Text

Safety Evaluation Supporting Amend 2 to License DPR-80
	ML20134P623
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Site:	Diablo Canyon
Issue date:	08/26/1985
From:	; Office of Nuclear Reactor Regulation
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SAFETY EVALUATION AMENDMENT NO. 2 TO DPR-80 DIABLO CANYON UNIT 1 DOCKET NO. 50-275 INTRODUCTION The Technical Specifications for Diablo Canyon Unit 1 full-power operation were issued on November 2, 1984, as Appendix A to the Unit 1 Operating License DPR-80 (Ref. 1). The Technical Specifications for Diablo Canyon Unit 2 low-power operation were issued on April 26, 1985, as Appendix A to the Unit 2 Operating License DPR-81 (Ref. 2).

In May 1985 the staff identified a number of Unit 2 Technical Specifications that required further resolution for full-power opera-tion (Ref. 3).

The staff and PG&E have been pursuing the development of common and combined Technical Specifications for Unit 1 and Unit 2 since late 1984. Diablo Canyon Unit I and Unit 2 are nearly identical. There are some differences due to the mirror image, unit-specific layout, core thermal ratings, and equipment ratings.

The controls for both units are located in one common control room. As stated in SSER 31, the operating staff will be licensed on both units, with a desig-nated crew assigned and dedicated to each unit on each shift. PGAE concluded that the Technical Specifications for both units can and should be identical to a great extent. Accordingly, PG&E proposed one set of Technical Specifications, comen to both units, with unit-specific specifications due to differences in l

design and operations appropriately identified.

UNIT 1 TECHNICAL SPECIFICATION CONSIDERATIONS In January 1985 PG&E submitted License Amendment Request LAR 85-01 (Ref. 4) for Unit I to include the Unit 2 specific Technical Specifications information and make changes necessary for a two unit operation (i.e., minimum shift crew composition and diesel generator surveillance requirements).

The changes neces-sary for two unit operation were issued on April 26, 1985 as Amendment I to the Unit I license (Ref. 16). PG&E subsequently subnitted further chances to the Unit 1 Technical Specifications in Revisions 1 through 4 of LAR 85-01 to include Radiological Effluent Technical Specifications (RETS) and Diesel Fuel Oil Speci-fications, make typographical, editorial and similar type changes, improve con-sistency within the specifications, and to further update the specifications with common and unit-specific information (Ref. 5 through Ref. 8). PG&E requested further changes to the Technical Specifications for both units in LAR 85-03 regarding movable control assemblies (Ref. 8), LAR 85-04 regarding reactor trip system instrumentation (Ref. 9) and LAR 85-06 regarding a reor-ganization of the PG&E Nuclear Power Generation organization (Ref.11).

Finally, in LAR 85-05 (Ref.10) and LAR 85-07 (Ref.12 and 13) PG&E requested Unit 1 Technical Specifications changes as a result of certain issues that had been raised by the staff with regard to full-power operation of Unit 2 (Ref. 3).

The above Technical Specifications changes for Unit I have been evaluated by the staff and are being issued as an amendment to the Unit 1 operating license DPR-80. We expect to issue the Unit 2 full-power license shortly with the same Technical Specification provisions. These changes will revise the current Unit 1 Technical Specifications to be identical to those to be issued with the W M PJ$p P

Unit 2 full-power license with unit-specific information clearly identified.

These combined and common Technical Specifications will apply to both units.

i UNIT 2 TECHNICAL SPECIFICATIONS CONSIDERATIONS By letter dated May 14, 1985, PG&E submitted License Anendment Request LAR 85-02 (Ref. 8) which provided a proposed common set of Technical Specifications in the form of a marked up copy of the Technical Specifications for Unit 2 low-power operation. The staff has reviewed these proposed changes and concludes that they will improve the overall effective use of the specifications and will identify appropriate Unit 1 specific information in the common and combined full power Technical Specifications.

The staff addressed the matter of common and combined Technical Specifications in a letter to PGAE, dated May 15,1985 (Ref. 3).

In the letter the staff requested also further consideration of 17 issues that had been identified dur-ing the preparation of the Technical Specifications for the Unit 2 low-power license. The staff met with PG&E on May 15, 1985 to discuss the issues (Ref. 20),

and PG&E provided additional information (Ref.10, and 15). The staff reviewed the infnrmation and discussed further these issues with PG&E. As a result, these issues were resolved, appropriate Technical Specifications were developed, and will be included in the common and combined Technical Specifications for both units except as follows.

The final resolution for five speciiications was achieved after the required Federal Register Notice for the Technical Specifications changes for the Unit 1 i

license amendment had been prepared (Ref. 21) as discussed under Unit 1 Technical Specification considerations above.

The five specifications are:

(1) administrative controls for startup test reports (2) reactor coolant system relief valves (3) loose-part detection system instrumentation (4) bases for electrical power systers (5) reactor coolant system pressure /6emperature limits Rather than issuing the combined Technical Specifications with substantially different requirements for Unit 1 and Unit 2 the staff determined that for an interim period the current Unit 1 Technical Specifications and Unit 2 low-power Technical Specifications, as reflected in the combined Technical Specifications, provide sufficiently conservative requirements.

PG&E, by letter dated June 20, 1985 (Ref.14) identified the resolution to the five specifications and committed 4

to submit an appropriate license amendment request for both units within 90 days after issuance of the Unit 2 full-power license. The staff finds this acceptable, As a result of its review of the Diablo Canyon Unit 1 and Unit 2 Technical Specifications, the staff determined that specifications should be included for the polar crane in the containment buildings and for the two bridge crancs in the turbine building or appropriate justification should be provided for not including them.

3 At this time, the parking location for the polar crane is restricted to pre-clude jet impingement from a postulated pipe rupture. This restriction should i

i

be included in the specifications or PGAE shall demonstrate that the crane can withstand the jet impingement forces from a postulated pipe rupture.

l The two bridge cranes in the turbine building are currently restricted by rail clamps from entering the end bays in the turbine building. The restriction for placing the two cranes relative to each other relies on visual observations imposed by administrative procedure. These restrictions should be included in the specificiations or PG8E shall demonstrate that no restriction is required.

The staff has requested PG&E to submit, in addition to the above five items, a i

license amendment request within 90 days of issuance of the Unit 2 full-power license regarding these restrictions or provide a justification why they are not required (Ref. 23). The staff finds acceptable, for the interim, the use of the current administrative procedures.

PGAE requested further changes to the current Unit I and Unit 2 Technical Specifications regarding movable control assemblies (Ref. 8), reactor trip system instrumentation (Ref. 9) and reorganization of the PG&E Nuclear Power Generation organization (Ref. 10). The staff evaluation of these proposed changes, which are applicable to both units, is presented in separate sections 1

below.

I.

MOVABLE CONTROL ASSEMBLIES 4

Introduction By letter dated May 14, 1985 PG&E requested in LAR 85-03 changes to the Techni-j cal Specification 3.1.3.1, " Movable Control Assemblies" for Diablo Canyon Units 1 and 2 (Ref. 8). The changes will allow continued operation for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months (to perform repairs) for a condition in which there are multiple control rods which cannot be moved by their drives because of electrical power failure but which are still within alignment and trippable for insertion under gravity.

Evaluation The proposed Technical Specification changes contain Action Statements which address inoperable and misaligned movable control assemblies as follows:

Action a. is unchanged and addresses a mechanical failure of one or more full length rods that leaves the rod (s) imovable and/or untrippable.

This is the classic stuck rod situation, requiring application of the stringent fix-or-shutdown requirement.

Action b. has been deleted and replaced with the new Action c. and d.

Action c. has been relabeled to Action b. with the content unchanged and addresses one full-length rod trippable but inoperable due to causes other l

than addressed by Action a.

Action c.

A new Action c. has been added to address multiple inoperable rods and replaces the old Action b.

The change now allows operation to continue for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months (to perform repairs) for a condition in which there are multiple imrovable rods that are still trippable and within alignment.

Action d. has been added to replace that nart of the old Action b. that j

addressed multiple misaligned rods.

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Although these Action statements have been realigned and relabeled, there is no i~

alteration of the content and requirements of the present Technical Specifications except for the new Action c.

The present Technical Specifications require the reactor to be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for the condition of new Action c.,

J which is when there are multiple inoperable rods which are not stuck or j

untrippable, j

l The intent and basis for Technical Specification 3.1.3.1 is to provide require-i ments which either directly in the Specification or in support of other Specifi-

]

cations ensure that:

1) acceptable power distribution limits are maintained, 1

2) the minimum shutdown margin is maintained with allowance for a stuck rod, j

3) the potential effects of rod misalignment on associated accident analyses i

i are limited, and 4) the trip reactivity assumed in the accident analysis will i

be available. Appropriate provisions have been made in the proposed change to j

ensure that these requirements continue to be maintained. Specifically, the

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proposed change requires that within one hour after entering the Action state-j ment the remainder of the rods in the bank (s) with the inoperable rods are aligned to within i 12 steps of the inoperable rods. This relates to ensuring l

conformance to items 1 and 3 above.

Further, it is explicitly specified that the rod sequence and rod insertion limit Technical Specifications must be l

i l

maintained as applicable. This ensures conformance with items 2 and 4 above, t

The intent of the proposed Technical Specification change is to allow continued t

operation of the unit in the case of electrical failures which prevent moving of more than one control rod. The rods remain trippable, so they can provide their safety function in shutting down the reactor, if needed.

The staff finds the proposed change acceptable because, as indicated above, the f

safety requirements for the control rods remain intact.

Control of the power f

j plant can still be effected with boration systems, if the inoperable rods are in the control bank or the rod sequence requirements potentially could not be i

maintained.

If the first regulating banks or the shutdown banks become inoper-able while inserted in the core, the insertion limit Specification requires the power level to be 0 If the last two banks become inoperable while at part L

power the insertion limits allow operation only at 1 art power. Operation at l

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part power or at full power with the regulating banc at its insertion limit for extensive periods of time is not accounted for in the power distribution analy-sis, which is assumed as an input condition for the LOCA analysis. Therefore the time span of the proposed change is limited to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . Nonnal operation i

of the power plant assumes all of the control rods are out of the core, except the regulating bank may be inserted a small amount to allow for control of the axial flux difference and day to day reactivity burnout (which may also be controlled by boron concentration).

a In Attachment C to a letter dated December 21, 1984 from E. P. Rahe, Jr.

(Westinghouse) to C. O. Thomas (NRC), a description of the movable control I

i assembly mechanisms is provided, along with a discussion of failures which j

might occur (Ref.19). The attachment provides guidance on the type of elec-j trical failures which may render contrcl assemblies inoperable but still i

trippable.

It will be used in the preparation of the bases of the Technical

}

Specifications.

i In preparing the proposed change, PGAE also deleted the words " step counter" from the Unit 1 Technical Specifications to be consistent with Unit 2 and to i

1

allow othtr means for detemining group demand position to satisfy this Specifi-cation. Tais change is also acceptable because it does not alter the require-ment for the Specification, but only allows the possibility of other means for determining the group demand position.

II. REACTOR TRIP SYSTEM INSTRUMENTATION

Background

On February 21, 1985, the NRC issued its Safety Evaluation Report on the Westinghouse Technical Specification Optimization Program for increased sur-veillance intervals and out-of-service times for testing and maintenance of the reactor trip system (RTS) (Ref. 17). The Optimization Program proposal was set forth in WCAP-10271, " Evaluation of Surveillance Frequencies and Out-0f-Service Times for the Reactor Protection Instrumentation System" Supplement 1 (Ref.18).

By letter dated May 20, 1985, PG&E Company submitted prnposed Technical Specifi-cations as LAR 85-04 for Diablo Canyon Units 1 and Unit 2 based on the Optimiza-tion Program (Ref. 9).

Evaluation The proposed Technical Specification changes include an increase in the sur-veillance interval of reactor trip system channels from monthly to quarterly.

The proposed quarterly surveillance intervals are annotated to indicate that the quarterly tests are to be performed on a staggered test basis.

In addition the surveillance intervals for those reactor trip sytem channels which also provide signals to the Engineered Safeguards Actuation System (ESFAS) have been annctated to indicate that the more restrictive ESFAS surveillance requirements apply to these channels. These changes are consistent with the staff's generic evaluation of the Optimization Program and are, therefore, acceptable.

The surveillance frequency for channels which would be tested prior to plant startup was noted as prior to startup in lieu of quarterly as recommended under the Optimization Program. Also, the startup surveillance was annotated to indicate that it is to be perfomed prior to startup if not perfomed in the previous 31 days.

The channels for which testing prior to startup applies are:

A.

Power Range, Neutron Flux, Low Setpoint The low setpoint channels are required to be tested prior to entry into Mode 2 and for operation in Mode 1 below the P-10 interlock setpoint.

