ML072600318
| ML072600318 | |
| Person / Time | |
|---|---|
| Site: | Three Mile Island  |
| Issue date: | 09/27/2007 |
| From: | Peter Bamford NRC/NRR/ADRO/DORL/LPLI-2 |
| To: | Crane C AmerGen Energy Co |
| Bamford, Peter J., NRR/DORL 415-2833 | |
| Shared Package | |
| ML072600307 | List: |
| References | |
| GL-06-001, TAC MD0115, TAC MD1807 | |
| Download: ML072600318 (17) | |
Text
September 27, 2007Mr. Christopher M. CranePresident and Chief Executive Officer AmerGen Energy Company, LLC 200 Exelon Way, KSA 3-N Kennett Square, PA 19348
SUBJECT:
THREE MILE ISLAND NUCLEAR STATION, UNIT 1 - ISSUANCE OFAMENDMENT REGARDING STEAM GENERATOR TUBE INTEGRITY USING THE CONSOLIDATED LINE ITEM IMPROVEMENT PROCESS AND GENERIC LETTER 2006-01 (TAC NOS. MD1807 AND MD0115)
Dear Mr. Crane:
The Nuclear Regulatory Commission (NRC, Commission) has issued the enclosed AmendmentNo. to Facility Operating License No. DPR-50 for the Three Mile Island Nuclear Station, Unit 1 (TMI-1), in response to your application dated May 15, 2006, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML061420294), as supplemented by letters dated October 6, 2006 (ADAMS Accession No. ML062830331), December 12, 2006 (ADAMS Accession No. ML063480459), May 31, 2007 (ADAMS Accession No. ML071520233),
July 25, 2007 (ADAMS Accession No. ML072120469), and September 4, 2007 (ADAMS Accession No. ML072540254). The amendment consists of changes to various technical specifications (TSs) regarding steamgenerator tube integrity. It is based on Revision 4 to Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-449, "Steam Generator Tube Integrity" and is adapted for the custom technical specifications used at TMI-1. As stated in your letter dated February 16, 2006, the amendment is also the modification of the SG portion of the TSs requested in NRC Generic Letter (GL) 2006-01, "Steam Generator Tube Integrity and Associated Technical Specifications." The NRC staff considers the amendment to be an acceptable and complete response to GL 2006-01. This completes the NRC staff's efforts on TAC No. MD1807 (Amendment ) and TAC No. MD0115 (GL 2006-01).A copy of the related safety evaluation is also enclosed. Notice of Issuance will be included inthe Commission's biweekly Federal Register notice. Sincerely,/RA/Peter J. Bamford, Project ManagerPlant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor RegulationDocket No. 50-289
Enclosures:
- 1. Amendment No. 261 to DPR-50 2. Safety Evaluationcc w/encls: See next page September 27, 2007Mr. Christopher M. Crane President and Chief Executive Officer AmerGen Energy Company, LLC 200 Exelon Way, KSA 3-N Kennett Square, PA 19348
SUBJECT:
THREE MILE ISLAND NUCLEAR STATION, UNIT 1 - ISSUANCE OFAMENDMENT REGARDING STEAM GENERATOR TUBE INTEGRITY USING THE CONSOLIDATED LINE ITEM IMPROVEMENT PROCESS AND GENERIC LETTER 2006-01 (TAC NOS. MD1807 AND MD0115)
Dear Mr. Crane:
The Nuclear Regulatory Commission (NRC, Commission) has issued the enclosed AmendmentNo. to Facility Operating License No. DPR-50 for the Three Mile Island Nuclear Station, Unit 1 (TMI-1), in response to your application dated May 15, 2006, (Agencywide Documents Access and Management System (ADAMS) Accession No. ML061420294), as supplemented by letters dated October 6, 2006 (ADAMS Accession No. ML062830331), December 12, 2006 (ADAMS Accession No. ML063480459), May 31, 2007 (ADAMS Accession No. ML071520233),
