ML20245D589

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Amends 64 & 53 to Licenses NPF-10 & NPF-15,respectively, Revising Tech Specs,Including Tech Spec 3.1.3.6 Re Regulating Control Element Assembly Insertion Limits
ML20245D589
Person / Time
Site: San Onofre  Southern California Edison icon.png
Issue date: 08/03/1988
From: Knighton G
Office of Nuclear Reactor Regulation
To:
Shared Package
ML20245D591 List:
References
TAC-54737, TAC-54738, TAC-65547, TAC-65548, TAC-66816, TAC-66817, TAC-66820, TAC-66821, TAC-66822, TAC-66823, TAC-66826, TAC-66827, NUDOCS 8808310208
Download: ML20245D589 (58)
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SOUTHERN CALIFORNIA EDISON COMPANY SAN DIEGO GAS AND ELECTRIC COMPANY THE CITY OF RIVERSIDE, CALIFORNIA THE CITY OF ANAHEIM, CALIFORNIA DOCKET NO. 50-361 SAN ONOFRE NUCLEAR GENERATING STATION, UNIT 2 I AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 64 License No. NPF-10 1. The Nuclear Regulatory Comission (the Coeission) has found that: A. The applications for amendment to the license for San Onofre Nuclear-Generating Station, Unit 2 (the facility) filed by the Southern California. Edison Company (SCE) on behalf of itself and San Diego Gas and Electric Company, The City of Riverside and The City of Anaheim, California (licensees), dated April 27, 1984 and May 12, November 4, and December 14, 1987'(as supplemented April 14 and May 6, 1988), comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's regulations as set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the applications, as amended, the provisions of the Act, and the regulations of the Commission; 1 C. There is reasonable assurance: (i)thattheactivitiesauthorized by this amendment can be conducted without endangring the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations set forth in 10 CFR Chapter I; D. The issuance of this license amendment will not be inimical to the consnon defense and security or to the health and safety of the public; E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied. 880831GC6 MM l ete \\

! 2. ' Accordingly, The license is amended by changes to the Technical-Specifications as indicated in the attachment.to this amendment and amended to.Q(2) of Facility Operating License No. NPF-10 is hereby Paragraph 2 read as follows: (2) Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 64, are hereby incorporated in the license. SCE shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan. 3. The changes in Technical Specifications are to become effective within 30 days of issuance of the amendment, with the exception of the change to Technical Specification 3/4.4.10, which is to become effective prior to startup from the Cycle 5 Refueling Outage. In the period between issuance of the amendment and the effective date of the new Technical Specifications, the licensees shall adhere to the Technical Specifications existing at the time. The period of time during changeover shall be minimized. 4 This license amendment is effective as of the date of its issuance. FOR THE NUCLEAR REGUL TORY COMMISSION ieorge . Knighto, Project Director Project Directorate V i Division of Reactor Projects - III, j IV, Y and Special Projects

Attachment:

f Changes to the Technical Specifications Date of Issuance: August 3, 1988 e i i ^ i

. c,- p 1 F 1 - ATTACHMENT TO LICEtiSE AMENDMENT NO. 64 ~ FACILITY OPERATING LICENSE NO.'NPF-1C DOCKET NO. 50-361 Replace the.following pages of the Appendix A Technical Specifications with the enclosed pages. The revised pages are identified.by Amendment number and contain vertical lines indicating the area.of change. Also to be replaced are the following overleaf pages to the amended pages. Amendment =Pages Overleaf Pages^ y Xi Xii 2-4 2-3 3/4 1-22 3/4 1-21 3/4 1 -- 3/4'l-24 3/4 1-25 3/4 1-26 3/4 3-3' 3/4 3-4 3/4 3-6 3/4 3-5 3/4 3-28 3/4 3-27 3/4 4-35 3/4 4-36 3/4 5-1 3/4 5-2 3/4 10-2 3/4 10-1 B3/4 1-5 B3/4 4-9 l --__n--

m a. y l

.1 INDEX I

i LIMITING CONDITION FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE HOTSROTD0WN............................................ 3/4 4-3 COLD SHUTDOWN - Loops F111ed............................ 3/4 4-5 COLD SHUTDOWN - Loops Not Fi11ed........................ 3/4 4-6 3/4.4.2 SAFETY VALVES - 0PERATING............................... 3/4 4-7 3/4.4.3 PRESSURIZER............................................. 3/4 4-8 3/4.4.4 STEAM GENERATORS........................................ 3/4 4-9 i 3/4.4.5 REACTOR COOLANT SYSTEM LEAKAGE i LEAKAGE DETECTION SYSTEMS............................ 3/4 4-16 OPERATIONAL LEAKAGE.................................. 3/4 4-17 3/4.4.6 CHEMISTRY............................................... 3/4 4-20 3/4.4.7 SPECIFIC ACTIVITY....................................... 3/4 4-23 3/4.4.8 PRESSURE / TEMPERATURE LIMITS REACTOR COO LANT SYSTEM............................... 3/4 4-27 PRESSURIZER - HEATUP/C00LDOWN........................ 3/4 4-31 OVERPRESSURE PROTECTION SYSTEMS RCS TEMPERATURE < 235'F........................... 3/4 4-32 RCS TEMPERATURE I235'F............................. 3/4 4-33 3/4.4.9 STRUCTURA L INTEGRITY.................................... 3/4 4-34 3/4.4.10 REACTOR COOLANT GAS VENT SYSTEM......................... 3/4 4-35 l 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 SAFETY INJECTION TANKS.................................. 3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - T,,, g 35Wm.... - ~.. ~. m m m..... 3 4 5-3 3/4.5.3 ECCS SUBSYSTEMS - T,y,< 350*Fm. m.~ m.~ m.~ ~ ~ ~ ~ 3/4 5-7 3/4.5.4 REFUELING WATER STORAGE TANK............................ 3/4 5-8 i SAN Gn0FRE-UNIT 2 V AMENDMENT NO. 64

INDEX BASES SECTION PAGE 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION......... B 3/4 4-1 3/4.4.2 SAFETY VALVES......................................... B 3/4 4-1 3/4.4.3 PRESSURIZER........................................... B 3/4 4-2 3/4.4.4 STEAM GENERATORS...................................... B 3/4 4-2 3/4.4.5 REACTOR COOLANT SYSTEM LEAKAGE........................ B 3/4 4-4 3/4.4.6 CHEMISTRY............................................. B 3/4 4-4 3/4.4.7 SPECIFIC ACTIVITY..................................... B 3/4 4-5 3/4.4.B PRESSURE /TEMPE RATURE LIMITS........................... B 3/4 4-6 3/4.4.9 STRUCTURAL INTEGRITY.................................. B 3/4 4. 3/4.4.10 REACTOR COOLANT GAS VENT SYSTEM....................... B 3/4 4-9 l 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 SAFETY INJECTION TANKS................................ B 3/4 5-1 3/4.5.2 and 3/4.5.3 EtCS SUBSYSTEMS........................... B 3/4 5-1 3/4.5.4 REFUELING WATER TANK.................................. B 3/4 5-2 l SAN ONOFRE-UNIT 2 XI AMENDMENT NO. 64

f INDEX l BASES SECTION PAGE 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAI,*. MENT........ B 3/4 6-1 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS................. B 3/4 6-3 3/4.6.3 CONTAINMENT ISCLATION VALVES...........'............... B 3/4 6-4 -' 3/4.6.4 COMBUSTIBLE GAS'CONTR0L............................... B 3/4 6-5 i , ? i i ) l I SAN ONOFRE-UNIT 2 XII AMENDMENT NO.16 l

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e o,- REACTIVITY CONTROL SYSTEMS SHUTDOWN CEA INS' RTION LIMIT t LIMITING CONDITION FOR OPERATION 4 L 3.1.3.5 All. shutdown CEAs shall be withdrawn to greater than or equal to j 145 inches. APPLICABILITY: MODES I and 2"#, ACTION: With a maximum of one shutdown CEA withdrawn to less than 145 inches, except fer surveillance testing pursuant to Specification 4.1.3.1.2, within one hour either: Withdraw the CEA to greater than or equal to 145 inches, or a. b. Declare the CEA inoperable and apply Specification 3.1.3.1. i i SURVEILLANCE REQUIREMENTS 4.1.3.5 Each shutdown CEA shall be determined to be withdrawn to greater than or equal to 145 inches: Within 15 minutes prior to withdrawal of any CEAs in regulating a. groups during an approach to reactor criticality, and b. At least once per 12 hours thereafter. ~ "See Special Test Exception 3.10.2.

  1. With K,ff greater than or equal to 1.0.

SAN ONOFRE-UNIT 2 3/4 1-21

5,. REACTIVITY CONTROL' SYSTEMS REGULATING CEA' INSERTION LIMITS LIMITING CONDITION FOR-0PERATION 3.1. 3. 6 When COLSS is in-service, the regulating CEA groups shall be limited a. to the withdrawal sequence and to the insertion limits.shown on Figure 3.1-2. The CEA insertion between the.Long Term Steady State Insertion Limits and the Transient Insertion Limits is restricted to: s-1. Less than or equal to 4 hours per 24 hour interval, 2. Less than or equal.to 5 Effective Full Power Days per 30 i i Effective Full Pcwer Day Interval, and i 3. Less than or equal to 14 Effective Full Power Days per 365 Effective Full Power Day: Interval. b. When COLSS is out-of-service, the regulating CEA groups shall be limited to the Short Term-Steady State Insertion Limit shown on' Figure 3.1-2. The CEA insertion between the Long Term Steady State Insertion Limits and the Short Term Steady State Insertion Limits is restricted to: 1. Less than or equal to'4 hours per 24 hour interval, I 2. Less than or equal to 5 Effective Full Power Days per 30 Effective Full Power Day Interval, and q 3. Less than or equal to 14 Effective Full Power Days per 365 Effective Full Power Day Interval. APPLICABILITY: MODES 1* and 2*#. 1 ACTION: When COLSS is in service and With the regulating CEA groups inserted beyond the Transient a. Insertion Limits, except for surveillance testing pursuant to Specification 4.1.3.1.2, within two hours either: 1. Restore the regulating CEA groups to within the limits, or 2. Reduce THERMAL POWER to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the CEA group position l using the above figure. a l See.Special Test Exceptions 3.10.2 and 3.10.4.

  1. With K,ff greater than or equal to 1.0.

I SAN ONOFRE-UNIT 2 3/4 1-22 AMENDMENT NO. 64 I

s.. ' REACTIVITY CONTROL SYSTEMS l J ACTION: (Continued)- i With'the regulating'CEA groups inserted between the Long Term Steady b. Statr1nsertion Limits and the Transient Insertion Limits for intervals greater than 4 hours per 24 hour interval, operation may proceed provided either: 1. The Short Tara Steady State Insertion Limits of Figure 3.1-2 are not exceeded, or 2. Any subsequent increase in THERMAL POWER is restricted to less I than or equal to 5% of RATED THERMAL POWER per hour. With the regulating CEA groups inserted between the long Ters Steady c. State Insertion Limits and the Transient Insertion Limits for intervals greater than 5 EFPD per 30 EFPD interval or greeter than l 3 14 EFPD per 305 EFPD. interval, either: 1. Restore the regulating groups to within the Long Tara Steady i State Insertion Limits within two hours, or .I 2. Be in at least HOT STANDBY within 6 hours. I When COLSS.is out of service.and the regulating CEA groups are inserted beyond_ _ the Short Term Steady State Insertion Limit except for Surveillance testing pursuant to Specification 4.1.3.1.2, within two hours either: .a. -Restore the regulating CEA group to within the limit, er b. Reduce thermal power to less than or equal to the fraction of Rated Thermal Power which is allowed by the CEA group position and the

Short Tern Steady State Insertion Limit.

i SURVEILLANCE REQUIREMENTS 4.1.3.6 The position of each regulating CEA group shall be deterstned to be within the Transient Insertion Limits at least once per 12 hours except during time intervals when the PDIL Auctioneer Alars Circuit is inoperable, then verify the individual CEA positions at least once per 4 hours. The accumulated times during which the regulating CEA groups are inserted beyond the Long Term Steady State Insertion Limits but within the Transient Insertion Limits shall be determined at least once per 24 hours. l SAN ONOFRE-UNIT 2 3/4 1-23 AMENDMENT NO. 64

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j . REACTIVITY CONTROL SYSTEMS PART LENGTH CEA INSERTION LIMITS'- LIMITING CONDITIOR FOR OPERATION The pe[t' length CEA group shall be limited to the insertion limits 3.1.3.7 shown on Figure 3.1-3 with PLCEA insertion between the Long Term Steady State Insertion Limit and the Transient Insertion Limit restricted to: I < 7 Effective Full Power Days per 30 Effective Full Power Day a. interval,;and b. < i4 Effective Full Power Days per 365 Effective Full Power Day Tnterval. APPLICABILITY: MODE I above 20% of RATED THERMAL POWER

  • ACTION:

a. With the part length CEA groups inserted beyond the Transient Inser-tion Limit, except for surveillance testing pursuant to specification 4.1.3.1.2, within two hours either: 1. . Restore the part length CEA groups to within the limit, or 2. Reduce THERMAL POWER to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the PLCEA group position using Figure 3.1-3. l b. With the part length CEA groups inserted between the Long Term Steady 1 State Insertion Limit and the Transient Insertion Limit'for intervals > 7 EFPD per 30 EFPD interval or > 14 EFPD per 365 EFPD interval, either: 1. Restore the part length groups to within the Long Term Steady State Insertion Limit within two hours, or 2. Reduce THERMAL POWER to less than or equal to 20% of RATED THERMAL POWER within the next 4 hours. SURVEILLANCE REQUIREMENTS 4.1.3.7 The position of the part length CEA groups shall be detemined to be within the Transient Insertion Limit at least once per 12 hours. The accumu-lated time during which the part length CEA groups are inserted beyond the Long Term Steady State Insertion Limit but within the Transient Insertion Limit shall be determined at least once per 24 hours.

  • See Special Test Exception 3.10.2.

SAN ONOFRE-UNIT 2 3/4 1-25 AMENDMENT NO. 64

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TABLE 3.3-1 (Continued) TABLE NOTATION With the protective system trip breakers in the closed position, the CEA drive system capable of CEA withdrawal, and fuel in the reactor vessel.

