ML19241A243

License Amendment Request to Extend Containment Leakage Test Interval

ML19241A243
Person / Time
Site:	Davis Besse
Issue date:	08/26/2019
From:	Bezilla M FirstEnergy Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-19-189
Download: ML19241A243 (82)

Text

FENOC

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FirstEnergy Nuclear Operating Company Mark B. Bezi/la Site Vice President August26,2019 L-19-189 ATTN: Document Control Desk U. S. Nuclear Regulatory Commission Washington, ~C 20555-0001

SUBJECT:

Davis-Besse Nuclear Power Station, Unit No. 1 Docket No. 50-346, License No. NPF-3 10 CFR 50.90 5501 North State Route 2 Oak Harbor, Ohio 43449 419-321-7676 Fax: 419-321-7582 License Amendment Request to Extend Containment Leakage Test Interval Pursuant to 10 CFR 50.90, FirstEnergy Nuclear Operating Company (FENOC) hereby requests an amendment to the facility operating license for the Davis-Besse Nuclear Power Station, Unit No. 1 (DBNPS). The proposed change would revise DBNPS Technical Specification 5.5.. 15, "Containment Leakage Rate Testing Program," by replacing the reference to Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," (Reference 1) with a reference to Nuclear Energy Institute (NEI) topical report NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J," Revision 3-A (Reference 2), and the limitation and conditions specified in NEI 94-01, Revision 2-A with the same title (Reference 3), for Type A and B testing.

The FENOC evaluation of the proposed changes is enclosed. Approval of the proposed amendment is requested by August 29, 2020. Once approved, the amendment shall be implemented within 30 days of receipt.

There are no regulatory *commitments contained in this submittal. If there are any questions or additional information is required, please contact Mr. Phil H. Lashley, Acting Manager Nuclear Licensing and Regulatory Affairs, at 330-315-6808.

I declare under penalty of perjury that the foregoing is true and correct. Executed on August2'., 2019.

Davis-Besse Nuclear Power Station, Unit No. 1 L-19-189 Page 2

Enclosure:

FENOC Evaluation of the Proposed Amendment cc: NRC Region Ill Administrator NRC Resident Inspector NRC Project Manager Branch Chief, Ohio Emergency Management Agency, State of Ohio (NRC Liaison)

Utility Radiological Safety Board

FENOC Evaluation of the Proposed Amendment Page 1 of 76

Subject:

License Amendment Request to Revise Technical Specification 5.5.15 for Permanent Extension of Containment Type A Leak Rate Test Interval TABLE OF CONTENTS 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION 2.1 Primary Containment and Shield Building Design 2.2 Containment Vessel Isolation Systems 2.3 Current Technical Specification Requirements 2.4 Reason for the Proposed Change 2.5 Description of the Proposed Change

3.0 TECHNICAL EVALUATION

3.1 Emergency Core Cooling System Net Positive Suction Head 3.2 Containment Vessel Tests and Inspections Performed Following Major Modifications 3.3 Justification for the Technical Specification Change 3.3.1 Chronology of 10 CFR 50, Appendix J, Testing Requirements 3.3.2 Current Containment Leakage Rate Testing Program Requirements 3.3.3 Support for the Continued Use of Exceptions 1 and 2 to Technical Specification 5.5.15 3.3.4 DBNPS Containment Leak Test Licensing History 3.3.5 DBNPS Periodic Integrated Leakage Rate (Type A) Test Results 3.4 Plant Specific Confirmatory Analysis 3.4.1 Methodology 3.4.2 Technical Adequacy of the Probabilistic Risk Analysis (PRA) 3.4.2.1 Internal Events PRA Quality Statement for Permanent 15-Year Integrated Leakage Rate Test Extension 3.4.2.2 PRA Maintenance and Update 3.4.2.3 Applicability of Peer Review Facts and Observations 3.4.2.4 Internal Events PRA Model 3.4.2.5 Fire PRA Model 3.4.2.6 Seismic PRA Model 3.4.2. 7 Consistency with Applicable PRA Standards

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 2 of 76 3.4.3 Summary of Plant-Specific Risk Assessment Results 3.4.4 Previous Assessments 3.4.5 Regulatory Guide 1.17 4, Revision 3 - Defense-in-Depth Evaluation 3.5 Non-Risk Based Assessment

• 3.5.1 Nuclear Safety-Related Protective Coatings Program 3.5.2 Containment lnservice Inspection Program 3.5.3 Supplemental Inspection Requirements 3.5.4 Results of Recent Containment Examinations 3.5.5 Containment Leakage Rate Testing Program -

Type B and Type C Testing Program 3.5.6 Type B and Type C Local Leak Rate Testing Program Implementation Review 3.6 Operating Experience 3.6.1 Aging Management Operating Experience 3.6.2 NRG Containment Liner Corrosion Operating Experience Summary 3.6.3 Information Notice 2004-09, "Corrosion of Steel Containment and Containment Liner

3.6.4 Information Notice 2014-07, "Degradation of Leak-Chase Channel Systems for Floor Welds of Metal Containment Shell and Concrete Containment Metallic Liner" 3.6.5 Shield Building Laminar Cracking

3. 7 License Renewal Aging Management 3.8 NRG Safety Evaluation Limitations and Conditions 3.8.1 Limitations and Conditions Applicable to Nuclear Energy Institute Topical Report NEI 94-01, Revision 2-A 3.8.2 Limitations and Conditions Applicable to Nuclear Energy Institute Topical Report NEI 94-01, Revision 3-A 3.9 Conclusion

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 3 of 76

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements and Criteria 4.2 Precedent 4.3 No Significant Hazards Consideration 4.4 Conclusion

6.0 REFERENCES

ATTACHMENTS A.

Proposed Technical Specification Changes B.

Evaluation of Risk Significance of Permanent Integrated Leakage Rate Test Extension

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 4 of 76 1.0

SUMMARY

DESCRIPTION In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," FirstEnergy Nuclear Operating Company (FENOC) requests an amendment to renewed facility operating license number NFP-3 for Davis-Besse Nuclear Power Station (DBNPS).

The proposed change would revise DBNPS Technical Specification 5.5.15, "Containment Leakage Rate Testing Program," by replacing the reference to Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program," (Reference 1) with a reference to Nuclear Energy Institute (NEI) topical report NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J,"

Revision 3-A (Reference 2), and the limitation and conditions specified in NEI 94-01, Revision 2-A with the same title (Reference 3), for Type A and B testing.

These documents will be used by DBNPS to continue with the implementation of the performance-based leakage testing program in accordance with Option B, "Performance Based Requirements," of 10 CFR 50, Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors."

2.0 DETAILED DESCRIPTION 2.1 Primary Containment and Shield Building Design Primary Containment The containment for DBNPS consists of three basic structures: a steel containment vessel, a reinforced concrete shield building, and the internal structures. The containment vessel is a cylindrical steel pressure vessel with hemispherical dome and ellipsoidal bottom. It is completely enclosed by a reinforced concrete shield building having a cylindrical shape with a shallow dome roof. An annular space is provided between the wall of the containment vessel and the shield building, and clearance is also provided between the containment vessel and the dome of the shield building. The containment vessel and shield building are supported on a concrete foundation founded on a firm rock structure. With the exception of the concrete under the containment vessel there are no structural ties between the containment vessel and the shield building above the foundation slab. Above this, there is unlimited freedom of differential movement between the containment vessel and the shield building. The containment internal structures are constructed of reinforced concrete and structural steel. These structures are isolated from the containment vessel by steel grating panels with sliding supports which allows free differential movement between the internal structures and the vessel. The internal structures are supported by the massive concrete fill within the containment vessel bottom head.

The containment vessel inside diameter is 130 feet and the net free volume is approximately 2,834,000 cubic feet. The cylindrical shell and bottom head thickness, exclusive of reinforced areas, is 1 1/2 inches with a dome thickness of 13/16 inch.

Access to the containment is provided by an equipment hatch, a personnel air lock, and

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 5 of 76 an emergency air lock. Electrical and mechanical penetrations are provided for services to the containment.

Penetrations:

Electrical Penetrations Modular type penetrations are used for electrical conductors passing through the steel containment vessel. The penetration modules are hollow cylinders through which the conductors pass. Each electrical penetration assembly is pressurized with dry nitrogen to maintain and monitor integrity, and to prevent the intrusion of moisture into the penetration. The penetration assemblies are installed in penetration sleeves welded into the wall of the containment vessel which are provided with bolting flanges. Mounting of the assemblies to the sleeves is accomplished by bolting. The headers through which the cables pass are hermetically sealed. Materials used in the design are selected for resistance to possible environment conditions.

Piping Penetrations Piping penetrations are divided into two general groups:

a. Type 1 Large diameter, high energy, hot piping

b. Type 2 General piping, small diameter, lower energy piping Type 1 Penetrations Type 1 penetrations are the main steam and main feedwater lines.

Each main steam and main feedwater containment penetration consists of the following l

I major components:

1. Process Pipe (main steam or main feedwater pipe)

2. Guard Pipe

3. Flued Head

4. Penetration bellows assembly Process Pipe:

The process pipe is made of welded carbon steel and is welded to the flued head.

Guard Pipe:

The guard pipe is made of welded carbon steel and is designed to contain the full pressure of the process pipe including jet effects. In case of passive failure of the process pipe, the guard pipe contains the process fluid and discharges it into the containment past the secondary shield wall and so it protects the containment vessel and the penetration bellows assembly from jet effects and over pressurization. One end of the guard pipe is welded to the flued head while the other end is open to the containment.

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 6 of 76 Flued Head:

The flued head is made from forged carbon steel. It is designed to contain the full pressure of the process fluid in the areas adjoining the process pipe and guard pipe and full containment pressure in parts adjoining the penetration bellows assembly. In addition, the flued heads are anchored and restrained at the Auxiliary Building and they are designed and analyzed to be capable of carrying loads resulting from the failure of the process pipe. The anchor and restraint of the flued head prevents axial displacement of the guard pipe and the penetration bellows assembly.

Penetration Bellows Assembly:

The penetration bellows assembly allows differential movement between the containment vessel and the Auxiliary Building. The penetration bellows assembly is an extension of the containment and is designed for containment pressure and displacement resulting from thermal expansion and seismic movements.

Type 2 Penetrations Type 2 penetrations are welded directly to the containment vessel nozzle or through double flued head, butt weld cap, or flat plate.

Flued heads are used for penetrations, where the configuration is such that forces and moments are transmitted from the differential movement between the penetrating pipe and the containment vessel. Butt weld cap and direct welding the penetrating pipe are used, where there is no differential movement between the containment vessel and the penetrating process pipe. Flat plate is used for below grade penetration where there is no differential movement between the containment vessel and the penetrating process pipe, due to the concrete embedding of both..

Equipment and Personnel Access The equipment hatch is a welded steel assembly, with a double-gasketed, flanged and bolted cover. Provision is made to pressurize the space between the double gaskets to 40 pounds per square inch gauge (psig). One personnel air lock and one emergency air lock are provided. These are welded steel assemblies. Each air lock has two double gasketed doors in series. Provision is made to pressurize the space between the gaskets. The doors are mechanically interlocked to ensure that one door cannot be opened until the second door is sealed. Provisions are made for deliberately violating the interlock by the use of special tools and procedures under strict administrative control. Each door is equipped with quick acting valves for equalizing the pressure across the doors. The doors are not operable unless the pressure is equalized.

Pressure equalization is possible from every point at which the associated door can be operated. The valves for the two doors are interlocked so that only one door can be opened at one time, and only when the opposite door is closed and sealed. Each door is designed so that with the other door open, it can withstand and seal against design and testing pressures of the containment vessel. There is visual indication outside each door showing whether the opposite door is open or closed. Provision is made outside

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 7 of 76 each door for remotely closing and latching the opposite door so that in the event that one door is accidentally left open, it can be closed by remote control. The access air locks have nozzles installed which permit pressure testing of the air lock at any time.

An interior lighting system and a communications system are installed; these systems are capable of operating from the essential power supply.

Fuel Transfer Penetrations Two fuel transfer penetrations are provided to transport fuel rods between the refueling canal and the spent fuel pool during refueling operations of the reactor. Each penetration consists of a 30-inch diameter stainless steel pipe installed inside a 42-inch sleeve. The inner pipe acts as the transfer tube. Provisions are made to provide integrity of containment, allowance for differential movement between structures and prevent leakage through the transfer tubes in the event of an accident.

Flexible Closures at Penetrations The design criteria for the flexible closures of the personnel air lock, emergency air lock, and the equipment hatch allow for temperature and pressure transients that could be experienced during the life of the station, the postulated tornado phenomena, or the loss of coolant accident. The flexible closures also accommodate the differential movements caused by either earthquakes, expansion during normal operating cycles, or loss of coolant accident. The closure meets the single failure criterion. The closure is designed for a temperature range of 30 degrees Fahrenheit (°F) to 150°F, a pressure differential of plus 3 pounds per square inch (psi) between the annular space and the penetration, both acting concurrently with the maximum differential movements that could be experienced with either earthquake or loss of coolant accident, or earthquake and loss of coolant accident.

Containment Vessel Relief Valves Inside the containment annulus space, ten vacuum relief valves are arranged in two groups of five penetrations on the containment vessel. They open when a vacuum begins to develop within the containment vessel. The valves are of the swinging-disc type with a self-alignment feature to permit the disc to seat squarely after each opening.

When closed, the valve holds against a pressure and temperature of 45 psig and 264°F, respectively. The vacuum relief valves are designed to start opening at 0.15 pounds per square inch differential (psid) pressure. The pressure loss through this system is below the 0.50 psi maximum differential allowed across the containment vessel.

Shield Building The shield building is a reinforced concrete structure of right cylinder configuration with a shallow dome roof. An annular space is provided between the steel containment vessel and the interior face of the concrete shield building of approximately 4.5 feet to permit construction operations and periodic visual inspection of the steel containment vessel. The volume contained within this annulus is approximately 678,700 cubic feet.

The shield building has a height of 279.5 feet measured from the top of the foundation

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 8 of 76 ring to the top of the dome. The thicknesses of the wall and the dome are approximately 2.5 feet and 2 feet, respectively.

The shield building completely encloses the containment vessel, the personnel access openings, the equipment hatch, and that portion of penetrations that are associated with primary containment. The design of the shield building provides for (1) biological shielding, (2) controlled release of the annulus atmosphere under accident condition, and (3) environmental protection of the containment vessel.

2.2 Containment Vessel Isolation Systems The general design bases governing isolation valve requirements for containment piping penetrations are as indicated in the following paragraphs.

Leakage through penetrations not serving accident-consequence-limiting systems is minimized by a double barrier so that no single, credible failure or malfunction of an active component can result in loss-of-isolation. The installed double barriers take the form of closed piping systems, both inside and outside the containment, and various types of isolation valves or flanges.

Containment vessel isolation valves are provided in lines penetrating the containment vessel to ensure that no uncontrolled release of radioactivity from the containment can occur, particularly following an accident with a radiation release.

Containment vessel isolation occurs on a safety features actuation signal. The isolation system closes penetrations not required for operation of the engineered safety features system, reactor coolant system makeup, or special exceptions. In addition, pneumatically-operated isolation valves, with the exception of those that are part of the engineered safety features, will fail closed: Motor-operated isolation valves, upon loss of normal and reserve electric power, are supplied with power from the emergency power system. Motor-operated isolation valves also have a manual override to be used in case of motor operator failure.

Isolation valves located outside the containment vessel are located as close to the containment vessel as practical. Upon loss of actuating power, the isolation valves are designed to maintain their present position or to take the position that provides the greater safety.

Control room operated containment isolation valves are provided with position indicating lights in the control room and either control switches or control and safety features actuation block switches in the control room.

To ensure the added reliability of containment integrity, the following penetration systems are designed in accordance with the American Society of Mechanical Engineers, Boiler and Pressure Vessel Code (ASME Code), Section Ill, Class 2, designed and analyzed as Seismic Class I, protected against missiles and high energy piping, suitably restrained so that passive failure of one component does not damage

FE NOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 9 of 76 adjacent components, and subjected to a strict quality assurance program to ensure that material and workmanship meet specifications:

a. Piping between the inside and outside isolation valves up to and including the valves.

b. In a closed system having only one isolation valve outside the containment, the entire system inside the containment up to and including the isolation valve.

2.3 Current Technical Specification Requirements Technical Specification Surveillance Requirement SR 3.6.1.1 requires the following surveillance on a frequency in accordance with the Containment Leakage Rate Testing Program.

Perform required visual examinations and leakage rate testing except for containment air lock testing, in accordance with the Containment Leakage Rate Testing Program.

DBNPS Technical Specification 5.5.15, "Containment Leakage Rate Testing Program,"

currently states, in part:

a. A program shall establish the leakage rate testing of the containment as required by 10 CFR 50.54(0) and 10 CFR 50, Appendix J, Option B, as modified by approved exemptions. For Type C [containment isolation valve leakage] tests, this program shall be in accordance with the guidelines contained in Nuclear Energy Institute (NEI) topical report NEI 94-01, Revision 3-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," dated July 2012. For Type A and Type B tests, 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 exceptions:

1. A reduced duration Type A test may be performed using the criteria and Total Time method specified in Bechtel Topical Report BN-TOP-1, Revision 1.

2. The fuel transfer tube blind flanges (containment penetrations 23 and 24) will not be eligible for extended test frequencies. Their Type B test frequency will remain at 30 months. However, as-found testing will not be required.

2.4 Reason for the Proposed Change The proposed license amendment would revise Technical Specification 5.5.15, "Containment Leakage Rate Testing Program," to permit the following:

Increase the existing Type A integrated leakage rate test program test interval from 10 years to 15 years in accordance with NEI 94-01, Revision 3-A (Reference 2), and the conditions and limitations specified in NEI 94-01, Revision 2-A (Reference 3).

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 10 of 76 Adopt a more conservative allowable test interval extension of nine months, for Type A and Type B leakage rate tests in accordance with NEI 94-01, Revision 3-A. If the due date of the Type A test falls between scheduled outages, this test interval extension allows the test to be performed during the next scheduled outage.

2.5 Description of the Proposed Change The proposed change to DBNPS Technical Specification 5.5.15 will replace the reference to Regulatory Guide 1.163 with a reference to the limitations and conditions specified in NEI 94-01, Revision 2-A (Reference 3) for Type A and B testing. The proposed change also involves removing the words, "For Type C tests," and "For Type A and Type B tests," as the Technical Specification will be expanded to ensure the Appendix J Program is in accordance with NEI 94-01, Revision 3-A for Types A, B and C testing, and the Revision 2-A limitations and conditions for Type A testing.

The proposed change will revise Technical Specification 5.5.15.a to state:

A program shall be established to implement the leakage rate testing of the containment 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 NEI 94-01, Revision 3-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR 50, Appendix J," and the conditions and limitations specified in NEI 94-01, Revision 2-A, as modified by the following exceptions:

1. A reduced duration Type A test may be performed using the criteria and Total Time method specified in Bechtel Topical Report BN-TOP-1, Revision 1.

2. The fuel transfer tube blind flanges (containment penetrations 23 and 24) will npt be eligible for extended test frequencies. Their Type B test frequency will remain at 30 months. However; as-found testing will not be required."

NEI 94-01, Revision 3-A, Section 8.0, "Testing Methodologies For Type A, B and C Tests," first paragraph states that:

Type A, Type B and Type C tests should be performed using the technical methods and techniques specified in ANSI/ANS-56.8-2002, or other alternative testing methods that have been approved by the NRC.

FENOC has chosen to adopt the use of American National Standards Institute and American Nuclear Society (ANSI/ANS) Standard 56.8-2002, "Containment System Leakage Testing Requirements," (Reference 4) for the performance of Type A and Type B testing.