July 25, 2007 (ADAMS Accession No. ML072120469), and September 4, 2007 (ADAMS Accession No. ML072540254). The amendment consists of changes to various technical specifications (TSs) regarding steamgenerator tube integrity. It is based on Revision 4 to Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-449, "Steam Generator Tube Integrity" and is adapted for the custom technical specifications used at TMI-1. As stated in your letter dated February 16, 2006, the amendment is also the modification of the SG portion of the TSs requested in NRC Generic Letter (GL) 2006-01, "Steam Generator Tube Integrity and Associated Technical Specifications." The NRC staff considers the amendment to be an acceptable and complete response to GL 2006-01. This completes the NRC staff's efforts on TAC No. MD1807 (Amendment ) and TAC No. MD0115 (GL 2006-01).A copy of the related safety evaluation is also enclosed. Notice of Issuance will be included inthe Commission's biweekly Federal Register notice. Sincerely,/RA/Peter J. Bamford, Project Manager Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor RegulationDocket No. 50-289
Enclosures:
- 1. Amendment No. to DPR-50 2. Safety Evaluationcc w/encls: See next pageDISTRIBUTIONPUBLICLPL1-2 R/FRidsNrrDorlLPL1-2RidsNrrPMPBamfordRidsRgn1MailCenterGHill(2)RidsOgcRpRidsNrrDirsItsbRidsNrrLACSolaRidsNrrLAABaxterRidsDorlDprLMiller, NRR KKarwoski, NRRRidsNrrDraCsgbRidsAcrsAcnwMailCenterTWertz, NRRDBeaulieu, NRRAccession Nos.: Package/ML072600307 ; Amendment/ML072600318 TecSpecs/ MLOFFICELPL1-2/PMLPL2-2/LACSGB/BCITSB/BCLPL1-2/BCNAMEPBamfordCSolaAHiser w/editsTKobetzHChernoffDATE09/19/200709/19/200709/21/200709/24/20079/27/07Official Record Copy Three Mile Island Nuclear Station, Unit 1 cc:
Site Vice President - Three Mile Island Nuclear Station, Unit 1 AmerGen Energy Company, LLC P. O. Box 480 Middletown, PA 17057Vice President - Operations, Mid-Atlantic AmerGen Energy Company, LLC 200 Exelon Way, KSA 3-N Kennett Square, PA 19348Vice President - Licensing and Regulatory AffairsAmerGen Energy Company, LLC 4300 Winfield Road Warrenville, IL 60555Regional Administrator Region I U.S. Nuclear Regulatory Commission
475 Allendale Road King of Prussia, PA 19406ChairmanBoard of County Commissioners of Dauphin County Dauphin County Courthouse Harrisburg, PA 17120ChairmanBoard of Supervisors of Londonderry Township R.D. #1, Geyers Church Road Middletown, PA 17057Senior Resident Inspector (TMI-1)U.S. Nuclear Regulatory Commission P.O. Box 219 Middletown, PA 17057Director - Licensing and Regulatory AffairsAmerGen Energy Company, LLC 200 Exelon Way, KSA 3-E Kennett Square, PA 19348DirectorBureau of Radiation Protection Pennsylvania Department of Environmental Protection Rachel Carson State Office Building P.O. Box 8469 Harrisburg, PA 17105-8469Plant Manager - Three Mile Island Nuclear Station, Unit 1 AmerGen Energy Company, LLC P. O. Box 480 Middletown, PA 17057Regulatory Assurance Manager - Three Mile Island Nuclear Station, Unit 1 AmerGen Energy Company, LLC P.O. Box 480 Middletown, PA 17057Ronald Bellamy, Region IU.S. Nuclear Regulatory Commission
475 Allendale Road King of Prussia, PA 19406Michael A. SchoppmanFramatome ANP Suite 705 1911 North Ft. Myer Drive Rosslyn, VA 22209Dr. Judith JohnsrudNational Energy Committee Sierra Club 433 Orlando Avenue State College, PA 16803Eric EpsteinTMI Alert 4100 Hillsdale Road Harrisburg, PA 17112Correspondence Control DeskAmerGen Energy Company, LLC P.O. Box 160 Kennett Square, PA 19348 Three Mile Island Nuclear Station, Unit 1 cc:
Manager Licensing - Three Mile IslandNuclear Station, Unit 1 Exelon Generation Company, LLC 200 Exelon Way, KSA 3-E Kennett Square, PA 19348Mr. Christopher M. CranePresident and Chief Executive Officer AmerGen Energy Company, LLC 4300 Winfield Road Warrenville, IL 60555Assistant General CounselAmerGen Energy Company, LLC 200 Exelon Way Kennett Square, PA 19348Associate General CounselExelon Generation Company, LLC 4300 Winfield Road Warrenville, IL 60555 AMERGEN ENERGY COMPANY, LLCDOCKET NO. 50-289THREE MILE ISLAND NUCLEAR STATION, UNIT NO. 1AMENDMENT TO FACILITY OPERATING LICENSEAmendment No. 261License No. DPR-501.The Nuclear