ML14135A250

Forwards Safety Evaluation of SEP Topic VI-3, Containment Pressure & Heat Removal Capability & VI-2.D, Mass & Energy Release for Possible Pipe Break Inside Containment for Facility.Licensee Analysis Acceptable

ML14135A250
Person / Time
Site:	San Onofre
Issue date:	01/12/1982
From:	Crutchfield D Office of Nuclear Reactor Regulation
To:	Dietch R SOUTHERN CALIFORNIA EDISON CO.
References
TASK-06-02.D, TASK-06-03, TASK-6-2.D, TASK-6-3, TASK-RR LSO5-81-01-030, LSO5-81-1-30, TAC-42516, NUDOCS 8201210005
Download: ML14135A250 (32)

Text

January 12, 1982 Docket No. 50-206 LS05 01-030 Mr. R. Dietch, Vice President Nuclear Engineering and Operations Southern California Edison Company 2244 Walnut Grove Avenue Post Office Box 800 Rosemead, California 91770

SUBJECT:

SYSTEMATIC EVALUATION PROGRAM (SEP) FOR THE SAN ONOFRE NUCLEAR POWER PLANT UNIT 1 - ERALUATION 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. 50-206, with the criteria currently used by the regulatory staff for licensing new facilities.

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 6r if NRC criteria relating to this subject are modified before the integrated assessment is completed.

Sincerely, r7:-

Cc tes Dennis M. Crutchfield, Chief A

Operating Reactors Branch No. 5 8201 210065 Division of Licensing 0

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Enclosure:

Draft SEP Topics VI-2.D and VI-3 cc wtenclosure:

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SAFETY EVALUATION REPORT ON CONTAINMENT PRESSURE AND HEAT REMOVAL CAPABILITY SEP TOPIC VI-3 MASS AND ENERGY RELEASE FOR POSSIBLE PIPE BREAK INSIDE CONTAINMENT, SEP TOPIC VI-2.D FOR THE SAN ONOFRE NUCLEAR GENERATING STATION, UNIT 1 DOCKET NO.

50-206

CONTENTS I.

Introduction ILI Review Criteria III.

Related Safety Topics IV.

Review Guidelines V.

Evaluation.

VI.

Conclusions VII.

References

I.

Introduction The San Onofre Nuclear Generating Station, Unit 1 began commercial operations in 1970.

Since then the staff's safety review criteria have changed. As part of the Systematic Evaluation Program (SEP), the mass.

and energy release data for postulated pipe breaks inside containment (Topic VI-2.D), and the containment functional design capability (Topic VI-3) have been re-evaluated.

The purpose of this evaluation is to document the deviations from current safety criteria as they relate to the containment functional design and the mass/energy release methodology for postulated pipe breaks inside containment. The significance of the identified deviations, and recommended corrective measures to improve safety, will be the subject of a subsequent, integrated assessment of the San Onofre, Unit 1 plant.

II. Review Criteria The review criteria used in the current evaluation of SEP Topics VI-2.D and VI-3 for the San Onofre Unit 1 plant are contained in the following documents:

1) 10 CFR Part 50, Appendix A, General Design Criteria for Nuclear Power Plants:

a)'

GDC 16 -

Containment design; b)

GDC 38 - Containment heat removal; and c) GDC 50 -

Containment design basis.

2) 10 CFR Section 50.46, "Acceptance Criteria for Emergency Core Cooling System for Light Water Nuclear Power Reactors."

3) 10 CFR Part 50, Appendix K, "ECCS Evaluation.Models".

4)

NUREG 75/087, Standard Review Plan for the Review of Safety Analysis.

Reports for Nuclear Power Plants (SRP 6.2.1, Containment Functional Design).

III. Related Safety Topics The review areas identified below are not addressed in this report,.but are related to the subject SEP topics.

