Text

Response to NRC Request for Additional Information Regarding License Amendment Request 06-04, Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity (TSTF-449)
	ML063380448
Person / Time
Site:	Diablo Canyon
Issue date:	11/22/2006
From:	Becker J; Pacific Gas & Electric Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	DCL-06-130, LAR 06-04, OL-DPR-80, OL-DPR-82, TSTF-449
	Download: ML063380448 (84)
	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, CA 93424 Station Director 805.545.3462 November 22, 2006 Fax: 805.545.4234 PG&E Letter DCL-06-130 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Docket No. 50-275, OL-DPR-80 Docket No. 50-323, OL-DPR-82 Diablo Canyon Units 1 and 2 Response to NRC Request for Additional Information Re-garding License Amendment Request 06-04, "Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity (TSTF-449)"

Dear Commissioners and Staff:

PG&E Letter DCL-06-061, dated May 30, 2006, submitted License Amendment Request (LAR) 06-04, "Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity (TSTF-449)." LAR 06-04 proposes to revise Technical Specification (TS) requirements related to steam generator (SG) tube integrity.

On October 26, 2006, the NRC staff requested additional information required to complete the review of LAR 06-04. PG&E's responses to the staff's questions are provided in Enclosure 1. Enclosure 2 provides marked-up TS pages, Enclosure 3 provides retyped TS pages, and Enclosure 4 provides marked-up TS Bases pages.

These three enclosures supersede Enclosures 2, 3 and 4 of PG&E Letter DCL-06-061 in their entirety.

The TS changes add structural integrity performance criteria, clarify the accident induced leak rate limit, revise the SG alternate tube repair criteria for consistency with TSTF-449, clarify the provisions for SG tube inspections, and revise the TS 5.6.10 reporting requirements. The TS Bases changes clarify the accident analyses assumptions for primary to secondary leakage and the basis for the accident induced leakage performance criterion.

This information does not affect the results of the technical evaluation or the no significant hazards consideration determination previously transmitted in PG&E Letter DCL-06-061.

PG&E makes no regulatory commitments (as defined by NEI 99-04) in this letter.

This letter includes no revisions to existing regulatory commitments.

Aool A

member of the STARS (Strategic Teaming and Resource Sharing)

Alliance Callaway

Comanche Peak

Diablo Canyon

° Palo Verde e South Texas Project

Wolf Creek

Document Control Desk November 22, 2006 Page 2 PG&E Letter DCL-06-130 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.

Executed on November 22, 2006.

initials kjse/4328 Enclosures cc:

Edgar Bailey, DHS Terry W. Jackson Bruce S. Mallett Diablo Distribution cc/enc: Alan B. Wang A

member of the STARS (Strategic Teaming and Resource Sharing)

A[liance Callaway 9 Comanche Peak.

Diablo Canyon

Palo Verde

South Texas Project

Wolf Creek PG&E Letter DCL-06-130 ENCLOSURE 1 Response to NRC Request for Additional Information Regarding License Amendment Request 06-04, "Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity (TSTF-449)"

1.

Per your proposed structural integrity performance criterion, a safety factor of 1.4 against burst will be applied to the design basis accident primary to secondary pressure differentials. However, Generic Letter 95-05, "Voltage-Based Repair Criteria for Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking," indicated that there is a possibility that a tube may have a burst pressure less than 1.4 times the steam line break pressure differential (given the uncertainties associated with the various correlations); therefore, the GL 95-05 alternate repair criteria (ARC) imposed a limit on the probability of burst (POB) of lx10-2. As a result, it is not clear from your submittal that the structural integrity performance criteria is complete since it does not fully address all the performance criteria for implementation of the voltage-based ARC. Please discuss your plans to modify the performance criteria to fully address the voltage-based ARC. For example, discuss your plans for modifying the structural integrity performance criteria to indicate that for predominantly axially-oriented outside diameter stress corrosion cracking (ODSCC) at the tube support plate elevations, the POB of one or more indications given a steam line break shall be less than lx10-2.

Upon incorporation of this criterion into the structural integrity performance criterion, please discuss your plans to eliminate the associated reporting requirement in proposed Technical Specification (TS) 5.6. 10.b.3 since operation in excess of this limit will not be permitted.

Wording such as the following may address the above questions/comments:

"This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary to secondary pressure differential and, except for flaws addressed through application of the alternate repair criteria discussed in Specification 5.5.8.c. 1, a safety factor of 1.4 against burst applied to the design basis accident primary to secondary pressure differentials."

"When alternate repair criteria discussed in Specification 5.5.9. c. 1 are applied to axially-oriented outside diameter stress corrosion cracking indications at tube support plate locations, the probability that one or more of these indications in a SG will burst under postulated main steam line break conditions shall be less than lx 10-2."

1 PG&E Letter DCL-06-130 The staff notes that similar modifications may need to be made to reflect the performance criteria associated with implementation of the axial primary water stress corrosion cracking (PWSCC) depth-based repair criteria.

Incorporation of these criteria may also eliminate the need for the reporting requirements in proposed TS 5.6. 10. e. 1.

PG&E Response:

PG&E has revised proposed TS 5.5.9.b.1, as provided below in italicized wording, to relocate the 1x10-2 probability of burst reporting criteria for PWSCC ARC and ODSCC ARC into the structural integrity performance criteria (SIPC) section of the TS.

"This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and, except for flaws addressed through application of the alternate repair criteria discussed in Specification 5.5.9.c. 1 and 5.5.9.c.3, a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials."

"When alternate repair criteria discussed in Specification 5.5.9. c. 1 are applied to axially-oriented outside diameter stress corrosion cracking indications at tube support plate locations, the probability that one or more of these indications in a SG will burst under postulated main steam line break conditions shall be less than lx10-2.

"When alternate repair criteria discussed in Specification 5.5.9. c. 3 are applied to axially-oriented primary water stress corrosion cracking indications at tube support plate locations, the probability that one or more of these indications in a SG will burst under postulated main steam line break conditions shall be less than lx 10-2.

PG&E has also deleted proposed TS 5.6.10.b.3 (current TS 5.6.10.d.3) and proposed TS 5.6.10.e.1 (current TS 5.6.10.g.1)to remove NRC reporting requirements related to exceeding the 1x10-2 probability of burst criteria for ODSCC ARC and PWSCC ARC.

2.

Proposed TS 5.5.9.b.2 describes the accident-induced leak rate limit. There is an exception allowed to the 1 gallon per minute (gpm) limit which references paragraph c of this TS section. Since TS paragraph c addresses all degradation mechanisms, please discuss your plans to modify your submittal to be more specific with respect to which degradation mechanisms the I gpm limit does not apply. In addition, as currently written it appears that the second sentence of the accident-induced leakage performance criteria essentially limits leakage from all sources to 10.5 gpm, which is inconsistent with TSTF-449 which is intended to limit leakage from non-alternate tube 2

PG&E Letter DCL-06-130 repair criteria sources to I gpm. Please discuss your plans to modify the submittal to address this issue.

In addition, it appears that you have made the second sentence an interpretation of your current design and licensing basis. The staff notes that the second sentence in the accident-induced leakage performance criteria in TSTF-449 is not intended to be an interpretation of the current design and licensing basis. Rather, the second sentence (in the accident-induced performance criteria in TSTF-449) is intended to ensure that the potential for induced leakage during severe accidents will be maintained at a level that will not increase risk. This is discussed in the staff's model safety evaluation on TSTF-449. As a result, please discuss your plans to modify your proposal to be consistent with TSTF-449 or justify why only a limit on the "faulted" steam generator is considered necessary to maintain risk at acceptable levels during the severe accident scenarios discussed in the staff's model safety evaluation on TSTF-449.

Wording such as the following may address the above questions/comments:

"Except during a steam generator tube rupture, leakage from all sources excluding the leakage attributed to the degradation described in TS Section [insert appropriate Sections] is also not to exceed I gallon per minute per SG."

The staff notes that reference to the 10.5 gpm limit in your current proposal is not needed since this should be consistent with your current accident analysis (which is addressed by the first sentence in your proposed accident induced leakage performance criterion). The staff also notes that your Bases may also need to be revised to clarify this issue.

PG&E Response:

PG&E has revised the second sentence of proposed TS 5.5.9.b.2 to use the following wording as recommended by the NRC: "Except during a steam generator tube rupture, leakage from all sources, excluding the leakage attributed to the degradation described in Specification 5.5.9.c.1, 5.5.9.c.2, and 5.5.9.c.3, is also not to exceed 1 gallon per minute per SG."

Revisions to the TS Bases are discussed in response to question 14.

3.

With respect to proposed TS 5.5.9.c. 1, please address the following:

a.

The second sentence of the introductory paragraph in this section states that the plugging (repair) limit at tube support plate intersections is based on maintaining SG tube serviceability. Please discuss your plans for removing the phrase "maintaining SG tube serviceability,"

3 PG&E Letter DCL-06-130 since serviceability is not defined in your proposed TS. In addition, this sentence references the "plugging limit." Since this term is no longer defined, discuss your plans to replace it with "repair criteria."

PG&E Response:

PG&E has revised proposed TS 5.5.9.c.1 to replace "plugging limit" with "repair criteria," and to remove the phrase "based on maintaining steam generator tube serviceability," as follows: "At tube support plate intersections, the repair criteria is described below:"

b.

In several locations in this section, the proposed TS use the phrase "the lower voltage repair limit (Note 1)." Please discuss your plans for removing this phrase and replacing it with "2.0 volts," since this is the value applicable to Diablo Canyon. Keeping Note 1 complicates the proposed TS.

PG&E Response:

In proposed TS 5.5.9.c.l.a, b, and c, the phrase "the lower voltage repair limit (Note 1)" has been replaced with "2.0 volts." Note 1 has been deleted from proposed TS 5.5.9.c.1.

c.

Proposed TS 5.5.9.c. 1.b and TS 5.5.9.c. 1.c reference the repair of tubes. Since Diablo Canyon does not have an option for tube repair (i.e., sleeving), please discuss your plans to remove the reference to tube repair from your proposed TSs.

PG&E Response:

In proposed TS 5.5.9.c.l.b and TS 5.5.9.c.1.c, "repaired" has been deleted. In addition, in TS 5.6.10.c.1 "or repaired" has been deleted and in TS 5.6.10.f.1 "repaired" has been replaced by "plugged".

d.

Proposed TS 5.5.9. c. 1. c specifies a particular eddy current probe (i.e.,

"rotating pancake coil inspection'). The proposal would require you to use this technology even if other, more advanced, probes were developed for detecting ODSCC at tube support plates. Please discuss your plans for modifying the TSs to avoid this limitation (e.g.,

"rotating pancake coil inspection or a comparable inspection technique').

PG&E Response:

In proposed TS 5.5.9.c.1.c, the phrase "or comparable inspection technique" has been added after "rotating pancake coil inspection."

4 PG&E Letter DCL-06-130

e.

Proposed TS 5.5.9.c. 1.c references "NOTE 2." To simplify the reading of the TSs, please discuss your plans to incorporate the wording from NOTE 2 directly into the TS (e.g., "(calculated according to the methodology in Generic Letter 95-05 as supplemented)').

PG&E Response:

In proposed TS 5.5.9.c.1.c, the phrase "Note 2" has been deleted and replaced with "calculated according to the methodology in Generic Letter 95-05 as supplemented." Note 2 has been deleted.

f.

As currently proposed, you may elect to implement alternate tube repair criteria. As a result, it would appear that the requirement contained in proposed TS 5.5.9.c.3 to plug tubes which contain a tube support plate intersection with both an axial ODSCC and an axial PWSCC indication should be listed under proposed TS 5.5.9.c. 1.

Please discuss your plans to incorporate this requirement under TS 5.5.9.c,1.

PG&E Response:

Proposed TS 5.5.9.c.3 has been retained, and TS 5.5.9.c.l.e has been added as follows: "A tube which contains a tube support plate intersection with both an axial ODSCC indication and an axial PWSCC indication will be plugged."

4.

With respect to proposed TS 5.5.9.c.2, please address the following:

a.

Proposed TS 5.5.9. c. 2. e references "serviceability." Since serviceability is not defined in your proposed TS, please discuss your plans for removing it. In addition, this sentence references the "plugging (repair) limit." Since this term is no longer defined, discuss your plans to replace it with "repair criteria."

PG&E Response:

PG&E has revised proposed TS 5.5.9.c.2.e to replace "plugging limit" with "repair criteria", and to remove the phrase "based on maintaining steam generator serviceability", as follows: "Within the hot leg tubesheet, the repair criteria is described below:"

b.

Proposed TSs 5.5.9.c.2.e.4 and 5 reference tube repair. Since Diablo Canyon does not have an option for tube repair (i.e., sleeving), please discuss your plans to remove the reference to tube repair from your proposed TSs.

5 PG&E Letter DCL-06-130 PG&E Response:

In proposed TS 5.5.9.c.2.e.4 and 5, "or repaired" is deleted.

c.

Proposed TS 5.5.9. c. 2 describes the W* repair criteria. Given that your current TS do not address the cold-leg tubesheet region, there were only a few instances in your current TS that indicate the W* repair criteria applies only to the hot-leg. However, in your current proposal, inspections in the cold-leg tubesheet would be required by your TS.

As a result, please discuss you plans to modify your proposal to clearly indicate that the W* alternate repair criteria could be applied only to the hot-leg side of the tubesheet (i.e., by referencing the hot-leg throughout proposed TS 5.5.9. c. 2).

PG&E Response:

In proposed TS 5.5.9.c.2, TS 5.5.9.c.2.b, and TS 5.5.9.c.2.e, "hot leg" is added before "tubesheet" to clarify that W* ARC is only applied to the hot leg side of the tubesheet.

5.

Proposed TS 5.5.9. c.3 indicates that under certain circumstances tubes should be "removed from service." Since plugging is referenced throughout your proposed TS, please discuss your plans to replace this phrase with "plugged".

PG&E Response:

In proposed TS 5.5.9.c.3, "removed from service" is replaced by "plugged."

6.

Proposed TS 5.5.9. d. 4 references the "Tube Support Plate Voltage-Based Repair Criteria" discussed in TS 5.5.9.c. 1 using different terminology. Please discuss your plans to consistently refer to the repair criteria and for referencing the applicable specification in this section (i.e., reference TS 5.5.9.c. 1).

PG&E Response:

In proposed TS 5.5.9.d.4, "tube/tube support plate repair criteria" is replaced with "tube support plate voltage-based repair criteria," and references to Specification 5.5.9.c.1 are added.

7.

Please discuss your plans to reference TS 5.5.9.c.3 in TS 5.5.9.d.6 (e.g.,

"...to implement the axial PWSCC depth-based repair criteria discussed in Technical Specification 5.5.9.c.3).

6 PG&E Letter DCL-06-130 PG&E Response:

In proposed TS 5.5.9.d.5, the phrase "in Specification 5.5.9.c.2" is added after "Implementation of the W* repair criteria." In proposed TS 5.5.9.d.6, the phrase "in Specification 5.5.9.c.3" is added after "to implement axial PWSCC depth-based repair criteria."

8.

Proposed TS 5.6. 10.b refers to the "Tube Support Plate Voltage-Based Repair Criteria" in TS 5.5.9. c. I as "voltage-based repair criteria to tube support plate intersections." Please discuss your plans to modify this proposed TS to indicate that for implementation of the "alternate tube repair criteria in TS 5.5.9.c. 1" the indicated report will be made.

PG&E Response:

Proposed TS 5.6.10.b is revised as follows to respond to this question and question 10: "For implementation of the tube support plate voltage-based repair criteria in Specification 5.5.9.c.1, notify the NRC prior to the initial entry into MODE 4 should any of the following arise:"

9.

Given that your proposed TS do not allow operation when the accident-induced leakage criteria is exceeded, please discuss your plans to omit TS Section 5.6. 10.b. 1.

PG&E Response:

The proposed TS 5.6.10.b.1 reporting requirement has been deleted.

10.

To be consistent with the other reporting requirements, discuss your plans to remove the reference to 10 CFR 50.4 in proposed TS 5.6. 10.c., d., and g. In addition, discuss your plans to indicate that the reports in these sections (and proposed TS 5.6. 10. 0 will be submitted "after the initial entry into MODE 4 following completion of an inspection performed in accordance with Specification 5.5.9, Steam Generator (SG) Program. This will make these reporting requirements consistent with the TSTF-449 reporting requirement.

PG&E Response:

In proposed TS 5.6.10.c, TS 5.6.10.d, TS 5.6.10.f, and TS 5.6.10.g, the wording has been revised to delete the reference to 10 CFR 50.4 and to state that the reports will be submitted within 90 or 120 days, as appropriate, after the initial entry into MODE 4 following completion of an inspection performed in accordance with Specification 5.5.9, Steam Generator (SG) Program.

