View source for SBK-L-13030, LAR 13-02 - Application to Revise Technical Specifications to Adopt TSTF-510, Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection, Using the Consolidated Line Item Improvement Process

{{Adams
| number = ML13093A151
| issue date = 03/27/2013
| title = LAR 13-02 - Application to Revise Technical Specifications to Adopt TSTF-510, Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection, Using the Consolidated Line Item Improvement Process
| author name = Walsh K
| author affiliation = NextEra Energy Seabrook, LLC
| addressee name = 
| addressee affiliation = NRC/Document Control Desk, NRC/NRR
| docket = 05000443
| license number = 
| contact person = 
| case reference number = SBK-L-13030
| document type = Letter type:SBK, License-Application for Facility Operating License (Amend/Renewal) DKT 50, Technical Specifications
| page count = 31
| project = 
| stage = Request
}}

=Text=
{{#Wiki_filter:NEXTera ENERGY6_
March 27, 2013 10 CFR 50.90 SBK-L-13030 Docket No. 50-443 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Seabrook Station License Amendment Request 13-02 Application to Revise Technical Specifications to Adopt TSTF-5 10, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection," Using the Consolidated Line Item Improvement Process Pursuant to 10 CFR 50.90, NextEra Energy Seabrook, LLC is submitting a request for an amendment to the Technical Specifications (TS) for Seabrook Station. The proposed amendment would modify TS requirements regarding steam generator tube inspections and reporting as described in TSTF-5 10, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection."
Attachment 1 provides a description and assessment of the proposed changes, the requested confirmation of applicability, and plant-specific verifications. Attachment 2 provides the existing TS pages marked up to show the proposed changes. Attachment 3 provides revised (clean) TS pages. Attachment 4 provides existing TS Bases pages marked up to show the proposed changes. The changes to the TS Bases are provided for information only and will be implemented in accordance with TS 6.7.6.j, TS Bases Control Program, upon implementation of the license amendment.
Approval of the proposed amendment is requested by March 30, 2014. Once approved, the amendment shall be implemented within 60 days.
In accordance with 10 CFR 50.91, a copy of this application with attachments is being provided to the New Hampshire State Liaison Officer.
NextEra Energy Seabrook, LLC, P.O. Box 300, Lafayette Road, Seabrook, NH 03874

United States Nticlear Regulatory Commission SBK-L-13030/ Page 2 Should you have any questions regarding this letter, please contact Mr. Michael O'Keefe, Licensing Manager, at (603) 773-7745.
I declare under penalty of perjury that the foregoing is true and correct..
Executed on     Ao6l/3,'/   ,.
Sincerely, Kevin T. Walsh Site Vice President NextEra Energy Seabrook, LLC Attachments: 1.      Description and Assessment
: 2. Proposed Technical Specification Changes (Mark-Up)
: 3. Revised Technical Specification Pages
: 4. Proposed Technical Specification Bases Changes (Mark-Up) cc:     NRC Region I Administrator J. G. Lamb, NRC Project Manager, Project Directorate 1-2 NRC Senior Resident Inspector Perry E. Plummer, Acting Director, Homeland Security and Emergency Management New Hampshire Department of Safety Division of Homeland Security and Emergency Management Bureau of Emergency Management 33 Hazen Drive Concord, NH 03305 John Giarrusso, Jr., Nuclear Preparedness Manager The Commonwealth of Massachusetts Emergency Management Agency 400 Worcester Road Framingham, MA 01702-5399

