ML20012B523

Amends 51 & 50 to Licenses DPR-80 & DPR-82,respectively, Revising Tech Spec to Allow Removal of Boron Injection Tank from Each Unit

ML20012B523
Person / Time
Site:	Diablo Canyon
Issue date:	02/26/1990
From:	Rood H Office of Nuclear Reactor Regulation
To:
Shared Package
ML20012B524	List: ML20012B523 ML20012B525
References
NUDOCS 9003150187
Download: ML20012B523 (24)

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WASHINGTON, D. C. 20555 '+9 *... + / PACIFIC GAS AND ELECTRIC COMPANY l DIABLO CANYON NUCLEAR POWER PLANT, UNIT 1 DOCKET NO. 50-275 _ AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 51 License No. DPR-80 l-1. The Nuclear Regulatory Commission (the Commission) has found that: l A. The application for amendment by Pacific Gas & Electric Company (thelicensee),datedMay 15, 1989, as modified by letters dated i~ July 3, Septeober 15, and November 30, 1989, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's regulations set forth in 10 CFR Chapter I; B. The facility will o>erate in conformity with the ap)11 cation, the provisions of t.1e Act, and the regulations of tie Comission;. C. There is reasonable assurance (1) that the activities authorized l-by this amendment can be conducted without endangering the health L ~ and safety of the public, and (ii) that such activities.will be l conducted-in compliance with the Comission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; L. and L 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. 9003150107 900226 PDR ADOCK 0500 5 g P 4 m m

^ 4 , 2. Accordingly, the license is amended by changes to the Technical l Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility. Operating License - No. DPR-80 is hereby amended to read as follows: (2) Technical Specifications i The Technical Specifications contained in Appendix A and the 4 . Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 51, are hereby incorporated in the license. Pacific Gas & Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan, except where otherwise stated in specific license l conditions, i-3. This license amendment becomes effective at the date of its issuance. FOR THE NUCLEAR REGULATORY COMMISSION C i. Harry Rood, Acting Director Project Directorate ~V Division of Reactor Projects - III, IV, V and Special Projects Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications l Date of Issuance: February 26, 1990 L. i G e

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WASHINGTON, D. C. 20555 \\...../ PACIFIC GAS AND ELECTRIC COMPANY DIABLO CANYON NUCLEAR POWER PLANT, UNIT 2 DOCKET NO. 50-323 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 50 Licerise No. DPR-82 1. The Nuclear Regulatory Comission (the Comission) has found that: A. The application for amendment by Pacific Gas & Electric Company (the licensee), dated May 15, 1989, as modified by letters dated i July 3 September 15, and November 30, 1989, complies with the l standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's regulations. set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the ap)11 cation, L the provisions of the Act, and the regulations of tie L Comission; L C. Thereisreasonableassurance(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; D.- The issuance of this amendment will not be inimical to' the-comon defense and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51-of the Comission's regulations and all applicable requirements have e been satisfied. l l L

j q 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-82 is hereby amended to 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. 50, are hereby incorporated in the license. Pacific Gas & Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental-Protection _ Plan, except where otherwise stated in specific license conditions. 3. This license amendment becomes effective at the date of its issuance. FOR THE NUCLEAR REGULATORY COMMISSION. l Harry d, Acting Director Project Directorate V-Division of Reactor Projects - III, IV, V and Special Projects Office of Nuclear Reactor Regulation

Attachment:

l Changes to the Technical-E Specifications Date of Issuance: February 26, 1990 b 4

l e 4 4 ~ g. ATTACHMENT TO LICENSE AMENDMENT NOS. 51 AND 50 FACILITY OPERATING LICENSE NOS. DPR-80 and DPR-82.- DOCKET NOS. 50-275 AND 50-323 Replace the following ) ages of the Appendix "A" Technical Specifications with the attached pages. Tie. revised pages are identified by amendment number and contain vertical. lines indicating the areas of change. Overleaf pages are also included, as appropriate. Remove Page Insert Page I i 3/4'3-28 3/4 3-28 l. 3/4 3-29 3/4 3-29 3/4 3-31 3/4 3-31 3/4 5-9 3/4 5-9 3/4 5-10 3/4 5-10 L 3/4 6-23 3/4 6-23 l 3/4 6-24 3/4 6-24 3/4 8-20 3/4 8-20 o l: 3/4 8-21 3/4 8-21 B 3/4 3-1 B 3/4 3-1 B 3/4 3-la B 3/4 5-2 B 3/4 5-2 e w- -m - =