11.

Since failure of the steam generator performance criteria is required (by 10 CFR 50. 72/50.73) to be reported and future operation is not permitted 7

PG&E Letter DCL-06-130 when the performance criteria (i.e., tube integrity) are not maintained, discuss your plans to delete proposed TS 5.6. 10.e.2.

PG&E Response:

The proposed TS 5.6.10.e.2 reporting requirement has been deleted.

12.

One of the purposes of TSTF-449 is to allow licensees to update their TSs to accurately reflect their SG tube integrity program. For implementation of the voltage-based tube repair criteria, licensees have submitted "90-day reports" providing information concerning tube pulls and condition monitoring/operational assessment results. Consistent with the philosophy of TSTF-449, please discuss your plans to modify TS Section 5.6.10 to include a requirement to provide the information described in Section 6b of Attachment I of GL 95-05 to the U.S. Nuclear Regulatory Commission.

PG&E Response:

Proposed TS 5.6.10.e has been revised to add a requirement to submit the voltage-based repair criteria 90-day reporting information described in Section 6b of Attachment 1 of GL 95-05 to the NRC.

13.

In several locations in the proposed TS 5.5.9 (e.g., 5.5.9.d.4, 5.5.9.d.5) you reference a "refueling outage inspection." Under the proposed TS, inspections need not be performed during a refueling outage. They only need to be performed at intervals not to exceed 24 effective full power months or one operating interval between refueling outages (whichever is less). As a result, if you were to elect to perform inspections at times other than refueling outages, you would still be required to inspect during the refueling outage. Please discuss your plans to modify your proposal to require the inspections every 24 effective full power months or one refueling outage (whichever is less).

PG&E Response:

Proposed TS 5.5.9.d.4 and TS 5.5.9.d.5 for ODSCC ARC and W* ARC have been revised to replace the phrase "during all future refueling outages" with "every 24 effective full power months or one refueling outage, whichever is less."

14.

With respect to Insert 1 to the Reactor Coolant System Operational Leakage Bases Section, the proposed wording does not match the wording in TSTF-449. For example, the reference that the leak rate may increase as a result of accident-induced conditions was not included. In addition, the wording is not clear (i.e., there is I gpm primary-to-secondary leakage to all steam generators). Please discuss your plans to modify your Bases to 8

PG&E Letter DCL-06-130 indicate that the leak rate can increase due to accident conditions and to clarify whether the assumption on accident induced leakage is 1 gpm from all steam generator or I gpm from each steam generator (resulting in 4 gpm).

Similarly, the wording for Insert 2 is not clear (i.e., I gpm total primary to secondary leakage to all steam generators). Please discuss your plans to clarify how much primary-to-secondary leakage was assumed in your accident analysis. Similar comments apply to the Steam Generator Tube Integrity Bases Section (first two paragraphs in the Applicable Safety Analyses Section, and second from last paragraph in the Limiting Condition for Operation Section). With respect to the discussion on the accident-induced leakage in the Limiting Condition for Operation Section in the Steam Generator Tube Integrity Bases Section, this discussion may need to be modified depending on the response to question 2 above.

PG&E Response PG&E has revised the applicable sections of the proposed Reactor Coolant System Operational Leakage Bases and the Tube Integrity Bases that discuss primary to secondary leakage. Wording from TSTF-449 was used to the extent possible. The revised Bases provide the following primary to secondary leak rate assumptions and clarifications:

" The safety analysis for the SGTR event assumes that the total primary to secondary leakage is 1 gpm from the intact SGs plus the leakage rate associated with a double-ended rupture of a single tube.

The safety analysis for the SLB event assumes that the primary to secondary leakage is 10.5 gpm in the faulted SG, or increases to 10.5 gpm as a result of accident induced conditions, and is 0.1 gpm for each intact SG.

The safety analyses for events other than SLB and SGTR assume that the total primary to secondary leakage is 0.75 gpm (approximately 0.19 gpm per SG) from the SGs.

15.

Please discuss the reasons for deleting the sentence in TSTF-449 that indicates, "The accident induced leakage rate includes any primary to secondary LEAKAGE existing prior to the accident in addition to primary to secondary LEAKAGE induced during the accident." In addition, discuss your plans to include this sentence in your TS Bases.

PG&E Response:

PG&E will include the following TSTF-449 sentence in the proposed TS 3.4.17 Bases for Tube Integrity LCO section: "The accident induced leakage rate includes any primary to secondary LEAKAGE existing prior to the accident in addition to primary to secondary LEAKAGE induced during the accident."

9 PG&E Letter DCL-06-130 Proposed Technical Specification Changes (marked-up)

Definitions 1.1 primary to secondary I

(primary to secondary LEAKAGE) 1.1 Definitions LEAKAGE (continued)

MASTER RELAY TEST MODE

3.

eactor Coolant Sstq (RCS) LEAKAGE thg h

a team generatorfSG to the Secondary Syste.

b.

Unidenti d LEAKAGE All LEAKA (except RCP seal water injection or leakoff) that is ot identifie LEAKAGE.

c.

Pressure Bound LEA LEAKAGE (excep S LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.

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

A MODE shall correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation. These tests are:

a.

Described in Chapter 14 of the FSAR;

b.

Authorized under the provisions of 10 CFR 50.59; or

c.

Otherwise approved by the Nuclear Regulatory Commission.

OPERABLE-OPERABILITY PHYSICS TESTS (continued)

DIABLO CANYON - UNITS 1 & 2 1.1-4 Unit 1 - Amendment No. 4-,5, Unit 2 - Amendment No. 4-35,

RCS Operational LEAKAGE 3.4.13 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.13 RCS Operational LEAKAGE LCO 3.4.13 RCS operational LEAKAGE shall be limited to:

a.

No pressure boundary LEAKAGE;

b.

1 gpm unidentified LEAKAGE;

c.

10 gpm identified LEAKAGE; and

d.

150 allons per day primary to secondary LEAKAGE through any oneE9.,...* -

steam generator (SG)

I I1-1ý Ioperational APPLI ABILITY:

ACTIO MODES 1, 2, 3*, and 4*.

or primary to secondary LEAKAGE CONDITION REQUIRED ACTION COMPLETION TIME A.

RCS LEAKAGE not wi A.1 Reduce LEAKAGE to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months limits for reasons

.er than within limits.

pressure bou ary LEAKAGE.

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 of Condition A not met.

OR Pressure boundary B.2 Be in MODE 5.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months LEAKAGE exists.

OR Primary to secondary LEAKAGE not within limit.

For MODES 3 and 4, if steam generator water samples indicate less than the minimum detectable activity of 5.0 E-7 microcuries/ml for principal gamma emitters, the leakage requirement of specification 3.4.13.d. may be considered met.

DIABLO CANYON - UNITS 1 & 2 3.4-27 Unit 1 - Amendment No. 4-35 442, Unit 2 - Amendment No. 35 4-42,

2. Not applicable to primary to secondary LEAKAGE.

SURVEILF 1NE REQUIREMENTS RCS Operational LEAKAGE 3.4.13 SURVEILLANCE FREQUENCY SR 3.4.13.1

NOT ot required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after Insert stablishment of steady state operation.

P......m.ROS.watr..entr..

baLne,.... _.

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months I

(SR 3.4.1322 erf angnerator tube integrity is in accordlance*

with the Steam Generator Tube Surveillance Program.

In accordance with the Steam Generator Tube Surveillance Program.

N SR 3.4.11ý erify steam gen rat ary-to-se-con-dary leýaage equal to or less than Ilowable.

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Insert 3 I nsert2 DIABLO CANYON - UNITS 1 & 2 3.4-28 Unit 1 - Amendment No. 4-35, Unit 2 - Amendment No. 1-45,

There are no changes to the page.

Page included for information only.

RCS Specific Activity 3.4.16 E

0) 250 0

200 I-0 0

a-U)150 I-_

z

.100 lo w

zW 50

}--

D a

W o

C')

0o) 0

((I]

I I I

I I

UNACCEPTABLE OPERATION I I I I I I I I I 1 1 1 1 1 1 1 ;

I I

1 1 1 ;

1 36 1 1 1 1 1 1 1 1 1 1 1 1 1 1 Il I----1-ACCEPTABLE{

~~OPEATO 20 30 40 50 60 70 80 90 100 PERCENT OF RATED THERMAL POWER 3-401 Figure 3.4.16-1 DOSE EQUIVALENT 1-131 REACTOR COOLANT SPECIFIC ACTIVITY LIMIT VERSUS PERCENT OF RATED THERMAL POWER WITH THE REACTOR COOLANT SPECIFIC ACTIVITY > 1 pCI/GRAM DOSE EQUIVALENT 1-131.

Insert Insert 4 for new TS 3.4.17, SG Tube Integrity, on next page DIABLO CANYON - UNITS 1 & 2 3.4-37 Unit 1 - Amendment No. 435, 184 Unit 2 - Amendment No. 1-35, 186

Programs and Manuals 5.5 Unit 1 5.5 Programs and Manuals (continued) 5.5.9 Steam Generator (SG) TubeS-ureillancdProqram SG tube integrity shall be demnstrfated by performance o-f the foNllowi^;ng augmented

.nc'c inpection proegram.

The p

.roVisioRs of SR 3.0.2 are applicable to the SG Tube u..ei. aNc.e-Pro'gram test freqU8RG~es.

a.

SG Sample Selection) and InpcinSG tube integrity shall be determined durin shutdown by selecting and inspecting at least the minimum number.a SGs specified in Table 5.5.9 1.

b.

SG Tubahe-Sample Selection aqndl Inspectinf The SG tube minimumn sample size inspection resiut classificGat*Eo n, and the correspond.ig action required shall be as specified in Table 5.5.9 2. The inser*Vice inspection of SG tubes shall be Insert 5 performed at the frequencies specified in Specification~ 6.5.9.G and the inspected

Note, tubes shall be verified acceptable per the acceptance criteria of Specification Insert 5 5.5.9.d. The tu.bes stelect...edd foir e..ach s

inspectionshall iRnlude at least 3%

changes TS of the total number. Of tubes in all SGs; the tubes selected for these Section shall be selected on a random basis except:

5.5.9 in its

1.

Where experience in similar: plants with similar water chemistry indicates entirety.

crtial areas to be inspected, then at least 50% of the tubesinpce The shall be from these ritical areas; changes

2.

The first sample of tubes selected for each inse..i.e inspection below (subsequent to the preserVice inspection) of each SG shall identify current a)

All nonplugged tubes, that pr-eviously had detectable wall Section penetFrations (greater than 20%),

5.5.9 text b)

Tubes in those areas where experience has indi*cate tential which is deleted or modified.

G)

A tube inspection (pursuant to Specification 5.5.9.d. 1.h) shall be performed on each selected tube. if any selected tube does not in Seciicaionpermit the passage of the eddy curr~ent probe for a tube in Specification no-......*.

7.

V.

inspectio*n this shall be rFcorded and ann a4daceRt tbe shall be 5.5.9..c.1I

.J.

Note, this is

(@))

Indications left in sernie sult of application of the tube moved to 5.5.9.d.4 support plate voltage-based repair cri e shall be inspected by yý. bobbin coil probaidrn l i rfeigotgs Noe hi sTubes identifie as W* tubes having a previously identified Note, this is indication wit in the flexible W* length shall be inspected using moved to 5.5.9.d.5 RPC) probe or equivalent for the full length o f t h e Io n

.........i n

/

n\\-.

~

efeigetg6 (continued) every 24 effective full power months or one refueling outage, whichever is less q9-DIABLO CANYON - UNITS 1 & 2 5.0-10 Unit 1 - Amendment No

-3.5, 1-54, 182 Unit 2 - Amendment No. 4-5, 4Z4, 184

Programs and Manuals 5.5 Unit 1 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Tube Sur-eilance Program (continued)

3.

The tubes selected as the second and third san iler, (of required by may be subjected to a Table 5.5.9 2) driRg each inservice IR&DeeI DRI inSeifcto partial tube inspection provided:

in Specification 5.5.9.c.1 a)

The tubes seleGted for these samples include the tubes frrm these areas of the tube sheet array where tuibes with imnperfccewr previously foun.d, and voltage-based 9-

~imperfections were previously founRd.

I Impl entation of the steam generator-tube/tube support plate repair criteri requires a 100% bobbin coil inspection for hot-leg and cold-leg Note, this is support plate intersections down to the lowest cold-leg tube support plate moved to 5.5.9.d.4 with known outside diameter stress corrosion cracking (ODSCC) indications. The determination of the lowest cold-leg tube support plate intersection having ODSCC indications shall be based on the performance o

I E3 sta 20% random sampling of tubes inspected over their full length.

9.c.3 5 5dtintersections will be performed in water stress corrosion cracking (PWSCC) depth-based repair c"'

. The.

K Liii extent a) a.

b.

b) of required inspection is:

Note, TS 5.5.9.b.5 is moved to 5.5.9. d. 6 100 percent bobbin coil inspection of all tube support plate (TSP) intersections.

Plus Point coil inspection of all bobbin coil indications at dented TSP intersections.

Plus Point coil inspection of all prior PWSCC indications left in service.

If bobbin coil is relied upon for detection of axial PWSCC in less than or equal to 2 volt dents, then on a SG basis perform Plus Point coil inspection of all TSP intersections having greater than 2 volt dents up to the highest TSP for which PWSCC has been detected in the prior two inspections or current inspection and 20%

of greater than 2 volt dents at the next higher TSP. If a circumferential indication is detected in a dent of "x" volts in the prior two inspections or current inspection, Plus Point inspections will be conducted on 100% of dents greater than "x - 0.3" volts up to the affected TSP elevation in the affected SG, plus 20% of dents greater than "x - 0.3" volts at the next higher TSP. "x" is defined as the lowest dent voltage where a circumferential crack was detected.

(continued)

Unit 1 - Amendment No. 145,152 Unit 2 - Amendment No. 44-5, 152 DIABLO CANYON - UNITS 1 & 2 5.0-11

Programs and Manuals 5.5 Unit 1 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Tube Su-reillance Program (continued) ziv If bobbin coil is not relied upon for detection of axial PWSCC in less than or equal to 2 volt dents, then on a SG basis perform Plus Point coil inspection of all dented TSP intersections (no lower dent voltage threshold) up to the highest TSP for which PWSCC has been detected in the prior two inspections or current inspection and 20% of all dents at the next higher TSP.

f7)

For any 20% dent sample, a minimum of 50 dents at the TSP elevation shall be inspected. If the population of dents is less than 50 at the TSP elevation, then 100% of the dents at the TSP elevation shall be inspected.

The results ofechsmle inpcinshall be classified into one of the following three Gatege&es

~~I no-',r, r.ir,.- Dr*,

,1 G4 Less than 5% of the total tubes inspected are degraded tubes and noe of the inspected tubes are defectiVe-.

PeOne or more tubes, but entar more than 1% of the total tue inspected are defective, or between 5.24 an00% of the total tubes inspected are degraded tbs 0-3-More than 10% of the total tubes inspected are degraded tubes or mrn~e than 1-0, of the inspected tubes are defective Note: In all inspections, previously degraded tubes must exhibit significant (greater th 10%) fufher wall penetratonR to be nclnthe above perfentage GaIGUlatiGR6.

G.

lspect*..

Frec.uencies The above required inse.Vice inspectio.s of SG tubes shall be performed at the following frequencie-sn LThe first npti shall be performed after 6 Effective Full Power Months but within 24 calenda moUnths of initial criticaitye Subsequent nsrceipctosshall be perfo~rmed at intervals Of not less than 12 nor mor~e than 24 calendar monRths after the previou inspection. If two) consecutive inspections not including the preservice inspection, result in all inspection results falling into the C 1 category or i two conRsecutive inspectiOns demon~strate that previously observed degradation-has inot con-.tinued and no additional degradation ha occurred, the inspection interv-al May be extended to a maximumA Of Onc peF 40 enGth-s-(continued)

Note, page number will change due to repagination of section 5.5.9 pages DIABLO CANYON - UNITS 1 & 2 5.0-11 a Unit 1 - Amendment No. 41-5, 152 Unit 2 - Amendment No. 1-5,15

Programs and Manuals 5.5 Unit 1 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Tube Surveillance Program (continued)

2.

if the results of the Onscr-ce inspectien of a SG cenducted in accordance with w Table 5.5.9 2 at 40- Month itrlsfall inCategory' C 3, the inspecto freq61eRy sha*l be ireased to at least once per 20 months. The inc in iRspection frequeRcy shall apply until the subsequen* iRspections satisfy the criteria of Specification 5.5.9.*.1. The interval

.ma" then be extended to a maximum Of once per 10 months; and

3.

