DCL-06-143, License Amendment Request 06-10 Revision to Technical Specification 5.5.8, Inservice Testing Program

License Amendment Request 06-10 Revision to Technical Specification 5.5.8, Inservice Testing Program

ML070160265
Person / Time
Site:	Diablo Canyon
Issue date:	12/29/2006
From:	Becker J Pacific Gas & Electric Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
DCL-06-143, LAR 06-010, OL-DPR-80, OL-DPR-82
Download: ML070160265 (25)

Text

PacificGas and ElectricCompany' James R. Becker Diablo Canyon Power Plant Vice President P. 0. Box 56 Diablo Canyon Operations and Avila Beach, CA 93424 Station Director 805.545.3462 Fax: 805.545.4234 December 29, 2006 PG&E Letter DCL-06-143 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 06-10 Revision to Technical Specification 5.5.8, "Inservice Testing Program" Pursuant to 10 CFR 50.90, Pacific Gas and Electric Company (PG&E) hereby requests an amendment to Facility Operating License Nos. DPR-80 and DPR-82 for Units 1 and 2 of the Diablo Canyon Power Plant (DCPP), respectively. The enclosed License Amendment Request (LAR) requests revision to Technical Specification (TS) 5.5.8, "Inservice Testing Program."

The proposed changes revise TS 5.5.8 to indicate that the Inservice Testing Program shall include testing frequencies applicable to the ASME Code for Operations and Maintenance (ASME OM Code), and to indicate that there may be some non-standard frequencies specified as 2 years or less in the Inservice Testing Program to which the provisions of SR 3.0.2 are applicable. The proposed changes are consistent with NRC-approved Technical Specification Task Force (TSTF) Travelers TSTF-479, Revision 0, "Changes to Reflect Revision of 10 CFR 50.55a," and TSTF-497, Revision 0, "Limit Inservice Testing Program SR 3.0.2 Application to Frequencies of 2 Years or Less." PG&E has determined that this LAR does not involve a significant hazard consideration as determined per 10 CFR 50.92.

PG&E's technical and regulatory evaluation of this LAR, the TS changes, the TS Bases changes (for information only), and a list of regulatory commitments are enclosed.

The changes in this LAR are not required to address an immediate safety concern. PG&E requests approval of this LAR no later than December 31, 2007.

Once approved, the amendment will be implemented within 90 days from the A member of the STARS (Strategic Teaming and Resource Sharing) Alliance CaLtaway

• Comanche Peak

• Diabto Canyon e Palo Verde

• South Texas Project e Wolf Creek 1--

Document Control Desk PG&E Letter DCL-06-143 December 29, 2006 Page 2 date of issuance. If you have any questions or require additional information, please contact Stan Ketelsen at 805-545-4720.

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

on December 29, 2006.

MJR/4557

Enclosures:

1. Licensee Evaluation

2. TS Page Markups

3. Changes to TS Bases (for information only)

4. Retyped TS Pages cc: Edgar Bailey, DHS Bruce S. Mallett Terry W. Jackson Diablo Distribution cc/enc: Alan B. Wang A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway

• Comanche Peak 9 Diablo Canyon

• Palo Verde

• South Texas Project

• Wolf Creek

Enclosure 1 PG&E Letter DCL-06-143 1.0

SUMMARY

DESCRIPTION This evaluation supports a request to amend Operating License Facility Operating License Nos. DPR-80 and DPR-82 for Units 1 and 2 of the Diablo Canyon Power Plant (DCPP), respectively, to update references to the source of requirements for inservice testing of ASME Code Class 1, 2, and 3 pumps and valves and to address the applicability of Surveillance Requirement (SR) 3.0.2 to some non-standard pump and valve testing frequencies.

The proposed changes are consistent with NRC-approved Technical Specification Task Force (TSTF) Travelers TSTF-479, Revision 0, "Changes to Reflect Revision of 10 CFR 50.55a," (Reference 1) and TSTF 497, Revision 0, "Limit Inservice Testing Program Application to Frequencies of 2 Years or Less," (Reference 2).

