ML17309A205

Forwards Draft Safety Evaluation of SEP Topics VI-2.D, Mass & Energy Release for Possible Pipe Break Inside Containment, & VI-3, Containment Pressure & Heat Removal Capability & Lll Draft Technical Evaluation

ML17309A205
Person / Time
Site:	Ginna
Issue date:	11/03/1981
From:	Crutchfield D Office of Nuclear Reactor Regulation
To:	Maier J ROCHESTER GAS & ELECTRIC CORP.
References
TASK-06-02.D, TASK-06-03, TASK-6-2.D, TASK-6-3, TASK-RR 8111103, LSO5-81-11-004, LSO5-81-11-4, NUDOCS 8111200828
Download: ML17309A205 (95)

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. y~~/PQ f4(Q November 3, 1981 Docket No. 50-244 LS05 11-004 Mr. John E. Maier, Vice President Electric and Steam Production Rochester Gas 5 Electric Corporation 89 East Avenue Rochester, New York 14649

SUBJECT:

SYSTEMATIC EVALUATION PROGRAII (SEP) FOR THE R. G. GINNA NUCLEAR POWER PLANT - EVALUATION REPORT ON TOPICS VI-2.D AND VI-3 Enclosed is a copy of our draft evaluation of SEP Topics VI-2.D, "Mass and Energy Release for Possible Pipe Break Inside Containment," and VI-3, "Containment Pressure and Heat Removal Capability." This evaluation compares your facility, as described in Docket No. 50-244, with the criteria curr ently used by the regulatory staff for licensing new faci-lities. Appendix A to our draft evaluation is a draft Technical Evaluation Report from our contractor, Lawrence Livermore National Laboratory. Please inform us if your as-built facility differs from the licensing basis assumed in our assessment. Comments are requested within 30 days of the receipt of this letter so that they may be considered in our final evaluation. This evaluation will be a basic input to the integrated safety assessment for your facility unless you identify changes needed to reflect the as-built conditions at your facility. This assessment may be revised in the future if your facility design is changed or if NRC criteria re'Iating to this subject are modified before the integrated assessment is completed. Sincerely,

Enclosure:

Draft SEP Topics VI-2.D and VI-3 ~ ~o~ Dennis M. Crutchfield, Chief Operating Reactors Branch No. 5 I / Division of Licensing QblS (o> AaP 'ATE P cc w/enclosure: See next page OFF ICE II SURNAME$ ~" 8iii200828 8iii03'DR ADQCK 05000244 PDR ~ ~ ~<<p ~ <<<< NRC FORM 318 (10-80) NRCM 0240 (See previous concurrance sheet) OFFICIAL RECORD COPY USGPO: 1991~~960

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g~8 RECO ~Ci+ 0 Cp 're &0 UNITED STATES NUCLfAR REGULATORY COMMISSION WASHINGTON D. C. 20555 November 0, 1981 Docket No. 50-244 LS05-81-11-004 ' ~ Mr. John E. Maier, Vice President Electric and Steam Production Rochester Gas Il Electric Corporation 89 East Avenue Rochester, New York 14649

SUBJECT:

SYSTEMATIC EVALUATION PROGRAM (SEP) FOR THE R. G. GINNA NUCLEAR POMER PLANT.- EVALUATION REPORT ON TOPICS Vl-2.D AND VI-3 Enclosed is a copy of our draft evaluation of SEP Topics VI-2.0, "Mass and Energy Release for Possible Pipe Break Inside Containment," and VI-3, "Containment Pr essure and HeatRemoval Capability." This evaluation compares your facility, as described in Docket No. 50-244, with the criteria currently used by the regulatory staff for licensing new faci-lities. Appendix A to our draft evaluation is a draft Technical Evaluation Report from our contractor, Lawrence Livermore National Laboratory. Please inform us if your as-built facility differs from the licensing basis assumed in our assessment. Comments are requested within 30 days of the receipt of this letter so that they may be considered in our final evaluation-. This evaluation will be a basic input to the integrated safety assessment for your facility unless you identify changes needed to reflect the as-built conditions at your facility. This assessment may be revised in the futur e if your facility design is changed or if NRC criteria relating to this subject are modified before the integrated assessment is completed. Sincerely,

Enclosure:

Draft SEP Topics VI-2.D and VI-3 cc w/enclosure: See next page'ennis M. Crutchfield, Chief Operating. Reactors Branch No. 5 Division of Licensing

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Wj /P Mr. John E. Maier CC Harry H. Yoigt, Esquire

LeBoeuf, Lamb, Leiby and MacRae 1333 New Hampshire Avenue, N. M;-,

Suite 1100 Mashington, D. C. 20036 Mr. Michael Slade 12 Trailwood Circle Rochester, New Yank 14618 Ezra Bialik Assistant Attorney General Environmental Protection Bureau New York State Department of Law 2 Morld Trade Center New York, New York 10047 Jeffrey Cohen New York State Energy Office Swan Street Building Core 1, Second Floor Empire State Plaza

Albany, New York 12223 Director, Bureau of Nuclear Operations State of New York Energy Office Agency Building 2 Empire State Plaza

Albany, New York 12223 Rochester Public Library 115 South Avenue Rochester, New York 14604 Supervisor of the Town of Ontario 107 Ridge Road Mest

Ontario, New York 14519 Resident Inspector R. E. Ginna Plant c/o U. S.

NRC 1503,Lake Road

Ontario, New York 14519 Mr. Thomas B. Cochran Natural Resources Defense Council, Inc.

1725 I Street, N. M. Suite 600 Mashington, D. C. 20006 U. S. Environmental Protection Agency Region II Office ATTN: Regional Radiation Representative 26 Federal Plaza New York, New York 10007 Herbert Grossman, Esq., Chairman Atomic Safety and Licensing Board U. S. Nuclear Regulatory Comtission Mashington, D. C. 20555 Dr. Richard F. Cole Atomic Safety and Licensing Board U. S. Nuclear Regulatory Comission Mashington, D. C. 20555 Dr. Emmeth A. Luebke Atomic Safety and Licensing Board U. S. Nuclear Regulatory Coneission Mashington, D. C. 20555 h h P h, ~

SAFETY EVALUATION REPORT ON CONTAINMENT PRESSURE AND HEAT REMOYAL CAPABILITY SEP-TOP IC= YI-3. AHD MASS AHD ENERGY RELEASE FOR POSSISL. PIPE BREAK INSIDE CONTAINMENT, SEP TOPIC YI-2.0 FOR THE R. E. GINNA NUCLEAR PO'<E'ER PL.'NT 00CKE" '<0 "0-"-44 gNNEL. . Lsckstk So- ~Vg Cantrele 8I(I2Onsz.8 ates os >n-.-< E TORY tjOQQ