Attachment 1 Description and Assessment

DESCRIPTION AND ASSESSMENT

==1.0       DESCRIPTION==

The proposed change revises Specifications 3.4.5, "Steam Generators;" 6.7.6.k, "Steam Generator (SG) Program;" and 6.8.1.7, "Steam Generator Tube Inspection Report." The proposed changes are needed to address implementation issues associated with the inspection periods, and address other administrative changes and clarifications. The proposed amendment is consistent with TSTF-5 10, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection."
2.0      ASSESSMENT 2.1 Applicability of Published Safety Evaluation NextEra Energy Seabrook, LLC (NextEra) has reviewed TSTF-5 10, Revision 2, "Revision to Steam Generator Program Inspection Frequencies and Tube Sample Selection," and the model safety evaluation dated October 27, 2011 as part of the Federal Register Notice for Comment. As described in the subsequent paragraphs, NextEra has concluded that the justifications presented in TSTF-5 10 and the model safety evaluation prepared by the NRC staff are applicable to Seabrook Station and justify this amendment for the incorporation of the changes to the Seabrook TS.
2.2 Optional Changes and Variations NextEra is proposing the following administrative variations from the TS changes described in TSTF-5 10, Revision 2, or the applicable parts of the NRC staff s model safety evaluation dated October 27, 2011.
    " The Seabrook TS utilize different numbering and titles than the Standard Technical Specifications on which TSTF-5 10 was based. Specifically, Seabrook TS 3.4.5, "Steam Generators," is equivalent to TS 3.4.20, "Steam Generator (SG) Tube Integrity," in TSTF-510. Seabrook TS 6.7.6.k contains the Steam Generator (SG) Program, which is designated TS 5.5.9 in TSTF-5 10; and Seabrook TS 6.8.1.7, "Steam Generator Tube Inspection Report," is equivalent to TS 5.6.7 in TSTF-5 10. These differences are administrative and do not affect the applicability of TSTF-510 to the Seabrook TS.
* The proposed change corrects an administrative inconsistency in TSTF-5 10, Paragraph d.2 of the Steam Generator Tube Inspection Program. In Section 2.0, "Proposed Change," TSTF-5 10 states that references to "tube repair criteria" in Paragraph d are revised to "tube plugging [or repair] criteria." However, in the following sentence in Paragraph d.2, this change was inadvertently omitted, "If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location 1

and that may satisfy the applicable tube repair criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated" (Emphasis added). NextEra does not have an approved tube repair criteria; therefore, the sentence is revised to state "tube plugging" criteria.

==3.0     REGULATORY ANALYSIS==

3.1 No Significant Hazards Consideration Determination NextEra requests adoption of an approved change to the standard technical specifications (STS) into the plant specific technical specifications (TS), to revise Specifications 6.7.6.k, "Steam Generator (SG) Program;" 6.8.1.7, "Steam Generator Tube Inspection Report;" and LCO 3.4.5, "Steam Generators," to address inspection periods and other administrative changes and clarifications.
As required by 10 CFR 50.91 (a), an analysis of the issue of no significant hazards consideration is presented below:
: 1. Does the proposed change involve a significant increase in the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change revises the Steam Generator (SG) Program to modify the frequency of verification of SG tube integrity and SG tube sample selection. A steam generator tube rupture (SGTR) event is one of the design basis accidents that is analyzed as part of a plant's licensing basis. The proposed SG tube inspection frequency and sample selection criteria will continue to ensure that the SG tubes are inspected such that the probability of a SGTR is not increased. The consequences of a SGTR are bounded by the conservative assumptions in the design basis accident analysis. The proposed change will not cause the consequences of a SGTR to exceed those assumptions. Therefore, it is concluded that this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.
: 2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?
Response: No.
The proposed changes to the Steam Generator Program will not introduce any adverse changes to the plant design basis or postulated accidents resulting from potential tube degradation. The proposed change does not affect the design of the SGs or their method of operation. In addition, the proposed change does not impact any other plant system or component.
2

Therefore, it is concluded that this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.
: 3. Does the proposed change involve a significant reduction in a margin of safety?
Response: No.
The SG tubes in pressurized water reactors are an integral part of the reactor coolant pressure boundary and, as such, are relied upon to maintain the primary system's pressure and inventory. As part of the reactor coolant pressure boundary, the SG tubes are unique in that they are also relied upon as a heat transfer surface between the primary and secondary systems such that residual heat can be removed from the primary system. In addition, the SG tubes also isolate the radioactive fission products in the primary coolant from the secondary system. In summary, the safety function of a SG is maintained by ensuring the integrity of its tubes.
Steam generator tube integrity is a function of the design, environment, and the physical condition of the tube. The proposed change does not affect tube design or operating environment. The proposed change will continue to require monitoring of the physical condition of the SG tubes such that there will not be a reduction in the margin of safety compared to the current requirements.
Therefore, it is concluded that the proposed change does not involve a significant reduction in a margin of safety.
Based on the above, NextEra concludes that the proposed change presents no significant hazards consideration under the standards set forth in 10 .CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.
4.0     ENVIRONMENTAL EVALUATION The proposed change would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed change does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed change meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9).
Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed change.
3

