ML092310530

Issuance of Amendment Technical Specification Changes to Reflect Steam Generator Replacement

ML092310530
Person / Time
Site:	Three Mile Island
Issue date:	09/15/2009
From:	Peter Bamford Plant Licensing Branch 1
To:	Pardee C Exelon Generation Co
Bamford, Peter J., NRR/DORL 415-2833
References
TAC MD9923
Download: ML092310530 (18)

Text

UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 September 15, 2009 Mr. Charles G. Pardee President and Chief Nuclear Officer Exelon Generation Company 4300 Winfield Road Warrenville, IL 60555 THREE MILE ISLAND NUCLEAR STATION, UNIT 1 -ISSUANCE OF AMENDMENT RE: TECHNICAL SPECIFICATION CHANGES TO REFLECT STEAM GENERATOR REPLACEMENT (TAC NO. MD9923)

Dear Mr. Pardee:

The Commission has issued the enclosed Amendment No. 271 to Facility Operating License No. DPR-50 for the Three Mile Island Nuclear Station, Unit 1 (TMI-1), in response to your application dated October 9,2008 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML082890539), as supplemented by letter dated April 2, 2009 (ADAMS Accession No. ML090920761).

The amendment removes portions of the technical specifications that are not applicable to the new steam generators and reflects changes related to the new thermally treated Inconel Alloy 690 tubing design of the replacement steam generators. A copy of the related safety evaluation is also enclosed. Notice of Issuance will be included in the Commission's biweekly Federal Register notice. Sincerely, Peter J. Bamford, Project Manager Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-289

Enclosures:

1. Amendment No. 271 to DPR-50

2. Safety Evaluation cc: Distribution via Listserv UNITED NUCLEAR REGULATORY WASHINGTON, D.C. 20555-0001 EXELON GENERATION COMPANY, LLC DOCKET NO. 50-289 THREE MILE ISLAND NUCLEAR STATION, UNIT NO.1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.271 License No. DPR-50 The Nuclear Regulatory Commission (the Commission or NRC) has found that: The application for amendment by Exelon Generation Company, LLC (the licensee, formerly AmerGen Energy Company, LLC), dated October 9, 2008, supplemented by letter dated April 2, 2009, 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 10 CFR Chapter I; The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; 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; 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 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.

-2Accordingly, 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: Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment NO.271 , are hereby incorporated in the license. The Exelon Generation Company shall operate the facility in accordance with the Technical Specifications. This license amendment is effective with the installation of the replacement steam generators and shall be implemented prior to exceeding cold shutdown following the Three Mile Island, Unit 1 steam generator replacement refueling outage (T1R18), which is scheduled to begin in the fall of 2009.

FOR THE NUCLEAR REGULATORY COMMISSION Harold K. Chernoff, Chief Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the License and Technical Specifications Date of Issuance:

September 1 5, 2009 ATTACHMENT TO LICENSE AMENDMENT N0271 FACILITY OPERATING LICENSE NO. DPR-50 DOCKET NO. 50-289 Replace the following page of the Facility Operating License with the revised page. The revised page is identified by amendment number and contains marginal lines indicating the areas of change. Remove Page 3 Page 3 Replace the following pages of the Appendix A Technical Specifications with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Insert 3-12 6-19 6-20 6-27 6-28 6-29 3-12 6-19 6-20 6-27 6-28 6-29

-3 Exelon Generation Company, pursuant to the Act and 10 CFR Parts 30, 40 and 70 to receive, possess and use at any time any byproduct, source and special nuclear material as reactor fuel, sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required for reactor operation; Exelon Generation Company, pursuant to the Act and 10 CFR Parts 30, 40 and 70 to receive, possess at either TMI-l or TMI-2, and use in amounts as required for TMI-l any byproduct, source or special nuclear material without restriction to chemical or physical form, for sample analysis, testing, instrument calibration, or associated with radioactive apparatus or components.

