Text

License Amendment Request 268, Request to Extend Containment Leakage Rate Test Frequency
	ML20034D803
Person / Time
Site:	Turkey Point
Issue date:	01/27/2020
From:	Stamp B; Florida Power & Light Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	L-2020-003
	Download: ML20034D803 (129)
	v • d • e

17 27 2020 January 17, L-2020-003 10 CFR 50.90 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington DC 20555-0001 RE: Turkey Point Nuclear Plant, Unit 3 and 4, Docket Nos. 50-250 and 50-251 Subsequent Renewed Facility Operating Licenses DPR-31 and DPR-41 License Amendment Request 268. Request to Extend Containment Leakage Rate Test Frequency Pursuant to 10 CFR Part 50.90, Florida Power & Light Company (FPL) hereby requests amendments to Subsequent Renewed Facility Operating Licenses DPR-31 and DPR-41 for Turkey Point Nuclear Plant Units 3 and 4 (Turkey Point), respectively. The proposed license amendments revise Turkey Point Technical Specification (TS) 6.8.4.h, "Containment Leakage Rate Testing Program," to require a program in accordance with 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" and the conditions and limitations specified in NEI 94-01, Revision 2-A. to this letter provides FPL's evaluation of the proposed change. Enclosure 2 provides a markup of the current TS showing the proposed change. Enclosure 3 provides the plant-specific risk

impact assessment. Appendix A 1 to Enclosure 3 provides documentation related to the technical adequacy of the Turkey Point probabilistic risk assessment (PRA). No Turkey Point TS Bases changes are proposed.

FPL has determined that the proposed changes do not involve a significant hazards consideration pursuant to 10 CFR 50.92(c), and there are no significant environmental impacts associated with the proposed changes. The Turkey Point Onsite Review Group has reviewed the proposed license amendments. In accordance with 10 CFR 50.91 (b)(1 ), a copy of the proposed license amendments is being forwarded to the State designee for the State of Florida.

FPL requests that the proposed changes are processed as a normal license amendment request with approval within one year of the submittal date. Once approved, the amendments will be implemented within 90 days.

This letter contains no regulatory commitments.

Should you have any questions regarding this submission, please contact Mr. Robert Hess, Turkey Point Licensing Manager, at 305-246-4112.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on the -Z-=if.

-a day of January 2020.

Sincerely, Brian Stamp Site Director, Turkey Point Nuclear Plant Florida Power & Light Company 9760 SW 344th Street, Homestead, FL 33035

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Page 2 of 2 Enclosures cc: USNRC Regional Administrator, Region II USNRC Project Manager, Turkey Point Nuclear Plant USNRC Senior Resident Inspector, Turkey Point Nuclear Plant Ms. Cindy Becker, Florida Department of Health

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 1 Page 1 of 43 ENCLOSURE 1 EVALUATION OF THE PROPOSED CHANGES Turkey Point Nuclear Plant, Unit 3 and 4 License Amendment Request 268, Request to Extend Containment Leakage Rate Test Frequency 1.0

SUMMARY

DESCRIPTION............................................................................................................... 2 2.0 DETAILED DESCRIPTION ............................................................................................................... 2 2.1 System Design and Operation ............................................................................................ 2 2.2 Current Requirements ......................................................................................................... 4 2.3 Description of the Proposed Change .................................................................................. 4 2.4 Reason for the Proposed Change ...................................................................................... 5

3.0 TECHNICAL EVALUATION

.............................................................................................................. 6 3.1 Containment Leak Rate Testing History ............................................................................. 6 3.1.1 Integrated Leak Rate (Type A) Testing ............................................................................... 6 3.1.2 Local Leak Rate Testing - Type B & C................................................................................ 7 3.1.3 Extension of Type B and C Intervals................................................................................. 11 3.2 Containment Inspections .................................................................................................. 23 3.2.1 IWE Program..................................................................................................................... 23 3.2.2 IWL Program ..................................................................................................................... 25 3.2.3 Containment Visual Inspection ......................................................................................... 27 3.3 NRC Information Notices .................................................................................................. 28 3.4 NRC Limitations and Conditions ....................................................................................... 29 3.5 Plant-Specific Confirmatory Analysis ................................................................................ 34 3.5.1 Methodology...................................................................................................................... 34 3.5.2 Probabilistic Risk Assessment (PRA) Acceptability .......................................................... 36 3.5.3 Conclusions of the Plant-Specific Risk Assessment Results............................................ 37 3.6 Conclusion ........................................................................................................................ 38

4.0 REGULATORY EVALUATION

....................................................................................................... 39 4.1 Applicable Regulatory Requirements/Criteria ................................................................... 39 4.2 Precedents ........................................................................................................................ 40 4.3 No Significant Hazards Consideration .............................................................................. 40 4.4 Conclusion ........................................................................................................................ 41

5.0 ENVIRONMENTAL CONSIDERATION

.......................................................................................... 41

6.0 REFERENCES

................................................................................................................................ 42

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 1 Page 2 of 43 1.0

SUMMARY

DESCRIPTION Florida Power & Light Company (FPL) hereby requests amendments to Subsequent Renewed Facility Operating Licenses DPR-31 and DPR-41 for Turkey Point Nuclear Plant Units 3 and 4 (Turkey Point), respectively. The proposed license amendments revise Turkey Point Technical Specification (TS) 6.8.4.h, Containment Leakage Rate Testing Program, to require a program in accordance with 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 (Reference 6.1) and the conditions and limitations specified in NEI 94 01, Revision 2-A (Reference 6.2).

The NRC determined that NEI 94-01, Revision 3-A, describes an acceptable approach for implementing the optional performance based requirements of Option B to 10 CFR 50, Appendix J, as specified by the conditions and limitations in its Safety Evaluation (Reference 6.6). This proposed change will allow extension of the Type A test interval up to one test every 15 years and extension of the Type C test interval up to 75 months, based on acceptable performance history as defined in NEI 94 01, Revision 3-A.

2.0 DETAILED DESCRIPTION 2.1 System Design and Operation The Turkey Point Unit 3 and 4 containment buildings are post-tensioned, reinforced concrete structures comprised of vertical cylinders with hemispherical domed roofs supported on conventionally reinforced concrete base slabs. The containment completely encloses the reactor and reactor coolant system to ensure, with certain engineered safeguards, that an acceptable upper limit for leakage of radioactive materials to the environment will not be exceeded, even if the Maximum Hypothetical Accident were to occur. The design assures that the integrity of the reactor containment is maintained under normal and accident conditions.

The post-tensioning system utilizes un-bonded steel tendons nominally consisting of:

180 vertical tendons in the cylindrical wall anchored at the top surface of the ring girder and at the bottom of the base slab; 489 hoop tendons in the cylindrical wall anchored alternately at six vertical buttresses. Each tendon encloses 120 degrees of arc; 165 dome tendons anchored at the vertical face of the dome ring girder. The tendons are arranged in three groups of 55 tendons oriented at 120 degrees to each other.

Each tendon nominally consists of ninety 1/4-inch diameter wires with buttonhead Berkemeier, Brandestini, Ros and Vogt (BBRV) type anchorages. Tendons are housed in spirally-wrapped corrugated thin-wall sheathing and are capped at each end with a sheathing filler cap. After tensioning, the tendon sheathing and caps were filled with corrosion preventing grease. Sufficient post-tensioning is applied on the cylinder and dome to more than balance the internal pressure, leaving a margin of external pressure beyond that required to resist the design accident pressure.

The inside surface of the concrete structure is covered by a continuous steel liner that provides a high degree of leak tightness. A 1/4-inch thick liner plate is attached to the walls and dome by means of an angle grid system stitch welded to the liner plate and embedded in the concrete. A floor liner is installed on top of the containment base slab and is covered

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 1 Page 3 of 43 with a 1-6 thick concrete cover slab. The liner is not relied upon to assist the concrete in maintaining structural integrity. The liner plate, with the exception of the floor liner, is coated on the inside surface for corrosion protection. There is no paint on the outside surface which is in contact with the concrete shell. The floor liner is coated with a bond breaker to allow free thermal expansion of the cover slab.

Principal dimensions of the Turkey Point containment building are as follows:

Inside diameter................................................ 116 feet Inside height (including dome) ........................ 169 feet Vertical wall thickness .................................. 3-3/4 feet Dome thickness ............................................ 3-1/4 feet Foundation slab thickness ......................... 10-1/2 feet Internal Free Volume Minimum Estimated ................................ 1.45E+06 cu. ft.

Maximum Estimated ............................... 1.60E+06 cu. ft Penetrations through the containment include access openings, mechanical penetrations, and electrical penetrations. All are designed for leak tightness by being welded to the liner or through the use of gasketed seals.

The equipment hatch is a 14-foot diameter opening that is utilized during refueling outages to move personnel, equipment and materials into the containment interior at the mezzanine level. The steel hatch door bolts to a flange that is welded to the containment liner plate all around the opening. Double gaskets are provided at the seal between the door and flange.

Two hatches are provided for personnel to access the containment via airlocks that preserve the pressure boundary. The personnel access hatch and the emergency escape hatch are steel barrels provided with doors on each end to create an airlock along with an equalization line with a single valve through each side. Each door is mechanically latched closed against two redundant seal rings, though the door and equalization valve are considered the isolation barriers. The barrel is supported by the containment wall and welded to the steel liner.

Mechanical penetrations (for piping and ventilation systems) are of the rigid welded type.

Steel penetration sleeves are solidly anchored to the containment wall precluding any requirement for expansion bellows. External guides and stops are provided as required to limit displacements and resist bending and torsional moments to prevent rupture of the penetrations and the adjacent liner plate.

For the piping of non-essential lines that pass through these penetrations, containment isolation is provided by means of barriers. In some cases, a single integral barrier, such as a welded or threaded pipe cap, has been credited as the means of isolation. In other cases, redundant barriers consisting of valves, closed systems or other closure devices are utilized. Flanges used as penetration barriers either utilize double gaskets, such as on the fuel transfer tube, or, if provided only a single gasket, are utilized with a redundant barrier.

Electrical penetrations consist of shop-fabricated assemblies that consist of a carbon steel pipe canister with stainless steel or carbon steel header plates and segments of electrical conductors passing through. The header plates are welded to the canister and the electrical conductors are sealed where they pass through both header plates. For high voltage conductors the seal is made by welding hermetically sealed glass or ceramic bushings. A flange on each canister is welded to a steel sleeve that embedded in the containment wall and welded to the liner.

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 1 Page 4 of 43 The containment structure is licensed and designed to withstand a pressure of 55 psig and 283°F. The original transient analysis calculated a peak accident pressure of 49.9 psig and a peak accident temperature of 276°F. The higher licensed design pressure and temperature are based on the AEC's guidelines for containment design at the time of the original SER in 1972. Since these AEC guidelines suggested that the design pressure of containment should be at least 10% higher than the calculated peak accident pressure, 55 psig was found to be an acceptable design pressure. The containment designs were re-evaluated under extended power uprate (EPU) conditions at 2644 MWt core power. Peak containment pressure for LOCA and MSLB events was reanalyzed for EPU conditions.

The results are within the 55 psig containment design pressure and 283°F design temperature.

The containment structure, including access openings and penetrations, is designed to a maximum allowable leak rate of 0.20 percent by weight of containment air per day at the Containment design pressure of 55 psig under EPU conditions. Under maximum hypothetical accident conditions, the site boundary and off-site doses are below the guidelines of 10 CFR 50.67.

2.2 Current Requirements TS 6.8.4.h, Containment Leakage Rate Testing Program, specifies requirements for implementing 10CFR50.54(o) and 10 CFR 50, Appendix J Option B, in accordance with the guidelines contained in Regulatory Guide (RG) 1.163, Performance-Based Containment Leak- Test Program, dated September 1995. Approved deviations and exemptions include Type A testing in accordance with Bechtel Topical Report BN-TOP-1 and testing airlock door seals by vacuum test in lieu of pressure test. RG 1.163 allowed for an Integrated Leak Rate Test (ILRT or Type A test) frequency of once every 15 years and extended intervals for Local Leak Rate Testing (LLRT) of containment isolation valves (Type C test) of up to 60 months.

2.3 Description of the Proposed Change The proposed license amendments revise TS 6.8.4.h to require establishment of the Containment Leakage Rate Testing Program in accordance with NEI 94-01, Revision 3-A, Industry Guidance for Implementing Performance Based Option of 10 CFR 50, Appendix J, and the conditions and limitations specified in NEI 94-01, Revision 2-A, in lieu of RG 1.163. The current exemption to allow vacuum testing of airlock door seals, which was approved by Amendment Nos. 73 and 67 (Reference 6.21), would be preserved under the proposed change. The proposed change is as follows:

TS 6.8.4, ADMINSTRATIVE CONTROLS, PROCEDURES AND PROGRAMS (Continued)

h. Containment Leakage Rate Testing Program A program shall be established to implement the leakage rate testing of the containment as required by 10 CFR 50.54(o) and 10 CFR 50, Appendix J, Option B, and as modified by approved exemptions. This program shall be in accordance with the guidelines contained in Regulatory Guide 1.163, Performance-Based Containment Leak-Test Program, dated September 1995, as modified by Nuclear Energy Institute (NEI) 94-01, Revision 3-A, Industry Guidance for Implementing Performance Based Option of 10 CFR 50 Appendix J, and the conditions and limitations specified in NEI 94-01, Revision 2-A, with the following deviations or exemptions:

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 1 Page 5 of 43

1) Type A tests will be performed either in accordance with Bechtel Topical Report BN-TOP-1, Revision 1, dated November 1, 1972, or the guidelines of Regulatory Guide 1.163.

2) Type A testing frequency in accordance with NEI 94-01, Revision 0, Section 9.2.3, except:

a) For Unit 3, the first Type A test performed after the November 1992 Type A test shall be performed no later than November 2007.

b) For Unit 4, the first Type A test performed after October 1991 shall be performed no later than October 2006.

3)1) A vacuum test will be performed in lieu of a pressure test for airlock door seals at the required intervals (Amendment Nos. 73 and 77, issued by NRC November 11, 1981).

The purpose of NEI 94-01, Revision 3-A guidance is to assist licensees in the implementation of Option B to 10 CFR 50, Appendix J, Leakage Rate Testing of Containment of Light Water Cooled Nuclear Power Plants, (hereafter referred to as Appendix J, Option B). Revision 2-A of NEI 94-01 (Reference 6.2) added guidance for extending containment ILRT surveillance intervals beyond ten years, and Revision 3-A of NEI 94-01 (Reference 6.1) adds guidance for extending LLRT Type C test surveillance intervals beyond 60 months. NEI 94-01, Revision 3-A incorporates, by reference, the provisions of ANSI/ANS-56.8-2002, Containment System Leakage Testing Requirements (Reference 6.5).

The technical basis for the proposed license amendments utilizes risk-informed analysis augmented with non-risk related considerations. FPL conducted a risk impact evaluation (Enclosure 3) which concluded that the increases in large early release frequency (LERF) are within the limits set forth by the applicable guidance contained in RG 1.174, An Approach for Using Probabilistic Risk Assessment in Risk Informed Decisions on Plant Specific Changes to the Licensing Basis (Reference 6.9), NUREG-1493, Performance Based Containment Leak-Test Program (Reference 6.10), and EPRI Report No. 1009325 (Reference 6.4).

2.4 Reason for the Proposed Change The reason for the proposed change is to reduce the number of ILRTs and Type C LLRTs performed over the licensed period of operation resulting in significant savings in radiation exposure to personnel, and critical path time during refueling outages.

Under the current TS requirements, the interval for ILRT is no longer than 10 years. The next ILRT for Unit 3 is required to be performed during the fall 2021 refueling outage. The ILRT for Unit 4 is required to be performed during the spring 2022 refueling outage. With the proposed change, FPL would no longer perform the ILRTs in 2021 and 2022, but would extend the maximum surveillance interval for the ILRTs to no longer than 15 years from the last ILRT based on satisfactory performance history. The proposed change would require performance of the next Unit 3 ILRT by no later than July 2027 and the next Unit 4 ILRT by no later than February 2028.

Under the current TS requirements, the interval for Type C LLRT testing can be extended up to 60 months based on satisfactory performance, which allows testing of eligible

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 1 Page 6 of 43 penetrations every third refueling outage. With the proposed change this interval would be extended up to 75 months with satisfactory performance, which would accommodate Type C LLRT testing every fourth refueling outage.

3.0 TECHNICAL EVALUATION

3.1 Containment Leak Rate Testing History Turkey Point has established a performance based Containment Leakage Rate Testing Program, as specified by 10 CFR Part 50, Appendix J, Option B, to ensure that leakage through the containment does not exceed allowable leakage rates specified in the technical specifications and, along with the containment inspection requirements, that the integrity of the containment structure is maintained during its service life. The program is implemented by performance of the following series of tests; Type A integrated leak rate tests (ILRT) that measure the containment system overall integrated leakage rate under conditions representing design basis loss of coolant accident containment peak pressure.

General visual inspection of the accessible interior and exterior surfaces of the containment structures and components prior to any Type A test and during two other refueling cycles before the next Type A test.

Type B tests that measure leakage across pressure-retaining or leakage-limiting boundaries of 1) containment penetrations whose design incorporates resilient seals, gaskets, sealant compounds, expansion bellows, or flexible seal assemblies; 2) seals, including door operating mechanism penetrations, which are part of the primary containment; and 3) doors and hatches with resilient seals or gaskets, except for seal-welded doors.

Type C tests that measure leakage rates from containment isolation valves which are potential gaseous leakage pathways from containment during a design-basis LOCA.

3.1.1 Integrated Leak Rate (Type A) Testing Turkey Point TS 6.8.4.h requires the performance of Type A testing in accordance with 10 CFR 50, Appendix J, Option B. Leak rate test methods have followed the guidance of Bechtel Topical Report BN-TOP-1, Testing Criteria for Integrated Leakage Rate Testing of Primary Containment Structures for Nuclear Power Plants (Reference 6.22) or ANSI/ANS-56.8. Both approaches use the Absolute Method for calculation of the containment air mass but the determination of leakage rate differs. Under BN-TOP-1, leakage rate determination has utilized the Total Time technique of data analysis. Under ANSI/ANS-56.8, data analysis has been by the Mass Point Analysis technique. Both approaches have criteria for abbreviated test periods (i.e. less than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ); BN-TOP-1 a minimum of 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and ANSI/ANS-56.8 a minimum of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . Testing is performed at peak containment accident pressure and acceptance criteria is based on the TS limits for leakage.

The preoperational containment pressure testing was performed at a pressure of 63 psig (115% of the 55 psig design pressure) and was accepted by the Atomic Energy Commission (AEC) as sufficient proof of the structural integrity of the containment.

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 1 Page 7 of 43 Prior to EPU, peak containment accident pressure, Pa, was 49.9 psig. The maximum containment leakage rate at Pa allowed by TS was 0.25% of containment air weight per day.

Following implementation of EPU, peak containment accident pressure, Pa, has increased to 55.0 psig. The maximum allowable containment leakage at Pa has changed to 0.20% of containment air weight per day.

In all cases, the acceptance criteria for Type A testing has been 1.0 La for as-found leakage rate and 0.75 La for as-left leakage rate.

Tables 3.1.1-1 and 3.1.1-2 present the results of past Type A tests.

Table 3.1.1 Turkey Point Unit 3 Type A Testing History As-Found As-Left Peak Test As-Found As-Left Acceptance Acceptance Test Accident Pressure Leakage Leakage Criteria, Criteria, Date Pressure (Note 1) Rate Rate 1.0 La 0.75 La Pa (psig) (psig) (wt %/day) (wt %/day)

(wt % /day) (wt %/day)

Nov.

49.9 51.35 0.159 0.250 0.139 0.1875 1992 Nov.

2004 49.9 51.37 0.092 0.250 0.081 0.1875 July 2012 55.0 53.13 0.128 0.200 0.124 0.150 Table 3.1.1 Turkey Point Unit 4 Type A Testing History As-Found As-Left Peak Test As-Found As-Left Acceptance Acceptance Test Accident Pressure Leakage Leakage Criteria, Criteria, Date Pressure (Note 1) Rate Rate 1.0 La 0.75 La Pa (psig) (psig) (wt %/day) (wt %/day)

(wt % /day) (wt %/day)

Oct.

49.9 51.43 0.093 0.250 0.057 0.1875 1991 May 2005 49.9 51.45 0.111 0.250 0.110 0.1875 Feb.

2013 55.0 54.04 0.081 0.200 0.068 0.150 (1) Test pressure at the end of the Type A test.

The tables above show that Type A test results were well within the TS limits of 0.75 La. Furthermore, whereas NEI 94-01, Revision 3-A Section 9.2.3, defines acceptable performance history as successful completion of two consecutive periodic Type A tests where the calculated performance leakage rate was less than 1.0 La, these results meet the criteria for the extension of Type A test intervals to 15 years based on acceptable performance history.

3.1.2 Local Leak Rate Testing - Type B & C The Turkey Point containment leakage rate testing program requires testing of electrical penetrations, airlocks, hatches, flanges, and containment isolation valves

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 1 Page 8 of 43 in accordance with 10 CFR 50, Appendix J, Option B and RG 1.163. In accordance with Turkey Point TS 6.8.4.h, the combined as-left leakage rate determined on a maximum pathway leakage rate (MXPLR) basis for all penetrations shall be less than 0.60 La or 166,355 standard cubic centimeters per minute (sccm). This criterion shall be met prior to entering a mode where containment integrity is required following a refueling outage or following a shutdown that included Type B or Type C testing. Similarly, the as-found leakage rate determined on a minimum pathway leakage rate (MNPLR) basis shall be less than 0.60 La or 166,355 sccm at all times when containment integrity is required.

As discussed in NUREG-1493 and NEI 94-01, Revision 3-A, Type B and C tests can identify the vast majority of all containment leakage paths. Type B and Type C testing will continue to provide a high degree of assurance that containment integrity is maintained.

Review of the Combined Type B & C test results from 2009 through 2019 (see Tables 3.1.2-1 and 3.1.2-2 below) shows generally very good performance of the penetrations.

The as-found minimum pathway leak rate for Turkey Point Unit 3 shows an average of 10.8% of 0.6 La with a high of 25.0%.

The as-left maximum pathway leak rate for Turkey Point Unit 3 shows an average of 19.9% of 0.6 La with a high of 25.7%.

The as-found minimum pathway leak rate for Turkey Point Unit 4 shows an average of 10.7% of 0.6 La with a high of 20.8%.

The as-left maximum pathway leak rate for Turkey Point Unit 4 shows an average of 23.0% of 0.6 La with a high of 29.2%.

With one exception, the testing in each refueling outage demonstrated substantial margin between the as-found and as-left summations and the TS leakage rate acceptance criteria of 0.60 La. The one exception was in refuel outage PT4-29 where the as-found combined leakage rate was 250,000 sccm or 150% of the acceptance criteria. The majority (94%) of this leakage was through a single check valve (CK-4-312C), which is the secondary barrier for the CVCS Charging system penetration. The closed system outside containment (CSOC) is the primary barrier supporting containment integrity for CVCS Charging. With consideration of the intact CSOC as the primary barrier, the combined minimum pathway leakage was determined to be well within acceptance limits and thus containment integrity was not compromised. Check valve CK-4-312C had been retained in the Containment Leakage Rate Testing Program for defense in depth purposes, though was not required, and has subsequently been removed from the program.

Table 3.1.2-1 provides LLRT combined leakage summaries for Unit 3 outages from 2009 to 2019.

Table 3.1.2 Turkey Point Unit 3 Type B & C Combined Leakage As-Found Percent of As-Left Percent of Refuel Outage/ Minimum Acceptance Maximum Acceptance Year Pathway Criteria, Pathway Criteria, (sccm) 0.6 La (sccm) 0.6 La

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 1 Page 9 of 43 PT3-24 / 2009 26,257 15.8% 29,032 17.5%

PT3-25 / 2011 7,638 4.6% 23,314 14.0%

PT3-26 / 2012 9,174 5.5% 27,499 16.5%

PT3-27 / 2014 13,269 8.0% 32,498 19.5%

PT3-28 / 2015 14,957 9.0% 35,397 21.3%

PT3-29 / 2017 12,709 7.6% 41,688 25.1%

PT3-30 / 2018 41,520 25.0% 42,804 25.7%

LLRTs for Unit 3 over the past three outages found one (1) Type B penetration and three (3) Type C penetrations to exceed their administrative leakage limits. Test results were unsatisfactory for valve CV-3-855 in PT3-28, valve 3-40-204 in PT3-29, and the spent fuel pool transfer tube penetration and valve POV-3-2602 in PT3-

30. In all cases, however, the as-found minimum pathway combined leakage was acceptable and within administrative limits.

Valve CV-3-855 (PT3-28 / 2015), Nitrogen Supply to SI Accumulators outboard isolation, exhibited leakage only slightly over its administrative limit (2,300 sccm vs 2,200 sccm). The valve was overhauled and as-left test results were satisfactory (31 sccm).

Valve 3-40-204 (PT3-29 / 2017), Stop Valve for Service Air Supply to Containment, had unsatisfactory leakage (10,500 sccm vs the 8,800 sccm administrative limit) that was attributed to seat leakage. The valve was repaired and as-left test results were satisfactory (18 sccm).

The spent fuel pool transfer tube penetration (PT3-30 / 2018) had as-found leakage over the administrative limit. A faulty test fitting was suspected and repairs were made. Following reinstallation of the flange after refueling the as-left test results were satisfactory (85 sccm). Subsequent to that test, water leakage was noted at the flange. Investigation revealed that the flange bolting had not been installed properly and head pressure from the spent fuel pool was sufficient to cause distortion of the flange and allow the leakage. Additional bolts were installed in the flange and subsequent LLRT results were evaluated as acceptable.

POV-3-2602 and POV-3-2603 (PT3-30 / 2018), Containment Purge Exhaust Isolation Valves, had unsatisfactory leakage (21,000 sccm vs.

13,860 sccm acceptance limit) whose primary cause was determined to be leakage at two locations on the POV-3-2602 disc/seat interface as well as several additional locations. Hardening of the valve seat on POV 2603 was considered to be a possible contributing factor due to its age.

Valve POV 2602 was adjusted and the seat was replaced on valve POV-3-2603. As-left LLRT results were satisfactory (2,400 sccm).

Table 3.1.2-2 provides LLRT combined leakage summaries for Unit 4 outages from 2009 to 2019.

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 1 Page 10 of 43 Table 3.1.2 Turkey Point Unit 4 Type B & C Combined Leakage As-Found Percent of As-Left Percent of Minimum Acceptance Maximum Acceptance Refuel Outage/

Pathway Criteria, Pathway Criteria, Year (sccm) 0.6 La (sccm) 0.6 La PT4-25 / 2010 13,931 8.4% 39,313 23.6%

PT4-26 / 2011 16,730 10.1% 33,446 20.1%

PT4-27 / 2013 18,570 11.2% 39,295 23.6%

PT4-28 / 2014 35,121 21.1% 48,584 29.2%

PT4-29 / 2016 250,288(1) 8.8%(1) 43,535 26.2%

PT4-30 / 2017 11,467 6.9% 22,282 13.4%

PT4-31 / 2019 14,578 8.8% 48,422 29.1%

(1) An adjusted as-found leakage value of 14,570 sccm was determined based on removal of the result for CK-4-312C (see discussion on CK-4-312C below).

LLRTs for Unit 4 over the past three outages found all Type B penetrations to be satisfactory. Five (5) Type C penetrations were found to exceed their administrative leakage limits. valves CK-4-312C, CK-4-890B, CV-4-2822 and CK-4-11-003 in PT4-29 and valve CK-4-890B again in PT4-31. In all cases, however, the as-found minimum pathway combined leakage was acceptable and within administrative limits.

CK-4-312C (PT4-29 / 2016), Charging Pump to Regen Heat Exchanger Check Valve, had significant unsatisfactory leakage (235,718 sccm) the cause of which was indeterminate. The bonnet was removed to perform an assessment and the valve condition was found to be satisfactory. The valve was reassembled with no further action and was retested with satisfactory results (3,450 sccm). It was subsequently determined that this valve did not need to be included in the Containment Leakage Rate Testing Program and has since been removed (via EC 292168) as allowed by closed loops. The containment isolation features for the subject penetration rely on a closed system outside containment (CSOC) and the subject check valve inside containment as containment isolation barriers.

The closed piping system is used as the containment isolation barrier. The basis for excluding seat leakage testing on valves used in conjunction with a closed system is that there are no credible failure modes for a passive closed system boundary that would require leak testing the second isolation barrier.

CV-4-2822 (PT4-29 / 2016), Containment Sump Discharge Isolation Valve, had unsatisfactory leakage (10,587 sccm) that was attributed to seating surfaces that were in need of repair. The seating surfaces were reconditioned and the as-left leakage was satisfactory (1,650 sccm).

CK-4-890B (PT4-29 / 2016 & PT4-31 / 2019), Containment Spray Pump B Discharge Check Valve, had unsatisfactory leakage (60,000 sccm) in PT4-

29. An engineering inspection of the valve internals found all parts to be in satisfactory condition with respect to mating surfaces. However, a

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 1 Page 11 of 43 significant accumulation of rust deposits was found around the valve seats. The inspection concluded that rust due to upstream piping was the likely cause of the unsatisfactory leakage. The parts were cleaned and reassembled and the as-left test results were satisfactory (4,300 sccm).

Again, in PT4-31 this valve had unsatisfactory leakage (40,000 sccm) and in response, a corrective overhaul that was performed on the valve.

Subsequent as-left testing was satisfactory (16,000 sccm).

