Text

Response to Request for Additional Information - Technical Specification Change Request No. 331 Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity
	ML071520233
Person / Time
Site:	Three Mile Island
Issue date:	05/31/2007
From:	Cowan P; AmerGen Energy Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	5928-07-20072, TAC MD1807
	Download: ML071520233 (51)
	v • d • e

May 31,2007 5928-07-20072 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 to AmerGen Energy Company, LLC dated May 17,2007, Request for Additional Information Regarding Proposed Steam Generator Tube Integrity Technical Specification Changes (TAC No. MD1807).

2) AmerGen Energy Company, LLC letter to NRC dated May 15,2006 (5928-06-20390), 7echnical Specification Change Request No. 331 -

Application for Technical Specification Improvement Regarding Steam Generator Tube Integrity.

3) AmerGen Energy Company, LLC letter to NRC dated December 12,2006 (5928-06-20559), 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).

This letter provides additional information in response to the NRC request for additional information (RAI), dated May 17, 2007 (Reference l ) , 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 Enclosure 1 responses, the proposed Technical Specification page markups have been revised from our submittal of December 12,2006 (Reference 3) to incorporate additional clarifications, consistent with the NRC approved TSTF-449, Revision 4. Reference 1 also provided NRC staff observations for consideration regarding the TMI Unit 1 proposed TS Bases page markups incorporating TSTF-449, Revision 4 Bases changes. These observations have been evaluated and incorporated, as applicable, into the revised proposed TS page markups provided in Enclosure 2.

U.S. Nuclear Regulatory Commission May 31,2007 Page 2 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 3.

We suggest that a meeting be scheduled to facilitate resolution if any significant open issues remain regarding TMI Unit 1 implementation of TSTF-449, Rev. 4.

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 31

day of May, 2007.

Respectfully, Pamela B. C&an Director - Licensing & Regulatory Affairs AmerGen Energy Company, LLC

Enclosures:

1) Response to Request for Additional Information

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

ENCLOSURE 1 TMl UNIT 1 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION TECHNICAL SPECIFICATION CHANGE REQUEST No. 331 APPLICATION FOR TECHNICAL SPEC1FICATION IMPROVEMENT REGARDING STEAM GENERATOR TUBE INTEGRITY

Enclosure 1 5928-07-20072 Page 1 of 9 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)

TMI UNIT 1 TECHNICAL SPECIFICATION CHANGE REQUEST No. 331 APPLICATION FOR TECHNICAL SPECIFICATION IMPROVEMENT REGARDING STEAM GENERATOR TUBE INTEGRITY

1. NRC Question The cover page of the October 6,2006, document (AmerGen response to RAI) indicates that the SG tube integrity TSs are going to be made consistent with Revision 1 of TSTF-449 and Page 1 of Enclosure 1 of the same document indicates that changes to the SG tube integrity TSs will be consistent with Revision 3 of the TSTF-449. Please clarify that the modifications being made are consistent with TSTF-449, Revision 4.

Response

As stated in the cover letter of this RAI response submittal, the modifications are being made consistent with TSTF-449, Revision 4.

2. NRC Question The Actions and Note in TS Section 3.1.1.2 should be rearranged to indicate that TS Section 3.1.1.2.a.(3) should only be entered if TS Section 3.1.1.2.a.(2) was not met. As a result, discuss your plans to remove the last sentence of the Note and relocate it as part of TS Section 3.1.1.2.a.(3). This sentence should also be modified to remove reference to TS Section 3.1.1.2.a.(1) since entry into TS Section 3.1.1.2.a.(3) is not permitted if TS Section 3.1 .I .2.a.(I) is not satisfied. For example, If the requirements of Section 3.1.1.2.a.(2) are not met for one or more tubes then perform the following.

TS 3.1.1.2 has been revised to remove the last sentence of the Note and relocate it as part of TS 3.1.1.2.a.(3). The relocated sentence has also been modified to remove the reference to TS 3.1.1.2.a.(1).

3. NRC Question It is the staffs understanding from discussions during a conference call on May 7, 2007, that you assume that the primary-to-secondary leakage during a steam line break (SLB) is more limiting than for the other design basis accidents (DBAs) that assume primary-to-secondary leakage exists. As a result of this assumption, you only determine the amount of leakage during a SLB and compare this to the assumptions in your accident analyses.

Please provide the technical basis for why it is only necessary to assess the leakage during a SLB (i.e., demonstrate that by satisfying the leakage limit associated with a SLB you will meet the leakage limit for the other DBAs).

Enclosure 1 5928-07-20072 Page 2 of 9

Response

Analyses of the SLB event were used as a basis for the plants derivation of examination and flaw dispositioning criteria for the steam generator kinetic expansions. These criteria, documented in ECR 02-01121, were recently reviewed and approved by the NRC staff on November 8,2005 (TAC MC7001). Assessment of hypothetical SLB leakage is a part of the dispositioning criteria.

Analyses of the SLB event were also used as a basis for the derivation of examination and flaw dispositioning criteria for the ID Volumetric IGA indications in the plants steam generator tubes. These criteria, documented in ECR TM 01-00328, were also reviewed and approved by the NRC staff in TMI Unit 1 License Amendment No. 237, dated October 5, 2001 (TAC NO. MB0664) and are incorporated into the plants current steam generator Technical Specifications. Assessment of hypothetical SLB leakage is also a part of those dispositioning criteria.