1. III-1, Classification of Structures, Components and Systems (Seismic and Quality)

2. III-7B, Design Codes, Design Criteria, Load Combinations, and Reactor Cavity Design Criteria

3. VI-7.B, ESF Switchover from Injection to Recirculation Mode (Automatic ECCS Realignment)

4. IX-3, Station Service and Cooling Water Systems

5. X, Auxiliary Feedwater System

6. USI-A24, Qualification of Class IE Safety Related Equipment IV. Review Guidelines General Design Criterion (GDC) 16 of Appendix A to 10 CFR Part 50.

requires that a reactor containment and associated systems shall be provided to establish an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment and to assure that.the containment design conditions important to safety are not exceeded for as long as the postulated accident conditions require. GDC 38 requires a containment heat removal system be provided whose system safety function shall be to reduce the containment pressure and temperature following any loss-of-coolant accident (LOCA) and maintain A

them at acceptably low levels; furthermore, the system safety function shall be achievable assuming a'single failure. GDC 50 requires that the containment structure and the containment heat removal system shallbe designed so that the structure can accommodate,.with sufficient margin, the calculated pressure and temperature conditions resulting from any LOCA. This margin as obtained from the conservative calculation of mass/energy release and the containment model is discussed in the Standard Review Plan (SP).Section 6.2.1, Containment Functional Design.

The containment design basis includes the effects of stored and generated energy in the accident. Calculations of the energy.available for.release should be done in accordance with the requirements of 10 CFR Part 50, Section 50.46 and Appendix K, paragraph I.A, and the conservatism as specified in SRP 6.2.1.3.

The mass and energy release.to the containment from a LOCA should be considered in terms of blowdown,.reflood, and post reflood. The mass and energy release for postulated secondary system pipe ruptures should be calculated in accordance with SRP 6.2.1.4. The review also includes the analysis of postulated single active failures of components in the secondary system.

V.

Evaluation San Onofre 1 is a 1347 MWt Westinghouse pressurized water reactor (PWR) nuclear power plant. The containment consists of a 140 foot diameter; steel containment sphere surrounded by a cylindrical concrete building.

The containment houses the reactor pressure vessel, the reactor coolant piping, the steam generators, and containment heat removal systems. The reactor coolant system consists of three closed loops connected in parallel to the reactor vessel. Each loop contains a steam generator, a circulating pump, loop piping and instrumentation. A pressurizer is connected to the hot -leg of one of the loops.

Containment Peak-Pressure Analysis-Due to a LOCA In the event of a postulated loss-of-coolant accident (LOCA),

the release of coolant from the pipe break will cause the high temperature and

. pressure fluid to flash to steam.

This release of mass and energy raises the temperature and pressure of the containment atmosphere.

The licensee has submitted an analysis of the containment response due to a LOCA (Reference 1).

The mass.and energy release data in Reference 1 were developed by Westinghouse in accordance with the methodology of WCAP-8312A (Reference 2), which has been found acceptable by the staff for LOCA calculations in Reference 3.

The calculational model for the LOCA analysis is divided into five phases, which are:

o Blowdown - This phase includes the period from initial break occurrence until zero break flow is first calculated;

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o Refill - Following blowdown, the lower plenum of the reactor vessel is refilled with water.

The calculations conservatively assume the lower plenum is instantaneously filled with saturated water at the end of blowdown; therefore, this phase is neglected.

o Reflood - In this phase the reactor core is being reflooded by safety injection system water. Water entering the core is converted to steam which also entrains water.

Water continues to enter the core and release stored energy until the water level is two feet from the core top. The core is then considered quenched, leaving only decay heat to generate steam; o

Post Reflood - After reflood, energy remaining in the steam generator secondary is removed, along with wall heat sources and decay heat, by boiling off a portion of the two phase flow mixture passing through the primary system. Energy release is terminated when the steam generators come into equilibrium with the containment pressure; o

Long Term Phase - In this-phase, reactor system water boils at the containment pressure. Energy sources during this-phase include decay heat generation, and energy from residual thick metal and steam generator cooldown.

The above approach accounts for all sources of energy as specified in SRP 6.2.1.3 and is considered acceptable for use in calculations of the peak containment. pressure.

The calculations considered various locations of the break and its effect on.the reflood transient. For a cold leg break (between reactor coolant pump and vessel), all of the fluid which leaves the core must vent through a steam generator and becomes superheated. However, relative to breaks at other locations, the core flooding rate (and therefore the rate

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of fluid leaving the core) is low because the core vent paths include the resistance of.the reactor coolant pump. For a hot leg -pipe break the vent path resistance is-relatively low, which results in a high core flooding rate, but the majority of.the fluid which exits the core bypasses the steam generators in venting to the containment. The pump suction break combines the effects of the relatively high. core flooding rate,-as in the hot.leg break, and steam generator heat addition as in the cold leg-break. As a result, the pump suction breaks yield the highest energy flow rates during the post blowdown period. Because of the phenomena of reflood as discussed above, the pump suction break location is the limiting case., For the same break.location, as the break area increases, the peak containment pressure increases. Thus the analyzed double-ended guillotine break in the cold leg pump suction side is considered acceptable from the standpoint of producing peak containment.pressure.