CK-4-11-003 (PT4-29 / 2016), Containment Air Sample Return Check Valve, had unsatisfactory leakage (121,000 sccm) that was attributed to debris on the disc. The valve was cleaned but did not require lapping of the seating surfaces. It was retested with satisfactory results (160 sccm).

3.1.3 Extension of Type B and C Intervals In accordance with NEI 94-01, Revision 0, Section 11.3.1, prior to determining and implementing extended test intervals for Turkey Point Type B and Type C components, an assessment of the plants containment penetrations and valve performance shall be performed and documented. Consistent with NEI 94-01, the following factors were considered:

Past Component Performance - Specific component performance of two successful consecutive as-found tests.

Service - The environment and particular types of components that influence likelihood of failure based on their performance history.

Design - Valve type and penetration design attributes that contribute to leakage characteristics.

Safety Impact - The relative importance of a penetration in terms of the potential impact of a failure in limiting releases from containment under accident conditions.

Cause Determination - Cause determinations performed for failures identified during an extended test interval should include corrective actions that identify and address common-mode failure mechanisms.

Type B and Type C components found to be acceptable for extended intervals based on this assessment may have their test interval extended to 120 and 60 months, respectively, based on the completion of two consecutive as-found tests where the results of each test are within the allowable administrative limit.

Table 3.1.3-1 presents a tally of Type B and C penetrations for each unit and indicates the number of penetrations on extended intervals in relation to the number that are eligible for extension to 60 months.

Table 3.1.3 Summary of Extended Test Frequencies Number of Penetrations Test Eligible for Pens. on Unit Elect. Mech. Total Type Extended Extended Pens. Pens. Pens.

Interval Interval B U3 48 11 59 57 54

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 1 Page 12 of 43 U4 47 13 60 58 55 U3 -- 28 28 26 17 C

U4 -- 28 28 26 16 The percentages of eligible penetrations on extended intervals are:

Type B Tests: Unit 3 .............. 54/57 = 95%

Unit 4 .............. 55/58 = 95%

Type C Tests: Unit 3 .............. 17/26 = 65%

Unit 4 .............. 16/26 = 62%

Unit 3 Type B Penetrations:

Table 3.1.3-2 identifies the Type B penetrations for Unit 3 and their test intervals.

Table 3.1.3 Unit 3 Type B Penetrations - Test Intervals Eligible for Currently on Pen Current Test Service Extended Extended No. Freq.(2)

Interval (1) Interval (1) 17 Safety Injection Test & Purge Yes Yes 3R 37 Spare Yes Yes 3R 38 Electrical Pens (42 canisters) Yes Yes 3R 39 Fuel Transfer Tube Yes No (1b) 1R 40 Equipment Hatch Yes No (1b) 1R 41 Personnel Airlock No (1a) No (1a) 1R 48 Electrical Pens (6 canisters) Yes Yes 3R 49 Emergency Escape Airlock No (1a) No (1a) 1R 61B Dead Weight Transfer Yes Yes 3R 62A Cont. Pressure Sensors Yes Yes 3R 62B Cont. Pressure Sensors Yes Yes 3R 62C Cont. Pressure Sensors Yes Yes 3R 65A ILRT Service Penetration Yes No (1b) 1R (1) Interval notes:

(a) Penetration is an airlock and thus not eligible per NEI 94-01 (b) Penetration is opened each refueling outage and thus requires LLRT (2) Test frequency:

1R - 1 refueling cycle (~18 months) 3R - 3 refueling cycles (< 60 months)

The total population of Unit 3 penetrations requiring Type B testing is 59. Two of the 59 penetrations are containment airlocks (Pen. Nos. 41 and 49), which per NEI 94-01, Section 10.2.2, are limited to intervals of 30 months, thus these are considered not eligible for extended test interval. Of the 57 remaining (i.e., eligible) penetrations, 48 are electrical and nine (9) are mechanical.

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 1 Page 13 of 43 The 48 electrical penetrations (Pen. No. 38 with 42 canisters and Pen. No. 48 with six (6) canisters) are all on extended intervals.

Of the nine (9) eligible mechanical penetrations, six (6) are on extended intervals.

The other three (3) are the fuel transfer tube (Pen. No. 39), equipment hatch (Pen.

No. 40), and a service penetration (Pen. No. 65A), which are opened each outage and thus remain on a one-cycle interval.

Thus, there are no Type B penetrations on base interval due to performance. (Note that the fuel transfer tube, Penetration 39, which is on a 1R interval because it is opened each outage as noted above, is also restricted because leakage exceeded administrative limits in the PT3-30 LLRT.)

Table 3.1.3-3 presents the most recent as-found test results for the Unit 3 Type B penetrations in comparison to their administrative limits. The administrative limits were assigned by FPL, based on individual penetration performance history, as an indicator of penetration degradation. The exception is where a TS limit is established for a particular component.

Table 3.1.3 Unit 3 Type B Penetrations - Most Recent Testing Admin. Most Recent Test Previous Test Pen Valves Limit Refueling Leakage Refueling Leakage No.

(sccm) Outage (sccm) Outage (sccm) 17 3-895V 2,200 PT3-29 18 PT3-28 18 37 3-2023 440 PT3-29 90 PT3-28 150 38 Canisters 18,480 PT3-28 358 PT3-26 283 39 O-ring seals 1,100 PT3-30 [2,200] PT3-29 210 40 O-ring seals 13,860 PT3-30 18 PT3-29 18 3S8 13,860 PT3-30 9,000 PT3-29 5,500 41 Equalizing valve 2,000 PT3-30 800 PT3-29 640 48 Canisters 2,640 PT3-28 21 PT3-26 58 3S9 13,860 PT3-30 2,000 PT3-29 480 49 Equalizing valve 2,000 PT3-30 18 PT3-29 823 61B 3-2024 4,400 PT3-29 18 PT3-28 22 PS-3-2008 62A 440 PT3-30 18 PT3-29 110 PS-3-2057 PS-3-2009 62B 440 PT3-30 18 PT3-29 30 PS-3-2058 PS-3-2007 62C 440 PT3-30 18 PT3-29 18 PS-3-2056 65A Flanges 8,800 PT3-30 18 PT3-29 18 Penetration No. 39, Fuel Transfer Tube, is the only recent test that exceeded administrative limits. Section 3.1.2 provides a discussion of the circumstances of that leakage.

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 1 Page 14 of 43 Unit 4 Type B Penetrations:

Table 3.1.3-4 identifies the Type B penetrations for Unit 4 and their test intervals.

Table 3.1.3 Unit 4 Type B Penetrations - Test Intervals Eligible for Currently on Pen Current Test Service Extended Extended No. Freq.(2)

Interval (1) Interval (1) 17 Safety Injection Test & Purge Yes Yes 3R 37 Spare Yes Yes 3R 38 Electrical Pens (41 canisters) Yes Yes 3R (1b) 39 Fuel Transfer Tube Yes No 1R 40 Equipment Hatch Yes No (1b) 1R (1a) (1a) 41 Personnel Airlock No No 1R 46A Cont. Pressure Sensors Yes Yes 3R 46B Cont. Pressure Sensors Yes Yes 3R 46C Cont. Pressure Sensors Yes Yes 3R 48 Electrical Pens (6 canisters) Yes Yes 3R (1a) (1a) 49 Emergency Escape Airlock No No 1R 56 Spare Yes Yes 3R 61A Spare Yes Yes 3R 61B Dead Weight Transfer Yes Yes 3R (1b) 65A ILRT Service Penetration Yes No 1R (1) Interval notes:

(a) Penetration is an airlock and thus not eligible per NEI 94-01 (b) Penetration is opened each refueling outage and thus requires LLRT (2) Test frequency:

1R - 1 refueling cycle (~18 months) 3R - 3 refueling cycles (< 60 months)

The total population of Unit 4 penetrations requiring Type B testing is 60. Two of the 60 penetrations are containment airlocks (Pen. Nos. 41 and 49), which per NEI 94-01, Section 10.2.2, are limited to intervals of 30 months; thus these are considered not eligible for extended test interval. Of the 58 remaining (i.e., eligible) penetrations, 47 are electrical and 11 are mechanical.

The 47 electrical penetrations (Pen. No. 38 with 41 canisters and Pen. No. 48 with six (6) canisters are all on extended intervals.

Of the 11 eligible mechanical penetrations, eight (8) are on extended intervals. The other three (3) are the fuel transfer tube (Pen. No. 39), equipment hatch (Pen. No.

40), and a service penetration (Pen. No. 65A), which are opened each outage and thus remain on a one-cycle interval.

Thus there are no Type B penetrations on base interval due to performance.

Table 3.1.3-5 presents the most as-found recent test results for the Unit 4 Type B penetrations in comparison to their administrative limits.

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 1 Page 15 of 43 Table 3.1.3 Unit 4 Type B Penetrations - Most Recent Testing Admin. Most Recent Test Previous Test Pen Valves Limit Refueling Leakage Refueling Leakage No.

(sccm) Outage (sccm) Outage (sccm) 17 4-895V 2,200 PT4-30 25 PT4-27 18 4-10-879 440 PT4-30 15 PT4-27 18 37 Cap 440 PT4-30 14 PT4-27 18 38 Canisters 18,480 PT4-29 375 PT4-27 292 39 O-ring seals 1,100 PT4-31 18 PT4-30 11 40 O-ring seals 13,860 PT4-31 18 PT4-30 0 4S8 13,860 PT4-31 5,400 PT4-30 920 41 Equalizing valve 2,000 PT4-31 18 PT4-30 15 PS-4-2008 46A 440 PT4-30 0 PT4-29 18 PS-4-2057 PS-4-2009 46B 440 PT4-30 10 PT4-29 18 PS-4-2058 PS-4-2007 46C 440 PT4-30 5 PT4-29 18 PS-4-2056 48 Canisters 2,640 PT4-29 98 PT4-28 10 4S9 13,860 PT4-31 720 PT4-30 180 49 Equalizing valve 2,000 PT4-31 225 PT4-30 38 56 Spare 440 PT4-29 10 PT4-27 18 61A Spare 440 PT4-29 18 PT4-27 18 61B 4-2024 4,400 PT4-29 18 PT4-27 18 65A Flanges 8,800 PT4-31 87 PT4-30 18 All recent tests of Unit 4 Type B penetrations were within the administrative limits.

Unit 3 Type C Penetrations:

Table 3.1.3-6 identifies the Type C penetrations for Unit 3 and their test intervals.

Table 3.1.3 Unit 3 Type C Penetrations - Test Intervals Eligible for Currently on Pen Current Test Service Extended Extended No. Freq.(2)

Interval (1) Interval (1) 5 PRT Gas Analyzer Sample Yes Yes 3R (1b) 6 PRT Nitrogen Supply Yes No 1R 7 Primary Water to PRT & RCP Yes Yes 3R 8 Pressurizer Steam Sample Yes Yes 3R 9 Pressurizer Liquid Sample Yes Yes 3R 10 RCDT Vent & Nitrogen Supply Yes Yes 3R 14 CVCS Normal Letdown Yes Yes 3R 16 Post-Accident Cont. Air Sampling Yes Yes 3R

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 1 Page 16 of 43 19A Containment Spray (Header A) Yes No 1R 19B Containment Spray (Header B) Yes No 1R 20 RCS Hot Leg Sample Yes Yes 3R 23 Containment Sump Discharge Yes No 1R 25 RCP Seal Return Yes Yes 3R 29 Instrument Air Supply Yes No (1b) 1R (1c) 31 RCDT Gas Analyzer Sample Yes No 1R (1b) 32 Cont. Air Sample Return Yes No 1R 33 Containment Air Sample Yes Yes 3R 34 Service Air Supply Yes No 1R (1a) (1a) 35 Containment Purge Supply No No 9 mo.(3) 36 Containment Purge Exhaust No (1a) No (1a) 9 mo. (3) 42 Nitrogen Supply to SI Accumulators Yes No (1b) 1R 47 Primary Water to Wash Header Yes Yes 2R 52 RCDT Pump Discharge Yes Yes 3R 53 Post-Accident Cont. Air Sampling Yes Yes 3R 55 Accumulator Sample Yes Yes 3R 63 Instrument Air Bleed Yes Yes 3R 65B ILRT Service Penetration Yes Yes 3R 65C ILRT Service Penetration Yes Yes 3R (1) Interval notes:

(a) Barriers are purge valves which are limited by NEI 94-01.

(b) Penetration is tested each refueling outage due to IST requirements for check valves.

(c) Interval set to 1R following valve replacement or maintenance and until satisfactory test history is established.

(2) Test frequency:

1R - 1 refueling cycle (~18 months) 2R - 2 refueling cycles 3R - 3 refueling cycles (< 60 months)

(3) Frequency for the Containment Purge valves is currently set at 9 months in accordance with the Surveillance Frequency Control Program.

The total population of Unit 3 penetrations requiring Type C testing is 28. Two of the 28 are purge valves (Pen. Nos. 35 and 36), which per NEI 94-01 Section 10.2 are limited to intervals of 30 months, thus these are considered not eligible. Of the 26 remaining (i.e., eligible) penetrations, 17 are on extended test intervals, most at 60 months.

Of the nine (9) eligible penetrations not on extended intervals, only one (1) has recent LLRT performance issues (Pen. No. 34). Four penetrations (Pen. Nos. 6, 29, 32 and 42) remain on one-cycle intervals due to ASME Section XI inservice testing (IST) requirements for the associated check valves. One other penetration (Pen. No. 31) was reset to one-cycle due to recent maintenance that could affect leak-tightness and will remain there until acceptable test history is established. The remaining three (3) penetrations are kept on one-cycle test intervals as a conservative measure and may be put on extended frequencies in the future.

For the Containment Purge Supply and Exhaust penetrations (Pen. Nos. 35 & 36) the surveillance test interval is currently set to 9 months, in accordance with the

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 1 Page 17 of 43 Surveillance Frequency Control Program requirements. This frequency exceeds the Appendix J program requirements of 30 months. These penetrations are 48 inch and 54-inch ventilation piping, respectively, with inboard and outboard butterfly valves. These normally-closed valves are air actuated to open, spring closed. The valves close automatically on Containment Ventilation Isolation signal and fail closed on loss of electrical power or pneumatic supply. Under power, these valves may be open for purge system operation for pressure control, for environmental conditions control, for ALARA considerations, and respirable air quality considerations. Other than for these reasons the valves shall be sealed closed to the maximum extent practicable or the penetrations isolated. The valves are to be demonstrated operable by verifying that measured leakage rate is less than or equal to 0.05 La. These penetrations are currently on a 9-month leakage test interval.

The outboard valves in particular have experienced corrosion due to their location outside the Containment on the Auxiliary Building roof. Following a 2008 short notice outage during which containment purge was placed in service, the purge supply penetration exceeded leakage limits during LLRT testing because the disc on outboard isolation valve POV-3-2600 was not fully closed. Corrosion in the actuator was suspected to have prevented the actuator from delivering the seating torque needed to provide an adequate seal. Similarly, the purge exhaust penetration exceeded leakage limits in a 2018 LLRT because outboard isolation valve POV-3-2602 did not seat properly. Both of these conditions were entered into the Turkey Point Corrective Action Program.

Mid-cycle LLRT testing on purge exhaust Penetration 36 identified leakage exceeding the established limits. Investigation found leakage through one of the taper pin assemblies that that hold the disc to the valve shaft on inboard isolation valve POV-3-2603. Installation of a washer and gasket under the taper pin nut corrected the leakage. Combined leakage remained within program limits. This condition was entered into the Turkey Point Corrective Action Program.

It should be noted that in Reference 6.23, FPL requested changes to the Turkey Point Technical Specifications which relocates the purge valve leakage limit of 0.05 La to licensee control and requires the purge valves to be administratively sealed closed and deactivated, or the associated penetration isolated by blind flange, during MODES 1 through 4.

Table 3.1.3-7 presents the most recent as-found test results for the Unit 3 Type C penetrations in comparison to their administrative limits.

Table 3.1.3 Unit 3 Type C Penetrations - Most Recent Testing Most Recent Test Previous Test Pen Admin. Limit Valves Refueling Leakage Refueling Leakage No. (sccm)

Outage (sccm) Outage (sccm)

SV-3-6385 4,400 PT3-30 18 PT3-27 18 5

CV-3-516 4,400 PT3-30 18 PT3-27 50 CK-3-518 11,000 PT3-30 2,600 PT3-27 5,250 6

CK-3-519 11,000 PT3-30 18 PT3-29 443

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 1 Page 18 of 43 CV-3-519A CV-3-519B 7 CV-3-522A 8,800 PT3-29 380 PT3-28 18 CV-3-522B CV-3-522C CV-3-951 4,400 PT3-28 10 PT3-26 18 8

CV-3-956A 4,400 PT3-28 580 PT3-26 600 CV-3-953 4,400 PT3-28 0 PT3-27 25 9 CV-3-956B 4,400 PT3-28 75 PT3-27 60 RV-3-300 CV-3-4658A 8,800 PT3-29 440 PT3-28 500 CV-3-4658B 8,800 PT3-29 1,200 PT3-28 175 10 3-4639 8,800 PT3-29 270 PT3-28 315 3-3449 3-4656 8,800 PT3-29 4,300 PT3-28 1,400 CV-3-200A CV-3-200B 13,200 PT3-30 45 PT3-29 380 CV-3-200C 14 CV-3-204 13,200 PT3-30 18 PT3-29 18 RV-3-203 HV-3-1 16 13,200 PT3-28 18 PT3-26 18 PAHM-3-002A CK-3-890A 26,400 PT3-30 16,000 PT3-29 250 19A MOV-3-880A 22,000 PT3-30 17,000 PT3-29 260 3-883M CK-3-890B 26,400 PT3-30 18 PT3-29 35 19B MOV-3-880B 22,000 PT3-30 1,020 PT3-29 160 3-883N SV-3-6427A 4,400 PT3-29 0 PT3-28 0 SV-3-6427B 4,400 PT3-29 450 PT3-28 400 20 SV-3-6428 4,400 PT3-29 330 PT3-28 240 CV-3-2821 8,800 PT3-30 2,500 PT3-29 200 23 CV-3-2822 8,800 PT3-30 75 PT3-29 0 MOV-3-6386 8,800 PT3-29 0 PT3-27 18 25 MOV-3-381 8,800 PT3-29 19 PT3-28 18 RV-3-303 CK-3-40-340A 17,600 PT3-30 750 PT3-29 200 29 CK-3-40-336 17,600 PT3-30 5,000 PT3-29 5,100 CV-3-4659A 31 2,200 PT3-30 18 PT3-29 18 CV-3-4659B CK-3-11-003 8,800 PT3-30 990 PT3-29 213 SV-3-2912 32 PAHM-3-001A 8,800 PT3-30 2,500 PT3-29 1,800 PAHM-3-001B SV-3-2911 13,200 PT3-29 600 PT3-28 800 33 SV-3-2913 13,200 PT3-29 460 PT3-28 700 3-40-205 8,800 PT3-30 140 PT3-29 0 34 3-40-204 8,800 PT3-30 30 PT3-29 [10,500]

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 1 Page 19 of 43 POV-3-2600 35 13,860 09/26/19 185 PT3-30 480 POV-3-2601 POV-3-2602 36 13,860 09/26/19 [27,500] PT3-30 [21,000]

POV-3-2603 CK-3-945E 13,200 PT3-30 200 PT3-29 250 42 CV-3-855 2,200 PT3-30 18 PT3-29 75 3-10-582 17,600 PT3-29 2,805 PT3-28 0 47 3-10-567 17,600 PT3-29 500 PT3-28 75 RV-3-302 CV-3-4668A 8,800 PT3-29 293 PT3-28 180 52 CV-3-4668B 8,800 PT3-29 18 PT3-28 18 HV-3-3 53 13,200 PT3-28 64 PT3-26 18 PAHM-3-002B CV-3-955C 4,400 PT3-30 1,025 PT3-29 900 CV-3-955D 4,400 PT3-30 18 PT3-29 0 55 CV-3-955E 4,400 PT3-30 18 PT3-29 0 CV-3-956D 4,400 PT3-30 18 PT3-29 18 RV-3-301 CV-3-2819 8,800 PT3-29 127 PT3-28 0 63 CV-3-2826 8,800 PT3-29 40 PT3-28 20 65B 3-2025 8,800 PT3-28 18 PT3-26 18 65C 3-2026 8,800 PT3-28 32 PT3-26 18 Recent Type C tests exceeding administrative limits are limited to Penetration No.

34, service air supply valve 3-40-204, in PT3-29, and Penetration No. 36, Containment Purge Exhaust.

The previous test history on Penetration 34 valve 3-40-204 had been good with low leakage rates allowing the penetration to be placed on extended interval. The PT3-29 test revealed some degradation in the valve seat but repair restored the valve to a low leakage rate. The penetration has been placed back on a one-cycle interval until satisfactory test history is reestablished.

Penetration No. 36 includes the Containment Purge Exhaust butterfly valves. The issues surrounding these valves are discussed in the paragraphs preceding Table 3.1.3-7.

Unit 4 Type C Penetrations:

Table 3.1.3-8 identifies the Type C penetrations for Unit 4 and their test intervals.

Table 3.1.3 Unit 4 Type C Penetrations - Test Intervals Eligible for Currently on Pen Current Test Service Extended Extended No. Freq.(2)

Interval (1) Interval (1) 5 PRT Gas Analyzer Sample Yes Yes 3R (1b) 6 PRT Nitrogen Supply Yes No 1R

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 1 Page 20 of 43 7 Primary Water to PRT & RCP Yes Yes 3R 8 Pressurizer Steam Sample Yes Yes 3R 9 Pressurizer Liquid Sample Yes Yes 3R 10 RCDT Vent & Nitrogen Supply Yes Yes 3R 14 CVCS Normal Letdown Yes Yes 3R 16 Post-Accident Cont. Air Sampling Yes Yes 3R 19A Containment Spray (Header A) Yes No 1R 19B Containment Spray (Header B) Yes No 1R 20 RCS Hot Leg Sample Yes No (1c) 1R (1c) 23 Containment Sump Discharge Yes No 1R 25 RCP Seal Return Yes Yes 3R (1b) 29 Instrument Air Supply Yes No 1R 31 RCDT Gas Analyzer Sample Yes Yes 3R 32 Cont. Air Sample Return Yes No (1b) 1R (1c) 33 Containment Air Sample Yes No 1R 34 Service Air Supply Yes Yes 3R 35 Containment Purge Supply No (a) No (1a) 9 mo.(3) 36 Containment Purge Exhaust No (a) No (1a) 9 mo. (3) 42 Nitrogen Supply to SI Accumulators Yes No (1b) 1R 47 Primary Water to Wash Header Yes Yes 2R 51 Post-Accident Cont. Air Sampling Yes Yes 3R 52 RCDT Pump Discharge Yes Yes 3R 55 Accumulator Sample Yes Yes 3R 63 Instrument Air Bleed Yes Yes 3R 65B ILRT Service Penetration Yes Yes 3R 65C ILRT Service Penetration Yes Yes 3R (1) Interval notes:

(a) Barriers are purge valves which are limited by NEI 94-01.

(b) Penetration is tested each refueling outage due to IST requirements for check valves.

(c) Interval set to 1R following valve replacement or maintenance and until satisfactory test history is established.

(2) Test frequency:

1R - 1 refueling cycle (~18 months) 2R - 2 refueling cycles 3R - 3 refueling cycles (< 60 months)

(3) Frequency for the Containment Purge valves is currently set at 9 months in accordance with the Surveillance Frequency Control Program.

The total population of Unit 4 penetrations requiring Type C testing is 28. Two of the 28 are purge valves (Pen. Nos. 35 and 36), which per NEI 94-01 Section 10.2 are limited to intervals of 30 months, thus these are considered not eligible. Of the 26 remaining (i.e., eligible) penetrations, 17 are on extended test intervals, most at 60 months.

Of the nine (9) eligible penetrations not on extended intervals, only one has recent LLRT performance issues (Pen. No. 19B). Four penetrations (Pen. Nos. 6, 29, 32 and 42) remain on one- cycle intervals due to IST requirements for the associated

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 1 Page 21 of 43 check valves. Three other penetrations (Pen. Nos. 20, 23 and 33) were reset to one-cycle intervals due to recent valve replacement or maintenance and will remain there until acceptable test history is established. One penetration (Pen. No.

19A) is on a one-cycle interval as a conservative measure.

As noted for Unit 3, the Containment Purge Supply and Exhaust penetrations are currently on 9-month test intervals under the Surveillance Frequency Control Program. TS 3.6.1.7 requires the valves to be sealed closed to the maximum extent practicable and TS 4.6.1.7.2 limits leakage rate to 0.05 La.

Similar to Unit 3, the outboard valves have experienced some corrosion. The purge exhaust penetration exceeded leakage limits in a 2006 LLRT because the disc on outboard isolation valve POV-4-2602 was not fully closed. Corrosion in the actuator bearing surfaces created friction that exceeded the actuator closing capability. The condition was entered into the Turkey Point Corrective Action Program.

It should be noted that in Reference 6.23, FPL requested changes to the Turkey Point Technical Specifications which relocates the purge valve leakage limit of 0.05 La to licensee control and requires the purge valves to be administratively sealed closed and deactivated, or the associated penetration isolated by blind flange, during MODES 1 through 4.

Table 3.1.3-9 presents the most recent as-found test results for the Unit 4 Type C penetrations in comparison to their administrative limits.

Table 3.1.3 Unit 4 Type C Penetrations - Most Recent Testing Most Recent Test Previous Test Pen Admin. Limit Valves Refueling Leakage Refueling Leakage No. (sccm)

Outage (sccm) Outage (sccm)

SV-4-6385 4,400 PT4-29 18 PT4-27 18 5

CV-4-516 4,400 PT4-29 68 PT4-27 75 CK-4-518 11,000 PT4-31 480 PT4-30 550 6

CK-4-519 11,000 PT4-31 470 PT4-30 755 CV-4-519A CV-4-519B CV-4-522A 8,800 PT4-30 18 PT4-29 170 7 CV-4-522B CV-4-522C CV-4-951 4,400 PT4-30 0 PT4-29 460 8

CV-4-956A 4,400 PT4-30 105 PT4-29 115 CV-4-953 4,400 PT4-31 18 PT4-29 0 9 CV-4-956B 4,400 PT4-31 18 PT4-29 18 RV-4-300 CV-4-4658A 8,800 PT4-30 18 PT4-27 35 CV-4-4658B 8,800 PT4-30 18 PT4-27 18 4-4639 10 8,800 PT4-30 18 PT4-27 18 4-3449 4-4656 8,800 PT4-30 18 PT4-27 35

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 1 Page 22 of 43 CV-4-200A CV-4-200B 13,200 PT4-31 3,200 PT4-30 2,010 14 CV-4-200C CV-4-204 13,200 PT4-31 105 PT4-30 750 RV-4-203 HV-4-1 16 13,200 PT4-30 0 PT4-27 18 PAHM-4-002A CK-4-890A 26,400 PT4-31 145 PT4-30 18 19A MOV-4-880A 22,000 PT4-31 18 PT4-30 40 4-883M CK-4-890B 26,400 PT4-31 [40,000] PT4-30 4,600 19B MOV-4-880B 22,000 PT4-31 750 PT4-30 1,500 4-883N SV-4-6427A 4,400 PT4-31 18 PT4-30 0 SV-4-6427B 4,400 PT4-31 18 PT4-30 0 20 SV-4-6428 4,400 PT4-31 1,300 PT4-30 355 CV-4-2821 8,800 PT4-31 2,800 PT4-30 3,500 23 CV-4-2822 8,800 PT4-31 2,900 PT4-30 30 MOV-4-6386 8,800 PT4-30 155 PT4-29 90 25 MOV-4-381 8,800 PT4-30 18 PT4-29 18 RV-4-303 CK-4-40-340A 17,600 PT4-31 250 PT4-30 355 29 CK-4-40-336 17,600 PT4-31 1,200 PT4-30 18 CV-4-4659A 31 2,200 PT4-29 18 PT4-27 25 CV-4-4659B CK-4-11-003 8,800 PT4-31 1,240 PT4-30 320 SV-4-2912 32 PAHM-4-001A 8,800 PT4-31 1,300 PT4-30 350 PAHM-4-001B SV-4-2911 13,200 PT4-31 1,500 PT4-30 650 33 SV-4-2913 13,200 PT4-31 1,700 PT4-30 420 4-40-205 8,800 PT4-31 1,400 PT4-28 1,800 34 4-40-204 8,800 PT4-31 18 PT4-28 18 POV-4-2600 35 13,860 PT4-31 5,300 08/16/18 3,200 POV-4-2601 POV-4-2602 36 13,860 PT4-31 140 08/16/18 600 POV-4-2603 CK-3-945E 13,200 PT4-31 250 PT4-30 795 42 CV-3-855 2,200 PT4-31 140 PT4-30 485 4-10-582 17,600 PT4-30 20 PT4-29 60 47 4-10-567 17,600 PT4-30 22 PT4-29 18 RV-4-302 HV-4-3 51 13,200 PT4-30 20 PT4-27 18 PAHM-4-002B CV-4-4668A 52 8,800 PT4-31 18 PT4-28 18 CV-4-4668B CV-4-955C 4,400 PT4-30 5 PT4-29 101 CV-4-955D 4,400 PT4-30 700 PT4-29 680

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 1 Page 23 of 43 55 CV-4-955E 4,400 PT4-30 640 PT4-29 560 CV-4-956D 4,400 PT4-30 18 PT4-29 18 RV-4-301 CV-4-2819 8,800 PT4-31 400 PT4-30 275 63 CV-4-2826 8,800 PT4-31 750 PT4-30 3,600 65B 4-2025 8,800 PT4-29 18 PT4-27 18 65C 4-2026 8,800 PT4-29 18 PT4-27 110 Penetration No. 19B, Containment Spray Header B, is the only recent test that exceeded administrative limits. As discussed in Section 3.1.2, valve CK-4-890B had leakage issues in two of its last three most recent tests. In PT4-31, a corrective overhaul was performed. Penetration No. 19B is currently on a one-cycle interval.