Other than radioactive apparatus and components to be used at TMI Unit 2 in accordance with the TMI-2 License, the radioactive apparatus and components that may be moved from TMI Unit 1 to TMI Unit 2 under this provision shall be limited to: (1) outage-related items (such as contaminated scaffolding, tools, protective clothing, portable shielding and decontamination equipment); and (2) other equipment belonging to TMI Unit 1 when storage of such equipment at TMI-2 is deemed necessary for load handling or contamination control considerations; Exelon Generation Company, pursuant to the Act and 10 CFR Parts 30 and 70, to possess at the TMI Unit 1 or Unit 2 site, but not separate, such byproduct and special nuclear materials as may be produced by the operation of either unit. Radioactive waste may be moved from TMI Unit 2 to TMI Unit 1 under this provision for collection, processing (including decontamination), packaging, and temporary storage prior to disposal.

Radioactive waste that may be moved from TMI Unit 1 to TMI Unit 2 under this provision shall be limited to: (1) dry active waste (DAW) temporarily moved to TMI Unit 2 during waste collection activities, and (2) contaminated liquid contained in shared system piping and tanks.

Radioactive waste that may be moved from TMI Unit 1 to TMI Unit 2 under this provision shall not include spent fuel, spent resins, filter sludge, evaporator bottoms, contaminated oil, or contaminated liquid filters. The storage of radioactive materials or radwaste generated at TMI Unit 2 and stored at TMI Unit 1 shall not result in a source term that, if released, would exceed that previously analyzed in the UFSAR in terms of offsite dose consequences.

The storage of radioactive materials or radwaste generated at TMI Unit 1 and stored at TMI Unit 2 shall not result in a source term that, if released, would exceed that previously analyzed in the PDMS SAR for TMI Unit 2 in terms of off-site dose consequences. This license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations 10 CFR Chapter I: Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Section 50.54 and 50.59 of Part 50, and Section 70.32 of Part 70; is subject to all applicable provisions of the Act and to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below: Maximum Power Level Exelon Generation Company is authorized to operate the facility at steady state reactor core power levels not in excess of 2568 megawatts thermal. Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.271 are hereby incorporated in the license. The Exelon Generation Company shall operate the facility in accordance with the Technical Specifications.

Amendment No.271 3.1.6 LEAKAGE Applicability 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 If the total reactor coolant leakage rate exceeds 10 gpm, the reactor shall be placed in hot shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of detection. If unidentified reactor coolant leakage (excluding normal evaporative losses) exceeds one gpm or if any reactor coolant leakage is evaluated as unsafe, the reactor shall be placed in hot shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of detection. If the primary-to-secondary leakage through anyone (1) steam generator exceeds 150 GPO, the reactor shall be placed in hot shutdown within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and in cold shutdown within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. 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 shutdown condition shall be initiated within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of detection. 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. 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. 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. When the reactor is critical and above 2 percent power, two reactor coolant leak detection systems of different operating principles shall be in operation for the Reactor Building with one of the two systems sensitive to radioactivity. The systems sensitive to radioactivity may be out-of-service for no more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> provided a sample is taken of the Reactor Building atmosphere every eight hours and analyzed for radioactivity and two other means are available to detect leakage. 3-12 Amendment No. 47, 180, 246, 261, 271 (12-22-78) 6.9.6 6.9.5 6.9.5.1 6.9.5.2 6.9.5.3 6.9.5.4 CORE OPERATING liMITS REPORT The core operating limits addressed by the individual Technical Specifications be established and documented in the CORE OPERATING LIMITS REPORT to each reload cycle or prior to any remaining part of a reload The analytical methods used to determine the core operating limits addressed by the individual Technical Specifications shall be those previously reviewed and approved by the NRC for use at TMI-1, specifically: BAW-10179 P-A, "Safety and Methodology for Acceptable Cycle Analyses." The current revision level shall be specified in the TR-078-A, "TMI-1 Transient Analyses Using the RETRAN Code", Revision O. NRC SER dated TR-087-A, "TMI-1 Core Thermal-Hydraulic Methodology Using VIPRE-01 Computer Code", Revision O. NRC SER dated TR-091-A, "Steady State Reactor Physics Methodology for Revision O. NRC SER dated TR-092P-A, "TMI-1 Reload Design and Setpoint Revision O. NRC SER dated BAW-10227P-A, "Evaluation of Advanced Cladding and Material (M5) in PWR Reactor Fuel", NRC SER dated February The core operating limits shall be determined so that all applicable limits (e.g., thermal-mechanical limits, core thermal-hydraulic limits, ECCS limits, limits such as shutdown margin, and transient/accident analysis limits) of the analysis are The CORE OPERATING LIMITS REPORT, including any mid-cycle revisions supplements thereto, shall be provided upon issuance for each reload cycle to NRC Document Control Desk with copies to the Regional Administrator Resident STEAM GENERATOR TUBE INSPECTION REPORT A report shall be submitted within 180 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, 6-19 Amendment No. 72,77,129,137,141,149,160,168,173,178,202,233,261,271

