Text

Amends 70 & 69 to Licenses DPR-80 & DPR-82,respectively, Revising TS to Support Comprehensive Program to Upgrade Plant RMS
	ML20095H091
Person / Time
Site:	Diablo Canyon
Issue date:	04/20/1992
From:	Gagliardo J; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20095H095	List: ML20095H091; ML20095H120;
References
	NUDOCS 9204290239
	Download: ML20095H091 (20)
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UNITED STATES E'

1 NUCLEAR REGULATORY COMMISSION

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9 l

W ASHINGTON, D. C. 20565

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PACIFIC GAS MD ELECTRIC COMPANY DIAELO CANYON NUCLEAR POWER PLANT. UNIT NO. 1 DOCKET NO. 50-275 l

AMENDMENT TO FACILITYfpMTING LICENSE Amendment No. 70 License No. DPR-80 1.

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

A.

The applicatior, for amendment by Pacific Gas & Electric Company (the licensee) dated Juna 5, 1991, 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 1;

B.

The facility will operate in conformity with the application, the provisions of the Act, and the regulations of the Comission; C.

There is reasonable assurance (i) ti4at the activities auth:rized 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 Commission'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 Comission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technit.a1 Specif t-cations 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:

9204290239 920420 PDR ADOCK 05000275 P

PDR

~ (2) lethnical Specilicaljmn The leci;nical Specifications contained in Appendix A and the Environmen-tal Protc: tion Plan contained in Appendix B, as revised through Amend-ment No. 70, are hereby incorporated in the license.

Pacific Gas &

Electric Cottpany shall operate the facility in accordance with the Technical Specifications and the Environ, mental Protection Plan, except where otherwise stated in specific license conditions.

3.

This license amendment becomes effective as of the date of its issuance, f R T E NUCLEAR REGULATORY COMMISSION t

1 o

)ames

. Gagliardo, Acting Director Proje Directorate V Division of Reactor Projects Ill/IV/V Office of Nuclear Reactor Regulation

Attachment:

Lhanges to the Technical Specifications Date of Issuance: April 20, 1992 s

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3 7 UNITED STATES

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NUCLEAR REGULATORY COMMISSION

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W ASHINoToN, D.C. 20f46 o.g

%.... +,.e PACIFIC GAS AND E RCTRIC COMPANY i

l DIABLO CANYON NUCLEAR POWER PLANT. UNIT NO. 2 DOCKEl' NO. 50-3r3 AMENDMEN1 TO FACILITY OPERATING LIC,G1E Amendmer.t No. 69 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 June 5, 1951, 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;

5 B.

The facility will operate in conformity with the application, the provisions of the Act, and the 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 Commission'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 Specifi-cations as indicated in the :,ttachment 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 Specificationt The Technical Specifications contained in Appendix A and the Environmen-tal Protection Plan contained in Appendix 0, as revised through Amend-ment No. 69, 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 as of the date of its issuance.

FR HE NdCLEAR REGULATORY COMMISSION b/l x I/(.

a f

Jame E. Gagliardo, Acting Director Proj ct Directorate V Division of Reactor Projects Ill/IV/V Office of Nuclear Reactor Regulation

)

Attachment:

Changes to the Technical t

Specifications Date of Issuance: April 20, 1992

ATTACHMENT TO LICENSE AMENDMENT NOS. 70 AND 69 FACILITY OPERATING LICENSE N05. DPR-80 AND 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 the captioned amendment number and contaih marginal lines indicating the area of change.

Overleaf pages are also included, as appropriate.

REMOVE PAGE INSERT PAGE 3/4 3-17 3/4 3-17 3/4 3-25 3/4 3-25 3/4 3-26 3/4 3-26 3/4 3-30 3/4 3-30 3/4 3-33 3/4 3-33 3/4 3-37 3/4 3-37 3/4 3-39 3/4 3-39 3/4 9-10 3/4 9-10 B 3/4 9-3 B 3/4 9-3 l

.4 n-

ks%

}

k('k TABLE 3.3-4 (Continued)

%Q ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS o

fi$

5 h w$

~

W C

FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES

/3 S$

Coincident With Either

"?N 4 O

Aj

,add M 1)

T,yg-tcw-Low, or

> 543"f

> 540.2*

g izi 2)

Steam Line Pressure-Low

> 600 psig

> 580 psig fff) g, 5.

