ML20063L371

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Requests Revision to Bases for TS 3/4.7.1.2,AFW for Units 2 & 3.Change Will Replace Statement in Existing Bases Re AFW Pumps Capability of Supplying 700 Gpm at SG Pressure of 1170 Psig
ML20063L371
Person / Time
Site: San Onofre  Southern California Edison icon.png
Issue date: 02/25/1994
From: Marsh W
SOUTHERN CALIFORNIA EDISON CO.
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20063L373 List:
References
NUDOCS 9403040307
Download: ML20063L371 (9)
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"; t:w rw Southem Califomia Edison Company 23 PAf4KEH STHEET IHVINE, CALIFOF4NBA 92718 wAacHe uxnss February 25, 1994 me, ,He MAP 4 AGE H (W PA#C4.E AM Mf. LAULA f(.HW Af F AlH% (714) 4%4-4403 U. S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, D. C. 20555 Gentlemen:

Subject:

Docket Hos. 50-361, and 50-362 Bases for Technical Specification 3/4.7.1.2, Auxiliary feedwater System San Onofre Nuclear Generating Station Units 2 and 3

References:

1. June 17, 1991 letter from H. B. Ray (SCE) to Document Control Desk (NRC).

Subject:

Amendment Application Nos.

106 and 91, " Auxiliary Feedwater," San Onofre Nuclear Generating Station, Units 2 and 3

2. March 5,1993 letter from H. B. Ray (SCE) to Document Control Desk (NRC).

Subject:

Amendment Application Nos.

130 and 114, " Main Steam Safety Valves," San Onofre Nuclear Generating Station, Units 2 and 3 This is a request to revise the Bases for Technical Specification (TS) 3/4.7.1.2, Auxiliary Feedwater (AFW) System, for the San Onofre Nuclear-Generating Station, Units 2 and 3. This change will replace a statement in the existing Bases that the AFW Pumps are capable of supplying 700 gpm at a Steam Generator (SG) pressure of 1170 psig with a statement of system design requi re.ments. Additional changes are also added concerning shutdown cooling entry conditions and single failure of a Motor Driven AFW Pump. The existing and proposed Bases for TS 3/4.7.1.2 are provided in Enclosures 1 and 2, respectively. t This Bases change closes out a commitment made in Reference 1 to provide the NRC with corrected AFW flow values.

Background

The original Safety Analyses performed in support of the San Onofre Units 2 and 3 Final Safety Analysis Report (FSAR) used an AFW flow rate of 700 gpm whenever the AFW system is operating. This value was used regardless of actual SG pressures. The original Safety Analyses demonstrated compliance with various design criteria.

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Document Control Desk However, because the AFW system design does not provide significant margin above the 700 gpm requirement during the early stages of some Design Basis Events (DBEs), in 1991 ABB-Combustion Engineering (ABB-CE) under contract with '

Southern California Edison (SCE) performed an analysis which demonstrates that during any DBE an AFW flow of 500 gpm to the SGs from one AFW pump would be adequate to prevent design criteria from being exceeded and accommodate stabilization of the plant. This work did not address AFW system requirements during the subsequent period of cooldown to shutdown cooling entry conditions. i The 700 gpm value for plant cooldown was previously reviewed by the NRC and determined to be acceptable based on review of our 1980 AFW System Design and Reliability Evaluation for San Onofre Units 2 & 3 as documented in the NRC Safety Evaluation Report (SER) NUREG 0712, Supplement 1, section II.E.1.1.

This evaluation was not revised as part of the 1991 ABB-CE analysis, i Therefore, the requirement that 700 gpm be available for cooldown remains unchanged. This evaluation states that the minimum required pump capacity of 700 gpm is based on a 75 F per hour cooldown rate, and that 700 gpm is always available under normal cooldown conditions.

AFW Reauirements for Post-Accident Conditions The safety analysis limit for SG pressure following a DBE is 1107 psia.