d. Location, orientation (if linear), and measured sizes (if available) of service induced indications, e. Number of tubes plugged during the inspection outage for each active degradation mechanism, f. Total number and percentage of tubes plugged to date, g. The results of condition monitoring, including the results of tube pulls and in-situ testing, h. The effective plugging percentage for all plugging in each SG, 6.10 RECORD RETENTION 6.10.1 The following records shall be retained for at least five years: Records of normal station operation including power levels and periods of operation at each power level. Records of principal maintenance activities, including inspection, repairs, substitution, or replacement of principal items of equipment related to nuclear safety. All REPORTABLE EVENTS. Records of periodic checks, tests and calibrations. Records of reactor physics tests and other special tests related to nuclear safety. Changes to procedures required by Specification 6.8.1. Deleted Test results, in units of microcuries, for leak tests performed on licensed sealed sources. Results of annual physical inventory verifying accountability of licensed sources on record. Control Room Log Book. Control Room Supervisor Log Book.

6-20 Amendment No.72, 77,129,137,141,150,173,180,219,261,271 Accident induced leakage performance criterion: The primary to secondary accident induced leakage rate for any design basis accident, other than a SG tube rupture, shall not exceed the leakage rate assumed in the accident analysis in terms of total leakage rate for all SGs and leakage rate for an individual SG. Leakage is not to exceed 1 gpm per SG. The operational leakage performance criterion is specified in TS 3.1.6, "LEAKAGE." Provisions for SG tube repair criteria. Tubes found by inservice inspection to contain lIaws with a depth equal to or exceeding 40% of the nominal tube wall thickness shall be plugged.

6-27 Amendment No.

2&+, 271 Provisions for SG tube inspections. Periodic SG tube inspections shall be performed.

The number and portions of the tubes inspected and methods of inspection shall be performed with the objective of detecting flaws of any type (e.g., volumetric flaws, axial and circumferential cracks) that may be present along the length of the tube, from the tube-to-tubesheet weld at the tube inlet to the tube-to-tubesheet weld at the tube outlet, and that may satisfy the applicable tube repair criteria. The tube-to-tubesheet weld is not part of the tube. In addition to meeting the requirements of d.1, d.2, and d.3 below, the inspection scope, inspection methods, and inspection intervals shall be such as to ensure that SG tube integrity is maintained until the next SG inspection. An assessment of degradation shall be performed to determine the type and location of flaws to which the tubes may be susceptible and, based on this assessment, to determine which inspection methods need to be employed and at what locations. Inspect 100% of the tubes in each SG during the first refueling outage following SG replacement. Inspect 100% of the tubes at sequential periods of 144, 108, 72, and, thereafter, 60 effective full power months. The first sequential period shall be considered to begin after the first inservice inspection of the SGs. In addition, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. No SG shall operate for more than 72 effective full power months or three refueling outages (whichever is less) without being inspected. If crack indications are found in any SG tube, then the next inspection for each SG for the degradation mechanism that caused the crack indication shall not exceed 24 effective full power months or one refueling outage (whichever is less). If definitive information, such as from examination of a pulled tube, diagnostic non-destructive testing, or engineering evaluation indicates that a crack-like indication is not associated with a crack(s), then the indication need not be treated as a crack.

e. Provisions for monitoring operational primary to secondary leakage.

