000 '4000..
6000 - 8000 . .
' Time (s)
^
. Figure A-322..' Sequ.ence 9-78 reactor vessel downcomer fluid temperature.'
( A 164 ,
-m, y 7 *- -g -9 -m ue+-g y { g .& y 99 g gg , y9p.g 9 , ,
16000 , , ,
-2200
, 14000 ,- --2000 m U o k,
- o. -
A C.
-1800 8 12000 - -
- 3 3
- m E
- -1600 g
' L Q- O.
10000 -
- 1400 8000 O 2000 4000 .
6000 8000 Time (s) Figure A-323. Sequence 9-79 reactor vessel downcomer fluid pressure.
~
600 . . . 600 m 550 - m v M l'- 500 v c e L L 3 3
} SCC L
]L
-400 a E E v e H
450 - -
-300 400 O 2000 4000 6000 8000 Time (s)
Figure A-324. Sequence 9-79 reactor vessel downcomer fluid temperature. A-165
1 l ts000 i i > i t i i
~
-2200
/
14000 , , g j- , _-2000 g O_ 6 o
-1800 t 12000 - . e 3 w m a M M
* -1600 y 10000 . ,
-1400
' ' ' ' , , , 1200 B000 0 1000 2000 3000 4000 5000 .6000 7000 8000 Time (s)
Figure A-325. Sequence 9-80 reactor vessel downcomer fluid pressure. 600 , , , , , , , 600 l m S50 . 500 e o u ' 3 " a f 5% L 0 e ' 400 $ E v 0 H H 450 -' . i
-300 .
' ' ' ' ' ' 1-400 0 1000 2000 3000 4000 5000 6000 -7000 8000 -
Time (s) l Figure A-326. Sequence 9-80 reactor vessel downcomer fluid temperature. 1 A-166
16000 i i i
-2300 15500 -- --2250 n
0 g Q- ca d - i
-2200 O
[ 15000 - -
,8 3 3 y -
-2150 y e e L L G- G.
(" 14500 - --2100
-2050 14000 O 2000 4000 .
6000 8000 Time (s) Figure A-327. Sequence 9-81 reactor vessel downcomer fluid pressure. 564 i i i 554 m ^ M l^- v 562 -- --552 v e e L L 3 3 o L
-550 &L e e C. c.
E Ss0 - E o -
-548 e W H
-546 558 O 2000 4000 6000 8000 Time (s)
Figure A-328. Sequence 9-81 reactor vessel downcomer fluid temperature. A-167 l
- . .- . . . . . . - . . _ . .~ . -. .
16500 . , , ' ~ 16000 - - i - J 2300 . Q - I o i. n.. 7n. 6 15500 - -
'e n
-2200 g
{ E 15000. - E e e
. L g o.' g, 14500 - --2100 14000 '
= 0 2000 4000 6000 8000 i ' Time (s) ; i Figure A-329. Sequence 9-82 reactor vessel downcomer fluid pressure. ] 1 9 A l q 580 , ., , > 560 - . -550 m M b. !
. v v l = e i u u i
, 3 3 , t 0 540 w j
.. u 500 .o 4 e-.
- c. 1a '
E E e e
~H 520 - -
- i i
l-
-450
- i . . .;
3,o ,
- l. O. .2000- 4000- ~6000 8000 .
I i Time -(s) 1. Figure A-330. Sequence 9-82 reactor vessel'downcomer fidd temperature. i i 1- s j
- A-168 r !
, 1 1 .. . -
... . . . . . . ._. - .m . . _ .. . _ _ . .
, 16500 , , , 16000 - -
-2300
< 0 ,0 D- m 6 15500 - - v ca.
.e ,
-2200 ' ,
m 15000 $
- e
' L CL - g 14500 - --2100 ,
l 14000 0 2000 4000 6000 8000 Tim. (s) I j Figure A-331. Sequence 9-83 reactor vessel downcomer fluid pressure. 4 i l 564 . , , , r 1
-554 i
I n a M b v 562 -- --552 v . . 1 V e
; ' L
- 3 3
-i 0 -550 g ; L
'e o C. o. ,
E Sao - E e -
-548 e-H H 3
e i
-546- i
^>
! 558 t
+ 0 -2000 4000 6000 8000'
~
.- Time ..(s)-
f 4 Figure A-332; Sequence 9-83 reactor vessel downcomer fluid temperature. l. i. 4 A 169 , d r e--- t <= -r - r c 2 % -_y 4 - - .we, e + .,
. . . .- . . . - ~. . _ . . . - _ - - . .
