ML062830331

Response to Request for Additional Information, Technical Specification Change Request No. 331: Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity

ML062830331
Person / Time
Site:	Three Mile Island
Issue date:	10/06/2006
From:	Cowan P AmerGen Energy Co, Exelon Corp
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
5928-06-20492, TAC MD1807
Download: ML062830331 (51)

Text

AmerCen Energy Coriymny. L?C wwiAJ.exeloncorp.cori An Exelori Company 2 0 0 Exelor Way i<ennett Square, PA :gy$

10 CFR 50.90 October 6, 2006 5928-06-20492 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 Three Mile Island, Unit 1 (TMI Unit 1)

Facility Operating License No. DPR-50 NRC Docket No. 50-289

Subject:

Response To Request For Additional Information -

Technical Specification Change Request No. 331: Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity (TAC No. MD1807)

References:

1) USNRC Letter dated August 14, 2006, Three Mile Island, Unit 1 - Request for Additional Information Regarding the Steam Generator Tube Integrity Technical Specification Amendment (TAC No. MD1807)

2) AmerGen Energy Company, LLC letter to NRC dated May 15, 2006 (5928-06-20390), Technical Specification Change Request No. 331 -

Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity This letter provides additional information in response to the NRC request for additional information (RAI) issued by letter dated August 14, 2006 (Reference I ) , regarding TMI Unit 1 Technical Specification Change Request No. 331, submitted to NRC for review on May 15, 2006 (Reference 2). The additional information is provided in Enclosure 1. As described in the attached responses, the proposed Technical Specification page markups have been revised to incorporate additional requirements and clarifications consistent with the NRC approved TSTF-449, Revision 1. These changes have no impact on the conclusions of the original safety analysis or no significant hazards consideration evaluation provided in Reference 2. The revised proposed Technical Specification pages are provided in Enclosure 2. Enclosure 2 provides a complete replacement set of the proposed Technical Specification pages previously submitted in Reference 2.

It is noted that this response is submitted subsequent to the requested response date.

Additional NRC questions and discussions regarding sleeves, and additional NRC email RAI questions on sleeves received September 13, 2006, resulted in a delay in finalizing this RAI response as several of these questions concern sleeve repairs. As discussed with the NRC during a telecon on September 20,2006, the August 14,2006 NRC RAI questions concerning sleeve repairs will be addressed in a separate submittal that also provides the responses to the additional September 13, 2006 email draft RAI questions on sleeves. These separate submittal responses on sleeves may result in additional revisions to the enclosed TS page markups.

U.S. Nuclear Regulatory Commission October 6,2006 Page 2 No new regulatory commitments are established by this submittal. If any additional information is needed, please contact David J. Distel at (610) 765-5517.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 6'h day of October, 2006.

Respectfully, Pamela 6.cowan Director - Licensing &. Regulatory Affairs AmerGen Energy Company, LLC

Enclosures:

1) Response to Request for Additional ,,i,a-mation

2) Revised TS Page Markups cc: S. J. Collins, USNRC Administrator, Region I F. E. Saba, USNRC Project Manager, TMI Unit 1 D. M. Kern, USNRC Senior Resident Inspector, TMI Unit 1 File No. 06007

ENCLOSURE 1 TMI UNIT 1 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION TECHNICAL SPECIFICATION CHANGE REQUEST No. 331 APPLICATION FOR TECHNICAL SPECIFICATION IMPROVEMENT REGARDING STEAM GENERATOR TUBE INTEGRITY

Enclosure 1 5928-06-20492 Page 1 of 9 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION TMI Unit 1 TECHNICAL SPECIFICATION CHANGE REQUEST No. 331 APPLICATION FOR TECHNICAL SPECIFICATION IMPROVEMENT REGARDING STEAM GENERATOR TUBE INTEGRITY

1. NRC Question The Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler (TSTF-449), Steam Generator Tube Integrity, uses MODES. Please provide the justification for using 250 degrees Fahrenheit (OF) to represent TSTF MODE 4 (HOT SHUTDOWN) and greater than 250°F to represent TSTF MODES 1 (POWER OPERATION), 2 (STARTUP), and 3 (HOT STANDBY), in your proposed Technical Specifications (TS). Alternatively, discuss your plans to modify your proposed TS to be consistent with TSTF-449 in which MODE 4 has an average temperature of greater than 200°F and less than 330F, and MODE 5 (COLD SHUTDOWN) has an average temperature of less than or equal to 200°F.

Response

The proposed TSs have been modified to revise the temperature threshold from 250°F to 200°F in the following proposed TS Sections:

3.1.1.2.a Insert 3.1.1.2.a.(3.)b 3.1 Bases (Page 3-2) 4.1 9 Applicability 4.1 9.2 4.19 Bases (Applicability Section on Page 4-80) 4.19 Bases (Actions Section on Page 4-81) 4.1 9 Bases (Surveillance Requirement SR 4.19.2 Section on Page 4-83)

Since the TMI Unit 1 operating mode definitions differ from the TSTF Standard Technical Specification MODE definitions, TMI Unit 1 has used the 200°F value to be consistent with the lower bound of the Standard Technical Specification MODE 4 used in TSTF-449. Associated with the above changes, existing TS Section 3.1 .I .2.a is being deleted and replaced with the originally proposed TS Section 3.1.1.2.b Insert from Reference 2, and TS 3.1 Bases is revised to describe that both steam generators must have tube integrity before heatup of the reactor coolant. This change provides further consistency with the wording and format of TSTF-449, Revision 3. An existing reference to TS Section 3.1.1.2.a in the Bases for TS Section 3.4 (TS Page 3-26c) is also being deleted for consistency. The revised TS page markups are provided in Enclosure 2.

2. NRC Question In proposed TS Section 3.1.1.2.b, it would appear that TS Section 3.1.1.2.b(3) would permit you to elect not to plug a tube provided the conditions in TS Section 3.1.1.2.b(1) and TS Section 3.1.1.2.b(2) were met. This is not consistent with TSTF-449. In TSTF-449, the required actions are intended to apply only in the event a tube was inadvertently identified as not being plugged rather than electing not to plug a tube. Please discuss

Enclosure 1 5928-06-20492 Page 2 of 9 your plans to clarify your TS in this regard and clearly indicate that separate condition entry is only allowed for TS Section 3.1.1.2.b(3). For example, If the requirements of TS 3.1.1.2.b were not met for one or more tubes then perform the following ... In addition, discuss your plans to modify TS Section 3.1.1.2.b(3) to include the AND statement associated with Conditions a and b so that the TS section is consistent with TSTF-449.

Response

Section 3.1.1.2.b in the prior submittal (Reference 2) is now numbered Section 3.1.1.2.a, since the previous Section 3.1.1.2.a is being proposed for deletion as described in response to Question No. 1 above.

The Note in proposed TS Section 3.1.1.2.a has been modified to clarify entry into Section 3.1.1.2.a.(3.) and Section 3.1.1.2.a.(4.), which follow the Note. In addition, the word Whas been added between Conditions a. and b. of proposed TS Section 3.1.1.2.a.(3.). The revised TS page markup is provided in Enclosure 2.

3. NRC Question Please discuss your plans for modifying proposed TS Section 3.1.1.2.b(2) to remove the parenthetical or at a minimum remove Repair from the title of TS Section 6.19.

Response

Section 3.1.1.2.b.(2.) in the prior submittal (Reference 2) is now numbered Section 3.1.1.2.a.(2.), since the previous Section 3.1.1.2.a is being proposed for deletion as described in response to Question No. 1 above.

The proposed TS Section 3.1.1.2.a.(2.) has been modified to delete the word Repair from the parenthetical statement. The revised TS page markup is provided in Enclosure 2.

4. NRC Question It is unclear why of detection was added to proposed TS Sections 3.1.1.2.b(4) and TS Section 3.1.6.3. Please provide justification for adding of detection to the TS sections, or alternatively discuss your plans to modify these TS sections by removing of detection .

