ML18351A006

License Amendment Request (LAR) for One-Cycle Extension of Appendix J Type a Integrated Leakage Rate Test

ML18351A006
Person / Time
Site:	Crane
Issue date:	12/14/2018
From:	Jim Barstow Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TMl-18-077
Download: ML18351A006 (77)

Text

Exelon Generation TMl-18-077 December 14, 2018 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555 Three Mile Island Nuclear Station, Unit 1 Renewed Facility Operating License No. DPR-50 NRC Docket No. 50-289 200 Exelon Way Kennett Square, PA 19348 www.exeloncorp.com 10 CFR 50.90

Subject:

License Amendment Request (LAR) for One-Cycle Extension of Appendix J Type A Integrated Leakage Rate Test In accordance with 10 CFR 50.90, "Application for amendment of license or construction permit," Exelon Generation Company, LLC (Exelon) requests the following amendment to the Technical Specifications, Appendix A, of Renewed Facility Operating License No. DPR-50 for Three Mile Island Nuclear Station, Unit 1 (TMI) to allow for a one-cycle extension to the 10-year frequency of the TMl-1 containment leakage rate test (i.e., Integrated Leakage Rate Test (ILRT) or Type A test). This test is required by Technical Specification (TS) 6.8.5, "Reactor Building Leakage Rate Testing Program." The proposed change would permit the existing I LAT to be extended from 1 O years to 11. 75 years.

Currently, the ILRT is to be performed approximately 3 months prior to the 101h year anniversary of the completion of the last I LRT (January 16, 2010). If granted, this revision would extend the period from 1 O years to 11. 75 years between successive tests. In terms of refueling outages (RFOs), this extension would move the performance of the next ILRT from the scheduled Fall 2019 end of cycle T1 R23 RFO to the Fall 2021 End of Cycle T1 R24 RFO. provides the Evaluation of Proposed Change. Attachment 2 provides the Proposed Technical Specification Marked-Up Page.

The proposed change has been reviewed by the TMI Plant Operations Review Committee in accordance with the requirements of the Exelon Quality Assurance Program.

There are no regulatory commitments contained in this submittal.

U.S. Nuclear Regulatory Commission License Amendment Request -

One-Cycle Extension of Appendix J Type A Integrated Leakage Rate Test December 14, 2018 Page 2 Exelon requests approval of the proposed amendment by September 1, 2019, to support the extension of the Unit 1 ILRT. Once approved, the amendment shall be implemented within 30 days.

Using the standards in 10 CFR 50.92, "Issuance of amendment," Exelon has concluded that the proposed change does not constitute a significant hazards consideration as described in the enclosed analysis performed in accordance with 1 O CFR 50.91 (a)(1 ).

In accordance with 1 O CFR 50.91, "Notice for public comment; State consultation," Exelon is notifying the Commonwealth of Pennsylvania of this application for a change to the TS by transmitting a copy of this letter and its attachments to the designated state official.

Should you have any questions concerning this submittal, please contact Mr. Frank J. Mascitelli at (610) 765-5512.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 14th day of December 2018.

Respectfully,

~QN'\\b ~

James Barstow Director - Licensing & Regulatory Affairs Exelon Generation Company, LLC Attachments: 1) Evaluation of Proposed Change

2) Proposed Technical Specification Marked-Up Page cc:

USNRC Regional Administrator, Region I USNRC Project Manager, TMl-1 USNRC Senior Resident Inspector TMl-1 Director, Bureau of Radiation Protection - PA Department of Environmental Resources

ATTACHMENT 1 License Amendment Request Three Mile Island Nuclear Power Station - Unit 1 Docket No. 50-289 EVALUATION OF PROPOSED CHANGE

Subject:

License Amendment Request (LAR) for One-Cycle Extension of Appendix J Type A Integrated Leakage Rate Test 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedence 4.3 No Significant Hazards Consideration 4.4 Conclusion

6.0 REFERENCES

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 1 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 1.0

SUMMARY

DESCRIPTION Exelon Generation Company (Exelon) requests an amendment to Renewed Facility Operating License No. DPR-50 for the Three Mile Island Nuclear Station, Unit 1 (TMI-1) to allow for a one-cycle extension to the 10-year frequency of the TMI-1 containment leakage rate test (i.e.,

Integrated Leakage Rate Test (ILRT) or Type A test). This test is required by Technical Specification (TS) 6.8.5, Reactor Building Leakage Rate Testing Program. The proposed change would permit the existing ILRT to be extended from 10 years to 11.75 years.

Currently, the ILRT is to be performed approximately 3 months prior to the 10th year anniversary of the completion of the last ILRT (January 16, 2010). If granted, this revision would extend the period from 10 years to 11.75 years between successive tests. In terms of refueling outages (RFOs), this extension would move the performance of the next ILRT from the scheduled Fall 2019 end of cycle T1R23 RFO to the Fall 2021 End of Cycle T1R24 RFO.

2.0 DETAILED DESCRIPTION The TMI-1 TS 6.8.5, "Reactor Building Leakage Rate Testing Program," currently states, in part:

"The Reactor Building Leakage Rate Testing Program shall be established, implemented, and maintained as follows:

A program shall be established to implement the leakage rate testing of the Reactor Building as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, Performance-Based Containment Leak-Test Program, dated September 1995, as modified by the following exception to NEI 94-01, Rev. 0, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J:

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.

The proposed change to TMI-1 TS 6.8.5 will be an administrative change to add the performance of the next Type A test no later than End of Cycle T1R24 RFO. The proposed change, shown in bold text, will revise TS 6.8.5, as follows, to state, in part:

"The Reactor Building Leakage Rate Testing Program shall be established, implemented, and maintained as follows:

A program shall be established to implement the leakage rate testing of the Reactor Building as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, Performance-Based Containment Leak-Test Program, dated September 1995, as modified by the following exception to NEI 94-

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 2 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 01, Rev. 0, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J:

a.

Section 9.2.3: The first Type A test performed after the January 2010 Type A test shall be performed no later than prior to startup from the T1R24 refueling outage in 2021.

A markup of TS 6.8.5 is provided in Attachment 2.

3.0 TECHNICAL EVALUATION

3.1 Justification for the Technical Specification Change 3.1.1 Current TMI-1 Reactor Building Leakage Rate Testing Program Requirements 10 CFR 50, Appendix J, was revised, effective October 26, 1995, to allow licensees to choose containment leakage testing under either Option A, "Prescriptive Requirements," or Option B, "Performance-Based Requirements." On May 27, 1997, the NRC approved TS Amendment 201 for TMI (Reference 1) authorizing the implementation of 10 CFR 50, Appendix J, Option B, for Types A, B and C tests.

Current TS 6.8.5 requires that a program be established to implement the leakage rate testing of the Reactor Building as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. The program is required to be in accordance with the guidelines contained in Regulatory Guide (RG) 1.163 (Reference 2). RG 1.163 endorses, with certain exceptions, NEI 94-01, Revision 0, Industry Guideline for Implementing Performance-Based Option to 10 CFR 50, Appendix J (Reference 3), as an acceptable method for complying with the provisions of Appendix J, Option B.

RG 1.163, Section C.1 states that licensees intending to comply with 10 CFR 50, Appendix J, Option B, should establish test intervals based upon the criteria in Section 11.0 of NEI 94-01 rather than using test intervals specified in ANSI/ANS 56.8-1994 (Reference 4). NEI 94-01, Section 11.0 refers to Section 9, which states that Type A testing shall be performed during a period of reactor shutdown at a frequency of at least once per ten years based on acceptable performance history. Acceptable performance history is defined as completion of two consecutive periodic Type A tests where the calculated performance leakage was less than 1.0La (where La is the maximum allowable leakage rate at design pressure). Elapsed time between the first and last tests in a series of consecutive satisfactory tests used to determine performance shall be at least 24 months.

Adoption of the Option B performance-based containment leakage rate testing program altered the frequency of measuring primary containment leakage in Types A, B, and C tests but did not alter the basic method by which Appendix J leakage testing is performed. The test frequency is based on an evaluation of the "as found" leakage history to determine a frequency for leakage testing, which provides assurance that leakage limits will not be exceeded. The allowed

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 3 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 frequency for Type A testing as documented in NEI 94-01 is based, in part, upon a generic evaluation documented in NUREG-1493 (Reference 5). The evaluation documented in NUREG-1493 includes a study of the dependence or reactor accident risks on containment leak tightness for differing types of containment types, including a dry, ambient containment similar to the TMI-1 containment structure. NUREG-1493 concludes in Section 10.1.2 that reducing the frequency of Type A tests (ILRT) from the original three (3) tests per 10 years to one (1) test per 20 years was found to lead to an imperceptible increase in risk. The estimated increase in risk is very small because ILRTs identify only a few potential containment leakage paths that cannot be identified by Types B and C testing, and the leaks that have been found by Type A tests have been only marginally above existing requirements. Given the insensitivity of risk to containment leakage rate and the small fraction of leakage paths detected solely by Type A testing, NUREG-1493 concludes that increasing the interval between ILRTs is possible with minimal impact on public risk.

3.1.2 TMI 10 CFR 50, Appendix J, Option B Licensing History May 27,1997 The NRC issued Amendment No. 201, which revised TS 6.8.5 to permit use of the 10 CFR Part 50, Appendix J, Option B, Performance-Based Containment Leakage Rate Testing. (Reference

1)

August 14, 2003 The NRC issued Amendment No. 244, which revised TS 6.8.5 to reflect a one-time deferral of the scheduled performance of the next Type A Containment ILRT from October 2003 to no later than September 2008. This change increased the test frequency from once every 10 years to once every 15 years (Reference 6).

June 29, 2007 The NRC issued Amendment No. 259, which revised TS 6.8.5 to allow a one-time deferral of the next Type A, containment leak rate test from "no later than September 2008" to "prior to startup from the T1R18 refueling outage. The T1R18 refueling outage was scheduled to begin no later than November 1, 2009." This change added approximately 15 months to the previously approved 15-year interval and allowed the Type A ILRT to be performed during a steam generator replacement outage in the fall of 2009 (Reference 7).

3.2 Description of the Primary Containment System 3.2.1 Reactor Building The Reactor Building is a reinforced concrete structure with cylindrical walls, a flat foundation mat, and a shallow dome roof. The foundation slab is reinforced with conventional mild steel reinforcing. The cylindrical walls are pre-stressed with a post-tensioning system in the vertical and horizontal directions. The dome roof is pre-stressed using a three-way post-tensioning tendon system. The inside surface of the Reactor Building is lined with a carbon steel liner to

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 4 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 ensure a high degree of leak-tightness for containment during operating and accident conditions. Normal line plate thickness is 3/8 inch for the cylinder and dome and 1/4 inch for the base.

The foundation mat is bearing on sound rock and is 9 ft. thick with a 2 ft. thick concrete slab above the bottom liner plate. The cylindrical portion has an inside diameter of 130 ft., wall thickness of 3 ft. 6 inches, and a height of 157 ft. from top of foundation slab to the spring line.

The shallow dome roof has a large radius of 110 ft., a transition radius of 20 ft. 6 inches, and a thickness of 3 ft, and an overall height of 32 ft 4 1/8 inches.

The Reactor Building has been designed to contain radioactive material, which might be released from the core following a loss of coolant accident (LOCA), at a maximum leak rate of 0.1 percent by weight of contained atmosphere in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at the design pressure. The pre-stressed concrete shell ensures that the structure has an elastic response to all loads and that the structure strains are limited such that the integrity of the liner is not compromised. The liner has been anchored so as to ensure composite action with the concrete shell.

The general construction sequence for the Reactor Building has been as follows:

a. Immediately after excavation of the rock for the foundation and a dewatering system was established, a lean concrete fill was placed to seal the rock to prevent weathering.

b. After the foundation was poured, the knuckle and bottom plate of the liner was installed and tested. Concrete was then placed on top of the base of the liner.

c. The liner was erected and individual welds tested prior to placing of reinforcement, tendon conduit, and concrete. Concrete work on the cylinder proceeded prior to completion of the cylindrical portion of the liner.

d. The dome liner was erected and individual welds tested prior to the placing of dome reinforcement, tendon circuit, and concrete.

e. The pre-stressing tendons were installed, stressed, and sealed off with end caps.

To support the 2009 Steam Generator Replacement (SGR) Project, a containment opening was made in the Reactor Building concrete wall and liner plate at the 290° azimuth, and directly above the existing Equipment Hatch to gain building access for rigging and handling of the steam generators. The design for this activity ensured that the opening area was restored to a condition meeting Reactor Building design requirements.

3.2.2 Liner and Plate Penetrations The Reactor Building is a steel-lined concrete shell in the form of a vertical right cylinder with an ellipsoidal dome and flat base. The concrete thickness is 3 ft. 6 inches for the cylindrical wall and 3 ft. for the dome. The liner has been designed as a free-standing vessel during erection.

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 5 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 Temporary erection loads, including loads resulting from using the liner as a form for concrete work, have been considered in the design. The liner is an element of the composite steel and concrete shell. The degree of leak tightness ensures a containment leak rate of no greater than 0.1 percent weight of containment atmosphere in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at 50.6 pounds per square in gauge (psig).

The steel plate for the liner main shell, including the dome, cylindrical wall, and base, conforms to:

a. ASTM A283-67, Grade C. Plate thickness 1/4 inch and 3/8 inch.

b. ASTM A516-67, Grade 55. Plate thickness 3/4 inch.

c. ASTM A36-67. Liner attachments, anchors for polar crane support and rolled sections including test channels and stiffeners.

The materials for penetration sleeves, including the personnel and equipment access hatches as well as the mechanical and electrical penetrations, conform to the requirements of the American Society of Mechanical Engineers (ASME) Nuclear Vessels Code for Class B Vessels. Materials for penetrations exhibit impact properties as required by Class B Vessels.

The equipment hatches and personnel access lock material is SA-516 Grade 60 modified to SA-300. Other penetrations materials are SA-516 Grade 60 modified to SA-300 for insert plates and A333 Grade 1 and SA-516 Grade 60 modified to SA-300 for sleeves, or equal.

The header plate for electrical penetration 201E is AS240 Grade 304.

To support the 2009 SGR Project, a containment opening was made in the Reactor Building concrete wall and liner plate at the 290° azimuth, and directly above the existing Equipment Hatch to gain building access for rigging and handling of the steam generators. A section of the liner plate approximately 23-6 wide and 22-0 tall, centered about the 290° azimuth and Elevation 378. The removed section of 3/8 liner plate was restored via full penetration welds with backing. The vacuum box test and 50.6 psig ILRT confirmed leak tightness.

3.2.3 Test Channels Steel channels have been provided along weld seams that were inaccessible. The channels have been segmented so as to ensure no length of weld covered by any one channel segment exceeds the greatest dimension of one plate. One plug fitting has been provided in each channel segment and extends through any covering material, including concrete. The fittings on the base have been protected by sleeves at the base weld to ensure no weld failure during placing of concrete. Test channels have been located on the face of the liner inside containment. Flanged heads have also been installed to cover penetration sleeve to liner plate welds.

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 6 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 3.2.4 Penetrations and Openings Personnel Access Locks Two personnel access locks are provided, one of which penetrates the dished head of the equipment hatch. Each personnel hatch is a welded steel assembly with double doors equipped with double gaskets to provide an air space that can be pressurized to the Reactor Building design pressure for leak testing or fluid blocking.

Equipment Access Hatch An equipment hatch with an inside diameter of 22 ft. 4 inches has been provided to enable passage of large equipment and components into the Reactor Building during a plant shutdown.

The items brought into the vessel include, in part, the reactor coolant pumps and motors, and reactor vessel O-rings.

A steel test channel was provided over the field weld of the penetration ring to the sleeve for replacement penetration 201E, to allow for pressure testing of the weld.

The following applies to both the equipment and personnel openings in the containment liner:

Flanged joints have been designed for the use of a double gasketed seal. This seal between the gaskets has been designed to a pressure equal to the design pressure.

The material used in the construction of the openings is compatible with the liner metallurgical characteristics.

The personnel opening doors are interlocked to prevent both doors being opened simultaneously. Interlocks are so connected that one door must be completely closed before the opposite door can be opened.

For the personnel openings, remote indication is provided in the Control Room to indicate whenever the interlock mechanism is defeated and the doors are not locked closed.

Personnel locks have an interior lighting system, which is capable of operating from the emergency power supply.

Personnel locks have an emergency communication system.

Provisions on personnel locks have been provided to permit bypassing the door interlocking system to allow doors to be left open when the plant is shut down.

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 7 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 The floor system of the personnel locks has been designed so that it can be easily removed.

The personnel locks have been designed, fabricated, tested, and stamped in accordance with the ASME Nuclear Vessels Code as Class B vessels.

Personnel locks have been designed to be capable of withstanding a test of 63.3 psig in the interspace between doors.

Personnel lock hinges are capable of a three-dimensional adjustment to assist proper seating. Hinges are capable of independent adjustment.

Seals, gaskets, O-rings, or other seating materials are suitable to withstand design temperature conditions.

Personnel lock equalizing valves are of the quick-acting type.

Mechanical Penetrations The following applies to the fabrication and testing of penetrations:

For hot pipelines (operating temperature equal to or greater than 150° F), an expansion joint was provided between the pipe and sleeve at the second barrier to accommodate the calculated axial and lateral pipe motions.

The penetration material is compatible with other liner materials.

Unibestos material was used where thermal insulation was required.

Pipe insertions are of greater wall thickness than wall thickness of the pipe elsewhere in the system.

Bellows, expansion joints, gaskets, canopies, protectors, or other flexible members were designed for a minimum of 250 cycles for the movement associated with each penetration.

The locations of penetrations with regard to azimuth location are within 1/2 inch measured on the circular section. The horizontal and vertical dimensions associated with the radial dimensions are 1/2 inch for the pipelines.

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 8 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 Penetrations were installed in the respective plate sections prior to testing.

Mechanical penetrations have double barriers designed for pressurization with air to 63.3 psig for leak testing and 60 psig during accident conditions.

Mechanical spares have both ends of the sleeves capped. Each sleeve is equipped with a 3/4 inch test connection outside containment. These sleeves are 12-inch, Schedule 80 pipe.

The containment penetration boundary welds are Category C, corner groove welds per paragraph N-461, ASME Section III, 1968 Edition. Corner groove welds were not considered suitable for radiography. Weld acceptance was by surface examination in accordance with Subparagraph N-1350(b), ASME Section III.

Electrical Penetrations Electrical Penetrations have been designed as follows:

Penetration cartridges have been installed in the penetration sleeves.

The penetration sleeves are 12-inch, Schedule 80 carbon steel pipe. Penetration sleeves were shop-welded to the reinforcement plate.

The weight of the liner cartridges is not more than 500 lb.

Leak Testing The penetrations have been tested as follows:

A proof test has been applied to each penetration by pressurizing the penetration annulus to 63.3 psig. The pressure was reduced to 55 psig and held at this pressure to soap bubble test any mated surfaces. Leaks were repaired and retested. This procedure was followed until no leaks existed.

Tests were conducted in accordance with ANS-7.60, Appendix A.

After installation and prior to the initial integrated leakage rate test of the Reactor Building, the spaces between the dual resilient seals on each of the doors of the personnel access air locks were pressurized with air at 50.6 psig and both sides of the seal checked with a sonic leak detector and/or soap bubble. Subsequently, with the

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 9 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 inner and outer doors both closed, the space between both doors was pressurized to 50.6 psig with air and sonic leak detection and/or soap bubble applied to the seals in the non-pressurized areas (i.e., inside and outside the Reactor Building).

Penetration 201E was replaced by a single header plate design, utilizing mechanically swaged ferrule seals with a qualified polysulfone resilient seal. The 24 feed-through ports and the field weld between the sleeve extension ring and the penetration sleeve, which has an applied leak test channel, were tested as noted above.