2.0 DETAILED DESCRIPTION The proposed changes would revise the requirements in Technical Specification (TS) 5.5.8, "Inservice Testing Program," to update references to the American Society of Mechanical Engineers (ASME)

Boiler and Pressure Vessel Code,Section XI as the source of requirements for the inservice testing of ASME Code Class 1, 2, and 3 pumps and valves. The proposed changes delete reference to Section XI of the Code and incorporate reference to the ASME Code for Operation and Maintenance of Nuclear Power Plants (ASME OM Code)

(Reference 3) and address the applicability of Surveillance Requirement (SR) 3.0.2 to some non-standard frequencies specified as 2 years or less in the Inservice Testing (IST) Program.

2.1 Proposed Changes TS 5.5.8, "Inservice Testing Program," is revised to indicate that the IST shall have testing Frequencies applicable to the ASME OM Code.

TS 5.5.8.b. is revised to indicate that there may be some nonstandard frequencies utilized in the IST Program to which the provisions of SR 3.0.2 are applicable. Specifically, TS 5.5.8.b. is revised to state:

"The provisions of SR 3.0.2 are applicable to the above required Frequencies and to other normal and accelerated Frequencies specified as 2 years or less in the Inservice Testing Program."

1

Enclosure 1 PG&E Letter DCL-06-143 Various sections of the TS Bases are also revised for consistency with the requirements of 10 CFR 50.55a(f)(4). The changes to the affected TS Bases pages will be incorporated in accordance with TS 5.5.14, "Technical Specifications (TS) Bases Control Program."

2.2 Background In 1990, the ASME published the initial edition of the ASME OM Code that provides rules for inservice testing of pumps and valves.

The ASME OM Code replaced Section XI of the Boiler and Pressure Vessel Code for inservice testing of pumps and valves.

The 1995 edition of the ASME OM Code was incorporated by reference into 10 CFR 50.55a(b). Since 10 CFR 50.55a(f)(4)(ii) requires that inservice testing during successive 10-year intervals comply with the requirements of the latest edition and addenda of the Code incorporated into 10 CFR 50.55a(b), TS 5.5.8 must be revised to reference the ASME OM Code.

3.0 TECHNICAL EVALUATION

The purpose of the IST Program is to assess the operational readiness of pumps and valves, to detect degradation that might affect component OPERABILITY, and to maintain safety margins with provisions for increased surveillance and corrective action. 10 CFR 50.55a defines the requirements for applying industry codes to each licensed nuclear powered facility.Section XI of the ASME Code has been revised on a continuing basis over the years to provide updated requirements for the inservice inspection and inservice testing of components. Until 1990, the ASME Code requirements addressing the inservice testing of pumps and valves were contained in Section XI, Subsections IWP (pumps) and IWV (valves). In 1990, the ASME published the initial edition of the OM Code that provides the rules for inservice testing of pumps and valves. Since the establishment of the 1990 Edition of the OM Code, the rules for inservice testing are no longer being updated in Section XI. As identified in NRC SECY-99-017 (Reference 4), the NRC has generally considered evolution of the ASME Code to result in a net improvement in the measures for inspecting piping and components and testing pumps and valves.

By final rule issued on September 22, 1999 (Reference 5), the NRC amended 10 CFR 50.55a(f)(4)(ii) to require licensees to update their IST Program to the latest approved edition of the ASME OM Code incorporated by reference 5 into 10 CFR 50.55a(b). TS 5.5.8 currently references the ASME Boiler and Pressure Vessel Code, Section Xl, as the 2

Enclosure 1 PG&E Letter DCL-06-143 source of the IST Program requirements for ASME Code 1, 2, and 3 components. The Code of record for the ongoing Third 10-Year IST Program interval is the 2001 Edition including the OMa-2002 and OMb-2003 Addenda of the ASME OM Code. The proposed changes to TS 5.5.8 are necessary for consistency with the IST requirements of 10 CFR 50.55a.