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TABLE OF COi'<T ITS ~ ~ ~ Introduction Rev'.ew Cr iver a IV. Re 1 a ted 5 afet J 7op ics Review Guidelines V. =valua:ion Vi. Conclusions Appendix A SEP Containment Analysis and E'valuation for the R. E. Ginna nuclear Power Plant 4g I ~ ~

Introduction The R. E. Ginna Nuclear ?ower ?lant began commercial operations in 1970. Since then the sta f 's safety review criteria have

changed, As part of the Systematic Evaluation

?. ogi'am (SEP), the containment pressure and heat removal capability (Topic VI-3) and the mass and energy release for possible pipe break inside containment (Topic yl-z.p) have been re-evaluated. The purpose o, this evaluation is.o document the deviations .rom curren. safety criteria as they relate to the containment pressure and heat removal capability and the mass/energy release for possible pipe break irside contain-

ment, Furthermore, independent analyses in accordance with current criteria were performed to determine the adequacy of the containment design basis (e.g.,

design pressure and tampe. ature) and to provide input for Unresolved Safety Issue (US !} A-24, guaiification o, Class lE Safety Related Equipment. The sig-nificance of the identified deviations, and recommended corre tive measures to improve safety, will be the subject of a subsequent, integrated assessment of the Ginna Plan-. II. Rev ew Criteria The ". eview "ri'.eria used in the c"rrent evaluation of SE? Topics V.'-2.0 and 'll-3 -, or she Ginna plant are contained in :he following documents: 10 CFR ?art 50, Apoendixc A, General les gn Cr teria for Nuclear ?owr ~lants: AA( P cue io ',b) GDC 3.S ontainment des gn; - Containment hea-. remcv=l; and (21 c) GOC ."0 - Containment desi.=r. bas s. 10 "."-R Section 0.-", "Ac eotance "." :aria -'or Emergency Core Cooling systems for Light Water Nuclear Power Reactors."

(3) 10 CFR Par 50, Appendix '<, "ECCS Evaluation Models." (4) NUREG 75/087, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear..~ow:er ?lants (SRP 5.2.1, Containment Functional 1 Oesign). III. Related Safet Topics The review areas iden ified be'.ow are not addressed in this report, but are related to the SEP topics of mass and energy release for possible pipe break inside containment, and/or containment pressure and heat removal capab i 1 ity. ( 1) III-l, Classification of Structures, Components and Systems (Seismic and guality) ( 2) III-78, Oesign

Codes, Oesign Criteria, Load Combinations, and Reactor Cavity Oesign Criteria (3)

VI-7.S, ESF Switchover from Injection to Recirculation i>lode (Automatic ECCS Realignmen ) (4) IX-3, Station Service and Cooling 'Rater Systems (5) X,.Auxiliary Feedwater System (6) USI-'A24, gualificat'on of Class 1E Safety Related Equipment IV. Review Gu!aelines Gene. al Oesign Criterion ',GOC) 15 of Apoendix A to 10 "FR Part 50 requires that a reac.or containmen and associated systems shall be provided 'o es.ablish an essentially leak-ight barrier against the uncontrolled release of radioac'ivity to the ervironment and.o assure that the containment design conditions important o safety are not exceeded for as long as the postulated acciden. conditions ", equire. GOC 38 requires a containment heat removal V ~ pl lt I y

system be provided whose sys em safety function shall be to reduce the containment pressure and temperature following any loss-of-coolant acci-dent yLQCA) apd ma'ingrain:&em at acceptably low levels; furthermore, the system sa,ety,unction shall be achievable assuming a single failure. GDC 60 requires that the containment structure and the containment heat re-movai system sha'.1 'e 'assigned so that the structure can accommodate, with sufficient margin, the calculated pressure and temperature conditions re-sulting from any LOCA. This margin as obtained ,rom the conservative calcu-lation of mass/energy release and the containment model is discussed in the Standard Review Plan (SRP) Section 6.2.1, Containmen. Functional "esign. The containment design basis includes the ef,ects of stored and generated energy in the accident. Calculations of the energy available for release should oe done in accordance with the requirements of 10 CFR Part 6C, Section 60.45 and Appendix '<, paragraoh '..A, and the conservatism as specified in SPP 6.2.1.3. The mass and energy release .o the containment ,rom a LOCA should be considered in terms of blowdown, ref'.ood, and post-reflood. The mass and energy release as a result of postulated secondary system pipe ruptures should be cà culated in accordance with SRP 5 the analysis of postulated single active fail .2.1.4. Our review also included ures of components in '.he second-ary system ln rev i wl n hc censee s ana'.ys i s, deviations from current criter > a E have been identified. independent

analyses, as requ'ired, were performed o

C evaluate the s igni, icanc of these deviations. The valua "ion was compl eted by comparing the resul s wi ,", the licensee's containmen: design basis. 1 V. Evaluation A review of the existing containment analysis for Ginna, as described in the Final Safety Analysis Report (FSAR), indicated two basic deviations from the current safety.,";criteria. First, for the LOCA analysis, the licensee had not considered a cold leg pump suction break location, the core reflood phase of mass and energy rel 'ase, or the release of secondary system energy to the containment. These aspec s of the LOCA analysis are addressed in Standard Review Plan ( SRP) 6.2. 1.3.

Second, the main steam line break (HSLB) analysis had not been performed for the Ginna Plant, SRP Sections 5.2,1.1.A and 6.2.1.4 address the NSLB analysis.