  1. The provisions of Specification 3.0.4 are not applicab N.

-4 (a). Trip may be manually bypassed above 10 % of RATED THERMAL POWER; bypass shall_g%'ofRATEDTHERMALPOWER.e automatically removed when THERMAL PO to 10 (b) Trip may be manually bypassed below 400 psia; bypass shall be automatically removed whenever pressurizer pressure is greater than or equal to 400 psia. ~4 (c) Trip may be manually bypassed below 10 % of RATED THERMAL POWER; bypass-shall be auy% of RATED THERMAL POWER.matically removed when THERMAL P equal to 10 During testing pursuant to Special Test Exception 3.10.2 or 3.10.3, trip may be manually bypassed below 5% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is greater than or equal to 5% of RATED THERMAL POWER. (d) Trip may be bypassed during testing pursuant to Special Test Exception 3.10.3. (e) See Special Test Exception 3.10.2. (f) Each channel shall be comprised of two trip breakers; actual trip logic shall be one-out-of-two taken twice. (g) Trip may be bypassed below 55% RATED THERMAL POWER. ACTION STATEMENTS ACTION 1 With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or be in at least HOT STANDBY within the next 6 hours and/or open the protective system trip breakers. ACTION 2 With the number of channels OPERABLE one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may continue provided the inoperable channel is placed in the bypassed or taipped condition within I hour. If the inoperable channel is bypassed, the desirability of maint.aining this channel in the i bypassed condition shall be reviewed in accordance with Specification 6.5.1.6e. The channel shall be returned to OPERABLE status no later than duri:rg the next COLD SHUTDOWN. i SAN ONOFRE-UNIT 2 3/4 3-4 AMENDMENT NO. 64

TABLE 3.3-1 (Continued) ACTION STATEMENTS With a channel process measurement circuit that affects multiple functional units inoperable or in test, bypass or trip all associated functional' units as listed below: Process Measurement Circuit Functional Unit Bypassed 1. Linear Power-Linear Power Level - High (Subchannel or Linear) Local Power Density - High DNBR - Low 2. Pressurizer Pressure - High Pressurizer Pressure - High Local Power Density - High DNBR - Low 3. Containment Pressure - High Containment Pressure - High (RPS) Containment Pressure - High (ESF) 4. Steam Generator Pressure - Steam Generator Pressure - Low Low Steam Generator AP 1 and 2 (EFAS 1 and 2) 5. . Steam Generator Level Steam Generator Level - Low. i Steam Generator Level - High Steam Generator AP (EFAS) 6. Core Protection Calculator Local Power Density - High DNBR - Low ACTION 3 - With the. number of channels DPERABLE one less than the Minimum Channels OPERABLE requirement,.STARTUP and/or POWER OPERATION may continue provided the following conditions are satisfied: a. Verify that one of the inoperable channals has been bypassed and place the other channel in the tripped condition within I hour, and i b. All functional unit affected by the bypassed / tripped channel shall also be placed in the bypassed / tripped condition as listed below: Process Measurement Circuit Functional Unit Bypassed / Tripped 1. Linear Power Linear Power Level - High (Subchannel or Linear) Local Power Density - High DNBR - Low l SAN ONOFRE-UNIT 2 3/4 2-5

t TABLE 3.3-1 (Continued) ACTION STATEMENTS 2. Pressurizer Pressure - Pressurizer Pressure - High High Local Power Density - High DNBR - Low 3. Containment Pressure - Containment Pressure - High (RPS) High Containment Pressure - High (ESF) 4. Steam Generator Steam Generator Pressure - Low ) Pressure - Low Steam Generator AP 1 and 2 i (EFAS 1 and 2) 5. Steam Generator Level Steam Generator Level - Low Steam Generator Level - High Steam Generator AP (EFAS) 6. Core Prctection Local Power Density - High Calculator DNBR - Low STARTUP and/or POWER OPERATION may continue until the performance of the next required CHANNEL FUNCTIONAL TEST. Subsequent H STARTUP and/or POWER OPERATION may continue if one channel is restored to OPERABLE status and the provisions of ACTION 2 are satisfied. 4 k ACTION 4 With'the number of channels OPERABLE one less than required by i the Minimum Channels OPERABLE requirement, suspend all operations involving positive reactivity changes. ACTION 5 With the number of channels OPERABLE one less than required by the Minimum s < annels OPERABLE requirement, be in at least HOT l STANDBY within 6 hours. ACTION 6 With~one CEAC inoperable, operation may continue for up a. to 7 days provided that at least once per 4 hours, each CEA is verified to be within 7 inches (indicated position) of all other CEA's in its group. After 7 days, operation may continue provided that Action 6.b is met.* If the exemption to Specification 3.0.4 is used, Action 6.b must be met. d b. With both CEACs inoperable, operation may continue provided that:" 1. Within 1 hour the DNBR margin required by Specifica' tion 3.2.4.b (COL 55 in service) or l Specification 3.2.4.d (COLSS out of service) is satisfied.

  • Note:

Requirements for CEA position indication given in Technical Specification 3.1.3.2. i SAN ONOFRE-UNIT 2 3/4 3-6 AMENDMENT NO. 64 1 1

TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME (SEC) 1. Manual ~~ a. SIAS Safety Injection Not Applicable Control Room Isolation Not Applicable Containment Isolation (3) Not Applicable Containment Emergency Cooling Not Applicable b. CSAS Containment Spray Not Applicable c. CIAS Containment Isolation Not Applicable d. MSIS Nain Steam Isolation Not Applicable e. RAS ~~ Containment Sump Recirculation Not Applicable f. CCAS Containment Emergency Cooling Not Applicable g. EFAS Auxiliary Feedwater Not Applicable h. CRIS Control Room Isolation Not Applicable 1. TGIS Toxic Gas Isolation Not Applicable j. FHIS Fuel Handling Building Isolation Not Applicable k. CPIS Containment Purge Isolation Not Applicable SAN ONOFRE - UNIT 2 3/4 3-27 l 1

l Table 3.3-5'(:ontinued) l INITIATING SIGNAL AND FUNCTION RESPONSE TIME (SEC) 2. Pressurizer Pressure-Low f a. SIAS (1) Safety Injection (a) High Pressure Safety Injection 31.2* (b) Low Pressure Safety Injection 41.2* (c) Charging Pumps 31.2* (2) Control Room Isolation Not Applicable (3) Containment Isolation (NOTE 3) 11.2* (NOTE 2) (4) Containment Spray (Pumps) 25.6* (5) Containment Emergency Cooling (a) CCW Pumps 31.2* (b) CCW Valves (Note 4b) 23.2* (c) Emergency Cooling Fans 21.2* 3. Containment Pressure-High a. SIAS (1) Safety Injection (a) High Pressure Safety Injection 41.0* (b) Low Pressure Safety Injection 41.0* (2) Control Room Isolation Not Applicable (3) Containment Spray (Pumps) 25.4* (4) Containment Emergency Cooling (a) CCW Pumps 31.0* (b) CCW Valves (Note 4b) 23.0* (c) Emergency Cooling Fans 21.0* b. CIAS (1) Containment Isolation 10.9* (NOTE 2) (2) Main Feedwater Isolation 10.9 and Backup Isolation Valves (HV 4048, HV 4052, HV 1105, HV 1106, HV 4047, HV 4051) (3) CCW Valves (Note 4a) 20.9 (4) Mainsteam Isolation Valves R V 8204, 8.9 HV 8205) (5) Minipurge Isolation Valves 5.9 4. Containment Pressure - High-High CSAS Containment Spray 23.0* SAN ONOFRE - UNIT 2 3/4 3-28 AMENDMENT NO. 64

e REACTOR C00LANT' SYSTEM- .l -3/4.4;10 REACTOR COOLANT GAS VENT SYSTEM l LIMITING CONDITION FOR OPERATION i 3.4.10.The Resetbr Coolant Gas Vent System shall be OPERABLE with: At least one of valves 2HV0296A or 2HV02968 capable of being powered a. from an emergency bus and providing a vent path from the reactor vessel head; and,. b. At least one of valves 2HV0297A or 2HV02978 capable of being powered from an emergency bus and providing a vent path from the pressurizer steam space; and, At least one of valves 2HV0298, capable of being powered from an c. emergency bus and providing a vent path to the containment atmosphere, or 2HV0299, capable of being powered from an emergency bus and providing a vent path to the quench tank; and

d..

Valves 2HV0296A, 2HV02968, 2HV0297A, 2HV02978, 2HV0299 and 2HV0298 all closed. APPLICABILITY: MODES 1, 2, 3 and 4 ACTION: a. .With any of valves 2HV0296A, 2HV0296B, 2HV0297A or 2HV02978 inoperable, operation may continue provided that: 1) power is removed from the inoperable valve (s) within 4 hours;

and, ii) valves 2HV0299 and 2HV0298 are maintained closed and power is removed within 4 hours; and, iii) the inoperable valve (s) is restored to OPERABLE status during the next COLD SHUTOOWN.

b. With any of valves 2HV0299 or 2HV0298 inoperable, operation may continue provided that: 1)- Fower is removed from the inoperable valve (s) within 4 hours;

and,
11) valves 2HV0296A, 2HV02968, 2HV0297A and 2HV02978 are all maintained closed and power is removed within 4 hoursTand i

9 SAN ONOFRE'- UNIT 2 3/4 4-35 AMENDMENT NO. 64

l . REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION iii) the inoperable valve (s) is restored to OPERABLE status during the next COLD SHUTDOWN. c. The provisions of 3.0.4 are not applicable for entry into MODES 3, 2 and 1. L 4 I SURVEILLANCE REQUIREMENTS f 4.4.10 Each reactor coolant system vent path shall be demonstrated OPERABLE at least once per 18 months by: 1. Verifying all manual isolation valves in each vent path are locked in the open position. 2. Cycling each valve in the vent path through at least one complete cycle of full travel from the control room during COLD SHUTDOWN or REFUELING. 3. Verifying flow through the reactor coolant vent system vent paths during venting during COLD SHUTDOWN. i SAN ON0FRE - UNIT 2 3/4 4-36 AMENDMENT NO. 64 l

.w i i 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4. 5.1 SAFETY 1NJECTION TANKS LIMITING CONDITION FOR OPERATION 3.5.1 Each reactor coolant system safety injection tank shall be OPERABLE with: The, isolation valve open and power to the valve removed, a. b. A contained borated water voltme of between 1680 and 1807 cubic

feet, Between 1850 and 2800 ppe of boron, and c.

d. A nitrogen cover pressure of between 615 and 655 psia. APPLICABILITY: MODES 1, 2 and 3.* ACTION: With one safety injection tank inoperable, except as a result of a ~ a. closed isolation valve, restore the inoperable tank to OPERABLE status within one hour or be in at least HOT STANDBY within the next 6 hours and in HOT SHUTDOWN within the following 6 hours. b. With one safety injection tank inoperable due to the isolation valve being closed, either immediately open the isolation valve or be in at least HOT STANDBY within one hour and be in HOT SHLTTDOWN within the next 12 hours. L-SURVEILLANCE REQUIREMENTS l 4.5.1 Each safety injection tank shall be demonstrated OPERABLE: a. At least once per 12 hours by: 1. Verifying that the contained borated water volume and nitrogen cover pressure in the tanks is within the above limits, and 2. Verifying that each safety injection tank isolation valve is open. "With pressurizer pressure greater than or equal to 715 psia. l l SAN ONOFRE-UNIT 2 3/4 5-1 AMENDMENT NO.64

' ] 1 ' EMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Cantinued) l I b. At least once per 31 days and within 6 hours after each solution volume increase of greater than or equal to 1% of tank volume by verifying the boron concentration of the safety injection tank solution. j j At least once per 31 days by verifying the fuses removed from each c. safety injection tank vent valve. d. At least once per 31 days when the RCS pressure is above 715 psia, ~ by verifying that the isolation valve operator breakers are padlocked in the open position. At least once per 18 months by verifying that each safety injection e. tank isolation valve opens automatically under each of the following conditions: 1. Before an actual or simulated RCS pressure signal exceeds 715 psia, and 2. Upon receipt of an SIAS test signal. t 1 l l l l t SAN ONOFRE-UNIT 2 3/4 5-2 Y