Additionally, TS 5.5.8 is revised to indicate that the provisions of SR 3.0.2 are applicable to other IST frequencies that are not specifically listed in the testing frequencies identified in TS 5.5.8. The IST Program may have frequencies for testing that are based on risk or other factors and do not conform to the standard testing Frequencies specified in TS 5.5.8. The Frequency of the Surveillance may be determined through a mix of risk informed and performance based means in accordance with the IST Program. Application of SR 3.0.2 to other IST Frequencies specified as 2 years or less is consistent with the guidance in NUREG-1482, paragraph 3.1.3 (Reference 6). This response would indicate that the 25 percent tolerance specified in SR 3.0.2 is applicable to any IST Frequency specified as 2 years or less.

4.0 REGULATORY EVALUATION

The proposed changes revise the requirements in TS 5.5.8, "Inservice Testing Program," to update references to the ASME Boiler and Pressure Vessel Code,Section XI as the source of requirements for the inservice testing of ASME Code Class 1, 2, and 3 pumps and valves. The proposed changes delete reference to Section XI of the Code, incorporate reference to the ASME OM Code, and address the applicability of Surveillance Requirement (SR) 3.0.2 to other normal and accelerated frequencies specified as 2 years or less in the IST Program.

4.1 Sigqnificant Hazards Consideration PG&E has evaluated whether or not a significant hazards consideration is involved with the proposed amendment(s) by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

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

Response: No.

3

Enclosure 1 PG&E Letter DCL-06-143 The proposed changes revise TS 5.5.8, "Inservice Testing Program," for consistency with of 10 CFR 50.55a(f)(4) requirements regarding inservice testing of pumps and valves. The proposed change incorporates revisions to the ASME Code that result in a net improvement in the measures for testing pumps and valves.

The proposed changes do not impact any accident initiators or analyzed events or assumed mitigation of accident or transient events. They do not involve the addition or removal of any equipment, or any design changes to the facility. Therefore, the proposed changes do not represent 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 changes do not involve a modification to the physical configuration of the plant (i.e., no new equipment will be installed) or change in the methods governing normal plant operation. The proposed change will not impose any new or different requirements or introduce a new accident initiator, accident precursor, or malfunction mechanism.

Additionally, there is no change in the types or increases in the amounts of any effluent that may be released off-site and there is no increase in individual or cumulative occupational exposure. Therefore, this proposed change does not create the possibility of an accident of a different kind than previously evaluated.

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

Response: No The proposed changes revise TS 5.5.8, "Inservice Testing Program," for consistency with the requirements of 10 CFR 50.55a(f)(4) regarding the inservice testing of pumps and valves. The proposed change incorporates revisions to the ASME Code that result in a net improvement in the 4

Enclosure 1 PG&E Letter DCL-06-143 measures for testing pumps and valves. The safety function of the affected pumps and valves will be maintained.

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

Based on the above, PG&E concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

4.2 Applicable Regulatory Requirements/Criteria NRC regulation, 10 CFR 50.55a, defines the requirements for applying industry codes to each licensed nuclear powered facility.

The regulations require that during successive 120-month intervals, programs be developed utilizing the latest edition and addenda incorporated into paragraph (b) of 10 CFR 50.55a 12 months before the start of the 120-month interval of the operating license subject to the limitations and modifications identified in paragraph (b).

There are no changes being proposed such that compliance with any of the regulatory requirements above would come into question. The evaluations documented above confirm that PG&E will continue to comply with all applicable regulatory requirements.

4.3 Precedent The NRC accepted TSTF-479 in December 2005 (Reference 7) and TSTF-497 in October 2006 (Reference 8).

A similar change was approved for the Cooper Nuclear Station in Amendment No. 223 on September 6, 2006. However, even though TSTF-497 was not approved at this time, this amendment addressed the application of SR 3.0.2 to other IST Frequencies with the same wording that is now included in the approved TSTF-497.

