ML071650519

Issuance of License Amendment 259, One-Time Type a Test Interval Extension

ML071650519
Person / Time
Site:	Three Mile Island
Issue date:	06/29/2007
From:	Peter Bamford NRC/NRR/ADRO/DORL/LPLI-2
To:	Crane C AmerGen Energy Co
Bamford, Peter J., NRR/DORL 415-2833
Shared Package
ML071650511	List: ML071650519 ML071800319
References
TAC MD3027
Download: ML071650519 (16)

Text

June 29, 2007 Mr. 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 OF AMENDMENT REGARDING ONE-TIME TYPE A TEST INTERVAL EXTENSION (TAC NO. MD3027)

Dear Mr. Crane:

The Commission has issued the enclosed Amendment No. 259 to Facility Operating License No. DPR-50 for the Three Mile Island Nuclear Station, Unit 1 (TMI-1), in response to your application dated September 15, 2006, as supplemented by letters dated February 26, 2007, May 22, 2007, and June 5, 2007.

The amendment consists of changes to the technical specifications (TSs) to allow a one-time deferral of the next Type A, containment integrated leak rate test (ILRT) from no later than September 2008" to prior to startup from T1R18 refueling outage. The T1R18 refueling outage will begin no later than November 1, 2009. This change would add approximately 15 months to the previously approved 15-year interval and allow the Type A ILRT to be performed during a steam generator replacement outage in the fall of 2009.

A copy of the related safety evaluation is also enclosed. A Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely,

/ra/

Peter Bamford, Project Manager Plant Licensing Branch I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-289

Enclosures:

1. Amendment No. 259 to DPR-50

2. Safety Evaluation cc w/encls: See next page

June 29, 2007 Mr. Christopher M. Crane President and Chief Nuclear Officer AmerGen Energy Company, LLC 200 Exelon Way, KSA 3-N Kennett Square, PA 19348

SUBJECT:

THREE MILE ISLAND NUCLEAR STATION, UNIT 1 ISSUANCE OF AMENDMENT REGARDING ONE-TIME TYPE A TEST INTERVAL EXTENSION (TAC NO. MD3027)

Dear Mr. Crane:

The Commission has issued the enclosed Amendment No. 259 to Facility Operating License No. DPR-50 for the Three Mile Island Nuclear Station, Unit 1 (TMI-1), in response to your application dated September 15, 2006, as supplemented by letters dated February 26, 2007, May 22, 2007, and June 5, 2007.

The amendment consists of changes to the technical specifications (TSs) to allow a one-time deferral of the next Type A, containment integrated leak rate test (ILRT) from no later than September 2008" to prior to startup from T1R18 refueling outage. The T1R18 refueling outage will begin no later than November 1, 2009. This change would add approximately 15 months to the previously approved 15-year interval and allow the Type A ILRT to be performed during a steam generator replacement outage in the fall of 2009.

A copy of the related safety evaluation is also enclosed. A Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely,

/ra/

Peter Bamford, Project Manager Plant Licensing Branch I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-289

Enclosures:

1. Amendment No. 259 to DPR-50

2. Safety Evaluation cc w/encls: See next page DISTRIBUTION PUBLIC LPLI-2 R/F RidsNrrDorlLPLI-2 RidsNrrPMPBamford RidsRgn1MailCenter G. Hill (2) RidsOgcRp RidsNrrDirsItsb RidsNrrPMEMiller RidsNrrLACSola RidsDorlDpr RidsAcrsAcnwMailCenter RidsNrrDraApla RidsNrrDssScvb RidsNrrDeEmcb GThomas, NRR HAshar, NRR RPalla, NRR Accession Number: ML071650519; TS: ML071800319; Package: ML071650511

• SE provided, no substantive changes.