L 3/4.10 SPECIAL' TEST EXCEPTIONS 3/4.10.1.SHUTOOWN MARGIN ' LIMITING CONDITION FOR OPERATION 3.10.1' The SHUTDOWN' MARGIN requirement of Specification 3.1.1.1 may be. suspended for measurement of CEA worth ~and shutdown margin provided reactivity equivalent to at least the highest estimated CEA worth is available for trip insertion from OPERABLE CEA(s). APPLICABILITY: MODES 2 and 3* ACTION: a. With any full length CEA not_ fully inserted and with less than the above reactivity equivalent available for trip insertion, immedi- - ately initiate and continue boration at greater than or equal to 40 gpm of a solution containing greater than or equal to 2350 ppm l boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored. b. With all full length CEAs fully inserted and the reactor suberitical-by less than the above reactivity equivalent, innediately~ initiate and continue boration at greater than or equal to 40 gom of a solution containing greater than or equal to 2350 ppe boron or its l equivalent until the SHUTDOWN MARGIN required by Specifiestion 3.1.1.1 is restored. SURVEILLANCE REQUIREMENTS 1 4.10.1.1 The position of each full length and part length CEA required either partially or fully withdrawn shall be determined at least once per 2 hours. 4.10.1.2 Each CEA not fully inserted shall be demonstrated capable of full j insertion when tripped from at least the 50% withdrawn position within 7 days l prior to reducing the SHUTDOWN MARGIN to less than the limits of ] Specification 3.1.1.1. ] l 1 l " Operation in MODE 3 shall be limited to 6 consecutive hours. SAN ON0FRE-UNIT 2 3/4 10-1 AMENDMENT NO. 61 1 a

y SPECIAL TEST EXCEPTIONS 1

3/4.10.2 GRO.UP HEIGHT,- INSERTION AND POWER DISTRIBUTION LIMITS-LIMITING CONDITION FOR OPERATION 3.10.2 The moderator temperature coefficient; group height, insertion, and. power distribution limits of Specifications 3.1.1.3, 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.1.3.7, 3.2.2, 3.2.3, 3.2.7, footnote C of Table 3.3-1 and footnote 5 of Table'2.2-1, and the Minimum Channels OPERABLE requirement of Functional Unit 15 { -of Table 3.3-1 may be suspended during the performance of PHYSICS TESTS provided: 1 i The THERMAL POWER is restricted to the test power plateau which a. shall not exceed 85% of RATED THERMAL POWER, and b. The limits of Specification 3.2.1 are maintained and determined as specified in Specification 4.10.2.2 below. APPLICABILITY: MODES 1 and 2. ACTION: With any of the limits of Specification 3.2.1 being exceeded while the requirements of Specifications 3.1.1.3, 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.1.3.7, 3.2.2, 3.2.3, 3.2.7, footnote C of Table 3.3-1 and footnote 5 of Table 2.2-1, and the Minimum Channels OPERABLE requirement of Functional Unit 15 of - Table 3.3-1 are' suspended, either: Reduce THERMAL POWER sufficiently to satisfy the requirements of a. Specification 3.2.1, or b. Be in HOT STANDBY within 6 hours. SURVEILLANCE REQUIREMENTS 4.10.2.1 The THERMAL POWER shall be determined at least once per hour during PHYSICS TESTS in which the requirements of Specifications 3.1.1.3, 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.1.3.7, 3.2.2, 3.2.3, 3.2.7, footnote C of Table 3.3-1 and footnote 5 of Table 2.2-1, or the Minimum Channels OPERABLE requirement of Functional Unit 15 of Table 3.3-1 are suspended and shall be verified to be within the test power plateau. 4.10.2.2 The linear heat rate shall be determined to be within the limits of Specification 3.2.'1 by monitoring it' continuously with the Incore Detector Monitoring System pursuant to the requirements of Specifications 4.2.1.2 and 3.3.3.2 during PHYSICS TESTS above 5% of RATED THERMAL POWER in which the requirements of Specifications 3.1.1.3, 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.1.3.7, 3.2.2, 3.2.3, 3.2.7, footnote C of Table 3.3-1 and footnote 5 of Table 2.2-1, or the Minimum Channels OPERABLE requirement of Functional Unit 15 of Table 3.3-1 are suspended. SAN ON0FRE-UNIT 2 3/4 10-2 AMENDMENT NO. 64

7 m l REACTIVITY CONTROL SYSTEMS BASES MOVABLE CONTROL ASSEMBLIES (Continued) The establishment of LS$5 and LCOs require that the expected long and . short term behavior of the radial peaking factors be determined. The long term behavior relates to the variation of the steady state radial peaking factors with core burnup and is affected by the amount of CEA insertion assumed, the portion of a burnup cycle over which such insertion is assumed and the expected power level variation throughout the cycle. The short tem behavior relates to transient' perturbations to the steady-state radial peaks due to radial xenon redistribution. The magnitudes of such perturbations depend upon the expected use of the CEAs during anticipated power reductions and load maneuvering. Analyses are performed based on the expected mode of operation of the NSSS (base load maneuvering, etc.) and from these analyses CEA insertions are deter-mined and a consistent set of radial peaking factors defined. The Long Tarm i Steady State and Short Tem Insertion Limits are determined based upon the { ' assumed mode of operation used in the analyses and provide a means of preserv- { ing the assumptions on CEA insertions used. The limits specified serve to limit the behavior of'the radial peaking factors within the bounds determined from analysis. The actions specified serve to limit the extent of radial menon-redistribution effects to those accommodated in the analyses. The Long and Short Term Insertion Limits of Specification 3.1.3.6 are specified for the plant which has been designed for pris,arily base loaded operation but which has i the ability to accommodate a limited amount of load maneuvering. The Transient Insertion Limits of Specification 3.1.3.6 and the Shutdown CEA Insertion Limits of Specification 3.1.3.5 ensure that 1) the minimum SHUT-DOWN MARGIN is maintained, and 2) the potential effects of a CEA ejection acci-j i dont are limited to acceptable levels. Long ters' operation at the Transient Insertion Limits is not pemitted since such operation could have effects on -the core power distribution which could-invalidate asstaptions used to deter-I mine the behavior of the radial peaking factors. The Part Length CEA Insertion Limits of Specification 3.1.3.7 ensure that adverse power shapes and rapid local power changes which affect radial peaking factors and DNB considerations do not occur as a result of a part length CEA group covering the same axial segment of the fuel assemblies for an extended '3 period of time during operation. The CEA fully withdrawn position is defined to be greater than or equal to 145 inches. The extreme limits of CEA travel, fully withdrawn and fully inserted, may be described as the upper electrical limit ar,d lower electrical limit respectively. ) SAN ON0FRE-UNIT 2 8 3/4 1-5 AMENDMENT NO.6 j \\

.= ' REACTOR COOLANT SYSTEM-8ASES ~; PRESSURE / TEMPERATURE LIMITS (Continued) The OPERABILITY of the Shutdown Cooling System relief valve or a RCS vent opening of greater than 5.6 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits' of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 235'F. The Shutdown Cooling System relief valve has adequate relieving capability to protect the RCS from overpressurization when the transient is. l limited to either (1) the start of an idle RCP with the secondary water temperature of the steam generator less than or equal to 100'F above the RCS cold leg temperatures or (2) inadvertent safety injection actuation with two HPSI pumps injecting into a water-solid RCS with full charging capacity and letdown isolated. 3/4.4.9 STRUCTURAL INTEGRITY j i The inservice inspection and testing programs for ASME Code Class 1, 2 and 3 components ensure that the structural integrity and operational readiness of these components will be maintained at an acceptable level throughout the life of.the plant. These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda ' l as required by 10 CFR Part 50.55a(g) except where specific written relief has - been granted by the Commission pursuant to 10 CFR Part 50.55a (g) (6) (1). Components of the reactor coolant system were designed to provide access to permit inservice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code,1974 Edition and Addenda through Sumner 1975. 3/4.4.10 REACTOR C0OLANT GAS VENT SYSTEM -Reactor coolant system gas vents are pruvided to exhaust noncondensible gases from the primary system that could inhibit natural circulation core cooling following a non-design bases accident. The OPERABILITY of at least one reactor coolant system vent path from the reactor vessel head and the pressurizer steam space ensures the capability exists to perform this function. The design redundancy of the Reactor Coolant Gas Vent Systes serves to minimize the probability of inadvertent or irreversible actuation while ensuring that a single failure of a vent valve, or control systes does not prevent isolation of the vent path. ) The function, capabilities, and testing requirements of the Reactor Coolant Gas Vent System are consistent with the requirements of Itaa II.b.1 of NUREG-0737, " Clarification of TMI Action Plan Requirements," November 1980. l \\ ' SAN ONOFRE-UNIT 2 B 3/4 4-9 AMENDMENT NO.64 lo j

_ -, _ _ _ _ = - - _ _ _ - - _ - - _ _ - _ _ _ _ _ - _ _ - _ _ - _ - _ _ _ _ - _ _ - - - _ _ _ _ _ - _ _ _ _ _ _ _ . _ - _ _ - _ _ = _ _ _ _ _ _ - _ - _ _ - - j,, UNITED STATES l- [ g NUCLEAR REGULATORY COMMISSION i i, j WASHING TON, D. C. 20555 l I SOUTHERN CALIFORNIA EDISON COMPANY SAN DIEGO GAS AND ELECTRIC COMPANY THE CITY OF RIVERSIDE, CALIFORNIA THE CITY OF ANAHEIM, CALIFORNIA DOCKET NO. 50-362 SAN ONOFRE NUCLEAR GENERATING STATION, UNIT 3 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 53 License No. NPF-15 1. The Nuclear Regulatory Comission (the Comission) has found that: A. The applications for amendment to the license for San Onofre Nuclear Generating Station, Unit 3 (the facility) filed by the Southern California Edison Company (SCE) on behalf of itself and San Diego Gas and Electric Company, The City of Riverside and The City of Anaheim, California (licensees), dated April 27, 1984 and May 12, November 4, and December 14, 1987 (as supplemented April 14 and May 6, 1988), comply with the standards and req)uirements of the Atomic Energy Act of 1954, as amended (the Act and the Comission's regulations as set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the applications, as amended, the provisions of the Act, and the regulations of the Comission; C. There is reasonable assurance: (1)thattheactivitiesauthorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations set forth in 10 CFR Chapter I; D. The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Comission's regulations and all applicable requirements have been satisfied. ~

j ] o- -.. u 2. Accordingly,"the license is amended by changes to the Technical i Specifications as indicated-in the attachment to this amendment and 1 Paragraph 2.C(2) of Facility Operating License No. NPF-15 is hereby amended to read as follows: (2) Technical Specifications The Technical Specifications contained in Appendix A and the i Environmental Protection Plan contained in Appendix B, as revised i through Amendment No. 53, are hereby incorporated in the license. SCE shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan. 3. The changes in Technical Specifications are.to become effective within~30 days of issuance of the amendment..In the period between issuance of the amendment and the effective date of the new Technical Specifications, the licensees shall adhere to the Technical Specific'ations existing at the time. The period of time during changeover shall be minimized. 4. This license amendment is effective as of the date of its issuance. FOR THE NUCLEAR REGULATORY COMMISSION eorge W Knighton, oject Director Project irectorate V. Division of Reactor Projects - III, IV, V and Special Projects

Attachment:

Changes to the Technical Specifications Date of Issuance: August 3, 1988

3

,c b -

ATTACHMENT TO LICENSE AMENDMENT NO. 53 ~ ' ' FACILITY OPERATING LICENSE NO. NPF-15 DOCKET NO. 50-362 Replace the following pages of.the Appendix A. Technical Specifications with the enclosed pages. -The revised pages are identified by Amendment number and contain vertical lines indicating the area of change. Also to be replaced are.the following overleafLpages to the amended pages. Amendment-Page Overleaf Page v vi xi-xii 2-4 2-3 3/4 1-22 3/4 1-21 3/4 1-23 -3/4 1 -- 3/4_ 1-25 3/4 1-26 3/4 3-3 3/4 3-4 3/4 3-6 3/4 3-5 3/4 '3-28 3/4 3-27 3/4'14-37 l 3/4' 4-38 I ~3/4 5-1 3/4 5-2 3/4 10-2 3/4 10-1 B3/4 1-5 B3/4 4-10 B3/4 4-9 1

m. y INDEX LIMITING CONDITION FOR'0PERATION AND SURVEILLANCE-REQUIREMENTS +.- SECTION PAGE HOT SHUT 00WN............................................ 3/4 4 COLD SHUTDOWN - LOOPS FILLED............................ 3/4 4-5 COLD SHUTDOWN - LOOPS NOT FILLED........................ 3/4 4-6 3/4.4.2 SAFETY-VALVES - 0PERATING............................... 3/4 4-7 3/4.4.3 - PRESSURIZER.................,............................ 3/4 4-8 3/4.4.4 STEAM GENERATORS........................................ 3/4 4-9 3/4.4.5 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS............................ 3/4 4-17 OPERATIONAL LEAKAGE.................................. 3/4 4-18 3/4.4.6 CHEMISTRY............................................... 3/4 4-21 3/4.4.7 SPECIFIC ACTIVITY....................................... 3/4 4-24 ~ 3/4.4.8 PRESSURE / TEMPERATURE ~ LIMITS REACTOR C0OLANT SYSTEM............................... 3/4 4-28 PRESSURIZER - HEATUP/C00LDOWN........................ 3/4 4-32 OVERPRESSURE. PROTECTION SYSTEMS RCS TEMPERATURE 3/4 4-33 RCS TEMPERATURE >$ 285'F............................ i 285'F............................ 3/4 4-35 i f 3/4.4.9-STRUCTURAL INTEGRITY.................................... 3/4 4-36 3/4.4.10 REACTOR COOLANT GAS VENT SYSTEM......................... 3/4 4-37 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 SAFETY IKIECTION TANKS.................................. 3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - T,,,> 350*F.......................... 3/4 5-3 ) 3/4.5.3 ECCS SUBSYSTEMS - Tavg <350*F.......................... 3/4 5-7 1 3/4.5.4 REFUELING WATER STORAGE TANK............................ 3/4 5-8 1 J J l i SAN DNOFRE - UNIT 3 V AMENDMENT N0'.53 l F w___ .__.__.__-m_.__-_.-m.mm___m_m_.____m____ _ _ _______ _ _. _.._- _ _ _. -. _ _ _ _ _ _ _ _ _ _ _ _ _ _. _ _ _