Additional, unscheduled inserVice inspectionsG shall be performed on each SG inR accordane with the first sample ispectio specified in Table 5.5.9-2 during the shutdown subsequent to any of the folloWing coRnditions:

a)

ReactoAr*.÷A to, s ecod ay tube leaks (not including leaks originating fromn tuibe to tube sheet welds) in ex-essq o-f the limits ot Specification 311;o b)

A seismi ocurene greater tha;n the Double Design Ea~thquake, G)

A loss Of coolant accident requiring actuat Safety Features, or AX A rn.on

+~-,

Hn f

Aar *+r kn kraI

'ORn o, tre.-Lqnlneerea d..

Acceptance Criteria.

1I As. used in this Specification:

a)

ImPerfection mreaRns aR eXception to the dimensions, finish or contour of a tube from that required by fabrication drawings or specifications. Eddy current testing indications below 200% of the n oin1a tube wall thick*es,*

if detectable, may be considered as mpe4feGtiens; general corrosion occurring On either inside or outside of a tube; G)

Deraded Tube m.eans a tube containing !mpeFfections greater than or equal to 20%0 Of the nominal wall thickness caused by degraatk)ý d)

% Deqradation means the percentage of the tube wall thficGkneqss affec-ted or removed by degradation.

e)

Dfcmenaniperfection of sucoh severity that it exceeds the P'lugging limit. A tube containing a defect is defective; f%

'i Pluggingq Limit mneans the imnperfection depth at or beyond which the tube shall be removed fromn service and is equal to 40% of the APomfin~I tubeP wIIM thir-kneq1;

1) This definitionR does not apply to tube support plate intersections for which the voltage based repair criteria are being applied.

Refer to 5.6.9.d. 1.j for the repair limit applicable to these inRtersectGions.

Note, page number will change due to repagination of section 5.5.9 pages (continued)

DIABLO CANYON - UNITS 1 & 2 5.0-12 Unit 1 - Amendment No. 4Z5 142 Unit 2 - Amendment No. 435 1

P r The axial PWSCC depth-based repair criteria are used for disposition and Manuals 5.5 Unit 1 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Tube Surveilanc,ýe Pro~>am (continued)

Note, this is moved to 5.5.9.c.3 3)-This definition does not apply to axial PWSCC indications, or portions thereof, which are located within the thickness of dented tube support plates which exhibit a maximum depth greater than or equal to 40 percent of the initial tube wall thickness. WCAP-1 5573, Revision 1, provides repair limits 0 aDDlicable to these intersections.

Note, this is A tube which contains a tube support plate intersection with both moved to an axial ODSCC indication and an axial PWSCC indication will 5.5.9.c.3 be Femo'ed frem serVicl pluggedY g)

Uns~erViceable deGcribes the co ndition of au iwiaks Or contains a defect large enuht#fect its structural integrity inthe eVent of a Do IIeuble Des*ig E r

.e, a loss f* Gcooant accident, or a steam. line o 1 4 A

feedwater line break as specified in 5.5.9.G.3, above;..

ijnspection of the SG tube from the tubcenpd-i~niint finn H k-nr i +n. +hn +^~r ~irr r+. f +k-1,4I J*n n*

Li n*

U U*N NU LU LI n*

@ *i @*

uulu The tube support plate voltage-based repa criteria are Note, TS 5.5.9. d. 1 moved tc 5.5.9.c.1

J I..

th e

,,*,*,*R Wh

,,i,*,**

t,

,R t..*

h e, tubesheet bl oew -the Flexible W* Le]*

or elw8 inches~fr-em the hot leg top of tubesheet, Pr se dn

'nr*t n

mean

nIi, h

tif ea-h Wn nk establish a* b se~liR arnnclotnn" I"-'h -

Wnah g

This

n se~ rt;GR sh a*ll be

(/*~

~

~F Plat

,1 P...

"D.'

u i,-;+.

used fo fe disposition of an allo,,

600 steam generator tube for contin ervice that is experiencing

]

predominantly axially oriented o~d diameter stress corrosion cracking jis /

confined within the thickne of the tulle support plates. At tube support

/

- '*ge'er'to t'be

's**

ha ba_*it.P.4..

_,e-.*1ty as describ ~d below:

2.

volt Stea generato,i, Htues 1,

whos e alr eda ton i atri uedainin tser ide.

diamterstres crroson rackng i inthebouns (ohetiubed 1k

)

I Note, page number will change due to repagination of section 5.5.9 pages Implementation of the W* repair criteria in Specification 5.5.9.c.2 requires a 100 percent RPC probe or equivalent 5.0-13 Unit 1 - Amendment No. 135.151 4-Q 18 DIABLO CANYON - UNITS 1 & 2 Unit 2 - Amendment No..35,11,4-52,-184

Programs and Manuals 5.5 Unit 1 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Tube Su-vei*lance Proqram (continued) 2.0 volts (iSteam generator tubes, whose degradation is attrib e to outside K

diameter stress corrosion cracking within the bou s of t e tube support plate with a bobbin voltage greater than I+---

repair limit (NOTE 1), will be repaired er plugged, except as oted in 5.5.9.c. i. c.5.9d1.j (iii) below.

,(iii Steam generator tubes, with indication of potential degradation attributed to outside diameter stress corrosion cracking within th

~bounds of the tube support plate with a bobbin voltage greater thal

c.

the lower vo ltage repair lim it (NOTE

1) but less than or equal to the upper voltage repair limit*(,NOT.E-2), may remain in service if a rotating pancake coil inspectio r~d'es not detect degradation. Steam generator tubes, with i~di~ations of outside diameter stress corrosion cracking degradation saith a

bobbin voltage greater than the upper voltage repair limit (

(,OTE4) will be plugged-es hr-epalr-et.

I Note. oane number will chanae due to reDaaination of section 5.5.9 Daaes I

I 5.0-13a DIABLO CANYON - UNITS 1 & 2

- -- I Unit 1-Amendment No. IM3If1-,1-5,182 Unit 2 - Amendment No. !r3 R5!

e. A tube which contains a tube support plate intersection with both an axial ODSCC indication and an axial PWSCC indication will be plugged.

Programs and Manuals 5.5 Unit 1 5.5 Programs and Manuals 5.5.9 Steam Generator (SG-Tube Sur'eillance Proqram (continued)

RvCe rain intersections as identified in PG&E Letter DCL-03-174, dated De'ember 19, 2003, will be excluded from application of the voltage-

d.

ba ed repair criteria as it is determined that these intersections may col apse or deform following a postulated LOCA + SSE event.

If an unscheduled mid-cycle inspection is performed, the following mid-cycle repair limits apply instead of the limits identified in 5.5.9 c.1.a,5.5.41.

4

(

,.d 5

.5.4.j. The mid-cycle repair 5.5.9.c.l.b, and limits are determined from the following equations:

5.5.9.c.1.c

\\

/

VSL VMURL 1.0 + NDE + Gr (CL - At)

CL VMLRL

= VMURL (VURL -

VLRL) (CL-At)

C L where:

VURL

= upper voltage repair limit VLRL

= lower voltage repair limit VMURL

- mid-cycle upper voltage repair limit based on time into cycle VMLRL

- mid-cycle lower voltage repair limit based on VMURL and time into cycle At

= length of time since last scheduled inspection during which VURL and VLRL were implemented CL

= cycle length (the time between two scheduled steam generator inspections)

VSL

= structural limit voltage Gr NDE

= average growth rate per cycle length

= 95% cumulative probability allowance for nondestructive examination uncertainty (i.e., a value of 20% has been approved by the NRC) 5.5.9.c. 1.a, Implementation of these mid-cycle repair limits should follow the same 5.5.9.c.1.b, and approach as in T 5.5.9..41.j (0), 5.5.941.j (ii), and 5.5.9.4d1.j (iii).

5.5.9.c.1.c (continuled)

Note, page number will change due to repagination of section 5.5.9 pages DIABLO CANYON - UNITS 1 & 2 5.0-14 Unit 1 -Amendment No. -35, 176 Unit 2 - Amendment No. -35, 178

Programs and Manuals 5.5 Unit 1 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Tube S,-'-'Reilance Program (continued)

The W*

.voltage repair limit 1i 2.

0.. olt.. for...8.inch.

TepWaE tubing at DCPP Units 1 and -2.

repair criteria are NOTE 2-Th.u pe. Ivotage repair limit i6 cGalulated acco.ding to the mnethodology in Generic Letter 95 05 as supplemented.

I used for disposition of an alloy 600 steam generator tube for continued service that is experiencing predominately axially Note, TS oriented inside diameter stress corrosion cracking confined within the 5.5.9.d. 1.k is tubesheet, below the bottom of the WEXTEX transition (BWT). As used hot leg moved to i

this specification:

5.5.9.c.2

, (i)) Bottom of WEXTEX Transition (BWT) is the highest of cont between the tube and tubesheet at, or below op-of-tubesh' t as

[a.

determined by eddy current testing.

ii) W* Length is the distance in th ubesheet below the B that precludes tube pull out in the event of the complete ci umferential separation of the tub,evw the W* length. The length is conservatively set a: 1 an undegraded hot le ube length 9

inches for Zone A tubes and 7.0 inches for e B tubeZand 7.5 inchos for ZoneB B tubes. Informati provided in WCAP-14797-P, Revision 2, defines the boundaries one A and Zone B.

J-- (iii)lexible W* Length is the W* le h adjusted for any cracks found within the W* region. The Fl ible W* Length is the total RPG--'

inspected length as meas ed downward from the BWT, nd includes NDE uncertainties and ack lengths within W* as adjust d for growth.

(iv)W* Tube is a tube h degradation within or below the

length that

-is left in service, nd degraded within the limits specified Specificati

()

eWiBm t

e ubesheet, th PlUgging (repair) limit is b low:

1) For tubes to which e W* criteria are applied, the len th of non-5.5.9.0.2.e

degraded tube bel BWT shall be greater than or eq al to the W* length plus ND uncertainties and crack growth fo the 1[

operating cycle.

rotating pancake coil (RPC)

Irepair criteria (continued)

Note, page number will change due to repagination of section 5.5.9 pages DIABLO CANYON - UNITS 1 & 2 5.0-15 Unit 1 - Amendment No. 435,182 Unit 2 - Amendment No.

18e

Programs and Manuals 5.5 Unit 1 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Tube Sur-veillance Program (continued)

2) -Axial cracks in tubes returned to service using W* shall have the upper crack tip below the BWT by at least the NDE measurement uncertainty and crack growth allowance, such that at the end of

-r the subsequent operating cycle the entire crack remains below the BWT.

Gj esolvable, single axial indications (multiple indications must Zj7*

return to the null point between individual cracks) within the flexible W* length can be left in service. Alternate RPC coils or an ultrasonic test (UT) inspection can be used to demonstrate return to null point between multiple axial indications or the absence of circumferential involvement between axial indications.

4) Tubes with inclined axial indications less than 2.0 inches long

.(including the crack growth allowance) having inclination angles relative to the tube axis of < 45 degrees minus the NDE uncertainty, ANDECA, on the measurement of the crack angle can be left in service. Tubes with two or more parallel (overlapping elevation), inclined axial cracks shall be plugged-eorrepaiFed. For application of the 2.0 inch limit, an inclined indication is an axial crack that is visually inclined on the RCP C-scan, such that an angular measurement is required, and the measured angle exceeds the measurement uncertainty of ANDECA.

5) Circumferential, volumetric, and axial indications with inclination

[angles greater than (45 degrees - ANDECA) within the flexible W*

ength shall be plugged-e9F-epaired.

F-T ' 6) Any type or combination of tube degradation below the flexible W*

LU length is acceptable.

2.The SG tube integrity shall be determined after comFpleting the corresponRding actions (plug all tubes eXceeding the plugging limnit) required by Table 5.5.9 2.

The contents and frequency of reporFts concerning the SG tube sur~efillance program shall be in accordance with Specificatien 5.6.10.

(continued)

Note, page number will change due to repagination of section 5.5.9 pages DIABLO CANYON - UNITS 1 & 2 5.0-16 Unit 1 - Amendment No 4_5, 442,182 Unit 2 - Amendment No. 1-3,, 1i42,184

Programs and Manuals 5.5 Unit 1 THIS PAGE NOT USED Note, page number will change due to repagination of section 5.5.9 pages NYON - UNITS 1 & 2 5.0-17 Unit 1 - Amendment No. 435,1-54,182 Unit 2 - Amendment No.

i-11184 DIABLO CA

Programs and Manuals 5.5 Unit 1 5.5 Programs and Manuals (continued)

TABLE 5-.5.9 1 RMIkIIMIIM NUMBIIIR t'E eTmAM GENERATODR I(S*

TI' *E INSPECTED DURING !NSERVICE INSPECTION Preser'.'je Inspection NO Yes Ne,,,f S+t.,m, I ir,,.

Two Three F

Two T-hFee Foý First I.se.ViGe In.spectki All one TwO TWo Secon.d & Subsequent Insereice epe 2

gne3 TABLE NOTATIONS

1.

The nevc inspection may be limited to one SG an a rotating schedule encomnpassing 3 N % of the tubes (where N is the number of SGs in the plant) if the results of the firsto preiou inpections indicate that all SGs are pe~fGormin in a like mnanner. Note that under somne circumstances, the operating conditionsin one-or more SGs ma e on to be more severe than tho~se finother SGs. UJnder sucGh circumIRstances the samnple sequence shall be modified to isetthe most severe conRditions.

2.

The other SG not isetdduorig the first inevc npcinshall be inspei'Aed. The tnhir an' surse1ueRt iseins shoul IId A tta o

i" nstrwifinniq Aes'i1 in 1 4"o*o

-1.........

r

3.

Each of the other twoR126 RGsnt inspected dur~ing the first ieriensctosshall be inRspected durFing the second and third inspections. The fourth and subseguent I

I I

II II al I

i tl i t I

inspections shall fallow the instruc-tions daesonoeci~ in I Ahnvtz r

(rni ued)

Note, page number will change due to repagination of section 5.5.9 pages 5.0-18 DIABLO CANYON - UNITS 1 & 2 Unit 1 - Amendment No.

Unit 2 - Amendment No.1

Programs and Manuals 5.5 Unit 1 5.5 Programs and Manuals (continued)

STErAM GENERATOR (SG2N TSAMPI IRDSPAIMPL SAPL NSETIN2NDSAIVRLE3R AML SAPLE SPECTINSPECTION NSPECTI Sam__

____en Result R

esult A

A-monomum ef-S T-ubes pef S-G7.

Q4 N -A-.

N -.A-N-A7 4

fl i

fl i.

Plug-defo tive tubes anLd ispe additiORE-g 2S tubesF G4 N-.A-N-.A-.

I

41 G-2 Plug defe~tVe tubes-n 4S-t~ibe 0 R this -SG7 G-4 G2*2 Pg-defetive feF-G 3 Fesult of first sanve tu+bes-tiR defeotp,~e tubes a~ie

nspeot2-25 tubes fi eaeh othef Ngtifieat*o pursuantt (2)-of 10-CFR Part 50 G-c3 Peform NC-Fesult-of fwst sagml AlltheN4 S-G-s Somie Ne,4-m N7AA S.G.6 actionR for C 2 G-2-but result-ef ne-add-seGGREI tti~a4 sampe S.G. aFe Add-tG*3 InspeGt all tubes*OiReaoh S..G.and plug defeGtPVe tubes.

N,*of*GatkeR puFsua+~t tG

§50.72 (b) (2) of4O-CGFR Part 50 N-.A-.

-. 3 N %

...h...re.

the number of SGs in the unit, and n is the number of SGs inspected n

duriRg a* irspectOir Note, page number will change due to repagination of section 5.5.9 pages (continued) 5.0-19 DIABLO CANYON - UNITS.1 & 2 Unit 1 - Amendment No. 135 Unit 2 - Amendment No. 135

Programs and Manuals 5.5 Unit 1 5.6 Reporting Requirements (continued) 5.6.10 Steam Generator (SG) Tube Inspection Report

a.

Within 15 days follo wing the m**letion o-f eac--h iO...

i

-nspectin o*fS tube--,

I hthe Rumrber of tubes plugged in each SG shall be reported to the Comm6s-in.,

b.

The complete reuls f the S-G tubi-e inevc npcinshall be submitted to the Commission in a report within 12 mnonths following completion Ofth

]

inspection. This Special Report shall

1)

Number and extent of tubes inspected,

2)

Location and percent o~f wall thickness penetration for each indication of an impr-fecio*n, ad

3)

IdentificatiOnR f tubes plugged.

