ML20035A262

Amends 77 & 76 to Licenses DPR-80 & DPR-82,respectively, Adding TS Re Main Feedwater Regulating,Bypass & Isolation Valves to Plant TS

ML20035A262
Person / Time
Site:	Diablo Canyon
Issue date:	03/18/1993
From:	Peterson S Office of Nuclear Reactor Regulation
To:
Shared Package
ML20035A263	List: ML20035A262 ML20035A265
References
NUDOCS 9303250066
Download: ML20035A262 (16)

Text

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91 UNITED STATES i.6

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NUCLEAR REGULATORY COMMISSION

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WASHINGTON. D.C. 206664001 -

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i PACIFIC GAS AND ELECTRIC COMPANY

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DOCKET NO. 50-275 DIABLO CANYON NUCLEAR POWER PLANT. UNIT NO. 1 i

AMENDMENT TO FACILITY OPERATING LICENSE f

Amendment No. 77 i

License No. DPR-80 i

1.

The Nuclear Regulatory Commission (the Commission) has found that:

i A.

The application for amendment by Pacific Gas & Electric Company (the licensee) dated December 21, 1990, as supplemented November 22, 1991 and October 26, 1992, complies with the j

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,

.r B.

The facility will operate in conformity with the application, the-i provisions of the Act, and the rules and regulations of the l

Commission; C.

There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health

-l and safety of the public, and (ii) that such activities will be i

conducted in compliance with the Commission's regulations; i

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-j and E.

The issuance of this amendment is in accordance with:10 CFR Part-51 of the Commission's regulations and all applicable requirements i

have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, l

and paragraph 2.C.(2) of Facility Operating License No. DPR-80 is hereby amended to read as follows:

9303250066 930318 PDR ADOCK 05000275 P

PDR

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(2)

Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 77, 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 is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

.44b Theodore R. Quay, Director Project Directorate V i

Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation i

Attachment:

Changes to the Technical Specifications Date of Issuance:

March 18,1993

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A*t-UNITED STATES j

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- NUCLEAR REGULATORY COMMISSION -

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i PACIFIC CAS AND ELECTRIC COMPANY DOCKET NO. 50-323 DIABLO CANYON NUCLEAR POWER PLANT. UNIT NO. 2 6MENDMENT TO FACILITY OPERATING LICENSE Amendment No.76 License No. DPR-82 1.

The Nuclear Regulatory Commission (the Commission) has found that:-

A.

The application for amendment by Pacific Gas & Electric Company (the licensee) dated December 21, 1990, as supplemented November 22, 1991 and October 26, 1992, 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; i

B.

The facility will operate in conformity with the application, the-i provisions of the Act, and the rules and regulations of the

.j Commission;

.ts C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be l

conducted in compliance with the Commission's regulations; i

t D.

The issuance of this amendment will not be inimical to the common j

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 Commission's regulations and all applicable requirements have been satisfied.

[

2.

Accordingly, the license is amended by changes to the Technical i

Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-82 is hereby l

amended to read as follows:

+

4 i-l 1

4 ;

(2)

Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 76, are hereby incorporated in the license. Pacific Gas & Electric Company shall operate the facility in accordance with the Technical Specifications and the i

Environmental Protection Plan, except where otherwise stated in specific license conditions.

3.

This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION f&n S.

m Theodore R. Quay, Director Project Directorate V Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications t

Date of Issuance:

March 18, 1993 P

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l ATTACHMENT TO LICENSE AMENDMENTS l

l AMENDMENT NO. 77 TO FACILITY OPERATING LICENSE NO. DPR-80 j

i AND AMENDMENT NO. 76 TO FACILITY OPERATING LICENSE NO. DPR-82 i

4 DOCKET NOS. 50-275 AND 50-323 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by i

the captioned amendment number and contain marginal lines indicating the area

)

of change. Overleaf pages are also included, as appropriate.

l i

REMOVE INSERT i

X X

3/4 3-28 3/4 3-28 l

3/4 3-29 3/4 3-29 i

3/4 3-30 3/4 3-30 3/4 3-31 3/4 3-31 l

3/4 7-9b B 3/4 7-2b B 3/4 7-2c l

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INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4.5.1 ACClMJLAT0RS............................................

3/4 5-1 3/4.5.2 ECCS SUBSYSTEMS - T,, GREATER THAN OR EQUAL TD 350'F...