e o-TABLE-3.3-4 (Continued) I ENGINEERED SAFETY FEATURES ACTUATION SYSTEN INSTRIMENTATION TRIP SETPOINTS o FUNCTIONAL UNIT TRIP SETPOINT ALLOWA8tE VALUES j 7. . Loss of Power [o (4.16 kV Emergency Bus Undervoltage) g a. First level Z 1) Diesel Start > 0 volts with a > 0 volts with a [ 30.8secondtimedelay 3.0.8secondtimedelay and and > 2583 volts with a > 2583 volts with a 310secondtimedelay 5 10 second time delay 2) Initiation of Load Shed 'One relay. One relay > 0 volts with a > 0 volts with a q 34secondtimedelay 34secondtimedelay and and w > 2583 volts with a > 2583 volts with a 325secondtimedelay 5 25 second time delay. with one relay with one relay m > 2870 volts, instantaneous > 2870 volts, instantaneous { b. Second Level C 1) Diesel Start > 3600 volts with a > 3600 volts with a 5 310secondtimedelay 510secondtimedelay $ g% > 2) Initiation of Load Shed > 3600 volts with a > 3600 volts with a l 320secondtimedelay 320secondtimedelay 8. Engineered Safety Features Actuation System Interlocks o et,[g* f;; a. Pressurizer Pressure, P-11 5 1915 psig 5 1925 psig "T b. Low-Low Tavg, P-12 increasing 543*F < 545.8'F (Units 1 and 2 wy {g Cycle 4 and after) < 545'F (Unit 2 Cycle 3) if" decreasing. 543*F > 540.2'F (Units 1 and 2 wg Cycle 4 and after) > 541*F (Unit 2 Cycle 3) c. Reactor Trip, P-4 M.A.- N.A.

TABLE 3.3-5 ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS - 1. Manual Initiation a. Safety Injection (ECCS) N.A. 1) Feedwater Isolation P.A. 2) Reactor Trip N.A. 3) Phase "A" Isolation N.A. 4) Containment Ventilation Isolation N. A. 5) Auxiliary Feedwater N.A. 6)- Component Cooling Water N.A. 7) Containment Fan Cooler Units N. A. 1 8) Auxiliary Saltwater Pumps N.A. b. Phase "B" Isolation i 1) Containment Spray (Coincident with SI Signal)- N.A. 2) Containment Ventilation Isolation N.A. c'. Phase "A" Isolation 1) Containment Ventilation Isolation N.A. d. Steam Line Isolation N.A. 2.' - Containment Pressure-High L a. Safety Injection (ECCS) 5 27I7)/25(4) l 3 1) Reactor Trip <2 2) Feedwater Isolation I 63(2) 3) Phase "A" Isolation i 18(4)/28(5) 4) Containment Ventilation Isolation R.A.(1) l 5) Auxiliary Feedwater < 60 i 38(4) l i 40(1)/48(5) 6) Component Cooling Water 7) Containment Fan Cooler Units J 8) Auxiliary Saltwater Pumps 548(4)/58(5) } 3. Pressurizer Pressure-Low a. Safety Injection-(ECCS) $ 27(7)/25(4)/35(5) l p 1). Reactor Trip <2 I 63(1) 2) 2) Feedwater Isolation 1 3) Phase "A" Isolation i 18( 4) Containment Ventilation Isolation R.A.(3) 5)- Auxiliary Feedwater < 60 i 6) Component Cooling Water i 48(3)/38II) 3) i 40C 7) Containment Fan Cooler Units 8) Auxiliary Saltwater Pumps 558(3)/48(1) I' DIABLO CANYON - UNITS 1 & 2 3/4 3-28 Amendment Nos. 51 and 50 1 iL