G.

Results of SG tube finspections, which fall into Category C 3, shall be reported in a Special Repeot to the Commission within 30 days and prior to resumption o plant operation. This report shall provide a description of investigations conducted to determn cas4f the tubep deadto and corrective measures taken to prevent reurr.n.e..

tube support plate I in Specification 5.5.9.c.1 For implementation of the voltage-based repair criteria to Wtbe support plate i=Rtesestienfs, notify the NRC prior

.t ge r to should any of harpseo:I 1.

if estimated leakage based On the projected end Of cycle (or if not practical, siRng the actual measured eRnd of y'cle) voltage distribution, increased by estimated leakage by all other sour.es (a.te.Rate repair criteria and non alternate repair criteria indications), exceeds the leak limit detefmined from the licensing basis dose calculation for the postulated main steamlin~e break rl thle nextL operatiiiy Gyue.

If ODSCC indications are identified that extend beyond the confines of the tube support plate.

3. If the calculated conditional burst probability based On the projected end o~f cycle (or if no~t practical, using the actual measur~ed end Of cycle) voltage, distribution exceeds 1 x 10-, notify the NRC and previde an assessment of I'I~

trle sa'etvT*TR1 1-1GR*ne+ T Tne ncIcurrence.

(continued)

DIABLO CANYON - UNITS 1 & 2 5.0-29 Unit 1 - Amendment No. 1-45, 4-82, Unit 2 - Amendment No. 1-45, 4-84,

Programs and Manuals 5.5 Unit 1 5.6 Reporting Requirements 5.6.10 Steam Generator (SG) Tube Inspection Report (continued)

e.

The results of the inspection of W* tubes shall bre ed te the Com,,mss~en

c.

geI-. ate

" This report shall include:

Identification of W* tub dications and indications that do not meet W*

/*

~r~equirements and d.

re plugged-re are l including the following F-*information:*/J number of indications, the locations of the indications 1

(relative the BWT and TTS), the orientation (axial, circumferential, vol etric, inclined), the radial position of the tube within the tubesheet,

.. ___-_ e W* Zone of the tube, the severity of each indication (estimated depth),

submitted within 90 the side of the tube in which the indication initiated (inside or outside days after the initial diameter), the W* inspection distance measured with respect to the BWT entry into MODE 4 or TTS (whichever is lower), the length of axial indications, the angle of following completion inclination of clearly skewed axial cracks (if applicable), verification that of an inspection the upper crack tip of W* indications returned to service in the prior cycle performed in remain below the BWT by at least the 95% confidence NDE uncertainty accordance with on locating the crack tip relative to the BWT, updated 95% growth rate for Specification 5.5.9, u

.n operational assessment, the cumulative number of indications Steam Generator (SG) detec in the tubesheet region as a function of elevation within the Program tubesheet, the condition monitoring and operational assessment main steamline ak leak rate for each indication and each SG in accordance with the rate methodology described in PG&E Letter

2.

DCL-05-018, dated March 2005, as supplemented by PG&E Letter I

__*

_1*- DCL-05-090, dated August 05,.0

2)

Assessment of whether the results we consistent with expectations and, if not consistent, a description of the propo corrective action.

f.

The aggregate calculated steam line break leakage from lication of all alternate repair criteria and non-alternate repair criteria shall b AI of the stm gnrtors.

Foriplmntto of the repair criteria for axial PVWSCC at dented TSPs, the V, r' I

ý^nr4i+i^nn +ký+

f-H. -

_f

_r;+-;_-

1)

The calculated SG probability of bur~st for condition monitoring eXceeds

2)

The calculated SG leakage for condition monitoring froM all sources (all Insert 7 alternate repair criteria and non alternate repair criteria indGcations) ex-eeds the leakage limit determined from the licensing basis steam line break dose calcul1ation.

(continued)

DIABLO CANYON - UNITS 1 & 2 5.0-30 Unit 1 - Amendment No. 1-35,4--1,4,524-K, Unit 2 - Amendment No. 4-5,45A,-1-2,4-84,

submitted within 120 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with Specification 5.5.9, Steam Generator (SG) Program 5.6 Reporting Requirements Programs and Manuals 5.5 Unit 1 5.6.10 Steam Generator (SG) Tube Inspection Report (continued)

h.

For implementation of the repair criteria for axial PWSCC at dente SPs, the results of the condition monitoring and operational assessments will Fb* Q4Gd to the NRC within 120 days folloWing completion of the inspection. The report will include:

u e

pg ged Tabulations of indications found in the inIesPA-elUeS and tubes left in service under the ARC.

Growth rate distributions for indications found in the inspection and growth rate distributions used to establish the tube repair limits.

Plus Point confirmation rates for bobbin detected indications when bobbin is relied upon for detection of axial PWSCC in less than or equal to 2 volt dents.

For condition monitoring, an evaluation of any indications that satisfy

.burst margin requirements based on the Westinghouse burst pressure model, but do not satisfy burst margin requirements based on the combined ANL ligament tearing and throughwall burst pressure model.

(continued)

DIABLO CANYON - UNITS 1 & 2 5.0-30a Unit 1 - Amendment No. 45,41-54-452,,4-2, Unit 2 - Amendment No. 41-3,4441-,-2,4-84,

Programs and Manuals 5.5 Unit 1 5.6 Reporting Requirements 5.6.10 Steam Generator (SG) Tube Inspection Report (continued)

5.

5)

Performance evaluation of the operational assessment methodology for predicting flaw distributions as a function of flaw size.

Evaluation results of number and size of previously reported versus new PWSCC indications found in the inspection, and the potential need to E

account for new indications in the operational assessment burst evaluation.

Identification of mixed mode (axial PWSCC and circumferential) indications found in the inspection and an evaluation of the mixed mode indications for potential impact on the axial indication burst pressures or Q) leakage.

8)

Any corrective actions found necessary in the event that condition monitoring requirements are not met.

For implementation of the probability of prior cycle detection (POPCD) method, for the voltage-based repair criteria at tube support plate intersections, if the end-of-cycle conditional main steamline break burst probability, the projected main steamline break leak rate, or the number of indications are underpredicted by the previous cycle operational assessment, the following shall bereoted4t

+kh m

ir'r; -rý,

,"-,,1-,r1 + +, -

I r' C'[=D rC) A..,;+k;-

Or ).A-,

f-11r-.4-,

r, r +

Th 1 ý erieof he-ste-am generators:

1)

The assessment of the probable causes for the underpredi ations, proposed corrective actions, and any recommended chan es to probability of detection or growth methodology indicated y potential

~

methods assessments.

2)/

An assessment of the potential need to revise the altern te repair criteria analysis methods if: the burst probability is underpredic ed by more than 0.001 (i.e., 10% of the reporting threshold) or an order f magnitude; or the leak rate is underpredicted by more than 0.5 gpm r an order of magnitude.

3)

An assessment of the potential need to increase the umber of predicted low voltage indications at the beginning of cycle if th total number of as-found indications in any SG are underestimated y greater than 15%

or by greater than 150 indications.

I submitted within 90 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with Specification 5.5.9, Steam Generator (SG)

Program DIABLO CANYON - UNITS 1 & 2 5.0-30b Unit 1 - Amendment No. 4-35,4-54-1-52,4--77,482, Unit 2 - Amendment No. 4-*5,4-5452,1-7-9,49-4,

Technical Specification Inserts Insert 1 Verify RCS operational LEAKAGE is within limits by performance of RCS water inventory balance.

Insert 2 Verify primary to secondary LEAKAGE is < 150 gallons per day through any one SG.

Insert 3 NOTE 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.

Technical Specification Inserts (continued)

Insert 4 SG Tube Integrity 3.4.17 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.17 Steam Generator (SG) Tube Integrity LCO 3.4.17 SG tube integrity shall be maintained.

AND All SG tubes satisfying the tube repair criteria shall be plugged in accordance with the Steam Generator Program.

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

ACTIONS K 1Tr

~

I Separate Condition entry is allowed for each SG tube.

CONDITION REQUIRED ACTION COMPLETION TIME A.

One or more SG tubes A. 1 Verify tube integrity of the 7 days satisfying the tube repair affected tube(s) is criteria and not plugged in maintained until the next accordance with the Steam refueling outage or SG Generator Program.

tube inspection.

AND A.2 Plug the affected tube(s) in Prior to entering accordance with the Steam MODE 4 following the Generator Program.

next refueling outage or SG tube inspection 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 of 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 OR SG tube integrity not maintained.

Technical Specification Inserts (continued)

Insert 4 (continued)

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.17.1 Verify SG tube integrity in accordance with the Steam In accordance with Generator Program.

the Steam Generator Program SR 3.4.17.2 Verify that each inspected SG tube that satisfies the Prior to entering tube repair criteria is plugged in accordance with the MODE 4 following Steam Generator Program.

a SG tube inspection Insert 5 A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following provisions:

a.

Provisions for condition monitoring assessments.

Condition monitoring assessment means an evaluation of the "as found" condition of the&.

tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes. Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged to confirm that the performance criteria are being met.

b.

Performance criteria for SG tube integrity.

SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.

1.

Structural integrity performance criterion: All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down and all anticipated transients included in the design specification) and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and, except for flaws addressed through application of the alternate repair criteria discussed in Specification 5.5.9.c.1 and 5.5.9.c.3, a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the

Technical Specification Inserts (continued)

Insert 5 (continued) assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

When alternate repair criteria discussed in Specification 5.5.9.c.1 are applied to axially-oriented outside diameter stress corrosion cracking indications at tube support plate locations, the probability that one or more of these indications in a SG will burst under postulated main steam line break conditions shall be less than lx10-2.

When alternate repair criteria discussed in Specification 5.5.9.c.3 are applied to axially-oriented primary water stress corrosion cracking indications at tube support plate locations, the probability that one or more of these indications in a SG will burst under postulated main steam line break conditions shall be less than lxi 0-2.

2.

Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG.

Except during a steam generator tube rupture, leakage from all sources, excluding the leakage attributed to the degradation described in Specification 5.5.9.c.1, 5.5.9.c.2, and 5.5.9.c.3, is also not to exceed 1 gallon per minute per SG.

3.

The operational LEAKAGE performance criterion is specified in LCO 3.4.13, "RCS Operational LEAKAGE."

c.

Provisions for SG tube repair criteria.

Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.

The following alternate tube repair criteria may be applied as an alternative to the 40%

depth based criteria:

1.

Tube Support Plate Voltage-Based Repair Criteria The tube support plate voltage-based repair criteria are used for the disposition of an alloy 600 steam generator tube for continued service that is experiencing predominantly axially oriented outside diameter stress corrosion cracking confined within the thickness of the tube support plates. At tube support plate intersections, the repair criteria is described below:

a.

Steam generator tubes, whose degradation is attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with bobbin voltages less than or equal to 2.0 volts, will be allowed to remain in service.

Technical Specification Inserts (continued)

Insert 5 (continued)

b.

Steam generator tubes, whose degradation is attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than 2.0 volts, will be plugged, except as noted in 5.5.9.c.1.c below.

c.

Steam generator tubes, with indication of potential degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than 2.0 volts but less than or equal to the upper voltage repair limit (calculated according to the methodology in Generic Letter 95-05 as supplemented), may remain in service if a rotating pancake coil inspection or comparable inspection technique does not detect degradation. Steam generator tubes, with indications of outside diameter stress corrosion cracking degradation with a bobbin voltage greater than the upper voltage repair limit will be plugged.

d.

Certain intersections as identified in PG&E Letter DCL-03-174, dated December 19, 2003, will be excluded from application of the voltage-based repair criteria as it is determined that these intersections may collapse or deform following a postulated LOCA + SSE event.

e.

A tube which contains a tube support plate intersection with both an axial ODSCC indication and an axial PWSCC indication will be plugged.

f.

If an unscheduled mid-cycle inspection is performed, the following mid-cycle repair limits apply instead of the limits identified in 5.5.9.c.1.a, 5.5.9.c.1.b, and 5.5.9.c.l.c. The mid-cycle repair limits are determined from the following equations:

VMURL VSL (CL-At) 1.0 + NDE + Gr CLCL VMLRL MR VR-VLL (CL-At)

MURL(VUL VRL)

CL where:

VURL

= upper voltage repair limit VLRL

= lower voltage repair limit VMURL = mid-cycle upper voltage repair limit based on time into cycle VMLRL = mid-cycle lower voltage repair limit based on VMURL and time into cycle At

= length of time since last'scheduled inspection during which VURL and VLRL were implemented

Technical Specification Inserts (continued)

Insert 5 (continued)

CL

= cycle length (the time between two scheduled steam generator inspections)

VSL

= structural limit voltage Gr

= average growth rate per cycle length NDE

= 95% cumulative probability allowance for nondestructive examination uncertainty (i.e., a value of 20% has been approved by the NRC)

Implementation of these mid-cycle repair limits should follow the same approach as in TS 5.5.9.c.l.a, 5.5.9.c.1.b, and 5.5.9.c.1.c.

2.

W* Repair Criteria The W* repair criteria are used for disposition of an alloy 600 steam generator tube for continued service that is experiencing predominately axially oriented inside diameter stress corrosion cracking confined within the hot leg tubesheet, below the bottom of the WEXTEX transition (BWT). As used in this specification:

a.

Bottom of WEXTEX Transition (BWT) is the highest point of contact between the tube and tubesheet at, or below the top-of-tubesheet as determined by eddy current testing.

b.

W*. Lpength is the distance in the hot leg tubesheet below the BWT that precludes tube pull out in the event of the complete circumferential separationr of the tube below the W* length. The W* length is conservatively set at an undegraded hot leg tube length bf 5.2 inches for Zone A tubes and 7.0 inches for Zone B tubes. Information provided in WCAP-14797-P, Revision 2, defines the boundaries of Zone A and Zone B.

c.

Flexible W* Length is the W* length adjusted for any cracks found within the W* region. The Flexible W* Length is the total rotating pancake coil (RPC) inspected length as measured downward from the BWT, and includes NDE uncertainties and crack lengths within W* as adjusted for growth.

d.

W* Tube is a tube with degradation within or below the W* length that is left in service, and degraded within the limits specified in Specification 5.5.9.c.2.e.

e.

Within the hot leg tubesheet, the repair criteria is described below:

1.

For tubes to which the W* criteria are applied, the length of non-degraded tube below BWT shall be greater than or equal to the W*

length plus NDE uncertainties and crack growth for the operating cycle.

2.

Axial cracks in tubes returned to service using W* shall have the upper crack tip below the BWT by at least the NDE measurement uncertainty and crack growth allowance, such that at the end of the subsequent operating cycle the entire crack remains below the BWT.

Technical Specification Inserts (continued)

Insert 5 (continued)

3.

Resolvable, single axial indications (multiple indications must return to the null point between individual cracks) within the flexible W* length can be left in service. Alternate RPC coils or an ultrasonic test (UT) inspection can be used to demonstrate return to null point between multiple axial indications or the absence of circumferential involvement between axial indications.

4.

Tubes with inclined axial indications less than 2.0 inches long (including the crack growth allowance) having inclination angles relative to the tube axis of < 45 degrees minus the NDE uncertainty, ANDECA, on the measurement of the crack angle can be left in service. Tubes with two or more parallel (overlapping elevation), inclined axial cracks shall be plugged. For application of the 2.0 inch limit, an inclined indication is an axial crack that is visually inclined on the RCP C-scan, such that an angular measurement is required, and the measured angle exceeds the measurement uncertainty of ANDECA.

5.

Circumferential, volumetric, and axial indications with inclination angles greater than (45 degrees - ANDECA) within the flexible W* length shall be plugged.

6.

Any type or combination of tube degradation below the flexible W*

length is acceptable.

3.

Axial Primary Water Stress Corrosion Cracking (PWSCC) Depth-Based Repair Criteria The axial PWSCC depth-based repair criteria are used for disposition of axial PWSCC indications, or portions thereof, which are located within the thickness of dented tube support plates which exhibit a maximum depth greater than or equal to 40 percent of the initial tube wall thickness. WCAP-15573, Revision 1, provides repair limits applicable to these intersections.

A tube which contains a tube support plate intersection with both an axial ODSCC indication and an axial PWSCC indication will be plugged.

Technical Specification Inserts (continued)

Insert 5 (continued)

d.

Provisions for SG tube inspections.

Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, d.3, d.4, d.5, and d.6 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. An assessment of degradation shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

1.

Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement.

2.

Inspect 100% of the tubes at sequential periods of 60 effective full power months.

The first sequential period shall be considered to begin after the first inservice inspection of the SGs. No SG shall operate for more than 24 effective full power months or one refueling outage (whichever is less) without being inspected.

3.