Regulatory Commission (the Commission or NRC) has found that:A.The application for amendment by AmerGen Energy Company, LLC (thelicensee), dated May 15, 2006, as supplemented by letters dated October 6, 2006, December 12, 2006, May 31, 2007, July 25, 2007 and September 4, 2007, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in Title 10 of the Code of Federal Regulations (10 CFR) Chapter I;B.The facility will operate in conformity with the application, the provisions of theAct, and the rules and regulations of the Commission;C.There is reasonable assurance: (i) that the activities authorized by thisamendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations;D.The issuance of this amendment will not be inimical to the common defense andsecurity or to the health and safety of the public; andE.The issuance of this amendment is in accordance with 10 CFR Part 51 of theCommission's regulations and all applicable requirements have been satisfied. 2.Accordingly, the license is amended by changes to the Technical Specifications asindicated in the attachment to this license amendment, and paragraph 2.c.(2) of Facility Operating License No. DPR-50 is hereby amended to read as follows:(2)Technical SpecificationsThe Technical Specifications contained in Appendix A, as revised throughAmendment No. 261, are hereby incorporated in the license. The AmerGen Energy Company, LLC, shall operate the facility in accordance with the Technical Specifications. 3.This license amendment is effective as of its date of issuance and shall be implementedwithin 30 days of issuance. Implementation of the amendment shall include updating the UFSAR in accordance with 10CFR50.71(e). This update shall include the effects of all safety analyses and evaluations performed by the licensee in support of the amendment. Specifically, the UFSAR update shall include, but not be limited to:
a statement that the NRC review of this amendment regarding sleeve parent tube inspections was limited to operation for only one cycle beyond refueling outage T1R17, scheduled for the fall of 2007.FOR THE NUCLEAR REGULATORY COMMISSION/RA/Harold K. Chernoff, ChiefPlant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation
Attachment:
Changes to License No. DPR-50and the Technical SpecificationsDate of Issuance: September 27, 2007 ATTACHMENT TO LICENSE AMENDMENT NO. 261FACILITY OPERATING LICENSE NO. DPR-50DOCKET NO. 50-289Replace the following pages of the Facility Operating License No. DPR-50 with the attachedrevised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.RemoveInsertPage 3Page 3Page 6Page 6 Page 7Page 7Replace the following pages of the Appendix A Technical Specifications with the attachedrevised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.RemoveInsertiviv vv vivi 3-1a3-1a
3-1b 3-123-12 4-84-8 4-774-774-78---- *4-79---- *4-80---- *4-81---- *4-82---- *4-84----4-85----
6-196-19 6-206-20 6-266-26
6-27
6-28
6-29*Page was formerly a specification page whose information has been deleted withthis amendment. The NRC recognizes that this page number may becomeassociated with a bases page under the control of the TMI-1 TS Bases Control Program.
SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATIONRELATED TO AMENDMENT NO. 261 TO FACILITY OPERATING LICENSE NO. DPR-50AMERGEN ENERGY COMPANY, LLCTHREE MILE ISLAND NUCLEAR STATION, UNIT 1DOCKET NO. 50-28
91.0 INTRODUCTION
By application dated May 15, 2006 (Agencywide Documents Access and Management System(ADAMS) Accession No. ML061420294), as supplemented by letters dated October 6, 2006 (ADAMS Accession No. ML062830331), December 12, 2006 (ADAMS Accession No.
ML063480459), May 31, 2007 (ADAMS Accession No. ML071520233), July 25, 2007 (ADAMS Accession No. ML072120469), and September 4, 2007 (ADAMS Accession No.