6-28 Amendment No. 2&+, 271 Refer to Section 6.9.6 for reporting requirements for periodic SG tube inspections. Control Room Envelope Habitability Program A Control Room Envelope (CRE)

Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Emergency Control Room Air Treatment System, CRE occupants can control the reactor safely under normal conditions and maintain it in a safe condition following a radiological event, hazardous chemical release, or a smoke challenge. The program shall ensure that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem total effective dose equivalent (TEDE) for the duration of the accident. The program shall include the following elements:

a. The definition of the CRE and the CRE boundary.

b. Requirements for maintaining the CRE boundary in its design condition including configuration control and preventive maintenance.

c. Requirements for (i) determining the unfiltered air inleakage past the CRE boundary into the CRE in accordance with the testing methods and at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, "Demonstrating Control Room Envelope Integrity at Nuclear Power Reactors," Revision 0, May 2003, and (ii) assessing CRE habitability at the Frequencies specified in Sections C.1 and C.2 of Regulatory Guide 1.197, Revision O. Measurement, at designated locations, of the CRE pressure relative to all external areas adjacent to the CRE boundary during the pressurization mode of operation by one train of the Control Room Ventilation System, operating at the design flow rate, at a Frequency of 24 months. The results shall be trended and used as part of the 24 month assessment of the CRE boundary. The quantitative limits on unfiltered air inleakage into the CRE.

These limits shall be stated in a manner to allow direct comparison to the unfiltered air inleakage measured by the testing described in paragraph c. The unfiltered air inleakage limit for radiological challenges is the inleakage flow rate assumed in the licensing basis analyses of DBA consequences. Unfiltered air inleakage limits for hazardous chemicals must ensure that exposure of CRE occupants to these hazards will be within the assumptions in the licensing basis. The provisions of Section 1.25 are applicable to the frequencies for assessing CRE habitability, determining CRE unfiltered inleakage, and measuring CRE pressure and assessing the CRE boundary as required by paragraphs c and d, respectively.

6-29 Amendment No. 261, 264,271 REGUl UNITED STATES

"'r 0-s> NUCLEAR REGULATORY COMMISSION

!:J ... ,L C'l WASHINGTON, D.C. 20555-0001

<< 0 ... III ****-l< SAFETY EVALUATION BYTHE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 27110 FACILITY OPERATING LICENSE NO. DPR-50 EXELON GENERATION COMPANY, LLC THREE MILE ISLAND NUCLEAR STATION, UNIT 1 DOCKET NO. 50-289

1.0 INTRODUCTION

By application dated October 9, 2008 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML082890539), as supplemented by letter dated April 2, 2009 (ADAMS Accession No. ML090920761), Exelon Generation Company (Exelon, or the ticensee)'

requested changes to the Technical Specifications (TSs) for Three Mile Island Nuclear Station, Unit 1 (TMI-1). The supplement provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC or Commission) staff's original proposed no significant hazards consideration determination as published in the Federal Register on March 10,2009 (74 FR 10310).

The proposed amendment requests revisions to TSs 3.1.6, "Leakage," 6.9.6, "Steam Generator Tube Inspection Report," and 6.19, "Steam Generator (SG) Program," and is in support of the planned replacement of the once through SGs at TMI-1 during the fall 2009 outage.

Specifically, the proposed amendment requests elimination of the existing requirements associated with tube sleeve repairs and alternate repair criteria which are not applicable to the replacement SGs. In addition, the proposed changes incorporate revised tube integrity surveillance frequency requirements applicable to Inconel Alloy 690 tubing, incorporate a revised primary-to-secondary leakage criteria, and change the required reporting period for SG inspection results.