To assess the significance of these two deviations, our consultant, the Lawrence Livermore National Laboratory (LLNL) performed independent LOCA and NSLS analyses ~hich are presented in Appendix A to this report. >lass and energy release rates utilized in the analysis were calculated using RELAP-4 ,"IOO 7, and the calculation of the containment oressure and temperature re-sponse was done using CONTEMPT-LT/-28. For the pr>mary system (LOCA analysis), a double-ended break at the pump suction of the "old leg was analyzed in Appendix A since it typically is the design basis LOCA for a P'AR plant. Blowdown, reflood and post-reflood phases were considered in the calculation of mass and energy release data; the release of the secondary sys;em energy ~as a'.so factored into the calculation. The calculated ransient resu!ts show a peak containment pressure of 74 psia. and a oeak containment atmosphere temperatur of 282'F. The containmert design pres-sure and temperature for Ginna are 75 psia and 286'F. There is, therefore, a slight margin between the calculated containment pressure and temperature and the design values. .;u

Sased on our review of the LLNL LOCA analysis, we concur with their findings and conclude that the Ginna containment design basis is adequate for postulated LOCAs. For the secondary

system, the worst case

!1SLB accident identified in'ppen-dix A is that postulated to occur during hot standby concurrent with the failure of one spray line. The peak calculated containment pressure is 85.8 psia which exceeds the containment design pressure by 11 psi. The worst peak calculated containment temperature is 421'F occurring at 102.'f full power with the single failure of one spray line. It is acknowledged that the HSLB accident analysis was done in a conservative manner. This was promoted by the unavailability of a water entrainment model to more explicitly define the mass and energy release during the blowdown of a steam generator. An appropriate water entrainment model is necessary to determine the maximum break size that results in a pure steam blowdown, which is smaller than a double-erded

break, Without the water entrainmen model, it was necessary

?o as-sume a doub'.e-ended steam line break with pure steam blowdown. Also, heat rans-fer from.he primary system to the steam generator secondary side was treated conservatively'. Stean addition from the unaf;ected steam,enerator and feedwater add:t:on ".0 he affected steam generator were treated in a inner hat uas con-sisten 'w th the performance of system isolation valves or single active failure assumpt.'ons. The hot standby condition was assured for the aralysis since .he steam gener-ator would have the largest inventory of water. Calculations were also performed assuming the reactor was at 102>> of full power, a condition of low steam genera-tor inventory, to bracket the results of the analysis. V I. Conclusions Our review of the con ai nment functional design of the Ginna plant, as re-ported in Appendix A, identified deviations from current safety criteria. To assess the significance of these deviations independent containment analyses were per ormed. The results of the analyses show that the containment design conditions are not exceeded for postulated LOCAs, but are exceeded for the MSLB accident. For the MSLS analysis, the peak calculated pressure exceeded <<he con-tainment design pressure by 11 psi; i.e., 85.8 psia. It should be noted, how-

ever, that a more refined MSLB analysis may show that the containment design pressure would not be exceeded.

Moreover, the Structural Integrity Test for the Ginna containment was performed at a pressure 1.15 times the containment design

pressure, or 69 psig (84 psia).

Therefore, as conservative as the MSLB analysis is, the peak calculated containment pressure is very close to the test "ressur~ for the Structural Integrity Test. The implications of exceeding the containment design pressure

are, therefore, not of grea. concern.

Therefore, the need for the licensee to upgrade the Ginna containmen analysis (for both loss of coolant and MSLB accidents) to ~eflec the application of current VASSS vendor analytical capa-bilities to the Ginna plant for future reference in 1 censing actions, will be I 'eferred to .he outcome of the integrated assessment of the Ginna'lant. The results of LLNL analyses are extremely con'servative, especially for equipment qualification purposes. However, if the licensee chooses to use the results of this report then the MSLB temperature profile in Figure 10 of Appendix A and the LOCA temperature profile in Figure 2 of Appendix A may be used to assess the environmental qualification of Class IE safety-related electrical equipment ("SI A-24) Alternatively, the licensee may choose-to perform a more realistic analysis to establish environmental conditions for postulated LOCA and steam line breaks. Appendix A SEP -Containment Analysis and Evaluation \\ for the R. E. Ginna Nuclear Power Plant Contents Pacae 1.0 Introduction and Background 2.0 Containment Functional Design 2.1 Review of Analysis of Ginna Containment, Functional Design 2.2 Primary System Pipe Break 2.3 Secondary Systems Pipe Break 2.4 Reanalysis of Ginna Containment Functional Design 3.0 Primary System Pipe Break 3.1 Initial and Boundary Conditions 3.2 Blowdown Phase 3.3 Reflood Phase 3.4 Post-Ref lood and Containment

Response

Calculation 3.5 Containment

Response

Results 14 14 15 15 16 16 17 18 19 20 f, )'p4g 7 0 r

4.0 Secondary System P'pe Breaks

4. 1 Analytical Model 4.2 Initial Conditions and Other Assumptions 4.3 Containment Response Calculation 4.4 Steam Generator BLowdown'nd Containment

Response

Results 21 22 24 5.0 References 25

I l

List of Tables Table Pacae Initial Conditions for Conainment Analysis Engineered Safety Systems Containment Structural Heat Sinks Blowdown Energy Balance 5 Blowdown Mass and Energy Release Rates Reflood Energy Balance Reflood Mass and Energy Release Rates Description of Steam Line Break Cases Hain Steam Line Break Mass and Energy Release Rates - Case 1 26 27 28 32 34 10 Main Steam Line Release Rates Break Mass and Energy - Case 2 Main Steam Line Break Mass and Energy Release Rates - Case 3 37 Hain Steam Line Break Mass and Energy Release Rates Case 4 38 Main Steam Line Break Mass and Energy Release Rates - Case 5 -10>>

List of Figures ~Fi vres Paae Ccntainmeng Atmosphere Pressure, Ginna -Double-Ended Suction Leg Break 40 ~ 5 Containment Atmosphere Temperature, Ginna Double Ended Suction Leg Break Schematic of Analytical Model for Steam Line Break at 102X of Full Power Schematic of Analytical Model for Steam Line Break at Hot Standby Conditions Ginna MSLB, Case 1 102X Power with MSIY Failure Ginna MSLB, Case 1 - 102% Power with MSIV Failure 41 42 43 45 10 Ginna MSLB, Case 2 - 102K Power with MFIV Failure Ginna MSLB, Case 2 - 102K Power with MFIY Failure Ginna MSLB, Case 3 - 102% Power with Containment Spray Failure Ginna MSLB, Case 3 - 102X Power with Containment Spray Failure 46 47 49

List of Figures (Continued) Ginna MSLQ, Case 4 - Hot Standby with MSIV Failure Ginna MSLB, Case 4 - Hot Standby with MSlV Failure Ginna MSLB, Case 5 - Hot Standby with Containment Spray Failure Ginna MSLB, Case 5 - Hot Standby with Containment Spray Failure 50 51 52 ~ "~iv ~ ~

