ML061420294

Technical Specification Change Request No. 331 - Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity

ML061420294
Person / Time
Site:	Three Mile Island
Issue date:	05/15/2006
From:	Cowan P B AmerGen Energy Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
5928-06-20390, BL-06-001
Download: ML061420294 (47)

Text

Amner Gen SM AmerGen Energy Company, LLC www.exelonCorp.COM An Exelon Company 200 Exelon Way Kennett Square, PA 19348 10 CFR 50.90 May 15, 2006 5928-06-20390 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Three Mile Island Nuclear Station, Unit 1 Facility Operating License No. DPR-50 NRC Docket No. 50-289

Subject:

Technical Specification Change Request No. 331 -Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity In accordance with 10 CFR 50.90, "Application for amendment of license or construction permit," AmerGen Energy Company, LLC (AmerGen) proposes changes to Appendix A, Technical Specifications (TS), of the Three Mile Island Nuclear Station, Unit 1 (TMI Unit 1), Facility Operating License.The proposed change would revise the TS requirements related to steam generator tube integrity.

The change is generally consistent with NRC-approved Revision 4 to Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-449,"Steam Generator Tube Integrity." The availability of this TS improvement was announced in the Federal Register, on May 6, 2005 (70 FR 24126) as part of the consolidated line item improvement process (CLIIP). This amendment request satisfies the TMI Unit 1 commitment to modify the steam generator sections of the TS, consistent with TSTF-449, Revision 4, as described in the TMI Unit 1 response to NRC Generic Letter 2006-01, dated February 16, 2006.The subject changes are generally consistent with the changes outlined in TSTF-449, Revision 4. Minor differences between the proposed changes and those of TSTF-449, Revision 4 are described in Enclosure

1. The TMI Unit 1 TS are currently based on a custom TS format rather than on NUREG-1430, which is the format of TSTF-449, Revision 4. Therefore, adaptation of TSTF-449, Revision 4 was required.As is prescribed by the TSTF-449, Revision 4, the TMI Unit 1 proposed Technical Specification Section 6.19, "Steam Generator (SG) Program," allows continued use of the existing TMI Unit 1 kinetic expansion inspection and repair criteria, ID Volumetric IGA inspection and repair criteria, and existing kinetic expansion and sleeve repairs.The TMI Unit 1 SGs are planned to be replaced at the end of operating cycle 17 in the Fall of 2009, and a separate license amendment request will be submitted to remove the sections of the plant's TS's that are not required, or are not applicable, for the replacement SGs.44o (

5928-06-20390 May 15, 2006 Page 2 This proposed amendment request is subdivided as follows: 1. Enclosure 1 provides a description of the proposed change and confirmation of applicability

2. Enclosure 2 provides "mark-ups" of the existing License, Technical Specifications, and Bases pages to show the proposed changes.Using the standards in 10 CFR 50.92, AmerGen has concluded that these proposed changes do not constitute a significant hazards consideration, as described in the enclosed analysis performed in accordance with 10 CFR 50.91 (a)(1). Pursuant to 10 CFR 50.91 (b)(1), a copy of this Technical Specification Change Request is provided to the designated official of the Commonwealth of Pennsylvania, Bureau of Radiation Protection, as well as the chief executives of the township and county in which the facility is located.We request approval of the proposed change by May 31, 2007, with the amendment being implemented within 120 days of issuance.

This will allow an orderly implementation of these changes prior to the Fall 2007 refueling outage (1 R1 7) for TMI Unit 1. Note that the proposed amendment is not currently expected to increase or decrease the planned TMI Unit 1 1 R1 7 outage steam generator inspection scope.These proposed changes have been reviewed by the Plant Operations Review Committee and approved by the Nuclear Safety Review Board.No new regulatory commitments are established by this submittal.

If you have any questions or require additional information, please contact David J. Distel at (610) 765-5517.I declare under penalty of perjury that the foregoing is true and correct.Respectfully, Executed on P61l5C NOI"l I E Pamela B. Co an Director -Licensing and Regulatory Affairs AmerGen Energy Company, LLC

Enclosures:

1) TMI Unit 1 Technical Specification Change Request No. 331 Description and Assessment

2) TMI Unit 1 Technical Specification Change Request No. 331 Markup of Proposed License, Technical Specifications, and Bases Page Changes cc: S. J. Collins, Administrator, USNRC Region I D. M. Kern, USNRC Senior Resident Inspector, TMI Unit 1 F. E. Saba, USNRC Project Manager, TMI Unit 1 D. Allard, Director, Bureau of Radiation Protection

-PA Department of Environmental Resources Chairman, Board of County Commissioners of Dauphin County Chairman, Board of Supervisors of Londonderry Township File No. 06007 ENCLOSURE 1 TMI Unit 1 Technical Specification Change Request No. 331 Application for TS Improvement Regarding Steam Generator Tube Integrity Using the Consolidated Line Item Improvement Process Description and Assessment ENCLOSURE 1 DESCRIPTION AND ASSESSMENT

1.0 INTRODUCTION

In accordance with 10 CFR 50.90, "Application for amendment of license or construction permit," AmerGen Energy Company, LLC (AmerGen) proposes changes to the License and the License's Appendix A, Technical Specifications (TS), for the Three Mile Island Nuclear Station, Unit 1 (TMI Unit 1).The proposed license amendment revises the requirements in TS related to steam generator tube integrity.

The changes are generally consistent with NRC approved Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-449,"Steam Generator Tube Integrity," Revision 4. The availability of this TS improvement was announced in the Federal Register on May 6, 2005, as part of the Consolidated Line Item Improvement Process (CLIIP).The proposed amendment also revises the TMI Unit 1 Facility Operating License to delete several out-of-date items that also pertain to the plant's steam generator tube integrity.

These license conditions proposed for deletion were originally incorporated into the license in the 1980's and are no longer required.

The license condition that prescribes the plant's primary-to-secondary leak rate limit is moved from the license to Section 3.1.6 of the TS. This relocation is consistent with the TSTF-449, Revision 4.

2.0 DESCRIPTION

OF PROPOSED AMENDMENT Consistent with the NRC-approved Revision 4 of TSTF-449, the proposed TMI Unit 1 TS changes include:* Revised License Pages to Delete License Condition 2.c.(8), Repaired Steam Generators," in order to remove out of date requirements.

The plant's primary-to-secondary leak rate limit from this section is moved to TS Section 3.1.6.* Revised TS 3.1.6, "LEAKAGE," to revise that Section's primary-to-secondary leak rate limit to be consistent with that from the License Condition described above.* Revised TS Section 3.1.1.2, "Steam Generators and Steam Generator (SG) Tube Integrity"* Revised TS Section 4.19, "Steam Generator (SG) Tube Integrity"* New TS 6.9.6, "Steam Generator Tube Inspection Report"* New TS 6.19, "Steam Generator (SG) Program" The proposed TS changes are generally consistent with TSTF-449, Revision 4. Minor differences between the proposed changes and those contained in TSTF-449, Revision 4 include:* Changes were made to several paragraphs, formats, Section numbers, and plant operating mode descriptions because the TSTF was written for plants that have Standard Technical Specifications.

The TMI Unit 1 Technical Specifications are not written in the format or nomenclature of the Standard Technical Specifications.

Enclosure 1 Description and Assessment Page 2 of 7* The primary-to-secondary leak rate limit of TSTF-449, Revision 4 was not adopted since the TMI Unit 1 plant currently has a leak rate limit of 144 gallons per day (GPD) total for both steam generators.

