Text

Amend Application 187 to License NPF-10,consisting of Proposed Change Number 504,requesting NRC Approval to Repair Degraded But Operable Shutdown Cooling Sys Check Valves at Unit 2 in Normal Operation (Mode 1)
	ML20205T284
Person / Time
Site:	San Onofre
Issue date:	04/24/1999
From:	Nunn D; SOUTHERN CALIFORNIA EDISON CO.
To:	;
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ML20205T271	List: ML20205T268; ML20205T284;
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	NUDOCS 9904270210
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMISSION Application of SOUTHERN CALIFORNIA ) Docket No. 50-361 EDISON COMPANY, H R. for a Class 103 )

License to Acquire, Possess, and Use )

a Utilization Facility as Part of ) Amendment Application Unit No. 2 of the San Onofre Nuclear )

Generating Station ) No. 187 SOUTHERN CALIFORNIA EDISON COMPANY, H R . pursuant to 10 CFR 50.90, hereby submit Amendment Application No. 187.

This amendment application consists of Proposed Change Number (PCN)-504 to Facility Operating i

License No. NPF-10. PCN-504 is a request for NRC approval to repair degraded but operable shutdown cooling system check valves at San Onofre Unit 2 with the unit in normal operation.

This is a request for an emergency license amendment to support repair of the valves as soon as possible.

Subscribed on this day of kl , 1999.

Respectfully submitted, SOUTHERN CALIFORNIA EDISON COMPA0Y f

Dwight E. Nunn Vice President j State f California n 4 pers'oiilly' 'appea' red form og 7A)l(M1VV

( A/1f_ k 'L9 .

Ali A personally kn n to me to be the person Gliose naui Ti subier16ed io the within instrument and acknowledged to me that he executed the same in his authorized capacity, and tha by his signature on the instrument the person, or the entity upon behalf

/

A ch the person acted, executed the instrument. -

WITNESShy ha 11a official' seal. i MARIANE SANCHEZ h

/ A Commission #1196482 E Signat >

$ WN-N I son Diego County f 9904270210 990424 {

4 MyComntE@usOct142D2y PDR ADOCK 05000361 <--------------

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Description of Proposed Change Number 504 San Onofre Nuclear Generating Station, Unit 2 Summary At San Onofre Units 2 and 3, a common Shutdown Cooling (SDC) System suction line connects Reactor Coolant System (RCS) hot leg to the two Low Pressure Safety Injection (LPSI) pumps.

There are two manual isolation valves (MU015 and MU018; one for each train) between the SDC system suction header and each LPSI pump. As originally designed, these isolation valves were norma!!y closed to preclude the possibility of both LPSI pumps drawing suction from one source (either a Refueling Water Storage Tank [RWST) or the containment emergency sump recirculation line) on certain single failures. If both LPSI pumps did draw suction from one source, net positive suction head (NPSH) could be reduced below pump limits and cause both LPSI pumps to be inoperable. This design also required an operator to manually open the valves to initiate SDC.

In the early 1980s, modifications were made to allow initiating SDC remotely from the control room. The modifications included swing check valves MU200 and MU202 (one check valve upstream of each isolation valve) so MU015 and MU018 can remain open. During normal operation, the check valves are normally closed. Their safety function is to remain closed during the injection and recirculation phases of Emergency Core Cooling System (ECCS) operation, and to open to allow remote initiation of shutdown cooling. This design change was made to comply with Branch Technical Position RSB 5-1, " Design Requirements for the Residual Heat Removal System."

During the current Unit 3 refueling outage, these check valves (S3120lMU202 and S3120lMU200) at Unit 3 were inspected. Southern California Edison (SCE) discovered that the disc nut was missing but the nut staking pin was in place. It is suspected that use of an incorrect material (carbon steel instead of stainless steel) allowed the nut to corrode away in the normal boric acid environment of the valve. This discovery caused SCE to question the status of these same valves at Unit 2. (The Unit 2 valves have passed their periodic Inservice Testing (IST),

including the last IST performed during the Unit 2, Cycle 10 outage in January 1999).

On April 22,1999, Southern California Edison (SCE) radiographed S2120lMU202 (MU202) and S2120lMU200 (MU200) at Unit 2 and discovered the valves were similarly degraded as follows:

MU202 - Dise nut partially corroded, staking pin in-place. )

MU200 - Disc nut missing, staking pin in-place. I 1

Description of Proposed Change Number 504 San Onofre Nuclear Generating Station, Unit 2 An operability assessment, based on the radiograph and information from the physical examination of the as-found condition of the Unit 3 valves, determined that both Unit 2 valves are operable in the as-found condition. However, a restriction was imposed such that SDC via LPSI beyond 4 days of continuous operation requires either: 1) verification that the valve internals are in proper alignment and the retaining pin is intact, or 2) at least one Containment Spray pump is available to support SDC as a backup system. Therefore, SCE desires to restore these valves to a condition equivalent to the original design as soon as possible, rather than waiting until the next refueling outage (currently scheduled for Spring 2001). To complete this repair, with Unit 2 in Mode 1 operation, SCE will need to close MU015 and MU018 to isolate the check valves and drain the isolated section of the shutdown cooling line. This action will place Unit 2 outside the licensing basis of the SDC system. SCE evaluated various repair options, including the insights provided by the plant's probabilistic safety assessment, and concluded repairing these valves now, and in Mode 1 operation is the most prudent course of action. (SCE estimates repairs will take between 30 and 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months .)