Response

Section 3.1.1.2.b.(4.) in the prior submittal (Reference 2) is now numbered Section 3.1.1.2.a.(4.), since the previous Section 3.1.1.2.a is being proposed for deletion as described in response to Question No. 1 above.

Enclosure 1 5928-06-20492 Page 3 of 9 The proposed TS Sections 3.1.1.2.a.(4.) and 3.1.6.3 have been modified to remove the phrase of detection. The revised TS page markups are provided in Enclosure 2.

5. NRC Question In proposed TS Section 3.1.1.2.b(4) and TS Section 3.15.3, HOT STANDBY is referenced. For TMI-1 HOT STANDBY has the reactor critical with an average temperature greater than 525°F. This is inconsistent with TSTF-449. Please discuss your plans to modify these TS sections to be consistent with TSTF-449 in which HOT STANDBY (MODE 3) has a keffless than 0.99 (i.e., reactor not critical) and an average temperature greater than 330°F.

Response

Section 3.1.1.2.b.(4.) in the prior submittal (Reference 2) is now numbered Section 3.1.1.2.a.(4.), since the previous Section 3.1.1.2.a is being proposed for deletion as described in response to Question No. 1 above.

The proposed TS Sections 3.1.1.2.a.(4.), 3.1.6.3, and 4.19 Bases (3.1.1.2.a.(4.) Section on Page 4-81) have been modified to change HOT STANDBY to HOT SHUTDOWN.

The TMI Unit 1 TS define HOT SHUTDOWN condition as subcritical by at least one percent delta k/k and Tavg 2525°F. The revised TS page markups are provided in Enclosure 2.

6. NRC Question In the proposed Bases for TS Section 3.1.1.2, TS Section 3.1 6 3 , and the proposed Bases for TS Section 3.1.6, through the steam generator tubes is an unnecessary qualifier when referring to primary-to-secondaryleakage. Please discuss your plans to modify the Bases and TS sections to remove this unnecessary qualifier. In addition, please discuss your plans to modify the Bases for TS Section 3.1.1.2 to spell out steam generator.

Response

The proposed TS Bases for TS Section 3.1.1.2, TS Section 3.1.6.3, and the proposed Bases for TS Section 3.1.6 have been modified to delete the phrase through the steam generator tubes. In addition, the Bases for TS Section 3.1.1.2 have been modified to spell out the acronym SGas steam generator. The revised TS page markups are provided in Enclosure 2.

7. NRC Question Please clarify the leakage limits in proposed TS Section 3.1.6.3. For example, if the 0.1 gallons per minute (gpm) (144 gallons per day (gpd)) is the sum of the leakage for both steam generators (SG), the TS section may be modified to read, if the sum of the

Enclosure 1 5928-06-20492 Page 4 of 9 primary-to-secondary leakage from both steam generators exceeds 0.1 gpm (144 gpd) ... Please discuss your plans to modify the TS to address this issue.

The proposed TS Section 3.1.6.3 and associated Bases have been modified to read, if the sum of the primary-to-secondary leakage from both steam generators exceeds 0.1 gpm (144 GPD) .... The revised TS page markups are provided in Enclosure 2.

8. NRC Question Please discuss your plans to modify the proposed Bases for TS Section 3.1.6 by replacing the term Tube leakage in the third paragraph with primary-to-secondary leakage. In addition, please confirm that the proposed insert for the Bases for TS Section 3.1.6 is consistent with the current design and licensing basis for TMI-1.

Response

The proposed Bases for TS 3.1.6 have been modified to replace the term tube leakage in the third paragraph with primary-to-secondary leakage. The proposed insert for the Bases for TS 3.1.6 is consistent with the current design and licensing basis for TMI Unit 1. The revised TS page markup is provided in Enclosure 2.

9. NRC Question On Page 4-78, the proposed Limiting Condition of Operation (LCO) for TS Section 3.1.1.2.b, third paragraph states that ...a SG tube is defined as the entire length of the tube, including the tube wall and any repairs made to it,... Please discuss your plans to modify the proposed LCO to remove and any repairs made to it given that TMI-1 does not have approved SG tube repair methods.

Response

This question affects the TS requirements for the TMI Unit 1 currently installed sleeves and will be addressed in a separate submittal.

10. NRC Question The proposed Bases for TS Section 3.1 .I .2.b LCO (Pages 4-79 to 4-80) states that ?he accident analysis assumes that accident induced leakage does not exceed 1 gpm per SG, except for specific types of degradation at specific locations where the U.S. Nuclear Regulatory Commission (NRC) has approved greater accident induced leakage. Please discuss your plans to modify this proposed LCO to further clarify this statement by including a reference to the specific types of degradation. In addition, discuss your plans

Enclosure 1 5928-06-20492 Page 5 of 9 to modify proposed TS Section 6.19.c to further clarify the accident induced leakage limits for the specific types of degradation. For example, Leakage from all sources, excluding the leakage attributed to the degradation described in TS Section 6.19.c, is not to exceed 1 gpm per SG.

Response

Section 3.1.1.2.b in the prior submittal (Reference 2) is now numbered Section 3.1 .I .2.a, since the previous Section 3.1 .I .2.a is being proposed for deletion as described in response to Question No. 1 above.

The proposed Bases for TS Section 3.1.1.2.a on new TS Page 4-80 have been modified to add a parenthetical reference to TS Section 6.19.c for approved repair criteria. In addition, the previously proposed TS Section 6.19.b.2. contained a sentence that read, Leakage is not to exceed 1 gpm per SG, except for specific types of degradation at specific locations as described in paragraph c of the Steam Generator Program below.

This sentence in proposed TS Section 6.19.b.2 (Insert To TS Page 6-26, page 1 of 3) has been underlined for emphasis, and provides the requested clarification. The revised TS page markups are provided in Enclosure 2.

11. NRC Question Proposed TS Section 4.1 9 states, 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 or is assumed to increase as a result of accident induced conditions. Please provide justification for removing ...1 gallon per minute or is assumed to increase to 1 aallon rser minute... or alternatively discuss your plans to modify your TS to be consistent with TSTF-449.

Response

The proposed TS Section 4.19 Bases (Applicable Safety Analyses - Page 4-78) have been modified to reference the leakage rates described in the proposed TS 6.19.c.2 as a result of accident-induced conditions. A similar change has been made to the proposed Bases Insert to TS Page 3-15a (Bases for Section 3.1.6). The leakage rates described in the proposed TS 6.1 9.c.2 are consistent with the current TMI Unit 1 design and licensing basis. The revised TS page markup is provided in Enclosure 2.

12. NRC Question Proposed TS Section 4.19 states (Page 4-78), For accidents that do not involve fuel damage, the primary coolant activity level of DOSE EQUIVALENT 1-131 is conservatively assumed to be equal to, or greater than, the TS 3.1.4, Reactor Coolant System Activity, limits. The addition or greater than is unclear, please provide justification for adding this statement or alternatively discuss your plans to modify your TS to be consistent with TSTF-449.

Enclosure 1 5928-06-20492 Page 6 of 9

Response

The TMI Unit 1 Reactor Coolant System allowable Dose Equivalent (DE) 1-131 concentration specified in existing TS 3.1.4 is 0.35 uCi/g. This decreased concentration was implemented in TS Amendment No. 204, dated October 2, 1997 associated with the reanalysis of the Main Steam Line Break Accident with Accident-Induced SG Tube Leakage. Some of the TMI Unit 1 Design Basis Accident Analyses described in the UFSAR, were based on either an assumed DE 1-131 concentration of 1.O uCi/g or a 1%

failed fuel assumption, which are more conservative than the current TS Section 3.1.4 limit. Since there are analyses that assumed a greater concentration than that allowed under TS 3.1.4, and those analyses are conservative with respect to TS 3.1.4, the phrase or greater than was added in the proposed TS Section 4.1 9 Bases wording to more accurately reflect the current TMI Unit 1 design and licensing basis as stated above.

