Text

License Amendment Request 07-04, Proposed Technical Specifications Change to Relocate Surveillance Test Intervals to a Licensee-Controlled Program (Risk Informed Technical Specifications Initiative 5b)
	ML072950183
Person / Time
Site:	Diablo Canyon
Issue date:	10/15/2007
From:	Becker J; Pacific Gas & Electric Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	DCL-07-097
	Download: ML072950183 (292)
	v • d • e

Pacific Gas and Electric Company' James R.Becker Diablo Canyon Power Plant Vice President P.0. Box 56 Diablo Canyon Operations and Avila Beach, CA93424 Station Director October 15, 2007 805.545.3462 PG&E Letter DCL-07-097 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Docket No. 50-275, OL-DPR-80 Docket No. 50-323, OL-DPR-82 Diablo Canyon Units 1 and 2 License Amendment Request 07-04 Proo~osed Technical Specifications Change to Relocate Surveillance Test Intervals to a Licensee-Controlled Program (Risk Informed Technical Specifications Initiative 5b)

References:

1. Technical Specification Task Force (TSTF) Traveler number TSTF-425, Revision 1, "Relocate Surveillance Frequencies to Licensee Control - RITSTF Initiative 5," dated April 2007.

2. NEI 04-10, Revision 1, "Risk-Informed Technical Specifications Initiative 5b, Risk-informed Method for Control of Surveillance Frequencies," dated April 2007.

Dear Commissioners and Staff:

In accordance with 10 CFR 50.90, enclosed is an application for amendment to Facility Operating License Nos. DPR-80 and DPR-82 for Units 1 and 2 of the Diablo Canyon Power Plant (DCPP) respectively.

Pacific Gas and Electric Company (PG&E) proposes to relocate all periodic surveillance frequencies from the technical specifications (TS) and place the frequencies under licensee control in accordance with a new program, the Surveillance Frequency Control Program.

The enclosed license amendment request (LAR) is being submitted as a pilot submittal in support of Risk Informed TS Initiative Sb, "Relocate Surveillance Test Intervals to Licensee Control." On September 28, 2006, the NRC approved the lead plant submittal for Initiative Sb for Limerick Generating Station. On that same date, the NRC provided the final Safety Evaluation for the methodology document, NE104-10, Revision 0.

A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway e Comanche Peak

Diablo Canyon e Palo Verde o South Texas Project

Wolf Creek j

Document Control Desk PG&E Letter DCL-07-097 October 15, 2007 Page 2 The NRC's approval of NEI 04-10 stated that it was only applicable to boiling water reactors and that it did not approve relocation of requirements to perform surveillances on a staggered test basis. Revision 1 of TSTF-425 (Reference 1) and NEI 04-10 (Reference 2) address the above two restrictions and support the changes proposed in this LAR. This LAR uses the methodology in Reference 2, and requests similar changes to those proposed in Reference 1, currently under review by the NRC.

By letter dated October 2, 2007, "Request for Additional Information (RAI) Regarding TSTF Traveler 425, Revision 1, 'Relocate Surveillance Frequencies to Licensee Control - RITSTF Initiative 5,' PROJ 0753, (TAG No. MD571 1)," the NRC requested additional information from the TSTF regarding Reference 1. This LAR does not address these RAls, some of which are unlikely to apply to this application or may be resolved generically through the TSTF.

This LAR proposes a change to several TS pages that are expected to be revised prior to approval of this LAR (due to other in-process LARs). Because of this, retyped TS pages are not included in this submittal. PG&E will provide fresh marked up and retyped TS pages prior to approval of this LAR. contains a description of the proposed changes, the supporting technical analyses, and the no significant hazards consideration determination. contains marked-up TS pages. Enclosure 3 contains marked-up TS Bases pages. TS Bases changes are provided for information only and will be implemented pursuant to TS 5.5.14, "Technical Specifications Bases Control Program," at the time this amendment is implemented.

Individual surveillance test intervals (STIs) will not be revised as part of this LAR; however, a summary of the results of example STI evaluations using the risk informed methodology as detailed in Reference 2 are provided in Enclosure 4. Once this LAR and the revised risk-informed process and methodology are approved by the NRC, future changes to the STIs will be evaluated in accordance with the licensee-controlled program. Enclosure 5 discusses the technical adequacy and scope of DCPP's probabilistic risk assessment (PRA) Model.

PG&E has determined that this LAR does not involve a significant hazard consideration as determined per 10 CFR 50.92(c). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs to be prepared in connection with the issuance of this amendment.

The changes in this LAR are not required to address an immediate safety concern.

PG&E requests approval of this LAR no later than October 2008. PG&E requests the license amendment(s) be made effective upon NRC issuance, to be implemented within 180 days from the date of issuance. An implementation period A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway

Comanche Peak

Diablo Canyon 9 Palo Verde

South Texas Project *Wolf Creek

Document Control Desk PG&E Letter DCL-07-097 October 15, 2007 Page 3 of 180 days would provide a reasonable amount of time for implementation of TS and TS bases changes, training, and new procedures necessary to implement the surveillance frequency control program.

This communication contains new commitments to be implemented in support of this LAR. The commitments are contained in Enclosure 6.

If you have any questions or require additional information, please contact Stan Ketelsen at 805-545-4720.

1state under penalty of perjury that the foregoing is true and correct.

Executed on October 15, 2007.

Sincerely, James 1R. ecker Vice President - Diablo Canyon Operations and Station Director mjrm/4557 A069431 5 cc: Gary W. Butner, DPH Elmo E. Collins, NRC Region IV Michael S. Peck, DCPP NRC Senior Resident Inspector Diablo Distribution cc/enc: Alan B. Wang, NRC Project Manager A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway

Comanche Peak

Diablo Canyon

Palo Verde

South Texas Project e Wolf Creek

Enclosure 1 PG&E Letter DCL-07-097 EVALUATION 1.0

SUMMARY

DESCRIPTION This letter is a request to amend the Facility Operating Licenses DPR-80 and DPR-82 for Units 1 and 2 of the Diablo Canyon Power Plant (DCPP),

respectively.

The proposed amendment would implement changes consistent with Technical Specification Task Force (TSTF) Traveler number TSTF-425, Revision 1, "Relocate Surveillance Frequencies to Licensee Control - RITSTF Initiative 5,"

dated April 2007, which relocates the surveillance frequencies, also known as surveillance test intervals (STIs), of various Technical Specification (TS) surveillance requirements (SRs) from the TSs to a licensee-controlled document.

TSTF-425 is currently under review by the NRC.

The relocated STIs would be controlled in accordance with the requirements stipulated in a new program, the Surveillance Frequency Control Program (SFCP), which is being added to the Administrative Controls Section of the TSs.

Enclosure 2 provides the marked up TS pages. Enclosure 3 provides the marked-up TS Bases pages for information only. Enclosure 4 provides sample STI evaluation forms. Enclosure 5 discusses the technical adequacy and scope of DCPP's probabilistic risk assessment (PRA) Model.

NEI has developed a risk-informed methodology, documented in NEI 04-10, Revision 1, "Risk-Informed Technical Specifications Initiative 5b, Risk-informed Method for Control of Surveillance Frequencies," dated April 2007, which provides a method to evaluate and revise STIs, where appropriate, within the SFCP. This methodology document will be referenced in the SFCP and will be incorporated by reference into the Administrative Controls Section of the TSs.

Individual STIs will not be revised as part of this license amendment request (LAR). However, once this LAR is approved by the NRC, future changes to the STIs will be evaluated in accordance with the licensee-controlled program,-and

.the STIs may be revised, as appropriate, based on the evaluation results without prior NRC approval. Examples of STI evaluations using the ris,k-informed process and methodology of NEI 04-10, Revision 1, are provided in Enclosure 4.

Various TS surveillance requirements, including, in some cases, their associated STIs, were established based on commitments to Regulatory Guides, or based on implementation of NRC-approved Licensing Topical Reports. The SRs themselves will not be relocated to the SFCP and will continue to be performed in accordance with the applicable Regulatory Guide or Topical Report, as appropriate; however, the associated STIs relocated to the SFCP may be modified in accordance with the SFCP. Where the associated STIs were established based on commitments documented in the plant's safety analysis, 1

Enclosure 1 PG&E Letter DCL-07-097 those commitments would be subject to review by an independent decision-making panel .(IDP) using the guidance of NEI 99-04, "Commitment Control," and could potentially be changed by the licensee-controlled program without prior NRC approval.

2.0 DETAILED DESCRIPTION This proposed change will result in the following:

1. Revised Index, as appropriate, to reflect the TS changes proposed below.

2. Replace various STIs specified within individual SRs with a reference to the licensee-controlled program, e.g., with the words, "in accordance with the Surveillance Frequency Control Program." The proposed change applies primarily to surveillances that are performed on a fixed periodicity. Existing STIs that are not proposed to be relocated as part of this request are associated with SRs that:

a. Have no time component but are purely event driven, e.g., "prior to thermal power exceeding 75% RTP"; or

b. Are event-driven but have a time component for performing the surveillance on a one time basis once the event occurs, e.g., "within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after thermal power reaching > 95% RTP;" or

c. Reference an already established and approved licensee program rather than a specific interval, e.g., SRs that refer to the Inservice Testing Program.

3. Add a new TS Section 5.5.18, "Surveillance Frequency Control Program,"

which will describe the basic means for licensee control of surveillance frequencies within the licensee-controlled program and includes the following requirements:

a. The Surveillance Frequency Control Program shall contain a list of Frequencies of those SRs for which the Frequency is controlled by the program,

b. Changes to 'the Frequencies listed in the SFCP shall be made in accordance with NEI-04-10, and

c. The provisions of SR 3.0.2 and 3.0.3 are applicable to the Frequencies established in the SFCP.

4. Revise the TS Bases, as appropriate, relative to the proposed TS changes described above. The TS Bases changes are included for information only.

2

Enclosure 1 PG&E Letter DCL-07-097

5. Remove a note from TS 3.3.5.2, SR 3.3.5.2 under Frequency that is no longer applicable.

3.0 TECHNICAL EVALUATION

The changes requested by this amendment are programmatically the same as those granted by the NRC for the Limerick Generating Station on September 28, 2006. These changes allowed the relocation of surveillance frequencies, also known as STIs, of various TS SRs from the TSs to a licensee-controlled document, which would be controlled in accordance with the requirements stipulated in a new program, the SFCP in the Administrative Controls Section of the TSs. Since the relocation of the STIs from the TS to a licensee-controlled program does not affect the plant design, hardware, or system operation and will not affect the ability of the plant to perform its design function in mitigating the consequences of a postulated design basis accident, the proposed TS changes are administrative in nature. The new program, SFCP, will control any changes to the relocated STIs and ensure the SRs specified in TS are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met. The program contains the following requirements:

" The SFCP shall contain a list of Frequencies of those SRs for which the Frequency is controlled by the program.

" Changes to the Frequencies listed in the SFCP shall be made in accordance with NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequencies," Revision 1.

" The provisions of Surveillance Requirements 3.0.2 and 3.0.3 are applicable to the Frequencies established in the SFCP.

NEI 04-10 provides the detailed process requirements for controlling surveillance frequencies of the TS SRs that have been relocated from the TSs to the SFCP.

The methodology described in NEI 04-10 provides a risk-informed process to support a plant expert panel assessment of proposed changes to surveillance frequencies, assuring appropriate consideration of risk insights and other deterministic factors, which may impact surveillance frequencies, along with appropriate performance monitoring of changes and documentation requirements.

The NRC issued a Final Safety Evaluation for NEI 04-10 Revision 0, "Risk-Informed Method for Control of Surveillance Frequencies," on September 28, 2006. The Staff found that NEI 04-10, Revision 0, was acceptable for referencing by licensees proposing to amend their TSs to establish an SFCP, provided that the following conditions are satisfied:

3

Enclosure 1 PG&E Letter DCL-07-097

1. The licensee submits documentation with regard to PRA technical adequacy consistent with the requirements of RG 1.200, Section 4.2.

2. When a licensee proposes to use PRA models for which NRC-endorsed standards do not exist, the licensee submits documentation, which identifies the quality characteristics of those models, consistent with RG 1.200, Sections 1.2 and 1.3. Otherwise, the licensee identifies and justifies the methods to be applied for assessing the risk contribution for those sources of risk not addressed by PRA models.

PG&E has performed sample STI evaluations to verify the NEI 04-10 process can be properly implemented. A discussion of the process and the STI evaluation forms are provided in Enclosure 4.

A discussion of the PRA quality and scope is provided in Enclosure 5.

4.0 REGULATORY ANALYSIS

4.1 Applicable Regulatory Requirements/Criteria Section 182a of the Atomic Energy Act of 1954, as amended (the Act) requires applicants for nuclear power plant operating licenses to include the TS as part of the license. The Commission's regulatory requirements related to the content for the TS are set forth in 10 CFR 50.36. That regulation requires that the TS include items in eight specific categories.

The categories are: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation; (3) SRs; (4) design features; (5) administrative controls; (6) decommissioning; (7) initial notification; and (8) written reports. However, the regulation does not specify the particular requirements to be included in a plant's TS.

The proposed change relocates the intervals for the performance of various SRs from TS to a licensee-controlled program using an NRC approved methodology for control of the surveillance intervals once in the licensee program. The SRs themselves will remain in TS. The In-Service Test Program (IST Program) for ASME components governed by TS 5.5.8 established a precedent for relocation of STIs to a licensee-controlled program. In several instances, the DCPP TS do not specify a particular surveillance interval but rather state: "In accordance with the Inservice Testing Program."

TS Section 5.5.8 references the ASME Operations and Maintenance (OM)

Code for inservice testing intervals. The testing intervals are based on the plant's IST Program which implements the ASME OM Code. Within the 4

Enclosure 1 PG&E Letter DCL-07-097 IST program, the actual testing intervals vary based on the performance of the individual components.

The proposed TS changes are administrative in nature. Relocation of the STIs from the TS to a licensee-controlled program does not affect the plant design, hardware, or system operation and will not affect the ability of the plant to perform its design function in mitigating the consequences of a postulated design basis accident. The program for changing the intervals establishes criteria and requirements for surveillance test frequencies adequate to demonstrate operability of the TS components.

Therefore, the proposed change does not adversely affect nuclear safety or plant operations.

4.2 Precedent On September 28, 2006, a similar change, Limerick Generation Station, -

Relocate Surveillance Test Intervals to Licensee-Controlled Program (TAO NOs. MC3567 AND MC3568) was approved. The program proposed by PG&E is administratively and technically identical to the program approved for Limerick with the following exceptions:

1. The DCPP submittal is based on Revision 1 of TSTF-425

2. The DCPP submittal references Revision 1 of NEI 04-10.

3. Limerick and DCPP are different reactor types, which is not relevant to the requested change.

4.3 Significant Hazards Consideration PG&E has evaluated whether or not a significant hazards consideration is

.involved with the proposed amendments by focusing on the three standards set forth in 10 CER 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed change involves the relocation of various surveillance test intervals from TSs to a licensee-controlled program and is administrative in nature. The proposed change does not involve the modification of any plant equipment or affect basic plant operation. The proposed change will have no impact on any safety related structures, systems or components.

Surveillance test intervals are not assumed to be an initiator of any 5

Enclosure 1 PG&E Letter DCL-07-097 analyzed event, nor are they assumed in the mitigation of consequences of accidents. The SRs themselves will be maintained in the TS along with the applicable Limiting Conditions for Operation (LCOs) and Action statements. The surveillances performed at the intervals specified in the licensee-controlled program will assure that the affected system or component function is maintained,!that the facility operation is within the Safety Limits, and that the LOOs are met.

Therefore, the proposed change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed change does not involve any physical alteration of plant equipment and does not change the method by which any safety-related structure, system, or component performs its function or is tested. As such, no new or different types of equipment will be installed, and the basic operation of installed equipment is unchanged. The methods governing plant operation and testing remain consistent with current safety analysis assumptions.

Therefore, the proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The proposed change is administrative in nature, does not negate any existing requirement, and does not adversely affect existing plant safety margins or the reliability of the equipment assumed to operate in the safety analysis. As such, there are no changes being made to safety analysis assumptions, safety limits or safety system settings that would adversely affect plant safety as a result of the proposed change. Margins of safety are unaffected by relocation of the surveillance test intervals to a licensee-controlled program.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, PG&E concludes that the proposed amendment presents no significant hazards consideration under the standards set 6

Enclosure 1 PG&E Letter DCL-07-097 forth in 10 CFR 50.92(c), and accordingly, a finding of "no significant hazards consideration" is justified.

4.4 Conclusions Based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

PG&E has evaluated the proposed amendment and has determined that the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51 .22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

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Enclosure 1 PG&E Letter DCL-07-097

6.0 REFERENCES

1) Letter from R. V. Guzman, Office of Nuclear Reactor Regulation, to C. M.

Crane, Exelon Nuclear, dated September 28, 2006 (ML062420047) (TAC NOS. MC3567 and MC3568).

2) Letter from H. Neih, Office of Nuclear Reactor Regulation, to A.

Pietrangelo, Nuclear Energy Institute, dated September 28, 2006 (ML062700012) (TAC NOS. MB2531 and MD3077).

3) NEI 04-10, Revision 0, dated July 2006.

4) NEI 04-10, Revision 1, "Risk-Informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies," dated April 2007.

5) Technical Specification Task Force Traveler number TSTF-425, Revision 1, "Relocate Surveillance Frequencies to Licensee Control - RITSTF Initiative 5," dated April 2007.

6) PG&E Letter DCL-06-141, "Request to be Considered PWR Pilot Plant for

-Risk-Informed Initiative 5b, Relocate Surveillance Frequencies to Licensee Control," dated December 28, 2006.

7) RG 1.200, Revision 1, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk- Informed Activities," dated January 2007.

8) American Society of Mechanical Engineers (ASME) RA 7Sa-2003, Addenda to ASME RA-S-2002, "Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications," dated December 5, 2003.

9) NEI 00-02, Revision A3, "PRA Peer Review Process Guidance," dated March 20, 2000.

8

Enclosure 2 PG&E Letter DCL-07-.097 Proposed Technical Specification Changes (marked-up) 1

Enclosure 2 PG&E Letter DCL-07-097 Insert 1 In accordance with the Surveillance Frequency Control Program Insert 2 5.5.18 Surveillance Frequency Control Progiram This program provides controls for Surveillance Frequencies. The program shall ensure that Surveillance Requirements specified in the Technical Specifications are performed at intervals sufficient to assure the associated Limiting Conditions for Operation are met.

a. The Surveillance Frequency Control Program shall contain a list of Frequencies of those Surveillance Requirements for which the Frequency

-is controlled by the program.

b. Changes to the Frequencies listed in the Surveillance Frequency Control Program shall be made in accordance with NEI 04-10, "Risk-Informed Method for Control of Surveillance Frequencies," Revision 1.

c. The provisions of Surveillance Requirements 3.0.2 and 3.0.3 are applicable to the Frequencies established in the Surveillance Frequency Control Program.

2

Definitions 1.1 1.1 Definitions (continued)

SLAVE RELAY TEST A SLAVE RELAY TEST shall consist of energizing all slave relays and verifying the OPERABILITY of each required slave relay. The SLAVE RELAY TEST shall include a continuity check of associated required testable actuation devices. The SLAVE RELAY TEST may be performed by means of any series of sequential, overlapping, or total steps.

STAGGERED TE BASIS the testing of A STAGG RED TEST BASIS 5hRconsist of one ersystems, subsysins, channels, or te Survillnceciomponei ignated during the inte pecified by the re cyso that a~ll ems sbystems,u channels, or er designated ponents are tested during Z 11uvel ce Frequency i vals, where n is the total numb pof systems, sub=stms channels, or other designated components in the associated function.

THERMAL POWER THERMAL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

TRIP ACTUATING DEVICE- A TADOT shall consist of operating the trip actuating device OPERATIONAL TEST and verifying the OPERABILITY of all devices in the channel (TADOT) required for trip actuating device OPERABILITY. The TADOT shall include adjustment, as necessary, of the trip actuating device so that it actuates at the required setpoint within the necessary accuracy. The TADOT may be performed by means of any series of sequential, overlapping or total channel steps.

DIABLO CANYON - UNITS 1 & 2 1.1-6 Unit 1-Amendment No. 13-8S9IDD07.DOA -R7 7 Unit 2 - Amendment No.4-.

SDM 3.1.1 3.1 REACTIVITY CONTROL SYSTEMS 3.1.1 SHUTDOWN MARGIN (SDM)

LCO 3.1.1 SDM shall be within the limits provided in the COLR.

APPLICABILITY: MODE 2 with keff < 1.0, MODES 3, 4, and 5.

ACTIONS _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

CONDITION REQUIRED ACTION COMPLETION TIME A. SDM not within limit. A.1. Initiate boration to 15 minutes restore SDM to within limit.

SURVEILLANCE REQUIREMENTS ________

SURVEILLANCE FREQUENCY SR 3.1.1.1 Verify SDM to be within limits.

DIABLO CANYON - UNITS 1 & 2 3.1-1 Unit 1 - Amendment No. 445,469, 8S91 DG05. DOA - R5 1 Unit 2 - Amendment No. 43~5,+715

Core Reactivity 3.1.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

+

SR 3.1.2.1 --------------------------

NOTE----------------

The predicted reactivity values may be adjusted (normalized) to correspond to the measured core reactivity prior to exceeding a fuel burnup of 60 effective full power days (EFPD) after each fuel loading.

Verify measured core reactivity is within +/- 1% Ak/k of Once prior to predicted values. entering MODE 1 after each refueling AND NOTE----- -

Only required after 60 EFPD I

I DIABLO CANYON - UNITS 1 & 2 3.1-3 Unit 1 - Amendment No.+--*&,

8S91 DG05. DOA - R5 3 Unit 2 - Amendment No. +357;

Rod Group Alignment Limits 3.1.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.4.1 Verify individual rod positions within alignment limit.

SR 3.1.4.2 Verify rod freedom of movement (trippability) by 7 n-,-r 4'1L moving each rod not fully inserted in the core Ž! 10 steps in either direction.

SR 3.1.4.3 Verify rod drop time of each rod, from the fully Prior to reactor withdrawn position, is <5:2.7 seconds from the criticality after each beginning of decay of stationary gripper coil removal of the reactor voltage to dashpot entry, with: head

a. Tavg Ž: 500 OF; and

b. All reactor coolant pumps operating.

DIABLO CANYON - UNITS 1 & 2 3.1-9 Unit 1 - Amendment No. I-M, 8S91DG05.DOA -R5 9 Unit 2 -Amendment No. +37S5

Shutdown Bank Insertion Limits 3.1.5 3.1 REACTIVITY CONTROL SYSTEMS 3.1.5 Shutdown Bank Insertion Limits LCO 3.1.5 Each shutdown bank shall be within insertion limits specified in the COLR.

APPLICABILITY: MODE 1, MODE 2 with any control bank not fully inserted.

NOTE----

This LCO is not applicable while performing SR 3.1.4.2.

ACTIONS ____________________

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more shutdown A. 1.1 Verify SDM to be within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months banks not within limits, the limits provided in the COLR.

OR A.1.2 Initiate boration to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months restore SDM to within limit.

AND A.2 Restore shutdown banks 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to within limits.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met.II SURVEILLANCE REQUIREMENTS ________

SURVEILLANCE FREQUENCY SR 3.1.5.1 Verify each shutdown bank is within the limits Z specified in the COLR.

DIABLO CANYON - UNITS 1 & 2 3.1-10 Unit 1 - Amendment No. 1113, 8S9IDG05.DOA -R5 10 Unit 2 - Amendment No. 44t

Control Bank Insertion Limits 3.1.6 SURVEILLANCE REQUIREMENTS _______

SURVEILLANCE FREQUENCY SR 3.1.6.1 Verify estimated critical control bank position is within Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months the limits specified in the COLR. prior to achieving criticality SR 3.1.6.2 Verify each control bank insertion is within the limits specified in the COLR.

SR 3.1.6.3 Verify sequence and overlap limits specified in the COLR are met for control banks not fully withdrawn from the core. I -

DIABLO CANYON - UNITS 1 & 2 3.1-12 Unit 1 - Amendment No. 't15' 8S91DG05. DOA - R5 12 Unit 2 -Amendment No. t36"

PHYSICS TESTS Exceptions - MODE 2 3.1.8 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.8.1 Perform a CHANNEL OPERATIONAL TEST on Prior to initiation of power range and intermediate range channels per PHYSICS TESTS SR 3.3.1.7, SR 3.3.1.8, and Table 3.3.1-1.

SR 3.1.8.2 Verify the RCS lowest operating loop average0 temperature is Ž: 5310 F.

SR 3.1.8.3 Verify THERMAL POWER is!*5% RTP.

SR 3.1.8.4 Verify SDM is within the limits provided in the COLR.

-T-- -Vk -I DIABLO CANYON - UNITS 1 & 2 3.1-17 Unit 1 - Amendment No..4ý,5 8S91 DG05. DOA - R5 17 Unit 2 - Amendment No. +35,-

FQ(Z) 3.2.1 SURVEILLANCE REQUIREMENTS

NOTE------------------------------------------------

During power escalation following shutdown, THERMAL POWER may be increased until an equilibrium power level has been achieved, at which a power distribution'map is obtained.

SURVEILLANCE FREQUENCY SR 3.2.1.1 Verify Fc (Z) is within limit. Once after each refueling prior to THERMAL POWER exceeding 75%

RTP AND Once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after achieving equilibrium conditions after exceeding, by Ž 20% RTP, the THERMAL POWER at which IFc(Z) was last verified AND 3=D If (continued)

DIABLO CANYON - UNITS 1 & 2 8S9IDH02.DOA - 2 3.2-2 Unit 1 - Amendment No. +95.-

R2 Unit 2 - Amendment No. M,

FQ(Z) 3.2.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.1.2 (continued) Once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after achieving equilibrium conditions after exceeding, by Ž 20% RTP, the THERMAL POWER at which F w(Z) was last verified AND I

DIABLO CANYON - UNITS 1 & 2 8S9IDH02.DOA - 4 3.2-4 Unit 1 - Amendment No.15 R2 Unit 2 -Amendment No. +35-

FAH 3.2.2 SURVEILLANCE REQUIREMENTS

NOTE ----------------------------------------------

During power escalation following shutdown, THERMAL POWER may be increased until an equilibrium power level has been achieved, at which a power distribution map is obtained.

SURVEILLANCE FREQUENCY SR 3.2.2.1 Verify FN is within limits specified in the COLR. Once after each AH refueling prior to THERMAL POWER exceeding 75%

RTP AND DIABLO CANYON - UNITS 1 & 2 8S91DH02.DOA - 7 3.2-7 Unit 1 - Amendment No. +96,-

R2 Unit 2 - Amendment No. +3~&T-

AFD 3.2.3 3.2 POWER DISTRIBUTION LIMITS 3.2.3 AXIAL FLUX DIFFERENCE (AFD)

LCO 3.2.3 The AFD in % flux difference units shall be maintained within the limits specified in the COLR.

L irvr1 -----------------------------------------------

The AFD shall be considered outside limits when two or more OPERABLE excore channels indicate AFD to be outside limits.

APPLICABILITY: MODE 1 with THERMAL POWER Ž!50% RTP.

ACTIONS _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

CONDITION REQUIRED ACTION COMPLETION TIME A. AFD not within limits.

______________________POWER A.1 Reduce THERMAL to < 50% RTP.

j30 minutes SURVEILLANCE REQUIREMENTS ______

SURVEILLANCE FREQUENCY SR 3.2.3.1 Verify AED within limits for each OPERABLE excore channel.

DIABLO CANYON - UNITS 1 & 2 8S9lDH02.DOA - 8 3.2-8 Unit 1 - Amendment No.44&,-

R2 Unit 2 - Amendment No. +Y5-,

QPTR 3.2.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.4.1-- ------------------------- NOTES-----------------------

1. With input from one Power Range Neutron Flux channel inoperable and THERMAL POWER *ý 75% RTP, the remaining three power range channels can be used for calculating QPTR.

2. SR 3.2.4.2 may be performed in lieu of this Ttve4w i L Surveillance.

Verify QPTR is within limit by calculation.s SR 3.2.4.2---------------------------- NOTE -----------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after the input from one or more Power Range Neutron Flux channels is inoperable with THERMAL POWER

> 75% RTP.

Verify QPTR is within limit using core power distribution measurement information.

DIABLO CANYON - UNITS 1 & 2 8S9IDH02.DOA - 11 3.2-1 1 Unit 1 - Amendment No. 4-35,464-r R2 Unit 2 -Amendment No. 445,+66-

RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS

NOTE---------------------------------------------

Refer to Table 3.3.1-1 to determine which SRs apply for each RTS Function.

SURVEILLANCE FREQUENCY SR 3.3.1.1 Perform CHANNEL CHECK.

SR 3.3.1.2- ------------------------- NOTE --------------------------

Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after THERMAL POWER is Ž: 15% RTP, but prior to exceeding 30% RTP.

Compare results of calorimetric heat balance calculation to power range channel output. Adjust power range channel output if calorimetric heat ~A balance calculation results exceed power range channel output by more than + 2% RTP.

SR 3.3.1.3- ------------------------- NOTE -------------------------

Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after THERMAL POWER is 2:50% RTP. £

Compare results of incore power distribution [31 etf~ctive full measurements to Nuclear Instrumentation System pD< r (NIS) AED. Adjust NIS channel if absolute difference (EFP~

is 2: 3%.

SR 3.3.1.4 ------------------------- NOTE -------------------------

This Surveillance must be performed on the reactor trip bypass breaker, for the local manual shunt trip only, prior to placing the bypass breaker in service.

Perform TADOT. 6 so-n SR 3.3.1.5 Perform ACTUATION LOGIC TEST. ao JT (continued)

DIABLO CANYON - UNITS 1 & 2 3.3-8 Unit 1 - Amendment No. 1-35,57,4664, +7-9§,

8S9IDI14.DOA -R14 10 Unit 2 - Amendment No. 135,-157,4-66, +8+,

RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.6 NOTE--------------------------

Not required to be performed until 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after THERMAL POWER Ž!75% RTP.

19-Calibrate excore channels to agree with incore power distribution measurements.

SR 3.3.1.7 NOTE--------------------------

1. Not required to be performed for source range instrumentation prior to entering MODE 3 from MODE 2 until 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after entry into MODE 3.

2. For source range instrumentation, this Surveillance shall include verification that interlocks P-6 and P-10 are in their required state for existing unit conditions.

Perform COT.

(ýw (continued)

P DIABLO CANYON - UNITS 1 & 2 3.3-9 Unit 1 - Amendment No. 4-35,4-64,4-7-9*,64, 8S9IDI14.DOA-R14 11 Unit 2 - Amendment No. 4-M,4-69,4-84,+-86,-

RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.8 NOTE--------------------------

NOTE----- -

This Surveillance shall include verification that Only required interlocks P-6 and P-10 are in their required state for when not existing unit conditions. performed within previous 184 days Perform COT. Prior to reactor startup AND 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reducing power below P-10 for power and intermediate instrumentation AND Four hours after reducing power below P-6 for source range instrumentation AND vejr8 a SR 3.3.1.9 NOTE---------

Verification of setpoint is not required.

Perform TADOT. (

SR 3.3.1.10-------------------------- NOTE--------------------------

This Surveillance shall include verification that the time constants are adjusted to the prescribed values.

Perform CHANNEL CALIBRATION.

(continued)

DIABLO CANYON - UNITS 1 & 2 3.3-10 Unit 1 - Amendment No. 4-35,+79-7fr 8S91DI14.DOA -R14 12 Unit 2 -Amendment No. 4-35, 44&l

RTS Instrumentation 3.3.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.11 ------------------------ NOTE --------------------------

1. Neutron detectors are excluded from CHANNEL CALIBRATION.

2. This Surveillance shall include verification that the time constants are adjusted to the prescribed values.

3. Power and Intermediate Range detector plateau voltage verification is not required to be performed until 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after achieving equilibrium Conditions with Thermal Power Ž! 95% ~w~wk 4 RTP.

Perform CHANNEL CALIBRATION.

SR 3.3.1.12 SR 3.3.1.13 Perform CHANNEL CALIBRATION.

Perform COT. _____ 4 9 SR 3.3.1.14 ------------------------- NOTE -------------------------

Verification of setpoint is not required. A-.4.

Perform TADOT.

SR 3.3.1.15 ------------------------- NOTE -------------------------- Prior to exceeding Verification of setpoint is not required. the Pr-9 interlock

whenever the unit has been in MODE 3, if not performed in th6 previous 31 days.

Perform TADOT _______

SR 3.3.1.16 ------------------------- NOTE -------------------------

Neutron detectors are excluded from response time testing. 2Y~~~

Verify RTS RESPONSE TIMES are within limits. on a LFB2 t DIABLO CANYON - UNITS 1 & 2 3.3-11 Unit 1 - Amendment No. t1357-8S9IDI14.DOA -R14 13 Unit 2 -Amendment No. 45.

ESFAS Instrumentation 3.3.2 SURVEILLANCE REQUIREMENTS

NOTE --------------------

Refer to Table 3.3.2-1 to determine which SRs apply for each ESFAS Function.

SURVEILLANCE FREQUENCY SR 3.3.2.1 Perform CHANNEL CHECK. v -k£ SR 3.3.2.2 Perform ACTUATION LOGIC TEST. f2/a s on

~T S-SR 3.3.2.3 Not used.

SR 3.3.2.4 Perform MASTER RELAY TEST. ~~n 5 EE§I[-BAS18 SR 3.3.2.5 Perform COT. ~ ~4~

SR 3.3.2.6 Perform SLAVE RELAY TEST. ~c SR 3.3.2.7 Not used.

SR 3.3.2.8- ---------------------- NOTE----------------------------

Verification of setpoint not required for manual initiation functions.

-- --- --- -- --- - Perform TADOT.

SR 3.3.2.9------------------------NOTE----------------------------

This Surveillance shall include verification that the time constants are adjusted to the prescribed values.

Perform CHANNEL CALIBRATION.

SR 3.3.2.10----------------------- NOTE----------------------------

Not required to be performed for the turbine driven AFW pump until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after SG pressure is Žý 650 4 psig.

Verify ESF RESPONSE TIMES are within limits. 4 -monhs on E§T BqASf U (continued)

DIABLO CANYON - UNITS 1 & 2 3.3-25 Unit 1 - Amendment No. -3~5, 44G 8S9IDI14.DOA -R14 29 Unit 2 - Amendment No. 4-35, 1484-

.1

ESFAS Instrumentation 3.3.2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.2.11----------------------- NOTE----------------------------

Verification of setpoint not required.

Perform TADOT.

SR 3.3.2.12 Perform ACTUATION LOGIC TEST SR 3.3.2.13-----------------------NOTE----------------------------

Verification of setpoint not required for manual initiation functions.

Perform TADOT DIABLO CANYON - UNITS 1 & 2 3.3-26 Unit 1 - Amendment No. +35-;

8S9lDI14.DOA -R14 30 Unit 2 -Amendment No. -.

PAM Instrumentation 3.3.3 SURVEILLANCE REQUIREMENTS

N r--------------------------------------------------------------

SR 3.3.3.1 and SR 3.3.3.2 apply to each PAM instrumentation Function in Table 3.3.3-1.

SURVEILLANCE FREQUENCY SR 3.3.3.1 Perform CHANNEL CHECK for each required instrumentation channel that is normally energized. T 4:-

SR 3.3.3.2 ------------------------ NOTE --------------------------

Neutron detectors are excluded from CHANNEL CALIBRATION. -K Perform CHANNEL CALIBRATION.

DIABLO CANYON - UNITS 1 & 2 3.3-36 Unit 1 -Amendment No. 45,442-, &ý 8S9IDI14.DOA -R14 40 Unit 2 - Amendment No. 4-35,44-2,46,t

Remote Shutdown System 3.3.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.3.4.1 Perform CHANNEL CHECK for each required instrumentation channel.

SR 3.3.4.2 Verify each required control circuit and transfer switch 4-is capable of performing the intended function.

i SR 3.3.4.3 ------------------------- NOTE ------------------------

Reactor Trip Breaker position is excluded from CHANNEL CALIBRATION.

-- ýj Perform CHANNEL CALIBRATION for each required instrumentation channel.

DIABLO CANYON - UNITS 1 & 2 3.3-39 Unit 1 - Amendment No. +35-,

8S9IDI14.DOA -R14 43 Unit 2 - Amendment No. +t

LOP DG Start Instrumentation 3.3.5 3.3 INSTRUMENTATION 3.3.5 Loss of Power (LOP) Diesel Generator (DG) Start Instrumentation LCO 3.3.5 One channel per bus of loss of voltage DG start Function; and two channels per bus of degraded voltage Function shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4, When associated DG is required to be OPERABLE by LCO 3.8.2, "AC Sources-Shutdown."

ACTIONS NOTE------------------------------------------------

Separate Condition entry is allowed for each Function.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more Functions with A.1---------- NOTE ------------

one or more channels per One channel may be bus inoperable, bypassed for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing.

Enter applicable Immediately Condition(s) and Required Action(s) for the associated DG made inoperable by LOP DG start instrumentation.

SURVEILLANCE REQUIREMENTS SURVEILLANCE SR 3.3.5.1 Not used SR 3.3.5.2 Perform TADOT.

(continued)

I" DIABLO CANYON - UNITS 1 & 2 3.3-4 1 Unit 1 - Amendment No. 135,-1-42,465&,

8S9IDI14.DOA - R14 45 Unit 2 - Amendment No. 4.35, 44 LOP DG Start Instrumentation 3.3.5 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY I. __________

SR 3.3.5.3 Perform CHANNEL CALIBRATION with Allowable Value setpoints as follows:

a. Loss of voltage Diesel Start Allowable Value 3v¶AA'(k.

>0 V with a time delay of *ý 0.8 seconds and

>2583 V with a *ý 10 second time delay.

Loss of voltage initiation of load shed with one relay Allowable Value Ž! 0 V with a time delay of

<4 seconds and Ž: 2583 V with a time delay

<25 seconds and with one relay Allowable Value

>2870 V, instantaneous.

b. Degraded voltage Diesel Start Allowable Value

>! 3785 V with a time delay of *! 10 seconds.

Degraded voltage initiation of Load Shed Allowable Value Ž: 3785 V with a time delay of

<!ý20 seconds.

DIABLO CANYON - UNITS 1 & 2 3.3-42 Unit 1 - Amendment No. 135, 142, 't6&6 8S91DI14.DOA -R14 46 Unit 2 - Amendment No. 446, 14%,

Containment Ventilation Isolation Instrumentation 3.3.6 SURVEILLANCE REQUIREMENTS

NOTE -------------------------------------------------

Refer to Table 3.3.6-1 to determine which SRs apply for each Containment Ventilation Isolation Function.

SURVEILLANCE FREQUENCY SR 3.3.6.1 Perform CHANNEL CHECK. i SR 3.3.6.2 ----------------------- NOTE------------------------------ days n This surveillance is only applicable to the actuation STA G~E logic of the ESFAS Instrumentation. ETB SIS Perform ACTUATION LOGIC TEST.

SR 3.3.6.3 ----------------------- NOTE------------------------------ day n This surveillance is only applicable to the master T G~

relays of the ESFAS Instrumentation. EES:T e Perform MASTER RELAY TEST. ~v z I SR 3.3.6.4 Perform CFT.

SR 3.3.6.5 Perform SLAVE RELAY TEST.

SR 3.3.6.6, Not used SR 3.3.6.7 Perform CHANNEL CALIBRATION.

SR 3.3.6.8 Verify ESF Containment Ventilation Isolation RESPONSE TIME is within limits.

)

DIABLO CANYON - UNITS 1 & 2 3.3-45 Unit 1 - Amendment No. 1-35,-1i-7 8S9lDI14.DOA -R14 49 Unit 2 - Amendment No. 1-35, 1-84.

CRVS Actuation Instrumentation 3.3.7 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.I Suspend movement of Immediately associated Completion recently irradiated fuel Time for Condition A or B assemblies. ~L) not met during movement of recently irradiated fuel assemblies.

E. Required Action and E.1 Initiate action to restore Immediately associated Completion one CRVS train to Time for Condition A or B OPERABLE status.

not met in MODE 5 or 6.

SURVEILLANCE REQUIREMENTS

NOTE------------------------------------------------

Refer to Table 3.3.7-1 to determine which SRs apply for each CRVS Actuation Function.

SURVEILLANCE FREQUENCY SR 3.3.7.1 Perform CHANNEL CHECK. r SR 3.3.7.2 Perform CFT.

Verification of setpoint is not required.

Perform TADOT.

SR 3.3.7.7 Perform CHANNEL CALIBRATION8 DIABLO CANYON - UNITS 1 & 2 3.3-48 Unit 1 - Amendment No. 435,+847-8S9IDI14.DOA -R14 52 Unit 2 - Amendment No. 95,44&.

FBVS Actuation Instrumentation 3.3.8 SURVEILLANCE REQUIREMENTS

NOTE---------------------------------------------------

Refer to Table 3.3.8-1 to determine which SRs apply for each FBVS Actuation Function.

SURVEILANCEFREQUENCY SR 3.3.8.3 Not used SR 3.3.8.4---------------------------NOTE------------------------

Verification of setpoint is not required.

Perform TADOT.

SR 3.3.8.5 Perform CHANNEL CALIBRATION.8 DIABLO CANYON - UNITS 1 & 2 3.3-51 Unit 1 - Amendment No. 135 8S9IDI14.DOA -R14 55 Unit 2 - Amendment No. 135

RCS Pressure, Temperature, and Flow DNB Limits 3.4.1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 RCS Pressure, Temperature, and Flow Departure from Nucleate Boiling (DNB) Limits LCO 3.4.1 RCS DNB parameters for pressurizer pressure, RCS average temperature, and RCS total flow rate shall be within the limits specified below:

a. Pressurizer pressure is greater than or equal to the limit specified in the COLR;

b. RCS average temperature is less than or equal to the limit specified in the COLR; and

c. RCS total flow rate within limits shown on Table 3.4.1-1 for Unit 1 and Table 3.4.1-2 for Unit 2.

APPLICABILITY: MODES 1.

I. IPVT~I

I'J r---------------------------------------------

Pressurizer pressure limit does not apply during:

a. THERMAL POWER ramp > 5% RTP per minute; or

b. THERMAL POWER step > 10% RTP.

ACTIONS CONDITION, REQUIRED ACTION COMPLETION TIME A. One or more RCS DNB A.1 Restore RCS DNB 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months parameters not within limits, parameter(s) to within limit.

B. Required Action and B.1 Be in MODE 2. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.1.1 Verify pressurizer pressure is greater than or equal to the limit specified in the COLR. N SR 3.4.1.2 Verify RCS average temperature is less than or equal to the limit specified in the COLR. 4-ý

{3~

SR 3.4.1.3 Verify RCS total flow rate is within limits.

SR 3.4.1.4 Verify measured RCS total flow rate is within limits. 4 DIABLO CANYON - UNITS 1 & 2 3.4-1 Unit 1 -Amendment No. 4-3-5,+95-8S91DJ08.doa- R8 1 Unit 2 - Amendment No. 4-35,143t--

RCS Minimum Temperature for Criticality 3.4.2 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.2 RCS Minimum Temperature for Criticality LCO 3.4.2 Each operating RCS loop average temperature (Tavg) shall be Ž! 541 OF.

APPLICABILITY: MODES 1, MODE 2 with keffŽý 1.0.

ACTIONS _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

CONDITION REQUIRED ACTION COMPLETION TIME A. Tavcg in one or more A.1 Be in MODE 2, with Keff 30 minutes operating RCS loops not <1.0 within limit.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.2.1 Verify RCS Tavg in each operating loop Ž! 541 OF.

DIABLO CANYON - UNITS 1 & 2 3.4-4 Unit 1 - Amendment No. t57%

8S91DJ08.doa- R8 4 Unit 2 - Amendment No. +S5

RCS P/T Limits 3.4.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.3.1-------------------------- NOTE --------------------------

Only required to be performed during RCS heatup and cooldown operations and RCS inservice leak and hydrostatic testing.

Verify RCS pressure, RCS temperature, and RCS ~ I heatup and cooldown rates are within the limits specified in the PTLR.

~ 'i-DIABLO CANYON - UNITS 1 & 2 3.4-6 Unit 1 - Amendment No. +&S-8S91DJ08.doa- R8 6 Unit 2 - Amendment No. 1-3 RCS Loops - MODES 1 and 2 3.4.4 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.4 RCS Loops-MODES 1 and 2 LCO 3.4.4 Four RCS loops shall be OPERABLE and in operation.

APPLICABILITY: MODES 1 and 2.

ACTIONS _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of LCO not A.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.4.1 Verify each RCS loop is in operation.

DIABLO CANYON - UNITS 1 & 2 3.4-7 Unit 1 - Amendment No. 4Ef5-,

8S91DJ08.doa- R8 7 Unit 2 - Amendment No. .6

RCS Loops - MODE 3 3.4.5 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. Four RCS loops inoperable. D.1 Place the Rod Control Immediately OR System in a condition OR incapable of, rod withdrawal.

No RCS loop in operation. AND D.2 Suspend operations that Immediately would cause introduction of coolant into the RCS with boron concentration less than required to meet SDM of LCO 3.1.1.

AND D.3 Initiate action to restore Immediately one RCS loop to OPERABLE status and operation.

SURVEILLANCE REQUIREMENTS _______

SURVEILLANCE FREQUENCY SR 3.4.5.1 Verify required RCS loops are in operation.

SR 3.4.5.2 Verify steam generator secondary side water levels (j are Žý15% for required RCS loops.

SR 3.4.5.3 Verify correct breaker alignment and indicated power are available to the required pump that is not in operation.!1

.F~w-se k-- "

DIABLO CANYON - UNITS 1 & 2 3.4-9 Unit 1 - Amendment No. 4135,158%

8S9IDJ08.doa- R8 9 Unit. 2 - Amendment No. -5,b,

RCS Loops - MODE 4 3.4.6 ACTIONS (continued) __________

CONDITION REQUIRED ACTION COMPLETION TIME B. Two required loops 8.1 Suspend operations that Immediately inoperable, would cause introduction OR of coolant into the RCS with boron concentration less than required to meet SDM of LCO 3.1.1.

AND No RCS or RHR loop in B.2 Initiate action to restore Immediately operation. one loop to OPERABLE status and operation.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.6.1 Verify one RHR or RCS loop is in operation.

SR 3.4.6.2 Verify SG secondary side water levels are Žý-15% for 12-required RCS loops.

SR 3.4.6.3 Verify correct breaker alignment and indicated power are available to the required pump that is not in operation.

DIABLO CANYON - UNITS 1 & 2 3.4-11 Unit 1 - Amendment No. 1-35 4-42 +-5&

8S9IDJ08.doa- R8 11 Unit 2 - Amendment No. 1-351442 +

RCS Loops - MODE 5, Loops Filled 3.4.7 ACTIONS ________

CONDITION REQUIRED ACTION COMPLETION TIME A. One RHR loop inoperable. A.1 Initiate action to restore Immediately AND a second RHR loop to OPERABLE status.

OR Required SGs secondary A.2 Initiate action to restore Immediately side water levels not within required SG secondary limits, side water levels to within limits.

-B. Required RHR loops B.1 Suspend operations that Immediately inoperable, would cause introduction OR of coolant into the RCS with boron concentration less than required to meet SDM of LCO 3.1.1.

AND No RHR loop in operation. B.2 Initiate action to restore Immediately one RHR loop to OPERABLE status and I ~operation.I SURVEILLANCE REQUIREMENTS _______

SURVEILLANCE FREQUENCY SR 3.4.7.1 Verify one RHR loop is in operation.

SR 3.4.7.2 Verify SG secondary side water level is Ž! 15% in required SGs.

SR 3.4.7.3 Verify correct breaker alignment and indicated power are available to the required RHR pump that is not in operation.

DIABLO CANYON - UNITS 1 & 2 3.4-13 Unit 1 -Amendment No. 1-351-8S91DJ08.doa- R8 13 Unit 2 - Amendment No. -3.5 +9

RCS Loops - MODE 5, Loops Not Filled 3.4.8 SURVEILLANCE REQUIREMENTS _______

SURVEILLANCE FREQUENCY SR 3.4.8.1 Verify one RHR_ loop is in operation.

SR 3.4.8.2 Verify correct breaker alignment and indicated power7 are available to the required RHR pump that is not in operation.

DIABLO CANYON - UNITS 1 & 2 3.4-15 Unit 1 - Amendment No. +3!

8S91DJ08.doa- R8 15 Unit 2 -Amendment No. I-at

Pressurizer 3.4.9 SURVEILLANCE REQUIREMENTS _______

SURVEILANCEFREQUENCY SR 3.4.9.2 Verify capacity of each required group of pressurizer 2-SR 3.4.9.3 Verify by transferring power, that required pressurizert DIABLO CANYON - UNITS 1 & 2 3.4-17 Unit 1 - Amendment No. 4-35r-8S91DJ08.doa- R8 17 Unit 2 - Amendment No. t35

Pressurizer PORVs 3.4.11 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.11.1 ------------------------ NOT r -----------------------------------

in accordance with the Required Actions of this LCO.

Perform a complete cycle of each block valve. 0ý SR 3.4.11.2------------------------- NOTE --------------------------

Required to be performed during MODES 3 or 4.

Perform a complete cycle of each PORV. In accordance with the IST Plan.

SR 3.4.11.3 Demonstrate OPERABILITY of the safety relatedo nitrogen supply for the Class I PORVs.

SR 3.4.11.4 Perform a COT on each required Class 1 PORV, "9 41 excluding actuation.A SR 3.4.11.5 Perform CHANNEL CALIBRATION for each requirec 4 Class 1 PORV actuation channel.I DIABLO CANYON - UNITS 1 & 2 3.4-22 Unit 1 - Amendment No. 4-35, +4H4 8S91DJ08.doa- R8 22 Unit 2 -Amendment No. 145, ti2-

LTOP System 3.4.12 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME G. (continued)

LTOP System inoperable for any reason other than Condition A, B, C, D, E, or F.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.12.1 Verify a maximum of zero safety injection pumps are capable of injecting into the RCS.

SR 3.4.12.2 Verify a maximum of one centrifugal charging pump is capable of injecting into the RCS.

SR 3.4.12.3 Verify each accumulator is isolated when accumulator pressure is greater than or equal to the maximum (90-ýD7 I RCS pressure for the existing RCS cold leg temperature allowed by the P/T limit curves provided -ilnewt. I in the PTLR.

SR 3.4.12.4 Not usedI SR 3.4.12.5 Verify required RCS vent Ž! 2.07 square inches open. (iurs for rinlocke d openn ven~t va s).

AV 31 days for vent valve(s ocked ope , sealed, or otnerwise secure in the, opel.

positio' SR 3.4.12.6 Verify PORV block valve is open for each required Class I PORV.