ML072540254), AmerGen Energy Company, LLC (the licensee), requested changes to the technical specifications (TSs) for Three Mile Island Nuclear Station, Unit 1 (TMI-1). The supplements provided additional information that clarified the application, did not expand thescope of the application as originally noticed, and did not change the staff's original proposed no significant hazards consideration determination as published in the Federal Register on July18, 2006 (71 FR 40744).The proposed changes would revise various TSs regarding steam generator tube integrity. Theproposed changes are based upon Revision 4 to Technical Specification Task Force (TSTF)
Standard Technical Specification Change Traveler, TSTF-449, "Steam Generator Tube Integrity" and are adapted for the custom technical specifications used at TMI-1. The amendment is related to the TMI-1 response to Nuclear Regulatory Commission (NRC) Generic Letter 2006-01, "Steam Generator Tube Integrity and Associated Technical Specifications,"
dated February 16, 2006. Specifically, the proposed changes would delete license condition 2.c.(8), "Repaired SteamGenerators," as well as modify TS 3.1.6, "LEAKAGE," TS 3.1.1.2, "Steam Generators and Steam Generator (SG) Tube Integrity," and TS Section 4.19, "Steam Generator (SG) Tube Integrity." The changes would also add new TS 6.9.6, "Steam Generator Tube Inspection Report" and new TS 6.19, "Steam Generator (SG) Program."
2.0 REGULATORY EVALUATION
The background, description, and applicability of the proposed changes associated with theSG tube integrity issue and the applicable regulatory requirements were included in the NRC staff's model safety evaluation (SE) published in the Federal Register on March 2, 2005(70 FR 10298). The "Notice of Availability of Model Application Concerning Technical Specification; Improvement To Modify Requirements Regarding Steam Generator Tube Integrity; Using the Consolidated Line Item Improvement Process," was published in theFederal Register on May 6, 2005 (70 FR 24126), and made the model SE available forlicensees to reference.
3.0 TECHNICAL EVALUATION
3.1 OverviewIn its May 15, 2006 application, and supplements listed above, the licensee proposed changesto the TSs that are modeled after TSTF-449, "Steam Generator Tube Integrity." There were minor differences between TSTF-449 and the licensee's application. These included differences in the facility licensing basis (than that discussed in TSTF-449), differences in TSformat and numbering, and changes that went beyond TSTF-449. These differences are discussed below.With respect to the differences in the facility licensing basis, the differences did not invalidatethe technical evaluation on TSTF-449; rather they resulted in the licensee having to slightly deviate from some of the modifications discussed in TSTF-449. For example, TSTF-449 indicates that analysis of an SG tube rupture event assumes a bounding primary-to-secondary leakage rate equal to the operational leakage rate limit plus the leakage rate associated with a double-ended rupture of a single tube. Since the licensee's current licensing basis only assumes the leak rate associated with the double-ended rupture of a single tube, the licensee did not incorporate the part of the sentence in TSTF-449 that indicates they added the leakage rate equal to the operational leakage rate limit. Another example is that TSTF-449 has specific definitions for the modes of operation. For example, in TSTF-449, Mode 3 is referred to as hot standby and the average reactor coolant temperature during this mode must be greater than or equal to 330-degrees Fahrenheit and the reactor must be subcritical. For TMI-1, the definition of hot standby indicates that the average reactor coolant temperature must be greater than 525 degrees Fahrenheit and the reactor must be critical. These differences in modes required the licensee to deviate from the changes discussed in TSTF-449 to remain consistent with the intent of TSTF-449 (e.g., referencing hot shutdown rather than hot standby since TMI-1's TS definition of hot shutdown is similar to the definition of hot standby in TSTF-449 or by simply referencing a temperature within the TSs rather than the mode of operation).Another example of a difference between TMI-1's licensing basis and that assumed inTSTF-449 is that the licensee's current licensing basis assumes that the sum of the primary-to-secondary leakage from both SGs is less than or equal to 144 gallons per day.