The calculations assumed an initial reactor power level.of 102% of full power or 1374 MWt. For the safety injection system, maximum safety injection flow was assumed which is conservative for calculating containment peak pressures.

Nucleate boiling was assumed for core heat transfer, and liquid entrainment during the reflood phase was based on Athe FLECHT experimental results,(Reference 4).

Critical flow calculations are made with the Moody correlation for saturated and two phase fluid conditions, and the Zaloudek correlation for the subcooled blowdown regime.

The SATAN-V and LOCTA.computer codes were used to analyze the blowdown phase,.and the W-REFLOOD code was used to analyze the reflood phase.

These codes have been found acceptable for calculating mass and energy release to the containment during a LOCA (Reference 3).

Assumptions and initial conditions for calculations of mass and energy release data are given in Table 1.

The containment response analysis is based upon ;the existing plant design. A conservative prediction of LOCA consequences was assured by determining appropriate bounding values of'containment initial conditions, geometric parameters and thermodynamic properties and applying these values in a manner producing maximum pressure results.

Table 2 gives the initial conditions prior to accident initiation.

For. the peak containment pressure analysis, the containment heat removal systems (containment spray system and recirculating system) were assumed to be affected by the most restrictive single active failure,.which-has been determined to be. the loss of onespray sytem pump..

Containment fan coolers are assumed to not be in operation since turbine plant cooling water is required and no credit is taken for its availability.

The spray headers are assumed to be filled after 44.1 seconds; this is conservative in that it includes a delay time based on a loss of offsite power and diesel generator loading which exceeds the delay time with offsite power available.

Table 3 gives the heat removal system operating assumptions.

The containment heat sinks, including their thermal properties, and heat transfer coefficients used in the analysis are given in Table 4. Node spacing in each material is fine enough to ensure an accurate representation of the thermal gradient in each slab.

The containment analysis was performed in. accordance with the Bechtel Topical Report BN-TOP-3 (Reference 5);

the Bechtel COPATTA computer code derived from the CONTEMPT program used by the-staff, was used to calculate the containment response. The COPATTA code has been used by applicants and licensees in the past, and based on numerous staff confirmitory analyses, we find it acceptable to use the COPATTA code.

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The CPATTA program uses a three region containment model consisting of:

the containment atmosphere (vapor region),. the sump (liquid region), and the water contained in the reactor vessel. Mass and energy are.

transferred between the liquid and-vapor regions by boiling, condensation, or liquid dropout. Evaporation is not considered. A convective heat transfer coefficient can be specified between the sump liquid and atmosphere vapor regions. However, since any heat transfer in this mode is-small, a conservative coefficient of zero was assumed. Each region is assumed homogeneous, but a temperature difference can exist between regions. Any moisture condensed in the vapor region during a time increment is assumed to fall immediately into. the liquid region.

Noncondensible gases are included in the vapor region.

In summary, the design oasis LOCA is identified as a double-ended cold leg (pump suction) break at 102% of full power. Loss of offsite power was assumed and the limiting single active failure was the loss of a spray system pump. The peak containment pressure and temperature of 49.4 psig and 291u F do not exceed the design limits of 51 psig and 3000 F, and therefore are acceptable. The resultant post-LOCA.pressure and temperature profile are presented in Figures 3 and 4.

One exception to the SRP guidelines was found. It concerns the reduction of containment pressure. Containment pressure should be reduced to less than 50% of the peak calculated pressure within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the accident. The calculated pressure after 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> was 34 psig, or 76% of the peak pressure.

Containment Peak Pressure and Temperature Analysis Due to Secondary System Pipe Breaks In the event of a postulated main steam line break (MSLB), high.energy fluid is released to the containment,.causing an increase in the pressure and temperature of the containment atmosphere.

The utility has analyzed the peak containment pressure and temperature for a spectrum of postulated main steam line breaks (MSLB) at various power levels and for various single active'failures. These are presented in References 6 and 7. Main feedwater line breaks are not analyzed since such breaks result in a blowdown less limiting than the MSLB due.to lower fluid enthalpy.

As.shown in Figure 1, the three individual main steam lines (20" diameter) from the three steam generators feed into a common header (24" diameter) inside containment.