4.4 Conclusions In conclusion, 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 5

Enclosure 1 PG&E Letter DCL-06-143 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 for environmental considerations. The review has determined that the proposed amendment would change requirements with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, and would change an inspection or surveillance requirement. However, 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 effluent 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.

6.0 REFERENCES

1. Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-479, Revision 0, "Changes to Reflect Revision of 10 CFR 50.55a"

2. Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-497, Revision 0, "Limit Inservice Testing Program SR 3.0.2 Application to Frequencies of 2 Years or Less"

3. ASME Operation and Maintenance Code for Operation and Maintenance of Nuclear Power Plants, 2001 Edition including the OMa-2002 and OMb-2003 Addenda

4. SECY-99-017, "Proposed Amendment to 10 CFR 50.55a,"

January 13, 1999

5. Federal Register Notice: Industry Codes and Standards; Amended Requirements, published September 22, 1999 (64 FR 51370)

6. NUREG-1482, Revision 1, "Guidelines for Inservice Testing at Nuclear Power Plants," January 2005 6

Enclosure 1 PG&E Letter DCL-06-143

7. Letter dated December 6, 2005, from USNRC to Technical Specifications Task Force

8. Letter dated October 4, 2006, from USNRC to Technical Specifications Task Force 7

Enclosure 2 PG&E Letter DCL-06-143 TS PAGE MARKUPS 1

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.7 Reactor Coolant Pump Flywheel Inspection Program This program shall provide for the inspection of each reactor coolant pump flywheel per the recommendations of Regulatory Position c.4.b of Regulatory Guide 1.14, Revision 1, August 1975.

In lieu of Position c.4.b(1) and c.4.b(2), a qualified in-place UT examination over the volume from the inner-bore of the flywheel to the circle one-half of the outer radius or a surface examination (MT and/or PT) of exposed surfaces of the removed flywheels may be conducted at approximately ten year intervals coinciding with the Inservice Inspection schedule as required by ASME Section XI.

5.5.8 Inservice Testing Progqram This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components. The program shall include the following:

a. Testin frequencies sp-ecified in S, ion Xl o-e ASME.,5ler-a(d Pressur (Vye~s0i;,1Cj~gtand applicable Addenda as follows:

f* -- .

• ASME Boi~er ancrbesur e oIVg C.,* e. Adden*da-nd applicable+ -- Required Frequencies for

-K%. Av CO&-t Addenda terminology for performing inservice testing

, inservice testing activities activities

. ,L.. Weekly At least once per 7 days 0tkl Monthly At least once per 31 days

'WA *Quarterly or every 3 months At least once per 92 days AEC Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days

b. The provisions of SR 3.0.2 are applicable to the above required Frequencies for performing inservice testing activities;

c. The provisions of SR 3.0.3 are applicable to inservice testing activities; and

d. Nothing in the ASM r pre'ure 7etss ode shall be construed to supersede the requirements of any TS.

DIABLO CANYON - UNITS 1 & 2 5.0-9 Unit 1 - Amendment No.-3"5-rad9ODE4.doc - R1 7 9 Unit 2 - Amendment No. 1-5"

Enclosure 3 PG&E Letter DCL-06-143 CHANGES TO TS BASES Changes to Technical Specification Bases Pages (For information only) 1

Enclosure 3 PG&E Letter DCL-06-143 Bases Insert 1 ASME Code for Operation and Maintenance of Nuclear Power Plants.

2

Pressurizer Safety Valves B 3.4.10 BASES ACTIONS B.1 and B.2 (continued) below LTOP arming temperature specified in the PTLR, overpressure protection is provided by the LTOP System. The change from MODE 1, 2, or 3 to MODE 4 reduces the RCS energy (core power and pressure), lowers the potential for large pressurizer insurges, and thereby removes the need for overpressure protection by three pressurizer safety valves.