2.0 REGULATORY EVALUATION

SG tubes function as an integral part of the reactor coolant pressure boundary (RCPB) and serve to isolate radiological fission products in the primary coolant from the secondary coolant and the environment. For the purposes of this safety evaluation, tube integrity means that the tubes are capable of performing these functions in accordance with the plant design and licensing basis. 1 The application dated October 9, 2008, was submitted by AmerGen Energy Company, LLC. Effective January 8, 2009, the license for TMI-1 was transferred from AmerGen Energy Company, LLC to Exelon Generation Company, LLC. By letter dated January 9, 2009, (ADAMS Accession No. ML090120538)

Exelon Generation Company adopted and endorsed docketed submittals that requested specific licensing actions that were made by AmerGen, and requested that the NRC staff continue to process those pending actions on the schedules previously agreed to by AmerGen.

The construction permit for TMI-1 was issued by the Atomic Energy Commission (AEC) on May 18,1968, and an operating license was issued on April 19, 1974. The plant design approval for the construction phase was based on the proposed General Design Criteria (GDC) published by the AEC in the Federal Register (32 FR 10213) on July 11, 1967 (hereinafter referred to as "draft GDC"). The AEC published the final rule that added Appendix A to Title 10 of the Code of Federal Regulations (10 CFR) Part 50, "General Design Criteria for Nuclear Power Plants," in the Federal Register (36 FR 3255) on February 20, 1971 (hereinafter referred to as "final GDC" or just "GDC"). Differences between the draft GDC and final GDC included a consolidation from 70 to 64 criteria. In accordance with an NRC staff requirements memorandum from S. J. Chilk to J. M. Taylor, "SECY-92-223

-Resolution of Deviations Identified During the Systematic Evaluation Program," dated September 18, 1992 (ADAMS Accession No. ML003763736), the Commission decided not to apply the final GDC to plants with construction permits issued prior to May 21, 1971, which includes TMI-1. The TMI-1 Updated Final Safety Analysis Report (UFSAR), Section 1.4 provides an evaluation of the design bases of TMI-1 against the draft GDC. The final GDC, criterion 14 (GDC-14) states that the RCPB shall have "an extremely low probability of abnormal leakage ... and gross rupture." Draft GDC, criterion 9, contains similar requirements.

Furthermore, other final GDC references state that the RCPB "shall be designed with sufficient margin," (GDC 15 and GDC 31); shall be of "the highest quality standards possible," (GDC-30); and shall be designed to permit "periodic inspection and testing ... to assess ... structural and leak tight integrity" (GDC-32). Draft GDC, criterion 33; draft GDC, criterion 1; and draft GDC, criterion 36; respectively, have comparable provisions. To this end, 10 CFR 50.55a, "Codes and Standards," specifies that components which are part of the RCPB must meet the requirements of Class 1 components in Section III of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code). Section 50.55a further requires that, throughout the service life of a pressurized water reactor (PWR) facility, ASME Code Class 1 components meet the requirements in Section XI of the ASME Code, "Rules for Inservice Inspection of Nuclear Power Plant Components," to the extent practical. This requirement includes the inspection and repair criteria of Section XI of the ASME Code.Section XI requirements pertaining to inservice inspection of SG tubing are augmented by additional SG tube surveillance requirements in the TSs. As part of the plant licensing basis, applicants for PWR licenses are required to analyze the consequences of postulated design-basis accidents, such as a SG tube rupture and main steam line break.

These analyses consider the primary-to-secondary leakage through the tubing that may occur during these events.

Furthermore, the analyses must show that the offsite radiological consequences do not exceed the applicable limits of the 10 CFR Part 100, "Reactor Site Criteria," guidelines for offsite doses (or 10 CFR 50.67, "Accident Source Term," as appropriate), GDC-19, "Control Room," criteria for control room operator doses, or some fraction thereof as appropriate to the accident, or the NRC-approved licensing basis. The current and proposed TMI-1 TSs are modeled after Technical Specification Task Force Traveler (TSTF)-449, Revision 4, "Steam Generator Tube Integrity," dated April 2005. TS 6.19 for TMI-1 requires that a SG program be established and implemented to ensure that SG tube integrity is maintained.