A. Background 1

At San Onofire Units 2 and 3, a common Shutdown Cooling System (SDC) suction line connects one of the RCS hot legs to the two LPSI pumps. There are two manual isolation valves (MU015 and MU018; one for each train) between the SDC system suction header and each LPSI pump. As originally designed, these isolation valves were normally closed to preclude the possibility of both LPSI pumps drawing suction from one source (RWST or the containment emergency sump recirculation line) on certain single failures.

If both LPSI pumps did draw suction from one source, net positive suction head (NPSH) could be reduced below pump limits and cause both LPSI pumps to be inoperable. This design also required an operator to manually open the valves to initiate SDC. Figure 1 provides a simplified diagram of the involved systems.

NRC Branch Technical Position RSB 5-1, " Design Requirements for the Residual Heat Removal System," requires, among other things, that all operations required to transition from the safety injection alignment to SDC be capable of being performed from the Control Room, including considerations of a single failure, except where approved by the NRC, This requirement is reiterated in Updated Final Safety Analysis Report (UFSAR)

Sec. 5.4.7.1.2, Design Criteria for SDC, and is part of the SDC system licensing basis. To meet this requirement, modifications were implemented in the early 1980s whereby two swing check valves, MU200 and MU202, and other components, were added to the SDC !

system. (The NRC did approve manual operator action to restore / remove power from electrical components in the 50' safety related switchgear rooms.) The check valves fulfill the isolation function of MU015 and MU018; therefore, MU015 and MU018 can remain 2 ,

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. Description of Proposed Change Number 504 San Onofre Nuclear Generating Station, Unit 2 normally open, allowing SDC to be initiated remotely from the control room. During normal operation, the check valves are closed. Their safety function is to remain closed ',

during the injection and recirculation phases of ECCS operation. Upon SDC initiation, the check valves open, establishing the SDC suction flow path to the LPSI pumps.

On March 27,1999, during the current Unit 3 refueling outage, Unit 3 swing check valves S31201MU200 (3MU200) and S31201MU202 (3MU202) had indeterminate results from their leak rate tests. Valves 3MU200 and 3MU202 were radiographed to determine the condition of the check valves. SCE discovered that the disc nut was missing from each of q these valves, but the nut staking pin was still in place (see Figure 2). It is suspected that use of an incorrect material (carbon steel instead of stainless steel) allowed the nut to corrode away in the normal boric acid environment of the valve. However, it was determined the r"issing nut was not the likely cause of the small valve leakage, rather the test results were de to test boundary valve leakage. Both valves were determined to have been operabli iin the "as found" condition. The Unit 3 valves have been restored to their original confguration.

This discovery caused SCE to question the status of these same valves at Unit 2. An initial operability assessment (OA) was performed for the Unit 2 valves. The OA concluded there was a reasonable expectation that the Unit 2 check valves were unaffected, i.e., had the proper stainless steel components to perform their safety functions. To confirm the assumption, the valves were radiographed on April 22,1999.

The radiographs showed that the Unit 2 check valves were degraded as follows:

MU202 - Disc nut partially corroded, staking pin in-place.

MU200 - Disc nut missing, staking pin in-place.

Based on the radiograph results and information gained from the physical examination of the similar valves in Unit 3, the check valves were determined to be operable. However, a restriction was imposed such that SDC via LPSI beyond 4 days of continuous operation requires either: 1) verification that the valve internals are in proper alignment and the retaining pin is intact, or 2) at least one Containment Spray pump is available to support SDC as a backup system. ,

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. Description of Proposed Change Number 504 l San Onofre Nuclear Generating Station, Unit 2 B. Proposed Activity l The proposed activity is to repair (restore to a condition equivalent to the original design)

M~U200 and MU202 in Mode 1. (The plans outlined here are based on the best available information. Once into the activity, the plan may be modified as new additional information becomes available.) To perform the repairs, SCE will isolate the affected check valves by closing MU015 and MU018. The SDC line between MU015/MU018 and the normally closed containment isolation valves (fW-9379 and HV-9336) will be drained.

This portion of the repair evolution is estimated to require 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months . After the line is drained, the current plan is to:

1. Remove valve bonnets;

2. Remove staking pins;

3. Clean valve disc threads and replace missing / degraded disc nuts);

4. Tack weld the new nuts to the disc stem (but not replace the staking pin ),

5. Re-install valve bonnets;

6. Re-fill and vent the SDC line - End of repair.

Steps 1 through 5 are estimated to take about 9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months to complete. Step 6 should take about 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . After the repair is complete MU015 and MU018 will be reopened. SCE therefore expects the total repair time to be between 30 and 40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months .

