Text

Amends 115 & 113 to Licenses DPR-80 & DPR-82,respectively, Consisting of Changes to TS Responding to 941114 Application & Supplemented by 951207,960202,0528 & 0730 Ltrs
	ML20117E547
Person / Time
Site:	Diablo Canyon
Issue date:	08/19/1996
From:	Steven Bloom; NRC (Affiliation Not Assigned)
To:	;
Shared Package
ML20117E550	List: ML20117E547; ML20117E554;
References
	NUDOCS 9608300056
	Download: ML20117E547 (13)
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UNITED STATES b

y NUCLEAR REGULATORY COMMISSION i

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WASHINGTON, D.C. N1

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PACIFIC GAS AND ELECTRIC COMPANY DOCKET NO. 50-275 DIABLO CANYON NUCLEAR POWER PLANT. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE l

i Amendment No. 115 License No. DPR-80 1.

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

A.

The application for amendment by Pacific Gas and Electric Company j

(the licensee) dated November 14, 1994, as supplemented by letters dated December 7, 1995, February 2, 1996, May 28, 1996, and July 30, 1996, complies with the 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; l

B.

The facility will operate in conformity with the application, the 4

provisions of the Act, and the rules and regulations of the Comission;

)

C.

There is reasonable assurance (1) th'at 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 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 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.

2.

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

and paragraph 2.C.(2) of Facility Operating License No. DPR-80 is hereby l

amended to read as follows:

l i

9608300056 960819 PDR ADOCK 05000275 P

PDR

]

(2)

Technical Snecifications The Technical Specifications contained in Appendix A and the

+

Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 115, are hereby incorporated in the j

license.. Pacific Gas and 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 its date of issuance to be i

implemented within 30 days of issuance.

i FOR THE NUCLEAR REGULATORY Co m ISSION I

Steven D. Bloom, Project Manager 4

i Project Directorate IV-2 1

Division of Reactor Projects III/IV i

i

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Office of Nuclear Reactor Regulation i

Attachment:

Changes to the Technical l

Specifications i

Date of Issuance:

August 19, 1996 w---

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4 UNITED STATES

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j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20666-0001 o

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PACIFIC GAS AND ELECTRIC COMPANY DOCKET NO. 50-323 DIABLO CANYON NUCLEAR POWER PLANT. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 113 License No'. DPR-82 1.

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

A.

The application for amendment by Pacific Gas and Electric Company (the licensee) dated November 14, 1994, as supplemented by letters dated December 7, 1995, February 2, 1996, May 28, 1996, and July 30,1996, complies with the 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 application, the provisions of the Act, and the rules and regulations of the.

Commission; C.

There is reasonable assurance (1) 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 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; 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.

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 Soecifications

}

The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 113, are hereby incorporated in the license.

Pacific Gas and Electric Company shall operate the facility in accordance with the Technical Specifications and the 1

Environmental Protection Plan, except where otherwise stated in i

specific license conditions.

l i

\\

3.

This license amendment is effective as of its date of issuance to be implemented within 30 days of issuance.

i FOR THE NUCLEAR REGULATORY COMMISSION even D. Bloom, Project Manager Project Directorate IV-2 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

August 19, 1996 i

v

ATTACHMENT TO LICENSE AMENDMENTS AMENDMENT NO.115 TO FACILITY OPERATING LICENSE NO. DPR-80 AND AMENDMENT NO.113 TO FACILITY OPERATING LICENSE NO. DPR-82 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 Amendment number and contain marginal lines indicating the areas of change.

The corresponding overleaf pages are also provided to maintain document completeness.

REMOVE INSERT 3/4 3-32 3/4 3-32 3/4 3-33 3/4 3-33 3/4 3-34 3/4 3-34 3/4 3-35 3/4 3-35 B 3/4 3-la B 3/4 3-la

TABLE 3.3 5 (Continuno TABLE ISTATIGtS (1) Diesel generator starting delay not included because offsite power available.