- - u - - .. -. . - . . ~
l i L 17000 ' , , ,
-2400 16000 -= -
- n. ^
O O
'1 .
-2200 Yn
. 6 150C0 .- - -
v
- c. ,
~e ,
' 6
/ 3 '
- wo 14000 -- -
-2000 e
' .c CL - a,
'13000 - -
-1800 12000 0 2000 4000 6000 8000 Time (s)- !
Figure A-333. Sequence 9-84 reactor vessel downcomer fluid pressure. P 580 , , , n 560 --550 m M la- -
- v. v e e b 6 3 3
]w-540 - -
g w t e -
-500 o
- c. '- o.
-E- E
- e. .
520 ;
- 450 g e i i 0 2000 4000 - :6000 - 8000 '
Time (s) - Figure A-334. Sequence 9-84 reactor vessel downcomer fluid temperature.
-A 170
_- , . - . . . _ _ . . _ . . . . _ . . . . ... . .._-_...m , i f me0 , , .
. -2300 15500 -- -2250
.m ^
J 0 0
- . .3. -
]
.x
.' .v -
-2200 O I.15000 - - e ,
o u -
.e 3
-2150~ m o "
u i e
! L
$ G,, . 14500 - -
--210 0 J -
-2050 2 14000 O 2000 4000 6000 8000 ;
Time (s)
+
4 l Figure A-335. Sequence 9-85 reactor vessel downcomer fluid pressure. b 1 i.* 1 564 , i i 554 :'-
. 563 - -
^ ^
M
--552 l'-
v
" 562--
e- e L L 3- 3 j ' ]L 561. - 550 5 ,' e b. o C. Q. ! E 56o - E s- - i -548 e H w i 559 - -
-546 i'
558
' O- 2000 4000' 6000 8000 Time' (s) 5
- Figure A-336.' Sequence 9-85 reactor vessel downcomer fluid temperature.
f 9 4 -- A-171
16000 , i ,
-2200 14000, -
m 2000 ^ O U n_ *g 6 *
- o. *
-1800
[ 12000 - a a e -
-1600 E / L 10000 -
-1400 2000 O 2000 4000 6000 8000 Time (s)
Figure A-337. Sequence 9-86 reactor sessel downcomer fluid pressure. 600 , , ,
-600 m S50 -
n M l'- v - 500 v c e L L 3 3
]' 500 -
g e w
- c. e
-400 a.
E E v e F
- 450 -
-300
- 400 O 2000 4000 6000 8000 ,
Time (s) Figure A-338. Sequence 9-86 reactor vessel downcomer fluid temperature. A-172
1 WMO , , ,
-2200
'., 14000 _
' ^ -2000 ^
j O S
=
a f -1800 I 12000 - - ' u 3 s M- g
-1600 g L u L' a, 10000 - -
-1400 8000 .
O 2000- 4000 6000 8000 Time (s) Figure A-339. Sequence 9-87 reactor vessel downcomer fluid pressure. I I I 600 m 550 - m M v l'- 500 v ef' e
' L 3 3
]'500 - -
g L 400< a E E o . e H gg - - H 300
-400- ~~
O 2000 4000 - - 6000 - 8000-- Time (s) Figure A-340. Sequence 9-87 reactor vessel downcomer fluid temperature. A-173
16000 i i i 14000 -
~_ poco ^
o c2 a 6 12000 - ~ v C- ' e e L f L
@ 1500 g n 10000 ~
m e - 5 e L L a- cL 80C0 - ~
-1000 6000 O 2000 4000 6000 8000 Time (s)
Figure A-141. Sequence 101 reactor vessel downcomer fluid pressure. 560 i i i l 540 - ~ m -500 ^ M h v v 8 520 - ~
- 3 3
[ - 450 8 [500 - ~ E E E N 480,- f ~ 400
?
