View source for DCL-06-061, License Amendment Request 06-04, Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity (TSTF-449)

{{Adams
| number = ML061570107
| issue date = 05/30/2006
| title = License Amendment Request 06-04, Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity (TSTF-449)
| author name = Becker J
| author affiliation = Pacific Gas &amp; Electric Co
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000275, 05000323
| license number = DPR-080, DPR-082
| contact person = 
| case reference number = DCL-06-061, LAR 06-04, OL-DPR-80, OL-DPR-82, TSTF-449
| document type = Letter type:DCL, License-Application for Facility Operating License (Amend/Renewal) DKT 50, Technical Specification, Amendment
| page count = 82
| project = 
| stage = Request
}}

=Text=
{{#Wiki_filter:Pacific Gas and ElectricCompany-James R. Becker               Diablo Canyon Power Plant Vice President                P. 0. Box 56 Diablo Canyon Operations and  Avila Beach, CA 93424 Station Director 805.545.3462 May 30, 2006                                                                         Fax: 805.545.4234 PG&E Letter DCL-06-061 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Docket No. 50-275, OL-DPR-80 Docket No. 50-323, OL-DPR-82 Diablo Canyon Units 1 and 2 License Amendment Request 06-04 Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity (TSTF-449)

==Dear Commissioners and Staff:==

In accordance with 10 CFR 50.90, enclosed is an application for amendment to Facility Operating License Nos. DPR-80 and DPR-82 for Units 1 and 2 of the Diablo Canyon Power Plant, respectively.
The license amendment request (LAR) proposes to revise Technical Specification (TS) requirements related to steam generator tube integrity. The change is consistent with NRC-approved Revision 4 to Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-449, "Steam Generator Tube Integrity." The availability of this TS improvement was announced in the Federal Register on May 6, 2005 (70 FR 24126), as part of the consolidated line item improvement process. contains a description of the proposed changes, the supporting technical analyses, and the no significant hazards consideration determination.
Enclosures 2 and 3 contain marked-up and retyped (clean) TS pages, respectively. provides the marked-up TS Bases changes. TS Bases changes will be implemented pursuant to TS 5.5.14, "Technical Specifications Bases Control Program," at the time this amendment is implemented.
Pacific Gas and Electric Company (PG&E) has determined that this LAR does not involve a significant hazard consideration as determined per 10 CFR 50.92.
Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs to be prepared in connection with the issuance of this amendment.
A member   of the  STARS   (Strategic Teaming  and    Resource     Sharing)   ALLiance Callaway
* Comanche Peak
* Diablo Canyon 9 Palo Verde 9 South Texas Project e Wolf Creek

Document Control Desk                                       PG&E Letter DCL-06-061 May 30, 2006 Page 2 The changes in this LAR are not required to address an immediate safety concern.
PG&E requests the license amendment(s) be made effective upon NRC issuance, and to be implemented within 120 days of the date of issuance.
This communication contains no new commitments.
If you have any questions or require additional information, please contact Stan Ketelsen at 805-545-4720, Sincerely, James     . Becker Vice President- Diablo Canyon Operationsand Station Director 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) Alliance Callaway
* Comanche Peak e Diablo Canyon
* Palo Verde
* South Texas Project
* Wolf Creek

PG&E Letter DCL-06-061 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION
                                                   ) Docket No. 50-275 In the Matter of                                   )    Facility Operating License PACIFIC GAS AND ELECTRIC COMPANY)                       No. DPR-80
                                                   ))
Diablo Canyon Power Plant                               Docket No. 50-323 Units I and 2                                       ) Facility Operating License I No. DPR-82 AFFIDAVIT James R. Becker, of lawful age, first being duly sworn upon oath states that he is Vice President - Diablo Canyon Operations and Station Director of Pacific Gas and Electric Company; that he has executed License Amendment Request 06-04 on behalf of said company with full power and authority to do so; that he is familiar with the content thereof; and that the facts stated therein are true and correct to, the best of his knowledge, information, and belief.
Jamesiet    Becker Vice President - Diablo Canyon 0 erations and Station Director Subscribed and sworn to before me this 30th day of May, 2006, by James R. Becker, personally known to me or proved to me on the basis of satisfactory evidence to be the person who appeared before me.
A KAREN WASON, Notary Public                                             CQMMbdW # 1426454 WARY Pubk --CaWbnVa County of San Luis Obispo                                 IM Lub Obbpo Cot"y State of California

Enclosure 1 PG&E Letter DCL-06-061 EVALUATION

==1.0  DESCRIPTION==

This letter is a request to amend Operating Licenses DPR-80 and DPR-82 for Units 1 and 2 of the Diablo Canyon Power Plant (DCPP), respectively.
The proposed changes would revise the Operating Licenses to revise the requirements in Technical Specification (TS) related to steam generator tube integrity. The changes are consistent with NRC approved Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-449, "Steam Generator Tube Integrity," Revision 4. The availability of this technical specification improvement was announced in the Federal Register on May 6, 2005, as part of the consolidated line item improvement process (CLIIP).

==2.0  PROPOSED CHANGE==
S Consistent with the NRC-approved Revision 4 of TSTF-449, the proposed TS changes include:
* Revised TS definition of LEAKAGE
         "   Revised TS 3.4.13, "RCS Operational Leakage" o   New TS 3.4.17, "Steam Generator (SG) Tube Integrity"
* Revised TS 5.5.9, "Steam Generator (SG) Tube Surveillance Program"
         "   Revised TS 5.6.10, "Steam Generator (SG) Tube Inspection Report" Proposed revisions to the TS Bases are also included in this application. As discussed in the NRC's model safety evaluation, adoption of the revised TS Bases associated with TSTF-449, Revision 4, is an integral part of implementing this TS improvement. The changes to the affected TS Bases pages will be incorporated in accordance with the TS Bases Control Program.
The proposed TS changes are noted on the marked-up TS page provided in Enclosure 2. The proposed retyped TS is provided in Enclosure 3. The revised TS Bases is contained for information only in Enclosure 4.

==3.0 BACKGROUND==

The background for this application is adequately addressed by the NRC Notice of Availability published on May 6, 2005 (70 FR 24126), the NRC Notice for Comment published on March 2, 2005 (70 FR 10298), and TSTF-449, Revision 4.
1

Enclosure I PG&E Letter DCL-06-061 4.0  REGULATORY REQUIREMENTS AND GUIDANCE The applicable regulatory requirements and guidance associated with this application are adequately addressed by the NRC Notice of Availability published on May 6, 2005 (70 FR 24126), the NRC Notice for Comment published on March 2, 2005 (70 FR 10298), and TSTF-449, Revision 4.

==5.0 TECHNICAL ANALYSIS==

Pacific Gas and Electric Company (PG&E) has reviewed the safety evaluation (SE) published on March 2, 2005 (70 FR 10298), as part of the CLIIP Notice for Comment. This included the NRC staffs SE, the supporting information provided to support TSTF-449, and the changes associated with Revision 4 to TSTF-449.
PG&E has concluded that the justifications presented in the TSTF proposal and the SE prepared by the NRC staff are applicable to DCPP, and justify this amendment for the incorporation of the changes to the DCPP TS.
In order to provide consistency with the new TSTF-449, TS 5.5.9.d, provisions for SG tube inspections which include the entire length of the SG tube, the current TS 5.5.9.d.1 .k (ii)definition for the W-Star (W*) length is revised to delete the cold leg W* length, thereby excluding application of W* repair criteria to cold leg degradation. The current TS 5.5.9 W* repair criteria for DCPP do not include an NRC approved leak methodology for cold leg degradation. Therefore, the TS revision to delete the cold leg W* Length will ensure that all degradation left in service under the W* repair criteria will be supported by an NRC approved leak methodology.