OFFICE LPLI-2/PM LPLI-2/PM LPLI-2/LA DRA/BC SCVB/BC EMCB/BC OGC (NLO) LPLI-2/BC NAME PBamford EMiller CSola MRubin* RDennig KManoly MBaty HChernoff DATE 6/14/2007 6/27/2007 6/27/07 03/07/2007 6/28/07 6/28/07 6/28/2007 6/29/07 Official Record Copy

Three Mile Island Nuclear Station, Unit 1 cc:

Site Vice President - Three Mile Island Nuclear Director Station, Unit 1 Bureau of Radiation Protection AmerGen Energy Company, LLC Pennsylvania Department of P. O. Box 480 Environmental Protection Middletown, PA 17057 Rachel Carson State Office Building P.O. Box 8469 Vice President - Operations, Mid-Atlantic Harrisburg, PA 17105-8469 AmerGen Energy Company, LLC 200 Exelon Way, KSA 3-N Plant Manager - Three Mile Island Nuclear Kennett Square, PA 19348 Station, Unit 1 AmerGen Energy Company, LLC Vice President - Licensing and Regulatory Affairs P. O. Box 480 AmerGen Energy Company, LLC Middletown, PA 17057 4300 Winfield Road Warrenville, IL 60555 Regulatory Assurance Manager - Three Mile Island Nuclear Station, Unit 1 Regional Administrator AmerGen Energy Company, LLC Region I P.O. Box 480 U.S. Nuclear Regulatory Commission Middletown, PA 17057 475 Allendale Road King of Prussia, PA 19406 Ronald Bellamy, Region I U.S. Nuclear Regulatory Commission Chairman 475 Allendale Road Board of County Commissioners King of Prussia, PA 19406 of Dauphin County Dauphin County Courthouse Michael A. Schoppman Harrisburg, PA 17120 Framatome ANP Suite 705 Chairman 1911 North Ft. Myer Drive Board of Supervisors Rosslyn, VA 22209 of Londonderry Township R.D. #1, Geyers Church Road Dr. Judith Johnsrud Middletown, PA 17057 National Energy Committee Sierra Club Senior Resident Inspector (TMI-1) 433 Orlando Avenue U.S. Nuclear Regulatory Commission State College, PA 16803 P.O. Box 219 Middletown, PA 17057 Eric Epstein TMI Alert Director - Licensing and Regulatory Affairs 4100 Hillsdale Road AmerGen Energy Company, LLC Harrisburg, PA 17112 200 Exelon Way, KSA 3-E Kennett Square, PA 19348 Correspondence Control Desk AmerGen Energy Company, LLC P.O. Box 160 Kennett Square, PA 19348

Three Mile Island Nuclear Station, Unit 1 cc:

Manager Licensing - Three Mile Island Nuclear Station, Unit 1 Exelon Generation Company, LLC 200 Exelon Way, KSA 3-E Kennett Square, PA 19348 Mr. Christopher M. Crane President and Chief Executive Officer AmerGen Energy Company, LLC 4300 Winfield Road Warrenville, IL 60555 Assistant General Counsel AmerGen Energy Company, LLC 200 Exelon Way Kennett Square, PA 19348 Associate General Counsel Exelon Generation Company, LLC 4300 Winfield Road Warrenville, IL 60555 Mr. Christopher M. Crane President and Chief Nuclear Officer Exelon Nuclear Exelon Generation Company, LLC 200 Exelon Way, KSA 3-E Kennett Square, PA 19348

AMERGEN ENERGY COMPANY, LLC DOCKET NO. 50-289 THREE MILE ISLAND NUCLEAR STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 259 License No. DPR-50

1. The Nuclear Regulatory Commission (the Commission or NRC) has found that:

A. The application for amendment by AmerGen Energy Company, LLC (the licensee), dated September 15, 2006, as supplemented by letters dated February 26, 2007, May 22, 2007, and June 5, 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 the Act, and the rules and regulations of the Commission; C. There is reasonable assurance: (i) that the activities authorized by this amendment 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 and security or to the health and safety of the public; and E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated 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 Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 259, 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 implemented within 60 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

/ra/

Harold K. Chernoff, Branch Chief Plant Licensing Branch I-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Facility Operating License No. DPR-50 and the Technical Specifications Date of Issuance: June 29, 2007

ATTACHMENT TO LICENSE AMENDMENT NO. 259 FACILITY OPERATING LICENSE NO. DPR-50 DOCKET NO. 50-289 Replace page 3 of Facility Operating License No. DPR-50 with the attached revised page 3.

The revised page is identified by amendment number and contains a marginal line indicating the area of change.

Replace the following page of the Technical Specifications with the attached page. The revised page is identified by amendment number and contains marginal lines indicating the area of change.