460 O 2000 4000 6000 8000 ,
' Time (s)
Figure A 342. Sequence 101 reactor vessel downcomer fluid temperature. A 174
t t F 16000 i i ,
-14000 ,-
_-2000 t ; ,
' ^ m O .O.
- a. .
, - 6 12000 - _
v { , e e
- u 6 3
1500 'j E 10000 ' - _ e L e 2 L u
- n. n.
.8000 --
i 1
-1000 ,
j 6000 , O 2000 4000 . 6000 3000 l 1 Time (s)- ; 1
! Figure A-343. Sequence 10-2 reactor vessel downcomer fluid pressure. ,
! t s 1
- i. .
3 600 , , ,
-600 i
l 4
.n 550 i
b b ' i 500 v
; e e u '
- s 1
]u500 - . 3 ,
o ' 4 6 g-
-400 - E'
{ E 1 e 450 .
. H
.m
' i i 400 I-0 2000 4000 .
.6000 8000
- Time (s)
T i Figure A-344. Sequence 10-2 reactor vessel downcomer fluid temperature. 4' ' A 175 4 . a - _______me._____._______--____mm__m__
20000 , , ,
-2500 15000 --
m 9 -
-2000 0 0-3 7c.
v v . 8 10000~- - 1500 e a 58 s vi 1000 C s - O.. - $ 5000 [
- -500 0 O O 2000 4000 6000 8000 Time (s)
Figure A-345. Sequence 10-3 reactor sessel downcomer fluid pressure. 560 , , i 540 - - a 500 ^ M v l'- v
- 520 - - @
3
- .3-O 450 o
6 - g [500 - - { E E e e H >=- 480,- --400 460 O 2000 4000 6000 8000 ' Time (s) l Figure A-346. Sequence 10-3 reactor vessel downcomer fluid temperature. I A-176 i i
16000 i i i 14000 -
--2000 n 9 e i , 6 12000 - -
S e h e L L
@ 1500 3 m 10000 - -
e e e L L Q. 0 8000 - - 1000 6000 O 2000 4000 6000 8000 Time (s) Figure A-347. Sequence 10-4 reactor vessel downcomer fluid pressure. 560 - i , i m S40 - - a v M l'- 500 v c e 6 6 3 3 O 520 O 6 L e e C. Q. i h - 450 h 500 - - { 480 O 2000 4000 6000 8000 Time (s) Figure A-348. Sequence 10-4 reactor vessel downcomer fluid temperature, l A-177 ! i
16000 i i i 14000 -
-2000 m
O y Q- e 6 12000 - - Q. , e c
' L @ -1500 g m 10000 - -
e e (' e s CL Q. 8000 - -
-1000 6000 O 2000 4000 6000 8000 Time (s)
Figure A-349. Sequence 10-5 reactor vessel downcomer fluid pressure. 600 , 3 i 600 m 550 ^ M l'- V v $50 - - v e
$ 500 -$
_ t j 0 O L L e e C. . 450 a E 500 - - E v e b- >- 400 450 O 2000 4000 6000 8000
~
Time (s) Figure A.350. Sequence 10-5 reactor vessel downcomer fluid temperature. A 178
w 20000 i i i j- -
-2500 m ~
g { 15000 - i., 6 -
-2000 v a
- . e- e
< L L i 3 3 a m
- M e
~-1500 e
' 10000 - L A o.
1 k i i . -1000 i ' ' ' l '5000
- j. 0 2000 4000 6000 8000 j- Time (s) i
{ Figure A-351. Sequence 11 1 reactor vessel downcomer fluid pressure. i i + i-1 I I 560 i i i i j 555'- 540 m m i M k'- J v v H, . e L e L j- 3 3 a 5 550 L
--530 $L e o Q- Q.
E E e e
& H^
$45 - -
-520 i !
i . 540 O 2000 4000 6000 8000 Time (s) 1 Figure A-352. Sequence 111 reactor vessel downcomer fluid temperature. ' r
'A 179
._____ - - _ _ _ -____- - _-____- _ - _ _ _ - _ _ _ _ _ _ _ _ _ _ _ - _ - _ = _ _ _ _ -
20000 i i i
- r
-2500 m n
[ 15000 6 -
-2000 O .
e e L L 3 3 m m M M e --1500 e
' 10000 . -
L Q- n. i
- -1000 5000 O 2000 4000 6000 8000 Time (s)
Figure A 353. Sequence 11-2 reactor sessel downcomer fluid pressure. 1 l l 550 , , , 540 i m n l M k'~
" 550 -- V 530 e e L L 3 3 D -
-520 0 L L e o Q. O.