The technical basis for why the SLB is utilized for leakage assessment is described in the above referenced ECRs. Additional information was also provided in RAI question responses during the staffs review of the ECRs. The following single sentence excerpt from ECR 02-01121 provides a summary as to why the SLB event was used to derive the conservative kinetic expansion repair criteria: ...MSLB is the design-basis accident for the kinetic expansions since it represents a hypothetical accident where tube stresses are relatively high, and the potential exists for offsite dose consequences from tube leakage resulting from significant primary-to-secondary pressure drop. Other accidents were also addressed in the above ECRs and their associated RAI responses, for example the Question No. 3 response in AmerGen letter to the NRC dated August 11,2005.

The proposed Technical Specification change proposes to insert the above, previously approved, tube repair criteria consistent with the TSTF.

4. NRC Question In TS Table 4.1.2, the Test for primary to secondary leakage is listed as Evaluate. The meaning of this term is not clear. Isnt the Test for primary to secondary leakage, continuous monitoring of the effluent (steam and feedwater systems) for radioactive isotopes or performing radiochemical analyses of grab samples of the steam and feedwater systems? Similarly, isnt the Test for reactor coolant system leakage, a water inventory balance rather than Evaluate? Please discuss your plans to modify your proposal to more accurately reflect the Test for monitoring primary-to-secondary leakage.

In addition, discuss your plans to modify the Note in Table 4.1 -2 to make it more consistent with the TSTF-449. The TSTF wording is Note: Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

Response

The proposed TS Table 4.1 -2 has been revised to include the TSTF wording: Note: Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation. In addition, the Frequency required for primary-to-secondary leakage evaluation has been revised from 24 to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months in accordance with the TSTF.

Enclosure 1 5928-07-20072 Page 3 of 9 The TS Table 4.1-2 entry for Test of both primary-to-secondary leakage and Reactor Coolant System leakage remains Evaluate. This is consistent with the TSTF and other US plants, where continuous monitoring is utilized to assess leakage of both primary-to-secondary leakage and leakage from the Reactor Coolant System, but the Technical Specifications require a periodic evaluation. The proposed Bases state that leakage shall be evaluated using the EPRI Guidelines.

Note that the TS Bases on Page 4-2b were also modified to be consistent with the above changes. However, the TSTF Bases statement that states, 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 was deleted from the Bases. (Several of these parameters affect the RCS leakage determination via mass balance, but do not affect primary to secondary leakage determination.)

5. NRC Question Even though TMI-1 will not have approved repair methods, please discuss your plans to describe the existing kinetic expansions and sleeves where the following note appears in the TSTF-449, [Steam generator tube repairs are only performed using approved repair methods as described in the Steam Generator Program.]

Response

The kinetic expansions and sleeves were approved and installed many years ago, but under the proposed TSs TMI Unit 1 will be required to obtain NRC approval prior to installation of new kinetic expansions, sleeves, or other repair methods. TS 6.1 9.f has been revised to clearly state this requirement.

The subject note appears in the TSTF page for the Bases section for the SG inspection surveillance. A small paragraph referring to the kinetic expansions and sleeves has been added to the subject Bases Section on proposed TS Bases Page 4-83.

6. NRC Question Please discuss your plans to modify TS Section 6.9.6.f to be consistent with TSTF-449 by reading, Total number and percentage of tubes plugged or repaired to date.

TS Section 6.9.6.f has been modified as suggested.

7. NRC Question Given that TMI-1 has tubes that are sleeved, please discuss your plans to modify TS Section 6.19.c to clearly indicate the repair criteria for the non-sleeved region of the sleeved tubes and the repair criteria for the sleeved region of the tube. In addition,

Enclosure 1 5928-07-20072 Page 4 of 9 discuss your plans to clearly indicate that the alternative to the 40-percent depth based criteria can only be applied to non-sleeved tubes.

Response

TS Section 6.19.c was also revised in the response to Question 13, below. Refer also to that response.

TS 6.19.c.l has been revised to indicate that the alternative to the 40-percent depth based criteria can only be applied to the non-sleeved regions of tubes. Note that the alternative repair criteria for Volumetric ID IGA described under TS Section 6.19.c.l .a is also applied to the non-sleeved region of sleeved tubes per ECR TM 01-00328. For that reason, TS Section 6.19.c.l uses the phrase non-sleeved regions of tubes rather than Inon-sleeved tubes.

8. NRC Question Please discuss your plans to include the definition of inside diameter intergranular attack from TS Section 4.1 9.4.a.9 in TS Section 6.19.c.l (current proposed numbering). In addition, please discuss your plans to remove repaired or from the fourth sentence of this TS since there will be no approved repair methods at TMI-1.

The definition of inside diameter intergranular attack (ID IGA) from current TS Section 4.1 9.4.a.9 was inserted (via a new sentence) into proposed TS Section 6.19.c.l. The phrase repaired or was removed from the referenced sentence of TS 6.19.c.l.a.

9. NRC Question The reason for referencing the accident-induced leakage criteria in proposed TS Section 6.19.c.l and 6.19.c.2 is not apparent. If it is consistent with your design and licensing basis it would seem that it is addressed by the first sentence under TS Section 6.19.b.2.

In addition, please discuss your plans to clearly indicate that the only exception to the 1 gpm per SG limit is for leakage attributed to the kinetic expansions. For example, Leakage from all sources excluding the leakage attributed to the degradation described in TS Section 6.19.c.2 is also not to exceed 1 gpm per SG.

Response

Proposed TS 6.19.b.2 is modified to reference TS 6.19.c, which provides the accident-induced leakage criteria exception for kinetic expansions.