During the time MU015 and MU018 are closed, both LPSI trains will be operable for safety injection. To transition from LPSI injection to SDC, or to establish SDC during a plant shutdown, a contingency plan has been established as part of the repair package which allows " backing out" of the repairs and restoring SDC within approximately 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (thus allowing the plant to continue to meet the requirement for SDC availability, even assuming the Condensate Storage Tanks (CSTs) are at their minimum required level).

C. Discussion

, SCE has concluded that repair of these valves now in Mode 1 operation is the most l prudent course of action. The following sections discuss the availability of backup equipment, compensatory measures including limited administrative controls, the safety function and events protected against, and an assessment of the risk associated with the repair.

a. Availability of redundant or backup equipment.

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Description of Proposed Change Number 504 San Onofre Nuclear Generating Station, Unit 2 The proposed activity temporarily removes SDC from immediate availability for service. During Mode 1, heat is removed from the RCS by the steam generators i and main feedwater. Backup heat removal capability is provided by the auxiliary feedwater system and the CSTs. During a non-emergency plant shutdown, auxiliary feedwater and the main condensate system can cool the RCS from full power to SDC entry conditions and maintain those conditions indefinitely.

For emergency conditions with off-site power unavailable, auxiliary feedwater would be provided from the CSTs, T-120 and T-121. These two tanks provide sufficient inventory (over 500,000 gallons) to cool the plant from full power to SDC entry conditions and maintain those conditions for a minimum of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . l Compensatory measures, discussed below, will increase T-120's useful inventory !

to provide at least an additional 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months of steaming at SDC entry conditions.

In addition, for non-seismic events, the CSTs can be replenished (refilled) from the three demineralized water storage tanks, with a capacity of 535,000 gallons each.

Another alternative would be filling the CSTs with demineralized water supplied from an outside source. A flanged fill connection in the piping from the Fire Protection System to the condensate storage system would allow the outside demineralized water supply to be pumped into T-120 or T-121. Additionally, cross feeding from Unit 3's condensate storage and transfer system can be ,

performed by realignment of the condensate transfer pump suction and discharge I headers. Unit 3 is currently in a refueling outage and would be unaffected by this action.

The use of the Fire Protection System to supply water to the CSTs could be performed under emergency conditions. The use of non-demineralized water can cause a corrosion and scaling problem. However, if the safety of the plant is in question, the use of Fire Protection Water would be permitted.

1 When RCS pressure is less than 200 psi, the Containment Spray pumps can be aligned to perform the SDC function. This evolution would only require restoring the pressure boundary integrity of MU200 and MU202. Thus, even if MU200 and MU202 were completely blocked, the SDC function would still be available.

The proposed activity does not adversely impact any other system required or credited to prevent or mitigate accidents or transients.

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Description of Proposed Change Number 504-San Onofre Nuclear Generating Station, Unit 2

b. Compensatory measures including limited administrative controls.

Because of the degraded condition of the internal components ofMU200 and j MU202, the valves will be restored to a condition equivalent to their original design. With MU200 and MU202 inoperable due to the repair evolution, the ,

following compensatory actions will be taken .

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1) Even with MUO15 and MU018 closed during the repairs, LPSI will bc operable for safety injection. The repair plan allows " backing out" of the

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j repairs and restoring SDC within approximately 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months if determined l necessary by plant operators or management. The contingency plan includes provisions for restoring the SDC path with MU200 and MU202 inoperable by restoring the valves' pressure boundary integrity.

2) A temporary instruction associated with Operating Instruction SO23-3-2.7.2,

" Safety Injection Removal / Return To Service," will be in place to provide {

guidance to operators to perform the required actions to restore SDC if

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required. Operators will be made aware of the temporary operating j instructions during "tailboards" for the planned repair work. )

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3) Work activities will be controlled to minimize high risk activities during the i repair period.

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4) The available volume in T-120 will be increased by isolating (or staging an operator to isolate) non-seismically qualified connections to the CST when the j valve repair is initiated. This will preclude the loss of about 80,000 gallons of j water assumed to occur following postulated seismic events. This volume of 1 water is sufficient to steam at SDC entry conditions for about 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . j

c. Safety function and events for which protection is required.

SDC provides normal long term cooling of the RCS in Modes 5 and 6. In l addition, SDC is used for long term cooling following small break Loss Of Coolant Accidents (LOCAs), Steam Generator Tube Rupture (SGTR), and seismic events. l The impacts of the proposed change relative to these events sie discussed in Section D, below.

di Probability of needing the safety function.

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i Description of Proposed Change Number 504 San Onofre Nuclear Generating Station, Unit 2 The expected duration of the repair is approximately 31 hours3.587963e-4 days 0.00861 hours 5.125661e-5 weeks 1.17955e-5 months . As described in Section C.c, above, the shutdown cooling safety function is required in the following events: small break LOCA, SGTR, and seismic events. The probability {

of any one of these initiating events occurring during this repair duration is estimated to be 5 E-5.