If crack indications are found in any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.

4.

Indications left in service as a result of application of the tube support plate voltage-based repair criteria in Specification 5.5.9.c.1 shall be inspected by bobbin coil probe every 24 effective full power months or one refueling outage, whichever is less.

Implementation of the steam generator tube support plate voltage-based repair criteria in Specification 5.5.9.c.1 requires a 100% bobbin coil inspection for hot-leg and cold-leg support plate intersections down to the lowest cold-leg tube support plate with known outside diameter stress corrosion cracking (ODSCC) indications.

The determination of the lowest cold-leg tube support plate intersection having ODSCC indications shall be based on the performance of at least a 20% random sampling of tubes inspected over their full length.

Technical Specification Inserts (continued)

Insert 5 (continued)

5.

Tubes identified as W* tubes having a previously identified indication within the flexible W* length shall be inspected using an RPC probe or equivalent for the full length of the W* region every 24 effective full power months or one refueling outage, whichever is less.

Implementation of the W* repair criteria in Specification 5.5.9.c.2 requires a 100 percent RPC probe or equivalent inspection of the hot leg tubesheet Flexible W*

Length, or 8 inches below the hot leg top of tubesheet, whichever is bounding.

6.

Inspection of dented tube support plate intersections will be performed in accordance with WCAP-1 5573, Revision 1, to implement axial PWSCC depth-based repair criteria in Specification 5.5.9.c.3. The extent of required inspection is:

a.

100 percent bobbin coil inspection of all tube support plate (TSP) intersections.

b.

Plus Point coil inspection of all bobbin coil indications at dented TSP intersections.

c.

Plus Point coil inspection of all prior PWSCC indications left in service.

d.

If bobbin coil is relied upon for detection of axial PWSCC in less than or equal to 2 volt dents, then on a SG basis perform Plus Point coil inspection of all TSP intersections having greater than 2 volt dents up to the highest TSP for which PWSCC has been detected in the prior two inspections or current inspection and 20% of greater than 2 volt dents at the next higher TSP. If a circumferential indication is detected in a dent of 'x" volts in the prior two inspections or current inspection, Plus Point inspections will be conducted on 100% of dents greater than "x - 0.3" volts up to the affected TSP elevation in the affected SG, plus 20% of dents greater than "x - 0.3" volts at the next higher TSP. "x" is defined as the lowest dent voltage where a circumferential crack was detected.

e.

If bobbin coil is not relied upon for detection of axial PWSCC in less than or equal to 2 volt dents, then on a SG basis perform Plus Point coil inspection of all dented TSP intersections (no lower dent voltage threshold) up to the highest TSP for which PWSCC has been detected in the prior two inspections or current inspection and 20% of all dents at the next higher TSP.

f.

For any 20% dent sample, a minimum of 50 dents at the TSP elevation shall be inspected. If the population of dents is less than 50 at the TSP elevation, then 100% of the dents at the TSP elevation shall be inspected.

e.

Provisions for monitoring operational primary to secondary LEAKAGE.

Technical Specification Inserts (continued)

Insert 6

a.

A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.9, Steam Generator (SG) Program. The report shall include:

1.

The scope of inspections performed on each SG,

2.

Active degradation mechanisms found,

3.

Nondestructive examination techniques utilized for each degradation mechanism,

4.

Location, orientation (if linear), and measured sizes (if available) of service induced indications,

5.

Number of tubes plugged during the inspection outage for each active degradation mechanism,

6.

Total number and percentage of tubes plugged to date, and

7.

The results of condition monitoring, including the results of tube pulls and in-situ testing.

Insert 7

e.

For implementation of tube support plate voltage-based repair criteria in Specification 5.5.9.c.1, a report shall be submitted within 90 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.9, Steam Generator (SG) Program. The report shall include the information described in Section 6b of Attachment 1 of NRC Generic Letter 95-05.

PG&E Letter DCL-06-130 Proposed Technical Specification Changes (retyped)

Remove Page Insert Page 1.1-4 3.4-27 3.4-28 5.0-10 5.0-11 5.0-11 a 5.0-12 5.0-13 5.0-13a 5.0-14 5.0-15 5.0-16 5.0-17 5.0-18 5.0-19 5.0-29 5.0-30 5.0-30a 5.0-30b 1.1-4 3.4-27 3.4-28 3.4-38 3.4-39 5.0-10 5.0-11 5.0-12 5.0-13 5.0-14 5.0-15 5.0-16 5.0-17 5.0-18 5.0-19 5.0-29 5.0-30 5.0-30a 5.0-30b

Definitions 1.1 1.1 Definitions LEAKAGE (continued)

MASTER RELAY TEST MODE

3.

Reactor Coolant System (RCS) LEAKAGE through a steam generator to the Secondary System (primary to secondary LEAKAGE).

b.

Unidentified LEAKAGE All LEAKAGE (except RCP seal water injection or leakoff) that is not identified LEAKAGE.

c.

Pressure Boundary LEAKAGE LEAKAGE (except primary to secondary LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.

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

A MODE shall correspond to any one inclusive combination of core reactivity condition, power level, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified in Table 1.1-1 with fuel in the reactor vessel.

A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related instrumentation. These tests are:

a.

Described in Chapter 14 of the FSAR;

b.

Authorized under the provisions of 10 CFR 50.59; or

c.

Otherwise approved by the Nuclear Regulatory Commission.

I OPERABLE-OPERABILITY PHYSICS TESTS (continued)

DIABLO CANYON - UNITS 1 & 2 1.1-4 Unit 1 - Amendment No. 435, Unit 2 - Amendment No. 43,5,

RCS Operational LEAKAGE 3.4.13 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.13 RCS Operational LEAKAGE LCO 3.4.13 RCS operational LEAKAGE shall be limited to:

a.

No pressure boundary LEAKAGE;

b.

1 gpm unidentified LEAKAGE;

c.

10 gpm identified LEAKAGE; and

d.

150 gallons per day primary to secondary LEAKAGE through any one steam generator (SG).

APPLICABILITY:

MODES 1, 2, 3*, and 4*.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A.

RCS operational LEAKAGE A.1 Reduce LEAKAGE to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months not within limits for reasons within limits.

other than pressure boundary LEAKAGE or primary to secondary LEAKAGE.

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 of Condition A not met.

OR Pressure boundary B.2 Be in MODE 5.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months LEAKAGE exists.

OR Primary to secondary LEAKAGE not within limit.

II I

For MODES 3 and 4, if steam generator water samples indicate less than the minimum detectable activity of 5.0 E-7 microcuries/ml for principal gamma emitters, the leakage requirement of specification 3.4.13.d. may be considered met.

DIABLO CANYON - UNITS 1 & 2 3.4-27 Unit 1 - Amendment No. 45 1-42, Unit 2 - Amendment No. 4-35 --42,

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.

Verify RCS operational LEAKAGE is within limits by 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 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 establishment of steady state operation.

Verify primary to secondary LEAKAGE is < 150 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months gallons per day through any one SG.

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

SG Tube Integrity 3.4.17 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.17 Steam Generator (SG) Tube Integrity LCO 3.4.17 SG tube integrity shall be maintained.

AND All SG tubes satisfying the tube repair criteria shall be plugged in accordance with the Steam Generator Program.

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

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

Separate Condition entry is allowed for each SG tube.

CONDITION REQUIRED ACTION COMPLETION TIME A.

One or more SG tubes A.1 Verify tube integrity of the 7 days satisfying the tube repair affected tube(s) is criteria and not plugged in maintained until the next accordance with the Steam refueling outage or SG Generator Program.

tube inspection.

AND A.2 Plug the affected tube(s) in Prior to entering accordance with the Steam MODE 4 following the Generator Program.

next refueling outage or SG tube inspection 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 of 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 OR SG tube integrity not maintained.

DIABLO CANYON - UNITS 1 & 2 3.4-38 Unit 1 - Amendment No.

Unit 2 - Amendment No.

SG Tube Integrity 3.4.17 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.17.1 Verify SG tube integrity in accordance with the Steam In accordance with Generator Program.

the Steam Generator Program SR 3.4.17.2 Verify that each inspected SG tube that satisfies the Prior to entering tube repair criteria is plugged in accordance with the MODE 4 following Steam Generator Program.

a SG tube inspection DIABLO CANYON - UNITS 1 & 2 3.4-39 Unit 1 - Amendment No.

Unit 2 - Amendment No.

Programs and Manuals 5.5 5.5 Programs and Manuals (continued) 5.5.9 Steam Generator (SG) Program A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained. In addition, the Steam Generator Program shall include the following provisions:

a.

Provisions for condition monitoring assessments.

Condition monitoring assessment means an evaluation of the "as found" condition of the tubing with respect to the performance criteria for structural integrity and accident induced leakage. The "as found" condition refers to the condition of the tubing during an SG inspection outage, as determined from the inservice inspection results or by other means, prior to the plugging of tubes.

Condition monitoring assessments shall be conducted during each outage during which the SG tubes are inspected or plugged to confirm that the performance criteria are being met.

b.

Performance criteria for SG tube integrity.

SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational LEAKAGE.

1.

Structural integrity performance criterion: All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down and all anticipated transients included in the design specification) and design basis accidents. This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and, except for flaws addressed through application of the alternate repair criteria discussed in Specification 5.5.9.c.1 and 5.5.9.c.3, a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials. Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse. In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.

When alternate repair criteria discussed in Specification 5.5.9.c.1 are applied to axially-oriented outside diameter stress corrosion cracking indications at tube support plate locations, the probability that one or more of these indications in a SG will burst under postulated main steam line break conditions shall be less than lx10-2.

(continued)

DIABLO CANYON - UNITS 1 & 2 5.0-10 Unit 1 - Amendment No.

Unit 2 - Amendment No.

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Progqram (continued)

When alternate repair criteria discussed in Specification 5.5.9.c.3 are applied to axially-oriented primary water stress corrosion cracking indications at tube support plate locations, the probability that one or more of these indications in a SG will burst under postulated main steam line break conditions shall be less than lx10-2.

2.

Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. Except during a steam generator tube rupture, leakage from all sources, excluding the leakage attributed to the degradation described in Specification 5.5.9.c.1, 5.5.9.c.2, and 5.5.9.c.3, is also not to exceed 1 gallon per minute per SG.

3.

The operational LEAKAGE performance criterion is specified in LCO 3.4.13, "RCS Operational LEAKAGE."

(continued)

DIABLO CANYON - UNITS 1 & 2 5.0-11 Unit 1 - Amendment No.

Unit 2 - Amendment No.

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Proqram (continued)

c.

Provisions for SG tube repair criteria.

Tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.

The following alternate tube repair criteria may be applied as an alternative to the 40% depth based criteria:

1.

Tube Support Plate Voltage-Based Repair Criteria The tube support plate voltage-based repair criteria are used for the disposition of an alloy 600 steam generator tube for continued service that is experiencing predominantly axially oriented outside diameter stress corrosion cracking confined within the thickness of the tube support plates. At tube support plate intersections, the repair criteria is described below:

a.

Steam generator tubes, whose degradation is attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with bobbin voltages less than or equal to 2.0 volts, will be allowed to remain in service.

b.

Steam generator tubes, whose degradation is attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than 2.0 volts, will be plugged, except as noted in 5.5.9.c.l.c below.

c.

Steam generator tubes, with indication of potential degradation attributed to outside diameter stress corrosion cracking within the bounds of the tube support plate with a bobbin voltage greater than 2.0 volts but less than or equal to the upper voltage repair limit (calculated according to the methodology in Generic Letter 95-05 as supplemented), may remain in service if a rotating pancake coil inspection or comparable inspection technique does not detect degradation. Steam generator tubes, with indications of outside diameter stress corrosion cracking degradation with a bobbin voltage greater than the upper voltage repair limit will be plugged.

d.

Certain intersections as identified in PG&E Letter DCL-03-174, dated December 19, 2003, will be excluded from application of the voltage-based repair criteria as it is determined that these intersections may collapse or deform following a postulated LOCA

+ SSE event.

(continued)

DIABLO CANYON - UNITS 1 & 2 5.0-12 Unit 1 - Amendment No.

Unit 2 - Amendment No.

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Program (continued)

e.

A tube which contains a tube support plate intersection with both an axial ODSCC indication and an axial PWSCC indication will be plugged.

f.

If an unscheduled mid-cycle inspection is performed, the following mid-cycle repair limits apply instead of the limits identified in 5.5.9.c.l.a, 5.5.9.c.1.b, and 5.5.9.c.1.c. The mid-cycle repair limits are determined from the following equations:

VMURL =

VSL (CL - At) 1.0 + NDE + Gr CL VMLRL :

VMURL (VuRL-VLRL) (CL -

At)

CL where:

VURL

= upper voltage repair limit VLRL

= lower voltage repair limit VMURL = mid-cycle upper voltage repair limit based on time into cycle VMLRL = mid-cycle lower voltage repair limit based on VMURL and time into cycle At

= length of time since last scheduled inspection during which VURL and VLRL were implemented CL

= cycle length (the time between two scheduled steam generator inspections)

VSL

= structural limit voltage Gr

= average growth rate per cycle length NDE

= 95% cumulative probability allowance for nondestructive examination uncertainty (i.e., a value of 20% has been approved by the NRC)

Implementation of these mid-cycle repair limits should follow the same approach as in TS 5.5.9.c.1.a, 5.5.9.c.1.b, and 5.5.9.c.1.c.

(continued)

DIABLO CANYON - UNITS 1 & 2 5.0-13 Unit 1 - Amendment No.

Unit 2 - Amendment No.

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Progqram (continued)

2.

W* Repair Criteria The W* repair criteria are used for disposition of an alloy 600 steam generator tube for continued service that is experiencing predominately axially oriented inside diameter stress corrosion cracking confined within the hot leg tubesheet, below the bottom of the WEXTEX transition (BWT). As used in this specification:

a.

Bottom of WEXTEX Transition (BWT) is the highest point of contact between the tube and tubesheet at, or below the top-of-tubesheet as determined by eddy current testing.

b.

W* Length is the distance in the hot leg tubesheet below the BWT that precludes tube pull out in the event of the complete circumferential separation of the tube below the W* length. The W* length is conservatively set at an undegraded hot leg tube length of 5.2 inches for Zone A tubes and 7.0 inches for Zone B tubes. Information provided in WCAP-14797-P, Revision 2, defines the boundaries of Zone A and Zone B.

III I

I

c.

Flexible W* Length is the W* length adjusted for any cracks found within the W* region. The Flexible W* Length is the total rotating pancake coil (RPC) inspected length as measured downward from the BWT, and includes NDE uncertainties and crack lengths within W* as adjusted for growth.

d.

W* Tube is a tube with degradation within or below the W* length that is left in service, and degraded within the limits specified in Specification 5.5.9.c.2.e.

(continued)

DIABLO CANYON - UNITS 1 & 2 5.0-14 Unit 1 - Amendment No.

Unit 2 - Amendment No.

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Progqram (continued)

e.

Within the hot leg tubesheet, the repair criteria is described below:

1.

For tubes to which the W* criteria are applied, the length of non-degraded tube below BWT shall be greater than or equal to the W* length plus NDE uncertainties and crack growth for the operating cycle.

2.

Axial cracks in tubes returned to service using W* shall have the upper crack tip below the BWT by at least the NDE measurement uncertainty and crack growth allowance, such that at the end of the subsequent operating cycle the entire crack remains below the BWT.

3.

Resolvable, single axial indications (multiple indications must return to the null point between individual cracks) within the flexible W* length can be left in service.

Alternate RPC coils or an ultrasonic test (UT) inspection can be used to demonstrate return to null point between multiple axial indications or the absence of circumferential involvement between axial indications.

4.

Tubes with inclined axial indications less than 2.0 inches long (including the crack growth allowance) having inclination angles relative to the tube axis of < 45 degrees minus the NDE uncertainty, ANDECA, on the measurement of the crack angle can be left in service. Tubes with two or more parallel (overlapping elevation), inclined axial cracks shall be plugged. For application of the 2.0 inch limit, an inclined indication is an axial crack that is visually inclined on the RCP C-scan, such that an angular measurement is required, and the measured angle exceeds the measurement uncertainty of ANDECA.

5.

Circumferential, volumetric, and axial indications with inclination angles greater than (45 degrees - ANDECA) within the flexible W* length shall be plugged.

6.

Any type or combination of tube degradation below the flexible W* length is acceptable.

(continued)

DIABLO CANYON - UNITS 1 & 2 5.0-15 Unit 1 - Amendment No.

Unit 2 - Amendment No.

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Progqram (continued)

3.