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.6 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT CONTAINMENT INTEGRITY................................ 3/4 6-1 C06iAINMENT LEAKAGE.................................. 3/4 6-2 CONTAINMENT AIR LOCKS............... 3/4 6-5 INTERNAL ESSURE.................................... 3/4 6-7 AIR TEMPE 4TURE...................................... 3/4 6-8 CONTAINMENT STRUCTURAL INTEGRITY..................... 3/4 6-9 CONTAINMENT VENTILATION SYSTEM....................... 3/4 6-14 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS CONTAINMENT SPRAY SYSTEM............................. 3/4 6-15 IODINE REMOVAL SYSTEM................................ 3/4 6-17 CONTAINMENT COOLING SYSTCM........................... 3/4 6-18 3/4.6.3 CONTAINMENT ISOLATION VALVES............................ 3/4 6-19 3/4.6.4 COMBUSTIBLE GAS CONTROL HYOR0 GEN MONIT0F.S.................................... 3/4 6-27 ELECTRIC HYDROGEN REC 0MBINERS........................ 3/4 6-28 CONTAINMENT DOME AIR CIRCULAT0RS..................... 3/4 6-29 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE SAFETY VALVES........................................ 3/4 7-1 AUXILIARY FEEDWATER SYSTEM........................... 3/4 7-4 CONDENSATE STORAGE TANKS............................. 3/4 7-6 ACTIVITY............................................. 3/4 7-8 MAIN STEAM LINE ISOLATION VALVES..................... 3/4 7-10 i . SAN ON0FRE-UNIT 3 VI

t V g INDEX iI BASES SECTION g 3/4.4.4 STEAM GENERATORS...................................... B 3/4 4-2 1, 3/4.4.5 REACTOR COOLANT SYSTEM LEAKAGE........................ B 3/4 4-4 3/4.9.E CHEMISTRY............................................. B 3/4 4-4 3/4.4.7 SPECIFIC ACTIVITY..................................... B 3/4 4-5 3/4.4.8 PRESSURE / TEMPERATURE LIMITS........................... B 3/4 4-6 3/4.4.9 STRUCTURAL INTEGRITY.................................. B 3/4 4-10 3/4.4.10 RE ACTOR COC' NT GAS VENT SYSTEM....................... B 3/4 4-10 l 3/4.5 _ EMERGENCY CORE COOLING SYSTEMS I 3/4.5.1 SAFETY INJECTION TANKS................................ B 3/4 5-1 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS.................'.......... B 3/4 5-1 3/4.5.4 REFUELING WATER STORAGE TANK.......................... B 3/4 5-2 3/4.6 CONTAININMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT................................... B 3/4 6-1 _ 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS.................. B 3/4 6-3 ~ 3/4.6.3 CONTAINMENT ISOLATION VALVES.......................... B 3/4 6-4 3/4.6.4 COMBUSTIBLE GAS CONTR0L............................... B 3/4 6-5 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE......................................... B 3/4 7-1 3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITATION....... B 3/4 7-3 3/4.7.3 COMPONENT COOLING WATER SYSTEM........................ B 3/4 7-3 3/4.7.4 SALT WATER COOLING SYSTEM.....,....................... B 3/4 7-3 3/4.7.5 CONTROL ROOM EMERGENCY AIR CLEANUP SYSTEM............. B 3/4 7-4 3/4.7.6 5NUBBERS.............................................. B 3/4 7-5 3/4.7.7 SEALED SOURCE CONTAMINATION........................... B 3/4 7-6 3/4.7.8 FIRE SUPPRESSION SYSTEMS.............................. B 3/4 7-6 3/4,7.9 FIRE RATED ASSEMILIES................................. B 3/4 7-7 SAN ONOFRE - UNIT 3 XI AMENDMENT NO.53

'4: INDEX BASES SECTION' PAGE-3/4.8 ELECTRICAL POWER SYSTEMS 3/4.8.1, 3/4.8.2 and 3/4.8.3 A.C. SOURCES, D.C. SOURCES and ONSITE POWER DISTRIBUTION SYSTEMS................ B 3/4 8-1 3/4.8.4 ELECTRICAL EQUIPMENT' PROTECTIVE DEVICES............... B 3/4 8-3 3/4.9 REFUELING OPERATIONS 3/4.9.1 BORON CONCENTRATION................................... B 3/4 9-1 3/4.9.2 INSTRUMENTATION....................................... B 3/4 9-1 3/4.9.3 DECAY TIME............................................ B 3/4 9-1 1 3/4.9.4 CONTAINMENT BUILDING PENETRATIONS..................... B 3/4 9-1 3/4.9.5 COMMUNICATIONS........................................ B 3/4 9-1 j 3/4.9.6 REFUELING MACHINE..................................... B 3/4 9-2 1 3/4.9.7 FUEL HANDLING MACHINE - SPENT FUEL STOR' AGE BUILDING... B 3/4 9-2 3/4.9.8 SHUTDOWN COOLING AND COOLANT CIRCULATION.............. B 3/4 9-2 i 3/4.9.9 CONTAINMENT PURGE ISOLATION SYSTEM.................... B 3/4 9-3 1 1 1 3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL and J J STORAGE POOL......................................... B 3/4 9-3 3/4.9.12 FUEL HANDLING BUILDING POST-ACCIDENT. CLEANUP FILTER i SYSTEM.............................................. B 3/4 9-3 J 3/4.10 SPECIAL TEST EXCEPTIONS 3/4.10.1 SHUTDOWN MARGIN....................................... B 3/4 10-1 3/4.10.2 GROUP HEIGHT, INSERTION AND POWER DISTRIBUTION LIMITS........................... B 3/4 10-1 l i 3/4.10.3 REACTOR COOLANT L00PS................................. B 3/4 10-1 l' 3/4.10.4 CENTER CEA MISALIGNMENT............................... B 3/4 10-1 SAN ONOFRE-UNIT 3 XII

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REACTIVITY CONTROL SYSTEMS SHUT 00WNCEAINSERiIONLIMIT I ~. - LIMITING CONDITION FOR OPERATION l l 3.1.3.5 All shutdown CEAs shall be withdrawn to greater than or equal to 145 inches. APPLICABILITY' MODES 1 and 2*f. ACTION: With a maximum of one shutdown CEA withdrawn to less than 145 inches, except for surveillance testing pursuant to Specification 4.1.3.1.2, within one hour either: Withdraw the CEA to greater than or equal to 145 inches, or a. b. Declare the CEA inoperable and apply Specification 3.1.3.1. SURVEILLANCE REQUIREMENTS 4.1.3.5 Each shutdown CEA shall be determined to be withdrawn to greater than or equal to 145 inches: Within 15 minutes prior to withdrawal of any CEAs in regulating a. groups during an approach to reactor criticality, and b. At least once per 12 hours thereafter. "See Special Test Exception 3.10.2.

  1. With K,ff greater than or equal to 1.0.

SAN ONOFRE-UNIT 3 3/4 1-21

,r3, ' REACTIVITY CONTROL SYSTEMS REGULATING CEA INSERTION LIMITS-LIMITING CONDITION FOR OPERATION 3.1.3.6-When COLSS is in-service,:the regulating CEA groups shall be limited a. to the withdrawal. sequence and to.the insertian limits shown on Figure'3.1-2. The CEA insertion between the Long Term Steady State Insertion Limits and the Transient. Insertion Limits is restricted to: 1. Less than:or equal to 4 hours per 24. hour interval, 2. Less than or equal to 5 Effective Full Power Days per 30 Effec-tive' Full Power Day interval, and 3. Less than or equal to 14 Effective Full Power Days per 365-Effective Full Power Day interval, a. When COLSS is out-of service, the regulating CEA groups shall be limited-to the Short Term Steady State Insertion Limit shown on 1 Figure 3.1-2. The CEA insertion between the'Long Term Steady State Insertion Limits and the Short Term Steady State Insertion Limits is restricted to: 1. Less than or equal to 4 hours per 24 hour interval, 2. Less than or equal to 5 Effective Full Power Days per 30 Effec-tive Full Power Day, interval, and 3. Less than or equal to 14 Effective Full Power Days per 365 Effective' Full Power Day interval. APPLICABILITY: MODES 1* and 2*#. ACTION: When COLSS is in service and a. With the regulating CEA groups inserted beyond the Transient Insertion Limits, except for surveillance testing pursuant to Specification 4.1.3.1.2, within two hours either: 1. Restore the regulating CEA groups to within the limits, or 2. Reduce THERMAL POWER to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the CEA group position using the above figure.

  • See Special Test Exceptions 3.10.2 and 3.10.4.
  1. With K,ff greater than or equal to 1.0.

9 SAN ONOFRE-UNIT 3 3/4 1-22 AMENDMENT NO. 53

ve; hEACTIVITYCONTROLSYSTEMS ACTION': (Continued) b. With the regulating CEA groups inserted between the Long Term Steady State Insertion Limits and the Transient Insertion Limits for inter-t valspeater than 4 hours per 24 hour interval, operation may nroceed provided either: 1. The Short Tem Steady State Insertion Limits of Figure 3.1-2 are not exceeded, or 2. Any subsequent increase in THERMAL POWER is restricted to less than or equal to 5% of RATED THERMAL POWER per hour. With the regulating CEA groups inserted between the Long Ters Steady c. State Insertion Limits and the Transient Insertion Limits for intervals greater than 5 EFPD per 30 EFPD interval or greater than 14 EFPD per 365 EFPD interval, either: l 1. Restore the regulating groups to within the Long Term Steady State Insertion Limits within two hours, or 2. Be in at least HOT STANDBY within 6 hours. When COLSS is out of service and the regulatiing CEA groups are inserted beyond' the Short Ters Steady State Insertion Limit except for surveillance testing pursuant to Specification 4.1.3.1.2, within two hours either: Restore the regulating CEA group to within the limit, or a. b. Reduce thermal power to less than or equal to that fraction of Rated Thermal Power which is allowed by the CEA group position and the Short Ters Steady State Insertion Limit. SURVEILLANCE REQUIREMENTS 4.1.3.6 The position of each regulating CEA group shall be determined to be within the Transient Insertion Limits at least once per 12 hours except during time intervals when the PDIL Auctioneer Alam Circuit is inoperable, then verify the individual CEA positions at least once per 4 hours. The accumulated times during which the regulating CEA groups are inserted beyond the Long Ters Steady State Insertion Limits but within the Transient Insertion Limits shall be determined at least once per 24 hours. SAN ONOFRE-UNIT 3 3/4 1-23 AMENDMENT NO. 53 l

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REACTIVITY CONTROL SYSTEMS PART LENGTH CEA INSERTION LIMITS LIMITING CONDITION FOR OPERATION t 3.1.3.7 The part length CEA group shall be limited to the insertion limits shown on Figure 3.1-3 with PLCEA insertion between the Long Term Steady State Insertion Limit and the Transient Insertion Limit restricted to: < 7 Effective Full Power Days per 30 Effective Full Power Day a. interval, and b. < 14 Effective Full Power Days per 365 Effective Full Power Day Tnterval. APPLICABILITY: MODE I above 20% of RATED THERMAL POWER

  • 3

) ACTION: 1 With the part length CEA groups inserted beyond the Transient Inser-a. tion Limit, except for surveillance testing pursuant to specifica-tion 4.1.3.1.2, within two hours either: 1 1. Restore the part length CEA groups to within the limit, or 2. Reduce THERMAL POWER to less than or equal to that fraction of RATED THERMAL POWER which is allowed by the PLCEA group posi-tion using Figure 3.1-3. b. With the part length CEA groups inserted between the Long Term Steady State -Insertion Lir.it and the Transient Insertion Limit for intervals > 7 EFPD per 30 EFPD interval or > 14 EFPD per 365 EFPD interval, either: 1. Restore the part length groups to within the Long Ters Steady ) State Insertion Limit within two hours, or 2. Reduce THERMAL POWER to less than or aqual to 20% of RATED THERMAL POWER withth the next 4 hours. SURVEILLANCE REQUIREMENTS 4.1.3.7 The position of the part length CEA group shs11 be determined to be within the Transient Insertion Limit at least once per 12 hours. The accumu-lated time during which the part length CEA groups are inserted beyond the Long Ters Steady State Insertion Limit but within the Transient Insertion Limit shall be determined at least once per 24 hours.

  • See Special Test Exception 3.10.2.

SAN ONOFRE-UNIT 3 3/4 1-25 AMENDMENT NO. 53

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9 TABLE 3.3-1 (Continued) j TABLE NOTATION With the protective system trip breakers in the closed position, the CEA drive system capable of CEA withdrawal, and fuel in the reactor vessel.

  1. The provisions of Specification 3.0.4 are not applicable.

~4 (a) Trip may bc manually bypassed above 10 % of RATED THERMAL POWER; bypass shallg%ofRATEDTHERMALPOWER.e automatically removed when THERMAL POW to 10 (b) Trip rev '; manually bypassed below 400 psia; bypass shall be automatically removed whenever pressurizer pressure is greater than or equal to 400 psia. (c) Trip may be manually bypassed below 10 % of RATED THERMAL POWER; bypass shall_g%ofRATEDTHERMALPOWER.e automatically removed when THERMAL POW to 10 During testing pursuant to Special Test Exception 3.10.2 or 3.10.3, trip may be manually bypassed below 5% of RATED THERMAL POWER; bypass shall be automatically removed when THERMAL POWER is greater than or equal to 5% of RATED THERMAL POWER. (d) Trip may be bypassed during testing pursuant to Special Test Exception 3.10.3. (e) See Special Test Exception 3.10.2. (f) Each channel shall be comprised of two trip breakers; actual trip logic shall be one-out-of-two taken twice. (g) Trip may be bypassed below 55% RATED THERMAL POWER. ACTION STATEMENTS ACTION 1 With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoper-able channel to OPERABLE status within 48 hours or be in at least HOT STANDBY within the next 6 hours and/or open the pro-tective system trip breakers. ) ACTION 2 With the number of channels OPERABLE one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may continue provided the inoperable channel is placed the bypassed or tripped condition within 1 hour. If the inoperable channel is l bypassed, the desirability of maintaining this channel in the I bypassed condition shall be reviewed in accordance with Specifi-cation 6.5.1.6e. The channel shall be returned to OPERABLE status no later than during the next COLD SHUTDOWN. 4 SAN ON0FRE - UNIT 3 3/4 3-4 AMENDMENT NO. 53

' ' l ' J. TABLE 3.3-1 (Continued ACTION STATEMENTS With a channel process' measurement circuit that affects multiple functional units inoperable or in test, bypass or trip all associated functional units as listed below: Process Measurement Circuit Functional Unit' Bypassed 1. Linear Power Linear Power Level - High (Subchannel or Linear) Local Power Density - High DNBR Low 2. Pressurizer Pressure - High Pressurizer Pressure - High Local Power Density - High DNBR - Low 3. Containment Pressure - High Containment Pressure - High (RPS) Containment Pressure - High (ESF) 4. Steam Generator Pressure - Steam Generator Pressure - Low Low Steam Generator AP 1 and 2 (EFAS 1 and 2) 5. Steam Generator Level Steam Generator Level - Low Steam Generator Level - High Steam Generator AP (EFAS) 6. Core Protection Calculator Local Power Density - High DNBR - Low ACTION 3 - With the number of channels OPERABLE one less than the Minimum Channels OPERABLE requirement, STARTUP and/or POWER OPERATION may continue provided the following conditions are satisfied: a. Verify that one of the inoperable channels has been bypassed and place the other channel in the tripped condi, tion within I hour, and b. All functional units affected by the bypassed / tripped channel shall also be placed in the bypassed / tripped condition as listed below: Process Measurement Circuit Functional Unit Bypassed / Tripped 1. Linear Power Linear Power Level - High (Subchannel or Linear) Local Power Density - High DNBR - Low 1 SAN ONOFRE - UNIT 3 3/4 3-5 1

6 TABLE 3.3-1 (Continued) ACTION STATEMENTS 2. Pressurizer Pressure - Pressurizer Pressure - High High Local Power Density - High DNBR - Low 3. Containment Pressure - Containment Pressure - High (RPS) High Containment Pressure - High (ESF) 4. Steam Generator Steam Generator Pressure - Low Pressure - Low Steam Generator AP 1 and 2 (EFAS 1 and 2) 5. Steam Generator Level Steam Generator Level - Low Steam Generator Level - High Steam Generator AP (EFAS) 6. Core Protection Local Power Density - High Calculator DNBR - Low STARTUP and/or POWER OPERATION may continue until the perfomance of the next required CHANNEL FUNCTIONAL TEST. Subsequent STARTUP and/or POWER OPERATION may continue if one channel is restored to OPERABLE status and the provisions of ACTION 2 are satisfied. ACTION 4 With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, suspend all operations involving positive reactivity changes. ACTION 5 With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, be in' at least HOT STANDBY within 6 hours. ACTION 6 With one CEAC inoperable, operation may continue for up to a. 7 days provided that at least once per 4 hours, each CEA is verified to be within 7 inches (indicated position) of all other CEA's in its group. After 7 days, operation may continue provided that ACTION 6.b is met.* If the exemption to Specification 3.0.4 is used, Action 6.b must be met. b. With both CEACs inoperable, operation may coninue provided that:* 1. Within 1 hour the DNBR margin reqired by Specification 3.2.4.b (COLSS in service) or Specification 3.2.4.d (COLSS out of service) is satisfied.