As discussed in Section C.b., above, the proposed activity includes a contingency ,

plan to restore SDC to operable within approximately 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , if required. '

e. Probabilistic Risk Assessment (PRA) results that determine how operating the facility in the manner proposed in this amendment application will impact the core damage frequency.

SCE completed a PRA of the proposed repair plan. The assessment included all events requiring the shutdown cooling function to mitigate core damage and large early release: sma9 break LOCAs, SGTRs, and seismic events. The increase in core damage a:.d large early release risk are estimated to be 7.lE-6 and 1.7E-7, respectively.

The dominant contributor to core damage risk during the repair is from a seismic )

event of a magnitude greater than 0.3g pga. A seismic event of this magnitude or greater is assumed to fail the condensate makeup function to the condensate storage tanks. In this case, the condensate storage tank inventory limits the time I available for restoring shutdown cooling to service. The compensatory measures such as use of firewater to replenish the condensate tanks described in Section C.b, l above, are not credited in the risk assessment. The dominant contributor to the I large early release risk during the repair is from a steam generator tube rupture event assuming unsuccessful depressurization of the reactor coolant system prior to refueling water storage tank inventory depletion.

Other repair options, such as performing the repair in Mode 4 (decay heat removal via steaming at reduced reactor coolant system pressure) and in a defueled condition during the next refueling outage, were considered. The risk of repairing the valves in Mode 4 is on the same order of magnitude as repair in Mode 1. Long term plant operation without repairing the valves until the next refueling outage was considered undesirable due to the degraded condition of the valves and the desire to do the repair in a planned and controlled manner, rather than attempt recovery actions in the unlikely event of an event requiring SDC.

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Description of Proposed Change Number 504 San Onofre Nuclear Generating Station, Unit 2 Based upon the PRA results and planned contingency measures (not considered explicitly in the PRA), the overall risk of the repair plan is small. Based upon Regulatory Guide 1.174, these increases in risk are also characterized as "small."

While the degraded check valves are operable, the proposed activity will render the che ,

valves and SDC inoperable during repairs. There is no Technical Specification in Mode 2, or 3 which governs or requires SDC to be operable, or MU015 or MU018 to be op Neither is SDC considered a necessary support system for any System Structure, or component (SSC) required to be operable in Modes 1,2, or 3, and Limiting Condit Operation (LCO) 3.0.6 would not apply. Below is a brief discussion of potentially applicable Technical Specifications.

a.

LCO 3.4.6, RCS Loops-MODE 4, requires two loops or trains consisting of any combination of RCS loops and SDC trains to be operable and at least one loop or train shall be in operation. The proposed activity does not impact the operability of the RCS loops.

Although the RCS loops would not be impacted, LCO 3.4.6, Condition A would be entered when one RCS loop is inoperable and both SDC trains are inoperable.

The Required Action is to immediately initiate action to restore a second loop or train to operable status. This Condition's Required Action and Completion Time can be met during the proposed activity because of the contingency plan which allows for immediately taking the action to restore SDC to operable within approximately 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Condition C is entered when the required RCS loop (s) or SDC train (s) are inoperable or no RCS loop or SDC train is in operation. The Required Action is to immediately suspend all operations involving reduction of RCS boron concentration and initiate actions to restore one loop or train to operable status and operating. The proposed activity does not adversely affect the ability to meet the Required Action and Completion Time.

If Condition C is exited by restoring one train of SDC, Condition B is entered which requires being in Mode 5 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The proposed activity does not adversely affect meeting this Required Action and Completion Time.

b.

LCO 3.5.2, ECCS - Operating, requires two ECCS trains to be operable in Modes 1 and 2, and 3, with pressurizer pressure 2400 psia. A train of ECCS is dermed as one train each of High Pressure Safety Injection (HPSI), LPSI and Charging. The 8 l

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. Description of Proposed Change Number 504 San Onofre Nuclear Generating Station, Unit 2 proposed activity does not impact either HPSI or Charging. For LPSI to be considered operable, the suctions of the two LPSI pumps must be isolated from each other. This is normally performed by MU200 and MU202. The proposed activity satisfies this requirement by closing the SDC suction line isolation valves MU015 and MU018. Hence, LPSI will remain operable during the proposed activity.

c. LCO 3.5.3, ECCS-Shutdown, requires one HPSI train to be operable in Mode 3 with pressurizer pressure < 400 psia, and Mode 4.

These ECCS requirements do not include LPSI because the RCS makeup requirements are less in Modes 3 and 4. During an event in these Modes requiring ECCS actuation, water from the RWST is delivered to the RCS via the HPSI pumps and their respective supply header (s) to each of the four cold leg injection nozzles. In the event of RWST depletion, this flow path will be switched to take its supply from the containment emergency sump and deliver its flow to the RCS hot and/or cold legs. The proposed activity does not impact HPSI operability.

d. LCO 3.6.6.1, Containment Spray and Cooling System, requires two containment spray trains and two containment cooling trains to be operable in Modes I, 2, and 3. The proposed activity does not affect Containment Spray and Cooling System operability.

e. LCO 3.4.12.2, Low Temperature Overpressure Protection (LTOP) System, requires at least one of the following overpressure protection systems to be operable when the RCS temperature is greater than 256 F in Mode 4, 5, or 6 (with certain constraints):