Flow measuring venturis (14.32" I.D.) are located in each of the 20" main steam lines between the steam generator nozzle and its connection to the 24" header.

The MSLB is assumed to.

occur anywhere in the 24" common header inside containment.

A double-ended break of this line has a total break area of 5.32 ft 2 which results in maximum blowdown from all three steam generators limited by their respective flow venturis with an effective break area of 1.12 2

ft per steam generator.

The analysis was performed assuming a variety of break locations and sizes for reactor power levels of hot standby, 20%, 25%, 72% and 103% of full power.

Break sizes.ranged from 0.25 ft 2 to 5.32 ft 2.

The most severe containment response was determined to correspond to a double ended break in the 24'" steam header.

The spectrum of break sizes and locations and reactor power levels investigated is considered acceptable in that they result in identifying the highest containment pressure response.

The MSLB break-flow is assumed to be pure steam. Steam generator water mass is calculated for a water level corresponding to the programmed level plus 5% instrument error and 8% void-fraction uncertainty. The hot standby inventory is 69,710 lbs of water per steam generator and 43,500 lbs of water at 103% of full power. Heat transfer from the steam generator tubes to the secondary side water is assumed to occur in the nucleate boiling regime, thus conservatively maximizing containment pressure response...

Mass and energy releases to the containment were calculated using the Westinghouse MARVEL computer code, (Reference 8), which'has been used for previous San 01ofre 1 analyses of core response following a MSLB.

This code has been used by Westinghouse in the past, and we have found that acceptable results are obtained if no entrainmnt is assumed. Since the analysis assumes pure steam blowdown, i.e., no entrainment, the method of analysis is acceptable. The following sources of mass and energy release to the containment were considered:

o Initial steam generator inventory o

Steam piping inventory o

Feedwater system pumping and flashing o

Auxiliary feedwater flow

• o Primary system to secondary system heat transfer As shown in Figure 1, since there are no steam isolation valves inside the containment, the main steam system piping volume, between the steam generators and the main steam stop valves, discharges through the break 3

into containment. This volume (1965 ft ) includes branch lines 6" in diameter or above (relief and safety valve header, pressure equalizing crosstie, and reheater supply lines with extension to the condenser dump valves.)

The mass of steam is calculated based.on the density of dry.

steam at the steam generator pressure for the power level being evaluated and is assumed to precede the reverse flow blowdown from two of the three.

steam generators. At full power, the steam mass is 3052 lb, and at no load the mass is 3979 lb.

Auxiliary feedwater flow was assumed to be manually initiated 10 minutes after.the time of break with a flow rate-of 250 gpm.

Figure 2.shows the feedwater system configuration. Following a MSLB, feedwater. is continuously pumped into the steam generator until feedwater isolation valve closure. The feedwater isolation valve starts to close 3 seconds after a MSLB and is fully closed 8 seconds after a MSLB. Should this valve fail to close, the backup feedwater regulating valve will close, with complete closure occurring 10 seconds after a MSLB.

For the hot standby MSLB, the feedwater system is manually controlled with flow limited to 5% of full power.

The mass of water added to the steam generators-pribr to feedwater line isolation is based on feedwater pump flow characteristics as a function of steam generator pressure decay. This flow continues until the safety injection activation signal (SIAS) is received (3.0 seconds after the.

MSLB) with flow thereafter linearly ramped to zero in 5 seconds as the feedwater isolation valves are closing. At full power, main feedwater pumping adds 12760 lbs while at no load main feedwater pumping adds only, 633 lbs. These are total inputs, assumed to be equally divided among all three steam generators.

Following main feedwater isolation, the depressurization of the steam generators causes flashing of the water in the feedwater piping between the steam generator and the feedwater regulating valves (as shown in Figure 2 of this report). The available volume of feedwater piping is 246 ft and is equivalent to a mass of 13030 lbs of water at full power based on the feedwater temperature of 4170F.

This mass is added to the steam generators where it is heated by primary to secondary heat transfer.

At no load, the feed system is assumed to be cool (700F) and does not flash Upon steam generator depressurization.

The most limiting single active failure in the analysis is assumed to be loss of a containment cooling line (i.e., containment spray pump). The effect of feedwater isolation valve failure as discussed above was analyzed but because of backup isolation capability this failure is not as severe.