==6.0 REGULATORY ANALYSIS==

A description of this proposed change and its relationship to applicable regulatory requirements and guidance was provided in the NRC Notice of Availability published on May 6,2005 (70 FR 24126), the NRC Notice for Comment published on March 2, 2005 (70 FR 10298), and TSTF-449, Revision 4.
6.1     VERIFICATION AND COMMITMENTS The following information is provided to support the NRC staff's review of this amendment application:
2

Enclosure I PG&E Letter DCL-06-061 Diablo Canyon Power Plant, Units 1 and 2 Steam Generator Model(s):      Westinghouse Model 51 Effective Full Power Years     Unit 1: 17.2 EFPY at end of cycle 13 (October (EFPY) of service for                  2005) currently installed SGs Unit 2: Projected 17.4 EFPY at end of Cycle 13 (April 2006)
Tubing Material                Inconel 600 Mill annealed Number of tubes per SG         3388 Number and percentage of       Unit 1 SG 1 - 232 tubes plugged (6.85%)
tubes plugged in each SG       Unit 1 SG 2 - 315 tubes plugged (9.30%)
Unit I SG 3 - 92 tubes plugged (2.72%)
Unit 1 SG 4 - 186 tubes plugged (5.49%)
Unit 2 SG 1 - 125 tubes plugged (3.69%)
Unit 2 SG 2 - 268 tubes plugged (7.91%)
Unit 2 SG 3 - 130 tubes plugged (3.84%)
Unit 2 SG 4 - 367 tubes plugged (10.8%)
Number of tubes repaired       None in each SG Degradation mechanism(s)
* Axial outside diameter stress corrosion identified                       cracking (ODSCC) at tube support plate (TSP) intersections
* Volumetric indications at TSP intersections
* Several types of stress corrosion cracking at dented TSP intersections: Axial primary water stress corrosion cracking (PWSCC), axial ODSCC, circumferential PWSCC, circumferential ODSCC, combination axial PWSCC and axial ODSCC, combination axial and circumferential indications
* Axial PWSCC in Row 1 and Row 2 U-bends
* Circumferential PWSCC in Row 1 U-bends o Circumferential PWSCC in greater than Row 2 U-bends
* Several types of Hot Leg Tubesheet Region Degradation: Axial ODSCC, axial PWSCC, circumferential PWSCC, circumferential ODSCC, volumetric indications 3

Enclosure 1 PG&E Letter DCL-06-061 Diablo Canyon Power Plant, Units 1 and 2
                             " Axial PWSCC and circumferential PWSCC in Hot Leg Deplugged Tube Plug Expansion Zone
                             " Axial ODSCC in free span dings
                             " Cold Leg Thinning at cold leg TSPs
                             " Anti-Vibration bar wear
                             " Ligament cracking in TSPs Current                      Per SG: 150 gallons per day (gpd) through any primary-to-secondary                   one SG leakage limits:              Leakage is evaluated at: Room temperature Approved Alternate Tube Repair Criteria (ARC):
: 1. Voltage-Based Repair      Approved by: Amendment Nos. 124 (Unit 1)
Criteria                  and 122 (Unit 2) dated 03/12/98.
Applicability (e.g., degradation mechanism, location): Axial ODSCC within TSPs.
Special limits on allowable accident leakage:
None.
Exceptions or clarifications to the structural performance criteria that apply to the ARC:
None.
: 2. W* Criteria               Approved by: Amendment Nos. 182 (Unit 1) and 184 (Unit 2) dated 10/28/05.
Applicability (e.g., degradation mechanism, location): Predominately axially oriented inside diameter stress corrosion cracking confined within the tubesheet, below the bottom of the WEXTEX transition; any combination of tube degradation below the flexible W* length is acceptable.
Special limits on allowable accident leakage:
None.
4

Enclosure 1 PG&E Letter DCL-06-061 Diablo Canyon Power Plant, Units 1 and 2 Exceptions or clarifications to the structural performance criteria that apply to the ARC:
None.
Approved by: Amendment Nos. 152 (Unit 1) and 152 (Unit 2) dated 05/01/02.
: 3. Axial PWSCC               Applicability (e.g., degradation mechanism, Depth-Based Repair        location): Axial PWSCC within the thickness of Criteria                  dented TSPs.
Special limits on allowable accident leakage:
None.
Exceptions or clarifications to the structural performance criteria that apply to the ARC:
None.
Approved SG Tube Repair      None.
Methods Performance criteria for     Primary-to-secondary leak rate values assumed accident leakage             in licensing basis accident analysis, including assumed temperature conditions: The primary-to-secondary leak rate values assumed in the licensing basis main steam line break accident radiological consequences analysis is 10.5 gallons per minute (gpm) in the faulted SG and 150 gpd in each of the three intact SGs.
These leak rates are at room temperature conditions.
Performance criteria: For combined alternate repair criteria and non-alternate repair criteria degradation, accident induced leakage is not to exceed 10.5 gpm (at room temperature) in the faulted SG. For only non-alternate repair criteria degradation, accident induced leakage in the faulted SG is not to exceed 1 gpm (at hot conditions), or 0.72 gpm at room temperature conditions, because the accident analyses assume a pre-existing I gpm total leakage to all SGs prior to the accident.
5

Enclosure I PG&E Letter DCL-06-061 7.0  NO SIGNIFICANT HAZARDS CONSIDERATION PG&E has reviewed the proposed no significant hazards consideration determination published on March 2, 2005 (70 FR 10298), as part of the CLIIP.
PG&E has concluded that the proposed determination presented in the notice is applicable to DCPP, and the determination is hereby incorporated by reference to satisfy the requirements of 10 CFR 50.91 (a).
8.0  ENVIRONMENTAL EVALUATION PG&E has reviewed the environmental evaluation included in the model SE published on March 2, 2005 (70 FR 10298), as part of the CLIIP. PG&E has concluded that the staffs findings presented in that evaluation are applicable to DCPP and the evaluation is hereby incorporated by reference for this application.
9.0  PRECEDENT This application is being made in accordance with the CLIIP. PG&E is not proposing variations or deviations from the TS changes described in TSTF-449, Revision 4, or the NRC staffs model SE published on March 2, 2005 (70 FR 10298).

==10.0 REFERENCES==

Federal Register Notices:
Notice for Comment published on March 2, 2005 (70 FR 10298)
Notice of Availability published on May 6, 2005 (70 FR 24126) 6

Enclosure 2 PG&E Letter DCL-06-061 Proposed Technical Specification Changes (marked-up)

Definitions 1.1 primary to secondary I (primary to secondary LEAKAGE)I 1.1 Definitions 3.ector Coolant S           (CS) LEAKAGE .,,.,,h LEAKAGE a's~Iam generat (S* t he Secondary Sys~tem.
(continued)
: b. Un~identif 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                            I--o nonisolable fault in an RCS component body, pipe wall, or vessel wall.
MASTER RELAY TEST                 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.
MODE                              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.
OPERABLE-OPERABILITY              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                     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.
(continued)
DIABLO CANYON - UNITS 1 & 2                       1.1-4         Unit 1 - Amendment No. 435, Unit 2 - Amendment No. 435,

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 one@9S         ---                                                      I1~
IoperationalI                                    steam generator (SG)
APPLI ABILITY:          MODES 1, 2, 3" and 4*.              or primary to secondary LEAKAGE ACTIO CONDITION                ,        REQUIRED ACTION                 COMPLETION TIME A. RCS LEAKAGE not wi                 A.1       Reduce LEAKAGE to            4 hours limits for reasons &#xfd;ier than               within limits.
pressure bopeary LEAKAGE.
B. Required Action and               B.1      Be in MODE 3.                6 hours associated Completion            AND Time of Condition A not met.
OR Pressure boundary                B.2       Be in MODE 5.                36 hours 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 I - Amendment No. 4-35 442, Unit 2 - Amendment No. 435 44.2,

L 2. Not applicable to primary to secondary LEAKAGE.           I      RCS Operational LEAKAGE 3.4.13 SURVEIL"NE REQUIREMENTS                                 E&#xfd;1 SURVEILLANCE                                     FREQUENCY SR 3.4.13.1                                  ---
                    '     ot required to be performed until 12 hours after Iinsert     I          stablishment of steady state operation.
       --                     o
                       .............. R        intc;    ......... .. ,
* 72 hours
(&#xfd;        &#xfd;413-2                                                                                           -I1 ei T ysearn generator tube integrity is in accordance     In accordance the  Steam      with  _
with the Steam Generator Tube Surveillance Program.                                                    Generator Tube Surveillance I                                                                      Program. _
           ,ell S2-SR 3.4.1i            Verify steam generator primary-to-secondary leakage          72 hours equal to or less than Ilowable.
I I 2 rInsert 2 Insert 3 DIABLO CANYON - UNITS 1 & 2                          3.4-28          Unit I - Amendment No. 435, Unit 2 - Amendment No. 435,