3.2 Containment Inspections Other containment inspections that are performed to fulfill the requirements of 10 CFR 50.55a and 10 CFR 50 Appendix J such as the ASME Section XI, Subsection IWE Program (IWE Program), the ASME Section XI, Subsection IWL Program (IWL Program) and the Containment Leakage Rate Testing Program Containment Building Visual Inspections will provide continuing supplemental means of identifying potential containment degradation that may affect leak tightness. These inspections examine the containment system for signs of degradation, damage, and other irregularities and provide assurance that degradation of the containment is detected and corrected.

3.2.1 IWE Program The IWE Program provides periodic inspection and examination of the containment liner and its integral attachments. The program is conducted in accordance with ASME Boiler and Pressure Vessel Code,Section XI, Subsection IWE subject to the limitations and modifications in 10 CFR 50.55a(b)(2), 10 CFR 50.55a(g)(4), and 10 CFR 50.55a(g)(6).

Under the IWE program, surfaces are examined for evidence of cracking, discoloration, wear, pitting, excessive corrosion, arc strikes, gouges, surface discontinuities, dents, or other signs of surface irregularities. Pressure-retaining bolting is examined for loosening and material conditions that affect either containment leak-tightness or structural integrity. Moisture barriers are visually inspected for degradation.

Coated surfaces are visually inspected, without removal of coating, for evidence of conditions that indicate degradation of the underlying base metal. These include flaking, blistering, peeling, discoloration, and other signs of distress. The coatings themselves are not the subject of the inspection thus coincidental coating failures are treated as a maintenance item if the substrate (primer) is intact and the underlying base metal is not degraded.

Acceptability of inaccessible areas of the liner is evaluated when conditions found in accessible areas indicate the presence of, or could result in, flaws or degradation in inaccessible areas.

Table 3.2.1-1 presents the IWE Inspection Intervals to date. Turkey Point is currently in the Third 10-Year Inspection Interval. This interval will be conducted in

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 1 Page 24 of 43 accordance with the 2007 Edition of ASME Section XI with Addenda through 2008 as modified by 10 CFR 50.55a.

Table 3.2.1 IWE Inspection Intervals Interval/Period Start Date End Date Sect XI Code Edition 1992 Edition First Interval 09/09/1996 09/09/2008 w/ 1992 Addenda 2001 Edition Second Interval 09/09/2008 07/14/2018 thru 2003 Addenda Third Interval 07/15/2018 07/14/2028 1st Period 07/15/2018 07/14/2021 2007 Edition thru 2008 Addenda 2nd Period 07/15/2021 07/14/2025 3rd Period 07/15/2025 07/14/2028 Components are scheduled for examination in accordance with ASME Section XI, Table IWE-2411-1 and Table IWE-2500-1. Turkey Point utilizes an IWE data management system (IDDEAL) to schedule and track the inspections for individual components. The IWE database management system includes tables that have brief descriptions of each component subject to examination, the required Code references, and any other pertinent information that is useful for determining examination requirements. Table 3.2.1-2 reflects the inspections scheduled for the Third Interval.

Table 3.2.1 IWE Examinations - Third Interval Item Exam Examinations Scheduled by Period Parts Examined Numbers Method 1 2 3 Examination Category E-A, Containment Surfaces E1.11 Accessible Surface Areas GV 100%(1) 100%(1) 100%(1)

The only submerged area is in Wetted Surfaces of E1.12 the reactor cavity sump. Exam is Submerged Areas VT-3 performed once per interval.

E1.30 Moisture Barrier GV 100% 100% 100%

Examination Category E-C, Augmented Examination E4.11 Visible Surfaces VT-1 None at this time Surface Area Grid Minimum E4.12 UT None at this time Wall Thickness Locations Examination Category E-G, Pressure-Retaining Bolting Each bolted penetration is examined once per interval. If a penetration is E8.10 Bolted Connections VT-1 disassembled during the interval the inspection is performed with the connection disassembled.

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 1 Page 25 of 43 (1) In addition, VT-1 examinations are performed for several areas of the liner designated by Engineering. Although these are VT-1 examinations, they are not augmented examinations since they are not dictated by IWE-1240. They are included by Engineering discretion due to CCW sweating that occurs in Containment during outages.

In November 2006, a small hole was found in the floor of the Unit 4 reactor cavity sump liner plate when one of the sump pumps was moved. The hole was found below a steel plate that supports the pump. Corrosion that caused the hole was attributed to a combination of boric acid and galvanic corrosion due to shim material used for the support plate. The hole was plugged and welded, the area was left with steel shims, and the steel support plate returned. The repair was leak tested successfully. Subsequent inspections found the area acceptable.

During the Unit 3 refueling outage in October 2010, a scheduled visual examination of the containment liner plate in the reactor cavity sump revealed significant corrosion in a localized region of the vertical wall section, immediately adjacent to the concrete floor. Augmented visual and ultrasonic examinations revealed an area measuring approximately 3 x 36 with wall thickness measurements less than minimum wall, including twelve through wall holes. Metallurgical investigation determined that the corrosion was caused by chemical attack from boric acid. The root cause of the event was determined to be degradation of a coating system that was not designed for immersion service. Periodic exposure of the coating system to immersion conditions in the reactor pit area caused accelerated degradation of the coating and corrosive attack of the underlying carbon steel liner plate.

Corrective actions were taken to remove the degraded portion of the liner plate and weld in replacement plate in accordance with the rules of the ASME Code. A coating system suitable for immersion service was applied. Similar but less severe conditions were found on Unit 4 and similar remediation made. The condition was entered into the Turkey Point Corrective Action Program.

Visual inspections of the moisture barrier in 2015 on Unit 3 and in 2016 on Unit 4 noted various areas of damaged or degraded sealant. The moisture barrier is a water tight seal installed between the 14 elevation concrete cover slab and the containment wall liner. At the surface the concrete abuts a toe plate rather than the liner itself. The toe plate is a steel angle that covers a weld in the liner and was used during plant construction to pressure test the weld. The gap between the concrete and the toe plate is filled with a joint sealant. The sealant was found to have physical damage, loss of adhesion and tearing. The damage was attributed to ordinary outage activities and the sealant was repaired in accordance with ASME Section XI Subsection IWE. In one instance a small hole was identified in the toe plate. The hole was enlarged to afford inspection of the normally inaccessible liner plate. The liner was found to be in satisfactory condition and the hole was repaired in accordance with approved generic repair details.

Other minor deterioration has been found in various inspections. Degraded coatings (e.g., chipped and flaking paint, exposed bare metal, discoloration) and surface corrosion have been observed. Engineering evaluations determined that these conditions did not constitute degradation of the liner.

3.2.2 IWL Program The IWL Program provides periodic inspection and examination of the concrete containment and post-tensioning system. The program is conducted in accordance with ASME Boiler and Pressure Vessel Code,Section XI, Subsection IWL

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 1 Page 26 of 43 (respectively) subject to the limitations and modifications in 10 CFR 50.55a(b)(2),

10 CFR 50.55a(g)(4), 10 CFR 50.55a(g)(6) and TS 3.6.1.6.

The containment tendon examination program has been conducted since initial startup of both Units 3 and 4 on 5-year intervals. The initial containment tendon surveillance examination requirements incorporated the general criteria and requirements of NRC RG 1.35, Inservice Inspection of Ungrouted Tendons in Prestressed Concrete Containments. The program transitioned to the requirements of ASME Code Section XI, Subsection IWL, due to the incorporation of the code into 10 CFR 50.55a.

Under the IWL program, accessible concrete surfaces are subject to visual inspections to detect deterioration and distress. Tendon wires and tendon anchorage hardware surfaces are inspected for loss of material, cracking, and mechanical damage. The tendon corrosion protection medium is tested for the pH, presence of free water, and soluble ion concentration. Individual wires are removed from selected sample tendons and tested for yield strength, ultimate tensile strength, and elongation Table 3.2.1-2 presents the IWL Inspection Intervals to date. Turkey Point has recently completed the Second 10-Year Inspection Interval. This interval included the 40th Year Surveillance in 2012 and the 45th Year Surveillance in 2017.

Table 3.2.2 IWL Inspection Intervals Interval/Period Start Date End Date Sect XI Code Edition First Interval 09/09/1996 09/09/2008 1992 Ed. w/ 1992 Addenda Second Interval 09/09/2008 09/08/2018 2001 Ed. thru 2003 Addenda Turkey Point has exercised the option allowed by IWL-2421(b) to modify the inspection requirements of IWL-2420. Turkey Point satisfies the requirements of IWL-2421(a) (i.e., dual unit site with containments utilizing the same pre-stressing system and essentially identical in design, and post-tensioning operations completed not more than two years apart) with Unit 3 Initial Structural Integrity Test (ISIT) in July 1971 and Unit 4 ISIT in February 1972. Therefore, the 40th Year Surveillance consisted of a physical inspection of Unit 3 and a visual inspection of Unit 4. The 45th Year Surveillance consisted of a visual inspection of Unit 3 and a physical inspection of Unit 4.

The scope of the surveillance includes visual inspection of 100% of the accessible concrete surface areas (IWL Category L-A) and inspection and testing of a sampling of post-tensioning tendons (IWL Category L-B). Examination tendons for each surveillance consist of five (5) hoop tendons (2% of 489 with maximum of 5), four (4) vertical tendons (2% of 180) and four (4) dome tendons (2% of 165) if acceptance criteria are met during each of the last three inspections. Among the tendons to be examined, one (1) common tendon (historical) of each type (hoop, vertical and dome) is selected for partial inspection.

Also included in the 40th and 45th surveillances were augmented examinations required by Table IWL-2521-2 for tendons affected by repair/replacement activities associated with the Reactor Vessel Closure Head (RVCH) replacement project which created a temporary opening in the containment wall. While, the Code of

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 1 Page 27 of 43 Record for the 35th Year inspection period, 1992 Edition with 1992 Addenda did not require any additional tendon inspection due to the RVCH project, as a conservative measure, additional tendons were inspected during the 35th Year inspection period. The additional tendons were selected based on IWL 2521.2 of the ASME Boiler & Pressure Vessel Code,Section XI, 2001 Edition with 2002 Addenda. These inspections continued during the 40th and 45th Year inspection periods in accordance with IWL 2521.2 and Table IWL-2521-2.

Table 3.2.2-2 depicts the scope of the 2nd Interval IWL examinations.

Table 3.2.2 IWL Examinations Item Surveillance Period Parts Examined Numbers 40 th Year 45th Year Examination Category L-A, Concrete L1.11 All Accessible Surface Areas U3/U4 U3/U4 L1.12 Suspect Areas U3/U4 U3/U4 Examination Category L-B, Unbonded Post-Tensioning System L2.10 Tendon U3(1) U4(2)

L2.20 Wire or Strand U3(1) U4(2)

Anchorage Hardware and L2.30 U3/U4 U3/U4 Surrounding Concrete L2.40 Corrosion Medium U3/U4 U3/U4 L2.50 Free Water U3/U4 U3/U4 (1) Plus U4 tendon samples associated with RVCH repair/replacement activities.

(2) Plus U3 tendon samples associated with RVCH repair/replacement activities.

The inspection reports for both surveillances concluded that the containment structure had experienced no abnormal degradation of the post-tensioning systems. Measured tendon forces were all within acceptance limits and tendon test wires had acceptable results. Visual examinations of the concrete found no significant issues. Some loss of corrosion medium (grease) was detected for a number of tendons. No active corrosion was found on tendon ends and anchorage components. All indications were documented and will continue to be monitored for evidence of degradation.

3.2.3 Containment Visual Inspection The Containment Leak Rate Testing Program includes requirements for visual inspection of accessible interior and exterior surfaces of the containment structure in order to identify evidence of deterioration that may affect the containment structural integrity or leak tightness in accordance with the following:

TS Surveillance Requirement (SR) 4.6.1.6.3 for structural integrity of the containment requires, in part, visual examinations in accordance with the Containment Leakage Rate Testing Program.

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 1 Page 28 of 43 TS 6.8.4.h requires visual examinations by reference to the guidelines contained in RG 1.163 (Reference 6.8).

Regulatory Position 3 of RG 1.163 stipulates that these examinations should be conducted prior to initiating a Type A test and during two other refueling outages before the next Type A test, if the interval for the Type A test has been extended to 10 years, in order to allow for early uncovering of evidence of structural deterioration. The Containment Leak Rate Testing Program currently requires these examinations prior to initiating a Type A test and during two other refueling outages before the next Type A test.

With the implementation of the proposed change, TS 6.8.4.h will be revised by replacing the reference to RG 1.163 with reference to NEI 94-01, Revision 3-A.

NEI 94-01, Revision 3-A requires that a general visual examination of accessible interior and exterior surfaces of the containment for structural deterioration that may affect the containment leak-tight integrity must be conducted prior to each Type A test and during at least three other outages before the next Type A test if the interval for the Type A test has been extended to 15 years. The Containment Leakage Rate Testing Program requires a general visual inspection during at least three outages between Type A tests when the frequency for Type A testing is extended to 15 years. Examinations performed under the IWE and IWL Programs may be credited for the subject visual inspections of the containment.

3.3 NRC Information Notices The NRC has issued several information notices concerning containment corrosion, and FPL reviewed these notices to determine the impact on the Turkey Point containment.

3.3.1 Information Notice (IN) 92-20 IN 92-20, Inadequate Local Leak Rate Testing, dated March 3, 1992 (Reference 6.11), stated that problems exist with testing of stainless steel containment penetration bellows. Specifically, in-leakage through such bellows may not be readily detectable by LLRTs. The testing deficiency can occur if the test tap pressurizes between the two sheets of bellows materials. Turkey Point piping and ventilation penetrations are of the rigid welded type and are solidly anchored to the containment wall, thus precluding any requirement for expansion bellows.

3.3.2 IN 2011-15 IN 2011-15, Steel Containment Degradation and Associated License Renewal Aging Management Issues, dated August 1, 2011 (Reference 6.14) describes mechanisms that can lead to degradation of coatings and pitting of containment liner plates due to long term exposure to water and moisture. Similar degradation mechanisms were described in IN 2004-09, Corrosion of Steel Containment and Containment Liner, dated April 27, 2004 (Reference 6.12), which stated that over time, the existing floor-to-containment seal can degrade, allowing moisture into the crevice between the containment liner plate and floor and that small amounts of stagnant water behind the floor seal area promote pitting corrosion.

Although not referenced in IN 2011-15, IN 2010-12, Containment Liner Corrosion, dated June 18, 2010 (Reference 6.13) provided additional examples of containment liner degradation caused by corrosion. The operating experience described in IN 2011-15 relates to containment liner corrosion that results from the liner plates being in contact with objects and materials that are lodged between or

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 1 Page 29 of 43 embedded in the containment concrete. Liner locations that are in contact with objects made of an organic material are susceptible to accelerated corrosion because organic materials can trap water that combined with oxygen will promote carbon steel corrosion.

At Turkey Point there has been no operating experience of liner corrosion attributed to the inaccessible (concrete) side of the liner.

3.3.3 IN 2014-07 IN 2014-07, Degradation of Leak-Chase Channel Systems for Floor Welds of Metal Containment Shell and Concrete Containment Metallic Liner, dated May 5, 2014 (Reference 6.15), provided examples of operating experience at some plants of water accumulation and corrosion degradation in the leak-chase channel system that has the potential to affect the leak-tight integrity of the containment shell or liner plate. Turkey Point does have an air chase system inside the Unit 3 and Unit 4 containment structures, similar to the leak chase system discussed in IN 2014-

07. Walk-downs for accessible air chase test connection condition were conducted during 2014 (Unit 4) and 2015 (Unit 3) refueling outages. Test connection (grouted pipe cap) condition was determined to be satisfactory or indeterminate (inaccessible). There has been no evidence of moisture intrusion through the accessible air-chase system test connections to inaccessible portions of the containment liner plate. The IWE inspection plan will be updated to formally include general visual inspection of 100% of the accessible air chase system test connections at the containment floor-level interfaces along with opening of any identified loose or degraded test connections for internal inspection of the test connection and channel/angle to ensure no moisture intrusion to the air chase.

3.4 NRC Limitations and Conditions 3.4.1 NRC Safety Evaluation for NEI 94-01, Revision 2 The limitations and conditions from the NRCs Safety Evaluation for NEI 94-01, Revision 2 (Reference 6.2) are presented below with FPLs response.

Table 3.4 NEI 94-01, Revision 2 NRC Safety Evaluation (SE)

Limitations and Conditions Limitation/Condition Response for Turkey Point (From Section 4.1 of Safety Evaluation)

1. For calculating the Type A leakage Turkey Point utilizes the definition in NEI rate, the licensee should use the 94-01, Revision 3-A, Section 5.0. This definition in NEI TR 94-01, Revision definition has remained unchanged from 2, in lieu of that in ANSI/ANS-56.8- Revision 2 to Revision 3-A of NEI 94-01.

2002. (Refer to SE Section 3.1.1.1).

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 1 Page 30 of 43

2. The licensee submits a schedule of [Reference Section 3.2.3]

containment inspections to be performed prior to and between Type General visual observations of the A tests. (Refer to SE Section 3.1.1.3) accessible interior and external surfaces of the containment structure shall continue to be performed in accordance with the Turkey Point Appendix J Testing Program to meet the requirements of the proposed revision to TS 6.8.4.h, the inspection requirements of ASME Code Section XI, subsection IWE, and NEI 94-01, Revision 3-A, Sections 9.2.1 and 9.2.3.2.

3. The licensee addresses the areas of [Reference Section 3.2.1 through 3.2.3]

containment structure potentially subjected to degradation. (Refer to General visual observations of the SE Section 3.1.3). accessible interior and external surfaces of the containment structure shall continue to be performed in accordance with the Turkey Point Appendix J Testing Program to meet the requirements of the proposed revision to TS 6.8.4.h, the inspection requirements of ASME Code Section XI, subsection IWE and NEI 94-01, Revision 3-A, Sections 9.2.1 and 9.2.3.2.

4. The licensee addresses any tests and In general, the NRC staff considers the inspections performed following major cutting of a large hole in the modifications to the containment containment for replacement of steam structure, as applicable. (Refer to SE generators or reactor vessel heads, Section 3.1.4). replacement of large penetrations, as major repairs or modifications to the containment structure.

Turkey Point created containment openings for purposes of Reactor Vessel Closure Head replacement in 2004 (Unit 3) and 2005 (Unit 4).

Following closure of each opening, a Type A test was completed with successful results.

5. The normal Type A test interval In the event an extension beyond the 15 should be less than 15 years. If a year interval is necessary, FPL will seek licensee has to utilize the provisions prior NRC approval pursuant to 10 CFR of section 9.1 of NEI TR 94-01, 50.90, consistent with the staffs position Revision 2, related to extending the in Regulatory Issue Summary (RIS)

ILRT interval beyond 15 years, the 2008-27 (Reference 6.20).

licensee must demonstrate to the NRC staff that it is an unforeseen emergent condition. (Refer to SE Section 3.1.1.2).

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 1 Page 31 of 43

6. For plants licensed under 10 CFR Not applicable. Turkey Point was not Part 52, applications requesting a licensed under 10 CFR Part 52.

permanent extension of the ILRT surveillance interval to 15 years should be deferred until after the construction and testing of containments for that design has been completed and applicants have confirmed the applicability of NEI TR 94- 01, Revision 2, and [Electric Power Research Institute] EPRI No.

1009325, Revision 2, [Risk-Impact Assessment of Extended Integrated Leak Rate Testing Intervals,]

including the use of past containment ILRT data.

3.4.2 NRC Safety Evaluation for NEI 94-01, Revision 3 The two conditions from Section 4.0 of the NRCs Safety Evaluation for NEI 94-01, Revision 3 (Reference 6.6) are stated below with the FPL response.

Condition 1 NEI TR 94-01, Revision 3, is requesting that the allowable extended interval for Type C LLRTs be increased to 75 months, with a permissible extension (for non-routine emergent conditions) of nine months (84 months total). The staff is allowing the extended interval for Type C LLRTs to be increased to 75 months with the requirement that a licensees post-outage report include the margin between Type B and Type C leakage rate summation and its regulatory limit. In addition, a corrective action plan shall be developed to restore the margin to an acceptable level. The staff is also allowing the non-routine emergent extension out to 84 months as applied to Type C valves at a site, with some exceptions that must be detailed in NEI 94-01, Revision 3. At no time shall an extension be allowed for Type C valves that are restricted categorically (e.g., BWR MSIVs [Main Steam Isolation Valves]), and those valves with a history of leakage, or any valves held to either a less than maximum interval or to the base refueling cycle interval. Only non-routine emergent conditions allow an extension to 84 months. This is Topical Report Condition 1.

Response to Condition 1:

Condition 1 presents three issues that are addressed as follows:

Issue 1 - The allowance of an extended interval for Type C LLRTs of 75 months carries the requirement that a licensee's post-outage report include the margin between the Type B and Type C leakage rate summation and its regulatory limit.

Response to Condition 1, Issue 1:

Consistent with NEI 94-01, Revision 3-A, the post-outage report shall include the margin between the Type B and Type C minimum pathway leak rate summation value, as adjusted to include the estimate of

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 1 Page 32 of 43 applicable Type C leakage understatement, and its regulatory limit of 0.60 La.

Issue 2 - A corrective action plan shall be developed to restore the margin to an acceptable level.

Response to Condition 1, Issue 2:

Consistent with NEI 94-01, Revision 3-A, when the potential leakage understatement adjusted Type B and Type C minimum pathway leak rate total is greater than the Turkey Point administrative leakage summation limit of 0.40 La, but less than the regulatory limit of 0.60 La, then an analysis and determination of a corrective action plan shall be prepared to restore the leakage summation margin to less than the administrative leakage limit.

The corrective action plan shall focus on those components which have contributed the most to the increase in the leakage summation value and the manner of timely corrective action (as deemed appropriate) that best focuses on the prevention of future component leakage performance issues.

Issue 3 - Use of the allowed 9-month extension for eligible Type C valves is only authorized for non-routine emergent conditions.

Response to Condition 1, Issue 3:

Consistent with NEI 94-01, Revision 3-A, Turkey Point will apply the 9-month grace period only to eligible Type C components and only for non-routine emergent conditions. Such occurrences will be documented in the record of tests.

Condition 2:

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. The containment leakage condition monitoring regime involves a portion of the penetrations being tested each refueling outage, nearly all LLRTs being performed during plant outages. For the purposes of assessing and monitoring or trending overall containment leakage potential, the as-found minimum pathway leak rates for the just tested penetrations are summed with the as-left minimum pathway leak rates for penetrations tested during the previous one, two, or even three refueling outages. Type C tests involve valves which, in the aggregate, will show increasing leakage potential due to normal wear and tear, some predictable and some not so predictable.

Routine and appropriate maintenance may extend this increasing leakage potential. Allowing for longer intervals between LLRTs means that more leakage rate test results from farther back in time are summed with fewer just tested penetrations and that total used to assess the current containment leakage potential. This leads to the possibility that the LLRT totals calculated understate the actual leakage potential of the penetrations. Given the required margin included with the performance

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 1 Page 33 of 43 criterion and the considerable extra margin most plants consistently show with their testing, any understatement of the LLRT total using a 5-year test frequency is thought to be conservatively accounted for. Extending the LLRT intervals beyond five years to a 75-month interval should be similarly conservative provided an estimate is made of the potential understatement and its acceptability determined as part of the trending specified in NEI 94-01, Revision 3, Section 12.1.

When routinely scheduling any LLRT valve interval beyond 60-months and up to 75- months, the primary containment leakage rate testing program trending or monitoring must include an estimate of the amount of understatement in the Type B & C total, and must be included in a licensees post-outage report. The report must include the reasoning and determination of the acceptability of the extension, demonstrating that the LLRT totals calculated represent the actual leakage potential of the penetrations. This is Topical Report Condition 2.

Response to Condition 2 Condition 2 presents two issues that are addressed as follows:

Issue 1 - Extending the LLRT intervals beyond five years (60 months) to a 75-month interval should be similarly conservative provided an estimate is made of the potential understatement and its acceptability determined as part of the trending specified in NEI 94-01, Revision 3, Section 12.1.

Response to Condition 2, Issue 1:

The change in going from a 60-month extended test interval for Type C tested components to a 75-month interval, as authorized under NEI 94-01, Revision 3-A, represents an increase of 25 percent in the local leak rate test periodicity. Consistent with NEI 94-01, Revision 3-A, Turkey Point will conservatively apply a potential leakage understatement adjustment factor of 1.25 to the as-left leakage total for each Type C component currently on the greater than 60-month (up to 75 month) extended test interval. This will result in a combined conservative Type C total for all 60-75 month local leak rate tests being carried forward and included whenever the total leakage summation is required to be updated (either while operating on-line or following an outage). When the potential leakage understatement adjusted leak rate total for those Type C components being tested on a greater than 60-month (up to 75 month) extended interval is summed with the non-adjusted total of those Type C components being tested at less than the 60-75 month interval and the total of the Type B tested components, if the minimum pathway leak rate is greater than the Turkey Point administrative leakage summation limit of 0.5 La, but less than the regulatory limit of 0.60 La, then an analysis and corrective action plan shall be prepared to restore the leakage summation value to less than the administrative leakage limit. The corrective action plan shall focus on those components that have contributed the most to the increase in the leakage summation value.

Issue 2 - When routinely scheduling any LLRT valve interval beyond 60 months and up to 75 months, the primary containment leakage rate testing program trending or monitoring must include an estimate of the amount of understatement in the Type B & C total, and must be included in a

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 1 Page 34 of 43 licensees post-outage report. The report must include the reasoning and determination of the acceptability of the extension, demonstrating that the LLRT totals calculated represent the actual leakage potential of the penetrations.

Response to Condition 2, Issue 2:

If the potential leakage understatement adjusted minimum pathway leak rate is less than the administrative leakage summation limit of 0.5 La, then the acceptability of the 75-month local leak rate test extension for all affected Type C components has been adequately demonstrated and the calculated local leak rate total represents the actual leakage potential of the penetrations.

In addition to Condition 1, Issues 1 and 2, which deal with the minimum pathway leak rate Type B and Type C summation margin, NEI 94-01, Revision 3-A, also has the following margin related requirement contained in Section 12.1, Report Requirements.

A post-outage report shall be prepared presenting results of the previous cycle's Type B and Type C tests, and Type A, Type B, and Type C tests, if performed during that outage. The technical contents of the report are generally described in ANSl/ANS-56.8-2002 and shall be available on-site for NRC review. The report shall show that the applicable performance criteria are met, and serve as a record that continuing performance is acceptable. The report shall also include the combined Type B and Type C leakage summation, and the margin between the Type B and Type C leakage rate summation and its regulatory limit. Adverse trends in the Type B and Type C leakage rate summation shall be identified in the report and a corrective action plan developed to restore the margin to an acceptable level.

In the event an adverse trend in the potential leakage understatement adjusted Type B and Type C summation is identified, an analysis and a corrective action plan shall be prepared to restore the margin to an acceptable level thereby eliminating the adverse trend. The corrective action plan shall focus on those components that have contributed the most to the adverse trend in the leakage summation value and what manner of timely corrective action, as deemed appropriate, best focuses on the prevention of future component leakage performance issues.

An adverse trend is defined as three consecutive increases in the Type B and Type C minimum pathway leak rate summation value adjusted to include the estimate of applicable Type C leakage understatement, as expressed in terms of La.

3.5 Plant-Specific Confirmatory Analysis

3.5.1 Methodology

An evaluation has been performed to assess the risk impact of extending the Turkey Point Type A test interval from the current 10 years to 15 years. A simplified bounding analysis consistent with the Electric Power Research Institute (EPRI)

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 1 Page 35 of 43 approach was used for evaluating the change in risk associated with increasing the test interval to fifteen years. The approach is consistent with that presented in:

Appendix H of Electric Power Research Institute, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals: Revision 2-A of 1009325, EPRI Topical Report TR-1018243, dated October 2008 (Reference 6.4);

Electric Power Research Institute, Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals, EPRI Topical Report TR-104285, dated August 1994; Nuclear Regulatory Commission, Performance-Based Containment Leak-Test Program, NUREG-1493, dated September 1995; and, Calvert Cliffs liner corrosion analysis described in a letter to the NRC dated March 27, 2002 (Reference 6.17).

The analysis uses results from Turkey Point's analysis of core damage scenarios (Level 1) and subsequent containment responses (Level 2) resulting in various fission product release categories (including intact containment or negligible release).

In the safety evaluation issued by NRC for NEI 94-01, Revision 2 (Reference 6.7),

the NRC concluded that the methodology in EPRI Report No. 1009325, Revision 2 (Reference 6.3), is acceptable for referencing by licensees proposing to amend their TS to permanently extend the Type A surveillance test interval to 15 years, subject to the conditions noted in Section 4.2 of the safety evaluation. The following table addresses each of the four conditions for the use of EPRI Report No.

1009325, Revision 2.

Table 3.5 EPRI Report No. 1009325, Revision 2, Limitations and Conditions Conditions (Section 4.2 of NRC Safety Response for Turkey Point Evaluation dated June 25, 2008)

1. The licensee submits documentation The Turkey Point PRA technical that the technical adequacy of their adequacy is addressed in Section (probabilistic risk assessment) PRA is 3.5.2.

consistent with the requirements of

[Regulatory Guide] RG 1.200 relevant to the [integrated leakage rate test]

ILRT extension application.