Axial Primary Water Stress Corrosion Cracking (PWSCC) Depth-Based Repair Criteria The axial PWSCC depth-based repair criteria are used for disposition of axial PWSCC indications, or portions thereof, which are located within the thickness of dented tube support plates which exhibit a maximum depth greater than or equal to 40 percent of the initial tube wall thickness.

WCAP-15573, Revision 1, provides repair limits applicable to these intersections.

A tube which contains a tube support plate intersection with both an axial ODSCC indication and an axial PWSCC indication will be plugged.

d.

Provisions for SG tube inspections.

Periodic SG tube inspections shall be performed. The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, d.3, d.4, d.5, and d.6 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. An assessment of degradation shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

1.

Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement.

2.

Inspect 100% of the tubes at sequential periods of 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. No SG shall operate for more than 24 effective full power months or one refueling outage (whichever is less) without being inspected.

3.

If crack indications are found in any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.

(continued)

DIABLO CANYON - UNITS 1 & 2 5.0-16 Unit 1 - Amendment No.

Unit 2 - Amendment No.

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Progqram (continued)

4.

Indications left in service as a result of application of the tube support plate voltage-based repair criteria in Specification 5.5.9.c.1 shall be inspected by bobbin coil probe every 24 effective full power months or one refueling outage, whichever is less.

Implementation of the steam generator tube support plate voltage-based repair criteria in Specification 5.5.9.c.1 requires a 100% bobbin coil I

inspection for hot-leg and cold-leg support plate intersections down to the lowest cold-leg tube support plate with known outside diameter stress corrosion cracking (ODSCC) indications. The determination of the lowest cold-leg tube support plate intersection having ODSCC indications shall be based on the performance of at least a 20% random sampling of tubes inspected over their full length.

5.

Tubes identified as W* tubes having a previously identified indication within the flexible W* length shall be inspected using an RPC probe or equivalent for the full length of the W* region every 24 effective full power months or one refueling outage, whichever is less.

Implementation of the W* repair criteria in Specification 5.5.9.c.2 requires a 100 percent RPC probe or equivalent inspection of the hot leg tubesheet Flexible W* Length, or 8 inches below the hot leg top of tubesheet, whichever is bounding.

6.

Inspection of dented tube support plate intersections will be performed in accordance with WCAP-1 5573, Revision 1, to implement axial PWSCC depth-based repair criteria in Specification 5.5.9.c.3. The extent of required inspection is:

a.

100 percent bobbin coil inspection of all tube support plate (TSP) intersections.

b.

Plus Point coil inspection of all bobbin coil indications at dented TSP intersections.

c.

Plus Point coil inspection of all prior PWSCC indications left in service.

(continued)

DIABLO CANYON - UNITS 1 & 2 5.0-17 Unit 1 - Amendment No.

Unit 2 - Amendment No.

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Progqram (continued)

d.

If bobbin coil is relied upon for detection of axial PWSCC in less than or equal to 2 volt dents, then on a SG basis perform Plus Point coil inspection of all TSP intersections having greater than 2 volt dents up to the highest TSP for which PWSCC has been detected in the prior two inspections or current inspection and 20% of greater than 2 volt dents at the next higher TSP. If a circumferential indication is detected in a dent of "x" volts in the prior two inspections or current inspection, Plus Point inspections will be conducted on 100% of dents greater than "x - 0.3" volts up to the affected TSP elevation in the affected SG, plus 20% of dents greater than "x - 0.3" volts at the next higher TSP. "x" is defined as the lowest dent voltage where a circumferential crack was detected.

e.

If bobbin coil is not relied upon for detection of axial PWSCC in less than or equal to 2 volt dents, then on a SG basis perform Plus Point coil inspection of all dented TSP intersections (no lower dent voltage threshold) up to the highest TSP for which PWSCC has been detected in the prior two inspections or current inspection and 20% of all dents at the next higher TSP.

f.

For any 20% dent sample, a minimum of 50 dents at the TSP elevation shall be inspected. If the population of dents is less than 50 at the TSP elevation, then 100% of the dents at the TSP

.elevation shall be inspected.

e.

Provisions for monitoring operational primary to secondary LEAKAGE.

DIABLO CANYON - UNITS 1 & 2 5.0-18 Unit 1 - Amendment No.

Unit 2 - Amendment No.

Programs and Manuals 5.5 THIS PAGE NOT USED DIABLO CANYON - UNITS 1 & 2 5.0-19 Unit 1 - Amendment No.

Unit 2 - Amendment No.

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.10 Steam Generator (SG) Tube Inspection Report

a.

A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.9, Steam Generator (SG) Program. The report shall include:

1.

The scope of inspections performed on each SG,

2.

Active degradation mechanisms found,

3.

Nondestructive examination techniques utilized for each degradation mechanism,

4.

Location, orientation (if linear), and measured sizes (if available) of service induced indications,

5.

Number of tubes plugged during the inspection outage for each active degradation mechanism,

6.

Total number and percentage of tubes plugged to date, and

7.

The results of condition monitoring, including the results of tube pulls and in-situ testing.

b.

For implementation of the tube support plate voltage-based repair criteria in Specification 5.5.9.c.1, notify the NRC prior to the initial entry into MODE 4 should any of the following arise:

1.

If ODSCC indications are identified that extend beyond the confines of the tube support plate.

(continued)

DIABLO CANYON - UNITS 1 & 2 5.0-29 Unit 1 - Amendment No. 4-5, 4-12, Unit 2 - Amendment No. 1-35, 1-84,

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.10 Steam Generator (SG) Tube Inspection Report (continued)

c.

The results of the inspection of W*. tubes shall be submitted within 90 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with Specification 5.5.9, Steam Generator (SG) Program. This report shall include:

1.

Identification of W* tube indications and indications that do not meet W*

requirements and were plugged, including the following information: the number of indications, the locations of the indications (relative to the BWT and TTS), the orientation (axial, circumferential, volumetric, inclined), the radial position of the tube within the tubesheet, the W* Zone of the tube, the severity of each indication (estimated depth), the side of the tube in which the indication initiated (inside or outside diameter), the W* inspection distance measured with respect to the BWT or TTS (whichever is lower), the length of axial indications, the angle of inclination of clearly skewed axial cracks (if applicable), verification that the upper crack tip of W* indications returned to service in the prior cycle remain below the BWT by at least the 95% confidence NDE uncertainty on locating the crack tip relative to the BWT, updated 95% growth rate for use in operational assessment, the cumulative number of indications detected in the tubesheet region as a function of elevation within the tubesheet, and the condition monitoring and operational assessment main steamline break leak rate for each indication and each SG in accordance with the leak rate methodology described in PG&E Letter DCL-05-018, dated March 11, 2005, as supplemented by PG&E Letter DCL-05-090, dated August 25, 2005.

2.

Assessment of whether the results were consistent with expectations and, if not consistent, a description of the proposed corrective action.

d.

The aggregate calculated steam line break leakage from application of all alternate repair criteria and non-alternate repair criteria shall be submitted within 90 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with Specification 5.5.9, Steam Generator (SG)

Program.

e.

For implementation of tube support plate voltage-based repair criteria in Specification 5.5.9.c.1, a report shall be submitted within 90 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with the Specification 5.5.9, Steam Generator (SG) Program. The report shall include the information described in Section 6b of Attachment 1 'of NRC Generic Letter 95-05.

(continued)

DIABLO CANYON - UNITS 1 & 2 5.0-30 Unit 1 - Amendment No. 1-3-5,1454,45,1-82, Unit 2 - Amendment No. 4-3541-64,41-52,41-84,

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.10 Steam Generator (SG) Tube Inspection Report (continued)

f.

For implementation of the repair criteria for axial PWSCC at dented TSPs, the results of the condition monitoring and operational assessments will be submitted within 120 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with Specification 5.5.9, Steam Generator (SG) Program. The report will include:

1.

Tabulations of indications found in the inspection, tubes plugged, and tubes left in service under the ARC.

2.

Growth rate distributions for indications found in the inspection and growth rate distributions used to establish the tube repair limits.

3.

Plus Point confirmation rates for bobbin detected indications when bobbin is relied upon for detection of axial PWSCC in less than or equal to 2 volt dents.

4.

For condition monitoring, an evaluation of any indications that satisfy burst margin requirements based on the Westinghouse burst pressure model, but do not satisfy burst margin requirements based on the combined ANL ligament tearing and throughwall burst pressure model.

(continued)

I DIABLO CANYON - UNITS 1 & 2 5.0-30a Unit 1 - Amendment No. 4-35454,1-52,1-82, Unit 2 - Amendment No. 435,445-,442,4-84,

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.10 Steam Generator (SG) Tube Inspection Report (continued)

5.

Performance evaluation of the operational assessment methodology for predicting flaw distributions as a function of flaw size.

6.

Evaluation results of number and size of previously reported versus new PWSCC indications found in the inspection, and the potential need to account for new indications in the operational assessment burst evaluation.

7.

Identification of mixed mode (axial PWSCC and circumferential) indications found in the inspection and an evaluation of the mixed mode indications for potential impact on the axial indication burst pressures or leakage.

8.

Any corrective actions found necessary in the event that condition monitoring requirements are not met.

g.

For implementation of the probability of prior cycle detection (POPCD) method, for the voltage-based repair criteria at tube support plate intersections, if the end-of-cycle conditional main steamline break burst probability, the projected main steamline break leak rate, or the number of indications are underpredicted by the previous cycle operational assessment, the following shall be submitted within 90 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with Specification 5.5.9, Steam Generator (SG) Program:

1.

The assessment of the probable causes for the underpredications, proposed corrective actions, and any recommended changes to probability of detection or growth methodology indicated by potential methods assessments.

2.

An assessment of the potential need to revise the alternate repair criteria analysis methods if: the burst probability is underpredicted by more than 0.001 (i.e., 10% of the reporting threshold) or an order of magnitude; or the leak rate is underpredicted by more than 0.5 gpm or an order of magnitude.

3.

An assessment of the potential need to increase the number of predicted low voltage indications at the beginning of cycle if the total number of as-found indications in any SG are underestimated by greater than 15%

or by greater than 150 indications.

DIABLO CANYON - UNITS 1 & 2 5.0-30b Unit 1 - Amendment No. -5,4-54,452,4-77,1-82, Unit 2 - Amendment No. 4-35,454,452,47-9,484, PG&E Letter DCL-06-130 Changes to Technical Specification Bases Pages

RCS Loops - MODES 1 and 2 B 3.4.4 BASES APPLICABLE SAFETY ANALYSES (continued)

RCS flow, RCP rotor seizure, and RCP shaft break events. For each of these events, it is demonstrated that all the applicable safety criteria are satisfied. For the remaining accident/safety analyses, operation of all four RCS loops during the transient up to the time of reactor trip is assured thereby ensuring that all the applicable acceptance criteria are satisfied. Those transients analyzed beyond the time of reactor trip were examined assuming that a loss of offsite power occurs which results in the RCPs coasting down.

The plant is designed to operate with all RCS loops in operation to maintain DNBR above the Safety Limit value during all normal operations and anticipated transients. By ensuring heat transfer in the nucleate boiling region, adequate heat transfer is provided between the fuel cladding and the reactor coolant.

RCS Loops - MODES 1 and 2 satisfy Criterion 2 of 10 CFR 50.36 (c)(2)(ii).

LCO The purpose of this LCO is to require an adequate forced flow rate for core heat removal. Flow is represented by the number of RCPs in operation for removal of heat by the SGs. To meet safety analysis acceptance criteria for DNB, four pumps are required at rated power.

An OPERABLE RCS loop csis

_ýRABLE RCP for heat transport and the associated SG, OPERA0_P.BLE i.n accordace W'it~h the Ste nwith a water level within the limits specified in SR 3.4.5.2, except for operational transients. A RCP is OPERABLE if it is capable of being powered and is able to provide forced flow if required.

APPLICABILITY In MODES 1 and 2, the reactor is critical and thus has the potential to produce maximum THERMAL POWER. Thus, to ensure that the assumptions of the accident analyses remain valid, all RCS loops are required to be OPERABLE and in operation in these MODES to prevent DNB and core damage.

The decay heat production rate is much lower than the full power heat rate. As such, the forced circulation flow and heat sink requirements are reduced for lower, noncritical MODES as indicated by the LCOs for MODES 3, 4, and 5.

(continued)

DIABLO CANYON - UNITS 1 & 2 Revision 3 17

RCS Loops - MODE 3 B 3.4.5 BASES (continued)

LCO The purpose of this LCO is to require that at least two RCS loops be OPERABLE. In MODE 3 with the Rod Control System capable of rod withdrawal, two RCS loops must be in operation. Two RCS loops are required to be in operation in MODE 3 with the Rod Control System capable of rod withdrawal due to the postulation of a power excursion because of an inadvertent control rod withdrawal. The required number of RCS loops in operation ensures that the Safety Limit criteria will be met for all of the postulated accidents.

When the Rod Control System is not capable of rod withdrawal, only one RCS loop in operation is necessary to ensure removal of decay heat from the core and homogenous boron concentration throughout the RCS. An additional RCS loop is required to be OPERABLE to ensure that redundancy for heat removal is maintained.

The Note permits all RCPs to be removed from operation for _< 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months per 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period. The purpose of the Note is to perform tests that are required to be performed without flow or pump noise. One of these tests is validation of the pump coastdown curve used as input to a number of accident analyses including a loss of flow accident. This test is generally performed in MODE 3 during the initial startup testing program, and as such should only be performed once. If, however, changes are made to the RCS that would cause a change to the flow characteristics of the RCS, the input values of the coastdown curve must be revalidated by conducting the test again.

Utilization of the Note is permitted provided the following conditions are met, along with any other conditions imposed by test procedures:

a.

No operations are permitted that would cause introduction of coolant into the RCS with boron concentration less than required to meet the SDM of LCO 3.1.1, thereby maintaining the margin to criticality. Boron reduction with coolant at boron concentrations less than required to assure the SDM is maintained is prohibited because a uniform concentration distribution throughout the RCS cannot be ensured when in natural circulation; and

b.

Core outlet temperature is maintained at least 1 0°F below saturation temperature, so that no vapor bubble may form and possibly cause a natural circulation flow obstruction.

An OPERABLE RCS loop consists of one OPERABLE RCP and one OPERABLE SG in aGccrd-nce with the Steam Generator Tube Surveillance Program, which has the minimum water level specified in SR 3.4.5.2. An RCP is OPERABLE if it is capable of being powered and is able to provide forced flow if required.

(continued)

DIABLO CANYON - UNITS 1 & 2 Revision 3 20

RCS Loops - MODES 4 B 3.4.6 BASES LCO (continued) temperature difference limits the available relative energy source and the pressurizer level condition provides an expansion volume to accommodate possible reactor coolant thermal swell. These conditions are intended to prevent a low temperature overpressure event due to a thermal transient when a RCP is started.

An OPERABLE RCS loop comprises an OPERABLE RCP and an OPERABLE SG in accordance with the Steam Generator Tube Surveillance Program, which has the minimum water level specified in SR 3.4.6.2.

Similarly for the RHR System, an OPERABLE RHR loop comprises an OPERABLE RHR pump capable of providing forced flow to an OPERABLE RHR heat exchanger. RCPs and RHR pumps are OPERABLE if they are capable of being powered and are able to provide forced flow if required. A RHR loop is in operation when the pump is operating and providing forced flow through the loop. Because a loop can be operating without being OPERABLE, the LCO requires at least one loop OPERABLE and in operation.

APPLICABILITY In MODE 4, this LCO ensures forced circulation of the reactor coolant to remove decay heat from the core and to provide proper boron mixing. One loop of either RCS or RHR provides sufficient circulation for these purposes. However, two loops consisting of any combination of RCS and RHR loops are required to be OPERABLE to meet single failure considerations.

Operation in other MODES is covered by:

LCO 3.4.4, "RCS Loops - MODES 1 and 2";

LCO 3.4.5, "RCS Loops - MODE 3";

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 and A.2 If one required RCS loop is inoperable and two RHR loops are inoperable, redundancy for heat removal is lost. Action must be initiated to restore a second RCS loop or RHR loop to OPERABLE status. The immediate Completion Time reflects the importance of maintaining the availability of two paths for heat removal.

(continued)

DIABLO CANYON - UNITS 1 & 2 Revision 3 26

RCS Loops - MODES 5, Loops Filled B 3.4.7 BASES LCO (continued)

The Note specifies that a RCP may be started if the pressurizer water level is less than 50%. This option of RCP start with pressurizer water level less than 50% supports plant operational flexibility. The open volume in the pressurizer provides space to sustain reactor coolant thermal swell without incurring a possible excessive pressure transient due to energy additions from the S/G secondary water. The purpose of conditions to allow initial RCP start when none is running is to prevent a possible low temperature RCS overpressure event due to a thermal transient when a RCP is started. The condition of SG/RCS temperature difference limits the available relative energy source and the pressurizer level condition provides an expansion volume to accommodate possible reactor coolant thermal swell. These conditions are intended to prevent a low temperature overpressure event due to a thermal transient when a RCP is started.