RCS Specific Activity There are no changes to the page.                            3.4.16 Page included for information only.
           -   250 I-)
            >00 0
5U 0
U"150 z
0 0
0
           < 100 0~
cf)
I--
z W    50
          -J D
W 0
0 o     0 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 > I 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 I - Amendment No. 435, 184 Unit 2 - Amendment No. 435, 186

Programs and Manuals 5.5 Unit 1 5.5 Programs and Manuals (continued) 5.5.9   Steam Generator (SG Tube Surveillanc Program SG tube intogrity shall be d-emonstrae                       ' prforance of the follow.ing augmentod The pro'..'i9Ron of SIR 3.0.2 are applicable to the SG Tube Sur.'oillanee ProgrFam test ftequeReies.
duriRg shutdown*m          by sletng          and           SG  tube at  integ least th   shall beu deteu rmined rity o in    Sreleci ona       nd In p to Sapler
                 *.SGsSCspecified           Table      5.5.9 1.
: b. SCTubeSample Selecation and Insgpection The SG tube miiu aple size, inspection rosult classification, and thcrespnding action required shall be as pefredinatTable sert 5           ;s-t5specified              5.5.9-2. The ir.1i-               i - rnspAet*iof SG*tubes shall be the freq..encies  .i     specifid         Specifn.
A         5.5.9.G and the inspected fate,             tuber, shall be uerified acceptable per the acceptance creiteia Of Specification 7sert 5           5........ The. tube...elcte fore.ch ine.                            inspectn sha! inc.ude
                                                                                                          .e       at least 3%
fianges TS        ofekatheHk tot-al eaa1 numbeer
                                       # r4    n of tubes     in all Sgs;a;,a+the tubes selected-for theseinpcos r4,rrr k'ec
: 1. Where expeine in similar plants with similar: water chemistr,' indicates critical areas to be inspected, then at least 50% of the tubes inspected shall be from. t~hese critical ar~eas; 2.The first sample of tubes selected for e~acGh insrtavice kispectio (subsequent to the preser:c inpcion) of each SG shall icuc a)         All neonplugged tube tha priusly had detectable wall pen~etrations (greater than 20%)-,
b)         Tubes inthese are-as wAhe-re experiencGe has indicated poten;tia!
problems, G),
A tube inspecation (pursuant to Specification~ 5.5.9441.h) shall be peformed on each selected tube.If                  an;y seec*ted tube does not permit the passage of the eddy current probe for a tube inspection, this shall be r-ecorded and an adjacent tube shall be seieciea ana suojecteG Io              a tune inspection, Note, this is             i              Indications left in service as a result of application of the tube plate voltage-based repair criteria shall be inspected by moved to 5.5.9.d.4
____y _.support bobbin coil probe during all future refueling outages,                               I.
Note, this                              Tubes identified as W* tubes having a previously identified 5.... ~
movedtjmoved~            ~~               .GGH..4. .......within
                                       ~eatn toRak indication                       RPC-)W*
                                                                   .,l.flexible
                                                             *.."the                  length probe   or equivalent    for the fullusing shall be inspected         length of the W* region during all future refueling outages.
(continued)
DIABLO CANYON - UNITS 1 & 2                                    5.0-10           Unit I - Amendment No 435, 45-, 182 Unit 2 -Amendment No                4

Programs and Manuals 5.5 Unit 1 5.5 Programs and Manuals 5.5.9    Steam Generator (SG) Tube Sur-eilf;nco Program (continued)                                                I 3.the tubes~r selected as the second and third samples (if required by Tab!o 5.5.9 2) during each i"s..;'i inspection, may be subjected to a partial tube inspection provided:
I     *1    d I       I    I  I f    Jl                             il
* I   f    dR
                      -I             twunco n1oh      rs;1ArtOO Tor     mtneq        samnies!RGncIuao  tAP WUD~ fmro  mtena arFeascA of hoe tub--he A-hooet   R- IPIFIR1 where tub-es, with mpretoswore proiosly found, and b)       T-he inspections include these portions of the tubes wher imperfections wore pr~eviously found.
Implementation of the steam generator tube/tube support plate repair criteria 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 of at least a 20% random sampling of tubes inspected over their full length.
Inspection of dented tube support plate intersections will be performed in accordance with WCAP-1 5573, Revision 1, to implement axial pr4way water. tress corosion cracking (PWSCC) depth-based repair criteria. The extent of required inspection is:
100 percent bobbin coil inspection of all tube support plate (TSP)                  I intersections.                                                                       I Plus Point coil inspection of all bobbin coil indications at dented TSP intersections.
Plus Point coil inspection of all prior PWSCC indications left in                    I service.                                                                             I 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 Note, TS                 volt dents up to the highest TSP for which PWSCC has been 5.5.9.b.5                detected in the prior two inspections or current inspection and 20%
is moved                 of greater than 2 volt dents at the next higher TSP. If a to                       circumferential indication is detected in a dent of "x" volts in the 5.5.9.d.6                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)
DIABLO CANYON - UNITS 1 & 2                           5.0-11                 Unit 1 -Amendment No 435, 152 Unit 2 - Amendment No. -345, 152

Programs and Manuals 5.5 Unit I 5.5 Programs and Manuals 5.5.9  Steam Generator (SG) Tu-bo Sur,-ewilancp               Program (continued)
LI<                     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 kz been detected in the prior two inspections or current inspection and
[          20% of all dents at the next higher TSP.
fy       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.
of uiti   samriple iiispe~iu IeR   shal i be dasswe       inte GReu 94- uiu 4uilu:iI~g  thIUU rcategegies!
Loss than 5%of the total tubos inspected are dogradod tubes and none of the inspected tubes are dofective.
One or mere tubes, but net more than 1% of thc total t'ubes ins..pected are defectire or     ", bet...een 594 and Q%of the total tubes inspected are derddtbes.
M.1oe than 10% of the total tubes inspected are degraded tubes or mr.e than, % of the iOnspeted tuber, are defective.
Ngte*   In all inspections, previously degraded tubes must exhibit significant (greater                   than 10%) fu.h.. wall penet*ations to be included in the above peGRentage InsPection Freque~ncies The above required isr'c                       npcin of SG tubes shall be peformed at the following frequencies:
                    . Th      frs nsoiew inpection shall be performRed Power Monthsr bu within 21 calendarm                 t after 6 Effecntiwe of.initial c.1.a    .
Ful-1 Subsequent inSor'Ace         i 24p~~ahallM~h insetin                        t be per~frmed     at rev~eu intervals of not less than 1       r Mere tha 21 calendar monthG aevu finspection. If W~O consecutive inspections not including               the.peo~c inpection, resul1t in all1 1nspoctiGn results falling lnto the C- Gategory Or if W9~ consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occur.ed, the i       .spection interal may be extended to a maximum of once peF 40 menthsj-(continued)
Note, page number will change due to repaginationof section 5.5.9 pagesI DIABLO CANYON - UNITS 1 & 2                              5.0-1 Ia          Unit 1 - Amendment No 435,152 Unit 2 - Amendment No. 435,152