Attachment 2 Proposed Technical Specification Changes (Mark-Up)
The attached markups reflect the currently issued version of the TS and Facility Operating License. At the time of submittal, the Facility Operating License was revised through Amendment No. 132.
Listed below are the license amendment requests that are awaiting NRC approval and may impact the currently issued version of the Facility Operating License affected by this LAR.
LAR       Title                                                      NextEra Energy           Date Seabrook Letter          Submitted 10-02    Application for Change to the Technical Specifications     SBK-L-10074              05/14/2010 for the Containment Enclosure Emergency Air Cleanup System Changes to the Technical Specifications for New and 11-04    Spent Fuel Storage                                         SBK-L-1 1245             01/30/2012 Application to Revise the Applicability of the Reactor 11-06    Coolant System Pressure - Temperature Limits and the       SBK-L- 11186             11/17/2011 Cold Overpressure Protection Setpoints Application for Technical Specification Improvement 12-06    to Extend the Inspection Interval for Reactor Coolant      SBK-L-12235              12/20/2012 Pump Flywheels Using the Consolidated Line Item Improvement Process The following TS pages are included in the attached markup:
Technical                                       Title                                    Page Specification TS 3.4.5                           Steam Generators                              3/3 4-13 TS 6.7.6.k                   Steam Generator (SG) Program                            6-11 6-12 6-13 6-14 TS 6.8.1.7               Steam Generator Tube Inspection Report                      6-21 6-21a

REACTOR COOLANT SYSTEM 3/4.4.5    STEAM GENERATORS LIMITING CONDITION FOR OPERATION 3.4.5      SG tube integrity shall be maintained AND All SG tubes satisfying the tub r.                      criteria shall be plugged in ccordance with the Steam Generator Program.
APPLICAlBILITY: MODES 1, 2, 3, and 4.
ACTION:
k IfV"'P "
               ----------------------------------------- 4vJ     r------------------------------------------
Separate action entry is allowed for each steam generator tube.
: a. With one or more steam generator tubes satisfying the tube                                      cri .ria and not plugged in accordance with the Steam Generator Program,
: 1.        Within 7 days, verify tube integrity of the affected tube(s) is ma ntained until the next refueling outage or SG tube inspection, or be in &#xfd;OT STANDBY within 6 hours and COLD SHUTDOWN within the ext 30 hours, and
: 2.        Plug the affected tube(s) in accordance with the Steam Gen rator Program prior to entering HOT SHUTDOWN following the n xt refueling outage or steam generator tube inspection.
: b. With steam generator tube integrity not maintained, be in HOT                                  ANDBY within 6 hours and in COLD SHUTDOWN within the next 30 hours.
SURVEILLANCE REQUIREMENTS 4.4.5.1    Verify SG tube integrity in accordance with the Steam Gener or Program.
4.4.5.2Verify that each inspected SG tube that satisfies the tub                                    riteria is plugged in K
accordance with the Steam Generator Program prior to entering HOT SHUTDOWN following a SG tube inspection.
SEABROOK UNIT - 1                                       3/4 4-13                                Admendment No. q15

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS 6.7.6 (Continued)
: j. Technical Specification (TS) Bases Control Program This program provides a means for processing changes to the Bases of these Technical Specifications.
: a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.
: b. Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:
: 1. A change in the TS incorporated in the license or
: 2. A change to the updated FSAR (UFSAR) or Bases that requires NRC approval pursuant to 10 CFR 50.59.
: c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the UFSAR.
: d. Proposed changes that meet the criteria of Specification 6.7.6j.b above shall be reviewed and approved by the NRC prior to implementation.
Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).
: k. 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 followingS:
: 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.
SEABROOK - UNIT 1                        6-11       Amendment No. 34, 55, 67, 88, 104,44-