1.0 INTRODUCTION

ANO BACKGROUNO As part of the Systematic Evaluation Program (SEP), the containment functional design capability of the R. E. Ginna Nuclear Power Plant has been reevaluated. The puzpose of. this report is to document the resolution of SEP Safety Topic YI-2.0, Mass and Energy Release for Possible Pipe Break Inside Containment, and SEP Safety Topic V1-3, Containment Pressure and Heat Removal Capability, and deviations from current safety criteria as they relate to the containment functional design. The significance of the identified deviations and recommended corrective measures will be the subject of a subsequent integrated assessment of the R. E. Ginna plant. The containment structure encloses the reactor system and is the final barrier against the release of radioactive fission products in the event of an accident. The containment structure must, therefore, te capable of withstanding, without loss of function, the pressure and temperature ~ ~ conditions resulting from postulated loss-of-coolant (LOCA) and steam-line break accidents. Furthermore, equipment with a post-accident safety function must be environmentally qualified for the resulting adverse pressure and temperature conditions. 2.0 CONTAINMENT FUNCTIONAL 0ESIGN Ginna is a 1520-MWt Westinghouse pressurized water reactor (PWR) which uses a dry cylindrical reinforced concrete type ccntainment. It is very similar to the San Onofre Unit 1 power plant also designed by Westinghouse. The reactor coolant system of Ginna consists of 2 loops, compared with 3 loops'or San Onofre 1. The engineered safety sytems provided include the containment air recizculation system containment spray system, and safety injection system.: The safety'njection system consists of two passive accunulators, three

high-pressure

pumps, and two low-pressure pumps.

In the event of loss of off-site power and failure of one diesel generator, miniaem safety injection is provided by two high-pressure pumps and one low-pressure

pump, and minimum containment heat removal is provided by one containment spray pump and two fan coolers.

2.l Review of Anal sis of-Ginna Containment Functional Oesi n For PWR plants the high-energy line break types that must be analyzed include primary system pipe breaks and secondary system pipe breaks. A break on the primary side generally results in the most severe pressure response in the containment, while a break on the secondary side results in the most severe temperature conditions in the containment. There are two sepa"ate calculations which comprise the containment analysis for a postulated pipe break. The first calculation includes the mass and energy release analysis which, for primary system pipe breaks (LOCAs), includes blowdown, reflood and post-ref lood phases. The results are mass and energy release rates into the containment. The second calculation is the containment response

analysis, which results in the containment temperature and pressure response to the mass and energy release from the postulated break.

The acceptance criteria used to evaluate the Ginna containment functional design anal'ysis are based'on the Standard Review Plan (SRP), NUREG-75/087. In order for the containment analysis to be found acceptable, both the mass and P energy release and the containment response calculations must meet the acceptance criteria specified in the SRP. 2.2 Primar S stem Pipe Break In the Ginna FSAR, the most severe primary system pipe. beak was (le identified as a double-ended cold-leg discharge'reak. For the postulated

break, the reflood phase, and hence the energy in the secondary

system, was

~ ~ not included in the analysis. This analysis, therefore, does not meet the acceptance criteria specified in the SRP. Since the analysis of mass and energy release rates is unacceptable so too is the containment response calculation based on the mass and energy release rate data. 2.3 Secondar S stem Pi e Break In the Ginna FSPR, the licensee's secondary system pipe-break analysis consisted of analyzing the reactor response to a steam-line break occurring at various locations inside and outside the containment. The analysis was performed to demonstrate that: (a) with a stuck rod and minimum engineered safety features, the core remains in place and essentially intact so as not to impair effective cooling of the core; and (b) with no stuck rod and all equipment operating at design capacity, insignificant cladding rupture occurs. This analysis was not intended to be used to evaluate the containment functional design calculation, and the results would not be appropriate for that purpose. Therefore, an acceptable secondary system pipe-break analysis has not been performed. 2.4 Regnal sis of Ginna Containment Functional Desi n As mentioned above in Section 2.2, Review of Analysis of Ginna Containment = j Functional Design, the containment response analysis for primary system pipe-breaks (LOCA analysis) does not satisfy current criteria, and a MS'nalysis suitable for evaluating the containment functional design has not l$ been performed. The secondary system pipe-break (MSLB) analysis generally is the most limiting case for temperature conditions inside the containment. The primary system pipe-break (LOCA) analysis generally results in the limiting peak pressure condition inside the containment. Both of these analyses were performed and are discussed below. 3.0 PRIHARY SYSTEH PIPE BREAK ANALYSIS For a primary system pipe break, three phases are involved in the calculation of mass and energy release

rates, namely the blowdown, reflood, and post-ref lood phases.

The mass and energy release ra'te calculations weze based on the guidelines of Standard Review Plan Section 6.2.1.3; in the calculations the carryout rate fraction during reflood was set equal to 0.80 at the bottom of the core. In general, the analysis was done in a manner that conservatively establishes the containment design pressure; i.e., maximizes the post-accident containment pressure. The worst break location was determined to be at the pump suction side of the cold leg, because of the consideration of energy input, from the steam generator in the affected loop during the reflood phase. 3.1 Initial and Boundar Conditions The initial mass of water in the reactor coolant system was based on the system volume calculated for the temperature and pressure conditions existing at 102% of full power (safeguards design rating) or 1550.4 HWt. The initial conditions within the containment and the reactor coolant system prior to accident initiation are given in Table l. for the containment, peak pressure

analysis, a double-ended guillotine break at the pump suction with loss of off-site power, was postulated.