For these tran-sitional operating modes, the surveillance test would be conducted prior to startup and not on a routine basis as would be implied by specifying a quarterly test frequency. Thus, the surveillance frequency is noted as prior to startup, S/V(1).

In the Table 4.3-1 Notation, note (1) is revised to indicate if not perfomed in the previous 31 days, instead of 7 days.

B.

Intermediate Range, Neutron Flux The intemediate range channels are required to be tested for the same transi-tional modes as the power range low setpoint channels. On the same basis the surveillance frequency is noted as prior to startup, rather than quarterly, t

C.

Source Range, Neutron Flux The source range channels are required to be tested in Mode 2 below the P-6 interlock setpoint and in Modes 3, 4, and 5.

The source range channels, in addition to initiating reactor trip, provide the high flux alarms at shutdown which alert the operator of reactivity changes caused by a boron dilution i

event. Therefore, the surveillance requirements for the source range channels apply to Modes 3, 4, and 5.

The surveillance frequency is noted as prior to i

startup and as quarterly.

D.

Reactor Trip System Interlock (RTSI), Intermediate Range Neutron Flux, P-6 The P-6 interlock channels are required to be tested prior to entry into Mode 2 and for operation in this mode below the P-6 interlock setpoint.

For this transitional operating mode, the surveillance test wculd be conducted prior to startup and not on a routine basis as would be implied by specifying a quarterly test frequency. Thus, the surveillance frequency is noted as prior to startup.

E.

RTSI, Low Power Reactor Trips, Pleck, P-7 The surveillance frequency is noted as prior to startup, consistent with that specified for the P-10 and P-13 channels as noted below, a

F.

RTSI, Power Range Neutron Flux, P-8 The currently specified monthly test interval is annotated to indicate that l

when the plant is at a power level greater than the channel trip setpoint, the surveillance requirement is satisfied by verifying that the interlock permissive logic is in its required state. This provision was included in the Technical Specifications such that power reductions below the channel setpoint would not be required for the sole purpose of meeting the surveillance requirement.

For i

example, the previous monthly surveillance requirement for the P-8 channels would require a power reduction below its setpoint had this annotation to the surveillance requirement not been included. Likewise, this same annotation is used for the P-10 channels and precludes the necessity of power reductions for testing.

l This annotation to the surveillance requirement only verifies the status of the permissive logic and does not address verification of channel setting or oper-ability. Those aspects would be verified following a refueling shutdown and prior to startup. With a monthly surveillance interval the annotated surveil-lance at power would in all likelihood have expired during a refueling shutdown (i.e., exceeded 31 days) and testing would be required prior to entry into Modes 2 or 1, as applicable to the permissive channels. However, this situation may not be true if a quarterly surveillance interval is specified, i.e., the annotated surveillance perfomed at power may not have exceeded a 92 day sur-veillance interval during a refueling shutdown. Therefore, since the only comprehensive tests are actually performed prior to startup, the surveillance interval is stated as prior to startup.

The status of interlock permissives at the logic and channel level are individ-ually indicated on status monitoring displays in the control room. As such they are routinely checked and particular attention is given to this infomation i

during operational mode changes. The fact that the permissive status indication is readily available and can be routinely verified constitutes a different con-sideration with respect to the availability of trip channels which must change state on the occurrence of an event and for which the function unavailability is dependent on the surveillance interval.

It is concluded that maintaining the requirement for the verification of the permissive logic status is not safety significant. Therefore, since the surveillance frequency has been noted prior to startup, note (8) is no longer applicable to these channels.

G.

RTSI, Low Setpoint Power Range Neutron Flux, P-10 The surveillance frequency is noted as prior to startup, on the same basis as for the power range P-8 channels noted above.

H.

RTSI, Turbine Impulse Chamber Pressure, P-13 The surveillance frequency is noted as prior to startup. The bases with regard to note (8) is the same as for the power range interlock channels noted above.

However, unlike the power range channels which can only be tested when the measurement signal is below the setpoint, the turbine impulse chamber pressure channels include features which would permit them to be tested without reducing reactor power. However, since the P-13 channels and hence P-13 logic only pro-vide inputs to the P-7 logic which are diverse to the P-10 inputs to the P-7 logic, it is concluded that there is sufficient justification for excluding testing as currently noted by the provision on note (8) when operating above the P-13 setpoint. Further, since the channel state is indicated, its state can be readily determined. Thus, it is concluded that the additional surveil-lance as would occur by noting a quarterly surveillance frequency annotated with note (8) is not warranted.

By letter dated July 18, 1985, PG&E requested a change to exclude the require-ment to test the P-13 channels prior to startup from Modes 2 or 3 (Ref. 22).

A startup from Mode 2 (power less than 5 percent) or from Mode 3 (not standby) would occur only following plant operation at power and subsequent reduction in power to these modes. Such plant startup would be due to unusual circum-stances and not on a routine basis.

It is concluded that this change is accept-able since the additional testing which may result under these conditions would not have an impact on plant safety due to considerations related to the operability of interlocks noted above. This exception to the startup tests is included as " Note 8" to the surveillance requirements.

With regard to the inclusion of the startup test requirement, wherein channels would be tested prior to plant startup if not tested in the previous 31 days, the staff concludes that this is appropriate in view of the increase in the surveillance frequency from that currently required for startup tests if not performed in the previous 7 days.

I.

Out of Service Times for Testing and Maintenance For those RTS channels which provide input signals to ESFAS, the ACTION column of Table 3.3-1 is annotated, as applicable, with note (1) and the Table 3.3-1 notation includes note (1) to indicate that the applicable Modes and ACTION statement for these channels under the ESFAS specification requirements are applicable. This is consistent with the staff's evaluation for the Optimization Program and is, therefore, acceptable.

Conclusion The staff concludes that the proposed Technical Specifications are acceptable for Diablo Canyon Units 1 and 2.

III. GONPRAC COMPOSITION PG&E requested in License Amendment Request LAR 85-06 of May 30, 1985 (Ref. 61), changes to the Technical Specifications for Diablo Canyon Units 1 and 2.

The LAR proposes to change the composition of the PG&E General Office Nuclear Plant Review and Audit Committee (GONPRAC) to reflect changes to the PG&E Nuclear Power Generation Department. These organizational changes have been reviewed by the staff during the full-power license review for Diablo Canyon Unit 2 and were found acceptable.

The proposed changes to the composition of the GONPRAC delete two members (i.e., Project Manager - Diablo Canyon, and Technical Assistant to the Vice President - Nuclear Power Generation) and add three new members (i.e., Manager

- Nuclear Engineering and Construction Services, Director - Nuclear Administra-tion and Support Services, and Director - Nuclear Regulatory Affairs).

The staff finds that the proposed changes meet the acceptance criteria of Section 13.4 of NUREG-0800, the Standard Review Plan, and concludes that the changes to the composition of the GONPRAC are acceptable.

IV. RADIOLOGICAL EFFLUENT TECHNICAL SPECIFICATION Introduction PG5E has filed with the Comission plans and proposed Radiological Effluent Technical Specifications (RETS) developed for the purpose of keeping releases of radioactive materials to unrestricted areas during normal operations, specifications would reasonably assure compliance, in radioactive waste management, with the provisions of 10 CFR Part 50.36a, as supplemented by Appendix I to 10 CFR Part 50, with 10 CFR Parts 20.105(c), 106(g), and 405(c);

with 10 CFR Part 50, Appendix A, General Design Criteria 60, 63, and 64; and with 10 CFR Part 50, Appendix B.

Background and Discussion NUREG-0472 provides radiological effluent technical specifications for pressurized water reactors which the staff finds to be an acceptable standard for licensing actions.

Further clarification of these acceptable methods is provided in NUREG-0133, " Preparation of Radiological Effluent Technical Specifications for Nuclear Power Plants." NUREG-0133 describes methods found acceptable to the staff of the NRC for the calculation of certain key values required in the preparation of proposed radiological effluent technical speci-fications for light-water-cooled nuclear power plants. NUREG-0133 also provides guidance to licensees in preparing requests for changes to existing radiological effluent technical specifications for operating reactors.

It also describes current staff positions on the methodology for estimating radiation exposure due to the release of radioactive materials in effluents and on the administrative control of radioactive waste treatment systems.

The above NUREG documents address all of the radiological effluent technical specifications needed to assure compliance with the guidance requirements provided by the regulations previously cited. However, alternative approaches

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to the preparation of radiological effluent technical specifications and alternative radiological effluent technical specifications may be acceptable if the staff determines that the alternatives are in compliance with the regulations and with the intent of the regulatory guidance.

J Plant Specific Background i

In three different submissions (Ref. 4, 5, 6) PG&E has formally requested i

Technical Specification changes that would upgrade the Diablo Canyon-1 RETS i

to be in conformance with the latest NRC staff positions. The first submission j

was used as the starting point for the development of acceptable RETS for j

Diablo Canyon-2. After a number of discussions between PG&E staff and NRC staff a final set of RETS was developed for Diable Canyon-2 which met the 2

intent of current NRC guidance. These RETS were included as part of the published Diablo Canyon-2 Technical Specifications (Ref 2).

Partway through this development for Diablo Canyon-2, PG&E made a second RETS i

submission for Diablo Canyon-1 that contained some of the original markup i

pages from the earlier submission, and some of the retyped pages from the J

Diablo Canyon-2 RETS that were in process.

Further discussion with PG&E resulted in an agreement to make a final i

submission for Diablo Canyon-1 once the Diablo Canyon-2 RETS had been j

accepted technically.

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PG&E then formally submitted on April 24, 1985 (Ref. 6) proposed changes to

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their RETS for Diablo Canyon Unit I to make them consistent with the RETS for l

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Diablo Canyon Unit 2, which in turn have been developed to be consistent with current regulatory guidance.

1i Evaluation The proposed changes to the Diablo Canyon Unit 1 RETS as presented in the April 24, 1985 submission were found to meet the intent of the NRC staff's model RETS for PWRs, NUREG-0472, Revision 2, February 1,1980, and therefore e

i are acceptable.

t Sumary I

The proposed changes to the radiological effluent technical specifications for a

Diablo Canyon Unit I h u e been reviewed, evaluated, and found to be in compliance l

with the requirements of the NRC regulations and with the intent of NUREG-0133 and NUREG-0472 (Diablo Canyon Unit 1 is a pressurized water reactor) and thereby fulfill all the requirements of the regulations related to radiological effluent technical specifications.

The' proposed changes will not remove or relax any existing requirement needed i

to provide reasonable assurance that the health and safety of the public will i

not be endangered by operation in the proposed manner. The staff, therefore, finds the proposed changes acceptable.

V.

AUXILIARY FEEDWATER SYSTEM In SSER No. 4, the staff identified the need for technical specifications to s

j limit the steam generator refill rate via the auxiliary feedwater system in i

l

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_ to _

order to further limit the potential for adverse (damaging) feedwater hamer.

Since issuance of SSER No. 4 (May 1976), the staff has completed generic i

studies of waterhammer concerns including resolution of USI A-1 and has com-pleted study of auxiliary feedwater system reliability in connection with Item II.E.1.1 of the TMI-2 Task Action Plan, NUREG-0737. Based on the results of this work, we now conclude that throttling auxiliary feedwater flowrate to limit steam generator level recovery as a means of reducing waterhamer poten-tial is no longer necessary or desirable. Preoperational testing at Diablo i

Canyon Unit 1 as well as operating experience at other Westinghouse reactors has shown that modifications to the feedring and revised plant operating pro-cedures have effectively limited waterhamer was rereviewed with an intent i

to eliminating it in order not to adversely affect auxiliary feedwater reliability as indicated in the implementation of Item II.E.1.1 of NUREG-j 0737 (Recommendation GS-3). Diablo Canyon was previously reviewed against this recommendation and found acceptable.

l Based on the above, we conclude that technical specification to limit steam generator refill rate are not necessary for Diablo Canyon Units 1 and 2, as i

suitable measures for limiting feedwater hamer via the auxiliary feedwater system are provided.

VI. DIESEL FUEL OIL SURVEILLANCE RE0VIREMENTS In letters dated January 30, March 13, and May 14, 1985, (Ref 4, 5, 6) Pacific Gas and Electric Company proposed changes in the diesel fuel oil surveillance i

requirements (Technical Specification 4.8.1.1.2.d and 4.8.1.1.3) for Diablo Canyon Unit 1.

The proposed changes will result in the Diablo Canyon Unit 1 1

Technical Specifications being consistent with the Diablo Canyon Unit 2 j

Technical Specifications.