Note 4 provides for an orderly transition from MODE 5 to MODE 4 during a planned heatup by permitting removal of RHR loops from operation when at least one RCS loop is in operation. This Note provides for the transition to MODE 4 where an RCS loop is permitted to be in operation and replaces the RCS circulation function provided by the RHR loops.

RHR pumps are OPERABLE if they are capable of being powered and are able to provide flow if required: An OP-ERABL-E SG can perform as a heat sink via natural circulation when it has an adequate water level and is OPERABLE in accordance with the Steam Generater Tube Survieillance Programn.

APPLICABILITY "Loops filled" is a condition in which natural circulation can be used as a backup means of decay heat removal if forced circulation via RHR is lost. RCS loops are considered filled when the RCS is capable of being pressurized to at least 150 psig, no gas has been directly injected into the RCS, and the RCS has not been drained below 112 ft (Ref. 2).

In addition to these requirements, crediting heat removal via natural circulation requires at least two steam generators filled to >15% narrow range level and vented, or capable of being vented, to the atmosphere, and auxiliary feedwater available to add water to the relied-upon steam generators (Ref. 1). A loops filled condition is established at the completion of steam generator U-tube vacuum refill or after "bumping" RCPs.

In MODE 5 with RCS loops filled, this LCO requires forced circulation of the reactor coolant to remove decay heat from the core and to provide proper boron mixing. One loop of RHR provides sufficient circulation for these purposes. However, one additional RHR loop is required to be OPERABLE, or the secondary side water level of at least two SGs is required to be Ž_ 15%.

(continued)

DIABLO CANYON - UNITS 1 & 2 Revision 3 31

There are no changes to this page. Page included for information only.

RCS Operational LEAKAGE B 3.4.13 B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.13 RCS Operational LEAKAGE BASES BACKGROUND Components that contain or transport the coolant to or from the reactor core make up the RCS. Component joints are made by welding, bolting, rolling, or pressure loading, and valves isolate connecting systems from the RCS.

During plant life, the joint and valve interfaces can produce varying amounts of reactor coolant LEAKAGE, through either normal operational wear or mechanical deterioration. The purpose of the RCS Operational LEAKAGE LCO is to limit system operation in the presence of LEAKAGE from these sources to amounts that do not compromise safety. This LCO specifies the types and amounts of LEAKAGE.

10 CFR 50, Appendix A, GDC 30 (Ref. 1), requires means for detecting and, to the extent practical, identifying the source of reactor coolant LEAKAGE. Regulatory Guide 1.45 (Ref. 2) describes acceptable methods for selecting leakage detection systems.

The safety significance of RCS LEAKAGE varies widely depending on its source, rate, and duration. Therefore, detecting and monitoring reactor coolant LEAKAGE into the containment area is necessary.

Quickly separating the identified LEAKAGE from the unidentified LEAKAGE is necessary to provide quantitative information to the operators, allowing them to take corrective action should a leak occur that is detrimental to the safety of the facility and the public.

Possible leakage from a Control Rod Drive Mechanism (CRDM) canopy seal weld may be construed as either identified or unidentified LEAKAGE but not construed as pressure boundary LEAKAGE in accordance with Westinghouse letter PGE-88-622.

A limited amount of leakage inside containment is expected from auxiliary systems that cannot be made 100% leak tight. Leakage from these systems should be detected, located, and isolated from the containment atmosphere, if possible, to not interfere with RCS leakage detection.

This LCO deals with protection of the reactor coolant pressure boundary (RCPB) from degradation and the core from inadequate cooling, in addition to preventing the accident analyses radiation release assumptions from being exceeded. The consequences of violating this LCO include the possibility of a loss of coolant accident (LOCA).

(continued)

DIABLO CANYON - UNITS 1 & 2 72 Revision 3

RCS Operational LEAKAGE B 3.4.13 BASES (continued)

APPLICABLE Except for primary to secondary LEAKAGE, the safety analyses do not SAFETY address operational LEAKAGE. However, other operational LEAKAGE ANALYSES is related to the safety analyses for LOCA; the amount of leakage can affect the probability of such an event. The safety analysis for an eVent i

i resulting in steam diScharge to the atmosphere assumes al1 gpm:

Insert I primary to secondary LEAKAGE as the initial condition.

Primary to secondary LEAKAGE is a factor in the dose releases outside containment resulting from a steam line break (SLB) accident.

To a lesser extent, other accidents or transients involve secondary steam release to the atmosphere, such as a steam generator tube rupture (SGTR). The leakage contaminates the secondary fluid.

The SGTR (Ref. 3) is more limiting for radiological releases at the site boundary. The radiological dose analysis assumes loss of off-site fail power at the time of reactor trip with no subsequent conde o ing available. The steam generator (SG) PORV fo that has sustained the tube rupture is assumed to be open for 30 minutes, at operator which time+aRCS pressure s belew the left sett of the PORV. The closes the consequences resulting from the SGTR accident are within the block valve to limits defined in 10 CFR 100 (Ref. 6).

he safety analysis for RCS main loop piping for GDC-4 (Ref. 1)

Insert 2 assumes 1 gpm unidentified leaka~ge and monitoring per RG 1.45 (Ref. 2) are maintained (Ref. 4 and 5).

The RCS operational LEAKAGE satisfies Criterion 2 of 10 CFR 50.36 (c)(2)(ii).

LCO RCS operational LEAKAGE shall be limited to:

a.

Pressure Boundary LEAKAGE No pressure boundary LEAKAGE is allowed, being indicative of material deterioration. LEAKAGE of this type is unacceptable as the leak itself could cause further deterioration, resulting in higher LEAKAGE. Violation of this LCO could result in continued degradation of the RCPB. LEAKAGE past seals, gaskets, or the CRDM canopy seal welds is not pressure boundary LEAKAGE.

Pressure boundary leakage is defined as "non-isolable" leakage.

A "non-isolable" RCS leak is one that is not capable of being isolated from the RCS using installed automatic or accessible manual valves.

(continued)

DIABLO CANYON - UNITS 1 & 2 73 Revision 3

RCS Operational LEAKAGE B 3.4.13 BASES LCO

b.

Unidentified LEAKAGE (continued)

One gallon per minute (gpm) of unidentified LEAKAGE is allowed as a reasonable minimum detectable amount that the containment air monitoring and containment sump level monitoring equipment can detect within a reasonable time period. Violation of this LCO could result in continued degradation of the RCPB, if the LEAKAGE is from the pressure boundary.

c.

Identified LEAKAGE Up to 10 gpm of identified LEAKAGE is considered allowable because LEAKAGE is from known sources that do not interfere with detection of unidentified LEAKAGE and is well within the capability of the RCS Makeup System. Identified LEAKAGE includes LEAKAGE to the containment from specifically known and located sources, but does not include pressure boundary LEAKAGE or controlled reactor coolant pump (RCP) seal leakoff (a normal function not considered LEAKAGE). Identified LEAKAGE does not include LEAKAGE from portions of the Chemical and Volume Control System outside of containment that can be isolated from the RCS. LEAKAGE of this nature may be reviewed for. possible. impact on the Primary Coolant Sources Outside Containment program. Violation of this LCO could result in continued degradation of a component or system.

d.

Primary to Secondary LEAKAGE throuqh All Steam Generators Total primary to secondary LEAKAGE amounting to 1 gpm through all SGs produces acceptable offsite doses in the SLB accident analysis. Violation of this LCO could exceed the offsite dose limits

\\

for this accident. Primary to secondary LEAKAGE must be

\\included in the total allowable limit for identified LEAKAGE.

d Primary to Secondary LEAKAGE throucqh Any One SG The 150 gallons per day limit on one SG is based on'the assumption that a single crack leaking this amount would not propagate to a SGTR under the stress conditions of a LOCA or a main steam line rupture. If leaked through many cracks, the cracks Insert 3 F

are very small, and the above assumption is conservative.

The primary-to-secondary operational leakage limit of 150 gallons per day per steam generator is more restrictive than the standard operating leakage limits and is intended to provide an additional margin to accommodate a crack which might grow at a greater than expected rate or unexpectedly extent outside the thickness of the tube support plate. Hence, the reduced leakage limit, when (continued)

DIABLO CANYON - UNITS 1 & 2 74 Revision 3

RCS Operational LEAKAGE B 3.4.13 BASES C*

LCO

e. Primary to Secondary LEAKAGE through Any One SG (continued) combined with an effective leak rate monitoring program, provides L(

additional assurance that should a significant leak be experienced in service, it will be detected, and the plant shut down in a timelY manner.

Calculations for primary-to-secondary leakage are performed using approximate Standard Reference State of 250C. When determining primary-to-secondary leakage of 150 gallons per day, indeterminant inaccuracies associated with determination of leakage are not considered.

For MODES 3 and 4, the primary system radioactivity level (source term) may be very low, making it difficult to measure primary-to-d secondary leakage of 150 gallons per day. Therefore, if steam generator water samples indicate less than the minimum detectable activity of 5.0*E-7 microcuries/ml for each principal gamma emitter, the leakage requirement of Specification 3.4.1 e

may be considered met.

V APPLICABILITY In MODES 1, 2, 3, and 4, the potential for RCPB LEAKAGE is greatest when the RCS is pressurized.

In MODES 5 and 6, LEAKAGE limits are not required because the reactor coolant pressure is far lower, resulting in lower stresses and reduced potentials for LEAKAGE.

LCO 3.4.14, "RCS Pressure Isolation Valve (PIV) Leakage," measures leakage through each individual PIV and can impact this LCO. Of the two PIVs in series in each isolated line, leakage measured through one PIV does not result in RCS LEAKAGE when the other is leak tight. If both valves leak and result in a loss of mass from the RCS, the loss must be included in the allowable identified LEAKAGE.

A note has been added to the APPLICABILITY. For MODES 3 and 4, the primary system radioactivity level (source term) may be very low, making it difficult to measure primary-to-secondary leakage of 150

, -gallons per day. Therefore, if steam generator water samples indicate d

l :ess than the minimum detectable activity of 5.0*E 7 microcuries/ml for ri i al gamma emitter, the leakage requirement of Specification 3.4.13.

may be considered met.

ACTIONS A.1 or Unidentified LEAKAGInidentified LEAKAG,lEo pri,,m3ary to seconda,5' LEAKAGE in excess of the LCO limits must be reduced to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . This Completion Time allows time to verify leakage (continued)

DIABLO CANYON - UNITS 1 & 2 75 Revision 3

RCS Operational LEAKAGE B 3.4.13 or primary to secondary LEAKAGE is not within limit, BASES ACTIONS A.1 (continued) rates and either identify unidentified LEAKA within limits before the reactor must be shu, necessary to prevent further deterioration o B.1 and B.2 GE or reduce LEAKAGE to down. This action is the RCPB.

or If any pressure boundary LEAKAGE exists or if unidentified-LEAKAGE, identified LEAKAGE, or primary to secondary LEAKAGE cannot be reduced to within limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , the reactor must be brought to lower pressure conditions to reduce the severity of the LEAKAGE and its potential consequences. It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE. 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 . This action reduces the LEAKAGE and also reduces the factors that tend to degrade the pressure boundary.

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 MODE 5, the pressure stresses acting on the RCPB are much lower, and'further deterioration is much less likely.

SURVEILLANCE SR 3.4.13.1 REQUIREMENTS Verifying RCS LEAKAGE to be within the LCO limits ensures the integrity of the RCPB is maintained. Pressure boundary LEAKAGE would at first appear as unidentified LEAKAGE and can only be positively identified by inspection. It should be noted that LEAKAGE past seals and gaskets is not pressure boundary LEAKAGE.

Unidentified LEAKAGE and identified LEAKAGE are determined by 2-performance of an RCS water inventory balance. (Primary to secondary "

r-'iEAKAGE is also measured by performance of an RCS water inventory L,

balance in conjunction with effluent monitoring within the secondary

//

The RCS water inventory balance must be met with the reactor at steady state operating conditions (stable temperature, power le states pressurizer level, makeup and letdown, and RCP 5e ion and return flows TheFefOe-a Note is added alow-ing hat this SR is not The requir o be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishing steady state Surveillance is

ration. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowance provides sufficient time to collect and modified by process all necessary data after stable plant conditions are established.

two Notes.

ISteady state operation is required to perform a proper inventory balance since calculations during maneuvering are not useful. For RCS (continued)

DIABLO CANYON - UNITS 1 & 2 76 Revision 3

RCS Operational LEAKAGE B 3.4.13 BASES SURVEILLANCE REQUIREMENTS SR 3.4.13.1 (continued) operational LEAKAGE determination by water inventory balance, steady state is defined as stable RCS pressure, temperature (Tavg changes less than 50F per hour) power level, pressurizer and makeup tank levels, makeup and letdown (balanced with no diversion to LHUTS), and RCP seal injection and return flows.

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 Insert 4 pressure boundary LEAKAGE. These leakage detection systems are specified in LCO 3.4.15, "RCS Leakage Detection Instrumentation."

The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Frequency is a reasonable interval to trend LEAKAGE and recognizes the importance of early leakage detection in the prevention 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 been 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 and every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months thereafter during steady state operation.,

SSR 3.4.13.2 This SR provides the means necessary to determine SG OPERABILITY in an operational MODE. The requirement to demonstrate SG tube integrity in accordance with the Steam Generator Tube Surveillance Program emphasizes the importance of SG tube integrity, even though this Surveillance cannot be performed at normal operating conditions.

This surveillance does not tie directly to any of the leakage criteria in the LCO or of the conditions; therefore failure to meet this surveillance is Ie5rconsidered failure to meet the integrity goals of the LCO and LCO 3.0.3 S R 3.4. 1.3-L2_J This SR provides a irleinement to determine primary-to-secondary leakage once every 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months while operating in MODES 1, 2, 3, or 4.

During normal operation this will be done using a correlation of Iner **

radioactivity at the steam jet air ejectors. During periods ofsinfct

/

source term and mass flow rate changes or when the primary system radioactivity levels are low engineering judgment may be used to aid in determining leakage rates. The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months frequency is adequate to allow early detection of a significant primary-to-secondary leakage and allow the plant to shutdown in a timely manner reducing the risk of a tube r u p t u r e.

I - ---

-(continued)

-d DIABLO CANYON - UNITS 1 & 2 77 Revision 3

RCS Operational LEAKAGE B 3.4.13 BASES (continued)

REFERENCES

1.

10 CFR 50, Appendix A, GDC 4 and 30.

2.

Regulatory Guide 1.45, May 1973.

3.

FSAR, Section 15.

4.

FSAR, Section 3.

5.

NUREG-1 061, Volume 3, November, 1984.

6.

10 CFR 100.

FInsert 6 DIABLO CANYON - UNITS 1 & 2 78 Revision 3

RCS Specific Activity B 3.4.16 BASES SURVEILLANCE SR 3.4.16.2 REQUIREMENTS This Surveillance is modified by a Note. The Note modifies the (continued) surveillance to allow entry into and operation in MODE 3 > 500'F and MODE 2 prior to performing this Surveillance Requirement.

This Surveillance is performed to ensure iodine remains within limit during normal operation and following fast power changes when fuel failure is more apt to occur. The 14 day Frequency is adequate to trend changes in the iodine activity level, considering gross activity is monitored every 7 days. The Frequency, between 2 and 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months after a power change > 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 fuel failure; samples at other times would provide less indicative results.

SR 3.4.16.3 A radiochemical analysis for Edetermination is required every 184 days (6 months) with the plant operating in MODE 1 equilibrium (as defined in SR 3.4.16.3 NOTE) conditions. The E determination directly relates to the LCO and is required to verify plant operation within the specified gross activity LCO limit. The analysis for E is the qualitative measurement of the specific activity for each radionuclide, except for radionuclides with half-lives less than 10 minutes and all radioiodines which are identified in the reactor coolant. The specific activity for these individual radionuclides shall be used in the determination of E for the reactor coolant sample. Determination of the contributors to E shall be based upon those energy peaks identifiable with a 95%

confidence level. The Frequency of 184 days recognizes Edoes not change rapidly.

This SR has been modified by a Note that indicates sampling for E determination is required to be performed within 31 days after a minimum of 2 effective full power days and 20 days of MODE 1 operation have elapsed since the reactor was last subcritical for at least 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . This ensures that the radioactive materials are at equilibrium so the analysis for E is representative and not skewed by a crud burst or other similar abnormal event.