CH DX/C111 I AKWC': CD 'ý A It) I REQUIREMENTS SRs-are s decifie in he Inservice Testing Program. The AS Code Sfection XI (~f* requir ae4ft safety and re ie* ts be lerforme7in ,

accordancwith ANSI SME OM-a-1988 (R .i5) 'No additional requirements are specified. The pressurizer safety valve setpoint is

+2.3%/-3% for OPERABILITY; however the valves are reset to +/- 1% of nominal pressure of 2485 psig during the Surveillance to allow for drift.

REFERENCES 1. ASME, Boiler and Pressure Vessel Code, Section II1.

2. FSAR, Chapter 15.

3. WCAP-7769, Rev. 1, June 1972.

4. SME, 'er and Pressure Vessel Code ction L

• -' . " L b.-t 5. Option and M enance Code, 87 wib- -a-1988 A denda.

'4J DIABLO CANYON - UNITS 1 & 2 Revision 4 8S91DW04.DOC - R4 45

Pressurizer PORVs B 3.4.11 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 supply for the Class I PORV operates properly when called upon. The Frequency of 24 months is based on a typical refueling cycle and the Frequency of the other Surveillances used to demonstrate Class I PORV OPERABILITY.

SR 3.4.11.4 Performance of a COT is required on each required Class I PORV to verify and, as necessary, adjust its lift setpoint. PORV actuation could depressurize the 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 every 24 months to 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, 9c6id for Operation aqamantenance of ear

..-..-- Plante,1987, with 1988 Aienda, Part 10.

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 8S91DW04.DOC - R4 55

RCS PIV Leakage B 3.4.14 BASES (continued)

REFERENCES 1. 10 CFR 50.2.

2. 10 CFR 50.55a(c).

3. 10 CFR 50, Appendix A, Section V, GDC 55.

4. WASH-1400 (NUREG-75/014), Appendix V, October 1975.

5. NUREG-0677, May 1980.

6. Not Used

7. ASME Codeýe oOperation and Mai aenanc f N,ucle ower

-wLPIants, 1987, with1988Adden a*Part1

8. 10 CFR 50.55a(g).

9. AR A0569744.

DIABLO CANYON - UNITS 1 & 2 Revision 4 8S91DW04.DOC - R4 81

ECCS - Operating B 3.5.2 BASES SURVEILLANCE SR 3.5.2.8 (continued)

REQUIREMENTS Opening one containment recirculation sump level transmitter (LT) access hatch in MODES 1 through 3, or in MODE 4 prior to core offload, also requires entry into TS 3.0.3. The containment recirculation sump can be considered OPERABLE with one containment sump LT access hatch open only during the initial entry into MODE 4 following core reload. This is acceptable since core decay heat is low following core reload and automatic actuation of containment spray would not occur in the event of an RCS pipe break.

Without containment spray actuation, less RWST inventory is discharged into containment and the containment water level remains below the level of the access hatch. Additionally, with one LT~access hatch open in MODE 4, high risk foreign material exclusion procedures must be followed. (Ref 10)

REFERENCES 1. 10 CFR 50, Appendix A, GDC 35.

2. 10 CFR 50.46.

3. FSAR, Sections 6.3 and 7.3.

4. FSAR, Chapter 15, "Accident Analysis."

5. NRC Memorandum to V. Stello, Jr., from R.L. Baer, "Recommended Interim Revisions to LCOs for ECCS Components," December 1, 1975.

6. IE Information Notice No. 87-01.

7. BTP EICSB-118, Application of the Single Failure Criteria to Manually-Controlled Electrically-Operated Valves.

8. SME/ANSI OM-I9, "Operational Main nance of Nucl ar Pow Plants", includin gM-a-1988 addend , art 6, "Inse ' e Tes* g kf Pumps in L jt Water Reactor Pwer Plants:," d part

¶ S.-a,--F*"T -

• tInservice T(Kstina of Valves in ,ght Water Rea tor Po r Plants."