1 In general, the proposed changes involve replacing the diesel fuel oil tests presently required by the Technical Specifications with the McGuire Tests that (1) are more effective in detecting unsatisfactory fuel oil, (2) can be performed onsite, and (3) are simpler to perform. The McGuire Units 1 and 2 Technical Specifications surveillance requirements were approved on June 5, i,

1984. The effectiveness of the McGuire Technical Specifications is based upon j

the ability to remove all the water that may accumulate in fuel oil storage on a 31 day frequency. The Diablo Canyon fuel oil storage tanks were evaluated on this basis.

In a letter dated March 13, and May 14, 1985 (Ref. 5,6), PG&E submitted information on the fuel oil storage tank design with regards to water removal capability and the commitment to remove the water on a 31 day frequency. The 4

staff has evaluated the information and finds that the design of the fuel oil i

storage tanks allows for the removal of essentially all the water that may accumulate in the tank. The removal of accumulated water from the storage tanks every 31 days will reduce the possibility of bacteria contamination of the stored fuel, minimize the formation of corrosion products, and prevent water from contaminating the diesel generator fuel oil system.

The proposed changes to the present surveillance requirements for diesel fuel 4

oil in Technical Specification 4.8.1.1.2.d and 4.8.1.1.3, as proposed by PG8E for Diablo Canyon Unit 1, will result in a more conservative approach

.I

to fuel oil surveillance. The added conservatism coupled with the simplified testing of fuel oil will provide immediate assurance in acceptance of quality fuel oil on delivery and maintenance of high quality stored fuel; this should increase diesel generator availability. Therefore, the proposed Technical Specifications changes to the diesel fuel oil surveillance are acceptable.

VII. THERMAL HYDRAULIC DESIGN EFFECT OF FUEL ROD B0 WING ON DNB In Diablo Canycn SER Supplement #6, the staff stated that as an interim measure, we will include a burnup dependent penalty factor to be applied i

to the reactor operating limits in Section 3.2.3 of the Technical Specifi-cations to reflect the reduced DNB conditions caused by increasing fuel rod bowing.

Subsequent to issuance of the SER Supplement, the Westinghouse tropical report WCAP-8691, Revision 1, " Fuel Rod Bow Evaluation", has been approved by the staff. This rod bow penalty evaluation method applies statistical convolution of the critical heat flux test data and interfuel rod gap closure data to derive the rod bow penalty on DNBR. The use of this method results in a significantly lower rod bow penalty compared to the interim method previously used.

PG&E has submitted a table of rod bow penalty as a function of fuel burnup calculated with the approved Westinghouse method.

Since rod bow and gap closure increase with fuel burnup, the rod bow penalty on DNBR increases with burnup. However, even though the plant may be operated at higher burnup, the maximum fuel burnup used for the rod bow penalty calculation is 33, 000 MWD /MTU. The reason for using 33,000 MWD /MTU as a cutoff paint is because the physical burndown effect of the high peaking fuel rod will exceed a burnup of 33,000 MWD /MTU, it is not capable of achieving limiting peaking factors due to the decrease in fissionable isotoper and the buildup of fission product inventory. Therefore, the rod bow penalty value of less than 3%

DNBR at 33,000 MWD /NTU represents the maximum rod bow penalty for Diablo Canyon plants having 17x17 R-Grid fuel assemblies.

However, credit is available to offset this rod bow DNBR penalty. The generic margin (credit) totaling 9.1 percent DNBR is derived from the difference between the design and required values on the following items:

(a) design DNBR limit, (b) grid spacing multipliier coefficient, (d) DNBR spacer factor multiplier and fe()c) thermal diffusion pitch reduction.

Since the rod bow penalty is completely compensated by the available margin of 9.1 percent, no penalty is required in the reactor operating limit. We have incorporated this rod bow penalty compensation in the bases of the Diablo Canyon Technical Specifications.

VIII. ADDITIONAL UNIT I TECH SPEC CHANGES NECESSARY FOR COMBINED UNIT 1 & 2 TECHNICAL SPECIFICATIONS Introduction In accordance with the licensee's applications for amendment dated May 14 (LAR 85-01, Rev. 4 and LAR 85-03), May 20 (LAR 85-04), and May 30, 1985 (LAR 85-05 and LAR 85-06), Ref. (8, 9, 10) the proposed changes would revise the Diablo Canyon Unit 1 Technical Specifications to eliminate typographict..

errors, provide additional clarifications, improve consistency, adjust nomen-clature, modify the reporting requirements to bring then into confonnance with the guidance in Generic Letter 83-43 for the Licensee Event Report System, bring portions of the Specifications into conformance with current NRC staff positions, incorporate Unit 2 information where appropriate, and make other minor changes. A primary objective of the amendment request is to achieve a single Technical Specifications document that is connon for the nearly identical Units 1 and 2 with individual specifications for each unit clearly identified as appropriate.

In this regard, the proposed changes to the Unit 1 Technical Specifications are consistent with the Unit 2 Technical Specifications recently issued as Appendix A to Facility Operating License DPR-81 and proposed changes to the Unit 2 Technical Specifications included in the licensee's application for amendment to DPR-81 dated May 14, 1985. Furthermore, most of the Unit 1 proposed changes have already been incorporated by the Connission into the current Unit 2 Technical Specifications as a part of DPR-81. This becomes a pertinent consideration in evaluating whether or not the Unit 1 proposed changes involve a significant hazards consideration. Discussed below are the various types of changes to the Technical Specifications.

A.

Administrative Changes to Technical Specifications The proposed changes in the Unit 1 Technical Specifications for the items listed below are provided to eliminate typographical errors, correct punctuation, adjust nomenclature, add clarification, improve consistency, make minor changes, and include Unit 2 Speci-fications as necessary to reflect operation of Units 1 and 2.

The following is a description of the proposed changes to the current Unit 1 Technical Specifications:

1.

Page 1-2, item c of Definition 1.8, CONTAINMENT INTEGRITY, is changed from "c. Each air lock is OPERABLE pursuant to Specification 3.6.1.3" to "c.

Each air lock is in compliance with the requirements of Specification 3.6.1.3."

2.

Page 1-2, Definition 1.10, an "s" is added to the title of

" CORE ALTERATION" and " reactor pressure vessel" is changed to

" reactor vessel."

3.

Page 1-3, in Definition 1.11, Table E-7 of HRC Regulatory Guide 1.109, Revision 1, October 1977 is added as an additional source of thyroid dose conversion factors, and the definition of the AVERAGE DISINTEGRATION ENERGY, 1.12, is changed to "E shall be the average (weighted in proportion to the concen-tration of each radionuclide in the sample) of the sum of the average beta and gamma energies per disintegration (MeV/d) for the radionuclides in the sample."

4 Pages 1-3 and 1-5, the tables referenced in Definition 1.15, FREQUENCY NOTATION, and Definition 1.21, OPERATIONAL MODE, are reversed to Table 1.1 and Table 1.2, respectively; Table 1.1 on page 1-8 becomes Table 1.2, and Table 1.2 on page 1-9 becomes Table 1.1.

These changes are made for consistency with the order in which the tables are cited in the text.

5.

Page 1-4, in Definitions 1.16 and 1.17, change " primary coolant system" and " primary system" to " Reactor Coolant System," and change " secondary system" to " Secondary Coolant System." Made these changes, where applicable, throughout the document.

6.

Page 1-5, the last line of Definition 1.27 RATED THERMAL POWER, which only includes Unit I data, is changed to "..

3338 MW t

for Unit I and 3411 MW for Unit 2."

t 7.

Page 1-6, Definition 1.29 of REPORTABLE OCCURRENCE, is replaced with " REPORTABLE EVENT shall be any of the conditions specified in,Section 50.73 of 10 CFR Part 50. " and the term " REPORTABLE OCuaRENCE" is chaaged throughout the Technical Specifications to " REPORTABLE EVENT."

8.

Page 1-4, 1-5, 1-6 and 1-7, Definitions 1.19, 1.20, 1.21, 1.22, 1.23, 1.24, 1.25, 1.26, 1.27, 1.28, 1.29, 1.30, 1.31, 1.32, 1.33, 1.34,1.35,1.36,1.37, and 1.38 are renumbered to 1.20, 1.21, 1.22, 1.23, 1.24, 12.5, 1.26, 1.27, 1.28, 1.29, 1.30, 1.31, 1.33, 1.34, 1.35, 1.36, 1.37, 1.38, 1.39, 1.41 and 1.42, respectively, to account for the three new definitions 1.19, 1.32 and 1.40.

9.

Page 2-1, in Specifications 2.1.1 and 2.1.2, Reactor Core and Reactor Coolant System Pressure, add "... and comply with the requirements of Specification 6.7" at the end of the ACTION stitement.

10. Page 2-4, Specification 2.2.1. Table 2.2-1, the footnote is changed to " Design flow is 87,700 qpm per loop for Unit 1 and 88,500 gpm per loop for Unit 2" (Unit 2 design flow added).

11. Page 2-6 Specification 2.2.1 Table 2.2-1, item 22.d is changed from " Low Setpoint Power Range Neutron Flux, P-10" to

" Power Range Neutron Flux, P-10".

12. Page 2-7, Specification 2.2.1, Table 2.2-1, the table subtitle is modified to " TABLE NOTATIONS", and in note 1 the defigtion of T' is replaced with "T' = less than or equal to 576.6 F for 0

Unit I and less than or equal to 577.6 F for Unit 2 Reference T,yg atRATEDTHERMALPOWER"(Unit 2T,yg added).

13. Pages 2-8 and 2-9, Specification 2.2.1. Table 2.2-1, the table substitute is changed to " TABLE NOTATIONS" and in Notes 2 and 4

" trip point" is changed to " Trip Setpoint" (twice each).

14. Page 2-9, Specification 2.2.1 Table 2.2-1, in Note 3, the defin{ tion of T" is changed to "T" = less than or equal to g

576.6 for Unit 1 and less than or equal to 577.6 F for Unit 2 Reference T,yg at RATED THERMAL POWER" (Unit 2 T,yg added).

_ 15.

Page 2-9, Specification 2.2.1, Table 2.2-1, in Note 3, the definition of K is changed to "... average temperature, g

and 0 for decreasing..."

16. Page 3/4 0-1, Specification 3/4.0, Applicabiltty of Limiting Conditions for Operation, add a new specification:

"3.0.5 Limiting Conditions for Operation including the associated ACTION requirements as follows:

2 a.

Whenever the Limiting Conditions for Operation refers to systems or components which are shared by both units, the ACTION requirements will apply to both units simultaneously. This will be indicated in the ACTION section; b.

Whenever the Limiting Conditions for Operation applies to only one unit, this will be identified in the APPLICABILITY section of the specification; and c.

Whenever certain portions of a specification contain operating parameters, Setpoints, etc., which are different for each unit, this will be identified in parentheses, footnotes, or body of the requirement."

17. Page 3/4 0-3, Specification 3/4.0, Applicability of Surveillance i

Requirements, add a new surveillance requirement:

"4.0.6 Surveillance Requirements shall apply to each unit individually unless otherwise indicated as stated in Specification 3.0.5 for individual specifications or whenever certain portions of a specification contain surveillance parameters different for each unit, which will be identified in parentheses, footnotes, or body of the requirement."

j

18. Page 3/4 1-1, in Surveillance Requirement 4.1.1.1.1d, "Speci-fication of the factors of e, below", is replaced with "of the factors of Specification 4.1.1.le, below".

19. Page 3/4 1-1, the footnote to specification 3.1.1.1 is changed from *See Special Test Execption 3.10.6." to "*See Special Test Execption Specification 3.10.1."

20. Page 3/4 1-4, Specification 3.1.1.3, delete "in lieu of any other report required by specification 6.9.1" from ACTION a.3.

21. Page 3/4 1-6, the footnote to Specification 3.1.1.4 is changed from "*See Special Test Exception 3.10.3." to "*See Special Test Execption Specification 3.10.3."

22. Page 3/41-12, Specification 3.1.2.5a.2 is changed from "Between 20,000 and 22,500 ppm of boron, and" to "A boron concentration between 20,000 and 22,500 ppm, and".

23. Page 3/4 1-13, Specification 3.1.2.6a.2 is changed from "Between 20,000 and 22,500 ppm of boron, and" to "A boron concentration between 20,000 and 22,500 ppm, and".

t

T 24. Page 3/41-13, Specification 3.1.2.6b.2 is changed from "Between 2000 and 2200 ppm boron, and" to "A Boron concentration between 2000 and 2200 ppm, and".

25. Page 3/4 1-16, the first line in Surveillance Requirement 4.1.3.1.2 is changed from "Each full-length rod not fully inserted shall be determined to be..." to "Each full-length rod not fully inserted in the core shall be determined to be..."

26. Page 3/4 1-18, Specification 3.1.3.2, "The shutdown and control rod position indication system..." is replaced with "The Digital F.od Position Indication System..."