Since this is more restrictive than the TSTF-449 limit on primary-to-secondary leakage of 150 gallons per day per SG, the licensee needed to modify (including deleting) the TSTF-449 wording in several places (including the TS and Bases) to maintain consistency with their licensing basis. Another example is that the accident analyses for TMI-1 assume a primary-to-secondary leakage rate for some design basis accidents and assume a leakage volume in the main steam line break accident analyses. Since TSTF-449 only considered that licensees assumed a leakage rate (and not a volume), the licensee modified the wording for the accident induced leakage performance criterion in TSTF-449 (which only reflects that the accident analyses assumes a leakage rate) to maintain consistency with their licensing basis.Since the differences between the licensee's proposal and TSTF-449 as a result of the above differences in licensing bases were minor in nature, they were consistent with the plant'slicensing basis, or they were consistent with the intent of TSTF-449, the NRC staff determined they were acceptable.With respect to the differences in the numbering of the TSs, these differences wereadministrative in nature and did not affect the technical adequacy of the submittal. As a result, the NRC staff determined they were acceptable.With respect to the differences in the format of the TSs, this resulted in the licensee having torelocate many of the TSTF-449 requirements into the appropriate sections of their TSs (e.g., the limiting conditions for operation and action statements are in different sections than the surveillance requirements). In addition, the licensee does not have a TS section where various leakage terms are defined so it did not need to make changes to these definitions in order to adopt TSTF-449. These differences in TS format also resulted in listing the requirements in sentence format rather than tabular format and using slightly different terminology. Lastly, these differences in format also required a few minor changes that were not discussed in TSTF-449. Since these differences were administrative in nature or consistent with the intent of TSTF-449, the NRC staff determined they were acceptable since they did not affect the technical adequacy of the submittal.The licensee proposed to include previously-approved alternate repair criteria into theirproposed new TSs. The structure of TSTF-449 allows licensees to incorporate alternate repair criteria into the TSTF-449 format. By incorporating the previously-approved repair criteria into the TSTF-449 format, there were several additions, deletions, and changes to the requirements. These changes (including additions and deletions), were made as a result of the format, content, and performance-based approach of TSTF-449. The staff verified that: (a) the inspection criteria associated with these repair criteria weremoved, as appropriate, to the inspection section of the proposed SG TSs, (b) the repair criteria were moved, as appropriate, to the repair criteria section of the proposed SG TSs, and (c) the reporting requirements were moved to the reporting section of the proposed SG TSs. With respect to the deletion of requirements associated with the alternate repair criteria, most of these deletions were a result of no longer requiring (in TSTF-449), the characterization of inspection results (as Category C-1, C-2, or C-3). With respect to the addition of repair requirements into the TSs, the most significant addition was the incorporation of the repair criteria for degradation within the tubesheet regions. This addition was necessary since the previous version of the TMI-1 TSs did not require an inspection of the portion of the tubing within the tubesheets. Since TSTF-449 would require inspection of the portion of the tubing within the tubesheets, the licensee proposed to include repair criteria for this portion of tubing.
For the non-kinetically expanded portion of tubing within the tubesheet, the licensee proposed to apply the standard depth-based through-wall limit of 40 percent to all flaws. With respect to the non-sleeved, kinetically expanded portion of tubing, the licensee proposed to apply the alternate repair criteria discussed in Engineering Report ECR-02-01121. The NRC staff reviewed the repair criteria that the licensee proposed in their TS for the tubesheet region and determined it was acceptable since the repair criteria for the non-kinetically expanded region of the tubing within the tubesheet region was consistent with the repair criteria for the remainder of the tube (i.e., 40 percent through-wall) and the repair criteria for the non-sleeved kinetically expanded region was consistent with the repair criteria discussed in Engineering Report ECR-02-01121 (which the staff had previously reviewed as discussed below). The licensee also added inspection and reporting requirements associated with implementation of the repaircriteria for the kinetically expanded region of the tubing. Since these proposed inspection and reporting requirements were consistent with the reporting requirements identified during the NRC staff's original review, the NRC staff found the proposed addition of these requirements acceptable.In summary, the NRC staff determined that the previously approved repair criteria wereappropriately incorporated into the plant's TSs.With respect to tube repair methods, the licensee's previous TSs indicated that they couldremove tubes from service by plugging, or repair tubes by kinetic expansion, sleeving, or other methods. In adopting TSTF-449, the licensee proposed to permit removing tubes from service only by plugging rather than to permit repairing tubes by the methods listed in their current TS.