TS Sections 6.1 9.c.l and 6.1 9.c.2 specify the tube repair criteria for the SGs. TSTF-449, Rev. 4, in the last paragraph of Page 15, Section 10, SG Tube Repair Criteria, states that any alternate repair criteria, and any allowed accident induced leakage rates for specific types of degradation at specific locations associated with tube repair criteria, are to be listed in the same section. In addition, Page 2 of the TSTFs proposed insert for STS

Enclosure 1 5928-07-20072 Page 5 of 9 Section 5.5.9, in the tube repair criteria section Reviewers Note states that descriptions of.. .alternate tube repair criteria ...should also include any allowed accident-induced leak rates.... Based on the above, the proposed TS Sections 6.19.c.l and 6.19.c.2 are consistent with the TSTF.

The existing, underlined, sentence in proposed TS Section 6.1 9.b.2 (modified as described above), Leakaae is not to exceed 1 aom per SG, exceot for soecific tvoes of dearadation at soecific locations as described in oaraaraoh 6.1 9.c of the Steam Generator Proaram below is consistent with TSTF-449, and performs the same function as the example statement provided in the question (i.e., Leakage from all sources excluding the leakage attributed to the degradation described in TS Section 6.19.c.2 is also not to exceed 1 gpm per SG.).

10. NRC Question Please discuss your plans to modify TS Section 6.1 9.d.4 to make it more consistent with your current TSs. For example, AmerGen Engineering Report, ECR No. TM 01-00328 durina all subseauent SG inspections.

Response

The phrase during all subsequent SG inspections has been added to proposed TS Section 6.1 9.d.4.

11. NRC Question Please discuss your plans to modify TS Section 6.1 9.f to indicate that repairs were performed by kinetic expansion and sleeving but no new tube repairs can be made without prior NRC approval.

Response

TS Section 6.19.f has been revised to describe repairs that were performed by kinetic expansion. (Repairs performed by sleeving were already described in the previous version of TS Section 6.19.f.) The last sentence of this section was revised to read, Installation of new repair methods, additional kinetic expansions, or additional sleeves, requires prior NRC approval.

12. NRC Question In your proposed TS Section 6.19.d, you define the length of the tube as from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet. Based on this definition, the parent tube behind the upper sleeve joint is part of the pressure boundary and is required to be inspected. Please confirm that this inspection will be performed with the implementation of TSTF-449. As currently proposed the 40-percent repair criteria would apply to the parent tube behind the upper sleeve joint since proposed TS Section 6.19.c.2 does not address the repair criteria for the parent tube behind the sleeve upper joint. However, a more appropriate repair criteria for flaws in the parent

Enclosure 1 5928-07-20072 Page 6 of 9 tubing at the upper sleeve joint may be 20-percent through-wall since this appears to be consistent with the original sleeve qualification (please note that this is based on the licensees submittal and not a review of the original qualification report). As a result, please discuss your plans to clarify the acceptance criteria for any flaws in the parent tube behind (adjacent to) the sleeves upper joint.

In addition, it is not clear why the phrase in accordance with ECR No. TM 02-01121, Rev.

2 is needed in the specification. Please clarify the purpose for the last phrase in TS Section 6.1 9.c.2. If ECR No. TM 02-01121, Revision 2 does not clarify the repair criteria for sleeves, discuss your plans to delete it.

ECR No. TM 02-01121, Rev. 2 must be in the proposed TMI Unit 1 TS since it provides examination scope, repair criteria, technical analyses and reporting requirements for the SG kinetic expansion indications. In addition, this ECR prescribes sleeve examination and repair requirements. This document was recently approved by the NRC staff on November 8,2005 (TAC MC7001). In accordance with ECR 02-01121, Rev. 2, inspection of the parent tube behind the upper sleeve joints will not be performed with the implementation of the proposed TS change to implement TSTF-449. No proposed acceptance or repair criteria for flaws in the parent tube behind the sleeves upper joints are provided. The sleeves were installed to prevent additional primary-to-secondary leaker outages. Plugging the sleeved tubes would result in an overall reduction in plant safety due to an increase in susceptibility to degradation of neighboring tubes.

The current TMI Unit 1 TSs do not include the upper tubesheet kinetic expansions. ECR TM 02-01121 was developed, and approved, to provide examination and flaw dispositioning criteria for kinetically-expanded tubing in the upper tubesheets. Since the TSTF requires the TSs to address the SG tubes from end-to-end, adoption of the TSTF into the TMI-1 TSs requires that ECR 02-01121 be included.

Sleeve upper roll inspections are performed each refueling outage. Additional information addressing the structural and accident leakage integrity of the upper sleeve joint is provided in AmerGen letter to the NRC, dated August 11, 2005, entitled, Additional Information Regarding Kinetic Expansion Inspection and Repair Criteria (TAC No.

MB6475).

TS Section 6.1 9.d was revised to add the sentence In tubes repaired by sleeving, the portion of the original tube wall above the sleeves lower sleeve-to-tube joint is not an area requiring re-inspection. This sentence is consistent with ECR TM 02-01121 and similar to sentences added to the TSTF paragraphs by other U.S. PWRs with sleeves.

13. NRC Question In your proposed TS (and TSTF-449) 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. Given this definition, the proposed repair criteria in TS Section 6.1 9.c could be misinterpreted. Please discuss your plans to modify your TS to more clearly define the repair criteria for the sleeved portion of a tube. For example, this TS may be modified using the following:

Enclosure 1 5928-07-20072 Page 7 of 9

1. The non-sleeved region of a tube found by inservice inspection to contain flaws with a depth equal to or exceeding 40-percent of the nominal tube wall thickness shall be plugged or repaired except when alternate tube repair criteria permitted by technical specifications are satisfied.