Existing DBE analyses assumed a minimum AFW total delivery of 700 gpm per pump to the SGs. When SCE calculation M56.18, "AFW System Flow Capacity," was ,

updated in 1991 to account for addition of a new check valve, it was i determined that the system is only capable of providing 711 gpm at a SG i pressure of 1107 psia. This allows very little margin for expected pump degradation. Therefore, SCE has also pursued reducing the post-accident flow requirement to 500 gpm for each pump to the SGs. ,

The Updated Final Safety Analysis Report (UFSAR) DBEs affected by AFW flow ,

were evaluated, and the most limiting events were reanalyzed. The UFSAR '

events which required reanalysis were:

1. Loss of All Normal AC Power l
2. Loss of Normal Feedwater Flow  ;
3. Loss of Normal Feedwater Flow with an Active Single Failure
4. Feedline System Pipe Break  ;

Results of the reanalyses of these four events demonstrated that an AFW total delivery to the intact SG of 500 gpm is sufficient to remove decay heat. For '

all cases, the maximum Reactor Coolant System (RCS) pressure and the maximum  ;

secondary system pressure remain within allowable pressures, which are 2750  ;

psia and 1210 psia, respectively. -;

The reanalyses with reduced AFW flow were evaluated with assumptions of primary safety valve tolerance of +2% and main steam safety valve tolerance of i

+3%. The increased safety valve tolerances were used in the reanalysis in  !

support of a separate TS change for the Main Steam Safety Valves, Proposed l Change Number (PCN) 329, which was submitted by Reference 2. The safety valve tolerances used in the reduced AFW flow analysis conservatively bound those currently allowed by existing TSs 3/4.4.2 and 3/4.7.1.

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Document Control Desk The AFW system post-accident flow requirement to the SGs has been reduced from 700 gpm for each pump to 500 gpm for each pump. The reduction is based on the 1991 analysis and will provide adequate margin for potential future system modifications and wear. The flow reduction was demonstrated to be acceptable by an update of the AFW system flow capacity calculation, a 10CFR50.59 Safety Evaluation, and the DBE reanalyses discussed above. This has been documented in the UFSAR in the 1993 update (Revision 9). '

The reduction of the AFW initial flow requirement does not involve any. ,

physical-change to the AFW system. Hence, the changes to the UFSAR have no impact on plant operations. However, the Bases to TS 3/4.7.1.2 need to be revised to provide consistency with the UFSAR and the plant capabilities.

Technical Specification Bases The purpose of the TS Bases is to describe the design functions of a system which justify a Limiting Condition of Operation for that system. The design ,

function requirements of the AFW system are: 1) provide sufficient feedwater to stabilize the plant following a DBE, and 2) to remove decay heat at a nominal cooldown rate of 75"F/hr (the emergency limit is 100"F/hr) to a point where the Shutdown Cooling system may be put into operation. Although a 100*F/hr rate of cooldown would initially require slightly higher AFW flows compared to a 75"F/hr cooldown, this difference would be easily met by the AFW system due to the rapid reduction in steam generator pressures and the associated rapidly increasing AFW flow capacity to the steam generators. ,

Currently, the Bases to TS 3/4.7.1.2 state that each AFW pump is capable of delivering 700 gpm at a SG pressure of 1170 psig. This is a statement of information on pump capability, not a limiting. condition of operation. -The pressure value of 1170 psig was the analysis limit used by ABB-CE in performing the original accident analyses. Prior to Cycle 1, it was recognized that this pressure value exceeded actual system capabilities and the SG pressure at which the initial AFW flow delivery would occur. At this ,

time the UFSAR was revised to reflect an analysis limit of 1107- psia for SG pressure. The TS Bases should also have been revised to state that the AFW pumps can supply 700 gpm per pump to the SGs at 1107 psia. However, through  !

an oversight, the Bases were not revised at that time.

As discussed above, recent evaluations show that the AFW pumps show little margin to the 700 gpm at elevated SG pressures and that 500 gpm per pump is ,

sufficient for DBE recovery. >

. This change removes the statements regarding AFW pump flow capacity of 700 gpm from the TS Bases. These statements are replaced by a description of the safety function and design basis of the system. The flow requirements of the -

AFW pumps (500 gpm per pump for post-accident conditions and 700 gpm per pump for cooldown from hot standby) are outlined in UFSAR sections 10.4.9.2 and 10.4.9.3 and in the Chapter 15 Safety Analyses of the relevant DBEs. If the pumps satisfy analysis assumptions and the operating requirements based on i these assumptions, they are capable of completing their intended safety '

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l Document Control Desk function, and, therefore, meet the design bas"s outlined in the proposed TS' Bases.