3/4 5-3 3/4.5.3 ECCS SUBSYSTEMS - T,yg LESS THAN350'F..................

3/4 5-7 3/4.5.4 DELETED 3/4.5.5 REFUELING WATER STORAGE TANK............................

3/4 5-11 3/4. 6 CONTAIRMERT SYSTEKS 3/4. 6.1 CONTAINKEhT Co n ta i rme nt I nte gri ty...................................

3/4 6-1 Containment Leakage.....................................

3/4 6-2 Centainment Air Locks...................................

3/4 6-5 Internal Pressure.......................................

3/4 6-7 Air Temperature.........................................

3/4 6-8 Containment Structural Integrity........................

3/4 6-9 Containment Ventilation Systes..........................

3/4 6-10 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS Containment Spray System................................

3/4 6-11 Spray Additive System...................................

3/4 6-12 Contai rment Cooling Sys tem..............................

3/4 6-13 3/4.6.3 CONTAI RKEh7 I SOLATION YALYES............................

3/4 6-15 1

3/4.6.4 COMBUSTIBLE GAS CONTROL Hydrogen Analyzers / Monitors.............................

3/4 6-17 El e ctric Rydrog e n Recombine rs...........................

3/4 6-18 DIABLO CANYON - UNITS 1 & 2 ix Amendment Nos. 73 & 72

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS 3/4.7 PLANT SYSTEMS i

3/4.7.1 TURBINE CYCLE Safety Valves 3/4 7-1 TABLE 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES.... 3/4 7-2 TABLE 3.7-2 STEAM LINE SAFETY VALVES PER LOOP...........

3/4 7-3 Auxiliary Feedwater System...............

3/4 7-4 Auxiliary Feedwater Source...............

3/4 7-6 f

Specific Activity...................

3/4 7-7 TABLE 4.7-1 SECONDARY COOLANT SYSTEM SPECIFIC ACTIVITY SAMPLE AND ANALYSIS PROGRAM..............

3/4 7-8 Main Steam Line Isolation Valves............

3/4 7-9 Steam Generator 10% Atmospheric Dump Valves 3/4 7-9a Main Feedwater Regulating Bypass and Isolation Valves 3/4 7-9b l

3/4.7.2 STEAM GENERATOR PRESSURE / TEMPERATURE LIMITA110N 3/4 7-10 3/4.7.3 VITAL COMPONENT COOLING WATER SYSTEM 3/4 7-11 3/4.7.4 AUXILIARY SALTWATER SYSTEM 3/4 7-12 3/4.7.5 CONTROL ROOM VENTILATION SYSTEM 3/4 7-13 3/4.7.6 AUXILIARY BUILDING SAFEGUARDS AIR FILTRATION SYSTEM 3/4 7-16 3/4.7.7 SNUBBERS

......................... 3/4 7-18 FIGURE 4.7-1 SAMPLE PLAN 2) FOR SNUBBER FUNCTIONAL TEST....... 3/4 7-23 TABLE 4.7-2 SNUBBER VISUAL INSPECTION INTERVAL..........

3/4 7-23a 3/4.7.8 SEALED SOURCE CONTAMINATION 3/4 7-24 l

1 i

DIABLO CANYON - UNITS 1 & 2 x

Amendment Nos. 44 &-65

% L -f4-77 & 76

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TABLE 3. 3-4 (Continued) 33 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES r-7.

Loss of Power n

E (4.16 kV Emergency Bus E!

Undervoltage) x a.

First Level Si 1)

Diesel Start

> 0 volts with a

> 0 volts with a il 2 0.8 second time delay 7 0.8 second time delay ind and

> 2583 volts with a

> 2583 volts with a

-7 10 second time delay

< 10 second time delay na 2)

Initiation of Load Shed One relay One relay

> 0 volts with a

> 0 volts with a j4secondtimedelay 34secondtimedelay-and and o,

> 2583 volts with a

> 2583 volts with a 3:

7 25 second time delay

< 25 second time delay Gith one relay Qith one relay u,

/.

> 2870 volts, instantaneous

> 2870 volts, instantaneous u

b.

Second Level 1)

Diesel Start

> 3600 volts with a

> 3600 volts with a 310secondtimedelay 310secondtimedelay 2)

Initiation of Load Shed

> 3600 volts with a

> 3600 volts with a 320secondtimedelay 320secondtimedelay

!i 8.