Remove Insert 6-11c 6-11c

SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 259 TO FACILITY OPERATING LICENSE NO. DPR-50 AMERGEN ENERGY COMPANY, LLC THREE MILE ISLAND NUCLEAR STATION, UNIT 1 DOCKET NO. 50-289

1.0 INTRODUCTION

By application dated September 15, 2006 (Reference (Ref.) 7.1), as supplemented by letters dated February 26, 2007 (Ref. 7.2), May 22, 2007 (Ref. 7.7) and June 5, 2007 (Ref. 7.8),

AmerGen Energy Company, LLC (AmerGen or the licensee), requested changes to the technical specifications (TSs) for Three Mile Island Nuclear Station, Unit 1 (TMI-1). The supplemental letters provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the staffs original proposed no significant hazards consideration determination as published in the Federal Register on December 19, 2006 (71 FR 75989).

The proposed change would revise Technical Specification (TS) Section 6.8.5, Reactor Building Leakage Rate Testing Program, to allow a one-time deferral of the next Type A, containment integrated leak rate test (ILRT) from no later than September 2008 to prior to startup from T1R18 refueling outage. The T1R18 refueling outage will begin no later than November 1, 2009. The Nuclear Regulatory Commission (NRC or the Commission) has previously approved a one-time 5-year extension to the Type A ILRT schedule for TMI-1 by issuance of Amendment No. 244, dated August 14, 2003 (Ref. 7.4). Amendment No. 244 changed the TSs to state that the Type A ILRT shall be performed no later than September 2008. The proposed amendment would add approximately 15 months to the currently approved 15-year interval. This deferral would allow the Type A ILRT to be performed during a steam generator (SG) replacement outage in the fall of 2009.

2.0 REGULATORY EVALUATION

Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix J, Option B requires that a Type A test be conducted at a periodic interval based on historical performance of the overall containment system. TMI-1 TS 6.8.5, Reactor Building Leakage Rate Testing Program, requires that leakage rate testing be performed as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions and in accordance with the guidelines contained in Regulatory Guide (RG) 1.163, Performance-Based Containment Leak-Test Program, dated September 1995, with one exception (which is also pertinent to the current request and is discussed in a succeeding paragraph). This RG endorses, with certain exceptions, Nuclear Energy Institute (NEI), report NEI 94-01, Revision 0, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, dated July 26, 1995.

The Type A test measures an overall integrated leakage rate of the containment as a barrier against the release of fission products to the outside environment. NEI 94-01 specifies an initial test interval of 48 months, but allows an extended interval of 10 years, based upon two consecutive successful tests. There is also a provision for extending the test interval an additional 15 months in certain circumstances. The most recent two Type A tests at TMI-1 (December 1990 and September 1993) have been successful, so the current interval requirement would normally be 10 years. However, by letter dated September 2002 (Ref. 7.3),

as supplemented on March 19, 2003, the licensee requested a one-time extension of the test interval from 10 years to 15 years based on historical performance of its containment supported by a risk-informed analysis. On August 14, 2003, the NRC staff granted this request to TMI-1 via License Amendment No. 244, which revised the TS to state that the specified Type A test shall be performed no later than September 2008 (Ref. 7.4).

By letter dated September 15, 2006 and additional information letter dated February 26, 2007, the licensee requested a proposed change to TS 6.8.5, which would alter their exception from the guidelines of RG 1.163 and NEI 94-01, by adding approximately 15 more months to the one-time 15-year Type A test interval already in place, for a total interval of 16 years and 3 months. Specifically, the change request states that the first Type A test performed after the September 1993 Type A test shall be performed prior to startup from the T1R18 refueling outage which is currently scheduled to commence in October 2009. During the T1R18 outage scheduled for the fall of 2009, the licensee plans to replace the SGs.

The proposed TS change does not involve any other changes to licensing commitments or acceptance criteria. The local leakage rate tests (Type B and Type C tests) are not affected by this license amendment request.

3.0 TECHNICAL EVALUATION

3.1 General The TMI-1 containment pressure boundary consists of a post-tensioned reinforced concrete structure with metal liner (0.25 inch thick at the base and 0.375 inch thick elsewhere),

containment access penetrations (equipment hatch, air-locks), and other process piping and electrical penetrations. The leak-tight integrity of the penetrations and isolation valves are verified through Type B and Type C local leak rate tests (LLRTs) and the overall leak-tight integrity of the primary containment is verified through a Type A ILRT as required by 10 CFR 50, Appendix J. These tests are performed at the design basis accident (DBA) pressure. The most recent Type A ILRT for TMI-1 was performed in September 1993. By License Amendment No. 244, the NRC staff granted the licensee a one-time extension of the Type A test interval from 10 years (per guidelines of RG 1.163 and NEI 94-01) to 15 years based on historical performance of the containment supported by a risk-informed analysis. This amendment changed the TS to state that the specified Type A test shall be performed no later than September 2008.