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 1 Page 36 of 43

2. The licensee submits documentation EPRI Report No. 1009325, Revision indicating that the estimated risk 2-A, incorporates these population increase associated with permanently dose and conditional containment extending the ILRT surveillance interval failure probability acceptance to 15 years is small, and consistent with guidelines, and these guidelines have the clarification provided in Section been used for the Turkey Point plant 3.2.4.5(1) of this [safety evaluation] SE. specific risk assessment.

Specifically, a small increase in The increase in population dose is population dose should be defined as discussed in Section 3.5.3.

an increase in population dose of less than or equal to either 1.0 person-rem per year or 1 percent of the total population dose, whichever is less restrictive.

In addition, a small increase in The increase in conditional

[conditional containment failure containment failure probability is probability] CCFP should be defined as discussed in Section 3.5.3.

a value marginally greater than that accepted in previous one-time 15 year ILRT extension requests. This would require that the increase in CCFP be less than or equal to 1 5 percentage

3. The methodology in EPRI Report EPRI Report No. 1009325, Revision 2-No. 1009325, Revision 2, is acceptable A, incorporates the use of 100 La as except for the calculation in the increase the average leak rate for the pre-in expected population dose (per year of existing containment large leakage reactor operation). In order to make the rate accident case (accident class 3b),

methodology acceptable, the average and this value has been used in the leak rate for the pre-existing Turkey Point plant specific risk containment large leak rate accident assessment.

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

4. A [license amendment request] LAR is Turkey Point does not rely on required in instances where containment containment overpressure for ECCS overpressure is relied upon for performance, consistent with the

[emergency core cooling system] ECCS guidance in NEI 04-07, Revision 0, performance. Volume 1.

(1) The SE for EPRI Report No. 1009325, Revision 2, indicates that the clarification regarding small increases in risk is provided in Section 3.2.4.5; however, the clarification is actually provided in Section 3.2.4.6.

3.5.2 Probabilistic Risk Assessment (PRA) Acceptability The Turkey Point Internal Events, Internal Flood, and Fire PRA models have been peer reviewed and there are no PRA upgrades that have not been peer reviewed.

The PRA models credited in this request are the PRA models used in the NFPA 805 application (Reference 8.44 of Enclosure 3) with maintenance updates applied. Capability Category (CC) II of the NRC-endorsed ASME/ANS PRA Standard is the target capability level for the NFPA 805 application. The acceptability (previously referred to as technical adequacy or quality) of the PRA model was reviewed by the NRC for the license amendment request (LAR) to adopt a risk-informed, performance-based fire protection program in accordance with 10 CFR 50.48(c) and NFPA 805 (2001 Edition) and found to be technically acceptable (Reference 8.36 of Enclosure 3). The impact of recent NRC-approved LARs to adopt a risk-informed Technical Specification Completion Time (RICT) program and to revise reactor coolant pump (RCP) seal PRA credit is discussed

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 1 Page 37 of 43 in Appendix A1 of Enclosure 3. Note that no equipment or mitigating strategies involving portable Diverse and Flexible Coping Strategies (FLEX) equipment is credited in the PRA models used in this LAR.

As stated in the NRC Final Safety Evaluation for NEI 94-01, Revision 2 and EPRI Report No. 1009325, Revision 2, CC I of the ASME PRA Standard shall be applied as the standard for assessing PRA quality for ILRT extension applications, as approximate values of core damage frequency (CDF) and LERF and their distribution among release categories, are sufficient to support the evaluation of changes to ILRT frequencies. The NRC Safety Evaluation also states the assessment of external events can be taken from existing, previously submitted and approved analyses or other alternate method of assessing an order of magnitude estimate for contribution of the external event to the impact of the changed interval. Therefore, the ILRT interval extension risk assessment is allowed to use the existing Internal Flooding and Fire PRA models and other existing seismic and external hazard evaluations.

Appendix A1 of Enclosure 3 provides a more detailed discussion of the external hazard evaluations and the PRA acceptability for the ILRT interval extension risk impact assessment. The information in Appendix A1 of Enclosure 3 demonstrates that the PRA is of sufficient quality and level of detail to support this submittal, and has been subjected to a peer review process assessed against a standard or set of acceptance criteria that is endorsed by the NRC.

3.5.3 Conclusions of the Plant-Specific Risk Assessment Results The findings of the Turkey Point risk assessment confirm the general findings of previous studies that the risk impact associated with extending the Type A test interval from three in ten years to one in 15 years is small. The Turkey Point plant-specific results for extending the Type A test interval from the three in ten years to 15 years is summarized below.

CDF is not impacted by the proposed change. Turkey Point does not rely on containment overpressure to assure adequate net positive suction head is available for emergency core cooling system pumps taking suction from the containment sump following design basis accidents.

RG 1.174 (Reference 6.9) provides guidance for determining the risk impact of plant-specific changes to the licensing basis. RG 1.174 defines very small changes in risk as resulting in increases of CDF less than 1E-06 per reactor year and increases in LERF less than 1E-07 per reactor year.

There is no quantifiable change in CDF as a result of the proposed ILRT Type A test interval extension. Therefore, the RG 1.174 acceptance guideline for a very small change in CDF is considered to be met as the impact on CDF for the Type A test interval extension is negligible. Thus, the relevant acceptance criterion is LERF.

The increase in LERF with consideration of liner corrosion included resulting from a change in the Type A ILRT test interval from three in ten years to one in fifteen years is conservatively estimated for Unit 3 as 2.54E-09/yr due to internal events contribution and 5.59E-07/yr due to internal flood and external events. The total combined impact for Unit 3 is 5.62E-07/yr. For Unit 4, the impact is conservatively estimated as 2.55E-09/yr due to internal events contribution and 5.74E- 07/yr due

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 1 Page 38 of 43 to internal flood and external events. The total combined impact for Unit 4 is 5.77E-07/yr.

The impact due to an increase in the Type A ILRT interval to one in fifteen years is very small when considering only internal events. RG 1.174 (Reference 6.9) also states that when the calculated increase in LERF is in the small range of 1.0E-07 per reactor year to 1.0E-06 per reactor year, applications will be considered only if it can be reasonably shown that the total LERF is less than 1.0E-05 per reactor year. When including the impact from internal flood and external events, the change to LERF is in the small range for both Turkey Point Unit 3 and Unit 4. Therefore, the total LERF is evaluated. The resulting total LERF for Unit 3 is approximately 1.89E-06/yr. The resulting total LERF for Unit 4 is approximately 2.14E-06/yr. The total LERF for both Unit 3 and Unit 4 are below the RG 1.174 acceptance criteria for total LERF of 1.0E-05/yr and therefore this change satisfies both the incremental and absolute criteria with regard to the RG 1.174 LERF metric.

The calculated increase in the total 50-mile population dose risk for the proposed ILRT Type A interval change from three per ten years to once per 15 years is measured as an increase to the total integrated dose risk for all accident sequences influenced by Type A testing. The total 50-mile population dose increase (relative to the base case, with corrosion) is 0.88% of the total dose rate for Unit 3 and 0.81% of the total dose rate for Unit 4. (The percent increases remain the same whether considering internal events only or including the effect of internal flood and external events.) EPRI Report No. 1009325, Revision 2-A, states that a very 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 one percent of the total population dose, whichever is less restrictive. Thus, the estimated 50-mile population dose increase at Turkey Point is very small using the guidelines of EPRI Report No. 1009325, Revision 2-A.

The increase in the conditional containment failure probability from the three per ten years to once in 15 years Type A test interval including corrosion effects is 0.87% for Unit 3 and 0.88% for Unit 4. EPRI Report No. 1009325, Revision 2-A, states that increases in conditional containment failure probability of less than or equal to 1.5 percentage points are very small. Therefore, this increase is judged to be very small at Turkey Point.

In summary, based on the above results, the proposed 15-year Type A test interval represents a very small change in risk and is acceptable as a permanent change.

Details of the Turkey Point risk assessments are contained in Enclosure 3 of the LAR.

3.6 Conclusion NEI 94-01, Revision 3-A, describes an NRC accepted approach for implementing the performance-based requirements of Appendix J, Option B. It incorporates the regulatory positions stated in RG 1.163 and includes provisions for extending Type A test 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.

Based on the previous Type A tests conducted at Turkey Point, extension of the containment Type A test interval from ten to 15 years represents minimal risk to increased

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 1 Page 39 of 43 leakage. The risk is further minimized by continued Type B and Type C testing performed in accordance with Appendix J, Option B, and the overlapping inspection activities performed as part of the following Turkey Point inspection programs:

Containment lnservice Inspection Program Containment Coatings Inspection and Assessment Program This experience is supplemented by risk analysis studies, including the Turkey Point risk analysis provided in Enclosure 3. The findings of the risk assessment confirm the general findings of previous industry studies, on a plant-specific basis, that extending the Type A test interval from ten to 15 years results in a very small and acceptable change to the Turkey Point baseline risk.

4.0 REGULATORY EVALUATION

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

10 CFR 50.54(o) requires primary reactor containments for water-cooled power reactors to be subject to the requirements of Appendix J to 10 CFR 50, Primary Reactor Containment Leakage Testing for Water-Cooled Nuclear Power Reactors. Appendix J specifies containment leakage testing requirements, including the types required to ensure the leakage through the primary reactor containment and systems and components penetrating primary containment shall not exceed allowable leakage rate values and periodic surveillance of reactor containment penetrations and isolation valves is performed so that proper maintenance and repairs are made during the service life of the containment, and systems and components penetrating primary containment. In addition, Appendix J discusses leakage rate test methodology, frequency of testing, and reporting requirements for each type of test.

RG 1.163, Performance Based Containment Leak Test Program, (September 1995) provides a method acceptable to the NRC for implementing the performance based option (Option B) of 10 CFR 50, Appendix J. The regulatory positions stated in RG 1.163 (September 1995) as modified by NRC Safety Evaluations of June 25, 2008 (Reference 6.7) and June 8, 2012 (Reference 6.6) are incorporated in NEI 94 01, Revision 3-A, Industry Guideline for Implementing Performance Based Option of 10 CFR Part 50, Appendix J.

The proposed license amendment would revise Turkey Point TS 6.8.4.h, Containment Leakage Rate Testing Program, by changing the wording to indicate that the program shall be in accordance with NEI 94-01, Revision 3-A, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, and the conditions and limitations specified in NEI 94-01, Revision 2-A, instead of RG 1.163, Performance Based Containment Leak Test Program, and the listed Type A test exception. The purpose of NEI 94-01 is to assist licensees in the implementation of Option B to 10 CFR Part 50, Appendix J. The NRC staff has reviewed NEI 94-01, Revision 3, and found that this guidance, as modified to include two limitations and conditions, is acceptable for referencing by licensees proposing to amend their TS in regards to containment leakage rate testing.

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 1 Page 40 of 43 FPL has evaluated the proposed changes against the applicable regulatory requirements and acceptance criteria. Based on the foregoing, the proposed amendment will continue to ensure compliance with 10 CFR 50.54(o), and Option B of 10 CFR Part 50, Appendix J.

4.2 Precedents This license amendment request is similar to Amendment No. 153 which was approved for Seabrook Station on March 15, 2017 (Reference 6.18) and Amendments 265 and 268 which were approved for Point Beach Nuclear Plant, Units 1 and 2, respectively, on April 25, 2019 (Reference 6.19). These amendments authorized containment leakage rate test programs for Seabrook and Point Beach 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 FPL requests license amendments to revise Turkey Point Technical Specification (TS) 6.8.4.h, Containment Leakage Rate Testing Program, to require a program that is in accordance with 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. This change will allow extension of the Type A test interval up to one test in 15 years and extension of the Type C test interval up to 75 months.

FPL has evaluated whether a significant hazards consideration is involved with the proposed amendments by focusing on the three standards set forth in 10 CFR 50.92, Issuance of Amendment, as discussed below:

(1) Do the proposed amendments involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No The proposed amendments adopt the NRC-accepted guidelines of NEI 94-01, Revision 3-A, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, for development of the Turkey Point performance-based containment testing program. NEI 94-01 allows, based on risk and performance, an extension of Type A and Type C containment leak test intervals.

Implementation of these guidelines continues to provide adequate assurance that during design basis accidents, the primary containment and its components will limit leakage rates to less than the values assumed in the plant safety analyses.

The findings of the Turkey Point risk assessment confirm the general findings of previous studies that the risk impact associated with extending the containment leak rate is small. Per the guidance provided in RG 1.174, an extension of the leak test interval in accordance with NEI 94-01, Revision 3-A results in an estimated change within the very small change region. Note that the proposed change does not affect the probability (or frequency) of an accident occurring.

Since the change is implementing a performance-based containment testing program, the proposed amendment does not involve either a physical change to the plant or a change in the manner in which the plant is operated or controlled.

The requirement for containment leakage rate acceptance will not be changed by this amendment.

(2) Do the proposed amendments create the possibility of a new or different kind of

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 1 Page 41 of 43 accident from any accident previously evaluated?

Response: No The proposed change to implement a performance-based containment testing program, associated with integrated leakage rate test frequency, does not change the design or operation of structures, systems, or components of the plant.

The proposed change would continue to ensure containment integrity and would ensure operation within the bounds of existing accident analyses. There are no accident initiators created or affected by this change. Therefore, the proposed change will not create the possibility of a new or different kind of accident from any accident previously evaluated.

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

(3) Do the proposed amendments involve a significant reduction in a margin of safety?

Response: No The proposed change to extend Type A and Type C test intervals does not affect plant operations, design functions, or any analysis that verifies the capability of a structure, system, or component of the plant to perform a design function. In addition, this change does not affect safety limits, limiting safety system setpoints, or limiting conditions for operation.

The specific requirements and conditions of the TS 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 specified by TS is maintained. This ensures that the margin of safety in the plant safety analysis is maintained. The design, operation, testing methods and acceptance criteria for Type A, B, and C containment leakage tests specified in applicable codes and standards would continue to be met with the acceptance of this proposed change since these are not affected by extension of the test intervals for Type A and Type C testing.

Therefore, the proposed amendments do not involve a significant reduction in a margin of safety.

Based upon the above analysis, FPL concludes that the proposed license amendments do not involve a significant hazards consideration, under the standards set forth in 10 CFR 50.92, Issuance of Amendment, and accordingly, a finding of no significant hazards consideration is justified.

4.4 Conclusion In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will continue to be conducted in compliance with the Commissions regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 1 Page 42 of 43 FPL has evaluated the proposed amendment for environmental considerations. The review has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or would change an inspection or surveillance requirement. However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment needs to be prepared in connection with the proposed amendment.

6.0 REFERENCES

6.1 Nuclear Energy Institute (NEI) Topical Report, 94-01, Revision 3-A, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, dated July 2012 (ADAMS Accession No. ML12221A202).

6.2 Nuclear Energy Institute (NEI) Topical Report, 94-01, Revision 2-A, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, dated October 2008 (ADAMS Accession No. ML100620847).

6.3 Electric Power Research Institute, Report No. 1009325, Revision 2, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, dated August 2007 (ADAMS Accession No. ML072970208).

6.4 Electric Power Research Institute, Report No. 1009325, Revision 2-A, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, dated October 2008 (also identified as EPRI TR-1018243, which is publicly available and can be found at www.epri.com by typing 1018243 in the search field box).

6.5 ANSI/ANS-56.8-2002, Containment System Leakage Testing Requirements.

6.6 Letter from S. Bahadur (NRC) to B. Bradley (NEI), Final Safety Evaluation of Nuclear Energy Institute (NEI) Report, 94-01, Revision 3, Industry Guideline for Implementing Performance- based Option of 10 CFR Part 50, Appendix J, (TAC No. ME2164), dated June 8, 2012 (ADAMS Accession No. ML121030286).

6.7 Letter from M. Maxin (NRC) to J. Butler (NEI), 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 (ADAMS Accession No. ML081140105).

6.8 U.S. Nuclear Regulatory Commission, Regulatory Guide 1.163, Performance-Based Containment Leak-Test Program, dated September 1995 (ADAMS Accession No. ML003740058).

6.9 U.S. Nuclear Regulatory Commission, 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, dated January 2018 (ADAMS Accession No. ML17317A256).

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 1 Page 43 of 43 6.10 U.S. Nuclear Regulatory Commission, "Performance-Based Containment Leak-Test Program," NUREG-1493, September 1995.

6.11 U.S. Nuclear Regulatory Commission, Information Notice 92-20, Inadequate Local Leak Rate Testing, dated March 3, 1992 (https://www.nrc.gov/reading-rm/doc-collections/gen-comm/info- notices/1992/in92020.html).

6.12 U.S. Nuclear Regulatory Commission, Information Notice 2004-09, Corrosion of Steel Containment and Containment Liner, dated April 27, 2004 (ADAMS Accession No. ML041170030).

6.13 U.S. Nuclear Regulatory Commission, Information Notice 2010-12, Containment Liner Corrosion, dated June 18, 2010 (ADAMS Accession No. ML100640449).

6.14 U.S. Nuclear Regulatory Commission, Information Notice 2011-15, Steel Containment Degradation and Associated License Renewal Aging Management Issues, dated August 1, 2011 (ADAMS Accession No. ML111460369).

6.15 U.S. Nuclear Regulatory Commission, Information Notice 2014-07, Degradation of Leak-Chase Channel Systems for Floor Welds of Metal Containment Shell and Concrete Containment Metallic Liner, dated May 5, 2014 (ADAMS Accession No. ML14070A114).

6.16 U.S. Nuclear Regulatory Commission, Safety Evaluation, AC Nos. 63152 and 63153, Containment Liner Leak Chase Channel Venting, September 18, 1989 (ADAMS Legacy Library Accession Nos. 8909270236 and 8910020122).

6.17 Calvert Cliffs Nuclear Power Plant, Response to Request for Additional Information Concerning the License Amendment Request for One Time Integrated Leakage Rate Test Extension, dated March 27, 2002 (ADAMS Accession No. ML020920100).

6.18 U.S. Nuclear Regulatory Commission, Seabrook Station, Unit No. 1 - Issuance of Amendment Re: Extension of Containment Leakage Rate Test Frequency (CAC No.

MF7565), dated March 15, 2017 (ADAMS Accession No. ML17046A443).

6.19 U.S. Nuclear Regulatory Commission, Point Beach Nuclear Plant, Unit 1 & 2 - Issuance of Amendments to Extend Containment Leakage Rate Test Frequency (EPID L-2018-LLA-0097), dated April 25, 2019 (ADAMS Accession No. ML19064A904).

6.20 U.S. Nuclear Regulatory Commission, NRC Regulatory Issue Summary 2008-027 Staff Position on Extension of the Containment Type A Test Interval Beyond 15 Years Under Option B of Appendix J to 10 CFR Part 50, dated December 8, 2008 (ADAMS Accession No. ML080020394).

6.21 Turkey Point, Units 3 & 4, License Amendments 73 & 67, Revise Technical Specifications to Include the Air Lock Testing According to Appendix J to 10 CFR Part 50, Make Certain Corrections in Terminology to be Consistent with Appendix J, dated November 4, 1981 (ADAMS Accession No. ML013340299).

6.22 BN-TOP-1, Revision 1, Testing Criteria for Integrated Leakage Rate Testing of Primary Containment Structures for Nuclear Power Plants, dated November 1, 1972.

6.23 Florida Power and Light (FPL) letter L-2019-192, License Amendment Request 270, Modify Containment Atmosphere Radioactivity Monitoring, Containment Ventilation Isolation and RCS Leakage Detection System Requirements dated November 4, 2019 (ADAMS Accession No. ML19315A003)

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 ENCLOSURE 2 PROPOSED TECHNICAL SPECIFICATION PAGE (MARKUP)

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Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 2 ENCLOSURE 2 PROPOSED TECHNICAL SPECIFICATION PAGE (MARKUP)

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 1 of 82 TURKEY POINT NUCLEAR PLANT PERMANENT ILRT INTERVAL EXTENSION RISK ASSESSMENT (APPENDIX A TO S&L 2019-01368)

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Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 2 of 82 Table of Contents Section Page 1.0 Purpose of Analysis ...............................................................................................................3 1.1 Purpose .................................................................................................................................3 1.2 Background ..........................................................................................................................3 1.3 Criteria .................................................................................................................................4 2.0 Methodology .........................................................................................................................5 3.0 Ground Rules .........................................................................................................................6 4.0 Inputs .....................................................................................................................................7 4.1 General Resources Available ...............................................................................................7 4.2 Plant Specific Inputs ..........................................................................................................11 4.3 Impact of Extension on Detection of Component Failures That Lead to Leakage (Small and Large) .......................................................................................................................20 4.4 Impact of Extension on Detection of Steel Liner Corrosion that Leads to Leakage .........21 5.0 Results .................................................................................................................................24 5.1 Step 1 - Quantify the Base-Line Risk in Terms of Frequency Per Reactor Year ..............26 5.2 Step 2 - Develop Plant Specific Person-Rem Dose (Population Dose) Per Reactor Year 30 5.3 Step 3 - Evaluate Risk Impact of Extending Type A Test Interval From 10 to 15 Years .34 5.4 Step 4 - Determine the Change in Risk in Terms of Large Early Release Frequency (LERF) ............................................................................................................................39 5.5 Step 5 - Determine the Impact on the Conditional Containment Failure Probability (CCFP) ............................................................................................................................39 5.6 Summary of Results ...........................................................................................................41 6.0 Sensitivities .........................................................................................................................42 6.1 Sensitivity to Corrosion Impact Assumptions ...................................................................42 6.2 Sensitivity to Class 3B Contribution to LERF...................................................................44 6.3 Potential Impact from External Events and Internal Flooding Contribution .....................44 7.0 Conclusions .........................................................................................................................71 8.0 References ...........................................................................................................................73 Appendix A1 ..................................................................................................................................77 Disposition and Resolution of Open Peer Review Findings and Self-Assessment Open Items ....81 Page l 2

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 3 of 82 1.0 Purpose of Analysis 1.1 Purpose The purpose of this analysis is to provide a risk assessment for extending the currently allowed containment Type A Integrated Leak Rate Test (ILRT) frequency to a permanent fifteen years.

The extension would allow for substantial cost savings as the ILRT could be deferred for additional scheduled refueling outages for Turkey Point Nuclear Generating Station (PTN). The risk assessment follows the guidelines from Nuclear Energy Institute (NEI) 94-01, Revision 3-A1 (Reference 1), the methodology used in Electric Power Research Institute (EPRI) TR-104285 (Reference 8.2), the NEI Interim Guidance for Performing Risk Impact Assessments In Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals from November 2001 (Reference 8.3), the NRC regulatory guidance on the use of Probabilistic Risk Assessment (PRA) as stated in Regulatory Guide (RG) 1.200 (Reference 8.35) as applied to ILRT interval extensions, and risk insights in support of a request for a plants licensing basis as outlined in RG 1.174 (Reference 8.4), the methodology used for Calvert Cliffs to estimate the likelihood and risk implications of corrosion induced leakage of steel liners going undetected during the extended test interval (Reference 8.5), and the methodology used in EPRI Report No. 1009325, Revision 2-A2 (Reference 8.26).

1.2 Background Revisions to 10 CFR 50, Appendix J (Option B) allow individual plants to extend the ILRT surveillance testing frequency requirement from three in ten years to at least once in ten years.

The revised ILRT frequency is based on an acceptable performance history defined as two consecutive periodic ILRTs at least 24 months apart in which the calculated performance leakage rate was less than the limiting containment leakage rate of 1La3.

The basis for the current 10-year test interval for PTN is provided in Section 11.0 of NEI 94-01, Revision 0, and was established in 1995. Section 11.0 of NEI 94-01 states that NUREG-1493, Performance-Based Containment Leak Test Program, September 1995 (Reference 8.6),

provides the technical basis to support rulemaking to revise leakage rate testing requirements contained in Option B to Appendix J. The basis consisted of qualitative and quantitative assessments of the risk impact (in terms of increased public dose) associated with a range of extended leakage rate test intervals. To supplement the NRCs rulemaking basis, NEI undertook a similar study. The results of that study are documented in EPRI TR-104285, Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals.

The NRC 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 PWR 1

Note that there are no differences in the risk assessment criteria between Revisions 2-A and 3-A.

2 EPRI Report No. 1009325, Revision 2-A, is also identified as EPRI TR-1018243.

3 La (percent/24 hours) is the maximum allowable leakage rate at the calculated peak design containment internal accident pressure, Pa, as specified in the Technical Specifications.

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Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 4 of 82 plant (i.e., Surry) 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 ILRT interval will not lead to a substantial increase in risk from containment isolation failures for PTN.

The guidance provided in Appendix H of EPRI Report No. 1009325, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, (Reference 8.26) for performing risk impact assessments in support of ILRT extensions builds on the EPRI Risk Assessment methodology, EPRI TR-104285. This methodology is followed to determine the appropriate risk information for use in evaluating the impact of the proposed ILRT changes.

It should be noted that containment leak-tight integrity is also verified through periodic inservice inspections conducted in accordance with the requirements of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section XI. More specifically, Subsection IWE provides the rules and requirements for inservice inspection of Class MC pressure-retaining components and their integral attachments, and of metallic shell and penetration liners of Class CC pressure-retaining components and their integral attachments in light-water cooled plants. Furthermore, NRC regulations 10 CFR 50.55a(b)(2)(ix)(E) require licensees to conduct visual inspections of the accessible areas of the interior of the containment.

The associated change to NEI 94-01 will require that visual examinations be conducted during at least three other outages, and in the outage during which the ILRT is being conducted. These requirements will not be changed as a result of the extended ILRT interval. In addition, Appendix J, type B and C local leak tests are performed to verify the leak-tight integrity of containment penetration bellows, airlocks, seals, and gaskets and containment isolation valves.

These tests are not affected by the change to the ILRT frequency; however Type C test frequencies are being updated as a part of this LAR.

1.3 Criteria The acceptance guidelines in RG 1.174 are used to assess the acceptability of this permanent extension of the ILRT interval beyond that established during the Option B rulemaking of Appendix J. RG 1.174 defines very small changes in the risk-acceptance guidelines as increases in Core Damage Frequency (CDF) less than 10-6 per reactor year and increases in Large Early Release Frequency (LERF) less than 10-7 per reactor year. Since the ILRT does not impact CDF, the relevant criterion is the change in LERF. RG 1.174 also defines small changes in LERF as below 10-6 per reactor year. RG 1.174 discusses defense-in-depth and encourages the use of risk analysis techniques to help ensure and show that key principles, such as the defense-in-depth philosophy, are met. Therefore, the increase in the Conditional Containment Failure Probability (CCFP) that helps to ensure that the defense-in-depth philosophy is maintained is also calculated.

The criteria described below are taken from Section 3.2.4.6 of the NRC Final Safety Evaluation for NEI 94-01, Revision 2 and EPRI Report No. 1009325, Revision 2 (Reference 8.29).

Regarding CCFP, the NRC concluded that a small increase in CCFP should be defined as a value marginally greater than that accepted in previous one time fifteen year ILRT extension requests.

To this end the NRC has endorsed a small increase in CCFP as an increase in CCFP less than or equal to 1.5% (Reference 8.29).

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Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 5 of 82 In addition, the total annual risk (person-rem/yr population dose) is examined to demonstrate the relative change in this parameter. For purposes of assessing the risk impacts of the Type A ILRT extension in accordance with the EPRI methodology, the NRC concluded that a small increase in population dose should be defined as an increase in population dose of less than or equal to either 1.0 person-rem per year or 1% of the total population dose, whichever is less restrictive.

2.0 Methodology A simplified bounding analysis approach consistent with the EPRI approach is used for evaluating the change in risk associated with increasing the test interval to fifteen years. The approach is consistent with that presented in Appendix H of EPRI Report No. 1009325, Revision 2-A, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals (Reference 8.26), EPRI TR-104285 (Reference 8.2), NUREG-1493 (Reference 8.6) and the Calvert Cliffs liner corrosion analysis (Reference 8.5). The analysis uses results from the Level 2 PRA model of core damage scenarios from the PTN PRA model to establish the containment fission product release categories and associated release frequencies.

The six general steps of this assessment are as follows:

1. Quantify the baseline risk in terms of the frequency of events (per reactor year) for each of the eight containment release scenario types identified in EPRI Report No. 1009325, Revision 2-A (Reference 8.26).

2. Develop plant-specific person-rem (population dose) per reactor year for each of the eight containment release scenario types from plant specific consequence analyses.

3. Evaluate the risk impact (i.e., the change in containment release scenario type frequency and population dose) of extending the ILRT interval to fifteen years.

4. Determine the change in risk in terms of LERF in accordance with RG 1.174 (Reference 8.4) and compare with the acceptance guidelines of RG 1.174.

5. Determine the impact of the ILRT interval extension on the CCFP and the population dose and compare with the acceptance guidance of Reference 8.29.

6. Evaluate the sensitivity of the results to assumptions in the liner corrosion analysis, external events and to the fractional contribution of increased large isolation failures (due to liner breach) to LERF.

This approach is based on the information and approaches contained in the previously mentioned studies. Furthermore:

Consistent with the other industry containment leak risk assessments, the PTN assessment uses LERF and delta LERF in accordance with the risk acceptance guidance of RG 1.174. Changes in population dose and conditional containment failure probability are also considered to show that defense-in-depth and the balance of prevention and mitigation is preserved.

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Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 6 of 82 Containment overpressure is not necessary to satisfy emergency core cooling system (ECCS) pump net positive suction head (NPSH) requirements at PTN, thus CDF is not affected by a change in containment leakage and LERF remains the relevant risk metric for the analysis of ILRT frequency extension.