2. Tubes shall be plugged if the sleeved region of a tube is found by in-service inspection to contain (actual repair criteria is dependent upon resolution of questions 7 and 12).

3. The following alternate repair criteria may be applied as an alternative to the 40-percent depth based criteria in TS 6.1 9.c.l:

a. Volumetric Inside Diameter (ID) inter-Granular Attack (IGA) indications...

b. Upper tubesheet kinetic expansion indications...

Response

The first sentence of TS Section 6.19.c.l has been revised to clarify that the section applies to non-sleeved sections of tubes. TS Section 6.19.c.2 applies to sleeved tubes.

This proposed change should lessen the possibility of misinterpretation.

Refer also to the responses to Questions 7 and 12, above.

14. NRC Question Given that TS Section 6.19.d provides the inspection requirements for the tube (which by definition includes the sleeve) and the inspections must be performed to ensure SG tube integrity, it is not clear that the proposed requirements in TS Section 6.19.d.6 are needed.

As a result, please discuss your plans for removing these proposed requirements. If the proposed requirements are maintained, please discuss your plans to add TS Section 6.1 9.d.6 to the first paragraph of TS Section 6.19.d.

Response

The proposed requirements in TS Section 6.1 9.d.6 have been deleted.

Refer also to the response to Questions 12 and 13, above. The proposed TS Section 6.19.d has been modified to clarify the inspection requirements for sleeved tubes.

15. NRC Question As currently written, proposed TS Section 6.19.f is unclear. Please discuss your plans to modify this proposed requirement. For example, There are currently no approved repair methods; however, tubes repaired with sleeves and by kinetically expanding the tube in the upper tubesheet prior to 2006 may remain in service subject to the requirements of TS Sections 3.1.1.2, 4.1 9, and 6.1 9.

Enclosure 1 5928-07-20072 Page 8 of 9 The phrase subject to the requirements of TS Sections 3.1.1.2,4.19, and 6.19 has been added to TS Section 6.1 9.f at two locations.

Note that Section 6.1 9.f has also been revised to respond to Question 11, above.

The sleeves and kinetic expansions were repair methods approved many years ago; therefore, the phrase, There are currently no approved repair methods... was not used.

Rather, TS Section 6.1 9.f was revised to clarify that no additional kinetic expansions, sleeves, or other new repairs may be installed without prior staff approval.

16. NRC Question As currently proposed, any flaws in the parent tube between the sleeve lower end and the parent tubes kinetic expansion are required to be plugged on detection. Since several flaws were removed from service as a result of the sleeving campaigns (several imperfections and one inside diameter indication greater than 40-percent through-wall), it would appear that upon adoption of the new proposed TSs there will be several tubes that exceed the repair criteria. This will require a tube integrity assessment, possibly a forced shutdown, and will result in these tubes being plugged. Please clarify that this was your intent and provide the results of the tube integrity assessment.

Response

As was discussed on the phone call of March 23,2007, the scenario proposed in the question was not the intended interpretation of the proposed TSs, and no tube integrity assessment is currently required.

The proposed TSs are written to apply to future steam generator tube inspections (i.e.,

inspections performed after the proposed TSs are implemented.) They were not written to apply to past steam generator inspections. For example, the phrase tubes found by inservice inspection to contain.. . at two locations in TS Section 6.19.c refers to inservice inspections occurring after the proposed TSs are implemented.

All in service sleeves have been examined over the past three refueling outages (i.e.,

approximately 33% of the sleeves each outage). No degradation meeting the proposed sleeve inspection repair criteria was identified in the sleeve or the adjacent parent tube above the lower sleeve end.

17. NRC Question Proposed TS Section 6.9.6.i implies that repairs are authorized at TMI-1. Given that tubes have been repaired in the past but that no new repairs will be permitted, please discuss your plans to modify TS Section 6.9.6.i to indicate that you will provide the number of tubes remaining in service using repair methods previously implemented.

Enclosure 1 5928-07-20072 Page 9 of 9

Response

Refer to the response to Question 6, above. Based on Question 6, TS Section 6.9.6.f has been revised to provide the number of repaired tubes. Since this information will be provided to the NRC under proposed TS Section 6.9.6.f, the former TS Section 6.9.6.i was deleted. Steps j. k, and I in proposed TS Section 6.9.6 were re-lettered since TS Section 6.9.6.i was deleted.

ENCLOSURE 2 TMl 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 Paaes 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

- PGLGVD In order to confirm the leak-tight integrity of the Reactor Coolant System, inciudi 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 ed dunng the steam generator hot test progr line leakage rate by more than 0.1 gpm', the all be shut down span, the leaking tube(s) shall be om service. The tion 3.1.6.3 is not exceed

3. The licensee sha st program at each power range (0- 5 %, 5%-5O%,

Topical Report 008, available the results of that test program and a sum t review, pnor to ascension from each power range and prio power operation.