The subject phrase, or greater than, was not modified on proposed Page 4-78 of Enclosure 2.

13. NRC Question Proposed TS Section 4.19.1 states that each SG is determined to be operable by verifying SG tube integrity in accordance with, and at the frequency required by, the SG Program. Given that the SG Program only provides maximum inspection intervals, this statement is not appropriate. In addition, the maximum intervals provided in the SG Program may not be sufficient to ensure SG tube integrity and therefore, it may be necessary to inspect more frequently to ensure that SG tube integrity is being maintained.

Please discuss your plans to remove the statement regarding the SG tube inspection frequency.

Response

The proposed TS Section 4.19.1 has been modified to remove the phrase regarding the SG tube inspection frequency. The revised TS page markup is provided in Enclosure 2.

14. NRC Question Given that TMI-1 does not have approved SG tube repair methods, discuss your plans to remove TS Section 6.9.6.i. In addition, for the same reason, discuss your plans to modify TS Section 6.1 9 by deleting Section 6.19.f.

Response

This question affects the TS requirements for the TMI Unit 1 currently installed sleeves and will be addressed in a separate submittal.

Enclosure 1 5928-06-20492 Page 7 of 9

15. NRC Question TS Section 6.9.6, Steam Generator Tube Inspection Report, does not appear to carry over reporting requirement 4.19.5.b(3) or 4.19.5.b(6) from the previous TS Section 4.1 9.5, Reports. Please confirm that both of these reporting requirements are contained in Engineering Change Request (ECR) No. TM 01-00328. If not, please provide justification for not carrying these reporting requirements over or alternatively, discuss your plans to modify TS Section 6.9.6 to include these reporting requirements.

Proposed TS Sections 6.9.6. j and 6.9.6.k have been added to include the reporting requirements from the current TMI Unit 1 TS Section 4.19.5.b(3) and 4.19.5.b(6). The revised TS page markup is provided in Enclosure 2.

16. NRC Question Please discuss your plans to modify TS Section 6.9.6 to clearly indicate when the 90-day report should be submitted to the NRC. For example, A report shall be submitted within 90 days after the plant reaches MODE 4 (using the TSTF-449 definition of MODE 4). In addition, it is unclear why TS Section 6.19(d) is referenced rather than TS Section 6.19.

Please provide justification for referencing TS Section 6.19(d), or alternatively modify TS Section 6.9.6 to reference TS Section 6.19. In addition, discuss your plans to remove reference to and tube repairs in proposed TS Section 6.9.6.h.

Response

The proposed TS Section 6.9.6 has been modified to indicate that the SG Tube Inspection Report should be submitted to NRC within 90 days after the average reactor coolant temperature exceeds 200°F following completion of an inspection performed in accordance with Section 6.19, Steam Generator Program. This bounds the conditions defined in TSTF-449 MODE 4. TS Section 6.9.6 has also been modified to reference TS Section 6.19 in lieu of TS Section 6.19(d). The revised TS page markup is provided in Enclosure 2.

The reference to and tube repairs in proposed TS Section 6.9.6.h affects the TS requirements for the TMI Unit 1 currently installed sleeves and will be addressed in a separate submittal.

17. NRC Question The NRC staff is aware that sleeves were installed in the TMI-1 SGs to stiffen the tubes and not as a SG tube repair method. Please confirm that the tube repair criteria (240-percent through-wall) is being applied to the parent tube behind the sleeves including the sleeve-to-tube joint. If the repair criteria is not being implemented for the required length of defect free joint, discuss your plans for submitting the sleeving method for approval as a repair technique.

Enclosure 1 5928-06-20492 Page 8 of 9

Response

This question affects the TS requirements for the TMI Unit 1 currently installed sleeves and will be addressed in a separate submittal.

18. NRC Question Please discuss your plans for moving the inspection requirements for implementation of your alternate repair criteria from the repair criteria section to the inspection section (i.e.,

6.19.d).

The proposed TS Sections 6.19.c(l) and 6.19.c(2) have been modified to remove the inspection requirements for implementation of alternate repair criteria. These inspection requirements have been relocated to new proposed TS Sections 6.19.d(4) and 6.1 9 4 5 ) .

The revised TS page markups are provided in Enclosure 2.

19. NRC Question Please confirm that the repair limits in current TS Section 4.1 9.4.6 for inside diameter intergranular attack (IDIGA) are identical to those in ECR No. TM 01-00328.

The repair limits in the current TMI Unit 1 TS Section 4.1 9.4.6 for inside diameter intergranular attack (IDIGA) are identical to those in ECR No. TM 01-00328. The proposed TS Section 6.19.c(l) has been modified to include the repair limit criteria from the current TS Section 4.19.4.6. The revised TS page markup is provided in Enclosure 2.

20. NRC Question Please discuss why the proposed Bases section for TS Section 4.1 9 no longer discusses the IDIGA alternate repair criteria. Alternatively, modify your proposed Bases to include a discussion of this alternate repair criteria.

Response

The proposed TS Section 4.19 Bases (Page 4 Surveillance Requirement SR 4.19.2), has been modified to incorporate the discussion of the IDIGA alternate repair criteria from the current TS Section 4.1 9 Bases (Page 4-83). The revised TS page markup is provided in Enclosure 2.

Enclosure 1 5928-06-20492 Page 9 of 9

21. NRC Question Please discuss the purpose of the statement in TS Section 6.1 9 to refer to Section 6.9.6 for reporting requirements...] Alternatively discuss your plans to remove this statement.

Response

The statement was added to provide a clear reference to the location of the reporting requirements for inspections. The proposed TS Section 6.19 has been modified to relocate this statement to the end of proposed TS Section 6.19 and to designate this statement as a clarifying Note. The revised TS page markups are provided in Enclosure 2.

ENCLOSURE 2 TMI Unit 1 Technical Specification Change Request No. 331 Revised Markup of Proposed License, Technical Specifications, and Bases Page Changes Revised License Paqes 6

7 Revised Technical Specifications & Bases Paqes Table of Contents Page iv Table of Contents Page v Table of Contents Page vi 3-1a 3-2 3-12 3-15a 3-26~

4-2b 4-8 4-77 4-78 4-79 4-80 4-81 4-82

' 4-83 4-83a 4-84 4-85 6-19 6-26

(8) Repaired Steam Generators a

- DELGFD In order to confirm the leak-tight integrity of the Reactor Coolant System, includi the steam generators, operation of the facility shall be in accordance with the

1. Prior to initial criticality, the licensee shall submit to NRC the result line leakage rate by more than 0.1 gpm', the hall be shut down span, the leaking tube(s) shall be r from service. The

3. The licensee sha st program at each power range with the program described in available the results of that test t review, prior to ascension from power operation.

4. The licensee shall condu ent examinations, consistent with the extended inservice d in Table 3.3-1 of NUREG-1019, either 90 calendar ower, or 120 calendar days after mes first. In the event of plant 0% power, the licensee shall ent at the end of 180 operation at power levels rmation as to the necessity of a s dy-current testing n before the end of the refueling that assessment and determine the time of tion, consistent with the other provisions of t sence of such an assessment, a special ECT shutd an additional 30 days of operation at power above 8 operation, the facility shall be shutdown and leak tested. Operation at ainder of Cycle 8 operation. After the 9R refueling outage, the leakage limit and Amendment No, Amendment N o s * ,

(9) Lona Ranae Plannina Proaram - Deleted I I

Sale and License Transfer Conditions (10) Deleted (11) Deleted (12) Deleted (13) Deleted Amendment No. $4p, W ,H-0,E33,24!3,458; I