1) SDC Relief Valve (PSV9349) with:

1. A lift setting of 406

10 psig, it. Relief valve isolation valves 2HV9337, 2HV9339, 2HV9377, and 2HV9378 open, or,

2) A minimum of one pressurizer code safety valve with a lift setting of 2500 psia

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1 Description of Proposed Change Number 504 San Onofre Nuclear Generating Station, Unit 2 The proposed activity does not affect any function of LTOP or the ability to comply with this LCO.

f. LCO 3.7.6, Condensate Storage Tank (CST T-121 and T-120), requires T-121 contained volume be 2144,000 gallons and T-120 contained volume be 2280,000 gallons. (Note: SCE submitted an amendnient application (PCN-499) to the NRC by letter dated January 11,1999, requesting the total volume be increased to 360,000 gallons. Until that PCN is approved, the higher level is maintained administratively). This LCO applies in Mode 1,2, and 3, and in Mode 4 when a steam generator is relied upon for heat removal.

The combined volume of CSTs ensures that sufficient water is available to maintain the unit in Mode 3 for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , including cooldown to SDC initiation.

As described above, the proposed activity includes:

1) A contingency plan to restore SDC operability within approximately 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2) Isolation of various postulated CST leakage paths (or staging an operator to isolate) prior to initiating repairs. This compensatory action will increase the available water for euxiliary feedwater by 80,000 gallons, extending the time available to restore SDC by approximately 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . This compensatory measure recaptures some inventory currently postulated to be lost through leakage paths before local manual isolation under the licensing basis.

Therefore, SDC will be available prior to the time the CST volume is exhausted.

D. Evaluation of the Safety and Potential Consequences of the Repair Evaluations of accidents are described in UFSAR Chapter 15, Accident Analysis. LPSI and SDC are used to mitigate the consequences of accidents and transients evaluated in the UFSAR. The proposed activity does not impact the operability of LPSI for safety injection. Restoration of SDC system operability prior to needing SDC for RCS/ decay heat removal assures that this activity will not adversely affect SDC's ability to provide long term core cooling.

For accident evaluations considering inoperable SDC, the most limiting accidents are LOCAs. UFSAR Figure 6.3-24 shows the spectrum of LOCAs evaluated in the UFSAR.

For certain size LOCAs (breaks larger than 0.01 ft2), SDC is not required for long term 10

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Description of Proposed Change Number 504 San Onofre Nuclear Generating Station, Unit 2 l

l cooling and accident mitigation and thus this repair does not affect does consequences.

Long term cooling is provided by simultaneous hot leg / cold leg HPSI injection.

, For very small break LOCAs (0.01 ft: or smaller), SDC is required for long term cooling.

The major assumptions used in performing the long term cooling analysis are listed in UFSAR Section 6.3.3.4.2. The proposed activity does not change any of those assumptions. However, the analysis credited in the UFSAR only takes credit for the volume in T-121 and does not take credit for T-120 inventory. The proposed activity does take credit for T-120's water inventory (including compensatory actions to increase its useful volume above the Technical Specification minimum limit) to extend the water inventory available to reach SDC entry conditions and to maintain that condition prior to SDC initiation. The additional time provides reasonable assurance that SDC can be returned to operable prior to the time it is required for accident mitigation.

The plant can be maintained on auxiliary feedwater using T-120 and T-121 until SDC has been returned to service. Reactor coolant inventory can be maintained using HPSI, either from the RWST or recirculation.

As shown in UFSAR Figure 15.6-126, for LOCAs 0.01 ft2 or smaller, core uncovery (and fuel damage)is not postulated. Therefore, the areas required to restore SDC operability, and locally operate the Atmospheric Dump Valves (ADVs) (required to control the steam generators on auxiliary feedwater should offsite power and normal plant support systems be unavailable) will remain habitable.

The abUity to establish SDC following a SGTR is the same as that described for the 0.01 ft2 or smaller LOCAs (UFSAR Section 15.6.3.2).

Various UFSAR Chapter 15 non-LOCA transients including seismic events are evaluated for the assumed scenario of either a loss of condenser vacuum or a loss of normal AC power, either of which requires use of one or both ADVs to effect plant cooldown prior to placing SDC into service. As long as the ADV from the affected steam generator is open, secondary side steaming provides an activity release path to the environment. In accordance with the UFSAR, these non-LOCA transient event scenarios terminate several hours into the event with the initiation of SDC and the coincident Operator closure of the ADVs to isolate the activity release path.

Should SDC be unavailable, it will be necessary for the Operators to continue use of the ADVs to effect plant cooldown. Consequently additional radioactivity may be released to the environment thereby increasing offsite and control room operator event duration dose 11

. Description of Proposed Change Number 504 San Onofre Nuclear Generating Station, Unit 2 exposures. However, the Exclusion Area Boundary (EAB) doses which are evaluated for only the first two hours of a transient are not affected by initiating SDC later in the events.