Since this proposal is more restrictive than their current TSs, the NRC staff found it acceptable.
However, since the licensee had repaired tubes in the hot-leg tubesheet region by kinetic expansion and had sleeved some tubes (primarily as a preventive measure to stiffen the tubes limiting the potential for fatigue failure of the tubes), the licensee needed to reflect this condition in their proposed TSs (i.e., that tube repairs had been made, but that no additional tubes could be repaired). This required several deviations from the TSs in TSTF-449. The NRC staff reviewed the proposed wording regarding tube repairs, which is contained in TS 6.19.f, and found it acceptable. In addition, since TSTF-449 specifies the repair criteria for the tubes, including the tube repairs, the licensee needed to propose to incorporate repair criteria associated with the repairs that they had previously implemented to be consistent with TSTF-449. The repair criteria associated with the currently installed kinetic expansions and sleeves are discussed below.The NRC staff reviewed and approved the use of kinetic expansions as discussed inNUREG-1019, "Safety Evaluation Report Related to Steam Generator Tube Repair and Return to Operation - Three Mile Island Nuclear Station, Unit No. 1," dated November 1983. Similarly, the NRC staff reviewed and approved the repair criteria for the non-sleeved portion of the kinetic expanded region of the tubing when it reviewed Engineering Report ECR-02-01121.
The NRC staff's review is documented in a letter dated November 8, 2005 (ADAMS Accession No. ML052840138). The NRC staff's review of Engineering Report ECR-02-01121 only consisted of a review of the repair criteria associated with the non-sleeved, kinetically expanded region of the tubing. As discussed above, the NRC staff determined that these previously approved repair criteria were appropriately incorporated into the TSs.The sleeves installed at TMI-1 are 80 inches long and extend from the upper tube end downthrough the 15 th tube support plate. The licensee indicated that the sleeves were installed(a) in accordance with Title 10 of the Code of Federal Regulations (10 CFR) 50.59 and(b) consistent with the qualification report for the sleeves. The licensee further indicated that this qualification report was generically approved by the NRC staff for referencing in licensing applications. In essence, the sleeves used at TMI-1 are similar to those approved for use at other plants. There is, however, one notable difference between the conditions at TMI-1 and the conditions under which the original sleeves were generically qualified. That is, the upper sleeve joint near the upper tube end at TMI-1 was installed in a region of tubing that had experienced widespread inside diameter intergranular attack in the early 1980s. This was the region of tubing repaired by kinetic expansions (discussed above). Given the difference between the original qualification report (which was generically approvedby the NRC staff for referencing in licensing applications) and the site-specific conditions atTMI-1, the licensee indicated that additional testing was performed to evaluate the potential impact of the parent tube defects on sleeve integrity. The licensee indicated that this work is described in their 10 CFR 50.59 evaluation and it was concluded that parent tube defects were not detrimental to sleeve integrity. Given the successful use of similar sleeves at other facilities, the limited additional time the licensee plans to keep the existing sleeves in service (i.e., the licensee plans to replace these SGs in 2009), the removal of authorization to sleeve from the TSs, and the conclusions cited by the licensee in their 10 CFR 50.59 evaluation, the NRC staff did not review the qualification of the sleeve design for use at TMI-1 as part of this review (e.g.,
the NRC staff did not review the basis for the conclusion that parent tube defects at the upper sleeve joint are not detrimental to the integrity of the sleeve joint). However, the licensee did commit to revise their Updated Final Safety Analysis Report in the next required periodic updateto identify that the technical basis for the adequacy of the original SG tube sleeve installation will be submitted to the NRC if the existing TMI-1 SGs are not replaced in the 1R18 refueling outage (currently planned for Fall 2009). The staff finds this acceptable for the reasons cited above (e.g., successful use at other facilities and limited additional in service time of the existing sleeves).There are no inspection or repair criteria in the current TSs pertaining to these repairs. Withthe adoption of the general inspection requirements in TSTF-449, the licensee will have appropriate inspection requirements for the sleeves. With respect to the repair criteria for the sleeves, the licensee has proposed to plug tubes found by inservice inspection to contain a flaw in a sleeve, or in the parent tube adjacent to a sleeve between the lower sleeve end and the top of the middle sleeve roll. The NRC staff finds this proposal acceptable since it is consistent with(or more conservative than) what