TABLE OF CONTENTS Section Paoe 4.8 DELETED 4-51 4.9 DECAY HEAT REMOVAL (DHR) CAPABILITY - PERIODIC TESTING 4-52 4.9.1 REACTOR COOLANT SYSTEM (RCS) TEMPERATURE GREATER THAN 250 DEGREES F 4-52 4.9.2 RCS TEMPERATURE LESS THAN OR EQUAL TO 250 DEGREES F 4-52a 4.10 REACTIVITY ANOMALIES 4-53 4.1 1 REACTOR COOLANT SYSTEM VENTS 4-54 4.12 AIR TREATMENT SYSTEMS 4-55 4.12.1 EMERGENCY CONTROL ROOM AIR TREATMENT SYSTEM 4-55 4.12.2 REACTOR BUILDING PURGE AIR TREATMENT SYSTEM (DELETED) 4-55b 4.12.3 AUXILIARY AND FUEL HANDLING BUILDING AIR TREATMENT 4-55d SYSTEM (DELETED) 4.12.4 FUEL HANDLING BUILDING ESF AIR TREATMENT SYSTEM 4-55f 4.13 RADIOACTIVE MATERIALS SOURCES SURVEILLANCE 4-56 4.14 DELETED 4-56 4.15 MAIN STEAM SYSTEM INSERVICE INSPECTION 4-58 4.16 REACTOR INTERNALSVENT VALVES SURVEILLANCE 4-59 4.17 SHOCK SUPPRESSORS ISNUBBERSl 4-60 4.18 FIRE PROTECTION SYSTEMS (DELETED) 4-72 4 4.19 A 7-9 I ,

I 4.20 REACTOR BUILDING AIR TEMPERATURE 4-86 4.21 RADIOACTIVE EFFLUENT INSTRUMENTATION(DELETED) 4-87 4.21.1 RADIOACTIVE LIQUID EFFLUENT INSTRUMENTATION (DELETED) 4-87 4.21.2 RADIOACTIVE GASEOUS PROCESS AND EFFLUENT MONITORING 4-87 INSTRUMENTATION(DELETED) 4.22 RADIOACTIVE EFFLUENTS (DELETED) 4-87 4.22.1 LIQUID EFFLUENTS (DELETED) 4-87 4.22.2 GASEOUS EFFLUENTS (DELETED) 4-87 4.22.3 SOLID RADIOACTIVE WASTE (DELETED) 4-87 4.22.4 TOTAL DOSE (DELETED) 4-87 4.23.1 MONITORING PROGRAM (DELETED) 4-87 4.23.2 LAND US CENSUS (DELETED) 4-87 4.23.3 INTERLABORATORYCOMPARISON PROGRAM (DELETED) 4-87 iv

TABLE OF CONTENTS Section &gg 5 DESIGN FEATURES 5-1 5.1 -

SITE 5-1 5.2 CONTAINMENT 5-2 5.2.1 REACTOR BUILDING 5-2 5.2.2 REACTOR BUILDING ISOLATION SYSTEM 5-3 5.3 REACTOR 5-4 5.3.1 REACTOR CORE 5-4 5.3.2 REACTOR COOLANT SYSTEM 5-4 5.4 NEW AND SPENT FUEL STORAGE FACILITIES 5-6 5.4.1 NEW FUEL STORAGE 5-6 5.4.2 SPENT FUEL STORAGE 5-6 5.5 AIR INTAKE TUNNEL FIRE PROTECTION SYSTEMS 5-8 6 ADMINISTRATIVE CONTROLS 6-1 6.1 RESPONSIBILITY 6-1 6.2 ORGANIZATION 6-1 6.2.1 CORPORATE 6-1 6.2.2 UNIT STAFF 6-1 6.3 UNIT STAFF QUALIFICATIONS 6-3 6.4 TRAINING 6-3 6.5 REVIEW AND AUDIT 6-3 6.5.1 TECHNICAL REVIRN AND CONTROL 6-4 6.5.2 INDEPENDENT SAFETY REVIEW 6-5 6.5.3 AUDITS 6-7 6.5.4 DELETED 6-8 6.6 REPORTABLE EVENT ACTION 6-10 6.7 SAFETY LIMIT VIOLATION 6-10 6.8 PROCEDURES AND PROGRAMS 6-11 6.9 REPORTING REQUIREMENTS 6-12 6.9.1 ROUTINE REPORTS 6-12 6.9.2 DELETED 6-14 1;

6.9.3 ANNUAL RADIOLOGICAL ENVIRONMENTALOPERATING REPORT 6-17 6.9.4 ANNUAL RADIOACTIVEEFFLUENT RELEASE REPORT 6-18 6.9.5 CORE OPERATING LIMITS REPORT 6-19 RECORD RETENTION 6-20 RADIATION PROTECTIONPROGRAM 6-22 HIGH RADIATION AREA 6-22 PROCESS CONTROL PROGRAM 6-23 OFFSITE DOSE CALCULATION MANUAL (ODCM) 6-24 nci nrn ULLL I L W 6-24 DELETED 6-24

-

MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS 6-25

-

TlON (TXBASES CONTROL PROGRAM 6-25 I S e A A 4-26>

LIST OF 'I'ABLES m R L E: TITLE I .2 Frequency Notation 2.3-1 Rciictor Protcction System Trill Setting Liriiits 3.1.6.1 Prcssure Isolation Check V:ilvcs Betwcrn the Priniitry Coolant Systcni iind LPIS 3.5-1 Instruments Operatiiig Conditions 3-29 3.S-IA DELETED 3.5-2 Accident iMoni to ri ng Inst runicii t s 3-JOC 3.5-3 Pust A cc i den t Mon i t o ri 11 g 111st niincrit ii t it) 11 3-40d 3.5-4 Rcmo t e S11 ti t d OW 11 SFStern Ifist ninietltit t ion it11 tl Con t rol 3-4Oi I

3.21-1 DELETED 3.21-2 DELETED 3.23-1 DELETED 3.23-2 DELETED 4.1-1 lnstnimcnt Surveillance Requirements 4-3 4.1-2 Minimuni Equipment Test Frcqiiency 4-8 4.1-3 M ininium Sampling Frequency 4-9 4.1-4 Post Accident Monitoring Instrunicnt;cticin 4-I Oil 4.19-1 4-M-4.19-2 -.--.---...-....--

4.21-1 DELETED 4.21-2 DELETED 4.22-1 DELETED 4.22-2 DELETED 4.23- 1 DELETED

3.1 REACTOR COOLANT SYSTEM 3.1.1 OPERATIONAL CWONENTS P o o l i c a b i l it v Applies t o t h e m e r a t i n g S t a t u s of r e a c t o r ccolant system components.

Ob i es t ive

• To specify those l i m i t i n g conditions f o r operation of r e a c t o r coolant system components wnich must be met t o ensure safe r e a c t o r operations.

Soecif i c a t l o n 3.1.1.1 Reactor Czzlant Pwnos

a. Pump combinations permissible f o r given power l e v e l s s h a l l ae a s snown in S p e c i f i c a t i o n Table 2.3.1.

b. Power coeration w i t h one i d l e r e a c t o r coolant pump i n each l ~ o ps h a l l be r e s t r i c t e d t o 20 nours. I f it;e L.

r e a c t o r i s not retuned t o an acceptable RC pump operating combination a t t h e end of tne 24-hout pericc, the r e a c t o r s h a l l be i n a h o t shutcown condition w i m i n the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> .

c. The boron c o n c r n t r a t i o n i n t h e reactcr czolant ,system not be rzauced unless a t l t a s t one i e a c t j t ccolant pump o r one clecay heat removal ~ c m pis c i r c u l a t i n g reactor coolant.

3.1.1.2 Steam G e n e r a t o r e 3.1.1.3 Pressurizer Safety Valves

a. The r e a c t o r s h a l l not remain c r i t i c a l unless both pressurizer code s a f e t y valves a r e operable w i t h a l i f t

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s e t t i n g of 2500 p s i g 1%.L

b. When the r e a c t o r is s u b c r i t i c a l , a t l e a s t one p r o s s u r i t r r code s a f e t y valve m a l l be cperaole ifa l l reactor coolant system openings are closed, except for h y d r o s t a t i s tests in accordance w i t h ASME 9 o i l z r anc Pxssurz Vessel Ccde, Sectlcn I I i .