Penetration Appurtenances The following applies to the penetrations:

Reinforcing was designed to support the penetration in the liner for shop testing, shipping and field erection.

Bellows were fabricated from stainless steel having metallurgical characteristics compatible with mechanical design requirements. Bellows are suitably protected against field damage and are a part of the permanent installation.

Special Penetrations Two penetrations, which required special attention, are the containment supply and exhaust purge ducts. The following additional requirements were imposed on these penetrations:

Formed heads were supplied with the penetration for use during liner tests.

Each penetration is provided with two test connections, one 3/4-inch pipe size to test the annulus space and one 2-inch pipe size to test the purge duct.

3.2.5 Post-Tensioning System The pre-stressing system used for the Reactor Building is the BBRV system using a maximum of 169 1/4-inch diameter wires. The wires consist of a high tensile steel, bright, cold drawn, and stress-relieved conforming to ASTM A421-65T, Type BA. This type of wire is not susceptible to stress corrosion such as hydrogen cracking. The BBRV system uses parallel wires with cold-formed button heads at the ends, which bear upon a perforated steel anchor head, thus providing a positive mechanical means for transferring the pre-stress force. The anchorage hardware is designed and fabricated for the use of 170 wires. However, one hole located on the outer perimeter of the holes in the anchor head was used to accommodate a removable unstressed surveillance wire.

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 10 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 To support the 2009 SGR Project, a Containment Opening was made in the Reactor Building concrete wall and liner plate at the 290° azimuth, and directly above the existing Equipment Hatch to gain building access for rigging and handling of the steam generators. The Containment Opening was restored to its original design requirements using the following:

New vertical and horizontal tendon sheaths within the SGR construction opening area were installed at the original locations to replace those removed for creation of the opening. The tendon sheaths used carbon steel pipe rather than the spiral wound thin gauge corrugated steel sheaths originally used. The sheaths were secured and otherwise supported to allow installation of the tendons prior to placing concrete. The sheath-to-sheath connections were sealed to prevent concrete intrusion or grease leakage.

The following tendons were removed for creation of the opening and were replaced with new tendons, which were tensioned as a part of the closure process:

o Verticals - V131 through V140; o Horizontals - H-46-030 through H-46-039; and, o Horizontals - H-51-028 through H-51-039.

The following tendons were de-tensioned during creation of the opening, and re-tensioned as part of the closure process:

o Verticals - V113 through V130; o Verticals - V141 through V157; o Horizontals - H-46-028 through H-46-029; o Horizontals - H-46-040 through H-46-042; and, o Horizontals - H-51-040 through H-51-042.

Re-tensioning of the horizontal and vertical tendons commenced after a 72-hour cure time and the concrete reached a compressive strength of 5800 psi. Tendon re-tensioning was controlled and sequenced to preclude overstressing of the concrete. The tensioning force met the requirements of the UFSAR, and the minimum value required at the end of the plant life.

The tendons were greased after tendon tensioning using Visconorust 2090P-4 grease.

3.2.6 Corrosion Protection The containment structure is protected against external corrosive influences by the following means:

The Reactor Building is surrounded by a circular retaining wall extending from top of foundation to Elevation 304. The retaining wall provides access space for stressing and inspecting tendons. At the base of the retaining wall between the wall and the Reactor

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 11 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 Building, a drainage system is provided to prevent significant accumulation of water in the access space. Therefore, waterproofing of the lower portion of the cylinder was not required. There was no waterproofing or membrane used on the base slab. The retaining wall precludes contact of groundwater with the shell.

A cover of concrete was provided in excess of that for normal construction as exemplified by code requirements.

A galvanized steel conduit for tendons was used with the added precaution of a thicker-walled, rigid steel conduit in the base mat and extending immediately above into the cylinder.

An inboard-oriented haunch, which results in only nominal tensile stresses of the outer fibers; these stresses are within the capacity of concrete in tension.

The exposed surface of the liner has been given a protective coating. The tendons and included anchorages are coated with a temporary corrosion inhibitor for protection prior to bulk-filling the concrete. The casing filler provides an environment for the tendon where corrosion is highly unlikely. The retaining wall and drainage system eliminate groundwater corrosion problems on the liner.

The tendons were inserted in galvanized steel conduit embedded in the concrete, which provides an additional water barrier, as well as an electrical shield against stray currents. The inner surface of the steel conduit, as well as the tendons including anchorages, is protected with a heavy wax base corrosion inhibitor. The end anchorages are covered with a metal container.

The retaining wall and drainage system around the Reactor Building provide excellent protection for the liner and tendons against corrosion; therefore, no cathodic protection system was provided. Metallic components including the liner plate and tendon conduit are electrically connected to prevent stray current corrosion. The tendons are enclosed in a metallic tube so as to isolate them from outside electrical influences. The 2009 SGR Project replaced the existing tendon conduits within the opening with identical 4-3/4 outside diameter (OD) galvanized tubing with a 0.065 wall thickness. Similar to the existing tendon conduit configuration, a stainless-steel medium-pressure 4 diameter pipe repair clamp, 7-1/2 long with 3 bolts, was used as a tendon coupler to attach the existing and new galvanized tendon conduits.

Permanent reference electrode stations have been installed to facilitate measurements of structure potentials. These stations consist of a plastic pipe extending from the ground surface to the point at which the structure potential measurement was required. The plastic pipe functions as a salt bridge. Standard reference half cells placed into the pipe, down to the groundwater level, can be used to make structure potential measurements.

3.2.7 Containment Isolation System The fluid penetrations that require isolation after an accident are classified as follows:

Type I Each line connecting directly to the Reactor Coolant System has two Reactor Building isolation valves. One valve is external and the other is internal to the

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 12 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 Reactor Building. These valves may be either: a check valve and a remotely operated valve, or two remotely operated valves, depending on the direction of normal flow.

Type II Each line connecting directly to the Reactor Building atmosphere has two isolation valves. At least one valve is external and the other may be internal or external to the Reactor Building. These valves may be either: a check valve and a remotely operated valve, or two remotely operated valves, depending on the direction of normal flow.

Type III Each line not directly connected to the Reactor Coolant System or not open to the Reactor Building atmosphere has at least one valve, either a check valve or a remotely operated valve. This valve is located external to the Reactor Building.

A closed loop, which has a low probability of rupture during an accident, may be used as the second isolation barrier.

Type IV Lines that penetrate the Reactor Building and are connected to either the building or the Reactor Coolant System, but which are never opened during reactor operation, have two normally closed barriers. (e.g., blind flange, closed valve).

3.3 Traditional Engineering Considerations 3.3.1 ILRT History Type A testing is performed to verify the integrity of the containment structure in its LOCA configuration. Industry test experience has demonstrated that Types B and C testing detect a large percentage of containment leakages and that the percentage of containment leakages that are detected only by ILRT is very small.

TMI-1 has undergone eight (8) operational Type A tests in addition to the pre-operational Type A test. The results of these tests demonstrate that the TMI-1 containment structure remains an essentially leak-tight barrier and represents minimal risk to increased leakage. These plant-specific results support the conclusions contained in NUREG-1493. The TMI-1 Type A test historical results are presented in Table 3.3.1-1 below.

Table 3.3.1 TMI-1 Type A Test History Test Date 95% Upper Confidence Limit (wt.%/day)

As-Found Leakage (wt.%/day)

Acceptance Criteria (wt.%/day)

As-Left Leakage (wt.%/day)

Acceptance Criteria (0.75La)

(wt.%/day)

1. 1974 Note 1 Note 1 Note 1 0.043 0.075 1977 0.052 Note 2 0.075 0.059 0.075 1978 0.064 0.081*

0.075 0.071 0.075 1981 0.028 0.028 0.075 0.028 0.075 1984 0.0405 0.057 0.075 0.042 0.075 11/9/1986 0.034 Note 3 0.075 0.034 0.075

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 13 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 NOTES:

1. The 1974 ILRT was a Pre-Operational Test. Therefore, no As-Found data was recorded.

2. The initial leak rate testing performed between April 16, 1977, and April 18, 1977, was not successful. An extensive search failed to identify any significant sources of leakage; however, a shift in the trend of the containment mass points occurred after 16:00 hours on April 18, 1977.

3. The initial ILRT testing revealed leakage past Reactor Building purge valve AH-V-1A and AH-V-1B interspace isolation valves PP-V-101 and PP-V-102. This leak path was eliminated and the test was successfully re-performed. Currently, this leakage path is local leak rate tested on a quarterly frequency.

4. The first test was declared invalid due to once-through steam generator valve leakage.

These valves were out of their normal test position and outside the test envelope. Valve lineup guidance was added to the procedure and these valves were shut or isolated as required. The ILRT was re-performed successfully. Subsequently, all Hancock 5500 W instrument root and drain/vent skin valves on the once-through steam generators were replaced with a different design valve, which is much less prone to body-to-bonnet flange leakage.

5. During the 1993 ILRT, a combination ILRT/local leakage rate test (LLRT) as-found test was performed at the beginning of the refueling outage, which represented a different method from that used in the past. Approximately two-thirds of the LLRTs were performed just prior to the as-found ILRT during the stabilization period. Also, during the ILRT larger-than-normal variations in the Reactor Building temperatures were observed.

These variations, while within acceptable bands, may have also influenced the test results.

• Two issues prevented the respective containment isolation valves from being exposed to the ILRT pressure. One was that the industrial cooler system was in service during the test, and the other was that IC-V-4 could not be opened to allow draining of the associated piping. The ILRT result for these penetrations is 0.007 wt%/day, which adjusts the measured ILRT result to 0.07 1wt%/day. The Type B & C summation adjustment then makes the performance leak rate 0.081 wt%/day.

1/13/1990 0.0126 Note 4 0.075 0.01326 0.075 1993 0.0707 0.0718 (Note 5) 0.075 0.0718 0.075 2009 0.0564 0.05693 0.1 0.05693 0.075

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 14 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 TMI-1 ILRT Performance Leakage Rate Determination The definition of performance leakage rate was first introduced in NEI 94-01, Revision 0, following the addition of Option B to 10 CFR 50, Appendix J. On May 27, 1997, the NRC approved TS Amendment 201 for TMI-1 (Reference 1) authorizing the implementation of 10 CFR 50, Appendix J, Option B for Types A, B and C tests. Therefore, the 1974, 1977, 1981, 1984, 1986, 1990, and 1993 ILRTs were not performed under the performance-based standard as they were performed prior to the adoption of Appendix J, Option B. NEI 94-01, Revision 0, Section 9.1.1, Performance Criteria, states, The performance criteria for Type A test allowable leakage is less than 1.0La. This allowable leakage rate is calculated as the sum of the Type A Upper Confidence Limit (UCL) and As-Left Minimum Pathway Leakage Rate (MNPLR) leakage rate for all Type B and Type C pathways that were in service, isolated, or not lined up in their test position (i.e., drained and vented to containment atmosphere) prior to performing the Type A test. In addition, leakage pathways that were isolated during performance of the test because of excessive leakage must be factored into the performance determination. If the leakage can be determined by an LLRT, the As-Found MNPLR for that leakage path must also be added to the Type A UCL.

Table 3.3.1 Verification of Current Extended ILRT Interval for TMI-1 (Jan 2010)

Measured Leakage Rate at 95% UCL

(%wt./day)

Water Level Corrections

(%wt./day)

Corrections for valves not in Accident Positions during Test

(%wt./day)

Components Isolated During ILRT Due to Excessive Leakage

(%wt./day)

Performance Leakage Rate

(%wt./day)

(Acceptance Criteria 0.5 wt.%/day Acceptance Criteria (1.0 La)

Test Method 0.0564 0

0.00053 0

0.057 0.1 Mass Point Since the performance leakage rate using mass point leakage results for the 2010 ILRT was less than the performance criteria value of 1.0 La (0.1 wt.%/day), the ILRT may remain on the extended interval of at least once per ten years.

3.3.2 Containment Leakage Rate Program - Types B and C Testing Program TMI-1 Types B and C testing program requires testing of electrical penetrations, airlocks, hatches, flanges, and containment isolation valves (CIVs) in accordance with 10 CFR 50, Appendix J, Option B and RG 1.163 (Reference 2). The results of the test program are used to demonstrate that proper maintenance and repairs are made on these components throughout their service life. The Types B and C testing program provides a means to protect the health and safety of plant personnel and the public by maintaining leakage from these components below appropriate limits. In accordance with TS 6.8.5, the allowable maximum pathway total Types B and C leakage is 0.6 La (104,846 sccm) where La equals approximately 174,743 standard cubic centimeters per minute (sccm).

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 15 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 As discussed in NUREG-1493 (Reference 5), Type B and Type C tests can identify the vast majority of all potential containment leakage paths. Type B and Type C testing will continue to provide a high degree of assurance that containment integrity is maintained.

A review of the Type B and Type C test results from 2007 through 2017 for TMI-1 has shown substantial margin between the actual As-Found (AF) and As-Left (AL) outage summations and the regulatory requirements, as described below:

The As-Found minimum pathway leak rate average for TMI, Unit 1, shows an average of 7.73% of 0.6 La with a high of 12.94% of 0.6 La.

The As-Left maximum pathway leak rate average for TMI, Unit 1, shows an average of 25.57% of 0.6 La with a high of 36.96% of 0.6 La.

Table 3.3.2-1 provides Types B and C LLRT data trend summaries for TMI-1, inclusive of the 2009 ILRT.

Table 3.3.2 TMI-1 Types B and C LLRT Combined As-Found/As-Left Trend Summary RFO / Year 2007 2009 2011 2013 2015 2017 T1R17 T1R18 T1R19 T1R20 T1R21 T1R22 AF Min Path (sccm) 7,524 7,846.0 3,726.0 7,512.0 12,946.0 9,055.5 Fraction of 0.6 La (percent) 7.18 7.48 3.55 7.16 12.35 8.64 AL Max Path (sccm) 22,489 38,752.0 11,825.0 17,821.0 33,902.0 36,099.0 Fraction of 0.6 La (percent) 21.45 36.96 11.28 17.00 32.34 34.43 AL Min Path (sccm) 10,416 8,162.0 3,747.0 7,289.0 12,966.0 13,568.5 Fraction of 0.6 La (percent) 9.93 7.78 3.57 6.95 12.37 12.94

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 16 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 3.3.3 Type B and Type C LLRT Program Implementation Review Table 3.3.3-1 identifies the components that were on extended intervals and have not demonstrated acceptable performance during the previous two outages for TMI-1.

Table 3.3.3-1 -TMI-1 Type B and C LLRT Program Implementation Review T1R21 - Fall 2015 Component As-Found SCCM Admin Limit SCCM As-Left SCCM Cause of Failure Corrective Action Scheduled Interval NS-V-11 9000 2000 1050 Hinge pin body bore holes were slightly oval, causing the disc to hang lower than design specification Valve internal repairs completed with new parts as part of scheduled check valve preventative maintenance 30 Months T1R22 - Fall 2017 Component As-Found SCFH Admin Limit SCFH As-Left SCFH Cause of Failure Corrective Action Scheduled Interval WDL-V-534/535 4200 500 7000 Foreign Material on Valve Seat High-Pressure Flush Between Valves 30 Months CF-V-2B 380 150 63 Foreign Material on Valve Seat Seating Surface Flushed with Clean Water 30 Months CM-V-3 2300 300 30 Worn Seating Surface Replaced

Packing, Stem, Ball and Valve Seats 30 Months The percentage of the total number of TMI-1 Type B penetrations that are on 120-month extended performance-based test intervals is 4.2%.

The percentage of the total number of TMI-1 Type C penetrations that are on 60-month extended performance-based test intervals is 71.0%.

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 17 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 3.3.4 Containment Electrical Penetration Missed Surveillance TMI-1 previously exempted the containment electrical penetrations having epoxy sealant from LLRT under 10 CFR 50, Appendix J. This had been justified by station personnel when Appendix J testing was implemented because epoxy did not seem to meet the following Type B testing criteria from Appendix J: Containment penetrations whose design incorporates resilient seals, gaskets, or sealant compounds, piping penetrations fitted with expansion bellows, and electrical penetrations fitted with flexible metal seal assemblies; or, the following criteria from NEI 94-01: Containment penetrations whose design incorporates resilient seals, gaskets, sealant compounds, expansion bellows, or flexible seal assemblies. NEI 94-01, Revision 3-A, Section 11.3.1 provided clarification that epoxy sealant is sensitive to age-related degradation. The original Final Safety Analysis Report (FSAR) states the following after listing the gasketed penetrations required to be LLRT tested, All other pipe penetrations of containment have at least one non-resilient barrier, attached with full penetration structural welds, between the containment atmosphere and the environment. All electrical penetrations have at least one epoxy insulator, structurally bonded between wire and steel, and full penetration, steel to steel welds as the barrier between the containment atmosphere and the environment. Thus, the determination that the electrical penetrations could be excluded from Appendix J testing was based on the epoxy sealant not being a resilient material.

The FSAR specifically stated in Section 1.4.56, Criterion 56 - Provisions for Testing of Penetrations (Category A), The penetration pressurization system provides continuous pressurization and a means of continuously monitoring the leakage rate from the equipment hatch resilient seals. Continuous leakage rate monitoring of the electrical penetrations, although not required, is conducted at 30 psig. For penetrations having resilient seals that are not continuously pressurized (the purge isolation valves, the doors of the two personnel access airlocks, and the fuel transfer tube flange O rings), there are special provisions for conducting individual leakage rate tests at design pressure at any time. Furthermore, FSAR Table 5.7-2, Components Required to be Tested Using a Type B Test, which lists penetrations requiring LLRT, includes Electrical Penetration 201E as the only one requiring testing since it was recently replaced with a design having a resilient seal on the conductor. The other electrical penetrations were not considered when this modification was reviewed for impact.

Risk Assessment of Missed Surveillance: LLRT of Reactor Building Electrical Penetrations TMI-1 previously exempted the Reactor Building electrical penetrations with epoxy sealant from LLRT under 10 CFR 50, Appendix J. Therefore, these penetrations had not been locally leak rate tested at the 50.6 psig peak accident pressure required by Appendix J. The electrical penetrations were tested quarterly during a 30 psig leakage test used to monitor for long-term degradation of the epoxy sealants. However, this test did not meet the requirements of 10 CFR 50, Appendix J.

LLRT falls under TS Section 4.4.1, Containment Leakage Tests, which applies to containment leakage for the Reactor Building. The objective is to verify that leakage from the Reactor Building is maintained within allowable limits. TS 4.4.1.2 states, Local Leakage Rate Testing (LLRT) shall be conducted in accordance with the Reactor Building Leakage Rate Testing Program. LLRT shall be performed at a pressure not less than peak accident pressure Pac with the exception that the airlock door seal tests shall normally be performed at 10 psig and the periodic containment airlock tests shall be performed at a pressure not less than Pac. LLRT

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 18 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 frequencies shall be in accordance with the Reactor Building Leakage Rate Testing Program.

The aforementioned program is the stations Appendix J program. There is no Limiting Condition of Operation (LCO) or Action Statement that applies to the Appendix J Program or TS 4.4.1.

The lack of testing for the Reactor Building electrical penetrations was treated as a missed surveillance. TS 4.0.2 defines Surveillance Standards at TMI-1. This TS required risk impact to be evaluated. Though the LLRT had never been performed on these electrical penetrations, the quarterly 30 psig leak checks and the ILRT, last performed in January 2010, had not identified any leakage through the penetrations. During the ILRT, the penetrations were exposed to the 50.6 psig peak accident pressure.