  • Note:

Requirements for CEA position indication given in Technical Specification 3.1.3.2. SAN ON0FRE - UNIT 3 3/4 3-6 AMENDMENT NO. 53

TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME (SEC) 1.- Manual ~ a. SIAS i Safety Injection Not Applicable. { Control Room Isolation Not Applicable i Containment Isolation (3) Not Applicable ') Containment Emergency Cooling Not Applicable b. CSAS i Containment Spray Not Applicable c. CIAS Containment Isolation Not Applicable d. NSIS Nain Steam-Isolation Not Applicable e. RAS Containment Sump Recirculation Not Applicable f. CCAS Containment Emergency Cooling Not Applicable g. EFAS Auxiliary Feedwater Not Applicable h. CRIS Control Room Isolation Not Applicable 1. TGIS Toxic Gas Isolation Not Applicable j. FHIS Fuel Nandling Building Isolation Not Applicable k. CPIS Containment Purge Isolation Not Applicable SAN DNOFRE - UNIT 3 3/4 3-27

p h Table 3.3-5'(continued) ' INITIATING SIGNAL AND FUNCTION RESPONSE TIME (SEC) 2. Pressurizer Pressure-Low SIAS (1) Safety Injection .(a) High Pressure Safety Injection 31.2* (b) Low Pressure Safety Injection 41.2* (c) Charging Pumps 31.2* ( (2) Control Room Isolation Not Applicable (3) Containment-Isolation (NOTE 3) 11.2* (NOTE 2) (4) Containment Spray (Pumps) 25.6* .(5) Containment Emergency Cooling -(a) CCW Pumps 31.2* (b) CCW Valves (NOTE 4b) 23.2* (c) Emergency Cooling Fans 21.2* 4 3. Containment Pressure-High a. SIAS (1) Safety Injection (a)' High Pressure Safety Injection 41.0* (b) Low Pressure Safety Injection 41.0* (2) Control-Room Isolation Not Applicable I (3) Containment Spray (Pumps) 25.4* (4) Containment Emergency Cooling i (a) CCW Pumps 31.0* (b) CCW \\/alves (NOTE 4b) 23.0* (c) Emergency Cooling Fans 21.0* b. CIAS (1) Containment Isolation 10.9* (NOTE 2) (2) Main Feedwater Backup Isolation 10.9 and Backup Isolation Valves (HV 4048, HV 4052, HV 1105, HV 1106, HV 4047, .HV 4051) (3) CCW Valves (Note 4a) 20.9 (4) Mainsteam Isolatior! Valves (HV 8204, 8.9 HV 8205) (5) Minipurge Isolation Valves

5. 9 4.

Containment Pressure - High-High CSAS Containment Spray 23.0* SAN ONOFRE - UNIT 3 3/4 3-28 AMENDMENT NO. 53

<; 4* . REACTOR COOLANT SYSTEM-3/4.4.10 REACTOR COOLANT GAS VENT SYSTEM LIMITING CONDITION FOR OPERATION i 3.4.10 The Reactor Coolant Gas Vent System shall be OPERABLE with: At least one of valves 3HV0296A or 3HV0296B capable of being powered a. from an emergency bus and providing a vent path from the reactor vessel head; and, b. At least one of valves 3HV0297A or 3HV02978 capable of being powered from an emergency bus and providing a vent path from the pressurizer steam space; and, 9 c. .At least one of valves 3HV0298, capable of being powered from an emergency bus and providing a vent path to the containment atmosphere, ~ or 3HV0299, capable of-being powered from an emergency bus and pro-viding a vent path to the quench tank; and d. Valves 3HV0296A, 3HV0296B, 3HV0297A, 3HV02978, 3HV0299 and 3HV0298 all closed. 1 APPLICABILITY: MODES 1, 2, 3 and 4 { ACTION: a. With any of valves 3HV0296A, 3HV0296B, 3HV0297A, or 3HV02978 inoperable, operation may continue provided that: 1 i) power is removed from the inoperable valve (s) within 4 hours-

and,

] ii) valves 3HV0299 and 3HV0298 are maintained closed and power is removed within 4 hours; and, q iii) the inoperable valve (s) is restered to OPERABLE status during the next COLD SHUTDOWN. 1 l b. With any of valves 3HV0299 or 3HV0298 inoperable, operation may continue provided that: 1) power is removed from the inoperable valve (s) within 4 hours;

and,
11) valves 3HV0296A, 3HV02968, 3HV0297A and 3HV02978 are all

) maintained closed and power is removed within 4 hoursDnd i SAN ONOFRE - UNIT 3 3/4 4-37 AMENDHENT NO.53

e 'l d REACTOR COOLANT ~ SYSTEM LIMITING CONDITION FOR OPERATION iii) the inoperable valve (s) is restored'to OPERABLE status during the next COLD SHUTDOWN. The provisions of 3.0.4 are not. applicable for entry into MODEE 3, 2 c. and 1. SURVEILLANCE' REQUIREMENTS 4.4.10 Each reactor coolant system vent path shall be demonstrated OPERABLE at least once per 18 months by: 1. Verifying all manual isolation valves in each vent path are locked in the open position. 2. Cycling each valve'in the vent path through at least one complete i cycle of full travel from the control room during COLD SHUTDOWN or j REFUELING. 3. Verifying flow through the reactor coolant vent system vent paths a during venting during COLD SHUTDOWN. SAN ONOFRE - UNIT 3 3/4 4-38 AMEN 0 MENT NO. 53 l 1 1

3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 SAFETY INJECTION TANKS LIMITING CONDITIOW FOR OPERATION '3.5.1 Each reactor coolant system safety injection tank shall be OPERABLE with: The isolation valve open and power to the valve removed. a. b. A contained borated water volume of between 1680 and 1807 cubic feet. g Between 1850 and 2800 ppm of boron, and c. d. A nitrogen cover pressure of between 615 and 655 psia. APPLICABILITY: MODES 1, 2 and 3.* ACTION: With one safety injection tank inoperable, except as a result of a a. closed isolation valve, restore the inoperable tank to OPERABLE status within one hour or be in at least HOT STANDBY within the next-6 hours and in HOT SHUTDOWN within the following 6 hours. ~ b. With one safety injection tank inoperable due to the isolation valve being closed, either immediately open the isolation valve or be in at least HOT STANDBY within one hour and be in HOT SHUTDOWN within the next 12 hours. SURVEILLANCE REQUIREMENTS 4.5.1 Each safety injection tank shall be demonstrated OPERABLE: a. At least once per 12 hours by: 1. Verifying that the contained borated water volume and nitrogen cover pressure in the tanks is within the above limits, and 2. Verifying that each safety injection tank isolation valve is open.

  • With pressurizer pressure greater than or equal to 715 psia.

SAN ONOFRE - UNIT 3 3/4 5-1 AMENDMENT NO. 53

L EMERGENCY CORE COOLING SYSTEMS i SURVEILLANCE REQUIREMENTS (Continued) b. At least once per 31 days and within. 6 ' hours after each solution ~ volume increase of greater than or equal to 1% of tank volume by verifying the boren concentration of the safety injection tank solution. . At least once per 31 days by verifying the fuses removed from each c. -safety-injection tank vent valve. d. At least once per 31 days when the RCS pressure is above 715 psia, by verifying that the isolation valve operator breakers are padlocked in the open position. At least once per 18 months by verifying that each safety injection e. tank isolation varve opens automatically under each of the following conditions: 1. Before an actual or simulated RCS pressure signal exceeds 715 psia, and 2. Upon receipt of an SIAS test signal. ( j l ) I ) I' o': i l SAN ONOFRE-UNIT 3 1/8 4-2 I

.c 3/4.10 SPECIAL TEST EXCEPTIONS l 3/4.10.1 SHUTDOWN MARGIN ] s 1 LIMITING CONDITION FOR OPERATION ) -3.10.1 The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 may be suspended for sensurement of CEA worth and shutdown margin provided reactivity equivalent to at least the highest estimated CEA worth is available for trip insertion from OPERABLE CEA(s). APPLICABILITY: MODES 2 and 3*. ACTION: a. With any full length CEA not fully inserted and with less than the above reactivity equivalent available for trip insertion, immedi-ately initiate and continue boration at greater than or equal to 40 gpm of a solution containing greater than or equal to 2350 ppm l boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored. b. With all full length CEAs fully inserted and the reactor suberitical. by less than the above reactivity equivalent, immediately initiate and continue boration at greater than or equal to 40 gpa of a solution containing greater than or equal to 2350 ppe boron or its equivalent until the SHUTDOWN MARGIN required by Specification 3.1 is restored. SURVEILLANCE REQUIREMENTS 4.10.1.1 The position of each full length and part length CEA required either partially or fully withdrawn shall be detemined at least once per.2 hours. 4.10.1.2 Each CEA not fully inserted shall be demonstrated capable of full insertion when tripped from at least the 50% withdrawn position within 7 days j prior to reducing the SHUTDOWN MARGIN to less than the limits of Specification 3.1.1.1. " Operation in DODE 3 shall be limited to 6 consecutive hours. SAN ONOFRE - UNIT 3 3/4 10-1 AMENDMENT NO. 50 l l

r SPECIAL TEST EXCEPTIONS 13/4.10.2 GROUP HEIGHT,' INSERTION AND POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION 3.10.2 The moderator temperature coefficient, group height, insertion, and-power distribution limits of Specifications 3.1.1.3, 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.1.3.7, 3.2.2, 3.2.3, 3.2.7,- footnote C-of Table 3.3-1 and footnote 5 of Table 2.2-1, and the Minimum Channels OPERABLE requirement of Functional Unit 15 of Table 3.3-1 may be suspended during the performance of. PHYSICS TESTS provided: 1 The THERMAL POWER is. restricted to the. test power plateau which a. shall not exceed 85% of RATED THERMAL POWER, and b. The limits of Specification 3.2.1 are maintained and determined as specified in Specification 4.10.2.2 below. APPLICABILITY: MODES 1 and 2. ACTION: With any of the limits of Specification 3.2.1 being exceeded while the require-ments of Specifications 3.1.1.3, 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.1.3.7, 3.2.2, . 3.2.3, 3.2.7, footnote C of Table 3.3-1 and footnote 5 of Table 2.2-1, and the Minimum Channels OPERABLE requirement of Functional Unit 15 of Table 3.3-1 are suspended, either: a. Reduce THERMAL POWER sufficiently to satisfy the requirements of . Specification 3.2.1, or b. Be in HOT STANDBY within 6 hours. SURVEILLANCE REQUIREMENTS 4.10.2.1 The THERMAL POWER shall be determined at least once per hour during PHYSICS TESTS in which the requirements of Specifications 3.1.1.3, 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.1.3.7, 3.2.2, 3.2.3, 3.2.7, footnote C of Table 3.3-1 and footnote 5 of Table 2.2-1, or the Minimum Channels OPERABLE requirement of Functional Unit 15 of Table 3.3-1 are suspended and shall be verified to be I within the test power plateau. 4.10.2.2 The linear heat rate shall be determined to be within the limits of Specification 3.2.1 by monitoring it continuously with the Incore Detector Monitoring System pursuant to the requirements of Specifications 4.2c1.2 and { 3.3.3.2 during PHYSICS TESTS above 5% of RATED THERMAL POWER in which the requirements of Specifications 3.1'.1.3, 3.1.3.1, 3.1.3.5, 3.1.3.6, 3.1.3.7, 3.2.2, 3.2.3, 3.2.7, footnote C of Table 3.3-1 and footnote 5 of Table 2.2-1, or the Minimum Channels OPERABLE requirement of Functional Unit 15 of Table 3.3-1 are suspended. SAN ONOFRE - UNIT 3 3/4 10-2 AMENDMENT N0.53