Engineered Safety Features Actuation R

System Interlocks a.

Pressurizer Pressure, P-11 5 1915 psig i 1925 psig b.

Lnw-Low T,yg, P-12 increasing 543*F

< 545.8'F r

decreasing 543*F

> 540.2'F c.

Reactor Trip, P-4 H.A.

N.A.

w 9'

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1 TABLE 3.3-5

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ENGINEERED SAFETY FEATURES RESPONSE TIMES i

INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS I.

Manual Initiation a.

Safety Injection (ECCS)

N.A.

1)

Feedwater Isolation N.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.

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 l

1)

Containment Ventilation Isolation N.A.

d.

Steam Line Isolation N.A.

2.

Containment Pressure-High 1

a.

Safety Injection (ECCS) s 27'7'/25

1)

Reactor Trip s2 2)

Feedwater Isolation s 63 l

3)

Phase "A" Isolation s 18"'/28

4)

Containment Ventilation Isolation N.A.

5)

Auxiliary Feedwater s 60

6)

Component Cooling Water s 38"'/48' '

7)

Containment fan Cooler Units s 4 0

8)

Auxiliary Saltwater Pumps s 48"'/58

1 3.

Pressurizer Pressure-Low a.

Safety Injection (ECCS) s 27/2 5/35 '

5 1)

Reactor Trip s2 2)

Feedwater Isolation s 63 l

3)

Phase "A" Isolation s 18"'

f 4)

Containment Ventilation Isolation N.A.

5)

Auxiliary Feedwater s 60

6)

Component Ccoling Water s 48

s 40/38"'

7)

Containment Fan Cooler Units 8)

Auxiliary Saltwater Pumps s 58/48"'

DIABLO CANYON - UNITS I & 2 3/4 3-28 Amendment Nos. & % M & M, 77 & 76

TABLE 3.3-5 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 4.

Differential Pressure Between Steam Lines-High a.

Safety Injection (ECCS) s 25"'/35(5) 1)

Reactor Trip s2 2) feedwater Isolation s 63 3)

Phase "A" Isolation s 18"'/28(3' 4)

Containment Ventilation Isolation N.A.

5)

Auxiliary Feedwater s 60(3' 6)

Component Cooling Water s 38")/48(3)

)

7)

Containment Fan Cooler Units s 40"8 8)

Auxiliary Saltwater Pumps s 48 /58(3' 5.

Steam Flow in Two Steam Lines - High Coincident with T,,-Low-Low a.

Safety Injection (ECCS) s 25"'/35(5)

I)

Reactor Trip 54 l

2)

Feedwater Isolation s 65"'/30(3) 3)

Phase "A" Isolation s 20 4)

Containment Ventilation Isolation N.A.

5)

Auxiliary Feedwater s 60(3) 6)

Component Cooling Water 5 40"'/50(3) 7)

Containment Fan Cooler Units s 40(3 8)

Auxiliary Saltwater Pumps s 50")/60(3) b.

Steam Line Isolation 5 10 6.

Steam Flow in Two Steam Lines-High Coincident with Steam Line Pressure-Low a.

Safety Injection (ECCS) s 25"'/35(3) r 1)

Reactor Trip s2 1

2)

Feedwater Isolation s 63"'/28(3) 3)

Phase "A" Isolation s 18 4)

Containment Ventilation Isolation N. A. 3) 5)

Auxiliary Feedwater s 60"(

6)

Component Cooling Water s 38 '/48(33 3

7)

Containment Fan Cooler Units s 40 '

8)

Auxiliary Saltwater Pumps s 48"'/58(3) b.

Steam Line Isolation s8 DIABLO CANYON - UNITS 1 & 2 3/4 3-29 Amendment Nos. -Si & +0-77 & 76

TABLE 3.3-5 (Continued) i ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 7.

Contsinment Pressure-High-High a.

Containment Spray s 4 8. 5

b.

Phase "B" Isolation N.A.

c.

Steam Line Isolation es 7 8.

Steam Generator Water Level-High-High a.

Turbine Trip s 2.5 b.

Feedwater Isolation s 66 l

9.

Steam Generator Water Level Low-Low l

i a.

Motor-Driven Auxiliary Feedwater Pumps s 60

b.