Programs and Manuals 5.5 Unit 1 5.5 Programs and Manuals 5.5.9  Steam Generator (SG) T,'ub Su-.Rillance Program (continued)                                            I
               -2. Ifthe resuits of the insei imw Inspection       of a SG Gonducted in accor-dancV with Table 5..92;t-10 mo.nnth intc-6=r_'l falyinCt Go'C 3, the inspection frequency shall be increased to at least once per 20 mo=ths. The increase n inspection frequency shall apply until the subsequent inspections satisfy the criteria Of Specification 55.5.9-n-4. The inter'.al may then be oxtonded to a maximum Of once per 10 months; and
              -.      Additional, Unscheduled inser-ice inspections ,haUl be perf.mied on each SG in accor-dance W'ith the first sample inspectio specifed inTbe559-2 dur;in the shutdown subsequent to any of the followiRg conditions:
a)      Reacator to secon~dary tube leaks (not including leaks or~igiating from; tube to tube sheet we'lds) in exc-esss of the limits at Specification 3.1.13; or b)    A seismi ocuene greater than the Double Design Earthquake, G) A los f olat cident requir-ing actuation of the EnRgineered Safet Features, or I            d) A m~ain sbtea~m linem or feedwater line bPrek.
: d.          optvv mva     i auii I.      Ac, used inths Specification:
a)     lmP.eorctioneans an exception to- the dvimenRsfions, fini contur a tbe o fom tat rqurd by fabrication drawings-or specifications. Edd cuenmt testing indiction.lw 20% of the nominal.. tube.hkness,            if det impeAeeiGRs7, b)    D)egr ada-;tionn m.eanaseic indu-e-d c-rac-king, wastage, wear ot genera! corrosion ocu9 igoGct inside or outside of a tube; r.)    Degraded Tube mnean;s atube containinig imperfections greater-than or equal to 20% of the nmAinal wAll1 thickness caused by degaatien,-
d)    % Degradation mneans the percen~tage of the tube wallthcns affected Or removed by degradation;.
e)    Defccgt means an imperfection Of such severity that it exceeds the plugging limit. A tube containing a defect is defectiv.e;:
f)/       ....
                              .omg.   ,,,;""s "m
                                               ,,)
R . ....,f+)',
t,,,       e,,,..,
iv*"*"
                                                                            ;"v,,
depth a., 9F o beyond w'hich th.
tube shall be removed from 6eRc"         " and is equal to 40% of the nomminal tube wall thikIcess.
: 1) This definition does not apply to tube support plate intemrsetions for which the voltage based repair-Gcitera are beiRng applied.
Refer to 5.5.9.d. 11 for the repair limit applicable to these Note, page number will change due to repaginationof section 5.5.9 pages                   j(continued)
DIABLO CANYON - UNITS I & 2 5.0-12      Unit I - Amendment No 4-35 142 Unit 2 - Amendment No. 435 142

I Note, this s                                                            axial PWSCC indications, or                ~4Q portions thereof, which are located within the thickness of moved to                         dented tube support plates which exhibit a maximum depth 5.5.9.c.3                        greater than or equal to 40 percent of the initial tube wall thickness. WCAP-1 5573, Revision 1, provides repair limits a$applicable to these intersections.
Note, this is          (&#xfd;4-)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 removed from service.
g)UnSeRgipeah-1e describes the condition. of a tube if it leaks or contains a defect large enougIh to af et its strucura Iitgity        I         an the event of a
* A.
Note, TS 5.5.9.d. 1.1 is moved 71&#x17d; t ho/
DoubleU PeSICIR[ ma~t[K1uakeUIU~.
Tube I,;,-,e            means a es          ALUUIUII feedw.vat:r line break as eDeCified in 5.5.9.c.3. above:
inspection of the U.XIUUBLR1 9F a steamI III1U 9
                                                                                              ,G from the tube end tube 5.5.9.d.5                 (hot   legP,-dte1bedtthto                                                 uptofteGl tobleg ...           ".L.... .                  +... v...
used for thedispostion of aalo N
                    /            v._                                  ....
wh            &#xfd;iboundinor ube _; _8 inches '1 ervim SFlexible W*Lge.                                       the hot leg top of tubesheet, elw I The tube support plate voltage-       s            confnestabdish a baseline       ticn of"the tubin.         This irperti s.hat l bes based repa criteria are                predominantl axsiallyoientiedR                    diaete stes oroin        gttsif rckn irplae~f~interscions ube         PatethePu.iLit           used for the disposition oftaan alloy pilugiglmtibsdodanann 600 steam generator tube for contin ed servce that is experiencing I l         predominantly axially oriented outsid* diameter stress corrosion cracking Note, TS jiis l          confined within the thickness of the tulle support plates. At tube support 5.5.9.d.1 *" /           plate intersections, the plugging limit is based on maintaining steam moved tc             I      ,,generator tube serviceability as described below:
5.5.9.c. I
                          ,(i* Steam generator tubes, whose degr tation is attributed to outside diameter stress corrosion cracking wi in the bounds of the tube S       support plate with bobbin voltages les than or equal to the lower voltage repair limit (NOTE 1), will be al wed to remain in service.
sconti uedw pag                                                                                  )
nubrwl hnedet Note,~------                                                                                   I Implementation of the W* repair criteria requires a 100 percent RPC probe or Note, page number will change due to repaginationof section 5.5.9 pages                   I I
equivalent DIABLO CANYON - UNITS 1 & 2                             5.0-13    Unit I - Amendment No. 121-I4f2,182 Unit 2 - Amendment No. ii6i                      184

Programs and Manuals 5.5 Unit 1 5.5 Programs and Manuals 5.5.9  Steam Generator (SG) Tube Sur'.'veilnco Program (continued)                                  I Y--7r 2
Steam generator tubes, whose degradation is attributed to outside diameter   stress corrosion cracking within the bounds of the tube b      support plate with a bobbin voltage greater than the lower voltage repair limit (NOTE 1), will be repaired or plugged, except as noted in 55.9c.l.c9l.4                    (4) below.
                        )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 the lower voltage repair limit (NOTE 1) but less than or equal to the upper voltage repair limit (NOTE 2), may remain in service if a rotating (contnued pancake coil inspection 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 (NOTE 2) will be plugged or repaired.
(continued) 3ination of section 5.5.9 oaaes    I I Note, page number will change due to repag in........ .. .. ..i.n. ....9n-o--   I DIABLO CANYON - UNITS 1 & 2               5.0-13a      Unit 1 - Amendment No..13E-I ',452,182 Unit 2- Amendment No. *35.!5!.52,184

Programs and Manuals 5.5 Unit 1 5.5 Programs and Manuals 5.5.9 Steam Generator (SG) Tube Su'erilan-gec Program (continued)
                     &#xfd; Q ivCertain December   intersections   as identified in PG&E Letter DCL-03-174, dated 19, 2003, will  be excluded from application of the voltage-
[.j El based repair criteria as it is determined that these intersections may collapse or deform following a postulated LOCA + SSE event.
: e.            (v) Ifan unscheduled mid-cycle inspection is performed, the following Emid-cycle                        repair limits apply instead of the limits identified in 5.5.9.c.
55:
551a  ,4                    (ii)t-,      5.5.94    4     (iii).. The  mid-cycle repair 5.5.9.c.l.b, 5.5.9.c.1 .cI and           limits are determined from the          following equations:
S VMURL =(CL                                    - At) 1.0 + NDE + Gr CL (VU       -      ,
VLRL)     (CL - At)
VMLRL = VMURL      -
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 VLfL 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) 5.5.9.c.1.a,            Implementation of these mid-cycle repair limits should follow the same 5.5.9.c. I.c and 5.5.9.c.l.b,            approach as in T.5.9.d.1.j vl.,. , . . w.i), 5.5.94d.1.j
                                                                     . . . . . (ii),            . . . . . (iii).
                                                                                    ... and 5.5.9*.*,4j     . .
(continued)
Note, page number will change due to repagination of section 5.5.9 pages DIABLO CANYON -UNITS 1 & 2 5.0-14              Unit I - Amendment No. 4-35, 176 Unit 2 - Amendment No 4-5, 178