This existing TMI-1 limit is more conservative than the limit of 150 GPD for each steam generator proposed by TSTF-449, Revision 4. Note that the TMI-1 primary-to-secondary leak rate limit, while unchanged, is moved from the License to Technical Specification 3.1.6.3 by the proposed change.* The TSTF-449, Revision 4 changes to the TS "LEAKAGE" Definition term are addressed in TMI Unit 1 TS 3.1.6, where applicable, since the TMI Unit 1 TS do not contain a definition for "LEAKAGE." Accordingly, the term "Leakage" is not capitalized as a TS Definition throughout the TMI Unit 1 proposed changes.* The existing TMI Unit 1 TS Sections 3.1.6.3 and 3.1.6.4 define that SG leakage is considered primary-to-secondary leakage. The existing TMI Unit 1 TS Table 4.1-2 surveillance frequency for primary to secondary leakage is added, and is specified as Daily beginning no later than 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, which is consistent with current requirements and is more conservative than the TSTF-449, Revision 4 requirement of every 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.* The proposed TMI Unit 1 TS 4.19 Bases revises the corresponding TSTF-449, Revision 4 Bases statement to clarify that the TMI Unit 1 licensing basis Steam Generator Tube Rupture (SGTR) event, as evaluated in Updated Final Safety Analysis Report (UFSAR)Section 14.1.2.10, assumes a bounding primary-to-secondary leakage rate associated with the double-ended rupture of a single tube, and did not necessarily add the initial leak rate allowed by TS, which is insignificant compared to the leakage associated with the single tube rupture. (The TMI-1 double-ended tube rupture analysis assumed a leak rate of 435 gpm. The initial leak rate of up to 1 gpm is insignificant in comparison.)

• TMI-1 has retained its current TS requirement to submit its outage tube inspection report within 90 days of plant startup after an outage (main generator breaker closure), rather than adopting the 180 days after initial entry into Mode 4 (hot shutdown) reporting requirement of the TSTF. The 90-day reporting requirement is more conservative than the TSTF requirement.

TMI has retained the 90-day requirement in the TS since this duration is also documented in the plant's two alternate tube repair criteria for its current SGs.* TMI-1 has retained its more restrictive requirement that steam generator tube integrity apply whenever the reactor coolant average temperature is greater than 250 0 F. Under this change, the TSTF-449, Revision 4 SG tube integrity specifications have been incorporated into the existing TS Section 3.1.1.2 where the 250 0 F limit was previously described.

The existing TMI Unit 1 SGs are currently planned to be replaced at the end of operating cycle 17 in the Fall of 2009. A separate license amendment request will be submitted to revise the TMI Unit 1 TS for the replacement SGs.Proposed revisions to the TS Bases are also included in this application.

As discussed in the NRC's model safety evaluation, adoption of the revised TS Bases associated with TSTF-449, Revision 4 is an integral part of implementing this TS improvement.

The TSTF-449, Revision 4 Bases for SG tube integrity and associated surveillance requirements have been combined for Enclosure 1 Description and Assessment Page 3 of 7 clarity. The changes to the affected TS Bases pages will be incorporated in accordance with the TS Bases Control Program.

3.0 BACKGROUND

The background for the proposed TS changes is adequately addressed by the NRC Notice of Availability published on May 6, 2005 (70 FR 24126), the NRC Notice for Comment published on March 2, 2005 (70 FR 10298), and TSTF-449, Revision 4.Additional background applicable to the proposed License Condition 2.c.(8) change is as follows: There are five conditions from the plant's 1981 to 1984 steam generator repairs that are proposed for deletion:* The first condition stipulated that: "...the licensee shall submit to NRC the results of the steam generator hot test program." The hot test program for the steam generator repairs was run in 1983. The results of those tests were submitted to the NRC and are discussed in References 1 and 2, below.* The second condition stipulated that TMI-1 establish a "baseline" primary-to-secondary leakage rate for the repaired steam generators, and that the plant be shutdown if the leakage rate is increased greater than that baseline leakage rate by more than 0.1 gpm. The baseline leakage rate for the TMI-1 steam generators has been 0.0 gpm for the last several operating cycles, so the net result of this condition has been that the plant has operated with an allowable operational primary-to-secondary leakage rate limit of 0.1 gpm. As described above, this allowable leakage rate limit of 0.1 gpm is being retained by the proposed change to TS Section 3.1.6.3. The 0.1 gpm limit is retained; and is relocated to TS Section 3.1.6.3 in the proposed TS changes.* The third condition required that TMI-1 complete a post-critical test program during the plant's restart after the steam generator tube repairs. The test program was completed during the 1985 restart of the plant (Reference 7).* The fourth condition required eddy current examinations to be performed as specified in NUREG-1019, and that assessments be performed based on the results of those eddy current examinations.

The examinations were completed after the restart and the results were reported in Reference 3, below.* The fifth condition required that long-term corrosion "lead tests" results be reported on a regular basis, and any adverse results be reported in a timely fashion. These lead tests were completed and the results were reported to the NRC (References 4, 5, and 6, below).

References:

1. GPU Nuclear Technical Data Report (TDR) 488, Rev. 0, "TMI-1 OTSG Hot Testing Results and Evaluation," October 1983.

Enclosure 1 Description and Assessment Page 4 of 7 2. U.S. Nuclear Regulatory Commission Atomic Safety and Licensing Board, Memorandum and Order, (Three Mile Island Nuclear Station, Unit No. 1) (Steam Generator Repair), June 1, 1984.3. GPU Nuclear Letter # 5211-86-2109, H. D. Hukill to Office of Nuclear Reactor Regulation (Attn: J. F. Stolz), "Post Eddy Current Inspection Report," June 17,1986.4. GPU Nuclear Letter # 5211-85-2106, H. D. Hukill to Office of Nuclear Reactor Regulation, (Attn: J.F. Stolz), U.S. Nuclear Regulatory Commission, "Final Report on the Long Term Corrosion Test Program," July 29,1985.5. U.S. Nuclear Regulatory Commission Letter, G. Edison to H. D. Hukill, "Long Term Corrosion Test Program for Three Mile Island Unit 1 (TAC # 59435)," May 11, 1987.6. GPU Nuclear Letter # 5211-87-2158

/ 5000-87-1351, R. F. Wilson to U. S.Nuclear Regulatory Commission, "Long Term Corrosion Test Program for Three Mile Island Unit 1," August 24,1987.7. TMI Unit Test Procedure 800/1, "Controlling Procedure for Power Escalation", Revision STR-1, February 13, 1986.4.0 REGULATORY REQUIREMENTS AND GUIDANCE The applicable regulatory requirements and guidance associated with this application are adequately addressed by the NRC Notice of Availability published on May 6, 2005 (70 FR 24126), the NRC Notice for Comment published on March 2, 2005 (70 FR 10298), and TSTF-449, Revision 4.5.0 TECHNICAL ANALYSIS AmerGen has reviewed the safety evaluation (SE) published on March 2,2005 (70 FR 10298)as part of the CLIIP Notice for Comment. This included the NRC staff's SE, the supporting information provided to support TSTF-449, and the changes associated with Revision 4 to TSTF-449.

AmerGen has concluded that the justifications presented in the TSTF proposal and the SE prepared by the NRC staff are applicable to TMI Unit 1 and justify this amendment for the incorporation of the changes to the TMI Unit 1 TS, recognizing that the SE must address the plant specific differences discussed herein.6.0 REGULATORY ANALYSIS A description of this proposed change and its relationship to applicable regulatory requirements and guidance was provided in the NRC Notice of Availability published on May 6, 2005 (70 FR 24126), the NRC Notice for Comment published on March 2, 2005 (70 FR 10298), and TSTF-449, Revision 4.