9. NRC letter to PG&E, EA 89-241, April 5, 1990; CHRON 148598.

10. Technical Specification Interpretation 90-07, Revision 1, March 9, 1999.

DIABLO CANYON - UNITS 1 & 2 Revision 4 8S91DX04.DOC - R4 18

MSSVs B 3.7.1 BASES ACTIONS A.1 (continued)

The allowed Completion Time is reasonable base on operating experience to complete the Required Actions in an orderly manner without challenging unit systems.

B.1 and B.2 If THERMAL POWER and Power Range Neutron Flux Trip are not reduced as required by Table 3.7.1-1 within the associated Completion Time, or if one or more steam generators have less than two MSSVs OPERABLE, 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 <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. 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 3.7.1.1 REQUIREMENTS This SR verifies the OPERABILITY of the MSSVs by the verification of each MSSV lift setpoint in accordance 'th the Inservice Testing -0 Program. The ASME Code, o (Ref. 5), requires t afety and relief valve tests be performed in accordance with A- ASME Cc...

& -ef. 5--According to Reference 6, the following tests are required:

a. Visual examination;

b. Seat tightness determination;

c. Setpoint pressure determination (lift setting);

d. Compliance with owner's seat tightness criteria; and

e. Verification of the balancing device integrity on balanced valves.

The AN15ASME StafldaTr-requires that all valves be tested every 5 years, and a minimum of 20% of the valves be tested every 24 months.

The ASME Code specifies the activities and frequencies necessary to satisfy the requirements. Table 3.7.1-2 allows a +/- 3% setpoint (as-found lift point) tolerance on the valves for OPERABILITY (with the exception of the lowest set MSSV setpoint, which is (+3%/-2%);

however, the valves are reset to +/- 1% during the Surveillance to allow for drift. The lift settings, according to Table 3.7.1-2 in the accompanying LCO, correspond to ambient conditions of the valve at nominal operating temperature and pressure.

(continued)

DIABLO CANYON - UNITS 1 & 2 Revision 4 8S91DZ04. DOC - R4a 5

MSSVs B 3.7.1 BASES SURVEILLANCE SR 3.7.1.1 (continued)

REQUIREMENTS This SR is modified by a Note that allows entry into and operation in MODE 3 prior to performing the SR. The MSSVs may be either bench tested or tested in situ at hot conditions using an assist device to simulate lift pressure. If the MSSVs are not tested at hot conditions, the lift setting pressure shall be corrected to ambient conditions of the valve at operating temperature and pressure.

REFERENCES 1. FSAR, Section 10.3.1.

2. ASME Boiler and Pressure Vessel Code,Section III, 1968.

3. FSAR, Section 15.2 and 15.3.

4. NRC Information Notice IN-94-60, "Potential Overpressurization of the Main Steam System." Auaust 22. 1994.

5.~ ~~ ~ ~~

Prsli* ~~~~t*___ l.'..inn"x..'-

P'= Pr..R 56'. C L  !^A 1o'_r. '* ....

.* ,* Westinghouse Report WCAP-1 1082, Revision 5, "Westinghouse Setpoint Methodology for Protection Systems Diablo Canyon Units 1 and 2, 24 Month Fuel Cycle Program."

9' PG&E Design Calculation N-1 14, "Over-Pressure Study for One MSSV Per Loop Unavailable", dated 3/10/94.

PG&E Design Calculation N-1 15, "Reduced Power Levels for A Number of MSSVs Inoperable", dated 3/14/94.

DIABLO CANYON - UNITS 1 & 2 Revision 4 8S91DZ04.DOC - R4a 6

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 ar ..not tested at power, they are exempt from the ASME Code e (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, PaT, *,e 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 returning the plant to operation following a refueling outage. The frequency of MSIV testing is every 24 months. The 24 month Frequency is based on the refueling cycle. Operating experience has shown that these components usually pass the Surveillance when performed at the 24 month Frequency. Therefore, this Frequency is acceptable from a reliability standpoint.

REFERENCES 1. FSAR, Section 10.3.

2. FSAR, Section 6, Appendix 6.2 C.

3. FSAR, Section 15.4.2.

4. 10 CFR 100.11.