In addition the loss of one diesel generator was assumed as the worst. single "~ rr y p active failure. This assueption of postulated break and single active failure t pically results in the maximum calculated containment internal pressure. y ca The components of the available safety injection and containment heat P removal systems, if off-site power and one diesel generator aze lost, are shown in Table 2. The containment heat sink data used in the analysis are described in Table 3. 3.2 Blowdown Phase Following a postulated rupture of the Reactor Coolant System (RCS), steam and water are released into the containment. Enitially, the water in the RCS is sub-cooled at a high pressure. When the break occurs, the water passes through the break where a portion flashes to steam at the low pressure in the containment. Break flow rates are calculated with the Moody critical flow model for saturated flow and the Henry-Fauske model for sub-cooled flow. A discharge coefficient of 1.0 was used. Reactor scram'was assumed to occur with loss of off-site power, at the initiation of the break. The recirculation pumps were tripped off and the steam generators were isolated at the time of the break. The containment back-pressure was conservatively assumed to be constant throughout the accident at 14.7 psia. The end of blowdown was defined as the time when the primary system pressure dropped below the containment design pressure of 74.7 psia. Natural convection heat transfer was used for the secondary coolant in the steam generator foz tube surfaces iwezsed in water. The mass and energy release rate was. calculated with the code RELAP4-MpD7. The RELAP4 input deck was obtained from the NRC. and carefully reviewed for code options and for initial and boundary conditions. The plant physical description was assumed to be correct..Additional information reayired to perform the analysis was obtained from information on the Ginna v 7 docket-and conversations with personnel of Rochester Gas and Electric Corp. The results of-the blowdown analysis are summarized in Tables 4 and 5. Table 4 provides a detailed energy balance prior to the accident and at the end of the blowdown phase, which occurred 14.7 seccnds after break initiation. The total energy released to the containment during blowdown was approximately 211.9 million Btu. 'able 5 provides mass and energy release rates from the blowdown phase for use in the containment response analysis. 3.3 Reflood Phase Following blowdown, the lower plenum below the reactor core is refilled by water from the safety injection system. This phase, known as refill, was conservatively omitted and reflood was assumed to begin immediately after blowdown. Initial conditions for the start of the reflood phase were based on the end-of-blowdown (EGB) results. At the start of reflood, 14.7 seconds ~ after break initiation, the water remaining in the reactor vessel was assumed to be saturated at the design pressure of 74.7 psia and at the level of the bottom of the active core. At 14.7 seconds, the core power level dropped to 100.11 MWt or approximately 6X of the initial power. The accumulator flows had been initiated on low cold-leg pressure of 700 psia, which occurred at about 7 seconds into blowdown. At the start of reflood the accumulator flows totaled 4550 ibm/s. For numerical stability of the RELAP4 computer-code, the Emergency Core Cooling System (ECCS) flow was set at the saturation temperature of 272.9 F. The reactor coolant pumps had coasted down and the rotors were locked. In the reflood phase, Safety Injection (SI) water enters into the downcomer.:As the downcomer is filled, a driving head across the vessel II i) ~ ~ y

forces water into the core. SI water entering the core is converted to steam, which entrains water into the hot legs at a high velocity. Water continues to enter the core and releases the stored energy of the fuel and cladding as the i mixture level in the coze increases. The carryout rate fraction (CRF), which is the mass ratio of liquid exiting the core to liquid entering the core, is 1 assumed to be at a constant value of O.BO throughout the reflood phase. The core is assumed to be quenched when the liquid level is 2 feet from the top of the core. The flow split between the broken and unbroken loop and any steam quenching was calculated by RELAP4-M007 using the homogeneous equilibrium model. The heat. tzansfer from the secondary coolant to the steam generator tubes was based on natural convection heat transfer for tube surfaces immersed in water. For tubes not, immersed in water, condensing heat transfer is assumed. Steam leaving the steam generator was conservatively assumed to be superheated to the temperature of the secondary 'coolant. The 'results of the reflood analysis are summarized in Tables 6 and 7. Table 6 provides a getailed energy balance at the end of blowdown just prior to reflood and at the end of the reflood phase, which ocurred 20.1 seconds after the start of reflood. Table 7 provides mass and energy release rates from the reflood phase needed for input into the containment response analysis. 3.4 Post-Ref lood Phase The post-ref lood phase consists of removing all remaining stored energy in the primary and secondary systems and accounting for decay heat. This is done I by conservatively assuming that all the energy in the secondary system and I primary heat structures is released in one hour after the'end of reflood. The .amount of energy in the secondary system and primary heat structures was I

calculated by assuming these structures would return to 212 F in one hour following reflood. This is conservative since the containment pressure will not return to 14.7 psia within one hour, and therefore the saturation temperature will be hotter than 212 F. The decay heat released over the one I Il g /g hour duration was based on the AN8 standard 'decay heat curve plus 20%. 3.5 Containment Res onse Calculation The containment spray systems and containment structures available for energy removal were mentioned earlier in Section 3.1, Initial and Houndary Conditions; they are given in Tables 1 and 2. The Tagami and Uchida heat transfer correlations were used for all structural heat sinks. The Tagami correlation was used until the end of blowdown or 14.7 seconds; thereafter the Uchida correlation was used. ~ ~ The containment response calculation was done using the CQNTEMPT-LT/028 computer code. The program uses a three-region containment model consisting of the containment atmosphere (vapor region), the sunup {liquid region), and the water in the reactor vessel. Mass and energy are transferred between the liquid and vapor regions by boiling, =condensation, or liquid dropout. Each region is homogeneous, but a temperature difference can exist between regions. The physical model was obtained from references 1, 2 and 3. 3.'6 Containmeht Res onse Results The containment pressure and temperature response was calculated by . assuming that the blowdown, reflood, and post-ref lood energy is released directly to the containment. This method is conservative since it does not take into account the energy that may be required to heat the water in the primary system to. saturation. In addition, it was also assumed that the vE4 P ~ ~ P

reflood and post-ref lood energy were released as superheated steam at the temperature of the secondary side (approximately 500 F). The results are shown in Figures 1 and 2. The calculated transient reflects a post-accident 0 containment pressure of 74 psia and a temperature of 282 F. The containment design pressure and temperature are 74 psia and 2$ F, respectively. There is, therefore, a slight margin between the peak calculated pressure and tem-perature and the design values. 4.0 SECONOARY SYSTEM PIPE BREAK ANALYSIS The containment response to a secondary system pipe break was also analyzed. For PWRs, the most limiting break is a main steam-line break with pure steam blowdown. The steam-line break accident was analyzed for various plant conditions from hot standby to 102X of full power. A detailed parametric study is required to determine the most limiting combination of consistent initial conditions and system operation modes. To circumvent an extensive parametric study, the most limiting set of conditions was considered. The postulated accidents analyzed were a double-ended guillotine break in 1 a main steam line at 102X of full power and the same break at hot standby. i In both of these cases, the mass and energy release rates were calculated assuning that off-site power was available. Since no liquid entrainment was assumed during steam generator

blowdown, a spectrum of break sizes was not analyzed.

In addition, three different single active failures were considered for the 102X of full power case, and two different single active failures were ccnsidered for the hot standby case. These were a main steam isolation valve (MSIV) and main feed isolation valve (HFIV) failure and loss" Fg(

of one train of containment heat removal systems for the 102K of full power case and NSIV fai lure and loss of one train of containment'eat-removal system for the hot standby case. Thus, five different containment response calculations-were performed to de'termine the most limiting pressure and temperature conditions resulting from a steam-line

break, (see Table 8).