Enclosure 1 Description and Assessment Page 5 of 7 6.1 Verification and Commitments The following information is provided to support the NRC staff's review of this amendment application:

Plant Name, Unit No. Three Mile Island Nuclear Station, Unit No. 1 Steam Generator Model(s):

B&W 177 FA Effective Full Power Years (EFPY) of service for currently installed SGs Approximately 21 EFPY at last inspection In Fall, 2005 Tubing Material Alloy 600 MA Number of tubes per SG 15531 Number and percentage of tubes plugged "A" SG: 1661 tubes plugged = 10.9% (with sleeve contribution) in each SG "B" SG: 871 tubes plugged = 5.9 % (with sleeve contribution)

Number of tubes repaired in each SG "A" SG: 247 sleeved tubes In service"B" SG: 252 sleeved tubes in service Degradation mechanism(s) identified PWSCC, ID Volumetric IGA, IGSCC from primary sulfur Intrusion, OD IGA, High Cycle Fatigue, Outside Diameter Stress Corrosion Cracking (ODSCC), Tube-to-Tube Support Plate Wear Fretting, Severed Plugged Tube-to-Tube Wear Current primary-to-secondary leakage limits: per SG: N/A Total: 0.1 GPM (144 GPD)Leakage is evaluated at what temperature condition?

Room temperature Approved Altemate Tube Repair Criteria Provide for each: (ARC): -Approved by [amendment number dated 2I-Applicability (e.g., degradation mechanism, location)-any special limits on allowable accident leakage any exceptions or clarifications to the structural performance criteria that apply to the ARC 1. Kinetic Expansion Inspection and 1. Approved by NRC SER, dated November 8, 2005, USNRC Repair Criteria Letter (TAC No. MC7001)-Applicable to upper tubesheet kinetic expansions only.-Leakage limits are discussed on Page 8 of the above-referenced NRC letter and are 3228 gallons over the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of, or 9960 gallons over the duration of, a MSLB accident.(This leakage Is calculated at 579 F, and was reviewed under Amendment No. 204 of the TMI License.)-Structural limits were based on MSLB-induced delta P's and axial, tensile loads.2. ID Volumetric IGA Inspection and 2. -Approved by NRC TS Amendment No. 237, dated October 5, Repair Criteria 2001 (TAC No. MB0664)-Applicable to Indications with ID Volumetric IGA morphology only; applicable to unexpanded tubing and the lower 5" of a subset of 22" upper tubesheet kinetic expansions only.-Accident leakage within 1 gpm for MSLB. Leakage calculated at 600F.-Structural limits were based on MSLB-induced delta P's and axial, tensile loads.

Enclosure 1 Description and Assessment Page 6 of 7 Approved SG Tube Repair Methods 1. 80" Sleeves 2. Upper Tubesheet Kinetic Expansions Provide for each:-Approved by [amendment number dated l-Applicability limits, if any Sleeve repair criteria (e.g., 40% of the initial sleeve wall thickness)

1. -Alloy 690 non-welded, mechanical sleeves were installed as preventive measure In early 1990's. Sleeving was performed under 1ICFR50.59.

NRC approval of sleeve material was provided by USNRC letter, J. F. Stoltz to G. T. Broughton,"Request for Proposed Alternative to Requirements of ASME Section III for the Three Mile Island Unit I (TAC No. 75932)," dated February 26, 1991.-Applicable to 80" upper tubesheet sleeves only.-Sleeve repair criterion Is 'plug on detection'.

This was a commitment In ECR TM 02-01121, which was approved by NRC SER, dated November 8, 2005, USNRC Letter (TAC No.MC7001)-Note that these sleeve repairs are listed for historical accuracy; TMI-1 will not install additional sleeves without NRC approval. (There are no currently approved repair methods for Installation at TMI-1.)2. -Upper tubesheet kinetic expansions were Installed in all inservice tubes In the early 1980's. This was approved by NUREG 1019, "USNRC SER Related to Steam Generator Tube Repair and Return to Operation", November 1983.-Note that these kinetic expansion repairs are listed for historical accuracy; TMI-1 will not Install additional kinetic expansions without NRC approval. (There are no currently approved repair methods for Installation at TMI-1.)Performance criteria for accident leakage -Primary to secondary leak rate values assumed in licensing basis accident analysis, including assumed temperature conditions As described In "Approved Alternate Tube Repair Criteria (ARC)" row above, primary-to-secondary leak rate values and temperatures for MSLB are: 3228 gallons over the first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> of, or 9960 gallons over the duration of, a MSLB accident. (This leakage Is calculated at 579 F.)Other accidents analyzed In the UFSAR (where primary-to-secondary leak rate was pertinent) used 1 gpm. A 1 gpm limit was also used for the plant's ID Volumetric IGA Inspection and Repair Criteria, as described in the "Approved Alternate Tube Repair Criteria (ARC)" row, above.

Enclosure 1 Description and Assessment Page 7 of 7 7.0 NO SIGNIFICANT HAZARDS CONSIDERATION AmerGen has reviewed the proposed no significant hazards consideration determination published on March 2, 2005 (70 FR 10298) as part of the CLIIP. AmerGen has concluded that the proposed determination presented in the notice is applicable to TMI Unit 1, recognizing that the determination must address the plant specific differences discussed herein, and the determination is hereby incorporated by reference to satisfy the requirements of 10 CFR 50.91 (a).8.0 ENVIRONMENTAL EVALUATION AmerGen has reviewed the environmental evaluation included in the model SE published on March 2, 2005 (70 FR 10298) as part of the CLIIP. AmerGen has concluded that the staff's findings presented in that evaluation are applicable to TMI Unit 1 and the evaluation is hereby incorporated by reference for this application.

9.0 PRECEDENT

This application is being made in accordance with the CLIIP. AmerGen is not proposing significant variations or deviations from the TS changes described in TSTF-449, Revision 4, or the NRC staff's model SE published on March 2,2005 (70 FR 10298). AmerGen has determined that the minor differences between the proposed TS changes and the TSTF-449, Revision 4, described above, do not adversely impact the TS limiting conditions for operation, action statements, or surveillance requirements imposed by TSTF-449, Revision 4.Additionally, AmerGen has determined that these minor differences have no impact on the conclusions of the NRC No Significant Hazards Consideration or model SE, as identified above.The TMI Unit 1 TS are currently based on a custom TS format rather than on NUREG-1430, which is the format of TSTF-449, Revision 4. Therefore, adaptation of TSTF-449, Revision 4 was required.

10.0 REFERENCES

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

ENCLOSURE 2 TMI Unit 1 Technical Specification Change Request No. 331 Markup of Proposed License, Technical Specifications, and Bases Page Changes Revised License Pages 6 7 Revised Technical Specifications

& Bases Paaes Table of Contents Page iv Table of Contents Page v Table of Contents Page vi 3-la 3-2 3-12 3-15a 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 CONTROLLED COPY-6 -(8) Repaired Steam Generators

\ In order to confirm the leak-tight integrity of the Reactor Coolant System, includin the steam generators, operation of the facility shall be in accordance with the following:

1. Prior to initial criticality, the licensee shall submit to NRC the results o he steam generator hot test program and a summary of its management revi 2. he licensee shall confirm baseline primary-to-secondary leage rate e blished during the steam generator hot test program. I eakage exceeds the seline leakage rate by more than 0.1 gpm', the fac y shall be shut down and e tested. If any increased leakage above base*e is due to defects in the tube ee span, the leaking tube(s) shall be rem g ed from service. The baseline le age shall be re-established, provide hat the leakage limit of Technical Sp ification 3.1.6.3 is not exceede 3. The licensee shal omplete its post-critica est program at each power range (0-5%, 5%-50%, 500 100%) in confo nce with the program described in Topical Report 008, Re 3, and shal ave available the results of that test program and a summary its man ement review, prior to ascension from each power range and priorn n al power operation.