E 540 - - E 8 510
- H w l
-500 530 O 2000 4000 6000 8000 .
Time (s) Figure A-354. Sequence 112 reactor vessel downcomer fluid temperature. l A.180
. . _ _ . -_ _ , _m.. . _ ,_. - . .. __ _
t: I 20000 , , , i 2500
- n. m y 15000 - -
3 ;
. d -
-2000 0-L L i @ 3 M
m ' 1500 e
' 10000'-
, 11.
g
-1000 l
5000 O =2000 4000 6000 8000 Time (s) Figure A 355. Sequence 11-3 reactor vessel downcomer fluid pressure. ( 4 560 , , , , t 540 n 555~- - g , v o.-
- 6 3 3
}6 550 -
--530 $
.g ct Ic- E o
545 - -
-520
+
S40 0 2000' 4000 6000 8000
- Time (s) =
- Figure A-356. Sequence 11-3 reactor vessel downcomer fluid temperature.
i l- .A-181 i_.__________-.______________________.____.______________._____._______.__.____.__________-.__..____'__________-__________.________.m_
- m._ . . .
m k:
'..j.
20000" i. . i
-2500 e
^ o 0 - - -
, n. 15000 m
~6 -
-2000 O .
l e e ! L' L. 3
.n a M
a M
.e --1500 e .
t 10000 - - L
> n'.- a.
1 1000 l 5000 ; 0 2000 4000. 6000 8000 '
-T i me - (s) ,
Figure A-357. Sequence 114 reactor vessel downcomer fluid pressure. t i i i ( W i i i l u l l l 555~- - 540 m m-M v l'- v j ' 550-- --530 u s.
+
3 O O L ' L. E545
--520 E. E o- e W >=
540 - - 510 . l i S35'
- j. 0- 2000 4000- - 6000 8000 .
-Time (s)
Figure A-358. Sequence 114 reactor veneel downcomer fluid temperature. A-182
15000 , . i 2500 16000 - n [ 14000 _- - Q x 2000 ;; o V a. v 8 12000 - 3 8 e a m m 0 m
' 10000 - 1500 o C-Q.
8000 - 1000 I
' ' ' 4 6000 - -
2000- 4000 6000
)
C.- 8000 Time (s) Figure A-359.. Sequence 121 reactor vessel downcomer fluid pressure. a +x. *
+ ,/
560 ,
, , , 5- ',
- ~ . ,. .
I 540
,, e
\ 3p - s .
?
n m . _Y a v (- ~g"' g 520 fu,V u 540 L l . , ' . O 3* / u
./ 4 7 ,3 ,
8 '2. ., 500 8
$530 Et f r '
[ , v /?' F W -:' s '
/, , , "
<~ ~ ,' Ey s >--
v ,#' a / 480 ^
- W c,~,
520 +- - ' l , e .. ' , t
.. - r , , , s .
+ - '
1 f 1
- - 460 i , O F00 4000 6000 ty'. d
/
r* Tirne (s) >
'4 t ? b g*
i
<
- liare A-3@.' Sequence 12-1 reactor vessel downcamer fluid temperature. +
) /
< ,4-1, y' r
,. rr >
- a se
,,E
; fr *
>> o e*
- Y p
wf A 183 . f ' a s- ,
,N4 .,I #
- a; -
p
' z
~
- q. c, sj f ks -
* .')
l 16000 , , 14000 ~
~-2000
/ m
'O' O. p/ O
'O 6 12000 - -
a . v o 0 5 - en 1500 @ g 10000 - ,
' C Q- f u Q.