In its current submittals (Ref. 7.1 and Ref. 7.2), the licensee has stated that compliance with the current TS would require performing the Type A test during the T1R17 outage (scheduled for fall 2007) or during an unplanned mid-cycle outage in 2008; and additional post-repair containment pressure testing during the T1R18 outage (scheduled for fall 2009) due to a breach in containment for the planned SG replacement project. The planned SG replacement

will require creating an approximately 21 feet (ft) wide by 25 ft high opening in the containment.

Following the planned SG replacement during the T1R18 outage, the licensee will have to perform a pressure test as required by the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI, Subsection IWL 5000 and Subsection IWE 5000 to verify leak-tightness of the restored metal liner. Subsection IWL 5000 further requires a containment pressure test at the DBA pressure to verify the structural integrity of the restored containment. The licensee has stated that performing the Type A ILRT concurrent with the post repair containment pressure test prior to startup following SG replacement will provide a more comprehensive test of the restored containment for leak-tightness and structural integrity. The licensee also stated that the proposed TS amendment request to defer the next Type A ILRT by approximately 15 months would avoid hardship from duplication and additional resources necessary to perform two containment pressure tests within 24 months at a small change in risk to the public.

The licensee justifies the proposed change to further extend the current ILRT interval based on historical plant specific ILRT test results and containment inservice inspection (ISI) program results, supported by a risk-informed analysis. The staff evaluation is based on the current licensee license amendment request (LAR) submittals (Ref. 7.1 and 7.2) and responses to requests for additional information (RAIs), (Ref. 7.7 and 7.8), licensee Tendon Surveillance Reports (Ref. 7.5 and Ref. 7.6), the previous LAR for the initial five year deferral (Ref. 7.3), and safety evaluation for License Amendment No. 244. The staff evaluation is based on the understanding that the planned SG replacement outage (T1R18) during which the next ILRT is proposed to be conducted will begin no later than November 1, 2009, as indicated by the licensee.

3.2 Containment ISI Program and Structural Integrity Considerations The leakage rate testing requirements of 10 CFR 50, Appendix J, Option B and the ISI requirements mandated by 10 CFR 50.55a complement each other in ensuring the leak-tight and structural integrity of the containment during its service life. The TMI-1 containment is examined in accordance with the requirements of the ASME Code,Section XI, Subsections IWE and IWL, the Containment Leakage Testing program and the Containment Coatings program.

The licensee has stated that results from previous TMI-1 Type A ILRT testing, since TMI-1 went operational, demonstrate that the containment structure remains essentially a leak-tight barrier and represents minimal risk. The licensee presented the specific results from the eight previous Type A ILRTs (one pre-operational and seven post-operational) in AmerGen letter dated September 30, 2002 (Ref. 7.3) .

Furthermore, the licensee has emphasized that the TMI-1 ISI program based on 10 CFR 50.55a is implemented in accordance with the 1992 Edition with the 1992 Addenda of the ASME Code,Section XI, Subsections IWE and IWL. In accordance with 10 CFR 50.55a(b)(2)(ix)(E), a General Visual Examination of the containment liner as required by ASME Code,Section XI, Subsection IWE must be performed during each Section XI ISI period of the ten-year interval. TMI-1 has completed this exam for the first period of the current 10-year interval, which began in April 2001, and is scheduled to complete the second period examination in 2007. TMI-1 is also required to perform a VT-3 (ASME Code,Section XI, Examination Category E-A, Item No. E1.12) of the accessible areas of the containment liner.

This examination is required to be performed during the third period of this 10-year interval and is currently scheduled for 2011. An additional visual exam of the area adjacent to the moisture barrier (i.e., between liner and concrete floor) is performed during each refueling outage.

Containment inspections also include an examination of pressure retaining bolting in accordance with ASME Code,Section XI, Examination Category E-G, Item No. E8.10. This program requires a VT-1 exam of 100 percent of all pressure retaining bolting over the course of the 10-year interval.