Tube integrity is maintained by meeting specified performance criteria for structural and leakage integrity consistent with the plant design and licensing bases, TS 6.19 requires a condition monitoring assessment be performed during each outage during which the SG tubes are inspected or plugged to confirm that the performance criteria are being met.

TS 6.19 also includes provisions regarding the scope, frequency, and methods of SG tube inspections. In 10 CFR 50.36, the Commission established its regulatory requirements related to the content of TSs. In doing so, the Commission emphasized those matters related to the preventing of accidents and mitigating their consequences. As recorded in the Statements of Consideration, Technical Specifications for Facility Licenses: Safety Analysis Reports (33 FR 18610, December 17,1968).

the Commission noted that applicants are expected to incorporate into their TSs those items that are directly related to maintaining the integrity of the physical barriers designed to contain radioactivity.

Pursuant to 10 CFR 50.36, TSs are required to include items in five specific categories related to station operation. Specifically, those categories include: (1) safety limits, limiting safety system settings, and limiting control settings; (2) limiting conditions for operation (LCO); (3) surveillance requirements (SRs); (4) design features; and (5) administrative controls. However, the rule does not specify the particular requirements to be included in a plant's TS. The licensee's application contains changes that involve specifications relating to steam generator integrity, an important element of the physical barriers designed to contain radioactivity.

Additionally, 10 CFR 50.36(c)(2)(ii) sets forth four criteria to be used in determining whether an LCO is required to be included in the TS for a certain item.

These criteria are as follows: Installed instrumentation that is used to detect, and indicate in the control room, a significant abnormal degradation of the reactor coolant pressure boundary. A process variable, design feature, or operating restriction that is an initial condition of a design-basis accident or transient analysis that assumes either the failure of or presents a challenge to the integrity of a fission product barrier. A structure, system, or component that is part of the primary success path and which functions or actuates to mitigate a design-basis accident or transient that either assumes the failure of or presents a challenge to the integrity of a fission product barrier. A structure, system or component which operating experience or probabilistic risk assessment has shown to be significant to public health and safety. The NRC staff has reviewed the proposed changes to ensure that these changes conform with 10 CFR 50.36 as discussed herein. 3.0 TECHNICAL EVALUATION TMI-1 has two Babcock and Wilcox once-through SGs. Each SG contains 15,531 stress relieved, mill annealed, Alloy 600 tubes. Each tube has a nominal outside diameter of 0.625 inches and a minimum wall thickness of 0.034 inches. The tubes were originally mechanically roll expanded in both the hot-leg and cold-leg tubesheets for approximately 1 inch of the 24 inch tubesheet thickness. In the 1980s, the tubes were kinetically expanded in the upper tubesheet to make the length of engagement between the tube and the tubesheet either 17 inches or 22 inches. The tubes are supported along their length by a number of carbon steel support plates. One of the primary differences between the replacement SGs and the existing SGs is that the tube material is thermally treated (TT) Alloy 690 in the replacement SGs versus the annealed Alloy 600 in the existing SGs. The Alloy 690 TT material has been shown to be much more resistant to stress corrosion cracking than the original Alloy 600 tubing. The licensee proposes to remove the TS requirements associated with alternate tube repair criteria applicable to their original SGs. These requirements are contained in TSs 6.9.6 (reporting requirements), 6.19.b (performance criteria), 6.19.c (tube repair criteria), and 6.19.d (tube inspection criteria). In addition, the licensee is proposing to remove the TS requirements associated with tube repair methods. These requirements are contained in TSs 6.9.6 (reporting requirements), 6.19.d (tube inspection criteria), and 6.19.f (tube repair methods). Repair methods and alternate tube repair criteria analyses were developed for the licensee's current SGs and not for the replacement SGs. As a result, the analyses used to justify these alternate tube criteria and the repair methods were only approved for the current SGs. That is, no alternate tube repair criteria or repair methods are applicable or approved for the plant's replacement SGs. As a result, the NRC staff finds these changes acceptable. The licensee is also proposing to modify their inspection requirements contained in TS 6.19.d.2 to adopt those requirements in TSTF-449 applicable to SGs with thermally treated Alloy 690 tubes. The modifications involve revised inspection intervals. With respect to frequency of inspection, the current specification requires that SG inspections be performed at least every 24 effective full-power months (EFPM). The proposed specification for Alloy 690 TT tubing would require that 100 percent of the tubes be inspected at sequential periods of 144, 108, 72, and, thereafter, 60 EFPM, with the first sequential period being considered to begin at the time of the first in-service inspection of the SGs following SG replacement (TMI-1 TS 6.19.d.1 already requires a 100 percent tube inspection during the first refueling outage following SG replacement). This sliding scale is intended to address the increased potential for the initiation of stress corrosion cracking over time by increasing the frequency of required inspections as the tubes age. In addition, the licensee would be required to inspect 50 percent of the tubes by the refueling outage nearest the mid-point of the period and the remaining 50 percent by the refueling outage nearest the end of the period. However, no SG shall operate for more than 72 EFPM or three refueling outages (whichever is less) without being inspected. TMI-1 TS 6.19.d.3 already requires that regardless of the type of tubing, if crack indications are found in any tube, that the next inspection for each SG for the degradation mechanism causing the crack indication shall not exceed 24 EFPM or one refueling outage (whichever is less). The proposed prescriptive requirements contained in 6.19.d.2 are intended primarily to supplement the existing TS 6.19 performance-based requirement that inspection frequency, in conjunction with inspection scope and methods, be such as to ensure tube integrity is maintained. This performance-based requirement must be satisfied in addition to the prescriptive requirements.