27. Page 3/4 1-18, Specification 3.1.3.2, in ACTION items a and b.1, " rod position indicator" is replaced with " digital rod position indicator".

28. Page 3/4 1-18, in Surveillance Requirement 4.1.3.2, replace

" rod position indicator" with " digital rod position indicator",

and " Rod Position Indication System" with " Digital Rod Position IndicationSystem"(twice).

29. Page 3/4 1-19. Specification 3.1.3.3, " rod position indicator" is replaced with " digital rod position indicator".

30. Page 3/4 1-19, the "#" footnote is changed to "# See Special Test Exceptions Specification 3.10.4".

31. Page 3/41-22, Specification 3.1.3.6 is chanced to "The control banks shall be limited in physical insertion as shown in Figure 3.1-la for Unit I and Figure 3.1-lb for Unit 2" (Unit 2 figure added).

32. Page3/41-22,delet["efther"inthesecondlineofthe.

introductory sentence under the Specification 3.1.3.6 ACTION statement.

33. Page 3/4 1-23, the designation of Figure 3.1-1 is changed to "3.1.la" and an indication that it applies to Unit 1 is added. <

A new Figure 3.1-lb, which applies to Unit 2 is added in page 3/4 1-24.

34 Page 3/4 2-3, in the third line of Surveillance Requirement 4.2.1.4, insert " Specification" before "4.2.1.3".

35.

Page 3/4 2-5, Specification 3.2.2, in the definition of K(Z)

"is" is replaced by "=".

36. Page 3/4 2-7, in Surveillance Requirement 4.2.2e, the last line is changed from " Specification 6.9.1.14" to " Specification 6.9.1.8".

37 Page 3/4 2-9, in the third lina ~of Specification 3.2.3, " Figure 3.2-3" is replaced with " Figure 3.2-3a for Unit 1 and Figure 3.2-3b for Unit 2".

4

r.

38. Page 3/4 2-9, in Specification 3.2.3 thg third line of new item c is changed to "... values of F H shall be used to calculate R since Figure 3.2-3a for Unit 1 and Figure 3.2-3b for Unit 2... "

39. Page 3/4 2-9, in Specification 3.2.3, the second line of the introductory ACTION paragraph is changed to "... acceptable operation shown on Figure 3.2-3a for Unit 1 and Figure 3.2-3b for Unit 2".

40. Page 3/4 2-10, in Specification 3.2.3, in the seventh line of ACTION item c " Figure 3.2-3" is replaced with " Figure 3.2-3a for Unit 1 and Figure 3.2-3b for Unit 2".

41. Page 3/4 2-10, in Surveillance Requirement 4.2.3.2 "R, R " is replacedwith"R"and" Figure 3.2-3"with" Figure 3.213afor Unit I and Figure 3.2-3b for Unit 2".

42. Page 3/4 2-11, in Surveillance Requirement 4.2.3.3, " Figure 3.2-3" is replaced with " Figure 3.2-3a for Unit 1 and Figure 3.2-3b for Unit 2" and "R, R " is replaced with "R".

1 2

43. Page 3/4 2-12, Specification 3.2.3, Figure 3.2-3 is redesignated Figure 3.2-3a, the DNB limit curve and the designation that the other curve represents the "LOCA LIMIT" are deleted, and an indication that it applies to Unit 1 is added.

44. Page 3/4 2-13 Specification 3.2.3, Figure 3.2-4 is replaced with a new Figure 3.2-3b, which provides RCS Total Flowrate versus R for Unit 2.

45. Page 3/4 2-18, Specification 3.2.5, Table 3.2-1 is revised to incorporate the limits for the Unit 2 DNB parameters, and thus provide the applicable DNB parameters for each unit.

46.

Pages 3/4 3-2 to 3/4 3-7 Specification 3.3.1, Table 3.3-1 Table Notations, is replaced with revised Table Notations to incorporate administrative changes.

In addition, the second and third lin?s in ACTION 11 are changed to "... channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however one channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance..."

47 Page 3/4 3-8, Specification 3.3.1. Table 3.2-2, a response time "less than or equal to 0.5 second*", with a reference to the table footnote, is assigned to Functional Unit 6.

Source Range, Neutron Flux.

48. Page 3/4 3-13. Surveillance Requirements 4.3.1.1. Table 4.3-1, the Table Notation "*" is revised to incorporate administrative changes, and Table Notation (6) is changed to "Incore-Excore Calibration, above 75% of RATED THERMAL POWER" and "The provisions of Specificatior 4.0.4 are not app (licable for entry into MODE 2 or 1" is added at the end of (2), 3)and(6).

i

t.

49. Pages 3/4 3-15 to 3/4 3-2?. Specification 3.3.2, Table 3.3-3 is revised to incorporate administrative changes.

50. Pages 3/4 3-16 and 3/4 3-17 Table 3.3-3, the information in the CHANNELS TO TRIP COLUMN for the 2.a and 3.b.I entries is changed to "2 with 2 coincident switches", and a reference to Table Notation "*" in the ACTION column for the 3.c.1) and 3.c.2) entries added.

51. Page 3/4 3-18, Specification 3.3.2 Table 3.3-3, a reference to Table Notation "##" in the Applicable Modes column is added to the 4.d entry, the ACTION for the 5.a entry is changed to "25",

and "': TION 25 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, be in at least H0T STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however, one channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 'for surveillance testing per Specification 4.3.2.1, provided the other channel-is OPERABLE" is added in page 3/4 3-22.

52. Pages 3/4 3-23 to 3/4 3-27, Specification 3.3.2, Table 3.3-4 is revised to incorporate administrative changes, and in pages 3/4 3-23 and 3/4 3-26 the Allowable Values for entries 1.f.2) and 4.d.2) are changed from "less than or equal to 585 psig" to less than or equal to 580 psig".

53. Pages 3/4 3-28 and 3/4 3-31, Specification 3.3.2. Table 3.3-5 is revised to incorporate administrative changes, replace the entry under Initiating Signal 8 with " Turbine Trip", and add main feedwater bypass 3alves information to Table Notation (2) arid, consecuently, the' requirement to verify response time on these valves.

54. Pages 3/4 3-32 to 3/4 3-36, Surveillance Requirement 4.3.2.1, Table 4.3-2 is revised to incorporate administrative changes, in the the Functional Unit 1.c the Applicable Modes are changed from Modes "1, 2, 3" to="1, 2, 3, 4", and a reference is added to new Table Notation "(2) For the Plant Vent Activity - High Monitor, only a CHANNEL FUNCTIONAL TEST shall be perfomed at least once every 31 days" for entry 3.c.2).

55. Page 3/4 3-37, in Specification 3/4.3.3, replace the title with

" RADIATION MONITORING FOR PLANT OPERATION", and insert "for plant operations" after " instrument channelc" it. the first line of Specification 3.3.3.1, after " Alarm Trip Setpoint" in the first line of ACTION item a., after "monitcring channels" in the first line of ACTION item b., and after " instrumentation channels" in the first line of Surveillance Requirement 4.3.3.1.

56. Pages 3/4 3-38 and 3/4 3-39, Specification 3.3.3.1, in Table 3.3-6, add "FOR PLANT OPERATIONS" to the title, delete the headings "1.

AREA MONITORS" and "2.

PROCESS MONITORS"; change

" Control Room Ventilation Isolation" to " Control Room Ventilation Mode Change"; and change the number of " MINIMUM CHANNELS OPERABLE" for the same line from "1" to "2" with a reference to L

l'.

footnote "***", which is added as:

"*** One channel for each normal intake to the Control Room Ventilation hystem (common to both units)", and change ACTION 30 to "With less than the minimum Channels Operable requirement, operation may continue for up to 30 days provided an appropriate portable continuous monitor with the same Alarm Setpoint or an individual qualified in Radiation Protection Proc.?dures with a radiation does rate monitoring

+

,J t

device is provined in the Fuel Storage Pool Area. Restore the Inoperable monitors to OPERABLE status within 30 days or suspend all operations involving fuel movement in the fuel storage pool s y area."

57. Page 3/4 3-40, Specification 3.3.3.1, Table 4.3-3, is revised to be consistent with and incorporate the applicable changes made to Table 3.3-6 (see item 76 above).

58. Page 3/4 3-42, in Specification 3.3.3.3, a reference to footnote

"# The Seismic Monitoring instrumentation is common to both q:

units but located in Unit 1 or common areas" is added.

x

59. Page 3/4 3-45, in Specification 3.3.3.4, a reference to footnote

"# The meteorclogical monitoring instrumentation channels are comon to both units", is added.

60. Page 3/4 3-48, in Specification 3.3.3.5, ACTION item a is changed from "... channels less than required by Table 3.3-9..."

to "... channels less than the minimum channels OPERABLE requirement by Table 3.3-9,..."

61. Page 3/4 3-49, Specification 3.3.3.5. Table 3.3-9, delete the

" MEASUREMENT RANGE" column, change Instrument 5 to " Steam Generator Wide Range Water Level", and change Instrument 6 to " Condensate Storage Tank Water Level".

62. Page 3/4 3-50, Surveillance Requirement 4.3.3.5, Table 4.3-6 is revised to incorporate administrative changes that parallel those in Table 3.3-9 (see item 82 above).

63. Page 3/4 3-51, ACTION items a and b of Specification 3.3.36 are changed to indicate that the Reactor Vessel Level Indication System is exempted from their provisions. ACTION item e is redesignated as item g, and ACTION item e, which stipulates the provisions that apply to the Reactor Vessel level Indication System, and ACTION item f, stating that ACTION item e applies only to the first cycle, are added, and replace " containment sump" with " containment recirculation sump" in ACTION items b and c.

64. Page 3/4 3-52, Specification 3.3.3.6, Table 3.3-10 entries 9 and 10 are changed to "9.

Containment Reactor Cavity Sump Level - Wide Range" and "10.

Containment Recirculation Sump Level - Narrow Range", respectively, and a new entry added "20. ' Reactor Vessel Level Indication System" with the REQUIRED NUMBER OF CHANNELS AS 2 AND THE MINIMUM NUMBER OF CHANNELS OPERABLE as 1.

19 _

65. Page 3/4 3-53, in Table 4.3-7 entries 9 and 10 are changed to "9.

Containment Reactor Cavity Sump Level - Wide Range",

respectively, and a new entry "20.

Reactor Vessel Level Indication System" with "M" (monthly) channel check and "R" (every 18 months) channel calibration is added.

66. Page 3/4 3-54, in Specification 3.3.3.7, add a reference to footnote "# The Chlorine Detection System is common to both units installed in the normal intakes to the Control Room Ventilation System".

67. Page 3/4 3-55, in Specification 3.3.3.8 delete ACTION item a.2 ACTION item a.1 becomes part of the introductory statement to ACTION a, and ACTION item a.3 is redesignated as ACTION item c.

68. Pages 3/4 3-57 and 3/4 3-58, Table 3.3-11. Fire Detection Instruments, Panel B, Zone 6 data is changed to "6.

Fire Pump Area (5) 1 N.A.

Unit 2 Auxiliary Building Supply Fan Room 1

N.A.

Control Room Ventilation Equipment Room 1 N.A.";

the first item in Zone 16(1) is changed to " Unit 1 Auxiliary Building...", and note "(5) The Fire Pumps and Diesel Generator No. 3 are common to both units. Located on the Unit I side and on the Unit 1 Fire Detection Instrument Panel only" is added.

69. Page 3/4 3-69, in Specification 3.3.4.1, lines 3 through 7 under ACTION a are changed to "... steam line inoperable, restore the inoperable valve (s) to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or isolate the turbine from the steam supply within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ".

70. Page 3/4 4-2, in Specification 3.4.1.2 delete the reference to footnote "**" in the APPLICABILITY section and the footnote.

71. Page 3/4 4-3, in Specification 3.4.1.3, modify the introductory statement a to "At least two of the loops / trains listed below shall be OPERABLE and at least one of these loops / trains shall be in operation *", delete item b, reverse the order of the footnotes, and in ACTION items a and b, replace "RC loops and/or RHR trains" by " loops / trains".

72. Page 3/4 4-4, change the first two lines of Surveillance Requirements 4.4.1.3.1 to "... pump (s) and/or RHR pumps, if not in operation, shall be determined OPERABLE once..."

73. Page 3/4 4-4, the second line of Surveillance Requirement 4.4.1.3.2 is changed to "... secondary side water level to be..."

74. Page 3/4 4-5, in Specification 3.4.1.4.1., ACTION a is changed to "With one of the RHR trains inoperable and with less than the required steam generator water level, immediately initiate corrective action to return the inoperable RHR train to OPERABLE status or restore the required steam generator water level as soon as possible."