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS 6.7.6 (Continued)
: 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 dow nd all anticipated transients included in the design specificatiorw 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 1 gpm total or 500 gpd through any one SG.
: 3. The operational LEAKAGE performance criterion is specified in LCO 3.4.6.2, "Reactor Coolant System Operational Leakage."
: c. Provisions for SG tub           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.
SEABROOK - UNIT 1                           6-12        Amendment No. 34, 39, 101, 109, ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS 6.7.6 (CContinued)
The following alternate tub r           criteria shall be applied as an alternative to the 40% depth based criteria:
Tubes with service-induced flaws located greater than 15.21 inches below the top of the tubesheet do not require plugging. Tubes with service-induced flaws located in the portion of the tube from the top of the tubesheet to 15.21 inches below the top of the tubesheet shall be plugged upon detection.
: 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 tub criteria. The portion of the tube below 15.21 inches from the top of the tubesheet is excluded from this requirement. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 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. ,0           ,c   ,, 4 assessment            6r    i) 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 SGr             e       cIZi?&#xfd; of 120, 90, and, thereafter*,
Ilnspec 100% of the tubes at sequential periods
: 2.                                                   sequential .d"     shall be 60 effective full p;*6 months. Thh.
cons&#xfd;!id r *L* begin after the 5       inservice insp .etron of the SGs. In ad " &#xfd;ins c 0                   ues by th pge~ling outage near sd e g50% by the refueli      outage /
i ippi' oftep              adtermai*
nearest the end of the period. N- ,G shall operate for           re than 48 effective full power  months or two refueling outages (whichever is less) wi*thout being inspected.
SEABROOK - UNIT 1                       6-13              Amendment No. 34, 104, 109, 115, 123,-414.-

INSERT d.2 After the first refueling outage following SG installation, inspect each SG at least every 48 effective full power months or at least every other refueling outage (whichever results in more frequent inspections). In addition, the minimum number of tubes inspected at each scheduled inspection shall be the number of tubes in all SGs divided by the number of SG inspection outages scheduled in each inspection period as defined in a, b, and c below. If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube plugging criteria, the minimum number of locations inspected with such a capable inspection technique during the remainder of the inspection period may be prorated. The fraction of locations to be inspected for this potential type of degradation at this location at the end of the inspection period shall be no less than the ratio of the number of times the SG is scheduled to be inspected in the inspection period after the determination that a new form of degradation could potentially be occurring at this location divided by the total number of times the SG is scheduled to be inspected in the inspection period. Each inspection period defined below may be extended up to 3 effective full power months to include a SG inspection outage in an inspection period and the subsequent inspection period begins at the conclusion of the included SG inspection outage.
a) After the first refueling outage following SG installation, inspect 100% of the tubes during the next 120 effective full power months. This constitutes the first inspection period; b) During the next 96 effective full power months, inspect 100% of the tubes. This constitutes the second inspection period; and c) During the life of the SGs, inspect 100 % of the tubes every 72 effective full power months. This constitutes the third and subsequent inspection periods.

ADMINISTRATIVE CONTROLS PROCEDURES AN D PROGRAMS                                                                o&
6.7.6 (Continued)
: 3. If crack indications are found in portions of the S tube not excluded above, 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 i K ). If definitive information, such as from i n of a pulled tube, diagnostic non-destructive testing, or eng~ineering evaluation indicates that a crack-like indication is not sscassociated with a crack(s), then the indication need not be treated as a crack.
: e. Provisions for monitoring operational primary to secondary leakage.
I. Control Room Envelope Habitability Program A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Emergency Makeup Air and Filtration System (CREMAFS), CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem total effective dose equivalent (TEDE) for the duration of the accident. The program shall include the following elements:
: a. The definition of the CRE and the CRE boundary.
: b. Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance.
: c. Requirements for (i) determining the unfiltered air in-leakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," Revision 0, May 2003, and (ii)assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.
SEABROOK - UNIT 1                           6-14        Amendment No. 34, 78, 104, 115, 119,42S.