4 The licensee shall conduct examinations, consistent with the extended inservice ins ed in Table 3.3-1 of NUREG-1019, either 90 calendar da exceeding 50% p omes first. In the event of plant ent at the end of 180

%, such assessment tion as to the necessity of a I eddy-cunent testing t assessment and determine the time o

, consistent with the other provisions of ce of such an assessment, a special ECT shu additional 30 days of operation at power abov exceeds the baseline leakage rate by more than 0.1 gpm during the re operation, the facility shall be shutdown and leak tested. Operation at Amendment No. Amendment N0533?-,

(9) Lona Ranae Plannina Proaram - Deleted Sale and License Transfer Conditions (10) Deleted (11) Deleted (12) Deleted (13) Deleted Amendment No. W,W ete,i128,W,+W-/

, I

TABLE OF CONTENTS Section f&g,g 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 4.1 2.1 EMERGENCY CONTROL ROOM AIR TREATMENT SYSTEM 4-55 4.1 2.2 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 4.13 FUEL HANDLING BUILDING ESF AIR TREATMENT SYSTEM RADIOACTIVE MATERIALS SOURCES SURVEILLANCE 4-55f 4-56 1

4.14 DELETED 4-56 4.15 MAIN STEAM SYSTEM INSERVICE INSPECTION 4-58 4.1 6 REACTOR INTERNALS VENT VALVES SURVEILLANCE 4-59 4.17 SHOCK SUPPRESSORS fSNUBBERS1 4-60 4.18 FIRE PROTECTION SYSTEMS (DELETED) 4-72

+? 4.19 A 7

-4+H- 4-77 Y-/ I 4- m-4.20 REACTOR BUILDING AIR TEMPERATURE 4-86 4.21 RADIOACTIVE EFFLUENT INSTRUMENTATION (DELETED) 4-87 4.21.1 RADIOACTIVE LIQUID EFFLUENT INSTRUMENTATION(DELETED) 4-07 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 COMPARlSON PROGRAM (DELETED) 4-87 iv

TABLE OF CONTENTS

-Section Pasre 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 FAClLlTiES 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 REV~EW 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 APFA 6-22 6.13 PROCESS-CONTRSIL-PROGRAM 6-23 6.14 OFFSITE DOSE CALCULATION MANUAL {ODCM) 6-24 6.15 DELETED 6-24 6.16 DELETED 6-24 I 6.17 MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS 6-25 6.18 I T S 1 - ) BASES CONTROL PROGRAM 6-25 I

LIS'I 0 1 7 'I'AULES mru TITLE Pi\<; E I .2 1-8 2.3-1 Reitctor Protection System T r i p Setting Limits 2-9 3.1.6.1 Pressure Isolation Check V:iIves Betwcm the 3- 1 sii Primary Coolant Systeni itnd LPlS 3.5- 1 Instruments Operatitag Coiiditittris 3-29 3.5- 1A DELETED 3.5-2 Accident Monitoring lnstrunieiits 3-40c 3.5-3 Post Accident Monitoring Iiistniinciitittion 3-4Otl 3.5-4 3-Jl)i I

3.21-1 DELETED 3.21-2 DELETED 3.23-1 DELETED 3.23-2 DELETED 4.1-1 Instnrmcnt Surveillance Requirements 4.1-2 Rlinimuni Equipment Test Fruqueiicy 4.1-3 Miininturn Sampling Frequencv 1.1-4 Post Accitlent Monitoring Instrumcritation 4.19-1 4.19-2 d=

4.21-1 DELETED 1.21-2 DELETED 4.22-1 DELETED 4.22-2 DELETED 4.23- I DELETED

3.1 3.1.1 REACTOR COOLANT SYSTEM OPERATIONAL COMPONENTS I

P o o l i c a b f l it v Applies t o the c p 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 iective s

To specify those l i m i t i n g conditions f o r c p e r a t i o n of r e a c t o r coolant syscem components wnich must be met t o ensure s a f e r e a c t o r operations.

Soeci f ics t i on 3.1.1.1 Reactor Cczlant Pumcs

a. Pmp combinations permissible for given power levels Shall be a s snown i n S p e c i f i c a t i o n Table 2.3.1.

b. Power coeration with one i d l e reactor coolant pump i n each lcop shall be restricted to 24 nours. I f the reactor is not retuned to an acceptable RC pump operating combination a t t h e end of the 24-hour pericc, the r e a c t o r s h a l l be i n a hot shutcown condition w i m i n the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

c. The boron concentration i n t h e reactcr ccolant ,systm s n a l i not be rsauced unless a t l t a s t one iesCt3r ccolant Pump or one ciecay heat removal ~ c m pi s c i x u l a t i n g 3 . I. 1.2

7cn-Ld .

3.1.1.3 Pressurizer Safety Valves

a. The r e a c t o r shall not remain c r i t i c a l unless both p r e s s u r i z e r code safety valves a r e operable with a l i f t s e t t i n g of 2500 p s i g E3. -

b. When the r e a c t o r i s s u b c r i t i c a l , a t l e a s t one p r z s s u r i z c r code ssfety valve! snall be cperaalc if a l l reactor coolant system ouenings a r e closed, excest f3r bvdrostatic tests in accordance w i t h A S K SaF1?r anc Pressure \lessel Cede, Secticn I I i .

3-la I

INSERT TO TS PAGE 3-la (REVISED TS 3.1.1.21

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.)

(3.) Ifthe requirements of Section 3.1.1.2.a.(2.) are not met for one or more tubes then perform the following:

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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and be in COLD SHUTDOWN within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

1 of 1

Bascs The limitation on potter operation \iith one idle RC pump in each loop has been imposed since the ECCS cooling performance has not been calculated in accordancc nith the Final Acceptance Critena requirements specificall) for this mode of reactor operation A time penod of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months IS allo\<edfor 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for this mode of operation is acceptable since this mode is espected to haLe considerable margin for the peak cladding temperature limit and since the likelihood of a LOCA \tithin the 24-hour period is considered ver) remote A reactor coolant pump or deca) heat removal pump is required to be in operation before the boron concentration is reduced by dilution with makeup water Either pump \ \ i l l provide mixing which t\ 111 prevent sudden positive reactivitl, changes caused by dilute coolant reaching the reactor One deca) heat removal pump will circulate the equivalent of the reactor coolant sllsteni 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 I , 2, and 3)