The acceptability of the new prescriptive requirements is based on the historical performance of Alloy 690 TT material in SG tube applications. The Alloy 690 TT material has demonstrated increased resistance to cracking as compared to the original Alloy 600 material which was the basis for the existing prescriptive requirements in the TMI-1 TS. The NRC staff concludes that the proposed prescriptive requirements represent an adequate inspection interval for Alloy 690 TT material, represent an effective strategy for ensuring tube integrity, and will serve to ensure that tube integrity is maintained between SG inspections. The licensee also proposes to modify the TS requirements associated with reactor coolant system operational leakage. These requirements are contained in TS 3.1.6.3

secondary leakage). The licensee is proposing to change the primary-to-secondary leak rate limit from 144 gallons per day (gpd) for the sum of leakage from both SGs to a new leak rate limit of 150 gpd for each SG. The lower limit, 144 gpd for the sum of leakage from both SGs, was implemented into the TMI-1 operating license by license amendment 103, dated December 21, 1984 (ADAMS Accession No. 8501110093).

The requirement was transferred from the operating license to the TSs by license amendment 261, dated September 27,2007 (ADAMS Accession No. ML072600318). The 144 gpd limit was imposed as a result of SG tube degradation and kinetic expansion repairs performed in the 1980's. The replacement SGs have not been subject to the mechanism that caused the original SGs to degrade and require the kinetic expansion repair. Hence, the basis for the reduced 144 gpd limit is no longer applicable.

With the installation of the replacement SGs, the licensee is proposing to use the industry standard 150 gpd leak rate limit for each SG, consistent with TSTF-449.

Although no leakage limit, even if reduced to zero, can be totally effective in preventing SG tube ruptures, operating experience demonstrates that leakage limits are an important element of an overall approach to limiting the occurrence of tube rupture and for ensuring SG tube integrity. In addition, the proposed limit is significantly less than the conditions assumed in the safety analyses.

Therefore, the NRC staff finds the revised specification to be in accordance with 10 CFR 50.36(c)(2)(ii) and, thus, acceptable.

The licensee also proposes to submit SG tube inspection reports 180 days after reaching 200 degrees Fahrenheit average reactor coolant temperature following an inspection performed in accordance with TS 6.19, consistent with the timeframe listed in TSTF-449.