SR 3.4.12.7 Not used (continue)

DIABLO CANYON - UNITS 1 & 2 3.4-25 Unit 1 - Amendment No. 1-ý5-,

8S9IDJ08.doa- R8 25 Unit 2 - Amendment No. 1.4-

LTOP System 3.4.12 SURVEILLANCE REQUIREMENTS (continued) _______

SURVEILLANCE FREQUENCY SR 3.4.12.8 ------------------------- NOTE -------------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after decreasing any RCS cold leg temperature to *ý LTOP arming temperature specified in the PTLR.

-- --- - Perform a COT on each required Class 1 PORV, 3 s excluding actuation.

SR 3.4.12.9 Perform CHANNEL CALIBRATION for each required Class I PORV actuation channel.

DIABLO CANYON - UNITS 1 & 2 3.4-26 Unit 1-Amendment No. $45ý 8S9IDJ08.doa- R8 26 Unit 2 - Amendment No. 44&,

RCS Operational LEAKAGE 3.4.13 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.13.1 ------------------------ NOTES-------------------------

1. Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

2. Not applicable to primary to secondary LEAKAGE. 4-Verify RCS operational LEAKAGE is within limits by performance of RCS water inventory balance.

SR 3.4.13.2 ------------------------ NOTE --------------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after j establishment of steady state operation.

----4 ,

Verify primary to secondary LEAKAGE is:5 150 gallons per day through any one SG.

DIABLO CANYON - UNITS 1 & 2 3.4-28 Unit 1 - Amendment No. 4-35, +-94-,

8S91DJ08.doa- R8 28 Unit 2 - Amendment No. 1-35, 1.9&-

RCS PIV leakage 3.4.14 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2.1 Isolate the high pressure 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months portion of the affected system from the low pressure portion by use of a second closed manual, deactivated automatic, or check valve.

OR A.2.2 Restore RCS PIV to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months within limits.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time for Condition A not AND met.

B.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS _______

SURVEILLANCE [_FREQUENCY SR 3.4.14.1 NOTES ------------------------

1. Not required to be performed in MODES 3 and 4.

2. Not required to be performed on the RCS PIVs located in the RHR flow path when in the shutdown cooling mode of operation.

3. RCS PIVs actuated during the performance of this Surveillance are not required to be tested more than once if a repetitive testing loop cannot be avoided.

T?

Verify leakage from each RCS PIV is equivalent to In accordance witto

! 0.5 gpm per nominal inch of valve size up to a the Inservice 'W maximum of 5 gpm at an RCS pressure Žý 2215 psig and *ý2255 psig.

AND ays 4ar

.(continued)

DIABLO CANYON - UNITS 1 & 2 3.4-30 Unit 1 - Amendment No. 4-35 t42-8S91DJ08.doa- R8 30 Unit 2 - Amendment No. 4-35 142-

RCS Leakage Detection Instrumentation 3.4.15 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D. Required Action and D.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion AND Time not met.

D.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months E. All required monitors E.1 Enter LCO 3.0.3. Immediately inoperable.

SURVEILLANCE REQUIREMENTS _______

SURVEILLANCE FREQUENCY SR 3.4.15.1 Perform CHANNEL CHECK of the required containment atmosphere particulate and gaseous radioactivity monitors.

SR 3.4.15.2 Perform CHANNEL FUNCTIONAL TEST of the required containment atmosphere particulate and gaseous radioactivity monitors.

SR 3.4.15.3 Perform CHANNEL CALIBRATION of the required containment sump monitors.

SR 3.4.15.4 Perform CHANNEL CALIBRATION of the required containment atmosphere particulate and gaseous radioactivity monitors.

SR 3.4.15.5 Perform CHANNEL CALIBRATION of the required CFCU condensate collection monitors.

I : Ki DIABLO CANYON - UNITS 1 & 2 3.4-34 Unit 1 - Amendment No. 4e-5; 8S91DJ08.doa- R8 34 Unit 2 - Amendment No. 1ý5,

RCS Specific Activity 3.4.16 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.16.1-----------------------------NOTE--------------------

Only required to be performed in MODE 1.

-- --- -- --- -- .4 Verify reactor coolant DOSE EQUIVALENT XE-i133 specific activity

600.0 pCi/gm.

SR 3.4.16.2------------------------- NOTE--------------------------

Only required to be performed in MODE 1.

-- ---- -- ---- -- -- ---- 4-Verify reactor coolant DOSE EQUIVALENT 1-131 specific activity:!* 1.0 pCi/gm. AND Between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a THERMAL POWER change ofŽý 15% RTP within a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period.

DIABLO CANYON - UNITS 1 & 2 3.4-36 Unit 1 - Amendment No. 4-35,492-8S9IDJ08.doa- R8 36 Unit 2 - Amendment No. 3514-37r1

Accumulators 3.5.1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.1 Accumulators LCO 3.5.1 Four ECCS accumulators shall be OPERABLE.

APPLICABILITY: MODES 1 and 2, MODE 3 with RCS pressure > 1000 psig.

ACTIONS_________ ___

CONDITION REQUIRED ACTION COMPLETION TIME A. One accumulator A. 1 Restore boron 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months inoperable due to boron concentration to within concentration not within limits.

limits.

B. One accumulator B.1 Restore accumulator to 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months inoperable for reasons OPERABLE status.

other than Condition A.

C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A or B AND not met.

C.2 Reduce RCS pressure 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to < 1000 psig.

D. Two or more accumulators D.1 Enter LCO 3.0.3. Immediately inoperable.

SURVEILLANCE REQUIREMENTS _______

SURVEILLANCE FREQUENCY SR 3.5.1.1 Verify each accumulator isolation valve is fully open. 2 SR 3.5.1.2 Verify borated water volume in each accumulator is 1 0

> 814 f 3 and < 886 f 3 SR 3.5.1.3 Verify nitrogen cover pressure in each accumulator is

> 579 psig and < 664 psig. y7 I (continued)

DIABLO CANYON - UNITS 1 & 2 3.5-1 Unit 1 - Amendment No. .1-35, 44-7, 460; 8S91 DK06. DOA - R6 1 Unit 2 - Amendment No. 4-35, 447~, 46-1,

Accumulators 3.5.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY

.- e-SR 3.5.1.4 Verify boron concentration in each accumulator is

> 2200 ppmn and < 2500 ppm.

AND NOTE ---- -

Only required to be performed for affected accumulators.

Once within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after each 1,

solution volume increase of > 5.6%

of narrow range indicated level that is not the result of addition from the refueling water storage tank.

SR 3.5.1.5 Verify power is removed from each accumulator isolation valve operator when RCS pressure is > 1000 psig.

/

~ &

DIABLO CANYON - UNITS 1 & 2 3.5-2 Unit 1 - Amendment No. +-35%

8S9IDK06.DOA - R6 2 Unit 2 - Amendment No. 4Y&-,

ECCS - Operating 3.5.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SIR 3.5.2.1 Verify the following valves are in the listed position with power to the valve operator removed.

Number Position Function 8703 Closed RHR to RCS Hot Legs 8802A Closed Safety Injection to RCS Hot Legs 8802B Closed Safety Injection to RCS Hot Legs 8809A Open RHR to RCS Cold Legs 8809B Open RHR to RCS Cold Legs 8835 Open Safety Injection to RCS Cold Legs 8974A Open Safety Injection Pump Recirc.

to RWST 8974B Open Safety Injection Pump Recirc.

to RWST 8976 Open RWST to Safety Injection Pumps 8980 Open RWST to RHR Pumps 8982A Closed Containment Sump to RHR Pumps 8982B Closed Containment Sump to RHR Pumps 8992 Open Spray Additive Tank to Eductor 8701 Closed RHR Suction Tv~wk .L 8702 Closed RHR Suction SR 3.5.2.2 Verify each ECCS manual, power operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.

SR 3.5.2.3 Verify ECCS piping is full of water.1 7- (continued)

DIABLO CANYON - UNITS 1 & 2 3.5-4 Unit 1 - Amendment No. 1-35ý 8S9IDK06.DOA - R6 4 Unit 2 - Amendment No. 135

ECCS - Operating 3.5.2 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.5.2.4 Verify each ECCS pump's developed head at the test In accordance with flow point is greater than or equal to the required the Inservice developed head. Testing Program..

SR 3.5.2.5 Verify each ECCS automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.

SR 3.5.2.6 Verify each ECCS pump starts automatically on an actual or simulated actuation signal. ~A$~~AA ~3.

SR 3.5.2.7 Verify, for each ECCS throttle valve listed below, each mechanical position stop is in the correct position.

Charging Injection Safety Injection Throttle Valves Throttle Valves 8810OA 8822A 8810B 8822B 8810C 8822C 8810D 8822D SR 3.5.2.8 Verify, by visual inspection, each ECCS train containment recirculation sump suction inlet is not restricted by debris and the suction inlet trash racks and screens show no evidence of structural distress or abnormal corrosion.

DIABLO CANYON - UNITS 1 & 2 3.5-5 Unit 1 - Amendment No.-fe-a5-8S91 DK06. DOA - R6 5 Unit 2 - Amendment No. +35T

RWST 3.5.4 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.4 Refueling Water Storage Tank (RWST)

LCO 3.5.4 The RWST shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. RWST boron concentration A.1 Restore RWST to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months not within limits. OPERABLE status.

OR RWST borated water temperature not within limits.

B. RWST inoperable for B.1 Restore RWST to 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months reasons other than OPERABLE status.

Condition A.

C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion AND Time not met.

C.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCEREQUIREMENTS _______

SURVEILLANCE IFREQUENCY SR 3.5.4.1 ------------------------ NOTE --------------

Only required to be performed when ambient air temperature is < 351F.

Verify RWST borated water temperature is 94 hows > 35 0F.

SR 3.5.4.2 Verify RWST borated water volume is > 400,000 9ws gallons (81.5% indicated level).I SR 3.5.4.3 Verify RWST boron concentration is > 2300 ppm and 1xLa-W -

< 2500 ppm.

DIABLO CANYON - UNITS 1 & 2 3.5-7 Unit 1 - Amendment No. 135 radCA2DB.Doc -R6 7 Unit 2 - Amendment No. 135

Seal Injection Flow 3.5.5 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) 3.5.5 Seal Injection Flow LCO 3.5.5 Reactor coolant Jpump seal injection flow resistance shall be

> 0.2117 ft/gpm .

APPLICABILITY: MODES 1,2, and 3.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Seal injection flow A.1 Adjust manual seal 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months resistance not within limit, injection throttle valves ~49~

to give a flow resistance within limit.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion AN Time not met.AN B.2 Be in MODE 4. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SURVEILLANCE REQUIREMENTS _______

SURVEILLANCE FREQUENCY SR 3.5.5.1-- --------------------- NOTE ------------------------

Not required to be performed until 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after the Reactor Coolant System pressure stabilizes at

> 2215 psig and < 2255 psig.

Verify manual seal injection throttle valves are3 adjusted to give a flow resistance > 0.2117 ftlgpm2 .

DIABLO CANYON - UNITS 1 & 2 3.5-8 Unit 1 - Amendment No. 145, t48-1 8S9IDK06.DOA - R6 8 Unit 2 - Amendment No. 1-35, $48-)

Containment Air Locks 3.6.2 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. One or more containment C.1 Initiate action to evaluate Immediately air locks inoperable for overall containment reasons other than leakage rate per Condition A or B. LCO 3.6. 1.

AND C.2 Verify a door is closed in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months the affected air lock.

AND C.3 Restore air lock to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months OPERABLE status.

D. Required Action and D.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion AND Time not met.

D.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.2.1-------------------------- NOTES---------------------

1. An inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test.

2. Results shall be evaluated against acceptance criteria applicable to SR 3.6. 1.1 Perform required air lock leakage rate testing in In accordance with accordance with the Containment Leakage Rate the Containment Testing Program. Leakage Rate Testing Program SR 3.6.2.2 Verify only one door in the air lock can be opened at a time.I DIABLO CANYON - UNITS 1 & 2 3.6-4 Unit 1 - Amendment No. t"5-,

TAB 3.6 -R3 4 Unit 2 - Amendment No. it5-6

Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS_________

SURVEILLANCE FREQUENCY SR 3.6.3.1 Not used SR 3.6.3.2 Verify each 48 inch containment purge supply and exhaust and 12 inch vacuum/pressure relief valve is closed, except when these valves are open for pressure control, ALARA or air quality considerations for personnel entry, or for Surveillances that require the valves to be open.

SR 3.6.3.3 ------------------------- NOTE ---------------------

Valves and blind flanges in high radiation areas may be verified by use of administrative controls.

Verify each containment isolation manual valve and blind flange that is located outside .X-C4,cL1-containment and not locked, sealed or otherwise secured and required to be closed during accident conditions is closed, except for containment isolation valves that are open under administrative controls.

SR 3.6.3.4 ------------------------ NOTE ---------------------

Valves and blind flanges in high radiation areas may be verified by use of administrative means.

Verify each containment isolation manual valve Prior to entering MODE and blind flange that is located inside 4 from MODE 5 if not containment and not locked, sealed or otherwise performed within the secured and required to be closed during previous 92 days accident conditions is closed, except for containment isolation valves that are open under administrative controls.

SR 3.6.3.5 Verify the isolation time of each automatic power In accordance with the operated containment isolation valve is within- Inservice Testing limits. Program SR 3.6.3.6 Not usedI (continued)

DIABLO CANYON - UNITS 1 & 2 3.6-9 Unit 1 - Amendment No. 4-3St TAB 3.6 -R3 9 Unit 2 -Amendment No. 1.8&,

Containment Isolation Valves 3.6.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY

+

SR 3.6.3.7 ------------------------ NOTE ---------------------

This surveillance is not required when the penetration flow path is isolated by a leak tested blank flange.

Perform leakage rate testing for containment purge supply and exhaust and vacuum/pressure AND relief valves with resilient seals.

For containment purge supply and exhaust valves only, within 92 days after opening the valve SR 3.6.3.8 Verify each automatic containment isolation valve that is not locked, sealed or otherwise secured in position, actuates to the isolation position on an actual or simulated actuation signal.

SR 3.6.3.9 Not used SR 3.6.3. 10 Verify each 12 inch containment vacuum/pressure relief valve is blocked to restrict the valve from opening > 5Q0'.*__

SR 3.6.3. 11 Not used DIABLO CANYON - UNITS 1 & 2 3.6-10 Unit 1 - Amendment No. 1-35,17-Er, TAB 3.6 -R3 10 Unit 2 -Amendment No. 4-35,+7 Containment Pressure 3.6.4 3.6 CONTAINMENT SYSTEMS 3.6.4 Containment Pressure LCO 3.6.4 Containment pressure shall be > -1.0 psig and < +11.2 psig.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS ________

CONDITION REQUIRED ACTION COMPLETION TIME A. Containment pressure not A.1 Restore containment 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months within limits, pressure to within limits.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion AND Time not met.

B.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SIR 3.6.4.1 Verify containment pressure is within limits.

DIABLO CANYON - UNITS 1 & 2 3.6-11 Unit 1 - Amendment No. t35i TAB 3.6 -R3 11 Unit 2 - Amendment No.4-T

Containment Air Temperature 3.6.5 3.6 CONTAINMENT SYSTEMS 3.6.5 Containment Air Temperature LCO 3.6.5 Containment average air temperature shall be < 120*F.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Containment average air A. 1 Restore containment 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months temperature not within limit, average air temperature to within limit.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion AND Time not met.

B.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS ________

SURVEILLANCE FREQUENCY SR 3.6.5.1 Verify containment average air temperature is (ýýo 1s within limit.

DIABLO CANYON - UNITS 1 & 2 3.6-12 Unit 1 - Amendment No.1i TAB 3.6 -R3 12 Unit 2 - Amendment No. +35-

Containment Spray and Cooling Systems 3.6.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.6.6.1 Verify each containment spray manual, power 1i operated, and automatic valve in the flow path that is not locked, sealed, or otherwise secured in position is in the correct position.

SR 3.6.6.4 Verify each containment spray pump's developed In accordance with the head at the flow test point is greater than or Inservice Testing equal to the required developed head. Program SR 3.6.6.5 Verify each automatic containment spray valve in the flow path that is not locked, sealed, or k 4~h~

otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.

SR 3.6.6.6 Verify each containment spray pump starts automatically on an actual or simulated actuation signal.

SR 3.6.6.7 Verify each CFCU starts automatically on an actual or simulated actuation signal.

SR 3.6.6.8 Verify each spray nozzle is unobstructed.

SR 3.6.6.9 Verify each CFCU starts on low speed.

I]-s j A DIABLO CANYON - UNITS 1 & 2 3.6-15 Unit 1 - Amendment No. 4e5, TAB 3.6 -R3 15 Unit 2 - Amendment No. 1~

Spray Additive System 3.6.7 3.6 CONTAINMENT SYSTEMS 3.6.7 Spray Additive System LCO 3.6.7 The Spray Additive System shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS_____ ___

CONDITION REQUIRED ACTION COMPLETION TIME A. Spray Additive System A. 1 Restore Spray Additive 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months inoperable. System to OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion AND Time not met.

B.2 Be in MODE 5. 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months SURVEILLANCE REQUIREMENTS ________

SURVEILLANCE FREQUENCY SR 3.6.7.1 Verify each spray additive manual, power operated, and automatic valve in the flow path that is not locked, sealed, or otherwise secured in position is in the correct position.

SR 3.6.7.2 Verify spray additive tank solution volume is

> 46.2% and < 91.9%.

SR 3.6.7.3 Verify spray additive tank NaOH solution concentration is-> 30% and < 32% by weight.

SR 3.6.7.4 Verify each spray additive automatic valve in the flow path that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.

SR 3.6.7.5 Verify spray additive flow from each solution's flow path.

I~et~

DIABLO CANYON - UNITS 1 & 2 3.6-16 Unit 1 -Amendment No.4M-TAB 3.6 -R3 16 Unit 2 - Amendment No. 3E

MSIVs 3.7.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.2.1-- ------------------------ NOTE----------------------

Only required to be performed in MODES 1 and 2.

Verify closure time of each MSIV is *- 5 seconds. In accordance with the Inservice Testing Program SR 3.7.2.2-------------------------- NOTE ----------------------

Only required to be performed in MODES 1 and 2.

Verify each MSIV actuates to the isolation position0 on an actual or simulated actuation signal.___

.Lv~~-k

DIABLO CANYON - UNITS 1 & 2 3.7-5 Unit 1 - Amendment No. +a&,--

8S91DM09.DOA- R9 5 Unit 2 - Amendment No. e95-,

MFl Vs, MFRVs, MFRV Bypass Valves, MFWP Turbine Stop Valves 3.7.3 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.7.3.3 Verify each MFIV, MFRV, MFRV bypass valve,h and MFWP turbine stop valve actuates to the closed position on an actual or simulated actuation signal.

SR 3.7.3.4 Verify the closure time of each MFWP turbine At each COLD stop valve is *! 1 second. SHUTDOWN, but not more frequently than once per 92 days.

DIABLO CANYON - UNITS 1 & 2 3.7-7a Unit 1 - Amendment No. 435,440,t1847 8S9IDM09.DOA- R9 8 Unit 2 - Amendment No. 435,440,48&-

ADVs 3.7.4 SURVEILLANCE REQUIREMENTS_________

SURVEILLANCE FREQUENCY SR 3.7.4.1 Verify one complete cycle of each ADV.

SR 3.7.4.2 Verify one complete cycle of each ADV block In accordance with the valve. Inservice Testing Program SR 3.7.4.3 Verify that the backup air bottle for each ADV4 has a pressure Ž! 260 psig.

DIABLO CANYON - UNITS 1 & 2 3.7-9 Unit 1 - Amendment No. lt-5,-r 8S9IDM09.DOA- R9 10 Unit 2 - Amendment No. 1~-

AEW System 3.7.5 SURVEILLANCE REQUIREMENTS ________

SURVEILLANCEFRQEC SR 3.7.5.1 Verify each AFW manual, power operated, and J'31 automatic valve in each water flow path, and in both steam supply flow paths to the steam turbine4 driven pump, that is not locked, sealed, or otherwise secured in position, is in the correct position.

SR 3.7.5.2-- ------------------------ NOTE-----------------------

Not required to be performed for the turbine driven AFW pump until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after Ž: 650 psig in the steam generator.

Verify the developed head of each AFW pump at In accordance with the the flow test point is greater than or equal to the Inservice Test required developed head. Program.

SR 3.7.5.3-- ------------------------ NOTE-----------------------

Not applicable in MODE 4 when steam generator is relied upon for heat removal.

Verify each AFW automatic valve that is not ý locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.

SR 3.7.5.4-- ------------------------ NOTES ---------------------

1. Not required to be performed for the turbine driven AFW pump until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after Ž! 650 psig in the steam generator.

2. Not applicable in MODE 4 when generator is *1 relied upon for heat removal.

Verify each AFW pump starts automatically on an <jOýDf actual or simulated actuation signal.___________

SR 3.7.5.5 Not used.

SR 3.7.5.6 Verify the FWST is capable of being aligned to the 4 AFW system by cycling each FWST valve in the flow path necessary for realignment through at -E*%-e-ri':.

least one full cycle.I DIABLO CANYON - UNITS 1 & 2 3.7-12 Unit 1 - Amendment No. 435,t-88-)

8S9IDM09.DOA- R9 13 Unit 2 - Amendment No. 445jeft--

CST and FWST 3.7.6 3.7 PLANT SYSTEMS 3.7.6 Condensate Storage Tank (CST) and Fire Water Storage Tank (FWST)

LCO 3.7.6 The CST level shall be Ž41.3% and the FWST level shall be Ž!22.2% for one unit operation and Ž41.7% for two unit operation.

APPLICABILITY: MODES 1, 2, and 3, MODE 4 when steam generator is relied upon for heat removal.

ACTIONS_____ ___

CONDITION REQUIRED ACTION COMPLETION TIME A. CST or FWST level not A. 1 Verify by administrative 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months within limit, means OPERABILITY of backup water supply. AND Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter AND A-2 Restore CST or FWST 7 days level to within limit.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion AN Time not met.AN B.2 Be in MODE 4, without 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months reliance on steam generator for heat removal.

SURVEILLANCE REQUIREMENTS SURVEILANCEFREQUENCY SR~~eiy'hLS

~ ~eeli ~ !4.% ~3... ~ ~ v~v DIABLO CANYON - UNITS 1 & 2 3.7-13 Unit 1-Amendment No. +3 8S91DM09.DOA- R9 14 Unit 2 -Amendment No. $35-

CCW System 3.7.7 3.7 PLANT SYSTEMS 3.7.7 Vital Component Cooling Water (CCW) System LCO 3.7.7 Two vital CCW loops shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION~m REQUIRED ACTION COMPLETION TIME A. One vital CCW loop A.1 NOTE -----------

inoperable. Enter applicable Conditions and Required Actions of LCO 3.4.6, "RCS Loops - MODE 4,"

for residual heat removal loops made inoperable by CCW.

Restore vital CCW loop to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion AN Time of Condition A not AN met. B.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.7.1 --------------------------- NOTE ----------------------

Isolation of CCW flow to individual components 4fj does not render the CCW System inoperable Verify each CCW manual, power operated, and 1Ay automatic valve in the flow path servicing safety related equipment, that is not locked, sealed, or otherwise secured in position, is in the correct position.

(continued)

DIABLO CANYON - UNITS 1 & 2 3.7-14 Unit 1 - Amendment No. t35-&-,

8S9IDM09.DOA- R9 15 Unit 2 -Amendment No. 4-

CCW System 3.7.7 SURVEILLANCE REQUIREMENTS (continued) _______

SR 3.7.7.2 Verify each CCW automatic valve in the flow path that o is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal.

SR 3.7.7.3 Verify each COW pump starts automatically on an -r actual or simulated actuation signal.

Te-~

DIABLO CANYON - UNITS 1 & 2 3.7-15 Unit 1 - Amendment No. +3~5--

8S9IDM09.DOA- R9 16 Unit 2 - Amendment No. 46-,5

ASW 3.7.8 3.7 PLANT SYSTEMS 3.7.8 Auxiliary Saltwater (ASW) System LCO 3.7.8 Two ASW trains shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One ASW train inoperable. A.1----------- NOTE -----------

Enter applicable Conditions and Required Actions of LCO 3.4.6, "RCS Loops - MODE 4,"

for residual heat removal loops made inoperable by ASK.

Restore ASW train to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months

_____________________ OPERABLE status B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion AN Time of Condition A not AN met. B.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS ________

SURVEILLANCE FREQUENCY SR 3.7.8.1 Verify each ASW manual and power operated, 3t s valve in the flow path servicing safety related equipment, that is not locked, sealed, or otherwise Jv'~k~

secured in position, is in the correct position.

SR 3.7.8.2 Verify each ASW power operated valve in the flow In accordance with the path that is not locked, sealed, or otherwise Inservice Test secured in position, can be moved to the correct Program.

position.

SR 3.7.8.3 Verify each ASW pump starts automatically on an actual or simulated actuation signal.

DIABLO CANYON - UNITS 1 & 2 3.7-16 Unit 1-Amendment No. +et5 8S9IDM09.DOA- R9 17 Unit 2 - Amendment No. 135-,

CRVS 3.7.10 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME E. Two CRVS trains EA1 Suspend movement of Immediately inoperable in MODE 5 OR recently irradiated fuel 6, or during movement of assemblies.

recently irradiated fuel assemblies.

F. Two CRVS trains F.1 Enter LCO 3.0.3. Immediately inoperable in MODE 1, 2, 3, or 4 for reasons other than Condition B.

SURVEILLANCE REQUIREMENTS ________

SURVEILLANCE FREQUENCY SR 3.7.10.1 Operate each CRVS train for Ž! 15 minutes.

SR 3.7.10.2 Verify that each CRVS redundant fan is aligned to receive electrical power from a separate OPERABLE vital bus.

SR 3.7.10.3 Perform required CRVS filter testing in In accordance with accordance with the Ventilation Filter Testing VFTP Program (VFTP).

SR 3.7.10.4 Verify each CRVS train automatically switches into the pressurization mode of operation on an actual or simulated actuation signal.

I SR 3.7.10.5 Verify one CRVS train can maintain a positive y pressure of Ž! 0. 125 inches water gauge, relative to the outside atmosphere during the pressurization mode of operation.

I I

_!ýVse" I DIABLO CANYON - UNITS 1 & 2 3.7-19 Unit 1 -Amendment No. 4-35,442,463,t8t 8S9IDM09.DOA- R9 21 Unit 2 - Amendment No. 4-35,44-2,1-65,186f

ABVS 3.7.12 3.7 PLANT SYSTEMS 3.7.12 Auxiliary Building Ventilation System (ABVS)

LCO 3.7.12 Two ABVS trains shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS ____________________

CONDITION REQUIRED ACTION COMPLETION TIME A. The common HEPA filter A.1 Restore the common 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and/or charcoal adsorber HEPA filter and charcoal inoperable. adsorber to OPERABLE status.

B. One ABVS train inoperable. B.1 Restore ABVS train to 7 days OPERABLE status C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion AND Time not met.

C.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS ________

SURVEILLANCE FREQUENCY SR 3.7.12.1 --------------------- ------------

This surveillance shall verify that each ABVS train is aligned to receive electrical power from a separate OPERABLE vital bus. 4.

----4 .

Operate each ABVS train for Ž! 15 minutes.

SR 3.7.12.2 Perform required ABVS filter testing in In accordance with the accordance with the Ventilation Filter Testing VFTP Program (VFTP).

NOTE-------------------------

SR is not applicable to a specific ABVS train when that ABVS train is configured and performing its -14 Ir safety function.

SR 3.7.12.3 Verify each ABVS train actuates on an actual or simulated actuation signal and the system realigns to exhaust through the common HEPA filter and charcoal adsorber.

(continued)

DIABLO CANYON - UNITS 1 & 2 3.7-21 Unit 1 - Amendment No. 1-35,1tt-.

8S9IDM09.DOA- R9 23 Unit 2 - Amendment No. 4-W5,1.5-7

ABVS 3.7.12 SURVEILLANCE REQUIREMENTS (continued) _________

SURVEILLANCEFRQEC SR 3.7.12.4 Not Used.

SR 3.7.12.5 Not Used.

SR 3.7.12.6 Verifying that leakage through the ABVS 24 s Dampers M2A and M2B is less than or equal to 5 cfm when subjected to a Constant Pressure or Pressure Decay Leak Rate Test in accordance .Lie&( A with ASME N510-1989. The test pressure for the leak rate test shall be based on a maximum operating pressure as defined in ASME N510-1989, of 8 inches water gauge.

DIABLO CANYON - UNITS 1 & 2 3.7-22 Unit 1 - Amendment No. +&

8S9IDM09.DOA- R9 24 Unit 2 - Amendment No. 1t36--.

FHBVS, 3.7.13 SURVEILLANCE REQUIREMENTS _______

SURVEILLANCE FREQUENCY SR 3.7.13.1 Operate each FHBVS train for Ž! 15 minutes. 3 SR 3.7.13.2 Perform required FHBVS filter testing in accordance In accordancemwtith with the Ventilation Filter Testing Program (VFTP). the VFTP SR 3.7.13.3 Verify each FHBVS; train actuates on an actual or simulated actuation signal.

SR 3.7.13.4 Verify one FHBVS train can maintain a pressure

! -0. 125 inches water gauge with respect to atmospheric pressure during the post accident mode of operation.

SR 3.7.13.5 Verify damper M-29 can be closed.14 o 1

4N DIABLO CANYON - UNITS 1 & 2 3.7-24 Unit 1 - Amendment No. 4-35 *427 8S9IDM09.DOA- R9 26 Unit 2 - Amendment No. 4-35 142--

Spent Fuel Pool Water Level 3.7.15 3.7 PLANT SYSTEMS 3.7.15 Spent Fuel Pool Water Level LCO 3.7.15 fuel assemblies seated in the storage racks.

APPLICABILITY: During movement of irradiated fuel assemblies in the spent fuel pool.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Spent fuel pool water level A.1 NOTE -----------

not within limit. LCO 3.0.3 is not applicable.

Suspend movement of Immediately irradiated fuel assemblies in the spent fuel pool.

SURVEILLANCE REQUIREMENTS ________

SURVEILLANCE FREQUENCY SR 3.7.15.1 Verify the spent fuel pool water level is Ž! 23 ft !iý,

above the top of the irradiated fuel assemblies seated in the storage racks. _

DIABLO CANYON - UNITS 1 & 2 3.7-26 Unit 1 - Amendment No. t35-)

8S9IDM09.DOA- R9 28 Unit 2 - Amendment No. 1 Spent Fuel Pool Boron Concentration 3.7.16 3.7 PLANT SYSTEMS 3.7.16 Spent Fuel Pool Boron Concentration LCO 3.7.16 The spent fuel pool boron concentration shall be Ž! 2000 ppm.

APPLICABILITY: When fuel assemblies are stored in the spent fuel pool.

ACTIONS ___________

CONDITION REQUIRED ACTION COMPLETION TIME A. Spent fuel pool boron-- ---------------- NOTE -------------

concentration not within LCO 3.0.3 is not applicable.

lim it. ------------------ ----

A. 1 Suspend movement of Immediately fuel assemblies in the spent fuel pool.

AND A.2 Initiate action to restore Immediately spent fuel pool boron concentration to within limit.

SURVEILLANCE REQUIREMENTS ________

SURVEILLANCE FREQUENCY SR 3.7.16.1 Verify the spent fuel pool boron concentration is -h within limit.

DIABLO CANYON - UNITS 1 & 2 3.7-27 Unit 1 - Amendment No. 4-35,45t 8S9IDM09.DOA- R9 29 Unit 2 - Amendment No. 435,1i54,

Secondary Specific Activity 3.7.18 3.7 PLANT SYSTEMS 3.7.18 Secondary Specific Activity LCO 3.7.18 DOSE EQUIVALENT 1-131.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Specific activity not within A. 1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months limit.

AND A.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months SURVEILLANCE REQUIREMENTS ________

SURVEILLANCE FREQUENCY SR 3.7.18.1 Verify the specific activity of the secondary coolant is

0.10 jýiCi/gm DOSE EQUIVALENT I-131.

t7~k I.

DIABLO CANYON - UNITS 1 & 2 3.7-33 Unit 1 - Amendment No. 1-35-,

8S9IDM09.DOA- R9 35 Unit 2 - Amendment No. t3 AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS _______

SURVEILLANCE FREQUENCY SR 3.8. 1.1 Verify correct breaker alignment and indicated power 4!:jjý -e-availability for each required offsite circuit. bt- _mýA 1-SR 3.8.1.2 ------------------------ NOTES ------------------------

1. Performance of SR 3.8.1.7 satisfies this SR.

2. All DG starts may be preceded by an engine prelube period and followed by a warmup period prior to loading. 2-v"=-e-e k '4 Verify each DG starts from standby conditions and achieves speed Ž: 900 rpm, steady state voltage

Ž3785 V and *! 4400 V, and frequency Ž! 58.8 Hz and

61.2 Hz.

SR 3.8.1.3 ------------------------ NOTES ------------------------

1. DG loadings may include gradual loading as recommended by the manufacturer.

2. Momentary transients outside the load range do not invalidate this test.

3. This Surveillance shall be conducted on only one DG at a time.

4. This SR shall be preceded by and immediately follow without shutdown a successful performance of SR 3.8.1.2 or SR 3.8.1.7.

3 r' c Verify each DG is synchronized and loaded and operates for Ž: 60 minutes at a load Ž! 2340 kW and

< 2600 kW.

SR 3.8.1.4 Verify each day tank contains Ž! 250 gal of fuel oil.

SR 3.8.1.5 Check for and remove accumulated water from each day tank.

SR 3.8.1.6 Verify the fuel oil transfer system operates to transfer fuel oil from storage tanks to the day tank.

(continued)

DIABLO CANYON - UNITS 1 & 2 3.8-4 Unit 1 -Amendment No. +35; 8S91 DN07. DOA - R7 4 Unit 2 - Amendment No. 1.35-

AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.7 --------------------------

NOTE----------------

All DG starts may be preceded by an engine prelube period.

Verify each DG starts from standby condition and achieves:

a. in *10 seconds, speed Ž! 900 rpm; and JZ~e~~' ~z. L

b. in *13 seconds, voltage Ž3785 V and

4400 V, and frequency Ž58.8 Hz and

61.2 Hz.

~1~

SR 3.8.1.8 --------------------------

NOTE----------------

This Surveillance shall not normally be performed for automatic transfers in MODE 1 or 2. However, this Surveillance may be performed to reestablish OPERABILITY provided an assessment determines ý_ 2 I

the safety of the plant is maintained or enhanced.

4 Verify automatic and manual transfer of AC power sources from the normal offsite circuit to the alternate required offsite circuit and manual transfer from the alternate offsite circuit to the delayed access circuit.

SR 3.8.1.9 NOTES ------------------------

1. This Surveillance shall not normally be performed in MODE 1 or 2. However, this Surveillance may be performed to reestablish OPERABILITY provided an assessment determines the safety of the plant is maintained or enhanced.

2. If performed with the DG synchronized with D offsite power, it shall be performed at a power factor *ý 0.9.

Verify each DG rejects a load greater than or equal to its associated single largest post-accident load, and:

a. Following load rejection, the frequency is <ý63 Hz;

b. Within 2.4 seconds following load rejection, the voltage is Žý3785 V and *ý4400 V; and

c. Within 2.4 seconds following load rejection, the frequency is Žý58.8 Hz and *ý 61.2 Hz.

(continued)

DIABLO CANYON - UNITS 1 & 2 3.8-5 Unit 1 - Amendment No. 1-35,174t 8S9IDN07.DOA - R7 5 Unit 2 -Amendment No. 44~5,W.

AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.10 Verify each DG operating at. a power factor!* 0.87 ,pt does not trip and voltage is maintained *! 5075 V during and following a load rejection of > 2340 kW Zv-~rt A and *ý 2600 kW.

SR 3.8.1.11 ------------------------ NOTES ------------------------

1. All DG starts may be preceded by an engine prelube period.

2. This Surveillance shall not normally be performed in MODE 1, 2, 3, or 4. However, portions of the Surveillance may be performed to reestablish OPERABILITY provided an assessment determines the safety of the plant is maintained or enhanced.

Verify on an actual or simulated loss of offsite power signal:

a. De-energization of emergency buses;

£1

b. Load shedding from emergency buses;

c. DG auto-starts from standby condition and:

1. energizes permanently connected loads in

1l0 seconds,

2. energizes auto-connected loads through auto-transfer sequencing timers,

3. maintains steady state voltage Ž!3785 V and *54400 V,

4. maintains steady state frequency

Ž! 58.8 Hz and *5 61.2 Hz, and

5. supplies permanently connected and auto-connected loads for Ž! 5 minutes.

(continued)

DIABLO CANYON - UNITS 1 & 2 3.8-6 Unit 1 - Amendment No. 1-35,19*)

8S91DN07.DOA - R7 6 Unit 2 - Amendment No. 1-35,4q-

AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.8.1.12 ------------------------ NOTEý11%

1. All DG starts may be preceded by an engine prelube period.

2. This Surveillance shall not normally be performed in MODE 1 or 2. However, portions

.of the Surveillance may be performed to reestablish OPERABILITY provided an assessment determines the safety of the plant is maintained or enhanced.

Verify on an actual or simulated Safety Injection signal each DG auto-starts from standby condition and: 6ý * ŽL

a. In *ý 13 seconds after auto-start and during tests, achieves voltage 2!3785 V and <!ý4400 V;

b. In *! 13 seconds after auto-start and during tests, achieves frequency Ž: 58.8 Hz and

61.2 Hz;

c. Operates for Ž: 5 minutes;

d. Permanently connected loads are energized from the alternate offsite power source; and

e. Emergency loads are auto-connected through the ESE load sequencing timers to the alternate offsite power source.

SR 3.8.1.13 Verify each DG's automatic trips are bypassed when 4 the diesel engine trip cutout switch is in the cutout position and the DG is aligned for automatic operation except:

a. Engine overspeed;

b. Generator differential current; and

c. Low lube oil pressure; (continued)

DIABLO CANYON - UNITS 1 & 2 3.8-7 Unit 1 - Amendment No. 45+--

8S9IDN07.DOA - R7 7 Unit 2 -Amendment No. 1-35,1W46--

AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY 1-SR 3.8.1.14 NOTES -----------------------

1. Momentary transients outside the load and power factor ranges do not invalidate this test.

Verify each DG operating at a power factor:!*0.87 operates for Ž! 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months s:

a. For Ž! 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months loaded Ž! 2600 kW and Zý-6ýr 4C

! 2860 kW; and

b. For the remaining hours of the test loaded

Ž! 2340 kW and *ý 2600 kW.

1-SR 3.8.1.15 NOTES ------------------------

1. This Surveillance shall be performed within 5 minutes of shutting down the DG after the DG has operated Ž: 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months loaded -,: 2340 kW and *ý 2600 kW.

Momentary transients outside of load range do not invalidate this test.

2. All DG starts may be preceded by an engine prelube period.

i ýýP4 Verify each DG starts and achieves:

a. in *10 seconds, speed Ž! 900 rpm; and

b. in *13 seconds, voltage Ž3785 V, and

4400 V and frequency Ž58.8 Hz and

61.2 Hz.

SR 3.8.1.16 NOTE--------------------------

This Surveillance shall not normally be performed in MODE 1, 2, 3, or 4. However, this Surveillance may be performed to reestablish OPERABILITY provided an assessment determines the safety of the plant is maintained or enhanced.

Verify each DG:

a. Synchronizes with offsite power source while loaded with emergency loads upon a simulated restoration of offsite power; (continued)

DIABLO CANYON - UNITS 1 & 2 3.8-8 Unit 1 - Amendment No. 4.45,I77#--

8S9IDN07.DOA - R7 8 Unit 2 - Amendment No. 435,+7&6

AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS _______

SURVEILLANCE FREQUENCY SR 3.8.1.16 b. Transfers loads to offsite power source; and (continued)

c. Returns to ready-to-load operation.

SR 3.8.1.17------------------------- NOTE --------------------------

This Surveillance shall not normally be performed in MODE 1, 2, 3, or 4. However, portions of the Surveillance may be performed to reestablish OPERABILITY provided an assessment determines the safety of the plant is maintained or enhanced.

Verify, with a DG operating in test mode and connected to its bus, an actual or simulated Safety Injection signal overrides the test mode by:

a. Opening the auxiliary transformer breaker; and

b. Automatically sequencing the emergency loads onto the DG.

SR 3.8.1.18------------------------- NOTE --------------------------

This Surveillance shall not normally be performed in MODE 1, 2, 3, or 4. However, this Surveillance may be performed to reestablish OPERABILITY provided an assessment determines the safety of the plant is i maintained or enhanced.

Verify each ESF and auto-transfer load sequencing p s timer is within its limits.

SR 3.8.1.19------------------------- NOTES ------------------------

1. All DG starts may be preceded by an engine prelube period.

2. This Surveillance shall not normally be performed in MODE 1,2, 3, or 4. However, portions of the Surveillance may be performed to reestablish OPERABILITY provided an assessment determines the safety of the plant is maintained or enhanced.

(continued)

DIABLO CANYON - UNITS 1 & 2 3.8-9 Unit 1 - Amendment No.45,7-8S9IDN07.DOA - R7 9 Unit 2 - Amendment No. 4-35,146-

AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY

+ -

SR 3.8.1.19 Verify on an actual or simulated loss of offsite (continued) power signal in conjunction with an actual or simulated Safety Injection signal: A

a. De-energization of emergency buses;

b. Load shedding from emergency buses; and

c. DG auto-starts from standby condition and:

1. energizes permanently connected loads in *! 10 seconds,

2. energizes auto-connected emergency loads through load sequencing timers,

3. achieves steady state voltage Žý3785 V and *54400 V,

4. achieves steady state frequency Ž: 58.8 Hz and5*61.2 Hz, and

5. supplies permanently connected and auto-connected emergency loads for

>! 5 minutes.

+

SR 3.8.1.20 -------------NOT E--- -----------------

All DG starts may be preceded by an engine prelube period. 4.k Verify when started simultaneously from standby condition, each DG achieves:

a. in *10 seconds, speed Ž! 900 rpm; and

b. in *13 seconds, voltage Žt 3785 V and

4400 V, and frequency Ž: 58.8 Hz and

61.2 Hz.

DIABLO CANYON - UNITS 1 & 2 3.8-10 Unit 1 - Amendment No. 41-35,W"4; 8S9IDN07.DOA -R7 10 Unit 2 - Amendment No. 4-35,4ý6-

Diesel Fuel Oil, Lube Oil, Starting Air, and Turbocharger Air Assist 3.8.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.3.1 Verify fuel oil storage tanks contain combined storage within limits.

SR 3.8.3.2 Verify lubricating oil inventory is Ž: 650 gal.

SR 3.8.3.3 Verify fuel oil properties of new and stored fuel oil are In accordance with tested in accordance with, and maintained within the the Diesel Fuel Oil limits of, the Diesel Fuel Oil Testing Program. Testing Program SR 3.8.3.4 Verify each DG has at least one air start receiver with a pressure is Ž! 180 psig.

SR 3.8.3.5 Check for and remove accumulated water from each 1 s fuel oil storage tank. -

SR 3.8.3.6 Verify each DG turbocharger air assist air receiver 3 pressure is Ž! 180 psig.

I

~ I' DIABLO CANYON - UNITS 1 & 2 3.8-17 Unit 1 - Amendment No-4a5-,

8S9IDN07.DOA - R7 17 Unit 2 -Amendment No. 13-5

DC Sources - Operating 3.8.4 SURVEILLANCE REQUIREMENTS _______

SURVEILLANCE FREQUENCY SR 3.8.4.1 Verify battery terminal voltage is greater than or equal to the minimum established float voltage.

SR 3.8.4.2 Verify each battery charger supplies Ž!400 amps at greater than or equal to the minimum established float voltage for Ž! 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

OR ~'

Verify each battery charger can recharge the battery to the fully charged state within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months while supplying the largest combined demands of the various continuous steady state loads, after a battery discharge to the bounding design basis event discharge state.

SR 3.8.4.3--------------------------- NOTES-----------------------

1. The modified performance discharge test in SR 3.8.6.6 may be performed in lieu of SR 3.8.4.3.

2. This Surveillance shall not be performed in MODE 1, 2, 3, or 4. J ~ .j Verify battery capacity is adequate to supply, and maintain in OPERABLE status, the required emergency loads for the design duty cycle when subjected to a battery service test.

DIABLO CANYON - UNITS 1 & 2 3.8-19 Unit 1 -Amendment No. 4-35,+1-2; 8S91DN07.DOA - R7 20 Unit 2 - Amendment No. 3,7

Battery Parameters 3.8.6 SURVEILLANCE REQUIREMENTS _______

SURVEILLANCE FREQUENCY SR 3.8.6.1-- ------------------------ NOTE -------------------------

Not required to be met when battery terminal voltage v~~~-.A is less than the minimum established float voltage of SR 3.8.4. 1.

Verify each battery float current is :5 2 amps. Zý SR 3.8.6.2 Verify each battery pilot cell voltage is 2: 2.07 V.3 -F SR 3.8.6.3 Verify each battery connected cell electrolyte level is greater than or equal to minimum established design limits.

SR 3.8.6.4 SR 3.8.6.5 Verify each battery pilot cell temperature is greater than or equal to minimum established design limits.

Verify each battery connected cell voltage is 9 4-

- 2.07 V.

SR 3.8.6.6-------------------------- NOTE --------------------------

This Surveillance shall not be performed in MODE 1, 2, 3, or 4.

4~

Verify battery capacity is Ž! 80% of the manufacturer's rating when subjected to a performance discharge AND test or a modified performance discharge test.

24 months when battery shows degradation or has reached 85% of expected life with capacity < 100%

of manufacturer's rating.

AND 24 months when battery has reached 85% of the expected life with capacity

Ž! 100% of manufacturer's rating.

DIABLO CANYON - UNITS 1 & 2 3.8-24 Unit 1 - Amendment No. 4-35,17-Z-8S9IDN07.DOA -R7 26 Unit 2 - Amendment No. 4-35,+4ý

Inverters - Operating 3.8.7 3.8 ELECTRICAL POWER SYSTEMS 3.8.7 Inverters-Operating LCO 3.8.7 Four Class 1lE Vital 120 V UPS inverters shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One required inverter A.1 ----------NOTE------------

inoperable. Enter applicable Conditions and Required Actions of LCO 3.8.9, "Distribution Systems -

Operating" with any vital 120 V AC bus de-energized.

Restore inverter to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months OPERABLE status.

B. Required Action and B.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion AND Time not met.

B.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months

[

SURVEILLANCE REQUIREMENTS ________

SURVEILLANCE FREQUENCY SR 3.8.7.1 Verify correct inverter voltage and alignment to required AC vital buses.

DIABLO CANYON - UNITS 1 & 2 3.8-26 Unit 1 - Amendment No. 1IS6, 8S9IDN07.DOA - R7 28 Unit 2 - Amendment No. *67&

Inverters - Shutdown 3.8.8 SURVEILLANCEREQUIREMENTS _______

SURVEILLANCE ]_FREQUENCY SR 3.8.8.1 Verify correct inverter voltage, and alignments to required AC vital buses.

DIABLO CANYON - UNITS 1 & 2 3.8-28 Unit 1 - Amendment No. M35, 8S9IDN07.DOA - R7 30 Unit 2 - Amendment No. --

Distribution Systems - Operating 3.8.9 SURVEILLANCE REQUIREMENTS _______

SURVEILLANCE FREQUENCY SR 3.8.9.1 Verify correct breaker alignments and voltage to 7~y required AC, DC, and 120 VAC vital bus electrical power distribution subsystems.i DIABLO CANYON - UNITS 1 & 2 3.8-30 Unit 1-Amendment No. 4at5 8S9IDN07.DOA - R7 32 Unit 2 - Amendment No. 1.57

Distribution Systems-Shutdown 3.8.10 ACTIONS_____ ___

CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2.4 Initiate actions to restore Immediately required AC, DC, and 120 VAC vital bus electrical power distribution subsystems to OPERABLE status.

AND A.2.5 Declare associated Immediately required residual heat removal subsystem(s) inoperable and not in operation.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.10.1 Verify correct breaker alignments and voltage to required AC, DC, and 120 VAC vital bus electrical power distribution subsystems. k A DIABLO CANYON - UNITS 1 & 2 3.8-32 Unit 1 - Amendment No. +3--,5i 8S9IDN07.DOA - R7 34 Unit 2 - Amendment No.

Boron Concentration 3.9.1 3.9 REFUELING OPERATIONS 3.91 Boron Concentration LCO 3.9.1 Boron concentrations of all filled portions of the Reactor Coolant System, the refueling canal, and the refueling cavity, that have direct access to the reactor vessel, shall be maintained within the limit specified in the COLR.

APPLICABILITY: MODE 6 ACTI ONS ________

CONDITION REQUIRED ACTION COMPLETION TIME A. Boron concentration not A. 1 Suspend CORE Immediately within limit. ALTERATIONS.

AND A.2 Suspend positive Immediately reactivity additions.

AND A.3 Initiate action to restore Immediately boron concentration to within limit.

SURVEILLANCE REQUIREMENTS ________

SURVEILLANCE FREQUENCY SR 3.9.1.1 Verify boron concentration is within the limit specified in COLR. 25 DIABLO CANYON - UNITS 1 & 2 3.9-1 Unit 1 - Amendment No. 435,46,9--

8S91 DO03. DOA - R3 1 Unit 2 -Amendment No. 1-45,+7-e-

Nuclear Instrumentation 3.9.3 3.9 REFUELING OPERATIONS 3.93 Nuclear Instrumentation LCO 3.9.3 Two source range neutron flux monitors shall be OPERABLE.

APPLICABILITY: MODE 6 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One required source range A.1 Suspend CORE Immediately neutron flux monitor ALTERATIONS except inoperable, for latching control rod drive shafts and friction testing of individual control rods.

AND A.2 Suspend operations that Immediately would cause introduction of coolant into the RCS with boron concentration less than required to meet the boron concentration of LCO 3.9. 1.

B. Two required source range B.1 Initiate action to restore Immediately neutron flux monitors one source range inoperable, neutron flux monitor to OPERABLE status.

AND B.2 Perform SR 3.9.1.1. Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SURVEILLANCE REQUIREMENTS ________

SURVEILLANCE FREQUENCY SR 3.9.3.1 Perform CHANNEL CHECK.

SR 3.9.3.2 ------------------------ NOTE --------------

Neutron detectors are excluded from CHANNEL CALIBRATION.

Perform CHANNEL CALIBRATION.