The risk evaluation considered all Reactor Building electrical penetrations with epoxy sealant, simultaneously (there are 37 such penetrations). The applicable Maintenance Rule function was 774-R01, Maintain RB Integrity where electrical conductors pass through the RB wall at all times. The electrical penetrations were not included in the Full Power Internal Events PRA or in the PARAGON model, which are used in assessing risk per site procedures. Electrical penetrations were also considered outside of the scope of Configuration Risk Management per 10 CFR 50.65(a)(4) as configuration control lies in scope of the associated Appendix J Programmatic requirements. Since this function is not explicitly represented in the TMI-1 PRA model and PARAGON model, the Refined Risk Assessment Methodology Option No. 3 will be used. This method was used to evaluate the appropriate compensatory actions.

The Maintenance Rule database for TMI-1 classifies 774-R01 as High Safety Significant (HSS),

and it showed that 774-R01 is currently being monitored under Maintenance Rule section 10 CFR 50.65(a)(2). There were no functional failures during the 2015-2016 reporting period. The 30 psig periodic leak checks performed every quarter, to monitor for long-term degradation of the epoxy seals, and the ILRT performed in January 2010, did not identify any leakage of the electrical penetrations. That ILRT was measured to be 54% La, against the leakage limit of 75% La, where La is 174,743 sccm. Based on these results and based on discussions with the Appendix J Program Engineer, there is no reason to suspect an increase in unreliability.

All Reactor Building electrical penetrations are constantly pressurized to 30 psig with nitrogen through the Penetration Pressurization System Electrical subsystem. The piping and tubing connect to and end at the penetrations. Any leakage would occur though valve stem, piping/tubing fittings, and the penetration seals. There are four electrical subsystem nitrogen manifolds (D, E, F and G), with each manifold supplying at least eight electrical penetrations.

There are 37 canister-type penetrations. The epoxy is the sealing medium between the individual cables and the penetration canister jacket on both ends of the canister. The space between the ends is kept pressurized with 30 psig of nitrogen during normal operation and during emergency safeguards conditions. Per the Penetration Pressurization System Operating procedure, leakage through these seals is checked every quarter, by checking nitrogen flow through each manifold. The Reactor Building penetrations are not in scope of Fire (a)(4).

Due to the passive nature of the electrical penetrations, modification of the PARAGON model to include these penetrations is not warranted. The PARAGON model does include modeling for the Online Containment Conditions Safety Function, but that modeling is limited to CIVs, Reactor Building Emergency Cooling and Reactor Building Spray Trains.

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 19 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 Based on historical performance of leak rate testing of the Reactor Building electrical penetrations and their role in station risk, the risk of continued operation, until the LLRT on these penetrations can be performed, was acceptably small. Modification of the PARAGON model is not required. Maintaining the current monitoring strategy until the LLRT could be performed was recommended.

During T1R22 in the Fall of 2017, the containment electrical penetration manifolds first-time LLRT was performed. Tables 3.3.4-1 through 3.3.4-4 detail the test results.

Table 3.3.4 Electrical Penetration Manifold 'D' T1R22 First-Time LLRT Performance Test Results Penetration No.

Measured Leakage (SLPM)

Acceptance Criteria (SLPM)

Test Date 217E 0

0.02 9/18/2017 218E 0

0.02 9/18/2017 219E 0

0.02 9/18/2017 220E 0

0.02 9/18/2017 223E 0

0.02 9/18/2017 224E 0

0.02 9/18/2017 225E 0

0.02 9/18/2017 226E 0

0.02 9/18/2017 Table 3.3.4 Electrical Penetration Manifold 'E' T1R22 First-Time LLRT Performance Test Results Penetration No.

Measured Leakage (SLPM)

Acceptance Criteria (SLPM)

Test Date 228E 0

0.02 9/18/2017 229E 0

0.02 9/18/2017 230E 0

0.02 9/18/2017 231E 0

0.02 9/18/2017 232E 0

0.02 9/18/2017 233E 0

0.02 9/18/2017 234E 0

0.02 9/18/2017 235E 0

0.02 9/18/2017 236E 0

0.02 9/18/2017 237E 0

0.02 9/18/2017 238E 0

0.02 9/18/2017 239E 0

0.02 9/18/2017

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 20 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 Table 3.3.4 Electrical Penetration Manifold 'F' T1R22 First-Time LLRT Performance Test Results Penetration No.

Measured Leakage (SLPM)

Acceptance Criteria (SLPM)

Test Date 315E 0

0.02 9/18/2017 316E 0

0.02 9/18/2017 317E 0

0.02 9/18/2017 318E 0

0.02 9/18/2017 341E 0

0.02 9/18/2017 342E 0

0.02 9/18/2017 343E 0

0.02 9/18/2017 344E 0

0.02 9/18/2017 Table 3.3.4 Electrical Penetration Manifold 'G' T1R22 First-Time LLRT Performance Test Results Penetration No.

Measured Leakage (SLPM)

Acceptance Criteria (SLPM)

Test Date 201E 0

0.02 9/18/2017 202E 0

0.02 9/18/2017 203E 0

0.02 9/18/2017 204E 0

0.02 9/18/2017 205E 0

0.02 9/18/2017 206E 0

0.02 9/18/2017 311E 0

0.02 9/18/2017 312E 0

0.02 9/18/2017 313E 0

0.02 9/18/2017 314E 0

0.02 9/18/2017 3.3.5 Protective Coating Monitoring and Maintenance Program The Protective Coating Monitoring and Maintenance Program performs examinations on Service Level I coatings inside the containment. Service Level I coatings are used in areas where corrosion protection may be required and where coating failure could adversely affect the operation of post-accident fluid systems and thereby impair safe shutdown. The Protective Coating Monitoring and Maintenance Program provides for inspections, assessment, and repairs for any condition that adversely affects the ability of Service Level I coatings to function as intended.

Based on current (and previous) inspection results and supporting evidence, a technical evaluation is processed to evaluate any of the following conditions:

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 21 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 Coatings were previously acceptable, and new coating discrepancies have developed, or coating discrepancies noted during previous inspection have worsened.

Coatings were previously inspected, but inaccessible for inspection in the current period.

Coatings inside containment were previously identified as unqualified, and new coating discrepancies have developed, or coating discrepancies noted during previous inspection have worsened.

Coatings inside containment are newly reported as not matching known generic types, or are of unknown type/origin (i.e., are indeterminate).

For repair dispositions, the following corrective actions are considered:

Remove entire coating, or remove degraded area(s) of coating to sound coating (without re-coating or repair coating).

Remove degraded coating to sound coating and perform a proper coating repair.

Remove entire coating and re-coat with evaluated alternate coating system.

For containment coatings, allow coating to remain as-is, and add quantity of discrepant coating material to unqualified coatings inventory (if sufficient margin exists).

Unqualified/Degraded Coatings in Containment At TMI-1, the quantity of unqualified coatings inside containment is required to be maintained to ensure licensing basis limits for ECCS suction strainers are not exceeded. The total amount of unqualified alkyd, enamel, and epoxy inside the reactor building is 5,839 square feet. The limit for the total amount of unqualified coatings is 7,005 square feet. The margin for unqualified coating is 17 percent. The total amount of degraded qualified coatings inside the reactor building is 156 square feet. The limit for degraded qualified coating is 600 square feet. The margin for degraded qualified coating is 74 percent. The total amount of exposed concrete that would become submerged following a LOCA is 449 square feet. The limit for exposed concrete is 600 square feet. The margin for exposed concrete is 25 percent.

3.3.6 Containment Inservice Inspection Program The TMI-1 Containment ISI (CISI) Plan includes ASME Section XI Class MC pressure retaining components and their integral attachments (including the Class CC metal liner), and Class CC components and structures, and post-tensioning systems that meet the criteria of sub-article IWA-1300. This CISI Plan also includes information related to augmented examination areas, component accessibility and examination review.

The TMI-1 Second Interval CISI Program Plan was developed in accordance with the requirements of 10 CFR 50.55a including all published changes through October 10, 2008, and the 2004 Edition, No Addenda of ASME Section XI, subject to the limitations and modifications contained within paragraph (b) of the regulation.

The TMI-1 Second CISI Interval is effective from April 20, 2011, through April 19, 2021.

However, the end of the TMI-1 Second CISI Interval is extended for one year from April 20, 2021, to April 19, 2022, per paragraph IWA-2430(d)(1) of ASME Section XI. The interval date will be extended to allow refueling outage T1R24 to fall within the Second CISI Interval. The

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 22 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 start date of the Third CISI Interval is not revised, thus creating an overlap of the two intervals.

Examinations can be performed during this time for either interval; however, no single examination can be credited to both intervals.

TABLE 3.3.6-0 SECOND CISI INTERVAL/PERIOD/OUTAGE MATRIX (FOR CISI CLASS MC COMPONENT EXAMINATIONS)

Interval Period Outages Start Date to End Date Start Date to End Date Projected Outage Start Date or Outage Duration Outage Number 2nd 04/20/11 to 04/19/221 1st 04/20/11 to 04/19/14 Scheduled 10/11 T1R19 Scheduled 10/13 T1R20 2nd 04/20/14 to 04/19/18 Scheduled 10/15 T1R21 Scheduled 09/17 T1R22 3rd 04/19/18 to 04/19/18 Scheduled 09/19 T1R23 Scheduled 09/21 T1R24 Note 1: The end of the TMI Second CISI Interval was extended for one year from April 19, 2021 to April 19, 2022 per Paragraph IWA-2430(d)(1) of ASME Section XI. The interval date was extended to allow Refueling outage T1R24 to fall within the Second CISI Interval.

The start date of the Third CISI Interval is not revised, thus creating an overlap of the two intervals. Examinations can be performed during this time for either interval; however, no single exam can be credited to both intervals.

Code of Federal Regulations 10 CFR 50.55a Requirements The following paragraphs in 10 CFR 50.55a that are applicable to the TMI-1 CISI Program are as follows:

10 CFR 50.55a(b)(2)(viii)(E) - Concrete Containment (CC) Examinations - Fifth Provision For Class CC applications, the licensee shall evaluate the acceptability of inaccessible areas when conditions exist in accessible areas that could indicate the presence of or result in degradation to such inaccessible areas. For each inaccessible area identified, the licensee shall provide the following in the ISI Summary Report required by IWA-6000:

(1)

A description of the type and estimated extent of degradation, and the conditions that led to the degradation; (2)

An evaluation of each area, and the result of the evaluation, and; (3)

A description of necessary corrective actions.

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 23 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 10 CFR 50.55a(b)(2)(viii)(F) - Concrete Containment Examinations - Sixth Provision Personnel that examine containment concrete surfaces and tendon hardware, wires, or strands must meet the qualification provisions in IWA-2300. The "owner-defined" personnel qualification provisions in IWL-2310(d) are not approved for use.

10 CFR 50.55a(b)(2)(viii)(G) - Concrete Containment Examinations - Seventh Provision Corrosion protection material must be restored following concrete containment post-tensioning system repair and replacement activities in accordance with the quality assurance program requirements specified in IWA-1400.

10 CFR 50.55a(b)(2)(ix)(A) - Metal Containment (MC) Examinations - First Provision For Class MC applications, the following apply to inaccessible areas:

(1)

The applicant or licensee must evaluate the acceptability of inaccessible areas when conditions exist in accessible areas that could indicate the presence of or could result in degradation to such inaccessible areas.

(2)

For each inaccessible area identified, the applicant or licensee must provide the following in the ISI Summary Report as required by IWA-6000:

(i)

A description of the type and estimated extent of degradation, and the conditions that led to the degradation; (ii)

An evaluation of each area, and the result of the evaluation; and (iii)

A description of necessary corrective actions.

10 CFR 50.55a(b)(2)(ix)(B) - Metal Containment Examinations - Second Provision When performing remotely the visual examinations required by Subsection IWE, the maximum direct examination distance specified in Table IWA-2210-1 may be extended and the minimum illumination requirements specified in Table IWA-2210-1 may be decreased provided that the conditions or indications for which the visual examination is performed can be detected at the chosen distance and illumination.

10 CFR 50.55a(b)(2)(ix)(F) - Metal Containment Examinations - Sixth Provision VT-1 and VT-3 examinations must be conducted in accordance with IWA-2200. Personnel conducting examinations in accordance with the VT-1 or VT-3 examination method must be qualified in accordance with IWA-2300. The "owner-defined" personnel qualification provisions in IWE-2330(a) for personnel that conduct VT-1 and VT-3 examinations are not approved for use.

10 CFR 50.55a(b)(2)(ix)(G) - Metal Containment Examinations - Seventh Provision The VT-3 examination method must be used to conduct the examinations in Item E1.12 and E1.20 of Table IWE-2500-1, and the VT-1 examination method must be used to conduct the examination in Item E4.11 of Table IWE-2500-1. An examination of the pressure-retaining

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 24 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 bolted connections in Item E1.11 of Table IWE-2500-1 using the VT-3 examination method must be conducted once each interval. The "owner-defined" visual examination provisions in IWE-2310(a) are not approved for use for VT-1 and VT-3 examinations.

10 CFR 50.55a(b)(2)(ix)(H) - Metal Containment Examinations - Eighth Provision Containment bolted connections that are disassembled during the scheduled performance of the examinations in Item E1.11 of Table IWE-2500-1 must be examined using the VT-3 examination method. Flaws or degradation identified during the performance of a VT-3 examination must be examined in accordance with the VT-1 examination method. The criteria in the material specification or IWB-3517.1 must be used to evaluate containment bolting flaws or degradation. As an alternative to performing VT-3 examinations of containment bolted connections that are disassembled during the scheduled performance of Item E1.11, VT-3 examinations of containment bolted connections may be conducted whenever containment bolted connections are disassembled for any reason.

Augmented Examination Areas:

Metal containment surface areas subject to accelerated degradation and aging require augmented examination per Examination Category E-C and paragraph IWE-1240. Similarly, concrete surfaces may be subject to Detailed Visual examination in accordance with Item Number L1.12 and paragraph IWL-2310(b), if declared to be 'Suspect Areas'.

The moisture barrier is located on the interior surface of containment where the concrete floor slab meets the metal liner at Elevation 281-0. This is a cork and silicone sealant seal between the concrete floor surface and the metallic liner. For the First CISI Interval, TMI-1 had classified portions of the metallic liner near the moisture barrier under Examination Category E-C, Item Numbers E4.11 and E4.12. Successive examinations have since been performed and the areas no longer require the Examination Category E-C designation per ASME Section XI. The areas of corrosion identified during the 2007 and earlier examinations were examined in 2009 and determined to be unchanged from prior examinations. The areas of corrosion where the liner was reduced to <90% of nominal thickness were restored to >90% nominal thickness during the 2009 refueling outage. This corrosion was attributed to degradation of the moisture barrier, which has been replaced for 360 degrees around the liner. However, for the Second CISI Interval, TMI-1 will optionally continue to identify the area near the moisture barrier Examination Category E-C, Item Number E4.11 and will implement an owners augmented examination program. The owners augmented examination program will be to conduct a visual examination each refueling outage at the metallic liner to moisture barrier interface.

No other significant conditions are currently identified in the Second CISI Interval as requiring application of additional augmented examination requirements under paragraph IWE-1240 or IWL-2310, or as additional owners augmented examination requirements under paragraph IWE-1240 or IWL-2310, or as additional owners augmented examination programs.

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 25 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 Component Accessibility CISI Class MC and CC components subject to examination shall remain accessible for either direct or remote visual examination from at least one side, per the requirements of ASME Section XI, paragraph IWE-1230.

Paragraph IWE-1231(a)(3) requires 80% of the pressure-retaining boundary that was accessible after construction to remain accessible for either direct or remote visual examination, from at least one side of the vessel, for the life of the plant.

Portions of components embedded in concrete or otherwise made inaccessible during construction are exempted from examination, provided that the requirements of ASME Section XI, paragraph IWE-1232 have been fully satisfied.

In addition, inaccessible surface areas exempted from examination include those surface areas where visual access by line of sight with adequate lighting from permanent vantage points is obstructed by permanent plant structures, equipment, or components; provided these surface areas do not require examination in accordance with the inspection plan, or augmented examination in accordance with paragraph IWE-1240.

Responsible Individual and Engineer ASME Section XI, Subsection IWE requires the Responsible Individual to be involved in the development, performance, and review of the CISI examinations. The Responsible Individual shall meet the requirements of ASME Section XI, paragraph IWE-2320.

ASME Section XI, Subsection IWL requires the Responsible Engineer to be involved in the development, approval, and review of the CISI examinations. The Responsible Engineer shall meet the requirements of ASME Section XI, paragraph IWL-2320.

Inspection Periods First Interval CISI Program CISI examinations were originally invoked by amended regulations contained within a Final Rule issued by the USNRC. The amended regulation incorporated the requirements of the 1992 Edition through the 1992 Addenda of ASME Section XI, Subsections IWE and IWL, subject to specific modifications that were included in paragraphs 10 CFR 50.55a(b)(2)(ix) and 10 CFR 50.55a(b)(2)(x).

The final rulemaking was published in the Federal Register on August 8, 1996 and specified an effective date of September 9, 1996. Implementation of the Subsections IWE and IWL Program from a scheduling standpoint was driven by the five-year expedited implementation period per 10 CFR 50.55a(g)(6)(ii)(B), which specified that the examinations required to be completed by the end of the First Period of the First CISI Interval (per Table IWE-2412-1) be completed by the effective date (by September 9, 2001).

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 26 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 The TMI-1 First Interval CISI Program was aligned with the Third Interval ISI Program and was effective from April 20, 2001, through April 19, 2011. However, at the end of the First CISI Interval and in preparation for the Second CISI Interval, a one-year extension was taken per paragraph IWA-2430(d)(1) of ASME Section XI, which allowed an inspection interval to be extended or decreased by as much as one year. As permitted by this allowance, the TMI-1 Third ISI Interval was extended by one year from April 20, 2011, through April 19, 2012, to allow Refueling outage T1R19 to fall within the First ISI Interval. The start date of the Second ISI Interval was not revised, thus creating an overlap of the two intervals. Examinations could have been performed during this time for either interval; however, no single examination could be credited to both intervals.

Second Interval CISI Program Pursuant to 10 CFR 50.55a(g), licensees are required to update their CISI Programs to meet the requirements of ASME Section XI once every ten years or inspection interval. The CISI Program is required to comply with the latest Edition and Addenda of ASME Section XI incorporated by reference in 10 CFR 50.55a twelve months prior to the start of the interval per 10 CFR 50.55a(g)(4)(ii). Based on this date, the latest Edition and Addenda of ASME Section XI referenced in 10 CFR 50.55a(b)(2) twelve months prior to the start of the Second CISI Interval was the 2004 Edition, No Addenda.

The TMI-1 Second Interval CISI Program Plan was developed in accordance with the requirements of 10 CFR 50.55a including all published changes through October 10, 2008, and the 2004 Edition, No Addenda of ASME Section XI, subject to the limitations and modifications contained within paragraph (b) of the regulation. These limitations and modifications are detailed in Table 1.9-1 of this section. The Second Interval CISI Program Plan addresses Subsections IWE, IWL, approved ASME IWE/IWL Code Cases, approved alternatives through related relief requests and safety evaluation reports (SERs), and utilizes Inspection Program B as defined therein.

The TMI-1 Second CISI Interval is effective from April 20, 2011 through April 19, 2021.

However, the end of the TMI-1 Second CISI Interval is extended for one year from April 20, 2021, to April 19, 2022, per ASME Section XI, paragraph IWA-2430(d)(1). The interval date will be extended to allow Refueling outage T1R24 to fall within the Second CISI Interval. The start date of the Third CISI Interval is not revised, thus creating an overlap of the two intervals.

Examinations can be performed during this time for either interval; however, no single examination can be credited to both intervals.