75. Page 3/4 4-5, delete Surveillance Requirements 4.4.1.4.1.1 and renumber the following two Surveillance Requirements.

76.

Page 3/4 4-6, delete Surveillance Requirement 4.4.1.4.2.1 and renumber the remaining Surveillance Requirement.

77. Page 3/4 4-7, in Specification 3.4.2.1, change the last two lines of the ACTION statement from "... and place an OPERABLE residual heat removal train into operation in the shutdown rooling mode." to "... and place a residual heat removal train into operation".

78.

Page 3/4 4-10, in Specification 3.4.4, delete "and remove power from the block valve (s);"in the third line of ACTION a.

79. Page 3/4 4-10, the last two lines of Surveillance Requirement 4.4.4.2 are changed to "... the block valve is closed in order to meet the requirement of ACTION a, of Specification 3.4.4.".

80.

Page 3/4 4-13, in Surveillance Requirement 4.4.5.3c.1) replace

" Primary" with " Reactor".

81. Page 3/4 4-18, in Specification 3.4.6.lb, replace " containment sump level" with " Containment Structure Sumps and the Reactor Cavity Sump Level".

82. Page 3/4 4-18, in Surveillance Requirement 4.4.6.lb, replace

" containment sump level" with " Containment Structure Sumps and the Reactor Cavity Sump Level".

83.

Page 3/4 4-19, change the second line of Specification 3.4.6.2e to "2235 + 20 psig, and ".

84. Page 3/4 4-20, in the third line in Surveillance Requirement 4.4.6.2.lc change the RCS pressure to "2235 + 20 psig".

85. Page 3/4 4-25, in Specification 3.4.8b, ACTION, Modes 1, 2, and 3, item c., and third line of ACTION, Modes 1, 2, 3, 4, 5, item a., insert "of gross radioactivity" after "microcuries/ gram".

86. Page 3/4 4-25, in Specification 3.4.8 ACTION a. that applies only during MODES 1, 2, and 3 is divided into two ACTION items,

a. and b., with no change in their provisions, and ACTION items

b. and c. are redesignated c. and d., respectively.

87.

Page 3/4 4-27, Specification 4.4.8, Table 4.4-4, add reference to footnote "**" in entry 1. and add footnote "**" which provides detailed guidance on accepted methodology to determine "groes

- radioactivity"; and add reference to footnote "***" in entry 3.

and add footnote "**" which provides detailed guidance on methodology to determine the " average disintegration energy".

88. Page 3/4 4-32. Table 4.4-5, insert " Unit 1" under the title, and add the Unit 2 reactor vessel material surveillance withdrawal schedule to make the table applicable to both units.

89. Page 3/4 5-1, in Specification 3.5.1 change items a and c to "The isolation valve open, and power removed," and "A boron concentration of between 1900 and 2200 ppm, and", respectively; and change the last two lines of ACTION item b to "... or be in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ."

90. Page 3/4 5-2, in Surveillance Requirement 4.5.1.1c replace the RCS pressure with "1000 psig", and delete Surveillance Requirement 4.5.1.1d.

91. Page 3/4 5-8, in Surveillance Requirement 4.5.3.1 insert

" Surveillance" between " applicable" and " requirements".

92. Page 3/4 5-8, in the second line of Surveillance Requirement 4.5.3.2 replace " required" with " allowed," and delete "per Specification 4.0.5" in the footnote.

93. Page 3/4 5-9, change item b of Specification 3.5.4.1 to "A boron concentration of between 20,000 and 22,500 ppm, and".

94. Page 3/4 5-11, in Specification 3.5.5, the OPERABILITY require-ments on the RWST, and the ACTION statement are editorially modified without changing the applicable provisions.

95. Pages 3/4 6-1, 3/4 6-7, and 3/4 6-8, in Technical Specifi-cation 3/4.6.1.1 replace " primary containment" with " containment".

96.

Page 3/4 6-1, change Surveillance Requirement 4.6.1.lb to "By verifying that each containment air lock is in compliance with the requirements of Specification 3.6.1.3; and".

97. Page 3/4 6-4, replace "per" with "by the requirements of" in item e to Surveillance Requirement 4.6.1.2.

98. Page 3/4 6-6, the second footnote in Surveillance Requirement 4.6.1.3 is changed to "** This represents an exemption to Appendix J, Paragraph III.D.2 of 10 CFR Part 50."

99. Page 3/4 6-11, in Items C.1) and C.2) of Surveillance Requirement 4.6.2.1, " Containment isolation phase "B" is replaced with " Phase "B" Isolation".

100. Page 3/4 6-15, delete " containment" from the first line in items a and b of Surveillance Requirement 4.6.3.2, and insert "its isolation" between "to" and " position" in the second line of item b.

i 101. Pages 3/4 6-19, 3/4 6-20, and 3/4 6-21, Specification 3.6.3, 1

in Table 3.6-1, delete the "1" for all manual valves showing a Unit 1 designation, to make the table applicable to both units. On page 3/4 6-20, delete the D from Valve RCS-512 and change the Function to Isolating Valve FI-927 OC.

102. Page 3/4 6-24, Specification 3.6.3, Table 3.6-1, replace the last 16 entries with Diablo Canyon plant specific valve designations, add " Unit 1 only" to valve FP-180, and add valve "VAC-252 FP-867 Containment Fire Water IC-Unit 2 only."

103. Page 3/4 6-26, delete " containment" from the first line of Specification 3.6.4.2.

104. Page 3/4 7-1, the last line of ACTION item a in Specifications 3.7.1.1 is changed to "and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ."

105. Page 3/4 7-2, Specification 3.7.1.1, for each entry in Table 3.7-1 add a reference to a new footnote "* Unless the Reactor l

Trip system breakers are in the open position."

106. Page 3/4 7-6, Specification 3.7.1.3, footnote "*", which applies to MODE 3 in the APPLICABILITY Section is deleted.

107. Page 3/4 7-8, Surveillance Requirement 4.7.1.4, Table 4.7-1 is revised to incorporate administrative changes.

1 108. Page 3/4 7-9, in the ACTION for Specification 3.7.1.5, the end of the statement for MODE 1 is changed to "... within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; otherwise be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ."; and the ACTION statement for MODES 2 and 3 is changed editorially and eliminates the exemption from the requirements of Specification 3.0.4, when the plant is in MODE 2 or 3 and the MSIV becomes inoperable.

109. Page 3/4 7-9, insert "The provisions of Specification 4.0.4 are not applicable for entry into MODE 3." at the end of l

Surveillance Requirement 4.7.1.5.

110. Page 3/4 7-11, replace " containment isolation phase "B" with

" Phase "B" Isolation" in item b of Surveillance Requirement 4.7.3.1.

111. Page 3/4 7-13, in Specification 3.7.5.1 add a reference to footnote "*The Control Room Ventilation System is common to both units" after " System" in the first line, and in the existing footnote replace "on the plant site" at the end of the paragraph with "within the SITE BOUNDARY."

4 112. Pages 3/4 7-14 and 3/4 7-15, in Surveillance Requirements 4.7.5.lc.3, 4.7.5.le.4, 4.7.5.lf, and 4.7.5.19, which apply to the Control Room Ventilation System; page 3/4 7-17, in Surveil-lance Requirements 4.7.5.lb.4, 4.7.6.1d.3, 4.7.6.le, 4.7.6.lf, 1

which apply to the Auxiliary Building Safeguards Air Filtration System; pages 3/4 9-14 and 3/5 9-15, in Surveillance Requirements 4.9.12b.4, 4.9.12e and 4.9.12f which apply to the Fuel Handling Ventilation System; on pages B 3/4 7-13, which provides the basis to support Specifications 3.4 7.5 and 3.4 7.6; and B 3/4 9-3, which provides the basis for Specification 3/4.9.12 i

change the applicable standards from " ANSI Standard N510-1975" to " ANSI Standard N510-1980."

113. Page 3/4 7-27, Surveillance Requirement 4.7.9.1, add reference in item c to footnote: "*Except valves which are l

located inside the containment and are locked, sealed, or otherwise secured in position. These valves shall be j

erified in the correct position during each COLD ~ SHUTDOWN j

except such verification need not be performed more often i

than once per 92 days.", after " valve" in the first line, j

and item d is changed to "At least once per 6 months by 1

performance of a system flush of the outside distribution loop to verify no flow blockage."

I' 114. Page 3/4 7-28, in Specification 3.7.9.2, delete reference to l

pumps " and 3" in item e, and insert the following two addi-i tional Spray and/or Sprinkler systems:

"f.

Centrifugal Charging Pump Area, and "h.

Containment Penetration Area."

115. Page 3/4 7-31, Table 3.7-3 is replaced by a revised Table 3.7-3 which incorporates the CO system Unit 2 data.

3 2

116. Pages 3/4 7-34 and 3/4 7-35, Table 3.7-14 is replaced with a revised Table 3.7-4 which incorporates information on the i

fire hose stations for Unit 2.

117. Page 3/4 7-37, Specification 3.7.11 is changed to "3.7.11 the temperature of each area shown in Table 3.7-5 shgil not be exceeded for more than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or by more than 30 F".

i 118. Page 3/4 7-37, in Specification 3.7.11, the format of ACTION items a and b is changed, and ACTION item a is revised to exempt the licensee from the provisions of Specifications 3.0.3 and 3.0.4 when the temperature limits of Table 3.7-5 t

are exceeded in one or more areas for more than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

i 119. Page 3/4 7-38, revise entry 15 in Table 3.7-5 to indicate l

that the Diesel Generator No. 3 Room is common to both units.

~

120. Page 3/4.7-39, in Specification 3.7.12 add a reference to footnote "*The UHS is common to both units" after (UHS) in the first line.

121. Page 3/4 7-40, in Specification 3.7.13 add a reference to footnote "*Both breakwaters are common to both units" after

"(east and west)" in the.first line.

122. Page 3/4 8-1, Specification 3.8.1.1, the footnote is changed to "*0PERABILITY of the third (common) diesel generatoi shall include the capability of functioning as a power source for the required Unit upon automatic demand from that Unit" to make it unit nonspecific.

123. Page 3/4 8-3, in footnote "*" to Surveillance Requirement 4.8.1.1.2, replace "for Unit 1" in the second line with "for one Unit" and "for Unit 2" in the fourth line with "for the other Unit" to make it unit nonspecific, and change footnote

"***" to Surveillance Requirement 4.8.1.1.2 to "May be the associated bus in the other Unit if that Unit is in Mode 1, 2, 3 or 4".

124. Page 3/4 8-8, the last sentence in the footnote to the Table 4.8-1 is changed to "For the purpose of this schedule, only valid tests conducted after the completion of the preoperational test requirements of Regulatory Guide 1.108, Revision 1, August 1977, shall be included in the computation of the "last 100 valid tests".

125. Page 3/4 8-11, in Specification 3.8.1.2, item b.3) is deleted, and item b.2) is changed to "One supply train of the Diesel Fuel Oil Storage and Transfer System with 8000 gallons of fuel in addition to the fuel required for the other Unit" to make it unit nonspecific.

126. Page 3/4 8-11, in Surveillance Requirement 4.8.1.2 the last line is changed to "... for ESF timers, b.6), b.7), b.10) and b.11)."

127. Fage 3/4 8-12, Specification 3.8.2.1d, e, f, g, h, i, j, k, 1, m, n, o, which, in reference to onsite power distribution, describe the electric busses that must be energized, delete the "1" immediately following the words " Bank" and " Bus", to make the statements applicable to both units.

128. Page 3/4 8-15, in Surveillance Requirement 4.8.3.1 items b.2) and c.3), replace "250 micro-ohms" with "150 x 10-6 ohm *" and add the footnote "* The resistance of cell-to-cell connecting cables does not have to be included."

129. Page 3/4 8-15, Surveillance Requirement 4.8.3.1 item e, which addresses the 60 month surveillance test to verify battery capacity, insert " required by Specification 4.8.3.1;" after

" service test".

130. Page 3/4 8-18, in Surveillance Requirement 4.8.4.1, item 1, replace " CHANNEL FUNCTIONAL" with " TRIP ACTUATION DEVICE OPERATIONAL".

131. Pages 3/4 8-19, 3/4 8-20 and 3/4 8-21. Specification 3.8.4.1, Table 3.8-1, for all components identified by a two-digit designation (e.g. SI pump 11 and Accumulator 14), delete

the first "1", to make the data in the table applicable to both units. Add a footnote

indicating that FCV-601 is i

common to both units.

132. Page 3/4 9-1, Specification 3.9.1, delete "With the reactor

}

vessel head closure bolts less than fully tensioned or with the head removed," in the introductory statement.

4 133. Page 3/4 9-2, Surveillance Requirements 4.9.2 items b and c

}

are changed to "An ANALOG CHANNEL OPERATIONAL TEST..."