4. The licensee shall conduct y rrent examinations, consistent with the extended inservice inspe on plan fined in Table 3.3-1 of NUREG-1019, either 90 calendar day fter reachin ull power, or 120 calendar days after exceeding 50% pow operation, which er comes first. In the event of plant operation for an emended period at less th 50% power, the licensee shall provide an asse sment at the end of 180 da of operation at power levels between 5% d 50%, such assessment to con in recommendations and supporting iformation as to the necessity of a sp ial eddy-current testing (ECT) s tdown before the end of the refueling cyci (The NRC staff will evalu that assessment and determine the time of th next eddy-current exa nation, consistent with the other provisions of the Ii nse conditions.)

In absence of such an assessment, a special ECT shutdo shall take place efore an additional 30 days of operation at power above 5%.*If leakge exceeds the baseline leakage rate by more than 0.1 gpm during the rem *nder of the cle 8 operation, the facility shall be shutdown and leak tested. Operation at lea ge rags of up to 0.2 gpm above the baseline leakage rate shall be acceptable during the mainder of Cycle 8 operation.

After the 9R refueling outage, the leakage limit and accompanying shutdown requirements revert to 0.1 gpm above the baseline leakage rate.Amendment No. q09, 163 Amendment No-2W-14111, CONTROLLED COPY-7 -(9) Long Range Planning Program -Deleted Sale and License Transfer Conditions (10) Deleted (11) Deleted (12) Deleted (13) Deleted Amendment No. +9I, 2ff, ei, 2S8, N,-<- I CONTROLLED COPY TABLE OF CONTENTS Section Pace 4.8 4.9 4.9.1 4.9.2 4.10 4.11 4.12 4.12.1 4.12.2 4.12.3 DELETED DECAY HEAT REMOVAL (DHR) CAPABILITY

-PERIODIC TESTING REACTOR COOLANT SYSTEM (RCS) TEMPERATURE GREATER THAN 250 DEGREES F RCS TEMPERATURE LESS THAN OR EQUAL TO 250 DEGREES F REACTIVITY ANOMALIES REACTOR COOLANT SYSTEM VENTS AIR TREATMENT SYSTEMS EMERGENCY CONTROL ROOM AIR TREATMENT SYSTEM REACTOR BUILDING PURGE AIR TREATMENT SYSTEM (DELETED)AUXILIARY AND FUEL HANDLING BUILDING AIR TREATMENT SYSTEM (DELETED)FUEL HANDLING BUILDING ESF AIR TREATMENT SYSTEM RADIOACTIVE MATERIALS SOURCES SURVEILLANCE DELETED MAIN STEAM SYSTEM INSERVICE INSPECTION REACTOR INTERNALS VENT VALVES SURVEILLANCE SHOCK SUPPRESSORS (SNUBBERS)

FIRE PROTECTION SYSTEMS (DELETED)4-51 4-52 4-52 4-52a 4-53 4-54 4-55 4-55 4-55b 4-55d 4-55f 4-56 4-56 4-58 4-59 4-60 4-72 I 4.12.4 4.13 4.14 4.15 4.16 4.17 4.18--v* 4.19 U I 5U I ULD II _LI ,V~. I~s .Lj I lure 4 7 4.19.1 CrEAM GSPIERAT-OR SAMPLE SEL1ECTIONI ANID 01N12ECT!N 4-77 4.49.2 STE..AM GENERATOR TUBE SAMPLE GELEGTIO91 AND INSPECTiati C-77 4.10.3 INGPECTION FRfEQUENGIEZ 2+-19 4.1 .4 ACCEPTANCE CflITEfXIA 4 Be 4.9. EPOfRT- 4 01 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 USE CENSUS (DELETED) 4-87 4.23.3 INTERLABORATORY COMPARISON PROGRAM (DELETED) 4-87"d i,,I , STE'1 cvEA6E#LToA (SGL) 7C-4 /'I t65 irl'9-? ' I iv Amendment No. 11,22,30,11,147,65,72,78,95,97,119,1422,129, 137,11 17 4A7e 212,215, 24C1-eA8 CONTROLLED COPY TABLE OF CONTENTS Section Page 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 REVIEW 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 6.9.3 ANNUAL RADIOLOGICAL ENVIRONMENTAL OPERATING REPORT 6-17 6.9.4 ANNUAL RADIOACTIVE EFFLUENT RELEASE REPORT 6-18 6.9.5 CORE OPERATING LIMITS REPORT 6-19 6.10 RECORD RETENTION 6-20 6.11 RADIATION PROTECTION PROGRAM 6-22 6.12 HIGH RADIATION AREA 6-22 6.13 PROCESS CONTROL PROGRAM 6-23 6.14 OFFSITE DOSE CALCULATION MANUAL (ODCM) 6-24 6.15 DELETED 6-24 6.16 DELETED 6-24 6.17 MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS 6-25 6.18 TECHNICAL SPECIFICATION (TS) BASES CONTROL PROGRAM 6-25-v-Amendment No. 11, 17, 72, 77, 128, 160,173, 212,262,2-263 56-,. 6 Ase4v (7C- c Tue s delve V-/? I CONTROLED COPY LIST OF TABLES TABLE 1.2 2.3-1 3.1.6.1 3.5-1 3.5-1 A 3.5-2 3.5-3 3.5-4 3.21-1 3.21-2 3.23-1 3.23-2 4.1-1 4.1-2 4.1-3 4.1-4 4.19-1 4.19-2 4.21-1 4.21-2 4.22-1 4.22-2 4.23-1 TITLE Frequency Notation Reactor Protection System Trip Setting Limits Pressure Isolation Check Valves Between the Primary Coolant System and LPIS Instruments Operating Conditions DELETED Accident Monitoring Instruments Post Accident Monitoring Instrumentation Remote Shutdown System InstnnmenItation and Control DELETED DELETED DELETED DELETED Instrument Surveillance Requirements Minimum Equipment Test Frequency Minimum Sampling Frequcncv Post Accident Monitoring Instrumentation Minimum Numbe r e-f -Steamn Generatom ob XrD Inspected During Insen ice Inspection Stcaem Cencr-atr.

TUbe Inspeetion DELETED DELETED DELETED DELETED DELETED PAGE 1-8 2-9 3-15a 3-29 3-40c 3-40d 3-40i I 4-3 4-8 4-9 4444-vi Amendment No. 59. 1, I t n 11n 4 I 3, 4 4,l9-7f,

'CONTROLLED COPY 3.1 REACTOR COOLANT SYSTEM 3.1.1 OPEATIONAL COMPONENTS Acolicabilitv Applies to the operating status of reactor coolant system components.

Objective To specify those limiting conditions for operation of reactor coolant system components wnich must be met to ensure safe reactor operations.