Turbine-Driven Auxiliary Feedwater Pump s 60

10. RCP Bus Undervoltage Turbine-Driven Auxiliary feedwater Pump s 60 t

11. Plant Vent Noble Gas Activity-High*

1 Containment Ventilation Isolation s 11

12. Containmert Ventilation Exhaust Radiation-High" r

Containment Ventilation isolation s 11 "The requirements for Plant Vent Neble Gas Activity-High are not applicable following installation of RM-44A and 44B.

"The requirements for Containment Ventilation Exhaust Radiation-High are

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applicable following installation of RM-44A and 44B.

1 DIABLO CANYON - UNITS 1 & 2 3/4 3-30 Amendment Nos. M- & -69; 42- &-7+

77 & 76 3

r

TABLE 3.3-5 (Continued)

TABLE NOTATIONS (1) Diesel generator starting delay not included because offsite power available.

(2) Notation deleted.

l (3) Diesel generator starting and loading delays included.

(4) Diesel generator starting delay not included because offsite power _ lish SI is avail path able. Response time limit includes opening of valves to estab 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 i

power is not available. Response time limit includes opening of valves to establish SI path and attainment of discharge pressure for centrifugal charging pumps. Sequential transfer of charging pump suction from the 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-tainment pressure exceeds the High-High Setpoint until the spray pump is started and the discharge valve travels to the fully open position assum-ing 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 does not include the time from LOCA start to "P" signal.

(7) Diesel generator starting and sequence loading delays included. Sequen-tial 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 51 flow path and attainment of discharge pressure for centrifugal charging pumps, 51, and RHR pumps (where applicable).

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.i DIABLO CANYON - UNITS I & 2 3/4 3-31 Amendment Nos. f4F & 6+, 7G- & -Al i

77 & 76

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TABLE 4.3-2 E

8 ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION 50RVLILLANCL REQUIRLMENis 94 E

TRIP ANALOG ACTUATING CHANNEL DEVICE MODES FOR c

CHANNEL OPERA-OPERA-MASTER SLAVE milch h

CHANNEL Call-TIONAL TIONAL ACTUATION RELAY RELAY SURVEILLANCE m FUNCTIONAL UNIT CHECK DRATION TEST TEST LOGIC TEST TEST TEST 15 REQUIRED

[1.

Safety injection, (Reactor Trip Feedwater Isolation, Start y

Diesel Generators, Containment Fan Cooler Units, and Component Cooling Water) a.

Manual Initiation N.A.

N.A.

N.A.

R N.A.

N.A.

N.A.

1, 2, 3, 4 b.

Automatic Actuation N.A.

N.A.

N.A.

N.A.

M(1)

N(1)

Q 1,2,3,4 logic and Actuation Relays c.

Containment Pressure-S R

Q N.A.

N.A N.A.

N.A.

1, 2, 3, 4 High d.

Pressurizer Pressure-Low 5

R Q

N. A N.A.

N.A.

N.A.

1,2,3 e.

Differential Pressure 5

R Q

N.A.

N.A.

N.A.

N.A.

1, 2, 3 Between Steam Lines-High f.

Steam Flow in Two Steam 5

R Q

N.A.

N.A.

N.A.

N.A.

1,2,3 Lines-High Coincident

[

With Either y

1) T,yg-L e low, or S

R Q

N.A.

N.A.

N.A.

N.A.

1, 2, 3

2) Steam Line 5

R Q

N.A.

N.A.

N.A.

N.A.

1, 2, 3 l

Pressure-Low I$ 2.

Containment Spray

[

a.

Manual Initiation N.A.

N.A.

N.A.

R N.A.

M.A.

N.A.

1, 2, 3, 4 E

b.

Automatic Actuation N.A.

N.A.

N.A.

N.A.

M(1)

M(1)

Q 1,2,3,4 g

logic and Actuation Relays c.

Containment Pressure-5 R

Q N.A.

N.A.

N.A.

N.A.

1, 2, 3 High-High

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PLANT SYSTEMS MAIN FEEDWATER REGULATING. BYPASS AND ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.7.1.7 In each feedwater line, each Main feedwater Isolation Yalve (MFIV) shall be OPERABLE or closed.

Each Main Feedwater Regulating Valve (MFRV), and i

MFRV bypass valve shall be OPERABLE, closed, er isolated.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

With one MFIV, one MFRV, or one MFRV bypass valve inoperable either:

a.