4ýý DIABLO CANYON - UNITS 1 & 2 3.9-2 Unit 1 - Amendment No. 4-3,5 E 8S9ID003.DOA -R3 2 Unit 2 - Amendment No. 4435 4659-

Containment Penetrations 3.9.4 3.9 REFUELING OPERATIONS 3.9.4 Containment Penetrations LCO 3.9.4 The containment penetrations shall be in the following status:

a. The equipment hatch capable of being closed and held in place by four bolts;

b. One door in each air lock capable of being closed; and

c. Each penetration providing direct access from the containment atmosphere to the outside atmosphere either:

1. closed by a manual or automatic isolation valve, blind flange, or equivalent, or

2. capable of being closed by an OPERABLE Containment Purge and Exhaust Isolation valve.

--- --------------------------- NOTE ------------------------------------

Penetration flow path(s) providing direct access from the containment atmosphere to the outside atmosphere may be unisolated under administrative controls.

APPLICABILITY: During CORE ALTERATIONS, During movement of irradiated fuel assemblies within containment.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more containment A.1 Suspend CORE Immediately penetrations not in required ALTERATIONS.

status.

AND A.2 Suspend movement of Immediately irradiated fuel.

assemblies within containment.

SURVEILLANCE REQUIREMENTS ________

SURVEILLANCE FREQUENCY SR 3.9.4.1 Verify each required containment penetration is in the required status. LCeA+L SR 3.9.4.2 Verify each required containment purge and exhaust ventilation isolation valves actuates to the isolation position on an actual or simulated actuation signal.

DIABLO CANYON - UNITS 1 & 2 3.9-3 Unit 1 - Amendment No. 445i-5fr-~

8S91 DO03. DOA - R3 3 Unit 2 - Amendment No. 4-3$5M

RHR and Coolant Circulation - High Water Level 3.9.5 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.5.1 With the reactor subcritical less than 57 hours6.597222e-4 days 0.0158 hours 9.424603e-5 weeks 2.16885e-5 months , (Týý-

verify one RHR loop is in operation and circulating reactor coolant at a flow rate of Ž: 3000 gpm, OR With the reactor subcritical for 57 hours6.597222e-4 days 0.0158 hours 9.424603e-5 weeks 2.16885e-5 months or more, verify one RHR loop is in operation and circulating reactor coolant at a flow rate of Ž! 1300 gpm.

41 DIABLO CANYON - UNITS 1 & 2 3.9-5 Unit 1 - Amendment No. 44&5-,

8S91 DO03. DOA - R3 6 Unit 2 -Amendment No. 1-ý

RHR and Coolant Circulation - Low Water Level 3.9.6 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.9.6.1 With the reactor subcritical less than 57 hours6.597222e-4 days 0.0158 hours 9.424603e-5 weeks 2.16885e-5 months , verify one RHR loop is in operation and circulating reactor coolant at a flow rate of Ž! 3000 gpm, OR With the reactor subcritical for 57 hours6.597222e-4 days 0.0158 hours 9.424603e-5 weeks 2.16885e-5 months or more, XVA-bef IZ.

verify one RHR loop is in operation and circulating reactor coolant at a flow rate of Ž! 1300 gpm.

SR 3.9.6.2 Verify correct breaker alignment and indicated power available to the required RHR pump that is not in operation.

DIABLO CANYON - UNITS 1 & 2 3.9-7 Unit 1 - Amendment No. 446,-

8S91D003.DOA-R3 8 Unit 2 -Amendment No. *S57

Refueling Cavity Water Level 3.9.7 3.9 REFUELING OPERATIONS 3.9.7 Refueling Cavity Water Level LCO 3.9.7 Refueling cavity water level shall be maintained Žý23 ft above the top of reactor vessel flange.

APPLICABILITY: During movement of irradiated fuel assemblies within containment.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Refueling cavity water level A.1 Suspend movement of Immediately not within limit, irradiated fuel assemblies within containment.

I SURVEILLANCE REQUIREMENTS ________

SURVEILLANCE FREQUENCY SR 3.9.7.1 Verify refueling cavity water level is Ž: 23 ft above the top of reactor vessel flange. 01--

DIABLO CANYON - UNITS 1 & 2 3.9-8 Unit 1 - Amendment No. 49&yi 8S9ID003.DOA -R3 9 Unit 2 - Amendment No. 436j-

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.17 Battery Monitorinq and Maintenance Proqram This Program provides for restoration and maintenance, based on the recommendations of IEEE Standard 450, "IEEE Recommended Practice for Maintenance, Testing, and Replacement of Vented Lead-Acid Batteries for Stationary Applications," or of the battery manufacturer, of the following:

a. Actions to restore battery cells with float voltage < 2.13 V, and

b. Actions to equalize and test battery cells that have been discovered with electrolyte level below the top of the plates.

T___ -J\C_-4_( -'C E.

DIABLO CANYON - UNITS 1 & 2 5.0-24a Unit 1-Amendment No. t-+/-

8S91DQ19.DOA- R19 25 Unit 2 - Amendment No. 1W4-

Enclosure 3 PG&E Letter DCL-07-097 Changes to Technical Specification Bases Pages (For information only) 1

Enclosure 3 PG&E Letter DCL-07-097 Insert 1 The Surveillance Frequency is based on operating experience, equipment reliability, and plant risk and is controlled under the Surveillance Frequency Control Program.

2

SDM B 3.1.1 BASES E SR 3.1.1.1 (continued)

REQUIREMEN TS Using the ITC accounts for Doppler reactivity in this calculation because the reactor is subcritical, and the fuel temperature will be changing at the same rate as the RCS.

The Frequency of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is based on the generally slow changee in req(ui red oncentration and the low probability of an accident cor Ty\e;e'k (-I occurrin@,without the required S .This allows e for the operator to collect the required data, Wich includes pe ~rming a boron REFERENCES 1 10 CFR 50, Appendix A, GDC 26.

2 ESAR, Chapter 15, Section 15.4.2.1.

3 FSAR, Chapter 15, Section 15.2.4.

4 10OCFR 100.

5 FSAR, Chapter 15, Section 15.4.6.1.6.

DIABLO CANYON - UNITS 1 & 2 Revision 4 radFAC33. Doc - R4 5

Core Reactivity B 3.1.2 BASES ACTIONS B.1 (continued) If the core reactivity cannot be restored to within the 1% Ak/k limit, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . If the SDM for MODE 3 is not met, then the boration required by LCO 3.1.1 Required Action A.1 would occur. The allowed Completion Time is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.1.2.1 REQUIREMENTS Core reactivity is verified by periodic comparisons of measured and predicted RCS boron concentrations. The comparison is made, considering that other core conditions are fixed or stable, including control rod position, moderator temperature, fuel temperature, fuel depletion, xenon concentration, and samarium concentration. The Surveillance is performed prior to entering MODE 1 as an initial check on core conditions and design calculations at BOC. The SR is modified by a Note. The Note indicates that the normalization (adjustment, only if necessary) of predicted core reactivity to the measured value must take place within the first 60 effective full power days (EFPD) after each fuel loading. This allows sufficient time for core conditions to reach steady state, but prevents operation for a large fraction of the fuel cycle without establishing a benchark for the desi n calculatins.Th required s sequent Frequpe y of .31 EFPD, following ~e initial 60 FPD afe ntring MODE s ccept le, base~d the slow te of core ags due to eI d ept'and the lindicators (QPTR, AFD, etc.) for pro mt indication of pres 'nce of other

[an Ianomatly.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 26, GDC 28, and GDC 29.

2. FSAR, Chapter 15.

DIABLO CANYON - UNITS 1 & 2 Revision 4 radFAC33.Doc - R4 10

Rod Group Alignment Limits B 3.1.4 BASES ACTIONS D.2 (continued)

If more than one rod is found to be misaligned or becomes misaligned because of bank movement, the unit conditions fall outside of the accident analysis assumptions. Since automatic bank sequencing would continue to cause misalignment, the unit must be brought to a MODE or Condition in which the LCO requirements are not applicable.

To achieve this status, the unit must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

The allowed Completion Time is reasonable, based on operating experience, for reaching MODE 3 from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.1.4.1 REQUIREMENTS Verification that individual rod positions are within alignment limits sk-0 -'

FrF~g"ency,'of 12 14"IFS p~rovides a history that allows the operator to detect a rod that is beginning to deviate from its expected position.

Ihe spec-itled hr euency takes into account other rod position informatioq.at is continuously avgjia~fe to the operatorý control room, 5<that during actual rod detected.iton oindeations crl-m ditlyb car~imdaeyb SR 3.1.4.2 Verifying each rod is OPERABLE would require that each rod be tripped. However, in MODES 1 and 2, tripping each rod would result in radial or axial power tilts, or oscillations. Exercising each individual rod

-94prov2darpoides confidence that all rods continue to be OPERABLE without exceeding the alignment limit, even if they are not regularly tripped. Moving each rod by 10 steps will not cause radial control room and 5 3.1.4.1, which is pwormed more fregue:ntl nd during required perform an&e-s of SR 3.1.4.2 (determination of rod OPERABILITY by movement), if a rod(s) is discovered to be immovable, but remains trippable, the rod(s) is considered to be OPERABLE. At any time, if a rod(s) is immovable, a determination of the trippability (OPERABILITY) of the rod(s) must be made, and appropriate action taken.

SR 3.1.4.3 Verification of rod drop times allows the operator to determine that the maximum rod drop time permitted is consistent with the assumed rod drop time used in the safety analysis. Measuring rod drop times prior (continued)

DIABLO CANYON - UNITS 1 & 2 Revision 4 radFAC33. Doc - R4 24

Shutdown Bank Insertion Limits B 3.1.5 BASES ACTIONS B.1 (continued) If the shutdown banks cannot be restored to within their insertion limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , the unit must be brought to a MODE where the LCO is not applicable. The allowed Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable, based on operating experience, for reaching the required MODE from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.1.5.1 REQUIREMENTS Verification that the shutdown banks are within their insertion limits prior to an approach to criticality ensures that when the reactor is critical, or being taken critical, the shutdown banks will be available to shut down the reactor, and the required SDM will be maintained following a reactor trip. This SR and Frequency ensure that the shutdown banks are withdrawn before the control banks are withdrawn during a unit startup.

Since the shutdown banks are positioned manually by the control room operator, a verification of shutdown bank position ati ý 4-24iewfter the reactor is taken critical, is adequate to ensure that Fq Fr' ue n cy they are within their insertion limits. IAlso, th 12 hou5 ...

.. e 2-vx~eW t into account pTer i~nfo~rmation available in the co:ntro'l 40om for t (urpose of~Thonitoring the statGlsof shutdown ro REFERENCES 1. 10 CFR 50, Appendix A, GDC 10, GDC 26, and GDC 28.

2. 10 CFR 50.46.

3. FSAR, Chapter 15, Section 15.4.3.2.4.

DIABLO CANYON - UNITS 1 & 2 Revision 4 radFAC33.Doc - R4 29

Control Bank Insertion Limits B 3.1.6 BASES (continued)

SURVEILLANCE SR 3.1.6.1 REQUIREMENTS This Surveillance is required to ensure that the reactor does not achieve criticality with the control banks below their insertion limits.

The estimated critical position (ECP) depends upon a number of factors, one of which is xenon concentration. If the ECP was calculated long before criticality, xenon concentration could change to make the ECP substantially in error. Conversely, determining the ECP immediately before criticality could be an unnecessary burden. There are a number of unit parameters requiring operator attention at that point. Performing the ECP calculation within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months prior to criticality avoids a large error from changes in xenon concentration, but allows the operator some flexibility to schedule the ECP calculation with other startup activities.

SR 3.1.6.2 pdion limits at a Fr ~uency of 2Lvc",e y k 1hor sufficient to ensure PEAIIT o detect control ban ha1it may be approach ngthe insertion limits since, normally, very little rod motionn occurs in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months SR 3.1.6.3 When control banks are maintained within their insertion limits as checked by SR 3.1.6.2 above, it is unlikely that their sequence and overlap will not be in accordance with requirements provided in the COLR. The verification of compliance with the sequence and overlap limits specified in the COLR consists of an observation that the static rod positions of those control banks not fully withdrawn from the core are within the limits specified in the COLR. Bank sequence and overlap must also be maintained durnn rod movement, implicit within I ~ the LCO. (A Fregwency of 12 houC,%< consistent wit inisertion limit REFERENCES 1. 10 CFR 50, Appendix A, GDC 10, GDC 26, GDC 28.

2. 10 CFR 50.46.

3. FSAR, Chapter 4, Section 4.3.2.4.

4. FSAR, Chapter 4, Section 4.3.2.4.

5. WCAP-9273-A, "Westinghouse Reload Safety Evaluation Methodology," July 1985.

DIABLO CANYON - UNITS 1 & 2 Revision 4 radFAC33.Doc - R4 34

PHYSICS TESTS Exceptions - MODE 2 B 3.1.8 BASES ACTIONS D.l1 (continued) If the Required Actions cannot be cornpieted within the associated Completion Time, the plant must be brought to a MODE in which the requirement does not apply. To achieve this status, the plant must be brought to at least MODE 3 within an additional 15 minutes. The Completion Time of 15 additional minutes is reasonable, based on operating experience, for reaching MODE 3 in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.1.8.1 REQUIREMENTS A CHANNEL OPERATIONAL TEST is required on each power range and intermediate range nuclear instrument in MODES 1 and 2 by LCO 3.3.1, "Reactor Trip System (RTS) Instrumentation." SR 3.1.8.1 verifies that the above surveillances are current on all bistables, ensuring that the RTS is properly aligned to provide the required degree of core protection prior to initiation and during the performance of PHYSICS TESTS.

SR 3.1.8.2 Verification that the RCS lowest operating loop Tavg is Ž!531 OF will rvJa4,K '1 -61 TESTS will eruilre that the initial conditions of the safet analyses are SR 3.1.8.3 Verification that the THERMAL POWER is:* 5% RTP will ensure that the plant is not operating in a condition that could invalidate the saety analyses. -VerMifation of the THERMAL POWER at a Frequency of Eour uring the performance of the PHYSICS TESTS will ensure that Cou t the initial conditions of the safety anal ses are ýnot violate~d.

-. !-A SR 3.1.8.4 Verification that the SDM is within limits specified in the COLR ensures that, for the specific RCCA and RCS temperature manipulations performed during PHYSICS TESTS, the plant is not operating in a condition that could invalidate the safety analysis assumptions.

The SDM for physics testing during tests where traditional SDM monitoring techniques are not adequate, is determined for the most restrictive test based on design calculations. Plant conditions are monitored during these tests to verify adequate SDM.

(continued)

DIABLO CANYON - UNITS 1 & 2 Revision 4 rad9AA54.Doc - R4 44

PHYSICS TESTS Exceptions - MODE 2 B 3.1.8 BASES SURVEILLANCE SR 3.1.8.4 (continued)

REQUIREME TS(he Frequen~cy of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is basedd o~~ geerall slw ahp e in A~required boron centrat~ion and on Zhelo poaility of aea ient

Z..L ~ccurrino witi out the required SDM.

REFERENCES 1. 10 CFR 50, Appendix B, Section XI.

2. 10 CFR 50.59.

3. Regulatory Guide 1.68, Revision 2, August, 1978.

4. Not used.

5. WCAP-9273-NP-A, "Westinghouse Reload Safety Evaluation Methodology Report," July 1985.

6. WCAP-1 1618, including Addendum 1, April 1989.

DIABLO CANYON - UNITS 1 & 2 Revision 4 radFAC33. Doc - R4 45

Fa(Z)

B 3.2.1 BASES SURVEILLANCE at RTP (or any other level for extended operation). Equilibrium REQUIREMENTS conditions are achieved when the core is sufficiently stable such that (continued) the uncertainties associated with the measurement are valid. In the absence of these Frequency conditions, it is possible to increase power to RTP and operate for 31 days without verification of Fc(Z) and F*,(Z). The Frequency condition is not intended to require verification of these parameters after every 20% increase in power level above the last verification. It only requires verification after a power level is achieved for extended operation that is 20% higher than that power at which FQ(Z) was last measured.

SR 3.2. 1.1 Verification that F~c(Z) is within its specified limits involves increasing Fm(Z) to allow for manufacturing tolerance and measurement uncertainties in order to obtain F~c(Z). Specifically, FN(Z) is the measured value of FQ(Z) obtained from core power distribution measurement results and F~c(Z) = Fm,(Z) UFO (Ref. 2). The value of UFQ is determined using the formulation provided in the COLR. F~c(Z) is then compared to its specified limits.

The limit with which F,'(Z) is compared varies inversely with power

.above 50% RTP and directly with a function called K(Z) provided in the COLR.

Performing this Surveillance in MODE 1 prior to exceeding 75% RTP (and meeting the 100% RTP FQ(Z) limit) provides assurance that the F~c(Z) limit is met when RTP is achieved, because peaking factors generally decrease as power level is increased.

If THERMAL POWER has been increased by Ž>20% RTP since the last determination of F~c(Z), another evaluation of this factor is required 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after achieving equilibrium conditions at this higher power level to ensure that F~c(Z) values are being reduced sufficiently with power increase to stay within the LCO limits.

The Frequency of. 3 FPD is adequate omonitor the changeTof power distr with core burnu cause such changes low andwecontrolled when t 51ant is operated in accafdM-nce witsh the Technical Specifications- TS).

(continued)

DIABLO CANYON - UNITS 1 & 2 Revision 4 radFEO3B.Doc R4 7

FQ(Z)

B 3.2.1 BASES SURVEILLANCE SR 3.2.1.2 (continued)

REQUIREMENTS prevent FQ(Z) from exceeding its limit for any significant period of time without detection. Performing the Surveillance in MODE 1 prior to exceeding 75% RTP or at a reduced power at any other time, and meeting the 100% RTP FQ(Z) limit, provides assurance that the FQ(Z) limit will be met when RTP is achieved, because peaking factors are generally decreased as power level is increased.

FQ(Z) is verified at power levels Ž! 20% RTP above the THERMAL POWER of its last verification, 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after achieving equilibrium conditions to ensure that F0 (Z) is within its limit at higher power levels.

The Surveillance Freq ency of 31 EFPD is normally adequatet monitor the change power distribution with core burnup. The Surveillance /m e done more frequent* if required by the results of FQ(Z) evalua ns.

The Frequency of 31 EFPD is ad uate to monitor change of power distribution because su a change is suf~i~ently slow, when the plant is operated in accord /ce with the TS, to preclude adverse peaking factors between 31 day surveillances.

REFERENCES 1. 10 CFR 50.46,1974.

2. WCAP-7308-L-P-A, "Evaluation of Nuclear Hot Channel Factor Uncertainties," June 1988.

3. WCAP-12472-P-A, "BEACON Core Monitoring and Operations Support System," August 1994.

DIABLO CANYON - UNITS 1 & 2 Revision 4 radFEO3B.Doc R4 9

B 3.2.2 BASES SURVEILLANCE SR 3.2.2.1 (continued)

REQUIREMENTS After each refueling, FN, must be determined in MODE 1 prior to exceeding 75% RTP. This requirement ensures that FAH limits are met at the beginning of each fuel cycle. Performing this Surveillance in MODE 1 prior to exceeding 75% RTP, or at a reduced power level at any other time, and meeting the 100% RTP FAH limit, provides assurance that the FNH limit is met when RTP is achieved, because peaking factors generally decrease as power levell is increased.

PTh31PD Frequency is acceptable beca~ushe power distribution~

~e~( r j -U...fchans reaieysolvrti mount of el burnup. ~ordingly,)

jthis Frequency is soenough that the FN limit cannot b6exceeded Ifor any significant period of operation.

REFERENCES 1. Regulatory Guide 1.77, Rev. 0, May 1974.

2. 10 CFR 50, Appendix A, GDC 26.

3. 10 CFR 50.46.

4. WCAP-12472-P-A, "BEACON Core Monitoring and Operations Support System," August 1994.

DIABLO CANYON - UNITS. 1 & 2 Revision 4 radFE03B.Doc R4 16

AFD B 3.2.3 BASES LCO and bottom excore detectors in each detector well. For convenience, (continued) this flux difference is converted to provide flux difference units expressed as a percentage and labeled as % Aflux or % Al.

The AFD limits are provided in the COLR. The AFD limits for RAOC do not depend on the target flux difference. However, the target AFD may be used to minimize changes in the axial power distribution.

Violating this LCO on the AFD could produce unacceptable consequences if a Condition 11,Ill, or IV event occurs while the AFD is outside its specified limits.

APPLICABILITY The AFD requirements are applicable in MODE 1 greater than or equal to 50% RTP when the combination of THERMAL POWER and core peaking factors are of primary importance in safety analysis.

For AFD limits developed using RAOC methodology, the value of the AED does not affect the limiting accident consequences with THERMAL POWER < 50% RTP and for lower operating power MODES.

ACTIONS A.1 As an alternative to restoring the AFD to within its specified limits, Required Action A.1 requires a THERMAL POWER reduction to

< 50% RTP. This places the core in a condition for which the value of the AFD is not important in the applicable safety analyses. A SURVEILANCE Completion Time of 30 minutes is reasonable, based on operating experience, to reach 50% RTP without challenging plant systems.

SR 3.2.3.1 REQUIREMENTS This Surveillance verifies that the AED, as indicated by each OPERABLE NIS excore channel, is within its specified limits. The urvei ance Frequqpcy of 7 days is adeut consdern the AFD is monitored by,o mputer and an yd~viation from requirei ents is alarmed.

REFERENCES 1. WCAP-8403 (nonproprietary), "Power Distribution Control and Load Following Procedures," Westinghouse Electric Corporation, September 1974.

2. WCAP-1 0216-P-A, Revision 1lA, Relaxation of Constant Axial Offset Control, FQ Surveillance Technical Specification, February 1994 (Westinghouse Proprietary).

3. FSAR, Chapter 4.3.2.2.4.

DIABLO CANYON - UNITS 1 & 2 Revision 4 radFEO3B.Doc R4 19

QPTR B 3.2.4 BASES ACTIONS A.6 (continued)

Required Action A.6 is modified by a Note that states that the peaking factor surveillances must be completed when the excore detectors have been normalized to restore QPTR to within limit (i.e., Required Action A.5). The intent of this Note is to have the peaking factor surveillances performed at operating power levels, which are only required if the excore detectors were normalized to restore QPTR to within limit per Required Action A.5.

B.1 If Required Actions A.1 through A.6 are not completed within their associated Completion Times, the unit must be brought to a MODE or condition in which the requirements do not apply. To achieve this status, THERMAL POWER must be reduced to

50% RTP within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The allowed Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is reasonable, based on operating experience regarding the amount of time required to reach the reduced power level without challenging plant systems.

SURVEILLANCE SR 3.2.4.1 REQUIREMENTS SR 3.2.4.1 is modified by two Notes. Note 1 allows QPTR to be calculated with three power range channels if THERMAL POWER is

< 75% RTP and the input from one Power Range Neutron Flux channel is inoperable. Note 2 allows performance of SR 3.2.4.2 in lieu of SR 3.2.4. 1.

Input from a Power Range Neutron Flux channel is considered to be operable if the upper and lower detector currents are obtainable. The remaining portion of the channel (the electronics required to provide the channel input to the QPTR alarm) need not be operable.

This Surveillance verifies that the QPTR, as indicated by the Nuclear Ti~,c~eiz. 1-sFrequency of 7 cas takes in-to cont otne'rinform~.ien and =aar is For those causes of QPT that occu r quickly (e.g., a dropped rod), there typically are other indications of abnormality that prompt a verification of core power tilt.

SR 3.2.4.2 This Surveillance is modified by a Note, which states that it is not required until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after the input from one or more Power Range Neutron Flux channels is inoperable and the THERMAL POWER is

> 75% RTP.

(continued)

DIABLO CANYON -UNITS 1 & 2 Revision 4 radFEO3B.Doc R4 24

QPTR B 3.2.4 BASES SURVEILLANCE SR 3.2.4.2 (continued)

REQUIREMENTS With an NIS power range channel inoperable, tilt monitoring for a portion of the reactor core becomes degraded. Large tilts are likely detected with the remaining channels, but the capability for detection of small power tilts in some auadrat is dcresd.fPe orm-in-gý

-~ A D~ SR3.2.4.2 atrequency of 12 housprovides an a~pu rate

-'s" ~~ J~ ~ alternative n~bans for ensurring tha any t~ilt rem~ais nislmt.

For purposes of monitoring the QPTR when one or more power range channels are inoperable, the moveable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the indicated QPTR and any previous data indicating a tilt. The incore detector monitoring is performed with a full incore flux map or two sets of four thimble locations with quarter core symmetry. The two sets of four symmetric thimbles is a set of eight unique detector locations. These locations are C-8, E-5, E-1 1, H-3, H-I 3, L-5, L-1 1, and N-8.

The symmetric thimble flux map can be used to generate symmetric thimble "tilt." This can be compared to a reference symmetric thimble tilt, from the most recent full core flux map, to generate an incore QPTR. Therefore, incore QPTR can be used to confirm that QPTR is within limits.

With one NIS channel inoperable, the indicated tilt may be changed from the value indicated with all four channels OPERABLE. To confirm that no change in tilt has actually occurred, which might cause the QPTR limit to be exceeded, the incore tilt result may be compared against previous tilt values either using the symmetric thimbles as described above or a complete flux map. Nominally, quadrant tilt from the Surveillance should be within 2% of the tilt shown by the most recent power distribution measurement data.

REFERENCES 1. 10 CFR 50.46.

2. Regulatory Guide 1.77, Rev 0, May 1974.

3. 10 CFR 50, Appendix A, GDC 26.

4. WCAP-12472-P-A, "BEACON Core Monitoring and Operations Support System," August 1994.

DIABLO CANYON -UNITS 1 & 2 Revision 4 radFE03B.Doc R4 25

RTS Instrumentation B 3.3.1 BASES (continued)

SURVEILLANCE The SRs for each RTS Function are identified by the SRs column of REQUIREMENTS Table 3.3.1-1 for that Function.

A note has been added to the SR Table stating that Table 3.3.1 -1 determines which SRs apply to which RTS Functions.

Note that each channel of process protection supplies both trains of the RTS. When testing Channel 1,Train A and Train B must be examined.

Similarly, Train A and Train B must be examined when testing Channel 11,Channel Ill, and Channel IV (if applicable). The CHANNEL CALIBRATION and COTs are performed in a manner that is consistent with the assumptions used in analytically calculating the required channel accuracies.

Performance of the CHANNEL CHECK efee-eveiy-+2-s nensures that gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying that the instrumentation continues to operate properly between each CHANNEL CALIBRATION.

Agreement criteria are determined by the unit staff based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit.

SR 3.3.1.1 channel fa is rare. The ANNELexperience t~hat demonstrates CHECýZ1Jpplements less on operating op o

formal, LThe i-a mre FrequencyZ Iu based frq5,checks of chan~douring normal operational use of the displays associated with the LCO required channels.

(continued)

DIABLO CANYON - UNITS 1 & 2 Revision 4 radD3867,Doc - R4 52

RTS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.2 (continued)

REQUIREMENTS For example, to assure a reactor trip below the power range high SAL, the Power Range Neutron Flux-High trip Setpoint is reduced as necessary prior to adjusting the power range channel output in the decreasing power direction whenever the calorimetric power is

>! 15% RTP and <45% RTP. The maximum allowable Power Range Neutron Flux-High trip Setpoint may be increased with increasing RTP in accordance with surveillance procedures. Following a plant refueling outage, it is prudent to reduce the Power Range Neutron Flux-High trip.

Setpoint prior to startup.

Before the Power Range Neutron Flux-High trip Setpoint is re-set to its nominal full power value (*5109% RTP), the power range channel calibration must be confirmed based on a calorimetric performed at

>: 45% RTP.

The Note to SR 3.3.1.2 clarifies that this Surveillance is required only if reactor power is Ž!15% RTP and that 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is allowed for performing the first Surveillance after reaching 15% RTP, but prior to exceeding 30% RTP. A power level of 15% RTP is chosen based on plant stability, i.e., automatic rod control capability and the turbine generator synchronized to the grid. The 24-hour allowance after increasing THERMAL POWER above 15% RTP provides a reasonable time to attain a scheduled power plateau, establish the requisite conditions, perform the required calorimetric measurement, and make any required adjustments in a controlled, orderly manner and without int roducing the potential for extended operation at high power levels with instrumentation that has not been verified to be acceptable for subsequent use.

op~eFraigexpuence, onsieringhur insaeutrumentreiabityaned operaitin

~c~A i~~a expeein ay 24hor od indicationse a alar end sitodetect deiainstrimnt chnelbltanoupurts. g SR 3.3.1.3 SR 3.3.1.3 compares the incore system to the NIS channel output every 31 EFPD. If the absolute difference is Ž!3%, the NIS channel is still OPERABLE, but must be readjusted. The excore NIS channel shall be adjusted if the absolute difference between the incore and excore AFD is Ž!3%. The comparison checks for differences due to changes in core power distribution since the last calibration.

(continued)

DIABLO CANYON - UNITS 1 & 2 Revision 4 radD3867.Doc - R4 54

RTS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.3 (continued)

REQUIREMENTS If the NIS channel cannot be properly readjusted, the channel is declared inoperable. This Surveillance is performed to verify the f(AI) input to the overtemperature AT Function.

The Note to SR 3.3.1.3 clarifies that the Surveillance is required only if reactor power is Ž! 50% RTP and that 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is allowed for performing the first Surveillan~ce after reaching 50% RTP. This Note allows power ascensions and associated testing to be conducted in a controlled and orderly manner, at conditions that provide acceptable results and without introducing the potential for extended operation at high power levels with instrumentation that has not be verified to be acceptable for subsequent use. Due to such effects as shadowing from the relatively deep control rod insertion and, to a lesser extent, the dependency of the axial ly-de pendent radial leakage on the power level, the relationship between the incore and excore indications of axial flux difference (AFD) at lower power levels is variable. Thus, it is prudent to defer the calibration of the excore AFD against the incore AFD until more stable conditions are attained (i.e., withdrawn control rods and higher power level). The AFD is used as an input to the Overtemperature AT reactor trip function and for assessing compliance with ITS LCO 3.2.3, "AXIAL FLUX DIFFERENCE." Due to the DNB benefits gained by administratively restricting the power level to 50%

RTP, no limits on AFD are imposed below 50% RTP by LCO 3.2.3; thus, the proposed change is consistent with LCO 3.2.3. requirements below 50% RTP. Similarly, sufficient DNB margins are realized through operation below 50% RTP that the intended function of the Overtemperature AT reactor trip function is maintained, even though the excore AFD indication may not exactly match the incore AFD indication. Based on plant operating experience, 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is a reasonable time frame to limit operation above 50% RTP while completing the procedural steps associated with the surveillance in an orderly manner.

The EFrequency of every 31 EFPD is adequate. It is based on

=unit Soperating experie ce, considering i strument reliability a~ operating

~'" L history data for strumenitsdrift o, since the chang9d in neutron flux are slow duni g the fuel cycle, t expected change i the absolute difference between the incore nd excore AFD will be less than 3 percent AFD during this interval (continued)

DIABLO CANYON -UNITS 1 & 2 Revision 4 radD3867. Doc - R4 55

RTS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.4 REQUIREMENTS SR 3.3.1.4 is the performance of a TADOT er ý e (continued) S-T!'G-GE0RED TEzST. BA&I45. This test shall verify OPERABILITY by actuation of the end devices.

The RTB test shall include separate verification of the undervoltage and shunt trip mechanisms. Independent verification of RTB undervoltage and shunt trip Function is not required for the bypass breakers. No capability is provided for performing such a test at power. The independent test for bypass breakers is included in SR 3.3.1.14. The bypass breaker test shall include a local manual shunt trip only. A Note has been added to indicate that this test must be performed on the bypass breaker prior to placing it in service.

Te Frequencyof every 62 daysA*a STAGG ERED-?TtSTýBASIS isi SR 3.3.1.5 SR 3.3.1.5 is the performance of an ACTUATION LOGIC TEST. T9h-

.&eismic trip i-s tese o; 2d.cn- TGGRDT3TOSS The SSPS is tested e SAGERDTET ASS using the semiautomatic tester. The train being tested is placed in the bypass condition with the RTB bypass breaker installed, thus preventing inadvertent actuation. Through the semiautomatic tester, all possible logic combinations, with and without applicable permissives, are tested for each protection function including operation of the P-7 permissive which is a logic function only. The P-7 alarm circuit is excluded from this testing since it only mimics the actions of the SSPS

'~-t /'-.I~

SR 3.3.1.6 SR 3.3.1.6 is a calibration of the excore channels to the incore channels. If the measurements do not agree, the excore channels are not declared inoperable but must be calibrated to agree with the incore power distribution measurements. The incore power distribution measurements can be obtained using the movable incore detectors or an OPERABLE Power Distribution Monitoring System (PDMS)

(Reference 26). If the excore channels cannot be adjusted, the channels are declared inoperable. This Surveillance is performed to verify the f(AI) input to the overtemperature AT Function.

(continued)

DIABLO CANYON - UNITS 1 & 2 Revision 4 radD3867.Doc - R4 56

RTS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.6 (continued)

REQUIREMENTS A Note modifies SR 3.3.1.6. The Note states that this Surveillance is required only if reactor power is Ž! 75% RTP and that 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after thermal power is Ž! 75% RTP is allowed for performing the first surveillance after reaching 75% RTP. The SR is deferred until a scheduled testing plateau above 75% RTP is attained during the post-outage power ascension. During a typical post-refueling power ascension, it is usually necessary to control the axial flux difference at lower power levels through control rod insertion. After equilibrium conditions are achieved at the specified power plateau, a power distribution measurement must be taken and the required data collected. The data is typically analyzed and the appropriate excore calibrations completed within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months after achieving equilibrium conditions. An additional time allowance of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is provided during which the effects of equipment failures may be remedied and any required re-testing may be performed.

The allowance of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after equilibrium conditions are attained at the testing plateau provides sufficient time to allow power ascensions and associated testing to be conducted in a controlled and orderly manner at conditions that provide acceptable results and without introducing the potential for extended operation at high power levels with instrumentation that has not been verified to be acceptable for subsequent use.

The F~~qup~ of 92 EFPD is adequate.t- ~a ed on inc try Cope~rati:0ex~perie~nce,Econsi~dering inst~rtfment reli=iyab -nd operatin isto data for instrument drift.

SR 3.3.1.7 SR 3.3.1.7 is the performance of a COT every 184 days.

A COT is performed on each required channel to ensure the entire channel will perform the intended Function.

Setpoints; must be within the Allowable Values specified in Table 3.3.1-1.

The difference between the current "as found" values and the previous test "as left" values must be consistent with the drift allowance used in the setpoint methodology. The setpoint shall be left set consistent with the assumptions of the current unit specific setpoint methodology.

The "as found" and "as left" values must also be recorded and reviewed for consistency with the assumptions of Reference 7. The frequency of 184 days is justified in Reference 29.

(continued)

DIABLO CANYON - UNITS 1 & 2 Revision 4 rad70124.Doc - R4 57

RTS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.7 (continued)

REQUIREMENTS SR 3.3.1.7 is modified by two notes. Note 1 provides a 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months delay in the requirement to perform this Surveillance for source range instrumentation when entering MODE 3 from MODE 2. This Note allows a normal shutdown to proceed without a delay for testing in MODE 2 and for a short time in MODE 3 until the RTBs are open and SR 3.3.1.7 is no longer required to be performed. If the unit is to be in MODE 3 with the RTBs closed for > 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months this Surveillance must be performed prior to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after entry into MODE 3. Note 2 requires that the quarterly COT for the source range instrumentation shall include verification by observation of the associated permissive annunciator window that the P-6 and P-i 0 interlocks are in their required state for the existing unit conditions. If this surveillance or if SR 3.3.1.8 has been performed within the previous 184 days, the requirements of this surveillance are satisfied.

SR 3.3.1.8 SR 3.3.1.8 is the performance of a COT as described in SR 3.3.1.7 it is modified by the same Note that this test shall include verification that the P-6 and P-10 interlocks are in their required state for the existing unit conditions by observation of the associated permissive annunciator window. The Frequency is modified by a Note that allows this surveillance to be satisfied if it has been performed within 184 days of the Frequencies prior to reactor startup, 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reducing power below P-1 0, and four hours after reducing power below P-6, as discussed below. The Frequency of "prior to startup" ensures this surveillance is performed prior to critical operations and applies to the source, intermediate and power range low instrument channels. The Frequency of "12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reducing power below P-1 0" (applicable to intermediate and power range low channels) and "4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after reducing power below P-6" (applicable to source range channels) allows a normal shutdown to be completed and the unit removed from the MODE of Applicability for this surveillance without a delay to perform the tes~tj* required by this surveillance. The Frequencysf-l

.eve y4.&~dythereafter applies if the plant remains in the MODE of Applicability after the initial performances of prior to reactor startup, 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reducing power below P-1Q0, and four hours after reducing power below P-6. The MODE of Applicability for this surveillance is < P-1 0 for the power range low and intermediate range channels and < P-6 for the source range channels.

(continued)

DIABLO CANYON - UNITS 1 & 2 Revision 4 radD3867.Doc - R4 58

RTS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.8 (continued)

REQUIREMENTS. Once the unit is in MODE 3, this surveillance is no longer required. If power is to be maintained < P-10 for more than 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or < P-6 for more than 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , then the testing required by this surveillance must be performed prior to the expiration of the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months limit, as applicable. These time limits are reasonable, based on operating experience, to complete the required testing or place the unit in a MODE where this surveillance is no longer required. This test ensures that the NIS source, intermediate, and power range low channels are OPERABLE prior to taking the reactor critical and after reducing power into the applicable MODE (< P-i 0 or < P-6) for the periods discussed above. T+1 fiq~ct2; of 34dti jsuie i SR 3.3.1.9 SR 3.3.1.9 is the performance of a TADOT~aond i ofro c~Lft ..- 92 days, ars ju"tfod. R46 ~ O7 The SR is modified by a Note that excludes verification of setpoints from the TADOT. Since this SR applies to RCP undervoltage and underfrequency relays, setpoint verification requires elaborate bench calibration and is accomplished during the CHANNEL CALIBRATION.

1ACHAN N E ;;,A-UBP ION is complete check of the instrument loop, including the sensor. The test verifies that the channel responds to a measured parameter within the necessary range and accuracy.

CHANNEL CALIBRATIONS must be performed consistent with the assumptions of the DCPP setpoint methodology. The difference between the current "as found" values and the previous test "as left" values must be consistent with the drift allowance used in the setpoint methodology.

Whenever an RTD is replaced in Functions 6, 7, or 14, the next required CHANNEL CALIBRATION of the RTDs is accomplished by an inplace cross calibration that compares the other sensing elements with the recently installed sensing element.

.~The Frequency f 24imonths is based the assumed ca ira ion nterval in thdetermination of the a ni ui rrntdh drf in the ret oint methodolo SR 3.3. 1.10 is modified by a Note stating that this test shall include verification that the time constants are adjusted to the prescribed values where applicable.

(continued)

DIABLO. CANYON -UNITS 1 & 2 Revision 4 radD3867.Doc - R4 59

RTS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.11 REQUIREMENTS SR 3.3.1.11 is the performance of a CHANNEL CALIBRATION, as (continued) described in SR 3...0evfy2 mtf The CHANNEL CALIBRATION for the power range nuclear instruments includes a normalization of the detectors based on a power calorimetric and flux map performed above 15% RTP, and a test that shows allowed variances of detector voltage do not effect detector operation. The CHANNEL CALIBRATION for the intermediate range nuclear instruments includes a test that shows allowed variances of detector voltage do not effect detector operation. The CHANNEL CALIBRATION for the source range nuclear instruments includes a periodic test that optimizes detector high voltage and a conditional test for establishing baseline channel settings after maintenance. The baseline test includes obtaining detector high voltage and discriminator bias curves and using this data to evaluate detector and channel settings based on manufacturers' recommendations and industry operating experience.

This SR is modified by three Notes. Note 1 state that neutron detectors are excluded from the CHANNEL CALIBRATION. Note 2 states that the test shall include verification that the time constants are adjusted to the prescribed values where applicable. Note 3 states that, prior to entry into MODE 2 or 1, the power and intermediate range detector plateau voltage verification (as described above) is not required to be current until 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after achieving equilibrium conditions with THERMAL POWER Ž! 95% RTP. Equilibrium conditions are achieved when the core is sufficiently stable at intended operating conditions to perform a meaningful detector plateau voltage verification. The allowance of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after equilibrium conditions are attained at the testing plateau provides sufficient time to allow power ascension testing to be conducted in a controlled and orderly manner at conditions that provide acceptable results and without introducing the potential for extended operation at high power levels with instrumentation that has not been verified to be OPERABLE for subsequent use. The source range curves are obtained as required under the conditions that apply during a plant outage.

The 2:4mo~nth Fr uency is bbas,,d on past operating, ~peine, which

- has shown t 3~e c nnusually pass t e urveillance whený erformed on t on uenc The condition fr obtaining the source range curves and for verifying the power and intermediate range detector operation are described above. The other remaining portions of the CHANNEL CALIBRATIONS may be performed either during a plant outage or during plant operation.

(continued)

DIABLO CANYON - UNITS 1 & 2 Revision 4 radD3867.Doc - R4 60

RTS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.12 REQUIREMENTS SR 3.3. 1.12 is the performance of a CHANNEL CALIBRATION of the (continued) seismic trip,~eyey 244=A#4" For function 22, Seismic Trip, the calibration shall encompass, as a minimum, the sensor relays, the SSPS, and associated required alarms. Since it is impractical to routinely remove and ship the seismic trigger packages to an offsite facility to verify calibration on a shaker table, the sensors shall be verified by introducing a known acceleration to voltage relationship to the sensor and verifying the proper action, in accordance with the manufacturers recommendations.

the Frequency -jpustifie-d byteaas~o~nof an ý24 mop th Scalibration in al in the d~etyerminatii of the mag nit deof equipmeint I drift in the setpoint analysis.

SR 3.3.1.13 SR 3.3.1.13 is the performance of a COT of RTS interlocks eyw 24-wfe19+

The Frequency is basnd on the known rejia bility of the interlocks and

~ 4 -ath~e multichannel ýýundancy availabjl-a nd has been shown =to be

[acceptable through operating experience.ý SR 3.3.1.14 SR 3.3.1.14 is the performance of a TADOT of the Manual Reactor Trip, RCP Breaker Position, Seismic Trip and the SI Input from ESFAS.

This TADOT.-i-sppei faui i nd evei y 24 incr&thg The Manual Reactor Trip test shall independently verify the OPERABILITY of the undervoltage and shunt trip mechanisms for the Reactor Trip Breakers and Reactor Trip Bypass Breakers. Breaker actuation is verified using the local indicator since physical verification of the main contacts is not practical.

This is acceptable based on breaker design and industry operating and maintenance experience. The Seismic Trip TADOT shall, as a minimum, verify the OPERABILITY of the channel from the seismic sensor relays to the input logic of the SSPS. The remainder of the channel is tested under the SR 3.3.1.5 or 3.3.1.12 requirements.

SThe Frequencyý% based on the known reliability of Functiiopcaýdi the multichR~fel red undancy aeilable, and has-5een showntoh accptaluethrough operatKrg experience.

The SR is modified by a Note that excludes verification of setpoints from the TADOT. The Functions affected have no setpoints associated with them except for the Seismic Trip that is calibrated by SR 3.3.1.12 At the &amo 21 memth freqe"M?

(continued)

.DIABLO CANYON - UNITS 1 & 2 Revision 4 radD3867.Doc - R4 61

RTS Instrumentation B 3.3.1 BASES SURVEILLANCE SR 3.3.1.16 (continued)

REQUIREMENTS Response time may be verified by actual response time tests in any series of sequential, overlapping or total channel measurements, or by the summation of allocated sensor response times with actual response time tests on the remainder of the channel. Allocations for sensor response times may be obtained from: (1) historical records based on acceptable response time tests (hydraulic, noise, or power interrupt tests), (2) inplace, onsite, or offsite (e.g. vendor) test measurements, or (3) utilizing vendor engineering specifications.

WCAP-1 3632-P-A Revision 2, "Elimination of Pressure Sensor Response Time Testing Requirements" (Ref. 8) provides the basis and methodology for using allocated sensor response times in the overall verification of the channel response time for specific sensors identified in the WCAP. Response time verification for other sensor types must be demonstrated by test.

WCAP-1 4036-P-A, Revision 1, "Elimination of Periodic Protection Channel Response Time Tests," provides the basis and methodology for using allocated signal processing and actuation logic response times in the overall verification of the protection system channel response time." The allocations for sensor, signal conditioning, and actuation logic response times must be verified prior to placing the component in operational service and reverified following maintenance work that may adversely affect response time. In general, electrical repair work does not impact response time provided the parts used for repair are of the same type and value. Specific components identified in the WCAP may be replaced without verification testing. One example where response time could be affected is replacing the sensing assembly of a transmitter.

number rednant channel's ineaspos tific RTmunction.Testfind ofvter fimnal acuton s deie isG inluEdT E i eAIS verification.

ac ep times oplueratio is reqie to erainsuc ehresponsetrimes.aexperifenct ehastoc show tha mothese comone shusuallpass fnthisn surveiatllcancel when perfored at thoneevrNtis 24Frqenywherefore, months Frqunctas the cncudbed to edacceptcabnle fro a reliabiit sTfntandoint. Ts h (otinued DIABLO CANYON - UNITS 1 & 2 Revision 4 radD3867.Doc - R4 63

ESFAS Instrumentation B 3.3.2 BASES BACKGROUND Solid State Protection System (continued)

'The SAE RELAY TEST interval is 24 months. The test frequency is based on relay reliability sessments presented in WCAP-1 3878,7 "Reliability Assessm of Potter and Brumfield MDR Series Relays,"

WCAP-13900 " ension of Slave Relay gurveillance Test Intervals,"

and WCAP-1 17, "Reliability Assessrrnet of Potter and Brumfieldi MDR Serie, elay." These reliabilitytssessments are relay specific and apply only to Potter and Bru ield MDR series relay hich are the only relays used in the ESE ac ation system. Note t for normally energized applications, the lays may have to bp eplaced periodically in accordance with the gui ance given in WCAP-1 3878 for MDR relays.

APPLI CABLE Each of the analyzed accidents can be detected by one or more SAFETY ESFAS Functions. One of the ESFAS Functions is the primary ANALYSES, LCO, actuation signal for that accident. An ESFAS Function may be the and primary actuation signal for more than one type of accident. An APPLICABILITY ESFAS Function may also be a secondary, or backup, actuation signal for one or more other accidents. Functions such as manual initiation, not specifically credited in the accident safety analysis, are qualitatively credited in the safety analysis and the NRC staff approved licensing basis for the unit. These Functions may provide protection for conditions that do not require dynamic transient analysis to demonstrate Function performance. These Functions may also serve as backups to Functions that were credited in the accident analysis (Ref. 3).

The LCO requires all instrumentation performing an ESFAS Function to be OPERABLE. Failure of any instrument renders the affected channel(s) inoperable and reduces the reliability of the affected Functions.

The LCO generally requires OPERABILITY of four or three channels in each instrumentation function and two channels in each logic and manual initiation function. The two-out-of-three and the two-out-of-four configurations allow one channel to be tripped, cut-out or bypassed during maintenance or testing without causing an ESFAS initiation.

Two logic or manual initiation channels are required to ensure no single random failure disables the ESFAS.

(continued)

DIABLO CANYON - UNITS 1 & 2 Revision 4 radD3867.Doc - R4 71

ESFAS Instrumentation B 3.3.2 BASES (continued)

ACTIONS P.1 or P.2.1. and P.2.2 (continued)

Placing a second channei in the bypass condition for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for testing purposes is justified in Reference 17. The allowed testing configurations provide flexibility for testing, while assuring that during testing no configuration will cause an inadvertent actuation of the function or keep a valid signal from actuating the function or an associated function as designed. This note is not intended to allow simultaneous testing of coincident channels on a routine basis.

SURVEILLANCE The SRs for each ESFAS Function are identified by the SRs column of REQUIREMENTS Table 3.3.2-1.

A Note has been added to the SR Table to clarify that Table 3.3.2-1 determines which SRs apply to which ESFAS Functions.

Note that each channel of process protection supplies both trains of the ESFAS. When testing channel I, train A and train B must be examined. Similarly, train A and train B must be examined when testing channel 11,channel Ill, and channel IV (if applicable).

The CHANNEL CALIBRATION and COTs are performed in a manner that is consistent with the assumptions used in analytically calculating the required channel accuracies.

SR 3.3.2.1 Performance of the CHANNEL CHECK ee viy24ýensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.

Agreement criteria are established in STP I-A, based on a combination of the channel instrument uncertainties, including indication and reliability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit.

The Frequency . based on operating experience that demo trates 6L-lEhannel fa is rare. The CHANNEL CHECK su pplerpels less forma -tt more frequent, check's-channels during normal heannls use of the displays ersociated with the LCO required (continued)

DIABLO CANYON -UNITS 1 & 2 Revision 4 radD3867. Doc - R4 110

ESFAS Instrumentation B 3.3.2 BASES SURVEILLANCE S R 3.3.2.2 -1*T4.*- I~~"

REQUIREMENTS SR 3.3.2.2 is the performance of an ACTUATION LOGIC TEST. Ttnh A--t9i-(continued) SSP&s tested eve8Fl 92 dayrb Gn a STACCEFE+D TSib ý ijusing the semiautomatic tester. The train being tested is placed in the bypass condition, thus preventing inadvertent actuation. Through the semiautomatic tester, all possible logic combinations, with and without applicable permissives, are tested for each protection function. In addition, the master relay coil is pulse tested for continuity. This verifies that the logic modules are OPERABLE and that there is an

.e every 2 dgys on a STAGGE DT BASIS isjustifi~pein SR 3.3.2.3 - Not used SR 3.3.2.4 SR 3.3.2.4 is the performance of a MASTER RELAY TEST. The MASTER RELAY TEST is the energizing of the master relay, verifying contact operation and a low voltage continuity check of the slave relay coil. Upon master relay contact operation, a low voltage is injected to the slave relay coil. This voltage is insufficient to pick up the slave relay, but large enough to demonstrate signal path continuity. 9+hisýe~

-i pegrformd-e~vy 92 dayro , d STAeeGDED TEST BA8109 The time allowed for the testing (4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ) is justified in Reference 8. The-fr#o9961 2Jy~~&SACEE ET AI O utf~

-Refcencoemek .-

SR 3.3.2.5 SR 3.3.2.5 is the performance of a COT.

A COT is performed on each required channel to ensure the entire channel will perform the intended Function. Setpoints must be found within the Allowable Values specified in Table 3.3.1-1.

The difference between the current "as found" values and the previous test "as left" values must be consistent with the drift allowance used in the setpoint methodology. The setpoint shall be left set consistent with the assumptions of the current unit specific setpoint methodology.