4 ~.. - TABLE 3.3-5 (Continued) I ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONOS 4. Differential Pressure Between Steam Lines-High a. Safety Injection (ECCS) 5 25(4)/35(5) l 1) Reactor Trip. <2 2) Feedwater Isolation I 63(2) 3) Phase "A" Isolation i 18(3)/28II) l 4) Containment Ventilation Isolation N.A.II) 5) Auxiliary Feedwater < 60 i 38I1) i 40(3)/48(3) 6) Component Cooling Water 7)- Containment Fan Cooler Units 8) Auxiliary Saltwater Pumps 3'48(1)/58(3) 5. Steam Flow in Two Steam Lines - High coincident with T,yg-Low-Low Safety Injection (ECCS) 5 25(4)/35(5) l a. 1) Reactor Trip. <4 2) Feedwater Isolation 7 65(2) 3) Phase "A" Isolation i 20(1)/30(3) l 4) Containment Ventilation Isolation N.A.(3) 5) Auxiliary Feedwater < 60 6) Component Cooling Water 7 40(1) II) 7) Containment Fan Cooler Units i 40(3)/50 8)- Auxiliary Saltwater Pumps 550(1)/60I3) b. Steam Line Isolation 5 10 6. Steam-Flow in Two Steam Lines-High Coincident with Steam Line Pressure-Low a. Safety Injection (ECCS) 5 25(4)/35(5) l-1) Reactor Trip <2 2) Feedwater Isolation i G3(2) 3) Phase "A" Isolation i 18(1)/28(3) l 4) Containment Ventilation Isolation R.A.(3) 5) Auxiliary Feedwater < 60 i 38(1) i 40(3)/48(3) 6) Component Cooling Water 7) Containment Fan Cooler Units 8) Auxiliary Saltwater Pumps 348(1)/58(3) b. Steam Line Isolation 58 L DIABLO CANYON - UNITS 1 & 2 3/4 3-29 Amendment Nos. 51 and 50

TABLE 3.3-5-(Continued) ENGINEERED SAFETY' FEATURES RESPONSE TIMES -{ O INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 7.. Containment Pressure-High-High a. Containment Spray < 48.5(6) b. Phase "B" Isolation N. A. c.~ Steam Line. Isolation 57 8. Steam Generator Water Level-High-High a. Turbine Trip _766g2) < 2. b. Feedwater Isolation 9. Steam Generator Water Level Low-Low l a. Motor-Driven Auxiliary Feedwater Pumps < 60

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Turbine-Driven Auxiliary ~ Feedwater Pump i 60 10. RCP Bus Undervoltage . Turbine-Driven Auxiliary u i Feedwater Pump < 60 1-Ell. Plant Vent Noble Gas Activity-High Containment Ventilation Isolation < 11 h b k O DIABLO CANYON - UNITS 1 & 2 3/4 3-30

4 TABLE 3.3-5 (Continued) TABLE NOTATIONS (1) - Diesel generator starting delay not included because offsite power available. I (2) Feedwater System overall response time shall include. verification of each individual Feedwater System valve closure time as shown below: s closure Time (not including Valve instrumentation delays) FCV-438 < 60 seconds 439 7 60 seconds 440 7 60 seconds 441 7 60 seconds -510 7 5 seconds 520 7 5 seconds 530 7 5 seconds 540 7 5 seconds 1510 7 5 seconds 1520 7 5 seconds L 1530 7 5 seconds b 1540 7 5 seconds (3) Diesel generator-starting and loading delays included. h -(4) ~ Diesel generator starting delay not included because offsite power is available.. Response time limit includes opening of valves to establish SI path and attainment of. discharge pressure for centrifugal charging pumps (where applicable). Sequential transfer of charging pump suction from the VCT to the RWST (RWST valves open, then VCT valves close) is included. (5) ~ Diesel generator starting and sequence loading delays included. Offsite power is not available. Response time limit-includes opening of valves I to establish SI path and attainment of discharge pressure for centrifugal charging pumps. Sequential transfer of charging pump suction:from the L VCT to the RWST (RWST valves open, then VCT valves close) is included. (6) The maximum response time of 48.5 seconds is the time from when the con-E tainment pressure exceeds the High-High Setpoint until the spray pump is started and the discharge valve travels to the fully open position assuming off-site power is not available. The time of 48.5 seconds includes the 28-second maximum delay related to ESF' loading sequence. Spray riser piping fill time is not included. The 80-second maximum spray delay time L does not include the time from LOCA start to "P" signal. .(7) Diesel' generator starting and sequence loading delays included. Sequential transfer of charging pump suction from the VCT to the RWST (RWST valves open, then VCT valves close) is not included. Response time limit includes opening of valves to establish SI flow 3 of discharge pressure for centrifugal charging pumps, path and attainment SI, and RHR pumps (where applicable). ~DIABLO CANYON - UNITS 1 & 2 3/4 3-31 Amendment Nos. 51 and 50 ,m.-

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4 DIABLO CANYON - UNITS 1 & 2 3/4 3-32 Amendment Nos. 36 and 35 APR 2 5 gggy

1 EMERGENCY CORE COOLING SYSTEMS-3/4.5.4, BORON INJECTION SYSTEM L BORON INJECTION TANK LIMITING CONDITION FOR OPERATION 3.5.4.1 The boron injection tank shall be OPERABLE with: A minimum contained borated water volume of 900 gallons of borated' a.

water, b.