Programs and Manuals 5.5 Unit 1 5.5 Programs and Manuals 5.5.9    Steam Generator (SG) Tu'b SRuweoiRapnce Program (continued)                                           I The W*                   NOTE 1: The lower voltage repair limit is 2.0 volts for 7/8 inch diameter repair*                                tubing at DCPP Units 1 and 2.
criteria are                        2: The upper voltage repair limit is calculated according to the hodology in Generic Letter 95-05 as supplemented.
                                .^                     sed 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 moved to                 in this specification:
5.5.9.c.2.         ,     (i) Bottom of WEXTEX Transition (BWT) is the highest point of contact between the tube and tubesheet at, or below the top-of-tubesheet as Ia.I            ,       determined by eddy current testing.
W* Length is the distance in the tubesheet below the BWT that EV                      precludes tube pull out in the event of the complete circumferential separation of the tub eiw the W* length. The W* length is conservatively set a:i, an undegraded hot leg tube lengthcOL52.2_Q inches for Zone A tubes and 7.0 inches for Zone B tubea ung*,radod cold leg tube length of 5.5 inchoc for Zone      A.t^u's ;*;nd-7.5 inchec for Zocn B tubo*. Information orovided in WCAP-14797-P.
Revision 2, defines the boundaries of Zone A and Zone B.
[.             (iii lexible W* Length is the W* length adjusted for any cracks found within the W* region. The Flexible W* Length is the total RPG--
inspected length as measured downward from the BWT, nd includes DdNDE                       uncertainties and crack lengths within W* as adjust d for growth.
(iv W* Tube is a tube with degradation within or below the
* length that
                              'is left in service, and degraded within the limits specified I
Specificati 6-5-941k-(V4.
: e.       * (v) Wit            e tubesheet, the plugging (repair) limit is based n j- ."                    intaining steam generator serviceability as described b low:
: 1) For tubes to which the W* criteria are applied, the len th of non-9.c2.e        I              degraded tube below BWT shall be greater than or eq al to the W* length plus NDE uncertainties and crack growth fo the operating cycle,                           rotating pancake coil (RPC)
(continued)
(continued)
Note, page number will change due to repagination of section 5.5.9 pages DIABLO CANYON - UNITS 1 & 2                              5.0-15          Unit I - Amendment No 435,182 Unit 2 - Amendment No. 435,184

Programs and Manuals 5.5 Unit 1 5.5 Programs and Manuals 5.5.9 Steam Generator (SG)Tubo Surweiga*nce Program (continued)
EV-upper   a 2)-al        cracks in tubes returned to service using W* shall have the 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.
j      Resolvable, single axial indications (multiple indications must E7                 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 ncluding 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 or repaired. 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 "anglesgreater than (45 degrees - ANDEcA) within the flexible W*
an~'-
                              .ngth shall be plugged or repaired.
Any type or combination of tube degradation below the flexible W*
6length      is acceptable.
: 2.      Tho SG tubo integrity shall be detormined after completing the required by,'Table 5.5.92-.
The contents and f.equon...       of rpet. Gcon.erning the SG tube c...eillan.e progr~am rchal;l bhe in accor-dance w.ith SpecificationR 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 I - Amendment No 435, 442,182 Unit 2 - Amendment No ,3, 442,184

Programs and Manuals 5.5 Unit I
                                                                                          -d THIS PAGE NOT USED                                            I Note, page number will change due to repagination of section 5.5.9 pages DIABLO CANYON - UNITS 1 & 2            5.0-17     Unit I - Amendment No. 435,444182 Unit 2 - Amendment No. 4-35, 454184

Programs and Manuals 5.5 Unit 1 5.5 Programs and Manuals (continued)
IflftllIA!lI!         Ikll M/DID f~l* *TI=ARR            V*-=IMIDAT/flID)        IC*f-&I T/fl DI INSPECTED DURING INSERVICE INSPECTION ProesePoiGo RnGPctin                                                            NG                                         Ye
 ,i,. ,f-,r S**.,                        ......
e.....              Si+         Two       Three I-euO                     Two          T-hFee     F-euW Forst nser.'wice Inspectien                                                     All                          Oe            Twe       Twe Se~ond & Subsequent Inservice                                                             ~
TALEINOTATIONTK
: 1.       Tho n.....;i*"i*n.                   ctio .. may be limited to one G-On a rotating G*hdule.. n ...pasein 3N%of the tubes e.w.hero N is the number of SGs inthe plant) if the rosultS of the 46rt or proiou inpoction indicate that all SGs a~e pe~foRming in a like m~anner. Note that unde-r coImIe circumstances, the operatin@g coditionsmli                                in    e or m         r      may be found, to be mor~e severe than those in other SGs. Un~derF SUch crcumstaRces the Sample ceguonco &-all bhemo-dified to inepect the mest 6everc conditions.
: 21.      The other SG not in~e~oted duFrng the first inovc n~cinshall be inSe~oted. The firusoun ex      2? .  .   . .
nnoroccou-TT.10niciunn esroainiaoo pr----- I Ax   I-A DL   p*ofl  11    Sf&#xfd;l             r%^                 !i   nI SI   Ifl*     **.^MI   ^V  II ISrS   I US p V inmcreton          dui
                              "*d v the ecd and third in...ctions. The feouth and subseuoent I I      el  l @    A   I inspections snall TO!HiOW the insrGIuctons aeScried in 1 above.-
Note, page number will change due to repagination of section 5.5.9 pages U-e DIABLO CANYON - UNITS 1 & 2                                                 5.0-18                Unit I - Amendment Unit 2 - Amendment No.115

Programs and Manuals 5.5 Unit I 5.5 Programs and Manuals (continued)
TABLE 5.5.9 2 STEAM G          ^ENERA  I*-* TRM= I-Nt*SprDcl-lm IST  SAMRPRLE INSPET                       2ND SAMPLE                    3RD SAMPLE Sample     IRosult         AGVGR        R~utI           AGWeR          Result        AGtioR SiZe     I__           Rle&#xfd;           _              Requied         __jI        Requied NGmme         N-.A-Gf-S Tubes-peF        G-      Plug def-                          NoRe tw tu~bes-          G4       Plug defeG~ve        G4     NGRe and iRePe                    tubes-d              G4    1RugdefeGtive inspee 2S tubese4A                  Additina             G4     Perform actio 424t bhp                    fGeG 3 es6' 4nspeet all                                              of fiFe4-sample 4''4      Pe .=..             N-.A-         N-A-aetiRor Ge2 defeetive                     Fesdt ef-first tu.ber, an4
* sperat 2S     All-ethef          None          N-A-           NA&#xfd; eaeh GtheF SOMe         Pe~fem~             NA.            NA&#xfd; S.G.6        arafenf6 G-Netifioaten     C-2 -bu      Fesulef teNRG           Fie add-     seeend pu~sa44eO       iWenwa       sample S Q r4 (2) ef-410F Pad S5o Add-tubes-In-eaGh4
                                          &G4          deferGfie NatifGatiGA t-nhRC pu~suan44 49.72 I4b}R
                                                       &-f 10 G FR P44-50 C - 3 N q4              iS the number of SG' in;the unit, and Ris the number of SG, incpccted n Nt    durig an i     nspedtion Note, page number will change due to repaginatien of section 5.5.9 pages        j(ontinued)
DIABLO CANYON - UNITS 1 & 2                     5.0-19           Unit I - Amendment No1 Unit 2- Amendment No 1

Programs and Manuals 5.5 Unit 1 5.6 Reporting Requirements (continued) 5.6.10 Steam Generator (SG) Tube Inspection Report a      '! !ilthin 15 daI S follOv ina the cemiletion of eachIOIIins;  if         I     G tuboc     I thenumerof tubes plugged in each SGsalbe rcported to the Gommsin
: b. The complete results of the SG tube ino.ieisetion shall be 6ubmi;#ed to the Commissio inaeport w.ithin 12 Monmths foAwn opletion of the in~spectionR. This Special Report ch.al! incrlude:
: 1)          Number and e~dont_ of tubes inspectedI
: 2)           Location; and perGent of W.al! thickness penetration for eac;hincaono an impegrfetion, and
: 3)           I-dentification of tubes plugged-.
G. Results of SG tube inspections, W..hich fall into Categor,' C 3, shall berprtdi a Special Report to the Commission Within; 30 days an~d prior to rosumpino plant operation. This report shall pro.'do a description Of in'..csiga..n con~ducted to determn cueof the tube degradation and corrective measures t~a~ken ton preven~t recurrence.
For implementation of the voltage-based repair criteria to tube support plate             I intersections, notify the NRC prior to returning the steam generators to service should any of the following arise:
: 1.       If estimated leakage based on the projected end-of-cycle (or if not practical, using the actual measured end-of-cycle) voltage distribution, increased by estimated leakage by all other sources (alternate repair criteria and non-altemate repair criteria indications), exceeds the leak limit determined from the licensing basis dose calculation for the postulated main steamline break for the next operating cycle.
: 2.      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-of-cycle (or if not practical, using the actual measured end-of-cycle) voltage distribution exceeds I x 102, notify the NRC and provide an assessment of the safety significance of the occurrence.
(continued)
DIABLO CANYON - UNITS 1 & 2 5.0-29       Unit I - Amendment No. 435, 4-82, Unit 2 - Amendment No. 4-35, 484,