Both steam generators m u s t + e p w a b k - of the Reactor Coolant S)stem to insure system integnty against leakage under normal and transient conditions Only one steam generator is required for decay heat removal purposes 4

One pressunzer code safet? valve is capable of preventmg overpressunzation N hen the reactor is not cntical since its relieving capacity is greater than that required by the sum of the available heat sources which are pump energy, pressunzer heaters, and reactor decay heat Both pressunzer code safep valves are required to be in service pnor to cnticalih to conform to the slvstem design relief capabilities The code safety valves prevent overpressure for a rod withdra\\aal or feedwater line break accidents (Reference 4) The pressunzer code safety valve l i f t set point shall be set at 2500 psig 5 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-I safety analysis.

References (1) UFSAR, Tables 9.5 (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. 37 (12/22/78),W =.

/

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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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 3.1.6.3 -.

actor shall be placed in cold shutdqwn within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .-

A t i e , wAJzr sllQ/1 leyf-d rh L o t FAu jdOd a(JXh 4 Xaurs4 an

._)

3.1.6.4 If any reactor coolant leakage exists through a nonisolable 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 I shutdown condition shall be initiated within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 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 0.02 hours 1.190476e-4 weeks 2.7396e-5 months 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. 42, W ,1118,-

(12-22-78)

is established as a quantity which can be accurately f 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.

I very low activity levels and would show an increase in activity level shortly after a reactor coolant leak developed within the auxiliary building.

t

b. Water inventories around the auxiliary building sump.

d. 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.

3-15a Amendment No. 4-44, 11-

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 is assumed to increase to the leakage rates described in TS 6.19.c.l as a result of accident-induced conditions.

The TS requirement to limit the sum of the primary to secondary leakage from 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 on the sum of the primary to secondary leakage from both SGs of 144 gallons per day 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 Continuedl 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 EFW Pump or both E R N flowpaths to a single OTSG to be inoperable for up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months during surveillance testing requires 1) at least one motor-driven EFW Pump operable, and 2) an individual involved in the task of testing the ERN System must be in communication with the control room and stationed in the immediate vicinity of the affected ERN 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 ERN 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 ENV components to operable status, and all actions requiring shutdown or changes in Reactor Operating Condition are suspended. With less than two E R N 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 E W components to operable status. TS 3.0.1 is not applicable, as it could force the unit into a less safe condition.

The ERN 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.I.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 K/K for at least one hour subsequent to power operation above 5% full power. According to Specification 3.1 reactor coolant average temperature is MSSVs are available for redundancy.

MSSVs are inoperable, the power leve such that the remaining MSSVs can p 3-26~

Amendment No. 78, ?19, 125, ?aa,l5F-;12e 1

,]

Bases (Cont'dl 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-I0167A, Supplement 1, December 5,1988.

(3) BAW-10167,May 1986. - -c, B A W - 1 f/

Ep42, \*fpc,srpi tC4 Realtor secd"atq3 Leak Pn*rna~J Gu;$eiL.

1 I

4-2b Amendment No. ?44&?26,44%-,

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 the sum of the primary to secondary leakage from both SGs is less than or equal to 144 gallons per day. 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 the sum of the 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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after establishment of steady state operation.

The TS Table 4.1-2 primary to secondary leakage surveillance frequency of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is a reasonable interval to trend primary to secondary leakage and recognizes the importance of early leakage detection in the prevention of accidents. The primary to secondary leakage is determined using continuous process radiation monitors or radiochemical grab sampling in accordance with the EPRl guidelines (Ref. 5).

1 of 1

TABLE 4.1-2 MINIMUM EQUIPMENT TEST FREQUENCY Item Test Freauencv

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 Inservice Testing Program

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

5. Refueling System Functional Start of each Interlocks refueling period

6. (Deleted) -- I

\

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)

Amendment No. 545,643,?I, 4-76, a, W, 2u-,-f467.

I I f 1

f The object 8 inservice inspection program assurance ed i n t e g r i t y of t h e tube Through S t r a t o r s , while a t the same exposure to p 1 in the performance of Specification Each 8tm gmctPtOr

\ -6 be demonstrated

/LE by performance imerofce inspec rogram and the

a. Each steam g f l e d in Table 4.19.1 a t the

b. Inservice %nu tubing shall include nondee truc t i v detect defects of 20 percent o r J

Th steam generator tube minimum rample eize, c 8sificetion, and the corre8ponding action e as specified in Table 4.19.2. The of steam ~ u r e r s t o rtubes shall be specified in Specification 4.19.3

'shall be v e r i f i e d acceptable per Specification 4.19.4. The tubas 4-77

a.

service inspection) of each steam generator shall include:

ated potential problems.

a tube inspection, this shall be recor and subjected to a tube inspection.