When explicitly evaluated, the Low Population (LPZ) and Control Room doses are d

evaluated for the event duration. These doses will increase as a consequence of the increased event duration. However, the increased doses will be acceptable (i.e., below SRP,10CFR100.ll and General Design Criterion [GDC] 19 dose acceptance criteria) for the following reasons:

1) LPZ doses are typically one or more orders of magnitude less than EAB doses due i to the additional atmospheric dispersion between the activity release points and the dose receptor. Consequently, increased activity releases due to a delay in SDC initiation will sull yield dose consequences that are significantly less than EAB dose consequences. For example, in the event of a SGTR with a pre-existing iodine spike, the LPZ thyroid dose assuming SDC is placed into service at 3.12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months into the event is 0.081 rem, while the 2-hour EAB dose is 2.8 rem. Even with a hypothetical factor of ten increase (> 31.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to SDC), the LPZ dose will still be less than that evaluated at the EAB.

2) A relatively large portion of the control room thyroid dose is attributed to activity entering the control room prior to the initiation of the Control Room Emergency Air Cleanup System (CREACUS). The proposed activity does not affect initiating l CREACUS. I As such, increased activity releases due to delays in SDC initiation will not yield ;

I dose consequences that are significantly greater than currently calculated on a per hour basis. For example, for SGTR with a pre-existing iodine spike, the Control Room thyroid dose assuming SDC is placed into service at 3.12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months into the event is 0.67 rem while the 3-minute control room dose is 0.11 rem. Even with a l hypothetical factor of ten increase (> 31.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to SDC) of the additional 0.56 rem dose occurring after 3 minutes, the control room dose would increase to 5.7 rem, which is significantly less than the 30 rem GDC 19 dose criterion. !

3) In the case of SGTR concurrent with the primary side temperature decreasing below 350* F, the primary to secondary side pressure gradient forcing additional radioactivity across the Technical Specification leaking steam generator tubes and into the secondary side will be reduced. As such the rate of radioactivity release to the environment is greatly reduced.

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Description of Proposed Change Number 504 San Onofre Nuclear Generating Station, Unit 2 The above dosis are based on design RCS activities. The Technical Specification coolant activity limits are lower, and would result in lower doses. The actual RCS activity at this time is less than the Technical Specification limit, providing substantial margin to the calculated doses.

E. No Significant Ilarnrds ConsidtIslion The Commission has provided standards for determining whether a significant hazards consideration exists as stated in 10 CFR 50.92. A proposed amendment to a facility operating license involves no significant hazards consideration if operation of the facility in accordance with a proposed amendment would not: (1) involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) cre'te the possibility of a new or different kind of accident from any accident previously evaluated; or (3) involve a significant reduction in margin of safety. A discussion of these standards as they relate to this amendment request follows.

1) Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated.

No.

Initiating events for accidents and transients evaluated in the Updated Final Safety Analysis Report (UFSAR) are listed in Chapter 15, Table 15.0-2, Initiating Events.

Except for a Shutdown Cooling (SDC) line break in Mode 4, both SDC and Low Pressure Safety Injection (LPSI) systems are accident mitigators and not accident initiators. The proposed activity will not change the probability of occurrence of any of the listed initiating events. The SDC piping involved in the proposed activity is isolated from the piping associated with initiating events. The proposed activity will preclude a SDC line break because SDC will not be initiated with MU015 and MU018 closed.

Therefore, this amendment request does not significantly increase the probability of an accident previously evaluated.

Evaluations of accidents are described in UFSAR Chapter 15, Accident Analysis.

LPSI and SDC are used to mitigate the consequences of accidents and transients evaluated in the UFSAR. The proposed activity does not impact the operability of LPSI for safety injection. Restoration of SDC system operability prior to needing SDC for Reactor Coolant System (RCS)/ decay heat removal assures that this 13

^

Description of Proposed Change Number 504 San Onofre Nuclear Generating Station, Unit 2 activity will not adversely affect SDC's ability to provide long term core cooling.

For accident evaluations considering inoperable SDC, the most limiting accidents are Loss of Coolant Accidents (LOCAs). UFSAR Figure 6.3-24 shows the spectrum of LOCAs evaluated in the UFSAR. For certain rize LOCAs (breaks larger than 0.01 ft2), SDC is not required for long term cooling and accident mitigation and thus, this repair does not affect does consequences. Long tenn cooling is provided by simultaneous hot leg / cold leg High Pressure Safety Injection (HPSI).

For very small break LOCAs (0.01 ft2 or smaller), SDC is required for long term cooling. The major assumptions used in performing the long term cooling analysis are listed in UFSAR Section 6.3.3.4.2. The proposed activity does not change any of those assumptions. However, the analysis credited in the UFSAR only takes credit for the volume in T-121 and does not take credit for T-120 inventory. The proposed activity does take credit for T-120's water inventory (including compensatory actions to increase its useful volume above the Technical Specification minimum limit) to extend the water inventory available to reach SDC entry conditions and to maintain that condition prior to SDC initiation. The additional time provides reasonable assurance that SDC can be returned to operable prior to the time it is required for accident mitigation. 4 The plant can be maintained on auxiliary feedwater using T-120 and T-121 until SDC has been returned to service. Reactor coolant inventory can be maintained using HPSI, either from the Refueling Water Storage Tank (RWST) or recirculation.