ADMINISTRATIVE CONTROLS 6.8.1.6.c  The core operating limits shall be determined so that all applicable limits (e.g.,
fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as SHUTDOWN MARGIN, and transient and accident analysis limits) of the safety analysis are met. The CORE OPERATING LIMITS REPORT for each reload cycle, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, to the NRC Document Control Desk with copies to the Regional Administrator and the Resident Inspector.
STEAM GENERATOR TUBE INSPECTION REPORT 6.8.1.7    A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with Specification 6.7.6.k, Steam Generator (SG) Program. The report shall include:
: a. The scope of inspections performed on each SG,
: b.  (&#xfd;,M       egradation mechanisms found,
: c. Nondestructive examination techniques utilized for each degradation mechanism,
: d. Location, orientation (if linear), and measured sizes (if available) of service induced indications,
: e. Number of tubes plugged during the inspection outage for eachl           E degradation mechanism, number and percent ge of tubes plujged tdate, c7,7   zt
: g. The results of condition monitoring, including the results of tube pulls and in-situ testing, The         ive ply        erce ntapl         ugg inj~cb*S6*
The primary to secondary leakage rate observed in each SG (if it is not practical to assign the leakage to an individual SG, the entire primary to secondary leakage should be conservatively assumed to be from one SG) during the cycle preceding the inspection which is the subject of the report, J            The calculated accident induced leakage rate from the portion of the tubes below 15.21 inches from the top of the tubesheet for the most limiting accident in the most limiting SG. In addition, if the calculated accident induced leakage rate from the most limiting accident is less than 2.49 times the maximum operational primary to secondary leakage rate, the report should describe how it was determined, and SEABROOK - UNIT I                        6-21      Amendment No. 22, 66, 88, 104, 107, 115, .23,-4,*'

ADMINISTRATIVE CONTROLS 6.8.1.7       (Continued)
        ~     The results of monitoring for tube axial displacement (slippage). If slippage is discovered, the implications of the discovery and corrective action shall be provided.
SPECIAL REPORTS 6.8.2    Special reports shall be submitted to the U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, Attn: Document Control Desk, with a copy to the NRC Regional Administrator within the time period specified for each report.
6.9      (THIS SPECIFICATION NUMBER IS NOT USED)
SEABROOK - UNIT I                           6-21 a                  Amendment No. 423,43+

Attachment 3 Revised Technical Specification Pages

REACTOR COOLANT SYSTEM 3/4.4.5    STEAM GENERATORS LIMITING CONDITION FOR OPERATION 3.4.5      SG tube integrity shall be maintained AND All SG tubes satisfying the tube plugging criteria shall be plugged in accordance with the Steam Generator Program.
APPLICABILITY: MODES 1, 2, 3, and 4.
ACTION:
               ---------------------------              NOTE            ----------------------
Separate action entry is allowed for each steam generator tube.
: a. With one or more steam generator tubes satisfying the tube plugging criteria and not plugged in accordance with the Steam Generator Program,
: 1.       Within 7 days, verify tube integrity of the affected tube(s) is maintained until the next refueling outage or SG tube inspection, or be in HOT STANDBY within 6 hours and COLD SHUTDOWN within the next 30 hours, and
: 2.        Plug the affected tube(s) in accordance with the Steam Generator Program prior to entering HOT SHUTDOWN following the next refueling outage or steam generator tube inspection.
: b. With steam generator tube integrity not maintained, be in HOT STANDBY within 6 hours and in COLD SHUTDOWN within the next 30 hours.
SURVEILLANCE REQUIREMENTS 4.4.5.1        Verify SG tube integrity in accordance with the Steam Generator Program.
4.4.5.2        Verify that each inspected SG tube that satisfies the tube plugging criteria is plugged in accordance with the Steam Generator Program prior to entering HOT SHUTDOWN following a SG tube inspection.
SEABROOK UNIT- 1                                       3/4 4-13                                 Amendment No. 44-5,

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS 6.7.6 (Continued)
: j. Technical Specification (TS) Bases Control Proqram This program provides a means for processing changes to the Bases of these Technical Specifications.
: a. Changes to the Bases of the TS shall be made under appropriate administrative controls and reviews.
: b. Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:
: 1. A change in the TS incorporated in the license or
: 2. A change to the updated FSAR (UFSAR) or Bases that requires NRC approval pursuant to 10 CFR 50.59.
: c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the UFSAR.
: d. Proposed changes that meet the criteria of Specification 6.7.6j.b above shall be reviewed and approved by the NRC prior to implementation. Changes to the Bases implemented without prior NRC approval shall be provided to the NRC on a frequency consistent with 10 CFR 50.71(e).
: k. Steam Generator (SG) Progqram 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:
: 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.
SEABROOK - UNIT 1                         6-11      Amendment No. 34, 55, 67, 88, 104, 115,