Marked-up pages showing the proposed changes to Technical Specification 5.5.15 are provided in Attachment A.

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 11 of 76

3.0 TECHNICAL EVALUATION

The following technical evaluation provides supporting information used in evaluating and justifying the proposed Technical Specification change and addressing Nuclear Regulatory Commission (NRC) staff concerns discussed in similar applications. The technical evaluation provides information regarding credit for post-accident containment pressure on emergency core cooling system net positive suction head, containment vessel tests and inspections performed following major modifications, chronology of 10 CFR 50, Appendix J test requirements, current DBNPS containment leak test program requirements, Technical Specification 5.5.15.a exceptions 1 and 2, DBNPS containment leak test licensing history, DBNPS periodic Type A results, plant specific risk and non-risk assessments for extended test intervals, plant operating experience, license renewal aging management, and NEI 94-01, Revision 2-A, and Revision 3-A, limitations and conditions.

3.1 Emergency Core Cooling System Net Positive Suction Head The regulatory position of Regulatory Guide 1.1, "Net Positive Suction Head for Emergency Core Cooling and Containment Heat Removal System Pumps,"

November 1970, provides that emergency core cooling and containment heat removal systems should be designed so that adequate net positive suction head (NPSH) is provided to system pumps assuming maximum expected temperatures of pumped fluids and no increase in containment pressure from that present prior to postulated loss of coolant accidents. The emergency core cooling and containment heat removal system pumps at DBNPS are the high-pressure core injection pumps, low-pressure core injection pumps and containment spray pumps.

The NPSH for the low-pressure core injection pumps and the containment spray pumps, during the recirculation phase following a loss of coolant accident was calculated in accordance with the guidelines of Regulatory Guide 1.1. No credit has been taken for the containment pressure. It was assumed that the vapor pressure is equal to the containment pressure thus only a static head (difference in elevations between the pump centerline and emergency sump) was available. In addition, for added conservatism only a minimum water level inside containment was used.

If required in the recirculation mode, the high-pressure core injection pump takes suction from the discharge of the low-pressure core injection pump. Therefore, no credit has been taken for the containment pressure in determining the NPSH for the high-pressure core injection pumps..

3.2 Containment Vessel Tests and Inspections Performed Following Major Modifications The NRC's final safety evaluation for NEI 94-01, Revision 2, Section 3.1.4, "Major and Minor Containment Repairs and Modifications," states the following:

Section 9.2.4 of NEI TR 94-01, Revision 2, states that: "Repairs and modifications that affect the containment leakage integrity require LLRT

[local leakage rate tests] or short duration structural tests as appropriate

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 12 of 76 to provide assurance of containment integrity following the modification or repair. This testing shall be performed prior to returning the containment to operation." Article IWE-5000 of the ASME Code,Section XI, Subsection IWE (up to the 2001 Edition and the 2003 Addenda),

would require a Type A test after major repair or modifications to the containment. In general, the NRG staff considers the cutting of a large hole in the containment for replacement of steam generators or reactor vessel heads, replacement of large penetrations, as major repair or modifications to the containment structure. At the request of a number of licensees, the NRG staff has agreed to a relief request from the IWE requirements for performing the Type A test and has accepted a combination of actions consisting of ensuring that: (1) the modified containment meets the pre-service non-destructive evaluation (NOE) test requirements (i.e., as required by the construction code), (2) the locally welded areas are examined for essentially zero leakage using a soap bubble, or an equivalent, test, and (3) the entire containment is subjected to the peak calculated containment design basis accident pressure for a minimum of 10 minutes (steel containment) and 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> (concrete containment), and (4) the outside surfaces of concrete containments are visually examined as required by the ASME Code,Section XI, Subsection IWL, during the peak pressure, and that the outside and inside surfaces of the steel surfaces are examined as required by the ASME Code,Section XI, Subsection IWE, immediately after the test.

This is defined as a short duration structural test of the containment. For minor modifications (e.g., replacement or addition of a small penetration),

or modification of attachments to the pressure retaining boundary (i.e.,

repair/replacement of steel containment stiffeners), leakage integrity of the affected pressure retaining areas should be verified by a LLRT.

The NRC's final safety evaluation for NEI 94-01, Revision 2, Section 4.1, "Limitations and Conditions for NEI TR 94-01, Revision 2," states in item 4 that:

The licensee addresses any tests and inspections performed following major modifications to the containment structure, as applicable. (Refer to SE [safety evaluation] Section 3.1.4 ).

On three different occasions, a large hole has been cut in the DBNPS containment vessel to accommodate reactor head replacements (twice) and steam generator removal and replacement (once). The tests and inspections performed are described as follows:

The April 2003 Type A test was performed following th*e restoration of the containment vessel opening used in support of reactor head removal and replacement.

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 13 of 76 The November 2011 Type A test was performed following the restoration of the containment vessel opening used in support of reactor head removal and replacement.

FENOC replaced the DBNPS steam generators in the spring of 2014. These replacement activities required the opening of the containment vessel to provide access for the removal of the original steam generators as well as the installation of the replacement steam generators. Following replacement of these major components, the containment vessel was restored to its original design requirements.

Once the containment vessel had been restored, a leakage test in accordance with IWE-5223.4, as modified by 10 CFR 50.55a, Paragraph (b)(2)(ix)(J), was required.

However, due to the nature of the repair, which restores the containment vessel to ASME Code requirements, an effective post-repair test of containment structural and leaktight integrity was performed by an alternative containment leakage test in accordance with IWE-5223.4(a) as described in 10 CFR 50.55a relief request number RR-E1.

The alternative containment leakage test described in relief request RR-E1 was reviewed and authorized by the NRC in a letter dated May 8, 2013 (Reference 5).

This proposed alternative test was to be performed in lieu of the required Type A integrated leak rate test following restoration of the containment vessel pressure boundary.

Structural integrity and the leaktight integrity of the containment vessel repair were ensured by the performance of the alternative localized leakage bubble test. The containment vessel opening repair weld was bubble tested after pressurizing the entire containment to 38.0 psig (+1.50, -1 psig). The VT-2 bubble test of the repair weld was performed on the construction opening weld and 1/2 inch of base material (minimum) on both sides of the weld after a hold time of at least 15 minutes at a pressure of 38.58 psig. The test acceptance criterion was zero detectable leakage, which was determined by the absence of bubble formation, as observed from the annular space, using a leak detection medium in accordance with test procedures.

3.3 Justification for the Technical Specification Change 3.3.1 Chronology of 10 CFR 50, Appendix J, Testing Requirements The testing requirements of 10 CFR 50, Appendix J, provide assurance that leakage from the containment, including systems and components that penetrate the containment, do not exceed the allowable leakage values specified in the Technical Specification. 10 CFR 50, Appendix J also ensures that periodic surveillance of reactor containment penetrations and isolation valves are performed so that proper maintenance and repairs are made during the service life of the containment and those systems and components penetrating primary containment. The limitation on containment leakage provides assurance that the containment would perform its design

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 14 of 76 function following an accident up to and including the plant design basis loss of coolant accident.

Appendix J identifies three types of required tests: 1) Type A tests, intended to measure the primary containment overall integrated leakage rate; 2) Type B tests, intended to detect local leaks and to measure leakage across pressure-containing or leakage limiting boundaries (other than valves) for primary containment penetrations; and

3) Type C tests, intended to measure containment isolation valve leakage rates.

Type A tests focus on verifying the leakage integrity of a passive containment structure.

Type Band Type C tests (also referred to as local leakage rate tests or LLRTs) focus on assuring that containment penetrations are essentially leak tight. Type B and Type C tests identify the majority of potential containment leakage paths. Type A tests identify the overall (integrated) containment leakage rate and serve to ensure continued leakage integrity of the containment structure by evaluating those structural parts of the containment not covered by Type B and Type C testing.

In 1995, 10 CFR 50, Appendix J, was amended to provide a performance-based Option B for the containment leakage testing requirements. Option B requires that test intervals for Type A, Type B, and Type C testing be determined by using a performance-based approach. Performance-based test intervals are based bn consideration of the operating history of the component and resulting risk from its failure. The use of the term "performance-based" in 10 CFR 50, Appendix J refers to both the performance history necessary to extend test intervals as well as to the criteria necessary to meet the requirements of Option B.

Also in 1995, Regulatory Guide 1.163 (Reference 1) was issued. The Regulatory Guide endorsed NEI,94-01, Revision O (Reference 6), with certain modifications and additions.

Option B, in concert with Regulatory Guide 1.163 and NEI 94-01, Revision 0, allows licensees with a satisfactory Type A test performance history (that is, two consecutive, successful Type A tests) to reduce the test frequency for the containment Type A test from three tests in 10 years to one test in 10 years.

In 2008, NEI 94-01, Revision 2-A (Reference 3), was issued. This document describes an acceptable approach for implementing the performance-based requirements of Option B to 10 CFR 50, Appendix J, subject to the limitations and conditions noted in Section 4.0 of the June 25, 2008 NRC safety evaluation (Accession Number ML081140105) for NEI 94-01, Revision 2-A. NEI 94-01, Revision 2-A, delineates a performance-based approach for determining Type A, Type B, and Type C containment leakage rate surveillance testing frequencies. Justification for extending test intervals is based on the performance history and risk insights. The approach includes provisions for extending Type A test intervals to up to 15 years and incorporates the regulatory positions stated in Regulatory Guide 1.163 (Reference 1 ).

In 2012, NEI 94-01, Revision 3-A (Reference 2), was issued. The NRC staff determined that this document describes an acceptable approach for implementing the performance-based requirements of Option B to 10 CFR 50, Appendix J, as modified by

FE NOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 15 of 76 the conditions and limitations summarized in Section 4.0 of the June 8, 2012 NRC safety evaluation (Accession Number ML121030286) for NEI 94-01, Revision 3-A.

NEI 94-01, Revision 3-A, delineates a performance-based approach for determining Type A, Type B, and Type C containment leakage rate surveillance testing frequencies.

Justification for extending test intervals is based on the performance history and risk insights.

Extensions of Type B and Type C test intervals are allowed based upon completion of two consecutive periodic as-found tests where the results of each test are within a licensee's allowable administrative limits. Intervals may be increased from 30 months up to a maximum of 120 months for Type B tests (except for containment air locks) and up to a maximum of 75 months for Type C tests. If a licensee considers extended test intervals of greater than 60 months for TypeB or Type C tested components, the review should include the additional considerations of as-found tests, schedule and review as described in NEI 94-01, Revision 3-A, Section 11.3.2, "Programmatic Controls."

The NRC has provided guidance concerning the use of test interval extensions in the deferral of Type A tests beyond the 15-year interval, in the June 25, 2008 NRC safety evaluation for NEI 94-01, Revision 2-A. Section 3.1.1.2, "Deferral of Tests Beyond The 15:-Year Interval," of the safety evaluation states in part:

Section 9.2.3, NEI TR 94-01, Revision 2, states, "Type A testing shall be performed during a period of reactor shutdown at a frequency of at least once per 15 years based on acceptable performance history." However, Section 9.1 states that the "required surveillance intervals for recommended Type A testing given in this section may be extended by up to 9 months to accommodate unforeseen emergent.conditions but should not be used for routine scheduling and planning purposes." The NRC staff believes that extensions of the performance-based Type A test interval beyond the required 15 years should be infrequent and used only for compelling reasons. Therefore, if a licensee wants to use the provisions of Section 9.1 in TR NEI 94-01, Revision 2, the licensee will have to demonstrate to the NRC staff that an unforeseen emergent condition exists.

The NRC has also provided the following concerning the extension of Type A test (integrated leakage rate test or ILRT) intervals to 15 years in NEI 94-01, Revision 3-A, NRC safety evaluation, Section 4.0, Condition 2, which states, in part:

The basis for acceptability of extending the ILRT interval out to once per 15 years was the enhanced and robust primary containment inspection program and the local leakage rate testing of penetrations. Most of the primary containment leakage experienced has been attributed to penetration leakage and penetrations are thought to be the most likely location of most containment leakage at any time.

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 16 of 76 3.3.2 Current Containment Leakage Rate Testing Program Requirements Current Technical Specification 5.5.15.a requires that a program for the leakage rate testing of the containment be established to *comply with the requirements of 10 CFR 50.54(0) and 10 CFR 50, Appendix J, Option B, as modified by approved exceptions. The current DBNPS exceptions to 10 CFR Part 50, Appendix J, Option A, are specified in Technical Specifications 5.5.15.a.1 and 5.5.15.a.2 and listed below.

1. A reduced duration Type A test may be performed using the criteria and Total Time method specified in Bechtel Topical Report BN-TOP-1, Revision 1.

2. The fuel transfer tube blind flanges (containment penetrations 23 and 24) will not be eligible for extended test frequencies. Their Type 8 test frequency will remain at 30 months. However, as-found testing will not be required.

The current Technical Specification 5.5.15.a also requires that the containment leak rate testing program be in accordance with the guidelines contained in Reg4latory Guide 1.163, September 1995, for Type A and Type B testing. Regulatory Guide 1.163 endorses, with certain exceptions, NEI 94-01, Revision 0, as an acceptable method for complying with the provisions of 10 CFR 50, Appendix J, Option 8. For Type C tests, this program shall be in accordance with the guidelines contained in NEI 94-01, Revision 3-A.

Regulatory Positior.i C.1 of Regulatory Guide 1.163, states in part that licensees intending to comply with Option 8 should establish test intervals based upon the criteria

. in Section 11.0 of NEI 94-01, Revision O (Reference 6), rather than using test intervals specified in ANSI/ANS 56.8-1994. NEI 94-01, Section 11.1 indicates that extended test intervals in Sections 9.0 and 10.0 have been selected based on performance and have been assessed for risk impact using historical performance data.

• Section 9 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 1.0 La (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 8 performance-based containment leakage rate testing program altered the frequency of measuring primary containment leakage in Type A, 8, 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 frequency for Type A testing, as documented in NEI 94-01, is based in part upon a generic evaluation documented in NUREG-1493, "Performance-Based Containment Leak-Test Program - Final Report" (Reference 7). The evaluation documented in NUREG-1493 included a study of the dependence of reactor accident

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 17 of 76 risks on containment leak tightness for differing containment types. NUREG-1493 concluded in Section 10.1.2 that reducing the frequency of Type A tests from the original three tests per 10 years to one test per 20 years was found to lead to an imperceptible increase in risk. The estimated increase in risk is very small because Type A tests identify only a few potential containment leakage paths that cannot be identified by Type Band Type 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 concluded that increasing the interval between Type A tests is possible with minimal impact on public risk.

3.3.3 Support for the Continued Use of Exceptions 1 and 2 to Technical Specification 5.5.15 This license amendment request does not request a change to the Technical Specification 5.5.15 exceptions. The Bechtel Topical Report BN-TOP-1 testing methodology exception will be retained in Technical Specification 5.5.15.a.1, since this methodology is an acceptable option when used to perform Type A tests, and this methodology is not addressed in the Updated Safety Analysis Report. The leak testing provisions for fuel transfer tube blind flanges referenced in Technical Specification 5.5.15.a.2 continue to be applicable.

Based on the foregoing, the provisions of Technical Specification 5.5.15.a, exceptions 1 and 2, will be retained.

3.3.4 DBNPS Containment Leak Test Licensing History By letter dated February 22, 1996 (Reference 8), the NRC issued Amendment 205 to the DBNPS facility operating license to permit the irilplemen'tation of 1 O CFR 50, Appendix J, Option Bin accordance with the guidelines contained in Regulatory Guide 1.163, dated September 1995. Regulatory Guide 1.163 specifies a method acceptable to the NRC for complying with Option B. These changes are related only to Type A integrated leakage rate testing.

By letter dated March 28, 2000 (Reference 9), the NRC issued Amendment 240 to the DBNPS facility operating license to permit the implementation of 10 CFR 50, Appendix J, Option Band reference Regulatory Guide 1.163, dated September 1995. These changes are related only to Type Band Type C (local) leakage rate testing.

By letter dated May 8, 2013 (Reference 5), the NRC authorized the proposed alternative in relief request RR-E1 to bubble test a containment vessel opening repair weld after pressurizing the entire containment in lieu of the required Appendix J, Type A, integrated leak rate test, as described above in Section 3.2.

By letter dated October 9, 2015 (Reference 10), the NRC issued Amendment 288 to the DBNPS facility operating license to adopt the guidance in NEl-94-01, Revision 3-A, for Type C testing. This change allowed extension of the containment Type C test interval from 60 months up to 75 months, based on acceptable performance.

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 18 of 76 3.3.5 DBNPS Periodic Integrated Leakage Rate {Type A) Test Results Table 3.3.5-1 lists the past Periodic Type A test results for DBNPS.

Table 3.3.5-1, DBNPS Type A Test History Test Date Percent Containment Air Mass Per Day 1 September 1976 - Preoperational 0.11224 August 1980 0.16645 March 1985 0.08783 September 1988 0.052511 October 1991 0.062179 May 2000 0.0127 April 2003 0.16714 2 November 2011 0.0722 2 April 2014 No Evidence of Leakage Noted 3 Notes:

1. As specified in DBNPS Technical Specification 5.5.15, the maximum allowable containment leakage rate (La) at the calculated peak containment internal pressure for the design basis loss of coolant accident (Pa), shall be 0.50 percent of containment air weight per day.

2.' The Type A test also served to satisfy the containment vessel post modification testing required by NEI 94-01, Revision 0, paragraph 9.2.4, and described in Section 3.2.

3. The containment vessel was pressurized to perform a VT-2 bubble leak test of the containment vessel construction opening closure weld in accordance with NRC approved relief request RR-E1 described in Section 3.2.

The current Type A test interval for DBNPS is ten years. Verification of this interval is presented in Table 3.3.5-2. The acceptance criteria used for this verification is contained in NEI 94-01, Revision 2-A and Revision 3-A, Section 5.0, Definitions, set out below, except that the as-left Type A test acceptance criteria (less than or equal to 0.75 La) specified in Technical Specification 5.5.15.d.1 is used.

The performance 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 Band 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

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 19 of 76 the test because of excessive leakage must be factored into the performance determination. The performance criterion for Type A tests is a performance leak rate of less than 1.0 La.

The UCL, as defined in ANSI/ANS 56.8-1994, is calculated at the 95 percent confidence level in Table 3.3.5-2.

FE NOC Evaluation of the Proposed Amendment.

Davis-Besse Nuclear Power Station Page 20 of 76 Table 3.3.5-2, DBNPS Type A Test Results Verification of Current Extended Type A Test Interval As Left Min Test Method,

- Type A Test Upper95%

Level Pathway Penalty for Test Confidence Level Isolated Pathways Date Data Analysis Acceptance (wt.%/day)

Corrections and (Leakage Techniques Criteria (wt.%/day)

(Test Pressure 1)

Improvements)

(wt.%/day)

April Absolute Method, 0.75 La 0.1627 BN-TOP-1 R1 0.0 0.00444 2003 Total Time (0.375 wt.%/day)

(38.3454 psig)

November Absolute Method, 0.75 La 0.0675 0.0031 2011 BN-TOP-1 R1 (0.375 wt.%/day)

(38.2949 psig) 0.0 (0.0016)3 Total Time Notes:

Adjusted As-Left Leak Rate (wt.%/day) 0.16714 0.0706

1. The calculated peak containment internal pressure for the design basis loss of coolant accident (Pa) is 38 psig.

2. The abbreviation % means percent, and the abbreviation wt. %/day means weight-percent per day.

3. The leakage improvements are the net sum of positive differences between the as found minimum pathway leakage rate and the as left minimum pathway leakage rate for penetrations not aligned to the containment integrated leakage rate test volume and whose leakage was improved by maintenance or similar action.