134. Page'3/4 9-6, Surveillance Requirements 4.9.6.1 and 4.9.6.2, which apply to the manipulator crane and auxiliary hoist during refueling operations, respectively, in the third line i

of each replace "to the start of such operations" with "to j

removal of the reactor vessel head".

135. Page 3/4 9-7, Specification 3.9.7. delete the reference to

]

footnote "**" and the footnote.-

i 136. Page 3/4 9-7, Specification 3.9.7, the existing ACTION becomes ACTION item "a", and a new ACTION item b is added as "b".

The provisions of Specification 3.0.3 and 3.0.4 are not applicable.

137. Page 3/4 9-8, delete Surveillance Requirement 4.9.8.1.1 "The required RHR trains shall be demonstrated OPERABLE pursuant 1

- to Specification 4.0.5" and renumber the next surveillance j

requirement to "4.9.8.1".

I 138. Page 3/4 9-10, Specification 3.9.9, the existing ACTION j

becomes ACTION item a and "The provisions of Specification 3.0.4 are not applicable" in the last sentence is deleted, 5

and a new ACTION item b, "b.

The provisions of Specifi-cations 3.0.3 and 3.0.4 are not applicable" is added.

139. Page 3/4 9-11, in Specification 3.9.10, change the APPLICABILITY j

statement to "During movement of fuel assemblies or control

[

rods within the containment when either the fuel assemblies being moved or the fuel assemblies seated within the reactor vessel are irradiated while in MODE 6."

140. Page 3/4 9-11, in Surveillance Requirement 4.9.10 delete 1

]

"within the reactor vessel" at the end of the requirement.

t 141. Page 3/4 9-12, in Specification 3.9.11, the existing ACTION becomes ACTION item a and "above" is inserted before "specifi-i i

cation" in the first line, and a new ACTION item b "b.

The provisions of Specification 3.0.3 and 3.0.4 are not' applicable" j

is added.

i 142. Page 3/4 9-13,' Specification 3.9.12, ACTION item c is changed 4

. to "The provisions of Specifications 3.0.3 and 3.0.4 'are not applicable."

j 143. Page 3/4 9-16, Specification 3.9.13 is changed to "No spent fuel shipping cask handling operation near the spent fuel pool (i.e, any movement of a cask located north of column line 12.9 for Unit 1 or south of column line 23.1 for Unit 2) shall be performed unless spent fuel in all locations in Racks 5 and 6 has decayed for at least 1000 hours0.0116 days 0.278 hours 0.00165 weeks 3.805e-4 months since shutdown."

144. Page 3/4 10-1, in Surveillance Requirements 4.10.1.1 and 4.10.1.2, add " control" between " full-length" and " rod".

145. Page 3/410-2, editorial changes in Surveillance Require-ments 4.10.2.1 and 4.10.2.2 are made.

146. Page 3/4 10-3, in Surveillance Requirement 4.10.3.2, replace "a CHANNEL FUNCTIONAL TEST" with "an ANALOG CHANNEL OPERATIONAL TEST".

147. Page 3/4 10-5, Specification 3.10.5, item a is incorporated into the introductory statement, item b and the applicable footnote are deleted; add "and during surveillance of digital position indication for OPERABILITY" to the APPLICABILITY, and, to account for the deletion of Technical Specification 3/4.10.4 (see 195 above), redesignate this Specification to 3.10.4, and the corresponding Surveillance Requirements to 4.10.4.

148. Page 5-5, replace Specification 5.4.2 with "The total water andsteamvolumeofthereactorcoolantsfstemis"12,811

+ 100 cubic feet at a nominal T of576Fforgnit1and T2,903 + 100 cubic feet at a no d al T of 577 F for Unit 2."

3yg 149. Pa:;es 6-1 through 6-25, replace " unit" with " plant" throughout Specification 6.0, Administrative Controls, and on pages 6-6 and 6-8 delete the heading " AUTHORITY".

150. Pages 6-2, Specification 6.2.1, Figure 6.2-1, Offsite Organi-zation, replace "on site" by " plant," delete the position of " Manager, Nuclear Plant Operations"; " Technical Assistant to Vice President, Nuclear Power Generation", and " Quality Assurance Engineer," become " Manager, Nuclear Operations Support" and " Director, Quality Support," respectively.

Figure 6.2.2 Plant Organization, replace " Technical Assistant to Vice President" and " Quality Assurance Engineers" with

" Manager, Nuclear Operations Support" and " Director, Quality Support", respectively.

151. Page 6-5, in Specification 6.2.3.1, insert the statement in Specification 6.2.3.4:

"The OSRG shall make detailed reco rnendations for revised procedures, equipment modifi-cations, maintenance activities, operations activities or other means of improving plant safety to the Manager, Nuclear Operations Support.", which is deleted from its original location as Section 6.2.3.4 on page 6-6.

152. Page 6-6, Specification 6.3.1 is redesignated "6.3", revised editorially, and the minimum qualifications for licensed operator and senior operators are introduced.

153. Page 6-6, delete " identified by the OSRG" at the end of Specification 6.4.1.

154. Page 6-8 in Specification 6.5.1.8, replace " Manager of Nuclear Plant Operations" with "Vice-President Nuclear Power Generation",

and delete " Chairman of".

155. Page 6-10 in Specification 6.5.2.7, change item g. to "All REPORTABLE EVENTS", and replace item 1 with " Reports and meeting minutes of the Plant Staff Review Committee and

)

the Onsite Safety Review Group."

156. Page 6-12, Specification 6.6, in the title "0CCURRENCE" is replaced with " EVENT" and the specification is revised to incorprate the REPORTABLE EVENT requirements and clarify the role of the PSRC and GONPRAC, and reflect organization changes.

157. Page 6-12, Specification 6.7, is modified to reflect organization changes and administrative changes to be taken when safety limits are violated.

158. Page 6-13, change items b through g in Specification 6.8.1 to "b.The emergency operating procedures required to implement the requirements of NUREG-0737 and Supplement 1 to NUREG-0737 as stated in Generic Letter No. 82-33, c.

Security plan implementation, d.

Emergency Plan implementation, e.

PROCESS CONTROL PROGRAM implementation, f.

ODCP and ERMP implementation, and g.

Quality Assurance Program for effluent and environmental monitoring."

159. Page 6-15, in Specification 6.9.1, "AND REPORTABLE OCCURRENCES" is deleted from the title, " Director" is replaced with " Regional Administrator" and " Inspection and Enforcement" with "the NRC."; and the date for submittal of the Annual Reports is changed to March 31 in Specification 6.9.1.4.

160. Pages 6-19, 6-20 and 6-21, to reflect new reporting require-ments and the LER system Specifications 6.9.1.11, 6.9.1.12 and 6.9.1.13 are deleted, and Specification 6.9.1.14 which is revised administratively, is redesignated as "6.9.1.8."

161. All Technical Specifications are revised to correct typo-graphical errors (e.g., replacing "1" by "1", KV by kv and "GPM" by "gpm" throughout), and conform to updated nomen-clature (e.g., replacing " percent" by2 ",use of scientific

"% " " delta' by "

" " secs" by "s," " square feet" by "ft,

notation, and spelled out numbers by digits when referring to surveillance time intervals), and notation changes (e.g., replacing i), 11), iii)... by 1), 2), 3)... in page 3/4 3-19; a, b, c, d, and e by 1, 2, 3, 4, 5 on page b, c,... by 1, 2, 3,

... and 1, 2, 3,... by a, b, c,... on page 3/4 11-2, Table 4.11-1 and page 3/4 11-9, Table 4.11-2; a, b, c, d, by 1, 2, 3, 4 on page B 3/4 2-4, Bases for Specifications 3/4.2.2,3/4.2.3,and floating notation by decimal notation on page 3/4 12-5, Table 3.12-2 and page 3/4 12-6, Table 4.12-1).

l The staff has evaluated the changes and have concluded that they are purely administrative in nature and do not change any sub-I stantive requirement or safety parameter or otherwise affect operation of the facility and are, therefore, acceptable.

B.

Changes to the Technical Specifications Resulting in Additional Restrictions The following proposed changes to the Unit 1 Technical Specifications will result in additional operational controls and restrictions, surveillance testing and verification requirements, and more restrictive ACTION items than those presently included in the Technical Spections, i

1.

Page 3/4 1-19, Surveillance Requirement 4.1.3.3 is changed to "4.1.3.3.

Each of the above required digital rod position indicator (s) shall be determined to be OPERABLE by verifying that the digital rod position indicators agree with the i

demand position indicators within 12 steps when exercised over the full range of rod travel at least once per 18 months."

L 2.

Pages 3/4 2-1 and 3/4 2-2. Specification 3.21, Axial Flux Difference, is replaced with a revised Specification 3.2.1, which extends the range of applicability from "above 50%

l RATED THERMAL POWER" to "above 15% RATED THERMAL POWER" and incorporates the LIMITING CONDITION FOR OPERATION and ACTION items that apply within the range of 15% to 50% RATED THERMAL POWER.

3.

Pages 3/4 3-2 to 3/4 3-7, Specification 3.3.1. Table 3.3-1, Table Notations, is replaced with revised Table Notations to incorporate administrative changes.

In addition, the second and third lines in ACTION 11 are changed to

"... channels OPERABLE requirement, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; however one channel may be bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance..."

4.

Pages 3/4 3-15 and 3/4 3-18 Specification 3.3.2, Table 3.3-3 in the Functional Units 1.c. and 4.a. the Apolicable Modes are changed from Modes "1, 2, 3" to "1. 2, 3, 4", and the reference to Table Notation "H" in the Applicable Modes column for the Functional Unit 1.e entry is deleted.

5.

Pages 3/4 3-28 and 3/4 3-31, Specification 3.3.2 Table 3.3-5 is revised to incorporate administrative changes, replace the entry under Initiating Signal 8 with " Turbine Trip", and

l l

add main feedwater bypass valves information to Table Notation (2) and, consequently, the requirement to verify response time l

on these valves.

l l

6.

Pages 3/4 3-32 to 3/4 3-36, Surveillance Requirement 4.3.2.1, Table 4.3-2 is revised to incorporate administrative changes, in the the Functional Unit 1.c the Applicable Modes are changed from Modes "1. 2, 3" to "1. 2, 3, 4", and a reference isaddedtonewTableNotation"(2) For the Plant Vent Activity - High Monitor, only a CHANNEL FUNCTIONAL TEST shall be performed at least once every 31 days" for entry 1.c.2).

7.

Pages 3/4 3-38 and 3/4 3-39 Specification 3.3.3.1, in Table 3.3 6, add "FOR PLANT OPERATIONS" to the title, delete the heac'ings "1.

AREA MONITOR $" and "2.

PROCESS MONITORS";

change " Control Room Ventilation Isolation" to " Control Room Ventilation Mode Change"; and change the number of " MINIMUM CHANNELS OPERABLE" for the same line from "1" to "2" with a t

reference to footnote "***", which is added as: "*** One channel for each normal intake to the Control Room Ventilation System (common to both units)", and change ACTION 30 to "With less than the minimum Channels Operable requirement, operation may continue for up to 30 days provided an appropriate portable continuous monitor with the same Alarm Setpoint or an individual qualified in Radiation Protection Procedures with a radiation does rate monitoring device is provided in the Fuel Storage Pool Area.

Restore the Inoperable monitors to OPERABLE status within 30 days or suspend all operations involving fuel movement in the fuel storage pool area."

8.

Page 3/4 3-69, in Specification 3.3.4.1, changes lines 2 and 3 of ACTION b to "... within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months isolate the turbine from the steam supply".

9.

Pages 3/4 4-29, 3/4 4-30 and 3/4 4-31, add "and the setpoint of Technical SP'..ification 3.4.9.3a" at the end of Surveillance Requirenent 4.4.9.1.2, and change the period for the material l

curves in Figures 3.4-2 and 3.4-3, Specification 3.4.9.1, to i

"6 EFPY."

10. Page 3/t. 5-1, in Specification 3.5.1 change items a and c to "The isolation valve open, and power removed," and "A boron concentration of between 1900 and 2200 ppm, and", respectively; and change the last two lines of ACTION item b to "... or be l

in HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the l

following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ."

11. Page 3/4 5-2, in Surveillance Requirement 4.5.1.lc replace the RCS pressure with "1000 psig", and delete Surveillance Requirement 4.5.1.1d.

12. Page 3/4 6-7. Specification 3.6.1.4, replace the statenent

" Primary containment internal pressure shall be maintained between -3.5 and +0.3 psig." with " Containment internal pressure shall be maintained between -1.0 and +0.3 psig."

13.