For the analysis, offsite power is assumed to be available since loss of offsite power would result in tripping of the reactor coolant pumps and main feedwater pumps. Each of these trips aids-in mitigating the effects of a MSLB by either reducing the fluid available to feed the blowdown or reducing the energy transferred from the. primary coolant system to the steam generator secondary side. Loss of offsite power would result in slight delays in initiating safety injection, containment spray and main feedwater isolation. However, backup feedwater isolation time is not affected and.spray initiation times

-assumed in-the analysis were conservatively set to those corresponding to loss of offsite power. The use of the above assumptions is acceptable in that the use of these assumptions results in peak containment pressure and temperatures.

The containment response analysis was performed by Bechtel using the COPATTA code as described in Reference 5. The method of analysis is the same as for the LOCA.case previously described and is acceptable.

Table 5 give results of the containment pressure response for the 103%

full power and hot standby cases.

Inspection of the break spectra calculated indicates that the steam-line break with manual auxiliary feedwater.flow represents the most severe secondary system pipe break.

The case of a hot standby steam-line break resulted in a peak containment pressure and temperature of 53.0 psig and 4040F. The 103% of full power steam-line break case resulted in a peak containment temperature

• 0 and pressure of 406 F and 50.0 psig.

The containment pressure profiles are presented in Figure 5.

The containment design pressure of 51 psig is exceeded by 2 psi for the hot-standby case. In Reference 6, the licensee stated that the containment Structural Integrity Test was conducted at a pressure of 53.4 psig. Therefore, the implications of exceeding the containment design pressure by 2 psi are not of great concern.

The-worst peak containment temperature for the MSLB accident -is 4060 F.

The parametric.studies, performed by the licenseee in Reference 6, showed, that a better estimate model resulted in a peak-containment temperature

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of 405cF.

The licensee provided the temperature profile for this case, which is presented in Figure 6; the profile may be used for equipment qualification assessments.

Manual initiation of auxiliary feedwater (AFN) flow was assumed i-n the foregoing discussion of the MSLB accident analysis.

The licensee, how ever, has also performed analyses assuming automatic AFW initiation, to reflect-the current design of the auxiliary feedwater system. For a spectrum of power levels and break areas, auxiliary feedwater was assumed to be added to the steam generators at a rate of 500 gpm coincident with the pipe break, and then reduced to 250 gpm by operator action after 10 minutes. Since the licensee has also installed a pump trip for runout protection of the motor-driven AFW pump, the addition of auxiliary feed water at 500 gpm for 10 minutes conservatively bounds the flow conditions that would actually occur in the event of a large MSLB.

For the case of a MSLB at 103% of full power with automatic AFW addition, the peak calculated containment pressure was 50.5 psig, which is slightly greater (0.5 psi) than that calculated for the case with manual AFW addition. For the no load (zero power) case with automatic AFW addition, the peak calculated containment pressure was 47.6-psig, which is 5.4 psi less than that calcu lated for the corresponding no-load case with manual AFW addition.

Further more, -the results also show that peak containment vapor temperatures are not appreciably affected by automatic AFW initiation.

An additional analysis was performed by the licensee to determine the sensi tivity of the containment response to auxiliary feedwater addition rates.

For the analysis (103% power), auxiliary feedwater was assumed to be added to the steam generators at a rate.of 1000 gpm for 90 seconds and 500 gpm up to 10 minutes, with the.flow then being reduced to 250 gpm by operator action.

The peak containment pressure was calculated to be only 1 psi greater.

As noted above, the licensee has installed an automatic trip of the motor driven AFW pump on low discharge pressure for runout protection. However, 12

since the turbine driven AFW pump would rapidly become ineffective under MSLB accident conditions, automatically tripping the motor driven AFW pump may jeopardize the ability to remove decay heat from the reactor coolant system. The implications of this are-discussed under SEP Safety Topic.XV-2. This matter will also be considered further in conjunction with the licensee's response to I&E Bulletin 80-04, Analysis of a PWR Main Steam Line Break with Continued Feedwater Addition.

VI.

Conclusion We have reviewed the licensee's analysis of mass and energy release and containment heat removal capability for LOCA and MSLB analyses.

For the LOCA analysis, the calculated peak pressure and temperature are within the design limits, and therefore-are acceptable. However, one exception to the SRP requirements is that containment pressure is not reduced to less than 50% of the peak calculated pressure within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the ac cident.

For the MSLB analysis, the calculated peak containment pressure exceeds the design pressure by 2 psi.

This is acceptable since the con tainment Structural Integrity Test was performed at a pressure above the calculated peak pressure.