Soecification 3.1.1.1 Reactor CoOlant Pumps a. Pump combinations permissible for given power levels small be as snown in Specification Table 2.3.1.b. Power cperation with one idle reactor coolant pump in each loop shall be restricted to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. If tre -reactor is not returned to an acceptable RC pump operating combination at the end of the 24-hour perioc, the reactor shall be in a hot shutdown condition within 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 concentration in the reactor coolant system shall not be reduced unless at least one reactor coolant pump or one decay heat removal pump is circulating reactor coolant.3.1.1.2 Steam Generators n Star(5&) T-n -r a. Both steam generators shall be coerable wtenever tie reactor coolant average temperature is above 250 0 F.3.1.1.3 Pressurizer Safety Valves a. The reactor shall not remain critical unless both pressurizer code safety valves are operable with a lift setting of 2500 psig 1%.D. When the reactor is subcritical, at least one pressurize:

code safety valve snail De operable if all reactor coolant system openings are closed, except for hydrostatic tests in accordance with ASME Soiler and Pressure Vessel Ccde,Section III.3-la lAmendment No. Hi, 7, U, %7 , d---~

INSERT TO TS PAGE 3-1a (REVISED TS 3.1.1.2)b. Whenever the reactor coolant average temperature is above 250 0 F, the following conditions are required: (1.) SG tube integrity shall be maintained.

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

NOTE-----------------------------------------------------------

Separate Condition entry is allowed for each SG tube.(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.

b. Plug the affected tube(s) in accordance with the Steam Generator Program prior to exceeding a reactor coolant average temperature of 250 0 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 STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> of detection 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 /> of detection.

1 of 1 CONTROLLED COPY Bases The limitation on power operation with one idle RC pump in each loop has been imposed since the ECCS cooling performance has not been calculated in accordance 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 allowed for 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 expected to have considerable margin for the peak cladding temperature limit and since the likelihood of a LOCA within the 24-hour period is considered verv remote.A reactor coolant pump or decay heat removal pump is required to be in operation before the boron concentration is reduced by dilution with makeup water. Either pump will provide mixing which will 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 system can remove decay heat when the reactor coolant system is below this temperature (References 1, 2, and 3).Both steam generators must be operable before heatup of the Reactor Coolant System to insure system integrity against leakage under normal and transient conditions.

Only one steam generator is required for decay heat removal purposes.One pressurizer code safety valve is capable of preventing overpressurization when 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 criticality 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 +/- 1% allowance for error. Surveillance requirements are specified in the Inservice 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 safety analysis.Reference Sef-,en q,17 Her Aqs~es fc 6 ec ncr: (I) UFSAR, Tables 9.5-1 and 9.5-2 (2) UFSAR, Sections 4.2.5.1 and 9.5 -"Decay Heat Removal" (3) UFSAR, 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 Protection" 3-2 Amendment No. 4. (12/22/78),-44

-2e-CONTROLLED COPY 3.1.6 LEAKAGE Applicability Applies to reactor coolant leakage from the reactor coolant system and the makeup and purification system.Oblective 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 one gpm or if any reactor coolant leakage is evaluated as unsafe the reato r 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.

O./fm (IWqepL 3.1.6.3 If primary-to-secondary leakage through the steam generator t e total for both steam generators the reactor shall be placed in cold shutdw withii 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> of detection.

IC. /lac6da AV. )3.1.6.4 If any reactor coolant leakage e o a ble fault in an RCS strength 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 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.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, 429,48G, 246 (12-22-78)

CONTROLLED COPY Bases (Continued)

I-or co .The unidentified eakage limit of 1 gpm is established as a quantity which can be accurately measured while sufficiently low to ensure 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 presence will not interfere with the detection of unidentified leakage. (O.'/.so (4 *ei) 9 The primary to secondary leakage through the steam generator tubes is limited tol gpm total.This limit ensures that the dosage contribution from tube leakage will be limited to a small fraction of Part 100 limits in the event of a steam line break. Steam generator leakage is 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.c;f Periodic equipment inspections.

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

l< 'I) t6.r. fif7-of;, 1'5fe~.. 6;v-Kc0 A- 1 3-1 5a Amendment No. 444, Order dtd. 4120/81,-246-INSERT TO TS PAGE 3-15a (BASES FOR SECTION 3.1.6)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 increases 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 CONTROLLED COPY 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)

-Periodic Testing and Reliability (2) NRC SER for BAW-10167A, Supplement 1, December 5, 1988.(3) BAW-10167, May 1986.4 BA-1 67A up ement 3, Februar/ t W~ dr RRect6r Ptoanta)-^

-S- *I. Leak &Cwdeues.4-2b Amendment No. 182265J INSERT TO TS PAGE 4-2b- (BASES FOR SECTION 4.1)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 Generators and 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 EPRI guidelines (Ref. 5).1 of 1 CONTROLLED COPY TABLE 4.1-2 MINIMUM EQUIPMENT TEST FREQUENCY Item Test Frequency 1. Control Rods 2. Control Rod Movement Rod drop times of all full length rods Movement of each rod Each Refueling shutdown Every 92 days, when reactor is critical 3. Pressurizer Safety Valves 4. Main Steam Safety Valves Setpoint Setpoint In accordance with the Inservice Testing Program In accordance with the Inservice Testing Program 5. Refueling System Interlocks Functional Start of each refueling period 6. (Deleted)I 7. Reactor Coolant System Leakage Evaluate Daily, when reactor coolant system temperature is greater than 525 degrees F (I 8. (Deleted)9. Spent Fuel Cooling System 10. Intake Pump House Floor (Elevation 262 ft. 6 in.)Functional (a) Silt Accumulation

-Visual inspection of Intake Pump House Floor Each refueling period prior to fuel handling Not to exceed 24 months (b) Silt Accumulation Measurement of Pump House Flow Quarterly 11. Pressurizer Block Valve (RC-V2)Functional*

Quarterly* Function shall be demonstrated by operating the valve through one complete cycle of full travel.Ae~ l4k e-,.,, ,E k-At Amendment No. i6, 68, 78, 449, 476, 498,2 4, *4 CONTROLLED COPY i CS7F## 96AAt4774 Cs) 7TUBE /A)~nvr licabilit/

i Techn al Sppcification applies to the inservice inspection o the SG tube portion of the reactor coolant pressure boundary.Ob ectiv The objecti of this inservice inspection program is to evide assurance of vtinued integrity of the tube portion of e Once-Through Stem rators, while at the same time min ing radiation exposure to peru ci in the performance of the inap tion.Specification Each steam generator ah 1 be demonstrated OP LE by performance of the following augmente inservice inspec on program and the requirements of Specificati 3.1.6.3.4.19.1 Steam Generator Sam le electiod and Inspection Methods a. Each steam generator sha e determined OPERABLE during shutdown by selecting and ecting at least the minimum number of steam generat sp ified in Table 4.19.1 at the frequency specified i 4.19.3. -b. Inservice inspecti of steam gene tor tubing shall include nondestructive nation by eddy-c zent testing or other equivalent teciques.

The inspection quipment shall be calibrated to rovide a sensitivity that will detect defects with a pene ation of 20 percent or more o the mimimum allowable -manufactured tube wall thickne 4.19.2 Steam erator Tube Sample Selection and Ins e tion The steam generator tube minimum sample size, map tIon result c usification, and the corresponding action requir shall e as specified in Table 4.19.2. The inservice inspec on of steam generator tubes shall be performed at the frequ cies specified in Specification 4.19.3 and the inspected tubes shall be verified acceptable per the acceptance criteria of Specification 4.19.4. The tubes selected for 4-77 Amendment No. O*f (12-22-78) each inservice it tu Lbes in all steam7.enerators; the tubes s or hese inspe ;onss b Ed o a random basis except a. The first sample of tubes selected for each inservice inspection (subsequent to the reservice inspection) of each steam generator shall include: 1.\ All nonplugged tubes that previously had detectable wall penetrations(>0%).