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 21 of 76 3.4 Plant Specific Confirmatory Analysis A confirmatory analysis was performed to provide a DBNPS plant specific risk assessment of permanently extending the allowed containment Type A test interval from ten to fifteen years. Attachment B contains the plant specific risk assessment conducted to support this proposed change.

3.4.1 Methodology The risk assessment follows the guidelines from the following documents:

NEI 94-01, Revision 3-A (Reference 2),

the NEI "Interim Guidance for Performing Risk Impact Assessments in Support of One-Time Extensions for Containment lnt~grated Leakage Rate Test Surveillance Intervals," November 2001 (Reference 11 ),

the NRG regulatory guidance on the use of probabilistic risk assessment (PRA) as stated in Regulatory Guide 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk Informed Activities,"

Revision 2, as applied to Type A test interval extensions, risk insights in support of a request for a change to the plant's licensing basis as outlined in Regulatory Guide 1.17 4, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 3 (Reference 12),

the methodology used at Calvert Cliffs to estimate the likelihood and risk implications of corrosion-induced leakage of steel liners going undetected during the extended test interval (Reference 13), and the methodology used in Electric Power Research Institute (EPRI) Report Number 1018243, "Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals," (Revision 2-A of EPRI Report Number 1009325, Reference 14).

The NRG report on performance-based leak testing, NUREG-1493, analyzed the effects of containment leakage on the health and safety of the public, and the benefits realized from the containment leak rate testing. In that analysis, it was determined that for a representative pressurized water reactor (PWR) plant (that is, SurfY), containment isolation failures contribute less than 0.1 percent to the latent risks from reactor accidents. Consequently, it is desirable to show that extending the Type A test interval will not lead to a substantial increase in risk from containment isolation failures for DBNPS.

NEI 94-01, Revision 3-A supports using EPRI Report Number 1009325, Revision 2-A (EPRI Report Number 1018243, Reference 14), for performing risk impact assessments in support of Type A test interval extensions. The guidance provided in Appendix H of EPRI Report Number 1009325, Revision 2-A builds on the EPRI Risk Assessment methodology, EPRI R-104285, "Risk Impact Assessment of Revised Containment Leak

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 22 of 76 Rate Testing Intervals," dated August 1994. This methodology is followed to determine the appropriate risk information for use in evaluating the impact of the proposed Type A test changes.

In the NRC safety evaluation dated June 25, 2008 (Reference 15), the NRC concluded that the methodology in EPRI TR-1009325, Revision 2 (Reference 14), was acceptable for referencing by licensees proposing to amend their Technical Specifications to permanently extend the Type A test surveillance interval to 15 years, subject to the limitations and conditions noted in Section 4.2 of the safety evaluation. Table 3.4.1-1 below addresses each of the four (4) limitations and conditions from safety evaluation Section 4.2.

Table 3.4.1-1, EPRI Report No. 1009325, Revision 2, Limitations and Conditions Limitation and Condition DBNPS Response (From Section 4.2 of Safety Evaluation)

1. The licensee submits documentation DBNPS PRA technical adequacy is indicating that the technical adequacy addressed in Section 3.4.2 of this license of their PRA is consistent with the amendment request and Attachment B, requirements of RG [Regulatory "Evaluation of Risk Significance of Guide] 1.200 relevant to the ILRT Permanent ILRT Extension."

extension application.

2.a The licensee submits documentation Because the Type A test does not impact indicating that the estimated risk core damage frequency (CDF), the increase associated with permanently extending the ILRT surveillance relevant criterion is large early release frequency (LERF). The increase in '

interval to 15 years is small, and LERF resulting from a change in the consistent with the clarification Type A test interval from 3 in 10 years to provided in Section 3.2.4.5 of this 1 in 15 years is estimated as 4.83E-8 per safety evaluation.

year using the EPRI guidance. This value increases negligibly if the risk impact of corrosion-induced leakage of the steel liners occurring and going undetected during the extended test interval is included. Therefore, the estimated change in LERF is determined to be "very small" using the acceptance guidelines of RG 1.17 4.

The risk change resulting from a change in the Type A test interval from 3 in 10 years to 1 in 15 years bounds the 1 in 10 years to 1 in 15 years risk change.

Considering the increase in LERF resulting from a change in the Type A test interval from 1 in 10 years to 1 in 15

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 23 of 76 Table 3.4.1-1, EPRI Report No. 1009325, Revision 2, Limitations and Conditions Limitation and Condition DBNPS Response (From Section 4.2 of Safety Evaluation) years is estimated as 2.01 E-8, the risk increase is "very small" using the acceptance guidelines of RG 1.17 4.

When external event risk is included, the increase in LERF resulting from a change in the Type A test interval from 3 in 10 years to 1 in 15 years is estimated as 6.06E-7 per year using the EPRI guidance, and total LERF is 5.50E-6 per year. As such, the estimated change in_

LERF is determined to be small using the acceptance guidelines of RG 1.17 4.

The risk change resulting from a change in the Type A test interval from 3 in 10 years to 1 in 15 years bounds the 1 in 10 years to 1 in 15 years risk change.

When external event risk is included, the increase in LERF resulting from a change in the Type A test interval from 1 in 10 years to 1 in 15 years is estimated as 2.53~-7 and the total LERF is 5.15E-

6. Therefore, the risk increase is small using the acceptance guidelines of RG

\\ 1.17 4 (See Attachment B, Section 7.0 of this submittal.)

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 24 of 76 Table 3.4.1-1, EPRI Report No. 1009325, Revision 2, Limitations and Conditions Limitation and Condition DBNPS Response (From Section 4.2 of Safety Evaluation) 2.bSpecifically, a small increase in The effect resulting from changing the population dose should be defined as Type A test frequency to 1 per 15 years, an increase in population dose of less measured as an increase to the total than or equal to either 1.0 person-rem integrated plant risk for those accident per year or 1 percent of the total sequences influenced by Type A testing, population dose, whichever is less is 0.016 person-rem per year. NEI 94-01 restrictive.

states that a small population dose is defined as an increase of less than or equal to 1.0 person-rem per year, or less than or equal to 1 percent of the total population dose, whichever is less restrictive for the risk impact assessment of the extended Type A test intervals.

The results of this calculation meet these criteria. Moreover, the risk impact for the Type A test extension when compared to other severe accident risks is negligible (Attachment B, Section 7.0 of this submittal has additional information.)

2.c In addition, a small increase in The increase in the conditional conditional containment failur~

con,tainment failure probabil.ity from the probability (CCFP) should be defined 3 in 10-year interval to 1 in 15-year as a value marginally greater than that interval is 0.870 percent. NEI 94-01 accepted in previous one-time 15-year states, in part, that increases in CCFP of ILRT extension requests. This would less than or equal to 1.5 percent are require that the increase in CCFP be small. Therefore, this increase is judged less than or equal to 1.5 percentage to be small. (Attachment B, Section 7.0 point.

of this submittal has additional information.)

3. The methodology in EPRI Report No.

The representative containment leakage 1009325, Revision 2, is acceptable for Class 3b sequences is 100 La based except for the calculation of the on the guidance provided in EPRI Report increase in expected population dose No. 1009325, Revision 2-A (EPRI Report (per year of reactor operation). In No. 1018243) (Attachment B, Section 4.0 order to make the methodology of this submittal has additional acceptable, the average leak rate information.)

accident case (accident case 3b) used by the licensees shall be 100 La instead of 35 La.

FE NOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 25 of 76 Table 3.4.1-1, EPRI Report No. 1009325, Revision 2, Limitations and Conditions Limitation and Condition DBNPS Response (From Section 4.2 of Safety Evaluation)

4. A license amendment request (LAR) is DBNPS does not rely upon containment required in instances where over-pressure for ECCS performance.

containment over-pressure is relied (Reference Section 3.1 of this submittal.)

upon for ECCS [Emergency Core Cooling System] performance.

3.4.2 Technical Adequacy of the Probabilistic Risk Analysis (PRA) 3.4.2.1 Internal Events PRA Quality Statement for Permanent 15-Year Integrated Leakage Rate Test Extension The DBNPS PRA model of record and supporting documentation have been maintained as a living program, with updates directed every other refueling cycle (approximately every four years) to reflect the as-built, as-operated plant. The PRA model (Reference 16) currently includes internal events, internal flooding, and seismic. Level 1 and Level 2 results are provided via this model. A fire model, which is based on the internal events model, has also been developed for implementation at DBNPS to support risk-:informed applications, and will be subject to the same configuration controls described below.

Interim updates may be prepared and issued in between regularly scheduled model updates on an as needed basis. Typically, an interim revision would be used for an

, update that would cause a change in core damage frequency of greater than 10 percent, a change in large early release frequency of greater than 20 percent, or for changes that could significantly impact a risk-informed application. Interim models may also be released following focused peer reviews once the associated findings and suggestions have been addressed. Under the FENOC PRA Program, if a portion of the model has been upgraded to satisfy the PRA standard (for example, internal flooding models at DBNPS), that portion of the model will not be released until after a focused peer review has been conducted and any findings and suggestions have been addressed.

The DBNPS model is highly detailed and includes a wide variety of initiating events, modeled systems, operator actions, and common cause events. The PRA quantification process used is based on the large linked fault tree methodology, which is a well-known and accepted methodology in the industry. The model is maintained and quantified using the EPRI Integrated Risk Technologies suite of software programs.

FENOC employs a multi-faceted, structured approach for establishing and maintaining the technical adequacy and plant fidelity of the PRA models for FENOC nuclear generation sites, including DBNPS. This approach includes a proceduralized PRA maintenance and update process, as well as the use of self-assessments and independent peer reviews.

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 26 of 76 3.4.2.2 PRA Maintenance and Update The FENOC risk management process ensures that the DBNPS PRA Model is an accurate reflection of the as-built, and as-operated plant. This process is defined in the FENOC PRA Program, which consists of a governing procedure, titled "Probabilistic Risk Assessment Program," and subordinate implementation procedures.

The Probabilistic Risk Assessment Program procedure serves as the higher tier procedure and establishes the FENOC PRA Program and provides administrative requirements for the maintenance and upgrade of the FENOC PRA models and risk-informed applications. The objective of the PRA Program is to provide technically adequate PRA models and risk-informed applications for risk-informed decision-making, The objectives satisfy Regulatory Guide 1.200 requirements to provide technically adequate results for risk-informed applications.

Working in conjunction with the above procedure, a business practice document, titled "Probabilistic Risk Assessment Model Management," establishes the administrative and technical requirements for the maintenance and upgrade of the FENOC PRA models.

A procedurally controlled process is used to maintain configuration control of the DBNPS PRA Model, data, and software. In addition to model control, administrative mechanisms are in place to assure that plant modifications, procedure changes relevant to the PRA, changes to calculations, and industry operating experience are appropriately screened, dispositioned, and tracked for incorporation into the PRA model if that change would impact the model. As part of this process, if any proposed changes are identified, which are perceived to significantly increase or decrease risk, they are incorporated into a working model (given their known level of detail at the time), and the results are compared to the effective model of record to identify if the proposed change should be pursued. These processes help to assure that the DBNPS PRA reflects the as-built, as-operated plant within the limitations of the PRA methodology, and that the significance of future expected changes or enhancements are understood and m~naged.

The interfacing process involves an ongoing solicitation of review of any changes that may have an impact upon the PRA model. Any changes to the PRA model or its supporting documentation are captured within a tracking database for PRA implementation tracking and future disposition. Additionally, the PRA staff provides the top risk significant operator actions to the operations training staff, for simulator validation to ensure that the current human reliability modeling reflects actual expected response and timing.

3.4.2.3 Applicability of Peer Review Facts and Observations The technical acceptability of the DBNPS PRA models has been demonstrated by the peer review process. The purpose of the industry PRA peer review process is to provide a method for establishing the technical capability and adequacy of a PRA relative to expectations of knowledgeable practitioners, using a set of guidance that establishes a set of minimum requirements. PRA peer reviews continue to be performed as PRAs are updated (and upgraded) to ensure the ability to support risk-informed applications and

FE NOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 27 of 76 has proven to be a valuable process for establishing technical adequacy of nuclear power plant PRAs.

3.4.2.4 Internal Events PRA Model The DBNPS Internal Events PRA Model has been peer reviewed in accordance with the guidance and NRG-endorsed standards in effect at the time of each of the reviews as follows:

April 2008 - full scope peer review addressing technical elements excluding large early release and internal flooding based on Regulatory Guide 1.200, Revision 1, and American Society of Mechanical Engineers (ASME) Standard RA-Sb-2005 (Reference 17)

October 2011 - focused scope peer review addressing technical element large early release frequency based on Regulatory Guide 1.200, Revision 2 (Reference 18), and American Society of Mechanical Engineers and American Nuclear Society (ASME/ANS) Standard RA-Sa-2009 (Reference 19)

July 2012 - focused scope peer review addressing technical element internal flooding based on Regulatory Guide 1.200, Revision 2, and ASME/ANS RA-Sa-2009 October 2017 - conditional containment failure focused scope peer review for method upgrade identified during independent assessment October 2017 - independent assessment of facts and observations (F&Os) and closeout review based on Appendix X of NEI 05-04 (Reference 20); any re-assessed supporting requirements were based on Regulatory Guide 1.200, Revision 2, and ASME/ANS RA-Sa:..2009 The reviews and assessments listed above are discussed in more detail below.

The most recent full-scope peer review of the DBNPS Internal Events PRA Model was the April 2008 gap assessment against the ASME RA-Sb-2005, "Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications," and Regulatory Guide 1.200, Revision 1. Typically, a gap assessment is performed to review the differences between two versions of the PRA ASME standard. However, the purpose of the 2008 gap assessment was to assess the current status of the internal events PRA with regard to the supporting requirements, assign an indication of the significance of the gaps and F&Os identified (levels A to D), describe the scope of effort needed to close the gap and F&O to achieve capability category II of the supporting requirement, and estimate the resources necessary to accomplish closure of the gap and F&O. Thus, this gap assessment is far more encompassing than a typical peer review of the PRA model.

Requirements for conducting a peer review were met.

The review team consisted of five independent reviewers qualified in accordance with NEI 05-04, "Process for Performing Follow-on PRA Peer Reviewers Using the ASME PRA Standard," January 2005, and they evaluated the current status of the PRA against the requirements in ASME RA-Sb-2005 and Regulatory Guide 1.200, Revision 1.

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 28 of 76 The review was a full scope internal events level 1 review, excluding large early release and internal flooding, but covering initiating events, accident sequences, success criteria, system analysis, human reliability analysis, data analysis, quantification and results, and the maintenance and update technical areas of the ASME RA-Sb-2005 PRA standard. If the PRA model and documentation did not meet at least capability category II of ASME RA-Sb-2005, an F&O was identified.

Because the peer review in 2008 predated the current Regulatory Guide 1.200 endorsed internal events standard, a comparison was made between the ASME/ANS RA-Sb-2005 standard used for the peer review and the endorsed ASME/ANS RA-Sa-2009 standard.

For the majority of the supporting requirements, the wording was unchanged, or the intent remained the same. There were no changes to the supporting requirements that were found to have an impact on the Internal Events PRA Model. Therefore, the 2008 peer review and the review of the updated PRA standard combined to assure the current Internal Events PRA Model satisfies capability category II of the current Regulatory Guide 1.200 endorsed internal events standard ASME/ANS RA-Sa-2009.

The 2008 gap assessment did not review the technical areas of large early release frequency and internal flooding. These two technical areas had separate focused scope peer reviews in 2011 (large early release frequency) and 2012 (internal flooding) against the relevant portions of the ASME/ANS-RA-Sa-2009 standard.

F&Os originating from the internal events, large early release, and internal flooding peer reviews were addressed using FENOC's PRA program to disposition each individual F&O, thus ensuring the model satisfies the PRA standard requirements.

The October 2017 independent assessment and closeout review addressed FENOC's disposition to F&Os from the three peer reviews discussed above. The independent assessment was conducted consistent with "N El 05-04/07-12/12-06 Appendix X: Close Out of Facts and Observations (F&Os)," (Reference 20, 21 and 22). Each member of the independent assessment team met the ASME standard criteria for independence from the DBNPS PRA and the relevant peer reviewer qualifications for the F&Os being reviewed. The team reviewed each F&O documented closure to determine if the F&O had been adequately addressed and could therefore be closed out using the appropriate parts of the ASME/ANS RA-Sa-2009 PRA standard. The relevant supporting requirements were also re-assessed for cases where the peer review identified the supporting requirement as not meeting capability category II. The independent assessment for internal events PRA closed the F&Os and determined that each associated supporting requirement meets at least capability category II.

During the independent assessment, the F&O dispositions for two supporting requirements (SY-B4 and DA-D5) related to common cause failure modeling were determined by the independent assessment team to be an upgrade, rather than an update. A focused scope peer review was held, and the team concluded these two supporting requirements were met to at least capability category 11 with no new findings.

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 29 of 76 Ultimately, the independent assessment and subsequent focused scope peer review for internal events PRA closed all but one of the F&Os and determined that all but one of the supporting requirement met capability category II or better. Finding SY-B11, associated with supporting requirement SY-B10 was found to remain open. Supporting requirement SY-B10 requires that those systems that are required for initiation and actuation of a system be identified. Because the PRA model justifies excluding some actuation logic due to negligible contribution for the emergency diesel generator, instead of explicitly modeling this logic, this supporting requirement was concluded to meet capability category I only. Since NEI 94-01 endorses using PRA models conformed to capability category I of the ASME/ANS standard, the DBNPS PRA model is of sufficient quality to use for this Type A test analysis. However, the F&O remains open, since it is desired that all supporting requirements meet capability category II.

Therefore, each supporting requirement of the internal events standard, other than SY-B10, has been determined to meet at least capability category II by the peer review team or the independent assessment team. SY-B10 meets capability category I, with an open F&O to support meeting capability category II. There are no other open F&Os remaining.

3.4.2.5 Fire PRA Model The DBNPS Fire PRA Model was peer reviewed by the PWR Owners Group in June 2013. The review was performed against the requirements of ASME/ANS RA-Sa-2009, Part 4, including clarifications and qualifications provided in the NRC endorsement of the standard contained in Regulatory Guide 1.200, Revision 2. The peer review was performed using the process defined in NEI 07-12, Revision 1 (Reference 21 ).

F&Os originating from the Fire PRA peer review were addressed using FENOC's PRA program to disposition each individual F&O, thus ensuring the model satisfies the PRA standard requirements.

The October 2017 independent assessment and closeout review addressed FENOC's disposition to the F&Os from the Fire PRA peer review. The independent assessment was conducted consistent with "NEI 05-04/07-12/12-06 Appendix X: Close Out of Facts and Observations (F&Os)." Each member of the independent assessment team met the ASME standard criteria for independence from the DBNPS Fire PRA and the relevant peer reviewer qualifications for the F&Os being reviewed. The team reviewed each F&O documented closure to determine if the F&O had been adequately addressed, and therefore could be closed out using the appropriate parts of the ASME/ANS RA-Sa-2009 PRA standard. The relevant supporting requirements were also re-assessed for cases where the peer review identified the supporting requirement as not meeting at least capability category II. The independent assessment for Fire PRA determined that each associated supporting requirement meets at least capability category II.

During the independent assessment, the F&O dispositions associated with two supporting requirements (PRM-B14 and PRM-B15) related to new accident progressions were determined by the independent assessment team to be an upgrade, rather than an

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 30 of 76 update. In addition, at the request of FENOC, three supporting requirements (CF-A 1, CF-A2, and CF-B1) related to direct current hot short methodology were requested to be addressed by a focused scope review. A focused scope peer review was held, and the team concluded that these supporting requirements were met to at least capability category II with no new findings.