The MSLB temperature profile in Figure 6, and the LOCA temperature profile in Figure 4 may be used to assess the envi ronmehtal qualification of class IE Safety-related electrical equipment (USI A-24).

As noted in the previous section, the current auxiliary feedwater system design requires further staff evaluation (which is already in progress in conjunction with other staff review efforts) to assure that system per formance requirements for core cooling and containment integrity consider ations are compatible.

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VII. References

1.

Letter from K. P. Baskin, Southern California Edison Company, to A. Schwencer, NRC, DOR, dated January 19, 1977.

2.

Westinghouse, "Westinghouse Mass and Energy Release for Containment Design," WCAP-8312A, Rev. 2, August 1975.

A

3.

Letter from D. B. Vassallo, U. S. Nuclear Regulatory Commission, to C. Eicheldinger, Westinghouse Electric Corp., dated 12 March 1975.

4.

F. F. Cadek, D. P. Dominicis, and R.,/H. Leyse, "PWR FLECHT:

(Full Length Emergency Cooling Heat Transfer) Final Report," WCAP-7665, Westinghouse Electric Corporation (April,1971).

5.

Bechtel Power Corporation, "Performance and Sizing of Dry Pressure Containments," BN-TP-3, Revision 4, dated October.1977.

6.

Letter from. K. P. Baskin, Southern California Edison Company, to D. M.

Crutchfield, NRC, DOR dated June 10, 1980.

7.

Letter from K. P. Baskin, Southern California Edison Company, to D. M. Crutchfield, NRC, DOR dated March 6, 1981.

8.

3. M. Geets, "MARVEL - A Digital Computer Code for Transient Analysis of a Multiloop PWR System," WCAP-7907 (non-proprietary), June 1972.

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Table 1 ASSUMPTIONS AND INITIAL CONDITIONS FOR CALCULATIONS OF MASS AND ENERGY RELEASE DATA

1. Assumptions The following items ensure that the core energy release is conservatively.

analyzed. for maximum containment pressure.

a. Maximum expected operating reactor coolant system (RCS) temperature (599.30F)

b. Allowance in RCS temperature for instrument error and dead band

(+40F)

c.

Margin in RCS volume (1.4%)

d..

Allowance in RCS volume for thermal expansion (l:6%)

e. Allowance for RCS calorimetric error (2%)

f. Conservatively modified coefficients of heat-transfer

g. Allowance in core stored energy for effect of fuel densification

h.

Margin in core stored energy (+15%)

2. Initial Conditions Core Power (102% of licensed power level) 1374 MWt Inlet Temperature 557.0 OF Outlet Temperature 603.3 OF.

Steam Pressure 785 psia Rod Array 14 x 14 Assumed Containment Reference Pressure 61.1 psia Pumped Injection (assumed) Maximum 44.9 Ft3/sec aS Table 2 INITIAL CONDITIONS FOR CONTAINMENT PEAK PRESSURE ANALYSIS Reactor Coolant System Reactor power level, Mwt(a) 1374 Mass of reactor coolant system, lbm 298060 Liquid plus steam energy, 106 (Btu)(b) 174.45 Containment Net free volume of sphere, ft3 1.21 x16 Pressure, lb/in2a 14.7 Temperature, OF 120 Relative humidity, %

100 Component cooling water temperature, OF 95 Refueling water temperature, OF 80 Outside air temperature, OF 100 Stored Water Refueling water storage tank, gal 240,000

a. 102% of licensed power level

b. All energies are relative to 32F Table 3 ENGINEERED SAFETY FEATURE SYSTEMS OPERATING ASSUMPTIONS.FOR CONTAINMENT PEAK PRESSURE ANALYSIS (Sheet 1 of 2)

Value Used for Peak Pressure System/Item Full Capacity Analyses Safety Injection System Number of Trains 2

2 Number of Injection Lines 3

3 Number of pumps Safety Injection Pumps 2

2 Feedwater Pumps 2

2 Charging Pumps 2

1 Flowrate, gal/min/train 10,000 (a) 10,000(a) 213 (b) (c) 2 1 3 (b) (c)

Containment Spray System Number of Lines 1

1 Number of Pumps 2

1 Refueling Water Pumps Flowrate, gal/min/pump 1080 1080 (a)(b) 500 (c)

Recirculation System Number of Lines 1

1 Number of pumps 2

1 Recirculation Pumps Number of Heat Exchangers 1

1 Type Shell and U-Tube Shell and U-Tube f2 Heat Transfer Area, ft 595 595 Overall heat transfer 335 335 Coefficient, Btu/h-ft -OF Table 3 ENGINEERED SAFETY FEATURE SYSTEMS PERATING ASSUMPTIONS FOR CONTAINMENT PEAK PRESSURE ANALYSIS (ceet 2 of 2)