A boron concentration of between 20,000 and 22,500 ppa, and c. A minimum solution temperature of 145'F. APPLICABILITY: MODES 1, 2 and 3. For Unit 1 Cycle 4. Unit 2 Cycle 3. I ACTION: a, With the boron injection tank inoperable, restore the tank to OPERABLE status within 1 hour or be in HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to 1% Ak/k at 200'F within the next 6 hours; restore the tank to OPERABLE status within the next 7 days or be in HOT SHUTDOWN within the next 12 hours. SURVEILLANCE REQUIREMENTS 4.5.4.1 The boron injection tank shall be demonstrated OPERABLE by: Verifying the contained borated water volume through a recirculation a.. flow test at least once per 7 days, b. Verifying the boron concentration of the water in the tank at least once per 7 days, and Verifying the water temperature at least once per 24 hours. .c. DIABLO CANYON - UNITS 1 & 2 3/4 5-9 Amendment Nos. 51 and 50 l ~

j EMERGENCY CORE COOLING SYSTEMS HEAT TRACING i LIMITING CONDITION FOR OPERATION L l 3.5.4.2 At least two independent channels of heat tracing shall be OPERABLE for the boron injection tank and for the heat traced portions _ of the associated ] flow paths. I: APPLICABILITY: MODES 1, 2 and 3. For Unit 1 Cycle 4 Unit 2 Cycle 3. l ACTION: L With only one channel of heat tracing on either the boron injection tank or on L b the heat traced portion of an associated flow path OPERABLE, operation may continue for up to 30 days provided the tank and flow path temperatures are verified to be greater than or equal to 145'F at least once per 8 hours; otherwise, be in HOT STANDBY within 6 hours and in at least HOT SHUTDOWN within the following 6 hours, t L SURVEILLANCE REQUIREMENTS 1 4.5.4.2 Each heat tracing channel for the boron injection tank and associated flow path shall be demonstrated OPERABLE: At least once per 31 days by_ energizing each' heat tracing channel,- a.- and b. At least once per 24 hours by verifying the tank and flow path tem-peratures to be greater than or equal to~145'F. The tank temperature shall be determined by measurement. 'The flow path temperature shall be determined by either measurement-or recirculation flow until establishment'of equilibrium temperatures within the tank. L DIABLO CANYON - UNITS 1 & 2 3/4 5-10 Amendment Nos. 51 and 50 s

n s y TABLE 3.6-1 (Continued). o S 2 ISOLATION TIME y VALVE NO. FUNCTION (Seconds) [ 5. Power-Operated Valves (Continued) FCV-698* Containment Air Sample (Post LOCA) Supply IC N.A. FCV-699* Containment' Air Sample (Post LOCA) Supply OC N.A. e. FCV-700* Containment Air Sample (Post LOCA) Return OC N.A. PCV-19#- Steam Generator No. I 10% Atmosphere Steam Dump OC N.A. PCV-20# Steam Generator No. 2 10E Atmosphere Steam Dump OC N.A. PCV-21# Steam Generator No. 3 10% Atmosphere Steam Dump OC N.A. 1 PCV-22# Steam Generator No. 4 105 Atmosphere Steam Dump OC N.A. ] 8107# Charging Line Isolation DC N.A. 8700A# RCS Hot leg to RHR Pump 1 OC-N.A. 8700B# RCS Hot-Leg to RHR Pump 2 OC N.A. 8701# RCS Loop 4 Not Leg to RHR IC N.A. 8703# RHR to RCS Hot Legs 1 and 2 IC N.A. 8716A# RHR to RCS Hot legs OC N.A. l 8716B# RHR to RCS Hot-Legs'OC N.A. 8801Af g Charging Injection OC' N.A. 8 88018# Charging Injection OC N.A. k 8802A# Safety Injection to RCS Hot Legs CC N. A. 8802B# Safety Injection to RCS Hot Legs DC N.A. p 8809A# Residual Heat Removal to RCS Cold Legs 1 and 2 N.A. 88098# g Residual Heat Removal to RCS' Cold legs 3 and 4 N.A. 8823# Safety Injection Check Valve Test Line IC N.A. o, fo O i -- = =