79 adjacent to the open inspection ram fiom tube 66- 1 to tube 75- 15 and from 86-1 t (2) Group A-2: Tubes havin led opening in the 15th support plate.

b. The tubes selected as the second required by Table 4.19.2) during each inservice inspection may be The tubes selected e second and thi es include the tubes fiom imperfections were cludes those portions of the tubes imperfections were C.

each sample inspection shall be classified into one of the following three categorie C m Inspection Resu 1ts 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. 43;-153,-23tj (1 2-22-78)

19 2 Specification (Continued) all inspections, previously degraded tubes whose d dation size measurement (> 0.24 volt bobbi nt to 4.19.2.a.4.defective or n shall be included in not be included Results Category for the general 4 19 3 Inspection Frequencies The required inservice inspections o or tubes shall be performed at the following frequencies:

24 calendar months of ini performed not more than two consecutive ins calendar months all reviously observed rred, the inspection

\

0 months of tubes means:

4- 79 Amendment No. 4+4&+4% +*?

accordance with the first sample inspection specified in Table 4.19-2 during the shu subsequent to any of the following conditions:

A seismic occurrence greater than the Operating Basis Earthquake.

major main steam line or feedwater line break.

of the limits of Specification 3.1.6.3 ection of the affected stearr ffected steam generator will be into the C-3 cate will be performed in the same Group

2. If the leaking tube is not Section 4.19.3.d. 1, then an inspection will be performed on the aEe ations. Eddy current testing either inside or outside of a tube.

means a tube containing:

(a) an inside diameter (I.D.) IGA indication with a bobbin c YODeeradation means the percentage of the tube wall thickness affected or remov by degradation.

4-80 Amendment -N.o 9 ,*,

5. Defect means an imperfection of such severity that it exceeds the repair limit. A tube containing a defect is defective.

. Repair Limit means the extent of degradation a shall be repaired or removed from service beca unserviceable prior to the next inspection.

limit i s equal to 40% of the nominal tube ter IGA indications shall be reuaired or 1 extent of 0.25 inchei, or a through wall degradatio

7. of a tube if it leaks or contains a integrity in the event of an lant accident, or a steam line or feedwater line break a d in 4.19.3.c., above.

8. of the steam generator tube from the to the top of the lower

9. I ECT to have a volumetric shail be determined (removal &om and all tubes s

4.19

/

Reports

a. DELETED

\I \

4-8 1 Amendment No. w, !5?, 286;28A 335

b. The compkte results of the seam generator rubs inssn ice

\

to the NRC within 90 days follolving completion of the generator breaker closure). The report shall include:

Number and extent of tubes inspected.

3 ach indication of an

?

J. mined). bobbin coil amplitude diameter IGBi in 4.

5. The number o f t d from service in each steam

6. of growth of inside diamete radation in accordance lumetric ID IGA management p ntained in Amerijen

\

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

\

, esults of steam generator tube inspections which fall into Category C-3 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 ?oprevent recurrence in accordance with 10 CFR 50.73.

4-82 Amendment No. 9 ,-28B,._2BA*,

rements for inspection ofthe steam generator tubes cnsurt: that the a1 integrity of this portion of the RCE will be maintained.

ce inspection of s t e m generator tubes is bas de 1.83, Revision 1. In-service inspection of steam gener rveillance of the conditions of the tubes in the event or progressive degradation due to design, manufac inspection of steam generator tubing also provides a e and cause of any tube degradation so that co taken.

\

The Unit is expected to be maintained within those try limits found to result in Ie corrosion of the steam generator tubes. If the pri chemistry limits, localized The extent of steam generator tube 1 o d d be limited by the secondary coolant activity, Specification 3.1.6.3.

The extent of cracking during plant operatio e limited by the limitation of total steam generator tube leakage between the prim (primary-to-secondary leakage = 1 gpm).

shutdown and an unscheduled inspection, repaired or removed fiom service.

scheduled inservice steam tube examinations. For tub ID IGA indications, he tube. For ID IGA indications 11 dimension will wall dimension. Ste rator tube inspections of operating plants r degradation equal to or in excess of 40% of the tube nominal wal 4-83 Amendment No.47, 29, -.*,

Where experience in similar p o of the tubes inspected should be from these critical areas Firs ns sample size may be modified subject to NRC review and approval.

Whenever the results of an ection fall into Category C-3 on the first sample inspection orted to hTRC pursuant to the requirements of Spe 19.5 c Such cases will be c y the NRC on a case-y current examination voltages referred to in this section (section 483a Amendment No. fi4&-&29+3+?09 +39

/

\

may be limited t o ssfng 6%o f the tubes i rteaa gentrator

% md rubsequent irkpecff conditions in more seTere than those i n the other steam generator. Under su circtrrnstsrllces t h e sample seguenet s h a l l be m o d i f k d t o inspett t h e most

/ severe conditions.

I-

. I

\ TABLE 4.19-2 S T E M GENERATION TUBE INSPfCTION(2)

/

1 s

- INSPECTION ii 2N0 SAMPLE INSPECTIOAl II 3RD SAMPLE I F I O Y 1 e Size 1 R e w t 1 Acti on Required 1 1 ii Result I Action Required I I ii Result I AcflGn Required I

i I i I I I I 1 I/ I I A minima of IS T u R j per IS.6.

I I

I I inspect 1 S tubes1 I i n other S.G.

I Provide n o t i f i -

I cation t o WRC notification to N I

I I

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

tubes inspected pursuant t o 4.19.2.a.4: No action i s required f o r C-u l t s i n one o r both steam generators plug or repair defective tubes. For C-3 or both steam generators, plug o r repair defective tubes and provide n o t i f i c a t i o n pursuant t o 10 CFR 50.72.6.2.1 followed by a written report pursuant t o 10

INSERT TO TS PAGE 4-77 (REVISED TS 4.191 4.19 STEAM GENERATOR (SG) TUBE INTEGRITY Aoolicability: Whenever the reactor coolant average temperature is above 200°F Surveillance Reauirements (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.

4.19.2Verify 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. The SG heat removal function is addressed by TS Section 3.4.

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 6.19. 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 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.l 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(~)(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. A portion of the parent tube length has been 4-78 2 of 7

BASES LCO (continued) removed from service in the sleeved tubes, so examination requirements for sleeved and unsleeved tubing lengths are described in the Specification.