Value Used for Peak Pressure System/Item Full Capacity Analyses Flowrates Recirculation Side, 1450 713 gal/min A

Exterior Side, gal/min 1000 1000 Source of Cooling Water Component Cooling Component Cooling Water Water Flow Begins, Sec 1250 1250 (a)

During safety injection (b)

Following safety injection prior to recirculation from sump (c)

During recirculation Table 4 CONTAINMENT STRLCTURAL HEAT SINKS A.

Material Properties THERMAL VOLUMETRIC HEAT CONDICTIVItY CAPACITY MATERIAL (BTU/HR FT OF (BTU/FT 3 OF Organic Pain 0.2 20 Carbon Steel 26 54 Concrete 0.8 30 Stainless Steel 10 54 Zinc Galvanizing 65 41 Aluminum 119 35.2 AI (a)

AIR 4.5 0.017 (a)

Air Between Containment Sphere and Enclosure Building; Only Radiative Heat Transfer has been Assumed; Radiative Heat-Transfer through the Air has been Modeled as a Pseudo-Thermal Conductivity with-a Rate of 1.5 BTLVHR FT2 OF through an Average 3-Foot Air Gap.

B.

Heat Sink Descriptions

1.

Containment Sphere (and Sphere Enclosure Building as Appropriate)

Geometry SPHERE Inside Radius, FT.

70.0 Surface-Area Factor 0.744 Compositioh, FT.

Organic Paint 0.000417 Carbon Steel 0.0854 Organic Paint 0.000417 AIR 3

Concrete 3

Boundary Conditions Inside Surface Exposed to Containment Atmosphere; Outside Surface Exposed to External Atmosphere.

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2. Concrete Below Sump Water Geometry SLAB Surface Area, FT2 4072 Composition, FT Organic Paint 0.00083 Concrete 4.76 Boundary Conditions Inside Surface Exposed to Sump Water; Outside Surface Assumed Insulated.

3. Concrete Above Sump Water Geometry SLAB Surface Area, *FT 2

34722 Composition, FT Organic Paint

.0.000417 Concrete 4.56 Organic Paint 0.000417 Boundary Conditions Both Sides Exposed to Containment Atmosphere.

4. Refueling Canal Walls Geometry SLAB Surface Area, FT2 4489 Composition, FT Stainless Steel 0.00696 Concrete 2.66 Organic Paint 0.000417 Boundary Conditions Both Sides Exposed to Containment Atmosphere.

5.

Thin Galvanized Steel Geometry SLAB Surface Area, FT2 15032 Composition, FT Zinc Galvanizing 0.000317 Carbon Steel 0.000853 Zinc Galvanizing 0.000317 Boundary Conditions Both Sides Exposed to Containment Atmosphere 20

6.

Galvanized Steel Gratings Geometry SLAB Surface Area, FT2 5390 Composition, FT Zinc Galvanizing 0.000317 Carbon Steel 0.02021 Zinc Galvanizing 0.000317 Boundary Conditions Both Sides Exposed to Containment Atmosphere.

7. Carbon Steel (Greater than 2 in. Thick)

Geometry SLAB Surface Area, FT2 1542 Composition, FT Organic Paint 0.000417 Carbon Steel 0.3172 Organic Paint 0.000417 Boundary Conditions Both Sides Exposed to Containment Atmosphere.

8.

Carbon Steel (1 to 2 in. Thick)

Geometry SLAB Surface Area, FT2 8657 Composition, FT Organic Paint 0.000417 Carbon Steel 0.1182 Organic Paint 0.000417 Boundary Conditions Both Sides Exposed to Containment Atmosphere

9.

Carbon Steel (Less than 1 in. Thick)

Geometry SLAB Surface Area, FT2 22154 Composition, FT Organic Paint 0.000417 Carbon Steel 0.04181 Organic Paint 0.000417 Boundary Conditions Both Sides Exposed to Containment Atmosphere.

- 1

10.