(2) Notation deleted.

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

Diesel generator starting delay not included because offsite (4)

Response time limit includes opening of valves to establish $ power it available.

1 path and attainment of discharge pressure for centrifuge 1 charging pumps (uhere applicable). Sequon-tial transfer of charging pug suction fra the VCT to the 267 (2d5T valves apen, then VCT valves close) is included.

(5) Diesel generator starting and sequence loaditG delays included. Offsite power is not available. Response time limit includes opening of valves to establish $1 path and attainment of discharge pressure for centrifugal charging pumps. Sequential transfer of charging pup suction from the VCT to the fel5T (RW5T valves open then VCT valves close) is included.

(6) The maximum response time of 48.5 seconds is the time from when the containment pressure exceeds the Nigh t:1gh 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 tier is not included. The 80-second maximum spray delay time does not include the time from LDCA start to "P' signal.

(7) Diesel generator starting and sequence loading delays included. Sequential trans.

fer of chargin valves close) g pump suction from the VCT to the RLf5T (RWST valves open then VCT is not included. Response time limit includes ooening of valees to establish SI flow path and attainment of discharge pressare for centrifugal charging pumps, SI. and RHR pugs (where applicable).

(8) Does nat include Trip Time Delays. Response times include the transmitters. Eagle-21 Process Protection cabinets. Solid State Protection System cabinets and actus-tien devices only. This reflects the response times necessary for TERMAL PCWER in excess of 502 RTP.

DIABLO CANYDN tal!TS 1 & 2 3/4 3 31 Amendmei.c Nos. 70 & 69. M 4 N.

M&D

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1 TABLE 4.3-2 l

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION o

SURVEILLANCE REQUIREMENTS Eg TRIP o

ACTUATING n

CHANNEL DEVICE MODES FOR h

CHl.NNEL OPERA-OPERA-MASTER SLAVE WHICH g

CHANNEL CALI-TIONAL TIONAL ACTUATION RELAY RELAY SURVEILLANCE-FUNCTIONAL UNIT CHECK BRATION TEST TEST LOGIC TEST TEST TEST..

IS REQUIRED

! 1. Safety Injection, (Reactor Trip d

Feedwater Isolation. Start Diesel Generators, Containment Fan Cooler Units, and Component Cooling Water) m a.

Manual Initiation N.A.

N.A.

R N.A.

N.A.

1, 2, 3, 4 b.

Automatic Actuation N.A.

N.A.

M(1)

R 1, 2, 3, 4 l

Logic and Actuation Relays c.

Containment Pressure-High S

R Q

N.A.

.N.A.

N.A.

1, 2, 3, 4 d.

Pressurizer Pressure-Low S

R Q

N.A N.A.

N.A.

1, 2, 3 e

LELrTED f.

Steam Line S

R Q

N.A.

1, 2, 3 Pressure-Low

$ $ 2. Containment Spray (coincident with SI signal)

"7 a.

Manual Initiation N.A.

N.A.

R N.A.

N.A.

1, 2, 3, 4 b.

Automatic Actuation Logic N.A.

N.A.

M(1)

R 1, 2, 3, 4 l

g and Actuation Relays 3,

c.

Containment Pressure-S R

Q N.A.

N.A.

1, 2, 3, 4 High-High S.t

.a.t:

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..____m TABLE 4.3-2 (Continued).

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION a

SURVEILLANCE REQUIREMENTS 5'E TRIP o

ACTUATING k"

CHANNEL DEVICE MODES FOR CHANNEL OPERA-OPERA-MASTER SLAVE WHICH E

CHANNEL CALI-TIONAL TIONAL ACTUATION RELAY RELAY SURVEILLANCE FUNCTIONAL UNIT CHECK BRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED 5

! 3. Containment Isolation d

a.

Phase "A" Isolation

1) Manual N.A.

N.A.

R N.A.

N.A.