3-la I Amendpent 12. 17, 28, P7, # ,

INSERT TO TS PAGE 3-1a (REVISED TS 3.1.1.2)

a. Whenever the reactor coolant average temperature is above 200°F,the following conditions are required:

(1.) SG tube integrity shall be maintained.

(2.) All SG tubes satisfying the tube repair criteria shall be plugged in accordance with the Steam Generator Program. (The Steam Generator Program is described in Section 6.19.)

Entry into Sections 3.1.1.2.a.(3.) and (4.), below, is allowed for each SG tube. If the requirements of Sections 3.1.1.2.a.(1.) or 3.1.1.2.a.(2.) were not met for one or more tubes then perform the following.

(3.) With one or more SG tubes satisfying the tube repair criteria and not plugged in accordance with the Steam Generator Program:

a. Verify within 7 days that tube integrity of the affected tube(s) is maintained until the next refueling outage or SG tube inspection, AND

b. Plug the affected tube(s) in accordance with the Steam Generator Program prior to exceeding a reactor coolant average temperature of 200°F following the next refueling outage or SG tube inspection.

(4.) If Action 3., above, is not completed within the specified completion times, or SG tube integrity is not maintained, be in HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and be in COLD SHUTDOWN within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

I nf I

Bascs The limitation on power operation ivith one idle RC pump in each loop has been imposed since thc ECCS cooling performance has not been calculated in accordancc with the Final Acceptance Criteria requirements specifically for this mode of reactor operation. A time period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is allo~vedfor operation with one idle RC pump in each loop to. effect repairs of the idle pump(s) and to return the reactor to an acceptable combination of operating RC pumps. The 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for this mode of operation is acceptable since this mode is espected to have considerable margin for the peak cladding temperature limit and since the likelihood of a LOCA ivithin the 24-hour period is considered very remote.

A reactor coolant pump or decay heat removal pump is required to be in operation before the boron concentration is reduced by dilution tvith makeup water. Either pump tvill provide mixing \vhich tvill prevent sudden positive reactivity changes caused by dilute coolant reaching the reactor. One decay heat removal pump will circulate the equivalent of the reactor coolant system volume in one-half hour or less.

The decay heat removal system suction piping is designed for 300°F and 370 psig; thus, the s>stern can remove decay heat when the reactor coolant system is below this temperature (References I , 2, Gfle +,bc i.fa and 3).

Both steam generators must bef6rE heatup of the Reactor Coolant System to insure system integrity against le&=ai and transient conditions. Only one steam generator is required for decay heat removal purposes.

4 One pressurizer code safety valve is capable of preventing overpressurization \vhen the reactor is not critical since its relieving capacity is greater than that required by the sum of the available heat sources which are pump energy, pressurizer heaters, and reactor decay heat. Both pressurizer code safety valves are required to be in service prior to criticalit) to conform to the system design relief capabilities. The code safety valves prevent,overpressure for a rod withdrawal or feedwater line break accidents (Reference 4). The pressurizer code safety valve lift set point shall be set at 2500 psig 2 I % allowance for error. Surveillance requirements are specified in the Insenice Testing Program. Pressurizer code safety valve setpoint drift of up to 3% is acceptable in accordance with ASME Section XI (Reference 5 ) and the assumptions of TMI-1 s References (1) UFSAII, Tables 9.5 (2) U F S a Sections 4.2.5.1 and 9.5 - Decay Heat Removal (3) UFSpJi Section 4.2.5.4 - Secondary System (4) UFSAR Section 4.3.10.4 - System Minimum Operational Components (5 ) UFSAR, Section 4.3.7 - Overpressure Prorection 3-2 Amendment No. 37 (12/22/78),-W -222-

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3.1.6 LEAKAGE Applicabilitv Applies to reactor coolant leakage from the reactor coolant system and the makeup and purification system.

Objective To assure that any reactor coolant leakage does not compromise the safe operation of the facility.

Specification 3.1.6.1 If the total reactor coolant leakage rate exceeds 10 gpm, the reactor shall be placed in hot shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of detection.

3.1 '6.2 If unidentified reactor coolant leakage (excluding normal evaporative losses) exceeds 3.1.6.3 3.1.6.4 If any reactor coolant leakage boundary (such as the reactor vessel, piping, valve body, etc., except the steam generator tubes), the reactor shall be shutdown, and a cooldown to the cold I shutdown condition shall be initiated within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of detection.

3.1.6.5 If reactor shutdown is required by Specification 3.1.6.1, 3.1 -6.2, 3.1.6.3, or 3.1.6.4, the rate of shutdown and the conditions of shutdown shall be determined by the safety evaluation for each case.

3.1.6.6 Action to evaluate the safety implication of reactor coolant leakage shall be initiated within four hours of detection. The nature, as well as the magnitude, of the leak shall be considered in this evaluation. The safety evaluation shall assure that the exposure of offsite personnel to radiation is within the dose rate limits of the ODCM. I 3.1.6.7 If reactor shutdown is required per Specification 3.1.6.1, 3.1.6.2, 3.1.6.3 or 3.1.6.4, the reactor shall not be restarted until the leak is repaired or until the problem is otherwise corrected.

3.1.6.8 When the reactor is critical and above 2 percent power, two reactor coolant leak detection systems of different operating principles shall be in operation for the Reactor Building with one of the two systems sensitive to radioactivity. The systems sensitive to radioactivity may be out-of-service for no more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> provided a sample is taken of the Reactor Building atmosphere every eight hours and analyzed for radioactivity and two other means are available to detect leakage.

3-12 Amendment No. 47,449, W , Y (12-22-78)

-

Bases (Continued)

The is established as a quantity which can be accurately early detection of leakage. Leakage of this magnitude can be reasonably detected within a matter of hours, thus providing confidence that cracks associated with such leakage will not develop into a critical size before mitigating actions can be taken.

Total reactor coolant leakage is limited by this specification to 10 gpm. This limitation provides allowance for a limited amount of leakage from known sources whose is limited to quantified by analysis of secondary plant activity.

If reactor coolant leakage is to the auxiliary building, it may be identified by one or more of the following methods:

a. The auxiliary and fuel handling building vent radioactive gas monitor is sensitive to very low activity levels and would show an increase in activity level shortly after a reactor coolant leak developed within the auxiliary building.

\

b. Water inventories around the auxiliary building sump.

I c.

d.

Periodic equipment inspections.

In the event of gross leakage, in excess of 13 gpm, the individual cubicle leak I detectors in the makeup and decay heat pump cubicles, will alarm in the control room to backup "a", "b", and "c' above.

When the source and location of leakage has been identified, the situation can be evaluated to determine if operation can safely continue. This evaluation will be performed by TMI-1 Plant Operations.

3-1 5a Amendment No. 4-44, Jf46,

INSERT TO TS PAGE 3-15a (BASES FOR SECTION 3.1.61 Except for primary to secondary leakage, the safety analyses do not address operational leakage. However, other operational leakage 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 that primary to secondary leakage from all steam generators (SGs) is one gallon per minute or is assumed to increase to the leakage rates described in TS 6.1 9.c.2 as a result of accident-induced conditions.

The TS requirement to limit primary to secondary leakage through both SGs to less than or equal to 144 gallons per day is significantly less than the conditions assumed in the safety analysis.

The limit of 144 gallons per day total for both SGs bounds the TSTF-449, Rev. 4 limit of 150 gallons per day per SG, which is based on the operational leakage performance criterion in NEI 97-06, Steam Generator Program Guidelines (Ref. 1). 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.

1 of 1

3.4 DECAY HEAT REMOVAL (DHR) CAPABILITY (Continued)

Bases (Continued)

If EFW were required during surveillance testing, minor operator action (e.g., opening a local isolation valve or manipulating a control switch from the control room) may be needed to restore operability of the required pumps or flowpaths. An exception to permit more than one ERN Pump or both EFW flowpaths to a single OTSG to be inoperable for up to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> during surveillance testing requires 1) at least one motor-driven EFW Pump operable, and 2) an individual involved in the task of testing the EFW System must be in communication with the control room and stationed in the immediate vicinity of the affected EFW flowpath valves. Thus the individual is permitted to be involved in the test activities by taking test data and his movement is restricted to the area of the EFW Pump and valve rooms where the testing is being conducted.