8000 _
~
-1000 6000 ' ' i 0 2000 4000 6000 8000 Time (s)
Figure A-361. Sequence 12-2 reactor vessel downcomer Duid pressure. 560 , , , 540 550 - m ^ Y 520 vb [$40 - u 3 3 a o
' 500 0 g $530 - _
a E E o v H 480 520 - _
' ' ' 460 510' 0 2000 4000 6000 8000 -
Time (s) Figure A-362. Sequence 12-2 reactor vessel downcomer fluid temperature. A-184 j
20000 -r - r - - --~ - r--- - j 2$00 15000 - o -
-2000 o
- c. *g
. 6 8 8 10000 ' -
--5 00 e a '
m 3 m M o " 1000
- U-0 5000 -
500 0 O O 2000 4000 6000 8000 Time (s) Figure A-363. Sequence 12-3 reactor vessel downcomer fluid pressure. l l 600 , , , 600 l I l
^ -
-550 ^ '
M b-V 550 - 0 o 5 -
- 500 -$ I O 1 O l
'v L y
l
\
C- - 450 C-E l o 500 E ' b y 1 F-400 450
, 0 2000 4000 6000 8000 Time (s)
Figure A-364. Sequence 12-3 reactor vessel downcomer fluid temperature. 1 1 1 l r l I A-185 l
r 20000 i i r- -
-2$00 15000 -
3 a 2000 3, _. 6 5 . 8 10000 ~ - 1500 e a - v1 g 3
- e. M e
L -
-1000
- a ,
5000 - a. 500 0 ' ' ' O O 2000 4000 6000 8000 Time (s) Figure A-365. Sequence 124 reactor vessel downcomer fluid pressure. 5000 , , ' 8000 4000 - 2 v 6000v [ 0
' 3000 o
3 u 3' 4000 3 y u
$ 2000 -
E { D E o
-2000 w 1000 -
' ' .o 0 '
O 2000 4000 6000 8000 Time (s)
- Figure A-366. Sequence 124 reactor vessel downcomer fluid temperature.
A2335 A-186 s
.o n cs ... .. .. , - ....o.,-..-,~-
62 a
"?'M
. sisuoanAPHIC DATA SHEET NUREG/CR-3935
.. c,.on o_, ..w.a EGG-2335 u.....
Thermal-H raulic Analyses of Overcooling [ Sequences r the H. B. Robinson Unit 2 f .,,,,..,c_,,,,
. Pressurized .ermal Shock Study "
o.. .... g
.v May 1985 C. Don Fletche f[
oa 'i .~' **o Cliff B. Davis " '" l Donald M. Oncion ! June 1985
. -.-. ,. r.o . .s <. -
.oo.. . .
. ou m .- -. - > -. .
EG&G Idaho, Inc. I,
' '"'''''~
P. O. Box 1625 l Idaho Falls, ID 83415 L
.e sm-
\ se ca.w ,
O[8 A6047
* *. , v ,e o. ..po.,
l j U.S. Nuclear Regulatory Co. , sion Division of Accident Evaluati Technical
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Office of Nuclear Regulatory Re archj Washington, DC 20555 u .u t. . . . .. . o . .
+r Oak Ridge National Laboratory (ORNL) a party the Nuclear Regulatory Commission's (NRC's) pressurized thermal shock (PTS) integration s y for the rs lution of Unresolved Safety issue A49, dentified overcooling sequences of interest to the H. .' Robinson PT udy. For each sequence. reactor vessel down-comer fluid pressure and temperature hi rics were require or the two-hour period following the ini-tiating event. Analyses previously perfor ed at the Idaho Natt 11 Engineering Laboratory (INEL) fully investigated a limited number of the ences using a detailed LAPS model of the H. B. Robinson, Unit 2 (HBR-2) plant. However, a full i estigation of all sequences the detailed model was not econom-ically practical. New methods were r ired to generate results for th emaining sequences. Pressure and temperature histories for these rema' ing sequences were generated at t NEL through a process combin-ing: (a) partial-length caletlations ing the detailed RELAPS model,( 'ull-length calculations using a simplified RELAP5 model, and hand calculations. This report docum bts both the methods used in this process and the results.
The sequences investigated ntain significant conservatisms concerning eq ment failures, operator actions, or both. Consequen , care should be taken in applying the results pre nted herein without an understanding of the conse tisms and assumptions. Theresultsof thetherma ydraulic analyses presented here, along with additional analyses of multidimen-sional and fracture mech ics effects, will be utilized by ORNL to assist the NRC in resolving the PTS _ unresolved safety issue.
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