There were five finding level F&Os (ES-A 1-01, ES-A 1-02, ES-A 1-03, ES-D1-01, and FQ-A 1-01) that remained open after the independent assessment, which are described in Attachment B, Table A-1, "DBNPS Fire PRA Focused-Scope Peer Review Facts &

Observations," of this submittal. The independent assessment also identified a suggestion level F&O to improve the PRA documentation.

F&Os ES-A1-01, ES-A1-02, ES-A1-03, and FQ-A1-01 have been resolved by making appropriate changes to the PRA model as identified in the F&O. The associated supporting requirements ES-A 1 and FQ-A 1 continue to be met, as identified in the independent assessment team report.

The remaining F&O ES-D1-01, is identified by the independent assessment team as documentation issue only. Supporting requirement ES-D1 continues to be met, as identified in the independent assessment team report.

Therefore, each supporting requirement of the Fire PRA standard has been determined to meet at least capability category II by the peer review team or the independent assessment team. The remaining five open finding level F&Os for the Fire PRA have been resolved or involve only a documentation update.

3.4.2.6 Seismic PRA Model I

The DBNPS Seismic PRA Model was peer reviewed by the PWR Owners Group peer review program in July 2014. The review was performed against the requirements of ASME/ANS RA-Sa-2009, Part 5, including clarifications and qualifications provided in the NRC endorsement of the standard contained in Regulatory Guide 1.200, Revision 2. The peer review was performed using the process defined in NEI 12-13, Revision 0 (Reference 22).

F&Os originating from the Seismic PRA peer review were addressed using FENOC's PRA program to disposition each individual F&O, thus ensuring the model satisfies the PRA standard requirements.

The October 2017 independent assessment and closeout review addressed FENOC's disposition to the F&Os from the Seismic PRA peer review. The independent assessment was conducted consistent with "N El 05-04/07-12/12-06 Appendix X: Close Out of Facts and Observations (F&Os)." Each member of the independent assessment team met the ASME standard criteria for independence from the DBNPS Seismic PRA and the relevant peer reviewer qualifications for the F&Os being reviewed. The team reviewed each F&O documented closure to determine if the F&O had been adequately addressed and could therefore be closed out using the appropriate parts of the ASME/ANS RA-Sa-2009 PRA standard. The relevant supporting requirements were also

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 31 of 76 re-assessed for cases where the peer review identified the supporting requirement as not meeting capability category II. The independent assessment for Seismic PRA closed the F&Os and determined that each associated supporting requirement meets at least capability category II.

During the independent assessment, the F&O dispositions associated with four supporting requirements (SHA-G1, SHA-H1, SHA-11, and SPR-E6) related to seismic hazard assessment and large early release analysis, were determined by the independent assessment team to be an upgrade rather than an update. A focused scope peer review was held, and the team concluded these supporting requirements were met to at least capability category II with no new findings. Therefore, each supporting requirement of the seismic PRA standard has been determined to meet at least capability category II by the peer review team or the independent assessment team, and there are no remaining open F&Os.

3.4.2.7 Consistency with Applicable PRA Standards Based on the peer reviews, independent assessment of F&O resolutions, the focused scope peer reviews, and the disposition of the remaining five open findings for the Fire PRA, FENOC concludes that the current DBNPS Internal Events, Fire, and Seismic PRA Models conform to capability category II of ASME RA-Sb-2009, ASME/ANS Standard 1for Probabilistic Risk Assessment of Nuclear Power Plant Applications, (Reference 23), as endorsed by Regulatory Guide 1.200, Revision 2, for supporting requirements except SY-B10, which conforms to capability category I. Since NEI 94-01 (Reference 3) endorses using PRA models conformed to capability category I of the ASME/ANS standard, and supporting requirements meet or exceed capability category I, using these models for this Type A test analysis meets technical adequacy requirements.

3.4.3 Summary of Plant-Specific Risk Assessment Results The findings of the DBNPS Risk Assessment contained in Attachment B of this submittal, confirm the general findings of previous studies that the risk impact associated with extending the Type A test interval from 3 in 10 years to 1 in 15 years is small.

Based on the results from Attachment B, Section 5.2, and the sensitivity calculations presented in Attachment B, Section 5.3, the following conclusions regarding the assessment of the plant risk are associated with extending the Type A test frequency to 15 years:

Regulatory Guide 1.17 4 (Reference 12) provides guidance for determining the risk impact of plant-specific changes to the licensing basis. Regulatory Guide 1.17 4 defines very small changes in risk as resulting in increases of core damage frequency less than 1.0E-06 per year and increases in large early release frequency less than 1.0E-07 per year. Since the Type A test does not impact core damage frequency, the relevant criterion is large early release frequency. The increase in large early release frequency resulting from a change in the Type A test interval from 3 in 10 years to 1 in 15 years is estimated as 4.83E-8 per year using the EPRI guidance; this value increases negligibly if the risk impact of corrosion-induced leakage of the steel liners occurring and going undetected during the extended test interval is included.

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 32 of 76 Therefore, the estimated change in large early release frequency is determined to be very small using the acceptance guidelines of Regulatory Guide 1.17 4. The risk change resulting from a change in the Type A test interval from 3 in 10 years to 1 in 15 years bounds the 1 in 10 years to 1 in 15 years risk change. Considering the increase in large early release frequency resulting from a change in the Type A test interval from 1 in 10 years to 1 in 15 years is estimated as 2.01 E-8, the risk increase is very small using the acceptance guidelines of Regulatory Guide 1.17 4.

When external event risk is included, the increase in large early release frequency resulting from a change in the Type A test interval from 3 in 10 years to 1 in 15 years is estimated as 6.06E-7 per year using the EPRI guidance, and total large early release frequency is 5.50E-6 per year. As such, the estimated change in large early release frequency is determined to be small using the acceptance guidelines of Regulatory Guide 1.17 4. The risk change resulting from a change in the Type A test interval from 3 in 10 years to 1 in 15 years bounds the 1 in 10 years to 1 in 15 years risk change. When external event risk is included, the increase in large early release frequency resulting from a change in the Type A test interval from 1 in 10 years to 1 in 15 years is estimated as 2.53E-7 and the total large early release frequency is 5.15E-

6. Therefore, the risk increase is small using the acceptance guidelines of Regulatory Guide 1.174.

The effect resulting from changing the Type A test frequency to 1-per-15 years, measured as an increase to the total integrated plant risk for those accident sequences influenced by Type A testing, is 0.016 person-rem per year. NEI 94-01 (Reference 3) states that a small population dose is defined as an increase of less than or equal to 1.0 person-rem per year, or less than or equal to 1 percent of the total population dose, whichever is less restric;tive for the risk impact assessment of the extended Type A test intervals. The results of this calculation meet these criteria.

Moreover, and the risk impact for the Type A test interval extension when compared to other severe accident risks is negligible.

The increase in the conditional containment failure probability from 3 in a 10-year interval to 1 in a 15-year interval is 0.870 percent. NEI 94-01 (Reference 3) states, in part, that increases in conditional containment failure probability of less than or equal to 1.5 percent are small. Therefore, this increase is judged to be small.

Therefore, the risk impact of increasing the Type A test interval to 15 years is considered to be small since it represents a small change to the DBNPS risk profile.

3.4.4 Previous Assessments In NUREG-1493, the NRC previously concluded that:

Reducing the frequency of Type A tests from three per 10 years to one per 20 years was found to lead to an imperceptible increase in risk. The estimated increase in risk is very small because Type A tests identify only a few potential containment leakage paths that cannot be identified by Type B or Type C testing, and the leaks that have been found by Type A tests have been only marginally above existing requirements.

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 33 of 76 Given the insensitivity of risk to containment leakage rate and the small fraction of leakage paths detected solely by Type A testing, increasing the interval between integrated leakage-rate tests is possible with minimal impact on public risk. The impact of relaxing the Type A test frequency beyond one in 20 years has not been evaluated. Beyond testing the performance of containment penetrations, Type A tests also test containment structure integrity.

The conclusions for DBNPS confirm these general conclusions on a plant-specific basis considering the severe accidents evaluated for DBNPS, the DBNPS containment failure modes, and the local population surrounding DBNPS.

Details of the DBNPS risk assessment are contained in Attachment B of this submittal.

3.4.5 Regulatory Guide 1.17 4, Revision 3 - Defense-in-Depth Evaluation Regulatory Guide 1.17 4, Revision 3 (Reference 12), describes an approach that is acceptable for developing risk-informed applications for a licensing basis change that considers engineering issues and applies risk insights. One of the considerations included in Regulatory Guide 1.17 4, Revision 3, is defense-in-depth. Defense-in-depth is a safety philosophy that employs successive compensatory measures to prevent accidents or mitigate damage if a malfunction, accident, or naturally caused event occurs at a nuclear facility.

The seven considerations presented in Regulatory Guide 1.17 4, Revision 3, Section 2.1.1.2, "Considerations for Evaluating the Impact of the Proposed Licensing Basis Change on Defense in Depth," are used to evaluate the proposed licensing basis change for overall impact on defense-in-depth. Each of the seven considerations are presented below and are followed by two FENOC responses. TJ,e first response addresses defense-in-depth considerations and the second response addresses PRA considerations.

Current Technical Specification 5.5.15 indicates that for Type C tests the containment leakage rate testing program shall be in accordance with the guidelines contained in NEI 94-01, Revision 3-A. NEI 94-01, Revision 3-A, guidelines allow extension of the containment Type C test interval up to 75 months, based on acceptable performance.

The impact of Type C testing in accordance with NEI 94-01, Revision 3-A, was considered in the following defense-in-depth evaluation.

1. Preserve a reasonable balance among the layers of defense.

A reasonable balance of the layers of defense (that is, minimizing challenges to the plant, preventing any events from progressing to core damage, containing the radioactive source term, and emergency preparedness) helps to ensure an apportionment of the plant's capabilities between limiting disturbances to the plant and mitigating their consequences. The term "reasonable balance" is not meant to imply an equal apportionment of capabilities. The NRC recognizes that aspects of a plant's design or operation might cause one or more of the layers of defense to be adversely affected. For these situations, the balance between the other layers of defense becomes especially

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 34 of 76 important when evaluating the impact of the proposed licensing basis change and its effect on defense-in-depth.

Defense-In-Depth Consideration 1 Response:

Layers of defense-in-depth described in Regulatory Guide 1.17 4, Revision 3, were evaluated to determine if the proposed change continues to preserve a balance among the layers. The proposed change to containment leakage rate testing program requirements:

Does not affect the design or operation, or create new failure modes for plant systems, structures, and components, and does not significantly increase the likelihood of initiating events or create new significant initiating events.

Does not impact the availability and reliability of systems, structures, and components providing the safety functions that prevent plant challenges from progressing to core damage.

Does not reduce the effectiveness of the emergency preparedness program, including the ability to detect and measure releases of radioactivity, notify offsite agencies and the public, and make recommendations to shelter or evacuate the public as necessary.

The containment leakage rate testing program requirements specified in plant technical specifications are provided to ensure the DBNPS containment structure, penetrations, isolation valves, and mechanical seal systems, continue to perform their intended safety function. The proposed change to these containment leakage rate testing program requirements does not significantly impact the containment function or systems, structures, and components supporting that function, as described below.

The purpose of the proposed change is to extend the interval of the Type A test from 10 years to 15 years. As described in Section 3.3.2 of this submittal, NUREG-1493 concluded that reducing the frequency of Type A tests to a 20-year test interval was found to lead to an imperceptible increase in risk.

Several programmatic factors can also be cited that help ensure the continued safety function of the DBNPS containment pressure boundary. NEI 94-01, Revision 3-A, requires sites adopting the 15-year extended Type A test interval to perform visual examinations of the accessible interior and exterior surfaces of the containment structure for structural degradation that may affect the containment leak-tight integrity at the frequency prescribed by the guidance. Additionally, several measures are put in place to ensure integrity of the Type B and Type C tested components. NEI 94-01, Revision 3-A, limits large containment penetrations such as air locks, purge, and vent valves; to a maximum 30-month testing interval. For those valves that meet the performance standards defined in NEI 94-01, Revision 3.;.A and are selected for test intervals greater than 60 months, a leakage understatement penalty is added to the minimum pathway leakage rate prior to the frequency being extended beyond 60-months. Finally, identification of adverse trends in the overall Type B and Type C leakage rate summations and available margin between the Type B and Type C leakage fate

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 35 of 76 summation and its regulatory limit are required by NEI 94-01, Revision 3-A, to be shown in the DBNPS post-outage report(s).

Therefore, the proposed change does not significantly impact the containment function or systems, structures, and components supporting that function, the layers of defense, such that a reasonable balance among the layers of defense is preserved.

PRA Response:

The usage of the risk metrics of large early release frequency, population dose, and conditional containment failure probability collectively ensures the balance between prevention of core damage, prevention of containment failure, and consequence mitigation is preserved. The change in large early release frequency is small per Regulatory Guide 1.17 4, and the change in population dose and conditional containment failure probability are small as defined in Attachment B of this submittal and consistent with NEI 94-01, Revision 3-A.

2. Preserve adequate capability of design features without an overreliance on programmatic activities as compensatory measures.

Nuclear power plant licensees implement a number of programmatic activities, including programs for quality assurance, testing and inspection, maintenance, control of transient combustible material, foreign material exclusion, containment cleanliness, and training. In some cases, activities that are part of these programs are used as compensatory measures; that is, they are measures taken to compensate for some reduced functionality, availability, reliability, redundancy, or other feature of the plant's design to ensure sc;1fety functions (for example, reactor vessel inspections that provide assurance that reactor vessel failure is unlikely). NUREG-2122, "Glossary of Risk-Related Terms in Support of Risk-Informed Decision Making," (Reference 24), defines "safety function" as those functions needed to shut down the reactor, remove the residual heat, and contain any radioactive material release.

A proposed licensing basis change might involve or require compensatory measures.

Examples include hardware (for example, skid-mounted temporary power supplies);

human actions (for example, manual system actuation); or some combination of these measures. Such compensatory measures are often associated with temporary plant configurations. The preferred approach for accomplishing safety functions is through engineered systems. Therefore, when the proposed licensing basis change necessitates reliance on programmatic activities as.compensatory measures, the licensee should justify that this reliance is not excessive (that is, not overly reliant). The intent of this consideration is not to preclude the use of such programs as compensatory measures.

but to ensure that the use of such measures does not significantly reduce the capability of the design features (for example, hardware).

Defense-In-Depth Consideration 2 Response:

The purpose of the proposed change is to extend the interval of the Type A test from 10 years to 15 years. This proposed licensing basis change to the containment leakage rate testing program requirements specified in plant technical specifications does not

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 36 of 76 substitute programmatic activities for design features. The proposed change does not alter the configuration of the DBNPS containment pressure boundary.

Several programmatic factors were described in the response to Item 1 above, which are required when adopting NEI 94-01, Revision 3-A. These factors are conservative in nature and are designed to generate corrective actions if the required testing or inspections are deemed unsatisfactory to ensure the continued safety function of the containment is maintained. The programmatic factors are designed to provide differing ways to test and examine th_e containment pressure boundary in a manner that verifies the DBNPS containment pressure boundary will perform its intended safety function.

PRA Response:

The adequacy of the design feature (the containment boundary subject to Type A testing) is preserved as evidenced by the overall small change in risk associated with the Type A test frequency change.

3. Preserve system redundancy, independence, and diversity commensurate with the expected frequency and consequences of challenges to the system, including consideration of uncertainty.

The defense-in-depth philosophy has traditionally been applied in plant design and operation to provide multiple means to accomplish safety functions. System redundancy, independence, and diversity result in high availability and reliability of the function and also help ensure that system functions are not reliant on any single feature of the design.

Redundancy provides for duplicate equipment that enables the failure or unavailability of at least one set of equipment to be tolerated without loss of function. Independence of equipment implies that the redundant equipment is separate such that it does not rely on the same supports to function. Jhis independence can sometimes be achieved by the use of physical separation or physical protection. Diversity is accomplished by having equipment that performs the same function rely on different attributes such as different principles of operation, different physical variables, different conditions of operation, or production by different manufacturers which helps reduce common-cause failure.

A proposed change might reduce the redundancy, independence, or diversity of systems.

The intent of this consideration is to ensure that the ability to provide the system function is commensurate with the risk of scenarios that could be mitigated by that function. The consideration of uncertainty, including the uncertainty inherent in the PRA, implies that the use of redundancy, independence, or diversity provides high reliability and availability and also results in the ability to tolerate failures or unanticipated events.

Defense-In-Depth Consideration 3 Response:

The proposed change to extend the frequency of the Type A test from 10 years to 15 years does not reduce the redundancy, independence or diversity of safety related systems. As described in Section 3.3.2 of this submittal, NUREG-1493 concluded that reducing the frequency of Type A tests to a 20-year test interval was found to lead to an imperceptible increase in risk.

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 37 of 76 In addition, NEI 94-01, Revision 3-A, Section 11.3.2 requires a schedule of tests be developed, for components on a test interval greater than 60 months, so that unanticipated random failures and unexpected common-mode failures are avoided. This is typically accomplished by implementing test intervals at approximately evenly distributed intervals. Therefore, the proposed change ensures high reliability and availability of the containment structure to perform its safety function in the event of unanticipated events.

PRA Response:

The redundancy, independence, and diversity of the containment subject to the Type A test is preserved, commensurate with the expected frequency and consequences of challenges to the system, as evidenced by the overall small change in risk associated with the Type A test frequency change.

4. Preserve adequate defense against potential common-cause failures.

An important aspect of ensuring defense-in-depth is to guard against common-cause failures. Multiple components may fail to function because of a single specific cause or event that could simultaneously affect several components important to risk. The cause or event may include an installation or construction deficiency, accidental human action, extreme external environment, or an unintended cascading effect from any other operation or failure within the plant. Common-cause failures can also result from poor design, manufacturing, or maintenance practices.

Defenses can prevent the occurrence of failures from the causes and events that could allow simultaneous multiple component failures. Another aspect of guarding against common-cause failure is to ensure that an existing defense put in place to minimize the impact of common-cause failure is not significantly reduced; however, a reduction in one defense can be compensated for by adding another.

Defense-In-Depth Consideration 4 Response:

As part of the proposed change, the performance-based testing standards outlined in NEI 94-01, Revision 3-A, along with ANSI/ANS 56.8-2002 for Type A and Type B testing will be adopted. NEI 94-01, Revision 3-A, Section 11.3.2, requires a schedule of tests be developed, for components on test intervals greater than 60 months, such that unanticipated random failures and unexpected common-mode failures are avoided. This is typically accomplished by implementing test intervals at approximately evenly distributed intervals. For those components that have demonstrated satisfactory performance and have had their testing limits extended, administrative testing limits are assigned on a component-by-component basis and are used to identify potential valve or penetration degradation. Administrative limits are established at a value low enough to identify and should allow early correction in advance of total valve failure. Should a component exceed its administrative limit during testing, NEI 94-01, Revisions 3-A requires a cause determination. The cause determination is designed to identify and address common-cause (including common-mode) failure mechanisms through appropriate corrective actions. The proposed change also imposes a requirement to address margin management (that is, margin between the current containment leakage

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 38 of 76 rate and its pre-established limit). As a result, adoption of the performance-based testing standards proposed by this change ensures adequate barriers exist to preclude failure of the containment pressure boundary due to common-mode failures and therefore continues to guard against common-cause failure.