The allowed action times are reasonable, based on operating experience, to reach the required plant operating conditions from full power in an orderly manner and without challenging plant systems. Without at least two EFW Pumps and one EFW flowpath to each OTSG operable, the required action is to immediately restore ERN components to operable status, and all actions requiring shutdown or changes in Reactor Operating Condition are suspended. With less than two ERN pumps or no flowpath to either OTSG operable, the unit is in a seriously degraded condition with no safety related means for conducting a cooldown. In such a condition, the unit should not be perturbed by any action, including a power change, which might result in a trip.

The seriousness of this condition requires that action be started immediately to restore ERN components to operable status. TS 3.0.1 is not applicable, as it could force the unit into a less safe condition.

The E W system actuates on: 1) loss of all four Reactor Coolant Pumps, 2) loss of both Main Feedwater Pumps, 3) low OTSG water level, or 4) high Reactor Building pressure. A single active failure in the HSPS will neither inadvertently initiate the EFW system nor isolate the Main Feedwater system. OTSG water level is controlled automatically by the HSPS system or can be controlled manually, ifnecessary.

The MSSVs will be able to relieve to atmosphere the total steam flow if necessary. Below 5%

power, only a minimum number of MSSVs need to be operable as stated in Specifications 3.4.1.2.1 and 3.4.1.2.2. This is to provide OTSG overpressure protection during hot functional testing and low power physics testing. Additionally, when the Reactor is between hot shutdown and 5% full power operation, the overpower trip setpoint in the RPS shall be set to less than 5%

as is specified in Specification 3.4.1.2.2. The minimum number of MSSVs required to be operable allows margin for testing without jeopardizing plant safety. Plant specific analysis shows that one MSSV is sufficient to relieve reactor coolant pump heat and stored energy when the reactor has been subcritical by 1% delta WK for at least one hour. Other plant analyses show that two (2) MSSVs on either OTSG are more than sufficient to relieve reactor coolant pump heat and stored energy when the reactor is below 5% full power operation but had been subcritical by 1% delta WK for at least one hour subsequent to power operation above 5% full MSSVs are inoperable, the power level must be reduced, as stated in Specification 3.4.1.2.3 such that the remaining MSSVs can prevent overpressure on a turbine trip.

3-26~

Amendment No. p, 157,229

Bases Cont'd)

The equipment testing and system sampling frequencies specified in Tables 4.1-2, 4.1-3, and 4.1-5 are considered adequate to maintain the equipment and systems in a safe operational status.

REFERENCE

-

(1) UFSAR, Section 7.1.2.3(d) "PeriodicTesting and Reliability' (2) NRC SER for BAW-l0167A, Supplement 1, December 5,1988.

- -- - ?

semA.&3 Leak G u r ' d ~ k e s .

- I 4-2b Amendment No. 4-8+226,-&6--,

INSERT TO TS PAGE 4-2b (BASES FOR SECTION 4.11 The primary to secondary leakage surveillance in TS Table 4.1 -2, Item 12, verifies that primary to secondary leakage is less than or equal to 144 gallons per day total through both SGs. Satisfying the primary to secondary leakage limit ensures that the operational leakage performance criterion in the Steam Generator Program is met. If this surveillance is not met, compliance with TS 3.1.1.2, Steam Generator (SG) Tube Integrity, and TS 3.1.6.3, should be evaluated. The 144 gallons per day limit is measured at room temperature. The operational leakage rate limit applies to leakage through both SGs.

The TS Table 4.1-2 primary to secondary leakage surveillance is modified by a Note, which states that the initial surveillance is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after establishment of steady state operation. For RCS primary to secondary leakage determination, steady state is defined as stable RCS pressure, temperature, power level, pressurizer and makeup tank levels, makeup and letdown, and RCP seal injection and return flows.

The TS Table 4.1 -2 primary to secondary leakage surveillance frequency of Daily 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 EPRl guidelines (Ref. 5).

1 of 1

TABLE 4.1-2 MINIMUM EQUIPMENT TEST FREQUENCY

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Item -

Test Freauency

1. Control Rods Rod drop times of all Each Refueling shutdown full length rods

2. Control Rod Movement of each rod Every 92 days, when Movement reactor is critical

3. Pressurizer Setpoint In accordance with the Safety Valves lnservice Testing Program

4. Main Steam Setpoint In accordance with the Safety Valves lnservice Testing Program

5. Refueling System Functional Start of each Interlocks refueling period

6. (Deleted) -- --

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7. Reactor Coolant Evaluate Daily, when reactor System Leakage coolant system temperature is greater than 525 degrees F

8. (Deleted) -- --

9. Spent Fuel Functional Each refueling period Cooling System prior to fuel handling

10. Intake Pump (a) Silt Accumulation - Not to exceed 24 months House Floor Visual inspection (Elevation of Intake Pump 262 ft. 6 in.) House Floor (b) Silt Accumulation Quarterly Measurement of Pump House Flow

11. Pressurizer Block Functional* Quarterly Valve (RC-V2)

• Function shall be demonstrated by operating the valve through one complete cycle of 2 -

Amendment No. 56,68,78,448,#6,#8,~,446.

J

.

f cbiC.1 Sppcification applier to the inservice inapection f this inservice inspection program icl Through Steam atora, while a t the same time mi exponure to p e l i n the perforaunca of the in8 Specification Eolch 8turm generator I be dewnrtrattd 0 LE by performance of the following augmen service inepe rogrsm and the requirement8 of Specifi shutdown by selecting t i n g a t least the minimum fled in Table 4.19.1 a t the i

pment s h a l l be 11 detect defects 4-7 7

a. he first sample of tubes selected for each inservice inspection (subsequent to the service inspection) of each steam generator shall include:

ted potential problems.

tion (pursuant to Specification 4.19.4.a.8) tube. If any selected tube does not permit a tube inspection, this shall be recor and subjected to a tube inspection.

the first random sample if 0th steam generators pected. No credit will be (1) Group A-1: Tub 79 adjacent to the open inspection drawn fiom tube 66- I to tube 75- 15 and from 86-1 to (2) Group A-2: Tubes havin led opening in the 15th support plate

b. The tubes selected as the second

1. The tubes selected those areas oft C.

each sample inspection shall be classified into one of the following three Catenory Inspection Results Less than 5% of the total tubes inspected in a steam generator are degraded tubes and none of the inspected tubes are defective.

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4-78 Amendment No. 4+W, e,

( 12-22-78)

19.2 Specification (Continued)

One or more tubes, but not more than 1% of the total tubes inspected in a ste More than 10% of the total tubes inspected in a steam generator a re than 1% of the inspected tubes are defective.

all inspections, previously degraded tubes whose d ion size measurement (> 0.24 volt bobbi ant to 4.19.2.a.4,defective or on shall be included in not be included i 4.19.3 Inspection Frequencies The required inservice inspections o r tubes shall be performed at the following frequencies:

two consecutive ins assing not less than 18 tinued and no additional degradation urred, the inspection 1 p 40months.

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4-79 Amendment No. 4?+3+!% *-I

Additional. unscheduled insenice inspections shall be performed on each steam gen I I1 accordance with the first sample inspection specified in Table 4.19-2 during the sh subsequent to any of the following conditions:

A seismic occurrence greater than the Operating Basis Earthquake.

major main steam line or feedwater line break.

tion of the affected stean rmed in accordance Lvith the follo

2. If the leaking tube is not n Section 4.19.3.d. 1, then an inspection will be performed on the aEe enerator@)in accordance with Table 4.19-2 ns. finish, or contour of a tube n degraded tube criteria sp means a tube containing:

extent, or (b) imperfections 2 20% of the nominal wall thickness caused by 4.

by degradation.