1, 2, 3, 4

2) Automatic Actuation N.A.

N.A.

M(1)

R 1, 2, 3, 4 l

Logic and Actuation

~

Relays i

3) Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.

b.

Phaso "B" Isolation

1) Manual N.A.

N.A.

R N.A.

N.A.

1, 2, 3, 4

2) Automatic Actuation N.A.

N.A.

M(1)

R 1, 2, 3, 4 l

w1 Logic and Actuation Relays w

de

3) Containment S

R Q

N.A.

1, 2, 3, 4 Pressure-High-High c.

Containment Ventilation Isolation

1) Automatic Actuation N.A.

N.A.

M(1)

R 1, 2, 3, 4 l

FF Logic and Actuation

??

Relays

2) Deleted m-

3). Safety Injection See Item 1. above for all Safety Injection Surveillance Requirements.

ii

a. m.

4) Containment Ventilation EE Exhaust Radiation-High l

EE (RM-44A and 44B)

S R

Q N.A.

N.A.

1, 2, 3, 4 RR an S2 N.N.

aa

$%8 tr om

TABLE 4.3-2 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS

(

E TRIP E'

ACTUATING o

CHANNEL DEVICE MODES FOR E

CHANNEL OPERA-OPERA-MASTER SLAVE WHICH CHANNEL CALI-TIONAL-TIONAL ACTUATION RELAY RELAY SURVEILLANCE FUNCTIONAL UNIT CHECK BRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED

! 4. Steam Line Isolation U

a.

Manual N.A.

N.A.

R N.A.

N.A.

1, 2, 3 b.

Automatic Actuation Logic N.A.

N.A.

M"8 M"3 R

1, 2, 3 I

~

and Actuation Relays m

c.

Containment Pressure-S R

Q N.A.

N.A.

1, 2, 3 High-High d.

Steam Line Pressure-Low S

R Q

N.A.

1, 2, 3 w

e.

Negative Steam Line S

R Q

N.A.

3(3) 1 Pressure Rate-High

5. Turbine Trip and Feedwater Isolation a.

Automatic Actuation N.A.

N.A.

M"3 M"3 R

1, 2 l

Logic and Actuation Relays cc b.

Steam Generator Water S

R Q

N.A.

1, 2 yh Level-High-High

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6. Auxiliary Feedwater a.

Manual N. Y.

N.A.

R N.A.

N.A.

1, 2, 3

((

E 5-b.

Automatic Actuation N.A.

N.A.

M"'

M")

R 1, 2, 3 l

f Logic and Actuation Relays en nan c.

Steam Generator Water g'

Level-Low-low caso

((

1) Steam Generator S

R Q

N.A.

1, 2, 3(5) gg Water Level-Low-Low

2) RCS Loop AT N.A.

R Q

N.A.

1, z wm

TABLE 4.3-2 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIRENENTS O

5 TRIP g

ACTUATING CHANNEL DEVICE MODES FOR ng CHANNEL OPERA-OPERA-MASTER SLAVE WHICH CHANNEL CALI-TIONAL TIONAL ACTUATION RELAY RELAY SURVEILLANCE E FUNCTIONAL UNIT CHECK BRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED i

H 6. Auxiliary Feedwater (Continued)

I d.

Undervoltage - RCP N.A.

R N.A.

N.A.

M.A.

I e.

Ssfety Injection See Item 1. above for all Safety Injection Surveillance Requirements.

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7.

Loss of Power a.

4.16 kV Emergency Bus N.A.

R N.A.

N.A.

1, 2, 3, 4 Level 1 b.

4.16 kV Emergency Bus N.A.

R N.A.

N.A.

-1,2,3,4 Level 2 w

8. Engineered Safety Feature Actuation System Interlocks gg a.

Presscrizer Pressure, M.A.

R --

Q N.A.

N.A.

1, 2, 3 P-ll 77 b.

Deleted kk c.