Programs and Manuals 5.5 Unit 1 5.6 Reporting Requirements 5.6.10 Steam Generator (SG) Tube Inspection Report (continued) eJ     The results of the inspection of W* tubes shall be reported to the Commission            I  -?
pursuant to 10 CFR 50.4 within 90 days following return to service of the steam generators. This report shall include:                                                   I Identification of W* tube indications and indications that do not meet W*
requirements and were plugged or repaired, including the following information: the number of indications, the locations of the indications (relative to the BWT and U-S), 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.
Assessment of whether the results were consistent with expectations and,          I-fe if not consistent, a description of the proposed corrective action.
The aggregate. calculated steam line break leakage from application of all                If?
alternate repair criteria and non-alternate repair criteria shall be reported to the Commission pursuant to 10 CFR 50.4 within 90 days following return to service of the steam generators.
For implementation of the repair criteria for axial PWSCC at dented TSPs, the NRC shall be notified prior to startup, pursuant to IOCFR50.72, of the following conditions that indicate a failure of performance criteria:
SThe calculated SG probability of burst for condition monitoring exceeds DI1         x 10-2.
:       The calculated SG leakage for condition monitoring from all sources (all alternate repair criteria and non-alternate repair criteria indications) exceeds the leakage limit determined from the licensing basis steam line break dose calculation.
(continued)
DIABLO CANYON - UNITS 1 & 2 5.0-30     Unit I - Amendment No. 435,41-5-,442,482, Unit 2 - Amendment No. 4-35,4441-,452,484,

Programs and Manuals 5.5 Unit 1 5.6 Reporting Requirements 5.6.10 Steam Generator (SG) Tube Inspection Report (continued)
[1        For implementation of the repair criteria for axial PWSCC at dented TSPs, the          I results of the condition monitoring and operational assessments will be reported to the NRC within 120 days following completion of the inspection. The report will include:
Tabulations of indications found in the inspection, tubes repaired, and       I tubes left in service under the ARC.                                          I 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       I 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. 435,4-5-,4-52,492, Unit 2 - Amendment No. 445,154-,452,4-4,

Programs and Manuals 5.5 Unit 1 5.6 Reporting Requirements 5.6.10 Steam Generator (SG) Tube Inspection Report (continued)
Performance evaluation of the operational assessment methodology for           I predicting flaw distributions as a function of flaw size.                      I 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.
Identification of mixed mode (axial PWSCC and circumferential)                 I indications found in the inspection and an evaluation of the mixed mode indications for potential impact on the axial indication burst pressures or
: 8.             leakage.
: 8)      Any corrective actions found necessary in the event that condition              I monitoring requirements are not met.                                            I 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 reported to the Commission pursuant to 10 CFR 50.4 within 90 days following return to service of the steam generators:
The assessment of the probable causes for the underpredications,                I EI-7             proposed corrective actions, and any recommended changes to probability of detection or growth methodology indicated by potential methods assessments.
An assessment of the potential need to revise the alternate repair criteria E 2. 2)4     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. &#xfd;d 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 I - Amendment No. 4-3,45-1-,-452,1-77,482, Unit 2 - Amendment No. 435,4*5-*,462,-7-9,484,

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 hours 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 S tf d e cNOTE.
il 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 hours associated Completion Time of Condition A not          AND met.                             B.2       Be in MODE 5.                36 hours 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 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

Technical Specification Inserts (continued)
Insert 5 (continued) loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.
: 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.
Leakage is not to exceed I gpm for the faulted SG, except for specific types of degradation at specific locations as described in paragraph c of the Steam Generator Program in which leakage from application of all alternate tube repair criteria degradation and non-alternate tube repair criteria degradation is not to exceed 10.5 gpm.
: 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:
I1. 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 plugging limit is based on maintaining steam generator tube serviceability as 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 the lower voltage repair limit (NOTE 1),
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 the lower voltage repair limit (NOTE 1), will be repaired or plugged, except as noted in 5.5.9.c.1.c below.

Technical Specification Inserts (continued)
Insert 5 (continued)
: 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 the lower voltage repair limit (NOTE 1) but less than or equal to the upper voltage repair limit (NOTE 2),
may remain in service if a rotating pancake coil inspection 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 (NOTE 2) will be plugged or repaired.
: 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. 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.1 .c. The mid-cycle repair limits are determined from the following equations :
Vs, VMURL                           (CL-At) 1.0 + NDE + Gr          CL
                                           =v~o.,(CL                      -  At)
VMLRL              -  (VURL - VLRL)  (CL-At
                                         =VMURL                          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 Vuj 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)

Technical Specification Inserts (continued)
Insert 5 (continued)
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.
NOTE 1: The lower voltage repair limit is 2.0 volts for 7/8 inch diameter tubing at DCPP Units I and 2.
NOTE 2: The upper voltage repair limit is calculated according to the methodology in Generic Letter 95-05 as supplemented.
: 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 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 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-1 4797-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 tubesheet, the plugging (repair) limit is based on maintaining steam generator serviceability as 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 or repaired. 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 or repaired.
: 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-1 5573, 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 removed from service.

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. Ifcrack 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). Ifdefinitive 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 shall be inspected by bobbin coil probe during all future refueling outages.
Implementation of the steam generator tube/tube support plate repair criteria 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 during all future refueling outages.
Implementation of the W* repair criteria 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. 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 Yx" 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.                            I

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.

Enclosure 3 PG&E Letter DCL-06-061 Proposed Technical Specification Changes (retyped)
Remove Page                               Insert Page 1.1-4                                     1.1-4 3.4-27                                    3.4-27 3.4-28                                    3.4-28 3.4-38 3.4-39 5.0-10                                    5.0-10 5.0-11                                    5.0-11 5.0-11a                                   5.0-12 5.0-12                                    5.0-13 5.0-13                                    5.0-14 5.0-13a                                   5.0-15 5.0-14                                    5.0-16 5.0-15                                    5.0-17 5.0-16                                    5.0-18 5.0-17                                    5.0-19 5.0-18 5.0-19 5.0-29                                    5.0-29 5.0-30                                    5.0-30 5.0-30a                                   5.0-30a 5.0-30b                                   5.0-30b

Definitions 1.1 1.1 Definitions LEAKAGE                          3. Reactor Coolant System (RCS) LEAKAGE through (continued)                           a steam generator to the Secondary System (primary to secondary LEAKAGE).                          I
: 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, I
pipe wall, or vessel wall.
MASTER RELAY TEST           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.
MODE                        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.
OPERABLE-OPERABILITY        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               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.
(continued)
DIABLO CANYON - UNITS 1 & 2               1.1-4          Unit I - Amendment No. 1-35, Unit 2 - Amendment No. -45,

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. I gpm unidentified LEAKAGE;
: c. 10 gpm identified LEAKAGE; and                                       I
: d. 150 gallons per day primary to secondary LEAKAGE through any         I 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 hours                I 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 hours associated Completion            AND Time of Condition A not met.
OR Pressure boundary                B.2      Be in MODE 5.               36 hours 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. 1-35 442, Unit 2 - Amendment No. 435 442,

RCS Operational LEAKAGE 3.4.13 SURVEILLANCE REQUIREMENTS SURVEILLANCE                                    FREQUENCY SR 3.4.13.1                           NOTES-
: 1. Not required to be performed until 12 hours after establishment of steady state operation.
: 2. Not applicable to primary to secondary LEAKAGE.
Verify RCS operational LEAKAGE is within limits by      72 hours performance of RCS water inventory balance.
SR 3.4.13.2                          .NOTE ------
Not required to be performed until 12 hours after establishment of steady state operation.
Verify primary to secondary LEAKAGE is < 150           72 hours gallons per day through any one SG.
DIABLO CANYON - UNITS 1 & 2               3.4-28        Unit I - Amendment No. 4-*,
Unit 2 - Amendment No. 4-5,

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 I &#xa3;'*'It'*
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 hours associated Completion Time of Condition A not          AND met.                             B.2     Be in MODE 5.                  36 hours OR SG tube integrity not maintained.
DIABLO CANYON - UNITS I & 2                      3.4-38         Unit I - 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 Unit 1 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 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.
: 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. Leakage is not to exceed I gpm for the faulted SG, except for specific types of degradation at specific locations as described in paragraph c of the Steam Generator Program in which leakage from application of all alternate tube repair criteria degradation and non-alternate tube repair criteria degradation is not to exceed 10.5 gpm.
(continued)
DIABLO CANYON - UNITS 1 & 2                  5.0-10      Unit I - Amendment No.
Unit 2 - Amendment No.