As shown in UFSAR Figure 15.6-126, for LOCAs 0.01 ft: or smaller, core uncovery (and fuel damage) is not postulated. Therefore, the areas required to restore SDC operability, and locally operate the Atmospheric Dump Valves (ADVs) (required to control the steam generators on auxiliary feedwater should offsite power and normal plant support systems be unavailable) will remain habitable.

The ability to establish SDC following a Steam Generator Tube Rupture (SGTR) is the same as that described for the 0.01 A2 or smaller LOCAs.

Various UFSAR Chapter 15 non-LOCA transients, including seismic events, are I evaluated for the assumed scenario of either a loss of condenser vacuum or a loss 14 l

l I

Description of Proposed Change Number 504 San Onofre Nuclear Generating Station, Unit 2 of normal AC power, either of which requires use of one or both Atmospheric Dump Valves (ADVs) to effect plant cooldown prior to placing SDC into service.

As long as the ADV from the affected steam generator is open, secondary side steaming provides an activity release path to the environment. In accordance with the UFSAR, these non-LOCA transient event scenarios terminate several hours into the event with the initiation of SDC and the coincident Operator closure of the ADVs to isolate the activity release path.

Should SDC be unavailable, it will be necessary for the Operators to continue use of the ADVs to effect plant cooldown. Consequently additional radioactivity may be released to the environment thereby increasing offsite and control room operator event duration dose exposures. However, the Exclusion Area Boundary (EAB) doses which are evaluated for only the first two hours of a transient are not affected by initiating SDC later in the events.

When explicitly evaluated, the Low Population (LPZ) and Control Room doses are evaluated for the event duration. These doses will increase as a consequence of the increased event duration. However, the increased doses will be acceptable (i.e., below Standard Review Plan,10CFR100.11 and General Design Criterion

[GDC] 19 dose acceptance criteria) for the following reasons:

1) LPZ doses are typically one or more orders of magnitude less than EAB doses due to the additional atmospheric dispersion between the activity release points and the dose receptor. Consequently, increased activity releases due to a delay in SDC initiation will still yield dose consequences that are significantly less than EAB dose consequences. For example, in the event of a SGTR with a pre-existing iodine spike, the LP7, thyroid dose assuming SDC is placed into service at 3.12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months into the event is 0.081 rem, while the 2-hour EAB dose is 2.8 rem. Even with a hypothetical factor of ten increase (> 31.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to SDC), the LPZ dose will still be less than that evaluated at the EAB.

2) A relatively large portion of the control room thyroid dose is attributed to activity entering the control room prior to the initiation of the Control Room Emergency Air Cleanup System (CREACUS). The proposed activity does not affect initiating CREACUS.

As such, increased activity releases due to delays in SDC initiation will not yield dose consequences that are significantly greater than currently 15

c Description of Proposed Change Number 504 )

San Onofre Nuclear Generating Station, Unit 2 calculated on a per hour basis. For example, for SGTR with a pre-existing !

iodine spike, the Control Room thyroid dose assuming SDC is placed into i service at 3.12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months into the event is 0.67 rem while the 3-minute control room dose is 0.11 rem. Even with a hypothetical factor of ten increase (>

31.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months prior to (SDC) of the additional 0.56 rem dose occurring after 3 3 minutes, the control room dose would increase to 5.7 rem, which is I significantly less than the 30 rem GDC 19 dose criterion.

3) In the case of SGTR concurrent with the primary side temperature decreasing below 350* F, the primary to secondary side pressure gradient forcing additional radioactivity across the Technical Specification leaking steam generator tubes and into the secondary side will be reduced. As such the rate of radioactivity release to the environment is greatly reduced.

The above doses are based on design RCS activities. The Technical Specification coolant activity limits are lower, and would result in lower doses. The actual RCS activity at this time is less than the Technical Specification limit, providing substantial margin to the calculated doses.

In the case of a seismic event, SDC will be able to perform its decay heat removal function discussed in UFSAR section 5.4.7.1.2, based on the Technical i Specification volume in T-120 and T-121 is sufficient to allow steaming for 24 l hours, compensatory measures which will increase the available CST volume to allow an additional 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months of steaming, and the repair plan includes provisions to back out of the repair and restore SDC within approximately 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . l 1

Therefore, this amendment request does not significantly increase the probability or consequences of any accident previously evaluated.

2) Does the amendment request create the possibility of a new or different kind of accident from any accident previously evaluated?

No.

UFSAR Section 15.0.1, Identification of Causes and Frequency Classification, describes how incidents are considered in the UFSAR. The initiating events are each placed in one of the categories of process variable perturbations listed in Table 15.0-1. The initiating events for which analyses are presented are listed in Table 15.0-2 along with their respective section designations. Certain initiating 16

Description of Proposed Change Number 504 San Onofre Nuclear Generating Station, Unit 2 events which are suggested for consideration are not explicitly analyzed. These initiating events, along with the reasons for omission of their analyses, are provided in the appropriate paragraphs in Chapter 15.