Turbine Trip and Feedwater Isolation (p.:

pj a.

Q m

a.

Autoraatic Actuation Logic N.A.

N.A.

and Actuation Relays p

.y b.

Steam Generator Water level-

< 67% of narrow range

< 68% of narrcw range High-High Instrument span each steam Instrument span each steam R

generator.

y 6.

Auxiliary Feedwater M

a.

flanual N.A.

N. A.

b.

Automatic Actuation Logic N.A.

N.A.

and Actuation Relays y,

c.

Steam Generator

> 7.2% of narrow range

> 6.2% of narrow range g

Water Levei-Low-Low instrunent span each Instrument span each Eag steam generator.

steam generator.

$k d.

Undervoltage - RCP

> 8050 volts

> 7935 volts

[

e.

Safety Injection See Item 1. above for all Safety Injection Trip Setpoints and Allowable Values.

N n.

tna

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) 1 25(4)/35(5) l 1)

Rearter Trip

<2 2)

Feedwater Isolation I 63(2) 3)

Phase "A" Isolation 7 18(3)/28(I) l 4)

Containment Ventilation Isolation N.A.(3) 5)

Auxiliary Feedwater

< 60 6)

Component Cooling Water I 38II)

I3)/48(3) 7)

Containn.ent Fan Cooler Units I 40 8)

Auxiliary Saltwater Pumps 348(3)/58I3) 5.

Steam Flow in Two Steam Lines - High coincident with T,y-Low-Low Safety Injection (ECCS) a.

i 25I4)/35(5) l 1)

Reactor Trip

<4 2)

Feedwater Isolation I 65(2) 3)

Phase "A" Isolation i 20(3)/30I3) l 4)

Containment Ventilation Isolation 5)

Auxiliary Feedwater R.A.53)

< 60 6)

Component Cooling Water I 400}

0)/50(3) 7)

Containment Fan Cooler Units i 40 8)

Auxiliary Saltwater Pumps

{500)/60I3) b.

Steam Line Isolation i 10 6.

Steam Flow in Two Steam Lines-High Coincident with Steam tine Pressure-Low Safety Injection (ECCS) a.

1 25I4)M',(5) l 1)

Reactor Trip

<2 2)

Feedwater Isolation i 630) 3)

Phase "A" Isolation i 180)/28II) 4)

Containment Ventilation Isolation 5)

Auxiliary Feedwater R.A.I3)

< 60 6)

Component Cooling Water I 3800)/48I3) 7)

Containment Fan Cooler Units I 40

)

8) - Auxiliary Saltwater Pumps 5480)/58I3) b.

Steam Line Isolation i8 DIABLO CANYON - UNITS 1 & 2 3/4 3-29 Amendment Nos. Si and 50

3BLE3.3-5(Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIAYING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 7.

Containment Pressure-High-High a.

Containment Spray

< 48.5(6) b.

Phase "B" Isolation R.A.

c.

Steam Line Isolation

<7 8.

Steam Generator Water Level-High-High a.

Turbine Trip

< 2. 5 66(2) b.

Feedwater Isolation 9.

Steam Generator Water Level Low-Low a.

Motor-Driven Auxiliary Feedwater Pumps

< 60 b.

Turbine-Oriven Auxiliary Feedwater Pump

< 60 10.

RCP Bus Undervoltage Turbine-Driven Auxiliary Feedwater Pump

< 60 11.

Plant Vent Noble Gas Activity-High(a)

Containment Ventilation Isolation

11 12.

Containment Ventilation Exhaust Radiation-Higb(b)

Containment Ventilation Isolation

< 11 l

l'

[

I l

l (a)The requirements for Plant Vent Noble Gas Activity-High are not applicable.

i following installation of itM-44A and 44B.