Component Summary Tables Tables 3.3.6-1 and 3.3.6-2 provide a summary of the ASME Section XI containment structures, post-tensioning systems and augmented program components for the Second CISI Interval at TMI-1.

Tables 3.3.6-3 through 3.3.6-6 provide the specific examination criteria for the CISI Class MC components at TMI-1 for the Second CISI Interval. The methodology for selection and scheduling of Examination Categories E-A and E-C components is in accordance with ASME Section XI, Table IWE-2500-1. Similarly, the methodology for selection and scheduling of

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 27 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 Examination Category L-A and L-B components is in accordance with ASME Section XI, Table IWL-2500-1. (Note: IWL components are scheduled through random selection prior to the performance of surveillance activities. Components are not prescheduled or contained within the program scheduling database). Augmented examinations following post-tensioning system repair/replacement activities are scheduled in accordance with Table IWL-2521-2. Therefore, a detailed selection methodology is not required.

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 28 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 Table 3.3.6 IWE Component Examination Table Examination Category (with Examination Category Description)

Item Number Description Exam Requirements Total Number of Components by System Relief Request /

TAP Number Notes E-A Containment Surfaces E1.11 Containment Vessel Pressure Retaining Boundary - Accessible Surface Areas General Visual 5 (5 Areas/Zones) Liner Dome, Walls, Penetrations, Hatches, Transfer Tubes (Not Requiring Examinations per IWE) and Attachments N/A N/A E1.11 Containment Vessel Pressure Retaining Boundary - Bolted Connections, Surface Visual, VT-3 33 N/A 1

E1.12 Containment Vessel Pressure Retaining Boundary - Wetted Surfaces of Submerged Areas Visual, VT-3 2 (2 Penetrations - Fuel Transfer Tubes)

N/A 2

E1.30 Moisture Barriers General Visual 1 (1 Seal, 360° at the 281' Elevation)

N/A N/A E-C Containment Surfaces Requiring Augmented Examination E4.11 Containment Surface Areas -

Visible Surfaces Visual, VT-1 1 (1 Area 360°)

Adjacent to the 281' Elevation N/A 3

E4.12 Containment Surface Areas -

Surface Area Grid Minimum Wall Thickness Locations Ultrasonic Thickness 0

N/A 4

Note 1: Bolted connections examined per Item Number E1.11 require a General Visual examination each period and a VT-3 visual once per interval and each time the connection is disassembled during a scheduled Item Number E1.11 examination. Additionally, a VT-1 visual examination shall be performed if degradation or flaws are identified during the VT-3 visual examination. These modifications are required by 10 CFR 50.55a(b)(2)(ix)(G) and 10 CFR 50.55a(b)(2)(ix)(H).

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 29 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 Note 2: Item Number E1.12 requires VT-3 visual examination in lieu of General Visual examination, as modified by 10 CFR 50.55a(b)(2)(ix)(G).

Note 3: Item Number E4.11 requires VT-1 visual examination in lieu of Detailed Visual examination, as modified by 10 CFR 50.55a(b)(2)(ix)(G).

Note 4: The ultrasonic examination acceptance standard specified in paragraph IWE-3511.3 for CISI Class MC pressure-retaining components must also be applied to metallic liners of CISI Class CC pressure-retaining components, as modified by 10 CFR 50.55a(b)(2)(ix)(I).

Table 3.3.6 IWL Component Examination Table Examination Category (with Examination Category Description)

Item Number Description Exam Requirements Total Number of Components by System Relief Request /

TAP Number Notes L-A Concrete Surfaces L1.11 Concrete Surfaces - All Accessible Surface Areas General Visual 3 Areas/Zones Containment Dome, Walls and Basemat N/A 5

L1.12 Concrete Surfaces - Suspect Areas (Suspect Areas are identified during examination of items in L1.11 and as follow-up examinations from previous surveillance report)

Detailed Visual Areas Identified From Above N/A 5

L-B Unbonded Post-Tensioning System L2.10 Tendon Physical IWL-2522 643 N/A 5

L2.20 Tendon - Wire or Strand Visual and Physical IWL-2523.2 1 Each Type Per Selection N/A 5

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 30 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 Table 3.3.6 IWL Component Examination Table Examination Category (with Examination Category Description)

Item Number Description Exam Requirements Total Number of Components by System Relief Request /

TAP Number Notes L2.30 Tendon - Anchorage Hardware and Surrounding Concrete Detailed Visual 643 x 2 = 1286 N/A 5

L2.40 Tendon - Corrosion Protection Medium Physical IWL-2525.2(a) 643 x 2 = 1286 N/A 5

L2.50 Tendon - Free Water Physical IWL-2525.2(b)

(Note IWL-2524.2)

If Present N/A 5

Note 5: IWL components are scheduled through random selection prior to the performance of surveillance activities.

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 31 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 Table 3.3.6 TMI-1 4th CISI Interval Selection Component Summary -

Category E-A Components[DH1]

Category Item No.

No. of Comp.

Total Selected Period 1

Period 2

Period 3

E-A E1.11 5

5 5

5 5

E-A E1.11B 33 33 24 42 33 E-A E1.12 2

2 0

0 2

E-A E1.30 1

1 1

1 1

Notes and Comments E-A:

E1.11 (1)

This selection represents the Second Interval for the CISI Program, which is a different interval cycle than the ISI Program (Fourth Interval).

E-A: E1.11B (1)

Bolted connections require VT-3 visual examination followed by VT-1 visual examination if degradation or flaws are identified, as modified by 10 CFR 50.55a(b)(2)(ix)(G) and 10 CFR 50.55a(B)(2)(ix)(H). (Note that the VT-3 visual examinations of the pressure-retaining bolted connections are only required to be conducted once each interval.)

(2)

This selection represents the Second Interval for the CISI Program, which is a different interval cycle than the ISI Program (Fourth Interval).

E-A:

E1.12 (1)

These components require a VT-3 visual examination in lieu of General Visual examination, as modified by 10 CFR 50.55a(b)(2)(ix)(G).

(2)

This selection represents the Second Interval for the CISI Program, which is a different interval cycle than the ISI Program (Fourth Interval).

E-A E1.30 (1)

This selection represents the Second Interval for the CISI Program, which is a different interval cycle than the ISI Program (Fourth Interval).

Table 3.3.6 TMI-1 4th CISI Interval Selection Component Summary -

Category E-C Components Category Item No.

No. of Comp.

Total Selected Period 1

Period 2

Period 3

E-C E4.11 1

1 2

2 2

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 32 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 Notes and Comments E-C:

E4.11 (1)

This Item Number requires VT-1 visual examination in lieu of Detailed Visual Examination, as modified by 10 CFR 50.55a(b)(2)(ix)(G). No significant conditions are currently identified in the Second Interval as requiring application of additional augmented examination requirements under paragraph IWE-1240.

(2)

This selection represents the Second Interval for the CISI Program, which is a different interval cycle than the ISI Program (Fourth Interval).

3.3.7 Results of Recent Containment Inspections 2013 - T1R20 Reactor Building Level 1 Coatings Inspection and Evaluation Coatings that were acceptable upon initial application, but were discovered in good overall condition, with at most, only minor localized discrepancies are considered minor discrepancies. For these minor discrepancies, no further action is required if it is documented in a technical evaluation that reasonable assurance exists that virtually the entire coating will continue to function to adhere under normal or accident conditions.

For coatings considered to have minor discrepancies, localized areas of the coating have been flaked/peeled/chipped away, with no loose coating identified at the defect area. The remainder of the coating is tightly adhered to the surface and reasonable assurance exists that the coating will remain intact under normal and accident conditions.

For coatings issues that are newly identified and are considered beyond minor discrepancies or previously identified coatings issues that have worsened beyond a minor discrepancy, a condition report is generated with an action to repair the coating and the impact on the coatings calculation is degraded. The calculation is updated with impacts, when necessary.

The following items were identified during the 2013 T1R20 inspection:

281 Elevation A D-ring floor - The D-ring floor was in a degraded condition and required repair.

This was previously identified by the T1R18 inspection and is already included in the coatings calculation. Repairing this floor during a refueling outage proves difficult due to outage traffic in the area.

Missing coating on inside wall of the A D-ring at 0 degrees azimuth - This coating is missing and the remaining coating is deemed intact and reasonable assurance exists that it will continue to adhere under normal or accident conditions.

Conservative estimates for exposed concrete area included in the coatings calculation. Repair should be performed in a future outage.

Various 8-inch support columns in the D-ring basement - These locations were first identified by the T1R18 inspection. The remaining coatings are deemed intact and

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 33 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 reasonable assurance exists that they will continue to adhere under normal or accident conditions. Repair should be performed in a future outage.

Ductwork supports all along the interior A D-ring wall - This newly identified coating degradation during T1R20 included approximately 10 square feet of coating on the ductwork. Issue Report (IR) 1589028 was issued documenting the degraded coating. Repair should be performed in a future outage.

Corrosion on AH-E-4A - The minor pinpoint rust on this coating has not worsened since the T1R19 inspection. The remaining coating is tightly adhered and reasonable assurance exists that the coating will remain intact under normal and accident conditions. The coating condition should be monitored and repair should be performed in a future outage if deemed necessary.

Corrosion on AH-E-4B - The minor pinpoint rust on this coating has not worsened since the T1R19 inspection. The remaining coating is tightly adhered and reasonable assurance exists that the coating will remain intact under normal and accident conditions. The coating condition should be monitored and repair should be performed in a future outage if deemed necessary.

4-inch conduit to Junction Box 19 along the Reactor Building wall at A stairs at Elevation 300 - This degraded coating was scraped in T1R18. The remaining coating condition has worsened in that flaking and peeling is continuing along the edges of the scraped area. IR 1589028 documents the degraded condition. The amount of degraded coatings was estimated to be approximately 10 square feet and was added to the coatings calculation. Repair should be performed in a future outage.

Liner at 292, 0-degree azimuth - This liner coating degradation is limited to the topcoat only, the primer is still intact. This degradation was repaired in T1R20.

Liner at 290 and 90-degree azimuth - This liner coating degradation was limited to the topcoat only; the primer was still intact. This degradation was repaired in T1R20.

Liner at 294 and 100-degree azimuth - This liner coating degradation was limited to the topcoat only; the primer was still intact. This degradation was repaired in T1R20.

Liner at 298 and 115-degree azimuth - This liner coating degradation was limited to the topcoat only; the primer was still intact. This degradation was repaired in T1R20.

381 Elevation Flaking and peeling coatings on the B stairwell, Elevation 308 - This degraded coating condition includes approximately 5 square feet of degraded coatings along the 308 floor support steel along the east side of the B stairwell. The additional degraded coating was added to the coatings calculation. Repair should be performed in a future outage.

Equipment Hatch - This general corrosion on the interior surface was first identified during the T1R18 CISI inspections. The remaining coating is tightly adhered to the surface and reasonable assurance exists that the coating will remain intact under normal and accident conditions. The corrosion appears to be surface metal corrosion with no significant metal loss. The coating on the hatch needs to be repaired in a future outage before the metal degradation becomes significant.

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 34 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 Scheduling of this work proves difficult as the hatch area is a primary access point during the outage for personnel and equipment.

Corrosion on the 54-inch purge supply duct, Elevation 308, azimuth 340 degrees -

This is minor corrosion along the base of the purge supply duct. There is reasonable assurance that the coating will remain intact under normal and accident conditions.

This should be considered for repair during future outages.

Repair on pipe penetration No. 102, Elevation 308 - This appears to be surface corrosion on the piping penetration. This was repaired in T1R20.

Penetration No. 413, Elevation 308 - This is minor surface corrosion on the piping penetration. There is no significant metal loss and there is reasonable assurance that the coating will remain intact under normal and accident conditions. This should be considered for repair during future outages.

RB-V-3A Valve, Elevation 308 - This is minor surface corrosion on the valve and piping. There is no significant metal loss and there is reasonable assurance that the coating will remain intact under normal and accident conditions. This should be considered for repair in future outages.

Elbow near RR-V-25A, Elevation 308 - This corrosion was limited to the surface of the piping. This was repaired during T1R20.

Penetration No. 410 near RR-V-23A, Elevation 308 - This surface corrosion on the piping penetration was repaired during T1R20.

8-inch cooling water piping from AH-E-1B - This degradation included flaking and peeling of approximately 15 square feet of coating on the piping. The degraded coating was added to the coatings calculation. Repair should be performed in a future outage.

Penetration No. 421 at 332 and 341-degree azimuth - This was identified by the T1R20 ISI inspection. The additional 2 square feet of degraded coatings identified was added to the coatings calculation. Repair should be performed in a future outage.

331 Elevation RC-P-1C prior to coating repair, Elevation 331 - This coating degradation on the RC-P-1C component cooling piping was previously identified and repaired during T1R20.

RC-P-1C flange joint prior to coating repair, Elevation 331 - This coating degradation on the RC-P-1C component cooling piping was previously identified and repaired during T1R20.

360 Elevation and Above Liner at Elevation 360, azimuth 200-230 degrees - The liner coating degradation appeared to be limited to the topcoat, with the primer appearing to be intact with no evidence of liner metal degradation. This was approximately 5 square feet of degraded coating. The additional degraded coating was added to the coatings calculation. Repair should be performed in a future outage.

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 35 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 Liner at A stairs, Elevation 365 - This was previously identified and repaired during T1R20.

Liner at 375, 0-degree azimuth - The liner coating degradation appeared to be limited to the topcoat, with the primer appearing to be intact with no evidence of liner metal degradation. This was approximately 10 square feet of degraded coatings.

The additional degraded coating was added to the coatings calculation. Repair should be performed in a future outage.

The degraded items identified above that were considered to be more than minor discrepancies were added to the tracking inventory of the degraded qualified coatings calculation. A total of 62 square feet of degraded coatings was added to the calculation as a result of this inspection.

2015 - T1R21 Unit 1 Containment Safety-Related Coating Assessment and Coating Repair An assessment of the Service Level I primary containment coated surfaces was performed at TMI-1 during the 2015 T1R21 refueling outage in accordance with ASTM D4537 (Reference 16) and ANSI/ASME N45.2.6 (Reference 17). National Association of Corrosion Engineers (NACE) Level III certified coatings inspectors performed the assessment. The general scope of the assessment and coating work performed inside of the Unit 1 Primary Containment included:

Qualitative inspection of accessible surfaces to assess the current condition of the protective coatings applied to areas of primary containment pressure boundary, structural steel, stairways and landings, piping, tanks, systems and components (valves, vessels, pumps), concrete walls, concrete floors and miscellaneous equipment.

Visual inspection and evaluation of coating deficiencies identified during previous outages.

Visual inspection and evaluation of new coating deficiencies.

Visual inspection of exposed substrates (if any) both steel and concrete to assess corrosion and spalling concrete if present.

Photographic documentation of representative conditions.

Assure all areas of identified loose flaking coatings are removed back to sound, tightly adherent coating. Loose coatings that are inaccessible will be reported to the coating engineer for future remediation and monitoring. This is required to limit the possibility of loose coatings completely separating from the liner wall, floors, and other components, thereby potentially clogging and restricting drains/strainer flow.

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 36 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 Inspection Findings The overall coating system on the interior surfaces of primary containment wall appears to be generally in good condition. Small (1 square inch) to large (greater than 1 square foot) sized, localized to random frequency of coating defects were identified on the liner wall, components, pipes, miscellaneous structures, and concrete floors and walls. Coating deficiencies include: mechanical damage, burnt coatings, cracked/flaking coating and delamination. Numerous areas of rust staining and coating discoloration were identified.

The majority of coating deficiencies were the result of mechanical damage. This was not age-related degradation; however, there are still significant coating issues that should be monitored and remediated as deemed necessary by the site coating engineer.

Basement Elevation 281 Inside and Outside of D-Ring The coating in the basement at Elevation 281, including D-Ring (A & B) appear to be generally in fair to good condition. Small (1 square inch) to medium (less than 1 square foot) sized, localized to random frequency coating defects were identified on the liner wall, pipes, miscellaneous structures, handrails and stair landings, concrete floors and walls.

Coating defects included: mechanical damage, burnt coatings, delamination and flaking.

Exposed carbon steel areas exhibit light uniform surface corrosion. No evidence of pitting corrosion was observed on aforementioned bare substrate areas.

The majority of coating defects (mechanical damage) were located on the basement floor at Elevation 281. Numerous gouges were identified in the concrete coating on the floor.

Significant coatings deficiencies were photographed and placed on a tracking log for future monitoring and remediation. Gouges in the concrete appeared to be the result of mechanical damage caused by equipment movement and scaffolding impact. Coating deficiencies identified during previous inspections were revisited and do not appear to have changed significantly. Areas of loose coating that can be accessed without erecting scaffolding were systematically removed backed to tightly adhered coating. The majority of coating on the D-Ring floor was in good condition. However, some areas exhibited mechanical damage and delamination to substrate. It appeared that the remaining coating on the floors was tightly adhered and would continue protecting the substrate.

Elevation 308 The coating on Elevation 308 appeared to be in fair to generally good condition, especially the liner wall. Small (1 square inch) to medium (less than 1 square foot) sized, localized coating defects were identified on the liner wall, components pipes, miscellaneous structures, and concrete floors and walls. Coating defects included:

mechanical damage, burnt coatings, flaking coating and delamination. No evidence of pitting corrosion was observed. Coating repairs that were performed during previous outages were found to be in good condition. Coating defects on steel and concrete were documented and prioritized for future monitoring and remediation.

Elevation 346 The coating system at Elevation 346 appeared to be generally in good condition.

Small (1 square inch) to large (greater than 1 square foot) sized, localized to random frequency coating defects were identified on the liner wall, components pipes,

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 37 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 miscellaneous structures and concrete walls and floors. Coating defects included:

mechanical damage caused by equipment and material handling, burnt coatings caused by grinding and welding operation, cracking and flaking coating. Exposed carbon steel areas exhibited light surface corrosion, however, no evidence of pitting corrosion was observed. Coating deficiencies were documented and prioritized for future monitoring and remediation.

Containment Dome and Down Liner Wall to Elevation 365 The containment dome and down the liner wall to Elevation 365 was inspected remotely from the floor and atop the D-Rings by the site Non-Destructive Examination (NDE) department and coatings inspectors. Flashlights and binoculars were used to conduct the inspection. The coating appeared to be in good condition. However, several localized areas of cracked coating were identified, photographed and documented. The subject coating deficiencies were located in previously repaired areas of the liner wall. The coating was lifting and rolling back in the area of disbondment; substrate was not exposed. The coating deficiencies were not removed back to tightly adhered coating because they were not accessible without erecting scaffolding.

Conclusions The overall coating system inside the primary containment building was in good condition.

The coating assessment identified areas of coating degradation that should be repaired to mitigate corrosion. Typical degraded areas have been identified above. No current coating conditions were observed that could impact structural integrity, plant operations, or safe shutdown. There was a significant amount of mechanical damage to the concrete floors throughout the containment building.

2017 - T1R22 Unit 1 Containment Safety-Related Coating Assessment and Coating Repair An assessment of the Service Level I primary containment coated surfaces was performed at TMI-1 during the 2017 T1R22 refueling outage. An ASTM D4537 and ANSI/ASME N45.2.6 qualified and certified inspector and a NACE Level III certified coatings inspector performed the assessment. The on-site coatings engineer assisted with the walk down inspection, and provided background information for areas and components that were inspected. The general scope of the assessment performed inside the Unit 1 Primary Containment included:

Visual inspection of accessible surfaces to assess the current condition of the protective coatings applied to areas of primary containment pressure boundary, structural steel, stairways and landings, piping, tanks, systems and components (valves, vessels, pumps), concrete walls, concrete floors and miscellaneous equipment.

Visual inspection and assessment of degraded coating areas identified and documented during previous outages.