The evaluation for PTN uses ground rules and methods to calculate changes in risk metrics that are similar to those used in Appendix H of EPRI Report No. 1009325, Revision 2-A, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals.

3.0 Ground Rules The following ground rules are used in the analysis:

The PTN Level 1 and Level 2 PRA models are used to conservatively estimate the impact of the proposed ILRT extension on population dose and LERF risk metrics.

The PTN Level 1 and Level 2 internal events PRA models provide representative results.

As discussed in Appendix A1, the acceptability of the PTN Internal Events, Internal Flooding PRA, and Fire PRA are consistent with the requirements of RG 1.200, Revision 2 (Reference 8.35), as is relevant to this ILRT interval extension. The results from screenings for other external hazards are also used.

The analysis includes a quantitative assessment of the contribution of external events (fire and internal flooding) to the risk impact assessment for extended ILRT intervals.

Although the seismic risk at PTN is low [Reference 8.17, Section 3.2.3], the ILRT risk impacts associated with seismic events are assessed using plant specific analyses to estimate the order of magnitude for contribution of seismic events to the overall impact of the changed interval.

It is assumed that the distribution of releases for each hazard (internal flood, fire, seismic, other external events) are consistent with the distribution calculated for the internal events PRA results. Sensitivity cases for different distributions are performed in order to verify this assumption. Additionally, it is noted that the impact from the ILRT extension on the percent increase in population dose would not be expected to change when accounting for the population dose contribution from external events.

Per EPRI Report No. 1009325, Revision 2-A, Section 4.2.2 (Reference 8.26), the order of preference for population dose information shall be plant-specific best estimate, Severe Accident Mitigation Alternative (SAMA) for license renewal, and scaling of a reference plant population dose. Therefore, PTN specific dose results from the Level 3 PRA results in Table F.1-4 of the environmental report for license extension (Reference 8.30) are used for this risk assessment.

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Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 7 of 82 Accident classes describing radionuclide release end states are defined consistent with EPRI methodology (Reference 8.2) and are summarized in Section 4.2.

The representative containment leakage for Class 1 sequences is 1 La. Class 3 accounts for increased leakage due to Type A inspection failures.

The representative containment leakage for Class 3a sequences is 10 La based on the previously approved methodology performed for Indian Point Unit 3 (References 8.8, 8.9). Class 3a represents intact containments with leakages somewhat larger than La as discussed in EPRI Report No. 1009325, Revision 2-A.

The representative containment leakage for Class 3b sequences is 100 La. based on the guidance provided in EPRI Report No. 1009325, Revision 2-A.

Class 3b can be very conservatively categorized as LERF based on the previously approved methodology (References 8.8, 8.9).

The impact on population doses from containment bypass scenarios is not altered by the proposed ILRT extension, but is accounted for in the EPRI methodology as a separate entry for comparison purposes. Since the containment bypass contribution to population dose is fixed, no changes on the conclusions from this analysis will result from this separate categorization.

The reduction in ILRT frequency does not impact the reliability of containment isolation valves to close in response to a containment isolation signal.

4.0 Inputs This section summarizes the general resources available as input (Section 4.1) and the plant specific resources required (Section 4.2).

4.1 General Resources Available Various industry studies on containment leakage risk assessment are briefly summarized here:

1. NUREG/CR-3539 (Reference 8.10)

2. NUREG/CR-4220 (Reference 8.11)

3. NUREG-1273 (Reference 8.12)

4. NUREG/CR-4330 (Reference 8.13)

5. EPRI TR-105189 (Reference 8.14)

6. NUREG-1493 (Reference 8.6)

7. EPRI TR-104285 (Reference 8.2)

8. NUREG-1150 (Reference 8.15) and NUREG/CR-4551 (Reference 8.7)

9. NEI Interim Guidance (Reference 8.3, Reference 8.20)

10. Calvert Cliffs Liner Corrosion Analysis (Reference 8.5)

11. EPRI Report No. 1009325, Revision 2-A, Appendix H (Reference 8.26)

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Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 8 of 82 The first study is applicable because it provides one basis for the threshold that could be used in the Level 2 PRA for the size of containment leakage that is considered significant and is to be included in the model. The second study is applicable because it provides a basis of the probability for significant pre-existing containment leakage at the time of a core damage accident. The third study is applicable because it is a subsequent study to NUREG/CR-4220 that undertook a more extensive evaluation of the same database. The fourth study provides an assessment of the impact of different containment leakage rates on plant risk. The fifth study provides an assessment of the impact on shutdown risk from ILRT test interval extension. The sixth study is the NRCs cost-benefit analysis of various alternative approaches regarding extending the test intervals and increasing the allowable leakage rates for containment integrated and local leak rate tests. The seventh study is an EPRI study of the impact of extending ILRT and LLRT test intervals on at-power public risk. The eighth study provides an ex-plant consequence analysis for a 50-mile radius surrounding a plant that can be used as the bases for the consequence analysis of the ILRT interval extension for PTN when plant-specific information is not available. The ninth study includes the NEI recommended methodology (promulgated in two letters) for evaluating the risk associated with obtaining a one-time extension of the ILRT interval. The tenth study addresses the impact of age-related degradation of the containment liners on ILRT evaluations. Finally, the eleventh study builds on the previous work and includes a recommended methodology and template for evaluating the risk associated with a permanent 15-year extension of the ILRT interval.

4.1.1 NUREG/CR-3539 (Reference 8.10)

Oak Ridge National Laboratory (ORNL) documented a study of the impact of containment leak rates on public risk in NUREG/CR-3539. This study uses information from WASH-1400 (Reference 8.16) as the basis for its risk sensitivity calculations. ORNL concluded that the impact of leakage rates on LWR accident risks is relatively small.

4.1.2 NUREG/CR-4220 (Reference 8.11)

NUREG/CR-4220 is a study performed by Pacific Northwest Laboratories for the NRC in 1985.

The study reviewed over two thousand LERs, ILRT reports and other related records to calculate the unavailability of containment due to leakage.

4.1.3 NUREG-1273 (Reference 8.12)

A subsequent NRC study, NUREG-1273, performed a more extensive evaluation of the NUREG/CR-4220 database. This assessment noted that about one-third of the reported events were leakages that were immediately detected and corrected. In addition, this study noted that local leak rate tests can detect essentially all potential degradations of the containment isolation system.

4.1.4 NUREG/CR-4330 (Reference 8.13)

NUREG/CR-4330 is a study that examined the risk impacts associated with increasing the allowable containment leakage rates. The details of this report have no direct impact on the modeling approach of the ILRT test interval extension, as NUREG/CR-4330 focuses on leakage rate and the ILRT test interval extension study focuses on the frequency of testing intervals.

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Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 9 of 82 However, the general conclusions of NUREG/CR-4330 are consistent with NUREG/CR-3539 and other similar containment leakage risk studies: the effect of containment leakage on overall accident risk is small since risk is dominated by accident sequences that result in failure or bypass of containment.

4.1.5 EPRI TR-105189 (Reference 8.14)

The EPRI study TR-105189 is useful to the ILRT test interval extension risk assessment because it provides insight regarding the impact of containment testing on shutdown risk. This study contains a quantitative evaluation (using the EPRI ORAM software) for two reference plants (a BWR-4 and a PWR) of the impact of extending ILRT and LLRT test intervals on shutdown risk.

The conclusion from the study is that a small but measurable safety benefit is realized from extending the test intervals.

4.1.6 NUREG-1493 (Reference 8.6)

NUREG-1493 is the NRCs cost-benefit analysis for proposed alternatives to reduce containment leakage testing intervals and/or relax allowable leakage rates. The NRC conclusions are consistent with other similar containment leakage risk studies:

Reduction in ILRT frequency from three per ten years to one per twenty years results in an imperceptible increase in risk.

Given the insensitivity of risk to the containment leak rate and the small fraction of leak paths detected solely by Type A testing, increasing the interval between integrated leak rate tests is possible with minimal impact on public risk.

4.1.7 EPRI TR-104285 (Reference 8.2)

Extending the risk assessment impact beyond shutdown (the earlier EPRI TR-105189 study), the EPRI TR-104285 study is a quantitative evaluation of the impact of extending ILRT and LLRT test intervals on at-power public risk. This study combined IPE Level 2 models with NUREG-1150 Level 3 population dose models to perform the analysis. The study also used the approach of NUREG-1493 in calculating the increase in pre-existing leakage probability due to extending the ILRT and LLRT test intervals.

EPRI TR-104285 uses a simplified Containment Event Tree to subdivide representative core damage frequencies into eight classes of containment response to a core damage accident:

1. Containment intact and isolated

2. Containment isolation failures dependent upon the core damage accident

3. Type A (ILRT) related containment isolation failures

4. Type B (LLRT) related containment isolation failures

5. Type C (LLRT) related containment isolation failures

6. Other penetration related containment isolation failures

7. Containment failures due to core damage accident phenomena

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Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 10 of 82 Consistent with the other containment leakage risk assessment studies, this study concluded:

[T]he proposed CLRT (containment leak rate tests) frequency changes would have a minimal safety impact. The change in risk determined by the analyses is small in both absolute and relative terms. For example, for the PWR analyzed, the change is about 0.04 person-rem per year 4.1.8 NUREG-1150 (Reference 8.15) and NUREG/CR 4551 (Reference 8.7)

NUREG-1150 and the technical basis, NUREG/CR-4551, provide an ex-plant consequence analysis for a spectrum of accidents including a severe accident with the containment remaining intact (i.e., Tech Spec leakage). This ex-plant consequence analysis is calculated for the 50-mile radial area surrounding Surry. The ex-plant calculation can be delineated to total person-rem for each identified Accident Progression Bin (APB) from NUREG/CR-4551. However, these references were not used in this analysis. PTN plant-specific dose results from the Level 3 PRA results in the environmental report for license extension (Reference 8.30) are used for this risk assessment.

4.1.9 NEI Interim Guidance for Performing Risk Impact Assessments In Support of One-Time Extensions for Containment Integrated Leakage Rate Test Surveillance Intervals (Reference 8.3, Reference 8.20)

The guidance provided in this document builds on the EPRI risk impact assessment methodology (Reference 8.2) and the NRC performance-based containment leakage test program (Reference 8.6), and considers approaches utilized in various submittals, including Indian Point 3 (and associated NRC SER) and Crystal River.

4.1.10 Calvert Cliffs Response to Request for Additional Information Concerning the License Amendment for a One-Time Integrated Leakage Rate Test Extension (Reference 8.5)

This submittal to the NRC describes a method for determining the change in likelihood, due to extending the ILRT, of detecting liner corrosion, and the corresponding change in risk. The methodology was developed for Calvert Cliffs in response to a request for additional information regarding how the potential leakage due to age-related degradation mechanisms were factored into the risk assessment for the ILRT one-time extension. The Calvert Cliffs analysis was performed for a concrete cylinder, dome and a concrete basemat, each with a steel liner.

Licensees may consider approved LARs for one-time extensions involving containment types similar to their facility. The PTN assessment has addressed the plant-specific differences from the Calvert Cliffs design, and how the Calvert Cliffs methodology was adapted to address the specific design features.

4.1.11 EPRI Report No. 1009325, Revision 2-A, Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals (Reference 8.26)

This report provides a generally applicable assessment of the risk involved in extension of ILRT test intervals to permanent 15-year intervals. Appendix H of this document provides guidance for performing plant specific supplemental risk impact assessments and builds on the previous EPRI Page l 10

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 11 of 82 risk impact assessment methodology (Reference 8.2) and the NRC performance-based containment leakage test program (Reference 8.6), and considers approaches utilized in various submittals, including Indian Point 3 (and associated NRC SER) and Crystal River.

The approach included in this guidance document is used in the PTN assessments to determine the estimated increase in risk associated with the ILRT extension. This document includes the bases for the values assigned in determining the probability of leakage for the EPRI Classes 3a and 3b scenarios in this analysis as described in Section 5.

4.2 Plant Specific Inputs The plant specific information used to perform the PTN ILRT Extension Risk Assessments include the following:

Level 1 Model results Level 2 Model results Release category definitions used in the Level 2 Model ILRT results to demonstrate adequacy of the administrative and hardware issues Containment design and fragility data 4.2.1 Level 1 Model The Level 1 model is a linked fault tree model, and was quantified for internal and external initiating events as shown in Table 4.2-1 (Internal Events, Internal Fire, Internal Flood, Seismic Events, and Other External Events) with the total Core Damage Frequency (CDF) = 6.42E-05/yr for Unit 3 and 6.59E-05/yr for Unit 4 (Reference 8.18). The Internal Events, Internal Flood, and Internal Fire results are from the Revision 11 model, which includes fire PRA changes for the NFPA-805 project (Reference 8.44). Credit for FLEX mitigating strategies involving portable equipment is not included in the Revision 11 model.

Table 4.2-1: PTN CDF Results by Hazard Model Unit 3 CDF Unit 4 CDF Source Internal Events 2.91E-07 2.91E-07 Peer reviewed plant-specific PRA model Internal Flood 1.37E-08 1.27E-08 Peer reviewed plant-specific PRA model Internal Fire 6.22E-05 6.39E-05 Peer reviewed plant-specific PRA model Seismic 6.98E-07 6.98E-07 Alternate approach (Reference 8.24)

Other External 1.00E-06 1.00E-06 Assumption per Reference 8.18 Hazards Total 6.42E-05 6.59E-05 Page l 11

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 12 of 82 4.2.2 Level 2 Model The Level 2 PRA model of record was developed to calculate LERF. The following table summarizes the pertinent PTN results for LERF in terms of the initiating hazard (Reference 8.18). The Internal Events, Internal Flood, and Internal Fire results are from the Revision 11 model, which includes fire PRA changes for the NFPA -805 project (Reference 8.44).

Table 4.2-2: PTN LERF Results by Hazard Model Unit 3 LERF Unit 4 LERF Source Internal Events 1.11E-08 1.11E-08 Peer reviewed plant-specific PRA model Internal Flood 5.44E-11 1.44E-10 Peer reviewed plant-specific PRA model Internal Fire 1.25E-06 1.49E-06 Peer reviewed plant-specific PRA model Seismic 2.66E-08 2.66E-08 Assumption per Reference 8.18 Other External 3.81E-08 3.81E-08 Assumption per Reference 8.18 Hazards Total 1.33E-06 1.57E-06 Although full Level 2 quantification results are not maintained, the model of record contains the gates needed to quantify the EPRI release classes with minor modification. To define the frequency of the EPRI release classes, each containment event tree (CET) sequence is assigned to one of the EPRI release classes in the PTN model. The model is then quantified to define the frequency for the release classes. These additional release class quantifications are used to define the distribution of CDF between the EPRI release categories as shown in Table 4.2-3 (Reference 8.18).

Table 4.2-3: EPRI Release Class Distributions EPRI Release Class Percent of CDF Unit 3 Unit 4 1 41.67% 36.88%

2 0.005% 0.005%

7 52.16% 57.14%

8 6.16% 5.97%

Applying the release class distributions of Table 4.2-3 to the CDF results from Table 4.2-1 yields the hazard specific release class distributions shown in Tables 4.2-4 and 4.2-5 (see Section 3.0 Ground Rules).

Table 4.2-4: Unit 3 EPRI Release Class Frequencies by Hazard EPRI Release Frequency by Hazard Type (per year)

Class Distribution (% of CDF) Internal Internal Internal Seismic Other Events Flood Fire CDF 100.00% 2.91E-07 1.37E-08 6.22E-05 6.98E-07 1.00E-06 1 41.67% 1.21E-07 5.71E-09 2.59E-05 2.91E-07 4.17E-07 2 0.00% 1.39E-11 6.53E-13 2.97E-09 3.33E-11 4.77E-11 Page l 12

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 13 of 82 Table 4.2-4: Unit 3 EPRI Release Class Frequencies by Hazard EPRI Release Frequency by Hazard Type (per year)

Class Distribution (% of CDF) Internal Internal Internal Seismic Other Events Flood Fire 7 52.16% 1.52E-07 7.15E-09 3.24E-05 3.64E-07 5.22E-07 8 6.16% 1.79E-08 8.44E-10 3.83E-06 4.30E-08 6.16E-08 Table 4.2-5: Unit 4 EPRI Release Class Frequencies by Hazard EPRI Release Frequency by Hazard Type (per year)

Class Distribution (% of CDF)

Internal Internal Internal Seismic Other Events Flood Fire CDF 100.00% 2.91E-07 1.27E-08 6.39E-05 6.98E-07 1.00E-06 1 36.88% 1.07E-07 4.68E-09 2.36E-05 2.57E-07 3.69E-07 2 0.005% 1.59E-11 6.92E-13 3.48E-09 3.80E-11 5.45E-11 7 57.14% 1.66E-07 7.26E-09 3.65E-05 3.99E-07 5.71E-07 8 5.97% 1.74E-08 7.58E-10 3.82E-06 4.17E-08 5.97E-08 4.2.3 Population Dose Calculations The baseline population dose for PTN is given in Table 4.2-6a. As discussed in EPRI 1009325, Revision 2-A, Section 4.2.2 (Reference 8.26), the order of preference for population dose information shall be plant-specific best estimate, Severe Accident Mitigation Alternative (SAMA) for license renewal, and scaling of a reference plant population dose. Therefore, PTN plant specific dose results based on the Level 3 PRA results in Table F.1-4 of the environmental report for license extension (Reference 8.30) are used for this risk assessment. This data is consistent with that submitted in the one-time extension request for PTN ILRT (Reference 8.31).

Table 4.2-6a: PTN Population Dose PTN CET End State Description Population Dose Dose EPRI Class

[-] [-] [SV] [Person - REM] [-]

NCF No containment failure - 1.69E+04 1 No CCI, early containment failure, in-vessel fission product release mitigated, D1-L (iso) leak, containment isolation failure 5.63E+03 5.63E+05 2 No CCI, early containment failure, in-vessel fission product release mitigated, D1-R (iso) rupture, containment isolation failure 8.20E+03 8.20E+05 2 No CCl, early containment failure, in-vessel fission product release not mitigated, leak, containment isolation D2-L (iso) failure 3.04E+04 3.04E+06 2 Page l 13

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 14 of 82 PTN CET End State Description Population Dose Dose EPRI Class

[-] [-] [SV] [Person - REM] [-]

No CCI, early containment failure, in-vessel fission product release not mitigated, rupture, containment isolation D2-R (iso) failure 2.87E+04 2.87E+06 2 Significant CCI occurs, early containment failure, in- and ex-vessel fission product release mitigated, leak, E1-L (iso) containment isolation failure 5.70E+03 5.70E+05 2 Significant CCI occurs, early containment failure, in- and ex-vessel fission product release mitigated, E1-R (iso) rupture, containment isolation failure 8.38E+03 8.38E+05 2 Significant CCl occurs, early containment failure, ex-vessel fission product release mitigated by overlying pool, in-vessel fission product release not mitigated, leak, containment E2-L (iso) isolation failure 3.20E+04 3.20E+06 2 Significant CCI occurs, early containment failure, ex-vessel fission product release mitigated by overlying pool, in-vessel fission product release not mitigated, rupture, containment E2-R (iso) isolation failure 3.05E+04 3.05E+06 2 Recovered in-vessel, late containment failure, in-vessel fission product release A1 mitigated 5.46E+01 5.46E+03 7 Recovered in-vessel, late containment failure, in-vessel fission product release A2 not mitigated 2.48E+04 2.48E+06 7 Recovered ex-vessel, late containment failure, in-vessel fission product release B1 mitigated 8.46E+02 8.46E+04 7 Recovered ex-vessel, late containment failure, in-vessel fission product release B2-L not mitigated, leak 2.64E+04 2.64E+06 7 Recovered ex-vessel, late containment failure, in-vessel fission product release B2-R not mitigated, rupture 3.73E+04 3.73E+06 7 No CCI, late containment failure, in-vessel fission product release mitigated B3-L by sprays, leak 8.46E+02 8.46E+04 7 No CCI, late containment failure, in-vessel fission product release mitigated B3-R by sprays, rupture 1.92E+03 1.92E+05 7 No CCI, late containment failure, in-vessel fission product release not B4-L mitigated, leak 2.64E+04 2.64E+06 7 No CCI, late containment failure, in-vessel fission product release not B4-R mitigated, rupture 3.78E+04 3.78E+06 7 Page l 14

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 15 of 82 PTN CET End State Description Population Dose Dose EPRI Class

[-] [-] [SV] [Person - REM] [-]

No CCI, late containment failure, in-vessel fission product release mitigated B5-L by sprays, leak 1.54E+03 1.54E+05 7 No CCI, late containment failure, in-vessel fission product release mitigated B5-R by sprays, rupture 3.28E+03 3.28E+05 7 No CCI, late containment failure, in-vessel and late fission product release B6-L not mitigated, leak 2.29E+04 2.29E+06 7 No CCl, late containment failure, in-vessel and late fission product release B6-R not mitigated, rupture 2.74E+04 2.74E+06 7 CCI occurs, late containment failure, ex-vessel fission product release mitigated by overlying pool, in-vessel release C1-L mitigated by sprays, leak 8.46E+02 8.46E+04 7 CCI occurs, late containment failure, ex-vessel fission product release mitigated by overlying pool, in-vessel release C1-R mitigated by sprays, rupture 1.92E+03 1.92E+05 7 CCI occurs, late containment failure, ex-vessel fission product release mitigated by overlying pool, in-vessel release not C2-L mitigated, leak 2.65E+04 2.65E+06 7 CCI occurs, late containment failure, ex-vessel fission product release mitigated by overlying pool, in-vessel release not C2-R mitigated, rupture 3.79E+04 3.79E+06 7 Significant CCI occurs, late containment failure, in- and ex-vessel fission product C3-L release mitigated by sprays, leak 8.46E+02 8.46E+04 7 Significant CCI occurs, late containment failure, in- and ex-vessel fission product C3-R release mitigated by sprays, rupture 1.92E+03 1.92E+05 7 Significant CCI occurs, late containment failure, in- and ex-vessel fission product C4-L release not mitigated, leak 2.65E+04 2.65E+06 7 Significant CCI occurs, late containment failure, in- and ex-vessel fission product C4-R release not mitigated, rupture 3.79E+04 3.79E+06 7 Moderate CCI occurs, late containment failure, in- and ex-vessel fission product C5-L release mitigated by sprays, leak 1.54E+03 1.54E+05 7 Moderate CCI occurs, late containment failure, in- and ex-vessel fission product C5-R release mitigated by sprays, rupture 3.28E+03 3.28E+05 7 Moderate CCI occurs, late containment failure, in- and ex-vessel fission product C6-L release not mitigated, leak 2.29E+04 2.29E+06 7 Moderate CCI occurs, late containment failure, in- and ex-vessel fission product C6-R release not mitigated, rupture 3.27E+04 3.27E+06 7 Page l 15

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 16 of 82 PTN CET End State Description Population Dose Dose EPRI Class

[-] [-] [SV] [Person - REM] [-]

No CCI, early containment failure, in-vessel fission product release mitigated, leak, leak, non-containment isolation D1-L (non-iso) failure 5.63E+03 5.63E+05 7 No CCI, early containment failure, in-vessel fission product release mitigated, rupture, non-containment isolation D1-R (non-iso) failure 8.20E+03 8.20E+05 7 No CCl, early containment failure, in-vessel fission product release not mitigated, leak, non-containment D2-L (non-iso) isolation failure 3.04E+04 3.04E+06 7 No CCI, early containment failure, in-vessel fission product release not mitigated, rupture, non-containment D2-R (non-iso) isolation failure 2.87E+04 2.87E+06 7 No CCI, early containment failure, in-vessel and late fission product release D3-L mitigated, leak 1.63E+04 1.63E+06 7 No CCI, early containment failure, in-vessel and late fission product release D3-R mitigated, rupture 1.89E+04 1.89E+06 7 No CCl, early containment failure, in-vessel and late fission product release D4-L not mitigated, leak 2.73E+04 2.73E+06 7 No CCI, early containment failure, in-vessel and late fission product release D4-R not mitigated, rupture 2.69E+04 2.69E+06 7 Significant CCI occurs, early containment failure, in- and ex-vessel fission product release mitigated, leak, E1-L (non-iso) non-containment isolation failure 5.70E+03 5.70E+05 7 Significant CCI occurs, early containment failure, in- and ex-vessel fission product release mitigated, rupture, non-containment isolation E1-R (non-iso) failure 8.38E+03 8.38E+05 7 Significant CCl occurs, early containment failure, ex-vessel fission product release mitigated by overlying pool, in-vessel fission product release not mitigated, leak, non-containment E2-L (non-iso) isolation failure 3.20E+04 3.20E+06 7 Significant CCI occurs, early containment failure, ex-vessel fission product release mitigated by overlying pool, in-vessel fission product release not mitigated, rupture, non-containment E2-R (non-iso) isolation failure 3.05E+04 3.05E+06 7 Significant CC[ occurs, early containment failure, in- and ex-vessel fission product release mitigated by E3-L sprays, leak 5.70E+03 5.70E+05 7 Page l 16

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 17 of 82 PTN CET End State Description Population Dose Dose EPRI Class

[-] [-] [SV] [Person - REM] [-]

Significant CCI occurs, early containment failure, in- and ex-vessel fission product release mitigated by E3-R sprays, rupture 8.38E+03 8.38E+05 7 Significant CCI occurs, early containment failure, fission product E4-L release not mitigated, leak 3.20E+04 3.20E+06 7 Significant CCI occurs, early containment failure, fission product E4-R release not mitigated, rupture 3.05E+04 3.05E+06 7 Moderate CCI occurs, early containment failure, in- and ex-vessel fission product release mitigated by sprays, no late E5-L fission product release, leak 1.63E+04 1.63E+06 7 Moderate CCI occurs, early containment failure, in- and ex-vessel fission product release mitigated by sprays, no late E5-R fission product release, rupture 1.89E+04 1.89E+06 7 Moderate CCI occurs, early containment failure, ex-vessel and late fission product E6-L release not mitigated, leak 2.83E+04 2.83E+06 7 Moderate CCI occurs, early containment failure, ex-vessel and late fission product E6-R release not mitigated, rupture 2.83E+04 2.83E+06 7 BP-V Containment bypass, ISLOCA sequences 4.46E+04 4.46E+06 8 BY-SGTR Containment bypass, SGTR sequences 8.07E+03 8.07E+05 8 Note that population dose for the NCF (EPRI Class 1) sequence is estimated in the one-time license extension (Reference 8.31). The dose was estimated to be 1/100 of the dose for EPRI Class 7. This was calculated by totaling the Class 7 population dose frequencies and dividing by the sum of the Class 7 release frequencies, then dividing by 100.

The subsequent license renewal application shows an increase in the permanent resident population within a 50 mile radius of PTN between 2000 and 2010 as well as an increase in power level. Therefore, the population dose is adjusted for population growth and power level increase. EPRI Report No. 1009325 (Reference 8.26) provides a method for approximating population dose based a reference plant if it is corrected for allowable containment leak rate (La),

reactor power level, and population density. Using this method, the population dose data in Table 4.2-6a is adjusted to account for population increase and power uprate at PTN.

The total population within a 50-mile radius of PTN estimated in 2053 is 4,916,069 people (Reference 8.54, pg. 3-313). The population of the area at the time of the initial license renewal application and at which the Level 3 population dose date given in Table 4.2-6a was generated was 3,952,697 people (Reference 8.30, Table F.1-3). The increase in population is taken into account by applying a population dose factor:

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Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 18 of 82 Population density adjustment FPopulation =

2053 PTN Population within 50 miles = 4,916,069 people 2025 PTN Population within 50 miles = 3,952,697 people FPopulation = 4,916,069/3,952,697 = 1.244 Note that at the time of the Level 3 analysis, La at PTN was 0.25 wt%/day (Reference 8.31),

while currently La at PTN is 0.2 wt%/day (References 8.53). If the EPRI report scaling methodology was applied, this reduction in La would reduce the PTN intact containment dose rates by a factor of 0.25/0.2=1.25. Conservatively, this adjustment due to reduced La at PTN is not credited.

Power level adjustment:

The power level at the time of the initial license renewal application and at which the Level 3 population dose date given in Table 4.2-6a was generated was 2300 MWt (Reference 8.57). The subsequent license renewal application shows an increase in power to 2644 MWt (Reference 8.38).

Current rated power level of PTN MWt FPower SAMA rated power level for PTN MWt FPower = 2644 MWt / 2300 MWt FPower = 1.150 The maximum PTN population dose for each EPRI release class in Table 4.2-6a is used to define the total population dose for each release category and then adjusted by the population and power level factors above as shown in Table 4.2-6b.

Table 4.2-6b: PTN Adjusted Population Dose Applicable Total Dose from Adjusted Population Power Level EPRI Release Description Table 4.2-6a Dose Adjustment Adjustment Category (person-rem) (person-rem)

No Containment 1 1.69E+04 1.244 1.150 2.42E+04 Failure Large Isolation 2 Failures (Failure to 3.20E+06 1.244 1.150 4.58E+06 Close)

Failures Induced by 7 Phenomena (Early 3.79E+06 1.244 1.150 5.42E+06 and Late)

Bypass (Interfacing 8 4.46E+06 1.244 1.150 6.38E+06 System LOCA)

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Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 19 of 82 4.2.4 Type A Tests The three most recent Type A ILRTs at PTN Unit 3 and Unit 4 have been successful, so the current ILRT interval requirement is 10 years (Reference 8.42). A one-time extension of the ILRT interval to 15 years was previously approved (Reference 8.56). The interval between the prior three tests as shown in the ILRT Test reports (Reference 8.42) and described in the Subsequent License Renewal Application (Reference 8.38) for Unit 3 and Unit 4 demonstrates prior successful performance at the extended interval. Per EPRI Report 1009325 (Reference 8.29), the two most recent Type A tests are considered to demonstrate adequacy of the administrative and hardware performance at PTN.