PRA Response:

Adequate defense against common-cause failures is preserved. The Type A test detects problems in the containment, which may or may not be the result of a common-cause failure. Such a common-cause failure may affect failure of another portion of containment (that is, local penetrations) due to the same phenomena. Adequate defense against common-cause failures is preserved via the continued performance of the Type B and Type C tests and the performance of inspections. The change to the Type A test, which bounds the risk associated with containment failure modes including those involving common-cause failures, does not degrade adequate defense as evidenced by the overall small change in risk associated with the Type A test frequency change.

5. Maintain multiple fission product barriers.

Fission product barriers include the physical barriers themselves (for example, the fuel cladding, reactor coolant system pressure boundary, and containment) and any equipment relied on to protect the barriers (for example, containment spray). In general, these barriers are designed to perform independently so that a complete failure of one barrier does not disable the next subsequent barrier. For example, one barrier, the containment, is designed to withstand a double-ended guillotine break of the largest pipe in the reactor coolant system, another barrier.

A plant's licensing basis might contain events that, by their very nature, challenge multiple barriers simultaneously. Examples include interfacing-system loss-of-coolant accidents, steam generator tube rupture, or crediting containment accident pressure.

Therefore, complete independence of barriers, while a goal, might not be achievable for all possible scenarios.

Defense-In-Depth Consideration 5 Response:

The purpose of the proposed change is to extend the testing frequency of the Type A test from 10 years to 15 years. As part of the proposed change, the performance-based testing standards outlined in NEI 94-01, Revision 3-A, along with ANSI/ANS 56.8-2002 will be adopted. The only fission product barrier affected by the proposed change to the containment leak-test program is the containment.

The overall containment leakage rate calculations associated with the testing standards contain inherent conservatisms through the use of margin. Plant Technical Specifications require the overall containment leakage rate to be less than 1.0 La. NEI 94-01 requires the as-found Type A test leakage rate must be less than the acceptance criterion of 1.0 La given in the plant Technical Specifications. Prior to entering a mode where containment integrity is required, the as-left Type A leakage rate shall not exceed 0.75 La. The as-found and as-left values are as determined by the appropriate testing methodology specifically described in ANSI/ANS 56.8-2002. Additionally, the combined

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 39 of 76 leakage rate for all Type B and Type C tested penetrations shall be less than or equal to 0.6 La, determined on a maximum pathway basis from the as-left LLRT results prior to entering a mode where containment integrity is required. This regulatory approach results in a 25 percent and 40 percent margin, respectively, to the 1.0 La requirements.

For those local leakage rate tested components that have demonstrated satisfactory performance and have had their testing limits extended, administrative testing limits are assigned on a component-by-component basis and are used to identify potential valve or penetration degradation. Administrative limits are established at a value low enough to identify and allow early correction in advance of total valve failure. Should a component exceed its administrative limit during testing, NEI 94-01, Revision 3-A, requires a cause determination. The cause determination is designed to identify and address failure mechanisms through appropriate corrective actions. Therefore, the proposed change adopts requirements with inherent conservatisms to ensure the margin to safety limit is maintained, thereby, preserving the containment fission product barrier.

PRA Response:

Multiple Fission Product barriers are maintained. The portion of the containment affected by the Type A test extension is still maintained as an independent fission product barrier, albeit with a small change in the reliability of the barrier.

6. Preserve sufficient defense against human errors.

Human errors include the failure of operators to correctly and promptly perform the actions necessary to operate the plant or respond to off-normal conditions and accidents, errors committed during test and maintenance, and incorrect actions by other plant staff.

Human errors can result in the degradation or failure of a system to perform its function, thereby significantly reducing the effectiveness of one of the layers of defense or one of the fission product barriers. The plant design and operation include defenses to prevent the occurrence of such errors and events. These defenses generally involve the use of procedures, training, and human engineering; however, other considerations (for example, communication protocols) might also be important.

Defense-In-Depth Consideration 6 Response:

The PRA response below addresses defense-in-depth consideration 6.

PRA Response:

Sufficient defense against human errors is preserved. The probability of a human error to operate the plant, or to respond to off-normal conditions and accidents is not significantly affected by the change to the Type A testing frequency. Errors committed during test and maintenance may be reduced by the less frequent performance of the Type A test (less opportunity for errors to occur).

7. Continue to meet the intent of the plant's design criteria.

For plants licensed under 10 CFR Part 50 or 10 CFR Part 52, the plant's design criteria are set forth in the current licensing basis of the plant. The plant's design cdteria define minimum requirements that achieve aspects of the defense-in-depth philosophy; as a

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 40 of 76 consequence, even a compromise of the intent of those design criteria can directly result in a significant reduction in the effectiveness of one or more of the layers of defense.

When evaluating the effect of the proposed licensing basis change, the licensee should demonstrate that it continues to meet the intent of the plant's design criteria.

Defense-In-Depth Consideration 7 Response:

The purpose of the proposed change is to extend the testing interval of the Type A test from 10 years to 15 years. The proposed extension does not involve a physical change to the plant or a change in the manner in which the plant is operated or controlled. As part of the proposed change, the performance-based testing standards outlined in NEI 94-01, Revision 3-A, along with ANSI/ANS 56.8-2002 will be adopted. The leakage limits imposed by plant technical specifications remain unchanged when adopting the performance-based testing standards outlined in NEI 94-01, Revision 3-A, and ANSI/ANS 56.8-2002. Plant design limits imposed by the Updated Safety Analysis Report also remain unchanged as a result of the proposed change. Therefore, the proposed change continues to meet the intent of the plant's design criteria to ensure the integrity of the DBNPS containment pressure boundary.

PRA Response:

The intent of the plant's design criteria continues to be met. The extension of the Type A test does not change the configuration of the plant or the way the plant is operated. As the intent of the plant's design criteria will continue to be met, this consideration for evaluating the impact of the proposed change on defense in depth will not affect the risk associated with the proposed change.

==

Conclusion:==

I I

I l

The responses to the seven defense-in-depth questions above conclude that the existing defense-in-depth has not been diminished. Therefore, the proposed change does not comprise a reduction in safety.

3.5 Non-Risk Based Assessment Consistent with the defense-in-depth philosophy discussed in Regulatory Guide 1.174, non-risk-based considerations relevant to the proposed amendment have been assessed. Multiple inspection and testing programs are provided that: 1) ensure the containment structure continues to remain capable of meeting its design functions, and

2) are designed to identify any degrading conditions that might affect that capability.

These programs are discussed below.

3.5.1 Nuclear Safety-Related Protective Coatings Program Protective coating material has been applied to carbon steel and concrete surfaces within the containment. The function of the protective coating material is to resist exposure to conditions including ionizing radiation, high temperature, and impingement from sprays, that result from normal operating and design basis accident conditions.

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 41 of 76 The Updated Safety Analysis Report includes a commitment to the regulatory position of Regulatory Guide 1.54 (Revision 0, June 1973, Reference 25) with the following clarifications.

1. This Regulatory Guide and its associated ANSI Standard (ANSI N101.4, Reference 26) implies that a significant amount of coating work is required at the plant site. Although this is correct for construction sites, the coating work at an operating site generally consists of repair and touchup work following maintenance and repair activities or the initial coating of components such as hangers, supports, and piping during facility modifications. Therefore, in lieu of the full requirements of the Regulatory Guide and ANSI N101.4, the following requirements shall be imposed:

a. The quality assurance requirements of Section 3 of ANSI N 101.4 applicable to the coating manufacturer shall be imposed on the coating manufacturer through the procurement process.

b. Coating application procedures shall be developed based on the manufacturer's recommendations for application of the selected coating systems.

c. Coating applicators shall be qualified to demonstrate their ability to satisfactorily apply the coatings in accordance with the manufacturer's recommendations.

d. Quality control personnel shall perform inspections to verify conformance of the coating application procedure. Section 6 of ANSI N101.4 shall be used as guidelines in the establishment of the inspection program.

e. Quality control personnel shall be qualified to the requirement of Regulatory Guide

. 1.58 (Revision 1 ).

f.

Documentation demonstrating conformance to the above requirements shall be maintained.

2. The requirements of Item 1 above apply to surfaces within containment with the following exceptions:

a. Surfaces to be insulated.

b. Surfaces contained within a cabinet or enclosure.

c. Repair and touchup areas less than 30 square inches or surface areas such as:

cut ends; bolt heads; nuts and miscellaneous fasteners; and damage resulting from spot, tack or arc welding.

d. Small items such as small motors, handwheels, electrical cabinets, control panels, loud speakers, motor operators, and other similar things where special painting requirements would be impracticable.

e. Stainless steel or galvanized surfaces.

f.

Banding used for insulated pipe.

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 42 of 76 The Updated Safety Analysis Report also includes a commitment to the requirements of ANSI N101.4-1972 for activities comparable in nature and extent to construction phase activities as modified by the commitment to Regulatory Guide 1.54.

Two design basis accident qualified coating systems are currently used on ferrous metal and concrete surfaces within the containment The first system consists of an inorganic zinc prime coat with an epoxy top coat. This system is applied to the containment vessel, structural steel, and equipment.

The second system consists of an epoxy primer and epoxy topcoat. This system is applied to ferrous metal surfaces and concrete surfaces.

Non-design-basis-accident qualified coating materials have also been applied to structures and components within the containment. These materials are standard manufacturer's paints or unqualified coating systems and epoxy materials applied to structures or components with inadequate surface preparation. These materials have been quantified and are tracked by a non-design-basis-accident qualified protective coating inventory. Coating material exclusions to this inventory include surfaces that are insulated or are contained within a cabinet or enclosure. The documented quantity of non-design-basis-accident qualified coating material must remain below the limit of coating material debris identified by the emergency core cooling system emergency sump debris analysis.

Coating condition assessment inspections are performed each refueling outage to identify and correct degraded coating materials.

The Nuclear Safety-Related Protective Coatings Program requires that an inventory of unqualified coatings within containment be maintained, and that degraded coating conditions be documented using the corrective action process.

The current amount of degraded and unqualified coatings reported is identified in Table 3.5.1-1 with their associated limits.

Table 3.5.1-1, DBNPS Unqualified and Degraded Containment Coatings Type of Coating Amount of Degraded Degraded Coatings Coatings (ft2)

Limits (ft2)

Alkyd 4578 5418 One-Coat Epoxy 6377 10,679 Two-Coat Epoxy 2141 2297

FE NOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 43 of 76 3.5.2 Containment lnservice Inspection Program The lnservice Inspection Program is an on-going condition monitoring program, which monitors the integrity of the containment vessel, containment hatches and air locks, seals, gaskets, moisture barriers, and containment penetration bolting. The program is in accordance with the requirements of the 2007 Edition, 2008 Addenda of ASME Code,Section XI, as required by 10 CFR 50.55a. The IWE program is implemented by the lnservice Inspection Program. Subsection IWL is not applicable to DBNPS.

The lnservice Inspection Plan establishes the 4th Interval lnservice Inspection requirements for the DBNPS. This plan describes the applicability of the 2007 Edition, 2008 Addenda of ASME Code,Section XI, inspection requirements to DBNPS structures, systems, and components. The plan also contains the intended schedule for the inspection of components subject to inservice inspection requirements.

Inspection Intervals The first 10-Year inspection interval for DBNPS began on July 31, 1978, the first day of commercial operation. Extended outages of greater than 6 months that occurred during the first inspection interval caused the interval to be extended an additional 34 months as permitted by Subarticle IWA-2400 of ASME Code,Section XI.

The start of the second 10-Year inspection interval was September 21, 1990.

The start of the third 10-Year inspection interval was September 21, 2000. Due to the extended thirteenth refueling outage, the 10-year interval was extended by two years in accordance with Subparagraph IWA-2430(e).

. The start of the fourth 10-Year inspection interval was September 21, 2012*. The.fourth 10-Year inspection interval is divided into three inspection periods. The inspection periods are scheduled for time periods and refueling outages (RFOs) as follows:

Table 3.5.2-1, DBNPS Fourth 10-Year Inspection Interval Inspection Period Dates Number of Outages 1st Period 09/21/2012 - 09/20/2015 1 (18RFO) 2nd Period 09/21/2015 - 09/20/2019 2 (19RFO, 20RFO) 3rd Period 09/21/2019 - 09/20/2022 2 (21 RFO, 22RFO)

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 44 of 76 The planned inspection periods and outages for the fifth 10-Year inspection interval are as follows:

Table 3.5.2-2, DBNPS Fifth 10-Year Inspection Interval ln~pection Period Dates Number of Outages 1st Period 09/21/2022 - 09/20/2025 1 (23RFO) 2nd Period 09/21/2025 - 09/20/2029 2 (24RFO, 25RFO) 3rd Period 09/21/2029 - 09/20/2032 2 (26RFO, 27RFO)

Inaccessible Areas Paragraph 10 CFR 50.55a(b)(2)(ix), "Section XI condition: Metal containment examinations," requires, in part, that licensees applying Subsection IWE, 2007 Edition through the latest addenda incorporated by reference in paragraph (a)(1 )(ii) of this section, must satisfy the requirements of paragraphs (b)(2)(ix)(A)(2), (b)(2)(ix)(B) and (J) of this section set out below.

(A) For Class MC applications, the following apply to inaccessible areas.

(2) For each inaccessible area identified for evaluation, the licensee must provide the following in the ISi [lnservice Inspection]

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; 1

I ii. An evaluation of each area, and the result of the evaluation, and; iii. A description of necessary corrective actions.

(B) 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.

(J) In general, a repair/replacement activity such as replacing a large containment penetration, cutting a large construction opening in the containment pressure boundary to replace steam generators, reactor vessel heads, pressurizers, or other major equipment; or other similar modification is considered a major containment modification. When applying IWE-5000 to Class MC pressure-retaining components, any major containment modification or repair/replacement, must be followed by a Type A test to provide assurance of both containment

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 45 of 76 structural integrity and leak-tight integrity prior to returning to service, in accordance with 10 CFR part 50, Appendix J, Option A or Option B on which the licensee's Containment Leak-Rate Testing Program is based. When applying IWE-5000, if a Type A, B, or C Test is performed, the test pressure and acceptance standard for the test must be in accordance with 10 CFR Part 50, Appendix J.

ASME Code Class IWE Containment Examinations The following provides a summary of the application of the ASME Code,Section XI, 2007 Edition through the 2008 Addenda to the containment vessel. The application and distribution of examinations for the fourth interval is based upon Paragraph IWE-2411.

The schedule for the examination of the containment vessel is contained within Tables 3.5.2-3, 3.5.2-4 and 3.5.2-5.

Category E-A: Containment Surfaces General visual examination of the accessible surfaces and moisture barriers on the containment vessel will be performed each inspection period. The examination shall consider areas identified in Notes 1 and 2 of Examination Category E-A as applicable.

FENOC has not identified any wetted surfaces or submerged areas (E1.12) at DBNPS.

There are two containment moisture barriers, one on the inside at the lower concrete to vessel ledge and the other in the annulus sand pocket region.

Category E-C: Containment Surfaces Requiring Augmented Examination A condition report identified that a gap had formed at two areas between the containment vessel and the concrete ledge on the inside of containment at the 565-foot elevation. It

' ' was determined that these areas should be considered surface areas requiring augmented examination as required by IWE-1240.

Another condition report documented the occurrence of general rust and corrosion noted on the sleeves of the service water and component cooling water piping penetrations into containment. From 2002 through 2014, these components were examined under an owner elected category. These penetrations have been placed in the E-C examination category as they meet the criteria noted by IWE-1241(a). These locations shall have a VT-1 visual examination performed once per Period.

Category E-G: Pressure Retaining Bolting Pressure retaining bolting will be examined visually, VT-1, during the inspection interval.

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 46 of 76

• Table 3.5.2-3, E-A, Containment Surfaces Item Description Exam Extent of Examination Total Required Number Method Examination Percentage Population Exams Accessible General 100% Each E01.011 Surface Visual Inspection 100.0%

2 6

Areas1 Period Moisture General 100% Each E01.030 Barriers2 Visual Inspection 100.0%

2 6

Period Total 4

12 Cumulative Percentage Period 1st 2nd 3rd 2

2 2

2 2

2 4

4 4

33.3%

66.6%

100.0%

Note 1: Examination shall include all accessible interior and exterior surfaces of Class MC components, parts, and appurtenances.

Note 2: 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.

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 47 of 76 Table 3.5.2-4, E-C, Containment Surfaces Requiring Augmented Examination Item Description Exam Extent of Examination Total Required Period Number Method Examination Percentage Population Exams 1st 2nd 3rd 100% of E04.011 Visible Surfaces VT-1 3 Surface Areas 100.0%1 8

24 8

8 8

Identified by IWE-1242 100% of Surface Area Minimum Wall Grid Minimum UT3 Thickness E04.012 Wall Thickness Thickness Locations 100.0%

1 1

1 02 02 Location During Each Inspection Period Total1 9

25 9

8 8

Cumulative Percentage 36.0%

68.0%

100.0%

Note 1: Containment surface areas requiring augmented examination are those identified in IWE-1240. The extent of examination shall be 100 percent(%) for each inspection period until the areas examined remain essentially unchanged for the next inspection period. Such areas no longer require examination in accordance with IWE-1240(c).

Note 2: A condition report has designated the containment vessel interior gap as requiring augmented examinations. Examine in 15RFO and 17RFO of the third lnservice Inspection Interval and the first period in' the fourth lnservice Inspection interval. 13RFO examination serves as a baseline examination. If there is no change then successive examinations are not required in the second and third periods.

Note 3: UT is an ultrasonic examination conducted to detect internal flaws or to characterize materials, and VT-1 is a visual examination conducted to detect discontinuities and imperfections on the surface of components, including such conditions as cracks, wear, corrosion, or-erosion.

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 48 of 76 Table 3.5.2-5, E-G, Pressure Retaining Bolting Item Exam Extent of Examination Total Required Number Description Method Examination Percentage Population Exams Bolted

Visual, 100% Each EOB.010 Inspection 100.0%

57 57 Connections VT-1 Period1*2 Total 57 57 Cumulative Percentage Period 1st 2nd 3rd 21 11 25 21 11 25 36.8%

56.1%

100.0%

Note 1: Examination shall include bolts, studs, nuts, bushings, washers, and threads in base material and flange ligaments between fastener holes.

Note 2: Examination may be performed with the connection assembled and bolting in place under tension, provided the connection is not disassembled during the interval. If the bolted connection is disassembled for any reason during the interval, the examination shall be performed with the connection disassembled.

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 49 of 76 Repairs or Replacements Repair, replacement, and modifications of vessels, tanks, pumps, valves and piping, including their supports, that are located within the Class 1, 2, 3, and IWE boundaries are performed in accordance with the rules of Section XI, 2007 Edition through the 2008 Addenda. The rules of Section XI for repairs, replacements, and modifications also apply to pressure retaining materials, structural attachments to pressure retaining materials, and to supports as shown on the support detail drawings. _Pressure retaining materials include vessel shells, heads and nozzles; pipes and fittings; valve bodies bonnets and disks; pump casings and covers; and bolting which joins pressure retaining items.

Repair, replacements, and modifications within the scope of Section XI are performed under DBNPS's "NR" and "VR" Certificate of Authorization issued by the National Board of Boiler and Pressure Vessel Inspectors, Certificate No. 20 and 316, respectively.

The pressure test requirements following a repair or replacement activity are found in IWA-4550 for Class MC containments. 10 CFR 50.55a{b)(2)(ix), Paragraph J, "Metal containment examinations: Tenth provision," provides in part that:

In general, a repair or replacement activity such as replacing a large containment penetration, cutting a large construction opening in the containment pressure boundary to replace steam generators, reactor vessel heads, pressurizers, or other major equipment; or other similar modification is considered a major containment modification. When applying IWE-5000 to Class MC pressure-retaining components, any major containment modification or repair/replacement, must be followed by a Type A test to provide assurance of both containment structural integrity and leaktight integrity prior to returning to service, in accordance with 10 CFR Part 50, Appendix J, Option A or Option B on which the licensee's Containment Leak-Rate Testing Program is based. When applying IWE-5000, if a Type A, B, or C Test is performed, the test pressure and acceptance standard for the test must be in accordance with 10 CFR Part 50, Appendix J.