Reactor Trip, P-4 N.A.

N.A.

R N.A.

N.A.

1, 2, 3 oa I(

TABLE NOTATIONS

=sZ

")

Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS.

h; For the Containment Ventilation Exhaust Radiation-High monitor only, a CHANNEL FUNCTIONAL TEST shall be performed

- (2) 13 g at least once every 31 days.

Trip function automatically blocked above P-II (Pressurizer Pressure Interlock) setpoint and is automatically 03

-b; blocked below P-11 w' en Safety Injection on Steam Line Pressure-Low is not blocked.

d Deleted.

l

"5)

For Mode 3, the Trip Time Delay associated with the Steam Generator Water Level-Low-Low channel must be less than

(

f *'{w j

or equal to 464.1 seconds.

1NSTRUMENTATION 3/4.3.3 MONITORING INSTRUMENTATION RADIATION HONITORING FOR PLANT OPERATIONS LIMITING CONDITION FOR OPERATION 3.3.3.1 shown in Table 3.3-6 shall be CPIRABLE with their Alarm the specified limits.

APPLICABILITY: As shown in Table 3.3-8.

ACTION:

With a radiation monitoring channel Alara/ Trip 5etpoint for plant a.

operations exceeding the value shown in Table 3.3-6, adjust the Setpoint to within the limit within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or declare the channel inoperable.

b.

With one or more radiation monitoring channels for plant operations inoperable, take the ACTION shown in Table 3.3-6.

The provisions of Specification 3.0.3 are not applicable.

c.

l 3URVIILLANCE REQUIREMENTS 4 3.3 1 In:5 radiation monitoring instrumentation channel for plant operations shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, C CALIBRATION and CHANNEL FUNCTIONAL TEST for the MODES and at the fre shown in Table 4.3-3.

DIABlD EANYON - UNITS 1 & 2 3/4 3-36 Amendment Nos. 55 and 54 1

3/4.3 INSTRLMMATION i

RASES e N TOR TRIP SYSTEM and ENGINEERED SAFETY FEATURES 3/A.3J H F8 y ACTUATION 5YSTEM IN5TRUMENTATI(M The OPERABILITY of the Reactor Trip Systes and Engineered $sfety Features Actuation

)

(1) the associated ACT!W and/or System instrumentation and interlocks ensure that:

Reactor trip will be initiated when the parameter son 1 tared by each channel or combina.

tion thereof reaches its Setpoint. (2) tne specified coincidence logic and sufficient redundancy is maintained to permit a channel to be out of service for testing or asin-tenance consistent with natntaining an appropriate level of reliability of the Reactor Protection 6nd Engineered Safety Features instrumentation, and (3) sufficient redundancy 15 maintained to permit a channel to be out of service for testing or maintenance, and j

(4) sufficient system functional capability is available from diverse parameters.

l The OPERABILITY of these systems is required to provide the overall reliability, 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 assugtions used in the accidetit analyses. Tne l

Surveillance Requirements specified for these systems ensure that t periodic surveillance tests performed at the minimum frequencies are sufficie l

demonstrate this capability.

maintenance outage times have been determined in accordance with WCAP 10271.

Evaluation j

of Surveillance Frequencies and Out of Service Times for the Reactor Protection Instru-Surveillance intervals and out of-mentation $ystem." and supplements to that report. service times were determined of the Reactor Protection System.

i The Process Protection System is designed to permit any one channel to be tested I

If a channel has been bypassed for any and maintained at power in a bypassed mode.

purpose. the bypass is continuously indicated in the control roca as required by apphcable codes and standards. As an alternative to testing in the bypass mode.

j 2esting in the trip mode is also possible and permitted.

4 j

f.