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 38 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 Visual inspection and assessment of new degraded coating areas to identify and document coating deficiencies.

Visual inspection of exposed substrates (if any) both steel and concrete to assess corrosion and spalling concrete, if present.

Photographic documentation of representative degraded coating conditions.

Identified loose flaking coatings; verify they are removed back to sound tightly adherent coating. Loose coatings that are inaccessible will be reported to the coating engineer for future remediation and monitoring. This is required to limit the possibility of loose coatings completely separating from the liner wall, floors and other components; thereby, potentially clogging and restricting drains/strainer flow.

Inspection Findings The protective coating system on interior surfaces of the primary containment wall appears to be generally in good condition and has not significantly changed since the previous T1R21 inspection, performed in 2015.

Localized to random degraded coating areas, small (1 square inch) to large (greater than 1 square foot) in size were identified on the liner wall, components, pipes, miscellaneous structures, and concrete floors and walls. Specific coating defects include: mechanical damage, stress cracking, burnt coatings, cracked/flaked coating and delamination.

Numerous areas of rust staining and coating discoloration were also identified.

The majority of coating deficiencies were the result of mechanical damage. This was not age-related degradation; however, there are still significant coating issues that should be monitored and remediated as deemed necessary by the site coating engineer.

Basement Elevation 281 Inside and Outside of D-Ring The protective coating in the basement at Elevation 281, including D-Ring (A & B) are generally in fair to good condition. Localized to random degraded areas, small (1 square inch) to medium (less than 1 square foot) in size were identified on the liner wall, pipes, miscellaneous structures, handrails and stair landings, concrete floors and walls. Specific coating defects include: mechanical damage, stress cracking of concrete coating, burnt coatings, delamination and flaking. Exposed carbon steel areas exhibit light uniform surface corrosion. No evidence of pitting corrosion was observed on aforementioned bare substrate areas.

The majority of coating defects (mechanical damage) are located on the basement floor at Elevation 281. Numerous gouges were identified in the concrete coating on the floor.

Significant coating deficiencies were photographed and placed on a tracking log for future monitoring and remediation. Gouges in concrete coating appeared to be the result of mechanical damage caused by tool, equipment and scaffolding staging impact.

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 39 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 Coating deficiencies identified during previous inspections were re-inspected and do not appear to have changed significantly. Areas of loose coating that can be accessed without erecting scaffolding is systematically removed back to tightly adhered coating during each outage.

The majority of coating on the D-Ring floor is in fair to good condition. However, some degraded areas exhibit mechanical damage and delamination to substrate. It appeared that the remaining coating on the floors was tightly adhered and would continue protecting the substrate.

Elevation 308 The protective coating on Elevation 308 is generally in good to fair condition.

Localized degraded areas, small (1 square inch) to medium (less than 1 square foot) sized, were identified on the liner wall, components, pipes, miscellaneous structures, and concrete floors and walls. Coating defects include: mechanical damage, burnt coatings, flaking coating and delamination. No evidence of pitting corrosion was observed. Coating repairs that were performed during previous outages are in good condition. Coating defects on steel and concrete were documented and prioritized for future monitoring and remediation.

Elevation 346 The coating system at Elevation 346 appeared to be generally in good condition.

Localized to random degraded areas, small (1 square inch) to large (greater than 1 square foot) in size were identified on the liner wall, components, pipes, miscellaneous structures, concrete walls and floors. Coating defects included mechanical damage caused by equipment and material handling, burnt coatings caused by grinding and welding operation, cracking and flaking coating. Exposed carbon steel areas exhibited light surface corrosion; however, no evidence of pitting corrosion was observed. Coating deficiencies were documented and provided to on-site coating engineer for future monitoring and remediation.

Containment Dome (and liner wall down to Elevation 365)

The containment dome and the liner wall down to Elevation 365 was inspected remotely from the floor and top the D-Rings by the site coating engineer and contractor coatings inspectors. High power flashlights were used to conduct the inspection. The protective coating appeared to be in good condition. However, several localized areas of cracked coating were observed, photographed and documented. The subject coating deficiencies were located in previously repaired areas of the liner wall. The coating was lifted and rolling back in the area of disbondment; the substrate was not exposed. Coating deficiencies were not removed back to tightly adhered coating because they were not accessible without erecting scaffolding.

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 40 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 Conclusions and Recommendations The protective coating system inside the primary containment building was generally in good condition. The coating assessment identified few additional areas of coating degradation that should be repaired to mitigate corrosion. No coating conditions were observed that could impact structural integrity, plant operations, or safe shutdown. There is a significant amount of mechanical damage to the concrete floors throughout the Reactor Building.

Elevation 281 displayed the largest amount of mechanical damage to the coating.

Coating repair work was initiated and should continue to be planned and scheduled prior to future outages, to address the areas of coating degradation identified during 1R22 and previous outages.

2009 - T1R18 Refueling Outage IWE Examination During scheduled containment visual examinations of the Reactor Building containment liner, the following conditions were identified:

Liner Courses 1, 2 and 3 Elevations 281 to 305:

Liner Elevation 281 at 275 degrees azimuth: Blistering, corrosion/pitting. Minor corrosion with no material degradation noted. This area is scheduled to be recoated in a future outage.

Liner Elevation 281 at 300 degrees azimuth: Corrosion/Pitting, missing paint/coating. The loose coating/paint was removed and recoated during T1R18.

Liner Elevation 285 at 140 degrees azimuth: Concrete noted on surface of liner.

This indication was determined to be acceptable as-is based on the small amount of concrete material observed.

Liner Elevation 294 at 170 degrees azimuth: Blistering, corrosion/pitting, and peeling. Minor corrosion with no material degradation noted. The loose coating was removed. The area is scheduled to be recoated in a future outage.

Liner Elevation 294 at 175 degrees azimuth: Corrosion/Pitting, missing paint/coating, blistering. Minor corrosion with no material degradation noted. The loose coating was removed. The area is scheduled to be recoated in a future outage.

Liner Elevation 289 at 170 degrees azimuth: Corrosion/Pitting, missing paint/coating. Minor corrosion with no material degradation noted. The loose coating was removed. The area is scheduled to be recoated in a future outage.

Liner Elevation 290 at 5 degrees azimuth: Corrosion/Pitting, missing paint/coating.

Minor corrosion with no material degradation noted. The loose coating was removed. The area is scheduled to be recoated in a future outage.

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 41 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 Liner Elevation 294 at 120 degrees azimuth: Corrosion/Pitting, missing paint/coating. Minor corrosion with no material degradation noted. The estimated coating area was added to the degraded coating calculation since the loose coating was not removed. The area is scheduled to be cleaned and recoated in a future outage.

Liner Elevation 294 at 235 degrees azimuth: Corrosion/Pitting, missing paint/coating. Minor corrosion with no material degradation noted. The loose coating was removed. The area is scheduled to be recoated in a future outage.

Liner Courses 8, 9, 10, 11, 12, 13, 14, 15, 16 and 17 Elevations 346 to 436:

Liner Elevation 346 at 240 degrees azimuth: Coating Damage/Peeling. The loose coating was removed. The area is scheduled to be recoated in a future outage.

Liner Elevation 360 at 205 degrees azimuth: Blistering, peeling, and coating damage.

The loose coating was removed. The area is scheduled to be recoated in a future outage.

Containment Exterior Penetrations:

Penetration 202E: Paint/coating removed on first weld from penetration. Minor rust present. No metal degradation noted. The condition was determined to be acceptable for material loss and is scheduled to be recoated in a future outage.

Penetration 407: Corrosion - minor surface rust. No metal degradation. Missing paint/coating. The condition was determined to be acceptable for material loss and is scheduled to be recoated in a future outage.

Penetration 409: Corrosion - minor surface rust. No metal degradation. Missing paint/ coating. The condition was determined to be acceptable for material loss and is scheduled to be recoated in a future outage.

Penetration 410: Corrosion - minor surface rust. No metal degradation. Missing paint/ coating. The condition was determined to be acceptable for material loss and is scheduled to be recoated in a future outage.

Containment Internal Penetrations Elevation 281 to Elevation 305 Penetration 202E: Recordable Indications Corrosion/Pitting/Missing Paint/Coating.

Medium rust with <5% material degradation. Loose coating was removed. Condition was determined to be acceptable for material loss and will be recoated in a future outage.

Penetration 206E: Recordable Indications Corrosion/Pitting/Missing Paint/Coating.

Medium Corrosion with no degradation of the material. Condition was determined to be acceptable for material loss and will be recoated in a future outage.

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 42 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 Penetration 311E: Recordable Indications Corrosion/Pitting/Missing Paint/Coating.

Minor rust with no degradation of material. Loose coating was removed. Condition was determined to be acceptable for material loss and will be recoated in a future outage.

Penetration 312E: Recordable Indications Corrosion/Pitting/Missing Paint/Coating.

Minor corrosion with no degradation of material for all penetrations listed. Loose coating was removed. Condition was determined to be acceptable for material loss and will be recoated in a future outage.

Penetration 313E: Recordable Indications Corrosion/Pitting/Missing Paint/Coating.

Minor corrosion with no degradation of material for all penetrations listed. Condition was determined to be acceptable for material loss and will be recoated in a future outage.

Penetration 314E: Recordable Indications Corrosion/Pitting/Missing Paint/Coating.

Minor corrosion with no degradation of material for all penetrations listed. Loose coating was removed. Condition was determined to be acceptable for material loss and will be recoated in a future outage.

Penetration 323E: Recordable Indications Corrosion/Pitting/Missing Paint/Coating.

Minor rust with no degradation of material. Loose coating was removed. Condition was determined to be acceptable for material loss and will be recoated in a future outage.

Containment Internal Penetrations 305 to 326:

Penetration 102: Corrosion/Pitting, missing paint/corrosion. Minor corrosion with no material degradation noted. Loose coating was removed. Condition was determined to be acceptable for material loss and will be recoated in a future outage.

Penetration 213: Corrosion/Pitting, missing paint/corrosion. Material loss no greater than 1/32 noted. Condition was determined to be acceptable for material loss and will be recoated in a future outage.

Penetration 217E: Corrosion/Pitting, missing paint/corrosion. Minor corrosion with no material degradation noted. Loose coating was removed. Condition was determined to be acceptable for material loss and will be recoated in a future outage.

Penetration 226E: Corrosion/Pitting, missing paint/corrosion. Minor corrosion with no material degradation noted. Loose coating was removed. Condition was determined to be acceptable for material loss and will be recoated in a future outage.

Penetration 233E: Corrosion/Pitting, missing paint/corrosion. Minor corrosion with no material degradation noted. Loose coating was removed. Condition was determined to be acceptable for material loss and will be recoated in a future outage.

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 43 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 Penetration 337: Corrosion/Pitting, missing paint/coating, and blistering. Minor corrosion with no material degradation noted. Loose coating was removed.

Condition was determined to be acceptable for material loss and will be recoated in a future outage.

Penetration 338: Corrosion/Pitting, missing paint/coating, blistering, peeling. Minor corrosion with no material degradation noted. Loose coating was removed.

Condition was determined to be acceptable for material loss and will be recoated in a future outage.

Penetration 341E: Corrosion/Pitting, missing paint/corrosion. Minor corrosion with no material degradation noted. Condition was determined to be acceptable for material loss and will be recoated in a future outage.

Penetration 342E: Corrosion/Pitting, missing paint/corrosion. Minor corrosion with no material degradation noted. Condition was determined to be acceptable for material loss and will be recoated in a future outage.

Penetration 343E: Corrosion/Pitting, missing paint/corrosion. Minor corrosion with no material degradation noted. Loose coating was removed. Condition was determined to be acceptable for material loss and will be recoated in a future outage.

Penetration 344E: Corrosion/Pitting, missing paint/corrosion. Minor corrosion with no material degradation noted. Loose coating was removed. Condition was determined to be acceptable for material loss and will be recoated in a future outage.

Penetration 410: Corrosion/Pitting, missing paint/corrosion. Minor corrosion with no material degradation noted. Condition was determined to be acceptable for material loss and will be recoated in a future outage.

Penetration 413: Corrosion/Pitting, missing paint/corrosion. Minor corrosion with no material degradation noted. Loose coating was removed. Condition was determined to be acceptable for material loss and will be recoated in a future outage.

Penetration 418: Corrosion/Pitting, missing paint/corrosion. Pitting less than 1/32 in 3 wide by 6 long hatch annulus. Minor corrosion with no material degradation noted. Loose coating was removed. Condition was determined to be acceptable for material loss and will be recoated in a future outage.

Repair to the containment liner as part of the LR commitment is discussed in Section 3.8.2.

2013 - T1R20 Refueling Outage IWE Examination During scheduled containment, visual examination of the Reactor Building containment liner, the following conditions were identified and need evaluation:

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 44 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 Liner Elevation 292' at 0 degrees azimuth: Paint degradation and chipping with some light rust. This coating degradation was the topcoat only, with no liner metal degradation. This area was recoated during T1R20.

Liner Elevation 290 at 90 degrees azimuth: Paint degradation and rusting. This coating degradation was the topcoat only, with no liner metal degradation. This area was recoated during T1R20.

Liner Elevation 294 at 100 degrees azimuth: Paint blistering and cracking. Upon further inspection, this coating degradation was the topcoat only, with no liner metal degradation. This area was recoated during T1R20.

Liner Elevation 298 at 115 degrees azimuth: Paint degradation and rusting. This coating degradation was the topcoat only, with no liner metal degradation. This area was recoated during T1R20.

Liner Elevation 295 at 155 degrees azimuth: Rusting on Penetration 314. This is a 12-inch penetration for low-level process instrumentation. This coating degradation appears to be surface rusting on the penetration, with the remaining coating tightly adhered. This will continue to be monitored by the coatings program for possible future repair.

Liner Elevation 298 at 210 degrees azimuth: Rusting on Penetration 323S. This is a 12-inch spare piping penetration. There is minor surface rust with no significant coating degradation. This will continue to be monitored by the coatings program for possible future repairs.

Liner Elevation 290 at 250 degrees azimuth: Staining and light rust on Penetration 301. This is the 16-inch penetration for the 8-inch reactor building spray supply "A".

There is minor staining and light surface rust on the penetration with no appreciable coatings degradation. This will continue to be monitored by the coatings program for possible future repair.

Liner Elevation 298 at 250 degrees azimuth: Coating not present, rusting on Penetration 315. This is a 12-inch penetration for low voltage power control. The degradation appears to be surface rust and the coating is missing, therefore it does not affect the degraded coatings calculation. This will continue to be monitored by the coatings program for possible future repair.

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 45 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 Penetration 322: Coating missing. This is a 12-inch penetration for a 2-inch HPI line to reactor coolant pump "B" at 299' and 214-degree azimuth. The minor surface rusting on the penetration has no impact on the HPI line itself. The remaining coating is tightly adhered and does not impact the degraded coatings calculation.

This will continue to be monitored by the coatings program for possible future repairs.

Penetration 419E: Missing coating from 11 o'clock to 3 o'clock position. Exposed area is 12 inches wide and rusted. This is a 38-inch penetration for a 24-inch main steam line from steam generator "B" at 337' and 359-degree azimuth. The coating is missing; therefore, it does not affect the degraded coatings calculation. The rust appears to be on the surface with no appreciable metal loss. The coating repair was deferred and is being evaluated in the outage scope process for the next opportunity of repair.

Penetration 413E: Missing coatings, exposed areas are rusted. This is the 2-inch steam generator cavity drain at 317' and 359-degree azimuth. The rust appears to be on the surface and the remaining coating is tightly adhered. There is no impact to the degraded coatings calculation. This minor coating discrepancy will be tracked by the coatings program for possible future repair.

Penetration 237E: Blistered coating at 9 o'clock position, approximately 1.25-inch area exposed and rusted. This is a 19-inch penetration for the "B" reactor coolant pump power supply at 338' and 114-degree azimuth. The extremely small area is considered a negligible impact to the degraded coatings calculation. This will be tracked by the coatings program for future inspections.

Penetration 236E: Blistered coating at 8 o'clock position, 0.5 inch by 2-inch area exposed and rusted. This is a 19-inch penetration for the "B" reactor coolant pump power supply at 338' and 114-degree azimuth. The extremely small area is considered a negligible impact to the degraded coatings calculation. This will be tracked by the coatings program for future inspections.

Penetration 114E: Minor staining observed. This is a 24-inch main steam line from steam generator "B" at 337' and 7-degree azimuth. This degradation is considered minor with no impact to the protective function of the coating. This will continue to be monitored as part of the coating program inspections.

Penetration 407E: Missing coating (2 bands), exposed areas are rusted. This is a 12-inch emergency cooling coil water line at 314' and 343-degree azimuth. The remaining coating appears tightly adhered and does not impact the degraded coatings calculation. The rust appears to be surface rust with no significant metal degradation. The coating was re-inspected during 1R21 and found to be acceptable.

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 46 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 Penetration 410E: Missing coating, exposed area is rusted. This is a 12-inch emergency cooling coil water line at 317' and 343-degree azimuth. The remaining coating appears tightly adhered and does not impact the degraded coatings calculation. The rust appears to be surface rust with no significant metal degradation. The coating was re-inspected during 1R21 and found to be acceptable.

Penetration 421E: Missing (blistered) coating, exposed area is rusted. This is an 8-inch reactor building normal cooling water supply line at 332' and 341-degree azimuth. The remaining coating appears tightly adhered and does not impact the degraded coatings calculation. The rust appears to be surface rust with no significant metal degradation. The coating repair was deferred and is being evaluated in the outage scope process for the next opportunity of repair.

Penetration 217: Paint degradation (bare metal exposed), signs of corrosion. This is a 12-inch penetration for miscellaneous thermocouples at 324' and 127-degree azimuth. The remaining coating appears tightly adhered and does not impact the degraded coatings calculation. The rust appears to be surface rust with no significant metal degradation. The coating repair was deferred and is being evaluated in the outage scope process for the next opportunity of repair.

Penetration 227: Paint removed, completely exposed bare metal with corrosion.

Appears smooth and prepped to paint. This is surface rust on the 38-inch penetration sleeve for the 20-inch main feedwater line to steam generator "B" at 329' and 103-degree azimuth. There is no impact to the degraded coatings calculation.

There appears to be no significant metal degradation. The coating repair was deferred and is being evaluated in the outage scope process for the next opportunity of repair.

Penetration 218E: Paint degradation with bare metal exposed, signs of corrosion.

This is a 12-inch penetration for miscellaneous thermocouples at 324' and 127-degree azimuth. The remaining coating appears tightly adhered and does not impact the degraded coatings calculation. The rust appears to be surface rust with no significant metal degradation. The coating repair was deferred and is being evaluated in the outage scope process for the next opportunity of repair.

Penetration 226: Paint degradation with bare metal exposed, signs of corrosion.

This is a 12-inch electrical penetration at 328' and 92-degree azimuth. The remaining coating appears tightly adhered and does not impact the degraded coatings calculation. The rust appears to be surface rust with no significant metal degradation. The coating repair was deferred and is being evaluated in the outage scope process for the next opportunity of repair.

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 47 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 Penetration 223E: Paint degradation with bare metal exposed, slight corrosion. This is a 12-inch penetration for the CRDM power supply at 335' and 100-degree azimuth.

The remaining coating appears tightly adhered and does not impact the degraded coatings calculation. The rust appears to be surface rust with no significant metal degradation. The coating repair was deferred and is being evaluated in the outage scope process for the next opportunity of repair.

Penetration 112: Paint degradation with bare metal exposed, 50% coverage. This is the 38-inch penetration sleeve for the 24-inch main steam line from steam generator "A" at 337' and 41-degree azimuth. There is no impact to the degraded coatings calculation. There appears to be no significant metal degradation. The coating repair was deferred and is being evaluated in the outage scope process for the next opportunity of repair.