Programs and Manuals 5.5 Unit 1 5.5 Programs and Manuals 5.5.9  Steam Generator (SG) Program (continued)
: 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 plugging limit is based on maintaining steam generator tube serviceability as 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 the lower voltage repair limit (NOTE 1), 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 the lower voltage repair limit (NOTE 1), will be repaired or 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 the lower voltage repair limit (NOTE 1) but less than or equal to the upper voltage repair limit (NOTE 2), may remain in service if a rotating pancake coil inspection 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 (NOTE 2) will be plugged or repaired.
: 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                 5.0-11     Unit I - Amendment No.
Unit 2 - Amendment No.

Programs and Manuals 5.5 Unit I 5.5 Programs and Manuals 5.5.9  Steam Generator (SG) Program (continued)                                                     I
: e.      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.l.b, and 5.5.9.c.1.c. The mid-cycle repair limits are determined from the following equations:
VSL VMURL    ---
VSL
                                                                   -  At) 1.0 + NDE + GrGr(CL  (CL VLRL)  (CL - At)
VMLRL = VMURL - (VuRL                       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)
(continued)
DIABLO CANYON - UNITS I & 2                5.0-12     Unit 1 - Amendment No.
Unit 2 - Amendment No.

Programs and Manuals 5.5 Unit 1 5.5 Programs and Manuals 5.5.9  Steam Generator (SG) Program (continued)
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.
NOTE 1: The lower voltage repair limit is 2.0 volts for 7/8 inch diameter tubing at DCPP Units I and 2.
NOTE 2: The upper voltage repair limit is calculated according to the methodology in Generic Letter 95-05 as supplemented.
: 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 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 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.
: 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                  5.0-13      Unit I - Amendment No.
Unit 2 - Amendment No.

Programs and Manuals 5.5 Unit 1 5.5 Programs and Manuals 5.5.9  Steam Generator (SG) Program (continued)
: e. Within the tubesheet, the plugging (repair) limit is based on maintaining steam generator serviceability as 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 or repaired. 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 or repaired.
: 6. Any type or combination of tube degradation below the flexible W* length is acceptable.
(continued)
DIABLO CANYON - UNITS 1 & 2                  5.0-14         Unit I - Amendment No.
Unit 2 - Amendment No.

Programs and Manuals 5.5 Unit 1 5.5 Programs and Manuals 5.5.9  Steam Generator (SG) Program (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-1 5573, 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 removed from service.
: 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                    5.0-15         Unit I - Amendment No.
Unit 2 - Amendment No.

Programs and Manuals 5.5 Unit 1 5.5 Programs and Manuals 5.5.9  Steam Generator (SG) Program (continued)
: 4. Indications left in service as a result of application of the tube support plate voltage-based repair criteria shall be inspected by bobbin coil probe during all future refueling outages.
Implementation of the steam generator tube/tube support plate repair criteria 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.
: 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 during all future refueling outages.
Implementation of the W* repair criteria 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. 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                    5.0-16          Unit 1 - Amendment No.
Unit 2 - Amendment No.

Programs and Manuals 5.5 Unit 1 5.5 Programs and Manuals 5.5.9  Steam Generator (SG) Program (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.
I DIABLO CANYON - UNITS I & 2                     5.0-17        Unit I - Amendment No.
Unit 2 - Amendment No.

Programs and Manuals 5.5 Unit I THIS PAGE NOT USED DIABLO CANYON - UNITS 1 & 2         5.0-18     Unit I - Amendment No.
Unit 2 - Amendment No.

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

Reporting Requirements 5.6 Unit 1 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 voltage-based repair criteria to tube support plate intersections, notify the NRC prior to returning the steam generators to service should any of the following arise:
: 1. If estimated leakage based on the projected end-of-cycle (or if not practical, using the actual measured end-of-cycle) voltage distribution, increased by estimated leakage by all other sources (alternate repair criteria and non-alternate repair criteria indications), exceeds the leak limit determined from the licensing basis dose calculation for the postulated main steamline break for the next operating cycle.
: 2. 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-of-cycle (or if not practical, using the actual measured end-of-cycle) voltage distribution exceeds 1 x le, notify the NRC and provide an assessment of the safety significance of the occurrence.
(continued)
DIABLO CANYON - UNITS I & 2                    5.0-29      Unit I - Amendment No. 435, 482, Unit 2 - Amendment No. 435, 484,

Reporting Requirements 5.6 Unit I 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 reported to the Commission pursuant to 10 CFR 50.4 within 90 days following return to service of the steam generators. This report shall include:
: 1.      Identification of W* tube indications and indications that do not meet W*
requirements and were plugged or repaired, 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 reported to the Commission pursuant to 10 CFR 50.4 within 90 days following return to service of the steam generators.
: e. For implementation of the repair criteria for axial PWSCC at dented TSPs, the NRC shall be notified prior to startup, pursuant to IOCFR50.72, of the following conditions that indicate a failure of performance criteria:
: 1.      The calculated SG probability of burst for condition monitoring exceeds x 10-2.
: 2.       The calculated SG leakage for condition monitoring from all sources (all alternate repair criteria and non-alternate repair criteria indications) exceeds the leakage limit determined from the licensing basis steam line break dose calculation.
(continued)
DIABLO CANYON - UNITS I & 2                 5.0-30    Unit I - Amendment No. 1435,454-I,452,482, Unit 2 - Amendment No. 435,444,4-2,484,

Reporting Requirements 5.6 Unit I 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 reported I
to the NRC within 120 days following completion of the inspection. The report will include:
: 1.       Tabulations of indications found in the inspection, tubes repaired, 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)
DIABLO CANYON - UNITS 1 & 2                5.0-30a      Unit I - Amendment No. 4-35,454,452,4-2, Unit 2 - Amendment No. 43&,44-1,4&2,484,

Reporting Requirements 5.6 Unit 1 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 reported to the Commission pursuant to 10 CFR 50.4 within 90 days following return to service of the steam generators:
: 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 I & 2            5.0-30b Unit I - Amendment No. 4-35,464,452,477,482, Unit 2 - Amendment No. 435,4-1-,42,1-47,4-84,

Enclosure 4 PG&E Letter DCL-06-061 Changes to Technical Specification Bases Pages

RCS Loops - MODES 1 and 2 B 3.4.4 BASES APPLICABLE        RCS flow, RCP rotor seizure, and RCP shaft break events. For each of SAFETY           these events, it is demonstrated that all the applicable safety criteria ANALYSES          are satisfied. For the remaining accident/safety analyses, operation of (continued)     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             sis f n~ PRABLE RCP for heat transport and the associ~ate:d 2SG, OPERABLE ,-F,:cod-a;c "-it~h t~he Steam, ,,r-ti Tube                  ,       ,,,with a water level within the limits specified in SR 3.4.5.2, except for operational transients. A I
RCP is OPERABLE if it is capable of being powered and is able to provide forced flow if required.
APPLICABILITY    In MODES I 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 hour per 8 hour 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 10&deg;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 n aGccordance with the Steam Generator Tube Su...,w.illa.e 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.a..nc      rd.anc
                                                .. .With tho Steam Gn*..ator Tub, Su'!eillanc Program, which has the minimum water level specified in SR 3.4.6.2.
I 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              The Note specifies that a RCP may be started if the pressurizer water (continued)      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 I
are able to provide flow if required. An OPERABLE 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 GenoRator Tube i
Surveillance Progr~am.
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 _>15%.
(continued)
DIABLO CANYON - UNITS 1 & 2                                                       Revision 3 31