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS 6.7.6 (Continued)
: 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 1 gpm total or 500 gpd through any one SG.
: 3.       The operational LEAKAGE performance criterion is specified in LCO 3.4.6.2, "Reactor Coolant System Operational Leakage."
: c. Provisions for SG tube plugging 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.
SEABROOK - UNIT 1                          6-12        Amendment No. 34, 39, 104, 109, 115,

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS 6.7.6 (Continued)
The following alternate tube plugging criteria shall be applied as an alternative to the 40% depth based criteria:
Tubes with service-induced flaws located greater than 15.21 inches below the top of the tubesheet do not require plugging.
Tubes with service-induced flaws located in the portion of the tube from the top of the tubesheet to 15.21 inches below the top of the tubesheet shall be plugged upon detection.
: 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 plugging criteria. The portion of the tube below 15.21 inches from the top of the tubesheet is excluded from this requirement. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 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. A degradation assessment 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 installation.
: 2.      After the first refueling outage following SG installation, inspect each SG at least every 48 effective full power months or at least every other refueling outage (whichever results in more frequent inspections). In addition, the minimum number of tubes inspected at each scheduled inspection shall be the number of tubes in all SGs divided by the number of SG inspection outages scheduled in each inspection period as defined in a, b, and c below. If a degradation assessment indicates the potential for a type of degradation to occur at a location not previously inspected with a technique capable of detecting this type of degradation at this location and that may satisfy the applicable tube plugging criteria, the minimum number of locations inspected with such a capable inspection SEABROOK - UNIT 1                   6-13        Amendment No. 34, 104, 109, 115, 123, 131,

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS 6.7.6 (Continued) technique during the remainder of the inspection period may be prorated. The fraction of locations to be inspected for this potential type of degradation at this location at the end of the inspection period shall be no less than the ratio of the-number of times the SG is scheduled to be inspected in the inspection period after the determination that a new form of degradation could potentially be occurring at this location divided by the total number of times the SG is scheduled to be inspected in the inspection period. Each inspection period defined below may be extended up to 3 effective full power months to include a SG inspection outage in an inspection period and the subsequent inspection period begins at the conclusion of the included SG inspection outage.
a)     After the first refueling outage following SG installation, inspect 100% of the tubes during the next 120 effective full power months. This constitutes the first inspection period; b)     During the next 96 effective full power months, inspect 100% of the tubes. This constitutes the second inspection period; and c)      During the life of the SGs, inspect 100 % of the tubes every 72 effective full power months. This constitutes the third and subsequent inspection periods.
: 3. If crack indications are found in portions of the SG tube not excluded above, then the next inspection for each affected and potentially affected SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever results in more frequent inspections). 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.
: e. Provisions for monitoring operational primary to secondary leakage.
SEABROOK - UNIT 1                      6-14      Amendment No. 34, 78, 101, 115, 119, 123,

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS 6.7.6 (Continued)
: 1. Control Room Envelope Habitability Program A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Emergency Makeup Air and Filtration System (CREMAFS), CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem total effective dose equivalent (TEDE) for the duration of the accident. The program shall include the following elements:
: a.      The definition of the CRE and the CRE boundary.
: b.       Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance.
: c.        Requirements for (i) determining the unfiltered air in-leakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision 0.
: d.       Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation by one train of the CREMAFS, operating at a flow rate of less than or equal to 600 CFM at a Frequency of 18 months on a STAGGERED TEST BASIS. The results shall be trended and used as part of the 18 month assessment of the CRE boundary.
SEABROOK - UNIT 1                              6-14a                 Amendment No. 149-,426,