2. t least 50% of the tubes inspected shall be in those areas where ex rience has in cated potential problems.3. A tub nspection (pursuant to Specification 4.19.4.a.8) shall eperformed on each sel ted tube. If any selected tube does not permit th assage of the eddy current pr e for a tube inspection, this shall be recorde and an adjacent tube shall be sele d and subjected to a tube inspection.

4. Tubes in the fol ing groups may be excluded ft m the first random sample if all tubes in a groups both steam generators ar nspected.

No credit will be taken for these tubes meeting minimum sa pIe size requirements.

(1) Group A-i: Tubes rows 73 thro h 79 adjacent to the open inspection lane, and tubes betwe and on I'es drawn from tube 66-1 to tube 75-15 and from 86-1 to 77-1.5.(2) Group A-2: Tubes having lled opening in the 15th support plate.b. The tubes selected as the second an ird sames (if required by Table 4.19.2) during each inservice inspection may be ojected to a rial tube inspection provided: 1. The tubes selected for ese second and third mples include the tubes from those areas of the tu sheet array where tubes Mth imperfections were previously found.2. The inspectio ncludes those portions of the tubes w re imperfections were previously Cd C. Implementation f the repair criteria for Inside Diameter (ID) Inter- anular Attack (IGA) requir 00% bobbin coil inspection of all non-plugged tubes accordance with AmerGen gineering Report, ECR No. TM 01-00328, during all subse ent steam generato nspection intervals pursuant to Section 4.19.3. ID TGA indicatio detected by the bo in coil probe shall be characterized using rotating coil probes, as de ed in that repr The rest of each sample inspection shall be classified into one of the following three categories:

Category Inspection Results C-i Less than 5% of the total tubes inspected in a steam generator are degraded tubes and none of the inspected tubes are defective.

4-78 Amendment No. 47, 153, -P3~(12-22-78)

CONTROLED COPY 19.2 Specification (Continued)

C-2 One or more tubes, but not more than 1% of the total tubes inspected in a stea generator are defective, or between 5% and 10% of the total tubes inspected re degraded tubes.C-3 More than 10% of the total tubes inspected in a steam generator are d raded tubes or more than 1% of the inspected tubes are defective.

NOTES: (i In all inspections, previously degraded tubes whose degr ation has not een spanned by a sleeve must exhibit significant incr se in the applicable d radation size measurement

(> 0.24 volt bobbin c amplitude increase for side diameter IGA indications or > 10% furt r wall penetration for all ot degradation) to be included in the above p centage calculations.

(2) Where specia nspections are performed rsuant to 4.19.2.a.4, defective or degraded tubes und as a result of the ispection shall be included in determining the In ection Results C egory for that special inspection but need not be included in de rmining the spection Results Category for the general steam generator inspectn.4.19.3 Inspection Frequencies The required inservice inspections of St m genator tubes shall be performed at the following frequencies:

a. The first (baseline) inspection as performed after ifective full power months but within 24 calendar months of initia riticality.

The subseque inservice inspections shall be performed not more than calendar months after the preious inspection.

If the results of two consecutive inspect' ns for a given group of tubes en mpassing not less than 18 calendar months all f into the C-I category or demonstrate at previously observed degradation has not ntinued and no additional degradation has ccurred, the inspection interval for that g up may be extended to a maximum of once per 0 months.b. If the results f the inservice inspection of a steam generator conducte *n accordance with Table 4.19 at 40 month intervals for a given group of tubes* fall into tegory C-3 the inspecti fequency for that group shall be increased to at least once per 2 months. The increa in inspection frequency shall apply until the subsequent inspections tisfy the cit a of Specification 4.19.3.a; the interval may then be extended to a maxim I of once ps0 months.\A Sopof tubes means:\(a) All tubes inspected pursuant to 4.19.2.a.4, or (b) All tubes in a steam generator less those inspected pursuant to 4.19.2.a.4 4-79 Amendment No. 47.4i3-2O6

,-i9'-

-J 19.3 Inspection Fre ti uM LLrD COP Y /\ c. Additional.

unscheduled inservice inspections shall be performed on each steam genera rin accordance with the first sample inspection specified in Table 4.19-2 during the shut wn subsequent to any of the following conditions:

\1. A seismic occurrence greater than the Operating Basis Earthquake.

2. A loss of coolant accident requiring actuation of engineering safe ards, or 3. \A major main steam line or feedwater line break.d. After prima -to-secondary tube leakage (not including leaks orgating from tube-to-tube sheet welds) excess of the limits of Specification 3.1.6.3, spection of the affected steam generator will Performed in accordance with the followi criteria: 1. If the leak ibove the 14th tube support plate* a Group as defined in Section 4.19.2.a.4(l) of the tubes in this Group e affected steam generator will be inspected above e 14th tube support pla. If the results of this inspection fall into the C-3 catego additional inspec ns will be performed in the same Group in the other steam gen; tor.2. If the leaking tube is not efin in Section 4.19.3.d.

1, then an inspection will be performed on the affected am generator(s) in accordance with Table 4.19-2.4.19.4 Acceptance Criteria a. As used in this Specification:

I .Imperfection me an exception to the dim sions. finish, or contour of a tube from that requ d by fabrication drawing or s cifications.

Eddy current testing indications I s than degraded tube criteria specifd in a.3 below may be considere mperfections.

2. De2 tion means a service-induced cracking.

wastagevear or general corrosion occ g on either inside or outside of a tube.3. egraded Tube means a tube containing: (a) an inside diameter (I.D.) IGA indication with a bobbin coi dication/ 0.2 volt or 2 0.13 inches axial extent or 2 0.26 inches circuerential extent, or (b) imperfections 2 20% of the nominal wall thickness caused by degrada ion.4. % Degradation means the percentage of the tube wall thickness affected or remove by degradation.

4-80 AmendmentNo.

116,149,153, 206, 2099. si 4.19.4 Acceptance Criterdt C-5. Defect means an imperfection of such severity that it exceeds the repair limit. A tube containing a defect is defective.

6. Repair Limit means the extent of degradation at or beyond whi he tube shall be repaired or removed from service because it may be me unserviceable prior to the next inspection.

is limit is equal to 40% of the nominal tube wallickness.

Inside dia ter IGA indications shall be repaired or re ved from service if they excee n axial extent of 0.25 inches, or a cir inferential extent of 0.52 inch , or a through wall degradation mensions of> 40% if assigned.7. Unserviceable de ibes the condition of a tube if it leaks or contains a defect large enough affect it ructural integrity in the event of an Operating Basis Earthq ke loss of coolant accident, or a steam line or feedwater line break as s ed in 4.19.3.c., above.8. Tube Inspection m s an inspeon of the steam generator tube from the bottom of the upr tubesheet coin etely to the top of the lower tubesheet, ext as permitted by 4 9b.2, above.9. Inside meter Inter-Granular Attack (IG ndication means an indic on initiating on the inside diameter su e and confirmed by dinostic ECT to have a volumetric morphology aracteristic of IGA.b. The cm generator shall be determined OPERABLE after co leting the cpresponding actions (removal from service by plugging, or repaiy kinetic expansion, sleeving, or other methods, of all tubes exceeding the rep limit and all tubes containing throughwall cracks) required by Table 4.19-2.4.19 eports a. DELETED 4-81 Amendment No. 17., ,3, 91, 10a3, 129, 119, 153, 157, 206,20. -no X r\CONTROLLED COPY\ b. The complete results of the steam generator tube inservice inspection shall beported to the NRC within 90 days following completion of the inspection and reps (main generator breaker closure).