The model and assumptions that were used in analyzing the main steam-line break are given in the following discussion. The blowdown mass and energy release rates were calculated using a four-volume RELAP4 model. One volume models the primary side of the affected steam generator and the other three volumes model the feedwater line, secondary side of the steam generator, and the steam line. A schematic of the four-volume model is shown in Figures 3 and 4. A description of the four-volume model follows. Steam Generator - On the primary side of the steam generator, steady state flow conditions are conservatively assumed throughout the blowdown for both the 102K of full power and hot standby cases. On the secondary side of the steam generator, the actual plant conditions representing 102K of full power and hot standby are used in each case. An infinite bubble rise velocity was assumed on the secondary

side, which precludes moisture carryover and ensures a pure steam blowdown.

One heat slab was used in the steam generator model to model the heat transfer between the primary and secondary sides. The heat transfer coefficient on the primary side was calculated by RELAP4; forced convection was assumed. On the secondary

side, nucleate boiling heat transfer jf

~ ~ was assumed. - The height of the heat slab used to model the steam-generator-tube'surface area was set to a small value to ensure that the tubes remained covered during the entire transient. <<22>>

Steam Line and Feedwater Line - The blowdown of the steam line and feedwater line was accounted for by a one-volume RELAP4 model for each line. The size of each volume was adjusted to account for the mass of steam or water in the line up to their respective isolatiorrvalves. Both lines have redundant isolation valves. 'The specific isolation valve considered depends on the single-failure assumption being used in the analysis. The blowdown of the unaffected steam generator through the connecting steam header before isolation valves close was conservatively modeled by assuming a constant back-pressure fillfor this line. Feedwater flow before isolation valve closure was modeled as a constant mass flow rate fill. The main feedwater . isolation valves are assumed to start closing 10.54 seconds after a steam-line break; these valves require 5 seconds to fully close. The main steam isolation valves are assumed to start closing at zero seconds after a steam-line break and require 5 seconds to fully close. Auxiliar Feedwater In ection - The auxiliary feedwater injection was assumed to be 200 gpm at 80 F for each case calculated. For breaks at 102X of full power, injection is assumed to start 30 seconds after the line-break occurs. At hot standby conditions, injection is assumed to start at the time of the break. 4.2 Initial Conditions and Other Assumptions The initial conditions for the three cases analyzed at 102X of full power and hot standby are suoearized in Figures 2 and 3. In all cases the sources .of energy include the following: The stored energy in the affected steam-generator vessel tubing.

The stored energy in the water contained within the affected steam generator. The stored energy in generator before the The stored energy in the isolation valves the feedwater transferred to the affected steam isolation'alves in the feedwater line close. the steam from the unaffected steam generator before in the unaffected steam generator close. The energy transferred from the primary coolant to the water in the affected steam generator during blowdown. The stored energy in the auxiliary feedwater transferred to the affected steam generator after auxiliary feedwater system initiation. In addition, the mass release rate was calculated with the Hoody model.. 4.3 Containment Res onse Calculation The containment for secondary system line breaks was modeled in a similar manner as for primary system blowdown as described in Section 3.5 with initial conditions as in Table l. One exception is that the Tagami heat transfer correlation was used with a peak time of 100 seconds for all cases analyzed. The containment engineered safety systems are described in Table 2. For cases 1, 2, and 4, (see Table 8) full capacity of the systems is assumed. For cases 3 and 5, (see Table 8) loss of one containment spray line is assumed and all the four fan coolers are assumed to remain in full capacity. In each case, the containment sprays are initiated at a containment pressure of 30 psig and reouire 35 seconds to come on line. 4.4 Steam Generator Blowdown and Containment Res onse Results Three different cases at 102% of full power and two cases at hot standby conditions were analvzed;.as described in Table 8. The blowdown mass and t C energy release rates for the five cases are tabulated in Tables'9 - 13. The resultant containment pressure and temperature response is given in Figures 5 - 14. As shown by Figures 13 and 14, case 5 results in the highest containment pressure, 85.8 psia at 91 seconds after steam-line break, with a containment temperature of 413 F at 34 seconds after steam-line break. Case 5 represents a hot standby plant ccnfiguration failure of one containment spray pump. As shown by Figure 10, case 3 results in the highest containment temperature, 421 F at 34~seconds after steam line break, with a correspond-ing containment peak pressure of 75 psia at 60 seconds after steam line break. Case 3 represents a 102X of full power plant configuration with failure of one containment spray pimp. The containment design conditions are 74.7 psia and 286+; thus, both values are exceeded as a result of a main steam-line break.

5.0 REFERENCES

l. Ginna Nuclear Power Plant Unit 1, "Updated Final Facility Description and Safety Analysis Report," Docket No's. 50-244-Al to 50-244-A4. 2. Exxon Nuclear Company, Inc., ECCS Analysis for the R. E. Ginna Reactor with ENC WREM-2 PNR Evaluation Model," XN-NF-77-58 dated December 1977. 3. Memo from S. Srown to C. Tinkler, "Ginna Containment Analysis Data," dated 20 September 1979.

Table l. 'Initial conditions for, containment analysis '13 Parameter Value Reactor coolant system Reactor power level (a) Mass of RCS Total. Liquid Energy(b) 1550.4 MWt 392 x 10 ibm 386 MBtu Containment Net free volume Pressure Temperature Relative humidity Refueling water temperature Outside air temperature Refueling water storage tank 972000 ft 14.7 psia 100 F 50K 80 F 100 F 230,000 gal. a. 102X of full power b. all energies are relative to 32 F

Table 2. Engineered safety systems Operating Assumptions for Containment Peak Pressure Analysis ~ System/Item Full Capacity Value Used for Peak Pressure Analyses Safety injection system Number of trains Number of Injection lines Number of pumps 'High-pressure pumps Low-pressure pumps Flowrate, gal/min/train Containment spray system Number of lines Number of refueling water pumps Flowrate, gal/min Recirculation system Number of lines Number of refueling water pumps Number of heat exchangers Type Design UA Btu/hr F Flowrates Recirculation side, gal/min Exterior side, gal/min Source of cooling water 2 4020 2 1 2160 1 2 2400 1 1 1200 2 Shell &, U-Tube 750,000 2 Shell k U-Tube 750,000 3120 1560 5560 2780 Component cooling . Component water cooling water ~ Il j <<27>>

e

Table 3. Containment Structural Heat Sinks. A. Material Pro erties Material Thermal Conductivity .(ELtu/hr ft 0) Yolumetric Heat Capacity (Btu/ft~ F) Steel Concrete Insulation B. Heat Sink Descri tions I. Insulated dome and wall Surface Area, ft2 Composition, ft Steel Concrete Insulation 2. Uninsulated dome and wall Surface Area, ft2 Compositicn, ft Steel Concrete 3. Sump walls Surface Area, ft2 Composition, ft Steel Concrete 30.0 0.8 0.02 36,181 0.03125 2.5 0.10417 12,474 0.0315 2.5 2,342 0.03125 5.0 54.0 30.0 1.0 4. Refueling c:avity inside wall and floor Surface Area, ft2 Composition, ft Steel Concrete 6,900 0.02083 2.5