The "as found" and "as left" values must also be recorded and reviewed for consistency with the assumptions of the surveill ance interval extension analysis (Ref. 8) when applicable.

(continued)

DIABLO CANYON - UNITS 1 & 2 Revision 4 radD3867.Doc - R4 ill

ESFAS Instrumentation B 3.3.2 BASES SURVEILLANCE SR 3.3.2.6 REQUIREMENTS SR 3.3.2.6 is the performance of a SLAVE RELAY TEST. The SLAVE (continued)

RELAY TEST is the energizing of the slave relays. Contact operation is verified in one of two ways. Actuation equipment that may be operated in the design mitigation MODE is either allowed to function, or is placed in a condition where the relay contact operation can be verified without operation of the equipment. Actuation equipment that may not be operated in the design mitigation MODE is prevented from operation by the SLAVE RELAY TEST circuit. For this latter case, contact operation is verified by a continuity check of the circuit containing the slave relay. Ti oti ofr~de~y2 luIlb T he Erequency is adeguate, based on operating expcr~ieý,

conid~ingrely rlicilty amd operatim h;3tefy data (Rof. 7.

SR 3.3.2.7 - Not used SR 3.3.2.8 SR 3.3.2.8 is the performance of a TADOT. This test is a check of the Manual Actuation Functions (except AFW; see SR 3.3.2.13). +4s-

.per9formd-eve24 *menth9 Each Manual Actuation Function is tested up to, and including, the master relay coils. In some instances, the test includes actuation of the end device (i.e., pump starts, valve cycles, 3~~

,j~ ec.) r~eFequncyisd-equate, based on industry operat~in_'

modified by a Note that excludes verification of setpoints during the TADOT for manual initiation Functions. The manual initiation Functions have no associated setpoints.

SR 3.3.2.9 SR 3.3.2.9 is the performance of a CHANNEL CALIBRATION.

A CHANNEL IBrAiON is per'brm~ed every2

eufn CHANJNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to measured parameter within the necessary range and accuracy.

CHANNEL CALIBRATIONS must be performed consistent with the assumptions of the unit specific se 'tpoint methodology. The difference between the current "as found" values and the previous test "as left" vausmust be consistent with the drift allowance used in the setpoint (continued)

DIABLO CANYON - UNITS 1 & 2 Revision 4 radD3867.Doc - R4 112

ESFAS Instrumentation B 3.3.2 BASES SURVEILLANCE SR 3.3.2.9 (continued)

REQUIREMENTS Whenever an RTD is replaced in Function 6.d., the next required CHANNEL CALIBRATION of the RTDs is accomplished by an inpiace cross calibration that compares the other sensing elements with the recently installed sensing element.

IYA ~ A Thne Frequency of 4 months is based on the assumpti9Ia Of an~

k~ 24 month cali) tion interval in e determination of tt(e magnitude of

~equipment ifft in the setpoi ~methodology.

This SR is modified by a Note stating that this test should include verification that the time constants are adjusted to the prescribed values where applicable.

SIR 3.3.2.10 This SR ensures the individual channel ESF RESPONSE TIMES are less than or equal to the maximum values assumed in the accident analysis. RESPONSE TIME testing acceptance criteria and the individual Functions requiring RESPONSE TIME Verification are included in ECG 38.2. Individual component response times are not modeled in the analyses. The analyses model the overall or total elapsed time, from the point at which the parameter exceeds the Trip Setpoint value at the sensor, to the point at which the equipment in both trains reaches the required functional state (e.g., pumps at rated discharge pressure, valves in full open or closed position).

For channels that include dynamic transfer functions (e.g., lag, lead/lag, rate/lag, etc.), the response time test may be performed with the transfer functions set to one with the resulting measured response time compared to the appropriate FSAR response time. Alternately, the response time test can be performed with the time constants set to their nominal value provided the required response time is analytically calculated assuming the time constants are set~at their nominal values.

The response time may be measured by a series of overlapping tests such that the entire response time is measured.

Response time may be verified by actual response time tests in any series of sequential, overlapping or total channel measurements, or by the summation of allocated sensor response times with actual response time tests on the remainder of the channel. Allocations for sensor response times may be obtained from: 1) historical records based on acceptable response time tests (hydraulic, noise, or power interrupt tests), 2) inplace, onsite, or offsite (e.g., vendor) test measurements, or 3) utilizing vendor engineering specifications.

WCAP-1 3632-P-A, revision 2, "elimination of Pressure sensor (continued)

DIABLO CANYON -UNITS 1 & 2 Revision 4 radD3867.Doc - R4 113

ESFAS Instrumentation B 3.3.2 BASES SURVEILLANCE SR 3.3.2.10 (continued)

REQUIREMENTS Response time Testing requirements," dated January 1996, provides the basis and the methodology of using allocated sensor response times in the overall verification of the channel response time for specific sensors identified in the WCAP. Response time verification for other sensor types must be demonstrated by test.

WCAP-14036-P-A, Revision 1, "Elimination of Periodic Protection Channel Response Time Tests," provides the basis and methodology for using allocated signal processing and actuation logic response times in the overall verification of the protection system channel response time." The allocations for sensor, signal conditioning, and actuation logic response times must be verified prior to placing the component in operational service and reverified following maintenance work that may adversely affect response time. In general, electrical repair work does not impact response time provided the parts used for repair are of the same type and value. Specific components identified in the WCAP may be replaced without verification testing. One example where response time could be affected is replacing the sensing assembly of a transmitter.

EFR ETI ME tegst-a re conducted on a 24 mamth STAGGEREED,* EST BAS§ISý Eac veýrifcationi shall include at least one train such that both trains .1 areveiid at least once per 48 months and one channel per function such that all channel re tested at least once every N times 24 months wher N is e total number of redund t channels in a specific ESFAS function. esting of the final actuati devices, which make up the bulk of the sponse time, is included i the testing of each train.

Therefore, st ggered testing results in re ponse time verification of one train of devices every 24 months. he 24 month Frequency is consistent with the typical refueling cy e and is based on unit operating experience, which shows t t random failures of instrumentation components causing serious response time idegradation, but not chane faljpae0 ent occrrences.

This SR is modified by a Note that clarifies that the turbine driven AFW pump is tested within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after reaching 650 psig in the SGs.

SR 3.3.2.11 -

SR 3.3.2.11 is the performance of a TADOT as described in SR 3.3.2.8, except that it is performed for the P-4 Reactor Trip Interlock. c2 nt eqey adonpctg prin.

The SR is modified by a Note that excludes verification of setpoints during the TADOT. The Function tested has no associated setpoint.

(continued)

DIABLO CANYON - UNITS 1 & 2 Revision 4 radD3867.Doc - R4 114

ESFAS Instrumentation B 3.3.2 BASES SURVEILLANCE SR 3.3.2.12 REQUIREMENTS SR 3.3.2.12 is the performance of an ACTUATION LOGIC TEST.

(continued) k This SIR is applied to the RHR Pump Trip on RWST Level-Low Tis is perfo r ed every 24 months.

-test equipment rel Kability and historical d ta.

he freqluelncy is adequ ....

SR 3.3.2.13 SR 3.3.2.13 is the performance of a TADOT. This test is a check of the Manual Actuation Function for AFW. It is p -'- -- d evor; 18 -

4uanblJ Each Manual Actuation Function is tested up to, and including, the master relay coils. In some instances, the test includes 1^S4+_ i SR is modified by a Note that excludes verification of setpoints during the TADOT for manual initiation Functions. The manual initiation Functions have no associated setpoints.

REFERENCES 1. FSAR, Chapter 6.

2. FSAR, Chapter 7.

3. FSAR, Chapter 15.

4. IEEE Std.279-1971.

5. 10 CFR 50.49.

6. Blank

7. WCAP-1 3900, "Extension of Slave Relay Surveillance Test intervals", April 1994

8. WCAP-10271-P-A, Supplement 2, Rev. 1, June 1990.

(continued)

DIABLO CANYON - UNITS 1 & 2 Revision 4 radD3867.Doc - R4 115

PAM Instrumentation B 3.3.3 BASES ACTIONS E.1 and E.2 (continued)

The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

F.1 Alternate means of monitoring Reactor Vessel Water Level and Containment Area Radiation have been developed. These alternate means may be temporarily installed if the normal PAM channel cannot be restored to OPERABLE status within the allotted time. Monitoring the Core Exit Thermocouples, Pressurizer Level indication (07-LI-459A, 460A or 461), and RCS Subcooling Monitor indication (07-YI-31) provide an alternate means for RVLIS. These three parameters provide diverse information to verify there is adequate core cooling or RCS inventory. When Containment Area Radiation Level (High Range) monitors (R-30 and R-31) are inoperable, selected portable radiation monitoring equipment is made available for taking correlated readings at the equipment or personnel hatches as the alternate method. If these alternate means are used, the Required Action is not to shut down the unit but rather to follow the directions of Specification 5.6.8, in the Administrative Controls section of the TS.

The report provided to the NRC should discuss the alternate means used, describe the degree to which the alternate means are equivalent to the installed PAM channels, justify the areas in which they are not equivalent, and provide a schedule for restoring the normal PAM channels.

SURVEILLANCE A Note has been added to the SR Table to clarify that SR 3.3.3.1 and REQUIREMENTS SR 3.3.3.2 apply to each PAM instrumentation Function in Table 3.3.3-1.

SR 3.3.3.1 Performance of the CHANNEL CHECK ee vr 4ý nue that a gross instrumentation failure has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION. The high radiation instrumentation should be compared to similar unit instruments located throughout the unit.

(continued)

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PAM Instrumentation B 3.3.3 BASES SURVEILLANCE SR 3.3.3.1 (continued)

REQUIREMENTS Agreement criteria are determined by the unit staff, based on a combination of the channel instrument uncertainties, including isolation, indication, and readability. If a channei is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit. If the channels are within the criteria, it is an indication that the channels are OPERABLE.

As specified in the SR, a CHANNEL CHECK is only required for those channels that are normally energized.

_The Frequency of_31_days is based on opere ting experience that

-~$~(*.dedeonstrates th channel failure is rare. ,he CHANNEL CHECK supplements I s formal, but more fre dent, checks of channels during rnormal ope tional use of the display§ associated with the LCO required channels.J S R 3.3.3.2 ______ ____

IA H~ ýEL ALIBRýAJN is performed e~v~er4i< nths, or3

~ar~oiatelv at evr r0efulin. CHAýNNEL CALIBRATIONis a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to measured parameter with the necessary range and accuracy. CHANNEL CALIBRATION of the Neutron Flux Wide Range Function excludes the detectors. To ensure that the detectors are verified, the Neutron Flux Wide Range Channels are cross-correlated and normalized to reactor thermal power.

CHANNEL CALIBRATION of the Containment Radiation Level (High Range) Function may consist of an electronic calibration of the channel, not including the detector, for range decades above lOR/h and a one point calibration check of the detector below 10 R/h with an installed or portable gamma source. Whenever an RTD is replaced in Functions 3 or 4, the next required CHANNEL CALIBRATION of the RTDs is accomplished by an inplace cross calibration that compares the other sensing elements with the recently installed sensing element.

Whenever an incore thermocouple is replaced in Function 15, 16, 17, or 18 the next required CHANNEL CALIBRATION of the incore thermocouples is accomplished by an inplace cross calibration that compares the other sensing elements with the recently installed sensing element. For function 9, Containment Isolation Valve Position, the instrument loop consists of the position switch mounted on the valve, the indication lights in the monitor boxes and the interconnecting wiring. For the CHANNEL CALIBRATION to verify that the channel (continued)

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PAM Instrumentation B 3.3.3 BASES SURVEILLANCE SR 3.3.3.2 (continued)

REQU IREMENTS responds with the necessary range and accuracy, the test must verify that the proper indication is received when the valve is stroked to the fully closed position. Verification of intermediate position or actual percentage closed is not required, however, for OPERABILITY, the position indication must be able to communicate the proper isolation status of the containment penetration. Adjustments to the channel may be done as part of this surveillance or through other controlled instruction s. earency is basedTon_ 9perating expeie~n~ean consis encv wtihevocal industry PefTielina cvcl e.--

REFERENCES 1. FSAR, 7.5.

2. Regulatory Guide 1.97, Revision 3.

3. NUREG-0737, Supplement 1, "TMI Action Items."

4. Supplemental Safety Evaluation Report 14.

5. Supplemental Safety Evaluation Report 31.

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Remote Shutdown System B 3.3.4 BASES (continued)

SURVEILLANCE SR 3.3.4.1 REQUIREMENTS Performance of the CHANNEL CHECK oneeejensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying that the instrumentation continues to operate properly between each CHANNEL CALIBRATION.

Agreement criteria are determined by the unit staff, based on a combination of the channel instrument uncertainties, including indication and readability. If the channels are within the criteria, it is an indication that the channels are OPERABLE. If the channels are normally off scale during times when Surveillance is required, the CHANNEL CHECK will verify only that they are off scale in the same direction. Offscale low current loop channels are verified to be reading at the bottom of the range and not failed downscale.

The CHANNEL CHECK for the RTB serves to verify that the indication correctly indicates the position of the RTB.

operating experience which

~ ~. emon/str/:es that of 31 days channel is basedisupon failure rary The CHANNEL CHECK EThe Frequency supl e, ents less formal, but more fra uent, checks of channels during normal operational use of the displays associated with the LCO req uirechnls SR 3.3.4.2 SR 3.3.4.2 verifies each required Remote Shutdown System control circuit and transfer switch performs the intended function. This verification is performed from the hot shutdown panel and locally, as appropriate. Operation of the equipment from the remote shutdown panel is not necessary. The Surveillance can be satisfied by performance of a continuity check. This will ensure that if the control room becomes inaccessible, the unit can be placed and maintained in MODE 3 from the remote shutdown panel and the local control stationsJ.The 24 ýmon Frequency is based on the need to perform Mis Surveillanc~eu er the conditions that apply Oring a plant outage for n unlannd trnsi4r,tfife Surveillance=were 1and he ptent

ýerformed with the reactor at power. (However, this Surveillance is not (continued)

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Remote Shutdown System B 3.3.4 BASES SURVEILLANCE SR 3.3.4.2 (continued)

REQUIREMENTS required to be peired only during aun ud.Oprtg experien demonstrates that- r ote shutdown contro channels 7- 4 j & usuall; pass' the Surveillance Tswhen performed athe 24mont SR 3.3.4.3 CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. The test verifies that the channel responds to a measured parameter within the necessary range and accuracy.

The channel calibration is not applicable to the RTB indication.

Whenever an RTD is replaced in Function 3.a or 3.b, the next required CHANNEL CALIBRATION of the RTDs is accomplished by an inplace cross calibration that compares the other sensing elements with the recently installed sensing element.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 19 (associated with 1967 GDC 11 per FSAR Appendix 3.11A.).

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LOP DG Start Instrumentation B 3.3.5 BASES (continued)

APPLICABILITY The LOP DG Start Instrumentation Functions are required in MODES 1, 2, 3, and 4 because ESE Functions are designed to provide protection in these MODES. Actuation in MODE 5 or 6 is required whenever the required DG must be OPERABLE so that it can perform its function on an LOP or degraded power to the vital bus.

ACTIONS In the event a channel's Setpoint is found nonconservative with respect to the Allowable Value, or the channel is found inoperable, then the function that channel provides must be declared inoperable and the LCO Condition entered for the particular protection function affected.

Because the required channels are specified on a per bus basis, the Condition may be entered separately for each bus as appropriate.

A Note has been added in the ACTIONS to clarify the application of Completion Time rules. The Conditions of this Specification may be entered independently for each Function listed in the LCO. The Completion Time(s) of the inoperable channel(s) of a Function will be tracked separately for each Function starting from the time the Condition was entered for that Function.

A.1 Condition A applies when one or more of the loss of voltage or the degraded voltage channel functions (this includes both relays and timers) on a single bus are inoperable.

In these circumstances the Conditions specified in LCO 3.8.1, "AC Sources-Operating," or LCO 3.8.2, "AC Sources-Shutdown," for the DG made inoperable by failure of the LOP instrumentation are required to be entered immediately. The actions of those LCOs provide for adequate compensatory actions to assure unit safety.

A Note is added to allow bypassing one channel for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for surveillance testing. This allowance is made where bypassing the channel does not cause an actuation and where at least one other channel is monitoring that parameter.

SURVEILLANCE SR 3.3.5.1 not used REQUIREMENTS SR 3.3.5.2 SR 3.3.5.2 is the performance of a TADOT. Tlii Ltub! ISperf~inieu evor; 18 month. The test checks trip devices that provide actuation signals directly, bypassing the analog process control equipment. For these ýtests the relay Setpoints are verified and adiusted as neces ry.

Th e Frequency is basedd n the known reliability of th lays and 1.-controls and the multi annel redndny available and has beenr (continued)

DIABLO CANYON -UNITS 1 & 2 Revision 4 radD3867.Doc - R4 142

LOP DG Start Instrumentation B 3.3.5 BASES SURVEILLANCE SR 3.3.5.3 REQUIREMENTS SR 3.3.5.3 is the performance of a CHANNEL CALIBRATION.

(continued)

The setpoints, as well as the response to a loss of voltage and a degraded voltage test, shall include a single point verification that the trip occurs within the required time delay.

A~~~~~~0CAIDeTO CHisE Fa niermd er m h.

CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to a measured parameter within the necessary range and accuracy.

~ ........ ~The Frequency of 18 months is based on operating experienc~eand consistency wit ýe typical industry refuelin g,4Scle and is justified by the assumpt~d&n of 18 month calibration interal in the determination of

[the m~a nitud~e of =equipment drift in the setpoint analysis.

REFERENCES 1. ESAR, Section 8.3.

2. FSAR, Chapter 15.

3. Blank

4. Calculation 174A-DC, "Undervoltage Relay Settings for 4KV System (27HFB2 & 27HFT1 )."

5. Calculation 357P-DC, "SLUR and SLUR Timer Setpoints."

DIABLO CANYON - UNITS 1 & 2 Revision 4 radD3867.Doc - R4 143

Containment Ventilation Isolation Instrumentation B 3.3.6 BASES (continued)

ACTIONS A Note states that Condition C is applicable during movement of (continued) recently irradiated fuel assemblies within containment.

SURVEILLANCE A Note has been added to the SR Table to clarify that Table 3.3.6-1 REQU IREMENTS determines which SRs apply to which Containment Ventilation Isolation Functions.

SR 3.3.6.1 Performance of the CHANNEL CHECK eigee..e-a&ýQýwensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.

Agreement criteria are determined by the unit staff, based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit.

7-FI - channel failure is rare The CHANNEL CHECK su eents less-SR 3.3.6.2 SR 3.3.6.2 is the performance of an ACTUATION LOGIC TEST. The train being tested is placed in the bypass condition, thus preventing inadvertent actuation. Through the semiautomatic tester, all possible logic combinations, with and without applicable permissives, are tested for each protection function. In addition, the master relay coil is pulse tested for continuity. This verifies that the logic modules are OPERABLE and there is an intact voltage signal path to the master relycoils. kThis- test ispromdevery :12 1ay rno1 -a-STAGG ER ED V EST BAS Th V6r.eillance interval is jus;ifred in Reference 4.

(continued)

DIABLO CANYON - UNITS 1 & 2 Revision 4 radD3867. Doc - R4 148

Containment Ventilation Isolation Instrumentation B 3.3.6 BASES SURVEILLANCE SR 3.3.6.3 (contnued)SR 3.3.6.3 is the performance of a MASTER RELAY TEST. The (contnued)MASTER RELAY TEST is the energizing of the master relay, verifying contact operation and a low voltage continuity check of the slave relay coil. Upon master relay contact operation, a low voltage is injected to the slave relay coil. This voltage is insufficient to pick up the slave relay,-but large enough to demonstrate signal path continuitv. TTF-is -test A.is pertormed every 92 daysý a STAGGERED TEST BASiii The

..Y~ve~eA-I- urveillance interval is ji~tified in Reference! 4I.

SR 3.3.6.4 0%QA A CFT is performed evr 2dy i ahrequired channel to ensure the entire channel will perform the intended Fun~cti~on. The Pr ý,uency Jis ýased on Te staff r emmenda fionfo-r incesing the a~irlability of 1radiation monitors according to NUREG-1366 (Ref. 2). Thstt verifies; the capability of the instrumentation to provide the containment purge and vacuum/pressure relief system isolation.

To ensure complete end-to-end testing through the CVI mode selector switch, the CFT is only valid for the position in use during the test.

SR 3.3.6.5 SR 3.3.6.5 is the performance of a SLAVE RELAY TEST. The SLAVE RELAY TEST is the energizing of the slave relays. Contact operation is verified in one of two ways. Actuation equipment that may be operated in the design mitigation mode is either allowed to function or is placed in a condition where the relay contact operation can be verified without operation of the equipment. Actuation equipment that may not be operated in the design mitigation mode is prevented from operation by the SLAVE RELAY TEST circuit. For this latter case, contact operation is verified bv a continuitv chec ofte cir uit 4~ j....a..TheFreqenc is cceable based on instrumen, reliabili yand SR 3.3.6.6 There is no manual actuation of CVI except via SI, phase A or B. This testing is performed as part of SR 3.3.2.8 SR 3.3.6.7 A Cl ANNE uALb~TlN ; i~refmed eyor; 24 mcenth3-,-er

~ ~r'rof 'lig~CHANNEL CALIBRATION is a

~ppoxrntel complete check of the instrument loop, including the sensor.

(continued)

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Containment Ventilation Isolation Instrumentation B 3.3.6 BASES SURVEILLANCE SR 3.3.6.7 (continued)

REQUIREMENTS The test verifies that the channel responds to a measured parameter funthion are lessr thanaordequualctyth maxmu valesqy bassuedin-th valeratinthexpsrensor, to the pintat which theequipmen in bothtain reahesin thceurdfucinlsae egvle i ulcoe ovesRlappingrestshsuc thatte entiirea responsel tiESOE TIsMEaSufred.e REPfoNm Cnamests aurge onductedonGaeu 24amothcuStAGeRE and TETBAI.uahneifction shesthnoqall vatluesastsoned tthmaincude trin suhe tacietbohtainalss.aReerifised aietleastinoncepernc 48rmothsiandeo ichanluper funci such that. chanonels arepne testedatreasnce Idvuall moevery times 24aonths. wher analissthmotal numboeralof in redapsdan chanels atswciichEFA fntio frmtepint parTesting fced thefrina ctuation

-Tvý,;e'r 4L 17 deviices, whicmk ptebl responsefeeludsued time, isy inasriso thertesting ofst each train theetrefo sagretespneting rs mesurtsdi response es ti veiiatio ofondce ran of devce evet 24moths.E TETh 24IS month Ferequecytison sisllentld with thstoe typcrefeingscycl tand istbraise onuitoertn exerifida easnce, whic 4mowts tarand om copoentscuing soterfiou fhainlurso in astruentifc SaSftion epnsecutime dhe tadationg echtainne falueargeinred uestng buot occurrencs REFERENCES 1. 10OCFR 100.11.

2. NUREG-1 366, December 1992.

3. DCM No. T-16, Containment Function.

4. WCAP-1 5376-P-A, Revision 1, "Risk-Informed Assessment of the RTS and ESFAS Surveillance Test Intervals and Reactor Trip Breaker Test and Completion Times," March 2003.

5. License Amendment 184/1 86, January 3, 2006.

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CRVS Actuation Instrumentation B 3.3.7 BASES ACTIONS C.1 and 0.2 (continued)

The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

DAl and D.2 Condition D applies when the Required Action and associated Completion Time for Condition A or B have not been met when recently irradiated fuel assemblies are being moved. Movement of recently irradiated fuel assemblies must be suspended immediately to reduce the risk of accidents that would require CRVS actuation.

E.1 Condition E applies when the Required Action and associated Completion Time for Condition A or B have not been met in MODE 5 or 6. Actions must be initiated to restore the inoperable train(s) to OPERABLE status immediately to ensure adequate isolation capability in the event of a waste gas decay tank rupture.

SURVEILLANCE A Note has been added to the SR Table to clarify that Table 3.3.7-1 REQUIREMENTS determines which SRs apply to which CRVS Actuation Functions.

SR 3.3.7.1 Performance of the CHANNEL CHECK Ge-v ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.

Agreement criteria are determined by the unit staff, based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit.

T~khe Frequency is based on operating experience that de= nstrates hanelfalue i rre T~eCHANNEL CHECK supple nts less formal, but more freq onit, checks of channels dunin normal operational use of the displays associated with the CO required (continued)

DIABLO CANYON -UNITS 1 & 2 Revision 4 radD3867.Doc - R4 155

CRVS Actuation Instrumentation B 3.3.7 BASES SURVEILLANCE SR 3.3.7.2 REQtiuIeMENT A CFT is performed once e Ir~92-d1 on each required radiation (contnued)monitor to ensure the entire channel will perform the intended function.

This test verifies the capability of the instrumentation to provide the CRVS actuation. The CRVS pressurization system actuation relays are directly actuated by the CRVS atmosphere intake radiation monitors. This signal is not processed through the SSPS, but goes directly to the CRVS actuation relays. The pressurization system is also actuated by Phase A, however, this signal is processed via the SSPS and the testing of the associated relays is performed via SR

~ j known reliability of t oEni oring equipment a% as been shown to SR 3.3.7.3 SR 3.3.7.3 is the performance of an ACTUATION LOGIC TEST. This test verifies the signal path to the Master Relay Coil. Although there are no "Master Relays" as in the SSPS, this surveillance was maintained to preserve the format of the standard specification. The surveillance is intended to ensure that the complete logic is tested for, the function. Since the radiation monitors directly actuate the actuation relays, this test is performed evsy.92ydaas part of the performance of SR 3.3.7.2.

SR 3.3.7.4 SR 3.3.7.4 is the performance of a MASTER RELAY TEST. This test energizes the Master Relay and verifies the actuation signal injected into the Slave Relays. Although there are no "Master Relays" as in the SSPS, this surveillance was maintained to preserve the format of the standard specification. The surveillance is intended to ensure that the complete logic is tested for the function. Since the radiation monitors directly actuate the actuation relays, this test is performed eya'ej-929 ~ .daW*as part of the performance of SR 3.3.7.2.

SVW SR 3.3.7.5 SR 3.3.7.5 is the performance of a SLAVE RELAY TEST. This test energizes the Slave Relays and verifies actuation of the equipment to the pressurization mode. Although there are no "Slave Relays" as in the SSPS, this surveillance was maintained to preserve the format of the standard specification. The surveillance is intended to ensure that the actuation relays, downstream of the logic, function to actuate the pressurization mode equipment. Since the radiation monitors directly actuate the actuation relays, this test is performed ovaý492-dalas part of the performance of SR 3.3.7.2.

-Jý7k-ZSCA(continued)

DIABLO CANYON -UNITS 1 & 2 Revision 4 radOE3D9.Doc - R4 156

CRVS Actuation Instrumentation B 3.3.7 BASES SURVEILLANCE SR 3.3.7.6 REQUIREMENTS SR 3.3.7.6 is the performance of a TADOT. This test is a che~k of the (continued) Manual Actuation Functions aind ic peofrfnred every 18 11*cnth . Each Manual Actuation Function is tested up to, and including, the master relay coils. In some instances, the test includes actuation of the end device (i.e., pump starts, valve cycles, etc.).

'The Frequency is based on the known reliability of t* Function and the redundaincy aailable, and has been shown to-1e acceptable through oper i~narexre .TThe SR is modified by a Note that excludes verification of setpoints during the TADOT. The Functions tested have no setpoints associated with them.

SR 3.3.7.7 A61-61IANNEL CALIBRATI 3N *3 pofformoid every 1 nnh CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to a measured parameter within the necessary range and accuracy.

j~ji'~,~ik ~1. LThe Fr~equency is bafied on operating experience an conn~sistentwi the typicaullindustpyrefueling cycle.

REFERENCES 1. WCAP-1 3878, "Reliability of Potter & Brumfield MDR Relays",

June 1994.

2. WCAP-1 3900, "Extension of Slave Relay Surveillance Test Intervals", April 1994.

3. License Amendment 184/1 86, January 3, 2006.

DIABLO CANYON - UNITS 1 & 2 Revision 4 radD3867.Doc - R4 157

FBVS Actuation Instrumentation B 3.3.8 BASES ACTIONS A.1.1. A.1.2.1. A.1.2.2. and A.1.3 (continued) Condition A applies to the radiation monitor functions, and the manual function. Condition A applies to the failure of one or more radiation monitor channels, or a single manual channel. If one or more channels or trains are inoperable, movement of recently irradiated fuel may continue for a period of 30 days. If movement of recently irradiated fuel continues, an appropriate portable continuous monitor with the same setpoint, or an individual qualified in radiation protection procedures with a dose rate monitoring device must be in the spent fuel pool area immediately and, one FBVS train must be placed in the Iodine Removal mode of operation immediately. This effectively accomplishes the actuation instrumentation function and places the area in a conservative mode of operation or provides appropriate monitoring for continued fuel movement.

B.1 Condition B applies when the Required Action and associated Completion Time for Condition A has not been met and recently irradiated fuel assemblies are being moved in the fuel building.

Movement of recently irradiated fuel assemblies in the fuel building must be suspended immediately to eliminate the potential for events that could require FBVS actuation.

SURVEILLANCE A Note has been added to the SR Table to clarify that Table 3.3.8-1 REQUIREMENTS determines which SRs apply to which FBVS Actuation Functions.

SR 3.3.8.1 Performance of the CHANNEL CHECK o;@eve y-2ýbrs ensures that a gross failure of instrumentation has not occurred. A CHANNEL CHECK is normally a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that instrument channels monitoring the same parameter should read approximately the same value. Significant deviations between the two instrument channels could be an indication of excessive instrument drift in one of the channels or of something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.

Agreement criteria are determined by the unit staff, based on a combination of the channel instrument uncertainties, including indication and readability. If a channel is outside the criteria, it may be an indication that the sensor or the signal processing equipment has drifted outside its limit.

(continued)

DIABLO CANYON - UNITS 1 & 2 Revision 4 radOE3D9.Doc - R4 160

FBVS Actuation Instrumentation B 3.3.8 BASES SURVEILLANCE SR 3.3.8.1 (continued)

REURMNS The Frequ cy is b e o er *n ex ;i e:nce that d Rns~trates channe I i re. he CHANNEL CHECK suppjl~ nsls

~lure kforma, u mo frequent, checks of channels dun~ normal

-. operational use of the displays associated with the LCO required channelýs. -K SR 3.3.8.2 A CFT is performed-ee viý 2d on each required channel to ensure the entire channel will perform the intended function. This test verifies the capability of the instrumentation to provide the FBACS actu~ation. The Freque ncy o days is based on the nown reliabili

~k ~. - bo hetmonitoring uipment and has been shown ktbe acceptable

ýthrough operati rg experience.r SR 3.3.8.3 - Not used SR 3.3.8.4 SR 3.3.8.4 is the performance of a TADOT. This test is a cherPk of the manual actuation functions and is peirfcrmed evcr', 18 mnth3 'Each manual actuation function is tested up to, and including, the master relay coils. In some instances, the test includes actuation of the end anCý,9_f 4,, (J_ -- 4h, of setpoints during the TADOT. The Functions tested have no setpoints associated with them.

SR 3.3.8.5 A CHAISINEL CACIB~MAT Ii'0is perrormaid every M~mmoi.9-10 CHANNEL CALIBRATION is a complete check of the instrument loop, including the sensor. The test verifies that the channel responds to a measured parameter within the necessary range and accuracy. :e In k..rA, ----- rFequen~cyis b'ased~ oeaigxprence and is-onsistent wvith th~e Itvr~ical industry ref~eling cycle.

REFERENCES 1. 10CFR100.11.

2. License Amendment 184/1 86, January 3, 2006.

3. PG&E Letter DCL-05-124 DIABLO CANYON - UNITS 1 & 2 Revision 4 radD3867. Doc - R4 161

RCS Pressure, Temperature, and Flow DNB Limits B 3.4.1 BASES (continued)

SURVEILLANCE SR 3.4. 1.1 REQUIREMENTS Since Required Acýt n A1 allows a Completion Time of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to restore parame s that are not within limits, the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Surveillance Frequency f pressurizer pressure is sufficient ensure the prsue can be re red to a normal operation se state conditionf ilowing TneL&k I -.

load changes and other expected transi operations. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months interval has been shown by operatin rctcetob sffiett regularly assess for potential degroation and to verify operation is SR 3.4.1.2 Since Require) Action A. 1 allows a Completion Time of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to Srestore para eters that are not within limits, the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Surveillance Freque, for RCS average temperature is sufficient to ensure the tempe ature can be restored to ýamal operation, steady state condition following load change and other expected tran .~t operations. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months mnt al has been shown by oj:5erating practice to be sufficient to egularly assess for potential degradation and to verify operation is %rtnsafetyaay~ tion SR 3.4.1.3

he 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months rveinace 1 Frqec o hniae Cýota flowl rate is pe. red using t installed flow instrumen~t ýn Tohie 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

interval as been sho by operating practice tolb sufficient to regul ly assess po ntial degradation and to verify leaonwtu saf analysis assumptions. TThe term "indicated RCS total flow" is used to distinguish between the "measured RCS total flow" determined in SR 3.4.1.4.

SR 3.4.1.4 SR 3.4.1.4 has two surveillance requirements, one for the CHANNEL CALIBRATION of the RCS flow indicators and the other for measurement of RCS total flow rate. Measurement of RCS total flow rate by performance of a peionfwcarmticor ob using the cold leg elbow tap methodology efnee every 24-mianthg-allows the installed RCS flow instrumentation to be normalized and verifies the actual RCS flow rate is greater than or equal to the minimum required RCS flow rate.

(continued)

DIABLO CANYON - UNITS 1 & 2 Revision 4 rad2EC3D. Doc - R4D 5

RCS Pressure, Temperature, and Flow DNB Limits B83.4.1 BASES SR 3.4.1.4 (continued)

REQUIREMENTS The second part of this surveillance is the routine CHANNEL CALIBRATION of the RCS flow indication instrumentation. The routine calibration of the flow instrumentation ensures that the channels are within the necessary range and accuracy for proper flow indication.ThruieCHN LCAIATO f1 C lz 1vxsek. I, Th rq co rotmHN for the 4months eLaLsBrAmet of RSthe flow idcto a

instrumentation ispobased onoperatying exp aferiec and consisdtaenc wit th rfueingcyce.w tyica inusty REFERENCES 1. FSAR, Section 15.

2. Diablo Canyon Power Plant Unit 1 Cycle 9 Reload Safety Evaluation, August 1995.

3. Diablo Canyon Power Plant Unit 2 Cycle 8 Reload Safety Evaluation, Rev.1, April 1996.

4. FSAR, Table 4.1-1.

5. WCAP-1 5113, Revision 1, "RCS Flow Measurement Using Elbow Tap Methodology at Diablo Canyon Units 1 and 2," April, 2002.

DIABLO CANYON - UNITS 1 & 2 Revision 4 rad2EC3D.Doc - R4D 6

RCS Minimum Temperature for Criticality B 3.4.2 BASES SURVEILLANCE SR 3.4.2.1 (continued)

REQUIREMENTS RCS loop average temperature is required to be verified at or above 541'F ha, (The SR-to verify RCS loop average eemperatures eery 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months takes into account indiýdtions andd ala=

mperathat e cnti n uou#"avai [able to the ýerýý)r in the control ntrol re j ;t t other routine survtei ce s whichý re andturýsconsis en 0 and room hP, typicall erformed o e per shift. In addit4io, olýeratn,ý rained t istent

/aef ..rained and will be sensitive to RC emperature during a roach to criticality ensure that the n nimum temperature for criticality is met as criticality

,,-is approached.

REFERENCES 1. FSAR, Chapter 15.

DIABLO CANYON - UNITS 1 & 2 Revision 4 rad2EC3D.Doc - R4D 10

RCS P/T Limits B 3.4.3 BASES (continued)

SURVEILLANCE SR 3.4.3. 1 Verification that operation is within the PTLR limits is required e~very*--

20 owdte~when RCS pressure and temperature conditions are L-ASC4 k. -RCS status. Al , since tempera re rate of chana rimits are Surveillance for heatup, cooldown, or ISLH testing may be discontinued when the definition given in the relevant plant procedure for ending the activity is satisfied.

This SR is modified by a Note that only requires this SR to be performed during system heatup, cooldown, and ISLH testing. No SR is given for criticality operations because LCO 3.4.2 contains a more restrictive requirement.

REFERENCES 1. Not Used

2. 10 CFR 50, Appendix G.

3. ASME, Boiler and Pressure Vessel Code,Section XI, Appendix E.

4. NRC Generic Letter 96-03, "Relocation of the Pressure Temperature Curves and Low Temperature Overpressure Protection System Limits," January 31, 1996.

DIABLO CANYON - UNITS 1 & 2 Revision 4 rad2EC3D.Doc - R4D 16

RCS Loops - MODES 1 and 2 B 3.4.4 BASES APPLICABILITY Operation in other MODES is covered by:

(continued) LCO 3.4.5, "RCS Loops - MODE 3";

LCO 3.4.6, "RCS Loops - MODE 4";

LCO 3.4.7, "RCS Loops - MODE 5, Loops Filled";

LCO 3.4.8, "RCS Loops - MODE 5, Loops Not Filled";

LCO 3.9.5, "Residual Heat Removal (RHR) and Coolant Circulation -

High Water Level" (MODE 6); and LCO 3.9.6, "Residual Heat Removal (RHR) and Coolant Circulation -

Low Water Level" (MODE 6).

ACTIONS A.1 If the requirements of the LCO are not met, the Required Action is to reduce power and bring the plant to MODE 3. This lowers power level and thus reduces the core heat removal needs and minimizes the possibility of violating DNB limits.

The Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging safety systems.

SR 3.4.4.1 REQUIREMENTS -oe+rr that each RCS loop is.in This SR requires verification eyev operation. Verification may include flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providin heat removal while maintaining the margin to DNB. FThe Fre~quen~cyf 12 D\sv*-I-- issuffciet cpsdering other indicaý50ns and alar js-al able to hour the operator in the 6ontrol room to mopfo-r RCS loop p_ rormance.

REFERENCES 1. FSAR, Section 15.

DIABLO CANYON - UNITS 1 & 2 Revision 4 rad2EC3D. Doc - R4D 19

RCS Loops - MODE 3 B 3.4.5 BASES (continued)

SURVEILLANCE SR 3.4.5.1 REQUIREMENTS This SR requires verification every 4944e~a that the required loops are in operation. Verification may include flow rate, temperature, or pump status monitoring, which help ensure that forced flow is -

-r ~~ ....... ~ on'si ering o er i cations and alw s available Xle opertri

- the control arooo monitor RCS Idop performance.r SR 3.4.5.2 SR 3.4.5.2 requires verification of SG OPERABILITY. SG OPERABILITY is verified by ensuring that the secondary side narrow range water level is Ž!15% for required RCS loops. If the SG secondary side narrow range water level is < 15%, the tubes may become uncovered and the associated loop may not be capable of providing the heat sink for removal of the decay heat. IThe 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months A (available Frequencyin~e nsidered is c~control adeqyate in view of ot~hfr indi~catiorlsý roomlt alert te oper or&to a loss of SG level.

SR 3.4.5.3 Verification that the required RCPs are OPERABLE ensures that safety analyses limits are met. The requirement also ensures that an additional RCP can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is 27:-V\ 41 performed by verifying proper breaker alignment and power availability to the required RCPs.

REFERENCES None.

DIABLO CANYON - UNITS 1 & 2 Revision 4 rad2EC3D.Doc - R4D 24

RCS Loops - MODE 4 B 3.4.6 BASES ACTIONS A.1 and A.2 (continued)

If one required RHR loop is OPERABLE and in operation and there are no RCS loops OPERABLE, an inoperable RCS loop or RHR loop must be restored to OPERABLE status to provide a redundant means for decay heat removal.

If the parameters that are outside the limits cannot be restored, the unit must be brought to MODE 5 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Bringing the unit to MODE 5 is a conservative action with regard to decay heat removal.

With only one RHR loop OPERABLE, redundancy for decay heat removal is lost and, in the event of a loss of the remaining RHR loop, it would be safer to initiate that loss from MODE 5 (*!200 0F) rather than MODE 4 (> 200OF to < 3500F). The Completion Time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is a reasonable time, based on operating experience, to reach MODE 5 from MODE 4 in an orderly manner and without challenging plant systems.

B.1 and B.2 If no loop is OPERABLE or in operation, except during conditions permitted by Note 1 in the LCO section, all operations involving introduction of coolant into the RCS with boron concentration less than required to meet the minimum SDM of LCO 3.1.1 must be suspended and action to restore one RCS or RHR loop to OPERABLE status and operation must be initiated.

Boron dilution requires forced RCS circulation from at least one RCP for proper mixing, so that an inadvertent criticality may be prevented.

Suspending the introduction of coolant into the RCS with boron concentration less than required to meet the minimum SDM of LCO 3.1.1 is required to assure continued safe operation. With coolant added without forced circulation, unmixed coolant could be introduced to the core, however coolant added with boron concentration meeting the minimum SDM maintains acceptable margin to subcritical operations. The immediate Completion Times reflect the importance of maintaining operation for decay heat removal. The action to restore must be continued until one loop is restored to OPERABLE status and operation.

SURVEILLANCE SR 3.4.6.1 REQUIREMENTS This SR requires verification e ep,'4-2-hýithat one RCS loop or RHR loop is in operation. Verification may include flow rate, temperature, or pump status monitoring, which help ensure that forced T~~nse ~ .....- considering other i*nda)cations and alarms avifde totele~ ~ tri (continued)

DIABLO CANYON -UNITS 1 & 2 Revision 4 rad2EC3D.Doc - R4D 28

RCS Loops - MODE 4 B 3.4.6 BASES SURVEILLANCE S R 3.4.6.2 W~"j SR 3.4.6.2 requires verification of SG OPERABILITY. SG (continued) OPERABILITY is verified by ensuring that the secondary side narrow range water level is Ž: 15%. If the SG secondary side narrow range water level is < 15%, the tubes may become uncovered and the associated loop may not be capable of providing the heat sink Sconsid~ered adequate in vi w of other' idicaj availablite, SR 3.4.6.3 Verification that the required pump is OPERABLE ensures that an additional RCS or RHR pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation.

Verification is performed by verifying proper breaker alignment nda

~ ... ~considered reasonable in view of otw administraý control REFERENCES None.

DIABLO CANYON - UNITS 1 & 2 Revision 4 rad2EC3D. Doc - R4D 29

RCS Loops - MODE 5, Loops Filled B 3.4.7 BASES (continued)

ACTIONS A.1 and A.2 If one RHR loop is inoperable and the required SGs have secondary side water levels < 15%, redundancy for heat removal is lost. Action must be initiated immediately to restore a second RHR loop to OPERABLE status or to restore the required SG secondary side water levels. Either Required Action A.1 or Required Action A.2 will restore redundant heat removal paths. The immediate Completion Time reflects the importance of maintaining the availability of two paths for heat removal.

BA1 and B.2 If no RHR loop is in operation, except during conditions permitted by Notes 1 and 4, or if no loop is OPERABLE, all operations involving introduction of coolant into the RCS with boron concentration less than required to meet the minimum SDM of LCO 3.1.1 must be suspended and action to restore one RHR loop to OPERABLE status and operation must be initiated. To prevent inadvertent criticality during a boron dilution, forced circulation from at least one RHR pump is required to provide proper mixing and preserve the margin to criticality in this type of operation. Suspending the introduction of coolant into the RCS with boron concentration less than required to meet the minimum SDM of LCO 3.1.1 is required to assure continued safe operation. With coolant added without forced circulation, unmixed coolant could be introduced to the core, however coolant added with boron concentration meeting the minimum SDM maintains acceptable margin to subcritical operations. The immediate Completion Times reflect the importance of maintaining operation for heat removal.

SURVEILLANCE SR 3.4.7.1 REQUIREMENTS This SR requires verification e#efy 42 hetn that the required loop is in operation. Verification may include flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providdin hheat Lre~moval, he F quency o ors is syficient considern 0ohr indications a alarms available to th operator in the crallrol room to Imonitor RH loop performance.

(continued)

DIABLO CANYON - UNITS 1 & 2 Revision 4 rad2EC3D.Doc - R4D 34

RCS Loops - MODE 5, Loops Filled B 3.4.7 BASES SURVEILLANCE SR 3.4.7.2 REQUIREMENTS Verifying that at least two SGs are OPERABLE by ensuring their (continued) secondary side narrow range water levels are Ž! 15% ensures an alternate decay heat removal method via natural circulation in the event that the second RHR loop is not OPERABLE. If both RHR loops FLk.~requency -iscogsidered adequate in view~o other idica:~6 SR 3.4.7.3 Verification that a second RHR pump is OPERABLE ensures that an additional pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation. Verification is performed by verifying proper breaker alignment and power available to the RHR pump. If secondary side water level is Ž! 15% in at I

,V- See3 - considered reason pble in view of o!th administrative ntros REFERENCES 1. NRC Information Notice 95-35, "Degraded Ability of Steam Generators to Remove Decay Heat by Natural Circulation."

2. AR A0582812.

DIABLO CANYON - UNITS 1 & 2 Revision 4 rad2EC3D.Doc - R4D 35

RCS Loops - MODE 5, Loops Not Filled B 3.4.8 BASES ACTIONS B.1 and B.2 (continued)

If no required RHR loops are OPERABLE or in operation, except during conditions permitted by Note 1, all operations involving introduction of coolant into the RCS with boron concentration less than required to meet the minimum SOM of LCO 3.1.1 must be suspended and action must be initiated immediately to restore an RHR loop to OPERABLE status and operation. Boron dilution requires forced circulation from at least one RHR pump for proper mixing so that inadvertent criticality can be prevented. Suspending the introduction of coolant into the RCS with boron concentration less than required to meet the minimum SDM of LCO 3.1.1 is required to assure continued safe operation. With coolant added without forced circulation, unmixed coolant could be introduced to the core, however coolant added with boron concentration meeting the minimum SDM maintains acceptable margin to subcritical operations. The immediate Completion Time reflects the importance of maintaining operation for heat removal. The action to restore must continue until one loop is restored to OPERABLE status and operation.

SURVEILLANCE SR 3.4.8.1 REQUIREMENTS This SR requires verification eive:12 -heI that one loop is in operation. Verification may include flow rate, temperature, or pump status monitoring, which help ensure that forced flow is providing heat SR 3.4.8.2 Verification that the required number of pumps are OPERABLE ensures that additional pumps can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation.

Verification is performed by verifying proper breaker alignment and power available to the required pumpJs) The Frequencyof 7 days is

-&considered ýrpsonable in view of 9jher administrativ, 6ntrols available alhas been shown tdbe acceptable by operating iexgeiece.j REFERENCES 1. AR A0582812.

DIABLO CANYON - UNITS 1 & 2 Revision 4 rad2EC3D.Doc - R4D 38

Pressurizer B 3.4.9 BASES ACTIONS 0.1 and 0.2 (continued) If one required group of pressurizer heaters is inoperable and cannot be restored in the allowed Completion Time of Required Action 8.1, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.4.9. 1 This SR requires that during steady state operation, pressurizer level is maintained below the nominal upper limit to provide a minimum space for a steam bubble. The Surveillance is performed by observing 3~4 -bshown by operatingpactice to be sufficient to regularly a ess level pressurizer. Alarms are also available for early detection of abnormal SR 3.4.9.2 The SR is satisfied when the power supplies are demonstrated to be capable of producing the minimum power and the associated pressurizer heaters are verified to be at their design rating. This may be done by testing the power supply output and by performing an electrical check on heater element continuity and resistance. hee Frequency of Z4 months is considere~..adequate to detej hater degradatg'and has been shown byoperating expe rince to be Laccept~ab'le.

SR 3.4.9.3 This SR demonstrates that the heaters can be manually transferred from the normal to the emergency power suppl and energized. e

'3--b Frequency of 2,4-¶bnths is based n a typical fo~ cycle and 5

ýconsistent v~iti similar venifica~n of emergjerilcy power Wplies.

REFERENCES 1. FSAR, Section 15.

2. NUREG-0737, November 1980.

DIABLO CANYON - UNITS 1 & 2 Revision 4 rad2EC3D.Doc - R4D 42

Pressurizer PORVs B 3.4.11 BASES ACTIONS G1 . n .

(continued) G1 . n .

If the Required Actions of Condition F are not met, then the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems. In MODES 4, 5, and 6 with the reactor vessel head closure bolts not fully de-tensioned, maintaining Class I PORV OPERABILITY is required by LCO 3.4.12.

SURVEILLANCE SR 3.4.11.1 REQUIREMENTS Block valve cycling verifies that the valve(s) can be closed if needed.

CThe basis ~the Frequency of 92,0es is the ASMfr-OM Code,1

~ - SubsectiNiISTCSC (Ref.f). 3:

The Note modifies this SR by stating that it is not required to be performed with the block valve closed in accordance with the Required Action of this LCO. Opening the block valve in this condition increases the risk of an unisolable leak from the RCS since the PORV is already inoperable.

SR 3.4.11.2 SR 3.4.11.2 requires a complete cycle of each PORV. Operating a PORV through one complete cycle ensures that the PORV can be manually actuated for mitigation of an SGTR. Operating experience has shown that these valves usually pass the surveillance when performed at the required Inservice Testing Program frequency. The frequency is acceptable form a reliability standpoint.

The Note modifies this SR to allow entry into an operation in Mode 3 prior to performing the SR. This allows the surveillance to be performed in MODE 3 or 4.

The Note that modified this SR to allow entry into and operation in MODE 3 prior to performing the SR. This allows the test to be performed in MODE 3 under operating temperature and pressure conditions, prior to entering MODE 1 or 2. In accordance with Reference 4, administrative controls require this test be performed in MODE 3 or 4 to adequately simulate operating temperature and pressure effects on PORV operation.

(continued)

DIABLO CANYON -UNITS 1 & 2 Revision 4 rad2EC3D. Doc - R4D 59

Pressurizer PORVs B83.4.11 BASES SURVEILLANCE SR 3.4.11.3 REQUIREMENTS Verifying OPERABILITY of the safety related nitrogen supply for the (continued)

Class I PORVs may be accomplished by:

a. Isolating and venting the normal air supply, and

b. Verifying that any leakage of the Class I backup nitrogen system is within its limits, and

c. Operating the Class I PORVs through one complete cycle of full travel.

Operating the solenoid nitrogen control valves and check valves on the nitrogen supply system and operating the Class I PORVs through one complete cycle of full travel ensures the nitrogen backup su ly for the Class I PORV operates properly whec u on. The -

JvA-Zk~r . -1 Frequency -of24 moj~ s is based on a typical fueling cyclep the Frequency of tl-6Fher Surveillances used t iass eonsrt PORV OPERABILITY.