.(b)The requirements for Containment Ventilation Exhaust' Radiation-High are

[

applicable following installation of.RM-44A and.48.

f 4

S DIABLO CANYON - UNITS 1 & 2 3/4 3-30~

Amendment Nos. 70 & 69 f

4e mer's-

.-vw-m--e as-w w w-m,nymw-es'rrihM'

.944aT-"p'v-t*terFw-dNMy-f-w=Teytree '

h+++'r gv w ew. v9 r'g-g-%-.yy==g-#.-gyq m, r 4hpya a

y+e-os-re---w m twr ve 7-gyggyg-+-W m-t=gie gww vte'=g-g pN':""

9 e g rt' p-e'e' + y ' r 4ty.W M S 't-

TABLE 4.3-2 (Continued)

ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATIO_N_

og SURVEllLANCE REQUIREMENTS to o

n TRIP g

ANALOG ACTUATING g

CHANNEL DEVICE MODES FOR z

CHANNEL OPERA-OPERA-MASTER SLAVE WHICH CHANNEL CALI-TIONAL TIONAL ACTUATION RELAY RELAY SURVEILI.ANCE g

FUNCTIONAL UNIT CHECK BRATIGy TEST TEST LOGIC TEST TEST TEST IS REQUIRED U

3.

Containment Isolation a.

Phase "A" Isolation l

[

1) Manual N.A.

N.A.

R N.A.

N.A.

1,2,3,4

2) Automatic Actuation N.A.

N.A.

M(1)

Q 1, 2, 3, 4 Logic and Actuation Relays

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

b.

Phase "B" Isolation

1) Manual N.A.

N.A.

N. A.

R N.A.

N.A.

N A.

1, 2, 3, 4 w1

2) Automatic Actuation N.A.

N.A.

N. A.

N.A.

M(1)

Q 1,2,3,4 logic and Actuation w

Relays w"

3) Containment S

R Q

N.A.

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

Containment i ntilation Isolation

1) Automatic Actuation N.A.

N.A.

M(1)

Q 1, 2, 3, 4 Logic and Actuation Relays y>

2) Plant Ver.t Noble Gas S

R M(2)

N.A.

1, 2, 3, 4 05 Activity-High (RM-14A h

and 148)(a) s

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

cn "

4) Containment Ventilation E

Ext.aust Raolation-High (RM-44A and 448)I )

S R

M(2)

N.A.

1, 2, 3, 4 A

E (a)The requirerents for Plant Vent Noble Gas Activity-High (PM-14A and 148) are not applicable following installation m

of RM-44A and 448.

(b)The requirements for Containment Ventilation Exhaust r diation-High (RM-44A ar.d 448) are applicable following c

installation of RM-44A'and 448.

i c,

E TABLE 4.3-2 (Continued) of ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION

()

9 SURVEILLANCE REQUIREMENIS 5

c, z

TRIP ANALOG ACTUATING CHANNEL DEVICE t

si NODES FOR CHANNEL OPERA-OPERA-NASTER SLAVE WHICH CitANNEL CALI-TIONAL TIONAL ACTUATION RELAY RELAY SUP.VEILLANCE 0' FUNCTIONAL UNIT CHECK BRATION TEST TEST LOGIC TEST TEST TEST IS REQUIRED p.4 Steam Line Isolatfon w

r*

a.

Manual N.A.

N.A.

d. A.

R N.A.

N.A.

1, 2, 3 b.

Automatic Actuation Logic N.A.

N.A.

M(1)

Q 1, 2, 3 and Actuation Relays c.

Containment Pressure-S R

Q N.A.

N.A.

N. A.

1, 2, 3 50 High-High

[,'

d.

Steam Flow in Two Steam S

R Q

N. A.

N.A.

1, 2, 3 J,

Lines-H'-5 Coincident With Eitner

1) T,yg-Low-Low or S

R Q

N.A.

1, 2, 3 s

2) Steam Line S

R Q

N.A.

N. A.

1, 2, 3 Pressure-Low l[5.

Turbine Trip and Feedwater Isolation El t

a.

Automatic Actuation N.A.

N.A.

M(1)

N(1)

Q 1, 2

-(

Lc';ic and Actuation Relays

[

b.

Steam Generator Water 5

R Q

N.A.

M.A.

N.A.

N. A.

1, 2

[

Level-High-High o

6 I

'6.

Auxiliary feedwater ag a.

Manual N.A.

N.A.

R N.A.

N.A.

N. A.

1, 2, 3

{;

b.

Automatic Actuation N.A.

N.A.