, u. Sg . TABLE 3.6-1 (Continued). o n z ISOLATION TIME- [o VALVE NO. FUNCTION (Seconds) [ ' 5. Power-Operated Valves (Continued)' 8824# . Safety Injection Check Valve Test Line IC N.A.. [ 8843# Charging Injection IC N.A. l e. 8835# Safety Injection to RCS Cold Legs DC N.A. [ 8885Af RHR to Cold Leg Test Line IC N.A. l 8885B# RHR to Cold Leg Test Line IC N.A. l 8982Af Containment Sump to Residual Heat Removal Train 1 OC N.A. 1 89828# Containment Sump to Residual Heat Removal Train 2 OC N. A. i T 8980# Refueling Water Storage Tank to RHR OC N.A. m 9001A Containment Spray Pump No. 1 Isolation OC N.A. 9001B Containment Spray Pump No. 2 Isolation OC N.A. 9003Af Residual Heat Removal to Containment Spray DC N.A. 90038# Residual Heat Removal to Containment' Spray OC N.A. 6. Check Valves 8028 g Relief Valve Outlets to Pressurizer Relief Tank IC N.A. g 8046 Primary Water to Pressurizer Relief Tank IC N.A. o.g 8047 Nitrogen to Pressurizer Relief Tank IC N.A. N 8109 Seal Water Return IC M.A. k 8368A thru Seal Water to Reactor Coolant Pumps IC N.A. 83680 8916 Nitrogen Supply to Accumulators IC N.A. .E ~- .2.w s ->e-L - - ~ -"

, ELECTRICAL POWER SYSTEMS 3/4.8.4 ELECTRICAL EQUIPMENT PROTECTIVE DEVICES f. MOTOR-0PERATED VALVES THERMAL OVERLOAD PROTECTION AND BYPASS DEVICES LIMITING CONDITION FOR OPERATION 3.8.4.1 The thermal overload protection and bypass devices, integral with the motor starter, of each valve listed in Table 3.8-1 shall be OPERABLE. APPLICABILITY: Whenever the motor-operated valve is required to be OPERABLE. ACTION: With one or more of the thermal overload protection and/cr bypass devices inoperable, declare the affected_ valve (s) inoperable and apply the appropriate ACTION Statement (s) for the affected valves. SURVEILLANCE REQUIREMENTS 4.8.4.1 The above required thermal overload protection and bypass devices shall be demonstrated OPERABLE: a. At least once per 18 months, by the performance of a TRIP ACTUATION DEVICE OPERATIONAL TEST of the bypass circuitry for those thermal overload devices which are either: 1) Continuously bypassed and temporarily placed in force only when-the valve motors are undergoing periodic or maintenance testing, or 2) Normally in force during plant operation and bypassed under accident conditions. b. At least once per 18 months by the performance of a CHANNEL CALIBRATION of a representative sample of at least 25% of: 1) All thermal overload devices which are not bypassed, such that each non-bypassed device is calibrated at least once per 6 years, and 2) All thermal overload devices which are continuously bypassed, such that each continuously bypassed device is calibrated and each valve is cycled through at least one complete cycle of full travel with the motor-operator when the thermal overload device is OPERABLE and not bypassed at least once per 6 years. DIABLO CANYON - UNITS 1 & 2 3/4 8-19

[- TABLE 3.8-1 MOTOR-0PERATED VALVES THERMAL OVERLOAD-PROTECTION AND BYPASS DEVICES BYPASS DEVICE VALVE NUMBER FUNCTION (YES/N0) 8107 Charging Isolation Yes j 8105 Charging Pumps' Recirculation Yes i FCV-430 Component Cooling Heat Exch. Outlet No LCV-1128 Volume control Tank outlet Yes o 8801A Charging Injection Yes 8803A ChargingInjection Yes 8807A' Safety Injection / Charging Suction Crosstie No l 8805A RWST to Charging Pumps Yes FCV-437 Raw Water Supply to Auxiliary Feedpumps No FCV-441 S/G 4 Feedwater Isolation Yes FCV-750 RCP Thermal Barrier CCW Return Yes FCV-438 S/G 1 Feedwater Isolation Yes i ~8923A SI Pump 1 Suction No { FCV-38 Aux. FWP Turb. Steam Supply No 8980 RWST to RHR No 8974A. S1 Pumps' Recirculation No 8992 Spray Additive Tank Outlet Yes 8000A Pressurizer RV Isolation No FCV-601* Auxiliary Saltwater Pumps Crosstie No. i 8808A Accumulator 1 Isolation No i 8802A. SI Pump 1.to Hot Leg No 8821A-S1 Pump 1 to Cold Legs No I 8808D Accumulator 4 Isolation No 8808B Accumulator 2 Isolation No 8106 Charging Pumps' Recirculation Yes i l-8108 Charging Isolation Yes LCV-112C. Volume Control Tank Isolation Yes t

• FCV-601 is common to both units.