The licensee has discussed information and results of the previous inspections (Ref. 7.1 and Ref. 7.3) of the containment liner and moisture barrier interface. During the T1R13 (1999) outage, 100 percent of the accessible portions of the containment liner and moisture barrier interface were examined in accordance with ASME Code,Section XI, Examination Category E-A, E1.11. Additionally, an augmented exam was conducted in accordance with the ASME Code,Section XI, IWE-1240, Examination Category E-C, E4.12, which requires a volumetric wall thickness examination at selected areas to be performed each period. This examination identified five areas where coating degradation had resulted in localized liner metal loss of 1/16 inch at the liner-to-concrete slab and moisture barrier (seal) interface (Elevation (EL) 281 ft). An engineering evaluation demonstrated that the remaining liner wall thickness was adequate to ensure leak-tightness. Coating repairs were performed at the degraded areas during the T1R13 (1999) refueling outage. In T1R14 (2001) refueling outage, a supplemental VT-1 examination was completed and the five areas were found to remain acceptable. As an additional action for the augmented examination requirements, six areas were marked for ultrasonic testing (UT) examination at EL 281 ft inside the Reactor Building. The liner thickness of these six areas were found to be consistent with little to no wall loss. These six areas are scheduled for UT examination again during the T1R17 outage in fall 2007.

There was also one area where a round indication 1/16 inch x 1/16 inch deep was observed at EL 374 ft in the liner during the T1R13 (1999) refueling outage. The licensee has stated that this indication was not the result of corrosion. An engineering evaluation demonstrated that the remaining liner wall thickness was adequate to serve its design basis function (leak-tightness).

A supplemental VT-1 exam of this area was performed in the T1R14 (2001) outage and the area was found to be acceptable.

Visual Examination of 360 degrees of the circumference of the moisture barrier region during the T1R15 (2003) refueling outage identified areas of chipped, scratched, and pealed paint and broken moisture barrier sealing. An area of corrosion was identified and UT examined and no areas less than 0.300 inch liner thickness were identified. These areas were corrected and reexamined to verify that they were corrected. A similar 360 degree visual examination of the moisture barrier was completed again during the T1R16 (2005) refueling outage. This examination identified six areas of the moisture barrier region where the caulking to the liner was no longer bonded and required repair. These areas were also corrected and subsequently examined.

Pressure retaining bolting examinations have been completed on 16 bolted penetration connections in accordance with ASME Code,Section XI, Examination Category E-G, item E8.10. All of the bolted penetration connections are currently scheduled for visual examination in 2007 in order to coincide with the scheduled IWE General Visual Examination required for the examination period.

The licensee has stated in Ref. 7.2 that TMI-1 will perform the following actions during the upcoming T1R17 (2007) refueling outage prior to the end of the current 15-year ILRT time frame: (i) General visual examination of the containment liner in accordance with 10 CFR 50.55a(b)(2)(ix)(E); (ii) Visual exam of the area adjacent to the moisture barrier (i.e.,

between the liner and concrete); and (iii) Replace a significant portion of the containment liner moisture barrier that has not been replaced in recent outages. The licensee has noted that the moisture barrier is not a pressure boundary and is installed to protect the metal containment liner from corrosion.

The staff reviewed the results of TMI-1's most recent two ISI program implementation (at 5 year intervals) on the exterior concrete containment structure and post-tensioning system documented in the 30th year and 25th year Reactor Building Tendon Surveillance Reports (Ref. 7.5 and 7.6) previously submitted by the licensee to the NRC on the docket. The most recent surveillances conducted in 2004 and 1999, respectively, were performed in accordance with the requirements of 10 CFR 50.55a and ASME Code,Section XI, Subsection IWL. The licensee has stated that these surveillances consisted of testing and visual examination of a randomly selected sample of post-tensioning tendons and visual examination of the accessible containment exterior. On the basis of these surveillance results, the licensee concluded that the reactor building post-tensioned concrete containment system is in sound structural condition and is capable of performing its intended function with ample margin of safety. None of the documented conditions that did not meet specified threshold acceptance criteria and were accepted following an engineering evaluation represented a degradation of containment integrity. On the basis of observed trends, the licensee concluded that the reactor building will remain in sound structural condition and retain an ample margin of safety well beyond the latest completion date of March 2010 specified for the next tendon surveillance of the post-tensioning system of the containment. The staff agrees that the reactor building is in a sound condition and the licensee has an adequate ISI program and procedures in place to examine, monitor and correct potential age-related and environmental degradations of the pressure retaining components of the TMI-1 primary containment throughout the extended ILRT interval.