Table 3. Containment Structural Heat Sinks (cont'd) 5. Outside refueling cavity wall and steam generator compartment Surface Area, ft2 Composition, ft Concrete 21,800 1.25 6. Intermediate level floor Surface Area, ft2 Composition, ft Concrete 7. Operating floor Surface Area, ft2 Composition, ft Concrete 8. Heavy steel beam and crane structure Surface Area, ft2 Composition, ft Steel 6,170 0.25 9)162 1.0 9)174 0.0625 Steel beam Surface Area, ft2 Composition, ft Steel 10. Cylindrical supports and beam Surface Area, ft2 Composition, ft Steel 5,016 0.04167 8,586 0.02088 ll. Crane support colons Surface Area, ft2 Composition, ft Steel 12. Grating and stairs Surface Area, ft2 Composition, ft Steel

5) 756 0.03?25 7,000 0.0052 Note:

Boundary condi.tions on all heat slabs are adiabatic on the inside and Tagami/Uchida on the outside.. 0

Table 4. Blowdown Energy Balance, 1 Ginna Double-Ended Guillotine Suction Leg Break Inventory 6 0.0 s 10 10m 10 8tu Inventory O 14.7 s 10 ibm 10 Btu Decrease l0 ibm 10 Btu Reactor coolant system Accumjlator system (a) Core stored energy (b) Primary sensible energy Decay heat 312.4 13.67 211.9 7.93 10.76 85.64 136.9 10.84 5.87 6.28 5.55 83.1 255.1 2.83 ~ 206.00 0.63 2.38 ~ 2453 4.768 The SI water temperature was 272.9 F to prevent numerical instabilities. Actual value should be 90 F. Based on ANS + 20K decay heat curve.

Table 5. Slowdown Mass and Energy Release Rates Ginna Double-Ended Guillotine Suction Leg Break Time (s) Mass (ibm/s) Energy (Btv/ibm) 0.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0 9.0 10.0 11.0 12.0 13.0 14.0 14.7 8.05 x 10 4.57 x 10 4 3.77 x 104 2.81 x 104 1.93 x 104 1.87 x 10 4 1.62 x 10 4 1.42 x 10 4 1.26 x 104 9.82 x 10 7.88 x 10 6.25 x 10 4.07 x 104 2.81 x 103 2.21 x 10 1.99 x 103 535. 8 540.8 557.7 576.3 637.5 601.6 610.9 601.3 606.9 603.1 566.4 558.5 530.7 512.4 476.6 458.0

Table 6. Reflood Energy Balance. Ginna Double-Ended Gui)lotine Suction Leg Break. Inventory 8 14.7 s 10 ibm 10 Btu Inventory 8 40 s ~ Decrease 10 ibm 10 Btu.'0 ibm 10 Btu I Reactor coolant system 43.9 30.3 39.7 28.6 4.2 1.7 Accumulator system 10.84 6.28 0.0 0.0 ~ t ~ 10.84 6.28 Core stored energy 18.8 6.17 12.63 Decay heat 1.96 , Steam generator (secondary side) 17.8 97.7 17.8 24.6 73.1

Table 7. Reflood mass and energy release rate. Ginna Double-Ended Guillotine Suction Leg Break. Time (s) 14.7 15.0 16.0 17.0 18.0 19.0 20.0 21.0 22.0 23.0 24.0 25.0 26.0 27.0 28.0 29.0 30.0 .31.0 7200. 7201. 1.0 x 10 Mass (ibm/s) 2.11 x 10 2.98 x 10 3.15 x 10 3.60 x 103 1.87 x 10 9.39 x 102 5.69 x 102 1.21 x 102 1.06 x 10 9.82 x 10 1 8.99 x 10 1 7.58 x 10 1 7.21 x 10 6.89 x 10 1 5.63 x 10 5.40 x 10 4.86 x 10 1.08 x 10 1.08 x 10 0.0-0.0 Energy (Btu/ibm) 1282 1282 1282 1282 1282 1282 1282 1282 1282 1282 1282 1282 1282 1282 1282 1282 1282 1282 1282 1282 1282

Table 8 Oescription of steam-line break cases Power level 102X of ull ower Hot standb RCS primary flow Steam generator Feed flow Mixture level Water volume Pressure 982 LBM/s 869 LBM/s 24.8 ft 1681 ft3 779 psia 962 LBM/s 0.0. LBM/s 41.8 ft 2821 ft3 102Q. psia Failure Of Steam-line Mass (LBH) Case 1 HSIV 2,462 Case 2 HFIV 2,412 Case 3 1 s ra line 2,412 Case 4 HSIV 2,462 Case 5 1 s ra line 2,412 Feedwater line Hass (LBM) Temp. (oF) 18,300 432 76,895 432 )8,300 432 18,300 432 18,300 432 Auxiliary feedwater Flow (GPM) Temp (OF) 200 6 30 s 80 200630 s 80 200 9 30 s 80 20090 s 80 200 8 0 s 80 Containment heat removal system Full capacity Full capacity One spray line h 4 fan coolers Full capacity One spray line h 4 fan coolers

Table 9. Main steam-line break mass and energy release rates - case l. Time (s) 0.0 0.1 0.2 0.3 0.4 0.5 1.0 2.0 3.0 5.0 7.0 9.0 10.0 15.0 20.0 25.0 30.0 35.0 50.0 70.0 100.0 100.1+ Mass .51bm/s) 8894 8894 8551 8250 8004 7812 7104 5595 4637 3388 2262 2019 1940 1747 1699 1625 1588 1578 991 8 8 0 Energy (Btu/ibm) 1199 1199 1197 1198 1197 1197 1199 1194 1189 1182 1201 1200 1199 1198 1197 1197 1196 1196 1188 1174 1174 0 +At this time, the steam generator has reached a dryout condition, and the steam generator and containment are in pressure equilibrium. The continued injection of auxiliary feedwater will result in oscillation in the blowdown flow. However, the mass release rate will be less than 28 lbs/sec (200 gpm water) and will not significantly influence the course of the accident since the containment pressure and temperature have already passed their peak and are rapidly decreasing. The analysis was, therefore terminated.