Penetration 421: Blistered paint with signs of corrosion, 50% coverage. This is an 8"reactor building cooling water supply penetration at 332' and 338-degree azimuth.

The coating appears blistered/degraded on about 50% of the pipe over a length of approximately 1 foot. The surface area of degraded coatings is estimated to be approximately 2 square feet. The degraded coatings calculation will be updated to include this additional amount. Sufficient margin exists in the calculation to handle this additional amount of degraded coatings. This area was recoated during T1R21.

Penetration 202-E: 2-inch band of missing coating with surface rust. This is a 12-inch penetration for nuclear instrumentation at 294' and 176-degree azimuth. The remaining coating is tightly adhered and the rust appears to be on the surface. This will be monitored by the coatings program for future repair.

Penetration 206-E: 2-inch band of missing coating with surface rust. This is a 12-inch penetration for reactor building cooling fan power at 298' and 172 degrees azimuth. The remaining coating is tightly adhered and the rust appears to be on the surface. This will be monitored by the coatings program for future repair.

Penetration 312-E: 2-inch band of missing coating with surface rust. This is a 12-inch penetration for nuclear instrumentation at 294' and 183-degree azimuth. The remaining coating is tightly adhered and the rust appears to be on the surface. This will be monitored by the coatings program for future repair.

Penetration 338-I: Blistered and flaking coating, exposed areas are rusting. This is a 12-inch penetration for the 2.5-inch high pressure injection (HPI) line to reactor coolant pump "C" at 317' and 216-degree azimuth. The remaining coating is tightly adhered and the rust appears to be on the surface. This does not impact the function of the HPI line as the rusting is on the penetration itself. This will be monitored by the coatings program for future repair.

Penetration 337-I: Blistered and flaking coating, exposed areas are rusting. This is a 12-inch penetration for the 4-inch reactor coolant pump seal water supply at 317' and 218-degree azimuth. The remaining coating is tightly adhered and the rust

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 48 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 appears to be on the surface. This does not impact the function of the seal injection line as the rusting is on the penetration itself. This will be monitored by the coatings program for future repair.

Penetration 344-I: 2-inch band of missing coating with surface rust. This is a 19-inch penetration for reactor coolant pump "D" power supply at 322' and 228-degree azimuth. The remaining coating is tightly adhered and the rust appears to be on the surface. This will be monitored by the coatings program for future repair.

Penetration 343-I: 2-inch band of missing coating with surface rust. This is a 19-inch penetration for reactor coolant pump "D" power supply at 322' and 232-degree azimuth. The remaining coating is tightly adhered and the rust appears to be on the surface. This will be monitored by the coatings program for future repair.

Liner Course 9, Plate 7: Peeling paint. This is the Reactor Building liner at 360' and 210-degree azimuth. There is approximately 5 square feet of degraded coatings that will need to be added to the degraded coatings calculation. The primer is still intact under the topcoat and liner integrity is not an immediate concern. Sufficient margin is available to handle the additional degraded coatings. This area was recoated during T1R21.

Liner Course 9, Plate 8: Peeling paint. This is the Reactor Building liner at 360' and 230-degree azimuth. There is approximately 5 square feet of degraded coatings that will need to be added to the degraded coatings calculation. The primer is still intact under the topcoat and liner integrity is not an immediate concern. Sufficient margin is available to handle the additional degraded coatings. This area was recoated during T1R21.

2010 - Augmented Reactor Building (IWL) In-Service Inspection The following is a summary of the results of the TMI-1 2010 Augmented Reactor Building (IWL) In-Service Inspection. This augmented in-service inspection was a one-year follow-up inspection after the repair and replacement activities creating and closing a containment opening for the T1R18 Steam Generator Replacement (SGR) Project.

The SGR at TMI-1 commenced in October 2009 with RFO T1R18. During this outage, containment tendons were de-tensioned and/or removed from the containment structure supporting the creation of a construction opening in the containment wall. A rectangular area of concrete was removed from the containment wall and the inner steel containment liner was cut out of the opening. This rectangular opening was located between Buttresses 5 and

6. Prior to the removal of concrete, vertical and hoop tendons, which routed through the opening were cut out and discarded while a selection of tendons adjacent to the opening was de-tensioned.

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 49 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 Following the SGR, the steel liner was repaired, new tendons were installed, and a concrete patch was poured restoring the containment wall. The new tendons and adjacent tendons were re-tensioned after the concrete patch was cured.

The results of the 2010 Augmented Reactor Building Inservice Inspections are summarized below:

Post-Tensioning System The results of the post-tensioning system examination, measurements and tests met prescriptive acceptance criteria with one exception, which was shown to be acceptable by additional testing, examination and evaluation. A listing of specific results follows:

All tendon forces were above the predicted values.

Elongations measured during re-tensioning of de-tensioned tendons were within 10%

of previously measured values.

End anchorage hardware items were free of active corrosion. All anchorage hardware items were free of cracking and distortion.

The as-found button head conditions were as documented during the previous examinations for all examined tendons except for V118. Two button heads were protruding on the shop end of V118. Further investigation revealed the two wires were broken just below the shop anchor head.

The tensile strength and elongation (at failure) of all wire test samples, including the two broken wires from V118, exceeded the minimum required values.

The results of the wire continuity tests on V118 and examination of the adjacent tendons, V117 and V119, found no further damage or broken wires. Evaluation of the condition found V118 acceptable without repair.

Water content, corrosive ion concentration and reserve alkalinity of all corrosion protection medium samples met acceptance criteria.

Concrete adjacent to end anchorages of the surveillance tendons was free of cracks over 0.01 inches wide.

The differences between the quantities of CPM removed and the quantities replaced in all tendons (including V117 and V119) were all within 10% of the net duct volume.

All SGR tendon end anchorage covers were free of damage and none showed signs of significant grease leakage.

Containment Surface The concrete repair patch had no signs of structural degradation and no cracks were found in the patch or the adjoining original concrete after detailed inspection. The VT-1C examination was performed directly from a hanging platform, without the need for optical aids.

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 50 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 The buttress, base mat, and dome trench around the SGR tendon anchorages were free of damage / deterioration except for effectively unchanged previously reported conditions.

Water seepage through the tendon gallery outer wall was unchanged from what was observed during the 35th year surveillance.

Repairs and Follow-Up Examinations Repairs The results of the surveillance show that no repair to either the concrete or post-tensioning system is necessary at the present time. The two broken wires in tendon V118 were evaluated as acceptable without repair.

Follow-Up Examinations during the 40th Year Surveillance There were no new follow-up examinations as a result of this augmented surveillance. The follow-up examinations specified during the 35th year surveillance in shall be completed during the 40th year surveillance.

Conclusions The following conclusions are based on and supported by evaluation of the surveillance results:

The force in each individual sample tendon exceeds the lower acceptance limit (95%

of the predicted value).

The group mean forces exceed the requirements for minimum required pre-stress.

The group mean forces extrapolated exceed the requirements for minimum required pre-stresses at the end of the next surveillance period 2Q15 (3/9/15).

Elongations measured during re-tensioning of de-tensioned tendons are as expected and are within 10% of previously measured values.

All examined wire button heads are seated and meet acceptance criteria (as-left).

The difference between quantities of CPM removed from sample tendons and quantities replaced were all within 10% of net duct volume showing that tendon duct fill was adequate both as-found and as-left.

Corrosion protection medium samples meet specified limits on absorbed water content and concentrations of corrosive ions. The samples also meet the specified lower limit on reserve alkalinity.

No free water was detected at tendon end anchorages; therefore, it is concluded that water intrusion is not a problem.

Tendon wire samples meet the specified lower limits on ultimate strength and elongation at failure.

Concrete surrounding sample tendon bearing plates is free of damage, deterioration, and cracks that exceed 0.010 inches in width.

End anchorage hardware items were free of active corrosion. All anchorage hardware items were free of cracking, distortion, and damage.

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 51 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 Concrete surfaces are free of damage and degradation. Spalling of grout patches, as noted in various areas, has no structural significance. The few concrete cracks that exceed the threshold acceptance criteria of 0.010 inches in width were accepted based on responsible engineer and Authorized Nuclear Inservice Inspector (ANII) review of the data sheets and are of no structural significance.

The SGR opening patch is structurally sound without cracking and is adherent to the adjoining concrete.

The SGR tendon end caps are free of damage and any indications of tendon anchorage failure.

SGR tendon end caps are not leaking CPM to any significance degree and no corrective action is required at this time.

The broken wires identified in tendon V118 were evaluated to be caused by an external event that physically damaged the affected wires. The damage is isolated to the affected wires in V118.

Per evaluation, tendon V118 is acceptable without repair.

Overall, the repair of the containment structure and post-tensioning system from the SGR is successful with no deficiencies.

2013 - 40th Year Reactor Building Tendon Surveillance The 40th Year Surveillance was conducted on site from September to December 2013, with lab tests being performed during January and February of 2014 and the final visual examinations on April 15, 2014. The surveillance, in its entirety, was performed in the examination window between March 2013 and March 2015, time window defined by the earliest start and latest finish dates specified in the governing code, ASME Section XI, Subsection IWL, as cited in 10 CFR 50.55a.

The results of the 40th Year (Period 10) Reactor Building Surveillance are summarized below:

Post-Tensioning System The results of the post-tensioning system examinations, measurements and tests met all prescriptive acceptance criteria except as noted below. The results are summarized as follows:

All tendon forces were above 95% of predicted value. Hoop tendon H13-03 was lift-off tested from only one buttress. This test is valid as the tendon was originally single-end stressed from the tested buttress.

The vertical, hoop, and dome tendon normalized group mean forces were above the minimum required values. As the SGR tendon forces are significantly greater than those documented for other sample tendons, these are, for conservatism, not included in the vertical and hoop tendon group means.

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 52 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 The 95% lower confidence limits on vertical, hoop and dome tendon force trends projected through the latest date for the completion of the 45th year surveillance are above the minimum required forces.

Elongations measured during the re-tensioning of de-tensioned tendons were within 10% of the "as-installed" construction values.

End anchorage hardware items were free of active corrosion, cracking, and distortion. Some tendons (i.e., V84, V115, and V136 tendon gallery bearing plates),

exhibited minor rust, which does not show evidence of progression (not active). A number of tendon gallery bearing plates are frequently wetted by ground water seepage between the gallery walls and the reactor building base mat. This has caused minor corrosion. There is no significant loss of metal, and at the present time, no need to take remedial action other than re-examination during the 45th year surveillance.

As-found button head conditions were, with 3 exceptions, as documented during construction. Two button heads at the Buttress 4 end of H24-15 were found unseated and one button head was found missing at the gallery end of SGR tendon V-136 (all button heads were intact at the top end). These conditions were deemed to have no structural significance and were accepted by evaluation.

The tensile strength and elongation (at failure) of all wire test samples was above the minimum required values.

Water content, corrosive ion (chlorides, nitrates, and sulfides) concentration, and reserve alkalinity (base number) of all corrosion protection medium samples met acceptance criteria, except as noted below.

o 9 CPM samples had base numbers less than the 0.50 reporting threshold and required further tests for their acid number.

o 3 of the 10 acid number tests resulted in acid numbers in excess of the acceptance limit of 1. These 3 samples were accepted by evaluation.

No free water was found at tendon anchorages.

Concrete adjacent to the tendon end anchorages was free of cracks over 0.01 inches wide.

The differences between the quantities of corrosion protection medium removed from sample tendons and the quantities replaced were, with 3 exceptions, within 10% of net duct volume. The 3 exceptions, involving "as-found" under-fill of the dome tendon ducts, were accepted by evaluation.

All tendon end anchorage covers (end caps) were free of damage and, with one exception, free of corrosion. The single exception consisted of light, dry rust covering part of a dome tendon end cap. CPM was observed to be leaking from 6 dome and 3 hoop end caps. All leakage noted was minor (estimated quantities lost on the order of 0.1 liter or less) and was deemed to require no corrective action at the present time.

Several shim pairs were found to have gaps in excess of the specified maximum, and the gaps were restored to an acceptable condition or accepted by evaluation.

Technical evaluations showed that the "as-found" reduced bearing area was

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 53 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 sufficient to carry the maximum expected tendon load, with a bearing stress below the material allowable. These evaluations concluded that no further examination is required at this time.

Containment Surface The Reactor Building concrete surfaces and tendon anchorage bearing plates were generally free of damage and deterioration. Minor corrosion was noted on the V84, V115, and V136 bearing plates in the tendon gallery; a result of ground water seepage between the gallery outer wall and the Reactor Building base mat. As the loss of metal is not significant, remedial action, other than re-examination during the 45th year surveillance, is not considered necessary at this time.

Minor seepage of corrosion protection medium (CPM) through vertical cracks in the lower wall was found to be continuing at a nominal rate unchanged from that observed during the 35th year surveillance. This is expected since most of the CPM in the TMI-1 tendon ducts is of an older formulation that liquefies at relatively low temperatures. Most of the seepage was noted in the area below the equipment access hatch. Ducts in this area curve around the opening, and therefore, are flexible and not fabricated for leak tightness. The seepage, which has no structural significance, is monitored and corrective actions will be taken to top off the vertical tendon CPM levels when seepage rates warrant.

Repairs and Follow-Up Examinations Repairs The results of the surveillance show that no repair to either the concrete or the post-tensioning system is necessary at the present time.

Follow-Up Examination During 45th Surveillance Period (Period 11)

Detailed visual examinations will be performed in the following areas during the 45th Year Surveillance.

The grout overlay of previously exposed reinforcing steel on the vertical face of the ringer girder southeast quadrant for degradation and separation from the underlying concrete.

The tendon gallery ceiling area; including base mat concrete, tendon bearing plates, and tendon end caps for evidence of CPM leakage, the effects of ground water seepage onto concrete and steel items, deterioration of previously documented exposed reinforcing steel, and other damage or deterioration.

The lower wall above the base mat will be examined to determine if corrosion protection medium seepage through the previously documented vertical cracks is increasing.

Tendon anchorage assemblies, noted below, for evidence of corrosion. Pump-through of new CPM in the tendon ducts will be used to correct a possible degraded

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 54 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 CPM condition (test sample acid numbers >1).

o H13-08 Buttress 1 o H35-02 Buttress 3 o V-32 Gallery Conclusions The following conclusions are based on, and supported by, evaluation of the surveillance results.

The force in each individual sample tendon exceeds the lower acceptance limit (95%

of the predicted value); no sample tendon force exceeds the implied upper limit of 74% Guaranteed Minimum Ultimate Tensile Strength (GUTS) (the limit imposed during initial and SGR Project tensioning).

Vertical, hoop and dome sample tendon mean normalized forces are above the minimum required values.

The 95% lower confidence limits on vertical, hoop, and dome tendon force trends are forecast to remain above minimum required levels through March 2020, the deadline for completion of the 45th year surveillance. Additional analysis provided for the Exelon TMI Reactor Building Tendon Relief Request (Reference 18) concluded that the mean force in each of the tendon groups is projected by log-linear regression and 95% confidence limit computations to remain above the specified minimum until well after the March 2034 license expiration. This projection is supported by the low scatter trend of common tendon force data.

Control tendon lift-offs exhibit relatively little scatter; vertical, hoop and dome control tendon trends indicate that forces are decreasing more slowly than predicted.

Elongations measured during the re-tensioning of de-tensioned tendons are all within 10% of previously measured values.

All examined tendon end anchorage hardware is free of active corrosion, cracks and distortion. Observed corrosion is limited to light, dry rust on end caps, bearing plates and shims. No corrosion was found on wires or anchor heads.

With two exceptions, all examined wire button heads are seated. Two wires at the Buttress 4 end of H24-15 were found unseated. This condition, which has no structural significance, was accepted by evaluation and requires no corrective action or further examination.

With one exception, all button heads are intact. One wire was found to be missing at the bottom end of the SGR tendon V-136. This condition, which was no structural significance, was accepted by evaluation and requires no corrective action or further examination.

Several shim pairs were found to have gaps that exceeded the specified limit. These were corrected or evaluated for continued service. All observed gaps were such that bearing area was still sufficient to carry the maximum expected tendon load without generating bearing stresses in excess of the allowable values for the shim material.

No additional examinations are required at the present time.

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 55 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 The difference between the quantities of CPM removed from sample tendons and quantities replaced were, with 3 exceptions, within 10% of the tendon net duct volume. The 3 exceptions, all involving "as-found" under-fill of the dome tendon duct, were determined to be acceptable by evaluation.

Corrosion protection medium samples meet specified limits on water content and concentrations of corrosive ions (chlorides, nitrates, and sulfides). All CPM samples, except for nine (9), met the reserve alkalinity (base number) requirements. An acid number test was conducted on the nine samples, which were below 0.50 reserve alkalinity, and on one (1) additional sample with a base number of 0.52. Three were found with acid numbers greater than the acceptance limit of 1. The three sample failures were accepted by evaluation with follow-up actions during the 45th Year Surveillance.

No free water was detected at tendon end anchorages; showing that water intrusion is not an issue.

Tendon wire samples meet the specified limits on ultimate strength and elongation at failure.

Concrete, surrounding tendon bearing plates, is free of damage, deterioration, and cracks exceeding 0.010 inches in width.

Concrete surfaces are free of damage and degradation. The deterioration of grout patches, as noted in various areas, has no structural significance. The few concrete cracks that exceed the threshold width of 0.010 inches (none exceeded 0.015 inches) were previously documented, accepted by evaluation, and reported as unchanged from previous surveillances.

Tendon end caps are free from damage, and with one exception, free from corrosion.

The corrosion observed consisted of light, dry rust with no noticeable loss of metal.

Conditions in areas identified for detailed visual examination have not changed since the 35th year surveillance in 2009. Detailed visual examination of these areas is recommended again during the 45th Year Surveillance.

2015 - Reactor Building Concrete/Liner Inspection During the 2015 T1R21 refueling outage, a visual inspection of the accessible interior and exterior surfaces of the containment structure and its components was performed to uncover any obvious evidence of deterioration which could affect either containments structural integrity or leak-tightness. The procedure was performed in order to meet the requirements of 10 CFR 50, Appendix J. The inspection is performed on a two-year frequency to coincide with the two-year refueling cycle. The 2015 inspection yielded no findings that challenged the containments structural integrity or leak-tightness.

2017 Reactor Building Concrete/Liner Inspection During the 2017 T1R22 refueling outage, a visual inspection of the accessible interior and exterior surfaces of the containment structure and its components was performed to uncover any obvious evidence of deterioration, which could affect either containments structural integrity or leak-tightness. The procedure was performed in order to meet the requirements

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 56 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 of 10 CFR 50, Appendix J. The inspection is performed on a two-year frequency to coincide with the two-year refueling cycle. The 2017 inspection yielded no findings that challenged the containments structural integrity or leak-tightness.

3.3.8 Containment Liner Repair During installation of the Reactor Building Sump Strainer modification in support of the TMI-1 response to GL 2004-02, water was found under the stainless-steel sump liner. The source of the water under the stainless-steel reactor building sump lining was determined to be due to previous leakage from the seal plate of the reactor-to-refuel canal annulus region.

There is no evidence of on-going leakage. This was confirmed by chemical and radiochemistry analysis that indicated the water source was reactor coolant and that it was from the early 1990s. During that timeframe, there were several outages where there was significant leakage from the seal plate that sealed the annular gap between the fuel transfer canal and the reactor flange. The leakage flooded the in-core trench and submerged the moisture barrier in that trench with about 1 foot of reactor coolant system / fuel transfer canal water. The tritium concentration and physical data (reactor building pressure held during the run cycle) indicated that the water was not ground water. Once the initial water, and a short period of weepage, was removed, no additional water filled the boreholes or the sump. The gasketed, removable seal plate was replaced with a seal welded, permanent seal plate.