Additionally, the statement that the AFW system is designed to cool the RCS to 350*F is replaced with a statement that the AFW system is designed to cool the 1 RCS to shutdown cooling entry conditions. This change is made to avoid the '

implication that 350"F is a requirement for entry into shutdown cooling operation. A paragraph is also added that states that the AFW system is capable of recovering from a Feedwater System Pipe. Break assuming a single failure of the motor driven AFW pump aligned to the intact SG. This is 4 consistent with the assumptions of the DBE analysis requirements.

In summary, SCE requests changes to TS Bases 3/4.7.1.2 to correct the Bases and make them consistent with the plant capabilities.

I If you have any questions or comments, please let me know.

Sincerely, lbb 0

Enclosures:

cc: K. E. Perkins, Jr., Acting Regional Administrator, NRC Region V J. A. Sloan, NRC Senior Resident Inspector, San Onofre Units 2 & 3 H. B. Fields, NRC Project Manage." San Onofre Units 2 and 3.

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C ENCLOSURE 1 ,

i- Existing Bases Units 2 & 3 p

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PLANT SYSTEMS BASES 3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM The OPERABILITY of the auxiliary feedwater system ensures that the Reactor Coolant System can be cooled down to less than 350 F from normal operating conditions in the event of a total loss of off-site power. In addition, the flow paths are automatically aligned to support an Emergency Feedwater Actuation Signal or a Main Steam Isolation Signal.

Each electric driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 700 gpm at a pressure of 1170.psig to the entrance of the steam generators. The steam driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 700 gpm at a pressure of 1170 psig to the entrance of the steam generators. This capacity is sufficient to ensure that adequate feedwater flow is available to remove decay heat and reduce the Reactor Coolant System temperature to less than 350 F when the shutdown cooling system may be placed into operation.

Each electric driven auxiliary feedwater pump is powered from an independent 1E power supply, and feeds one steam generator through a set of valves powered frem the same IE source. The AC powered valves associated with the same train ele:,ric driven auxiliary feedwater pump defines that flow path. The steam- l driven auxiliary feedwater pump can feed each steam generator through two sets of valves powered from 125VDC IE power sources. Each set of valves aligned to l a steam generator from the steam driven auxiliary feedwater pump, are powered from the opposite train from the valves from the corresponding electric driven auxiliary feedwater pump. For purposes of identifying the appropriate action i statement, the steam-driven auxiliary feedwater pump flow path is defined '

as both sets of valves aligned to steam generators. Loss of Operability of one or more of the DC powered valves constitutes loss of the steam-driven auxiliary feedwater flow path.

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If the steam generators are used for decay heat removal in Mode 4 under the provisions of Technical Specifications 3/4.4.1.3, at least one motor-driven auxiliary feedwater pump and associated flow path per steam generator is required j to be OPERABLE to provide decay heat removal. l 3/4.7.1.3 CONDENSATE STORAGE TANKS The OPERABILITY of the condensate storage tank T-121 with the mimimum water volume ensures that sufficient water is available to maintain the RCS at H0T STANDBY conditions for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> followed by cooldown to shutdown cooling initia- )

tion, with steam discharge to atmosphere with concurrent loss of offsite power and most limiting single failure. The OPERABILITY of condensate storage tank T-120 in conjunction .with tank T-121 ensures that sufficient water is available

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to maintain the RCS at HOT STANDBY conditions for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> including cooldown to shutdown cooling initiation, with steam discharge to atmosphere with concur-rent loss of offsite power and most limiting single failure. The' contained water volume limits are specified relative to the highest auxiliary feedwate.r i

SAN ON0FRE-UNIT 2 B 3/4 7-2 AMENDHENT NO. 99

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4 PLANT SYSTEMS-BASES 3/4.7.1.3 CONDENSATE STORAGE TANKS (Continued) pump suction inlet in the tank for T-121, and to the T-121 cross connect siphon '

inlet for T-120. (Water volume below these datum levels is not considered recoverable for purposes of this specification.) Vortexing, internal structure, and instrument error are considered in determining the tank levels corresponding to the specified water volume limits.