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS 6.7.6 (Continued)
: e. The quantitative limits on unfiltered air in-leakage into the CRE.
These limits shall be stated in a manner to allow direct comparison to the unfiltered air in-leakage measured by the testing described in paragraph c. The unfiltered air in-leakage limit for radiological challenges is the in-leakage flow rate assumed in the licensing basis analyses of DBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis.
: f.      The provisions of SR 4.0.2 are applicable to the Frequencies for assessing CRE habitability, determining CRE unfiltered in-leakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c and d, respectively.
: m. Reactor Coolant Pump Flywheel Inspection Program In addition to the requirements of Specification 4.0.5, each reactor coolant pump flywheel shall be inspected at least once every 10 years. This inspection shall be by either of the following examinations:
: a.      An in-place examination, utilizing ultrasonic testing, over the volume from the inner bore of the flywheel to the circle of one-half the outer radius; or
: b.      A surface examination, utilizing magnetic particle testing and/or penetrant testing, of the exposed surfaces of the disassembled flywheel.
SEABROOK - UNIT 1                             6-14b                      Amendment No. 44-9,

ADMINISTRATIVE CONTROLS PROCEDURES AND PROGRAMS 6.8       REPORTING REQUIREMENTS ROUTINE REPORTS 6.8.1     In addition to the applicable reporting requirements of Title 10, Code of Federal Regulations, the following reports shall be submitted to the Regional Administrator of the Regional Office of the NRC unless otherwise noted.
STARTUP REPORT 6.8.1.1     A summary report of station startup and power escalation testing shall be submitted following: (1) receipt of an Operating License, (2) amendment to the license involving a planned increase in power level, (3) installation of fuel that has a different design or has been manufactured by a different fuel supplier, and (4) modifications that may have significantly altered the nuclear, thermal, or hydraulic performance of the station.
The Startup Report shall address each of the tests identified in the Final Safety Analysis Report and shall include a description of the measured values of the operating conditions or characteristics obtained during the test program and a comparison of these values with design predictions and specifications. Any corrective actions that were required to obtain satisfactory operation shall also be described. Any additional specific details required in license conditions based on other commitments shall be included in this report.
Startup Reports shall be submitted within: (1) 90 days following completion of the Startup Test Program, (2) 90 days following resumption or commencement of commercial power operation, or (3) 9 months following initial criticality, whichever is earliest. If the Startup Report does not cover all three events (i.e., initial criticality, completion of Startup Test Program, and resumption or commencement of commercial operation),
supplementary reports shall be submitted at least every 3 months until all three events have been completed.
SEABROOK - UNIT 1                               6-14c                             Amendment No. I

ADMINISTRATIVE CONTROLS 6.8.1.6.c  The core operating limits shall be determined so that all applicable limits (e.g.,
fuel thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as SHUTDOWN MARGIN, and transient and accident analysis limits) of the safety analysis are met. The CORE OPERATING LIMITS REPORT for each reload cycle, including any mid-cycle revisions or supplements thereto, shall be provided upon issuance, to the NRC Document Control Desk with copies to the Regional Administrator and the Resident Inspector.
STEAM GENERATOR TUBE INSPECTION REPORT 6.8.1.7    A report shall be submitted within 180 days after the initial entry into MODE 4 following completion of an inspection performed in accordance with Specification 6.7.6.k, Steam Generator (SG) Program. The report shall include:
: a.   *The scope of inspections performed on each SG,
: b. Degradation mechanisms found,
: c. Nondestructive examination techniques utilized for each degradation mechanism,
: d. Location, orientation (if linear), and measured sizes (if available) of service induced indications,
: e. Number of tubes plugged during the inspection outage for each degradation mechanism,
: f. The number and percentage of tubes plugged to date, and the effective plugging percentage in each steam generator,
: g. The results of condition monitoring, including the results of tube pulls and in-situ testing,
: h. The primary to secondary leakage rate observed in each SG (if it is not practical to assign the leakage to an individual SG, the entire primary to secondary leakage should be conservatively assumed to be from one SG) during the cycle preceding the inspection which is the subject of the report, The calculated accident induced leakage rate from the portion of the tubes below 15.21 inches from the top of the tubesheet for the most limiting accident in the most limiting SG. In addition, if the calculated accident induced leakage rate from the most limiting accident is less than 2.49 times the maximum operational primary to secondary leakage rate, the report should describe how it was determined, and SEABROOK - UNIT I                6-21        Amendment No. 22, 66, 88, 104, 107, 115, 123, 131,