The new seal plate is leak tight and the in-core trench remains dry.

During the T1R17 refueling outage, the entire moisture barrier around the perimeter of the Reactor Building was removed. The underlying metal was visually examined at the barrier, above the barrier, and below the barrier approximately 2-4 into the gap between the floor and the wall. Corrosion was identified at the reactor building liner to concrete interface in several locations around the perimeter. Specifically, characterization of the findings from the visual examinations of the reactor building liner (supplemented by ultrasonic thickness examinations when corrosion activity was exhibited) were performed above, at and at accessible portions below the moisture barrier located at the interface between the reactor building concrete slab and the liner revealed the following patterns:

Examinations immediately above the moisture barrier interface with the reactor building liner and concrete generally revealed the largest areas of corrosion (largest surface corrosion not depth).

The excavated (i.e., caulking removed) areas in and at the moisture barrier generally revealed lessened corrosion activity (in area or size) on the reactor building liner surface, when compared to areas immediately above the moisture barrier. However, this area contained the deepest pitting or lowest remaining wall thickness.

Examined accessible bands of the reactor building liner surface below the moisture barrier revealed lesser corrosion, to no corrosion activity, when compared to the areas in and above the moisture barrier.

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 57 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 TMI-1 performed an inspection of the containment liner in accordance with ASME Section XI, Subsection IWE during the 2007 outage. This examination included the supplemental ultrasonic thickness examinations of six 12 by 12 grid areas at the liner-moisture barrier interface at 281 Elevation, which were previously examined during the 1999 refueling outage. The 2007 readings for these areas all varied less than 0.1 inches from the 1999 results.

Ten additional 12 by 12 grids were chosen to be included in the 2007 IWE inspections.

The locations of these grids were selected based on the areas where significant corrosion was detected in 1999 and 2003. The results of these examinations are shown in Table 3.3.8-1 below:

Table 3.3.8-1 IWE Examination of 12 X 12 Grid Areas on Reactor Building Liner Grid No.

Indication No.

2007 Examination Data Minimum Thickness (in.)

Nominal Thickness / IWE-3122.4 Acceptance Criteria (in.)

IWE Examination Results 1

Random Point 0.356

.375 /.337 No Indications Noted 2

Random Point 0.388

.375 /.337 No Indications Noted 3

Random Point 0.396

.375 /.337 No Indications Noted 4

Random Point 0.388

.375 /.337 No Indications Noted 5

Random Point 0.382

.375 /.337 No Indications Noted 6

Random Point 0.409

.375 /.337 No Indications Noted 7

No. 29 0.393

.375 /.337 No Indications Noted 8

No. 30 0.39

.375 /.337 No Indications Noted 9

No. 31 0.392

.375 /.337 No Indications Noted 10 No. 32 0.386

.375 /.337 No Indications Noted 11 No. 33 0.387

.375 /.337 No Indications Noted 12 No. 34 0.665

.750 /.675 IWE Criteria Not Met 13 No. 34 0.697

.750 /.675 No Indications Noted 14 No. 34 0.771

.750 /.675 No Indications Noted 15 Location 12 0.322

.375 /.337 IWE Criteria Not Met 16 Location 36 0.358

.375 /.337 No Indications Noted During outage T1R17, the moisture barrier at the liner concrete interface (Reactor Building 281 Elevation) was replaced. During the timeframe that the moisture barrier was removed, a 100 percent VT-3 visual examination was performed of the excavated region of the liner.

Based on the results of the visual examinations, ultrasonic thickness data and pit depth measurements were collected on the areas exhibited the most severe corrosion, in order to establish the minimum wall thickness in areas of identified corrosion. The results of the ultrasonic thickness examinations and pit depth measurements of the corroded areas in the excavated areas are summarized in Table 3.3.8-2 below:

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 58 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 Table 3.3.8-2 Examination Results of Moisture Barrier Area of Reactor Building Liner Indication No.

Plate No.

2007 Examination Data Minimum Thickness (in.)

Nominal Thickness /

IWE-3122.4 Acceptance Criteria (in.)

Results No. 37 C1-P47 0.377

.375 /.337 Acceptable No. 38 C1-P44 0.332

.375 /.337

> 10% Wall Loss No. 39 C1-P10 0.365

.375 /.337 Acceptable No. 40 C1-P47 0.292

.375 /.337

> 10% Wall Loss No. 41 C1-P44 0.277

.375 /.337

> 10% Wall Loss No. 42 C1-P37 0.358

.375 /.337 Acceptable No. 43 C1-P36 0.348

.375 /.337 Acceptable No. 44 C1-P24 0.672

.750 /.675

> 10% Wall Loss No. 45 C1-P22 0.69

.750 /.675 Acceptable No. 46 C1-P21 0.582

.750 /.675

> 10% Wall Loss No. 47 C1-P34 0.361

.375 /.337 Acceptable No. 48 C1-P34 0.318

.375 /.337

> 10% Wall Loss No. 49 C1-P33 0.302

.375 /.337

> 10% Wall Loss No. 50 C1-P30 0.251

.375 /.337

> 10% Wall Loss No. 51 C1-P29 0.321

.375 /.337

> 10% Wall Loss No. 52 C1-P32 0.282

.375 /.337

> 10% Wall Loss No. 53 C1-P32 0.239

.375 /.337

> 10% Wall Loss No. 54 C1-P32 0.311

.375 /.337

> 10% Wall Loss No. 55 C1-P13 0.346

.375 /.337 Acceptable No. 56 C1-P12 0.324

.375 /.337

> 10% Wall Loss The evaluation of the NDE examination results at the moisture barrier area of the reactor building liner concluded that the containment liner had sufficient thickness and complied with all design requirements. Design calculations indicated that the liner thickness allowed at the 281 Elevation of the liner can be a minimum of 0.200 inches in the area that has experienced corrosion and 0.48 at the knuckle region where the nominal plate thickness is 0.750. Note that this calculation conservatively assumed voids behind the liner plate and that the corrosion is at the minimum values in a 3 wide band, 360 degrees all the way around the liner at the basemat (281 Elevation) floor level. Additional analysis would need to be performed to show that substantially less wall thickness would have been required.

The NDE results from the 1R17 inspections noted above, were all greater than the minimum liner thickness allowed.

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 59 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 Aging Management Commitment The TMI-1 License Renewal Application (LRA) discussed the corrosion and need to repair the reactor building containment liner in the IWE section of Appendix B - Aging Management Programs in a description of Operating Experience related to the program.

The LRA stated, In 2007, TMI-1 conducted the 2nd ISI period examinations of the reactor building liner in accordance with ASME Section XI, Subsection IWE. 100% of the accessible areas of the liner were visually inspected. The results of the inspection were acceptable and were similar to findings in previous outages. In addition, augmented UT examinations were performed, which resulted from previous inspections in 1999 and 2003.

The results of the inspection were acceptable and confirmed that sufficient containment liner thickness remains.

Also in 2007, the entire Reactor Building moisture barrier was replaced during the refuel outage and a 100 percent VT-3 inspection was performed on the excavated region of the liner. The VT-3 examinations indicated some localized corrosion in the exposed area. UT examinations of the liner were performed in these regions. After replacement of the moisture barrier was complete, the adjoining service level 1 coating system was replaced.

The results of the inspections were acceptable and confirmed that sufficient containment liner thickness remains.

During the 2007 refueling outage, primary system water was discovered between the Reactor Building sump stainless steel liner and the lowest point of the carbon steel containment liner. This space is filled with concrete. Based on radiological analysis, the water was determined to have resulted from primary system leakage about 15 years ago.

The cause of the water intrusion was most likely due to previous leakage past a degraded moisture barrier between the Reactor Building reinforced concrete floor and the carbon steel containment liner. Corrosion of the stainless-steel sump liner and carbon steel containment liner due to continued exposure to the water was determined not to be an acceptable aging mechanism because the pH of the water was greater than 11.5.

Future augmented inspections under the IWE program will include inspection of the previously corroded areas behind the moisture barrier in accordance with IWE (or other alternative approved by the NRC). In addition, a one-time inspection will be performed of the liner in the area of the cork down to the horizontal plate. The entire moisture barrier will continue to be visually inspected each refueling outage.

The SER for the LRA addressed this Operating Experience as follows:

Operating Experience. The staff reviewed the operating experience provided in LRA Section B.2.1.24 and also interviewed the applicants technical staff to confirm that the plant-specific operating experience did not reveal any aging effects not bounded by the GALL report. The staff confirmed that applicable aging effects and industry and plant-specific operating experience have been reviewed by the applicant and are evaluated in the GALL report. The staff also confirmed that the applicant has addressed operating experience identified after the issuance of the GALL report.

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 60 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 The staff noted that the liner thickness corrosion rate was noticeable from operating experience provided, especially at locations adjacent to the moisture barrier at Elevation 281 and 279-6. To ensure the essential leak-tight condition of the containment for the period of extended operation, the staff identified an issue concerning the restoration of degraded plate areas where additional information was needed to complete its review.

In the LRA, the applicant committed to replacing the existing steam generators with new once through steam generators (OTSG) prior to entering the period of extended operation.

The applicant stated that the repair/replacement of the reactor building liner plate, removed for access purposes, will be performed in accordance with ASME Section XI, Subsection IWE. The applicant indicated that the liner will be restored (weld repair) to full design thickness at all locations identified as less than 90% before entering the period of extended operation. In RAI B.2.1.24-2, dated October 7, 2008, the staff requested that the applicant provide additional information to confirm the repairs and provide the proposed schedule for completion.

In its response to the RAI dated October 30, 2008, the applicant stated that prior to the period of extended operation, the reactor building liner will be restored to its nominal plate thickness by weld repair for the previously identified corroded areas where the thickness of the base metal is reduced by more than 10% of the nominal plate thickness. The applicant added this information to LRA Table A.5, as Commitment No. 42.

As part of commitments pertaining to the renewal of the operating license, the following commitment was added to LRA Table A.5 and Table A-5, License Renewal Commitment List of the TMI-1 UFSAR as Commitment No. 42:

Containment Liner Repair Prior to the period of extended operation, TMI-1 will restore the Reactor Building liner to its nominal plate thickness by weld repair for the previously identified corroded areas of the Reactor Building liner where the thickness of the base metal is reduced by more than 10%

of the nominal plate thickness.

Implementation Schedule: Prior to the period of extended operation.

During refueling outage T1R18, 91.75 feet of the moisture barrier was removed from the Reactor Building basement floor to liner interface to support the liner restoration. Based on the initial moisture barrier removal, 59 feet of concrete was also removed to support additional examination/repair activities. In total, approximately 53 linear feet of weld repairs were performed on the identified degraded areas. Follow-up visual and volumetric examinations verified the weld repairs were satisfactory prior to reinstallation of the concrete floor and moisture barrier. Since these repairs were performed concurrent with the Steam Generator Replacement Outage, the ILRT verified the integrity of the weld repairs. The ILRT was completed satisfactory with a leakage rate of 0.05693 wt.%/day (Acceptance Criteria 0.1 wt.%/day).

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 61 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 3.4 NRC Information Notices For the TMI-1 Primary Containment, the following site specific and industry events have been evaluated for impact on TMI-1:

Information Notice (IN) 1992-20, "Inadequate Local Leak Rate Testing" IN 2004-09, Corrosion of Steel Containment and Containment Liner IN 2010-12, "Containment Liner Corrosion" IN 2014-07, "Degradation of Leak-Chase Channel Systems for Floor Welds of Metal Containment Shell and Concrete Containment Metallic Liner" Regulatory Issue Summary (RIS) 2016-07, Containment Shell of Liner Moisture Barrier Inspection Each of these areas is discussed in detail in Sections 3.4.1 through 3.4.5, respectively.

3.4.1 IN-92-20, Inadequate Local Leak Rate Testing The NRC issued IN 92-20 to alert licensees of problems with local leak rate testing two-ply stainless steel bellows used on piping penetrations at four different plant: Quad Cities, Dresden Nuclear Station, Perry Nuclear Power Plant, and the Clinton Station. Specifically, LLRTs could not be relied upon to accurately measure the leakage rate that would occur under accident conditions since, during testing, the two plies in the bellows were in contact with each other, restricting the flow of the test medium to the crack locations. Any two-ply bellows of similar construction may be susceptible to this problem. The common issue in the four events was the failure to adequately perform LLRT on different penetration configurations leading to problems that were discovered during ILRT tests in the first three cases.

In the event at Quad Cities, the two-ply bellows design was not properly subjected to LLRT pressure and the conclusion of the utility was that the two-ply bellows design could not be Type B LLRT tested as configured.

In the events at both Dresden and Perry, flanges were not considered a leakage path when the Type C LLRT test was designed. This omission led to a leakage path that was not discovered until the plant performed an ILRT test.

In the event at Clinton, relief valve discharge lines that were assumed to terminate below the suppression pool minimum drawdown level were discovered to terminate at a level above that datum. These lines needed to be reconfigured and the valves should have been Type C LLRT tested.

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 62 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 Discussion TMI-1 does not have double ply bellows that perform a containment isolation function that are tested by the LLRT program. Likewise, there are no double ply metal bellows elsewhere at TMI-1 whose leak test could be adversely affected as noted in the IN. Therefore, no further action is required.

The TMI-1 LLRT program would detect the incidents involving the flange joint leakage. The TMI-1 LLRT program defines the test boundaries in accordance with drawings contained in Surveillance Procedure 1303-11.18, RB Local Leak Rate Testing. The procedure specifies target leakage criteria and action to seek out leaking joints if leakage is unacceptable. The TMI-1 LLRT Program does test some CIVs in the reverse direction; however, the valves are not directionally dependent. Therefore, no further action is required.

TMI-1 does not have a designated seal water system that is an accompaniment as described in the IN that would be relied upon to perform a containment isolation function.

Therefore, no further action is required.

3.4.2 IN 2004-09, Corrosion of Steel Containment and Containment Liner The NRC issued IN 2004-09 to alert addressees to recent occurrences of corrosion in freestanding metallic containments and in liner plates of reinforced and pre-stressed concrete containments. Any corrosion (metal thinning) of the liner plate or freestanding metallic containment could change the failure threshold of the containment under a challenging environmental or accident condition. Thinning changes the geometry of the containment shell or liner plate and may reduce the design margin of safety against postulated accident and environmental loads. Recent experience has shown that the integrity of the moisture barrier seal at the floor-to-liner or floor-to-containment junctions is important in avoiding conditions favorable to corrosion and thinning of the containment liner plate material. Inspections of containment at the floor level, as well as at higher elevations, have identified various degrees of corrosion and containment plate thinning.

Discussion TMI-1 programs and procedures for inservice inspections of the containment liner and its components are established in accordance with ASME Boiler and Pressure Vessel Code,Section XI, 1992 Edition, 1992 Addenda through 1998 Edition. Previous examinations have found corrosion in the containment liner and degraded conditions in the moisture barrier.

Actions have been taken to correct the as-found conditions and the structural adequacy of the containment was not affected by the as-found conditions in accordance with the evaluations. Additional TMI-1 actions have been identified to examine non-conforming containment conditions. No follow-up actions were required as a result of this IN. The existing and additional TMI-1 NDE programs and procedures are adequate to identify corrosion of the liner and degraded conditions of the moisture barrier. In addition, the structural adequacy of the containment was not affected by the as-found conditions, in accordance with the evaluations.

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 63 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 3.4.3 IN 2010-12, Containment Liner Corrosion IN 2010-12 was issued to alert plant operators to three events that occurred where the steel liner of the containment building was corroded and degraded. At the Beaver Valley and Brunswick plants, material had been found in the concrete, which trapped moisture against the liner plate and corroded the steel. In one case, it was material intentionally placed in the building and in the other case, it was foreign material, which had inadvertently been left in the form when the wall was poured. But the result in both cases was that the material trapped moisture against the steel liner plate leading to corrosion. In the third case, Salem, an insulating material placed between the concrete floor and the steel liner plate absorbed moisture and led to corrosion of the liner plate.

Discussion TMI-1 has implemented periodic examinations of the metallic containment structure or liner during RFOs in accordance with ASME Section XI, Subsection IWE and 10 CFR 50, Appendix J. The applicable visual examination procedure requires the conditions described in the IN to be recorded. Conditions that may affect the surface integrity are then required to be evaluated by engineering or repaired/replaced prior to startup from the refueling outage.

Liner corrosion on metallic containment surfaces were identified at TMI-1 during periodic IWE examinations and during refueling outages. However, the conditions were dispositioned in accordance with the applicable rules of ASME Section XI. Rigorously implementing the examinations and tests in accordance with the rules of ASME Section XI, Subsection IWE and 10 CFR 50, Appendix J, and dispositioning observed conditions in accordance with the applicable guidance are existing barriers that ensure the integrity of metallic containment surfaces.

3.4.4 IN 2014-07, Degradation of Leak-Chase Channel Systems for Floor Welds of Metal Containment Shell and Concrete Containment Metallic Liner The NRC issued IN 2014-07 to inform the industry of issues concerning degradation of floor weld leak-chase channel systems of steel containment shell and concrete containment metallic liner that could affect leak-tightness and aging management of containment structures. Specifically, this IN provides examples of operating experience at some plants of water accumulation and corrosion degradation in the leak-chase channel system that has the potential to affect the leak-tight integrity of the containment shell or liner plate. In each of the examples, the plant had no provisions in its ISI plan to inspect any portion of the leak-chase channel system for evidence of moisture intrusion and degradation of the containment metallic shell or liner within it. Therefore, these cases involved the failure to perform required visual examinations of the containment shell or liner plate leak-chase systems in accordance with the ASME Code Section XI, Subsection IWE, as required by 10 CFR 50.55a(g)(4).

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 64 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 The containment basemat metallic shell and liner plate seam welds of pressurized water reactors are embedded in a 3-feet by 4-feet concrete floor during construction and are typically covered by a leak-chase channel system that incorporates pressurizing test connections. This system allows for pressure testing of the seam welds for leak-tightness during construction and also while in service, as required. A typical basemat shell or liner weld leak-chase channel system consists of steel channel sections that are fillet welded continuously over the entire bottom shell or liner seam welds and subdivided into zones, each zone with a test connection.

Each test connection consists of a small carbon or stainless-steel tube (less than 1-inch diameter) that penetrates through the back of the channel and is seal-welded to the channel steel. The tube extends up through the concrete floor slab to a small access (junction) box embedded in the floor slab. The steel tube, which may be encased in a pipe, projects up through the bottom of the access box with a threaded coupling connection welded to the top of the tube, allowing for pressurization of the leak-chase channel. After the initial tests, steel threaded plugs or caps are installed in the test tap to seal the leak-chase volume. Gasketed cover plates or countersunk plugs are attached to the top of the access box flush with the containment floor. In some cases, the leak-chase channels with plugged test connections may extend vertically along the cylindrical shell or liner to a certain height above the floor.

Discussion At TMI-1, the leak chase channel system is different from the sketch provided in the IN. The reference drawings document how multiple liner plates are formed into a continuous sheet.

Beneath every joint, WT sections are inverted, the liner plates placed onto the flange, and both the plates and WT flange are seal-welded together. Above the joint, an inverted channel is installed; also, seal-welded to the liner plate. The inverted channel is placed above and for the entire length of all liner plate joints below the floor slab. Moving outward from the center of the containment building, the concrete interface between the floor and wall has a gentle radius that is matched by each channel. Each channel follows the profile of the sloping wall up to the high of the concrete floor slab. At these locations, the channels are capped at the end with a moisture barrier to prevent moisture and debris from entering the channel space and fitted with test ports. A 2 feet thick layer of concrete was then placed over the entire surface (liner plate and channel system).