Aluminum (CRDM Cooler Pipes & Ihsulation Skin)

Geometry SLAB Surface Area, FT2 9460 Composition, FT Aluminum 0.0055 Boundary Conditions Outside Surface Exposed to Containment Atmosphere; Inside Surface Adiabatic.

11. Stainless Steel (Piping & Equipment)

Geometry SLAB Surface Area, FT2 1369 Composition, FT Stainless Steel 0.0240 Boundary Conditions Outside Exposed to Containment Atmosphere; Inside Surface Assumed Adiabatic.

C. Heat-Transfer Coefficients VALUE

1.

Containment Atmosphere to Heat Sink "Modified Tagami/Uchida" Surfaces for Condensation; 2 BTU/HR FT2. OF for Convection.

2.

Containment Atmosphere to Containment 0

Sump Water.

3.

Containment Sump Water to Concrete 0.4 BTU/HR-FT2 -oF Below Sump Water.

4.

Refueling Canal Liner Plate 100 BTU/HR-FT2.oF GAP Conductance.

5.

Containment Sphere Enclosure Building 2 BTU/HR-FT2-oF Wall to Outside Atmosphere.

22

Table 5 MAIN STEAM LINE BREAK - -CONTAINMENT RESPONSE SAN ONOFRE -

UNIT 1 Break:

24-in, double-ended guillotine rupture Break Area:

2 x 2.66 ft 2 Blowdown:

dry -steam; flow limited by 1.12 ft 2 area flow venturis in each steam generator outlet line.

Offsite Power:

available Safety Injection Signal at 2 psig Containment Spray Signal at 10 psig Auxiliary Feedwater on at 10 minutes Case 12 Mass/Energy Release Assumption Standard Standard Reactor Power (%)

103 0

Peak Containment Conditions Pressure (psig) 50.0 53.0 Time (Sec) 110 378 Vapor Temperature (OF) 406 404 Time (Sec) 32 31 Energy Integrals @ Peak Containment Pressure (106Btu)

Bfeak Flow 190.32 242.48 Passive Heat Sinks 35.79 70.44 Spray Heat Transfer 2.91

-10.76 23 -

SAFETY VALVES TYPICAL (5)

CONTAINMENT REllEATERS ATMOS DUMP COVElu0R VALVES (24A6---T VALVES 24" STOP 201 VALVE 20"11 STEAM GENERATORS HIGH PRESSURE TURBINE FLOW VENTURI CONDENSER DUmp VALVES FIGURE 1 MAIN STEAM SYSTEM SAN ONOFRE UNIT 1

CONTAINMENT TYPICAL OF TIREE FROM AUX FROM AUX FROH SIS TO SIS FW SYS j

SYS BYPASS FW REG VVALVE 101 BLOCK FW REG

-1 STEM CNEATORS F'Uf NPUMP 121 VALVE VALVE FROMM AUXN SUCTION DISCHARGE VALVE 1st POINT FW SYS VALVE FW PUMP FW HEATER 18" 10" SEE VALVING ABOVE 1

1211 10" SEE VALVING ABOVE FIGURE *2 MAIN FEEDWATER SYSTEM SAN OHOFRE UNIT 1

60 P

= 49.4 PSIG

@ 62 SECONDS 50 40 30 4J 4-J 20 10 0

1 1010 10 104 50 106 Time Following Break (Seconds)

Figure 3 LOCA CONTAINMENT PRESSURE DOUBLE ENDED PUMP SUCTION BREAK

300 (VAPOR) = 291 0 F

@47 SEC CONTAINMENT VAPOR 250 MAXIMUM SUM TEMP 250 0 F @

90000 SEC 200 r-CONTAINMENT StJMP WATER o

150 4-J 100 50 10 10 10 10 10 10 Time Following 11reak (Seconds)

Figure 4 LOCA CONTAINMENT TEMPERATURE

60 P

= 53,0 PSIG so S

@ 378 SEC 000 40 HOT STANDBY P.4:

30 (103% POWER)

S 20 o

IlILI III iII LLII 10 102 10 10 105 6

Figure 5 Time (Seconds)

MSLB CONTAINMENT PRESSURE

450 T

=405 0 F Q VAPOR MAX 832 SEC 400 rcA ko 0

250 (100% POWER) 200 150 I

lWl2LLII Ir II WLLL IJ WI I

IIW1 1

10 102 103 10 10 6

Figure 6 Time (Seconds)

MSLB CONTAINMENT TEMPERATURE