H. A.-

M(1)

Q 1, 2, 3 Logic and Actuation Relays c.

Steam Generator Water 5

R Q

N.A.

M.A.

N.A.

1, 2. 1 1-Low-Low

~

v n

9.

TABLE 3.3-6 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS o

Ep MINIMUM o

CHANNELS APPLICABLE ALARM / TRIP Q

INSTRUMENT OPERAPij MODES SETPOINT ACTION g

1.

Fuel Handling Building

c a.

Storage Area b

1) Spent Fuel Pool 1

< 75 mR/hr 30 & 32.,(a)

~

315 Whr 30 & 32**I")

E

2) New Fuel Storage 1

Z b.

Gaseous Activity Fuel Handling Building 1

Per the ODCP 32**

' Ventilation Mode Change (d) 2.

Control Room m

Ventilation Mode Change 2***

All 3.

Containment

- 2 mR/hr 34 a.

Gaseous Activity

1) Containment 1

6 Per the ODCP 33 R

Ventilation I olation Y

( & l4A or 14B)(b)

2) RCS Leakage 1

1,2,3,4 N.A.

31

3) Containment Venti-1 6

Per the ODCP 33 I

lation Isolatio I

(RM-44A or 445)gc) b.

Particulate Activity

1) Containment Venti-1 6

Per the 00CP 33 lationIsolatiogC)

Og (RM-44A or 448)

,g

2) RCS Leakage 1

1, 2, 3, 4 N.A.

31 E.E

$5

With fuel in the spent fuel pool or new fuel storage vault.

J

- With irradiated fuel in the spent fuel pool.

o.=

- - 0ne channel for each normal intake to the Control Room Ventilation System (common to both units).

g, (a) Action 32 is not applicable to the Fuel Storage Area Monitors following installation of RE-45A and 453.

(b)The requirements for Containment Ventilation Isolation (RM-14A or 148) are not applicable following o=

ma installation of RM-44A and 448.

E (c)The requirements for Containment Ventilation Isolation (RM-44A or 448) are applicable following

=

installation of RM-44A and 44B.

(d)The requirements for Fuel Handling Building Ventilation Mode Change are applicable following installation of RM-45A and 458.

TABLE 3.3 6 (Continued)

ACTION STATEMENTS I

ACTION 30 -

With less than the Minimum Channels CPERABLE requirement, opera-tion say continue for up to 30 days provided an appropriate portable continuous monitor with the same Alars Setpoint or an individual qualified in radiation protection procedures with a radiation dose rate sonitoring device is provided in the fuel storage pool area.

Restore the inoperable sonitors to CPERA8LE status within 33 days or suspend all operations involving fuel movement in the fuel storage pool areas.

ACTION 31 -

With the number of OPEMBLE channels less than required by the Minimus Channels OPERABLE requirement, comply with the ACTION requirements of Specification 3.4.6.1.

The provisions of Specification 3.0.4 are not applicable.

ACTION 32 -

With the number of CPERABLE channels less il9n required bv the Minimum Channels OPEMBLE renutrement, comply with the ACh0N requirements of specification' 3.9.12.

ACTION 33 -

With the number of OPEMBLE channels less than required by the Minimum Channels OPERABLE requirement, conply with the ACTION requirements of Specification 3.9.9.

t ACTION 34 -

With the number of OPERABLE channels less than required by the Minimum Channels OPERABLE requirement, within I hout initiate and enintain operation of the Control Room Ventilation Syster in a recirculation mode with the HEPA filter and charcoal adsort,er bank in operation.

I i

4 b

i l

1 t

i f

D?ABLO CANYON - UNITS 1 & 2, 3/4 3 38 Amendment Nos. 55 and 54 i

t t

.. u.

f l

!f:;tE. F 3 RADIATION MONITORING INSTRUMENTATION FOR PLANT OPERATIONS SURVEILLANCE REQUIREMENTS o

i E

CHANNEL MODES FOR WHICH I

CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE i

9 CHECK CALIBRATION TEST IS REQUIRED L

4 i

E 1.

Fuel Handling Building a.

Storage Area c

1)

Spent Fuel Pool S

R M

?!

2)

New Fuel Storage 5

R M

[

d b.