Page 3/4 7-1, the last line of ACTION item a in Specifi-cations 3.7.1.1 is changed to "and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ."

i 14 Page 3/4 7-6. Specification 3.7.1.3. footnote "*", which 1

applies to N0DE 3 in the APPLICABILITY Section is deleted.

15. Page 3/4 7-9, in the ACTION for Specification 3.7.1.5, the end of the statement for MODE 1 is changed to "... within 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months s: otherwise be in HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ."; and the ACTION statement for MODES 2 and 3 is changed editorially and eliminates the exemption from the requirements of S)ecification 3.0.4, when the plant is in MODE 2 or 3 and j

t.1e MSIV becomes inoperable, i

i

16. Page 3/4 8-2, Specification 3.8.1.1, ACTION C, add "If these conditions are not satisfied within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months be in at least i

HOT STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in COLD SHUTOOWN 1

within the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months .

17. Page 3/4 8-8, the last sentence in the footnote to the Table 4.8-1 is changed to "For the lurpose of this schedule, only valid tests conducted after t1e completion of the preoperational test requirements of Regulatory Guide 1.108, Revision 1, August 1977, shall be included in the computation of the "last 100 valid tests".

18.

Page 3/4 8-12, Specification 3.8.2.1, at the end of items m, n and o, add "and its associated full capacity charger".

f

19. Page 3/4 8-12a, after " associated Battery Bank" in the first j

and second lines of ACTION item c of Specification 3.8.2.1 add "and full-capacity charger".

l

20. Page 3/4 8-13, add "and full-capacity charger" at the end of item e of Specification 3.8.2.2.

21. Page 3/4 8-16, Surveillance Requirement 4.C.3.1. Table 4.8-3, in footnote (b) replace "5 amps" with "2 arps".

22. Pages 3/4 8-22 through 3/4 8-26, a new Specification 3.8.4.2, and corresponding Survelliance Requirement 4.8.4.2, which apply to Containment Penetration Conductor Overcurrent Protective Devices is added.

23.

Page 3/4 9-9, change Surveillence Requirement 4.9,8.2 to "At least one RHR loop shall be verified in operation and circulating reactor coolant at a flow rate greater than or equal to 3000 9pm at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ."

24 Page 3/4 10 4, Technical Specification 3/4.10.4, which i

addresses Special Test Exceptions for RCS loops during natural circulation tests is deleted.

25. Page 6-6 Specification 6.3.1 is redesignated "6.3", revised editorially, and the minimum qualifications for licensed operator and senior operators are introduced.

26 Page 3/4 3-44, Surveillance Requirement 4.3.3.3.1. Table 4.3-4, Seismic Monitoring Instrumentation Surveillance Requirements, and the requirement to perform a channel calibration in accordance with ANSI /ANS-2.2-1978 each refueling outage for the triaxial peak accelographs and the triaxial response-spectrun recorders.

27. Page 3/4 3-52, Specification 3.3.3.6, Table 3.3-10 Accident Monitoring Instrumentation, for the PORV Position Indicator, change the required number of channels from 1/ valve to 2/ valve as a requirement (cr backup instrumen-

)

tation. Add footnotes to identify indication as one direct, stem-mounted indicator per valve and one comon i

j temperature element.

28. Page 3/4 7-1. Specification 3.7.1.1, Turbine Cycle Safety Valves, delete Action b., which allows a maximum of 14 main steam line safety valves to be made inoperable in Mode 3 to permit insitu testing of the OPERABLE safety valve.

29. Page 3/4 7-18 Survet11ance Requirement 4.7.7.1, Snubber Surveillance, change "in lieu of" to "and" in the intro-ductory statement, thus requiring each snubber to be demon-strated OPERABLE by performance of both the augmented inservice program (described subsequently in Surveillance Requirement 4.7.7.1) and the requirements of Specification 4.0.5.

The Staff has reviewed each of the foregoing against their prior evaluations and has concluded that each is encompassed by a prior evaluation of the respective system or corrponent and are, therefore, acceptabis.

C.

Changes To Provide Relief From Current Cperating Restrictions, The proposed changes listed below provide request for relief from an operating restriction, control or limitation en the basis that acceptable operation under the proposed conditions has been demonstrated. The original restrictions had been imposed because acceptable operation had not been yet demonstrated.

1.

Page 3/4 2-9, in the second line of Specification 3.2.3.

i "R, R " is replaced with "R", and item a is changed to define"R".

2 2.

Page 3/4 2-9 in Specification 3.2.3, item b, which defines R in terms of R and the Rod Bow Penalty (RDP(BU)), and item 2

3 e, which defines RBP(BU), are deleted, and items "c" and "d" are redesignated as items "b" and "c", respectively.

m

l I '

i 3.

Page 3/4 2-9, in Specification 3.2.3, thg third line of new l

1-item c is changed to "... values of F H shall be used to l

calculate R since Figure 3.2-3a for Unit I and Figure 3.2-3b l

for Unit 2... "

4.

Pages 3/4 2-9 and 3/4 2-10, in Specification 3.2.3, ACTION items a.1, b, and c. "R, R " is replaced with "R".

g 2

5.

Page 3/4 2-10, in Specification 3.2.3, in the seventh line i

of ACTION item c " Figure 3.2-3" is replaced with " Figure

{

3.2-3a for Unit I and Figure 3.2-3b for Unit 2".

6.

Page 3/4 2-10, in Surveillance Requirement 4.2.3.2 "R,

i R"isreplacedwith"R"and" Figure 3.2-3"with"Figbre 23.2-3a for Unit I and Figure 3.2-3b for Unit 2".

7.

Page 3/4 2-11, in Surveillance Requirement 4.2.3.3, " Figure 1

3.2-3" is replaced with "Ffoure 3.2-3a for Unit I and Figure j

3.2-3b for Unit 2" and "R, R " is replaced with "R".

3 2

I 8.

Page 3/4 2-12, Specification 3.2.3, Figure 3.2-3 is i

I redesignated Figure 3.2-3a, the DNB limit curve and the i

designation that the other curve represents the "LOCA LIMIT" l

are deleted, and an indication that it applies to Unit 1 is added.

}

9.

.Page 3/4 2-13, Specification 3.2.3, Figure 3.2-4 is replaced with a new Figure 3.2-3b, which provides RCS Total Flowrate l

versus R for Unit 2.

i

10. Pages 3/4 3-16 and 3/4 3-17, Table 3.3-3, the infomation in the CHANNELS TO TRIP COLUMN for the 2.a and 3.b.1 entries is changed to "2 with 2 coincident switches", and a reference to l

Table Notation "*" in the ACTION column for the 3.c.1) and 3.c.2)entriesadded.

11. Page 3/4 3-18, Specification 3.3.2, Table 3.3-3, a reference to Table Notation " H" in the Applicable Modes column is i

added to the 4.d entry, the ACTION for the 5.a entry is changed to "25", and " ACTION 25 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE i

requirement, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ; j

however,

12. Page 3/4 3-40, Specification 3.3.3.1, Table 4.3-3, is revised i

to be consistent with and incorporate the applicable changes l

mace to Table 3.3-6.

1

13. Page 3/4 4-1, in Specification 3.4.1., delete the reference l

to the footnote in the APPLICABILITY section and the footnote, and change the ACTION statement to "With less than the above 1

required reactor coolant loops in operation, be in at least HOT STANDBY within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months ".

1 i

4

14 Page B 3/4 2-4, in Bases for Specifications 3/4.2.2 and 3/4.2.3, which support hot channel factor limits, in the bottom paragraph: replace "R," by "R" in the first line, and delete "R9 as defined, allows for the inclusion of a penalty for Rbd Bow on DNBR Only." in lines 5 and 6.

15. Pages 3-1 through 3-13 and page B 3/4 3-1, revise Section 3/4.3.1, Reactor Trip System Instrumentation, which specifies the serveillance test requirements for the Reactor Trip System Instrumentation channels and interlocks and the automatic trip logic. The changes incorporate new surveillance requirements as recomended by WCAP-10271. Specifically, the proposed changes increase the surveillance intervals of the Reactor Trip System and time that an inoperable Reactor Trip System analog channel may be maintained or bypassed to allow the testing of another channel.

Subsequent to the issuance of the Unit 1 Technical Specifications additional information has been made available to the staff.

Based on the staff's review of this information against previous evaluations, we find that they confonn with NRC requirements and are, therefore, acceptable. Morever, these changes will result in additional restrictions such that they enhance safe operation.

D.

Modification to Technical Specifications that Constitute Changes to Previously Analyzed Occurrences The proposed changes presented below constitute 'ar to previously analyzed occurrences, but the results are clear 13.ioin all accept-able criteria with respect to the system or compone:,as specified in the Standard Review Plan.

1.

Pages 3/4 1-15 and 3/4 1-16, in Specification 3.1.3.1, Movable Control Assemblies, delete ACTION item b and add new ACTION items c and d.

The previous ACTION item c is redesignated to item b.

This change would allow the use of rods which are imovable but retain their trip function and are properly positioned.

2.

Page 3/4 3-13, Surveillance Requirements 4.3.1.1 Table 4.3-1, the Table Notation "*" is revised to incorporate administrative changes, and Table Notation (6) is changed to "Incore-Excore Calibration, above 75% of RATED THERMAL POWER" and "The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1" is added at the end of (2), (3) and (6).

3.

Page 3/4 3-42, Surveillance Requirement 4.3.3.3.2, which applies to seismic monitoring instrumentation following a seismic event, the time allowed for CHANNEL CALIBRATION in the third line is extended from "5" to "10" days, and the time allowed for sub-mittal of a Special Report in the seventh line is extended for "10" to "14" days.

4.

Pages 3/4 3-65 and 3/4 3-66. Specification 3.3.3.10, in Table 3.3-13, the reference under the APPLICABILITY column for the Noble Gas Activity Monitor in entry 5 is changed to "*** Modes 1-4, also Mode 6 during CORE ALTERATIONS or movement of irradiated fuel within containment," and the Instrument designation "(RM-14A and 148)" is changed to "(RM-14A or 14B)."

5.

Pages 3/4 3-67 and 3/4 3-68, Surveillance Requirement 4.3.3.10, in Table 4.3-9, the reference under the MODES FOR WHICH SURVEILLANCE REQUIRED column for the Noble Gas Activity Monitor in entry 5 is changed to "*** Modes 1-4, also Mode 6 during CORE ALTERATIONS or movement of irradiated fuel within contain-ment", and the Instrument designation "(RM-14A and 148)" is changed to "(RM-14A or 14B)".

6.

Page 3/4 5-2, in Surveillance Requirement 4.5.1.1c replace the RCS pressure with "1000 psig", and delete Surveillance Requirement 4.5.1.1d.

7.

Page 3/4 6-2, in Specification 3.6.1.2, add "as applicable,"

at the end of item a.1) and delete "(a)" and "(b) with" in ACTION statement.

8.

Page 3/4 6-3, in Surveillance Requirement 4.6.1.2, insert "as applicable," after "0.75L}r"d lines of item c.1) of Surveillance in lines 1, 4 and 7 of item b, and replace the second and th Requirement 4.6.1.2 with "... supplemental test results L minus the sum of the Type A and the superimposed leak. L,' is c

equal to or less than 0.25L, or 0.25 L, as applicable.0; and changeitemc.3)to"Requir$sthatthefateatwhichgasis injected into the containment or bled from the containment during the supplemental test is between 0.75 L and 1.25 L '

a a

or 0.75 L and 1.25 L, as applicable."

t t

9.

Page 3/4 7-36, the APPLICABILITY statement in Specification 3.7.10 is changed to "Whenever the equipment protected by the fire barrier penetrations is required to be OPERABLE.", to conform to the current NRC position as presented in the April 5, 1984 Regional Appendix R Workshop.

10. Page 3/4 8-15, Surveillance Requirement 4.8.3.1 item f, the first and second lines are changed to "At least once per 18 months during shutdown by giving performance discharge tests of battery capacity to any battery that shows...".

11. Page 3/4 9-9, Specification 3.9.8.2, add a reference to footnote "*" at the end of the statement, add a new footnote

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

12. Page 3/4 10-5, Specification 3.10.5, item a is incorporated into the introductory statement, item b and the applicable footnote are deleted; add "and during surveillance of digital position indication for OPERABILITY" to the APPLICABILITY, and,

to account for the deletion of Technical Specification 3/4.10.4 (see 195 above), redesignate this Specification to 3.10.4, and the corresponding Surveillance Requirements to 4.10.4.

The staff has reviewed the above proposed changes and find them to be consirtent with the provisions of the Standard Review Plan, bounded by staff previous evalutations and, are, therefore, acceptable.

E.

Changes To Conform To Regulations The proposed changes listed below constitute changes made to conform to changes in the regulations.

1.