. +. V. ::

i F

REACTIVITY CONTROL SYSTEMS I BASES MOVABLE CONTROC A'SSEMBLIES (Continued). The establishment of LSSS and tCOs require that the expected long and short term behavior of the radial peaking factors be determined. The long ters behavior relates to the variation of the steady-state radial peaking factors with core burnup and is affected by the amount'of CEA insertion assumed, the portion of a burnup cycle over which such insertion is assumed J and the expected power level variation throughout the cycle. The short term behavior relates to transient perturbations to the. steady-state radial peaks e due to radial xenon redistribution. The magnitudes of such perturbations depend upon the expected use of'the CEAs during anticipated power reductions and load maneuvering. Analyses are performed based on the expected mode of operation of.the NSSS (base load maneuvering, etc.) and from these analyses CEA insertions are determined and a consistent set of radial peaking factors defined. The Long Tern Steady-State and Short Tern Insertion Limits are determined based upon the assumed mode of operation used in the analyses and provide a means of preserving the assumptions on CEA insertions used. The limits specified serve to limit the behavior of the radial peaking factors within the bounds determined from analysis. The actions specified serve to limit the extent of radial xenon redistribution effects to those accommodated ~ in the analyses. The Long and Short Tere Insertion Limits of Specifica-tion 3.1.3.6 are specified for the plant which has been designed for primarily base loaded operation but which has the ability to accommodate a limited amount of load maneuvering. The Transient Insertion Limits of Specification 3.1.3.6 and the Shutdown CEA Insertion ~ Limits of Specification 3.1.3.5 ensure that 1) the minisium SHUTDOWN MARGIN is maintained, and 2) the potential effects of a CEA ejection accident are limited to acceptable levels, Long ters operation at the Transient Insertion Limits is not permitted since such operation could have effects on the core power distribution which could invalidate assumptions used to determine the behavior of the radial peaking factors. The Part Length CEA Insertion Limits of Specification 3.1.3.7 ensure that adverse power shapes and rapid local power changes which affect radial peaking factors and DNB considerations do not occur as a result of a part length CEA group covering the same axial segment of the fuel assemblies for an extended period of time during operation. The CEA fully withdrawn position is defined to be greater than or equal to 145 inches. The extreme limits of CEA travel, fully withdrawn and fully inserted, may be described as the upper electrical limit and lower electrical limit respectively. l SAN ONOFRE-UNIT 3 8 3/4 1-5 AMENDMENT NO.53 4

mfLc il e nerr ih oi MSfD b l fhoc - 0000 AA A t0 4455 00 NN N eo5t rN f u9 tV a ry b e p0 I 55 00 pr3 2211 22 00 0 ma 11 - - 21 1 eh9 TC t f s tt hl 0000 00 00 0 pgu 2222 22 43 4 ) ois - - - - d ree e DWR u EE n EEEE // i //// SS t SSSS n gg o gggg nn C nnnn ii ( iiii gg gggg rr ll e 1 rrrr oo ee m n oooo FF ee o 4 o FFFF PP D i ee 4 t eeee ll dd d / a l l ll zz aa a 3 c zzzz zz ee e o zzzz oo HH H 8 L oooo NN NNNN ee e E l tt rr r L e ttt't ee uu u 8 s eeee ll ss s A s l l l l tt oo o T e nnnn uu l l l V IIII OO CC C 11 1 LL L l CC C a 111I 11 BB B i LLLL LL RR R r CCCC CC GG G e 8888 08 33 3 t 0000 00 33 3 a 5555 55 55 5 M AAAA AA AA A 1234 12 11 1 .o - - - - M 1111 22 34 5 3333 33 33 3 e 8888 88 88 8 d 6666 66 66 6 e o - - - - l C CCCC CC CC C ba l iav A .o t N o 3333 77 11 2 N e 0000 00 00 0 c - - - - = e 5555 55 11 1 0000 00 33 3 A i P 2222 22 22 2 M ,E,z mE d z 4 w ,R[a

i c REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) The OPERABILITY of the Shutdown Cooling System relief valve or an RCS vent opening of greater than 5.6 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs is less than or equal to 285 F. The Shutdown Cooling System relief valve has adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either (1) the start of an idle RCP with the secondary water temperature of the steam generator less than or equal to 100 F above the RCS cold leg temperatures or (2) inadvertent safety injection actuation with two HPSI pumps injecting into a water-solid RCS with full charging capacity and letdown isolated. 3/4.4.9 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASME Code Class 1, 2 and 3 components ensure that the structural integrity and operational readiness of these components will be maintained at an acceptable level throughout the life of the plant. These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50.55a(g) except where specific written relief has been granted by the Commission pursuant to 10 CFR Part 50.55a (g) (6) (i). Components of the Reactor Coolant System were des'igned to provide access to permit inservice inspections in accordance with Section XI of the ASME Boiler and Pressure Vessel Code, 1974 Edition and Addenda through Summer 1975. 3/4.4.10 REACTOR COOLANT GAS VENT SYSTEM Reactor coolant system gas vents are provided to exhaust noncondensible gases from the primary system that could inhibit natural circulation core cooling following a non-design bases accident. The OPERABILITY of at least one reactor coolant system vent path from the reactor vessel head and the pressurizer steam space ensures the capability exists to perform this function. The design redundancy of the Reactor Coolant Gas Vent System serves to minimize the probability of inadvertent or irreversible actuation while ensuring that a single failure of a vent valve, or control system does not prevent isolation of the vent path. The function, capabilities, and testing requirements of the Reactor Coolant Gas Vent System are consistent with the requirements of Item II.b.1 of NUREG-0737, " Clarification of TMI Action Plan Requirements," November 1980. SAN ON0FRE-UNIT 3 8 3/4 4-10 AMENDMENT NO. 53

a ara uk / UNITED STATES l P NUCLEAR REGULATORY COMMISSION i W ASHINGTON, D. C,20555 4,e SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION SUPPORTING AMENDMENT NO. 64 TO FACILITY OPERATING LICENSE NO. NFF-10 ANDEMINDMENTNO.53TOFACILITYOPERATINGLICENSEMO.NPF-15 SOUTHERN CALIFORNIA EDISON COMPANY, ET AL. SAN ONOFRE NUCLEAR GENERATING STATION, UNITS 2 & 3 DOCKET NOS. 50-361 AND 50-362

1.0 INTRODUCTION

Southern California Edison Company (SCE), on behalf of itself and the i other licensees, San Diego Gas and Electric Company, The City of Riverside, California, and The City of Anaheim, California, has submitted a number of applications for license amendments for San Onofre i Nuclear Generating Station (SONGS), Units 2 and 3. The NRC staff's evaluation of six of these applications (referred to as Proposed Change,, Numbers 197, 231, 233, 234, 239 and 155) is described below. 2.0 DISCUSSION PCN-197 By letter dated May 12, 1987 the licensees submitted proposed change PCN-197 to the SONGS Units 2 and 3 Technical Specifications. The proposed changes would revise Technical Specification 3.1.3.6, " Regulating CEA Insertion Limits," 3.1.3.7, "Part Length CEA Insertion Limits," and 3.10.2, " Group Height, Insertion and Power Distribution Limits," as well as Bases 3/4.1.3, " Movable Control Assemblies." Technical Specification 3.1.3.6 currently provides restrictions on control element assembly (CEA) insertion limits to periods less than or equal to 14 effective full power days (EFPD) per calendar year. The proposed change would replace " calendar year" with "365 EFPD interval." Since calendar year may be interpreted as the )eriod from January ~ through December, which is not the intent of tie specified interval, the proposed change provides greater clarification of the intended interval restriction. The staff finds this nomenclature change acceptable. The part length CEA insertion limits of Technical Specification 3.1.3.7 are intended to ensure that adverse power shapes and rapid local power changes which affect radial peaking factors and departure from nucleate boiling (DNB) considerations do not occur as a result of a part length -_ % o v i n, y~U m -~ f y u(p u v t v - I

fl .(, CEA group covering the same axial segment of the fuel assemblies for an extended period of time during cperation. However, the Specification does not clearly specify the allowable duration within the transient ' insertion limit nor does it clearly address operation within the lono term stealfyf state insertion limit. The long term s_teady state limit'is-l. based on the expected variation of the steady state radial peaking factors with burnup...It serves to limit the behavior of the radial-peaking factors within acceptable bounds determined from analyses. The transient: insertion limit aids in ensuring'that.the minimum shutdown margin is maintained and that the potential effects of a'CEA ejection accident are limited to acceptable levels. Long term operation at the transient insertion limit is not permitted since such operation could have effects on the core power distribution which could invalidate assumptions used to determine the behavior of the radial peaking factors. The proposed change would restrict the part length CEA group l. insertion to the insertion limits of Figure 3.1-3 with insertion between the long-term steady state insertion limit and the trar>sient insertion limit restricted to. intervals less than or equal to 7 EFPD per 30 EFPD interval and intervals less than or equal to 14 EFPD per 365 EFPD interval. This is acceptable since it clarifies the intent of Specification 3.1.3.7 to apply the 7 EFPD per 30 EFPD interval only to insertions beyond the long term steady state insertion limit. The proposed change would also revise the actions to be taken if part- - length CEA groups are inserted beyond the transient insertion limit or between the long term steady state insertion limit and the transient insertion limit for excessive periods of time. The staff finds these revised Action Statements acceptable. The applicability of Specification 3.1.3.7 would be revised from Modes 1 and 2 to Mode 1 above 20% of rated power. The staff finds this proposed change acceptable since plant operation is limited to no greater than 5% of rated power.in Mode 2, and below 20% of rated power part length CEA insertion is unrestricted. In addition the proposed change to Specification 4.1.3.7 would revise the Surveillance Requirement so that part length CEA group position is determined every 12 hours to be within the transient insertion limit and the accumulated time for insertion beyond the long term steady state insertion limit but within the transient insertion limit is determined every 24 hours. The staff considers these surveillance intervals ) adequate to allow the appropriate proposed actions to be taken in sufficient time. The proposed Surveillance Requirement changes are, therefore, acceptable. The proposed change would also revise Specification 3.10.2 to allow suspension of the insertion limits of Specification 3.1.3.7 during special physics tests. This is acceptable since the addition of power dependent insertion limits for the part length CEAs makes Specification 3.1.3.7 similar to Specification 3.1.3.6 which has previously been included as a special test exception in Specification 3.10.2.

[p i r > \\ l ~ Finally, the proposed change would revise' the Bases to Specification-f 3/4.1.3:to clarify the extreme limits of CEA travel (fully withdrawn and fullycinserted CEA positions).~ The terms " upper electrical limit" and " lower elect.rical-limit" are used to describe the fully withdrawn and fully inserted CEA position.. Furthermore, the withdrawn position would be defined as greater than or equal to 145 inches. This change is ~ acceptable since it is administrative in nature and provides clarification in terminology.which should avoid misunderstandings. PCN-231 l By-letter dated December 14, 1987 the licensees submitted proposed change J PCN-231 which'would revise Figure 3.1-2 of Technical Specification 3/4.1.3.6[CEA)gulatingCEAInsertionLimits"torelaxthecontrolelement "Re assembly powerdependentinsertionlimits(PDIL)atpowerlevelsof I 25% of rated thermal power or less. The licensee has requested this '{ change.in order to preclude the need to borate prior to startup to assure i that the estimated critical CEA position is within the zero power CEA ) insertion limits.- This will also help reduce the amount of waste water l generated during startup. 1 ~ The proposed change would allow more CEAs to be inserted at power levels - i below 25% of rated thermal power. This results in higher reactivity insertion rates in the event of a CEA initiated reactivity accident. The greater CEA insertion at low power also results in a decrease in shutdown margin. Therefore, the licensee reevaluated the three limiting events that are affected by individual CEA worth and required shutdown margin. These are the.CEA withdrawal at low power, the CEA ejection at zero power, and the steam line; break at zero power. The revised PDIL results in a maximum reactivity insertion rate of 1.7.x 10~4 delt 1.1 x 10~g k/k/sec during a CEA withdrawal event at low power compared to delta k/k/sec reported in the Cycle 3 reload analysis report. However, for the Cycle 3 relead, the licensee performed a parametric study on reactivity insertion rate in order to maximize peak reactor coolant system pressyre. This resulted in an intermediate insertion rate (less than'1.1 x 10" delta k/k/sec) producing the most adverse results. l The licensee has verified that this remains true even for the requested PDIL revision and, therefore, the' previous Cycle 3 analysis for the CEA withdrawal event remains bounding. The CEA ejection event at zero power is initiated by a higher ejected CEA vtorth than for Cycle 3 due to tie revised PDIL. However, the Cycle 1 reference analysis allowed greater CEA insertion at zero power and, therefore, an even higher ejected CEA worth. The CEA ejection event with the proposed revised PDIL is, therefore, bounded by the reference analysis of Cycle 1. 1he most restrictive shutdown margin requirement is based on the postulated steam line break event at no load operating temperature and the resulting I

=; hY [.; ,p 4 uncontrolled reactor coolant system cooldown. The reference analysis (Cycle 1) assumed a shutdown margin of 5.15% delta k/k for this event. The proposed revised PDIL results in a greater calculated shutdown margin than 5.15%, delta k/k and, thus, is bounded by the reference analysis. In sumary, although the proposed change allows' more CEA insertion at lower power relative to current Cycle 3 allowances, it is still less than was allowed for Cycle 1. A reevaluation of the limiting events which are adversely affected by greater CEA insertion have shown that the previous reference analysis results are still bounding.' The proposed relaxation of the PDIL-at 25% of rated thermal power and below is therefore acceptable. ! 6 PCN-233 By letter dated December 14, 1987 the licensees submitted proposed change PCN-233 which would revise Technical Specification 3/4.10.2, " Group Height, Insertion and Power Distribution Limits," and. Tables 2.2-1 and 3.3-1 for the San Onofre Nuclear Generating Station (SONGS) Units 2 and 3. The first part of_the proposed change would reference footnote 5 of Table 2.2-1, " Reactor Protective Instrumentation Trip Setpoint Limits," and footnote C of Table 3.3-1, " Reactor Protective Instrumentation," in the body of Special Test Exception 3.10.2. This part of the change would., also modify (1) footnote 5 of Table 2.2-1 to indicate that the bypass setpoint for core protection calculator (CPC) generated trips may be changed during testing pursuant to Special Test Exception 3.10.2, and (2) footnote C of Table 3.3-1 to indicate thr.t the trip may be manually bypassed below 5% of rated thermal-power during testing pursuant to Special Test Exception 3.10.2. The second part of the proposed change would modify Surveillance 4.10.2.2 to reference Specification 4.21.2. The final change would modify Action 6 of Table 3.3-1 by inserting an exemption to Specification 3.0.4 in the event of inoperability of one or both control element assembly calculators '(CEACs). CEA bank reactivity worth measurements which are performed during low power physics testing routinely generate abnormal CEA configurations which could cause a CPC trip. Therefore the bistable setpoint for these CPC trips (local power density and.DNBR),must be raised during physics testing as currently allowed by Special Test Exception 3.10.3. However, since this specification test exception was developed for partial reactor coolant system flow conditions which are not applicable during normal physics testing, the more appropriate test exception would be 3.10.2. Oneoftheproposedchgnges,therefore,allowsresettingoftheplant trip bistable from 10~ % to 55 power in Special Test Exception 3.10.2 to allow perfomance of the tests that are now accomplished pursuant to Exception 3.10.3. This change does not alter the physics test program l and is acceptable to the staff. Another proposed change would reference Surveillance Specification 4.2.1.2 in Surveillance 4.10.2.2. Since 4.2.1.2' specifies conditions l within which the linear heat rate is to be determined, it is the' i appropriate surveillance to reference. The change, therefore, ensures that the intended surveillance will be performed and is acceptable, j l I i l 1 (