DIABLO CANYON - UNITS 1 & 2 3/4 8-20 Amendment Nos. 51 and 50 0

. + TABLE 3.8-1 (Continued) MOTOR-0PERATED VALVES THERMAL OVERLOAD PROTECTION AND BYPAS$' DEVICES BYPASS DEVICE D VALVE NUMBER FUNCTION (YES/N0) 8809A-RHR to Cold Legs No 8801B Charging Injection Yes l 8805B RWST to Charging Pumps Yes 8700A RHR Pump 1 Suction No 8804A RHR to Charging Pumps No L-FCV-436 RWSR to Auxiliary Feedpump No 9001A Spray Isolation Yes FCV-363 RCP CCW Return Isolation Yes 8835 SI Pumps to RCS Cold Legs No L 8701-RCS to RHR System No 8100 RCP Seal Water Return Yes 8803B ChargingInjection' Yes FCV-431' CCW Heat Exchanger Outlet No FCV-641A RHR Recirculation No 8716A-RHR to Hot-Legs No 8000B Pressurizer RV Isolation No FCV-439 S/G 2 Feedwater Isolation. Yes 9003A RHR to Spray No FCV-95 -Turb. Feedpump_ Steam Supply Yes 8994A Spray Additive Tank Outlet Yes 8703 RHR to Hot Legs No 8104 Emergency Borate No 8982A Containment Sump RHR Recirculation No LCV-106 S/G 1 Aux. Feedwater Supply No LCV-107 S/G 2 Aux. Feedwater Supply No LCV-108 S/G 3 Aux. Feedwater Supply No LCV-109 S/G 4 Aux. Feedwater Supply No DIABLO CANYON - UNITS 1 & 2 3/4 8-21 Amendment Nos. 51 and 50

e L-4 TABLE 3.8-1 (Continued) I MOTOR-0PERATED VALVES THERMAL OVERLOAD PROTECTION AND BYPASS DEVICES BYPASS DEVICE VALVE NUMBER FUNCTION (YES/NO) FCV-356 RCP CCW Supply Isolation Yes 1 8808C Accumulator 3 Isolation No-1 FCV-641B RHR Recirculation No -) FCV-355 CCW Header.C Isolation-Yes-9001B Spray Isolation Yes 89828' Containment Sump RHR Recirculation No. 90038 RHR to Spray No 8976 RWST to SI Pumps No 8804B RHR to SI Pumps No 8802B SI Pump 2 to Hot Legs No 8112 RCP Seal' Water Return Isolation Yes FCV-357 RCP Barrier CCW Return Isolation Yes

f FCV-749 RCP Bearing Cooling H O Return Isolation Yes 2

[ 8702 RCS to RHR-Suction No L FCV-440 S/G 3 Feeo' water Isolation Yes j 87168 RHR to RCS Hot Legs No FCV-37 Aux. Feedpump Steam Supply No 8821B SI Pump 2 to Cold Legs No t 8807B Safety Injection / Charging Suction Crosstie-No 8000C Pressurizer RV Isolation No l 8994B Spray Additive Tank Outlet Yes FCV-495 ASW Crosstie No FCV-496 ASW Crosstie No ) 8700B RHR Pump 2 Suction No 8974B SI Pumps Recirculation No 88098 RHR to Cold Legs No 8923B SI Pump 2 Suction No i DIABLO CANYON - UNITS 1 & 2 3/4 8-22