Referencing Section 5.5.5.3, Preliminary Tests, in TMI-1s UFSAR, the test channel design with test plugs was to be used originally for post-construction weld inspection of the liner plate. Prior work has been completed in the past to modify and improve upon the design of the existing plugs for access to the test channel. In fulfillment of TMI-1s License Renewal Commitments, modified test ports were installed in support of the T1R18 refueling outage liner examination.

Inspections of the liner, through leak test channels where applicable (some locations are inaccessible due to obstructions), are completed on a routine interval. Following a thorough review of the completed inspection procedures, no liner issues were documented.

Therefore, no further actions are necessary.

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 65 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 3.4.5 NRC RIS 2016-07, Containment Shell or Liner Moisture Barrier Inspection The NRC staff identified several instances in which containment shell or liner moisture barrier materials were not properly inspected in accordance with ASME Code Section XI, Table IWE-2500-1, Item E1.30. Note 4 (Note 3 in editions before 2013) for Item E1.30 under the Parts Examined column states, Examination shall include moisture barrier materials intended to prevent intrusion of moisture against inaccessible areas of the pressure retaining metal containment shell or liner at concrete-to-metal interfaces and at metal-to-metal interfaces, which are not seal welded. Containment moisture barrier materials include caulking, flashing and other sealants used for this application.

Examples of inadequate inspections have included: licensees not identifying sealant materials at metal-to-metal interfaces as moisture barriers because they do not specifically match Figure IWE-2500-1; and, licensees not inspecting installed moisture barriers, as required by Item E1.30, because the material was not included in the original design or was not identified as a moisture barrier in design documents.

Discussion TMI-1 performed a review of the containment design when the program was developed and identified all areas that were designed to prohibit water intrusion into inaccessible areas of the containment liner. The current inspection procedure includes Reactor Building Liner Inspection Ports and has adequate detail to ensure all inspection ports through the concrete floor have been included. Additional Reactor Building Liner Inspection locations were added in the 2009 RFO for Life Extension and are included in the inspection procedure. Otherwise, there have been no changes/modifications to moisture barriers installed at TMI-1.

Preventative maintenance plans are in place to generate work orders so that moisture barriers are inspected in each inspection period. This ensures that these inspections do not get missed during RFOs.

Discussion with prior ISI personnel identified potential inadequate detail regarding inspection scope. Inspection ports in the Reactor Building floor are listed in the Reactor Building Concrete/Liner Inspection surveillance and the program procedure for Visual Examination of Section XI Class MC Surfaces and Class CC Liners. If new moisture barriers requiring visual inspection were to be added, then this would go through the Design Change Process and would be subject to review by the ISI/CISI Program Owner. This review barrier would evaluate the inclusion into the CISI program. Additionally, the ISI program software acts as a barrier to prevent the occurrence of a missed inspection.

3.5 Summary Analysis The testing requirements of 10 CFR 50, Appendix J, provide assurance that leakage from the primary containment, including systems, and components that penetrate the containment, does not exceed the allowable leakage values specified in TS. The limitation on containment leakage provides assurance that the primary containment will perform its design function following plant design basis accidents.

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 66 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 The allowed frequency for Type A testing, as documented in NEI 94-01, is based, in part, upon a generic evaluation documented in NUREG-1493 (Reference 5). As discussed in NUREG-1493, reducing the Type A ILRT testing frequency to one per 20 years was found to lead to an imperceptible increase in risk. Additionally, while Types B and C tests identify the vast majority (greater than 95%) of all potential leakage paths, performance-based alternatives are feasible without significant risk impacts. Since leakage contributes to less than 0.1% of overall risk under existing guidelines, the overall effect is small.

TMI-1 has undergone eight operational Type A tests in addition to the pre-operational Type A test. The results of these tests demonstrate that the TMI-1 containment structure remains an essentially leak-tight barrier and represents minimal risk to increased leakage. The most recent ILRT was performed in 2010 per the requirements of 10 CFR 50, Appendix J, Option B and NEI 94-01, Revision 0. Extensions of the ILRT from three tests in 10 years to one test in 10 years is permissible if the performance leakage rate is below 1.0 La (0.1 wt.%/day). Since the performance leakage rate using mass point leakage results was less than the performance criteria value of 1.0 La (0.1 wt.%/day), the TMI-1 ILRT remains on an extended interval of at least once per ten years.

As discussed in NUREG-1493 (Reference 5), Type B and Type C tests can identify the vast majority of all potential containment leakage paths. Type B and Type C testing will continue to provide a high degree of assurance that containment integrity is maintained. A review of the Type B and Type C test results from 2007 through 2017 for TMI-1 has shown substantial margin between the actual As-Found (AF) and As-Left (AL) outage summations and the regulatory requirements. The Types B and C pathway totals for the current operating cycle show 87 percent margin between the As-Left MNPLR totals and the regulatory limit of 0.6La.

In addition, the TMI-1 extended frequency component percentage in excess of 85 percent for the Type C tested components supports that the maintenance and corrective action programs are aggressive in addressing CIV failures.

Additionally, the ASME Section XI and Reactor Building Level 1 Coatings examinations provide a high degree of assurance that any degradation of the containment structure is identified and corrected before a containment leakage path is introduced. These inspections aided in the discovery of a degraded containment moisture barrier, which resulted in containment liner plate corrosion. The corrosion was caused by historical leakage from the seal plate of the reactor to the refuel canal annulus region. Extent of condition examinations were performed on the liner to identify areas of the liner with greater than 10% wall loss. During refueling outage T1R18 in 2009, 91.75 feet of the moisture barrier was removed from the reactor building basement floor to liner interface to support the liner restoration. Based on the initial moisture barrier removal, 59 feet of concrete was also removed to support additional examination/repair activities. In total, approximately 53 linear feet of weld repairs were performed on the identified degraded areas. Follow-up visual and volumetric examinations verified the weld repairs were satisfactory prior to reinstallation of the concrete floor and moisture barrier. Pressure testing of the repaired areas was performed by the 2009 ILRT with satisfactory results.

In conclusion, TMI-1 utilizes several testing and inspection methodologies to ensure the containment structure will perform its safety function, if called upon during a design basis accident. A one-time extension of the ILRT by 1.75 years is considered to be acceptable.

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 67 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria The proposed change has been evaluated to determine whether applicable regulations and requirements continue to be met. The requirements to perform testing of the primary reactor containment are set forth in 10 CFR 50.54(o) and 10 CFR 50, Appendix J. Both of these sections address criteria established in 10 CFR 50, Appendix A, General Design Criteria (GDC): GDC 50 (Containment Design Basis); GDC 51 (Fracture Prevention of Containment Pressure Boundary); GDC 52 (Capability for Containment Leakage Rate Testing); and, GDC 53 (Provisions for Containment Testing and Inspection. Exelon has determined that the proposed change does not require any additional exemptions or relief from regulatory requirements and does not affect conformance with any GDC as described in the Updated Final Safety Analysis Report (UFSAR). However, this change does propose a one-cycle extension of the frequency for performance of the Type A ILRT.

10 CFR 50.54(o) requires primary reactor containments for water-cooled power reactors to be subject to the requirements of Appendix J to 10 CFR Part 50, Leakage Rate Testing of Containment Water Cooled Nuclear Power Plants. Appendix J specifies containment leakage testing requirements, including the types required to ensure the leak-tight integrity of the primary reactor containment and systems and components which penetrate the containment. In addition, Appendix J discusses leakage rate acceptance criteria, test methodology, frequency of testing and reporting requirements for each type of test.

The adoption of the Option B performance-based containment leakage rate testing for Type A, Type B, and Type C testing did not alter the basic method by which Appendix J leakage rate testing is performed; however, it did alter the frequency at which Type A, Type B, and Type C containment leakage rate tests must be performed. Under the performance-based option of 10 CFR 50, Appendix J, the test frequency is based upon an evaluation that reviewed "as-found" leakage history to determine the frequency for leakage testing which provides assurance that leakage limits will be maintained. The change to the Type A test frequency will not directly result in an increase in containment leakage.

10 CFR 50.36, Technical Specifications, provides the regulatory requirements for the content required in a plants TS. 10 CFR 50.36(c)(5), Administrative controls, requires that provisions relating to organization and management, procedures, recordkeeping, review and audit, and reporting necessary to assure operation of the facility in a safe manner will be included in the plants TS.

10 CFR 50, Appendix J, Option B,Section V.B, Implementation subparagraph 3, requires that the regulatory guide or implementation document used to develop a performance-based leakage-testing program be included by general reference in the plants TS. The Appendix J Testing Program is included in the Administrative Controls section of the TMI-1 TS as TS 6.8.5, Reactor Building Leakage Rate Testing Program. This LAR does not remove this administrative control requirement, but simply revises the administrative controls TS 6.8.5 to include extending the frequency for performing the Type A ILRT from 10 years to 11.75 years. Therefore, this 10 CFR 50.36 requirement continues to be met by this change.

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 68 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 4.2 Precedence This LAR is similar in nature to the following license amendments to extend the Type A test frequency, which were previously authorized by the NRC in the associated SERs:

Three Mile Island Nuclear Station, Unit 1, Amendment No. 244, dated August 14, 2003 (Reference 6)

Three Mile Island Nuclear Station, Unit 1, Amendment No 259, dated June 29, 2007 (Reference 7)

Nine Mile Point Nuclear Station, Unit 1, Amendment No. 151, dated December 29, 1994 (Reference 8)

Vermont Yankee Nuclear Power Station, Amendment No. 215, dated June 2, 2003 (Reference 9)

Arkansas Nuclear One, Unit No. 2, Amendment No. 284, dated July 20, 2009 (Reference 10)

Palisades Nuclear Plant, dated August 23, 2010 (Reference 11)

Oconee Nuclear Station, Unit 1, dated October 1, 2012 (Reference 12)

Oconee Nuclear Station, Units 2 and 3, dated August 5, 2013 (Reference 13)

McGuire Nuclear Station, Units 1 and 2, dated September 26, 2016 (Reference 14)

Grand Gulf Nuclear Station, Unit 1, dated December 29, 2017 (Reference 15) 4.3 No Significant Hazards Consideration Exelon Generation Company, LLC (EGC) has evaluated whether or not a significant hazards consideration is involved with the proposed amendment to extend the Three Mile Island Nuclear Station, Unit 1 (TMI-1) Type A integrated leakage rate test (ILRT), by focusing on the three standards set forth in 10 CFR 50.92, "Issuance of amendment," as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

The proposed amendment to the Technical Specifications (TS) 6.8.5 involves a one-time extension of the Three Mile Island Nuclear Station, Unit 1 (TMI-1), Type A integrated leakage rate test (ILRT) from 10 years to 11.75 years, in accordance with the Nuclear Regulatory Commission (NRC)-accepted guidelines of Nuclear Energy Institute (NEI) 94-01, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, Revision 3-A. This change will extend the requirement to perform the Type A ILRT from the current requirement of prior to startup from the T1R18 refueling outage, to November 2009 Type A test shall be performed no later than prior to startup from the T1R24 refueling outage in 2021.

The proposed extension does not involve either a physical change to the plant or a change in the manner in which the plant is operated or controlled. The containment is designed to provide an essentially leak tight barrier against the uncontrolled release of

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 69 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 radioactivity to the environment for postulated accidents. Types B and C testing ensures that individual containment isolation valves (CIVs) are essentially leak tight. In addition, aggregate Types B and C leakage rates support the leakage tightness of primary containment by minimizing potential leakage paths. The proposed amendment will not change the leakage rate acceptance requirements. As such, the containment will continue to perform its design function as a barrier to fission product releases. In addition, the containment and the testing requirements invoked to periodically demonstrate the integrity of the containment exist to ensure the plants ability to mitigate the consequences of an accident, and do not involve the prevention or identification of any precursors of an accident previously evaluated. Therefore, this proposed interval extension does not involve a significant increase in the probability or consequences of an accident previously evaluated.

Therefore, the proposed change does not result in a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the proposed change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

The proposed amendment to the TS involves a one-time extension of the TMI-1 Type A ILRT from 10 years to 11.75 years. The containment and the testing requirements to periodically demonstrate the integrity of the containment exist to ensure the plants ability to mitigate the consequences of an accident do not involve any accident precursors or initiators.

The proposed change does not involve a physical change to the plant (i.e., no new or different type of equipment will be installed) or a change to the manner in which the plant is operated or controlled. This administrative change to extend the Type A ILRT for TMI-1 will not affect the control parameters governing unit operation or the response of plant equipment to transient or accident conditions.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed change involve a significant reduction in a margin of safety?

Response: No.

The proposed amendment to the TS involves the extension of the TMI-1 Type A ILRT interval to 11.75 years. This amendment does not alter the manner in which safety limits, limiting safety system set points, or limiting conditions for operation are determined. The specific requirements and conditions of the TS 6.8.5, Reactor Building Leakage Rate Testing Program, for containment leak rate testing exist to ensure that the degree of containment structural integrity and leak-tightness that is considered in the plant safety analysis are maintained. The overall containment leak rate limit specified by TS is maintained.

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 70 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289 The proposed change involves the extension of the interval for only the Type A containment leakage rate test at TMI-1. The proposed surveillance interval extension is bounded by the 15-year Type A test interval currently authorized within NEI 94-01, Revision 3-A. The design, operation, testing methods, and acceptance criteria for Types A, B, and C containment leakage tests specified in applicable codes and standards would continue to be met with the acceptance of this proposed change, since these are not affected by the proposed change to the Type A test interval.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, EGC concludes that the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of no significant hazards consideration is justified.

4.4 Conclusion In conclusion, based on the considerations discussed above, (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 Commissions 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.

5.0 ENVIRONMENTAL CONSIDERATION

A review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement.

However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 REFERENCES

1.

Letter from B. C. Buckley (NRC) to J. W. Langenbach (TMI), Three Mile Island -

Issuance of Amendment [No. 201] Re: 10 CFR 50, Appendix J, Option B (TAC No.

M96029), dated May 27, 1997 (ML003765721)

2.

Regulatory Guide 1.163, Performance-Based Containment Leak-Test Program, dated September 1995

3.

NEI 94-01, Revision 0, Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J, dated July 1995

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 71 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289

4.

ANSI/ANS 56.8-1994, Containment System Leakage Testing Requirements, dated August 4, 1994

5.

NUREG-1493, Performance-Based Containment Leak-Test Program, January 1995

6.

Letter from D. M. Skay (NRC) to J. L. Skolds (TMI), Three Mile Island Nuclear Station, Unit 1 (TMI-1), Re: Deferral of Containment Integrated Leakage Rate Test (TAC No. MB6487), dated August 14, 2003 (ML032050212)

7.

Letter from P. Bamford (NRC) to Mr. C. M. Crane (AmerGen Energy Company, LLC),

Three Mile Island Nuclear Station, Unit 1 - Issuance of Amendment [No. 259]

Regarding One-Time Type A Test Interval Extension (TAC No. MD3027), dated June 29, 2007 (ML071650519)

8.

Letter from D. S. Brinkman (NRC) to B. R. Sylvia (Niagara Mohawk), Issuance of Amendment [No. 151] for Nine Mile Point Nuclear Station Unit No.1 (TAC No.

M90278), dated December 29, 1994 (ML011080782) 9 Letter from R. M. Pulsifer (NRC) to J. K. Thayer (Entergy Nuclear Vermont Yankee, LLC), Vermont Yankee Nuclear Power Station - Issuance of Amendment [No. 215]

Re: One-Time Extension of Appendix J Type A Integrated Leakage Rate Test Interval (TAC No. MB6507), dated June 2, 2003 (ML031320686)

10.

Letter from N. K. Kalyanam (NRC) to Vice President, Operations (Entergy Operations, Inc.), Arkansas Nuclear One, Unit No. 2 - Issuance of Amendment [No.

284] Re: One-Time Extension to 10-Year Frequency of Integrated Leak Rate Test (TAC No. MD9502), dated July 20, 2009 (ML091540158)

11.

Letter from M. L. Chawla (NRC) to Vice President, Operations (Entergy Nuclear Operations, Inc.), Palisades Nuclear Plant - Issuance of Amendment [No. 240] Re:

One-Time Extension to the Integrated Leak Rate Test Interval (TAC No. ME2122),

dated August 23, 2010 (ML102090137)

12.

Letter from J. P. Boska (NRC) to P. Gillespie (Duke Energy Carolinas, LLC), Oconee Nuclear Station, Unit 1, Issuance of Amendment [No. 381] Regarding Extension of the Reactor Building Integrated Leak Rate Test (TAC No. ME8407), dated October 1, 2012 (ML12250A339)

13.

Letter from J. P. Boska (NRC) to S. Batson (Duke Energy Carolinas, LLC), Oconee Nuclear Station, Units 2 and 3, Issuance of Amendments [Nos. 383 and 382, respectively] Regarding Extension of the Reactor Building Integrated Leak Rate Test (TAC Nos. ME9777 and ME9778), dated August 5, 2013 (ML13193A329)

Evaluation of Proposed Changes One-Cycle Extension of Appendix J Page 72 of 72 Type A Integrated Leakage Rate Test Docket No. 50-289

14.

Letter from G. E. Miller (NRC) to S. D. Capps (Duke Energy Carolinas, LLC),

McGuire Nuclear Station, Units 1 and 2 - Issuance of Amendments [Nos. 290 and 269, respectively] Re: One-Time Extension of Appendix J Type A Integrated Leakage Rate Test Interval (CAC Nos. MF7407 and MF7408), dated September 26, 2016 (ML16236A053)

15.

Letter from S. P. Lingam (NRC) to Vice President, Operations (Entergy Operations, Inc.), Grand Gulf Nuclear Station, Unit 1 - Issuance of Amendment [No. 214] Re:

One-Cycle Extension of Appendix J Type A Integrated Leakage Test and Drywell Bypass Test Interval (CAC No. MF9461; EPID L-2016-LLA-0040), dated December 29, 2017 (ML17334A739)

16.

ADTM D4537, Standard Guide for Establishing Procedures to Qualify and Certify Personnel Performing Coating and Lining Work Inspection in Nuclear Facilities

17.

ANSI/ASME N45.2.6, Qualifications of Inspection, Examination, and Testing Personnel for Nuclear Power Plants

18. TMI-18-094, Submittal of Relief Request RR-18-01 Concerning Containment Unbonded Post-Tensioning System lnservice Inspection Requirements, dated October 16, 2018 (ML18289A363)

ATTACHMENT 2 License Amendment Request Three Mile Island Nuclear Power Station - Unit 1 Docket No. 50-289 PROPOSED TECHNICAL SPECIFICATION MARKED-UP PAGE Page 6-11c

6.8.5 Reactor Building Leakage Rate Testing Program The Reactor Building Leakage Rate Testing Program shall be established, implemented, and maintained as follows:

A program shall be established to implement the leakage rate testing of the Reactor Building as required by 10 CFR 50.54(0) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," dated September 1995, as modified by the following exception to NEI 94-01, Rev. 0, "Industry Guideline for Implementing Performance-Based Option of.

10 CFR Part 50, Appendix J":

a.

The peak calculated Reactor Building internal pressure for th coolant accident, Pac. is 50.6 psig.

The maximum allowable Reactor Building leakage rate, La, s II be 0.1 weight percent of containment atmosphere per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> at Pac*

Reactor Building leakage rate acceptance criteria is s 1.0 La. During the first plant startup following each test performed in accordance with this program, the leakage rate acceptance criteria ares 0.60 La for the Type Band Type C sts ands 0.75 La for the Type A tests.

January 2010 Type A test shall be performed no later than prior to startup from the T1 R24 refueling outage in 2021.

6-11c Amendment No. 201, 244, ~