Visual Examinations The General Visual Examination required by Table IWE-2500-1, Examination Category E-A, will be conducted in accordance with IWE-2311.

Leak Testing The bubble test required by IWE-5223.4(a) shall be performed in accordance with procedures meeting ASME Code Section V, Article 10, Appendix I or any other Article 10 leak test that can be performed in conjunction with the containment pneumatic test.

The alternate bubble test-vacuum box technique shall be performed in accordance with procedures meeting ASME Code,Section V, Article 10, Appendix II and the additional requirements of IWE-5224.

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 50 of 76 Qualification of Nondestructive Examination Personnel Personnel performing volumetric or surf;3ce examinations will be qualified in accordance with ANSI/ASNT CP-189-1995 as amended by the requirements of IWA-2300 and 10 CFR 50.55a. Personnel performing examinations required to meet ASME Code,Section XI, Appendix VIII, shall meet the additional qualification requirements of Appendix VIII and VII.

Personnel performing visual (VT-1, VT-2, and VT-3) will be qualified and certified in accordance with ASME Code,Section XI, IWA-2300 and Appendix VI.

The General Visual Examination required by Table IWE-2500-1, Examination Category E-A, will be performed by personnel with a visual acuity meeting the requirements of IWA-2321 (a) and who have been instructed, trained and approved by the responsible individual in accordance with IWE-2320.

Personnel performing the leak testing or bubble vacuum-box testing shall be qualified in accordance with ASME Code,Section V, Article 1 or ASME Code,Section XI, IWA-2300.

Relief Requests and Requests for Alternatives Relief Requests and Requests for Alternatives are required when there are situations where the Code requirements cannot be met or where an alternative is desired. Relief requests are only applicable for the 10-year inspection interval during which relief was requested and approval does not apply to subsequent inspection intervals except when the NRC specifically states they are for longer periods of time.

Table 3.5.2-6, Relief Requests Applicable to the 4th IWE Inspection Interval Request No.

General Description Approval Date Alternative Post-Repair RR-E1 Pressure Testing 05/08/2013 Requirements Relief request RR-E 1 was previously described in Section 3.2, "Containment Vessel Tests and Inspections Performed Following Major Modifications," of this evaluation.

3.5.3 Supplemental Inspection Requirements Inspections of the interior and exterior steel containment vessel surfaces are performed.

The visual examination of the steel containment vessel surfaces intended to fulfill the requirements of 10 CFR 50, Appendix J, Option B testing, will be performed in accordance with the requirements of and frequency specified in Table 3.5.2-3, E-A, Containment Surfaces. This will require the performance of a minimum of four (4) visual inspections in accordance with ASME Code,Section XI, Subsection IWE surpassing the

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 51 of 76 requirements of NEI 94-01, Revision 3-A Section 9.2.3.2, Supplemental Inspection Requirements, of three (3) inspections over a 15-year interval.

An additional visual inspection of the steel containment vessel is conducted prior to the conduct of each Type A test.

3.5.4 Results of Recent Containment Examinations 20RFO Visual Examination of IWE Surfaces Summary No.

E01.011.969400 Report No.

20-VT-090 Description Containment Vessel Interior Surfaces General Visual Paint scrapes in various areas around the entire reactor building with no corrosion noted on the 565-foot and 585-foot elevations. These indications extend from walking level up to eight feet above floor level on both elevations.

At the 558-foot elevation and 280-degree azimuth a 6-inch by 6-inch area of chipped paint was found.

Summary No.

E01.011.969400 Report No.

20-VT-115 Description Containment Vessel Interior Surfaces General Visual At 653-foot to 658-foot elevations and 89-degree to 90-degree azimuth, the top coat of paint was found to be flaking between the lower ladder and the containment,

recirculation fan. No degradation of the liner surface was detected.

At 654-foot to 722-foot elevations and 83-degree to 85-degree azimuth the top coat of paint was found to be peeling and flaking around and behind the crane access ladder.

No degradation of the inner liner surface was detected.

Numerous nicks, chips, and abrasions on painted liner surfaces was seen on the grating at 603-foot to 722-foot elevations and 0-degree to 360-degree azimuth. These appear to be scrapes, impacts, and material handling bumps. No significant corrosion was detected.

Visual examination of the 656-foot to 657-foot elevations and 250-degree azimuth noted areas of chipped paint, and bare metal at the 656-foot elevation ( 1-inch by 1-1 /4 inches in size), and at the 657-foot elevation (1-inch by 3-inches in size). Tightly adhered paint surrounded both areas with no significant corrosion noted.

At the 610-foot to 615-foot elevations and 125-degree azimuth, a large area of paint was found to be chipping in very small individual spots. No bare metal was exposed.

The conditions noted in the comments listed above were also seen in previous outages.

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 52 of 76 Summary No.

E01.011.010634 Report No.

20-VT-104 Description Containment Vessel Exterior Surfaces General Visual General visual of exterior surfaces only from the 603-foot elevation to the top of the containment dome. A 3-foot by 5-foot area of paint peeling was noted at the 785-foot elevation and 275-degree azimuth.

3.5.5 Containment Leakage Rate Testing Program -

Type Band Type C Testing Program DBNPS Type B and Type C testing program requires testing of electrical penetrations, air locks, hatches, flanges, and containment isolation valves in accordance with 10 CFR 50, Appendix J, Option Band Regulatory Guide 1.163 (Reference 1) for Type B testing and NEI 94-01, Revision 3-A for Type C testing. 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 Technical Specification 5.5.15, the allowable maximum pathway total Types B and C leakag~ is less than 0.6 La (599,400 standard cubic centimeters per minute [seem])

where La equals 999,000 seem.

As discussed in NUREG-1493 (Reference 7), Type Band Type C tests can identify the vast majority (greater than 95 percent) of 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 as-found (AF) and as-left (AL) test values for DBNPS can be summarized as:

As-found minimum pathway leak rate shows an average of 5.23 percent of 0.6 La (599,400 seem) with a high of 10.01 percent of 0.6 La.

As-left maximum pathway leak rate shows an average of 10.66 percent of 0.6 La (599,400 seem) with a high of 22.69 percent of 0.6 La.

Table 3.5.5-1 provides LLRT data trend summaries for DBNPS refueling outages (RFOs) since 2008 (the last Type A test was performed in 2011 ).

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 53 of 76 Table 3.5.5-1 DBNPS Types B and C LLRT Combined As-Found and As-Left Trend Summary Outage 15RFO 16RFO 17MID 17RFO 18RFO 19RFO 20RFO

&Year 2008 2010 2011 2012 2014 2016 2018 AF Min Path 30,099 14,945 17,495 29,469 27,603,,

39,801 60,012 (seem)

Fraction 5.03 2.50 2.92 4.92 4.61 6.64 10.01 of 0.6 La AL Max Path 45,477 30,000 50,056 29,279 60,357 95,666 136,016 (seem)

Fraction 7.60 5.01 8.35 4.90 10.10 15.96 22.69 of 0.6 La The as-found minimum pathway summations represent the high quality of maintenance of Type 8 and Type C tested components while the as-left maximum pathway summations represent the effective management of the Containment Leakage Rate Testing Program-by the program owner.

3.5.6 Type B and Type C Local Leak Rate Testing Program Implementation Review Table 3.5.6-1 (below) identifies DBNPS components, which were on Appendix J, Option B performance-based extended test intervals, but have not demonstrated acceptable performance during the previous two outages. The component test intervals for the components shown have been reduced to 30 months.

Table 3.5.6-1 DBNPS Types B and C LLRT Program Implementation Review 19RF0-2016 Admin Component As-Limit As-left Cause of Corrective Scheduled (Penetration found Alert/

Failure Action Interval No.)

Action seem seem seem CV5079 30,8671 12,200 9,230 Packing Packing 30 months (8J)

Leak Adjustment2

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 54 of 76 Table 3.5.6-1 DBNPS Types B and C LLRT Program Implementation Review 20RF0-2018 Admin Component As-Limit As-left Cause of Corrective Scheduled (Penetration found Alert/

seem Failure Action Interval No.)

seem Action seem Valve was Replaced CC1407C stuck (4) 6,482 300 4

partially with a new 30 months component open.

Valve was Replaced RC229C stuck (48) 6,175 1,750 43 partially with a new 30 months component open.

Note 1: Valve CV5079 exceeded its maximum allowable leak rate this outage (19RFO) largely attributed to a packing leak.

Note 2: During 20RFO, the LLRT of valve CV5079 measured leakage of 59,217 seem that exceeded the predetermined action limit (48,800 seem) for replacement of CV5079. Valve CV5079 was upgraded and replaced with a new valve.

The total population of penetration components that require Type B testing is 33. Of the 33 components, 18 mechanical penetration components are eligible for an extended test frequency and are tested on an extended frequency as shown in Table 3.5.6-2 below.

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 55 of 76 Table 3.5.6-2, Type B Penetration Test Frequency Item Penet.

Penetration Component Outage Use Eligible for Currently On No.

No.

Extended Extended Frequency Frequency 1

BB CV5071 Inboard Flange Not Used Yes Yes 2

80 CV5073 Inboard Flange Not Used Yes Yes 3

BF CV5075 Inboard Flange Not Used Yes Yes 4

81 CV5078 Inboard Flange Not Used Yes Yes 5

17 Flange As Required Yes Yes 6

23 FTT 1-2 Bellows /Guard Pipe Not Used Yes Yes 7

23 FTT 1-2 Flange Each Outage No No 8

24 FTT 1-1 Bellows /Guard Pipe Not Used Yes Yes 9

24 FTT 1-1 Flange Each Outage No No 10 30 DH9A Emergency Sump Not Used Yes Yes Guard Pipe 11 31 DH9B Emergency Sump Not Used Yes Yes Guard Pipe 12 37 Feedwater Inner Expansion Not Used Yes Yes Bellows 13 37 Feedwater Outer Expansion Not Used Yes Yes Bellows 14 38 Feedwater Inner Expansion Not Used Yes Yes Bellows 15 38 Feedwater Outer Expansion Not Used Yes Yes Bellows 16 39 Main Steam Inner Expansion Not Used Yes Yes Bellows 17 39 Main Steam Outer Expansion Not Used Yes Yes Bellows 18 40 Main Steam Inner Expansion Not Used Yes Yes Bellows Current Test Future Frequency Frequency Change 10 Years No 10 Years No 10 Years No 10 Years No 10 Years No 10 Years No R1 No 10 Years No R1 No 10 Years No 10 Years No 10 Years No 10 Years No 10 Years No 10 Years No 10 Years No 10 Years No 10 Years No

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 56 of 76 Table 3.5.6-2, Type B Penetration Test Frequency Item Penet.

Penetration Componen't Outage Use Eligible for Currently On No.

No.

Extended Extended Frequency Frequency 19 40 Main Steam Outer Expansion Not Used Yes Yes Bellows 20 59 Blind Flanges As Required Yes Yes 21 80 Emergency Air Lock (EAL)

Each Outage No No 22 80 EAL Inner Top Shaft Seal Not Used No No 23 80 EAL Inner Bottom Shaft Seal Not Used No No 24 80 EAL Outer Top Shaft Seal Not Used No No 25 80 EAL Outer Bottom Shaft Seal Not Used No No 26 81 Personnel Air Lock (PAL)

Each Outage No No 27 81 PAL Inner Top Shaft Seal Not Used No No 28 81 PAL Inner Bottom Shaft Seal Not Used No No 29 81 PAL Outer Top Shaft Seal Not Used No No 30 81 PAL Outer Bottom Shaft Seal Not Used No No 31 82 Equipment Hatch Each Outage No No 32 101 Electrical (East)

Not Used No No 33 102 Electrical (West)

Not Used No No Notes:

1. FTT is fuel transfer tube, and R1 is next refueling outage.

Current Test Future Frequency Frequency Change 10 Years No 10 Years No R1 No R1 No R1 No R1 No R1 No R1 No R1 No R1 No R1 No R1 No R1 No R1 No R1 No

FENOC Evaluation of the Proposed Amendment _

Davis-Besse Nuclear Power Station Page 57 of 76 Electrical penetrations 101 and 102 are tested on a refueling outage frequency, and are not on extended test intervals due to their inability to establish good performance.

Additionally, Type B testing is performed on penetrations 101 and 102 to support nitrogen check valve reverse flow testing that is required every outage.

The remaining 13 mechanical penetration components are not eligible for testing on an extended test frequency. The two flange components associated with penetrations 23 and 24 are not eligible for an extended interval due to technical specification requirements. Regulatory Guide 1.163 does not allow extension of the testing interval for the emergency air lock (Penetration 80), and the personnel air lock (Penetration 81 ).

The eight shaft seal components associated with penetrations 80 and 81 were installed in the Spring 2018 refueling outage (20RFO). Testing of the equipment hatch (Penetration 82) each refueling outage is required because the equipment hatch is removed each refueling outage.

Performance Summary For DBNPS, Type B and Type C components eligible for extended intervals, following implementation of 75-month Type C test frequencies in 2015, are on extended intervals except for valves CV5079, CC1407C and RC229C which failed the as-found LLRT in 19RFO, 20RFO, and 20RFO, respectively.

A potential adverse trend can be identified over the last three outages. However, substantial leakage margin exists and changes in leakage can be attributed to increased leakage from vacuum breaker penetrations (88 and 8J) as it pertains to as-found containment total leakage. Valves CV5079, and CC1407C were replaced in 20RF0-2018 with leakage returning to an acceptable level. Though it did not impact the 20RFO as-found minimum pathway containm~nt total leakage, RC229C was replaced and returned to acceptance leakage.

3.6 Operating Experience 3.6.1 Aging Management Operating Experience ASME Code,Section XI, Subsection IWE was incorporated into 10 CFR 50.55a in 1996.

Prior to this time, operating experience pertaining to degradation of steel components of containment was gained through the inspections required by 10 CFR Part 50, Appendix J and ad hoc inspections conducted by licensees and the NRC. NRC Information Notice numbers 86-99, 88-82, 89-79, 04-09, and NUREG-1522, "Assessment of lnservice Conditions of Safety-Related Nuclear Plant Structures,"

described occurrences of corrosion in steel containment shells. NRC Generic Letter 87-05, "Request for Additional Information -Assessment of Licensee Measures to Mitigate and/or Identify Potential Degradation of Mark I Drywells," addressed the potential for corrosion of Boiling Water Reactor Mark I steel drywells in the "sand pocket region."

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 58 of 76 NRC Information Notice 97-10 identified specific locations where concrete containments are susceptible to liner plate corrosion; Information Notice 92-20 described an instance of containment bellows cracking, resulting in loss of leak tightness.

Other operating experience indicates that foreign objects embedded in concrete have caused through-wall corrosion of the liner plate at a few plants with reinforced concrete containments.

The DBNPS lnservice Inspection Program considers the liner plate and containment shell corrosion and cracking concerns described in these generic communications.

Implementation of the inservice inspection requirements of Subsection IWE, in accordance with 10 CFR 50.55a, and augmented to consider operating experience, is a necessary element of aging management for steel components of steel and concrete containments through the period of extended operation.

Degradation of threaded bolting and fasteners in closures for the reactor coolant pressure boundary has occurred from boric acid corrosion, and stress corrosion cracking (NRC Bulletin 82-02, NRC Generic Letter 91-17). Stress corrosion cracking has occurred in high strength bolts used for nuclear steam supply system component supports as described in EPRI Report NP-5769 (Reference 27). Inspections in accordance with the augmented ASME Code,Section XI, Subsection IWE, incorporating recommendations documented in EPRI NP-5769 and TR-104213 (Reference 28), are necessary to ensure containment bolting integrity.

DBNPS containment examinations and tests required by the lnservice Inspection Program have been implemented in accordance with the established schedule.

There have been three conditions identified,. which h.ave required E;ngineering evaluation or repair or replacement activities.

1. Prior to the implementation date of IWE, the "sand pocket" in the annulus was found to hold moisture, which resulted in scale on the containment vessel surface in this region. The sand and scale were removed from this area and the containment vessel in this area was recoated. When the scale was removed, pitting of the containment vessel was identified. Ultrasonic thickness measurements verified that the minimum recorded vessel thickness was greater than the minimum required wall thickness. An engineering evaluation determined that the pitting was not detrimental to the containment vessel. The cause of the moisture in the sand pocket region was plugged floor drains.

2. During *the Cycle 12 refueling outage, seepage of water between the containment vessel and the floor of the sand pocket in the annulus was noted. Similar seepage was also noted during the Cycle 13 refueling outage and documented in condition reports. An engineering change request was implemented to add a moisture barrier to this region. The seepage wets the containment vessel at the interface between the containment vessel and the floor only. Access to perform examinations in this area is not available. Therefore, this area was addressed in the Cycle 12 and Cycle 13 refueling outages in accordance with 10 CFR 50.55a(b)(2)(ix)(A), which requires

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 59 of 76 that when conditions exist in accessible areas that could indicate the presence of, or result in, degradation to inaccessible areas, the following information be provided in the inservice inspection summary report required by IWA-6000:

A description of the type and estimated extent of degradation, and the conditions that led to the degradation; An evaluation of each area, and the result of the evaluation, and; A description *of necessary corrective actions.

3. Another condition report identified that gaps had formed at two areas between the containment vessel and the concrete ledge on the inside of containment at the' 565-foot elevation. This condition was evaluated, and corrective actions were developed. Although no actual degradation has been identified as a result of these gaps, the affected areas were designated as surface areas requiring augmented examination (Examination Category E-C) as required by IWE-1240. Access to these areas of the containment vessel is only available from one side in the annulus area.

Ultrasonic thickness readings were taken in these areas from the annulus in the Cycle 13 and Cycle 15 refueling outages in accordance with ASME Code Case N-605 requirements. The thicknesses in these areas have remained essentially unchanged since the initial Cycle 13 refueling outage ultrasonic thickness readings.

Examinations scheduled since the third period of the second inservice inspection interval have been completed. These examinations and tests performed to date have satisfied the acceptance standards contained within Article IWE-3000. lnservice inspection records are maintained in accordance with Article IWA-6000 and are maintained in the permanent plant file storage.

3.6.2 NRC Containment Liner Corrosion Operating Experience Summary An NRC containment liner corrosion operating experience summary report (Reference 29) noted that in July 2002 at DBNPS, corrosion was identified where the containment meets the floor. Ultrasonic test (UT) examinations were performed to confirm that the freestanding metal containment had not been corroded below the minimum design thickness. A moisture barrier was subsequently installed at the containment-to-floor junction to prevent moisture intrusion.

3.6.3 Information Notice 2004-09, "Corrosion of Steel Containment and Containment Liner" Information Notice 2004-09 references the corrosion identified, in the Cycle 13 refueling outage, on the DBNPS containment vessel as one of the industry occurrences that led to the issuance of the Information Notice. The information notice discussion refers to an amendment to Section 50.55a of Title 10 of the Code of Federal Regulations (10 CFR 50.55a) (61 FR 41303). This amendment requires inservice inspections be performed in accordance with the ASME Code,Section XI, Subsections IWE and IWL. The DBNPS containment vessel was subject to an intensive corrosion investigation during

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 60 of 76 the Cycle 13 refueling outage. This investigation included the use of inservice inspection methods to identify material loss due to corrosion.