Prior to achieving 100% RATED THERMAL POWER, Figure 3.7-1 is used to deter-  :

' mine the minimum required water volume for T-121 for the maximum power level (hence maximum decay heat) achieved.  ;

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SAN ONOFRE-UNIT 2 B 3/4 7-2a AMENDMENT NO. 99 i

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PLANT SYSTEMS BASES 3/4.7.1.2 AUXILIARY FEEDWATER SYSTEM The OPERABILITY of the auxiliary feedwater system ensures that the Reactor Coolant System can be cooled down to less than 350 F from normal operating conditions in the event of a total loss of offsite power. In addition, the flow paths are automatically aligned to support an Emergency Feedwater Actuation Signal or Main Steam Isolation Signal.

Each electric-driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 700 gpm at a pressure of 1170 psig to the entrance of the steam generators. The steam-driven auxiliary feedwater pump is capable of delivering a total feedwater flow of 700 gpm at a pressure of 1170 psig to the entrance of the steam generators. This capacity is sufficient to ensure that adequate feedwater flow is available to remove decay heat and reduce the Reactor Coolant System temperature to less than 350 F when the shutdown cooling system may be placed into operation. -

Each electric driven auxiliary feedwater pump is powered from an independent IE power supply, and feeds one steam generator through a set of valves powered from the same IE source. The AC powered valves associated with the same train electric driven auxiliary feedwater pump defines that flow path. The steam-driven auxiliary feedwater pump can feed each steam generator through two sets of valves powered from 125VDC 1E power sources. Each set of valves aligned to

- a steam generator from the steam driven auxiliary feedwater pump, are powered from the opposite train from the valves from the corresponding electric driven auxiliary feedwater pump. For purposes of identifying the appropriate action statement, the steam-driven auxiliary feedwater pump flow path is defined as both sets of valves aligned to steam generators. Loss of Operability of one or more of the DC powered valves constitutes loss of the steam-driven auxiliary l feedwater flow path.

If the steam generators are used f'or decay heat removal in Mode 4 under the provisions of Technical Specifications 3/4.4.1.3, at least one motor-driven auxiliary feedwater pump and associated flow path per steam generator is required I to be OPERABLE to provide decay heat removal. I 3/4.7.1.3 CONDENSATE STORAGE TANKS The OPERABILITY of the condensate storage tank T-121 with the minimum water volume ensures that sufficient water is available to maintain the RCS at HOT STANDBY conditions for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> followed by cooldown to shutdown cooling initiation, with steam discharge to atmosphere with concurrent loss of offsito i power and most limiting single failure. The OPERABILITY of condensate storage tank T-120 in conjunction with tank T-121 ensures that sufficient water is available to maintain the RCS at HOT STANDBY conditions for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> including cooldown to shutdown cooling initiation, with steam discharge to atmosphere

- with concurrent loss of offsite power and most limiting single failure. The contained water volume limits are specified relative to the highest auxiliary feedwater pump suction inlet in the tank for T-121, and to the T-121 cross connect siphon inlet for T-120. (Water volume below these datum levels.is h

f SAN ONOFRE-UNIT 3 B 3/4 7-2 AMENDHENT NO. 88 0

PLANT SYSTEMS

- BASES 3/4.7.1.3 CONDENSATE STORAGE TANKS (Continued) i not considered recoverable for purposes of this specification.) Vortexing, internal structure and instrument error are considered in determining the tank levels corresponding to the specified water volume limits.

Prior to achieving 100% RATED THERMAL POWER, Figure 3.7-1 is used to determine the minimum required water volume for T-121 for the maximum power level (hence maximum decay heat) achieved.

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SAN ONOFRE-UNIT 3 B 3/4 7-2a AMENDMENT NO. 86 i

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