is implemented at other plants with similar sleeves installed, where the staff has found provides reasonable assurance of tube integrity. With respect to the repair criteria for the parent tube adjacent to the sleeve upper tubesheet roll expansion, the licensee proposed to plug the tube if there is any change in the number, orientation, or size of the flaw(s) in the parent tube when compared to the first inservice inspection of this region with a rotating +Point TM probe. Although this assumes that the condition of the tubing wasacceptable at the time of the first inservice inspection of this region with a rotating +Point TMprobe, the NRC staff finds it acceptable given the reasons cited above (e.g, licensee's conclusion indicating that parent tube defects are not detrimental to tube integrity and limited additional in service time of the existing sleeves).In addition to the above, the licensee also proposed a few changes that adapt the TMI-1 customTS and facility operating license to the TSTF-449 intent or that apply a more restrictive and therefore conservative criteria. For example, the licensee proposed to delete License Condition 2.c(8), "Repaired Steam Generators." With respect to the requirements in this License Condition, the only requirement that was still pertinent was the limit on the rate of primary-to-secondary leakage. Since the limit on the primary-to-secondary leakage rate was incorporated into TS Section 3.1.6, the NRC staff found the proposal to delete License Condition 2.c(8) acceptable. As discussed above, the limit on the primary-to-secondary leakage rate incorporated in TS Section 3.1.6 is more conservative than that in TSTF-449. In addition, the licensee proposed a 90-day reporting requirement rather than the 180-day reporting requirement in TSTF-449. Since the licensee's proposal is more restrictive than that contained in TSTF-449, the NRC staff finds it acceptable. The remainder of the application was consistent with, or more limiting than, TSTF-449. The licensee proposed to change the frequency for verifying primary-to-secondary leakage iswithin acceptable limits from 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Since this proposed TS change provides a reasonable interval to trend primary-to-secondary leakage, recognizes the importance of early leakage detection in the prevention of accidents and is consistent with TSTF-449, the NRC staff determined that the proposed change was acceptable.During the review of the licensee's application, the NRC staff interacted with the licensee on anumber of issues related to the accident induced leakage performance criterion. These issues included wording of the proposed TS, wording of the Bases, and implementation of the performance criterion. With respect to the wording of the proposed TS, the NRC staff concludes that the licensee's proposal is consistent with the intent of TSTF-449 and its licensing basis. In addition to the above, the NRC staff also identified several issues related to how the licenseemay implement the accident induced leakage performance criteria. The proposed TS require that the primary-to-secondary leakage rate or volume for any design basis accident must be kept below the value assumed in the accident analyses. During each outage where SG tubes are inspected or plugged, condition monitoring assessments must be made against the specified performance criteria. In response to a request for additional information, the licensee described an approach for implementing the accident induced leakage performance criterion in its July 25, 2007 submittal (see response to first NRC question). Subsequent to that submittal, the NRC staff discussed with the licensee that the approach described in their response to the NRC staff's question did not appear to clearly demonstrate that the accident induced leakage performance criterion would always be met. During these discussions, the staff also indicated to the licensee that several approaches forimplementing this requirement could be taken. For example, the licensee could calculate the amount of primary-to-secondary leakage it would expect to have for each design basis accident that assumes primary-to-secondary leakage exists, other than an SG tube rupture, based on the flaws in the SG. The licensee could then compare these calculated leakage rates to the leakage rate assumed in the appropriate accident analyses. If the leakage rates are less than or equal to that assumed in the corresponding accident analyses, the licensee would satisfy the accident induced leakage performance criterion. Another approach for satisfying this criterion would be to try to determine whether one of the design basis accidents would always result in the highest leakage rate for the distribution of flaws in the SG. If this is the case, the calculated leakage rate for this accident could then be compared to the leakage rate for the design basis accident with the lowest assumed primary-to-secondary leakage rate. Since this ensures the primary-to-secondary leakage rate for all design basis accidents is within the limits assumed the accident analyses, the licensee would satisfy