-*..AME, Builei dd r*zrAo' , ssel Code. Srtinn Y_.lý

6. ANSi/ASivE Orv-I-1987, (linciudui,, ,-,M a 1088 ^DDnDAMn).

DIABLO CANYON - UNITS 1 & 2 Revision 4 8S91DZ04.DOC - R4a 11

MFIVs, 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 MFRV, 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 actuation signal. This Surveillance is normally performed upon returning the plant to operation following a refueling outage. The Frequency of MFIV, MFRV, MFRV bypass valve, and MFWP turbine stop valve testing is every 24 months. The 24 month Frequency is based on the refueling cycle. Operating experience has shown that these components are reliable and can be expected to pass the Surveillance when performed at the 24 month Frequency. Therefore, this Frequency is acceptable from a reliability standpoint.

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. AN*SiiASMF_ Oivi- -i-i987,(liicludi,, M aM-1030 ADDENDA5.*-

A>

DIABLO CANYON - UNITS 1 & 2 Revision 4 8S91DZ04.DOC - R4a 18

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-0-1 557 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.

The 24 month frequency, based on engineering judgement, is consistent with Referencesand9.

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.

3. DCM S-313.

A-- ASME GM-2001 (inng 2002 and 2003 DDNU) 4 *. 10 CFR 50.55a(b)(3)(vi).

DIABLO CANYON - UNITS 1 & 2 Revision 4 8S91DZ04.DOC - R4a 31

AC Sources - Operating B 3.8.1 BASES SURVEILLANCE SR 3.8.1.20 REQUIREMENTS This Surveillance demonstrates that the DG starting independence has (continued) 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 10 year Frequency is consistent with the recommendations of Regulatory Guide 1.108 (Ref. 9).

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 90°F but less than 1750 F.

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. wz,"

11. -ASE, BuiV9,,t, ,d Pressure V e*eCod,' Scion

.... X.-

12. Generic Letter 94-01, "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 166/167, 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 8S91EA04.DOC - R4a 30

Enclosure 4 PG&E Letter DCL-06-143 RETYPED TS PAGES 1

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.7 Reactor Coolant Pump Flywheel Inspection Program This program shall provide for the inspection of each reactor coolant pump flywheel per the recommendations of Regulatory Position c.4.b of Regulatory Guide 1.14, Revision 1, August 1975.

In lieu of Position c.4.b(1) and c.4.b(2), a qualified in-place UT examination over the volume from the inner-bore of the flywheel to the circle one-half of the outer radius or a surface examination (MT and/or PT) of exposed surfaces of the removed flywheels may be conducted at approximately ten year intervals coinciding with the Inservice Inspection schedule as required by ASME Section XI.

5.5.8 Inservice Testina Program This program provides controls for inservice testing of ASME Code Class 1, 2, and 3 components. The program shall include the following:

a. Testing frequencies applicable to the ASME Code for Operations and Maintenance of Nuclear Power Plants (ASME OM Code) and applicable Addenda as follows:

ASME OM Code and applicable Addenda terminology for Required Frequencies for inservice testing activities performing inservice testing activities Weekly At least once per 7 days Monthly At least once per 31 days Quarterly or every 3 months At least once per 92 days Semiannually or every 6 months At least once per 184 days Every 9 months At least once per 276 days Yearly or annually At least once per 366 days Biennially or every 2 years At least once per 731 days

b. The provisions of SR 3.0.2 are applicable to the above required Frequencies and to other normal and accelerated Frequencies specified as 2 years or less in the Inservice Testing Program for performing inservice testing activities;

c. The provisions of SR 3.0.3 are applicable to inservice testing activities; and

d. Nothing in the ASME OM Code shall be construed to supersede the requirements of any TS.

I (continued)

DIABLO CANYON - UNITS 1 & 2 TS Retypes-S.doc - 9 5.0-9 Unit 1 - Amendment No. 135 RXX Unit 2 - Amendment No. 4-35