The report shall include: I.\ Number and extent of tubes inspected.

2. L ion and percent of wall-thickness penetratio each indication of an imper ction.3. Location, bon coil depth estimate (if ermined), bobbin coil amplitude (if determined), d axial and circum rential extent for each inside diameter IGA indic n, and 4. Identification of tubes repa or removed from service.5. The number of tube aired or oved from service in each steam generator, 6. An assess nt of growth of inside diameter degradation in accordance with th olumetric ID IGA management progr contained in AmerGen En eering Report. ECR No. TM 01-00328, and 7. esults of in-situ pressure testing, if performed.

c. esults of steam generator tube inspections which fall into Category C- quire 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 to determine the cause of the tube degradation and corrective measures taken to prevent recurrence in accordance with 10 CFR 50.73.4-82 Amendment No. 17, 86, 116, 149, 153 06. 20, CONTROLLED COPY e Surveillance Requirements for inspection of the steam generator tubes ensure that the st tural integrity of this portion of the RCS will be maintained.

The pro am for inservice inspection of steam generator tubes is based on modificatio of Regulato Guide 1.83, Revision 1. In-service inspection of steam generator tubing essential in order to mai in surveillance of the conditions of the tubes in the event that there s evidence of mechanical d ge or progressives degradation due to design, manufacturing elrs, or inservice conditions.

Inse ce inspection of steam generator tubing also provides a m s of characterizing the n re and cause of any tube degradation so that correct' e measures can be taken.The Unit is expected to be eratec l in a manner such that the prim and secondary coolant will be maintained within those ch istr limits found to result in ne gible corrosion of the steam generator tubes. If the primary Aecondary coolant chemistryi not maintained within these chemistry limits, localized corrosi may likely result.The extent of steam generator tube le e due to cracki: gwould be limited by the secondary coolant activity, Specification

3. 1.6.3.The extent of cracking during plant operation o be limited by the limitation of total steam generator tube leakage between the primary coo t system and the secondary coolant system (primary-to-secondary leakage = I in). Le age excess of this limit will require plant shutdown and an unscheduled inspetion, ring whi the leaking tubes will be located and repaired or removed from service.Wastage-type defects are unlikely proper chemistry treaent of the primary or the secondary coolant. However, eve fa defect would develop in ervice. it will be found during scheduled inservice steam gener o tube examinations.

For tube *th ID IGA indications, additional conservatism is be' ap lied to evaluate circumferentiald axial dimensions for determining final dispositioof the ube. For ID IGA indications thro h wall dimension will continue to be assigned t ose indi cations where amplitude response peits measuring through wall dimension.

Stea enerator tube inspections of operating plants have emonstrated the capability to reliably etect degrad ion that has penetrated 20% of the origin tube wall thickness.

Removal fro ervice by plugging, 3r repair by kinetic expansion, sleeving, or othe ethods, will be req ed for degradation equ 1 to or in excess of 40% of the tube nominal wall 'ckness.Tubes wi LD. initiated intergranul, 3r degradation may remain in service without % T.W.zing if the d radation morphology has be-en characterized as not crack-like by diagnostic eddy curre inspection and the degradation is of limited circumferential and axial length to ensure tub structural integrity.

Additionally, serviceability for accident leakage under the limiting stulated Main Steam Line Break (MSLB) accident will be evaluated by determining that this D. initiated degradation mechanism is inactive (e.g. comparison of the outage examination 4-83 Amendment No. 17, 129, 206, 209, -Em e (continued)

-CONTROLLED COPY results h the results from past outages meets the requirements of AmerGen Enginel Report, ECNo. TM 01-00328) and by successful in-situ pressure testing of a sa e of these degraded tubes t aluate their accident leakage potential when in-situ presstests are performed.

Where experience in similar pla with similar water chemist s documented by USNRC Bulletins/Notices, indicate critical ar o be inspected, aist 50% of the tubes inspected should be from these critical areas. First le ins ions sample size may be modified subject to NRC review and approval.Whenever the results of any steam gen tor tubing inse inspection fall into Category C-3 on the first sample inspection (See T e 4.19.2), these results w e reported to NRC pursuant to the requirements of Specifican 4.19.5.c.

Such cases will be con red by the NRC on a case-by-case basis and may r t in a requirement for analysis, laboratory e inations, tests, additional eddy cu inspection, and revision of the Technical Specificat if necessary.

NOTE: T eddy current examination voltages referred to in this section (section 4. re based on a malization procedure that sets the bobbin coil prime frequency peak-to-peak respon the four 20% through-wall holes of an ASME calibration standard to 4 volts.4-83a Amendment No. 17, 129, 206, 209, E CONTROUID COPY TABLE 4. 19- i 5IMMMNLU4 zSEm R OF sEAM' GE%;WS TO N 7I9SPECTM DURING 1NSMvMCE INSPECI0 Prese Inspection Wo. of Ste Generators per Unit C C First Inservixeinspection X TAMLE WMTAT I ON: X. The Inservice Inspect may be limited to oeteam generator on a rotating schedule copassing 6X of the tubes in asteam generator If the results the first and subsequent inspectis odie that both steam gen ators are performing in a like matn~er. nIote tha der s*e circinces, the operating conditions in one steam generator be found t more severe than those in the other steam generator.

Under mu ciricustances the sample sequence shall be modified to inspect the most severe conditions.

._.1 Amendment Igo.<4 7 (12-22-78) 4.84 ,

I f I 3 W 9 fD_TABLE 4.19-2 STEAM GENERATION TUBE INSPECTION(2)A II _I 2ND SAMPLE INSPECTION 3RD SAMPLE I T, ON I IA minimum of S1 Tut per is. G .1 I I Prug or repair def ilve tubes and in Ctt additiona y's tubes in th S.G. 0 Result I Action Required N/A I N/A I C-1 I None--I IPlug or repair I defective tubes and I inspect additional 4S I___ _ I tubes in this S.G.T erform action f9 C-3 I C-3 result of rst I \ I sample. /I Other I geform action o IS.G. Is, C2'~suit of second_ -X_/'lsampl esut Ic Kon Required N/A N/A I I Nlk- I N/A lNone I C-2 I Plug or repair I I defective tubes. I V Perf orm action I C-3 I for C-3 result lI of first sample. I T I HI/A I N/A I N/A I N/A! !___ -Ln inspect all I tubes In this S.G., plug or repair defect-ive tubes and I Inspect PS tubesl In other S.G.Provide notifi-cation to NRC Ipursuant to lOCFR50.72.b.1 and submi report rsuant to 1 R50.73.-a .11. -i Ot~we G. is C -3 inspect a tubes in each S.G. a plug or I repair defect I tubes. Provide \notification to NRC I pursuant to 1OCFRSO.7 b.2. and submit a I report pursuant to lOCFR5O.73.a.2.ii.

N/A N/A , ..-.,-Notes: (1) S = 3 -I Where N is the number of steam generators in the unit, and n is number of steam generators inspected during an inspection.