.e-l The final portion of the proposed change would modify Action 6 of Table 3.3-1-by adding a Specification 3.0.4 exemption when one or both CEACs are inoperable. 'An exemption to Specification 3.0.4' allows plant startup in situations where a specified system or component is in an inoperable condition" ~Such exemptions are normally authorized in cases where system or component inoperability would be allowed indefinitely under the provis-ions cf the Action requirements. Since operation of the plant with both CEACs inoperable is currently allowed to continue indefinitely under the conditions of Action 6.b, the staff finds the proposed change to be consist-ent with the. generic intent of the exemption and is therefore acceptable. PCN-234 By letter dated December 14, 1987 the licensees submitted proposed change PCN-234 which would revise Technical Specification 3/4.5.1, " Safety Injection Tanks". The existing Limiting Condition for Operation (LCO) 1 3.5.1.d requires that each reactor coolant system safety injection tank be Operable with a nitrogen cover-pressure of between 600 and 625 psig. This requirement of the LCO ensures that a sufficient volume of borated water will be issnediately forced into the reactor core through the cold legs of the Reactor Coolant System (RCS) in the event that the RCS pressure falls below the pressure of the safety injection tanks (SITS). This surge of water into the core provides the initial cooling mechanism. during large pipe ruptures within the reactor coolant pressure boundary.- The proposed change would revise the required upper limit of the nitrogen cover-pressure from 625 psig to 640 psig. This change would prevent possible violation of the pressure limit in the other SITS due to inleakage from the common fill header when one of the tanks is being filled to maintain its pressure within limits. SCE has evaluated the effect of this higher SIT pressure u of coolant accident (LOCA)pon the previously evaluated large break loss By letters dated April 14 and May 6, 1988 SCE provided the results of this evaluation. The higher pressure would j cause an additional 160 cubic feet of water to flow out the break but j there would still be sufficient water remaining in the SITS after blowdown is complete to fill the downcomer. Also, the higher pressure would result l in core reflodd commencing sooner.' The net effect is to reduce the peak fuel clad temperature. The licensee has determined that a SIT sressure in excess of 1000 psig wo0Td be necessary to cause an unaccepta>1e amount i of flow loss during the blowdown period so that insufficient liquid would remain to fill the downcomer. Therefore, the proposed increase in the upper limit of the nitrogen cover-pressure is acceptable because its effect is bounded by the previous accident analysis. In addition, the proposed change would revise the units of pressure from ounds per square inch gauge (psig) to pounds per square inch absolute 1 p(psia). This will make the units of measurement consistent with other units on the control room panel. This change is accestable because it is administrative in nature and provides consistency wit 1 other pressure measurements displayed in the control room. I

y '3."t Finally, the proposed change would delete the Unit 3 Cycle 2' specific lower SIT boron concentration requirement from the Limiting Condition for 10peration. This change is acceptable because this requirement only applied tosCycle 2 operation of Unit 3, which has' been completed. Therefore,-this specification is no longer applicable. PCN-239 By letter dated November 4, 1987.the licensees submitted proposed change PCN-239 which would revise Technical Specification (TS) 3/4.3.2, " Engineered Safety Feature Actuation System Instrumentation" to correct an-editorial discrepancy in Table 3.3-5 of the specification. Table 3.3-5, " Engineered Safety Features Response Times," specifies the Main Steam Isolation Valve (MSIV) response time in two locations, one associated with the Main Steam Isolation Signal (MSIS)(and the other associated with the Containment Isolation Actuation Signal CIAS). Amendment 46 for Unit 2 and Amendment 35 for Unit 3, issued on May 16, 1986, approved an increase c in the MSIV response time from 6.9 to 8.9 seconds but only in association with the MSIS signal. The proposed change would increase the MSIV response time in association with the CIAS signal from 6.9 seconds to 8.9 seconds to make it consistent with that of the MSIS signal. The safety evaluation supporting the MSIS signal also applies to the CIAS response t ir.e. This change will provide consistency in Table 3.3-5 and it will not effect plant equipment or operation nor will it affect previously analyzed accidents. Therefore, this change is acceptable. PCN-155 By letter dated April 27, 1984 the Itcensees submitted preposed change PCN-155 which would add Section 3/4.4.10, " Reactor Coolant Gas Vent System" (RCGVS) and its bases to the Technical Specifications. This proposed change was subsequently revised to correct typographical errors by letter dated June 3, 1986. The primary safety concern with the RCGVS is.the ability to isolate the system. The proposed Technical Specification h would maintain the system isolated and, in the event of a malfunction of any valve in the system, would remove power from the redundant isolation valve (s)inthatventpath(s). The ACTION statement provides acceptable-time limits for returning the system to ope,rability since the RCGVS'is not used for nomal plant operations or for mitigation of any design basis accident. The exception'to Section 3.0.4 of the Technical Specifications is acceptable for the same reason. Therefore. the staff finds the proposed Section 3/4.4.10 to be acceptable. 3.0

SUMMARY

OF STAFF EVALUATION The staff has reviewed PCN 197, 231, 233, 234, 239 and 155. We find these changes to be acceptable as proposed and clarified by the licensee's submittals referenced herein. m_ __,,_,________,_,__,.____m.--___---___.s.___

s. . 4.0. CONTACT WITH-STATE OFFICIAL

The NRC. staff has advised the Chief'cf the Radiological Health Brsneh,.

State Department of Health Services, State of California..of the proposed determination of no significant hazards consideration. No comments were received.

5.0 ENVIRONMENTAL CONSIDERATION

Pursuant to 10 CFR 51.32, an environmental assessment related to PCN-231, ~ 233 and'234 has been published (53 FR 29291) in the Federal Register on August 3, 1988. The Commission has-determined that the issuance of this amendment for those changes will not have a significant effect on the quality of the human environment. The remaining changes, PCN-197, 239 and 155 covered by these amendments, involve changes in the installation or use of facility. components located within the restricted area. The staff has detenmined that those portions of the amendments' involve no significant increase in the amounts of any effluents that may be released offsite and that there is no significant increase in individual or cumulative occupation radiation exposure. The Commission has previously issued proposed findings that the changes involve. no significant hazards consideration, and there has been no public' comment on such findings. Accordingly, those portions.of the amendments meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no' environmental impact state-ment or environmental assessment need to be prepared in connection with the issuance of this portion of these amendments.

6.0 CONCLUSION

Based upon our evaluation of the propcsed changes to the San Onofre Units 2 and 3 Technical Specification, we have concluded that: (1)thereis reasonable assurance that the health and safety of the endangered by operation in the proposed manner; and (2)public will not be such activities. will be conducted in compliance with'the Commission's regulations and the issuance of the amendments will not be inimical to the common defense and security or to the health and safety of the public. Principal Contributors: L. Kopp and D. Hickman Dated: August 3, 1988 n

7590-01 1 _ UNITED STATES NUCLEAR REGULATORY COMMISSION SOUTHERN CALIFORNIA EDISON COMPANY. ET AL. DOCKET N05. 50-361 AND 50-362 ' ' ENVIRONMENTAL ASSESSMENT AND FINDING OF N0 SIGNIFICANT IMPACT i The United States Nuclear Regulatory Comunission (the Commission) is considering issuance of amendments to Facility Operating License Nos. NPF-30 and NPF-15, issued to Southern California Edison Company, San Diego Gas and Electric Company, The City of Riverside, California and The City of Anaheim, California (the licensees), for operation of the San Onofre Nuclear Generating Station, Units 2 and 3, located in San Diego County, California. ENVIRONMENTAL ASSESSMENT Identification of Proposed Action: The proposed amendments would incorporate proposed changes, identified as PCN-231, 233 and 234, as described below: Proposed Change PCN-231 is a request to revise Technical Specification 3/4.1.3.6, " Regulating CEA Insertion Limits.' The proposed change would revise Figure 3.1-2, relaxing the CEA insertion limits at low power levels to increase operating flexibility and to reduce the volume of radioactive waste water. Proposed Change PCN-233 is a request to revise Technical Specification 3/4.10.2.

  • Group Neight, Insertion and Power Distribution Lief ts,' and Tables 2.21 and 3.31. The proposed change would modify the existing Technical Specification to allow use of the correct Special Test Exception during the sensurement of various CEA reactivity worths at low power levels. Tables l

2.2-1 and 3.3-1 would incorporate this Special Test Exception in Footnotes (5) 4>p Qq93102 2 S 7

i, r, s and(C),respectively. Additionally, the proposed change would specify, in Table 3.31, an exemption from the requirements of Specification 3.0.4 if the I ~ CEACs are inoperable. PCN-233 would also correct a reference t'rror in Surveillance Requirement 4.10.2.2. Preposed Change PCN-234 is a request to revise Technical Specification 3/4.5.1, " Safety Injection Tanks." The proposed change would increase the upper limit on SIT cour gas pressure from 625 to 640 psig and change the designated units from psig to psia. The Need for the Proposed Action: The proposed changes would provide operational flexibility, clarify and correct the Technical Speciffr4tions, and reduce the volume of radioactive waste water. Environmental Impacts of the Proposed Action: The Commission has completed its evaluation of the proposed revisions to the Technical Specifications and has concluded that the proposed changes provide reasonable assurance that the facility caa be operated safely. The proposed changes do not increase the probability er consequences of accidents, no changes are being aasde in the types of arty effluents that may be released offsite, and there is no significant increase in the allowable individual or cumulative occupational radiation exposure. Changing the insertion limits will decrease the volume of radioactive waste water and changing the SIT cover gas pressure will result in core reflood after a large break LOCA commencing sooner. Accordingly, the Commission concludes that this proposed action would result i in no significant adverse radiological environmental impact. l With regard to potential nonradiologic61 1spects, the proposed change to the Technical Specifications involves systems located within the restricted

.y,.., i

4. l i

area as defined in.10 CFR Part 20. It does not affect nonradiological plant effluents and has no other environmental impact. Therefore, the Comission concludes that there are no significant nonradiological environmental impacts l associated with the proposed amenhents. The Notice of Consideration of Issuance of Amendments and Opportunity for Hearing in' connection with this action was published in the FEDERAL REGISTER on March 31,1988(53FR10452). No request for hearing or petition for-leave to intervene was filed following these notices. Alternative to the Proposed Action: Since the Comission concluded that there is no significant adverse environmental effect that would result from the proposed action, alternatives with equal or greater environmental impacts need not be evaluated. 4 The principal alternative would be to deny the requested amendments. Denial of the request would not reduce environmental impacts of plant operation and in fact would prevent a reduction in the volume of radioactive waste water generated at the facility. Alternative Use of Resources: This action does not involve the use of resources not previously considered in the Final Environmental Statement Related to the Operation of the San Onofre Nuclear Generating Station, tinits 2 and 3, dated April 1981. Agencies and persons Conculted: The NRC staff has reviewed the licensee's request and did not consult other agencies or persons. FINDING 0F NO $1GNIFICANT IMPACT The Commission has determined not to prepare an environmental impact statement for the proposed license amendment.

l. t a Based upon this environmental assessment, we conclude that the proposed action will not,.have a significant adverse effect on the quality of the human environment. For further details with respect to this action, see the application for amendments dated December 14, 1987 and the supplementary information provided by letters dated April 14 and May 6,1988, which.are available for pubite inspection at the Comission's Public Document Room,1717 H Street, N. W., Washington, D.C., and at the General Library, University of California at Irvine, Irvine, California 92713. ' Dated at Rockville, Maryland, this 29th day of July,1988. FOR TH NUCLEAR P.EGUL ORY COMISSION e eorge Knighto, Director Projec Directorate Y Division of Reactor Projects - III IV, Y and Special Projects Office of Nuclear Reactor Regulation $a 4 0 ________.____m..____

, 7, y. ?,.. L 7590-01 L UNITED STATES NUCLEAR REGULATORY COMMISSION ' SOUTHERN CALIFORNIA EDISON COMPANY. ET AL. DOCKET NOS. 50-361 AND 50-362 NOT'1CE OF ISSUANCE OF AMENDMENTS TO i FACILITY OPERATING LICENSES -{ The U.S. Nuclear Regulatory Commission (Cossiission) has issued Amendment No. 64 to Facility Operating License No. NPF-10 and Amendment No. 53 to Facility - ( Operating License No. NPF-15, issued to Southern California Edison Company, San Diego Gas and Electric Company, W Cf ty of Rivertide, California and The City of Anaheim, California (the t{cesees), which revised the Technical SpecP/1 cations for operation of the San Onofre Nuclear Generating Station, Units 2 and 3, located in San Diego County, California. ~. The amendment was effective as of the date of issuance. These amendments revise a number of Technical Specificaticas (TS), and are in partial response to applications for amendments designated as I'CN-231, 233 and 234. The Technical Specifications th'at are changed by each PCN are as follows: PCN-231 - Figure 3.1-2 of TS 3/4.1.3.6, " Regulating CEA Insertion Limits"; PCN-233 - TS 3/4.10.2, " Group Height Insertion and Power Distribution Limits," and Tables 2.2-1 and 3.3-1; and PCN-234 - TS 3/4.5.1, " Safety Injection Tanks." The applications for the amendments comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations. The Commission has made appropriate findings as required by the Act and the Constission's rules and regulations in 10 CFR Chapter I, which are set forth in the license amendments. i %808310926 Ry I 1