SR 3.4.11.4 Lv'c~e4 ,~Performance of a COT is required on each required Class I PORV to could verify and, as necessary, adjust its lift setpoint. PORV actuation deprssuizethe RCS and is not required.

SR 3.4.11.5 Performance of a CHANNEL CALIBRATION on each required Class I PORV actuation channel is required ey~ ~ 4 adjust the whole channel so that it responds and the valve opens within the required range and accuracy to known input.

REFERENCES 1. Not Used.

2. FSAR, Section 15.2.

3. ASME, Code for Operation and Maintenance of Nuclear Power Plants, 2001 Edition including 2002 and 2003 Addenda.

4. Generic Letter 90-06, "Resolution of Generic Issue 70, 'Power-Operated Relief Valve and Block Valve Reliability,' and generic issue 94, 'Additional Low-Temperature Overpressure Protection for Light-Water Reactors,' Pursuant to 10 CFR 50.54(f)," June 25, 1990.

DIABLO CANYON - UNITS 1 & 2 Revision 4 rad2EC3D.Doc - R4D 60

LTOP System B 3.4.12 BASES (continued)

SURVEILLANCE SR 3.4.12. 1. SR 3.4.12.2. and SR 3.4.12.3 REQUIREMENTS To minimize the potential for a low temperature overpressure event by limiting the mass input capability, a maximum of zero SI pumps and one ECCS COP are verified capable of injecting into the RCS and the accumulator discharge isolation valves are verified closed and their breakers open. Verification that each accumulator is isolated is only required when accumulator isolation is required as stated in Note 1 to the LCO. Further, CCP-3 must be realigned for LTOP operation during LTOP conditions.

The SI pumps and one ECCS COP are rendered incapable of injecting I into the RCS for example, through opening the DC knife switch supplying the pumps breakers control power or removing the power from the pumps by racking the breakers out under administrative control or by isolating the discharge of the pump by closed isolation valves with power removed from the operators or by a manual isolation valve secured in the closed position.

An alternate method of providing low temperature overpressure protection may be employed to prevent a pump start that could result in an injection into the RCS. An inoperable pump may be energized for test or for accumulator fill provided the discharge of the pump is isolated from the RCS by closed isolation valve(s) with power removed frmtheRCalven of Ž2.07 j squ marinches isoproven OPERABLE bylei shecursed pointheopn. p~h ion. ous!55 n, b.. Oncsderi~eveý&ry ndiaysfomtervn pathsý a tgh, oalertato is lh otoke, seale, oterwisyte secured inth)jo tor open positiont.) eoe

-LV-%s'e-ek 0. prssrier RSaft vavenof207suroen manwaaloeis tis crvnOEAtLEgory.

verifyng itsopen (continued)r-DIABLO CANYON - UNITS 1 & 2 Revision 4 rad2EC3D.Doc - R4D 72

LTOP System B 3.4.12 BASES SURVEILLANCE SR 3.4.12.5 (continued)

REQU IREMENTS Any passive vent path arrangement need only be open when required to be OPERABLE. This Surveillance is required to be performed if the vent is being used to satisfy the pressure relief requirements of LCO 3:4.12.b.

SR 3.4.12.6 The Class I PORV block valve must be verified open every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to provide the flow path for each required Class I PORV to perform its function when actuated. The valve must be remotely verified open in the main control room. This surveillance is performed if the PORV satisfies the [CO.

The block valve is a remotely controlled, motor operated valve. The power to the valve operator is not required removed, and the manual operator is not required locked in the inactive position. Thus, the block valve can be closed in the event the PORV develops excessive leakage or does not close (sticks open) after relieving an overpressure situation.

4administrativ ontrols available to the o~p tor in the contro som, SR 3.4.12.7 Not Used SR 3.4.12.8 The SR Note states that the SR is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after decreasing any RCS cold leg temperature to:* LTOP arming temperature specified in the PTLR.

)I~e~4L~ I The SR may be performed prior to reaching :< LTOP arming temperature and must be current (within 31 days) to meet this surveillance requirement. If not performed prior to reaching LTOP temperature, the test must be performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering the LTOP MODES. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowance considers the unlikelihood of a low temperature overpressure event during this time.

Following the initial SR, while remaining in the Applicable [TOP MODE, the SR will be performed AwF 4-ayflhereafter on each required Class I PORV to verify and, as necessary, adjust its lift setpoint. The COT will verify the setpoint is within the PTLR allowed limits in the PTLR. PORV actuation could depressurize the RCS and is not required.

-- t* (continued)

DIABLO CANYON - UNITS 1 & 2 Revision 4 rad4494C.Doc - R4D 66

LTOP System B 3.4.12 BASES SURVEILLANCE SR 3.4.12.9 (contnued)Performance of a CHANNEL CALIBRATION on each required Class I (cotined)PORV actuation channel is required every 24 mo~thro adjust the

~~ whole channel so that it responds and the valve opens within the Srequired range and accuracy to known input.

REFERENCES 1. 10 CER 50, Appendix G.

2. Generic Letter 88-11.

3. Not Used

4. FSAR, Chapter 5.

5. 10 CFR 50, Section 50.46.

6. 10 CFR 50, Appendix K.

7. Generic Letter 90-06.

8. Not Used

9. ASME Code Case N-514.

10. AR A0625429

11. AR A0589860 DIABLO CANYON - UNITS 1 & 2 Revision 4 rad2EC3D.Doc - R4D 74

RCS Operational LEAKAGE B 3.4.13 BASES SURVEILLANCE SR 3.4.13.1 (continued)

REQUIREMENTS An early warning of pressure boundary LEAKAGE or unidentified LEAKAGE is provided by the automatic systems that monitor the containment atmosphere radioactivity and by the containment structure sump level and flow monitoring system. It should be noted that LEAKAGE past seals, gaskets or CRDM canopy seal welds is not pressure boundary LEAKAGE. These leakage detection systems are specified in LCO 3.4.15, "RCS Leakage Detection Instrumentation."

Note 2 states that this SR is not applicable to primary to secondary LEAKAGE because LEAKAGE of 150 gallons per day cannot be measured accurately by an RCS water inventory balance.

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Fr ~uency is a reasonable interv. oEtrenýd LEK(Eand j--Je-4 ( recognizes t~i f al p ::dtcinventin rerac of accidents The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency after steady state operation has een achieved provides for those situations following a transient such that the 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months plus extension allowed by SR 3.0.2 would be exceeded. Under these circumstances, the SR would be due within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after steady state operation has been reestablished adwp-SR 3.4.`13.2 This SR verifies that primary to secondary LEAKAGE is less than or equal to 150 gallons per day through any one SG. Satisfying the primary to secondary LEAKAGE limit ensures that the operational LEAKAGE performance criterion in the Steam Generator Program is met. If this SR is not met, compliance with LCO 3.4.17, "Steam Generator Tube Integrity," should be evaluated. The 150 gallons per day limit is measured at room temperature as described in Reference 8.

The operational LEAKAGE rate limit applies to LEAKAGE through any one SG. If it is not practical to assign the LEAKAGE to an individual SG, all the primary to secondary LEAKAGE should be conservatively assumed to be from one SG.

The Surveillance is modified by a Note which states that the Surveillance is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation. For RCS primary to secondary LEAKAGE determination, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows.

prmr tose dry eEua ;cognze the imý aneOf early monitors or radiochemical grab sampling in accordance with the EPRI guidelines (Ref. 8).

(continued)

DIABLO CANYON -UNITS 1 & 2 Revision 4 rad4494C.Doc - R4D 74

RCS PIV Leakage B 3.4.14 BASES (continued)

SURVEILLANCE SR 3.4.14.1 REQUIREMENTS Performance of leakage testing on each RCS Ply or isolation valve used to satisfy Required Action A.1 and Required Action A.2 is required to verify that leakage is below the specified limit and to identify each leaking valve. The leakage limit of 0.5 gpm per inch of nominal valve diameter up to 5 gpm maximum applies to each valve.

Leakage testing requires a stable pressure condition.

For the two PIVs in series, the leakage requirement applies to each valve individually and not to the combined leakage across both valves.

This method results in testing each valve separately. If the PIVs are not individually leakage tested, one valve may have failed completely and not be detected if the other valve in series meets the leakage requirement. In this situation, the protection provided by redundant valves would be lost.

~l$A~fTesting is to be performed every 24 months, a typical refueling cycle.

The 24 month Fre uency is consistent with 10 CFR 50.55a(g) (Ref. 8) as contained i e Inservice Testing Pro ram, is within frequp~cy allowed by e ASME OM Code, Suobs tion ISTC (Ref. 7)",Knd is based on the need to perform such 9urveillances under fe conditions that apply during an outage and t e potential for an unplanned transient if the Surveillance were performed with the reactor at power.

Test pressures less than 2235 psig but greater than 150 psig are allowed for valves where higher pressures would tend to diminish leakage channel opening. Observed leakage shall be adjusted for actual pressure to 2235 psig assuming the leakage to be directly proportional to pressure differential to the one half power.

(continued)

DIABLO CANYON -UNITS 1 & 2 Revision 4 rad2EC3D.Doc - R4D 88

RCS PIV Leakage B83.4.14 BASES SURVEILLANCE SR 3.4.14.1 (continued)

REQUIREMENTS In addition, testing must be performed once after the valve has been opened by flow or exercised to ensure tight reseating. PIVs disturbed in the performance of this Surveillance should also be tested unless documentation shows that an infinite testing loop cannot practically be avoided. Testing must be performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the valve has been reseated. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is a reasonable and practical time limit for performing this test after opening or reseating a valve. For check valves 8956A-D, evaluation determined that trickle forward flow of 10 gpm or less through these valves would not challenge their seating integrity and therefore do not require testing to ensure tight reseating (Reference9).

The leakage limit is to be met at the RCS pressure associated with MODES 1 and 2. This permits leakage testing at high differential pressures with stable conditions not possible in the MODES with lower pressures.

Entry into MODES 3 and 4 is allowed to establish the necessary differential pressures and stable conditions to allow for performance of this Surveillance. In addition, this Surveillance is not required to be performed on the RHR System when the RHR System is aligned to the RCS in the shutdown cooling mode of operation. PIVs contained in the RHR shutdown cooling flow path must be leakage rate tested after RHR is secured and stable unit conditions and the necessary differential pressures are established.

!23y0eV U efrequently tha months as these valves are moaperated with s ystem high pressure alarms.

SR 3.4.14.2 and 3.4.14.3 Not Used (continued)

DIABLO CANYON - UNITS 1 & 2 Revision 4 rad2EC3D. Doc - R4D 89

RCS Leakage Detection Instrumentation 8 3.4.15 BASES ACTIONS C.1.1, C.1.2, C.2.1, and 0.2.2 (continued) analyzed or water inventory balances, in accordance with SR 3.4.13.1, must be performed to provide alternate periodic information.

The follow-up Required Action is to restore either of the inoperable required monitors to OPERABLE status within 30 days to regain the intended leakage detection diversity. The 30 day Completion Time ensures that the plant will not be operated in a reduced configuration for a lengthy time period.

DA1 and D.2 If a Required Action of Condition A, B, or C, cannot be met, the plant must be brought to a MODE in which the requirement does not apply.

To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

E.1 With all required monitors inoperable, (LCO a, b, and c) no means of monitoring leakage are available, and immediate plant shutdown in accordance with LCO 3.0.3 is required.

SURVEILLANCE SR 3.4.15. 1 REQUIREMENTS SR 3.4.15.1 requires the performance of a CHANNEL CHECK of the required containment atmosphere radioactivity monitors. The check gives reasonable-c-onfidence that the channels are op~erating proerly.0 C4~ 4 L4-~ he Frequenp of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is based,9o instrument reliabiity and is SR 3.4.15.2 SR 3.4.15.2 requires the performance of a CHANNEL FUNCTIONAL TEST (CFT) on the required containment atmosphere radioactivity monitors. The test ensures that the monitors can perform their manne buding alarm functions. The desired in the3-ays considers Tr-equ-e-ncy-o tfnction instriujxent reliabilit , and operating exeiec a's shown that it is pro p'r for detect~&g degrad ain.

SR 3.4.15.3. SR 3.4.15.4, and SR 3.4.15.5 These SRs require the performance of a CHANNEL CALIBRATION for each of the RCS leakage detection instrumentation channels. The calibration verifies the accuracy of the instrument string, including the instruments located inside containment. I;harr i aE- ee4mf 2tr*h (continued)

DIABLO CANYON -UNITS 1 & 2 Revision 4 rad2EC3D.Doc - R4D 95

RCS Leakage Detection Instrumentation B83.4.15 BASES SURVEILLANCE SR 3.4.15.3. SR 3.4.15.4. and SR 3.4.15.5 (continued)

REQ UIREM ENTS (e5xcept for thp required containment atmosphere particulate and gaseousjz6ioactivity monitors whi have a frequency of 8 month is consrstent with refueling cl`and considers ch anqaeliability.

Again, operating experie'nce ras proven that this Frequeac is acce Wle.

REFERENCES 1. 10 CFR 50, Appendix A, Section IV, GDC 30.

2. Regulatory Guide 1.45.

3. FSAR, Section 5.2.7.

4. NUREG-609, "Asymmetric Blowdown Loads on PWR Primary System," 1981.

5. Generic Letter 84-04, "Safety Evaluation of Westinghouse Topical Reports Dealing with Elimination of Postulated Breaks in PWR Primary Main Loops."

6. ESAR, Section 3.6B3.

DIABLO CANYON - UNITS 1 & 2 Revision 4 rad2EC3D.Doc - R4D 96

RCS Specific Activity B 3.4.16 BASES ACTIONS B.1 (continued) With the DOSE EQUIVALENT XE-133 in excess of the allowed limit, DOSE EQUIVALENT XE-133 must be restored to within limits within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . The allowed Completion Time of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is acceptable since it is expected that, if there were a noble gas spike, the normal coolant noble gas concentration would be restored within this time period. Also, there is a low probability of a SLB or SGTR occurring during this time period.

A Note permits the use of the provisions of LCO 3.0.4c. This allowance permits entry into the applicable MODE(S), relying on Required Action B.1 while the DOSE EQUIVALENT XE-133 LCO limit is not met. This allowance is acceptable due to the significant conservatism incorporated into the specific activity limit, the low probability of an event which is limiting due to exceeding this limit, and the ability to restore transient-specific activity excursions while the plant remains at, or proceeds to, power operation.

CA1 and C.2 If the Required Action and the associated Completion Time of Condition A or B is not met, or if the DOSE EQUIVALENT 1-131 is >

60.0 pLCi/gm, the reactor must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.4.16.1 REQUIREMENTS SR 3.4.16.1 requires performing a gamma isotopic analysis as a measure of the noble gas specific activity of the reactor coolant eF~e-as-4are eeyeaeVd This measurement is the sum of the degassed gamma activities and the gaseous gamma activities in the sample taken. This Surveillance provides an indication of any increase in the noble gas specific activity.

Trending the results of this Surveillance allows proper remedial action to be taken before reaching the LCO limit under normal operating

~ ~yconditiomOThe7 day Frequency considers the unlikelihood of a grooss If a specific noble gas nuclide listed in the definition of DOSE EQUIVALENT XE-133 in Specification 1.1, "Definitions," is not detected, it should be assumed to be present at the minimum detectable activity.

(continued)

DIABLO CANYON - UNITS 1 & 2 Revision 4 rad2EC3D.Doc - R4D 100

RCS Specific Activity B 3.4.16 BASES SURVEILLANCE The definition of DOSE EQUIVALENT XE-I 33 in Specification REQUIREMENTS 1.1, "Definitions," requires that the determination of DOSE (continued) EQUIVALENT XE-I133 shall be performed using the effective dose conversion factors for air submersion listed in Table Ill. 1 of EPA Federal Guidance Report No. 12, 1993, "External Exposure to Radionuclides in Air, Water, and Soil." These dose conversion factors are consistent with the dose conversion factors used in the applicable dose consequence analyses.

The Note modifies this SR to allow entry into and operation in MODE 4, MODE 3, and MODE 2 prior to performing the SR. This allows the Surveillance to be performed in those MODES, prior to entering MODE I.

SR 3.4.16.2 This Surveillance is performed to ensure iodine specific activity remains within the LCO limit during normal operation and following fast power

.i~

tj- Frquenc s ahequn e iodine agsi activn evemo Frequency, between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a power change Ž! 15% RTP within a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months period, is established because the iodine levels peak during this time following iodine spike initiation; samples at other times would provide inaccurate results.

The definition of DOSE EQUIVALENT 1-131 in Specification 1.1, "Definitions," specifies the thyroid dose conversion factors which may be used to determine DOSE EQUIVALENT 1-131. The thyroid dose conversion factors used to determine DOSE EQUIVALENT 1-131 are to be consistent with the dose conversion factors used in the applicable dose consequence analyses, or be conservative with respect to the dose conversion factors used in the applicable dose consequence analyses such that a higher DOSE EQUIVALENT 1-131 is determined.

The Note modifies this SR to allow entry into and operation in MODE 4, MODE 3, and MODE 2 prior to performing the SR. This allows the Surveillance to be performed in those MODES, prior to entering MODE 1.

(continued)

DIABLO CANYON - UNITS 1 & 2 Revision 4 rad2EC3D. Doc - R4D 101

Accumulators B 3.5.1 BASES ACTIONS B.1 (continued) power to the valve, or restore the proper water volume or nitrogen cover pressure ensures that prompt action will be taken to return the inoperable accumulator to OPERABLE status. The Completion Time minimizes the potential for exposure of the plant to a LOCA under these conditions. The 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months allowed to restore an inoperable accumulator to OPERABLE status is justified in WCAP-1 5049-A, Rev 1. (Ref. 7)

CA1 and C.2 If the accumulator cannot be returned to OPERABLE status within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and RCS pressure reduced to

<ý1000 psig within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

D.1 If more than one accumulator is inoperable, the plant is in a condition outside the accident analyses; therefore, LCO 3.0.3 must be entered immediately.

SURVEILLANCE SR 3.5. 1.1 REQUIREMENTS Each accumulator motor operated isolation valve (8808A, B, C, and D) should be verified to be fully open a"F 4.2Ihe. This verification ensures that the accumulators are available for injection and ensures timely discovery if a valve should be less than fully open. If an isolation valve is not fully open, the rate of injection to the RCS would be reduced. Although a motor operated valve position should not change with power removed, a closed valve could result in not SR 3.5.1.2 and SR 3.5.1.3 19 nL4 orated water volume and nitrogen cover pressure F-uei are verified for each accumulator. IIhis Frequency is sufficient to ensure adequate injeq n during a LOiCA. ~ecause of thA static design of the accuolator, a 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Fre <enCy usu Wallows the operator to identi changes before Ii~i s are reached. Operating experience hayshown this Frequency to be appropriate for early deeton and correction of off normal trends.

DIABLO CANYON - UNITS 1 & 2 Revision 4 radC87D1. Doc - R4c 6

Accumulators B 3.5.1 BASES SURVEILLANCE SR 3.5.1.4 (cont.~inued)L.I The boron concentration should be verified to be within required limits (cotined)for each accumulator ev'eiy8-4 d-dasince the static design of the 3~-~roL a ou croccuvfomn mechanims such as in-leakag .Sa'mplingthýe affected accumulator within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a solution volume increase of 5.6% (101 gallon) narrow range indicated level will identify whether in-leakage has caused a reduction in boron concentration to below the required limit. It is not necessary to verify boron concentration if the added water inventory is from the refueling water storage tank (RWST), and the RWST has not been diluted since verifying that its boron concentration satisfies SR 3.5.4.3, because the water contained in the RWST is nominally within the accumulator boron concentration requirements as verified by SR 3.5.4.3. This is consistent with the recommendation of GL 93-05 (Ref. 4).

SR 3.5.1.5 Verification aaQ!dwhtpower is removed from each accumulator isolation valve operator (8808A, B, C, and D) when the RCS pressure is greater than 1000 psig ensures that an active failure could not result in the undetected closure of an accumulator motor operated isolation valve. If this were to occur, only two accumulators would be available for injection given a single failure coincident with a

-10se-ft T This SR allows power to be supplied to the motor operated isolation valves when RCS pressure is less than or equal to 1000 psig, thus allowing the valves to be closed to enable plant shutdown without discharging the accumulators into the RCS.

REFERENCES 1. FSAR, Chapter 6.

2. 10.CFR 50.46.

3. ESAR, Chapter 15.

4. GL 93-05, Item 7.1.

5. DCM S-38A.

6. License Amendment 147/147, May 3, 2001.

7. WCAP-15049-A, Rev 1, April, 1999.

DIABLO CANYON - UNITS 1 & 2 Revision 4 radC87Dl.Doc - R4c 8

ECCS - Operating B 3.5.2 BASES ACTIONS A.1 (continued) single OPERABLE ECCS train is not available, the facility is in a condition outside the accident analysis. Therefore, LCO 3.0.3 must be immediately entered.

Opening the containment recirculation sump strainer system access ports, or lower plenum drain valve (SI-294) without pipe cap or inlet strainer (STR-440) installed, for Unit 1 or sump screen access hatch on the grating for Unit 2 in MODES 1 through 3 is considered to be a condition which is outside the accident analysis. Therefore, LCO 3.0.3 must be immediately entered (Ref. 9.)

B.1 and B.2 If the inoperable trains cannot be returned to OPERABLE status within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 4 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.5.2.1 REQUIREMENTS Verification of proper valve position ensures that the flow path from the ECCS pumps to the RCS is maintained. Valve position is the concern and not indicated position in the control room. Misalignment of these valves could render both ECCS trains inoperable. Securing these valves in position by removal of power ensures that they cannot change position as a result of an active failure or be inadvertently misaligned. The surveillance can be satisfied using indicated position in the control room but may also be satisfied using local observation.

These valves are of the type, described in References 6 and 7, that can disable the function of both ECCS trains and invalidate the accident ana~lyses. A 2hour Frequy cy is considered re3Wnable i~n

-Tv\.e,erck. '3.., -'view of otheradministrative controts that will ensure a rKfspositioned in LCO Note 1, both SI pump flow pDaths kalve is uirw1-ikeiy3 A-s -noted may each be isolated for two hours in MODE 3 by closure of one or more of these valves to perform pressure isolation valve testing.

In addition to the valves listed in SR 3.5.2.1, there are other ECCS related valves that must be appropriately positioned. Improper valve position can affect the ECCS performance required to meet the analysis assumptions. These valves are identified in plant documents and are listed in the following table.

(continued)

DIABLO CANYON -UNITS 1 & 2 Revision 4 radC87Dl.Doc - R4c 17

EGGS - Operating B 3.5.2 BASES SURVEILLANCE SR 3.5.2.2 (continued)

REQUIREMENTS The 31 day Frequgncy is appropriate because the valv~s area operatqd under administr ive control, /ard'an improper valve-psition wouAW

~pd ~only affect a, iingle train. Th' Frequency has been shown to beu

-acceptable through operating experience.

SR 3.5.2.3 With the exception of the operating COP, the EGOS pumps are normally in a standby, nonoperating mode. As such, flow path piping has the potential to develop voids and pockets of entrained gases.

Maintaining the piping from the EGGS pumps to the RCS full of water ensures that the system will perform properly, injecting its full capacity into the RCS upon demand. This will also prevent water hammer, pump cavitation, gas binding, and pumping of non-condensable gas (e.g., air, nitrogen, or hydrogen) into the reactor vessel following an I signal or during shutdown cooling. e 91day Fre~quen~ takes into (co-n-sieration the graduL nature of gas ~cumulation' the EGGS piping and the proceýVal controls go~erning system operation.

The intent of the SR is to assure the EGGS piping is adequately vented. Different means of verification, as alternates to venting the accessible system high points, can be employed to provide this assurance, such as ultrasonic testing the vent lines of the EGGS pump casings and accessible high point vents.

SR 3.5.2.4 Periodic surveillance testing of EGGS pumps to detect gross degradation caused by impeller structural damage or other hydraulic component problems is required by the ASME Code. (Ref. 8) This type of testing may be accomplished by measuring the pump developed head at only one point of the pump characteristic curve.

This verifies both that the measured performance is within an acceptable tolerance of the original pump baseline performance and that the performance at the test flow is within the performance assumed in the plant safety analysis. SRs are specified in the applicable portions of the Inservice Testing Program, which encompasses Subsection ISTB of the ASME Code for Operation and Maintenance of Nuclear Power Plants. (Ref. 8). This section of the ASME Code provides the activities and frequencies necessary to satisfy the requirements.

The following EGGS pumps are required to develop the indicated differential pressure when tested on recirculation flow:

COP Ž! 2400 psid SI pump Ž! 1455 psid RHR pump Ž! 165 psid DIABLO CANYON -UNITS 1 & 2 Revision 4 radO87D. .Doc - R4c 20

ECCS - Operating B 3.5.2 BASES SURVEILLANCE SR 3.5.2.5 and SR 3.5.2.6 REQUIREMENTS These Surveillances demonstrate that each automatic ECOS valve (continued) actuates to the required position on an actual or simulated SI signal and that each ECOS pump starts on receipt of an actual or simulated SI signal. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under eduimnitraiecnrl. acuaio The logic srqec testaedaspr S cutonth thednspervice thestin Srorvelam. es uneyh odtostapl SR 3.t. P T The correct position of throttle/runout valves in the ECOS flow paths is necessary for proper EGOS performance. These manual throttle/runout valves are positioned during flow balancing and have mechanical locks and seals to ensure that the proper positioning for restricted flow to a ruptured cold leg is maintained. The verification of proper position of a throttle/runout valve can be accomplished by confirming the seals have not been altered since the last performance of the flow balance test. Restricting the flow to a ruptured cold leg ensures tha't the other cold legs receive at least the required minimum those stated in SR3.5.2.5 a ýR3526 V-4 SR 3.5.2.8 Periodic inspections of the containment recirculation sump suction inlet ensure that it is unrestricted and stays in pro er o 2:-see potential for an nplanned transient /ihe Surveillance we~j efre sufficient t6 detect abnormal degradation and is confirmed by operating experience.1 Opening the containment recirculation sump strainer system access ports, or lower plenum drain valve (SI-294) without pipe cap or inlet strainer (STR-440) installed, for Unit 1 or sump screen access hatch on the grating for Unit 2 in MODES 1 through 4 is considered to be a condition which is outside the accident analysis. Therefore, LCO 3.0.3 must be immediately entered. (Ref. 9)

DIABLO CANYON - UNITS 1 & 2 Revision 4 radC87D1.Doc - R4c 22

RWST B 3.5.4 BASES ACTIONS A.1 (continued)

The requirement for RWST temperature is to be greater than or equal to the minimum required temperature. The expression "within the required limits", applied to RWST temperature is satisfied when the temperature is greater than or equal to the minimum.

8.1 With the RWST inoperable for reasons other than Condition A (e.g.,

water volume), it must be restored to OPERABLE status within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

In this Condition, neither the ECCS nor the CS System can perform its design function. Therefore, prompt action must be taken to restore the tank to OPERABLE status or to place the plant in a MODE in which the RWST is not required. The short time limit of 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to restore the RWST to OPERABLE status is based on this condition simultaneously affecting redundant trains and that borated water volume can be restored more rapidly than boron concentration or temperature.

CA1 and C.2 If the RWST cannot be returned to OPERABLE status within the associated Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant SURVELLIAsystems.

NCE SR 3.5.4.1 REQUIREM ENTS The RWST borated water temperature should be verified sfet"224ýL

-he be above th inimum assumed in the accident anal ses.

-IvNs-, k /....This Frequency is sufficient to identify a)!rperature ch~1ge that would appro~ac A'e limit and has been shown to be arcceptable through operýino ex erience The SR is modified by a Note that eliminates the requirement to perform this Surveillance when ambient air temperature is above the minimum temperature for the RWST. With ambient air temperature above the minimum temperature, the RWST temperature should not exceed the limit.

(continued)

DIABLO CANYON - UNITS 1 & 2 Revision 4 radC87Dl.Doc - R4c 35

RWST B 3.5.4 BASES SURVEILLA~qCE S R 3.5.4.2 REQUIREME .NTS The RWST water volume should be verified ev'e,'-Y to be above (continued) the required minimum level in order to ensure that a sufficient initial supply is available for ECCS injection and CS System pum o and to support contne ECCS on recirculation. ince the RWST

(~....bvo meis normal stable and the contained volume required is

- (roeced Zb computer alarm, a 7 Frequency is a opriate and

.,as!eeon to be acceptable tirough operating Zpperience.

The RWST water volume is administratively maintained at greater than the SR 3.5.4.2 limit in accordance with STP R-20. I S R 3.5.4.3 The boron concentration of the RWST should be verified to be within the required limits. This SR ensures that the reactor will remain subcritical following a LOCA. Further, it assures that the resulting sump pH will be maintained in an acceptable range so that boron precipitation in the core will not occur and the effect of chloride and caustic stress corrosion on mechanical systems and components will be minimized. v5nce The RvvS I volume is normally staole, a 7 day

-4 4- .sampling ýFrequenc vrfboron co entration is ap pre n REFERENCES 1. FSAR, Chapter 6 and Chapter 15.

2. Surveillance Test Procedure R-20, "Boric Acid Inventory." I DIABLO CANYON - UNITS 1 & 2 Revision 4 radC87Dl .Doc - R4c 36

Seal Injection Flow B 3.5.5 BASES ACTIONS B.1 and B.2 (continued) When the Required Actions cannot be completed within the required Completion Time, a controlled shutdown must be initiated. The Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months for reaching MODE 3 from MODE 1 is a reasonable time for a controlled shutdown, based on operating experience and normal cooldown rates, and does not challenge plant safety systems or operators. Continuing the plant shutdown begun in Required Action B.1, an additional 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is a reasonable time, based on operating experience and normal cooldown rates, to reach MODE 4, where this LCO is no longer applicable.

SURVEILLANCE SR 3.5.5.1 REQUIREMENTS Verification avso 31 dayethat the manual seal injection throttle valves are adjusted to give a hydraulic resistance within the limit ensures proper manual seal injection throttle valve position, and hence

..... nengineering judgment and is cnistent with othrnCCS valve The seal water injection filters can affect the system flow. As differential pressure across the filter increases over the life of the filter element, certain operating adjustments may be made to maintain the RCP seal flow within the allowed limits. The effect on the system flow resulting from valving in a clean standby filter, after having adjusted the system over time, could result in a resistance flow value outside the TS limit. Therefore, instructions are provided that when a filter is removed from or returned to service, that the procedure to ensure flow characteristics of the seal injection water flow path satisfy the accident analysis and TS may need to be performed.

As noted, the Surveillance is to be completed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months after the RCS (pressurizer) pressure has stabilized within the specified pressure limits at nominal operating pressure. The RCS (pressurizer) pressure requirement is specified since this configuration will produce the required pressure conditions necessary to assure that the manual valves are set correctly. The exception is limited to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to ensure that the Surveillance is timely.

REFERENCES 1. FSAR, Chapter 6 and Chapter 15.

2. 10 CFR 50.46.

3. License Amendment 148/148, May 7, 2001.

DIABLO CANYON - UNITS 1 & 2 Revision 4 radC87Dl. Doc - R4c 41

Containment Air Locks B 3.6.2 BASES (continued)

SURVEILLANCE SR 3.6.2.1 REQUIREMENTS Maintaining containment air locks OPERABLE requires compliance with the leakage rate test requirements of the Containment Leakage Rate Testing Program. This SR reflects the leakage rate testing requirements with regard to air lock leakage (Type B leakage tests).

The acceptance criteria were established during initial air lock and containment OPERABILITY testing. The periodic testing requirements verify that the air lock leakage does not exceed the allowed fraction of the overall containment leakage rate. The Frequency is required by the Containment Leakage Rate Testing Program.

The SR has been modified by two Notes. Note 1 states that an inoperable air lock door does not invalidate the previous successful performance of the overall air lock leakage test. This is considered reasonable since either air lock door is capable of providing a fission product barrier in the event of a DBA. Note 2 has been added to this SR requiring the results to be evaluated against the acceptance criteria which is applicable to SR 3.6. 1.1. This ensures that air lock leakage is properly accounted for in determining the combined Type B and C containment leakage rate.

SR 3.6.2.2 The air lock interlock is designed to prevent simultaneous opening of both doors in a single air lock. Since both the inner and outer doors of an air lock are designed to withstand the maximum expected post accident containment pressure, closure of either door will support containment OPERABILITY. Thus, the door interlock feature supports containment OPERABILITY while the air lock is being used for personnel transit in and out of the containment. Periodic testing of this interlock demonstrates that the interlock will function as designed and that simultaneous opening of the inner and outer doors will not intelock give oand thatheint erlocm echanismai n otur normally

~ ~ ct~al n e h nt e c n e t arl c o ri e o nht r n adtequokate giventa thatte interlockismotchallnged duringt useofmathe (cniud DIABLO CAYNUIS1&2Rvsok

~ n 8591D04.da - ~a 1

Containment Isolation Valves B 3.6.3 BASES SURVEILLANCE SR 3.6.3.2 (continued)

REQUIREMENTS environment following a LOCA. Therefore, these vail 42 dt be open for limited periods of time. he day Frequýepcy iss (onsistent E with ofiher containment isol i-on valve recrairements

\discussed in ýý3.6 .3.3.

SR 3.6.3.3 This SR requires verification that each containment isolation manual valve and blind flange located outside containment and not locked, sealed, or otherwise secured and required to be closed during accident conditions is closed. The SR helps to ensure that post accident leakage of radioactive fluids or gases outside of the containment boundary is within design limits. This SR does not require any testing or valve manipulation. Rather, it involves verification, through a system walkdown, which may include the use of local or remote indicators, that those containment isolation valves outside containment and capable of being mis o0 IsolaItio alves ouitside c t.ainment is relatively Te that containment isolation v~a yes that are o~pen u~nder administrative controls are not required to meet the SR during the time the valves are open. This SR does not apply to valves that are locked, sealed, or otherwise secured in a closed position since these were verified to be in the correct position upon locking, sealing, or securing.

The Note applies to valves and blind flanges located in high radiation areas and allows these devices to be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted during MODES 1,2, 3 and 4 for ALARA reasons. Therefore, the probability of misalignment of these containment isolation valves, once they have been verified to be in the proper position, is small.

SR 3.6.3.4 This SR requires verification that each containment isolation manual valve and blind flange located inside containment and not locked, sealed, or otherwise secured and required to be closed during accident conditions is closed. The SR helps to ensure that post accident leakage of radioactive fluids or gases outside of the containment boundary is within design limits. For containment isolation valves inside containment, the Frequency of "prior to entering MODE 4 from MODE 5 if not performed within the previous 92 days" is appropriate since these containment isolation valves are operated under (continued)

DIABLO CANYON - UNITS 1 & 2 Revision 4 8S91DY04.doa - R4a 25

Containment Isolation Valves B 3.6.3 BASES SURVEILLANCE SR 3.6.3.4 (continued)

REQUIREMENTS administrative controls and the probability of their misalignment is low.

The SR specifies that containment isolation valves that are open under administrative controls are not required to meet the SR during the time they are open. This SR does not apply to valves that are locked, sealed, or otherwise secured in a closed position, since these were verified to be in the correct position upon locking, sealing, or securing.

This Note allows valves and blind flanges located in high radiation areas to be verified closed by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted during MODES 1, 2, 3, and 4, for ALARA reasons. Therefore, the probability of misalignment of these containment isolation valves, once they have been verified to be in their proper position, is small.

SR 3.6.3.5 Verifying that the isolation time of each automatic power operated containment isolation valve is within limits is required to demonstrate OPERABILITY. The isolation time test ensures the valve will isolate in a time period less than or equal to that assumed in the safety analyses. The isolation time and Frequency of this SR are in accordance with the Inservice Testing Program.

SR 3.6.3.6 Not Used SR 3.6.3.7 Containment Purge supply and exhaust, and Containment Pressure/Vacuum Relief valves with resilient seals, are leakage rate tested beyond the test requirements of 10 CFR 50, Appendix J, Option B to ensure OPERABILITY. Industry operating experience has demonstrated that this type of seal has the potential to degrade in a TVNCC.A.~ .... ~observation and the importance of maintaining these penetrations leak Seal Det loration" (Ref. 4). Since then, Kereliability ofA ese valves has improved with very low incidence of leakage excedding the allowable administrative limits. This allows extending the leakage test (continued)

DIABLO CANYON - UNITS 1 & 2 Revision 4 8S91DY04.doa - R4a 26

Containment Isolation Valves B 3.6.3 BASES SURVEILLANCE SR 3.6.3.7 (continued)

REQUIREMENTS Additionally, this SR must be performed within 92 days after opening the containment purge supply and exhaust valves. The 92 day Frequency was chosen recognizing that cycling these valves could introduce additional seal degradation (beyond that occurring to a valve that has not been opened). Thus, decreasing the interval E~ffrIf-24*-

IlU~feisa prudent measure after a valve has been opened.

Because of proven reliability of the containment vacuum/pressure relief valves, no leakage testing is required after they are opened.

A Note is added to clarify that Leakage Rate testing is not required for containment purge valves with resilient seals when their penetration flow path is isolated by a leak tested blank flange.

SR 3.6.3.8 Automatic containment isolation valves close on a containment isolation (Phase A, Phase B, or CVI) signal to prevent leakage of radioactive material from containment following a DBA. This SR ensures that each automatic valve will actuate to its isolation position on a containment isolation signal. This surveillance is not required for valves that are locked, sealed, or otherwise securednth eUr position under administrative controls. rVThe 24 month Frequency is based on the need to perform this Surveillance under the conditions 4-. rthat apply dauring plant outage anda the otential for an unp nrned transient if the/urveillance were perf med with the reaco rtpwr Operating perience has shown t/hthese componeW usually pass this Survei lance when performed t the 2.4 month Frequency.

Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

SR 3.6.3.9 Not Used SR 3.6.3. 10 Verifying that each 12 inch containment pressure/vacuum relief valve is blocked to restrict opening to:* 500 is required to ensure that the valves can close under DBA conditions within the times assumed in the analyses of References 1 and 2. If a LOCA occurs, the containment pressure/vacuum relief valves must close to maintain containment leakage within the values assumed in the accident analysis.I I e~ mnh Frequency is app ropriate because the J~c,~c~k ý-Gblocking dpdices are not typicall y remooled except during algtr~ii~eance.y SR 3.6.3. 11 Not Used (continued)

DIABLO CANYON - UNITS 1 & 2 Revision 4 8S91DY04.doa - R4a 27

Containment Pressure B 3.6.4 BASES (continued)

SURVEILLANCE SR 3.6.4.1 Verifying that containment pressure is within limits ensures that unit operation remains wtin th liisasumed in the containment analysis. JThe 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency of this SR was developed based on operating expe nce related to trending of containment pressur

.- ~variations d g the aplcbeMODES. rthermore, the 1 or Frequenc is considered adequate in v' ohe indications available in the control room, includi~a~lwarms, to alert the operator t an ab~normal containment pressure condition.

REFERENCES 1. FSAR, Section 6.2.

2. 10 CFR 50, Appendix K.

DIABLO CANYON - UNITS 1 & 2 Revision 4 8S91DY04.doa - R4a 31

Containment Air Temperature B 3.6.5 BASES (continued)

ACTIONS A.1 When containment average air temperature is not within the limit of the LCO, it must be restored to within limit within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . This Required Action is necessary to return operation to within the bounds of the containment analysis. The 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time is acceptable considering the sensitivity of the analysis to variations in this parameter and provides sufficient time to correct minor problems.

B.1 and B.2 If the containment average air temperature cannot be restored to within its limit within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.6.5.1 REQUIREMENTS Verifying that containment average air temperature is within the LCO limit ensures that containment operation remains within the limit assumed for the containment analyses. In order to determine the containment average air temperature, an arithmetic average is calculated using four temperature measurements. The four temperature measurement locations are pre-selected from:

a. TE-85 or TE-86, approximately 100 ft elevation between crane wall and containment wall,

b. TE-87 or TE-88, approximately 100 ft elevation between steam generators,

c. TE-89 or TE-90, approximately 140 ft elevation near equipment hatch or stairs at 2700, respectively,

d. TE-91 or TE-92, approximately 184 ft elevation on top of steam generator missile barriers away from steam generators.

r~cd&. 'The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency of this SR is considered acceptable based on osred slw rates o temperature increase within co inment as a result of envionme al heat sources due to the laro volume of containment). F hermore, the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Freque4 is considered

[adequate in vi of other indic ions available A the control room, (continued)

DIABLO CANYON -UNITS 1 & 2 Revision 4 8S91DY04.doa - R4a 34

Containment Spray and Cooling Systems B 3.6.6 BASES (continued)

SURVEILLANCE SR 3.6.6.1 REQUIREMENTS Verifying the correct alignment for manual, power operated, and automatic valves in the containment spray flow path provides assurance that the proper flow paths will exist for Containment Spray System operation. The containment spray flow path consists of the direct flow path from the fluid source (e.g., RWST) to the supplied safety-related component (e.g., spray headers) and portions of any branch line flow path off the direct flow path that a valve misposition could result in degradation of the system safety function. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these were verified to be in the correct position prior to locking, sealing, or securing. This SR also does not apply to valves which are closed and secured by a cap or blind flange (e.g., manual test, vent, and drain valves), to valves that cannot be inadvertently misaligned (e.g., check valves), or to valves in instrument or sample lines. This SR does not require any testing or valve manipulation.

Rather, it involves verification, through a system walkdown, which may include the use of local or remove indicators, that those valves outside containment (only check valves are inside containment) and capable A.Th,~ofpotentially being mispositioned are in the correct position.

SR 3.6.6.2 Operating each required CFCU for Ž! 15 minutes ensures that all trains are OPERABLE and that all associated controls are functioning properly. It also ensures that blockage, fan or motor failure or

_: y., A 1 Vxersifyingrtatioeach requir CeedfCUis reciving t require he thequnyia designoflodratesasumedni safety analysesbilt the be achievedn baltsancd dunrin nshetora opration to ednsuehaatalaleastp 1650 gpi deivreda to e cachFCdugringadesignthbaseseven ( pt bA).thee hydrauicnsytebanc considers fowo 15 norma syoidste algmnssuande thet potential for any single active failure.

(continued)

DIABLO CANYON - UNITS 1 & 2 Revision 4 8S91DY04.doa - R4a 46

Containment Spray and Cooling Systems B 3.6.6 BASES SURVEILLANCE SR 3.6.6.3 (continued)

REQUIREMENTS Operation of the CFCUs is permitted with lower CCW flow to the CFCUs during ASME Section Xl testing or decay heat removal in MODE 4 with the residual heat removal heat exchangers in service.

To support this conclusion, a calculation was performed to evaluate containment heat removal with one train of containment spray OPERABLE and reduced CCW flow to three CFCUs. The calculation concluded that this configuration would provide adequate heat removal to ensure that the maximum design pressure of containment was not exceeded during a DBA in MODE 1. This analysis also determined that a single failure could not be tolerated during this condition and still assure that the maximum design pressure of

+ ; +tnvan 1.4, +i Ka ana A A~ d' D fa

%..AJ CaOIHIIII ILUVV.UIU I% VU VAV V VI~.'J Cooling Water trai' redundancy stem, the two considering the known reliability avajd6lei of th~e

and the

[The Frequencyofwas developed low probabiUy a significant dp radation of flow o&urring between surveillances.

SR 3.6.6.4 Verifying each containment spray pump's developed head at the flow test point is greater than or equal to the required developed head (205 psid) ensures that spray pump performance has not degraded during the cycle. Flow and differential pressure are normal tests of centrifugal pump performance required by the ASME O&M Code (Ref. 5). Since the containment spray pumps cannot be tested with flow through the spray headers, they are tested on recirculation flow.

During refueling operation, a containment spray pump can be aligned to take suction from the refueling water storage tank (RWST) and discharge into the residual heat removal system discharge line back to the RWST. Flow using this lineup can achieve full design flow of 2600 gpm. This test confirms one point on the pump design curve and is indicative of overall performance. Such inservice tests confirm component OPERABILITY, trend performance, and detect incipient failures by abnormal performance. The Frequency of the SR is in accordance with the Inservice Testing Program.

SR 3.6.6.5 and SR 3.6.6.6 These SRs require verification that each automatic containment spray valve actuates to its correct position and that each containment spray pump starts upon receipt of an actual or simulated actuation of a containment high-high pressure signal with a coincident "S"signal.

This Surveillance is not required for valves that are locked, sealed, or

_Tvhsee -.. Usually pass Ke Surveillancers w~h*performed at Th month Frequency /herefore, -the Fegency was concluded to be acceptable" from a reliability standpoint.

DIABLO CANYON - UNITS 1 & 2 Revision 4 8S91DY04.doa - R4a 47

Containment Spray and Cooling Systems B 3.6.6 BASES SURVEILLANCE SR 3.6.6.7 REQUIREMENTS This SR requires verification that each CFCU actuates upon receipt of (continued) an actual or simulate nection ýsin fhe ý24mo~nth

-~be accept ble through operating expe nce. See SR .. 6.5 and SR SR 3.6.6.8 With the containment spray inlet valves closed and the spray header drained of any solution, low pressure air or smoke can be blown through test connections. This SR ensures that each spray nozzle is unobstructed and provides assurance that spray coverage of the containment durfn aacient is not degraded LDue to the ssive

.- f hen~~ra tes ao 0 year inev~-~ sded adequae dsin dtct obtuctiono hiozeý to SR 3.fi6.6.9 The CFCUs are designed to start or restart in low speed upon receipt of an SI signal. This SSR ensures that this feature is functionina Tv\54,(k .. properly. he 31 d frequency is select based upon the pormalr operation of the CUs in oer oerý in.

REFERENCES 1. FSAR, Appendix 3.1A

2. 10 CFR 50, Appendix K.

3. ESAR, Section 6.2.1.

4. ESAR, Section 6.2.2.

5. ASME Code for Operation and Maintenance of Nuclear Power Plants, 2001 Edition including 2002 and 2003 Addenda.

6. License Amendment 89/88, April 16, 1996.

7. Calculation STA-075, "Minimum ECCS Flow and Minimum Recirculation Spray Flow During the Sump Recirculation Phases."

DIABLO CANYON - UNITS 1 & 2 Revision 4 8S91DY04.doa - R4a 49

Spray Additive System B 3.6.7 BASES (continued)

ACTIONS A.1 If the Spray Additive System is inoperable, it must be restored to OPERABLE within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . The pH adjustment of the Containment Spray System flow for corrosion protection and iodine removal enhancement is reduced in this condition. The Containment Spray System would still be available and would remove some iodine from the containment atmosphere in the event of a DBA. The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time takes into account the redundant flow path capabilities and the low probability of the worst case DBA occurring during this period.

B.1 and B.2 If the Spray Additive System cannot be restored to OPERABLE status within the required Completion Time, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within 84 hours9.722222e-4 days 0.0233 hours 1.388889e-4 weeks 3.1962e-5 months . The allowed Completion Time of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems. The extended interval to reach MODE 5 allows 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months for restoration of the Spray Additive System in MODE 3 and 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months to reach MODE 5. This is reasonable when considering the reduced pressure and temperature conditions in MODE 3 for the release of radioactive material from the Reactor Coolant System.

SURVEILLANCE SR 3.6.7.1 REQUIREMENTS Verifying the correct alignment of Spray Additive System manual, power operated, and automatic valves in the spray additive flow path provides assurance that the system is able to provide additive to the Containment Spray System in the event of a DBA. The spray additive flow path consists of the direct flow path from the fluid source (e.g., spray additive tank) to the supplied safety-related component (e.g., spray headers) and portions of any branch line flow path off the direct flow path that a valve misposition could result in degradation of the system safety function. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves were verifieid to be in the correct position prior to locking, sealing, or securing. This SR also does not apply to valves which are closed and secured by a cap or blind flange (e.g., manual test, vent, and drain valves), to valves that cannot be inadvertently misaligned (e.g., check valves), or to valves in instrument or sample lines. This SR does not require any testing or valve manipulation. Rather, it involves verification through a system walkdown, which may include the use of local or remote indicators, that those valves outside containment and capable of potentially being mispositioned are in the correct position.

(continued)

DIABLO CANYON -UNITS 1 & 2 Revision 4 8S9 IDYO4.doa - R4a 45

Spray Additive System B 3.6.7 BASES SURVEILLA~NCE SR 3.6.7.2 REQUIREME KMTQ To provide effective iodine removal, the containment spray must be an (continued) alkaline solution. Since the RWST contents are normally acidic, the volume of the spray additive tank must provide a sufficient volume of spray additive to adjust pH for all water injected. This SR is performed to verify the availability of sufficient NaOH solution in the Spray Additive System. The required volume may be surveilled using an indicated level band of 50 to 88% for the Spray Additive Tank which corresponds to the LCO 3.6.7 minimum and maximum limits adjusted

~j.bFreq1u-enc-y wasJeveloped based on the low probability of an SR 3.6.7.3 This SR provides verification of the NaOH concentration in the spray additive tank and is sufficient to ensure that the spyra Isolution being in the sray addit' tank remains within th stablished limi~tp .his is Iariance in tank volume will be detected.

SR 3.6.7.4 This SR provides verification that each automatic valve in the Spray Additive System flow path actuates to its correct position on a containment spray actuation signal. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in te reg position under administrative controls. rThe 2-4 month Frequency is base, on Fe to perform this Surveillance under the conditions

ýneed 1 -1. tha aplydun n~a plant outage and the potential for an unplanneds transient if thSuurveil lance w/ere rformed with t/he eactor at power.

the Surveilance when performt ed at the 24 montK Frequency.

from a the Frequency was concluded to be acceptable Therefore,standpoint.

teeliability (continued)

DIABLO CANYON - UNITS 1 & 2 Revision 4 8S91DY04.doa - R4a 53

Spray Additive System B 3.6.7 BASES SURVEILLANCE SR 3.6.7.5 REQUIREMENTS To ensure correct operation of the Spray Additive System, flow from (continued) the spray additive tank to the eductors is verified AGý This SR is performed by verifying that the solution flow path is not blocked from the spray additive tank through test valve 8993, from the RWST through test valve 8993 for each of the two flow paths, and from the RWST to the eductors. This SR provides assurance that NaCH will be metered into the flow path upon Containment S r System initiation. jDue to tde-passive nature of the spray addi ve flow controls, the 5~e r Frequency is swff cient to identify combon'ievnt deqradatio at may affect flow rate-r REFERENCES 1. FSAR, Chapter 6.2.

DIABLO CANYON - UNITS 1 & 2 Revision 4 8S91DY04.doa - R4a 55

MSIVs B 3.7.2 BASES SURVEILLANCE SR 3.7.2.1 (continued)

REQUIREMENTS analyses. This Surveillance is normally performed upon returning the unit to operation following a refueling outage. The MSIVs should not be tested at power, since even a part stroke exercise increases the risk of a valve closure when the unit is generating power.