Gaseous Activity j

Fuel Handling Building R

M Ventilation Mode Change (c) S g

e.

to 2.

Control Room Ventilation Mode Change S

R M

All 3.

Containment i

{

a.

Gaseous Activity 1)

Containment S

.R M

6

{

1

i Ventilation Isolation Y

(RM-14A or 148)(a) r 2)

RCS Leakage S

_R M

1,2,3,4 l

3)

Containment Venti-S R

M 6

(

1ation Isolation (RM-44A or 448)(b)

S R

M 6

b.

. Particulate Activity

[

'1)

Containment Venti-S R

M

. 6 i

lation Isolation

.(RM-44A or 448)(b)

^

2)

RCS Leakage S

R

.M 1, 2, 3, 4 l

F s

y

'k

With fuel in the spent fuel pool or new fuel storage vault.

?

(a)The requirements for Containment Ventilation Isolation (RM-14A or 148) are not applicable

[

t following installation of RM-44A 'and 448.

2

}

i g

(b)The requirements for Containment Ventilation Isolation (RM-44A or 448) are applicable following installation of RM-44A and 448.

(c)The requirements for Fuel Handling Building Ventilation Mode Change are applicable o

following insta11stion of RM-45A and 458.

o-I.

t i

.w

.---e.-

,,.--,n

- ~

INSTRUMENTATION MOVABLE INCORE DETECT 0P.S LIMITING CONDITION FOR OPERATION 3.3.3.2 The Movable Incore Detection System shall be OPERABLE with:

At least 75% of the detector thimbles, a.

b.

A minimum r.f two detector thimbles per core quadrant, and Sufficient movable detectors, drive, and readout equipment to map c.

these thimbles.

APPLICABILITY: When the Movable Incote Detection System is used for:

Reca11 oration of the Excore Neutron Flux Detection System, or a.

b.

Monitoring the QUADRANT POWER TILT RATIO, or MeasurementofF[g,F(Z)andF c.

q xy.

ACTION:

With the Movable Incore Detection System inoperable, do not use the system for the above applicable monitoring or calibration functions.

The provisions of Specification 3.0.3 are not applicable.

l SURVEILLANCE REQUIREMENTS 4.3.3.2 The Movable Incore Detection System shall be demonstrated OPERABLE at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months by normalizing each detector output when requirtd for:

Recalibration of the Excore Neutron Flux Detection System, er a.

b.

Monitoring the QUADRANT POWER TILT RATIO, or H

c.

Measurement of F H' I (Z) and Fxy.

Q 4

DIABtC CANYON - UNITS 1 & 2 3/4 3-40 Amendment Hos. 55.and 54 4

Q

\\

REFUELING OPERATIONS

.I LOW WHER LEVEL LIMITING CONDITIOM FOR OPERATION 3.9.8.2 Two independent residual heat retaoval (RHR) trains shall be OPERABLE and at least one RHR train shall be in operation.*

APPLICABILITY:

MODE 6, when the water level above the top of the reactor i

vessel flange is less than 23 feet.

ACTION:

With less than the required RHR trains OPERABLE, immediately initiate a.

corrective action to return the required RHR trains to OPERABLE status, or to establish at least 23 feet of water above the reactor vessel flange, as soon as possible.

b.

With no RHR-train in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required RHR train to operation.

Close all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVE)LLANCE REQUIREMENTS 4.9.8.2.1 With the reactor subcritical less than 57 hours6.597222e-4 days 0.0158 hours 9.424603e-5 weeks 2.16885e-5 months , at least one RHR l

train shall be verified in operation and circulating reactor coolant at a flow rate of greater than or equal to 3000 gpm at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .-

4.9.8.2.2 With the reactor suberitical for 57 hours6.597222e-4 days 0.0158 hours 9.424603e-5 weeks 2.16885e-5 months or more, at least one RHR train shall be verified in operation and circulating reactor coolant at a flow rate of greater than or equal to 1300 gpm at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Prior to initial criticality, the RHR train may be removed'from operation for up to I hour per 8-hour period during the performance of CORE ALTERATIONS in the vicinity of'the reactor vessel hot legs.