.. t.; , Notice of Consideration of Issuance of Amendment and Opportunity for Prior Hearing in* connection with this action was published in the FEDERAL REGISTER on March 31,1988(53FR1D452). No request for a hearing or petition for leave to intervene was filed following this notice. The Cossnission has prepared an Environmental Assessment related to the action and has determined that an environmental impact statement will not be prepared and that issuance of the amendments will not have a significant effect on the quality of the human environment. For further details with respect to the action see (1) the applications for amendments dated December 14, 1987 and supplemental letters dated April 14 and May 6,1988,(2) Amendment No. 64 to License No. NPF-10 and Amendment No. 53 License No. NPF-15, (3) the Cossnission's related Safety Evaluation and (4) the. Cossnission's Environmental Assessment. All of these items are available for public inspection at the Cossnission's Public Document Room,1717 H Street N.W., and at the General Library, University of California, P.O. Box 19557 Irvine, California 92713. A copy of items (2), (3) and (4) may be obtained upon request addressed to the U.S. Nuclear Regulatory Cossnission, Washington, D.C. 20555, Attention: Director, Division of Reactor Projects III, IV, Y and Special Projects. Dated at Rockville, Maryland this 3rd day of August 1988. FOR THE NUCLEAR REGULATORY COPMISSION = D. E. Hickman, Project Manager Project Directorate Y Division of Reactor Projects - III, IV, Y and Special Projects Office of Nuclear Reactor Regulation

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' UNITED STATES NUCLEAR REGULATORY COMMISSION l l ... SOUTHERN CALIFORNIA EDISON COMPANY, ET AL DOCKET NOS. 50-361/362 NOTICE OF PARTIAL DENIAL OF AMENDMENTS TO FACILITY OPERATING LICENSES AND OPPORTUNITY FOR HEARING 1 The U.S. Nuclear Regulatory Comission (the Comission) has denied part of a request by Southern California Edison Company, San Diego Gas and Electric Company, The City of Riverside, California and the City of Anaheim, California (licensees) for amendments to Facility Operating License Nos. NPF-10 and NFF-15, ~ issued to the licensees, for operation of the San Onofre Nuclear Generating Station (SONGS), Units 2 and 3, located in San Diego County, California. The Notice of Consideration of Issuance of Amendments was published in the FEDERAL REGISTER on March 31,1988(53FR10452). The amendments, as proposed by the licensees, would revise Technical Specification (TS) Surveillance Requirement 4.3.3.2(a) by changing the fre-quer.cy of performance of the Incore Detection System Channel Check from 7 days to 31 Effective Full Power Days. The proposed change would allow verifi-cation of incore detector operability to be performed in conjunction with other routine surveillance. The staff finds that there are no plant-unique characteristics of SONGS 2 or 3 which would justify deviation from the_ Incore Detection System Operability Requirements specified in the Standard Technical Specifications. In addition, the staff does not consider the reported error introduced by a 15055iO22? by'

.% % w. i r - single soft (undetected) detector failure to be insignificant. The degree of compromise to mea'surement quality which is acceptable is a judgment which l should be considered generically. By September 9, 1988, the licensees may demand a hearing with respect to the denial described above and any person whose interest may be affected by this proceeding may file a written petition for leave to intervene. A request for a hearing or petition for leave to intervene must be filed with the Secretary for the Comission, U.S. Nuclear Regulatory Comission, Washington, D.C. 20555, Attention: Docketing and Service Branch, or may be delivered to the Comission's Public Document Room,1717 H Street, N.W., Washington, D.C., by the above date. A copy of any petitions should also be sent to the Office of the General Counsel, U.S. Nuclear Regulatory Comission, Washington, D.C. 20555, and to Charles R. Kocher, Esq., Southern California Edison Company, 2244 Walnut Grove Avenue, P.O. Box 800, Rosemead, California 91770, and Orrick, Herrington & Sutcliff, Attn: David R. Pigott, Esq., 600 Montgomery Street, San Francisco, California 94111. For further details with respect to this action, see (1) the application for amendments dated December 14,1987, and (2) the Comission's Notification of Denial fonsarded to the licensees by letter dated April 19, 1988, which are available for public inspection at the Comission's Public Document Room, 1717. H Street, N.W., Washington, D.C., and at the General Library, University of California at Irvine, Irvine, California 92713. A copy of Item (2) may be obtained upon request addressed to the U.S. Nuclear Regulatory Commission,

j w e. % - '5,, 3-Washington, D.C. 20555, Attention: Director, Division of Reactor Projects - III, IV, V and Special Projects. Dated at Rockville, Maryland, this 3rd day of August 1988. FOR THE NUCLEAR REGULATORY COMMISSION &F Donald E. Hickman, Project Manager Project Directorate V Division of Reactor Projects - III, IV, Y and Special Projects e e i h

k / October 4, 1988 Docket Nos.: 50-361 and 50-362 fir. Kenneth P. Baskin l*r. Gary D. Cotton Vice President Senior Vice President SouGern California Edison Company Engineering and Operations 2244 Walnut Grove Avenue San Diego Gas and Electric Company Post Office Ecx 800 101 Ash Street Rosemead, California 91770 Post Office Box 1831 San Diego, California 92112 Gentlemen:

SUBJECT:

REISSUANCE OF AMEllDMENTS 64 AND 5310 FACILITY OFERATING LICENSES NPF-10 AND NPF-15, RESPECTIVELY (TAC NOS. 54737, 54738, 65547, 6E540, 66816, 66817, 668201, 66821, 66822, 66823, 66826 AND 66287) On August 3,1988, the Commission issued the above amenan':cnts. Through an error in reproduction, r.ct all the Technical Specifications pages were printed. Therefore, we have reissued those amendnents in their entirety. We have also found that a symbol was omitted in Item 15 of Table 3.3-1. The revised amendments correct this error also. We apologize for any inconvenience caused by these mistates. Sincerely, original signed by Donald E. Hickman, Project Manager Froject Directorate V Division of Reactor Projects - III, IV, Y and Special Projects

Enclosures:

As stated cc: See next page DISTRIBUTION Docket tise NRC & LPDRs PDV Reading JLee .DHickman" ACRS (10) DHagan GPA/PA OGC ?9!!C/C%?D-Vg ^ iG :DRSP: LV:LA :DR5P:PDV:Pl1 :DF5F:U:P V - - - - -. : - -. - - - -... - : - - - #2. 3-..:- DE :JL L.

DHickman:CW :GWKNIGHTON :

a---:h-7/88.f....--; k/ /88..-------:k/9/88 EE : / 0FFICIAL RECORD COPY ('

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/' %'i UMITED STATES 8' NUCLEAR REGULATORY COMMISSION l 5 g%' { -E WASHINGTON, D. C. 20555 l t t l '%...+ ! October 4, 1988 i Iccket hos.: 50-361 and 50-362 Mr. Kenneth P. Baskin Mr. Gary D. Cotton Vice President Senior Vice President Southern California Edison Company Engineering and Operations 2244 Walnut Grove Avenue San Diego Gas and Electric Company Post Office Box 800 101 Ash Street Rosetead, California 9177C Fest Office Ecx 1831 San Diego, California 92112 Gentlemen: SUEJECT: REISSUANCE OF A!!ENDMENTS 64 AFD 53 TO F ACILITY OPERATING LICE!JSES hPF-10 AND NPF-15, RESPECTIVftY (TAC N05. 54737, E4738, 655d7, CE548, ] 66816, 66817, 66P201, 66821, 66022, 66823, 6CE26 AND 662E7) ) l On August 3, 1988, the Commission issue 6 the above amendments. Through an error in reproduction, not all the Technical Specifications pages were printec. Therefore, we have reis, sued those amendments in their entirety. L'e have also found that a symbol was omitted in Item 15 of Table 3.3-1. The revised sraendments correct this error also. We apologize for any inconvenience caused by these n.istakes. Sincerely, \\ / ~' ' / / i) j'.. / D<r i Ecnald E. Hickman, Project Manager Project Directorate V Division of Reactor Projects - III, IV, V and Special Projects

Enclosures:

As stated cc: See next page ( Q Ib s Ip5,

,y ', / ,/ Mr. Kenneth P. Baskin San Onofre Nuclear Generating Southern California Edison Company Station, Units 2 ano 3 ~ cc: Mr. Gary D. Cotton Mr. Hans Kaspar, Executive Director Senior Vice President Marine Review Comittee Inc. Engineering and Operations 531 Encinitas Boulevard, Suite 105 San Diego Gas & Electric Company Encinitas, California 92024 101 Ash Street Post Office Box 1831 San Diego, California 92112 Mr. Mark Medford Southern California Edison Company Charles R. Kocher, Esq. 2244 Walnut Grove Avenue James A. Beoletto, Esq. P. O. Box 800 Southern California Edison Company Rosemead, California 91770 1 2244 Walnut Grove Avenue P. O. Box 800 Mr. Robert G. Lacy Rosemead, California 91770 Manager, Nuclear Department San Diego Gas & Electric Company Orrick, Herrington & Sutcliffe P. O. Box 1831 i ATTN: David R. Pigott, Esq. San Diego, California 92112 600 Montgomery Street i San Francisco, California 94111 Richard J. Wharton, Esq. University of San Diego School of_ Alan R. Watts, Esq. Law Rourke & Woodruff Environmental Law Clinic 701 S. Parker St. No. 7000 San Diego, California 92110 Orange, California 92668-4702 I Charles E. McClung, Jr., Esq. Attorney at Law Mr. S. McClusky 24012 Calle de la Plaza / Suite 330 Bechtsi Power Corporation Laguna Hills, California 92653 P. O. Box 60860, Terminal Annex Los Angeles, California 90060 Regional Administrator, Region V U.S. Nuclear Regulatory Comission Mr. C. B. Brinkman 1450 Maria Lane / Suite 210 Combustion Engineering, Inc. Walnut Creek, California 94596 7910 Woodmont Avenue, Suite 1310 Bethesda, Maryland 20814 Resident Inspector, San Onofre NPS c/o U. S. Nuclear Regulatory Comission Mr. Dennis F. Kirsh Post Office Box 4329 U.S. Nuclear Regulatory Coenission San Clemente, California 92672 Region Y 1450 Maria Lane, Suite 210 Mr. Sherwin Harris Walnut Creek, California 94596 Resource Project Manager Public Utilities Department Mr. Dennis M. Smith, Chief City of Riverside Radiological Programs Division City Hall Governor's Office of Emergency Services 3900 Main Street 2800 Headowview Road Riverside, California 92522 j Sacramento, California 95832 1-

N ? k 4^, Southern California Edison Company - 2_- SanOnofre2/3(whenspecified) i ~ cc:- I California State Library Government Publications Section l Library & Courts Building Sacramento,.CA 95841-ATIN:- Ms. Mary Schnell Mayor, City of San Clemente i L San.Clemente, CA 92672 f Chairman, Board Supervisors San Diego County 1600-Pacific Highway, Room 335 San Diego, CA 92101 California Department of Health ATTN: Chief. Environmental Radiation Control Unit Radiological Health Section 714 P Street, Room 498-Sacramento, CA-95814 q Mr. Jack McGurk, Acting Chief ~~ Radiological Health Branch State Department of Health Services 714 P Street-Building #8 Sacramento, California 95814 e ,___a____-_ _ - - - _ - - - - - - - - - - ^

4 g f Gt2 / 4* UNITED STATES 8' NUCLEAR REGULATORY COMMISSION ^ n .,E WASHINGTON, D. C. 20555 k....,o/ August 3, 1988 Docket Nos.: 50-361 and 50-262 s.. Mr. Kenneth P. Baskin Mr. Gary D. Cotton Vice President Senior Vice President Southern California Edison Company Engineering and Operations 2244 Walnut Grove Avenue San Diego Gas & Electric Company Post Office Box 800 101 Ash Street Rosemead, California 91770 Post Office Box 1831 San Diego, California 92112 Gentlemen:

SUBJECT:

ISSUANCE OF AMENDMENT NO. 64 TO FACILITY OPERATING LICENSE NPF-10 AND AMENDMENT NO. 53 TO FACILITY OPERATING LICENSE NPF-15 SAN ONOFRE NUCLEAR GENERATING STATION, UNITS 2 AND 3 (TAC NOS. 54737/54738/65547/65548/66816/66817/66820/66821/66822/ 66823/66826/66827) The Nuclear Regulatory Comission (the Commission) has issued the enclosed Amendment No. 64 to Facility Operating License No. NPF-10 and Amendment No. 53 to Facility Operating License No. NPF-15 for the San Onofre Nuclear Generating Station, Units 2 and 3, located in San Diego County, California. These amendments revise a number of Technical Specifications (TS), and are in partial response to your applications for amendments designated as PCN-197, 231, 233, 234, 239 and 155, dated May 12, December 14, and November 4,1987 and April 27, 1984. The Technical Specifications that are changed by each PCN are as follows: PCN-197 - TS 3.1.3.6, " Regulating CEA Insertion Limits," TS 3.1.3.7, "Part Length CEA Insertion Limits," TS 3.10.2, " Group Height, Insertion and Power Distribution Limits," and Bases 3/4.1.3, " Movable Control Assemblies"; PCN-231 - Figure 3.1-2 of TS 3/4.1.3.6, " Regulating CEA Insertion Limits"; PCN-233 - TS 3/4.10.2, " Group Height, Insertion and Power Distribution Limits," and Tables 2.2-1 and 3.3-1; PCN-234 - TS 3/4.5.1, " Safety Injection Tanks"; PCN-239 - TS 3/4.3.2, " Engineered Safety Features Actuation System Instrumentation"; and PCN-155 - TS-3/4.4.10. " Reactor Coolant Gas Vent System." A copy of the related safety evaluation supporting this amendment is enclosed. The enclosed Notice of Issuance will, which relates to approval of PCN 231, 233 and 234, be forwarded to the Office of the Federal Register for publication. Approval of PCN-155, 197 and 239 will be included in a separate Notice of Issuance in the Comission's bi-weekly Federal Register Notice. Proposed Change Number 232 concerning TS 3/4.3.3 " Monitoring Instrumentation," has been denied, as indicated in our letter to you dated April 19, 1988. 8'8083/0/77

c o l l ~2-A copy of the rel'ated Notice of Denial is enclosed for your information. The Notic.e has been forwarded to the Office of the Federal Register for publication, o The remaining undispositioned changes proposed in your letters dated May 12 and December 14, 1987 require further staff review which is currently underway. Si erely, N.c Donald E. Hickman, Project Manager i Project Directorate V Division of Reactor Project - III, IV, V and Special Projects

Enclosures:

1. Amendment No. 64 to NPF-10 2. Amendment No. 53 to NPF-15 3. Safety Evdluation 4 Environmental Assessment 5. Notice of Issuance 6. Notice of Denial cc: See next page l l l i I i )}}

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