4.2.5 Release Category Definitions Table 4.2-7 defines the accident classes used in the ILRT extension evaluation, which is consistent with the EPRI methodology (Reference 8.2). These containment failure classifications are used in this analysis to determine the risk impact of extending the Containment Type A test interval as described in Section 6 of this report.

Table 4.2-7: EPRI Containment Failure Classification (Reference 8.2)

Class Description 1 Containment remains intact including accident sequences that do not lead to containment failure in the long term. The release of fission products (and attendant consequences) is determined by the maximum allowable leakage rate values La, under Appendix J for that plant.

2 Containment isolation failures (as reported in the IPEs) include those accidents in which there is a failure to isolate the containment.

3 Independent (or random) isolation failures include those accidents in which the pre-existing isolation failure to seal (i.e., provide a leak-tight containment) is not dependent on the sequence in progress.

4 Independent (or random) isolation failures include those accidents in which the pre-existing isolation failure to seal is not dependent on the sequence in progress. This class is similar to Class 3 isolation failures, but is applicable to sequences involving Type B tests and their potential failures. These are the Type B-tested components that have isolated but exhibit excessive leakage.

5 Independent (or random) isolation failures include those accidents in which the pre-existing isolation failure to seal is not dependent on the sequence in progress. This class is similar to Class 4 isolation failures, but is applicable to sequences involving Type C tests and their potential failures.

6 Containment isolation failures include those leak paths covered in the plant test and maintenance requirements or verified per inservice inspection and testing (ISI/IST) program.

7 Accidents involving containment failure induced by severe accident phenomena. Changes in Appendix J testing requirements do not impact these accidents.

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Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 20 of 82 Table 4.2-7: EPRI Containment Failure Classification (Reference 8.2)

Class Description 8 Accidents in which the containment is bypassed (either as an initial condition or induced by phenomena) are included in Class 8. Changes in Appendix J testing requirements do not impact these accidents.

4.3 Impact of Extension on Detection of Component Failures That Lead to Leakage (Small and Large)

The ILRT can detect a number of component failures such as liner breach, failure of certain bellow arrangements and failure of some sealing surfaces, which can lead to leakage. The proposed ILRT test interval extension may influence the conditional probability of detecting these types of failures. To ensure that this effect is properly accounted for, the EPRI Class 3 accident class as defined in Table 4.2-7, it is divided into two sub-classes, Class 3a and Class 3b, representing small and large leakage failures, respectively.

The probability of the EPRI Class 3a and 3b failures is determined consistent with the EPRI Guidance. For Class 3a, the probability is based on the maximum likelihood estimate of failure (arithmetic average) from the available data (i.e., 2 small failures in 217 tests leads to 2/217=0.0092). For Class 3b, Jefferys non-informative prior distribution is assumed for no large failures in 217 tests (i.e., 0.5 / (217+1) = 0.0023).

In a follow on letter (Reference 8.20) to their ILRT guidance document (Reference 8.3), NEI issued additional information concerning the potential that the calculated delta LERF values for several plants may fall above the very small change guidelines of the NRC RG 1.174. This additional NEI information includes a discussion of conservatisms in the quantitative guidance for delta LERF. NEI describes ways to demonstrate that, using plant specific calculations, the delta LERF is smaller than that calculated by the simplified method.

The supplemental information states:

The methodology employed for determining LERF (Class 3b frequency) involves conservatively multiplying the CDF by the failure probability for this class (3b) of accident. This was done for simplicity and to maintain conservatism. However, some plant specific accident classes leading to core damage are likely to include individual sequences that either may already (independently) cause a LERF or could never cause a LERF, and are thus not associated with a postulated large Type A containment leakage path (LERF). These contributors can be removed from Class 3b in the evaluation of LERF by multiplying the Class 3b probability by only that portion of CDF that may be impacted by type A leakage.

The application of this additional guidance to the analysis for PTN, as detailed in Section 5, involves the following:

The Class 2 and Class 8 sequences are subtracted from the CDF that is applied to Class 3b.

To be consistent, the same change is made to the Class 3a CDF, even though these events are not considered LERF. Class 2 and Class 8 events refer to sequences with either large Page l 20

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 21 of 82 preexisting containment isolation failures or containment bypass events. These sequences are already considered to contribute to LERF in the PTN Level 2 PRA analyses.

Class 1 accident sequences may involve availability and or successful operation of containment sprays. It could be assumed that, for calculation of the Class 3b and 3a frequencies, the fraction of the Class 1 CDF associated with successful operation of containment sprays can also be subtracted. Containment sprays are conservatively not subtracted in the ILRT extension analysis for PTN.

Consistent with the NEI Guidance (Reference 8.3), the change in the leak detection probability can be estimated by comparing the average time that a leak could exist without detection. For example, the average time that a leak could go undetected with a three year test interval is 1.5 years (3 yr/2), and the average time that a leak could exist without detection for a ten year interval is five years (10 yr/2). This change would lead to a non-detection probability that is a factor of 3.33 (5.0/1.5) higher for the probability of a leak that is detectable only by ILRT testing. An extension of the ILRT interval to fifteen years can be estimated to lead to about a factor of 5.0 (7.5/1.5) increase in the non-detection probability of a leak compared to a three year interval.

It should be noted that using the methodology discussed above is very conservative compared to previous submittals (e.g., the IP3 request for a one-time ILRT extension that was approved by the NRC (Reference 8.9)) because it does not factor in the possibility that the failures could be detected by other tests (e.g., the Type B local leak rate tests that will still occur.) Eliminating this possibility conservatively over-estimates the factor increases attributable to the ILRT extension.

4.4 Impact of Extension on Detection of Steel Liner Corrosion that Leads to Leakage An estimate of the likelihood and risk implications of corrosion-induced leakage of the steel liners occurring and going undetected during the extended test interval is evaluated using the methodology from the Calvert Cliffs liner corrosion analysis (Reference 8.5). The Calvert Cliffs analysis was performed for a concrete cylinder, dome and a concrete basemat, each with a steel liner. PTN has a similar type of containment (Reference 8.32).

The following approach is used to determine the change in likelihood, due to extending the ILRT, of detecting corrosion of a containment steel liner. It should be noted that this computation is being applied to provide an upper bound approach to quantify corrosion induced risk. The Calvert Cliffs corrosion likelihood methodology is then used to determine the resulting change in risk. Consistent with the Calvert Cliffs analysis, the following issues are addressed:

Differences between the containment basemat and the upper containment (cylinder and dome regions in Calvert Cliffs evaluation)

The historical steel liner flaw likelihood due to concealed corrosion The impact of aging Page l 21

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 22 of 82 The corrosion leakage dependency on containment pressure The likelihood that visual inspections will be effective at detecting a flaw 4.4.1 Assumptions Consistent with the Calvert Cliffs analysis, a half failure is assumed for basemat concealed liner corrosion due to the lack of identified failures. (See Table 4.4-1, Step 1.)

The two corrosion events used to estimate the liner flaw probability in the Calvert Cliffs analysis are assumed to be applicable to this PTN containment analysis. These events, one at North Anna Unit 2 (September 1999) and one at Brunswick Unit 2 (April 1999),

were initiated from the nonvisible (backside) portion of the containment liner.

The Calvert Cliffs analysis used the estimated historical liner flaw probability of 5.5 years to reflect the years since September 1996 when 10 CFR 50.55a started requiring visual inspection. Additional success data was not used to limit the aging impact of this corrosion issue, even though inspections were being performed prior to this date. Since the time of the Calvert Cliffs submittal, two additional relevant liner corrosion events involving concealed corrosion (corrosion initiated on the inaccessible liner surface) were observed and are considered in the corrosion risk assessment. These events occurred at Beaver Valley Unit 1 (June 2009) and D.C. Cook Unit 2 (March 2001) (Reference 8.27 and Reference 8.28, respectively). Consistent with the addition of the two observed events, the historical liner flaw probability was established by incrementing the flaw observation time by 7.75 years (March 2002 to June 2009). This re-evaluation resulted in a reduction of the historical liner flaw likelihood to 4.3E-03/year ((2+2) / [70 * (5.5 +

7.75)] = 4.3E-03/year). This value is smaller than the value of 5.2E-03 which is used in the Calvert Cliffs analysis. The conservatively high value of 5.2E-03 will be used in this PTN report to remain consistent with the Calvert Cliffs analysis.

In the Calvert Cliffs analysis, the likelihood of the containment atmosphere reaching the outside atmosphere given that a liner flaw exists was estimated as 1.1% for the cylinder and dome and 0.11% (10% of the cylinder failure probability) for the basemat. These values were determined from an assessment of the probability versus containment pressure, and the selected values are consistent with a pressure that corresponds to the ILRT target pressure of 37 psig. For PTN, the containment failure probabilities are less than these values at the PTN ILRT target pressure of 55 psig (Reference 8.53, 8.18).

Conservative probabilities of 1% for the cylinder and dome and 0.1% for the basemat are used in this analysis, and sensitivity studies are included that increase and decrease the probabilities by an order of magnitude. (See Table 4.4-1, Step 4.)

Consistent with the Calvert Cliffs analysis, the likelihood of leakage escape (due to crack formation) in the basemat region is considered to be less likely than the containment cylinder and dome region. (See Table 4.4-1, Step 4.)

Consistent with the Calvert Cliffs analysis, a 5% visual inspection detection failure likelihood given the flaw is visible and a total detection failure likelihood of 10% is used.

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Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 23 of 82 To date, all liner corrosion events have been detected through visual inspection. (See Table 4.4-1, Step 5.) Sensitivity studies are included that evaluate total detection failure likelihood of 5% and 15%, respectively.

Consistent with the Calvert Cliffs analysis, all non-detectable containment failures are assumed to result in early releases. This approach avoids a detailed analysis of containment failure timing and operator recovery actions.

4.4.2 Analysis Table 4.4-1: Steel Liner Corrosion Analysis Step Description Containment Walls Containment Basemat Historical Steel Liner Flaw 1 Likelihood Events: 2 Events 0 (assume 0.5 failure)

Failure Data: 2/(70*5.5) = 5.2E-3 0.5/(70

5.5) = 1.3E-3 Age-Adjusted Steel Liner Flaw Year Failure Rate Year Failure Rate 2 1 2.1E-3 1 5.0E-4 Likelihood avg. 5-10 5.2E-3 avg. 5-10 1.3E-3 15 1.4E-2 15 3.5E-3 15 year average = 6.27E-3 15-year average = 1.57E-3 0.71% (1 to 3 years) 0.18% (1 to 3 years)

Flaw Likelihood at 3, 10, and 15 3 4.06% (1 to 10 years) 1.02% (1 to 10 years) years 9.40% (1 to 15 years) 2.35% (1 to 15 years)

Likelihood of Breach in 4 1% 0.1%

Containment Given Steel Liner Flaw 5 Visual Inspection Detection 10% 100%

Failure Likelihood 0.00071% 0.00018%

0.71% *1%

10% 0.18%

0.1%

100%

Likelihood of Non-Detected 0.0041% 0.0010%

6 Containment Leakage (Steps 3*4*5) 4.1%

1%

10% 1.0%

0.1%

100%

0.0094% 0.0024%

9.4%

1%

10% 2.4%

0.1%

100%

The total likelihood of the corrosion-induced, non-detected containment leakage is the sum of Step 6 for the leakages for the upper containment and the containment basemat as summarized below for PTN.

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Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 24 of 82 Total Likelihood of Non-Detected Containment Leakage Due To Corrosion for PTN:

At 3 years: 0.00071% + 0.00018% = 0.00089%

At 10 years: 0.0041% + 0.0010% = 0.0051%

At 15 years: 0.0094% + 0.0024% = 0.0118%

The above factors are applied to those core damage accidents that are not already independently LERF or that could never result in LERF. For example, the Unit 3 three in ten year case is calculated as follows:

Per Table 5-2a, the EPRI Unit 3 Class 3b frequency is 6.28E-10/yr.

As discussed in Section 5.1, the PTN Unit 3 CDF associated with accidents that are not independently LERF or could never result in LERF is CDF - Class2 Frequency - Class 8 Frequency = 2.91E 1.39E 1.79E-08 = 2.73E-07/yr.

The increase in the base case Class 3b frequency due to the corrosion-induced concealed flaw issue is calculated as 0.00089%*2.73E-07/yr = 2.43E-12/yr, where 0.00089% was previously shown above to be the cumulative likelihood of non-detected containment leakage due to corrosion at three years.

The three in ten year Class 3b frequency including the corrosion-induced concealed flaw issue is then calculated as 6.31E-10/yr.

5.0 Results The application of the approach based on the guidance contained in EPRI Report No. 1009325, Revision 2-A, Appendix H, EPRI-TR-104285 (Reference 8.2) and previous risk assessment submittals on this subject (References 8.5, 8.8, 8.21, 8.22) have led to the following results. The results are displayed according to the eight accident classes defined in the EPRI report. Table 5-1 lists these accident classes.

The analysis performed examined PTN specific accident sequences in which the containment remains intact or the containment is impaired. Specifically, the breakdown of the severe accidents contributing to risk is considered in the following manner:

Core damage sequences in which the containment remains intact initially and in the long term (EPRI TR-104285 Class 1 sequences).

Core damage sequences in which containment integrity is impaired due to random isolation failures of plant components other than those associated with Type B or Type C test components. For example, liner breach or bellows leakage. (EPRI TR-104285 Class 3 sequences).

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Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 25 of 82 Core damage sequences in which containment integrity is impaired due to containment isolation failures of pathways left opened following a plant post-maintenance test. For example, a valve failing to close following a valve stroke test. (EPRI TR-104285 Class 6 sequences). Consistent with the NEI Guidance, this class is not specifically examined since it will not significantly influence the results of this analysis.

Accident sequences involving containment bypassed (EPRI TR-104285 Class 8 sequences),

large containment isolation failures (EPRI TR-104285 Class 2 sequences), and small containment isolation failure-to-seal events (EPRI TR-104285 Class 4 and 5 sequences) are accounted for in this evaluation as part of the baseline risk profile. However, they are not affected by the ILRT frequency change.

Class 4 and 5 sequences are impacted by changes in Type B and C test intervals; therefore, changes in the Type A test interval do not impact these sequences.

Table 5-1: Accident Classes Accident Classes (Containment Release Description Type) 1 No Containment Failure 2 Large Isolation Failures (Failure to Close) 3a Small Isolation Failures (Liner Breach) 3b Large Isolation Failures (Liner Breach) 4 Small Isolation Failures (Failure to Seal-Type B) 5 Small Isolation Failures (Failure to Seal-Type C) 6 Other Isolation Failures (e.g., Dependent Failures) 7 Failures Induced by Phenomena (Early and Late) 8 Bypass (Interfacing System LOCA)

CDF All CET End states (including Very Low and No Release)

The steps taken to perform this risk assessment evaluation are as follows:

Step 1 - Quantify the base-line risk in terms of frequency per reactor year for each of the eight accident classes presented in Table 5-1.

Step 2 - Develop plant specific person-rem dose (population dose) per reactor year for each of the eight accident classes.

Step 3 - Evaluate risk impact of extending Type A test interval from three to fifteen and ten to fifteen years.

Step 4 - Determine the change in risk in terms of Large Early Release Frequency (LERF) in accordance with RG 1.174.

Step 5 - Determine the impact on the Conditional Containment Failure Probability (CCFP).

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Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 26 of 82 5.1 Step 1 - Quantify the Base-Line Risk in Terms of Frequency Per Reactor Year As previously described, the extension of the Type A interval does not influence those accident progressions that involve large containment isolation failures, Type B or Type C testing, or containment failure induced by severe accident phenomena.

For the assessment of ILRT impacts on the risk profile, the potential for pre-existing leaks is included in the model. (These events are represented by the Class 3 sequences in EPRI TR-104285). The question on containment integrity was modified to include the probability of a liner breach or bellows failure (due to excessive leakage) at the time of core damage. Two failure modes were considered for the Class 3 sequences. These are Class 3a (small breach) and Class 3b (large breach).

The frequencies for the severe accident classes defined in Table 5-1 were developed for PTN by first determining the frequencies for Classes 1, 2, 7 and 8 using the categorized sequences and the identified correlations shown in Table 4.2-3, scaling these frequencies to account for the uncategorized sequences, determining the frequencies for Classes 3a and 3b, and then determining the remaining frequency for Class 1. Furthermore, adjustments were made to the Class 3b and hence Class 1 frequencies to account for the impact of undetected corrosion per the methodology described in Section 4.4.

The Unit 3 total frequency of the categorized sequences is 2.91E-07/yr, the total CDF is 2.91E-07/yr (Table 4.2-4), and the scale factor is 1.0. The scaling factor is determined by dividing CDF by the total categorized release category frequency (2.91E-07/yr / 2.91E-07/yr =

1.0).

The Unit 4 total frequency of the categorized sequences is 2.91E-07/yr, the total CDF is 2.91E-07/yr (Table 4.2-5), and the scale factor is 1.0. The scaling factor is determined by dividing CDF by the total categorized release category frequency (2.91E-07/yr / 2.91E-07/yr =

1.0).

Tables 5-2a and 5-2b contain the frequencies from the scaling factor. The results are summarized below and in Tables 5-3a and 5-3b.

Table 5-2a: PTN Unit 3 Categorized Accident Classes and Frequencies Adjusted Frequency Using EPRI Class Class Frequency (-/yr)

Scale Factor of 1.0 (per yr) 1 1.21E-07 1.21E-07 2 1.39E-11 1.39E-11 7 1.52E-07 1.52E-07 8 1.79E-08 1.79E-08 Total 2.91E-07 2.91E-07 Frequency 3a =0.0092*(CDF-Class2-Class8) 2.51E-09 3b =0.0023*(CDF-Class2-Class8) 6.28E-10 Page l 26

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 27 of 82 Table 5-2b: PTN Unit 4 Categorized Accident Classes and Frequencies Adjusted Frequency Using EPRI Class Class Frequency (-/yr)

Scale Factor of 1.0 (per yr) 1 1.07E-07 1.07E-07 2 1.59E-11 1.59E-11 7 1.66E-07 1.66E-07 8 1.74E-08 1.74E-08 Total 2.91E-07 2.91E-07 Frequency 3a =0.0092*(CDF-Class2-Class8) 2.52E-09 3b =0.0023*(CDF-Class2-Class8) 6.29E-10 Class 1 Sequences. This group consists of all core damage accident progression bins for which the containment remains intact (modeled as Technical Specification Leakage). The frequency per year is initially determined from the Level 2 PRA model .

Class 2 Sequences. This group consists of all core damage accident progression bins for which a failure to isolate the containment occurs. The frequency per year for these sequences is obtained from the Level 2 PRA model.

Class 3 Sequences. This group consists of all core damage accident progression bins for which a pre-existing leakage in the containment structure (e.g., containment liner) exists. The containment leakage for these sequences can be either small (in excess of design allowable but

< 10 La) or large (> 100 La).

The respective frequencies per year are determined as follows:

PROBclass_3a = probability of small pre-existing containment liner leakage

= 0.0092 [see Section 4.3]

PROBclass_3b = probability of large pre-existing containment liner leakage

= 0.0023 [see Section 4.3]

As described in Section 4.3, additional consideration is made to not apply these failure probabilities on those cases that are already LERF scenarios (i.e., Class 2 and Class 8 contributions).

Unit 3 Class 3a Frequency = 0.0092 * (CDF - CLASS2 - CLASS8)

= 0.0092 * (2.91E 1.39E 1.79E-08) = 2.51E-09/yr Class 3b Frequency = 0.0023 * (CDF - CLASS2 - CLASS8)

=0.0023 * (2.91E 1.39E 1.79E-08) = 6.28E-10/yr Page l 27

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 28 of 82 Unit 4 Class 3a Frequency = 0.0092 * (CDF - CLASS2 - CLASS8)

= 0.0092 * (2.91E 1.59E 1.74E-08) = 2.52E-09/yr Class 3b Frequency = 0.0023 * (CDF - CLASS2 - CLASS8)

=0.0023 * (2.91E 1.59E 1.74E-08) = 6.29E-10/yr For this analysis, the associated containment leakage for Class 3a is 10 La and for Class 3b is 100 La. These assignments are consistent with the guidance provided in EPRI Report No. 1009325, Revision 2-A.

Class 4 Sequences. This group consists of all core damage accident progression bins for which containment isolation failure-to-seal of Type B test components occurs. Because these failures are detected by Type B tests which are unaffected by the Type A ILRT, this group is not evaluated any further in the analysis.

Class 5 Sequences. This group consists of all core damage accident progression bins for which containment isolation failure-to-seal of Type C test components occurs. Because the failures are detected by Type C tests which are unaffected by the Type A ILRT, this group is not evaluated any further in this analysis.

Class 6 Sequences. This group is similar to Class 2. These are sequences that involve core damage accident progression bins for which a failure-to-seal containment leakage due to failure to isolate the containment occurs. These sequences are dominated by misalignment of containment isolation valves following a test/maintenance evolution, typically resulting in a failure to close smaller containment isolation valves. All other failure modes are bounded by the Class 2 assumptions. Consistent with guidance provided in EPRI Report No. 1009325, Revision 2-A, this accident class is not explicitly considered since it has a negligible impact on the results.

Class 7 Sequences. This group consists of all core damage accident progression bins in which containment failure induced by severe accident phenomena occurs (e.g., overpressure). For this analysis, the frequency is determined from the Level 2 PRA model.

Class 8 Sequences. This group consists of all core damage accident progression bins in which containment bypass occurs. For this analysis, the frequency is determined from the Level 2 PRA model.

5.1.1 Summary of Accident Class Frequencies In summary, the accident sequence frequencies that can lead to radionuclide release to the public have been derived consistent with the definitions of accident classes defined in EPRI-TR-104285 the NEI Interim Guidance, and guidance provided in EPRI Report No. 1009325, Revision 2-A.

Tables 5-3a and 5-3b summarize these accident frequencies by accident class for PTN and provide the changes in frequency associated with the corrosion analysis.

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Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 29 of 82 Table 5-3a: Unit 3 Radionuclide Release Frequencies as a Function of Accident Class Accident Frequency (per Rx-yr)

Classes due to due to due to Description EPRI (Containment Corrosion (3 in Corrosion (1 Corrosion (1 in Release Type) Methodology 10 yr) in 10 yr) 15 yr)

No Containment 1 1.18E-07 -2.43E-12 -1.39E-11 -3.22E-11 Failure Large Isolation 2 Failures (Failure to 1.39E-11 0.00E+00 0.00E+00 0.00E+00 Close)

Small Isolation 3a Failures (liner 2.51E-09 0.00E+00 0.00E+00 0.00E+00 breach)

Large Isolation 3b Failures (liner 6.28E-10 2.43E-12 1.39E-11 3.22E-11 breach)

Small Isolation 4 Failures (Failure to N/A 0.00E+00 0.00E+00 0.00E+00 seal -Type B)

Small Isolation 5 Failures (Failure to N/A 0.00E+00 0.00E+00 0.00E+00 sealType C)

Other Isolation 6 Failures (e.g., N/A 0.00E+00 0.00E+00 0.00E+00 dependent failures)

Failures Induced by 7 Phenomena (Early 1.52E-07 0.00E+00 0.00E+00 0.00E+00 and Late)

Bypass (Interfacing 8 1.79E-08 0.00E+00 0.00E+00 0.00E+00 System LOCA)

CDF All CET end states 2.91E-07 0.00E+00 0.00E+00 0.00E+00 1

Based on data developed in Section 4.4. Only Classes 1 and 3b are impacted by the corrosion analysis.

The increase in Class 3b frequency leads to a reduction in Class 1 frequency to preserve overall CDF.

Table 5-3b: Unit 4 Radionuclide Release Frequencies as a Function of Accident Class Accident Classes Frequency (per Rx-yr)

(Containment Description EPRI Corrosion (3 Corrosion (1 Corrosion (1 Release Type) Methodology in 10 yr) in 10 yr) in 15 yr) 1 No Containment Failure 1.04E-07 -2.43E-12 -1.40E-11 -3.23E-11 Large Isolation Failures 2 (Failure to Close) 1.59E-11 0.00E+00 0.00E+00 0.00E+00 Small Isolation Failures 3a (liner breach) 2.52E-09 0.00E+00 0.00E+00 0.00E+00 Large Isolation Failures 3b (liner breach) 6.29E-10 2.43E-12 1.40E-11 3.23E-11 Small Isolation Failures 4 (Failure to seal -Type N/A 0.00E+00 0.00E+00 0.00E+00 B)

Small Isolation Failures 5 (Failure to sealType N/A 0.00E+00 0.00E+00 0.00E+00 C)

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Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 30 of 82 Table 5-3b: Unit 4 Radionuclide Release Frequencies as a Function of Accident Class Accident Classes Frequency (per Rx-yr)

(Containment Description EPRI Corrosion (3 Corrosion (1 Corrosion (1 Release Type) Methodology in 10 yr) in 10 yr) in 15 yr)

Other Isolation Failures 6 (e.g., dependent N/A 0.00E+00 0.00E+00 0.00E+00 failures)

Failures Induced by 7 Phenomena (Early and 1.66E-07 0.00E+00 0.00E+00 0.00E+00 Late)

Bypass (Interfacing 8 1.74E-08 0.00E+00 0.00E+00 0.00E+00 System LOCA)

CDF All CET end states 2.91E-07 0.00E+00 0.00E+00 0.00E+00 1

Based on data developed in Section 4.4. Only Classes 1 and 3b are impacted by the corrosion analysis.

The increase in Class 3b frequency leads to a reduction in Class 1 frequency to preserve overall CDF.

5.2 Step 2 - Develop Plant Specific Person-Rem Dose (Population Dose) Per Reactor Year Plant specific release analyses were performed to estimate the person-rem doses to the population within a 50 mile radius from the plant, and summarized in Table 4.2-6b. The results of applying these releases to the EPRI containment failure classification are as follows:

Class 1 = 2.42E+04 person-rem (Note 1)

Class 2 = 4.58E+06 person-rem (Note 2)

Class 3a = Class 1 Frequency x 10 La = 2.42E+05 person-rem (Note 3)

Class 3b = Class 1 Frequency x 100 La = 2.42E+06 person-rem (Note 3)

Class 4 = Not analyzed Class 5 = Not analyzed Class 6 = Not analyzed Class 7 = 5.42E+06 person-rem (Note 4)

Class 8 = 6.38E+06 person-rem (Note 5)

Notes:

(1) Class 1 is assigned the dose from No Containment Failure category from Table 4.2-6b.

(2) The Class 2, containment isolation failures is assigned the dose from the Large Isolation Failures category from Table 4.2-6b.

(3) The Class 3a and 3b doses are related to the leakage rate as shown. This is consistent with the guidance provided in EPRI Report No. 1009325, Revision 2-A.

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Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 31 of 82 (4) The Class 7 dose is assigned the dose from the Failures Induced by Phenomena category from Table 4.2-6b.

(5) Class 8 sequences involve containment bypass failures; as a result, the person-rem dose is not based on normal containment leakage. The releases for this class would include some fraction of Early SGTR, Late SGTR, and ISLOCA contributions. However, Class 8 is conservatively assigned the dose based entirely on the Bypass category from Table 4.2-6b because this bounds all the categories.

In summary, the population dose estimates derived for use in the risk evaluation per the EPRI methodology (Reference 8.2) containment failure classifications, and consistent with the NEI guidance (Reference 8.3) as modified by EPRI Report No. 1009325, Revision 2-A are provided in Table 5-4.

Table 5-4: PTN Population Dose Estimates for Population Within 50 Miles Accident Classes (Containment Release Description Person-Rem (50 miles)

Type) 1 No Containment Failure 2.42E+04 2 Large Isolation Failures (Failure to Close) 4.58E+06 3a Small Isolation Failures (liner breach) 2.42E+05 3b Large Isolation Failures (liner breach) 2.42E+06 4 Small Isolation Failures (Failure to seal -Type B) N/A 5 Small Isolation Failures (Failure to sealType C) N/A 6 Other Isolation Failures (e.g., dependent failures) N/A 7 Failures Induced by Phenomena (Early and Late) 5.42E+06 8 Bypass (Interfacing System LOCA) 6.38E+06 The above dose estimates, when combined with the results presented in Tables 5-3a and 5-3b, yield the PTN baseline mean consequence measures for each accident class. These results are presented in Tables 5-5a and 5-5b.

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Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 32 of 82 Table 5-5a: PTN Unit 3 Annual Dose as a Function of Accident Class; Characteristic of Conditions for ILRT Required 3/10 Years EPRI Methodology Plus Accident EPRI Methodology Change Due to Corrosion Classes Person-Rem Corrosion Description Person- Person-(Containment (50 miles) Frequency Frequency Person-Rem/yr (50 Rem/yr (50 Release Type) (per Rx-yr) (per Rx-yr) Rem/yr(1) miles) miles) 1 No Containment Failure (2) 2.42E+04 1.18E-07 2.86E-03 1.18E-07 2.86E-03 -5.88E-08 2 Large Isolation Failures (Failure to Close) 4.58E+06 1.39E-11 6.35E-05 1.39E-11 6.35E-05 0.00E+00 3a Small Isolation Failures (liner breach) 2.42E+05 2.51E-09 6.07E-04 2.51E-09 6.07E-04 0.00E+00 3b Large Isolation Failures (liner breach) 2.42E+06 6.28E-10 1.52E-03 6.31E-10 1.52E-03 5.88E-06 4 Small Isolation Failures (Failure to seal -Type B) N/A N/A N/A N/A N/A N/A 5 Small Isolation Failures (Failure to sealType C) N/A N/A N/A N/A N/A N/A 6 Other Isolation Failures (e.g.,

dependent failures) N/A N/A N/A N/A N/A N/A 7 Failures Induced by Phenomena (Early and Late) 5.42E+06 1.52E-07 8.23E-01 1.52E-07 8.23E-01 0.00E+00 8 Bypass (Interfacing System LOCA) 6.38E+06 1.79E-08 1.14E-01 1.79E-08 1.14E-01 0.00E+00 CDF All CET end states N/A 2.91E-07 9.43E-01 2.91E-07 9.43E-01 5.82E-06

1) Only release Classes 1 and 3b are affected by the corrosion analysis. The increase in Class3b frequency leads to a reduction in Class1 frequency to preserve overall CDF, thus the Person-Rem change for Class1 is negative.