DIABLO CANYON - UNITS 1 & 2-3/4 9-9 AMENDMEh7 WOS. 28 AND 27

REFUELING OPERATIONS 3/4.9.9 CONTAINMENT VENTILATION ISOLATION SYSTEM LIMITING CONDITION FOR OPERATION 3.9.9 The Containment Ventilation Isolation System shall be OPERABLE.

APPLICABIIITY:

Dur',g CORE ALTERATIONS or movement of irradiated fuel within containmenET~

ACTION:

With the Containment Ventilation Isolation System inoperable, close a.

each of the ventilation penetrations providing direct access from the containment atmosphere to the outside atmosphere.

b.

The provisions of Specification 3.0.3 are not applicable.

SURVEILLANCE REQUIREMENTS 4.9.9 The Containment Ventilation Isolation Syste:s shall be demonstrated OPERABLE within 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to the start of and at least once per 7 days during CORE ALTERATIONS by verifying that containment ventilation isolation occurs on a High Radiation test sign monitoringinstrumentationchannels.g})fromtheplantventnoblegasactivity l

(a) Following installation of RM-44A and 44B, the high radiation test signal shall come from the containment ventilation exhaust radiation monitoring instrumentation channels.

DIABu0 CANYON - UNITS 1 & 2 3/4 9-10 Amendment Hos. 55 and 54 70 and 69

9 F UELING OPERATIONS BASES 3/4.9.9 CONTAINMENT VENTILATION ISOLATION SYSTEM The OPERABILITY of this system ensures that the containment ventilation penetrations will be automatically isolated upon detection of high radiation levels within the containment.

The OPERABILITY of this system is required to restrict the releast of radioactive meterial frcm the containment atmosphere to the environment.

3/4.9.10 and 3/4.9.11 WATER LEVEL - REACTOR VESSEL tind SPENT FUEL POOL The restrictions on minimum water level ensure that sufficient water depth is available to remove 99% of the assumed 10% iodine gap activity released from the rupture of an irradiated fuel assembly. The minimum water depth is consistent with the assumptions of the safety analysis.

The minimum water level for movement of fuel assemblies (23 feet above the vessel flange) assures that sufficient water depth is maintained above fuel elements being moved to or from the vessel. With the upper internals in place, fuel assemblies and control rods cannot be removed frort the vessel.

Operations involving the unlatching of control rods with the vessel upper internals in place may proceed with less than 23 feet of water above tha vessel flange provided that 23 feet of water (12 feet above the flange) is maintained above all irradiated fuel assemblies within the reactor vessel.

3/4.9.12 FUEL HANDLING BUILDING VENTILATION SYSTEM The limitations on the Fuel Handling Building Ventilation System ensure that all radioactive material released from an irradiated fuel assembly will be filtered through the HEPA filters and charcoal adsorber prior to discharge to the atmosphere.

The OPERABILITY of this system and the resulting iodine removal capacity are consistent with the assumptions of the safety analyses.

Transfer of system operation into the iodine removal mode (exhaest through HEPA filters and charcoal adsorbers) is initiated automatically by either the new fuel storage or spent fuel pool area radiation monitors required by Specifi-cation 3.3.3.

Following installation of the fuel Handling Building Ventilation exhaust radiation monitors, the automatic function of the fuel storage area monitors will be removed.

Transfer of system operation into the iodine removal mode will be by either of the two Fuel Handling Building Ventilation exhaust radiation monitors required by Specification 3.3.3.

ANSI N510-1980 will be used as a procedural guide for surveillance testing.

3/4.9.13 SPENT FUEL SHIPPING CASK MOVEMENT The restriction on spent fuel shipping cask movement ensures that no fuel assemblies will be ruptured in the event of a spent fuel shipping cask accident.

The dose consequences of this accident are within the dose guideline values of 10 CFR Part 100.

3/4.9.14 SPENT iUEL ASSEMBLY STORAGE The restrictions placed on spent fuel assemblies stored in Region 2 of the spent fuel pool and the requirement for 2000 ppm boron concentration ensure that keff will not be greater than 0.95.

The spent fuel storage has been designed and analyzed for a maximum enrichment of 4.5 weight percent U-235.

DIABLO CANYON - UNITS 1 & 2 B 3/4 9-3 AMENDMENT NOS. # AND 6 70 AND 69

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ML20095H091

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