Table 10. Hain steam-line break mass and energy release rates - case 2. Time (s) 0.0 0.1 0.2 '.3 0.4 0.5 1.0 2.0 3.0 5.0 7.0 10.0 15.0 20.0 25.0 65.0 75.0 100.0 110.'0>> Hass ...(ibm/s) 8891 8891 8543 8245 8003 7M9 7103 5592 4632 3384 2261 1932 1735 16M 1661 1604 246 57 0 Energy {Btu/ibm) 1199 1199 1197 1197 1197 1197 1199 1194 1189 1182 1201 1199 1198 1197 1197 1196 1175 1174 1174 See footnote at the bottom of Table 9.

Table ll. Main steam-line break mass and energy release rates - case 3. Time (s) 0.0 0.1 0.2 0.3 0.4 0.5 1.0 2.0 3.0 5.0. 7.0 19.0 10.0 15.0 20.0 25.0 30.0 35.0 50.0 70.0 100.0 100.1% Mass .-. - (ibm/s) 8891 8891 8542 8244 8003 7810 7104 5592 4632 3384 2262 2019 1940 1747 1699 1625 1588 1578 911 8 8 0 (Btu/1tm) 1199 1199 1197 1197 1197 1197 1199 1194 1189 1182 1201 1200 1199 1198 1197 1197 1196 1196 1188 1174 1174 0 +See footnote at the bottom of Table 9 1<>>' ('gJ kg 37>>

Table 12. Main steam-line break mass and energy release rates - case 4. Time (s) 0.0 0.1 0.2 0.3 0.4 0.5 1.0 2.0 3.0 4.0 5.0 6.0 8,0 10.0 15.0 20.0 25.0 30.0 40.0 50.0 70.0 80.0 90.0 100.0 101.0 102.0 120.0 131.0 132.0>> Mass "':(ibm/s ) 11935 11825 11362 10973 10655 10400 9454 7373 6012 4963 4189 3245 2314 1956 1536 1370 1307 1269 1232 1214 1211 935 367 353 97 28 14 6 0 Energy (Btu/1bm) 1192 1191 1190 1190 1191 1191 1194 1190 1187 1183 1180 1182 1200 1199. 1196 1194 1193 1193 1192 1192 1191 1187 1176 1176 1174 1174 1174 1174 1174 +See footnote at the bottom of Table 9.

0

Table 13. Main steam-Line break mass and energy release rates - case 5. Time (s) 0.0 0.1 0.2 0.3 0.4 0.5 1.0 2.0 3.0 4.0 5.0 - 6.0 8.0 10.0 15.0 20.0 25.0 30.0 40.0 50.0 70.0 80.0 90.0 100.0 101.0 102.0 120.0 131.0 132.0* Mass ( ibm/s) 11935 11820 11355 10965 10649 10396 9454 7368 6005 4958 4183 3236 2297 1956 1536 1370 1307 1269 1232 1214 1211 935 367 353 97 28 14 6 0 Energy (Stu/ibm 1192 1191 1190 1190 1191 1191 1194 1190 1187 1183 1180 1182 1201 1199 1196 1194 1193 1193 1192 1192 1191 1187 1176 1176 1174 1174 1174 1174 1174

• See footnote at the bottom of Table 9

f ~ ~, 1 I

SQ. 9 7Q.Q BS. 8 l 8.8 I.QE-QI I.QE+88 I.QE+QI I.QE+92 I.QE~Q3 I.K+N I K+SS

time, s

Figure 1. Containment Atmosphere

Pressure, Ginna,0oubfe-Ended Suction Leg Break

368 9 MS. 9 159.9 1.8E"81 1.8E+88 1.SE+St 1.RE~82 f.SE+83 1.8E~84 1.SF+96

time, s

Figure 2. Containment Atmosphere Temperature, Ginna Double-Ended . Suction Leg Break Constant Containment Back Pressure Constant fill Mass flow rate, 982 LBH/s Temp., 603oF Steam Line Constant Mass in steam line 779 psia no HSIV failure, 2412 ibm g 514 F MSIV failure, 2462 ibm HSIV) Steam Generator Primary side 2250 psia 590oF Initial Conditions Secondary side 779 psia 514oF ft

water, 1681 Mixt., level, 24.8 Aux. feed, 200 gpm 8 30 s.,

Temp = 80oF Constant leak Mass flow rate, 982 ibm/s Feedwater line Mass in feedwater line: no HSIV failure, 18300 ibm HFIV failure, 76895 ibm HFIV~ water 869 ibm/sec 432oF Constant fill conditions Figure 3. Schematic of Analytical Model for Steam Line Break at 1025 of Full ~ ~ Power. f'~ t l

Constant Containment Back Pressure Constant fill Mass flow rate, 962 1bmfs Temp., 547oF Steam Line Mass in steam line no MSIV failure, 2412 ibm MSIV failure, 2462 LBM Constant 1020 psia HSEV 547oF Steam Primary side 2250 psia Enitial conditions Generator Secondary side 1020 psia 547oF ft3 water, 2821 Mixt., level, 41.8 ft Aux. feed, 200 GPM I 0 s., Temp. = 80oF Constant leak Mass flow rate, 962 ibm/s Feedwater line Mass in feedwater line: 18300 ibm Valve closed at hot standby ~MFIV Constant fill conditions Figure 4. Schematic of Analytical Model for Ste m-Line Break at, Hot Standby Conditions. I I 1" g" GINNR NSLB, t:RSE 1 102% PONER NITH NSIV F'RILURE Tent@ t5CCONOSl

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CI GINNR HSLB, CASE 1 102/ POWER HITH NSIV f RILURE a $J C7 Cg 8 f~~ ~ ~ CI QR CS Cl 8 10 10 T]VC lSECONOS) )0 10'

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GINNR HSLB, CASE 2 102% PONER WITH HF'IY FAILURE ~ i 10'1tK (SECONGS) 10'0'

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rl ID QINNR HSLB, CRSE 3 102% POHEB HATH CONT. 5PRRY F'RILURE n lD I P4 g n O Ul C ~l ge I Pl'5Ct)0'0 TltK (SECOhSS)

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GINNR HSLB, CASE 3 102/ PONER HITH CONT. SPRAY f RILURE ~ I TltK (SECOhGS)

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vl Joill CD CS .GINNR VSLB, CASE 4 HOT STWDB> NITH MSIV f.RILVRE: ~ I Cb IC 10 10 TltK tSECOMlS)

GI NNR HSLB, CRSE 5 HOT STRNDBY NITH CONT. SPRRY F RI LURE CI po C$ d )ON TltK tSECONDSl

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