2) Characterized as 1La release magnitude consistent with the derivation of the ILRT non-detection failure probability for ILRTs. Release Classes 3a and 3b include failures of containment to meet the Technical Specification leak rate.

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Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 33 of 82 Table 5-5b: PTN Unit 4 Annual Dose as a Function of Accident Class; Characteristic of Conditions for ILRT Required 3/10 Years EPRI Methodology Plus EPRI Methodology Change Due Accident Classes Corrosion Person-Rem to Corrosion (Containment Description Person-(50 miles) Frequency Person-Rem/yr Frequency (per Person-Release Type) Rem/yr (50 (per Rx-yr) (50 miles) Rx-yr) Rem/yr(1) miles) 1 No Containment Failure (2) 2.42E+04 1.04E-07 2.52E-03 1.04E-07 2.52E-03 -5.88E-08 2 Large Isolation Failures (Failure to Close) 4.58E+06 1.59E-11 7.26E-05 1.59E-11 7.26E-05 0.00E+00 3a Small Isolation Failures (liner breach) 2.42E+05 2.52E-09 6.08E-04 2.52E-09 6.08E-04 0.00E+00 3b Large Isolation Failures (liner breach) 2.42E+06 6.29E-10 1.52E-03 6.32E-10 1.53E-03 5.88E-06 4 Small Isolation Failures (Failure to seal -Type B) N/A N/A N/A N/A N/A N/A 5 Small Isolation Failures (Failure to sealType C) N/A N/A N/A N/A N/A N/A 6 Other Isolation Failures (e.g.,

dependent failures) N/A N/A N/A N/A N/A N/A 7 Failures Induced by Phenomena (Early and Late) 5.42E+06 1.66E-07 9.01E-01 1.66E-07 9.01E-01 0.00E+00 8 Bypass (Interfacing System LOCA) 6.38E+06 1.74E-08 1.11E-01 1.74E-08 1.11E-01 0.00E+00 CDF All CET end states N/A 2.91E-07 1.02E+00 2.91E-07 1.02E+00 5.82E-06

1) Only release Classes 1 and 3b are affected by the corrosion analysis. The increase in Class3b frequency leads to a reduction in Class1 frequency to preserve overall CDF, thus the Person-Rem change for Class1 is negative.

2) Characterized as 1La release magnitude consistent with the derivation of the ILRT non-detection failure probability for ILRTs. Release Classes 3a and 3b include failures of containment to meet the Technical Specification leak rate.

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Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 34 of 82 5.3 Step 3 - Evaluate Risk Impact of Extending Type A Test Interval From 10 to 15 Years The next step is to evaluate the risk impact of extending the test interval from its current ten year value to fifteen years. To do this, an evaluation must first be made of the risk associated with the ten year interval since the base case applies to a three year interval (i.e., a simplified representation of a three in ten interval).

5.3.1 Risk Impact Due to 10-year Test Interval As previously stated, Type A tests impact only Class 3 sequences. For Class 3 sequences, the release magnitude is not impacted by the change in test interval (a small or large breach remains the same, even though the probability of not detecting the breach increases). Thus, only the frequency of Class 3a and 3b sequences is directly impacted.

As it is assumed that the new Class 3 end states arise from previously intact containment states, the intact state frequency is reduced accordingly. The risk contribution is changed based on the NEI guidance as described in Section 4.3 by a factor of 3.33 compared to the base case values. The results of the calculation for a ten year interval are presented in Tables 5-6a and 5-6b.

5.3.2 Risk Impact Due to 15-Year Test Interval The risk contribution for a fifteen year interval is calculated in a manner similar to the ten year interval. The difference is in the increase in probability of leakage in Classes 3a and 3b. For this case, the value used in the analysis is a factor of 5.0 compared to the three year interval value, as described in Section 4.3. The results for this calculation are presented in Tables 5-7a and 5-7b.

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Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 35 of 82 Table 5-6a: PTN Unit 3 Annual Dose as a Function of Accident Class; Characteristic of Conditions for ILRT Required 1/10 Years EPRI Methodology Plus EPRI Methodology Change Due to Accident Classes Corrosion Person-Rem (50 Corrosion (Containment Description Frequency Person- Person-miles) Frequency Person-Release Type) (per Rx- Rem/yr (50 Rem/yr (50 (per Rx-yr) Rem/yr(1) yr) miles) miles) 1 No Containment Failure (2) 2.42E+04 1.11E-07 2.68E-03 1.11E-07 2.68E-03 -3.37E-07 2 Large Isolation Failures (Failure to Close) 4.58E+06 1.39E-11 6.35E-05 1.39E-11 6.35E-05 0.00E+00 3a Small Isolation Failures (liner breach) 2.42E+05 8.37E-09 2.02E-03 8.37E-09 2.02E-03 0.00E+00 3b Large Isolation Failures (liner breach) 2.42E+06 2.09E-09 5.06E-03 2.11E-09 5.09E-03 3.37E-05 4 Small Isolation Failures (Failure to seal -Type B) N/A N/A N/A N/A N/A N/A 5 Small Isolation Failures (Failure to sealType C) N/A N/A N/A N/A N/A N/A 6 Other Isolation Failures (e.g.,

N/A N/A N/A N/A N/A N/A dependent failures) 7 Failures Induced by Phenomena (Early and Late) 5.42E+06 1.52E-07 8.23E-01 1.52E-07 8.23E-01 0.00E+00 8 Bypass (Interfacing System LOCA) 6.38E+06 1.79E-08 1.14E-01 1.79E-08 1.14E-01 0.00E+00 CDF All CET end states N/A 2.91E-07 9.47E-01 2.91E-07 9.47E-01 3.33E-05

1) Only release Classes 1 and 3b are affected by the corrosion analysis. The increase in Class3b frequency leads to a reduction in Class1 frequency to preserve overall CDF, thus the Person-Rem change for Class1 is negative.

2) Characterized as 1La release magnitude consistent with the derivation of the ILRT non-detection failure probability for ILRTs. Release Classes 3a and 3b include failures of containment to meet the Technical Specification leak rate.

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Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 36 of 82 Table 5-6b: PTN Unit 4 Annual Dose as a Function of Accident Class; Characteristic of Conditions for ILRT Required 1/10 Years EPRI Methodology Plus Accident Classes EPRI Methodology Change Due to Person-Rem Corrosion (Containment Description Corrosion Person-(50 miles) Frequency Person-Rem/yr Frequency Person-Rem/yr Release Type) Rem/yr(1)

(per Rx-yr) (50 miles) (per Rx-yr) (50 miles) 1 No Containment Failure (2) 2.42E+04 9.68E-08 2.34E-03 9.68E-08 2.34E-03 -3.37E-07 2 Large Isolation Failures (Failure to Close) 4.58E+06 1.59E-11 7.26E-05 1.59E-11 7.26E-05 0.00E+00 3a Small Isolation Failures (liner breach) 2.42E+05 8.38E-09 2.03E-03 8.38E-09 2.03E-03 0.00E+00 3b Large Isolation Failures (liner breach) 2.42E+06 2.10E-09 5.06E-03 2.11E-09 5.10E-03 3.37E-05 4 Small Isolation Failures (Failure to seal -Type B) N/A N/A N/A N/A N/A N/A 5 Small Isolation Failures (Failure to sealType C) N/A N/A N/A N/A N/A N/A 6 Other Isolation Failures (e.g.,

dependent failures) N/A N/A N/A N/A N/A N/A 7 Failures Induced by Phenomena (Early and Late) 5.42E+06 1.66E-07 9.01E-01 1.66E-07 9.01E-01 0.00E+00 8 Bypass (Interfacing System LOCA) 6.38E+06 1.74E-08 1.11E-01 1.74E-08 1.11E-01 0.00E+00 CDF All CET end states N/A 2.91E-07 1.02E+00 2.91E-07 1.02E+00 3.34E-05

1) Only release Classes 1 and 3b are affected by the corrosion analysis. The increase in Class3b frequency leads to a reduction in Class1 frequency to preserve overall CDF, thus the Person-Rem change for Class1 is negative.

2) Characterized as 1La release magnitude consistent with the derivation of the ILRT non-detection failure probability for ILRTs. Release Classes 3a and 3b include failures of containment to meet the Technical Specification leak rate.

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Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 37 of 82 Table 5-7a: PTN Unit 3 Annual Dose as a Function of Accident Class; Characteristic of Conditions for ILRT Required 1/15 Years EPRI Methodology Plus EPRI Methodology Accident Classes Corrosion Change Due to Person-Rem (Containment Description Person- Person- Corrosion Person-(50 miles) Frequency Frequency Release Type) Rem/yr (50 Rem/yr (50 Rem/yr(1)

(per Rx-yr) (per Rx-yr) miles) miles) 1 No Containment Failure 2.42E+04 1.06E-07 2.55E-03 1.06E-07 2.55E-03 -7.79E-07 (2) 2 Large Isolation Failures (Failure to Close) 4.58E+06 1.39E-11 6.35E-05 1.39E-11 6.35E-05 0.00E+00 3a Small Isolation Failures (liner breach) 2.42E+05 1.26E-08 3.04E-03 1.26E-08 3.04E-03 0.00E+00 3b Large Isolation Failures (liner breach) 2.42E+06 3.14E-09 7.59E-03 3.17E-09 7.67E-03 7.79E-05 4 Small Isolation Failures (Failure to seal -Type B) N/A N/A N/A N/A N/A N/A 5 Small Isolation Failures (Failure to sealType N/A N/A N/A N/A N/A N/A C) 6 Other Isolation Failures (e.g., dependent failures) N/A N/A N/A N/A N/A N/A 7 Failures Induced by Phenomena (Early and 5.42E+06 1.52E-07 8.23E-01 1.52E-07 8.23E-01 0.00E+00 Late) 8 Bypass (Interfacing System LOCA) 6.38E+06 1.79E-08 1.14E-01 1.79E-08 1.14E-01 0.00E+00 CDF All CET end states N/A 2.91E-07 9.51E-01 2.91E-07 9.51E-01 7.71E-05

1) Only release Classes 1 and 3b are affected by the corrosion analysis. The increase in Class3b frequency leads to a reduction in Class1 frequency to preserve overall CDF, thus the Person-Rem change for Class1 is negative.

2) Characterized as 1La release magnitude consistent with the derivation of the ILRT non-detection failure probability for ILRTs. Release Classes 3a and 3b include failures of containment to meet the Technical Specification leak rate.

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Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 38 of 82 Table 5-7b: PTN Unit 4 Annual Dose as a Function of Accident Class; Characteristic of Conditions for ILRT Required 1/15 Years EPRI Methodology Plus EPRI Methodology Accident Classes Corrosion Change Due to Person-Rem (50 (Containment Description Frequency Frequency Corrosion miles) Person-Rem/yr Person-Rem/yr Release Type) (per Rx- (per Rx- Person-Rem/yr(1)

(50 miles) (50 miles) yr) yr) 1 No Containment 2.42E+04 9.16E-08 2.21E-03 9.15E-08 2.21E-03 -7.80E-07 Failure (2) 2 Large Isolation Failures (Failure to 4.58E+06 1.59E-11 7.26E-05 1.59E-11 7.26E-05 0.00E+00 Close) 3a Small Isolation Failures (liner breach) 2.42E+05 1.26E-08 3.04E-03 1.26E-08 3.04E-03 0.00E+00 3b Large Isolation Failures (liner breach) 2.42E+06 3.15E-09 7.60E-03 3.18E-09 7.68E-03 7.80E-05 4 Small Isolation Failures (Failure to N/A N/A N/A N/A N/A N/A seal -Type B) 5 Small Isolation Failures (Failure to N/A N/A N/A N/A N/A N/A sealType C) 6 Other Isolation Failures (e.g., N/A N/A N/A N/A N/A N/A dependent failures) 7 Failures Induced by Phenomena (Early and 5.42E+06 1.66E-07 9.01E-01 1.66E-07 9.01E-01 0.00E+00 Late) 8 Bypass (Interfacing System LOCA) 6.38E+06 1.74E-08 1.11E-01 1.74E-08 1.11E-01 0.00E+00 CDF All CET end states N/A 2.91E-07 1.02E+00 2.91E-07 1.02E+00 7.72E-05

1) Only release Classes 1 and 3b are affected by the corrosion analysis. The increase in Class3b frequency leads to a reduction in Class1 frequency to preserve overall CDF, thus the Person-Rem change for Class1 is negative.

2) Characterized as 1La release magnitude consistent with the derivation of the ILRT non-detection failure probability for ILRTs. Release Classes 3a and 3b include failures of containment to meet the Technical Specification leak rate.

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Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 39 of 82 5.4 Step 4 - Determine the Change in Risk in Terms of Large Early Release Frequency (LERF)

The risk increase associated with extending the ILRT interval involves the potential that a core damage event that normally would result in only a small radioactive release from an intact containment could in fact result in a larger release due to the increase in probability of failure to detect a pre-existing leak. With strict adherence to the EPRI guidance, 100% of the Class 3b contribution would be considered LERF.

RG 1.174 provides guidance for determining the risk impact of plant specific changes to the licensing basis. RG 1.174 defines very small changes in risk as resulting in increases of core damage frequency (CDF) below 10-6/yr and increases in LERF below 10-7/yr, and small changes in LERF as below 10-6/yr. Because the ILRT does not impact CDF, the relevant metric is LERF.

For PTN, 100% of the frequency of Class 3b sequences can be used as a very conservative first-order estimate to approximate the potential increase in LERF from the ILRT interval extension (consistent with the EPRI guidance methodology). Based on a ten year test interval from Tables 5-6a and 5-6b, the Class 3b large early release frequency contribution (conservatively including corrosion) is 2.11E-09/yr for Unit 3 and 2.11E-09/yr for Unit 4; and, based on a fifteen year test interval from Tables 5-7a and 5-7b, this LERF contribution increases to 3.17E-09/yr for Unit 3 and 3.18E-09/yr for Unit 4. Thus, the increase in the overall LERF due to Class 3b sequences that is due to increasing the ILRT test interval from three to fifteen years is 2.54E-09/yr for Unit 3 and 2.55E-09/yr for Unit 4 as shown in Tables 5-8a and 5-8b. Similarly, the increase in LERF due to increasing the ILRT interval from ten years to fifteen years is 1.07E-09/yr for Unit 3 and 1.07E-09/yr for Unit 4 as shown in Tables 5-8a and 5-8b.

As can be seen, even with the conservatisms included in the evaluation (per the EPRI methodology), the estimated change in LERF for PTN is below the threshold criteria for a very small change when comparing both the fifteen year results to the current ten year requirement, and when the fifteen year ILRT extension results are compared to the original three year requirement, the increase in LERF is below the threshold for a very small change.

5.5 Step 5 - Determine the Impact on the Conditional Containment Failure Probability (CCFP)

Another parameter that the NRC guidance in RG 1.174 states can provide input into the decision-making process is the change in the conditional containment failure probability (CCFP).

The change in CCFP is indicative of the effect of the ILRT on all radionuclide releases, not just LERF. The CCFP can be calculated from the results of this analysis. One of the difficult aspects of this calculation is providing a definition of the failed containment. In this assessment, the CCFP is defined such that containment failure includes all radionuclide release end states other than the intact state. The conditional part of the definition is conditional given a severe accident (i.e., core damage).

The change in CCFP can be calculated by using the method specified in the EPRI Report No. 1009325, Revision 2-A. The NRC has previously accepted similar calculations (Reference 8.9) as the basis for showing that the proposed change is consistent with the Page l 39

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 40 of 82 defense-in-depth philosophy. The list below shows the CCFP values that result from the assessment for the various testing intervals including corrosion effects.

Unit 3 CCFP = [1 - (Class 1 frequency + Class 3a frequency) / CDF]

100%

CCFP3 = [1 - (1.18E-07/yr + 2.51E-09/yr) / 2.91E-07/yr]

100% = 58.55%

CCFP3 = 58.55%

CCFP10 = [1 - (1.11E-08/yr + 8.37E-09/yr) / 2.91-07/yr]

100% = 59.05%

CCFP10 = 59.05%

CCFP15 = [1 - (1.06E-07/yr + 1.26E-08/yr) / 2.91E-07/yr]

100% = 59.42%

CCFP15 = 59.42%

CCFP3 to 15 = CCFP15 - CCFP3 = 0.87%

CCFP10 to 15 = CCFP15 - CCFP10 = 0.37%

CCFP3 to 10 = CCFP10 - CCFP3 = 0.51%

Unit 4 CCFP = [1 - (Class 1 frequency + Class 3a frequency) / CDF]

100%

CCFP3 = [1 - (1.04E-07/yr + 2.52E-09/yr) / 2.91E-07/yr]

100% = 63.33%

CCFP3 = 63.33%

CCFP10 = [1 - (9.68E-08/yr + 8.38E-09/yr) / 2.91-07/yr]

100% = 63.84%

CCFP10 = 63.84%

CCFP15 = [1 - (9.15E-08/yr + 1.26E-08/yr) / 2.91E-07/yr]

100% = 64.21%

CCFP15 = 64.21%

CCFP3 to 15 = CCFP15 - CCFP3 = 0.88%

CCFP10 to 15 = CCFP15 - CCFP10 = 0.37%

CCFP3 to 10 = CCFP10 - CCFP3 = 0.51%

The change in CCFP of approximately 0.88% from extending the test interval to fifteen years from the original three in ten year requirement is judged to be very small. It is below the acceptance criteria for increase in CCFP of <1.5%.

Page l 40

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 41 of 82 5.6 Summary of Results The results from this ILRT extension risk assessment for PTN are summarized in Tables 5-8a and 5-8b.

Table 5-8a:

PTN Unit 3 ILRT Cases: Base, 3 to 10, and 3 to 15 Yr Extensions (Including Age Adjusted Steel Liner Corrosion Likelihood)

EPRI Class DOSE Base Case 3 in 10 Years Extend to 1 in 10 Extend to 1 in 15 Years Person-Rem Years CDF/Yr Person- CDF/Yr Person- CDF/Yr Person-Rem/Yr Rem/Yr Rem/Yr 1 2.42E+04 1.18E-07 2.86E-03 1.11E-07 2.68E-03 1.06E-07 2.55E-03 2 4.58E+06 1.39E-11 6.35E-05 1.39E-11 6.35E-05 1.39E-11 6.35E-05 3a 2.42E+05 2.51E-09 6.07E-04 8.37E-09 2.02E-03 1.26E-08 3.04E-03 3b 2.42E+06 6.31E-10 1.52E-03 2.11E-09 5.09E-03 3.17E-09 7.67E-03 7 5.42E+06 1.52E-07 8.23E-01 1.52E-07 8.23E-01 1.52E-07 8.23E-01 8 6.38E+06 1.79E-08 1.14E-01 1.79E-08 1.14E-01 1.79E-08 1.14E-01 Total N/A 2.91E-07 9.43E-01 2.91E-07 9.47E-01 2.91E-07 9.51E-01 ILRT Dose Rate from 2.13E-03 7.11E-03 1.07E-02 3a and 3b From 3 yr N/A 4.80E-03 8.27E-03 Delta Total Dose Rate (Person-Rem/year) From 10 yr N/A N/A 3.47E-03 From 3 yr N/A 0.51% 0.88%

% change in dose rate from base From 10 yr N/A N/A 0.37%

3b Frequency (LERF) 6.31E-10 2.11E-09 3.17E-09 From 3 yr N/A 1.48E-09 2.54E-09 Delta LERF From 10 yr N/A N/A 1.07E-09 CCFP % 58.55% 59.05% 59.42%

Delta From 3 yr N/A 0.51% 0.87%

CCFP% From 10 yr N/A N/A 0.37%

Page l 41

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 42 of 82 Table 5-8b:

PTN Unit 4 ILRT Cases: Base, 3 to 10, and 3 to 15 Yr Extensions (Including Age Adjusted Steel Liner Corrosion Likelihood EPRI Class DOSE Base Case 3 in 10 Years Extend to 1 in 10 Years Extend to 1 in 15 Years Person- CDF/Yr Person- CDF/Yr Person- CDF/Yr Person-Rem Rem/Yr Rem/Yr Rem/Yr 1 2.42E+04 1.04E-07 2.52E-03 9.68E-08 2.34E-03 9.15E-08 2.21E-03 2 4.58E+06 1.59E-11 7.26E-05 1.59E-11 7.26E-05 1.59E-11 7.26E-05 3a 2.42E+05 2.52E-09 6.08E-04 8.38E-09 2.03E-03 1.26E-08 3.04E-03 3b 2.42E+06 6.32E-10 1.53E-03 2.11E-09 5.10E-03 3.18E-09 7.68E-03 7 5.42E+06 1.66E-07 9.01E-01 1.66E-07 9.01E-01 1.66E-07 9.01E-01 8 6.38E+06 1.74E-08 1.11E-01 1.74E-08 1.11E-01 1.74E-08 1.11E-01 Total N/A 2.91E-07 1.02E+00 2.91E-07 1.02E+00 2.91E-07 1.02E+00 ILRT Dose Rate from 2.13E-03 7.12E-03 1.07E-02 3a and 3b Delta Total From 3 yr N/A 4.81E-03 8.28E-03 Dose Rate (Person-Rem/year) From 10 yr N/A N/A 3.47E-03

% change in From 3 yr N/A 0.47% 0.81%

dose rate from base From 10 yr N/A N/A 0.34%

3b Frequency (LERF) 6.32E-10 2.11E-09 3.18E-09 From 3 yr N/A 1.48E-09 2.55E-09 Delta LERF From 10 yr N/A N/A 1.07E-09 CCFP % 63.33% 63.84% 64.21%

Delta From 3 yr N/A 0.51% 0.88%

CCFP% From 10 yr N/A N/A 0.37%

6.0 Sensitivities 6.1 Sensitivity to Corrosion Impact Assumptions The PTN results in Tables 5-5a through 5-7b show that including corrosion effects calculated using the assumptions described in Section 4.4 does not significantly affect the results of the ILRT extension risk assessment.

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Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 43 of 82 Sensitivity cases were developed to gain an understanding of the sensitivity of the results to the key parameters in the corrosion risk analysis. The time for the flaw likelihood to double was adjusted from every five years to every two and every ten years. The failure probabilities for the upper containment and the basemat were increased and decreased by an order of magnitude. The total detection failure likelihood was adjusted from 10% to 15% and 5%. The results are presented in Table 6-1 based on Unit 3 results. Due to the similarity of the Unit 3 and Unit 4 results this sensitivity is documented only for Unit 3. In every case the impact from including the corrosion effects is very minimal. Even the upper bound estimates with very conservative assumptions for all of the key parameters yield increases in LERF due to corrosion of only 8.92E-10/yr. The results indicate that even with very conservative assumptions, the conclusions from the base analysis would not change.

Table 6-1:

Unit 3 Steel Plate Corrosion Sensitivity Cases Visual Inspection Increase in Class 3b Frequency Containment (LERF) for ILRT Extension from 3

& Non-Visual Age (Step 3 in the Breach (Step 4 in to 15 years (per Rx-yr)

Flaws (Step 5 in corrosion analysis) the corrosion the corrosion analysis) Increase Due to analysis) Total Increase Corrosion Base Case Base Case Base Case Doubles (1% Upper (10% Upper every 5 yrs Containment, 0.1% Containment, 2.54E-09 3.21E-11 Basemat) 100% Basemat)

Doubles every 2 yrs Base Base 2.58E-09 6.39E-11 Doubles every 10 yrs Base Base 2.54E-09 2.86E-11 Base Base 15% 2.56E-09 4.49E-11 Base Base 5% 2.53E-09 1.93E-11 10% Upper Base Containment, 1% Base 2.83E-09 3.21E-10 Basemat 0.1% Upper Base Containment, Base 2.52E-09 3.21E-12 0.01% Basemat Lower Bound 0.1% Upper 5% Upper Doubles every Containment, Containment, 2.51E-09 1.14E-12 10 yrs 0.01% Basemat 1% Basemat Upper Bound 10% Upper 15% Upper Doubles every 2 yrs Containment, 1% Containment, 3.40E-09 8.92E-10 Basemat 100% Basemat Page l 43

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 44 of 82 6.2 Sensitivity to Class 3B Contribution to LERF The Class 3b frequency for the base case of a three in ten year ILRT interval including corrosion is 6.31E-10/yr for Unit 3 and 6.32E-10/yr for Unit 4 (Tables 5-5a and 5-5b). Extending the interval to one in ten years results in a frequency of 2.11E-09/yr for Unit 3 and 2.11E-09/yr for Unit 4 (Tables 5-6a and 5-6b). Extending it to one in fifteen years results in a frequency of 3.17E-09/yr for Unit 3 and 3.18E-09/yr for Unit 4 (Tables 5-7a and 5-7b), which is an increase of approximately 2.54E-09/yr from three in ten years to once in fifteen years.

Even when 100% of the Class 3b sequences are assumed to have potential releases large enough for LERF, then the increase in LERF due to extending the interval from three in ten to one in fifteen is below the RG 1.174 threshold for very small changes in LERF of 1.0E-07/yr.

6.3 Potential Impact from External Events and Internal Flooding Contribution As described in Section 4.2.1 above, the ILRT risk assessment quantitative results are based on PTNs Internal Events Level 1 and Level 2 PRA model. This model is used to define the distribution of releases amongst the different EPRI release class bins. This section summarizes the impact on the ILRT risk assessment of including external events (fire, seismic, other external events) and internal flooding core damage frequency.

The purpose of the external events evaluation is to determine whether there are any unique insights or important quantitative information that explicitly impact the risk assessment results obtained when considering only internal events.

The detailed external events and internal flooding risk assessment results are shown in Tables 6-3a to 6-9b. As described in Section 4.2.1, the PTN PRA includes fire and internal flooding models that are peer reviewed against the requirements of the PRA Standard and RG-1.200, Rev.

2. A bounding seismic risk estimate is based on a plant level fragility estimate rather than component-level fragilities used in a full seismic PRA (Reference 8.24), and other external events are estimated based on other external events CDF and LERF estimates and assumptions (Reference 8.18). Accordingly, seismic and other external event risk is qualitatively assessed below to further demonstrate that the proposed ILRT Type A test extension has a minimal risk impact from external events. Further, quantitative sensitivities are performed for seismic risk, and for external event release class distributions.

6.3.1 External Events and Internal Fire & Flooding Contribution - Qualitative Insights Internal Fire Events - Qualitative Risk Impacts on ILRT Extension The PTN Fire PRA includes an integrated quantitative assessment of at-power internal fire risk.

In the base case plant risk model, internal fire events contribute significantly to the total Unit 3 and Unit 4 at-power CDF and LERF. The base case quantitative CDF/LERF risk results for internal fire events are as provided in Tables 4.2-1 and 4.2-2 and are used in the quantitative external event assessment in Tables 6-3a to 6-9b.

The proposed extension of the ILRT interval does not impact the frequency of any fire initiating events nor does the ILRT extension impact the reliability of active equipment credited in fire Page l 44

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 45 of 82 initiating event sequences. PTN Fire PRA addressed critical fire vulnerabilities as part of the transition to a performance-based, risk-informed fire protection program in accordance with 10 CFR 50.48(c) and NFPA 805 (Reference 8.36).

The ILRT test is focused on performing a periodic validation of the containment liner leak tightness, which is a condition that is independent of fire events risk. Therefore the ILRT extension has no direct effect on the core damage from fire events.

Internal Flooding Events - Qualitative Risk Impacts on ILRT Extension The PTN PRA includes an integrated quantitative assessment of at-power internal flooding risk.

In the base case plant risk model, internal flooding events contribute <1% of the total Unit 3 and Unit 4 at-power CDF and LERF. The base case quantitative CDF/LERF results for internal flooding events are provided in Tables 4.2-1 and 4.2-2 and are used in the quantitative external event assessment in Tables 6-3a to 6-9b.

The proposed extension of the ILRT interval does not impact the frequency of any flooding initiating events nor does the ILRT extension impact the reliability of active equipment credited in flooding initiating event sequences. The ILRT test is focused on performing a periodic validation of the containment liner leak tightness, which is a condition that is independent of flooding events risk. Therefore the ILRT extension has no direct effect on the core damage from flooding events.

Seismic Events - Qualitative Risk Impacts on ILRT Extension The PTN seismic risk is conservatively estimated at the plant level based on seismic hazard curves for Turkey Point and a plant-level fragility curve (Reference 8.24). The CDF estimate does not credit any components designed to withstand seismic events with accelerations higher than the analyzed value. The base case quantitative CDF/LERF estimates for seismic events are provided in Tables 4.2-1 and 4.2-2 and are used in the quantitative external event assessment in Tables 6-3a to 6-9b. As shown in Tables 4.2-1 and 4.2-2, external events (including seismic events) are assumed to contribute approximately the same proportion of CDF to LERF as the internal events (~3.81%) (Reference 8.18). Therefore, the seismic contribution to LERF is estimated to be 2.66E-08/yr based on the conservative CDF estimate.