ADMINISTRATIVE CONTROLS 6.8.1.7         (Continued)
: j.      The results of monitoring for tube axial displacement (slippage). If slippage is discovered, the implications of the discovery and corrective action shall be provided.
SPECIAL REPORTS 6.8.2      Special reports shall be submitted to the U.S. Nuclear Regulatory Commission, Washington, D.C. 20555, Attn: Document Control Desk, with a copy to the NRC Regional Administrator within the time period specified for each report.
6.9        (THIS SPECIFICATION NUMBER IS NOT USED)
SEABROOK - UNIT I                             6-21 a                 Amendment No. 123, 131,

Attachment 4 Revised Technical Specification Bases Changes (Mark-Up) 1

REACTOR COOLANT SYSTEM BASES 3/4.4.5      STEAM GENERATORS (SG) TUBE INTEGRITY (Continued) rupture of a SG tube that relieves to the lower pressure secondary system. The analysis assumes that contaminated fluid is released to the atmosphere through the main steam safety valves or the atmospheric steam dump valves.
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 the total primary-to secondary leakage from all SGs of 1 gallon per minute and 500 gallons per day from any one SG or is assumed to increase to these values as a result of accident induced conditions. 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 3.4.8, "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 50.67 (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        i criteria be lugged in accordance with the Steam Generator Program.                   ,        101r    &#xfd;7 During a      i pection, any inspected tube that satisfies the Steam Generator Program
* t *c ria is removed from service by plugging. If a tube was determined to satisfy th          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 6.7.6.k, "Steam Generator (SG) 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.
SEABROOK - UNIT 1                                 B 3/4 4-5      BC-07 -01, 0-02, 09 04, 12-05

REACTOR COOLANT SYSTEM BASES 3/4.4.5      STEAM GENERATORS (SG) TUBE INTEGRITY (Continued)
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 MODES 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.
ACTIONS The ACTIONS are modified by a Note clarifying that the actions may be entered independently for each SG tube. This is acceptable because the actions provide appropriate compensatory actions for each affected SG tube. Complying with the actions may allow for continued operation, and subsequent affected SG tubes are governed by subsequent entry and application of associated actions.
a and b Action a applies if it is discovered t      ne or more SG tubes examined in an inservice inspection satisfy the tub           criteria but were not plugged in accordance with the Steam Generator Program as required by SR 4.4.5.2. 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 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, Action 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, Action a 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.
SEABROOK - UNIT 1                               B 3/4 4-7                               BC 07-01

REACTOR COOLANT SYSTEM BASES 3/4.4.5     STEAM GENERATORS (SG) TUBE INTEGRITY (Continued)
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 shutdown 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 REQUIREMENTS 4.4.5.1 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                 '
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 portion of the SG tubes below 15.21 inches from the top of the tubesheet is excluded from periodic inspections and plugging.
The Steam Generator Program defines the Frequency of SR 4.4.5.1. 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. In addition, Specification 6.7.6.k contains prescriptive requirements concerning inspection intervals to provide added assurance that the SG performance criteria will be met between scheduled inspections. 2,/          if',, ,    ,#     ,J.,     .    ,,
1041 0/ /,1                                            /r    4  v ,/k/4d&#xfd;Vt            1*
a//,0'4 W041/'
SEABROOK - UNIT 1                               B 3/4 4-8                       BC g-7-g-l-, 12-05

REACTOR COOLANT SYSTEM BASES 3/4.4.5 STEAM GENERATORS (SG) TUBE INTEGRITY (Continued)
SIR 4.4.5.2                                                                "-       /J-/
Durin          inspection, any inspected tube that satisfies the Stea       e erator Progra             criteria is removed from service by plugging. The tub              c iteria delineated in Specification 6.7.6.k are intended to ensure that tubes acce ed or continued service satisfy the SG performance criteria with allowance for error in          flaw size measurement and for future flaw growth. In addition, the tub               criteria, n conjunction with other elements of the Steam Generator Program, ensure that the SG p rformance 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 t tubes remaining in service will continue to meet the SG performance criteria.
The frequency of prior to entering MODE 4 following a SG in ection ensures that the Surveillance has been completed and all tubes meeting th                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 50.67
: 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."
SEABROOK - UNIT 1                              B 3/4 4-9                       BG 07 94-, BC 08-02}}