As the MSIVs are not tested at power, they are exempt from the ASME Code, Section Xl (Ref. 5), requirements during operation in MODE 1 or 2.

The Frequency is in accordance with the Inservice Testing Program.

This test may be conducted in MODE 3 with the unit at operating temperature and pressure. This SR is modified by a Note that allows entry into and operation in MODE 3 prior to performing the SR.

However, the test is normally conducted in MODE 5 as permitted by the cold shutdown frequency justification provided in the Inservice Testing Program (IST) and as permitted by Reference 6, Subsection ISTC-3521 (c). I SR 3.7.2.2 This SR verifies that each MSIV can close on an actual or simulated actuation signal. This Surveillance is normally performed upon

,e_. v, -~ frequency of MSIV testing is every 24 months. The 24 month 2.qunc FSAR Setion 6,e Appuelndi 6.2 C. rtige REERNC3 . FSAR, Section 15.4.2

4. 100CFR 100. 11.

5. ASME, Boiler and Pressure Vessel Code,Section XI.

6. ASME Code for Operation and Maintenance of Nuclear Power Plants, 2001 Edition including 2002 and 2003 Addenda.

DIAB3LO CANYON - UNITS 1 & 2 Revision 4 rad5l 0 E0.Doc - R4c 11

MFl Vs, MFRVs, MFRV Bypass Valves, MFWP Turbine Stop Valves B 3.7.3 BASES (continued)

SURVEILLANCE SR 3.7.3.1 and SR 3.7.3.2 REQUIREMENTS These SRs verify that the closure time of each MFIV is:!* 60 seconds and that each MERV, and MFRV bypass valves is:!* 7 seconds, not including the instrument delays. The MFIV and MFRV and MFRV bypass valve closure times are assumed in the accident and containment analyses. These Surveillances are normally performed upon returning the unit to operation following a refueling outage.

These valves should not be tested at power since even a part stroke exercise increases the risk of a valve closure with the unit generating power. This is consistent with the ASME Code (Ref. 2) stroke requirements during operation in MODES 1 and 2.

The Frequency for these SRs is in accordance with the Inservice Testing Program.

SR 3.7.3.3 This SR verifies that each MFIV, MFRV, MFRV bypass valve, and MFWP turbine stop valve can close on an actual or simulated whenrmll atainsaa.hsSurveillance atte4mnhFeuenypTeeore performed thiFequncisacertabl aoth reibii stndoint. re nyi SR 3.7.3.4 This SR verifies that the closure time of each MFWP turbine stop valve is:* 1 second, not including the instrument delays. The MFWP turbine stop valve closure times are assumed in the accident and containment analyses. These surveillances are normally performed on returning the unit to operation following a refueling outage. The Frequency is the same as that for the MFRVs and the MFRV bypass valves.

Preventive/predictive maintenance related to the MFWP turbine stop valves, and actions initiated in response to control oil cleanliness problems, shall be performed to ensure reliability of MFWP trip function.

REFERENCES 1. FSAR, Section 10.4.7.

2. ASME Code for Operation and Maintenance of Nuclear Power Plants, 2001 Edition including 2002 and 2003 Addenda.

DIABLO CANYON - UNITS 1 & 2 Revision 4 rad5lOEO. Doc - R4c 18

ADVs B 3.7.4 BASES (continued)

SURVEILLANCE Plant procedures which provide a 31 day verification that the 10%

REQUIREMENTS ADV manual block valves are open assures that the valves have not been inadvertently closed.

SR 3.7.4.1 To perform a controlled cooldown of the RCS, the ADVs must be able to be opened and closed remotely using the remote manual controls and the backup air bottles. This SR ensures that the ADVs are tested through a full control cycle at least once per fuel cycle. Performance Scomponents are expected to pass the2i rveillance when poormed at SR 3.7.4.2 The function of the block valve is to isolate a failed open ADV. Cycling

)( the block valve both closed and open demonstrates its capability to perform this function. Performance of inservice testing or use of the

~~> bock valve during unit cooldown may satisfy this reeurm~l b~~ain e~xp~eriý nce has shown that these components re exp~ected to pass the SVF~eillan ce when pe'formed "ttiespeciff6d frequency.

\The Freque rcy is acceptabl rrom a reliab6ility standpoint.

C~OyPk~

SR 3.7.4.3 The function of the back-up air bottles is to assure that the ADVs will be able to be opened as required to perform a controlled cooldown of the RCS in the event of a loss of the normal air supply system. The backup air bottle system was specifically installed to allow the RCS to

_!_y\C"e_ *1. be cooled for a SGTR coincident with a loss of offsite power.

Verification of the bottle pressure oms eyefyýe4-ýrrallows for timely bottle replacement and trending for leaks.

REFERENCES 1. FSAR, Section 15.

2. WCAP-1 1723

3. DCM S-25B, S-313, AND S-4.

DIABLO CANYON - UNITS 1 & 2 Revision 4 rad~l 0EO. Doc - R4c 22

AFW System B 3.7.5 BASES (continued)

SURVEILLANCE SR 3.7.5.1 REQUIREMENTS Verifying the correct alignment for manual, power operated, and automatic valves in the AFW System water and steam supply flow paths provides assurance that the proper flow paths will exist for AFW operation. The AFW System flow paths consist of the direct flow paths from the fluid source (e.g., CST, steam generators) to the supplied safety-related components (e.g., steam generator, turbine driven AEW pump) and portions of any branch line flow path off a direct flow path that a valve misposition could result in degradation of the system safety function. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since they are verified to be in the correct position prior to locking, sealing, or securing. This SR also does not apply to valves which are closed and secured by a cap or blind flange (e.g., manual test, vent, and drain valves), to valves that cannot be inadvertently misaligned (e.g., check valves), or to valves in instrument or sample lines. This Surveillance does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position.

The 31 day F~rqquency, based on engi~nenrng judg~ent, is consistent with proceok11aI controls governing vdfve operatin and ensures orrect valve positions.

The valves in the flowpath from the CST to the AFW pump suction are verified to be in the correct position prior to use of the AFW system for normal startup, and are subsequently controlled by a sealed valve checklist. Use of AFW for normal startups and shutdowns, and performance of the quarterly pump surveillance tests confirms that the CST flowpath to the AFW pump suction is properly aligned.

S R 3.7.5.2 Verifying that each AFW pump's developed head at the flow test point is greater than or equal to the required developed head ensures that AFW pump performance has not degraded during the cycle. Flow and differential head are normal tests of centrifugal pump performance required by the ASME OM Code (Ref 2). The ASME OM Code requires a comprehensive pump test on each AFW pump every two years. The comprehensive pump test is required to be performed at

+/- 20% of pump design flow. This test confirms one point on the pump design curve and is indicative of overall performance. In addition to the comprehensive pump test, the ASME OM Code also requires a less comprehensive test to be performed at 3-month intervals for each pump. These tests are performed at recirculation flow so as to limit thermal shocking of AFW/FW piping nozzles. The ability of steam traps 104, 105, and 106 to remove condensate in the steam supplies is verified during the inservice testing of the pumps.

Such inservice tests confirm component OPERABILITY, trend performance, and detect incipient failures by indicating abnormal DIABLO CANYON - UNITS 1 & 2 Revision 4 rad~l OEO. Doc - R4c 30

AFW System B 3.7.5 BASES SURVEILLANCE SR 3.7.5.2 (continued)

REQUIREMENTS This SR is modified by a Note indicating that the SR for the turbine-driven AFW pump should be deferred until suitable test conditions are established. This deferral is required because there is insufficient steam pressure to perform the test.

SR 3.7.5.3 This SR verifies that AFW can be delivered to the appropriate steam generator in the event of any accident or transient that generates an ESFAS, by demonstrating that each automatic valve in the flow path actuates to its correct position on an actual or simulated actuation generated by an auxiliary feedwater actuation signal. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in . under ad '*ati controls. The 24 month Frequency is based on the need to perform tis Surveillance under e conditions that a ply during a unit outage JLviL4zi A; 1 and the potential fo n unplanned trans if theSre nwr This SR is modified by a Note that states the SR is not required in MODE 4 when the steam generator is being relied upon for heat removal. In MODE 4, the required AFW train may already be aligned and operating.

SR 3.7.5.4 This SR verifies that the AF.W pumps will start in the event of any accident or transient that generates an ESFAS by demonstrating that

.each AFW pump starts automatically on an actual or simulated actuation generated by an auxiliary feedwater actuation signal in MODES 1, 2, and 3. In MODE 4, the required pump is already 2v~c~.vk '3- -&

arayoperating and the autostart function is not required.~h I4mntMOD 4,te haemvlrequirisbsdntemendts would behles puroveidiange modre time for perantoracsetion toemauavellyastartethe required Awit pump NaorUsed.

t j DIABLO CANYON - UNITS 1 & 2 Revision 4 radSI OEO. Doc - R4c 32

AFW System B 3.7.5 BASES SURVEILLANCE SR 3.7.5.6.

REQUIREMENTS This SR verifies that the FWST is capable of being aligned to the AFW (continued) pump suctions. This assures that this additional supply of required AFW is available from the seismically qualified FWST should it be needed for a natural circulation cooldown.

Since there is insufficient volume in the CST alone for long-term cooling needs, the NRC required in SSER 8 that the FWST have a seismically-qualified flow path to the AFW Pumps suction to withstand an assumed seismic failure of any single valve (valve jammed shut).

This means that valves MU-0-1557 and MU-1-297 and MU-2-298 should be maintained in their normal positions. If these valves are required to be out of position due to maintenance activities, then these activities should be treated as if entering the LCO action for TS 3.7.6.

Mnse,r(k. I.-bThes24 mont equ ency, base On engineering judgement, i A similar SR is not required for the CST alignment since the AFW system is used for startup and an AFW pump is tested each month.

This operation and the pump tests assure proper valve alignment.

REFERENCES 1. FSAR, Section 6.5 and Section 15.2.8.

2. ASME Code for Operation and Maintenance of Nuclear Power Plants, 2001 Edition including 2002 and 2003 Addenda.

3. DCM S-313.

4. 10 CFR 50.55a(b)(3)(vi). -I DIABLO CANYON - UNITS 1 & 2 Revision 4 rad5l OE. Doc - R4c 34

CST and FWST B 3.7.6 BASES ACTIONS B.1 and B.2 (continued) If the CST or FWST cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 4, without reliance on the steam generator for heat removal, within 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months .

The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

SURVEILLANCE SR 3.7.6.1 This SR verifies that the CST contains th a rpqi iffireri W e of coolin water. he 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is based on operating experience andd the need for operator awareness of unit evolutions 7that may affect the CST inventory

ýb een considerec ýequabe in checks.

view of other the 13,hourin Frequen Also,indý-ýtions the contry irloom, including alarms, to alert the operato d abnormal u deviatio;nsy inISthe coo'i sf T CST levels.

_IV\ eec +, SR 3.7.6.2 This SR verifie-sAh-at the FWST contain the required volume of cooli and hour Frequency is based on operating e anence water.

R for12

ýneed he oper3W awareness f iýevolutions that may affeict the IL FWST invei c r between checks. A0, the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Fre ncy is considered ý7equate in view of othi r indications in th;2oWntrrol room,

?I including alarms, to alert the operator to abnormal deviations in the REFERENCES 1. FSAR, Section 9.2.6 and 9.5.1.

2. FSAR, Chapter 6.

3. FSAR, Chapter 15.

4. DCM S-3B3.

DIABLO CANYON - UNITS 1 & 2 Revision 4 rad~l OEO. Doc - R4c 38

CCW System B 3.7.7 BASES ACTIONS B.1 and B.2 (continued)

If the vital CCW loop cannot be restored to OPERABLE status within the associated Completion Time, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

SURVEILLAIN'CE SR 3.7.7.1 REQUIREME.NTS This SR is modified by a Note indicating that the isolation of the CCW flow to individual components may render those components inoperable but does not affect the OPERABILITY of the CCW System.

A possible exception to this note, is isolation of CCW to the CECUs.

Isolation of CCW to the CFCUs could potentially affect the flow balance and requires evaluation to ensure continued operability.

Verifying the correct alignment for manual, power operated, and automatic valves in the CCW flow path provides assurance that the proper flow paths exist for CCW operation. The CCW flow path consists of the direct flow path servicing the safety related equipment (e.g., ECCS pump coolers, CFCUs, RHR heat exchanger) and portions of any branch line flow path off the direct flow path that a valve misposition could result in degradation of the system safety function. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves are verified to be in the correct position prior to locking, sealing, or securing. This SR also does not apply to valves which are closed and secured by a cap or blind flange (e.g., manual test, vent, and drain valves), to valves that cannot be inadvertently misaligned (e.g., check valves), or to valves in instrument or sample lines. This Surveillance does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position.

--,ýC"elr t -alThe 31 day Freque ncy is based on engineerrig judgment, is consistent with e procedural controls go ~rning valve operation, andl leonsures correct valve positionis.I SR 3.7.7.2 This SR verifies proper automatic operation of the COW valves on an actual or simulated Phase A or Phase B containment isolation actuation signal. The CCW System is a normally operating system that cannot be fully actuated as part of routine testing during normal operation. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under administrative controls. ~ e2 oth Frequency . based on the

- ' jn edto perform this Surveillaj e under the c9Ritions that apply Idrga nT outaae and tpotenia fraunlnned trans~ient if they DIABLO CANYON - UNITS 1 & 2 Revision 4 rad~l OEO. Doc - R4c 42

CCW System B 3.7.7 BASES SURVEILLANCE SR 3.7.7.2 (continued)

REQUIREMENTS SR 3.7.7.3 This SR verifies proper automatic operation of the CCW pumps on an reqireentyapplestinor ataorsimulanced the SoS auo-star aDother 4kVat auto-transfer automatic starts only. I Uperating experience has shown (that these coompr1 n1s usull pas ,he veillance when me*%red 2xvs-t rk -4 --

~at the 24 mprffh Frequency. Therefore the Frequency is a~eeptable Moma reli~ bility standpoint.

REFERENCES 1. FSAR, Section 9.2.2.

2. FSAR, Section 6.2.

3. WCAP-14282, Revision 1, "Evaluation of Peak CCW Temperature Scenarios for Diablo Canyon Units 1 and 2,"

dated December 1997.

DIABLO CANYON - UNITS 1 & 2 Revision 4 rad5I OEO. Doc - R4c 44

ASW B 3.7.8 BASES (continued)

ACTIONS A.1 If one ASW train is inoperable, action must be taken to restore OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . In this Condition, the remaining OPERABLE ASW train is adequate to perform the heat removal function. However, the overall reliability is reduced because a single failure in the OPERABLE ASW train could result in loss of ASW system function. The Note indicates that the applicable Conditions and Required Actions of LCO 3.4.6, "RCS Loops-MODE 4," should be entered if an inoperable ASW train results in an inoperable decay heat removal train. This is an exception to LCO 3.0.6 and ensures the proper actions are taken for these components. The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time is based on the redundant capabilities afforded by the OPERABLE train, and the low probability of a DBA occurring during this time period.

BA1 and B.2 If the ASW train cannot be restored to OPERABLE status within the associated Completion Times, the unit must be placed in a MODE in which the LCO does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

SURVEILLANCE SR 3.7.8. 1 REQUIREMENTS Verifying the correct alignment for manual and power operated valves in the ASW system flow path provides assurance that the proper flow paths exist for ASW system operation. The ASW system flow path consists of the direct flow path servicing the safety related equipment (e.g., CCW heat exchanger) and portions of any branch line flow path off the direct flow path that a valve misposition could result in degradation of the system safety function. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since they are verified to be in the correct position prior to being locked, sealed, or secured. This SR also does not apply to valves which are closed and secured by a cap or blind flange (e.g., manual test, vent, and drain valves), to valves that cannot be inadvertently misaligned (e.g., check valves), or to valves in instrument or sample lines. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.

The 31 day Frequency is based on en Jeering judgme t, Is consistent wite procedural contrc 7sgovernin va e operation, and rensur es or valve positionis.

DIABLO CANYON - UNITS 1 & 2 Revision 4 rad~l 0EO. Doc - R4c 48

ASW B 3.7.8 BASES SURVEILLANCE S R 3.7.8.2 REQUIREMENTS This SR verifies proper remote manual full stroke operation of the (continued) ASW valves. This Surveillance is not required for valves that are locked, sealed, or otherwise secured in the required position under program frequency anda i consistent with the ASME O&M Code testing requirements nd ensures the ability to crectly align the valves. Operatin experience has shown tha esecmont usually pass t e Surveillance when perfor dat the 92 day S R 3.7.8.3 This SR verifies proper automatic operation of the ASW pumps on an actual or simulated safety related actuation signal. The ASW is a normally operating system that c~annot ýbefful actuated as part of normal testing during no~rm~alo eration. The 24 month Frequency is b~ased-on tie peed to perform this Surveillapee under the conditions Sthat apply,0ring a unit outage and the pp~ential for an unplanned Itransient if the Surveillance were performed with the reactor at power.

This surveillance requirement applies to the SIS auto-start and the 4kV auto transfer automatic starts only. Uperating exxperjence has shown _,A.-

§Ice, I i tat these c ~onents usually pass the Surveilla ~e when performed

\t tthe 24 rthFrequency. Therefore, the Fr uency is acceptable trom a reliblt standpoint.I REFERENCES 1. FSAR, Section 9.2.7.

2. FSAR, Section 6.2.

3. NRC Generic Letter 91-13, "Request for Information Related to the Resolution of Generic Issue 130, 'Essential Service Water System Failures at Multi-unit Sites,' Pursuant to 10 CFR 50.54 (f),"dated September 19, 1991.

DIABLO CANYON - UNITS 1 & 2 Revision 4 radSIO0EO. Doc - R4c 50

CRVS B 3.7.10 BASES ACTIONS E.1 (continued) In MODE 5 or 6, or during movement of recently irradiated fuel assemblies, with two CRVS trains inoperable, action must be taken immediately to suspend activities that could result in a release of radioactivity that might enter the control room. This places the unit in a condition that minimizes accident risk. This does not preclude the movement of fuel to a safe position.

F.1 If both CRVS trains are inoperable in MODE 1, 2, 3, or 4, for reasons other than an inoperable control room boundary (i.e., Condition B), the CRVS may not be capable of performing the intended function and the unit is in a condition outside the accident analyses. Therefore, LCO 3.0.3 must be entered immediately.

SURVEILLANCE Once actuated due to a fuel handling accident the CRVS must be REQUIREMENTS protected against a single failure. This protection, although not required for immediate accident response, is assured by requiring that a backup power supply be provided as described above in Applicability. This back up is assured via the performance of surveillances that verify the ability to transfer power supplies.

The 31 day procedural verification of the separate vital power supplies for the redundant fans assures system reliability.

SR 3.7.10.1 Standby systems should be checked periodically for Ž! 15 minutes to ensure that they function properly. As the environment and normal operating conditions on this system are not too severe, testing each train once every month, by initiating, from the control room, flow through the HEPA filter and charcoal adsorber using either redundant set of booster and pressurization supply fans, provides an adequate Lvxse. k of--thi etadte two tra redundancy avai nilit SR 3.7.10.2 This SR assures that the emergency power alignment is appropriate for the operating conditions of the plant. With the power supply options available it is appropriate to verify that the redundant fans for each train are aligned to receive power from separate OPERABLE vital buses.(cniud DIABLO CANYON - UNITS 1 & 2 Revision 4 rad~l OEO. Doc - R4c 60

CRVS B 3.7.10 BASES SURVEILLANCE SR 3.7.10.3 REQUIREMENTS This SR verifies that the required CRVS testing is performed in (continued) accordance with the Ventilation Filter Testing Program (VFTP). The CRVS filter tests are in accordance with ANSI N510-1980 (Ref. 3).

The VFTP includes testing the performance of the HEPA filter, charcoal adsorber efficiency, minimum flow rate, and the physical properties of the activated charcoal. Specific test Frequencies and additional information are discussed in detail in the VFTP.

SR 3.7.10.4 This SR verifies that each CRVS train automatically starts and operates in the pressurization mode on an actual or simulated actuation signal generated from a Phase "A"lsola~tionrrI~e -Frequency SR 3.7.10.5 This SR verifies the integrity of the control room enclosure, and the assumed inleakage rates of the potentially contaminated air. The control room positive pressure, with respect to potentially contaminated adjacent areas, is periodically tested to verify proper functioning of the CRVS. During the pressurization mode of operation, the CRVS is designed to pressurize the control room -20.125 inches water gauge positive pressure with respect to the outside atmosphere in order to prevent unfiltered inleakage. The CRVS is designed to 40%4AW-& rmonths on aSTAGGERED TEST BASIS )i ased upontWe maintenance and op atmn history (Ref. 6 REFERENCES 1. FSAR, Section 9.4.1.

2. FSAR, Chapter 15.

3. ANSI N510-1980.

4. NUREG-0800, Section 6.4, Rev. 2, July 1981.

5. DCM S-23F.

6. License Amendment 119/117, April 14, 1997.

7. License Amendment 184/1 86, January 3, 2006.

DIABLO CANYON - UNITS I & 2 Revision 4 rad~l 0EO. Doc - R4c 61

ABVS 8 3.7.12 BASES (continued)

SURVEILLANCE SR 3.7.12.1 REQUIREMENTS Each ABVS train should be checked periodically to ensure that it functions properly. As the environment and normal operating conditions on this system are not severe, testing each train with flow through both the HEPA filter and charcoal adsorber bank once a month provides an adequate check on this system. Both ABVS trains shall be operated long enough (Ž!15 minutes) to verify all components are operating correctly. Monthly verification of the separate OPERABLE vital power supplies for the exhaust fans assures syse

~ #1-~. f eqipmet andiihe two train redundancy avoitble.

SR 3.7.12.2 This SR verifies that the required ABVS testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The ABVS filter tests are in accordance with References 3 and 4. The VFTP includes testing HEPA filter performance, charcoal adsorber efficiency, minimum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations).

Specific test Frequencies and additional information are discussed in detail in the VFTP.

SR 3.7.12.3 The SR is modified by a Note, which limits the applicability of this SR when the ABVS is already in its safety function configuration and is verified to be capable of providing that function. The intent of this change is only to address this specific condition and the SR is considered applicable and must be met whenever the ABVS is not in that configuration.

This SR verifies that each ABVS train actuates on an actual or simulated actuation signal by verifying that the exhaust fan starts and the associated dampers align to exhaust through the common HE k.~ filter and charcoal adsorber (Ref. 3 and 4). e 24 month FFre~que~ncy, isbased upon the mainleflance and operati~ge-tistory (Ref. 8).

SR 3.7.12.4 Not Used.

SR 3.7.12.5 Not Used.

(continued)

DIABLO CANYON -UNITS 1 & 2 Revision 4 rad~l OEO. Doc - R4c 66

ABVS B83.7.12 BASES SURVEILLANCE SR 3.7.12.6 REQUIREMENTS This SR verifies the leak tightness of dampers that isolate flow to the (continued) normally operating filter train. This SR assures that the flow from the auxiliary building passes through the HEPA filter and charcoal adsorber unit when the ABVS Buildings and Safeguards or Safeguards Only modes have been actuated colincid- -ihan-SL hre 24 monto Frequency is based uperrthe maintenance and]

joperahtgin3i11story REFERENCES 1. FSAR, Section 9.4.2.

2. FSAR, Section 15.5.

3. ASTM D 3803-1989

4. ANSI N510-1980

5. 10OCFR 100.11.

6. NUREG-0800, Section 6.5.1, Rev. 2, July 1981.

7. DCM S-23B, "Main Auxiliary Building Heating and Ventilation System".

. R I *rpnqeAmend ý.

DIABLO CANYON - UNITS 1 & 2 Revision 4 rad~l OEO. Doc - R4c 67

FHBVS B63.7.13 BASES ACTIONS C.1 (continued) When two trains of the FHBVS are inoperable during movement of recently irradiated fuel assemblies in the fuel handling building, suspend movement of recently irradiated fuel assemblies in the fuel handling building. This does not preclude the movement of fuel assemblies to a safe position.

SURVEILLANCE Once actuated due to a fuel handling accident the FHBVS must be REQUIREMENTS protected against a single failure coincident with a loss of offsite power. Protection against a loss of power, although not required for immediate accident response, is assured by requiring that a backup power supply be provided as described above in the LCO section.

This back up is assured via the performance of non-TS surveillances.

SR 3.7.13. 1 Standby systems should be checked periodically to ensure that they function properly. As the environmental and normal operating conditions on this system are not severe, testing each train once every month provides an adequate check on this system. This testing requires establishing air flow through both the HEPA filters and charcoal adsorbers.

Systems without heaters need only be operated for Ž! 15 minutes to fuctin o te sste- The 31 day F equency is demnstateth

-'based on the knownvliability of the equipment anthtwtri SR 3.7.13.2 This SR verifies that the required FHBVS testing is performed in accordance with the Ventilation Filter Testing Program (VFTP). The FHBVS filter tests are in accordance with References 5 and 6. The VFTP includes testing HEPA filter performance, charcoal adsorber efficiency, minimum system flow rate, and the physical properties of the activated charcoal (general use and following specific operations).

Specific test frequencies and additional information are discussed in detail in the VFTP.

SR 3.7.13.3 This SR verifies that each FHBVS train starts and operates on an actual or simulated actuation signal and directs its exhaust flowIl through the HEPA Filters and charcoal adsorber banks. T4Th-24-wonthj 2L4C'er~* ~-FrzguncI ic zniastzt with. Referecne" O.-

(continued)

DIABLO CANYON - UNITS 1 & 2 Revision 4 rad5lO0EO. Doc - R4c 72

FHBVS B 3.7.13 BASES SURVEILLANCE SR 3.7.13.4 REQUIREMENTS This SR verifies the integrity of the fuel handling building enclosure.

(continued) The ability of the fuel handling building to maintain negative pressure with respect to potentially uncontaminated adjacent areas is periodically tested to verify proper function of the FHBVS. During the post accident mode of operation, the FHBVS is designed to maintain a slight negative pressure in the fuel handling building, to prevent unfiltered LEAKAGE. The FHBVS is designed to maintain the building pressure *ý -0. 125 inches water gauge with respect to atmospheric pess The 2ý4 monhFrequ enc on a ~E<DET -

LA is based onthe maintena~nce ope~rating hilrRef. ý)

SR 3.7.13.5 Operation of damper M-29 is necessary to ensure that the system 4functions properly. The operability of damper M-29 is verified if it can eýs ho P, menth Frequ j znistert with Rofernco 0 REFERENCES 1.FSAR, Section 9.4.4.

2. FSAR, Section 15.5.

3. Regulatory Guide 1.25.

4. 10OCFR 100.

5. ASTM D 3802-1989

6. ANSI N510-1980.

7. NUREG-0800, Section 6.5.1, Rev. 2, July 1981.

8. DCM S-23D, "Fuel handling Building HVAC System."

Lcrdet19'1,

~ ~~ ~ ~ ~pi ~ 419.~

1~~~~0

0~'61.

10. License Amendment 184/1 86, January 3, 2006.

11. PG&E Letter DCL-05-124 DIABLO CANYON - UNITS 1 & 2 Revision 4 rad~l OEO. Doc - R4c 73

Spent Fuel Storage Pool Water Level B83.7.15 BASES (continued)

APPLICABILITY This LCO applies during movement of irradiated fuel assemblies in the spent fuel pool, since the potential for a release of fission products exists.

ACTIONS A.1 Required Action A. 1 is modified by a Note indicating that LCO 3.0.3 does not apply.

When the initial conditions for prevention of an accident cannot be met, steps should be taken to preclude the accident from occurring.

When the spent fuel pool water level is lower than the required level, the movement of irradiated fuel assembly in the spent fuel pool is immediately suspended. This does not preclude movement of a fuel assembly to a safe position.

If moving irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in MODES 1, 2, 3, and 4, the fuel movement is independent of reactor operations. Therefore, inability to suspend movement of irradiated fuel assemblies is not sufficient reason to require a reactor shutdown.

SURVEILLANCE SR 3.7.15. 1 REQUIREMENTS This SR is done during the movement of irradiated fuel assemblies as stated in the Applicability. This SR verifies sufficient fuel storage pool water is available in the event of a fuel handling accident. The water 4 b, Frequency is appropi te because the volu in tral

.te o During refueling operations, the level in the spent fuel pool is in equilibrium with the refueling canal, and the level in the refueling canal is checked daily in accordance with SR 3.9.7.1.

REFERENCES 1. FSAR, Section 9.1.2.

2. FSAR, Section 9.1.3.

3. FSAR, Section 9.1.4.3.4, 15.4.5 and 15.5.22.

4. Regulatory Guide 1.25, Rev. 0.

5. 10OCFRI100.11.

DIABLO CANYON - UNITS 1 & 2 Revision 4 rad~l OEO. Doc - R4c 76

Spent Fuel Pool Boron Concentration B 3.7.16 BASES (continued)

ACTIONS A.1 and A.2 The Required Actions are modified by a Note indicating that LCO 3.0.3 does not apply.

When the concentration of boron in the spent fuel storage pool is less than required, immediate action must be taken to preclude the occurrence of an accident or to mitigate the consequences of an accident in progress. This is most efficiently achieved by immediately suspending the movement of fuel assemblies and immediately taking actions to restore the spent fuel pool boron concentration to greater than or equal to 2000 ppm. This suspension of fuel movement does not preclude movement of fuel assemblies to a safe position.

If the LCO is not met while moving fuel assemblies LCO 3.0.3 would not be applicable since the inability to suspend movement of fuel assemblies is not sufficient reason to require a reactor shutdown.

SURVEILLANCE SR 3.7.16.1 REQUIREMENTS This SR verifies by chemical analysis that the concentration of boron in the spent fuel pool is at or above the required limit. As Iloný as this

]LM.,J *. AL -64requency is ýappropriate because no major re ishment of pool REFERENCES 1. Double contingency principle of ANSI N16.1-1975, as specified in the April 14, 1978 NRC letter (Section 1.2) and implied in the proposed revision to Regulatory Guide 1.13 (Section 1.4, Appendix A).

2. Not used.

3. "Criticality Safety Evaluation of Region 2 of the Diablo Canyon Spent Fuel Storage Racks with 5.0 % Enrichment," S.E.Turner, October 1993, Holtec Report HI-931077.

4. ESAR, Section 9.1, 15.4.5, and 15.5.22.

5. "Diablo Canyon Units 1 and 2 Spent Fuel Criticality Analysis,"

February 14, 2001, Paul F. O'Donnell, Westinghouse Doc. No.

A-DP1-FE-0001.

6. "Diablo Canyon Units 1 and 2 Spent Fuel Boron Dilution Analysis," January, 2001, Gary J. Corpora

7. License Amendment 154/154, September 25, 2002.

8. "Spent Fuel Storage Expansion at Diablo Canyon Units 1 & 2 for Pacific Gas & Electric Co.", October 2004, Holtec Report HI-2043162.

9. License Amendment 183/1 85, November 21, 2005.

DIABLO CANYON -UNITS 1 & 2 Revision 4 rad~l OEO. Doc - R4c 80

Secondary Specific Activity B 3.7.18 BASES ACTIONS A.1 and A.2 (continued) least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , and in MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

SURVEILLANCE SR 3.7.18. 1 REQUIREMENTS This SR verifies that the secondary specific activity is within the limits of the accident analysis. A gamma isotopic analysis of the secondary coolant, which determines DOSE EQUIVALENT 1-131, confirms the validity of the safety analysis assumptions as to the source terms in post accident releases. It also serves to identify and trend any unusual isotopic concentrations that might indicate changes in reactor elow theL 201imlimi._

REFERENCES 1. 10OCFRI100.11.

2. ESAR, Chapter 15.

DIABLO CANYON - UNITS I & 2 Revision 4 rad~l OEO. Doc - R4c 86

AC Sources - Operating B 3.8.1 BASES SURVEILLANCE Where the SRs discussed herein specify voltage and frequency REQUIREMENTS tolerances, the following is applicable. The minimum steady state (continued) output voltage of 3785 V is consistent with the second level undervoltage relay allowable values. This is the minimum steady state voltage needed on the 4160 volt vital buses to ensure adequate 4160 volt, 480 volt and 120 volt levels. The specified maximum steady state output voltage of 4400 V is equal to the maximum operating voltage for 4000 V motors specified in ANSI C84.1. The maximum steady state output voltage of 4400 V ensures that for a lightly loaded distribution system, the voltage at the terminals of 4000 V motors is no more than the maximum rated operating voltages. The specified minimum and maximum frequencies of the DG are 58.8 Hz and 61.2 Hz, respectively. These values are equal to +/- 2% of the 60 Hz nominal frequency and are derived from the recommendations given in Regulatory Guide 1.9 Rev. 2 (Ref. 16).

SR 3.8.1.1 This SR ensures proper circuit continuity for the offsite AC electrical power.supply to. the onsite distribution network and availability of offsite AC electrical power. The breaker alignment verifies. that each breaker is in its correct position to ensure that distribution buses and loads are

... operator being~A~are of it and bcý isstatus is dis yedith control room. bcuef SR 3.8.1.2 and SR 3.8.1.7 These SRs help to ensure the availability of the standby electrical power supply to mitigate DBAs and transients and to maintain the unit in a safe shutdown condition.

To minimize the wear on moving parts that do not get lubricated when the engine is not running, these SRs are modified by a Note (Note 2 for SR 3.8.1.2) to indicate that all DG starts for these Surveillances may be preceded by an engine prelube period and, for SR 3.8.1.2, followed by a warmup period prior to loading.

For the purposes of SR 3.8.1.2 and SR 3.8.1.7 testing, the DGS are started from standby conditions. Standby conditions for a DG means that the diesel engine coolant and oil temperature is being maintained consistent with manufacturer recommendations of equal to or greater than 90OF but less than 1750F. For the purposes of this SR, the diesel generator start will be initiated using one of the following signals:

1) manual, 2) simulated loss of offsite power, and 3) safety injection actuation test signal.

(continued)

DIABLO CANYON - UNITS 1 & 2 Revision 4 8S9 IEAO4.DOA - R4a 13

AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.2 and SR 3.8.1.7 (continued)

REQUIREMENTS SR 3.8.1.7 requires thatat-s++lý c 0 ~g~'the ý DG starts from standby conditions and achieves required speed within 10 seconds and required voltage and frequency within 13 seconds. The 10 second start requirement reflects the point during the DG's acceleration at which the DG is assumed to be able to accept load. The 13 second start requirement reflects the point at which the DG is assumed to have reached stable operation. These stability points represent the recovery of the DG and the power distribution system following a transient. This assures the ability of the system to undergo further transients. Actual steady state operation is expected to achieve a level of stability closer to the nominal 60 Hz value. The 10 and 13 second start requirements support the assumptions of the design basis LOCA analysis in the FSAR, Chapter 15 (Ref. 5).

Since SR 3.8.1.7 requires a timed start, it is more restrictive than SR 3.8.1.2, and it may be performed in lieu of SR 3.8.1.2. This is the intent of Note 1 of SR 3.8.1.2.

The 31 day Frequency for SR 3.8.1.2 is consistent with Generic Letter 94-01 (Ref. 12). The 184 dpyFrequency for SR 3.8.1.7 is aredu on~

in cold testing consistenjý*th Generic Letter -15 (Ref.' 7). r~e Frequencies provide aequate assurances ý DG OPERAB ILITY while minimizing degradation resulting from testing.

SR 3.8.1.3 This Surveillance verifies that the OGs are capable of synchronizing with the offsite electrical system and accepting loads greater than or equal to the equivalent of the maximum expected accident loads. A minimum run time of 60 minutes is required to stabilize engine temperatures, while minimizing the time that the DG is connected to the offsite source.

Although no power factor requirements are established by this SR, the DG is normally operated at a power factor between 0.8 lagging and 1.0. The 0.8 value is the design rating of the machine, while the 1.0 is an operational limitation to ensure circulating currents are minimized.

The load band is provided to avoid routine overloading of the DG.

OPERATION within the load range of 90% to 100% of rated full load without anomalies will provide adequate assurance of the machine's ability to carry 100% of rated full load if required.

The 31 day Frequency fo)-t is Surveýjijce is consisrt with Regulatory Guide 1. ev3R4)

(continued)

DIABLO CANYON - UNITS 1 & 2 Revision 4 8S9IEA04.DOA - R4a 14

AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.3 (continued)

REQUIREMENTS This SR is modified by four Notes. Note 1 indicates that diesel engine runs for this Surveillance may include gradual loading, as recommended by the manufacturer, so that mechanical stress and wear on the diesel engine are minimized. Note 2 states that momentary transients, because of changing bus loads, do not invalidate this test. Similarly, momentary power factor transients above the limit do not invalidate the test. Note 3 indicates that this Surveillance should be conducted on only one DG at a time per unit in order to avoid common cause failures that might result from offsite circuit or grid perturbations. Note 4 stipulates a prerequisite requirement for performance of this SR. A successful DG start must precede this test to credit satisfactory performance.

SR 3.8.1.4 This SR provides verification that the level of fuel oil in the day tank is a contained quantity sufficient for DG operation at full load for a nominal one-hour period. One hour is adequate time for an operator to take corrective action to restore the fuel oil supply to the affected day tank.

The level is expressed as an equivalent volume in gallons.

The 31 day F quency is adequate to assure that a sufficient supply of1 kzi.

a-' fue oiRisailable, since the tr~ansf pumps auto-starts ar a level above t?~e minimum containe~d v ume. Therefore, nor DG operation will not result in da tank level below the nimumn re uired volu Ad i ional assurance o su I ay ank contained volume is provided by a low level alarm.

SR 3.8.1.5 Microbiological fouling is a major cause of fuel oil degradation. There are numerous bacteria that can grow in fuel oil and cause fouling, but all must have a water environment in order to survive. Removal of water from the fuel oil day tanks &meew 14 eliminates the necessary environment for bacterial survival. This is the most effective means of controlling microbiological fouling. In addition, it eliminates the potential for water entrainment in the fuel oil during DG operation.

Water may come from any of several sources, including condensation, ground water, rain water, contaminated fuel oil, and breakdown of the fuel oil by bacteria. Frequent checking for and removal of accumulated water minimizes fouling and provides data regarding the watertight Xv~sevk '~~ esaished yvar Guid1.3 necessarily represent failure of this SR, provided the accumulated water is removed during the performance of this Surveillance.

(continued)

DIABLO CANYON - UNITS 1 & 2 Revision 4 8S9 IEAO4.DOA - R4a 15

AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.6 REQUIREMENTS This Surveillance demonstrates that each required fuel oil transfer (continued) pump operates and transfers fuel oil from the fuel oil storage tanks to each day tank. This is required to support continuous operation of standby power sources. This Surveillance provides assurance that the fuel oil transfer pump is OPERABLE, the fuel oil piping system is intact, the fuel delivery piping is not obstructed, and that controls are configured such that each unit will preferentially receive fuel from a different storage tank while using the other unit's preferred storage as its backup storage.

T he Frequency 31dy saequate aI to verify proper o eration of th I fuloil transferp ps and day tank supply valves jpI1ainta in the requiedum of fuel oil in the day tanks. The frequency has been proveacceable through operating experience.1 SR 3.8.1.7 See SR 3.8.1.2.

SR 3.8.1.8 Transfer of each 4.16 kV ESF bus power supply from the normal offsite circuit to the alternate offsite circuit, which is the immediate access 230 kV demonstrates the OPERABILITY of the alternate circuit distribution network to power the shutdown loads. Transfer of each 4.16 kV ESE bus power supply from the alternate offsite circuit (immediate access 230 kV) to the delayed access circuit (500 kV circuit) demonstrates the ability of the delayed access circuit. he 24 month Frequenicy of the Surveillance is based on engineering judgment, takin into consideration the unit conditions requiredTto perform ftheurveillance, and is intendedd to bDconsistent wit expecte uel cycle lengths. Operating exp ience has sh9 that these o~mrponents usually pass the SR en performe -atthve 24 month Frequency. Therefore, the req uency was concluded to be j acceptable from a reliability standpoint.

This SR is modified by a Note. The reason for the Note is that, during operation with the reactor critical, performance of this SR for automatic bus transfers could cause perturbations to the electrical distribution systems that could challenge continued steady state operation and, as a result, unit safety systems. The restriction applies only to automatic bus transfers where a unit trip and reactor trip will occur.

(continued)

DIABLO CANYON - UNITS 1 & 2 Revision 4 8S9 IEA04.DOA - R4a 16

AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.8 (continued)

REQUIREMENTS This restriction from normally performing the Surveillance in MODE 1 or 2 is further amplified to allow the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines plant safety is maintained or enhanced. This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed Surveillance, a successful Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independenty for the Surveillance; as well as the operator procedures available to cope with these outcomes. These shall be measured against the avoided risk of a plant shutdown and startup to determine that plant safety is maintained or enhanced when the Surveillance is performed in MODE 1 or 2. Risk insights or deterministic methods may be used for this assessment. This restriction does not apply to manual bus transfers which are a normal action required during a plant startup or shutdown.

Preplanned maintenance that would require the performance of this SR to demonstrate operability following the maintenance shall only be performed in Modes 3, 4, 5, or 6.

SR 3.8. 1.9 Each DG is provided with an engine overspeed trip to prevent damage to the engine. Recovery from the transient caused by the loss of a large load could cause diesel engine overspeed, which, if excessive, might result in a trip of the engine. This Surveillance demonstrates the DG load response characteristics and capability to reject the largest single load without exceeding predetermined voltage and frequency and while maintaining a specified margin to the overspeed trip. The single largest DG load is a centrifugal charging pump (CCP), which is rated at 600 hp. The CCP has a maximum demand, based on the maximum expected horsepower input and motor efficiency, of 515 kW.

This Surveillance may be accomplished by:

a. Tripping the DG output breaker with the DG carrying greater than or equal to its associated single largest post-accident load while paralleled to offsite power, or while solely supplying the bus; or

b. Tripping its associated single largest post-accident load with the DG solely supplying the bus.

c. Simultaneously tripping a combination of loads equal to or greater than the DG's associated single largest post-accident load with the DG solely supplying the bus.

(continued)

DIABLO CANYON - UNITS 1 & 2 Revision 4 8S91EA04.DOA - R4a 17

AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.9 (continued)

REQUIREMENTS The time, voltage, and frequency tolerances specified in this SR are derived from Regulatory Guide 1.9 Rev. 2 (Ref. 16) recommendations for response during load sequence intervals. The 2.4 seconds specified is equal to 60% of a typical 4 second load sequence interval associated with sequencing of the largest load. The voltage and frequency specified are consistent with the design range of the equipment powered by the DG. SR 3.8.1 .9.a corresponds to the maximum frequency excursion, while SR 3.8.1.9.b and SR 3.8.1.9.c his SR is modifi~ed by two Notes. The reason for Note 1 is that during operation with the reactor critical, performance of this SR could cause perturbations to the electrical distribution systems that could challenge continued steady state operation and, as a result, unit safety systems.

This restriction from normally performing the Surveillance in MODE 1 or 2 is further amplified to allow the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines plant safety is maintained or enhanced. This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed Surveillance, a successful Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independenty for the Surveillance; as well as the operator procedures available to cope with these outcomes. These shall be measured against the avoided risk of a plant shutdown and startup to determine that plant safety is maintained or enhanced when the Surveillance is performed in MODE 1 or 2. Risk insights or deterministic methods may be used for this assessment.

Preplanned maintenance that would require the performance of this SR to demonstrate operability following the maintenance shall only be performed in Modes 3, 4, 5, or 6.

(continued)

DIABLO CANYON - UNITS 1 & 2 Revision 4 8S9 IEA04.DOA - R4a 18

AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.9_ (continued)

REQUIREMENTS In order to ensure that the DG is tested under load conditions that are as close to design basis conditions as possible, Note 2 requires that, if synchronized to offsite power, testing must be performed using a power factor

0.9 lagging. This power factor is chosen to be representative of the actual design basis inductive loading that the DG would experience.

SR 3.8.1.10 This Surveillance demonstrates the DG's capability to reject a full load without overspeed tripping or exceeding the predetermined voltage limits. The DG full load rejection may occur because of a system fault or inadvertent breaker tripping. This Surveillance ensures proper engine generator load response under the simulated test conditions.

This test simulates the loss of the total connected load that the DG would experience following a full load rejection and verifies that the DG does not trip upon loss of the load. These acceptance criteria provide for DG damage protection. While the DG is not expected to experience this transient during an event and continue to be available, this response ensures that the DG is not degraded for future application, including reconnection to the bus if the trip initiator can be corrected or isolated.

In order to ensure that the DG is tested under load conditions that are as close to design basis conditions as possible, testing must be performed using a power factor *ý 0.87 lagging. This power factor is chosen to be representative of the actual design basis inductive loading that the DG would experience.

=Regulatory EThe Guide,241. 8 (Ref.

month isintended

9) and is Frequency fuelccyle liengths.I to with consistent th ejiirn~tof be consistent with expected (continued)

DIABLO CANYON - UNITS 1 & 2 Revision 4 8S91EA04.DOA - R4a 19

AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.11 REQUIREMENTS As required by Regulatory Guide 1.108 (Ref. 9), paragraph 2.a.(1), this (continued) Surveillance demonstrates the as designed operation of the standby power sources during loss of the offsite source. This test verifies all actions encountered from the loss of offsite power, including shedding of the nonessential loads and energization of the emergency buses and respective loads from the DG. It further demonstrates the capability of the DG to automatically achieve the required voltage and frequency within the specified time.

The DG autostart time of 10 seconds is derived from requirements of the accident analysis to respond to a design basis accident. The 10 second requirement reflects the assumption of the accident analysis that the DG has reached the point in its acceleration where the DG is able to accept load. The Surveillance should be continued for a minimum of 5 minutes in order to demonstrate that all starting transients have decayed and stability is achieved. After energization of the loads, steady state voltage and frequency are required to be within their limits.

The requirement to verify the connection and power supply of permanent and autoconnected loads is intended to satisfactorily show the relationship of these loads to the DG loading logic. The permanently connected loads are the Class 1E 480 VAC buses. The permanently connected loads do not receive a load shed signal. In addition, the containment fan cooler units do not receive a load shed signal but are de-energized when their motor contactors drop out on undervoltage. The permanently connected loads are re-energized when the DG breaker closes to energize the bus. The auto-connected loads are those loads that are energized via their respective sequencing timer. In certain circumstances, many of these loads cannot actually be connected or loaded without undue hardship or potential for undesired operation. In lieu of actual demonstration of connection and loading of loads, testing that adequately shows the capability of the DG systems to perform these functions is acceptable.

This testing may include any series of sequential, overlapping, or total k

steps so that the entire connection and loading sequence is verified.

The Frequency of 24 months is consistent with the intent of Regulatory Gude 1198(Ref. 9), paragraph 2.a.(.I-) takes into consideration unit conditio pis required to perform the,.Errveillance, and is intendedEto be]

,consistent with expected fuel cycle lengths.

(continued)

DIABLO CANYON -UNITS 1 & 2 Revision 4 8S91EA04.DOA - R4a 20

AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.12 (continued)

REQUIREMENTS The requirement to verify the connection of permanent and autoconnected loads to the immediate access 230 kV offsite power system is intended to satisfactorily show the relationship of these loads to the DG loading logic. For a description of the permanent and auto-connected loads, see SR 3.8.1.11 Bases. In certain circumstances, many of these loads cannot actually be connected or loaded without undue hardship or potential for undesired operation. In lieu of actual demonstration of connection and loading of loads, testing that adequately shows the capability of the DG system to perform these functions is acceptable. This testing may include any series of sequential, overlapping, or total steps so that the entire connection and loading sequence is verified.

The Frequency of 24 months takes into consideration unit conditions

'~,evk 4~' required to perform the Surveillance and is intendd to be consistent' with the expect dfuel cycle lengths. Operatin ,Aperience has sh9~vn\

that these c9¶fI ponents usually pass the SR Wen performed at tHe 24 month Irrequency. Therefore, the Frequency was concluded to be acceptable fro ia tnpit This SR is modified by two Notes. The reason for Note 1 is to minimize wear and tear on the DGS during testing. For the purpose of this testing, the DGS must be started from standby conditions, that is, with the engine coolant and oil temperature maintained consistent with manufacturer recommendations of equal to or greater than 90'F but less than 175'F. The reason for Note 2 is that during operation with the reactor critical, performance of this Surveillance could cause perturbations to the electrical distribution systems that could challenge continued steady state operation and, as a result, unit safety systems.

This restriction from normally performing the Surveillance in MODE 1 or 2 is further amplified to allow portions of the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g. post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines plant safety is maintained or enhanced. This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed partial Surveillance, a successful partial Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the partial Surveillance; as well as the operator procedures available to cope with these outcomes.

These shall be measured against the avoided risk of a plant shutdown and startup to determine that plant safety is maintained or enhanced when portions of the Surveillance are performed in MODE 1 or 2. Risk insights or deterministic methods may be used for this assessment.

(continued)

DIABLO CANYON -UNITS 1 & 2 Revision 4 8S91EA04.DOA - R4a 22

AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.12 (continued)

REQUIREMENTS Preplanned maintenance that would require the performance of this SR to demonstrate operability following the maintenance shall only be performed in Modes 3, 4, 5, or 6.

SR 3.8.1.13 This Surveillance demonstrates that DG noncritical protective functions are bypassed when the diesel engine trip cutout switch is in the cutout position and the DG is aligned for automatic operation. The noncritical trips include directional power, loss of field, breaker overcurrent, high jacket water temperature, and diesel overcrank. These noncritical trips are bypassed during DBAs and provide an alarm on an abnormal engine condition. This alarm provides the operator with sufficient time to react appropriately. The DG availability to mitigate the DBA is more critical than protecting the engine against minor problems that are not immediately detrimental to emergency operation of the DG.

'The 24 rmonth Frequency is based on engineering judgment, taking k~ ~ ito cnsierWh unit conditions required to perform the Surveillance, and is interjn6d to be consistent with petdfe ccelnt Operatinr experience has shown at these components us ly pass the SR when performed at the 2ý4 month Frequency. The fore, the Frequency was concluded to be acceptable from a reliability SR 3.8.1.14 The refueling outage intent of Regulatory Guide 1.108 (Ref. 9),

paragraph 2.a.(3), requires demonstration nep-2Z A ath DGs can start and run continuously at full load capability for an interval of not less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , Žý2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of which is at a load equivalent to 110% of the continuous duty rating and the remainder of the time at a load equivalent to the continuous duty rating of the DG. The DG starts for this Surveillance can be performed either from standby or hot conditions. The provisions for prelubricating and warmup, discussed in SR 3.8.1.2, and for gradual loading, discussed in SR 3.8.1.3, are applicable to this SR.

(continued)

DIABLO CANYON -UNITS 1 & 2 Revision 4 8S91EA04.DOA - R4a 23

AC Sources - Operating B 3.8.1 BASES SR 3.8.1.14 (continuted).