In the LAR, TMI-1 proposes to perform its next periodic ILRT (to verify leak-tightness) in combination with the post-repair containment pressure test required per ASME Code,Section XI, IWL-5000 (to verify structural integrity of the restored containment following the major repair/replacement activity) during the T1R18 outage. In order to enable staff to gain an understanding of how the two tests would be performed in combination, the licensee was requested through an RAI to outline the main characteristics of the pressurization process and the extended surface examinations, additional examinations during pressurization, other examinations, and measurements of structural response to pressure of the affected areas/components of the post-repair containment structure, required by IWL-5250, that the licensee plans to conduct during the combined test. The licensee was also requested to provide information on how these examinations and the ILRT would be relatively scheduled/sequenced during the combined test. In their RAI response dated May 22, 2007 (Ref. 7.7), the licensee stated that two tests will be performed concurrently at Pa [peak calculated design basis accident pressure]= 50.6 pounds per square inch gauge (plus a small margin) following the completion of containment restoration work at the end of the SG replacement outage scheduled for late 2009. The licensee also summarized the bases, pressurization sequence and planned activities associated with both tests. The staff finds the approach and procedure indicated by the licensee for performing the two tests and associated examinations rational, systematic and consistent with industry standards and regulatory

guidance, and therefore, the RAI response is acceptable to staff.

In summary, to demonstrate acceptable performance of the containment, the licensee has cited the results of their previous Type A test and ASME Code,Section XI, Subsections IWE and IWL, ISI findings and actions taken. The licensee reports that visual examinations and augmented examination of the area adjacent to the moisture barrier were conducted in accordance with the ASME Code,Section XI. The containment liner areas which had experienced some degradation were identified, analyzed and repaired as necessary to ensure an acceptable containment barrier still exists. The results of the past ILRTs and the ISI programs demonstrate that the structural and leak-tight integrity of the containment structure is sound and adequately managed.

On the basis of the above discussion and the safety evaluation for License Amendment No. 244 (Ref. 7.4), the staff concludes that the containment structure is in a sound condition and the licensee has an adequate program and procedures in place to examine, monitor and take appropriate corrective action as necessary to ensure the leak-tight and structural integrity of the TMI-1 containment. Granting an additional one-time additional extension of approximately 15 months for performing the next ILRT will not adversely affect the leak-tight integrity of the primary containment and, therefore, is acceptable. This acceptance is based on the staffs understanding that the planned SG replacement outage (T1R18 refueling outage) during which the next ILRT is proposed to be conducted will begin no later than November 1, 2009. The licensee has agreed to incorporate this date into the proposed TS 6.8.5 amendment statement by modifying the language to read as follows:

a. Section 9.2.3: The first Type A test performed after the September 1993 Type A test shall be performed prior to startup from the T1R18 refueling outage. The T1R18 refueling outage will begin no later than November 1, 2009.

Based on the NRC staffs review of the licensees submittal of September 15, 2006, and supplemental information on February 26, 2007, response to the staffs request for additional information dated May 22, 2007, the 30th and 25th year Tendon Surveillance Reports, and safety evaluation for License Amendment No. 244 (which previously extended ILRT interval from 10 years to the current 15 years), the staff finds that the TMI-1 containment structure is in a sound condition and the licensee has an adequate ISI program and procedures in place to examine, monitor and correct potential age-related and environmental degradations of the pressure retaining components of the TMI-1 primary containment. Therefore, granting a one-time approximately 15-month extension to the current 15-year interval for performing the ILRT as proposed by the licensee in Section 6.8.5 of the proposed TS revision request is acceptable.

This acceptance is based on the staffs understanding that the planned SG replacement outage (T1R18 refueling outage) during which the next ILRT is proposed to be conducted will begin no later than November 1, 2009.

The NRC staff notes that a significant number of the tendons will be removed/de-tensioned in order to create the containment breach to facilitate SG replacement and then some new tendons installed and other affected existing ones re-tensioned after completion of SG replacement. In order to ensure that continued leak-tightness and structural integrity is maintained following this major repair/replacement activity, the NRC staff recommends augmenting the tendon surveillance program following SG replacement to include examination of tendons that are impacted by SG replacement. The use of the process established in

Subsection IWL of the 2001 Edition of the ASME Code,Section XI (2002 Addenda) is a method acceptable to the staff.