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 burstkollapse 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 Ill, 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 the sum of the primary to secondary leakage from both 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.

ACT1ONS 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.a.(3.)b.

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 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.a.(4.) applies.

4-80 4 of 7

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@) have tube integrity, Required Action 3.1.1 .2.aS(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@) 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.

If the Required Actions and associated Completion Times of Condition 3.1 .I .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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and COLD SHUTDOWN within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

The allowed Completion Times are reasonable, based on operating experience, to reach the desired plant conditions from full power conditions in an orderly manner and without challenging plant systems.

SURVEILLANCE 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 (Lea,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 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.

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, 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.

4-82 6 of 7

Steam generator tube repairs are described in TS Section 6.19.f. All in-service tubes were repaired by kinetic expansion in the early 1980s, and approximately 250 tubes in each SG were sleeved in the early 1990s.

Installation of additional kinetic expansions, sleeves, or other type of tube repair requires prior NRC approval. ECR 02-01121 prescribes examination requirements and flaw dispositioning criteria for the kinetic expansions and sleeves. NRC approval of ECR 02-01121 was provided under Reference 7.

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 Ill, Subsection NB.

5. Draft Regulatory Guide 1.I 21, Basis for Plugging Degraded Steam Generator Tubes, August 1976.

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

7. U.S.N.R.C. Letter, Three Mile Island Nuclear Station, Unit 1 - Steam Generator Tube Kinetic Expansion Inspection and Repair Criteria (TAC No.MC7001), November 8,2005.

4-83 (Pages 4-84 through 4-85 deleted) 7 of 7

6.9.5 CORE OPERATING LIMITS REPORT

)

6.9.5.1 The core operating limits addressed by the jndividual 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 Specificationsshall 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, "MI-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 Methodologyfor TMI-lnt Revision 0. NRC SER dated 2/21/96.

(5) TR-092P-A, YMI-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. I 6.9.5.3 The core operating limits shall be determined so that all applicable limits (e.g., fuel thermal-mechanicallimits, core thermal-hydraulic limits, ECCS limits, nuclear limits such as shutdown margin, and transientlamidentanalysis limits) of the safety analysis are met.

6.9.5.4 The CORE Of RATING LIMITS REPORT, including any mid-cycle revisions or supplements thereto, shall be provided upon.issuancefor each reload cycle to the NRC Document Control Desk with copies to the RegionalAdministrator and Resident Inspector.

-c #

6-19 i

Amendment NO.?^, ??, 7

-j

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 or repaired to date,

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

j. 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.

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

1 of 1

b. Licensees may make changes to Bases without prior NRC approval provided the changes do not require either of the following:

1. A change in the TS incorporated in the license or

2. A change to the updated FSAR (UFSAR) or Bases that requires NRC approval pursuant to 10 CFR 50.59.

c. The Bases Control Program shall contain provisions to ensure that the Bases are maintained consistent with the UFSAR.

d. Proposed changes that meet the criteria of Specification 6.18.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).

6-26 Amendment N0.456-1

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-secondarypressure differential and a safety factor of 1.4 against burst applied to the design basis accident primary-to-secondarypressure 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 aom oer SG, exceot for soecific twes of dearadation at soecific locations as described in oaraaraoh 6.19.c of the Steam Generator Proaram below.

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

1 Of3

c. Provisions for SG tube repair criteria.

1. The non-sleeved regions of tubes found by inservice inspection to contain flaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.

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

a. 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 indication means an indication initiating on the inside diameter surface and confirmed by diagnostic ECT to have a volumetric morphology characteristic of IGA. ID IGA indications shall be 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 2 40% if assigned.

b. 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 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , and 9960 gallons over the MSLB duration, under this report.

2. Tubes found by inservice inspection to contain a flaw in a sleeve, or in a sleeves parent tube adjacent to the sleeve between the lower sleeve end and the parent tube kinetic expansion transition, shall be plugged-on-detection in accordance with ECR No. TM 02-01121, Rev. 2.

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 tubes repaired by sleeving, the portion of the original tube wall above the sleeves lower sleeve-to-tube rolled joints is not an area requiring re-inspection. In addition to meeting the requirements of d.1, d.2, d.3, d.4, and d.5 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 2 of 3

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@),then the indication need not be treated as a crack.

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

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

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.

TMI-1s kinetic expansion repairs installed in the 198Os, and without flaws exceeding the criteria of 6.19.c.l.b, may remain in service subject to the requirements of TS Sections 3.1.1.2, 4.19, and 6.19.

TMI-1s 80 Inconel-690 rolled sleeves installed in 1991 and 1993, and without flaws exceeding the repair criteria of 6.1 9.c.2, may remain in service subject to the requirements of TS Sections 3.1.1.2, 4.1 9, and 6.1 9.

Installation of new repair methods, additional kinetic expansions, or additional sleeves, requires prior NRC approval.

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

3 of 3

v t e Letter Sequence Response to RAI
TAC:MD1807, Steam Generator Tube Integrity (Approved, Closed)
Initiation Request, Request Acceptance... Administration Questions 20 February 2007 9 July 2007 14 August 2006 9 November 2006 17 May 2007 20 July 2007 Answers 6 October 2006 12 December 2006 31 May 2007 25 July 2007 Results Approval Other: ML072710035
MONTHYEAR2007-05-17 [Table View]

ML071520233

Text

Subject:

References:

Enclosures:

Response

Response

Response

Response

Response

Response

Response

Response

Response

Response

Response

Response

Navigation menu

Search