The components involved in the proposed activity are passive in nature, and do not interact with other Systems, Structures or Components (SSCs) in such a way as to cause any of the initiating event categories listed in Table 15.0-1.

With isolation valves MU015 and MU018 open, the possible events are bounded s by existing analyses. With the isolation valves closed, the SDC system becomes !

inoperable, but this does not create the possibility of a new or difference kind of accident.

Therefore, this amendment request does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3) Does this amendment request involve a significant reduction in a margin of safety?

No.

SCE completed a probabilistic risk assessment (PRA) of the proposed repair plan.

The assessment included all events requiring the shutdown cooling function to mitigate core damage and large early release: small break LOCAs, SGTRs, and seismic events. The increase in core damage and large early release risk are estimated to be 7.1E-6 and 1.7E-7, respectively.

The dominant contributor to core damage risk during the repair is from a seismic event of a magnitude greater than 0.3g pga. A seismic event of this magnitude or I greater is assumed to fail the condensate makeup function to the condensate storage tanks. In this case, the condensate storage tank inventory limits the time available for restoring shutdown cooling to service. The compensatory measures !

such as use of firewater to replenish the condensate tanks are not credited in the risk assessment. The dominant contributor to the large early release risk during the repair is from a steam generator tube rupture event assuming unsuccessful depressurization of the, reactor coolant system prior to refueling water storage tank inventory depletion.

Other repair options, such as performing the repair in Mode 4 (decay heat removal via steaming at reduced reactor coolar.t system pressure) and in a defueled 17

. Description of Proposed Change Number 504 San Onofre Nuclear Generating Station, Unit 2 condition during the next refueling outage, were considered. The risk of repairing the valves in Mode 4 is on the same order of magnitude as repair in Mode 1. Long term plant operation without repairing the valves until the next refueling outage was considered undesirable due to the degraded condition of the valves and the desire to do the repair in a planned and controlled manner, rather than attempt recovery actions in the unlikely event of an event requiring SDC.

Based upon the PRA results and planned contingency measures (not considered explicitly in the PRA), the overall risk of the repair plan is small. Based upon Regulatory Guide 1.174, these increases in risk are also characterized as "small."

For very small break LOCAs (0.01 fl2or smaller), SDC is required for long term cooling. The major assumptions used in performing the long term cooling plan analysis are listed in UFSAR Section 6.3.3.4.2. The proposed activity does not change any of those assumptions. However, the analysis credited M the UFSAR only takes credit for the volume in T-121 and does not take credit for T-120 inventory. The proposed activity does take credit for T-120's water inventory (including compensatory actions to increase its useful volume above the Technical Specification minimum limit) to extend the water inventory available to reach SDC entry conditions and to maintain that condition prior to SDC initiation. The additional time provides reasonable assurance that SDC can returned to operable prior to the time it is required for accident mitigation.

The ability to establish SDC following a Steam Generator Tube Rupture (SGTR) is the same as that described for the 0.01 fl2 or smaller LOCAs.

Should SDC be unavailable, it will be necessary for the Operators to continue use of the ADVs to effect plant cooldown. Consequently additional radioactivity may be released to the environment thereby increasing offsite and control room operator event duration dose exposures. However, the Exclusion Area Boundary (EAB) doses which are evaluated for only the first two hours of a transient are not affected by initiating SDC later in the events.

When explicitly evaluated, the Low Population (LPZ) and Control Room doses are evaluated for the event duration. These doses will increase as a consequence of the increased event duration. However, the increased doses will be acceptable .

(i.e., below SRP,10CFR100.11 and General Design Criterion [GDC] 19 dose acceptance criteria). l I

18 l

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l 1

Description of Proposed Change Number 504 San Onofre Nuclear Generating Station, Unit 2 The calculated doses are based on design RCS activities. The Technical Specification coolant activity limits are lower, and would result in lower doses.

The actual RCS activity at this time is less than the Technical Specification limit, providing substantial margin to the calculated doses.

In the case of a seismic event, SDC will be able to perform its decay heat removal function discussed in UFSAR netion 5.4.7.1.2, based on the Technical Specification volume in T-120 and T-121 is sufficient to allow steaming for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , compensatory measures which will increase the available CST volume to i allow an additional 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months of steaming, and the repa:r plan includes provisions to back out of the repair and restore SDC within approximately 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Therefore, this amendment request does not involve a significant reduction in a margin of safety. I l

i Based on the negative responses to these three Commission criteria, SCE concludes that j the proposed amendment involves no significant hazards consideration. )

i F. Environmental Consideration l I

Southern California Edison has determined that the proposed Technical Specification change involves no changes in the amount or type of effluent that may be released offsite, and results in no increase in individual or cumulative occupational radiation exposure. As described above, the proposed Technical Specification amendment involves no significant hazards consideration and, as such, meets the eligibility criteria for categorical exclusion set forth in 10CFR51.22(c)(9).

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