3.3 Risk Impact Assessment As stated previously, in TMI-1 License Amendment No. 244, the NRC approved a one-time extension of the containment ILRT interval from 10 to 15 years for TMI-1. This test interval extension was supported by a licensee risk assessment. The NRCs review of the licensees risk assessment was documented in the safety evaluation report (SER) for the license amendment, and concluded that the combined risk impact of the test interval extensions, in terms of total integrated plant risk, large early release frequency, and conditional containment failure probability, is small and supportive of the change.

By letter dated September 15, 2006, the licensee requested that TS 6.8.5 regarding the Reactor Building Leakage Rate Testing Program be amended to effectively allow a one-time extension of the ILRT interval from 15 years to approximately 15 years plus 15 months for TMI-1. The licensee performed a risk assessment of the impact of extending the ILRT test frequency from the original three tests in 10 years to one test in 15 years plus 15 months, and reported the risk results in the September 15, 2006, application for license amendment. The risk assessment for this evaluation is based on the same methodology, input, and assumptions used to support License Amendment 244, with the exception of the revised test interval and the use of an updated version of the plant-specific probabilistic risk assessment.

Based on the analyses provided by the licensee, the risk impacts and risk comparisons for the proposed change are essentially unchanged from those reported in the previous SER, and the staff conclusions remain valid. Specifically, the increase in the total integrated plant risk is small and supportive of the proposed change, the increase in the test interval results in only a small change in large early release frequency consistent with the acceptance guidelines of RG 1.174, and the defense-in-depth philosophy is maintained based on the small magnitude of the change in the conditional containment failure probability.

Based on these conclusions, the staff finds that the increase in predicted risk due to the proposed change is within the acceptance guidelines while maintaining the defense-in-depth philosophy of RG 1.174 and, therefore, is acceptable.

4.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Pennsylvania State official was notified of the 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 facility component located within the restricted area as defined in 10 CFR Part 20 or changes an inspection or surveillance requirement. 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. The Commission 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 75989). 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 is reasonable 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.

7.0 REFERENCES

7.1 Letter 5928-06-20531 from Pamela B. Cowan (AmerGen) to NRC,?Technical Specifications Change Request No. 334 - One-time Type A Test Interval Extension, September 15, 2006 (Agencywide Documents Access and Management System (ADAMS) Accession Number ML062680040).

7.2 Letter 5928-07-20047 from Pamela B. Cowan (AmerGen) to NRC, ?Additional Information Concerning Technical Specifications Change Request No. 334 - One-time Type A Test Interval Extension, February 26, 2007 (ADAMS Accession Number ML070580301).

7.3 Letter 5928-02-221096 from Michael P. Gallagher (AmerGen) to NRC, ?License Amendment Request No. 318 Integrated Leak Rate Test Deferral, September 30, 2002 (ADAMS Accession Number ML022830707).

7.4 Letter from NRC to John L. Skolds (AmerGen), ?Three Mile Island Nuclear Station, Unit 1 (TMI-1), RE: Deferral of Integrated Leak Rate Test (TAC. No. MB6487),

August 14, 2003 (ADAMS Accession Number ML032050212).

7.5 Three Mile Island Unit 1 30th Year Reactor Building Tendon Surveillance (Period 8),

Topical Report No. 183, Revision 00, July 2005 (ADAMS Accession Number ML050760506) 7.6 Three Mile Island Unit 1 25th Year Reactor Building Tendon Surveillance (Period 7),

Topical Report No. 136, Revision 01, July 2001 (ADAMS Accession Number ML012640020) 7.7 Letter 5928-07-20117 from Russel G. West (AmerGen) to NRC, ?Response to Request for Additional Information Concerning Technical Specifications Change Request No. 334 - One-time Type A Test Interval Extension, May 22, 2007 (ADAMS Accession Number ML071430219).

7.8. Letter 5928-07-20130 from Pamela B. Cowan (AmerGen) to NRC, Response to Request for Additional Information Concerning Technical Specifications Change Request No. 334 - One-Time Type A Test Interval Extension, June 5, 2007 (ADAMS Accession Number ML071560602).

Principal Contributors: George Thomas Robert Palla Hans Ashar Date: June 29, 2007