Restore the inoperable valve to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or b.

Close the inoperable valve within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or c.

If the inoperable valve is a MFRV or MFRV bypass valve, isolate the inoperable valve with at least one closed valve within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or d.

Be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SVRVEILLANCE RE0UIREMENTS 4.7.1.7.1 Each MFRV and MFRV bypass valve shall be demonstrated OPERABLE by determining the isolation time of each valve to be less than or equal to 7 seconds (not including instrument delays) at least each COLD SHUTDOWN but not more frequently than once per 92 days.

4.7.1.7.2 Each MFIV shall be demonstrated OPERABLE by determining the isolation time of each valve to be less than or equal to 60 seconds (not including instrument delays) when tested pursuant to Specification 4.0.5.

t DIABLO CANYON - UNITS 1 & 2 3/4 7-9b Amendment Nos. 77 & 76 l

9 1

PLANT SYSTQiS BASES 3 /4.7.1.7 MAIN FEEDWATER REGULATING. BYPASS AND ISOLATION VALVES The OPERABILITY of the Main Feedwater Isolation Valves (MFIVs), Main feedwater Regulating Valves (MFRVS), and MFRV bypass valves ensures that the valves will be capable of performing their intended safety function. The safety function of these valves is to rapidly close following: (1) a steam line or feedwater line rupture, thereby limiting the Reactor Coolant System cooldown and limiting the total energy release to the containment; or (2) a feedwater system malfunction, thereby limiting Reactor Coolant System cooldown.

The analysis of excessive RCS heat removal due to a feedwater system malfunction assumes that a control system malfunction or operator error causes a MFRV and associated bypass valve to open fully, resulting in a step increase in feedwater flow to one steam generator. The analysis assumes a feedwater isolation signal is generated by a high-high steam generator level. Feedwater isolation is assumed to occur as a result of the MFRV and associated bypass valve closing as a result of the feedwater isolation signal.

Rupture of a steam line is analyzed to calculate the response of the reactor core and to determine the resulting mass and energy releases. Two separate analyses are performed since conservative assumptions for the core response analy5.is are different than the conservative assumptions for the mass and energy release analysis. The core response analysis credits feedwater isolation as a result of the safety injection signal which results in a feed-water isolation signal. Feedwater isolation is assumed to occur as a result of closure of all MFRVs and MFRV bypass valves.

The mass and energy release analysis consists of several cases. The analysis assumes feedwater isolation occurs as a result of the safety injec-tion signal which results in a feedwater isolation signal. Some cases are analyzed that assume a MFRV fails and feedwater isolation occurs as a result of closure of the MFIV. For cases with other single failure assumptions, feedwater isolation is assumed to occur as a result of closure of all MFRVs and MFRV bypass valves.

The core response and mass and energy releases that would result from a rupture of a main feedwater line are bounded by the analyses of the rupture of a main steam line.

l 1

DIABLO CANYON - UNITS 1 & 2 B 3/4 7-2b Amendment Nos. 77 & 76 l

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i PLANT SYSTEMS BASES 3/4.7.I.7 MAIN FEEDWATER REGULATING. BYPASS AND ISOLATION VALVES (continued)

The OPERABILITY of the MFIVs, MFRVs, and MFRV bypass valves within the closure time of the surveillance requirements is consistent with the assump-tions used in the safety analyses. When these valves are closed, they are performing their safety function.

The APPLICABILITY of this specification is MODES I, 2, and 3.

The basis for this is that in MODES I and 2 there is significant energy and in MODE 3 there may be significant energy in the Steam Generators. With significant energy in the Steam Generators the valves are needed for isolation of the Steam Generators in the event of a secondary system pipe rupture.

The ACTION statement requires that an inoperable valve either be restored to an OPERABLE condition or closed within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Closing the valve fulfills the safety function of feedwater isolation so the ACTION Statement can be exited.

If a MFRV or a MFRV bypass valve is inoperable, another option is available to isolate the inoperable valve with at least one closed valve within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. This option is not available for the MFIVs since the MFIVs are in the Class I feedwater piping and there are no other valves, other than i

check valves, in the Class I piping that could be closed to isolate the Class I portion of the feedwater line.

)

DIABLO CANYON - UNITS I & 2 B 3/4 7-2c Amendment Nos. 77 & 76