(2) or tubes inspected pursuant to 4.19.2.a.4:

No action is required for C-1 results. or C-2 results in one or both steam generators plug or repair defective tubes. For C-3 resu in one or both steam generators, plug or repair defective tubes and provide notification to NRC pursuant to 10 CFR 50.72.b.2.1 followed by a written report pursuant to 10 CFR S0.73.a.2.i I

INSERT TO TS PAGE 4-77 (REVISED TS 4.19)4.19 STEAM GENERATOR (SG) TUBE INTEGRITY Applicability:

Whenever the reactor coolant average temperature is above 250 0 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, and at the frequency required by, the Steam Generator Program.4.19.2 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 250 0 F following an SG tube inspection.

BASES: BACKGROUND Steam generator (SG) tubes are small diameter, thin walled tubes that carry primary coolant through the primary to secondary heat exchangers.

The SG tubes have a number of important safety functions.

Steam generator tubes are an integral part of the reactor coolant pressure boundary (RCPB) and, as such, are relied on to maintain the primary system's pressure and inventory.

The SG tubes isolate the radioactive fission products in the primary coolant from the secondary system. In addition, as part of the RCPB, the SG tubes are unique in that they act as the heat transfer surface between the primary and secondary systems to remove heat from the primary system. This Specification addresses only the RCPB integrity function of the SG.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 of 6 BASES BACKGROUND (continued) operational leakage. The SG performance criteria are described in Specification 6.19. 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 SAFETY ANALYSES The steam generator tube rupture (SGTR) accident is the limiting design basis event for SG tubes and avoiding an SGTR is the basis for this Specification.

The analysis of a SGTR event assumes a 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 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.b 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 6 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 burst/collapse condition to be established." For tube integrity evaluations, except for circumferential degradation, axial thermal loads are classified as secondary loads. For circumferential degradation, the classification of axial thermal loads as primary or secondary loads will be evaluated on a case-by-case basis.The division between primary and secondary classifications will be based on detailed analysis and/or testing.Structural integrity requires that the primary membrane stress intensity in a tube not exceed the yield strength for all ASME Code, Section 1I1, 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 1I1, 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 6 BASES LCO (continued) where the NRC has approved greater accident induced leakage. 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 250 0 F.RCS conditions are far less challenging when average temperature is at or below 250 0 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.b.(3.)a.

and 3.1.1.2.b.(3.)b.

3.1.1.2.b.(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 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, 3.1.1.2.b.(4.)

applies.4-80 4 of 6 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.b.(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 250 0 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.

3.1.1 .2.b.(4.)If the Required Actions and associated Completion Times of Condition 3.1.1.2.b.(3.)

are not met or if SG tube integrity is not being maintained, the reactor must be brought to HOT STANDBY within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> of detection and COLD SHUTDOWN within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> of detection.

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. 1), and its referenced EPRI Guidelines, establish the content of the Steam Generator Program. Use of the Steam Generator Program ensures that the inspection is appropriate and consistent with accepted industry practices.

During SG inspections a condition monitoring assessment of the SG tubes is performed.

The condition monitoring assessment determines the"as found" condition of the SG tubes. The purpose of the condition monitoring assessment is to ensure that the SG performance criteria have been met for the previous operating period.The Steam Generator Program determines the scope of the inspection and the methods used to determine whether the tubes contain flaws satisfying the tube repair criteria.

Inspection scope (i.e., which tubes or areas of tubing within the SG are to be inspected) is a function of existing and potential degradation locations.

The Steam Generator Program also 4-81 5 of 6 BASES 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 6.19 contains prescriptive requirements concerning inspection intervals to provide added assurance that the SG performance criteria will be met between scheduled inspections.

SURVEILLANCE REQUIREMENT SR 4.19.2: 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 6.19 are intended to ensure that tubes accepted for continued service satisfy the SG performance criteria with allowance for error in the flaw size measurement and for future flaw growth. In addition, the tube repair criteria, in conjunction with other elements of the Steam Generator Program, ensure that the SG performance criteria will continue to be met until the next inspection of the subject tube(s). Reference 1 provides guidance for performing operational assessments to verify that the tubes remaining in service will continue to meet the SG performance criteria.The frequency of "prior to exceeding an average reactor coolant temperature of 2500F 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 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".

4-82 (Pages 4-83 through 4-85 deleted)6 of 6 COcNTROLLED COPY 6.9.5 CORE OPERATING LIMITS REPORT) 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-1 0179 P-A, "Safety and Methodology for 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, -TMI-1 Core Thermal-Hydraulic Methodology Using the VIPRE-01 Computer Code", Revision 0. NRC SER dated 12/19/96.(4) TR-091-A, "Steady State Reactor Physics Methodology for TMI-1", Revision 0. NRC SER dated 2121/96.(5) TR-092P-A, -TMI-1 Reload Design and Setpoint Methodology", Revision 0. NRC SER dated 4/22/97.(6) BAW-10227P-A, "Evaluation of Advanced Cladding and Structural Material (M5) in PWR Reactor Fuel", NRC SER dated February 4, 2000.6.9.5.3 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 transient/accident 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.(t,4 S 57^ i*eA~eA11TbP 714,6 tUs#6cT(os QJoF 6-19 Amendment No.72,77, 1429, 137, 111, 144, 150, 168, 173, 178,02, -2e32 INSERT TO TS PAGE 6-19 6.9.6 STEAM GENERATOR TUBE INSPECTION REPORT A report shall be submitted within 90 days of plant startup after an outage (main generator breaker closure) following completion of an inspection performed in accordance with Section 6.19 (d). 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. The information specified for reporting in ECR No. TM 01-00328, and k. The information specified for reporting in ECR No. 02-01121, Rev.2.1 of 1 CONTROLLED COPY b. Licensees may make changes to Bases without prior NRC approval provided U 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.1 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).(G. 7 SEAM rrE6-IA-4rTO (54*) 4A~a-kPrSI

)8 6-26 Amendment No.-Q56--

INSERT TO TS PAGE 6-26 6.19 STEAM GENERATOR (SG) PROGRAM A Steam Generator Program shall be established and implemented to ensure that SG tube integrity is maintained.

In addition, the Steam Generator Program shall include the following provisions:

a. Provisions for condition monitoring assessments.

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

SG tube integrity shall be maintained by meeting the performance criteria for tube structural integrity, accident induced leakage, and operational leakage.1. Structural integrity performance criterion:

All in-service steam generator tubes shall retain structural integrity over the full range of normal operating conditions (including startup, operation in the power range, hot standby, and cool down and all anticipated transients included in the design specification) and design basis accidents.

This includes retaining a safety factor of 3.0 against burst under normal steady state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondary pressure differentials.

Apart from the above requirements, additional loading conditions associated with the design basis accidents, or combination of accidents in accordance with the design and licensing basis, shall also be evaluated to determine if the associated loads contribute significantly to burst or collapse.

In the assessment of tube integrity, those loads that do significantly affect burst or collapse shall be determined and assessed in combination with the loads due to pressure with a safety factor of 1.2 on the combined primary loads and 1.0 on axial secondary loads.2. Accident induced leakage performance criterion:

The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. Leakage is not to exceed 1 gpm per SG, except for specific types of degradation at specific locations as described in paragraph c of the Steam Generator Program 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.1 of 3 The following alternate tube repair criteria may be applied as an alternative to the 40%depth based criteria: 1. Volumetric ID IGA indications may be dispositioned in accordance with ECR No. TM 01-00328.

Implementation of these repair criteria for Inside Diameter (ID) Inter-Granular Attack (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. 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.)2. Upper tubesheet kinetic expansion indications may be dispositioned in accordance with ECR No. TM 02-01121, Rev. 2. Implementation of these 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. 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 the 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.Refer to Section 6.9.6 for reporting requirements for periodic SG tube inspections.

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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.)3 of 3