REQUIREMENTS In order to ensure that the DG is tested under load conditions that are as close to design conditions as possible, testing must be performed using a power factor of:* 0.87 lagging. This power factor is chosen to be representative of the actual design basis inductive loading that the DG would experience. The load band is provided to avoid routine overloading of the DG. Routine overloading may result in more frequent teardown inspections in accordance with vendor recommendations in order to maintain DG OPERABILITY.

The 24 month Fr uency is consistent with the intent of egulaor Guide 1.10ý,.ýef. 9), paragraph 2.a.(3), takes into c sideration unit conditions required to perform the Surveillance, d1is intended to beit consisaJ1wt xetdfIrrP1-qLj This Surveillance is modified by Note 1 which states tha't momentary transients due to changing bus loads do not invalidate this test.

Similarly, momentary power factor transients above the power factor limit will not invalidate the test.

Administrative controls for performing this SR in MODES 1 or 2, with the DG parelleled to an offsite power supply, ensure or require that:

a. Weather conditions are conducive to performing this SR.

b. The offiste power supply and switchyard conditions support performing this SR, including communicating with the transmission group responsible for the 230 kV and 500 kV switchyards to ensure that, during the DG testing, vehicle access to these switchyards is controlled and no elective maintenance or testing on the offsite power sources is performed potentially affecting:

0 230 kV and 500 kV systems (Exceptions are to be

.authorized by Operations Management) 0 Either units' 12 kV startup bus

Transformers or insulators

c. No equipment or systems assumed to be available for supporting the performance of the SR are removed from service.

(continued)

DIABLO CANYON - UNITS 1 & 2 Revision 4 8S9 IEAO4.DOA - R4a 24

AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.15 REQU IREMENTS This Surveillance demonstrates that the diesel engine can restart from (continued) a hot condition, such as subsequent to shutdown from normal Surveillances, and achieve stability by reaching the required voltage and frequency within 13 seconds. The 13 second time is derived from the requirements of the accident analysis to respond to a design basis accident. The acceptance criteria represents the recovery of the DG and the power distribution system following a start and load transient.

This assures the ability of the system to undergo further transients.

This SR is modified by two Notes. Note 1 ensures that the test is performed with the diesel sufficiently hot. The load band is provided to avoid routine overloading of the DG. Routine overloads may result in more frequent teardown inspections in accordance with vendor recommendations in order to maintain DG OPERABILITY. The requirement that the diesel has operated for at least 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months at full load conditions prior to performance of this Surveillance is based on test data and manufacturer recommendations for achieving hot conditions.

Momentary transients due to changing bus loads do not invalidate this test. Note 2 allows all DG starts to be preceded by an engine prelube period to minimize wear and tear on the diesel during testing.

SR 3.8.1.16 As required by Regulatory Guide 1.108 (Ref. 9), paragraph 2.a.(6), this Surveillance ensures that the manual synchronization and load transfer from the DG to the offsite source can be made and the DG can be returned to ready to load status when offsite power is restored. It also ensures that the autostart logic is reset to allow the DG to reload if a subsequent loss of offsite power occurs. The DG is considered to be in ready to load status when the DG is at rated speed and voltage, the output breaker is open and can receive an auto close signal on bus undervoltage, and the load sequencing timers are reset.

Frequncy. of 24 mnths is consiste~nt with the intent of Regulltory

- uie llKRef. 9), paragraph 2.a.(6), and kes into consi~rtion (continued)

DIABLO CANYON - UNITS 1 & 2 Revision 4 8S91EA04.DOA - R4a 25

AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.16 (continued)

REQUIREMENTS This SR is modified by a Note. The reason for the Note is that performing the Surveillance would remove a required offsite circuit from service, perturb the electrical distribution system, and challenge safety systems. This restriction from normally performing the Surveillance in MODE 1 or 2 is further amplified to allow the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines plant safety is maintained or enhanced. This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed Surveillance, a successful Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independenty for the Surveillance; as well as the operator procedures available to cope with these outcomes. These shall be measured against the avoided risk of a plant shutdown and startup to determine that plant safety is maintained or enhanced when the Surveillance is performed in MODE 1 or 2. Risk insights or deterministic methods may be used for this assessment.

Preplanned maintenance that would require the performance of this SR to demonstrate operability following the maintenance shall only be performed in Modes 5 or 6.

SR 3.8.1.17 Demonstration of the test mode override ensures that the DG availability under accident conditions will not be compromised as the result of testing. A Safety Injection signal, received while the DG is operating in a test mode, results in the auxiliary breaker opening and the emergency loads automatically sequencing onto the DG.

In lieu of actual demonstration of connection and loading of loads, testing that adequately shows the capability of the emergency loads to perform these functions is acceptable.

This testing may include any series of sequential, overlapping, or total steps so that the entire connection and loading sequence is verified.

\ 24 mp th Frequency is consistent with be intent of Regulatory hecniions required to perform the Surveillance, and is intended to be expected4 4nýtn-it fuel cycle lengths. -

(continued)

DIABLO CANYON - UNITS 1 & 2 Revision 4 8S9[EA04.DOA - R4a 26

AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.17 (continued)

REQUIREMENTS This SR is modified by a Note. The reason for the Note is that performing the Surveillance would remove a required offsite circuit from service, perturb the electrical distribution system, and challenge safety systems. This restriction from normally performing the Surveillance in MODE 1 or 2 is further amplified to allow portions of the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines plant safety is maintained or enhanced. This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed partial Surveillance, a successful partial Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the partial Surveillance; as well as the operator procedures available to cope with these outcomes. These shall be measured against the avoided risk of a plant shutdown and startup to determine that plant safety is maintained or enhanced when portions of the Surveillance are performed in MODE 1 or 2. Risk insights or deterministic methods may be used for this assessment.

Preplanned maintenance that would require the performance of this SR to demonstrate operability following the maintenance shall only be performed in Modes 5 or 6.

(continued)

DIABLO CANYON - UNITS 1 & 2 Revision 4 8S91EA04.DOA - R4a 27

AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.18 REQUIREMENTS Under accident and loss of offsite power conditions, loads are (contnued)sequentially connected to the bus by load sequencer timers. The sequencing logic controls the permissive and starting signals to motor breakers to prevent overloading of the DGs due to high motor starting currents. The load sequence time interval tolerances ensure that sufficient time exists for the DG to restore frequency and voltage prior to applying the next load and that safety analysis assumptions regarding ESE equipment time delays are not violated. The timing limits for the load sequence timers are found in table B3.8.1-1 (ESE Timers) and table B3.8.1-2 (Auto transfer Timers).

With an ESF timer found to be outside the range of acceptable settings, the corresponding DG shall be declared inoperable in MODES 1, 2, 3, and 4, and the corresponding CONDITION followed. With an Auto Transfer timer found to be outside the range of acceptable settings, the corresponding DG shall be declared inoperable for all MODES. This action is necessary only for that time required to open the breaker on the affected load.

with t .intent of Regulatory~

~ Guide 1.1 08 (Rf9), is consistent paragraph o~fmo~nths 2.a.(2), take;nto consideration unit EThe Frequency cciionditions reqFired t~o perform the Surveillance, and is intended to be consistent wiheecefulcceents This SR is modified by a Note. The reason for the Note is that performing the Surveillance would remove a required offsite circuit from service, perturb the electrical distribution system, and challenge safety systems. This restriction from normally performing the Surveillance in MODE 1 or 2 is further amplified to allow the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines plant safety is maintained or enhanced. This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed Surveillance, a successful Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independenty for the Surveillance; as well as the operator procedures available to cope with these outcomes. These shall be measured against the avoided risk of a plant shutdown and startup to determine that plant safety is maintained or enhanced when the Surveillance is performed in MODE 1 or 2. Risk insights or deterministic methods may be used for this assessment.

Preplanned maintenance that would require the performance of this SR to demonstrate operability following the maintenance shall only be performed in Modes 5 or 6.

(continued)

DIABLO CANYON -UNITS 1 & 2 Revision 4 8S9lEA04.DOA - R4a 28

AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.19 REQUIREMENTS In the event of a DBA coincident with a loss of offsite power, the DGs (continued) are required to supply the necessary power to ESE systems so that the fuel, RCS, and containment design limits are not exceeded.

This Surveillance demonstrates the DG operation, as discussed in the Bases for SR 3.8.1.11, during a loss of offsite power actuation test signal in conjunction with a Safety Injection signal. In lieu of actual demonstration of connection and loading of loads, testing that adequately shows the capability of the DG system to perform these functions is acceptable. This testing may include any series of sequential, overlapping, or total steps so that the entire connection and loading sequence is verified.

TeFrequency of 24 months takes into considyation unit conditpl*

required toye orm the Surveillance and is *endeid to be consten Swith an ex eecte~d fuel cylletho 24 months. J This SR is modified by two Notes. The reason for Note 1 is to minimize wear and tear on the DGs during testing. For the purpose of this testing, the DGs must be started from standby conditions, that is, with the engine coolant and oil temperature maintained consistent with manufacturer recommendations for DGs of equal to or greater than 90OF but less than 175*F. The reason for Note 2 is that the performance of the Surveillance would remove a required offsite circuit from service, perturb the electrical distribution system, and challenge safety systems. This restriction from normally performing the Surveillance in MODE 1 or 2 is further amplified to allow portions of the Surveillance to be performed for the purpose of reestablishing OPERABILITY (e.g., post work testing following corrective maintenance, corrective modification, deficient or incomplete surveillance testing, and other unanticipated OPERABILITY concerns) provided an assessment determines plant safety is maintained or enhanced. This assessment shall, as a minimum, consider the potential outcomes and transients associated with a failed partial Surveillance, a successful partial Surveillance, and a perturbation of the offsite or onsite system when they are tied together or operated independently for the partial Surveillance; as well as the operator procedures available to cope with these outcomes. These shall be measured against the avoided risk of a plant shutdown and startup to determine that plant safety is maintained or enhanced when portions of the Surveillance are performed in MODE 1 or 2. Risk insights or deterministic methods may be used for this assessment.

Preplanned maintenance that would require the performance of this SR to demonstrate operability following the maintenance shall only be performed in Modes 5 or 6.

(continued)

DIABLO CANYON - UNITS 1 & 2 Revision 4 8S91EA04.DOA - R4a 29

AC Sources - Operating B 3.8.1 BASES SURVEILL ANCE SR 3.8.1.20 REQUIREI VENTS This Surveillance demonstrates that the DG starting independence has (continue d) not been compromised. Also, this Surveillance demonstrates that each engine can achieve proper speed within the specified time when the DGs are started simultaneously.

... ~ The 110 yer Frequency frconsistent with the P66*mmendations of This SR is modified by a Note. The reason for the Note is to minimize wear on the DG during testing. For the purpose of this testing, the DGs must be started from standby conditions, that is, with the engine coolant and oil temperature maintained consistent with manufacturer recommendations of equal to or greater than 90OF but less than 175 0F.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 17.

2. FSAR, Chapter 8.

3. Regulatory Guide 1.9, Rev. 0, March 10, 1971 (Safety Guide 9).

4. FSAR, Chapter 6.

5. FSAR, Chapter 15.

6. Regulatory Guide 1.93, Rev. 0, December 1974.

7. Generic Letter 84-15, "Proposed Staff Actions to Improve and Maintain Diesel Generator Reliability," July 2, 1984.

8. 10 CFR 50, Appendix A, GDC 18.

9. Regulatory Guide 1.108, Rev. 1, August 1977.

10. Regulatory Guide 1.137, Rev. 1, Oct 1979.

11. ASME, Boiler and Pressure Vessel Code,Section XI.

12. Generic Letter 94-0 1,"Removal of Accelerated Testing and Special Reporting Requirements for Emergency Diesel Generators," May 31, 1994.-

13. Diesel Generator Allowed Outage Time Study, LA 44/43, October 4, 1989

14. License Amendment 44/43, October 4, 1989.

15. Regulatory Guide 1.9 Rev. 3, July 1993.

16. Regulatory Guide 1.9 Rev. 2, December 1979.

17. License Amendment 166/1 67, April 20, 2004.18. Calculation PRA 02-06, "Diesel Generator LAR for 14-day AOT."

19. License Amendment 174/176, September 28, 2004.

DIABLO CANYON - UNITS 1 & 2 Revision 4 8S9IEA04.DOA - R4a 30

Diesel Fuel Oil, Lube Oil, Starting Air, and Turbocharger Air Assist B 3.8.3 BASES ACTIONS F.1 (continued) adequate capacity for at least one start attempt, and the DG can be considered OPERABLE while the turbo air assist air receiver pressure is restored to the required limit. A period of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is considered sufficient to complete restoration to the required pressure prior to declaring the DG inoperable. This period is acceptable based on the remaining capacity, the fact that most DG starts are accomplished on the first attempt, and the low probability of an event during this brief period.

G.1 With a Required Action and associated Completion Time or Conditions E or F not met, or one or more DG's starting air, or turbocharger air assist subsystem not within limits for reasons other than addressed by Conditions E or F, the associated DG may be incapable of performing its intended function and must be immediately declared inoperable.

H.1. H.2. and H.3 With a Required Action and associated Completion Time not met, or the fuel oil storage tanks not within limits for reasons other than addressed by Conditions A, B, C, or D, the fuel oil storage tanks may be incapable of supporting the DGs in performing their intended function. This condition requires declaring inoperable, all the OGs on the unit(s) associated with either the inadequate fuel oil inventory, the fuel storage tank(s) having particulate outside the limit, and/or the fuel storage tank(s) having properties outside limits; and shutting down to MODE 3 in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and MODE 5 in 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months any associated unit(s) operating in MODE 1,2,3,or 4.

SURVEILLANCE SR 3.8.3.1 REQUIREMENTS This SR provides verification that there is an adequate inventory of fuel oil in the storage tanks to support DG operation for 7 days based on a realistic (minimum) ESF systems loading profile. The 7 day period is sufficient time to place the unit in a safe shutdown condition and to bring in replenishment fuel from an offsite location.

~ Th~e31 day Frequency is adequate to en ure that a sufficient =supply of SR 3.8.3.2 This Surveillance ensures that sufficient lube oil inventory is available to support at least 7 days of operation for each DG at minimum ESF systems loading. The 650 gal requirement is based on the DG (continued)

DIABLO CANYON -UNITS 1 & 2 Revision 4 8S91EA04.DOA -R4a 44

Diesel Fuel Oil, Lube Oil, Starting Air, and Turbocharger Air Assist B 3.8.3 BASES SURVEILLANCE SR 3.8.3.2 (continued)

REQUIREMENTS manufacturer consumption values for the run time of the DG at 1% of fuel oil consumption. The storage system used to meet this requirement is that located within the. warehouse where 650 gallons of lube oil is stored in drums.

~~ ~ A 31 day Feqenyi adequaet ensure that a sufimt lueoi supl is prsite , since DG sta and ruie aresc;eý2y onioeb the u SR 3.8.3.3 The tests listed below are a means of determining whether new fuel oil is of the appropriate grade and has not been contaminated with substances that would have an immediate, detrimental impact on diesel engine combustion. If results from these tests are within acceptable limits, the fuel oil may be added to the storage tanks without concern for contaminating the entire volume of fuel oil in the storage tanks. These tests are to be conducted prior to adding the new fuel to the storage tanks. The tests, limits, and applicable ASTM Standards are as follows:

a. Sample the new fuel oil in accordance with ASTM D4057-81 (Ref. 6);

b. Verify in accordance with the tests specified in ASTM D975-81 (Ref. 6) that the sample has an absolute specific gravity at 60/60'F of Ž!0.82 and *! 0.89 or an API gravity at 60OF of Ž! 270 and *5 420, a kinematic viscosity at 40*C of Ž1.9 centistokes and *ý4.1 centistokes, and a flash point of> 125 0F7; and

c. Verify that the new fuel oil has a clear and bright appearance with proper color when tested in accordance with ASTM D41 76 or a water and sediment content of *ý 0.05 volume percent when tested in accordance with ASTM D-1 796-83 (Ref. 6).

Failure to meet any of the above limits is cause for rejecting the new fuel oil, but does not represent a failure to meet the LCO concern since the fuel oil is not added to the storage tanks.

Within 31 days following the initial new fuel oil sample, the fuel oil is analyzed to establish that the other properties specified in Table 1 of ASTM D975-81 (Ref. 7) are met for new fuel oil when tested in accordance with ASTM D975-81 (Ref. 6), except that the analysis for sulfur may be performed in accordance with ASTM 0 1552-79 (Ref. 6) or ASTM D2622-82 (Ref. 6). The 31 day period is acceptable because the fuel oil properties of interest, even if they were not within stated limits, would not have an immediate effect on DG operation. This Surveillance ensures the availability of high quality fuel oil for the DGs.

(continued)

DIABLO CANYON -UNITS 1 & 2 Revision 4 8S91EA04.DOA - R4a 45

Diesel Fuel Oil, Lube Oil, Starting Air, and Turbocharger Air Assist B 3.8.3 BASES SURVEILLANCE SR 3.8.3.3_ (continued)

REQUIREMENTS If the analysis of the new fuel oil sample indicates that one or more of the other properties specified in Table 1 of ASTM D975-81 are not within limits, then Required Action DA1 shall be entered, allowing 31 days to restore fuel oil properties to within limits.

Fuel oil degradation during long term storage shows up as an increase in particulates, due mostly to oxidation. The presence of particulates does not mean the fuel oil will not burn properly in a diesel engine. The particulates can cause fouling of filters and fuel oil injection equipment, however, which can cause engine failure.

Particulate concentrations should be determined in accordance with ASTM D2276-78, Method A (Ref. 6). This method involves a gravimetric determination of total particulate concentration in the fuel oil and has a limit of 10 mg/I. It is acceptable to obtain a field sample for subsequent laboratory testing in lieu of field testing. Each tank must be considered and tested separately.

ASTM D 2276-78 was written specifically for aviation fuel. However, it is used in this SR to evaluate diesel fuel oil. Therefore, it may be necessary to perform this test as a modified method. For example, a 500 ml sample may be analyzed rather than a one gallon sample.

The Frequency of this test takes into consideration fuel oil degradation trends that indicate that particulate concentration is unlikely to change significantly between Frequency intervals.

SR 3.8.3.4 This Surveillance ensures that, without the aid of the refill compressor, sufficient air start capacity for each DG is available. The system design requirements provide for a minimum of three engine start cycles without recharging. Each start cycle is 15 seconds of cranking. The pressure specified in this SR is intended to reflect the lowest value at which three starts can be accomplished.

4: j 0 SR 3.8.3.5 Microbiological foulin g is a major cause of fuel oil degradation. There are numerous bacteria that can grow in fuel oil and cause fouling, but all must have a water environment in order to survive. Removal of water from the fuel storage tanks eige-. suei+/-'Udayeliminates the necessary environment for bacterial survival. This is the most effective (continued)

DIABLO CANYON - UNITS 1 & 2 Revision 4 8S9 1EA04.DOA - R4a 46

Diesel Fuel Oil, Lube Oil, Starting Air, and Turbocharger Air Assist B 3.8.3 BASES SURVEILLANCE SR 3.8.3.5 (continued)

~L~JFX~II~Pd means of controlling microbiological fouling. In addition, it eliminates the potential for water entrainment in the fuel oil during DG operation.

Water may come from any of several sources, including condensation, ground water, rain water, contaminated fuel oil, or from breakdown of the fuel oil by bacteria. Frequent checking for and removal of

~ ~ ae esablihed Re latoD uide 1.137 (Ref. 2. This SR iso preventiye maintenance. The presence of water does not necessarily represent failure of this SR, provided the accumulated water is removed during performance of the Surveillance.

SR 3.8.3.6 This Surveillance ensures that, without the aid of the refill compressor, sufficient turbocharger air assist air receiver capacity for each DG is available. The system design requirements provide for a minimum of three engine start cycles without recharging. Each start cycle is 15 seconds of cranking. The pressure specified in this SR is intended to reflect the lowest value at which three starts can be accomplished.

The 31 day Frequency takes into account the capacity, cppability, redundancy, and diversit the AC sources and nt~dications available in the control om, including alarms, to afert the operator t~o below a! r au e i sitarrciersue REFERENCES 1. FSAR, Section 9.5.4.2.

2. Regulatory Guide 1.137.

3. ANSI N195-1 976, Appendix B.

4. FSAR, Chapter 6.

5. FSAR, Chapter 15.

6. ASTM Standards: D4057-81; D975-81; D4176-82; D1796-83; D1 552-79; D2622-82; D2276-78, Method A.

7. ASTM Standards, D975, Table 1.

8. ASME, Boiler and Presser Vessel Code,Section XI.

9. License Amendment 74/73, August 12, 1992.

10. License Amendment 181/1 83, May 25, 2005.

11. AR A05661 59, AR A0512756, AR A0504056 DIABLO CANYON - UNITS 1 & 2 Revision 4 8S91EA04.DOA - R4a 47

DC Sources - Operating B 3.8.4 BASES ACTIONS D.l1 (continued) The design of the 125 VDC electrical power distribution system is such that a battery can have associated with it a dedicated full capacity charger powered from its associated 480 VAC vital bus or a backup full capacity charger powered from another 480 VAC vital bus. Use of the backup full capacity charger results in more than one full capacity charger receiving power simultaneously from a single 480 V vital bus and causes the requirements of independence and redundancy between subsystems to no longer be maintained. Thus, operation with two chargers powered by the same vital bus is limited to 14 days.

E.1 and E.2 If the inoperable DC electrical power subsystem cannot be restored to OPERABLE status within the required Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and to MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging plant systems. The Completion Time to bring the unit to MODE 5 is consistent with the time required in Regulatory Guide 1.93 (Ref. 7).

SURVEILLANCE SR 3.8.4.1 The minimum established float voltage provided by the battery manufacturer is 2.17 Vpc or 130.2 V at the battery terminals for a 60-cell battery. This voltage maintains the battery plates in a condition that supports maintaining the grid life (expected to be approximately 20 years). Verifying battery terminal voltage while on float charge for the batteries helps to ensure the effectiveness of the battery chargers, which support the ability of the batteries to perform their intended function. Float charge is the condition in which the charger is supplying the continuous charge required to overcome the internal losses of a battery and maintain the battery in a fully charged state, while supplying the continuous steady state loads of the associated DC electrical power subsystem. On float charge, batter cells will receive adequate current to op~timally charge the battery. rThe 7 day Frequency.

I-& (is consiste with man~ufa~ctutrer rec ~mendations aýpýEEE-450 (continued)

DIABLO CANYON - UNITS 1 & 2 Revision 4 8S91EA04.DOA - R4a 55

DC Sources - Operating B 3.8.4 BASES SURVEILLANCE SR 3.8.4.2 REQUIREMENTS This SR verifies the design capacity of the battery chargers. According (continued) to Regulatory Guide 1.32 (Ref. 9), the battery charger supply is recommended to be based on the largest combined demands of the various steady state loads and the charging capacity to restore the battery from the design minimum charge state to the fully charged state, irrespective of the status of the unit during these demand occurrences. The minimum required amperes and duration ensures that these requirements can be satisfied.

This SR provides two options. One option requires that each battery charger be capable of supplying 400 amps at the minimum established float voltage for greater than 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The ampere requirements are based on the output rating of the chargers. The voltage requirements are based on the charger voltage level after a response to a loss of AC power. The time period is sufficient for the.charger temperature to have stabilized and to have been maintained for at least 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

The other option requires that each battery charger be capable of recharging the battery after a service test coincident with supplying the largest combined demands of the various continuous steady state loads (irrespective of the status of the plant during which these demands occur). This level of loading may not normally be available following the battery service test and will need to be supplemented with additional loads. The duration for this test may be longer than the charger sizing criteria since the battery recharge is affected by float voltage, temperature, and the exponential decay in charging current.

The battery is recharged when the measured charging current is *ý 2 amps.

The Surveillance Frequency is acceptable, given the unit conditionsN required to perform t~e test and the other administr tve controls existing to ensurejadequate charger performance/during these7 24 month intervars. In addition, this Frequency is intended to be consistent with expected fuel cycle lengths.

(continued)

DIABLO CANYON - UNITS 1 & 2 Revision 4 8S91EA04.DOA - R4a 56

DC Sources - Operating B 3.8.4 BASES SURVEILL ANCE SR 3.8.4.3 REQUIRE~ EA -r (continue d A battery service test is a special test of battery capability, as found, to satisfy the design requirements (battery duty cycle) of the DC electrical power system. The discharge rate and test length should correspond to the design duty cycle requirements as specified in FSAR Chapter 8, (Ref. 4).

A bThe Surveillance Frequency of 24 months is consi 4ýnt with the intent of Regulatory GuideýY2 (Ref. 9) and Regulatpr Guide 1.129 (Reif. 10), which state that the battery serviW test should be erformed durnn :refueling operationortsmeOhrti This SR is modified by two Notes. Note 1 allows the performance of a modified performance discharge test in lieu of a service test.

The reason for Note. 2 is that performing the Surveillance would perturb the electrical distribution system and challenge safety systems.

REFERENCES 1. 10 CFR 50, Appendix A, GDC 17.

2. Regulatory Guide 1.6, March 10, 1971.

3. IEEE Std. 308-1971.

4. FSAR, Chapter 8.

5. FSAR, Chapter 6.

6. FSAR, Chapter 15.

7. Regulatory Guide 1.93, December 1974.

8. IEEE Std. 450-1995.

9. Regulatory Guide 1.32, February 1977.

10. Regulatory Guide 1.129, December 1974.

11. Electrical Design Calculations 235A-DC thru 23SF-DC.

DIABLO CANYON - UNITS 1 & 2 Revision 4 8S9 IEA04.DOA - R4a 57

Battery Parameters B 3.8.6 BASES (continued)

SURVEILLANCE SR 3.8.6.1 REQUIREMENTS Verifying battery float current while on float charge is used to determine the state of charge of the battery. Float charge is the condition in which the charger is supplying the continuous charge required to overcome the internal losses of a battery and maintain the battery in a charged state. The float current requirements are based on the float current indicative of a charged battery. Use of float current to determine the state of charge of the battery is consistent with IEEE-450

k. ~I.

.,~,e~q -. ~ (ef. 3j e a uecicnisp4i -50 .3 This SR is modified by a Note that states the float current requirement is not required to be met when battery terminal voltage is less than the minimum established float voltage of SR 3.8.4.1. When this float voltage is not maintained the Required Actions of LCO 3.8.4 Action A are being taken, which provide the necessary and appropriate verifications of the battery condition. Furthermore, the float current limit of 2 amps is established based on the nominal float voltage value and is not directly applicable when this voltage is not maintained.

SR 3.8.6.2 and 3.8.6.5 Optimal long term battery performance is obtained by maintaining a float voltage greater than or equal to the minimum established float voltage provided by the battery manufacturer, which corresponds to 130.2 V for 60 cells at the battery terminals, or 2.17 Vpc. This provides adequate over-potential, which limits the formation of lead sulfate and self discharge, which could eventually render the battery inoperable.

Float voltages in the range of less than 2.13 Vpc, but greater than 2.07 Vpc, are addressed in Specification 5.5.17. SRs 3.8.6.2 and 3.8.6.5 require verification that the cell float voltages are equal to or reater

"'\92 days for eap connected cell is conse ent with EEE-4E (Ref. 3).

SR 3.8.6.3 The electrolyte level minimum established design limit is the manufacturer minimum level indication mark on the battery case. The limit specified for electrolyte level ensures that the plates suffer no physical daae n maintains adequate electron transfer capability.

e Fre Ocy is consistent wt (continued)

DIABLO CANYON -UNITS 1 & 2 Revision 4 8S9 IEA04.DOA - R4a 67

Battery Parameters B 3.8.6 BASES SURVEILLANCE SR 3.8.6.4 (contInuEd)NT This Surveillance verifies that the pilot cell temperature .isgreater than (contnued)or equal to the minimum established design .limit (i.e. 601F). Pilot cell electrolyte temperature is maintained above this temperature to assure the battery can provide the required current and voltage to meet the design requirements. Temperatures lower than assumed in battery sizing calculations act to inhibit or reduce battery capacity. eTiT

_X-IZ ý c,-*: Frequency is consistept-with IEEE-450 (Ref. 3). -'

SR 3.8.6.6 A battery performance discharge test is a test of constant current capacity of a battery, normally done in the as found condition, after having been in service, to detect any change in the capacity determined by the acceptance test. The test is intended to determine overall battery degradation due to age and usage.

Either the battery performance discharge test or the modified performance discharge test is acceptable for satisfying SR 3.8.6.6; however, only the modified performance discharge test may be used to satisfy the battery service test requirements of SR 3.8.4.3.

A modified performance discharge test is a test of the battery capacity and its ability to provide a high rate, short duration load (usually the highest rate of the duty cycle). This will often confirm the battery's ability to meet the critical period of the load duty cycle, in addition to determining its percentage of rated capacity. Initial conditions for the modified performance discharge test should be identical to those specified for a service test. The modified performance discharge test and service test should be performed in accordance with IEEE-450 (Ref. 3).

It may consist of just two rates; for instance the one minute rate published for the battery or the largest current load of the duty cycle, followed by the test rate employed for the performance test, both of which envelope the duty cycle of the service test. Since the ampere-hours removed by a one minute discharge represents a very small portion of the battery capacity, the test rate can be changed to that for the performance test without compromising the results of the performance discharge test. The battery terminal voltage for the modified performance discharge test must remain above the minimum battery terminal voltage specified in the battery service test for the duration of time equal to that of the service test.

(continued)

DIABLO CANYON -UNITS 1 & 2 Revision 4 8S91EA04.DOA - R4a 68

Battery Parameters B 3.8.6 BASES SURVEILLANCE SR 3.8.6.6 (continued)

REQUIREMENTS The acceptance criteria for this Surveillance are consistent with IEEE-450 (Ref. 3) and IEEE-485 (Ref. 5). These references recommend that the battery be replaced if its capacity is below 80% of the manufacturer's rating. A capacity of 80% shows that the battery rate of deterioration is increasing, even if there is ample capacity to meet the load requirements. Furthermore, the battery is sized to meet the assumed duty cycle loads when the battery design capacity reaches this 80% limit.

3ur;3lk rqu, ic ruf this 1teStL- .ýol F-fe 0mrlS. f the battery shows degradation, or if the battery has reached 85% of its expected service life and capacity is < 100% of the manufacturer's rating, the Surveillance Frequency is reduced to 24 months. However, if the battery shows no degradation but has reached 85% of its expected life, the Surveillance Frequency is only reduced to 24 months for batteries that retain capacity Ž: 100% of the manufacturer's rating.

Degradation is indicated, according to IEEE-450 (Ref. 3), when the battery capacity drops by more than 10% relative to its capacity on the previous performance test or when it is < 90% of the manufacturer's rating. The Surveillance Frequency basis is consistent with IEEE-450

.(Ref. 3), except if accelerated testing is required, it will be performed at a 24-month frequency to coincide with a refueling outage.

This SR is modified by a Note. The reason for the Note is that performing the Surveillance would perturb the electrical distribution system and challenge safety systems.

REFERENCES 1. FSAR, Chapter 6.

2. FSAR, Chapter 15.

3. IEEE Std. 450-1995.

4. FSAR, Chapter 8.

5. IEEE Std. 485-1983.

DIABLO CANYON - UNITS 1 & 2 Revision 4 8S9lEA04.DOA - R4a 69

Inverters - Operating B 3.8.7 BASES ACTIONS A.1 (continued)

AC electrical power sources (offsite and onsite). The uninterruptible inverter source to the 120 VAC vital buses is the preferred source for powering instrumentation trip setpoint devices.

B.1 and B.2 If the inoperable devices or components cannot be restored to OPERABLE status within the required Completion Time, the unit must be brought to a MODE in which the LCO does not apply. To achieve this status, the unit must be brought to at least MODE 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months SURVEILANCE and to MODE 5 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging plant systems.

SR 3.8.7.1 REQUIREMENTS This Surveillance verifies that the inverters are functioning properly with all required circuit breakers closed and 120 VAC vital buses energized from the inverter. The verification of proper voltage output ensures that the required power is readily available for the instrumentation of the RPS and ESFAS connected to the AC vital buses. Ihe 7 day FTreq-ue-ncy =0~e into apount -theredundant capabilityp.1te inverters and other iniaipavailable in the control room thM' alert the REFERENCES 1. FSAR, Chapter 7.

2. FSAR, Chapter 6.

3. FSAR, Chapter 15.

DIABLO CANYON - UNITS 1 & 2 Revision 4 8S9 IEA04.DOA - R4a 72

Inverters - Shutdown B 3.8.8 BASES ACTIONS A. 1, A.2. 1, A.2.2, A.2.3, and A.2.4 (continued)

Suspension of these activities shall not preclude completion of actions to establish a safe conservative condition. These actions minimize the probability of the occurrence of postulated events. It is further required to immediately initiate action to restore the required Class 1E UPS inverters and to continue this action until restoration is accomplished in order to provide the necessary Class 1E UPS inverter power to the unit safety systems.

The Completion Time of immediately is consistent with the required times for actions requiring prompt attention. The restoration of the required inverters should be completed as quickly as possible in order to minimize the time the unit safety systems may be without power or powered from a constant voltage source transformer.

SURVEILLANCE SR 3.8.8.1 REQUIREMENTS This Surveillance verifies that the inverters are functioning properly with all required circuit breakers closed and AC vital buses energized from the inverter. The verification of proper voltage output ensures that the required power is readily available for the instrumentation connected to the 120 VAC vtlbses. [Te 7 day Frequency takeý into- accouint the rFedunid-ant capabiliyofthe inverters and other in ations avai5061,in the control roomnv-Mat alert the operator to inverr$'rr malfunctions.

REFERENCES 1. FSAR, Chapter 6.

2. FSAR, Chapter 15.

3. License Amendment 184/1 86, -January 3, 2006.

DIABLO CANYON - UNITS 1 & 2 Revision 4 8S91EA04.DOA - R4a 75

Distribution Systems - Operating B 3.8.9 BASES ACTIONS E.1 (continued) Condition E corresponds to two required Class 1lE AC, DC, or 120 VAC vital buses with inoperable distribution subsystems that result in a loss of safety function, adequate core cooling, containment OPERABILITY and other vital functions for DBA mitigation would be compromised, and immediate plant shutdown in accordance with LCO 3.0.3 is required.

SURVEILLANCE SR 3.8.9.1 I '~LA.,(~.JII ~L This Surveillance verifies that the required Class 1E AC, DC, and 120 VAC vital bus electrical power distribution systems are functioning properly, with the correct circuit breaker alignment. The correct breaker alignment ensures the appropriate separation and independence of the electrical divisions is maintained, and the appropriate voltage is available to each required bus. The verification of proper voltage availability on the buses ensures that the required voltage is readily available for motive as well as control functions for critical system loads co'nnected to these buses 7e 7 -dayFrequency account into vital a esVAC the redundant capability of the AC, DC, arV IN12 bus eJ 6 irical power distributýsubsystemp <,and other

-Týefrk Lindications avaifl;V<in the control room '1at alert the operator to subsystem malfuntions. _

Table B 3.8.9-1 The table on the next page defines the general features of the AC and DC Electrical Power Distribution System.

REFERENCES 1. FSAR, Chapter 6.

2. FSAR, Chapter 15.

3. Regulatory Guide 1.93, December 1974.

DIABLO CANYON - UNITS 1 & 2 Revision 4 8S91EA04.DOA - R4a 83

Distribution Systems - Shutdown B 3.8.10 BASES (continued)

SURVEILLANCE SR 3.8.10.1 REQUIREMENTS This Surveillance verifies that the Class 1lE AC, DC, and 120 VAC vital bus electrical power distribution subsystems are functioning properly, with all the buses energized. The verification of proper voltage availability on the buses ensures that the required power is readily available for motive as well as control functions for critical system loads connected to these buses. Whie 7 day Frequcytksioacun the caabiliýt of the electrical power disriobn su bsys>Pt* and other]

inica-to available in the control room that alertthoproro subsys emn malfunctions.

REFERENCES 1. ESAR, Chapter 6.

2. FSAR, Chapter 15.

3. License Amendment 184/1 86, January 3, 2006.

DIABLO CANYON - UNITS 1 & 2 Revision 4 8S91EA04.DOA - R4a 88

Boron Concentration B 3.9.1 BASES (continued)

ACTIONS A.1 and A.2 Continuation of CORE ALTERATIONS or positive reactivity additions (including actions to reduce boron concentration) is contingent upon maintaining the unit in compliance with the LCO. If the boron concentration of any coolant volume in the RCS, and when connected, the refueling canal or the refueling cavity is less than its limit, all operations involving CORE ALTERATIONS or positive reactivity additions must be suspended immediately.

Suspension of CORE ALTERATIONS and positive reactivity additions shall not preclude moving a component to a safe position. Operations that individually add limited positive reactivity (e.g., temperature fluctuations, inventory addition, or temperature control fluctuations), but when combined with all other operations affecting core reactivity (e.g.,

intentional boration) result in overall net negative reactivity addition, are not precluded by this action.

A.3 In addition to immediately suspending CORE ALTERATIONS and positive reactivity additions, boration to restore-the concentration must be initiated immediately.

In determining the required combination of boration flow rate and concentration, no unique Design Basis Event must be satisfied. The only requirement is to restore the boron concentration to its required value as soon as possible. In order to raise the boron concentration as soon as possible, the operator should begin boration with the best source available for unit conditions.

Once actions have been initiated, they must be continued until the boron concentration is restored. The restoration time depends on the amount of boron that must be injected to reach the required concentration.

SURVEILLANCE SR 3.9.1.1 REQUIREMENTS This SR ensures that the coolant boron concentration in the filled portions of the RCS, the refueling canal, and the refueling cavity that have direct access to the reactor vessel is within the COLR limits. The boron concentration of the coolant in each required volume is determined periodically by chemical analysis.

Aminimum Frequency oonce every 72 nours is a reasonable amount 3\$gAz~ 1 -b of time to verify the ron concentration of repres~e~pttive samples.

The Frequency *a sed on operating experien~e~which has shown 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to be adequate:

REFERENCES 1. 10 CFR 50, Appendix A, GDC 26.

2. FSAR, Chapter 15, Section 15.2.4 DIABLO CANYON -UNITS 1 & 2 Revision 4 8S91EB04.DOA - R4 4

Nuclear Instrumentation B 3.9.3 BASES ACTIONS A.1 and A.2 (continued)

The core coupling in this configuration would allow one source range detector to detect significant reactivity changes associated with control rod movement (Ref. 3). Performance of Required Action A.1 shall not preclude completion of movement of a component to a safe position or normal cooldown of a coolant volume for the purpose of system temperature control.

B.1 With no source range neutron flux monitor OPERABLE including no OPERABLE audible alarm and count rate functions, action to restore a monitor to OPERABLE status shall be initiated immediately. Once initiated, action shall be continued until a source range neutron flux monitor including no OPERABLE audible alarm and count rate functions is restored to OPERABLE status.

B.2 With no source range neutron flux monitor OPERABLE, there are no direct means of detecting changes in core reactivity. However, since CORE ALTERATIONS and boron concentration changes inconsistent with Required Action A.2 are not to be made, the core reactivity condition is stabilized until the source range neutron flux monitors are OPERABLE. This stabilized condition is determined by performing SR 3.9.1.1 to ensure that the required boron concentration exists.

The Completion Time of once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ensures that unplanned changes in boron concentration would be identified. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is reasonable, considering the low probability of a change in core reactivity during this time period.

SURVEILLANCE SR 3.9.3.1 REQUIREMENTS SR 3.9.3.1 is the performance of a CHANNEL CHECK, which is a comparison of the parameter indicated on one channel to a similar parameter on other channels. It is based on the assumption that the two indication channels should be consistent with core conditions.

Changes in fuel loading and core geometry can result in significant differences between source range channels, but each channel should be consistent with its local conditions. For core reload, the first CHANNEL CHECK for each channel may be performed using the first fuel assembly as a source, prior to unlatching it in the core.

A The Frequency oft2hours is consistent with theCHANNEL CHECK Frequency spdfe similarly for the same instruprifits in LCQ9.3-3r1.

(continued)

DIABLO CANYON - UNITS 1 & 2 Revision 4 8S91EB04.DOA - R4 7

Nuclear Instrumentation B 3.9.3 BASES SURVEI LLANCE SR 3.9.3.2 REQU IREMENTS SR 3.9.3.2 is the performance of a CHANNEL CALIBRATION eyeip-1`

(continued) P-4-,Pmith?-his SR is modified by a Note stating that neutron detectors are excluded from the CHANNEL CALIBRATION. The CHANNEL CALIBRATION for the normal N31 and N32 source range neutron flux monitors is described in B 3.3.1, "Reactor Trip System (RTS) Instrumentation." The CHANNEL CALIBRATION for the normal N31 and N32 audible alarm and count rate functions includes verification of the control room audible alarm and count rate functions using a simulated signal. TFhe 24 month Frequency is based on the

, n eed to performtthis Surveillance underfe conditions thalpply during REFERENCES 1. 10 CFR 50, Appendix A, GDC 13, GDC 26, GDC 28, and GDC 29.

2. FSAR, Section 15.2.4.

3. License Amendment 46/45, October 4, 1989.

4. NRC letter, "Diablo Canyon Nuclear Power Plant, Unit Nos.1 and 2 - Technical Specification Bases Change (TAC Nos.

M98430 and M98431)," June 9, 1998.

DIABLO CANYON - UNITS 1 & 2 Revision 4 8S9 IEB04.DOA - R4 8

Containment Penetrations B 3.9.4 BASES (continued)

SURVEILLANCE SR 3.9.4.1 REQU IREMENTS This Surveillance demonstrates by inspection or administrative means that each of the containment penetrations is closed or capable of being closed. The Surveillance on the open purge and exhaust valves will demonstrate that the valves are not blocked from closing. Also the Surveillance will demonstrate that each valve operator has motive power, which will ensure that each valve is capable of being closed by an OPERABLE automatic containment purge and exhaust isolation signal.

The Surveillance is performed every 7 days during CORE ALTERATIONS or mov~enent of irradiated fuel as 'mblies within I' cntaiment TheSu illance interval is slec d obe commensurate with e normal duration of t!i. to cComplete fuele handling op/erati s. A surveillance before e start of refueling operations wil rovide two or three surve' ance verifications during the applicable niod for this LCO. As suc , this Surveillance ensures that a postulated fuel handling accident t t releases fission product radioactivity within the containment will not result in a release of fission rodut rdioctiityto he nvionmnt hatexcedsacc tale limits.

SR 3.9.4.2 This Surveillance demonstrates that each containment purge and exhaust valve actuates to its isolation position on manual initiation or (Frequency maintains coni( nywt other similar ESFAS Xinstrumentation and va e testing requirements. In LCO 3.3.6, the Containmen Purge d.Exhaust Isolation instrumegtation requires a A ---

bCHANNEL CHEC every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and a CFT e ry 92 days to ensure

,T w ý,e4 t- the channel

/OP BILITY during refueling o rations. Every 24 months a ANNEL CALIBRATION is prformed. The system actuation re onse time is demonstrated ery 24 months, during refueling, a STAGGERED TEST BA S. SIR 3.6.3.5 demonstrates that the i olation time of each valve is~ accordance with the Inservice Testing Program requirements. Th e Survei~llances performed during MODE 6 will ensure that the valves are capable of closing after aI postulated fuel handling accident to limit a release of fission prod uct radioactivity from the containment.

fcontinued)

DIABLO CANYON - UNITS 1 & 2 Revision 4 8S9 IEB04.DOA - R4 14

RHR and Coolant Circulation - High Water Level B 3.9.5 BASES (continued)

SURVEILLAN' CE SR 3.9.5.1 REQUIREMEI *' This Surveillance demonstrates that the RHR loop is in operation and circulating reactor coolant. The flow rate of 3000 gpm is determined by the flow rate necessary to provide sufficient decay heat removal capability and to prevent thermal and boron stratification in the core prior to 57 hours6.597222e-4 days 0.0158 hours 9.424603e-5 weeks 2.16885e-5 months of core subcriticality. The second part of this Surveillance serves the same function but with 57 hours6.597222e-4 days 0.0158 hours 9.424603e-5 weeks 2.16885e-5 months or more of core subcriticality. The flow rate of 1300 gpm is determined by the flow rate necessary to provide sufficient decay heat removal capability and to prevent thermal and boron stratification in the core. Both of these flow rates are points of the same flo w rate verses decay heat. FTiTe A .~Frequency of 12 hrs is sufficient, considering tJ6 flow, temp rture pump control, a alarm indications available tethe operatorfn the control room fr monitoring the RHR sem Re. )

REFERENCES 1. FSAR, Section 5.5.7.

2. License Amendment 28/27, January 5, 1988.

DIABLO CANYON - UNITS 1 & 2 Revision 4 8S9 IEBO4.DOA - R4 19

RHR and Coolant Circulation - Low Water Level B 3.9.6 BASES ACTIONS B.3 (continued) If no RHR loop is in operation, all containment penetrations providing direct access from the containment atmosphere to the outside atmosphere must be closed within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . With the RHR loop requirements not met, the potential exists for the coolant to boil and release radioactive gas to the containment atmosphere. Closing containment penetrations that are open to the outside atmosphere ensures that dose limits are not exceeded.

The Completion Time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is reasonable at water levels above reduced inventory, based on the low probability of the coolant boiling in that time. At reduced inventory conditions or mid-loop operations, additional actions are taken to provide containment closure in a reduced period of time (Ref. 3). Reduced inventory is defined as less than Elev. 111 ft.

SR 3.9.6.1 REQUIREMENTS This Surveillance demonstrates that one RHR loop is in operation and circulating reactor coolant. The flow rate of more than 3000 gpm is determined by the flow rate necessary to provide sufficient decay heat removal capability and to prevent thermal and boron stratification in the core prior to 57 hours6.597222e-4 days 0.0158 hours 9.424603e-5 weeks 2.16885e-5 months subcritical. The second part of this Surveillance serves the same function but with 57 hours6.597222e-4 days 0.0158 hours 9.424603e-5 weeks 2.16885e-5 months or more of core subcriticality and provides a reduced flow rate of 1300 gpm based upon a reduced decay heat load. Both of these flow rates are points of the same flow rate verses decay heat curves. The 1300 gpm limit also precludes exceeding the 1675 gpm upper flow limit to prevent vortexing and air entrainment of the RHR piping system. RHR pump vortexing (failure to meet pump suction requirements) during mid-loop operation may result iný um Ifailur n n~-conservatv R Sle l

ýRrHR indication. te requenc of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is sufficient, consid i~rng the SR 3.9.6.2 Verification that the required pump is OPERABLE ensures that an additional RHR pump can be placed in operation, if needed, to maintain decay heat removal and reactor coolant circulation.

Verification is performed by verifying proper breae alig ne n an asben shown to) acceptable by o erating ex ience. )

(continued)

DIABLO CANYON - UNITS 1 & 2 Revision 4 8S9 1EB04.DOA - R4 22

Refueling Cavity Water LeveiB 3.9.7 BASES (continued)

APPLICABILITY LCO 3.9.7 is applicable during CORE ALTERATIONS, except during latching and unlatching of control rod drive shafts, and when moving irradiated fuel assemblies within containment. The LCO minimizes the possibility of a fuel handling accident in containment that is beyond the assumptions of the safety analysis. If irradiated fuel assemblies are not present in containment, there can be no significant radioactivity release as a result of a postulated fuel handling accident.

Requirements for fuel handling accidents in the spent fuel pool are covered by LCO 3.7.15, "Fuel Storage Pool Water Level."

ACTIONS A.1 With a water level of < 23 ft: above the top of the reactor vessel flange, all operations involving movement of irradiated fuel assemblies within the containment shall be suspended immediately to ensure that a fuel handling accident cannot occur.

The suspension of fuel movement shall not preclude completion of movement of a component to a safe position.

SURVEILLANCE SR 3.9.7.1 REQUIREMENTS Verification of a minimum water level of 23 ft above the top of the reactor vessel flange ensures that the design basis for the analysis of the postulated fuel handling accident during refueling operations is met. Water at the required level above the top of the reactor vessel flange limits the consequences of damaged fuel rods that are postulated to result from a fuel handling accident inside containment (Ref. 2).

The Frequencyo 24 ours is based on engineer~ing judgmen~t and is1 considered ýadequ in view of the large volume of ater and the, normal procedu controls of valve positions, wl^~ make signi rcant un lanned 1=v I cha~n es unlikelyj REFERENCES 1. Regulatory Guide 1.25, March 23, 1972.

2. FSAR, Section 15.4.5.

3. NUREG-0800, Section 15.7.4.

4. 10OCFR 100.10.

5. Malinowski, D. D., Bell, M. J., Duhn, E., and Locante, J.,

WCAP-828, Radiological Consequences of a Fuel Handling Accident, December 1971.

DIABLO CANYON - UNITS 1 & 2 Revision 4 8S91EB04.DOA - R4 25

Enclosure 4 PG&E Letter DCL-07-097 SURVEILLANCE TEST INTERVAL EVALUATION FORMS 1

Enclosure 4 PG&E Letter DCL-07-097 Surveillance Test Interval (STI) Evaluation Process The attached STI evaluation forms were prepared following the guidance in NEI 04-10, Revision 1.

Selection of the Example STIs The three attached STIs were selected to demonstrate the methodology for Risk Informed TS Initiative 5b.

Input was received from various site organizations, including Engineering, Operations, and Licensing. The following criteria, as listed in NEI 04-10, Revision 1, was considered.

Safety risk.

Reactivity management.

Maintaining dose as low as reasonably achievable (ALARA).

Burden reduction, including consideration of cost of the test (resources).

Outage impact (outage work control).

Work management. simplification (on-line work control).

Production risk.

Reducing wear and tear on the structure, system, or component.

Reducing potential for test-caused errors.

Difficulty of the test and potential for error during the test and its consequence.

Consideration of the role of the test on the reliability of the associated function.

Maintenance Rule Al item that has an associated action plan that necessitates more frequent testing.

Maintenance Rule and the associated corrective action process that necessitates more frequent testing.

A list of candidate STIs was compiled and grouped into the three categories below related to the DCPP PRA model:

1. PRA Modeled

2. PRA not modeled but could be modeled

3. PRA modeling not practical (qualitative evaluation only).

Figure 1 in NEI 04-10 provides a process flow map for the SFCP. The process includes three different branches to follow depending on how PRA is modeled, as shown in Figure 1 below in steps 8, 9, and 10:

2

Enclosure 4 PG&E Letter DCL-07-097 Three process paths are identified on the marked up figure below as Path (1), Path (2),

and Path (3)..