the criterion. Although the preceding discussionreferences the primary-to-secondary leakage rate, a similar approach can be used for accidents that assume a specific volume of primary-to-secondary leakage. In addition, there may be other approaches that could result in satisfying the accident induced leakage performance criterion (e.g., demonstrating that the amount of leakage expected during a specific accident will always be less than what is observed during normal operation - since the primary-to-secondary leakage rate assumed during an accident must be greater than or equal to the leakage rate permitted during normal operation). NRC Regulatory Issue Summary 2007-20, "Implementation of Primary-to-Secondary Leakage Performance Criteria," contains similar and additional guidance concerning implementing the accident induced leakageperformance criteria. In conclusion, since the licensee's proposed specification wording for their accident induced leakage performance criterion was acceptable, the staff determined theproposed TS changes were acceptable. The licensee is responsible for ensuring the approach they take for assessing accident induced leakage will ensure the requirement is fully met.In summary, the staff determined that the model SE is applicable to this review and finds theproposed changes acceptable. Consistent with TSTF-449, the proposed TS changes include:
(1) a revised TS Section 3.1.1.2, "Steam Generator (SG) Tube Integrity," (2) a revised TS Section 3.1.6, "Leakage", (3) a revised Table 4.1-2, "Minimum Equipment Test Frequency,"
(4) a revised TS 4.19, "Steam Generator (SG) Tube Integrity," (5) a new TS Section 6.9.6, "Steam Generator Tube Inspection Report," (6) a new TS Section 6.19, "Steam Generator (SG)
Program," and (7) revised Table of Content pages to reflect the proposed changes.3.2ConclusionThe proposed TS changes establish a programmatic, largely performance-based regulatoryframework for ensuring SG tube integrity is maintained. The NRC staff finds that it addresses key shortcomings of the current framework by ensuring that SG programs are focused on accomplishing the overall objective of maintaining tube integrity. It incorporates performance criteria for evaluating tube integrity that the NRC staff finds consistent with the structural margins and the degree of leak tightness assumed in the current plant licensing basis. The NRC staff finds that maintaining these performance criteria provides reasonable assurance that the SGs can be operated safely without increase in risk.The revised TSs will contain limited specific details concerning how the SG Program is toachieve the required objective of maintaining tube integrity; the intent being that the licensee will have the flexibility to determine the specific strategy for meeting this objective. However, the NRC staff finds that the revised TSs include sufficient regulatory constraints on the establishment and implementation of the SG Program such as to provide reasonable assurance that tube integrity will be maintained.Failure to meet the performance criteria will be reportable pursuant to the requirements in10 CFR Parts 50.72 and 50.73. The NRC reactor oversight process provides a process by which the NRC staff can verify that the licensee has identified any SG Program deficiencies that may have contributed to such an occurrence and that appropriate corrective actions have been implemented.In conclusion, the NRC staff finds that the TS changes proposed by the licensee in itsMay 15, 2006, application and the supplements listed above conform to the requirements of 10 CFR 50.36 and establish a TS framework that will provide reasonable assurance that SG tube integrity is maintained without undue risk to public health and safety.The licensee included in its application the revised TS Bases to be implemented with the TSchange. The NRC staff finds that the TS Bases Control Program is the appropriate process for updating the affected TS Bases pages and has, therefore, not specifically reviewed or included the affected Bases pages with this amendment unless, for the custom TS used at TMI-1, the Bases pages also include changes to the Specification.
4.0 STATE CONSULTATION
In accordance with the Commission's regulations, the Pennsylvania State official was notified ofthe proposed issuance of the amendment. The State official had no comments.
5.0 ENVIRONMENTAL CONSIDERATION
The amendment changes a requirement with respect to installation or use of a facilitycomponent located within the restricted area as defined in 10 CFR Part 20. The NRC staff has determined that the amendment involves no significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. TheCommission has previously issued a proposed finding that the amendment involves no significant hazards consideration, and there has been no public comment on such finding (71 FR 40744). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.
6.0 CONCLUSION
The Commission has concluded, based on the considerations discussed above that (1) there isreasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.Principal Contributors: L. MillerK. KarwoskiDate: September 27, 2007