RCS Operational LEAKAGE here are no changes to this page. Page                                       B 3.4.13 included for information only.
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 resulting in steam discharge to the atmosphere assumes a ,gpim Insert I         primnary to... Ocdary,LE=AK^AQ\.,E as thini*ti,.l 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.            oing available. The steam generator (SG) PORV fo             -that     has sustained the tube rupture is assumed to be-open for 30 minutes, at operator             w~hich time      Of"-' .........
                                                 -         below
                                                              .... he lift s     -, -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).
Iner 2,he safety analysis for RCS main loop piping for GDC-4 (Ref. 1)
Iner2assumes I gpm unidentified leakage 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 I & 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 Ss      to Secondary LEAKAGE through All Steam Generators*               t
                        *Total primary to secondary LEAKAGE amounting to I gpm through all SGs produces acceptable offsite doses in the SL13 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.         t.
                -'j       . Prima     to Secondary LEAKAGE through Any One SG The 150 gallons per day limit on one S is based on the assumption that a single crack leaking this amount would notion propagate to a SGTR under the stress conditions of a LOCA or a Insert 3     L-=     mainvery are    steam  lineand small,  rupture. Ifleaked the above      through ismany assumption                         o cracks, the cracks conservative.
I"  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 (con inued DIABLO CANYON - UNITS 1 & 2                       74                                    Revision 3

RCS Operational LEAKAGE B 3.4.13 BASES LCO              e. Primary to Secondary LEAKAGE through Any One SG (continued) cnOmbined with an effective leak rate monitoring prograpoie             1 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 250 C. 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.
I 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       ess than the minimum lesacb*i*.ial              detectable gamma emitter,     the activity  5.0*E7 microcuries/ml for leakageofrequirement of Specification 3.4.13. may be considered met.
ACTIONS          A.1 Unidentified LEAKAGEk,identified LEAKAG ,.or primay to secondary LEAKAGE in excess of the LCO limits must be reduced to within limits within 4 hours. This Completion Time allows time to verify leakage I
(continued)
DIABLO CANYON - UNITS I & 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 LEAKD GE or reduce LEAKAGE to within limits before the reactor must be shu down. This action is necessary to prevent further deterioration c the RCPB.
B.1 and B.2 If any pressure boundary LEAKAGE exists 'or if unidentified-L-,AKAG identified LEAKAGE, or pF-*ia.; to.      .,,nday LEAKAGE cannot be or reduced to within limits within 4 hours, 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 hours and MODE 5 within 36 hours. This action reduces the LEAKAGE and also reduces the factors that tend to degrade the pressure boundary.
The allowed Completion Times are reasonable, based 6n 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 determinedb
                         /oerormnce of an RCS water inventor bslanre-,f-Primary to secondary*
                    /"*LEAKAGE is also measured by performance of an RCS water inventory L.,balance in conjunction with effluent monitoring within the secondary             /
                     *,,team and feedae The RCS water inventory balance must be met with the reactor at steady state operating conditions (stable temperatur~e, power le            states pressurizer level, makeup and letdown, and RCP e               ion and      st return flowslr--herefere-a Note *i added allowingthat this SR is not The                 requir&#xfd;o be performed until 12 hours after establishing steady state Surveillance is           ration. The 12 hour 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 I & 2                      76                                    Revision 3

RCS Operational LEAKAGE B 3.4.13 BASES SURVEILLANCE       SR 3.4.13.1 (continued)
REQUIREMENTS       operational LEAKAGE determination by water inventory balance, steady state is defined as stable RCS pressure, temperature (Tavg changes less than 5&deg;F 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 hour Frequency is a reasonable interval to trend LEAKAGE and recognizes the importance of early leakage detection in the prevention of accidents. The 12 hour Frequency after steady state operation has been achieved provides for those situations following a transient such that the 72 hours plus extension allowed by SR 3.0.2 would be exceeded. Under these circumstances, the SR would be due within 12 hours after steady state operation has been reestablished and every 72 hours thereafter during steady state operation.
                *"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
                \considered         conditions; failure to meettherefore failure the integrity    to meet goals of thethis LCO surveillance  is ll and LCO 3.0.3" SR 3,4.1A            2 eaThis leakageSR once provides  a requirement every              to determine 72 hours while operating inprimary-to-secondary MODES 1, 2, 3, or 4.
* During normal operation this will be done using a correlation of Insert 5  -     radioactivity at the steam jet air ejectors. During periods of significant I             Ssource 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 hour 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 rupture.
(continued)
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-1061, Volume 3, November, 1984.
: 6. 10CFR100.
I Insert 6 DIABLO CANYON - UNITS I & 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&deg;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 hours after a power change >_15% RTP within a 1 hour 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 E determination 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 Efor 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 E does not change rapidly.
This SR has been modified by a Note that indicates sampling for 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 hours. 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)                                                                   I DIABLO CANYON - UNITS 1 & 2                  97                                         Revision 3

Technical Specification Bases Inserts Insert 1 that prior to the accident there is I gpm total primary to secondary LEAKAGE to all steam generators (SGs) as an initial condition. 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 safety analysis.
Insert 2 The safety analyses for the SLB accident (Ref. 3) assumes that prior to the accident there is 1 gpm total primary to secondary LEAKAGE to all SGs as an initial condition, which bounds the allowable operational LEAKAGE rate limits in LCO 3.4.13.d. The dose consequences resulting from the SLB accident are within a small fraction of the limits defined by 10 CFR 100.
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. Ifthis 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. Ifit is not practical to assign the LEAKAGE to an individual SG, all the primary to secondary LEAKAGE should be conservatively assumed to be from one SG.
The Surveillance is modified by a Note which states that the Surveillance is not required to be performed until 12 hours after establishment of steady state operation. For RCS primary to secondary LEAKAGE determination, steady state is defined as stable RCS pressure,

Technical Specification Bases Inserts (continued) temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows.
The Surveillance Frequency of 72 hours 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.

Technical Specification Bases Inserts (continued)           SG Tube Integrity B 3.4.17 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 system6 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  The steam generator tube rupture (SGTR) accident is the limiting SAFETY      design basis event for SG tubes and avoiding an SGTR is the basis for ANALYSES    this Specification. The analysis of a SGTR event assumes that prior to the accident there is 1 gpm total primary to secondary LEAKAGE to all SGs as an initial condition, which bounds the allowable operational LEAKAGE rate limits in LCO 3.4.13, "RCS Operational LEAKAGE".
The analysis assumes following the accident, there is 150 gpd leakage to each of the three intact SGs, and there is accident induced leakage to the ruptured SG 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.) In these analyses, the steam discharge to the atmosphere is based on an assumed 1 gpm total primary to secondary LEAKAGE to all SGs as an initial condition prior to the accident. For accidents that do not involve fuel damage, the primary coolant activity level of DOSE EQUIVALENT 1-131 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).
LCO         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.
(continued)

Technical Specification Bases Inserts (continued)
BASES LCO            During an SG inspection, any inspected tube that satisfies the Steam (continued)   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.
In the context of this Specification, a SG tube is defined as the entire 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.
(continued)

Technical Specification Bases Inserts (continued)
BASES LCO            For circumferential degradation, the classification of axial thermal loads (continued)  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.
Structural integrity requires that the primary membrane stress intensity 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 that the primary to secondary LEAKAGE caused by a design basis accident, other than a SGTR, is within the accident analysis assumptions. The main steam line break accident analysis assumes that accident induced leakage following the accident does not exceed 10.5 gpm to the faulted SG, except for specific types of degradation at specific locations as described in paragraph 5.5.9.c of the Steam Generator Program in which leakage from application of all alternate tube repair criteria degradation and non-alternate tube repair criteria degradation is not to exceed 10.5 gpm. To support the accident induced leakage performance criteria, the non-alternate repair criteria degradation accident induced leakage following the accident in the faulted SG is conservatively limited to 1 gpm at hot conditions.
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.
Ifthe 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 hours and MODE 5 within 36 hours.
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."}}