3

Enclosure 4 PG&E Letter DCL-07-097 Path (1) Path (2) Path (3)

Figure 1. Surveillance Frequency Control Program Change Process 4

Enclosure 4 PG&E Letter DCL-07-097 The surveillances selected to demonstrate the functionality of the three process paths are:

TS SR 3.3.5.2 - Diesel Generator Trip Actuating Device Operational Test (Path 1)

TS SR 3.6.3.3 - Containment Isolation Manual Valve Sealed Checklist (Path 2)

TS SR 3.1.2.1 - Verify Measured Core Reactivity (Path 3)

PG&E conducted an Independent Decision Making Panel (IDP) on September 24, 2007, using the guidance in NEI 04-10, Revision 1.

The IDP reviewed proposed changes to TS SR 3.3.5.2 and TS SR 3.1.2.1. The STI evaluation forms, including IDP comments and conclusions are attached. The STI evaluation form for TS SR 3.6.3.3 has not been reviewed by the IDP.

PG&E will conduct another IDP panel for the NRC to observe, that will review TS SR 3.6.3.3 and discuss the two previously reviewed STIs.

5

DIABLO CANYON POWER PLANT SURVEILLANCE TEST INTERVAL EVALUATION FORM

>X. S;5TRVE 11,TANCE TEST I~NFORMVATIOUN'

1. Unit(s): I &2

2. Surveillance Test (ST) Number (s) / Revision Number (s)

STP R-4, Revision 13A

3. Technical Specification Surveillance Requirement (SR) Number(s):

SR 3.1.2.1

4. Technical Specification SR (Text):

NOTE---------------------------

The predicted reactivity values may be adjusted (normalized) to correspond to the measured core reactivity prior to exceeding a fuel bumnup of 60 effective full power days (EFPD) after each fuel loading.

Verify measured core reactivity is within +/- I1%Ak/k of predicted values.

5. Technical Specification SR Bases (and Intent):

Core reactivity is verified by periodic comparisons of measured and predicted RCS boron concentrations. The comparison is made, considering that other core conditions are fixed or stable, including control rod position, moderator temperature, fuel temperature, fuel depletion, xenon concentration, and samarium concentration. The Surveillance is performned prior to entering MODE 1 as an initial check on core conditions and design calculations at BOC. The SR is modified by a Note. The Note indicates that the normalization (adjustment, only if necessary) of predicted core reactivity to the measured value must take place within the first 60 effective full power days (EFPD) after each fuel loading. This allows sufficient time for core conditions to reach steady state, but prevents operation for a large fraction of the fuel cycle without establishing a benchmark for the design calculations. The required subsequent Frequency of 31 EFPD, following the initial 60 EFPD after entering MODE 1, is acceptable, based on the slow rate of core changes due to fuel depletion and the presence of other indicators (QPTR, AFD, etc.) for prompt indication of an anomaly.

6. Recommended ST Frequency Change:

From: 31 EFPD To: 92 EFPD Pagel1 of8

Note: The terms Surveillance Test Interval (STI) and ST Frequency are used interchangeably.

7. Station Benefit:

Each performnance of STP R-4 requires approximately 5 reactor engineering man hours.

An RCS boron sample is required at nominal hot full power conditions for the surveillance test.

VA. SNSTEMt& MWIALNENCE RULE (MRttle)IIORIMATIO'N

1. SYSTEM NUMBER:

95

2. SYSTEM DESCRIPTION: Nuclear Fuel

3. CURRENT MRULE RISK SIGNIFICANCE (R-S) CLASSIFICATION:

Risk Significant

4. CURRENT MRULE R-S BASIS:

N.A. This is not modeled via the site Probalistic Risk Analysis (PRA) model. It is considered risk significant because it is related to a fission product barrier.

5. Current PRA RAW (System): Not Modeled (MRule R-S threshold: > 2.0)

6. Current PRA RRW (System): Not Modeled (MRule R-S threshold: > 1.005)

Page 2 of 8

7. Current PRA Limiting Sequences: Not Modeled (MRule R-S threshold: top 90%; Trigger value: N/A B. QUALITATIE ANALYSIS:

1. COMMITMENT REVIEW (Is STI credited in any commitments?) No.

The Procedure Commitment Database shows no active commitment data records for the implementing document STP R-4.

The updated Final Safety Analysis Report (FSAR) does not discuss the frequency of the core reactivity balance surveillance.

2. SURVEILLANCE TEST HISTORY OF THE COMPONENTS AND SYSTEM ASSOCIATED WITH THE STI ADJUSTMENT:

A review of STP R-4 surveillance test history was performed for Unit 1 and Unit 2 over cycles 12, 13, and 14. No surveillance tests failures were noted.

Unit and cycle-specific surveillance test histories are tabulated below by surveillance date and core average bumnup. Reactivity deviations are measured in units of pcm, where I% A"/= 1000 pcm.

The data for Unit 2 Cycle 12 show an outlier point indicating a reactivity deviation of 417 pcm on 6/25/04. This cause of this larger than expected reactivity deviation is attributed to not having acquired a reactor coolant system boron sample for 13- 1 analysis. Instead, the depleted (measured) concentration was compared to the undepleted predictions.

3. RELIABILITY REVIEW:

PERFORMANCE (OPERATION & MAINTENANCE) HISTORY OF THE COMPONENTS AND SYSTEM ASSOCIATED WITH THE STI ADJUSTMENT:

Maintenance Rule Train Actual Unreliability: A fuel leak on Unit 2 is below Action Level "1I"of TS6. IDl1.

Maintenance Rule Unreliability Performance Criteria: Fuel failure that results in Action Level 2 per TS6.I13l.

Additional component history review: N/A Page 3 of 8

4. UNAVAILABILITY REVIEW:

Maintenance Rule Train Actual Unavailability: 0 hours0 days 0 hours 0 weeks 0 months Maintenance Rule Unavailability Performance Criteria: 0 hours0 days 0 hours 0 weeks 0 months / year based on failed fuel integrity exceeding T.S. limits resulting in a plant shutdown.

5. PAST INDUSTRY AND PLANT-SPECIFIC EXPERIENCE WITH THE FUNCTIONS AFFECTED BY THE PROPOSED CHANGES Westinghouse Technical Bulletin TB-04-1 6, "Updated Reactivity Surveillance Policy for B' 0 Isotopic Concentration," discusses the deviations between measured and predicted critical boron concentration exacerbated by not accounting for boron- 10 0

depletion during the operating cycle and refueling outages. Not accounting for B'1 depletion can lead to errors in shutdown margin calculations, estimated critical conditions during a reactor startup, and routine reactivity balance calculations.

Consistent with the recommendations of the Technical Bulletin, Chemistry periodically samples the reactor coolant system (RCS), boric acid storage tank, and refueling water storage tank to determine the boron isotopic ratio. Reactor Engineering uses the measured B' 0 ratio to correct the measured RCS boron concentration for B' 0 depletion for use in reactivity balance calculations. Diablo Canyon implemented an improved B' 0 depletion methodology for reactivity balance calculations in July 2004 and each surveillance since this time has shown a measured-to-predicted deviation of no more than 200 pcm.

6. VENDOR-SPECIFIED MAINTENANCE FREQUENCY N/A.

7. TEST INTERVALS SPECIFIED IN APPLICABLE INDUSTRY CODES AND STANDARDS The test interval is only specified in the Technical Specifications.

Page 4 of 8

8. 8.OTHER QUALITATIVE CONSIDERATIONS From TS 3.1.2 Bases:

"Therefore, reactivity balance is used as a measure of the predicted versus measured core reactivity during power operation. The periodic confirmation of core reactivity is necessary to ensure that Design Basis Accident (DBA) and transient safety analyses remain valid. A large reactivity difference could be the result of unanticipated changes in fuel, control rod worth, or operation at conditions not consistent with those assumed in the predictions of core reactivity, and could potentially result in a loss of SDM or violation of acceptable fuel design limits. Comparing predicted versus measured core reactivity validates the nuclear methods used in the safety analysis and supports the SDM demonstrations (LCO 3. 1.1, "SHUTDOWN MARGIN (SDM)") in ensuring the reactor can be brought safely to cold, subcritical conditions."

9. QUALITATIVE ANALYSIS - CONCLUSIONS The surveillance test history for the reactivity balance has demonstrated that the core reactivity has been well within the bounds of the acceptance criterion of++/-1%Ak/k of predictions. Based on the testing history, estimated critical condition calculations and physics testing performed as part of the initial reactor startup, relaxation of the STI would not prevent reactor engineering from discovering gross anomalies in core design predictions.

10. PHASED IMPLEMENTATION REQUIREMENTS No phased implementation is recommended due to the relaxation of the STI from 31 EFPD to 92 EFPD.

11. PROPOSED SURROGATE MONITORING RECOMMENDATIONS:

No surrogate monitoring is feasible for this surveillance.

12. PREPARER:

Prepared by: Shane Guess Date 8/20/2007

_____(System or Component Specialist)

C. PRA ANALYSIS Page 5 of 8

1. OVERVIEW OF PRA MODELING OF STI PRA modeling of the surveillance requirement to verify measured core reactivity within +/- 1% Ak/k of predicted values is not practical and is not currently modeled in the DCPP PRA model. Current industry wide practices do not model the surveillance requirement to verify measured core reactivity.

Current PRA Model: N/A.

2.

IMPACTS (CDF Comparison against R.G 1.174 limits)

N/A, see Item 1.

3. FPIE LEVEL 2 PRA MODEL IMPACTS (LERF Comparison against R.G 1.174 limits)

N/A, see Item 1.

4. FIRE RISK IMPACTS (CDF & LERF Comparison against R.G 1.174 limits)

N/A, see Item 1.

5. SEISMIC RISK IMPACTS (CDF & LERF Comparison against R.G 1.174 limits)

N/A, see Item 1.

6. SHUTDOWN RISK IMPACTS (CDF & LERF Comparison against R.G 1.174 limits)

N/A, see Item 1.

7. OTHER PRA ISSUES (ex. Impacts from Other External Events excluding seismic

& Fire Risk Impacts, or changes in test strategy)

N/A, see Item 1.

8. TOTAL EFFECT OF THIS STI EXTENSION ON INTERNAL, EXTERNAL &

SHUTDOWN PRAs (CDF & LERF Comparison against R.G 1.174 limits)

N/A, see Item 1.

Page 6 of 8

9. CUMULATIVE EFFECT OF ALL RI-TS STI ADJUSTMENTS ON INTERNAL, EXTERNAL & SHUTDOWN PRAs. (CDF & LERF Comparison against R.G 1.174 limits)

N/A, see Item 1.

10. IMPACT ON DEFENSE-IN-DEPTH PROTECTION N/A.

11. PRA ANALYSIS - CONCLUSIONS N/A.

12. PREPARER Prepared by: Nathan Barber Date 8/16/2007 (Risk Management [PRA] Engineer)

D. INTEGRATED DECISION-MAKING PANEL REVIEW MEETING

1. Presenter(s): Shane Guess, Amir Afzahi (PRA)

2. Meeting Discussion Summary:

A quorum was verified. The IDP meeting followed the agenda fairly close, with the exception that questions and discussion occurred during the STI evaluation presentations, as well as after each conclusion. Actions and IDP required changes were captured during the process, and are documented verified below.

(Review of Qualitative and Quantitative analyses, and Cumulative Impact)

3. Meeting Results/Recommendations/Bases:

(Consider: phased implementation, additional performance monitoring of failure rates)

(include comment resolution)

4. Approval/Disapproval: Check one of the following:

11 STI Approved X STI Approved with Comments oSTI Disapproved Page 7 of 8

IDP/Expert Panel Members Listing of IDP attendees:

(signatures not required - see IDP meeting minutes)

1. Engineering* Ken Bych 2 Maintenance* Mark Frauenheim

3. Operations* ------

4. Risk Management (PRA)* Amir Afzali 5 Maintenance Rule Coordinator* Don Shelley

6. Surveillance Test Coordinator Chuck Dunlap

7. System manager or Component Shane Guess Engineer I___________________

Also Maintenance Rule Expert Panel Member

5. IDP COMMENT RESOLUTION Prepared by: Shane Guess Date: 10/9/2007 (System Manager or Component Specialist)

Prepared by: Amir Afzali Date: 10/4/2007

_____(Risk Management Engineering)

6. IDP/Expert Panel Coordinator Final Review/Closure:

(All IDP comments resolved): Chuck Dunlap Date: 10/11/07 (IDP Coordinator)

Page 8 of 8

DIABLO CANYON POWER PLANT SURVEILLANCE TEST INTERVAL EVALUATION FORM

__ S_"RV'LLANCE TEST INFORMATION

1. Unit(s): 1& 2

2. Surveillance Test (ST) Number (s) / Revision Number (s)

STP M-13F Revision 34 STP M-13G Revision 30 STP M-13H Revision 27

3. Technical Specification Surveillance Requirement (SR) Number(s):

SR 3.3.5.2

4. Technical Specification SR (Text):

Perform TADOT.

5. Technical Specification SR Bases (and Intent):

SR 3.3.5.2 is the performance of a TADOT. This test is performed every 18 months. The test checks trip devices that provide actuation signals directly, bypassing the analog process control equipment. For these tests, the relay Setpoints are verified and adjusted as necessary. The Frequency is based on the known reliability of the relays and controls and the multichannel redundancy available, and has been shown to be acceptable through operating experience.

6. Recommended ST Frequency Change:

From: 18 months To: 24 months Note: The terms Surveillance Test Interval (STI) and ST Frequency are used interchangeably.

7. Station Benefit:

Changing this STI to a 24 month frequency would provide for greater scheduling flexibility. A 24 month frequency is consistent with other surveillance tests on a once per refueling cycle frequency.

Page I of 10

A._ SYSTEM & MAINTENAkNCE RULE-(MRuI1e) INFO RIATFION

1. SYSTEM NUMBER: 63A

2. SYSTEM DESCRIPTION: 4kV Vital

3. CURRENT MRULE RISK SIGNIFICANCE (R-S) CLASSIFICATION:

Risk Significant

4. CURRENT MRULE R-S BASIS:

Modeled in PRA with a high RAW / RRW value.

5. Current PRA RAW (System): Vital 4kV = 126 (MRule R-S threshold: > 2.0)

6. Current PRA RRW (System): Vital 4 kV = 1.12 (MRule R-S threshold: > 1.005)

Page 2 of 10

7. Current PRA Limiting Sequences: 4713 saved sequences (MRule R-S threshold: top 90%; Trigger value: 363 )

The system in question appears in the top 90% of sequences.

B. QUALITATIVE ANALYSIS:

1. COMMITMENT REVIEW (Is STI credited in any commitments?)

A PCD search was performed for STP M- 13 F, STP M- 13 G, and STP M- 13 H for units 1 and 2. No PCD commitments exist that address the frequency of testing.

2. SURVEILLANCE TEST HISTORY OF THE COMPONENTS AND SYSTEM ASSOCIATED WITH THE STI ADJUSTMENT:

Review of the last 6 years of surveillance test history for these STPs show only one performance case (IR 13) when STP M-1I3G (EDG 12) did not meet its acceptance criteria for transfer of the 4 KV Bus G to EDG 12 on second l'evel undervoltage. The time from opening of the 4 KV Bus G feeder breaker to EDG 12 breaker closing on the Bus was exceeded by 7.1 sec. This failure was due to an Agastat Timing Relay (ETR14D) setpoint drift. The same type of relay caused a CCW Pump failure to start during performance of STP M-1 5. In this case the relay failed to actuateý The Agastat ETR1I4D timers have shown high reliability over the years and the above two incidents may have been the result of work activities in the panels by external contract workers during that outage.

3.1 RELIABILITY REVIEW:

PERFORMANCE (OPERATION & MAINTENANCE) HISTORY OF THE COMPONENTS AND SYSTEM ASSOCIATED WITH THE STI ADJUSTMENT:

Maintenance Rule Train Actual Unreliability: Ul Bus F = 1 MPFF (all other, U1 & 2 Busses = 0)

Maintenance Rule Unreliability Performance Criteria: < 1 MPFF in 2 years Additional component history review: The above review is for the past 2 years for Busses F, G and H for Units 1 and 2.

4. UNAVAILABILITY REVIEW:

Maintenance Rule Train Actual Unavailability: 0 hours0 days 0 hours 0 weeks 0 months Maintenance Rule Unavailability Performance Criteria: 0 hours0 days 0 hours 0 weeks 0 months / year Page 3 of 10

5. PAST INDUSTRY AND PLANT-SPECIFIC EXPERIENCE WITH THE FUNCTIONS AFFECTED BY THE PROPOSED CHANGES Review of industry operating experience on ETR14D Time-Delay Relays found 25 issues in the past 21 years. Most of the issues involved poor solder connections. No, applicability to DCPP experience was found. Most operating experience associated with Agastat Time-Delay Relays has been associated with the electro-pneumatic models. Search of operating experience with these relays at DCPP showed only 2 reliability issues in the past. PG&E does not consider the two failures to be indicative of a larger problem with the Agastat ETR14D Time - Delay Relays based on good performance by the relays in the past and successful testing of the other relays during 1R13, 2R13 and 1R14.

6. VENDOR-SPECIFIED MAINTENANCE FREQUENCY All DCPP 4 KV Bus Autotransfer relays are tested in accordance with vendor recommendations, which are part of DCPP's PM program.

27HFT1, 27HGTl, 27HHTI - Basler BEl-27 Basler relays are static devices which require no preventive maintenance other than a periodic operational check. The operational test procedure of Section 5,'Tests and Adjustments, provides an adequate check to verify proper operation of the relay.

Most components are on conformally-coated PC boards. In-house replacement of individual components may be difficult and should not be attempted unless appropriate equipment and qualified personnel are available.

The relay may be returned to the factory for repair. When returning the relay to the factory, ship the entire relay cradle assembly preferably in its case.

27HFT2, 27HGT2, 27HHT2 - West. SSV-T 27HFB3, 27HGB3, 27HHB3 27HFB4, 27HGB4, 27HHB4 Ref 663332-80-7 Check calibration and clean contacts every year.

27HFB2, 27HGB2, 27HHB2 - Basler BElI-GPS 100 Because the BE1I-GPS 100 has extensive internal test capabilities, periodic testing of the protection system can be greatly reduced. Relay operating characteristics are a function of programming instructions that do not drift over time. Thus, the user may wish to verify items that the relay's self-testing features cannot completely determine.

Periodic testing might consist of the following settings and function checks:

-Verify that the set points that were proven during commissioning have not been changed.

Page 4 of 10

-Verify that the inputs and outputs are interfacing properly with the rest of the protection and control system.

-Verify that the power system analog parameters used by the protection and control functions are being measured accurately.

27XHFB2, 27XGB2, 27XHB2 - West. SG 27YHFB2, 27YHGB2, 27YHHB2 27ZHFB2, 27ZHGB2, 27ZHHB2 No manufacturer maintenance recommendation found. Ref 663102-24 62HF3A, 62HG3A, 62HH3A - Agastat ETR 62HF3B, 62HG3B, 62HH3B Replacement Schedule - Series EGP/EML/ETR The qualified life of these relays is 25,000 electrical operations or 10 years from the date of manufacture, whichever occurs first.

7. TEST INTERVALS SPECIFIED IN APPLICABLE INDUSTRY CODES, AND STANDARDS There are no industry codes or standards that specify test intervals for bus autotransfer.

8. OTHER QUALITATIVE CONSIDERATIONS (include: comparison to Improved TS, alternate ST test list [retained], LCO review [optional],

assumptions in plant licensing basis, degree ST provides conditioning exercise for operability, etc.)

NONE

9. QUALITATIVE ANALYSIS - CONCLUSIONS Based on the review of the above tests performed over the past several years, it is recommended that these Surveillance Tests be extended from an 18-month to 24-month frequency.

10. PHASED IMPLEMENTATION REQUIREMENTS NONE Page 5 of 10

1. PROPOSED SURROGATE MONITORING RECOMMENDATIONS:

Maintenance Rule.

12. PREPARER:

Prepared by: Stefan Bednarz Date 8/29/2007 (System Manager or Component Specialist)

C. PRA ANALYSIS

1. OVERVIEW OF PRA MODELING OF STI The surveillance requirement of interest tests the actuation of undervoltage relays required for a start of the Emergency Diesel Generators (EDG). The DCPP PRA model considers these undervoltage relays in the modeling of EDG components.

The application base model DCO 15B3A was created by incorporating the standby failure rate component for the U/V relays into model DCO1, the current DCPP PRA model of record. Model DC01513A1I was created with the proposed surveillance interval (24 months) incorporated into the standby failure rate calculation for U/V relays.

Current PRA Model: DC015BA

2. FULL POWER INTERNAL EVENTS (FPIE) LEVEL 1 PRA MODEL IMPACTS (CDF Comparison against R.G 1. 174 limits)

FPIE CDF from DC015BA is 1.077E-05/year. EPIE CDF for DC015BAI is 1.0797E-05/year.

ACDF = 2.7E-08/year (< 1E-06/year 1.174 limits)

3. FPIE LEVEL 2 PRA MODEL IMPACTS (LERE Comparison against R.G 1.174 limits)

FPIE LERF from DC015BA is 1.4952E-06/year. FPIE LERF from DC015BAI is 1.4966E-06/year.

ALERF = 1.4E-09/year (< 1E-07/year 1.174 limits)

Page 6 of 10

4. FIRE RISK IMPACTS (CDF & LERF Comparison against R.G 1.174 limits)

Fire CDF from DCOI5153A is 1.7014E-05/year. Fire CDF from DCOI513BAI is 1.7015E-05/year.

ACDF = 1E-09/year (< 1E-06/year 1.174 limits).

Fire LERF is not quantified in the DCPP PRA model.

Assuming a constant ratio for LERF in non SGTR/ISLOCA sequences fire LERF was estimated to be: ALERF =ILERF/ICDF*FCDF=2.55E-02*1EO9/year =2.55E-1 1.

Additionally, the DCPP IPEEB did not identify any significant vulnerabilities to containment isolation as a result of fire.

5. SEISMIC RISK IMPACTS (CDF & LERF Comparison against R.G 1.174 limits)

Seismic CDF fr'om DC015BA is 3.7765-05/year. Fire CDF from DC015BAI is 3.7772-05/year.

ACDF=7E-09/year (< 1E-06/year 1.174 limits)

Seismic LERF from DC015BA is 1.8919E-06/year. Seismic LERF from DC015BAI is 1.8919-06/year. The change in seismic LERF is negligible.

6. SHUTDOWN RISK IMPACTS (CDF & LERF Comparison against R.G 1.174 limits)

DCPP currently uses a simplistic ORAM-Sentinel shutdown PRA model to characterize risk during outages. A typical hot midloop outage "accumulates" around 3E-06 CDF (3E-06/1 8 month refueling cycle = 1.67E-07/year shutdown CDF). When compared to the average CDF for the FPIE model, shutdown risk is nearly 2 orders of magnitude lower. Given the relatively low contribution from shutdown risk to total risk, a conservative assumption of this shutdown risk would be to double the ACDF and ALERF obtained from the FPIE model.

ACDF=5 .4E-08/year ALERF=2. 8E-09/year Additionally, STP M- 13 is currently performed during an outage. With the change to a 24 month period, the exposure time for outage risk will always be less than 18 months since the test will have been performed at some point online prior to the outage (With an 18 month refueling cycle). The ACDF between the current testing practice and the new with a shorter exposure time from online test to outage will be negative.

Page 7 of 10

7. OTHER PRA ISSUES (ex. Impacts from Other External Events excluding seismic

& Fire Risk Impacts, or changes in test strategy)

No other external events have been identified that would impact these results. Internal flooding is considered as part of the full power internal events model.

8. TOTAL EFFECT OF THIS STI EXTENSION ON INTERNAL, EXTERNAL &

SHUTDOWN PRAs (CDF & LERF Comparison against R.G 1.174 limits)

Total ACDF = 8.90E-08 (< 1E-06/year 1.174 limits)

Total ALERF = 4.20E-09 (< 1E-07/year 1.174 limits)

9. CUMULATIVE EFFECT OF ALL RI-TS STI ADJUSTMENTS ON INTERNAL, EXTERNAL & SHUTDOWN PRAs. (CDF & LERF Comparison against R.G 1.174 limits)

"The proposed change documented in this worksheet, is the first example that will have an impact on the PRA figures of merit. Therefore, the cumulative effect is equal to the effect calculated for this example."

10. IMPACT ON DEFENSE-IN-DEPTH PROTECTION DCPP has 3 EDGs and 2 offsite power circuits available to mitigate an accident. The small increase in the standby failure will not significantly degrade the emergency AC power defense in depth.

11. PRA ANALYSIS - CONCLUSIONS The results of the PRA analysis indicate that an extension in the surveillance interval for EDG undervoltage start relays is not risk significant based on a comparison to RG 1.174 limits for CDF and LERF.

12. PREPARER Prepared by: Nathan Barber Date 8/20/2007 (Risk Management [PRA] Engineer)

Page 8 of 10

D. INTEGRATED DECISION-MAKING PANEL REVIEW MEETING DATE:

09/24/2007

1. Presenter(s): Stephan Bednarz, Amir Afzahi (PRA)

2. Meeting Discussion Summary:

A quorum was verified. The IDP meeting followed the agenda fairly close, with the exception that questions and discussion occurred during the STI evaluation presentations, as well as after each conclusion. Actions and IDP required changes were captured during the process, and are documented verified below.

(Review of Qualitative and Quantitative analyses, and Cumulative Impact)

3. Meeting Results/Recommendations/Bases:

(Consider: phased implementation, additional performance monitoring of failure rates)

(include comment resolution)

4. Approval/Disapproval: Check one of the following:

0 STI Approved X STI Approved with Comments 0 STI Disapproved IDP/Expert Panel Members Listing of IDP attendees:

(signatures not required - see IDP meeting minutes)

1. Engineering* Ken Bych 2 Maintenance* Mark Frauenheim

3. Operations*

4. Risk Management (PRA)* Amir Afzali 5 Maintenance Rule Coordinator* Don Shelley

6. Surveillance Test Coordinator Chuck Dunlap

7. System manager or Component Stefan Bednarz Eneineer

Also Maintenance Rule Expert Panel Member Page 9 of 10

IDP COMMENT RESOLUTION Prepared by: Stefan Bednarz Date: 10/4/2 007 (System Manager or Component Specialist)

Prepared by: Amir Afzali Date: 10/4/2 007 (Risk Management Engineering)

6. IDP/Expert Panel Coordinator Final Review/Closure:

(All IDP comments resolved) Chuck Dunlap Date: 10/10/2007 (IDP Coordinator)

Page 10 of 10

DIABLO CANYON POWER PLANT SURVEILLANCE TEST INTERVAL EVALUATION FORM X. SU.RV.EILLANCE TES.T INFORMAT [ON

1. Unit(s): 1&2

2. Surveillance Test (ST) Number (s) /Revision Number (s)

STP V-6A, Unit I / Revision 1 STP V-6A, Unit 2 / Revision 1 STP I- ID, Unit 1 / Revision 76 STP 1-ID3, Unit 2 / Revision 58 STP I- IF, Unit 1 / Revision 5 STP I- IF, Unit 2 / Revision 6

3. Technical Specification Surveillance Requirement (SR) Number(s):

3.6.3.3

4. Technical Specification SR (Text):.

NOTE----------------------

Valves and blind flanges in high radiation areas may be verified by use of administrative controls.

Verify each containment isolation manual valve and blind flange that is located outside containment and not locked, sealed or otherwise secured and required to be closed during accident conditions is closed, except for containment isolation valves that are open under administrative controls.

5. Techni cal Specification SR Bases (and Intent):

This SR requires verification that each containment isolation manual valve and blind flange located outside containment and not locked, sealed, or otherwise secured and required to be closed during accident conditions is closed. The SR helps to ensure that post accident leakage of radioactive fluids or gases outside of the containment boundary is within design limits. This SR does not require any testing or valve manipulation. Rather, it involves verification, through a system walkdown, which may include the use of local or remote indicators, that those containment isolation valves outside containment and capable of being mispositioned are in the correct position. Since verification of valve position for containment isolation valves outside containment is relatively easy, the 31 day Frequency is based on engineering judgmient and was chosen to provide added assurance of the correct positions. The SR specifies that containment isolation valves that are open under administrative controls are not required to meet the SR during the time the valves are open. This SR does not apply to valves that are locked, sealed, or otherwise secured in a closed position since these were verified to be in the correct position upon locking, sealing, or securing.

The Note applies to valves and blind flanges located in high radiation areas and allows these devices to be verified closed by use of administrative means.

Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted during MODES 1, 2, 3 and 4 for ALARA reasons. Therefore, the probability of misaligniment of these containment isolation valves, once they have been verified to be in the proper position, is small.

6. Recommended ST Frequency Change:

From: 31 days To: 92 days Note: The terms Surveillance Test Interval (STI) and ST Frequency are used interchangeably.

7. Station Benefit:

Changing the frequency of SR 3.6.3.3 will reduce the man hours required to perform STP V-6A, STP I- ID, and STP I- IF, and will result in less radiation exposure.

A.. ~S.STEM & MVAJ1NT.ENA1NCE RULE (MRule) INFORNMATION

1. SYSTEM NUMBER: 45B

2. SYSTEM DESCRIPTION: Containment Isolation Vlvs

3. CURRENT MRULE RISK SIGNIFICANCE (R-S) CLASSIFICATION:

Risk Significant

4. CURRENT MRULE R-S BASIS:

Modeled in PRA.

5. Current PRA RAW (System): 1.0 (MRule R-S threshold: > 2.0)

6. Current PRA RRW (System): 1.0 (MRule R-S threshold: > 1.005)

7. Current PRA Limiting Sequence: Not applicable since this system's contribution is only to LERF. LERF sequences were not used for Maintenance Rule RS deternination.

(MRule R-S threshold: top 90%; Trigger value: n/a)

1. COMMITMENT REVIEW (Is STI credited in any commitments?)

A review of the PCD was performed for TS SR 3.3.6.3, and procedures STP V-6A, STP I-ID, and STP I-IF. Commitment T32700 will require an update to reference the revised frequency, or to remove the frequency from the description. T32700 makes reference to the 31 day frequency of SR 3.6.3.3.

2. SURVEILLANCE TEST HISTORY OF THE COMPONENTS AND SYSTEM ASSOCIATED WITH THE STI ADJUSTMENT:

An AR search was performed covering the past 3 years with no' cases of improperly positioned valves associated with SR 3.3.6.3 at power. Event database for the past two years (limit of current database) was reviewed for improperly positioned containment isolation-valves with none noted.

3. RELIABILITY REVIEW:

PERFORMANCE (OPERATION & MAINTENANCE) HISTORY OF THE COMPONENTS AND SYSTEM ASSOCIATED WITH THE STI ADJUSTMENT:

Maintenance Rule Train Actual Unreliability: 5MPFF in 2 years (unit 1), OMPFF in 2 years (unit 2)

Maintenance Rule Unreliability Performance Criteria: < 2MPFF in 2 years Note: None of the MPFF on Unit 1 are related to valve mispositions or missing seals.

4. UNAVAILABILITY REVIEW:

Maintenance Rule Train Actual Unavailability: 0 hrs (unit 1 and 2)

Maintenance Rule Unavailability Performance Criteria: 0 hrs / year

5. PAST INDUSTRY AND PLANT-SPECIFIC EXPERIENCE WITH THE FUNCTIONS AFFECTED BY THE PROPOSED CHANGES Review of INPO OE database revealed several instances, mostly in reports to the NRC, where small containment isolation valves were found open, rather than in their required position. In general, most of the instances reviewed were cases where the valves were not properly included in the required surveillance. These instances would not be applicable to extension of the surveillance interval as the equipment is already included in the required tests. In two instances found, valves were opened after the surveillance was performned and left open. In both cases, the implementing procedures did not have adequate administrative controls. In both cases, the problem was not found during reperformance of the surveillance, but rather due to review of either other completed tests or by personnel during routine activities. Increasing the frequency of the surveillance would prevent finding errors of this nature as soon as the current surveillance interval would require. None of the reports reviewed had any significant safety significance.

6. VENDOR-SPECIFIED MAINTENANCE FREQUENCY N/A. Verifying the containment isolation sealed valve checklist is not a maintenance activity. No frequency is specified by valve manufacturers for verifying sealed valve positions.

7. TEST INTERVALS SPECIFIED IN APPLICABLE INDUSTRY CODES AND STANDARDS None.

8. OTHER QUALITATIVE CONSIDERATIONS None.

9. QUALITATIVE ANALYSIS - CONCLUSIONS This change will not affect the condition of the valves, and will have no affect on their performance. No instances of mispositioned sealed valves or missing seals could be found reviewing ARs over the past three years. Based on review of the above information, it is recommended that the STI be revised from 31 days (monthly) to 92 days (quarterly).

10. PHASED IMPLEMENTATION REQUIREMENTS None recommended.

11. PROPOSED SURROGATE MONITORING RECOMMENDATIONS:

Use existing Maintenance Rule monitoring.

12. PREPARER:

Prepared by: Tony Chitwood Date 9/26/2007 (System Manager or Component Specialist)

C. PAANALYSIS

1. OVERVIEW OF PRA MODELING OF STI OVERVIEW OF PRA MODELING OF STI This Surveillance Requirement involves a monthly walkdown of unsealed, manual containment isolation valves to verify these valves are closed and capped.

The modeling approach used was to determine the likelihood that one of the subject valves was left open and contributes to Large Early Release Frequency (LERF). None of the valves reviewed are larger than 2" valves and would not, by themselves, contribute to LERF. Only cases in which two valves were open simultaneously were considered. A common cause coupling factor was used.

A review of plant history (AR search) for the valves of concern was performed to determine if any of these valves had been found open at power. Only one valve was found open and its condition was corrected prior to the end of a refueling outage. This single incidence was used to calculate the likelihood of a valve remaining open past the end of a refueling outage for the plant's 42 reactor-year operating history.

The periodic walkdown is assumed to discover the open valve either 1 month or 3 months after a refueling outage for the current and future STIs respectively. DC0 15BB calculates LERF for the 1 month surveillance interval and DCO 15BB 1 calculates LERF for the proposed 3 month surveillance interval.

Current PRA Model: DCO15BB/DCO15BBI

2. FULL POWER INTERNAL EVENTS (FPIE) LEVEL 1 PRA MODEL IMPACTS 0 - CIVs do not impact CDF.

___(CDF Comparison against R.G 1.174 limits)

3. FPIE LEVEL 2 PRA MODEL IMPACTS (LERF Comparison against R.G 1.174 limits)

ALERF = 2. 1E-09/year (<1I E-07/year 1.174 limits)

4. FIRE RISK IMPACTS (CDF & LERF Comparison against R.G 1.174 limits)

ACDF = 0 - CIVs do not impact CDF. (< 1E-06/year 1.174 limits)

ALERE = 3.79E-09/year - Calculated by assuming constant LERF/CDF ratio. FLERF is not calculated directly by model. (< 1E-07/year 1.174 limits)

5. SEISMIC RISK IMPACTS (CDF & LERF Comparison against R.G 1.174 limits)

ACDF = 0 - CIVs do not impact CDF. (< IE-06/year 1.174 limits)

ALERE = 8.5E-09/year (< 1E-07/year 1.174 limits)

6. SHUTDOWN RISK IMPACTS (CDF & LERF Comparison against R.G 1.174 limits)

DCPP currently uses a simplistic ORAM-Sentinel shutdown PRA model to characterize risk during outages. A typical hot midloop outage "accumulates" around 3E-06 CDF (3E-06/1 8 month refueling cycle = 1.67E-07/year shutdown CDF). When compared to the average CDF for the FPIE model, shutdown risk is nearly 2 orders of magnitude lower. Given the relatively low contribution from shutdown risk to total risk, a conservative assumption of this shutdown risk would be to double the ACDF and ALERF obtained from the FPIE model.

ACDF=O ALERF=4.2E-09/year

7. OTHER PRA ISSUES (ex. Impacts from Other External Events excluding seismic

& Fire Risk Impacts, or changes in test strategy)

No other external events have been identified that would impact these results. Internal flooding is considered as part of the full power internal events model.

8. TOTAL EFFECT OF THIS STI EXTENSION ON INTERNAL, EXTERNAL &

SHUTDOWN PRAs (CDF & LERF Comparison against R.G 1.174 limits)

Total ACDFO0.0 Total ALERF = 2.04E-08/year

9. CUMULATIVE EFFECT OF ALL RI-TS STI ADJUSTMENTS ON INTERNAL, EXTERNAL & SHUTDOWN PRAs. (CDF & LERF Comparison against R.G 1.174 limits)

Cumulative LERF ______

AILER ASLER AFLER AShutdown Total F F F LERF ALERF 3.50E- 1.03E-09 08 3.82E-09 7.OOE-09 2.46E-08 Cumulative CDF______

AICDF ASCDF AFCDF AShutdown Total ACDF CDF ____

2.70E-08 7.OE-09 1.OOE-09 5.40E-08 8.90E-08 All values are less than RG 1.174 limits.

10. IMPACT ON DEFENSE-IN-DEPTH PROTECTION Given the low likelihood that more than one of the subject valves is left open inadvertently (Only one of these valves was discovered open over DCPP's operating history), the impact on containment isolation defense-in-depth is minimal.

11. PRA ANALYSIS - CONCLUSIONS The results of the PRA analysis indicate that an extension in the surveillance interval for CIV verification is not risk significant based on a comparison to RG 1.174 limits for CDF and LERF.

12. PREPARER Prepared by: Nathan Barber Date 8/20/2007 (Risk Management [PRA] Engineer)

D. INTEGRATED DECISION-MAKING PANEL REVIEW MEETIN4G DATE:

1. Presenter(s):

2. Meeting Discussion Summary:

(Review of Qualitative and Quantitative analyses, and Cumulative Impact)

3. Meeting Results/Recommendations/Bases:

(Consider: phased implementation, additional performance monitoring of failure rates)

(include comment resolution)

4. Approval/Disapproval: Check one of the following:

o1 STI Approved ol STI Approved with Comments 0STI Disapproved IDP/Expert Panel Members Listing of IDP attendees:

(signatures not required - see IDP meeting minutes)

1. Engineering*___________________

2 Maintenance*

3. Operations*~_____________________

4. Risk Management (PRA)* ___________________

5 Maintenance Rule Coordinator*

6. Surveillance Test Coordinator

7. System or Component Engineer ___________________

Also Maintenance Rule Expert Panel Member

5. IDP COMMENT RESOLUTION Prepared by: Date:

(System or Component Specialist)

Prepared by: Date:

(Risk Management Engineering)

2. Meeting Discussion Summary:

(Review of Qualitative and Quantitative analyses, and Cumulative Impact)

3. Meeting Results/Recommendations/Bases:

(Consider: phased implementation, additional performance monitoring of failure rates)

(include comment resolution)

.4. Approval/Disapproval: Check one of the following:

0 STI Approved E3 STI Approved with Comments 13STI Disapproved IDP/Expert Panel Members Listing of IDP attendees:

(signatures not required - see 1IDP meeting minutes)

1. Engineering*___________________

2 Maintenance*

3. Operations* ____________________

4. Risk Management (PRA)* ______ ____________

5 Maintenance Rule Coordinator*

6. Surveillance Test Coordinator

7. System or Component Engineer ___________________

Also Maintenance Rule Expert Panel Member

5. IDP COMMENT RESOLUTION Prepared by: Date:

(System or Component Specialist)

Prepared by:- Date:

(Risk Management Engineering)

6. IDP/Expert Panel Coordinator Final Review/Closure:

(All IDP commnents resolved) ______________Date:

(IDP Coordinator)

Enclosure 5 PG&E Letter DCL-07-097 QUALITY AND SCOPE OF THE PRA MODEL 1

Enclosure 5 PG&E Letter DCL-07-097 Technical Adequacy and Scope of the PRA Model The technical adequacy of the probabilistic risk assessment (PRA) must be compatible with the safety implications of the proposed Technical Specification (TS) changes and the role the PRA plays in justifying the changes. The Nuclear Regulatory Commission (NRC) has developed regulatory guidance to address PRA technical adequacy, Regulatory Guide (RG) 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities" [Reference a], which addresses the use of the American Society of Mechanical Engineers (ASME) RA-Sa-2003, Addenda to ASME RA-S-2002, "Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications" [Reference b], and the NEI peer review process NEI 00-02, "PRA Peer Review Process Guidance" [Reference c].

NEI 04-10 [Reference d] requires an assessment of the PRA models used to support the Surveillance Frequency Control Program (SFCP) against the criteria of RG 1.200 to ensure that the PRA models are capable of determining the change in risk due to changes to surveillance frequencies of systems, structures, and components (SSCs),

using plant-specific data and models. Capability Category 11of ASME RA-Sb-2003 is applied as the standard, and any identified deficiencies relative to those requirements are assessed further in sensitivity studies to determine any impacts on proposed changes to surveillance frequencies. This level of PRA technical adequacy, combined with the proposed sensitivity studies, is sufficient to support the evaluation of changes to surveillance frequencies within the SFCP, and is consistent with Regulatory Position 2.3.1 of RG 1.177 [Reference e].

The Diablo Canyon Power Plant (DCPP) PRA model used in this evaluation is a full-scope Level 1 and Level 2 PRA model that addresses internal, seismic, and fire events at full power. The large early release frequency (LERF) figure of merit is calculated using the full Level 2 PRA model. The PRA model was developed for Unit 1, but it is equally applicable to Unit 2 since the two units are essentially identical.

The PRA model is based on the original 1988 Diablo Canyon PRA (DCPRA-1 988) model [Reference f] that was developed as part of the Long-Term Seismic Program (LTSP) [Reference g]. The DCPRA-1 988 was a full-scope Level 1 PRA that evaluated internal and external events. The NRC reviewed the LTSP and issued Supplement No.

34 to NUREG-0675 [Reference h] in June 1991, accepting the DCPRA-1 988.

Brookhaven National Laboratory (BNL) performed the primary review of the DOPRA-1988 for the NRC; their review is documented in NUREG/CR-5726 [Reference i].

The DCPRA-1 988 was subsequently updated to support the Individual Plant Examination (IPE) in 1991 and the Individual Plant Examination for External Events (IPEEE) in 1993. Since 1993, several other updates have been made to incorporate plant and procedure changes, update plant-specific reliability and unavailability data, improve the fidelity of the model, incorporate Westinghouse Owners Group (WOG) Peer 2

Enclosure 5 PG&E Letter DCL-07-097 Review comments [Reference j], and support other applications, such as On-line Maintenance, Risk-Informed In-Service Inspection (RI-ISI), Emergency Diesel Generator Completion Time Extension (EDG GTE), and Mitigating System Performance Index (MS P1).I The enhancements to the DCPRA-1 988 PRA model include:

" Modeling the probability of a loss-of-offsite power (LOSP) subsequent to non-LOSP initiating events

" Incorporating the sixth emergency diesel generator

" Upgrading the auxiliary saltwater system model to be more consistent with the Station Blackout submittal

" Allowing credit for cross-tie of the vital 4kV buses (i.e., one diesel generator (DG) feeds loads on two vital buses)

" Adding a 500kV switchyard model, to supplement the 230kV switchyard model

" Updating initiating event frequencies to reflect data from NUREG-5750

[Reference k]

" Using the Rhodes Model to characterize the reactor coolant pump (RCP) seal performance on loss of cooling and seal injection

" Modeling of the fire water storage tank (FWST) as a supplemental water source to the condensate storage tank (CST) as required

" Modification of the loss of offsite power initiating event. The modification was to breakdown the total loss of offsite power frequency into 4 different types of causes and to assign a separate initiating event frequency/offsite power recovery probability for each cause.

Updating the Level 2 PRA model to allow a more realistic assessment of the Large Early Release Frequency figure or merit.

The DCPRA is a living PRA, which is maintained through a periodic review and update process.

Peer Review (Certification) of the DCPP PRA model, using the WOG Peer Review Certification Guidelines, was performed in May 2000 [Reference j]. On the basis of its evaluation, the Certification Team determined that, with certain facts and observations (F&Os) addressed, the technical adequacy of all elements of the PRA would be sufficient to support risk significance evaluations with defense-in-depth input, for RI applications. The two "A"F&Os, related to the human reliability analysis (HRA) were addressed by upgrading the methodology used for the evaluation. The upgraded HRA was recently subjected to a focused peer review. All the findings of this focused review will be addressed prior to implementation of the proposed TS changes either by modifying the model or treatment of the issue via a sensitivity study.

3

Enclosure 5 PG&E Letter DCL-07-097 The "B" F&Os from the WOG Peer Review have also been addressed during model updates in support of the EDG GTE license amendment request (LAR), the LAR effort to extend the OTs for several emergency core cooling system (EGOS) components, and MVSPI calculations.

In addition to the Peer Review, three recent limited scope and independent assessments of the DGPP PRA Level 1 and Level 2 PRA models have been performed by leading industry PRA experts (i.e., Gap Analyses) to support several risk-informed applications, including the MVSPI calculations and DCPP's transition to the National Fire Protection Association (NFPA) 805 Standard. All the findings of these assessments will be addressed prior to implementation of the proposed TS changes and based on the system being subject to change either by modifying the model or treatment of the issue via a sensitivity study. This approach is consistent with the requirements of NEI 04-10

[Reference d]. Since Initiative 5b could be applied to many different systems, DGPP intends to disposition each open finding as it applies to the system for which its Surveillance Frequency is to be modified using the risk-informed application (i.e., at the sub-application level) consistent with the approach that will be proposed by the PWR Owners Group (PWROG).

Additionally, it should be noted that during the MVSPI industry cross-comparison review, the DGPP PRA model was not identified as an outlier.

External Events As stated above, the DCPP PRA model used in this evaluation is a full scope Level 1 and Level 2 PRA model that addresses internal, seismic, and fire events at full power.

Seismic PRA Model Again, as stated above, the original DCPP PRA model, including the seismic model, was reviewed by the NRC and found to have sufficient technical adequacy. Although DCP P's equipment fragilities and seismic hazard curves have not been updated, the plant response model, which is based on the internal events model, has been updated several times.

Fire PRA Model The original Fire PRA model was updated to support the 1993 IPEEE. Other than Control Room (CR) and Cable Spreading Room (CSR) fire scenarios, the Fire PRA quantifies the core damage frequency (CDF) associated with most internal fire initiating events using the same linked event tree models as the internal and seismic events analyses. Separate event trees using conservative assumptions were developed for evaluating CR and CSR fire scenarios. Currently, the Fire PRA model is being 4

Enclosure 5 PG&E Letter DCL-07-097 upgraded to a state-of-the-art model to support transitioning the fire protection program to the NFPA-805 Standard.

Other External Events The evaluation of high winds, external floods, and other external events, which was done as part of the IPEEE, revealed no potential vulnerabilities.

Conclusions on PRA Technical Adequacy and Scope Due to the sound basis of the original model as documented in NUREG-0675 Supplement No. 34 [Reference g] and NUREG/CR-5726 [Reference h] and the considerable effort to incorporate the latest industry insights into the PRA using self-assessments and peer reviews, PG&E is confident that the application of the DCPP PRA model will meet the expectations for PRA technical adequacy. PG&E will either close all identified RG 1.200 gaps, or will address the gaps through sensitivity studies for the surveillance test interval being evaluated using the NEI 04-10 [Reference d]

process and methodology. PG&E commits to evaluate the impact of future model updates (internal model or external model) on the conclusions of the assessments that are performed in support of this application. This approach is judged to be the most effective approach in assuring the appropriateness of the PRA technical adequacy and scope.

References a) RG 1.200, Revision 1, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk- Informed Activities," dated January 2007.

b) American Society of Mechanical Engineers (ASME) RA-Sa-2003, Addenda to ASME RA-S-2002, "Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications," dated December 5, 2003.

c) NEI 00-02, Revision A3, "PRA Peer Review Process Guidance," dated March 20, 2000.

d) NEI 04-10, Revision 1, "Risk-informed Technical Specifications Initiative 5b, Risk-Informed Method for Control of Surveillance Frequencies," dated April 15, 2007.

e) Regulatory Guide 1.177, "An Approach for Plant-Specific Risk-Informed Decisionmaking: Technical Specification," US Nuclear Regulatory Commission, dated August 1998.

5

Enclosure 5 PG&E Letter DCL-07-097 f) PLG, Inc., "Diablo Canyon Probablilistic Risk Assessment," prepared for Pacific Gas and Electric Company, PLG-0637, July 1998.

g) Pacific Gas and Electric Company, "Long Term Seismic Program Final Report," PG&E Letter No. DCL-88-1 92, July 31, 1988.

h) U.S. Nuclear Regulatory Commission, "Safety Evaluation Report,"

Supplement No. 34 to NUREG-0675, dated June 1991.

i) Bozoki, G., et al., "Review of the Diablo Canyon Probabilistic Risk Assessment," NUREGICR-5726 (DRAFT), June 1991.

j) "Diablo Canyon Power Plant Probabilistic Risk Assessment Peer Review Report - Final Report," dated August 2000.

k) NUREG/CR-5750, INEEL/EXT-98-OO4O1, "Rates of Initiating Events at U.S.

Nuclear Power Plants: 1987-1 995," February 1999.

1) Peer Review (Certification) of the DCPP PRA model, using the WOG Peer Review Certification Guidelines, was performed in May 2000.

6

Enclosure 6 PG&E Letter DCL-07-097 COMMITMENTS

1) PG&E will provide fresh marked up and retyped Technical Specification (TS) pages prior to approval of this LAR.

2) PG&E will conduct another Integrated Decisionmaking Panel (IDP) panel for the NRC to observe, that will review TS SIR 3.6.3.3 and discuss the two previously reviewed surveillance test intervals (STIs).

3) PG&E commits to evaluate the impact of future model updates (internal model or external model) on the conclusions of the assessments that are performed in support of this application.

4) The upgraded human reliability analysis was recently subjected to a focused peer review. All the findings of this focused review will be addressed prior to implementation of the proposed TS changes either by modifying the model or treatment of the issue via a sensitivity study.

5) Three recent limited scope and independent assessments of the DCPP PRA Level 1 and Level 2 PRA models have been performed by leading industry PRA experts (i.e., Gap Analyses) to support several risk-informed applications, including the MVSPI calculations and DCPP's transition to the National Fire Protection Association (NFPA) 805 Standard. All the findings of these assessments will be addressed prior to implementation of the proposed TS changes and based on the system being subject to change either by modifying the model or treatment of the issue via a sensitivity study.

6) PG&E will either close all identified RG 1.200 gaps, or will address the gaps through sensitivity studies for the surveillance test interval being evaluated using the NEI 04-10 process and methodology.

1

DCL-07-097, License Amendment Request 07-04, Proposed Technical Specifications Change to Relocate Surveillance Test Intervals to a Licensee-Controlled Program (Risk Informed Technical Specifications Initiative 5b)

Contents

Text

References:

Dear Commissioners and Staff:

SUMMARY

3.0 TECHNICAL EVALUATION

4.0 REGULATORY ANALYSIS

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

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