4-80 Amendment No.-I%+W, !53, ,?06+9 * ,*,

limit. A tube containing a defect is defective.

unserviceable prior to the next inspection.

xial extent of 0.25 inches, or a rential extent of r a through wall degradati ural integrity in the event of an Operating Basis Eart ss of coolant accident, or a steam line or feedwater line break in 4.19.3.c., above.

of the steam generator tube fiom the y to the top of the lower generator shall be determined OPERABLE after co actions (removal fiom service by plugging, or sleeving, or other methods, of all tubes exceeding containing throughwall cracks) required by

a. DELETED 4-8 1 Amendment No. =,!?I, #3+9+#, !53, !5?, B36+% 23tj

c

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. The compiete results of the s e a m generator tube: i n s e n k c to the NRC within 90 da:.s folloxving completion of the enerator breaker closure). The report shall include:

Number and extent of tubes inspected.

-.3 and percent of uall-thickness penetratio indication of an 3.

\

imper ction.

- / ined), bobbin coil amplitude xtent for each inside 4.

5. I 6.

generator,

/ of growth of inside

\ in accordance 1 ID IGA ECR No. TM 0 1-00328, and

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m esults of in-situ pressure testing, if performed.

esults of steam generator tube inspections which fall into Category notification in accordance with 10 CFR 50.72 prior to resumption of plant operation. The written follow-up of this report shall provide a description of investigations conducted 10 determine the cause of the tube degradation and corrective measures taken to prevent recurrence in accordance with 10 CFR 50.73.

irements for inspection of the steam generator tubes ensure that the is portion of the RCS will be maintained.

ce inspection of steam generator tubes is based on Revision 1. In-service inspection of steam gener llance of the conditions of the tubes in the event rogressive degradation due to design, manufact pection of steam generator tubing also provide and cause of any tube degradation so that corr taken.

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The Unit is expected to secondary coolant will be maintained within of the steam generator tubes. If chemistry limits, The extent of steam generator tube le would be limited by the secondary coolant activity, Specification 3.1.6.3.

(primary-to-secondary leakage = 1 gprn). xcess of this limit will require plant shutdown and an unscheduled inspectio repaired or removed from service.

secondary coolant. However, e a defect would develop ce, it will be found during scheduled inservice steam tube examinations. For tu lied to evaluate circumfere determining final disp the tube. For ID IGA indications thro 11 dimension will continue to be assi wall dimension.

for degradation equal to or in excess of 40% of the tube nominal wall

. initiated intergranular degradation may remain in service without % T.

ation morphology has been characterized as not crack-like by diagnostic edd ion and the degradation is of limited circumferential and axial length to e m u 4-83 Amendment No. ??, l??, we,

01-00328) and by successful in-situ pressure testing of Where experience in similar p ocurnented by USNRC Bulletins/'Notices.indicate criti 50?0 of the tubes inspected should be from these critical areas Firs ns sample size may be modified subject to NRC review and approval.

the first sample inspection the requirements of Spe spection, and revision of the Technical Specific y current examination voltages referred to in this section (section 4-83a Amendment No. 47, :29, W+!@ 9 - 3 3 ) ;

4

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\ WLE 4.194,

/

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rotating 6%o f the tubes +team generator the flrst and t f e more .evere than those in t h e other steam generator. Under a

IU ca,

/ circrrmstmcas the rPmple sequence t h n l l be rrrodif'iedto inspect t h e post

'Leacnammt 190.

(12-22-78) 0 L.

TABLE 4.19-2 STEAM GENERATION TUBE INSPECTICM(2)

/

I 1s INSPECTION I I

I I I

I 2ND SWLE INSPECTION I I I I I I d

3RD SAMPLE lNSP TIW I

1 I

I Sample s i z e I

I'Ia;Ly1t\I I ktion Required I I

I Result I I 1 Action Required I I Result I AcHon Required I I I

I I MIA I MIA I I c-1 None I I c-2 1 P1 ug o r r e p a i r I I I I defective tubes and I

I I

I I S.G.

II I I inspect additional 4s

by i n this e r om act on I

I C-2 I Plug o r r e p a i r I defective tubes.

I Perform a c t i o n I

I I I 4'. C-3 I C-3 r e s u l t of r s t I C-3 I f o r C-3 result I I I o f f t r s t sample. I I I c-3 I Inspect a l l I I I tubes i n t h i s I I I I S.G., plug o r I I I I r e p a i r defect- I I I i v e tubes and I I

I I r I I N/A I N/A I I

I I

I I

I I

Notes: (1) S = Yhere N i s the number of steam generators i n the unit, and n i s mber o f steam generators inspected during an inspection.

INSERT TO TS PAGE 4-77 (REVISED TS 4.1 91 4.19 STEAM GENERATOR (SG) TUBE INTEGRITY Armlicability: Whenever the reactor coolant average temperature is above 200°F Surveillance Requirements (SR):

Each steam generator shall be determined to be OPERABLE by performance of the following:

4.19.1 Verify SG tube integrity in accordance with the Steam Generator Program.

4.1 9.2 Verify that each inspected SG tube that satisfies the tube repair criteria is plugged in accordance with the Steam Generator Program prior to exceeding an average reactor coolant temperature of 200°F following an SG tube inspection.

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 systems pressure and inventory. The SG tubes isolate the radioactive fission products in the primary coolant from the secondary system. In addition, as part of the RCPB, the SG tubes are unique in that they act as the heat transfer surface between the primary and secondary systems to remove heat from the primary system. This Specification addresses only the RCPB integrity function of the SG.

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 6.19, Steam Generator (SG) Program, requires that a program be established and implemented to ensure that SG tube integrity is maintained. Pursuant to Specification 6.19, tube integrity is maintained when the SG performance criteria are met. There are three SG performance criteria: structural integrity, accident induced leakage, and 4-77 1 Of7

BASES BACKGROUND (continued) operational leakage. The SG performance criteria are described in Specification 6.1 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).

APPLICABLE The steam generator tube rupture (SGTR) accident is the limiting design SAFETY basis event for SG tubes and avoiding an SGTR is the basis for this ANALYSES Specification. The analysis of a SGTR event assumes a bounding primary to secondary leakage rate associated with a double-ended rupture of a single tube. The accident analysis for a SGTR assumes the contaminated secondary fluid is only briefly released to the atmosphere via safety valves and the majority is discharged to the main condenser.

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 or is assumed to increase to the leakage rates described in TS 6.19.c.2 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 conservatively assumed to be equal to, or greater than, the TS 3.1.4, Reactor Coolant System 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 TS 3.1.1.2.a The LCO requires that SG tube integrity be maintained. The LCO also requires that all SG tubes that satisfy the repair criteria be plugged in accordance with the Steam Generator Program.

During a SG inspection, any inspected tube that satisfies the Steam Generator Program repair criteria is removed from service by plugging. If a tube was determined to satisfy the repair criteria but was not plugged, the tube may still have tube integrity.

In the context of this Specification, a SG tube is defined as the entire length of the tube, including the tube wall and any repairs made to it, 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.

4-78 2 of 7

BASES LCO (continued)

A SG tube has tube integrity when it satisfies the SG performance criteria.

The SG performance criteria are defined in Specification 6.19, 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 burstlcollapse condition to be established. For tube integrity evaluations, except for circumferential degradation, axial thermal loads are classified as secondary loads. For circumferential degradation, the classification of axial thermal loads as primary or secondary loads will be evaluated on a case-by-case basis.

The division between primary and secondary classifications will be based on detailed analysis and/or testing.

Structural integrity requires that the primary membrane stress intensity in a tube not exceed the yield strength for all ASME Code,Section I l l ,

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 Ill, 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 accident analysis assumes that accident induced leakage does not exceed 1 gpm per SG, except for specific types of degradation at specific locations 4-79 3 of 7

BASES LCO (continued) where the NRC has approved greater accident induced leakage. (Refer to TS 6.19.c for specific types of degradation and approved repair criteria.)