The containment vessel corrosion assessment used UT thickness measurements of the vessel as one of the investigation methods. The UT measurements verified that the minimum recorded vessel wall thickness (1.404 inches) was greater than ttie minimum required wall thickness (1.35 inches), as documented in a plant calculation. It was determined that the containment vessel was operable for continued service during all plant modes.

The containment vessel is inspected in accordance with the requirements of IWE of ASME Code,Section XI. These inspections include a visual examination of the entire internal surface of the containment vessel every 3-1 /3 years as well as visual inspection of the internal moisture barrier at the concrete-to-steel interface. The internal moisture barrier is inspected each outage. The interior and exterior moisture barriers were installed to protect uncoated portions of the vessel and to minimize exposure to water.

These inspections exceed the ASME Code,Section XI inspection frequency requirements.

As a result of a condition report corrective action, the containment vessel area behind the* interior concrete structure was designated as an area susceptible to corrosion and the augmented examination requirements of IWE were imposed. The augmented inspection areas received UT thickness examinations every 3-1/3 years until 3 successive examinations show no evidence of on-going corrosion. Ultrasonic test readings taken in 2008 showed that the containment vessel wall thicknesses in these areas have remained essentially unchanged since the initial Cycle 13 refueling outage UT readings were taken. These augmented inspections were performed according to the third inservice inspection interval in 15RFO, 17RFO, and the first period of the fourth inservice inspection interval in 18RFO. UT results were acceptable. No further corrective actions are required.

3.6.4 Information Notice 2014-07, "Degradation of Leak-Chase Channel Systems for Floor Welds of Metal Containment Shell and Concrete Containment Metallic Liner" This information notice addresses issues identified by the NRG staff 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. This information notice indicates, in part, that since leak-chase channels have a configuration that could permit moisture intrusion and water accumulation into inaccessible containment shell and liner areas, inservice inspection of accessible components of the leak-chase channel system is to be performed as required by ASME Code,Section XI, Subsection IWE (Reference 30).

DBNPS is not designed with a leak chase system for floor welds; hence, this information notice is not applicable.

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 61 of 76 3.6.5 Shield Building Laminar Cracking Shield building cracking was initially discovered in the fall of 2011 during opening of the shield building to support installation of a new reactor vessel closure head. Core bores were performed at that time to quantify the cracking.

Investigation of the shield building cracking determined that moisture intrusion into the concrete along with severe cold temperatures was a probable cause for the crack development. The exterior surfaces of the shield building were coated to prevent further moisture intrusion.

An inspection plan has been established to periodically monitor the shield building for cracking and crack propagation. Inspections performed in 2016 identified areas where concrete cracks had grown. These areas have been evaluated and it has been determined that the shield building's overall integrity has not been adversely affected.

Further investigation into crack growth is in progress.

Aging Management Program Description The Shield Building Monitoring Program is a plant-specific prevention and condition-monitoring program for DBNPS. The program consists of inspections of the shield building concrete and reinforcing steel (rebar). The inspections conducted as part of the Shield Building Monitoring Program supplement the inspections conducted as part of the Structures Monitoring Program.

The program monitors for cracking, change of material properties, and loss of material of concrete. The program also monitors for corrosion of the concrete rebar. As a preventive action of this program, the shield building wall, shield building dome and shield building emergency air lock. enclosure wall exterior concrete coatings will be inspected at a 5-year interval for evidence of loss of effectiveness. Also, the shield building wall, shield building dome, and shield building emergency air lock enclosure wall exterior concrete coatings will be reapplied at a 15-year interval.

Visual inspections are performed on rebar (when exposed), core bore, and core bore sample (concrete core) surfaces in accordance with an implementing procedure by inspectors qualified as described in Chapter 7 of American Concrete Institute (ACI)

Report ACI 349.3R, "Evaluation of Existing Nuclear Safety-Related Concrete Structures." The quantitative acceptance criteria for coatings from Chapter 5, Sections 5.1.4 and 5.2.4, of ACI Report 349.3R-02 are used (Reference 31 ).

The Shield Building Monitoring Program includes periodic scheduled inspections to ensure that the existing environmental conditions are not causing material degradation that could result in loss of shield building intended functions during the period of extended operation.

Implementation of this program ensures that the intended functions of the shield building and shield building emergency air'lock enclosure are maintained during the period of extended operation.

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 62 of 76

. In NUREG-2193, Supplement 1, "Safety Evaluation Report Related to the License Renewal of DBNPS," dated April 2016 (Reference 32, page 3-58) the NRC concluded, in part, that:

FENOC has demonstrated that the effects of aging on the shield building laminar cracking will be adequately managed so that the intended function(s) will be maintained consistent with the current licensing basis for the period of extended operation, as required by 10 CFR 54.21 (a)(3). The staff also reviewed the USAR [Updated Safety Analysis Report] supplement for this AMP [aging management program], as amended by letter dated July 3, 2014

[Accession No. ML141848184], and concludes that it provides an adequate summary description of the program, as required by 10 CFR 54.21 (d).

3. 7 License Renewal Aging Management DBNPS Updated Final Safety Analysis Report, Chapter 18, Managing the Effects of Component Aging, contains the Updated Final Safety Analysis Report Supplement as required by 10 CFR 54.21 (d) for the DBNPS License Renewal Application. These programs and activities were developed to support renewal of the original operating license for DBNPS that was scheduled to expire on April 22, 2017.

An integrated plant assessment in support bf license renewal identified the aging management programs and activities necessary to provide reasonable assurance that components within the scope of license renewal will continue to perform their intended functions for the period of extended operation. The period of extended operation is the 20-year period ending April 22, 2037.

For each of the plant-specific time-limited aging analyses, the evaluations have determined that the analyses remain valid for the period of extended operat(on; the analyses have been projected to the end of the period of extended operation; or, that the effects of aging on the intended function(s) will be adequately managed for the period of extended operation.

The license renewal integrated plant assessment and evaluation of time-limited aging analyses identified existing and new aging management programs necessary to provide reasonable assurance that components within the scope of licP.nse renewal will continue to perform their intended functions consistent with the current licensing basis for the period of extended operation. The aging management programs identified as necessary in association with the evaluation of time-limited aging analyses are described in Sections 18.1.14 and 18.1.16 of the DBNPS Updated Safety Analysis Report.

Appendix A of NUREG-2193, "Safety Evaluation Report Related to the License Renewal of Davis-Besse Nuclear Power Station," dated April 2016, and Supplement 1 to NUREG-2193, dated April 2016 (Reference 32), identified commitments associated with the aging management programs and activities to manage aging effects for structures and components. These commitments are provided in DBNPS Updated Safety Analysis Report, Table 18-1, "License Renewal Commitments."

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 63 of 76 The following programs described in DBNPS Updated Safety Analysis Report Sections 18.1.22, and 18.1.42, and activities described in license renewal commitments, are credited with the aging management of the primary containment.

18.1.22, lnservice Inspection Program - IWE The lnservice Inspection Program - IWE establishes responsibilities and requirements for conducting ASME Code,Section XI, Subsection IWE (Reference 30) inspections as required by 10 CFR 50.55a. The lnservice Inspection Program - IWE includes examination and testing of accessible surface areas of the steel containment; containment hatches and air locks; seals, gaskets and moisture barriers; and containment pressure-retaining bolting in accordance with the requirements of IWE.

The program includes surface examinations to monitor for cracking of stainless-steel penetration sleeves, dissimilar metal welds, bellows, and steel components that are subject to cyclic loading but have no current licensing basis fatigue analysis. The inspection sample size includes 10 percent of the containment penetration population that are subject to cyclic loading but have no current licensing basis fatigue analysis.

Penetrations included in the inspection sample will be scheduled for examination in each 10-year inservice inspection interval that occurs during the period of extended operation. Should fatigue analyses be performed in the future for the subject containment penetrations, the surface examinations will no longer be required. In addition, the 10 CFR 50, Appendix J Program provides for verification that a general visual inspection of the accessible interior and exterior surfaces of the primary containment and components (includes penetrations) has been performed prior to the Type A test pressurization to identify evidence of structural deterioration that might affect either the primary containment structural integrity or leak tightness.

The inservice examinations 1 conducted throughout the service life of DBNPS will comply with the requirements of the ASME Code,Section XI, Edition and Addenda incorporated by reference in 10 CFR 50.55a(b) 12-months prior to the start of the inspection interval, subject to prior approval of the edition and addenda by the NRC.

18.1.42, Nuclear Safety-Related Coatings Program The Nuclear Safety-Related Protective Coatings Program monitors the performance of Service Level 1 coatings inside containment through periodic coating examinations, condition assessments, and remedial actions, including repair or testing. The Nuclear Safety-Related Protective Coatings Program defines roles, responsibilities, controls and deliverables for monitoring the condition of coatings in containment. This program also ensures that the design basis accident analysis limits with regard to debris loading from failed coatings will not be exceeded for the emergency core cooling systems suction strainers.

License Renewal Commitment Number 27 DBNPS surveillance test procedure titled "Containment Vessel and Shielding Building Visual Inspection," Subsection 2.1.2, was enhanced to state:

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 64 of 76 Personnel who perform general visual examinations of the exterior surface of the Containment Vessel and the interior and exterior surfaces of the Shield Building shall meet the requirements for a general visual examiner in accordance with NOP-CC-5708, Written Practice for the Qualification and Certification of Nondestructive Examination Personnel.

These individuals shall be knowledgeable of the types of conditions which may be expected to be identified during the examinations.

License Renewal Commitment Number 35 Perform the following actions for each of two examinations (Phase 1 [complete] and Phase 2) of the containment vessel in the sand pocket region:

Perform nondestructive examination of the containment vessel from the outer surface at five areas of previously identified groundwater in-leakage.

o Examine the vessel at a minimum of three vertical grid locations at 12 inches nominal horizontal spacing at each area. Examine the containment vessel at a minimum of three elevations:

1. approximately 3 inches below the existing grout-to-vessel interface in the sand pocket region;

2. at the existing grout-to-vessel interface level in the sand pocket region; and,

3. approximately 3 inches above the existing grout-to-vessel interface in the sand pocket region.

Compare the UT thickness readings to minimum ASME Code vessel thickness requirements anq to the results obtained 9u,ring previous UT examinations of the containment vessel. Determine the need for maintenance or repair of the containment vessel based on the results and evaluation of the examinations.

Document the results of each of the two examinations in the work order system.

Document and evaluate adverse conditions in accordance with the FENOC Corrective Action Program for an evaluation of potential degradation of the steel containment vessel thickness over the longer term.

License Renewal Commitment Number 36 Perform the following actions related to the containment vessel sand pocket region each refueling outage ( ongoing activity):

Perform visual inspection of 100 percent of the accessible areas of the wetted outer surface of the containment vessel in the sand pocket region.

Perform visual inspection of accessible dry areas of the outer surface of the containment vessel in the sand pocket region and the areas above the grout-to-steel interface up to Elevation 566 feet plus 3 inches, minus 1 inch.

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 65 of 76 Perform visual inspection for deterioration (for example, missing or damaged grout) of accessible grout and the containment exterior moisture barrier in the sand pocket area.

Perform opportunistic visual inspections of inaccessible areas of the containment vessel in the sand pocket region when such areas are made accessible.

Perform opportunistic visual inspections for deterioration (for example, missing or damaged grout) of inaccessible grout in the sand pocket region when such areas are made accessible. Inaccessible grout is the grout below the normally-exposed surface of the grout in the sand pocket area.

Address issues of pitting or microbiologically-influenced corrosion, and degraded grout, moisture barrier or sealant identified during the inspections using the FENOC Corrective Action Program.

Sample the water in the sand pocket region when sufficient volumes are available.

The number of sampled water volumes will be determined by the number of water volumes observed and the size of those water volumes. Analyze the sample(s) for pH, chlorides, iron and sulfates. Treat or wash (or a combination thereof) the sand pocket area to reduce measured chloride concentrations to less than 250 parts per million (ppm) if the concentration of chlorides in a sample exceeds 250 ppm.

Note: Water samples may be taken at different times during each outage.

Engineering judgment may be used to determine the priority of the chemical analyses to be performed if sufficient water is not available in a given sample for all analyses.

License Renewal Commitment Number 39 Address the potential for borated water degradation of the steel containment vessel through the following actions:

Access the inside surface of the embedded steel containment at a vertical height no greater than 10 inches above bottom dead center. A core bore was completed (Phase 1 ). If necessary, a second core bore will be completed by the end of 2020 (Phase 2). If there is evidence of the presence of borated water in contact with the steel containment vessel, conduct non-destructive testing (NOT) to determine what effect, if any, the borated water has had on the steel containment vessel. Based on the results of NOT, perform a study to determine the effect through the period of extended operation of any identified loss of thickness in the steel containment due to exposure to borated water.

The NRC evaluated the core bore results of Phase 1 of Commitment No. 39 and noted that there was no evidence of the presence of borated water in the concrete or degradation of the inaccessible portion of the steel containment vessel due to borated water at the locations of the core bores.

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 66 of 76 License Renewal Commitment Number 47 Enhance the lnservice Inspection Program - IWE to:

Include surface examinations to monitor for cracking of containment stainless steel penetration sleeves, dissimilar metal welds, bellows, and steel components that are subject to cyclic loading but have no current licensing basis fatigue analysis. The inspection sample size will include 10 percent of the containment penetration population that is subject to cyclic loading but has no current licensing basis fatigue analysis. Penetrations included in the inspection sample will be scheduled for examination in each 10-year inservice inspection interval that occurs during the period of extended operation. Should fatigue analyses be performed in the future for the subject containment penetrations, the surface examinations will no longer be required.

This commitment was closed based on a structural integrity calculation titled, "Davis-Besse Containment Penetration Fatigue Analyses," that concluded containment penetrations meet the ASME Code requirement for fatigue usage.

3.8 NRC Safety Evaluation Limitations and Conditions 3.8.1 Limitations and Conditions Applicable to NEI 94-01, Revision 2-A The NRG staff found that the use of NEI TR 94-01, Revision 2, was acceptable for referencing by licensees proposing to amend their Technical Specifications to permanently extend the Type A test surveillance interval to 15 years, provided the conditions from the Safety Evaluation (SE), as listed in Table 3.8.1-1, are satisfied.

Table 3.8.1-1 also provides the DBNPS response to each condition.

Table 3.8.1-1 NEI 94-01 Revision 2-A Limitations and Conditions Limitation/Condition DBNPS Response (From Section 4.0 of SE)

For calculating the Type A leakage rate, DBNPS will utilize the definition in the licensee should use the definition in NEI 94-01, Revision 3-A, Section 5.0.

the NEI TR 94-01, Revision 2, in lieu of This definition has remained unchanged that in ANSI/ANS-56.8-2002. (Refer to SE from Revision 2-A to Revision 3-A of Section 3.1.1.1.)

NEI 94-01.

The licensee submits a schedule of Reference Section 3.5.2 (Tables 3.5.2-:1, containment inspections to be performed 3.5.2-2, 3.5.2-3, 3.5.2-4 and 3.5.2-5) of prior to and between Type A tests. (Refer this license amendment request to SE Section 3.1.1.3.)

submittal.

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 67 of 76 Table 3.8.1-1 NEI 94-01 Revision 2-A Limitations and Conditions Limitation/Condition (From Section 4.0 of SE)

The licensee addresses the areas of the containment ~tructure potentially subjected to degradation. (Refer to SE Section 3.1.3.)

The licensee addresses any tests and inspections performed following major modifications to the containment structure, as applicable. (Refer to SE Section 3.1.4.)

The normal Type A test interval should be less than 15 years. If a licensee has to utilize the provision of Section 9.1 of NEI TR 94-01, Revision 2, related to extending the ILRT interval beyond 15 years, the licensee must demonstrate to the NRC staff that it is an unforeseen emergent condition. (Refer to SE Section 3.1.1.2.)

DBNPS Response Reference Section 3.5.2 and Tables 3.5.2-3 and 3.5.2-4 of this license amendment request submittal.

There have been three (3) major containment repairs or modifications performed on the DBNPS containment vessel. The first two were for reactor head replacement activities and the third was for steam generator replacements.

Reference Section 3.2 of this license amendment request submittal.

DBNPS will follow the requirements of NEI 94-01, Revision 3-A, Section 9.1.

This requirement has remained unchanged from Revision 2-A to Revision 3-A of NEI 94-01.

In accordance with Section 3.1.1.2 of the June 25, 2008 NRC safety evaluation for NEI 94-01, Revision 2-A, FENOC will demonstrate to the NRC staff that an unforeseen emergent condition exists in the event an extension beyond the 15-year interval is required.

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 68 of 76 Table 3.8.1-1 NEI 94-01 Revision 2-A Limitations and Conditions Limitation/Condition (From Section 4.0 of SE)

For plants licensed under 10 CFR Part 52, applications requesting a permanent extension of the ILRT surveillance interval to 15 years should be deferred until after the construction and testing of containments for that design have been completed and applicants have confirmed the applicability of NEI 94-01, Revision 2, and EPRI Report No. 1009325, Revision 2, including the use of past containment ILRT data.

DBNPS Response Not applicable. DBNPS was not licensed under 10 CFR Part 52.

3.8.2 Limitations and Conditions Applicable to NEI 94-01, Revision 3-A A safety evaluation dated June 8, 2012 (incorporated in Reference 2) docunients the NRC staff's evaluation and acceptance of NEI TR 94-01, Revision 3, subject to the limitations and conditions summarized in Section 4.0 of the safety evaluation.

By letter dated October 9, 2015 (Reference 10), the NRC issued Amendment 288 to the DBNPS facility operating license to adopt the guidance in NEl-94-01, Revision 3-A, for Type C testing. The NRC staff determined that the limitations and conditions sl.immadzed in Section 4.0 of the June 8, 2012 safety evaluation are adequately addressed as described in the following sections of the. October 9, 2015 safety evaluation.

Section 3.2, "NRC Condition 1" Section 3.3, "NRC Condition 2" Section 3.4, "Technical Summary" 3.9 Conclusion NEI 94-01, Revision 3-A, dated July 2012, and the limitations and conditions specified in NEI 94-01, Revision 2-A, dated October 2008, describe an NRG-accepted approach for implementing the performance-based requirements of 10 CFR 50, Appendix J, Option B. NEI 94-01, Revision 3-A, incorporates the regulatory positions stated in Regulatory Guide 1.163 and includes provisions for extending Type A intervals to 15 years and Type C test intervals to 75 months. NEI 94-01, Revision 3-A, delineates a performance-based approach for determining Type A, Type B, and Type C containment leakage rate surveillance test frequencies. FENOC proposes to adopt the guidance of

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 69 of 76 NEI 94-01, Revision 3-A, with the limitations and conditions specified in NEI 94-01, Revision 2-A, for the DBNPS, 10 CFR 50, Appendix J testing program plan.

Based on the previous Type A tests conducted at DBNPS, FENOC concludes that the permanent extension of the containment Type A test interval from 10 to 15 years represents minimal risk to increased leakage. The risk is minimized by continued Type B and Type C testing performed in accordance with Option B of 10 CFR 50, Appendix J, and the overlapping inspection activities performed as part of the following DBNPS inspection programs:

lnservice Inspection Program Nuclear Safety-Related Coatings Program Containment Inspections per Technical Specification Surveillance Requirement 3.6.1.1 This experience is supplemented by risk analysis studies, including the DBNPS risk analysis provided in Attachment 1. The risk assessment concludes that increasing the Type A test interval on a permanent basis to a one-in-fifteen-year frequency is not considered to beisignificant because it represents only a small change in the DBNPS risk profile.

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements and Criteria The proposed change has been evaluated to determine whether applicable regulations and requirements continue to be met.