The 180-day reporting requirement is adequate given that the failure of the SG program to maintain tube integrity as indicated by condition monitoring would be promptly reportable in accordance with 10 CFR 50.72, as clarified by NUREG-1022, Revision 2, "Event Reporting Guidelines 10 CFR 50.72 and 50.73," dated October 31,2000, and errata dated September 28,2004 (ADAMS Accession No. ML073050400), allowing the NRC staff to engage in any follow-up activities that it determines to be necessary.

Therefore, the NRC staff finds this change acceptable.

Regarding the 40 percent through-wall tube repair criterion in TS 6.19.c, the licensee indicated that it remains acceptable for the replacement SGs. This determination was based on the methodology contained in Nuclear Energy Initiative 97-06, "Steam Generator Program Guidelines," and the associated Electric Power Research Institute Guidelines.

This methodology is intended to be consistent with Regulatory Guide (RG) 1.121, "Bases for Plugging Degraded PWR Steam Generator Tubes." The NRC staff notes that the 40 percent through-wall repair criterion is consistent with the repair criterion at other plants with similar tube size and material. Since the repair criteria has been established consistent with RG 1.121, and TMI-1 is required to maintain tube integrity per TS 4.19 and TS 6.19 regardless of the TS tube repair criteria, the staff concludes that the 40 percent tube repair criterion remains acceptable. In summary, the NRC staff finds that the proposed changes to the SG TS requirements are acceptable because they maintain sufficient regulatory constraints on the establishment and implementation of the SG Program such as to provide reasonable assurance that tube integrity will be maintained.

The !\IRC staff concurs with the licensee that the requested changes properly reflect the Alloy 690 TT tube material in the replacement SGs. Based on the above, the staff concludes that the proposed changes to the SG tube integrity program TS requirements, including the associated reporting requirements, are 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. 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 (74 FR 10310). 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.

Principal Contributors:

A. Obodoako P. Bamford Date: September 15, 2009 September 15, 2009 Mr. Charles G. Pardee President and Chief Nuclear Officer Exelon Generation Company 4300 Winfield Road Warrenville, IL 60555 THREE MILE ISLAND NUCLEAR STATION, UNIT 1 -ISSUANCE OF AMENDMENT RE: TECHNICAL SPECIFICATION CHANGES TO REFLECT STEAM GENERATOR REPLACEMENT (TAC NO. MD9923)

Dear Mr. Pardee:

The Commission has issued the enclosed Amendment No. 271 to Facility Operating License No. DPR-50 for the Three Mile Island Nuclear Station, Unit 1 (TMI-1), in response to your application dated October 9, 2008 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML082890539), as supplemented by letter dated April 2, 2009 (ADAMS Accession No. IVIL090920761).

The amendment removes portions of the technical specifications that are not applicable to the new steam generators and reflects changes related to the new thermally treated Inconel Alloy 690 tubing design of the replacement steam generators. A copy of the related safety evaluation is also enclosed.

Notice of Issuance will be included in the Commission's biweekly Federal Register notice. Sincerely, Ira! Peter J. Bamford, Project Manager Plant Licensing Branch 1-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-289

Enclosures:

1. Amendment No. 271 to DPR-50 2. Safety Evaluation cc: Distribution via Listserv DISTRIBUTION PUBLIC LPLI-2 RtF AObodoako, NRR RidsAcrsAcnw_MailCTR Resource RidsNrrDirsltsb Resource RGrover, NRR RidsNrrDorlDpr Resource RidsNrrDorlLpl1-2 Resource KKarwoski, NRR RidsNrrCsgb Resource RidsNrrPMPBamford Resource RidsOgcRp Resource RidsNrrLAABaxter Resource RidsRgn1MailCenterResource ADAMS Accession No . ML092310530

.. OFFICE LPL1-2/PM LPLI-2/LA CSGB/BC ITSB/BC OGC LPL 1-2/BC NAME PBamford ABaxter MKavrilas (w/comment)

RElliolt LSubin (NLO w/comments)

HChernoff (REnnis for) DATE 8/24/09 9/1109 9/3/09 9/4/09 9/14/09 9/15/09 ..Official Record Copy