y 3/4.3 INSTRUMENTATION I* l MSES 3/4.3.: and 3/4.3.2 REACTOR TRIP $YSTEN and ENGINEERED SAFETY FEATURES ACTUAT.ON SYSTEM INSTRUMENTATION The OPERABILITY of the Reactor Trip System and Engineered Safety Features tsetuation System instrumentation and interlocks ensure that: (1) the associated ACTION and/or Reactor trip will be initiated when the parameter monitored by each channel or combination thereof reaches its Setpoint. (2) the specified coincidence logic and sufficient redundancy is maintained to pemit a channel to be out of service for testing or maintenance consistent with maintaining an appropriate level of reliability of the Reactor Protection and Engineered Safety Features instrumentation, and (3) sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance, and (4) sutficient system functional capability is available from diverse parameters. The OPERABILITY of these systems is required to provide the overall re-liability, redundancy, and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions. The integrated operation of each of these systems is consistent with the assump-tions used in the accident analyses. The Surveillance Requirements specified for these systems ensure that the overall system functional capability is main-tained comparable to the original design standards. The periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capability. Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with WCAP-10271, " Evaluation of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System, and supplements to that report. Surveillance intervals and out-of-service times were determined based on maintaining an appropriate level of reliability of the Reactor Protection System. The Engineered Safety Features Actuation System senses selected plant parameters r.nd determines whether or not predetermined limits are being exceeded. If they are, the signals are combined into logic matrices sensitive to combinations indicative of various accidents, events, and transients. Once the required logic combination is completed, the system sends actuation signals to those engineered safety features components whose aggregate function best serves the requirements of the condition. As an example, the following actions may be initiated by the Engineered safety Features Actuation System to mitigate the consequences of a steam line break or loss of coolant accident: (1) safety injection pumps start and automatic valves position. (2) Reactor trip. (3) feedwater isolation, (4) startup of the emergency diesel generators, (5) containment spray pumps start and automatic valves position. (6) containment isolation, (7) steam line isolation, (8) Turbine trip, (9) auxiliary feedwater pumps start and automatic valve position, (10) containment fan cooler units start and automatic valves pcsition, and (11) component cooling water pumps . start and automatic valves posit'on. ESF response times specified in Table 3.3-5, which include sequential operation of the RWST-and VCT valves (Table Notations 4 and 5), are based on values assumed in the non-LOCA safety analyses. These analyses take credit for injection of borated water from the RWST. Injection of borated water is DIABLO CANYON - UNITS 1&2 B 3/4 3-1 Amendment Nos. 51 and 50 I

1 i INSTRDMENTATION BASES i REACTOR PROTECTION SYSTEM and ENGINEERED SAFETY FEATURES ACTUAT!DN IN5TRUMENTATION (Continued) i assumed not to occur until the VCT charging pump suction isolation valves are closed following opening of the RWST charging pump suction isolation valves. When the sequential operation of the RWST and VCT valves is not included in the i response times (Table Notation 7), the values specified are based on the LOCA analyses. The LOCA analyses takes credit for injection flow regardless of the source. Verification of the response times specified in Table 3.3-5 will assure that the assumptions used for the LOCA and non-LOCA analyses with respect to the operation of the VCT and RWST valves are valid. e t W l l DIABLO CANYON - UNITS 1&2 8 3/4 3-la Amendment Nos. 51 and 50

i INSTRUMENTATION l i SASES i REACTOR PROTFCTION SYSTEN and ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued) i The Engineered Safety Features Actuation System interlocks perform the following functions: P-4 Reactor tripped - Actuates Turbine trip, closes main feedwater 1 valves on T,y, below setpoint, prevents the opening of the main feedwater valves which were closed by a safety Injection or High Steam Generator Water Level signal, allows safety Injection block so that components can be reset or tripped. Reactor not tripped - prevents manual block of Safety Injection. t l P-11 On increasing pressurizer pressure P-11 automatically reinstates Safety Injection actuation on low pressurizer pressure. On decreasing l pressure, P-11 allows the manual block of Safety Injection actuation on Low Pressurizer Pressure. P-12 On increasing reactor coolant loop temperature, P-12 automatically reinstates safety Injection actuation on High Steam Flow coincident L with either Low-Low T,y, or Low Steam Line Pressure, and provides an i arming signal to the Steam Dump System. On decreasing reactor coolant loop temperature P-12 allows the manual block of Safety Injection actuation on High Steam Flow coincident with either Low-Low i T,y, or Low Steam Line Pressure and automatically removes the arming j signal from the Steam Dump System. 3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3d RADIATION MONITORING FOR PLANT OPERATIONS l, l The OPERABILITY of the radiation monitoring channels ensures that: (1) the radiation levels are continually measured in the areas served by the individual channels and (2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded. 3/4.3.3.2 MOVABLE INCORE DETECTORS The OPERABILITY of the movable incore detectors with the specified minimum I complement of equipment ensures that the measurements obtained from us0 of this system accurately represent the spatial neutron flux distribution of the core. The OPERABILITY of this system is demonstrated by irradiating each detector used and determining the acceptabilitu of its voltage curve. ForthepurposeofmeasuringF(Z)orFhafullincorefluxmapisused. q Quarter-core flux maps, as defined in WCAP-8648, June 1976, may be used in recalibration of the Excore Neutron Flux Detection System, and full incore flux maps or symmetric incore thimbles may be used for monitoring the QUADRANT POWER TILT RATIO when one Power Range channel is inoperable. DIABLO CANYON - UNITS 1&2 8 3/4 3-2 ,-,.,,.,--,,,--,,-,,.,,,-,,,,g, - -., - - - - - - - - - ~ - -.,, -, -