10 CFR 50.54(0) requires primary reactor containments for water-cooled power reactors to be subject to the requirements of 10 CFR 50, Appendix J, "Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors." Testing requirements in 10 CFR Part 50, Appendix J, ensure that: (a) leakage through containments or systems and components penetrating containments does not exceed allowable leakage rates specified in the Technical Specifications; and (b) integrity of the containment structure is maintained during its service life.

10 CFR 50, Appendix J, includes two options: "Option A - Prescriptive Requirements,"

and "Option B - Performance Based Requirements," either of which can be chosen for meeting the requirements of the Appendix.

Option B of 10 CFR Part 50, Appendix J, specifies performance-based requirements and criteria for preoperational and subsequent leakage-rate testing. These requirements are met by performance of Type A tests to measure the containment system overall integrated leakage rate, Type B tests to detect and measure local leakage rates across pressure-retaining or leakage-limiting boundaries such as penetrations, and Type C tests to measure containment isolation valve leakage rates. After the preoperational

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 70 of 76 tests, these tests are required to be conducted at periodic intervals based on the historical performance of the overall containment system (for Type A tests) and based on the safety significance and historical performance of each boundary and isolation valve (for Types Band C tests) to ensure integrity of the overall containment system as a barrier to fission product release.

Paragraph V.B.3, of 10 CFR 50, Appendix J, Option B, requires that the regulatory guide or other implementation document used by a licensee to develop a performance-based leakage-testing program be included, by general reference, in the plant technical specifications. Furthermore, the submittal for technical specification revisions must contain justification, including supporting analyses, if the licensee chooses to deviate from methods approved by the Commission and endorsed in Regulatory Guide 1.163.

FENOC currently implements a performance-based containment leakage testing program at DBNPS to comply with Option B. The implementation documents currently referenced in Technical Specification 5.5.15 are Regulatory Guide 1.163, dated September 1995, for Type A and Type B tests, and NEI 94-01, Revision 3-A, for Type C tests. Revision O of NEI 94-01, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," dated July 26, 1995, is endorsed by Regulatory Guide 1.163 as a document that provides methods acceptable to the NRC staff for complying with the provisions of Option B, subject to four regulatory positions delineated in Section C of Regulatory Guide 1.163. Revision O of NEI 94-01 includes provisions that allow the performance-based Type A test interval to be extended to up to 10 years, based upon two consecutive successful tests.

In the NRC final safety evaluation for NEI 94-01, Revision 2, the NRC staff concluded that Revision 2 of NEI 94-01 describes an acceptable approach for implementing the

,optional performance-based requirements of Option B to 10 CFR 50, Appendix J, and is acceptable for referencing by licensees proposing to amend their Technical Specifications in regard to containment leakage rate testing, subject to the specific limitations and conditions listed in Section 4.0 of the final safety evaluation. NEI 94-01, Revision 2, incorporates the regulatory positions stated in Regulatory Guide 1.163 and includes provisions for extending Type A test intervals up to 15 years.

In the NRC final safety evaluation for Revision 3 of NEI 94-01, dated June 8, 2012, *the NRC staff concluded that Revision 3 of NEI 94-01 describes an acceptable approach for implementing the optional performance-based requirements of Option B, as modified by the conditions and limitations summarized in Section 4.0 of the safety evaluation. The NRC staff also stated in the safety evaluation that the guidance, as modified to include two limitations and conditions, is acceptable for referencing by licensees proposing to amend their Technical Specifications regarding containment leakage rate testing.

Revision 3 of NEI 94-01 includes guidance for extending Type C LLRT surveillance intervals up to 75 months. The adoption of NEI 94-01, Revision 3-A for the performance of Type C testing and the associated frequency change was previously approved for DBNPS in License Amendment 288 (Reference 10). Revision 3-A of NEI 94-01 and Revision 2-A of NEI 94-01 include their corresponding NRC staff safety evaluations.

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 71 of 76 The regulations in 10 CFR 50.55a, "Codes and Standards," specify the containment inservice inspection requirements that, in conjunction with the requirements of Appendix J, ensure the continued leak-tight and structural integrity of the containment during its service life. The proposed change would not revise the DBNPS Containment lnservice Inspection Program implemented to maintain compliance with the requirements of 10 CFR 50.55a.

The proposed change would revise Technical Specification 5.5.15 to follow the NRG approved guidance of NEI 94-01, Revision 3-A, with limitations and conditions specified in NEI 94-01, Revision 2-A, for Type A and Type B tests. Based on the foregoing, the proposed Technical Specification changes would continue to ensure compliance with 10 CFR 50.54(0), and Option B of 10 CFR 50, Appendix J.

4.2 Precedent This license amendment request is like Amendment No. 265 that was approved for Point Beach Nuclear Plant, Unit No. 1, on April 25, 2019 (Reference 33), except that it excludes changes to the Type C test provisions that were previously approved for DBNPS (Reference 10). The amendment for the Point Beach Nuclear Plant revised Technical Specifications to require a containment leakage testing program in accordance with the guidelines contained in NEI 94-01, Revision 3-A, and conditions and limitations specified in NEI 94-01, Revision 2-A.

4.3 No Significant Hazards Consideration The proposed license amendment would revise Technical Specification 5.5.15, "Containment Leakage Rate Testing Program," to follow guidance developed by the Nuclear Energy Institute (NEI) and presented in topical report, NEI 94-01, Revision 3-A, "Industry Guideline for Implementing Performance-Based Option of, 10 CFR Part 50, Appendix J," with conditions and limitations specified in NEI 94-01, Revision 2-A with the same title, for containment overall integrated leakage rate testing (Type A testing) and testing to detect local leaks and to measure leakage across pressure-containing or leakage limiting boundaries (other than valves) for primary containment penetrations (Type B testing). The NEI 94-01, Revision 3-A, guidelines would allow the Type A test interval to be changed from 10 years to 15 years and require a more conservative allowable test interval extension of nine months for Type A and Type B tests.

FirstEnergy Nuclear Operating Company has evaluated whether a significant hazards consideration is involved with the proposed amendment 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 test interval extensions do not involve either a physical change to the plant or a change in the way the plant is operated or controlled. The containment is designed to provide an essentially leak tight barrier against the uncontrolled release of radioactivity to the environment for postulated accidents. As such, the

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 72 of 76 containment and the testing requirements invoked to periodically demonstrate the integrity of the containment exist to ensure the plant's ability to mitigate the consequences of an accident and do not involve the prevention or identification of any precursors of an accident.

The change in Type A test frequency to once-per-fifteen years, measured as an increase to the total integrated plant risk for those accident sequences influenced by Type A testing, based on the internal events probabilistic risk analysis is 0.016 person-Roentgen Equivalent Man (rem) per year. In Section 3.2.4.6, "Acceptance Guidelines," of the final safety evaluation for NEI 94-01, Revision 2, the Nuclear Regulatory Commission staff concluded that for the purposes of assessing the risk impacts of the Type A test extension in accordance with the Electric Power Research Institute Report Number 1009325, Revision 2, methodology, a small increase in population dose should be defined as an increase in population dose of less than or equal to 1.0 person-rem per year or less than or equal to 1 percent of the total population dose, whichever is less restrictive. The risk impact for the integrated leak rate test interval extension when compared to other severe accident risks is negligible.

As documented in the Nuclear Regulatory Commission technical supportidocument NUREG-1493, "Performance-Based Containment Leak-Test Program," dated September 1995, Type B and Type C testing can detect a very large percentage of containment leakages, and the percentage of containment leakages that can be detected only by Type A testing is very small. The DBNPS Type A test history supports this conclusion.

Based on the above paragraphs, the proposed test interval extensions do not involve a significant increase in the probability of an accident previously evaluated.

The overall containment leak rate limit is maintained with the proposed test interval extension changes. Since the proposed changes do not result in a significant increase in containment leakage, the changes do not involve a significant increase in the consequences of a previously evaluated accident.

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 containment and the testing requirements to periodically demonstrate the integrity of the containment exist to ensure the plant's ability to mitigate the consequences of an accident and do not involve any accident precursors or initiators. The proposed change does not alter the design or configuration of the plant (that is, no physical change will be made to the plant and no new or different

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 73 of 76 type of equipment will be installed), nor does the proposed change alter the manner in which the plant is operated or controlled.

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.

This amendment does not alter the way safety limits, limiting safety system set points, or limiting conditions for operation are determined. The specific requirements and conditions of the Technical Specification 5.5.15, "Containment Leakage Rate Testing Program," exist to ensure that the degree of containment structural integrity and leak-tightness that is considered in the plant safety analysis is maintained. The overall containment leak rate limit is maintained. The design, operation, testing methods and acceptance criteria for Type A and B containment leakage tests specified in applicable codes and standards would continue to be met, with the acceptance of this proposed amendment, since they are not affected by implementation of a performance-based containment testing program.

The combination of the above factors ensures that the margin of safety in the plant safety analysis is maintained. Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, FirstEnergy Nuclear Operating Company 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 con~ideratio.n" is justified.,

, 1 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 Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

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. 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

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 74 of 76 environmental assessment need be prepared in connection with the proposed amendment.

6.0 REFERENCES

1.

Regulatory Guide 1.163, "Performance-Based Containment Leak-Test Program,"

September 1995 (Accession Number ML003740058).

2.

NEI 94-01, Revision 3-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," July 2012 (Accession Number ML12221A202).

3.

NEI 94-01, Revision 2-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," October 2008 (Accession Number ML100620847).

4. American Nuclear Society, ANSI/ANS 56.8-2002, "Containment System Leakage Testing Requirements," LaGrange Park, Illinois, November 2002.

5.

Letter from NRC (J. S. Bowen) to FirstEnergy Nuclear Operating Company (R. Lieb), Safety Evaluation in Support of Proposed Alternative Regarding Post-Repair Pressure Testing Requirements (TAC No. MF0537), dated May 8, 2013 (Accession Number ML13121A404).

6.

NEI 94-01, Revision 0, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J," dated July 21, 1995 (Accession Number ML11327A025).

7.

NUREG-1493, "Performance-Based.Containment Leak-Test Program - Final Report," September 1995 (Accession Number ML9510200161 ).

8.

Letter from NRC (L. L. Gundrum) to Centerior Service Company (J. P. Stetz),

Amendment No. 205 to Facility Operating License No. NPF Davis-Besse Nuclear Power Station, Unit 1 (TAC No. M94280), dated February 22, 1996 (Accession Number ML021210135).

9.

Letter from NRC (D. V. Pickett) to FirstEnergy Nuclear Operating Company (G. G.

Campbell), Davis-Besse Nuclear Power Station, Unit 1 - Issuance of Amendment

[No. 240] (TAC No. MA6093), dated March 28, 2000 (Accession Number ML003698061 ).

10. Letter from NRC (B. Purnell) to FirstEnergy Nuclear Operating Company (B. D.

Boles), Davis-Besse Nuclear Power Station, Unit 1 - Issuance of Amendment [No.

288] Related to Containment Leakage Rate Testing Program (TAC No. MF5433),

dated October 9, 2015 (Accession Number ML152399293).

FENOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 75 of 76

11. "Interim Guidance for Performing Risk Impact Assessments In Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals,"

Rev. 4, Developed for NEI by EPRI and Data Systems and Solutions, November 2001.

12. Regulatory Guide 1.174, Revision 3, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," January 2018 (Accession Number ML17317A256).

13. Letter from Constellation Nuclear (C. H. Cruse) to NRC (Document Control Desk),

Calvert Cliffs Nuclear Power Plant, Unit No. 1; Docket No. 50-317 - Response to Request for Additional Information Concerning the License Amendment Request for a One-Time Integrated Leakage Rate Test Extension, dated March 27, 2002 (Accession Number ML020920100).

14. EPRI TR-1018243, "Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals," Revision 2-A of 1009325, October 2008.

15. Letter from NRC (M. J. Maxin) to NEI (J. C. Butler), Final Safety Evaluation for Nuclear Energy Institute (NEI) Topical Report {TR) 94-01, Revision 2, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J" and Electric Power Research Institute (EPRI) Report No. 1009325, Revision 2, August 2007, "Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals" (TAC No. MC9663), dated June 25, 2008 (Accession Number ML081140105).

16. Davis-Besse Nuclear Power Station, "Davis-Besse Internal Events, Internal

,Flooding, and Seismic PRA Model," Revision 6:, PRA.-DB1-AL-R06.

17. American Society of Mechanical Engineers, "Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications," (ASME RA-S-2002), Addenda RA-Sb-2005, December 2005.

18. Regulatory Guide 1.200, Revision 2, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities,"

March 2009 (Accession Number ML090410014).

19. ASME/ANS, "Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," ASME/ANS RA-Sa-2009, dated March 2009. Addendum A to RA-S-2008.

20. NEI 05-04, Revision 2, "Process for Performing Internal Events PRA Peer Reviews Using the ASME/ANS PRA Standard," November 2008

21. NEI 07-12, "Fire Probabilistic Risk Assessment (FPRA) Peer Review Process Guidelines," Revision 1, June 2010.

22. NEI 12-13, "External Hazards PRA Peer Review Process Guidelines," August 2012.

I I

FE NOC Evaluation of the Proposed Amendment Davis-Besse Nuclear Power Station Page 76 of 76

23. ASME/ANS, "Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," ASME/ANS RA-Sb-2009, dated March 2009. Addendum B to RA-S-2008.

24. NUREG-2122, "Glossary of Risk-Related Terms in Support of Risk-Informed Decision Making" (Accession Number ML13311A353).

25. Regulatory Guide 1.54, "Quality Assurance Requirements for Protective Coatings Applied to Water-Cooled Nuclear Power Plants," Revision 0, dated June 1973 (Accession Number ML003740187).

26. ANSI N101.4-1972, "Quality Assurance for Protective Coatings Applied to Nuclear Facilities," Revision 0, dated June 1973.

27. EPRI NP-5769, "Degradation and Failure of Bolting in Nuclear Power Plants," dated April 1988 (Accession Number ML003727113).

28. TR-104213, "Bolted Joint Maintenance & Applications Guide," dated December 1995 (Accession Number ML003767012).

29. "Containment Liner Corrosion Operating Experience Summary," Technical Letter Report - Revision 1, by D. S. Dunn, A. L. Pulvirenti, and M. A. Hiser (Office of Nuclear Regulatory Research - NRG), dated August 2, 2011 (Accession Number ML112070867).

30. ASME Boiler & Pressure Vessel Code,Section XI, Subsection IWE, "Requirements for Class MC and Metallic Liners of Class CC Components of Light-Water Cooled Plants."

31. American Concrete Institute (ACI) Report ACI *349.3R-02, "Evaluation of Existing Nuclear Safety-Related Concrete Structures," June 17, 2002.

32. NUREG-2193, "Safety Evaluation Report Related to the License Renewal of Davis-Besse Nuclear Power Station," dated April 2016 (Accession Numbers ML16104A207, ML16104A301), and Supplement 1 to NUREG-2193, dated April 2016 (Accession Number ML16104A350).

33. Letter from NRG (D. J. Wrona) to NextEra Energy Point Beach, LLC (M. Nazar),

Subject:

"Point Beach Nuclear Plant, Units 1 and 2 - Issuance of Amendments to Extend Containment Leakage Rate Test Frequency," dated April 25, 2019, (Accession Number ML19064A904 ).

ATTACHMENT A Proposed Technical Specification Changes (2 Pages follow)

Programs and Manuals 5.5 5.5 Programs and Manuals 5.5.14 5.5.15 Safety Function Determination Program (continued)

3.

Provisions to ensure that an inoperable supported system's Completion Time is not inappropriately extended as a result of multiple support system inoperabilities; and

4.

Other appropriate limitations and remedial or compensatory actions.

b.

A loss of safety function exists when, assuming no concurrent single failure, no concurrent loss of offsite power, or no concurrent loss of onsite diesel generator(s), a safety function assumed in the accident analysis cannot be performed. For the purpose of this program, a loss of safety function may exist when a support system is inoperable; and

1.

A required system redundant to the system(s) supported by the inoperable support system is also inoperable; or

2.

A required system redundant to the system(s) in turn supported by the inoperable supported system is also inoperable; or

3.

A required system redundant to the support system(s) for the

• supported systems described in Specifications 5.5.14.b.1 and 5.5.14.b.2 above is also inoperable.

c.

The SFDP identifies where a loss of safety function exists. If a loss of safety function is determined to exi~_t by this program, the appropriate Conditions and Required Actions of the LCO in which the loss of safety f,unction exists are, required, tq be entered. When a loss of safety function is caused by the inoperability of a single Technical Specification support system, the appropriate Conditions and Required Actions to enter are those of the support system.

Containment Leakage Rate Testing Program

a.

, and the conditions and limitations specified in NEI 94-01, Revision 2-A, dated October 2008, A program shall establish the leakage rate testing of the ontainment as required by 10 CFR 50.54(0) and 10 CFR 50, Appen

• J, Option B, as modified by approved exemptions. Fer Ty!'e C tests, this program shall be in accordance with the guidelines contained in Nuclear Energy Institute (NEI) topical report NEI 94-01, Revision 3-A, "Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J,"

dated July 2012. For T~r,e A and Ty!'e B tests tl=lis pFOgraA1 shall be in aeeerdanee *ovith the guidelines eontained in Regulatery Guide 1.163, "Peffermetnee Betsed Centetinment Leetl< Test PregFBffi," detted Davis-Besse Septe111be1 1995, as modified by the following exceptions:

1.

A reduced duration Type A test may be performed using the criteria and Total Time method specified in Bechtel Topical Report BN-TOP-1, Revision 1.

5.5-12 Amendment~

Programs and Manuals 5.5

.------------------------------------------------------------~

j No change to this page is proposed. This page is included for context.. I L-----------------------------------------------------------~

5.5 Programs and Manuals 5.5.15 5.5.16 5.5.17 Davis-Besse Containment Leakage Rate Testing Program (continued)

2.

The fuel transfer tube blind flanges (containment penetrations 23 and*

24) will not be eligible for extended test frequencies. Their Type B test frequency will remain at 30 months. However, as-found testing will not be required.

b.

The calculated peak containment internal pressure for the design basis loss of coolant accident, Pa, is 38 psig.

c.

The maximum allowable containment leakage rate, La, at Pa, shall be 0.50% of containment air weight per day.

d.

Leakage rate acceptance criteria are:

1.

Containment leakage rate acceptance criterion is < 1.0 La. During the first unit startup following testing in accordance with this program, the leakage rate acceptance criteria are < 0.60 La for the Type B and C tests and~ 0.75 La for Type A tests.

2.

Air lock testing acceptance criteria are:

a)

Overall air lock leakage rate is~ 0.015 La when tested at~ Pa.

b)

For each door, leakage rate is~ 0.01 La when the volume between the door seals is pressurized to~ 10 psig.

e.

. The provisions of SR 3.0.3 are applicable to the Containment Leakage Rate Testing Program.

Battery Monitoring and Maintenance Program This Program provides for battery restoration and maintenance, including the following:

a.

Actions to restore battery cells with float voltage < 2.13 V;

b.

Actions to equalize and test battery cells that had been discovered with electrolyte level below the top of the plates; and

c.

Actions to verify that the remaining cells are > 2.07 V when a pilot cell or cells have been found to be < 2.13 V.

Control Room Envelope Habitability Program A Control Room Envelope (CRE) Habitability Program shall be established and implemented to ensure that CRE habitability is maintained such that, with an OPERABLE Control Room Emergency Ventilation System (CREVS), CRE 5.5-13 Amendment 279

ATTACHMENT B Evaluation of Risk Significance of Permanent

[Integrated Leakage Rate Test] ILRT Extension (45 Pages follow) 1 I