Tne Engineered Safety Features Actuation System senses selected plant parametersIf they are. thj eM 0** mines whether or not predetermined limits are being exceeded.

f signals are combined into logic astrices sensitive to combinations indicative of various accidents. events. and transients. Once the required logic combination is cogleted.

tne 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 (1) safety sitigste the consequences of a steam line break or loss of coolant accident:

i injection pugs start and automatic valves position, (2) Reactor trip. (3) feekster isolation (4) startup of the emergency diesel generators ($) containment spray pimps start and automatic valves position (6) containment isolation. (7) steam line isola.

tion. (8) Turbine trip. (9) auxiliary feedwater pumps start and automatic valve posi-l tion. (10) containment fan cooler units start, and (11) cogonent coolig water pumps start and automatic valves position.

The Engineered Safety Feature Actuation System Instrumentation Trip 5etpoints spect-fied in Table 3.3 4 are the nominal values at which the trips are set for each func.

If the functional unit is based on analog herbere, the setpoint is con-sidered to be adjusted consistent with the nominal value when the *as left* setpoint is tional unit.

For all setpoints in digital within the band allowed for calibration accuracy.

baroware. the setpoints are set at the nominal values.

/

i Amendment was. 4t and

DIABLO CANYON WITS 1 & 2 8 3/4 3 1 M&E

_.m.___

INSTRUMENTATION BASES REACTOR PROTECTION SYSTEM and ENGINEERED SAFETY FESTURES ACTUATION SYSTEM j

INSTRUMENTATION (Continued)

To accommodate the instrument drift that may occur between operational j

l tests and the accuracy to which setpoints can be me.sured and calibrated, l

Allowable Values for the setpoints have been specified in Table 3.3-4.

l Operation with setpoints less conservative than the Trip Setpoint, but within the Allowable Value, is acceptable.

I The methodology to derive the Trip Setpoints is. based upon combining all I

of the uncertainties in the channel.

Inherent to the determination of the Trip Setpoints are the magnitudes of these channel uncertainties. Sensor and 4

rack. instrumentation utilized in these channels are expected to be capable of operating within the allowances of these uncertainty magnitudes.

ESF response times specified in Table 3.3-5, which include sequential 8

operation of the RWST and VCT valves (Table Notations 4 and 5), are based on values assumed in the on-LOCA safety analyses. These analyses take credit for injection of borated water from the RWST.

Injection of borated water is assumed not to occur until the VCT charging pump suction isolation valves are closed following opening of the RWST charging pump section isolation valves.

When the sequential operation of the RWST and VCT valves is not included in the 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.

For slave relays in the ESF. actuation system circuit that are Potter &-

)

Brumfield type ER relays, the SLAVE RELAY TEST is performed on a refueling frequency. The test frequency is based on relay reliability assessments presented in WCAP-13878, " Reliability Assessment of Potter and Brumfield HDR Series Relays," WCAP-13900, " Extension of Slave Relay Surveillance Test Intervals," and WCAP-14117, " Reliability Assessment of Potter and Brumfield MDR Series Relays." These reliability assessments are relay specific and apply only to Potter 'and Bruafield MDR series relays. Note that for normally energized applications, the relays may have to be replaced periodically in accordance with the guidance given in WCAP-13878 for MOR relays.

Undervoltage protection will generate a loss of power diesel generator start in the event a loss of voltage or degraded voltage condition occurs.

The diesel generators provide a source of emergency power when offsite power is either unauilable or is insufficiently stable to allow safe unit operation. The first level undervoltage relays (FLURs)-detect the loss of bus voltage (less than 69% bus voltage). The second level undervoltage relays (SLURS) provide a second level of undervoltage protection which protects all Class IE loads from short or long term degradation in the offsite power

-system. The SLUR allowable value is the minimum steady state voltage needed on the 4160 volt vital bus to ensure adequate voltage is available for safety related equipment at the 4160. volt, 480 volt, and 120 volt levels.

DIABLO CANYON - UNITS 1 & 2 B 3/4 3-la Unit 1 - Amendment No. Sb 93,115 Unit 2 - Amendment No. Sh92,113 eo w

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