The accident induced leakage rate includes any primary to secondary leakage existing prior to the accident in addition to primary to secondary leakage induced during the accident.

The operational leakage performance criterion provides an observable indication of SG tube conditions during plant operation. The limit on operational leakage is contained in TS 3.1.6.3, LEAKAGE, and limits primary to secondary leakage through the SGs to 144 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 when the reactor coolant system average temperature is above 200°F.

RCS conditions are far less challenging when average temperature is at or below 200°F; 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 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.

3.1.1.2.a.(3.)a. and 3.1.1.2.ad3.Ib.

3.1.1.2.a.(3.) 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 Surveillance Requirement 4.19.2. 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 petformance 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, 3.1.1.2.a.(4.) applies.

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BASES ACTIONS (continued)

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

If the evaluation determines that the affected tube(s) have tube integrity, Required Action 3.1.1.2.a.(3.)b. 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 exceeding a reactor coolant average temperature of 200°F 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.

3.1 .I .2.a.(4.1 If the Required Actions and associated Completion Times of Condition 3.1.1.2.a.(3.) are not met or if SG tube integrity is not being maintained, the reactor must be brought to HOT SHUTDOWN within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and COLD SHUTDOWN within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

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 REQUIREMENT SR 4.19.1:

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. l ) , and its referenced EPRl 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 (Le., 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 4-81 5 of 7

BASES

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SURVEILLANCE REQUIREMENTS (continued) specifies the inspection methods to be used to find potential degradation.

Inspection methods are a function of degradation morphology, non-destructive examination (NDE) technique capabilities, and inspection locations.

The Steam Generator Program defines the frequency of SR 4.19.1. 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 6.1 9 contains prescriptive requirements concerning inspection intervals to provide added assurance that the SG performance criteria will be met between scheduled inspections.

SURVEILLANCE REQUIREMENT SR 4.1 9.2:

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 6.1 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.

Tubes with inside diameter (ID) initiated intergranular degradation may remain in service without percent throughwall sizing if the degradation has been characterized as not crack-like by diagnostic eddy current inspection and if the degradation is of limited circumferential and axial length to ensure tube structural integrity. Additionally, serviceability for accident leakage under the limiting postulated Main Steam Line Break (MSLB) accident will be evaluated by determining that this ID initiated degradation mechanism is inactive (e.g., comparison of the outage examination results with the results from past outages meets the requirements of AmerGen Engineering Report ECR No. TM 01-00328) and by successful in-situ pressure testing of a sample of these degraded tubes to evaluate their accident leakage potential when in-situ pressure tests are performed.

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The frequency of prior to exceeding an average reactor coolant temperature of 200°F following an SG tube 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 I l l , Subsection NB.

5. Draft Regulatory Guide 1.I21, Basis for Plugging Degraded Steam Generator Tubes, August 1976.

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

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'3 6.9.5 CORE OPERATING LIMITS REPORT i

i 6.9.5.1 The core operating limits addressed by the individual Technical Specifications shall be established and documented in the CORE OPERATING LIMITS REPORT prior to each reload cycle or prior to any remaining part of a reload cycle.

6.9.5.2 The analytical methods used to determine the core operating limits addressed by the individual Technical Specifications shall be those previously reviewed and approved by the NRC for use at TMI-1, specifically:

(1) BAW-10179 P-A, "Safety and Methodologyfor Acceptable Cycle Reload Analyses." The current revision level shall be specified in the COLR.

(2) TR-078-A, TMI-1 Transient Analyses Using the RETRAN Computer Code", Revision 0. NRC SER dated 2/10/97.

(3) TR-087-A, YMI-1 Core Thermal-Hydraulic Methodology Using the VIPRE-01 Computer Code", Revision 0. NRC SER dated l a 19/96.

(4) TR-09t-A, "Steady State Reactor Physics Methodology for TMI-1").

Revision 0. NRC SER dated 2/21/96.

(5) TR-092P-A, "TMI-1 Reload Design and Setpoint Methodology",

Revision 0. NRC SER dated 4/22/97.

(6) BAW-10227P-AI "Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel", NRC SER dated February 4,2000. I 6.9.5.3 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 transientiaccident analysis limits) of the safety analysis are met.

6.9.5.4 The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon.issuance for each reload cycle to the NRC Document Control Desk with copies to the Regional Administrator and Resident Inspector.

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d S.f E C T i O d R c P 4 T

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k.4, G S ~ E ~G H ~ A J ~ A TUBE~ R ~ IR I ,

6-19 j

Amendment N0.72~77,

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INSERT TO TS PAGE 6-19 6.9.6 STEAM GENERATOR TUBE INSPECTION REPORT A report shall be submitted within 90 days after the average reactor coolant temperature exceeds 200°F following completion of an inspection performed in accordance with Section 6.19, Steam Generator (SG) Program. The report shall include:

a. The scope of inspections performed on each SG,

b. Active 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 active degradation mechanism,

f. Total number and percentage of tubes plugged to date,

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

h. The effective plugging percentage for all plugging and tube repairs in each SG,

i. Repair method utilized and the number of tubes repaired by each repair method, if any,

j. Location, bobbin coil depth estimate (if determined), bobbin coil amplitude (if determined), and axial and circumferential extent for each inside diameter (ID) IGA indication.

k. An assessment of growth of inside diameter IGA degradation in accordance with the volumetric ID IGA management program contained in AmerGen Engineering Report, ECR No. TM 01-00328.

I. The information specified for reporting in ECR No. 02-01121, Rev.2.

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b. Licensees may make changes to Bases without prior NRC approval provided I

i 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 6.18.b.l or 6.18.b.2 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).

\

s r ~ k f GEtd&khTik l (SG) /%GRAM 6-26 Amendment No.*

1

INSERT TO TS PAGE 6-26 6.19 STEAM GENERATOR (SGI 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.O 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. Leakaae is not to exceed 1 aDm Der SG, exceDt for sDecific twes of dearadation at sDecific locations as described in DaraaraDh c of the Steam Generator Proaram below.

3. The operational leakage performance criterion is specified in TS 3.1.6, 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.

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

depth based criteria:

1. Volumetric Inside Diameter (ID) Inter-Granular Attack (IGA) indications may be dispositioned in accordance with ECR No. TM 01-00328. MSLB accident-induced leakage rates are limited to less than 1 gpm under the report. (ECR No. TM 01-00328 is not applicable to tube sleeves nor the parent tubing spanned by the sleeves.) ID IGA indications shall be repaired or removed from service if they exceed an axial extent of 0.25 inches, or a circumferential extent of 0.52 inches, or a through wall degradation dimension of 1 4 0 % if assigned.

2. Upper tubesheet kinetic expansion indications may be dispositioned in accordance with ECR No. TM 02-01121, Rev. 2. MSLB accident-induced leakage is limited to less than 3228 gallons for the initial 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, and 9960 gallons over the MSLB duration, under this report.

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, 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. An assessment of degradation shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations.

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

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

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

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

4. Implementation of the repair criteria for ID IGA requires 100% bobbin coil inspection of all non-plugged tubes in accordance with AmerGen Engineering Report, ECR No.

TM 01-00328. ID IGA indications detected by the bobbin coil probe shall be characterized using rotating coil probes, as defined in that report.

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5. Implementation of the repair criteria for kinetic expansion indications requires 100°/~

rotating probe inspection of the required lengths of the kinetic expansions in all non-plugged, non-sleeved, tubes in accordance with AmerGen Engineering Report, ECR No. TM 02-01121, Rev.2.

e. Provisions for monitoring operational primary to secondary leakage.

f. Provisions for SG tube repair methods. Steam generator tube repair methods shall provide the means to reestablish the RCS pressure boundary integrity of SG tubes without removing the tube from service. For the purposes of these Specifications, tube plugging is not a repair. All acceptable tube repair methods are listed below.

(None.)

NOTE: Refer to Section 6.9.6 for reporting requirements for periodic SG tube inspections.

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