REFERENCES

1.

1 OCFR1 00.11, 1973.

2.

FSAR, Sections 15.4.3 and 15.5.20.

Insert 7 (for new B 3.4.17)

DIABLO CANYON - UNITS 1 & 2 97 Revision 3

Technical Specification Bases Inserts Insert 1 Safety analyses for design basis events that model primary to secondary LEAKAGE result in steam discharge to the atmosphere. The safety analysis for the SLB event assumes that primary to secondary LEAKAGE is 10.5 gpm (room temperature conditions) from the faulted SG or increases to 10.5 gpm as a result of accident induced conditions, and 0.1 gpm (room temperature conditions) from each intact SG. The safety analyses for events resulting in steam discharge to the atmosphere, other than SGTR and SLB, assume that primary to secondary LEAKAGE from all SGs is 0.75 gpm (hot conditions). The LCO requirement to limit primary to secondary LEAKAGE through any one SG to less than or equal to 150 gallons per day is significantly less than the conditions assumed in the SLB safety analysis for the faulted SG.

Insert 2 The SLB is more limiting for site radiation releases for events other than SGTR. The safety analysis for the SLB accident assumes 10.5 gpm primary to secondary LEAKAGE is through the faulted SG. The dose consequences resulting from the SLB accident are well within the limits defined in 10 CFR 100 or the staff approved licensing basis (i.e., small fraction of these limits).

Insert 3 The limit of 150 gallons per day per SG is based on the operational LEAKAGE performance criterion in NEI 97-06, Steam Generator Program Guidelines (Ref. 7). The Steam Generator Program operational LEAKAGE performance criterion in NEI 97-06 states, "The RCS operational primary to secondary leakage through any one SG shall be limited to 150 gallons per day." The limit is based on operating experience with SG tube degradation mechanisms that result in tube leakage. The operational leakage rate criterion in conjunction with the implementation of the Steam Generator Program is an effective measure for minimizing the frequency of steam generator tube ruptures.

Insert 4 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.

Insert 5 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

Technical Specification Bases Inserts (continued) 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.

The Surveillance Frequency of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is a reasonable interval to trend primary to secondary LEAKAGE and recognizes the importance of early leakage detection in the prevention of accidents. The primary to secondary LEAKAGE is determined using continuous process radiation monitors or radiochemical grab sampling in accordance with the EPRI guidelines (Ref.

8).

Insert 6

7. NEI 97-06, "Steam Generator Program Guidelines."

8. EPRI, "Pressurized Water Reactor Primary-to-Secondary Leak Guidelines."

Insert 7 See next page for new TS 3.4.17 Bases.

B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.17 Steam Generator (SG) Tube Integrity BASES BACKGROUND Steam generator (SG) tubes are small diameter, thin walled tubes that carry primary coolant through the primary to secondary heat exchangers. The SG tubes have a number of important safety functions. Steam generator tubes are an integral part of the reactor coolant pressure boundary (RCPB) and, as such, are relied on to maintain the primary system's pressure and inventory. The SG tubes isolate the radioactive fission products in the primary coolant from the secondary system. In addition, as part of the RCPB, the SG tubes are unique in that they act as the heat transfer surface between the primary and secondary systems to remove heat from the primary system. This Specification addresses only the RCPB integrity function of the SG.

The SG heat removal function is addressed by LCO 3.4.4, "RCS Loops

- MODES 1 and 2," LCO 3.4.5, "RCS Loops - MODE 3," LCO 3.4.6, "RCS Loops - MODE 4," and LCO 3.4.7, "RCS Loops - MODE 5, Loops Filled."

SG tube integrity means that the tubes are capable of performing their intended RCPB safety function consistent with the licensing basis, including applicable regulatory requirements.

Steam generator tubing is subject to a variety of degradation mechanisms. Steam generator tubes may experience tube degradation related to corrosion phenomena, such as wastage, pitting, intergranular attack, and stress corrosion cracking, along with other mechanically induced phenomena such as denting and wear. These degradation mechanisms can impair tube integrity if they are not managed effectively. The SG performance criteria are used to manage SG tube degradation.

Specification 5.5.9, "Steam Generator (SG) Program," requires that a program be established and implemented to ensure that SG tube integrity is maintained. Pursuant to Specification 5.5.9, tube integrity is maintained when the SG performance criteria are met. There are three SG performance criteria: structural integrity, accident induced leakage, and operational LEAKAGE. The SG performance criteria are described in Specification 5.5.9. Meeting the SG performance criteria provides reasonable assurance of maintaining tube integrity at normal and accident conditions.

The processes used to meet the SG performance criteria are defined by the Steam Generator Program Guidelines (Ref. 1).

(continued)

Technical Specification Bases Inserts (continued)

BASES APPLICABLE SAFETY ANALYSES The steam generator tube rupture (SGTR) accident is the limiting design basis event for SG tubes and avoiding an SGTR is the basis for this Specification. The analysis of a SGTR event assumes a total primary to secondary LEAKAGE rate of 1 gpm from the intact SGs plus the leakage rate associated with a double-ended rupture of a single tube. The SGTR radiological dose analysis assumes loss of off-site power at the time of reactor trip with no subsequent condenser cooling available. The SG PORV for the SG that has sustained the tube rupture is assumed to fail open for 30 minutes, at which time the operator closes the block valve to the PORV. The SGTR radiological dose analysis assumes the contaminated secondary fluid is released briefly to the atmosphere from all the PORVs following reactor trip, is released from the ruptured SG PORV for 30 minutes, is released from the intact SG PORVs during the cooldown, and is released from all PORVs following cooldown until termination of the event.

The analysis for design basis accidents and transients other than a SGTR assume the SG tubes retain their structural integrity (i.e., they are assumed not to rupture.) For the SLB event, the primary to secondary LEAKAGE is 10.5 gpm from the faulted SG or is assumed to increase to 10.5 gpm as a result of accident induced conditions, and 0.1 gpm from each intact SG. For other events, the steam discharge to the atmosphere is based on the total primary to secondary LEAKAGE from all SGs of 0.75 gpm. For accidents that do not involve fuel damage, the primary coolant activity level of DOSE EQUIVALENT I-131 is assumed to be equal to the LCO 3.4.16, "RCS Specific Activity,"

limits. For accidents that assume fuel damage, the primary coolant activity is a function of the amount of activity released from the damaged fuel. The dose consequences of these events are within the limits of GDC 19 (Ref. 2), 10 CFR 100 (Ref. 3) or the NRC approved licensing basis (e.g., a small fraction of these limits).

Steam generator tube integrity satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

The LCO requires that SG tube integrity be maintained. The LCO also requires that all SG tubes that satisfy the repair criteria be plugged in accordance with the Steam Generator Program.

During an SG inspection, any inspected tube that satisfies the Steam Generator Program repair criteria is removed from service by plugging.

If a tube was determined to satisfy the repair criteria but was not plugged, the tube may still have tube integrity.

(continued)

LCO

Technical Specification Bases Inserts (continued)

BASES LCO In the context of this Specification, a SG tube is defined as the entire (continued) length of the tube, including the tube wall, between the tube-to-tubesheet weld at the tube inlet and the tube-to-tubesheet weld at the tube outlet. The tube-to-tubesheet weld is not considered part of the tube.

A SG tube has tube integrity when it satisfies the SG performance criteria. The SG performance criteria are defined in Specification 5.5.9, "Steam Generator Program," and describe acceptable SG tube performance. The Steam Generator Program also provides the evaluation process for determining conformance with the SG performance criteria.

There are three SG performance criteria: structural integrity, accident induced leakage, and operational LEAKAGE. Failure to meet any one of these criteria is considered failure to meet the LCO.

The structural integrity performance criterion provides a margin of safety against tube burst or collapse under normal and accident conditions, and ensures structural integrity of the SG tubes under all anticipated transients included in the design specification. Tube burst is defined as, "The gross structural failure of the tube wall. The condition typically corresponds to an unstable opening displacement (e.g., opening area increased in response to constant pressure) accompanied by ductile (plastic) tearing of the tube material at the ends of the degradation." Tube collapse is defined as, "For the load displacement curve for a given structure, collapse occurs at the top of the load versus displacement curve where the slope of the curve becomes zero." The structural integrity performance criterion provides guidance on assessing loads that have a significant effect on burst or collapse. In that context, the term "significant" is defined as "An accident loading condition other than differential pressure is considered significant when the addition of such loads in the assessment of the structural integrity performance criterion could cause a lower structural limit or limiting burst/collapse condition to be established." For tube integrity evaluations, except for circumferential degradation, axial thermal loads are classified as secondary loads.

For circumferential degradation, the classification of axial thermal loads as primary or secondary loads will be evaluated on a case-by-case basis. The division between primary and secondary classifications will be based on detailed analysis and/or testing.

(continued)

Technical Specification Bases Inserts (continued)

BASES LCO Structural integrity requires that the primary membrane stress intensity (continued) in a tube not exceed the yield strength for all ASME Code,Section III, Service Level A (normal operating conditions) and Service Level B (upset or abnormal conditions) transients included in the design specification. This includes safety factors and applicable design basis loads based on ASME Code,Section III, Subsection NB (Ref. 4) and Draft Regulatory Guide 1.121 (Ref. 5).

The accident induced leakage performance criterion ensures (a) that the primary to secondary LEAKAGE caused by a design basis accident, other than a SGTR, is within the accident analysis assumptions, and (b) that the primary to secondary LEAKAGE will not exceed 1 gpm per SG (except for specific types of degradation at specific locations where the NRC has approved greater accident induced leakage) to ensure that the potential for induced leakage during severe accidents will be maintained at a level that will not increase risk. The accident analysis for the SLB event assumes that accident induced leakage does not exceed 10.5 gpm in the faulted SG and 0.1 gpm in each intact SG. For the faulted SG in the SLB event, 10.5 gpm is the accident induced leakage limit for specific sources (as approved by the NRC), and 1 gpm is the accident induced leakage limit for all other sources (i.e., those not specifically exempted by the NRC). The accident analyses for events otherthan SGTR and SLB assume that accident induced leakage does not exceed 0.75 gpm total, equally partitioned among the four SGs (approximately 0.19 gpm from each SG). The accident induced leakage rate includes any primary to secondary LEAKAGE existing prior to the accident in addition to primary to secondary LEAKAGE induced during the accident.

The operational LEAKAGE performance criterion provides an observable indication of SG tube conditions during plant operation. The limit on operational LEAKAGE is contained in LCO 3.4.13, "RCS Operational LEAKAGE," and limits primary to secondary LEAKAGE through any one SG to 150 gallons per day. This limit is based on the assumption that a single crack leaking this amount would not propagate to a SGTR under the stress conditions of a LOCA or a main steam line break. If this amount of LEAKAGE is due to more than one crack, the cracks are very small, and the above assumption is conservative.

APPLICABILITY Steam generator tube integrity is challenged when the pressure differential across the tubes is large. Large differential pressures across SG tubes can only be experienced in MODE 1, 2, 3, or 4.

RCS conditions are far less challenging in MODES 5 and 6 than during MODES 1, 2, 3, and 4. In MODES 5 and 6, primary to secondary differential pressure is low, resulting in lower stresses and reduced potential for LEAKAGE.

(continued)

Technical Specification Bases Inserts (continued)

BASES ACTIONS The ACTIONS are modified by a Note clarifying that the Conditions may be entered independently for each SG tube. This is acceptable because the Required Actions provide appropriate compensatory actions for each affected SG tube. Complying with the Required Actions may allow for continued operation, and subsequent affected SG tubes are governed by subsequent Condition entry and application of associated Required Actions.

A.1 and A.2 Condition A applies if it is discovered that one or more SG tubes examined in an inservice inspection satisfy the tube repair criteria but were not plugged in accordance with the Steam Generator Program as required by SR 3.4.17.2. An evaluation of SG tube integrity of the affected tube(s) must be made. Steam generator tube integrity is based on meeting the SG performance criteria described in the Steam Generator Program. The SG repair criteria define limits on SG tube degradation that allow for flaw growth between inspections while still providing assurance that the SG performance criteria will continue to be met. In order to determine if a SG tube that should have been plugged has tube integrity, an evaluation must be completed that demonstrates that the SG performance criteria will continue to be met until the next refueling outage or SG tube inspection. The tube integrity determination is based on the estimated condition of the tube at the time the situation is discovered and the estimated growth of the degradation prior to the next SG tube inspection. If it is determined that tube integrity is not being maintained, Condition B applies.

A Completion Time of 7 days is sufficient to complete the evaluation while minimizing the risk of plant operation with a SG tube that may not have tube integrity.

If the evaluation determines that the affected tube(s) have tube integrity, Required Action A.2 allows plant operation to continue until the next refueling outage or SG inspection provided the inspection interval continues to be supported by an operational assessment that reflects the affected tubes. However, the affected tube(s) must be plugged prior to entering MODE 4 following the next refueling outage or SG inspection. This Completion Time is acceptable since operation until the next inspection is supported by the operational assessment.

(continued)

Technical Specification Bases Inserts (continued)

BASES ACTIONS B.1 and B.2 (continued)

If the Required Actions and associated Completion Times of Condition A are not met or if SG tube integrity is not being maintained, 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 desired plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.4.17.1 REQUIREMENTS During shutdown periods the SGs are inspected as required by this SR and the Steam Generator Program. NEI 97-06, Steam Generator Program Guidelines (Ref. 1), and its referenced EPRI Guidelines, establish the content of the Steam Generator Program. Use of the Steam Generator Program ensures that the inspection is appropriate and consistent with accepted industry practices.

During SG inspections a condition monitoring assessment of the SG tubes is performed. The condition monitoring assessment determines the "as found" condition of the SG tubes. The purpose of the condition monitoring assessment is to ensure that the SG performance criteria have been met for the previous operating period.

The Steam Generator Program determines the scope of the inspection and the methods used to determine whether the tubes contain flaws satisfying the tube repair criteria. Inspection scope (i.e., which tubes or areas of tubing within the SG are to be inspected) is a function of existing and potential degradation locations. The Steam Generator Program also specifies the inspection methods to be used to find potential degradation. Inspection methods are a function of degradation morphology, nondestructive examination (NDE) technique capabilities, and inspection locations.

The Steam Generator Program defines the Frequency of SR 3.4.17.1.

The Frequency is determined by the operational assessment and other limits in the SG examination guidelines (Ref. 6). The Steam Generator Program uses information on existing degradations and growth rates to determine an inspection Frequency that provides reasonable assurance that the tubing will meet the SG performance criteria at the next scheduled inspection.

(continued)

Technical Specification Bases Inserts (continued)

BASES SURVEILLANCE In addition, Specification 5.5.9 contains prescriptive requirements REQUIREMENTS concerning inspection intervals to provide added assurance that the SG (continued) performance criteria will be met between scheduled inspections.

SR 3.4.17.2 During an SG inspection, any inspected tube that satisfies the Steam Generator Program repair criteria is removed from service by plugging.

The tube repair criteria delineated in Specification 5.5.9 are intended to ensure that tubes accepted for continued service satisfy the SG performance criteria with allowance for error in the flaw size measurement and for future flaw growth. In addition, the tube repair criteria, in conjunction with other elements of the Steam Generator Program, ensure that the SG performance criteria will continue to be met until the next inspection of the subject tube(s). Reference 1 provides guidance for performing operational assessments to verify that the tubes remaining in service will continue to meet the SG performance criteria.

The Frequency of prior to entering MODE 4 following a SG inspection ensures that the Surveillance has been completed and all tubes meeting the repair criteria are plugged prior to subjecting the SG tubes to significant primary, to secondary pressure differential.

REFERENCES

1.

NEI 97-06, "Steam Generator Program Guidelines."

2.

10 CFR 50 Appendix A, GDC 19.

3.

10 CFR 100.

4.

ASME Boiler and Pressure Vessel Code,Section III, Subsection NB.

5.

Draft Regulatory Guide 1.121, "Basis for Plugging Degraded Steam Generator Tubes," August 1976.

6.

EPRI, "Pressurized Water Reactor Steam Generator Examination Guidelines."

v t e Letter Sequence Response to RAI
TAC:MD2132, Steam Generator Tube Integrity (Approved, Closed)
Initiation Request Acceptance... Administration Questions 26 October 2006 Answers 22 November 2006 Results Approval Other: ML070440494
MONTHYEAR2006-11-22 [Table View]

DCL-06-130, Response to NRC Request for Additional Information Regarding License Amendment Request 06-04, Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity (TSTF-449)

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DCL-06-130, Response to NRC Request for Additional Information Regarding License Amendment Request 06-04, Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity (TSTF-449)

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Dear Commissioners and Staff:

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