3/4.5 EMERGENCY CORE COOLING SYSTEMS I BASES I 3/4.5.1 ACCUMULATORS j The OPERABILITY of each Reactor Coolant System (RCS) accumulator ensures l that a sufficient volume of borated water will be immediately forced into the core through each of the cold legs in the event the RCS pressure falls below the pressure of the accumulators. This initial surge of water into the core i i provides the initial cooling mechanism during large RCS pipe ruptures. The limits on accumulator volume, boron concentration and pressure ensure that the assumptions used for accumulator injection in the safety analysis are i met. l The accumulator power operated isolation valves are considered to be " operating b<iasses in the context of IEEE Std. 279-1971, which requires that bypasses of a protective function be removed automatically whenever permissive conditions are not met. In addition, as these accumulator isolation valves fail to meet single failure criteria, removal of power to the valves is required. l The limits for operation with an accumulator inoperable for any reason except an isolation valve closed minimizes the time exposure of the plant to a LOCA event occurring concurrent with failure of an additional accumulator which may result in unacceptable peak cladding temperatures. If a closed isolation valve cannot be immediately opened, the full capability of one accumulator is not available and prompt action is required to place the reactor in a MODE where this capability is not required. P 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS The OPERABILITY of two ECCS subsystems ensures that sufficient emergency core cooling capability will be available in the event of a LOCA assuming the-loss of one subsystem through any single failure consideration. Either subsystem l operating in conjunction with the accumulators is capable of supplying sufficient core cooling to limit the peak cladding temperatures within acceptable limits for all postulated break sizes ranging from the double ended break of the largest RCS cold leg pipe downward. In addition, each ECCS subsystem provides i long term core cooling capability in the recirculation mode during the accident recovery period. I With the RCS temperature below 350'F, one OPERABLE ECCS subsystem is accept-able without single failure consideration on the basis of the stable reactivity + condition of the reactor and the limited core cooling requirements. i 1 I DIABLO CANYON - UNITS 1 & 2 B 3/4 5-1

i 4 EMERGENCY CORE COOLING $YSTEMS BASES l ECCS SUBSYSTEMS (Continued) The requirement to maintain the RHR Suction Valves 8701 and 8702 in the locked closed condition in MODES 1, 2 and 3 provides assurance that a fire could not cause inadvertent opening of these valves when the RCS is pressur-ized to near operating pressure. These valves are not part of an ECCS subsystem. The limitation for a maximum of one centrifugal charging pump to be 4 OPERABLE and the Surveillance Requirement to verify all centrifugal charging pumps and Safety Injection pumps except the required OPERABLE charging pump to be inoperable beinw 323'F provides assurance that a mass addition pressure transient can be relieved by the operation of a single PORV. The Surveillance Requirements provided to ensure OPERABILITY of each component ensures that, at a minimum, the assumptions used in the safety analyses are met and that subsystem OPERABILITY is maintained. Surveillance requirements for throttle valve position stops and flow balance testing provide assurance that proper ECCS flows will be maintained in the event of a LOCA. Maintenance of proper flow resistance and pressure drop in the piping system to each injection point is necessary to: (1) prevent total pump flow from exceeding runout conditions when the system is in its minimum resistance configuration, (2) provide the proper flow split between injection points in accordance with the assumptions used in the ECCS-LOCA analyses, and (3) provide an acceptable level of total ECCS flow to all injection points equal to or above that assumed in the ECCS-LOCA analyses. 3/4.5.4 BORON INJECTION SYSTEM The Boron Injection System is only required for Unit 1 Cycle 4 and Unit 2 Cycle 3. The OPERABILITY of the Boron Injection System as part of the ECCS ensures that sufficient negative reactivity is injected into the core to counteract any positive increase in reactivity caused by RCS cooldown. RCS cooldown can be caused by inadvertent depressurization, a loss-of-coolant accident or a steam line rupture. The limits on injection tank minimum contained volume and boron concentra-tion ensure that the assumptions used in the steam line break analysis are met. The contained water volume limit includes an allowance for water not usable i. because of tank discharge line location or other physical characteristics. The OPERABILITY of the redundant heat tracing channels associated with the boron injection system ensure that the solubility of the boron solution will i be maintained above the solubility limit of 135'F at 21,000 ppm boron. l l l DIABLO CANYON - UNITS 1 & 2 B 3/4 5-2 Amendment Nos. 51 and 50 .}}