Page 1-6, Definition 1.29 of REPORTABLE OCCURRENCE, is replaced with " REPORTABLE EVENT shall be any of the conditions specified in Section 50.73 of 10 CFR Part 50.." and the term " REPORTABLE OCCURRENCE" is changed throughout the Technical Specifications to " REPORTABLE EVENT."

2.

Page 1-4,1-5,1-6 and 1-7, Definitions 1.19, 1.20, 1.21, 1.22, 1.23, 1.24, 1.25, 1.26, 1.27, 1.28, 1.29, 1.30, 1.31, 1.32, 1.33, 1.34, 1.35, 1.36, 1.37, and 1.38 are renumbered to 1.20, 1.21, 1.22, 1.23, 1.24, 12.5, 1.26, 1.27, 1.28, 1.29, 1.30, 1.31, 1.33, 1.34, 1.35, 1.36, 1.37, 1.38, 1.39, 1.41 and 1.42, respec-tively, to account for the three new definitions 1.19, 1.32 and 1.40.

3.

Page 2-1, in Specifications 2.1.1 and 2.1.2, Reactor Core and Reactor Coolant System Pressure, add "... and comply with the requirements of Specification 6.7" at the end of the ACTION statement.

4 Page 3/4 3-55, in Specification 3.3.3.8 delete ACTION item a.2, ACTION item a.1 becomes part of the introductory statement to ACTION a, and ACTION item a.3 is redesignated as ACTION item c.

5.

Page 3/4 4-15, change lines 2, 3, and 4 of Surveillance Requirement 4.4.5.5c. to "... Category C-3 shall be reported in a Special Report pursuant to Specification 6.9.2 within 30 days prior to resumption of operation. This report shall provided...".

6.

Page 3/4 4-17, Surveillance Requirement 4.4.5, Table 4.4-2, the reporting requirements when the results of a first and second sample steam generator tube inspection fall in Categor C-3, are changed to " Notification to NRC pursuant to 50.72(y)(2) to b

10 CFR Part 50."

7.

Pages 3/4 4-25 and 3/4 4-26, the second sentence in the intro-ductory statement to the ACTION item a., which applies during MODES 1, 2, 3, 4, and 5, is replaced with "For this ACTION statement, prepare a Special Report to the Commission pursuant to Specification 6.9.2 within 30 days with a copy to the Director, Nuclear Reactor Regulation, Attention: Chief, Core Performance

Branch, and Chief, Accident Analysis Branch, U.S. Nuclear Regulatory Commission, Washington, D.C.,

20555.". The word "first" is deleted from item 1 under the same ACTION item a, and item 2 is replaced with "2.

Results of a) the last isotopic analysis for radionuclides performed prior to exceeding the limit, b) analysis while limit was exceeded, and c) one analysis after the radioiodine activity was reduced to less than the limit including each isotopic analysis,

~

and the date and time of sampling, and the radiciodine concentrations;"

8.

Page 3/4 4-34, in Specification 3.4.9.3, delete existing item 1, and combine the existing items b and c to create a new item a.

To correspond to the above change, delete ACTION item a and redesignate the letters of the following ACTION items. These changes are made to conform with NRC requirements as described in Diablo Canyon SSER-21, 9.

Page 3/4 4-35, delete items d and e in Surveillance Requirement 4.4.9.3.1.

10. Page 3/4 6-9, the third line in Surveillance Requirement 4.6.1.6.2 is changed to "... Commission is a Special Peport pursuant to Specification 6.9.2 within 15 days. This report shall include a..."

11. Page 3/4 7-26, ACTION items a and b in Specification 3.7.9.1 are changed to remove the requirements to submit Special Reports pursuant to Specification 6.9.2 when portions of the Fire Suppression Water System are inoperable.

12. Pages 3/4 7-28, 3/4 7-30, 3/4 7-32, 3/4 7-33 and 3/4 7-36, in ACTION a to Specifications 3.7.9.2, 3.7.9.3, 3.7.9.4, 3.7.9.5 and 3.7.10, all of which refer to fire protection, delete the requirement for a Special Report pursuant to Specification 6.9.2 in case diverse portions of the system cannot be restored to OPERABLE status within 14 days.

13. Page 3/4 7-36, the APPLICABILITY statement in Specification 3.7.10 is changed to "Whenever the equipment protected by the fire barrier penetrations is required to be OPERABLE.",

to conform to the current NRC position as presented in the April 5, 1984 Regional Appendix R Workshop.

14. Page 3/4 8-7, the second line in Surveillance Requirement 4.8.1.1.4 is changed to "be reported as a Special Report within 30 days to the Commission pursuant to Specification 6.9.2".

. 15. Page 6-7, Specification 6.5.1.6, replace " Manager of Nuclear Plant Operations" with "Vice-President, Nuclear Power Generation" in item c, and item f with " Review of all REPORTABLE EVENTS".

16. Page 6-10 in Specification 6.5.2.7, change item g. to "All REPORTABLE EVENTS", and replace item 1. with " Reports and meeting minutes of the Plant Staff Review Committee and the Onsite Safety Review Group."

17. Page 6-12, Specification 6.6, in the title "0CCURRENCE" is replaced with " EVENT" and the specification is revised to incorprate the REPORTABLE EVENT requirements and clarify the role of the PSRC and GONPRAC, and reflect organization changes.

j

18. Pages 6-19, 6-20 and 6-21, to reflect new reporting requirements and the LER system Specifications 6.9.1.11, 6.9.1.12 and 6.9.1.13 are deleted, and Specification 6.9.1.14 which is revised adminis-tratively, is redesignated as "6.9.1.8."

19. Page 6-22, change item c to "ALL REPORTABLE EVENTS" in Specifi-cation 6.10.1.

4 The proposed changes have already been incorporated by the Comission into the current Unit 2 Technical Specifications. The staff has reviewed these changes and finds that incorporation of these same changes into Unit 1 Technical Specifications will upgrade the specifications for Unit I and make the Unit 1 Technical Specications essentially equivalent in tems of content, style and format. Based on the staff review, wr find that the changes confom to the regula-tions and are, there, acceptable.

SUMARY AND CONCLUSIONS PG&E submitted a number of license amendment requests for changes to the Techni-cal Specifications for the Unit 1 full-power license and the Unit 2 low-power license in order to (1) establish common and combined Technical Specifications for both units (2) make editorial and fonnat changes (3) revise the specifications for movable control rod assemblies, reactor trip system instrumentation, PG&E General Office Nuclear Plant Review and Audit Comittee (GONPRAC) composition, Radiological Effluent Technical Specifi-cation, auxiliary feedwater system, diesel-fuel oil surveillance require-ments, and thermal hydraulic design effect of fuel rod bowing on DNB.

f (4) resolve certain full-power considerations that had been identified by the staff during the Unit 2 low-power licensing.

Based on the staff review of the infonnation provided by PG&E and based on further discussion with PG&E the staff has developed common and combined Technical Specifications for Units 1 and ? that include the above considera-tions. The Technical Specifications for six issues identified under Unit 2 Tech Spec considerations will be revised BY LAR for both units after issuance of the specifications.

. r,.

CONTACT WITH STATE OFFICIAL The NRC staff has advised the Chief of the Radiological Health Branch, State Department of Health Services State of California, of the proposed determina-tions of no significant hazards consideration. No coments were received.

ENVIRONMENTAL CONSIDERATION This amendment involves changes in administrative procedures and requirements and in the installation or use of facility components located within the re -

stricted area, respectively. The staff has determined that the amendment involves no significant increase in the amounts of any effluents that may be released offsite and that there is no significant increase in the individual or cumulative occupation radiation exposure. The Comission has previously issued proposed findings that the amendment involves no significant hazards consideration, and there has been no public coment on such findings. Ac-cordingly, the amendment meets the eligibility criteria for categorical exclu-sion set forth in 10 CFR Sec. 51.22(c)(9) and (10).

Pursuant to 10 CFR 51.22(b) no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

CONCLUSION Based upon our evaluation of the proposed changes to the Diablo Canyon, Unit 1 Technical Specifications, we have concluded that:

there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and such activities will be conducted in compliance with the Cornission's regulations and the issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public.

Dated: August 26, 1985

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1 REFERENCES 1.

NUREG-1102, November 1984, " Technical Specifications, Diablo Canyon Nuclear Power Plant, Unit No.1."

2.

NUREG-1132, April 1985, " Technical Specifications, Diablo Canyon Nuclear Power Plant, Unit No. 2."

i 3.

U.S. Nuclear Regulatory Commission letter from H. L. Thompson, Jr. (NRC) to J. D. Shiffer (PG&E), May 15, 1985, "Diablo Canyon Technical Specifications."

l 4.

PG&E letter DCL-85-028, January 29, 1985, "Diablo Canyon Unit 1 - License i

Amendment Request 85-01, Technical Specification Changes."

(

5.

PG&E letter DCL-85-148, April 12,1985, "Diablo Canyon Unit 1 - License Amendment Request 85-01, Revision 1, Radiological Environmental Monitoring."

6.

PG&E letter DCL-85-164, April 24,1985, "Diablo Canyon Unit 1 - License i

Amendment Request 85-01, Revision 2, Technical Specifications (Applicable to Radiological Effluents)."

7.

PG8E letter DCL-85-184, May 14,1985, "Diablo Canyon Unit 1 - License Amendment Request 85-01, Revision 3, Diesel Fuel Oil Specification."

8.

PG&E letter DCL-85-185, May 14, 1985, "Diablo Canyon Units 1 and 2 -

License Amendment Requests: 85-01, Revision 4 Unit 1 Technical Specifica-tions; 85-02, Unit 2 Full Power Technical Specifications; and 85-03, Units 1 and 2, Movable Control Rod Assemblies."

9.

PG&E letter DCL-85-187, May 20,1985, "Diablo Canyon Units 1 and 2 -

I License Amendment Request 85-04 Technical Specification 3/4.3.1, Reactor i

Trip System Instrumentation."

i i

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- 10 PG4E letter DCL-85-197, May 30,1985, "Diablo Canyon Unit 1 - License 1

Amendment Request 85-05, Technical Specification Changes Sections 6.8.1, 6.9.1.2, 4.3.3.3.1, 3.3.3.6, 3.7.1.1, and 4.7.7.1."

t 11.

PG4E letter DCL-85-198, May 30,1985, "Diablo Canyon Units 1 and 2 -

License Amendment Request 85-06, Technical Specification Change Section 6.5.2.2, GONPRAC Composition."

12.

PG&E letter DCL-85-199, May 31, 1985, "Diablo Canyon Unit 1 - License Amendment Request 85-07. Technical Specification Changes, Specifications 3.8.2.1 and 3.8.2.2."

I 13.

PG&E letter DCL-85-214, June 14, 1985, "Diablo Canyon Unit 1 - Supplemental Infonnation to License Amendment Request 85-07."

14 PG&E letter DCL-85-221, June 20,1985, "Diablo Canyon Units 1 and 2 -

Additional Information Regarding the Combined Units 1 and 2 Technical Specifications."

t l

15.

PG&E letter DCL-85-188 May 21, 1985. "Diablo Canyon Units 1 and 2 -

Combined Technical Specifications."

i 16.

U.S. Nuclear Regulatory Comission. April 26, 1985 " Amendment No. I to License No. DPR-80, Diablo Canyon Unit 1, Docket No. 50-275."

1 17.

U.S. Nuclear Regulatory Comission February 21, 1985, letter from C.O. Thomas (NRC)toJ.J.Sheppard(WOwnersGroup).

4 18.

Westinghouse Report WCAP 10271, Supplement 1. July 1983, " Evaluation of Surve11ance Frequencies and Out-Of-Service Times for the Reactor Protec-tion System.'

19.

Westinghouse letter, December 21, 1984, from E. P. Rahe, Jr. (W) to j

C. O. Thomas (NRC) re specification and bases for movable control rod 1

assemblies.

l 1

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41 -

I i

l 20.

U.S. Nuclear Regulatory Commission memorandum, May 30, 1985 from I

H. Schierling (NRC), "Diablo Canyon Units 1-and 2 - NRC/PGAE Meeting i

on May 15, 1985."

r 21.

U.S. Nuclear Regulatory Commission, letter from H. Schierling (NRC) to f

J. D. Shiffer (PG8E), June 14,1985. " Issuance of Notices of Considera-i tion of Issuance of Amendments."

22.

PG8E letter DCL-85-244, July 18,1985. "Diablo Canyon Units 1 and 2 Cer-tification of Diablo Canyon Units 1 and 2 Full Power Technical Specifica-tions."

[

23.

U.S. Nuclear Regulatory Commission Letter from G.W. Knighton (NRC) to J.D.Shiffer(PG&E), July 27, 1985, " Technical Specification for Cranes."

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