The proposed extension of the ILRT interval does not impact the frequency of any seismic initiating event nor does the ILRT extension impact the reliability of active equipment used in seismic initiating event sequences. The ILRT test is focused on performing a periodic validation of the containment liner leak tightness, which is a condition independent of the risk of seismic events. Therefore the ILRT extension has no direct effect on the core damage or release mitigation capability from seismic events.

As noted above, the PTN seismic risk is based on a conservative CDF estimate rather than a full seismic PRA. If a seismic PRA were developed using the latest industry seismic risk hazard and methods, the actual PTN seismic-induced CDF/LERF could be higher (although it is expected to be lower than the reported value). The seismic CDF stated in Table D-1 of Appendix D to the NRC Safety/Risk Assessment for Generic Issue 199 (GI-199), Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United States on Existing Plants Page l 45

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 46 of 82 (Reference 8.33), is 5.9E-06/yr for the simple average and is 1.0E-05/yr for the bounding weakest link model (both Units 3 and 4). A seismic CDF of 1.0E-05/yr is approximately 14 times higher than the estimated seismic CDF used in the base risk assessment. Therefore, a seismic risk sensitivity evaluation was performed by assuming that the current seismic-induced CDF is equivalent to 1.0E-05/yr and the proportion contributing to LERF is doubled (i.e. ~7.6%

of CDF). All other calculations in the assessment are unchanged. The results of this evaluation for Unit 3 are shown in Table 6-2. Due to the similarity of the results for Unit 3 and Unit 4, this sensitivity is documented only for Unit 3. This evaluation shows that the Class 3b frequency (i.e.

LERF) due to external events for a 15 year ILRT interval would become 7.98E-07/yr resulting in a delta-LERF of 6.40E-07/yr from a 3 in 10 year frequency and 2.68E-07/yr from a 10 year frequency. This represents an increase in the delta-LERF of approximately 8.06E-08/yr above the baseline Unit 3 external event delta-LERF of 5.58E-07/yr (Table 6-9a). This shows that the baseline delta-LERF is relatively insensitive to the assumed increase in seismic risk/contribution to LERF (~14% increase in LERF when the seismic CDF was increased by a factor of ~14).

Based on this seismic risk sensitivity, the baseline quantitative seismic CDF and LERF contributions are judged appropriate for use in the delta LERF impact of the ILRT Type A test interval extension.

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Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 47 of 82 Table 6-2: Sensitivity to GI-199 Seismic CDF for PTN Base Case 3 in 10 Years Extend to 1 in 10 Years Extend to 1 in 15 Years DOSE EPRI Class Person- Person- Person-Person-Rem CDF/Yr CDF/Yr CDF/Yr Rem/Yr Rem/Yr Rem/Yr 1 2.42E+04 2.97E-05 7.18E-01 2.79E-05 6.74E-01 2.65E-05 6.42E-01 2 4.58E+06 3.49E-09 1.60E-02 3.49E-09 1.60E-02 3.49E-09 1.60E-02 3a 2.42E+05 6.32E-07 1.53E-01 2.10E-06 5.09E-01 3.16E-06 7.64E-01 3b 2.42E+06 1.59E-07 3.83E-01 5.30E-07 1.28E+00 7.98E-07 1.93E+00 7 5.42E+06 3.82E-05 2.07E+02 3.82E-05 2.07E+02 3.82E-05 2.07E+02 8 6.38E+06 4.51E-06 2.88E+01 4.51E-06 2.88E+01 4.51E-06 2.88E+01 Total N/A 7.32E-05 2.37E+02 7.32E-05 2.38E+02 7.32E-05 2.39E+02 ILRT Dose Rate from 5.36E-01 1.79E+00 2.69E+00 3a and 3b Delta Total From 3 yr N/A 1.21E+00 2.08E+00 Dose Rate (Person-Rem/year) From 10 yr N/A N/A 8.72E-01

% change in From 3 yr N/A 0.51% 0.88%

dose rate from base From 10 yr N/A N/A 0.37%

3b Frequency (LERF) 1.59E-07 5.30E-07 7.98E-07 From 3 yr N/A 3.71E-07 6.40E-07 Delta LERF From 10 yr N/A N/A 2.68E-07 CCFP % 58.55% 59.05% 59.42%

Delta From 3 yr N/A 0.87% 0.87%

CCFP% From 10 yr N/A N/A 0.37%

Other External Events - Qualitative Risk Impacts on ILRT Extension Other external events include external flooding, high winds, tornado, transportation and nearby industrial facility hazards, turbine missile, aircraft crash, heavy load drop, etc. These other external events are screened out based on the low probability of occurrence and rigorous plant design features as described in the UFSAR and PTN-BFJR-17-051 (Reference 8.32, 8.37).

These other external events are judged to have a very small contribution to CDF/LERF, however, for the purposes of the quantitative evaluation in Tables 6-3a to 6-9b, the CDF and LERF values cited in Tables 4.2-1 and 4.2-2 are used.

The proposed extension of the ILRT test interval does not impact the initiating event frequencies of other external events nor does the ILRT extension impact the reliability of active equipment Page l 47

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 48 of 82 needed to mitigate these events. The ILRT test is focused on performing a periodic validation of the containment liner leak tightness, which is a condition that is independent of external event risk. Therefore the ILRT extension has no direct effect on the core damage and release mitigation capability from external events.

6.3.2 External Events and Internal Flooding Contribution - Quantitative Evaluation It is assumed that the distribution of the internal flooding, internal fire, and external events (IF/EE) contributions to core damage frequency will be similar to that of internal events (Section 3.0). The percent contribution of the total CDF to each accident class is provided in Table 4.2-3.

Additional release class distribution sensitivity cases are provided in Section 6.3.3.

The total contribution to CDF from IF/EE is:

U3: 1.37E-08/yr (IF) + 6.98E-07/yr (seismic) + 6.22E-05/yr (fire) + 1.0E-06/yr (other)=6.39E-05/yr U4: 1.27E-08/yr (IF) + 6.98E-07/yr (seismic) + 6.39E-05/yr (fire) + 1.0E-06/yr (other)=6.56E-05/yr Tables 4.2-4 and 4.2-5 provide the results of distributing the internal flooding and external events CDF contributions to the EPRI accident classes. Utilizing the combined external event input, the impact of ILRT interval extension due to external events is evaluated in Tables 6-3a to 6-9b.

Table 6-3a:

PTN Unit 3 Categorized Accident Classes and Frequencies (EE Sensitivity)

Adjusted Frequency Using EPRI Class Class Frequency (-/yr)

Scale Factor of 1.0 (per yr) 1 2.66E-05 2.66E-05 2 3.05E-09 3.05E-09 7 3.33E-05 3.33E-05 8 3.94E-06 3.94E-06 Total 6.39E-05 6.39E-05 Frequency 3a =0.0092*(CDF-Class2-Class8) 5.52E-07 3b =0.0023*(CDF-Class2-Class8) 1.38E-07 Page l 48

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 49 of 82 Table 6-3b:

PTN Unit 4 Categorized Accident Classes and Frequencies (EE Sensitivity)

Adjusted Frequency Using EPRI Class Class Frequency (-/yr)

Scale Factor of 1.0 (per yr) 1 2.42E-05 2.42E-05 2 3.58E-09 3.58E-09 7 3.75E-05 3.75E-05 8 3.92E-06 3.92E-06 Total 6.56E-05 6.56E-05 Frequency 3a =0.0092*(CDF-Class2-Class8) 5.68E-07 3b =0.0023*(CDF-Class2-Class8) 1.42E-07 Table 6-4a:

PTN Unit 3 Radionuclide Release Frequencies as a Function of Accident Class (EE Sensitivity)

Accident Frequency (per Rx-yr)

Classes Description EPRI Corrosion Corrosion (1 Corrosion (Containment Release Type) Methodology (3 in 10 yr) in 10 yr) (1 in 15 yr)

No Containment 1 2.59E-05 -5.34E-10 -3.06E-09 -7.08E-09 Failure Large Isolation 2 Failures (Failure to 3.05E-09 0.00E+00 0.00E+00 0.00E+00 Close)

Small Isolation 3a Failures (liner breach) 5.52E-07 0.00E+00 0.00E+00 0.00E+00 Large Isolation 3b Failures (liner breach) 1.38E-07 5.34E-10 3.06E-09 7.08E-09 Small Isolation 4 Failures (Failure to N/A 0.00E+00 0.00E+00 0.00E+00 seal -Type B)

Small Isolation 5 Failures (Failure to N/A 0.00E+00 0.00E+00 0.00E+00 sealType C)

Other Isolation 6 Failures (e.g., N/A 0.00E+00 0.00E+00 0.00E+00 dependent failures)

Failures Induced by 7 Phenomena (Early and 3.33E-05 0.00E+00 0.00E+00 0.00E+00 Late)

Bypass (Interfacing 8 System LOCA) 3.94E-06 0.00E+00 0.00E+00 0.00E+00 CDF All CET end states 6.39E-05 0.00E+00 0.00E+00 0.00E+00 1

Based on data developed in Section 4.4. Only Classes 1 and 3b are impacted by the corrosion analysis.

The increase in Class 3b frequency leads to a reduction in Class 1 frequency to preserve overall CDF.

Page l 49

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 50 of 82 Table 6-4b:

PTN Unit 4 Radionuclide Release Frequencies as a Function of Accident Class (EE Sensitivity)

Accident Frequency (per Rx-yr)

Classes Description EPRI Corrosion (3 Corrosion (1 Corrosion (1 (Containment Release Type) Methodology in 10 yr) in 10 yr) in 15 yr)

No Containment 1 2.35E-05 -5.49E-10 -3.15E-09 -7.28E-09 Failure Large Isolation 2 Failures (Failure to 3.58E-09 0.00E+00 0.00E+00 0.00E+00 Close)

Small Isolation 3a Failures (liner 5.68E-07 0.00E+00 0.00E+00 0.00E+00 breach)

Large Isolation 3b Failures (liner 1.42E-07 5.49E-10 3.15E-09 7.28E-09 breach)

Small Isolation 4 Failures (Failure to N/A 0.00E+00 0.00E+00 0.00E+00 seal -Type B)

Small Isolation 5 Failures (Failure to N/A 0.00E+00 0.00E+00 0.00E+00 sealType C)

Other Isolation 6 Failures (e.g., N/A 0.00E+00 0.00E+00 0.00E+00 dependent failures)

Failures Induced by 7 Phenomena (Early 3.75E-05 0.00E+00 0.00E+00 0.00E+00 and Late)

Bypass (Interfacing 8 3.92E-06 0.00E+00 0.00E+00 0.00E+00 System LOCA)

CDF All CET end states 6.56E-05 0.00E+00 0.00E+00 0.00E+00 1

Based on data developed in Section 4.4. Only Classes 1 and 3b are impacted by the corrosion analysis.

The increase in Class 3b frequency leads to a reduction in Class 1 frequency to preserve overall CDF.

Table 6-5:

PTN Population Dose Estimates for Population Within 50 Miles (EE Sensitivity)

Accident Classes Person-Rem (50 (Containment Release Description miles)

Type) 1 No Containment Failure 2.42E+04 2 Large Isolation Failures (Failure to Close) 4.58E+06 3a Small Isolation Failures (liner breach) 2.42E+05 3b Large Isolation Failures (liner breach) 2.42E+06 4 Small Isolation Failures (Failure to seal -Type B) N/A 5 Small Isolation Failures (Failure to sealType C) N/A 6 Other Isolation Failures (e.g., dependent failures) N/A 7 Failures Induced by Phenomena (Early and Late) 5.42E+06 8 Bypass (Interfacing System LOCA) 6.38E+06 Page l 50

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 51 of 82 Table 6-6a: PTN Unit 3 Annual Dose as a Function of Accident Class; Characteristic of Conditions for ILRT Required 3/10 Years (EE Sensitivity)

EPRI Methodology Plus Accident EPRI Methodology Corrosion Change Due to Classes Person-Rem Description Corrosion Person-(Containment (50 miles) Frequency Person-Rem/yr Frequency Person-Rem/yr Rem/yr(1)

Release Type) (per Rx-yr) (50 miles) (per Rx-yr) (50 miles) 1 No Containment Failure (2) 2.42E+04 2.59E-05 6.27E-01 2.59E-05 6.27E-01 -1.29E-05 2 Large Isolation Failures (Failure to Close) 4.58E+06 3.05E-09 1.39E-02 3.05E-09 1.39E-02 0.00E+00 3a Small Isolation Failures (liner breach) 2.42E+05 5.52E-07 1.33E-01 5.52E-07 1.33E-01 0.00E+00 3b Large Isolation Failures (liner breach) 2.42E+06 1.38E-07 3.33E-01 1.38E-07 3.35E-01 1.29E-03 4 Small Isolation Failures (Failure to seal -Type B) N/A N/A N/A N/A N/A N/A 5 Small Isolation Failures (Failure to sealType C) N/A N/A N/A N/A N/A N/A 6 Other Isolation Failures (e.g.,

dependent failures) N/A N/A N/A N/A N/A N/A 7 Failures Induced by Phenomena (Early and Late) 5.42E+06 3.33E-05 1.81E+02 3.33E-05 1.81E+02 0.00E+00 8 Bypass (Interfacing System LOCA) 6.38E+06 3.94E-06 2.51E+01 3.94E-06 2.51E+01 0.00E+00 CDF All CET end states N/A 6.39E-05 2.07E+02 6.39E-05 2.07E+02 1.28E-03

1) Only release Classes 1 and 3b are affected by the corrosion analysis. The increase in Class3b frequency leads to a reduction in Class1 frequency to preserve overall CDF, thus the Person-Rem change for Class1 is negative.

2) Characterized as 1La release magnitude consistent with the derivation of the ILRT non-detection failure probability for ILRTs. Release Classes 3a and 3b include failures of containment to meet the Technical Specification leak rate.

Page l 51

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 52 of 82 Table 6-6b: PTN Unit 4 Annual Dose as a Function of Accident Class; Characteristic of Conditions for ILRT Required 3/10 Years (EE Sensitivity)

EPRI Methodology Plus Accident EPRI Methodology Corrosion Change Due to Classes Person-Rem Description Corrosion Person-(Containment (50 miles) Frequency Person-Rem/yr Frequency Person-Rem/yr Rem/yr(1)

Release Type) (per Rx-yr) (50 miles) (per Rx-yr) (50 miles) 1 No Containment Failure (2) 2.42E+04 2.35E-05 5.68E-01 2.35E-05 5.68E-01 -1.33E-05 2 Large Isolation Failures (Failure to Close) 4.58E+06 3.58E-09 1.64E-02 3.58E-09 1.64E-02 0.00E+00 3a Small Isolation Failures (liner breach) 2.42E+05 5.68E-07 1.37E-01 5.68E-07 1.37E-01 0.00E+00 3b Large Isolation Failures (liner breach) 2.42E+06 1.42E-07 3.43E-01 1.42E-07 3.44E-01 1.33E-03 4 Small Isolation Failures (Failure to seal -Type B) N/A N/A N/A N/A N/A N/A 5 Small Isolation Failures (Failure to sealType C) N/A N/A N/A N/A N/A N/A 6 Other Isolation Failures (e.g.,

dependent failures) N/A N/A N/A N/A N/A N/A 7 Failures Induced by Phenomena (Early and Late) 5.42E+06 3.75E-05 2.03E+02 3.75E-05 2.03E+02 0.00E+00 8 Bypass (Interfacing System LOCA) 6.38E+06 3.92E-06 2.50E+01 3.92E-06 2.50E+01 0.00E+00 CDF All CET end states N/A 6.56E-05 2.29E+02 6.56E-05 2.29E+02 1.31E-03

1) Only release Classes 1 and 3b are affected by the corrosion analysis. The increase in Class3b frequency leads to a reduction in Class1 frequency to preserve overall CDF, thus the Person-Rem change for Class1 is negative.

2) Characterized as 1La release magnitude consistent with the derivation of the ILRT non-detection failure probability for ILRTs. Release Classes 3a and 3b include failures of containment to meet the Technical Specification leak rate.

Page l 52

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 53 of 82 Table 6-7a:

PTN Unit 3 Annual Dose as a Function of Accident Class; Characteristic of Conditions for ILRT Required 1/10 Years (EE Sensitivity)

Accident EPRI Methodology Plus EPRI Methodology Change Due to Classes Person-Rem Corrosion Description Corrosion Person-(Containment (50 miles) Frequency Person-Rem/yr Frequency Person-Rem/yr Rem/yr(1)

Release Type) (per Rx-yr) (50 miles) (per Rx-yr) (50 miles) 1 No Containment Failure (2) 2.42E+04 2.43E-05 5.88E-01 2.43E-05 5.88E-01 -7.39E-05 2 Large Isolation Failures 4.58E+06 3.05E-09 1.39E-02 3.05E-09 1.39E-02 0.00E+00 (Failure to Close) 3a Small Isolation Failures (liner 2.42E+05 1.84E-06 4.44E-01 1.84E-06 4.44E-01 0.00E+00 breach) 3b Large Isolation Failures (liner 2.42E+06 4.59E-07 1.11E+00 4.62E-07 1.12E+00 7.39E-03 breach) 4 Small Isolation Failures N/A N/A N/A N/A N/A N/A (Failure to seal -Type B) 5 Small Isolation Failures N/A N/A N/A N/A N/A N/A (Failure to sealType C) 6 Other Isolation Failures (e.g., N/A N/A N/A N/A N/A N/A dependent failures) 7 Failures Induced by 5.42E+06 3.33E-05 1.81E+02 3.33E-05 1.81E+02 0.00E+00 Phenomena (Early and Late) 8 Bypass (Interfacing System 6.38E+06 3.94E-06 2.51E+01 3.94E-06 2.51E+01 0.00E+00 LOCA)

CDF All CET end states N/A 6.39E-05 2.08E+02 6.39E-05 2.08E+02 7.32E-03

1) Only release Classes 1 and 3b are affected by the corrosion analysis. The increase in Class3b frequency leads to a reduction in Class1 frequency to preserve overall CDF, thus the Person-Rem change for Class1 is negative.

2) Characterized as 1La release magnitude consistent with the derivation of the ILRT non-detection failure probability for ILRTs. Release Classes 3a and 3b include failures of containment to meet the Technical Specification leak rate.

Page l 53

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 54 of 82 Table 6-7b:

PTN Unit 4 Annual Dose as a Function of Accident Class; Characteristic of Conditions for ILRT Required 1/10 Years (EE Sensitivity)

EPRI Methodology Plus Accident Classes EPRI Methodology Change Due to Person-Rem Corrosion (Containment Description Corrosion Person-(50 miles) Frequency Person-Rem/yr Frequency Person-Rem/yr Rem/yr(1)

Release Type)

(per Rx-yr) (50 miles) (per Rx-yr) (50 miles) 1 No Containment Failure (2) 2.42E+04 2.18E-05 5.28E-01 2.18E-05 5.28E-01 -7.60E-05 2 Large Isolation Failures 4.58E+06 3.58E-09 1.64E-02 3.58E-09 1.64E-02 0.00E+00 (Failure to Close) 3a Small Isolation Failures 2.42E+05 1.89E-06 4.57E-01 1.89E-06 4.57E-01 0.00E+00 (liner breach) 3b Large Isolation Failures 2.42E+06 4.72E-07 1.14E+00 4.76E-07 1.15E+00 7.60E-03 (liner breach) 4 Small Isolation Failures N/A N/A N/A N/A N/A N/A (Failure to seal -Type B) 5 Small Isolation Failures N/A N/A N/A N/A N/A N/A (Failure to sealType C) 6 Other Isolation Failures N/A N/A N/A N/A N/A N/A (e.g., dependent failures) 7 Failures Induced by 5.42E+06 3.75E-05 2.03E+02 3.75E-05 2.03E+02 0.00E+00 Phenomena (Early and Late) 8 Bypass (Interfacing System 6.38E+06 3.92E-06 2.50E+01 3.92E-06 2.50E+01 0.00E+00 LOCA)

CDF All CET end states N/A 6.56E-05 2.30E+02 6.56E-05 2.30E+02 7.53E-03

1) Only release Classes 1 and 3b are affected by the corrosion analysis. The increase in Class3b frequency leads to a reduction in Class1 frequency to preserve overall CDF, thus the Person-Rem change for Class1 is negative.

2) Characterized as 1La release magnitude consistent with the derivation of the ILRT non-detection failure probability for ILRTs. Release Classes 3a and 3b include failures of containment to meet the Technical Specification leak rate.

Page l 54

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 55 of 82 Table 6-8a:

PTN Unit 3 Annual Dose as a Function of Accident Class; Characteristic of Conditions for ILRT Required 1/15 Years (EE Sensitivity)

EPRI Methodology Plus Accident Classes EPRI Methodology Change Due to Person-Rem Corrosion (Containment Description Corrosion Person-(50 miles) Frequency Person-Rem/yr Frequency Person-Rem/yr Rem/yr(1)

Release Type)

(per Rx-yr) (50 miles) (per Rx-yr) (50 miles) 1 No Containment Failure (2) 2.42E+04 2.32E-05 5.60E-01 2.32E-05 5.60E-01 -1.71E-04 2 Large Isolation Failures (Failure to Close) 4.58E+06 3.05E-09 1.39E-02 3.05E-09 1.39E-02 0.00E+00 3a Small Isolation Failures (liner breach) 2.42E+05 2.76E-06 6.67E-01 2.76E-06 6.67E-01 0.00E+00 3b Large Isolation Failures (liner breach) 2.42E+06 6.90E-07 1.67E+00 6.97E-07 1.68E+00 1.71E-02 4 Small Isolation Failures (Failure to seal -Type B) N/A N/A N/A N/A N/A N/A 5 Small Isolation Failures (Failure to sealType C) N/A N/A N/A N/A N/A N/A 6 Other Isolation Failures (e.g., dependent failures) N/A N/A N/A N/A N/A N/A 7 Failures Induced by Phenomena (Early and 5.42E+06 3.33E-05 1.81E+02 3.33E-05 1.81E+02 0.00E+00 Late) 8 Bypass (Interfacing System LOCA) 6.38E+06 3.94E-06 2.51E+01 3.94E-06 2.51E+01 0.00E+00 CDF All CET end states N/A 6.39E-05 2.09E+02 6.39E-05 2.09E+02 1.69E-02

1) Only release Classes 1 and 3b are affected by the corrosion analysis. The increase in Class3b frequency leads to a reduction in Class1 frequency to preserve overall CDF, thus the Person-Rem change for Class1 is negative.

2) Characterized as 1La release magnitude consistent with the derivation of the ILRT non-detection failure probability for ILRTs. Release Classes 3a and 3b include failures of containment to meet the Technical Specification leak rate.

Page l 55

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 56 of 82 Table 6-8b:

PTN Unit 4 Annual Dose as a Function of Accident Class; Characteristic of Conditions for ILRT Required 1/15 Years (EE Sensitivity)

Accident EPRI Methodology Plus EPRI Methodology Change Due to Classes Person-Rem Corrosion Description Corrosion Person-(Containment (50 miles) Frequency Person-Rem/yr Frequency Person-Rem/yr Rem/yr(1)

Release Type) (per Rx-yr) (50 miles) (per Rx-yr) (50 miles) 1 No Containment Failure 2.42E+04 2.07E-05 4.99E-01 2.06E-05 4.99E-01 -1.76E-04 (2) 2 Large Isolation Failures (Failure to Close) 4.58E+06 3.58E-09 1.64E-02 3.58E-09 1.64E-02 0.00E+00 3a Small Isolation Failures (liner breach) 2.42E+05 2.84E-06 6.86E-01 2.84E-06 6.86E-01 0.00E+00 3b Large Isolation Failures (liner breach) 2.42E+06 7.09E-07 1.71E+00 7.17E-07 1.73E+00 1.76E-02 4 Small Isolation Failures (Failure to seal -Type N/A N/A N/A N/A N/A N/A B) 5 Small Isolation Failures (Failure to sealType N/A N/A N/A N/A N/A N/A C) 6 Other Isolation Failures (e.g., dependent N/A N/A N/A N/A N/A N/A failures) 7 Failures Induced by Phenomena (Early and 5.42E+06 3.75E-05 2.03E+02 3.75E-05 2.03E+02 0.00E+00 Late) 8 Bypass (Interfacing System LOCA) 6.38E+06 3.92E-06 2.50E+01 3.92E-06 2.50E+01 0.00E+00 CDF All CET end states N/A 6.56E-05 2.31E+02 6.56E-05 2.31E+02 1.74E-02

1) Only release Classes 1 and 3b are affected by the corrosion analysis. The increase in Class3b frequency leads to a reduction in Class1 frequency to preserve overall CDF, thus the Person-Rem change for Class1 is negative.

2) Characterized as 1La release magnitude consistent with the derivation of the ILRT non-detection failure probability for ILRTs. Release Classes 3a and 3b include failures of containment to meet the Technical Specification leak rate.

Page l 56

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 57 of 82 Table 6-9a:

PTN Unit 3 ILRT Cases: Base, 3 to 10, and 3 to 15 Yr Extensions Including Age Adjusted Steel Liner Corrosion Likelihood (EE Sensitivity)

EPRI DOSE Base Case 3 in 10 Years Extend to 1 in 10 Years Extend to 1 in 15 Years Class Person-Rem CDF/Yr Person- CDF/Yr Person- CDF/Yr Person-Rem/Yr Rem/Yr Rem/Yr 1 2.42E+04 2.59E-05 6.27E-01 2.43E-05 5.88E-01 2.32E-05 5.60E-01 2 4.58E+06 3.05E-09 1.39E-02 3.05E-09 1.39E-02 3.05E-09 1.39E-02 3a 2.42E+05 5.52E-07 1.33E-01 1.84E-06 4.44E-01 2.76E-06 6.67E-01 3b 2.42E+06 1.38E-07 3.35E-01 4.62E-07 1.12E+00 6.97E-07 1.68E+00 7 5.42E+06 3.33E-05 1.81E+02 3.33E-05 1.81E+02 3.33E-05 1.81E+02 8 6.38E+06 3.94E-06 2.51E+01 3.94E-06 2.51E+01 3.94E-06 2.51E+01 Total N/A 6.39E-05 2.07E+02 6.39E-05 2.08E+02 6.39E-05 2.09E+02 ILRT Dose Rate from 4.68E-01 1.56E+00 2.35E+00 3a and 3b Delta Total From 3 yr N/A 1.05E+00 1.82E+00 Dose Rate From 10 yr N/A N/A 7.61E-01

% change in dose From 3 yr 0.03% 0.51% 0.88%

rate from base From 10 yr N/A N/A 0.37%

3b Frequency (LERF) 1.38E-07 4.62E-07 6.97E-07 Delta From 3 yr N/A 3.24E-07 5.58E-07 LERF From 10 yr N/A N/A 2.34E-07 CCFP % 58.55% 59.05% 59.42%

Delta From 3 yr N/A 0.51% 0.87%

CCFP

% From 10 yr N/A N/A 0.37%

Page l 57

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 58 of 82 Table 6-9b:

PTN Unit 4 ILRT Cases: Base, 3 to 10, and 3 to 15 Yr Extensions Including Age Adjusted Steel Liner Corrosion Likelihood (EE Sensitivity)

EPRI DOSE Base Case 3 in 10 Years Extend to 1 in 10 Years Extend to 1 in 15 Years Class Person-Rem CDF/Yr Person- CDF/Yr Person- CDF/Yr Person-Rem/Yr Rem/Yr Rem/Yr 1 2.42E+04 2.35E-05 5.68E-01 2.18E-05 5.28E-01 2.06E-05 4.99E-01 2 4.58E+06 3.58E-09 1.64E-02 3.58E-09 1.64E-02 3.58E-09 1.64E-02 3a 2.42E+05 5.68E-07 1.37E-01 1.89E-06 4.57E-01 2.84E-06 6.86E-01 3b 2.42E+06 1.42E-07 3.44E-01 4.76E-07 1.15E+00 7.17E-07 1.73E+00 7 5.42E+06 3.75E-05 2.03E+02 3.75E-05 2.03E+02 3.75E-05 2.03E+02 8 6.38E+06 3.92E-06 2.50E+01 3.92E-06 2.50E+01 3.92E-06 2.50E+01 Total N/A 6.56E-05 2.29E+02 6.56E-05 2.30E+02 6.56E-05 2.31E+02 ILRT Dose Rate from 4.81E-01 1.61E+00 2.42E+00 3a and 3b Delta Total From 3 yr N/A 1.08E+00 1.87E+00 Dose Rate From 10 yr N/A N/A 7.83E-01 chang e in From 3 yr N/A 0.47% 0.81%

dose rate from base From 10 yr N/A N/A 0.34%

3b Frequency (LERF) 1.42E-07 4.76E-07 7.17E-07 Delta From 3 yr N/A 3.33E-07 5.74E-07 LERF From 10 yr N/A N/A 2.41E-07 CCFP % 63.33% 63.84% 64.21%

Delta From 3 yr N/A 0.51% 0.88%

CCFP

% From 10 yr N/A N/A 0.37%

RG 1.174 and EPRI Report 1009325 provide guidance for determining acceptable changes in risk due to extension of the ILRT interval as summarized in Tables 6-10a and 6-10b. 100% of the frequency of Class 3b sequences can be used as a conservative first-order estimate to approximate the potential increase in LERF from the ILRT interval extension (consistent with the EPRI guidance methodology). Based on a 3 in 10 year test interval from Tables 6-6a and 6-6b, the Class 3b LERF contribution (including corrosion) is 1.38E-07/year for Unit 3 and 1.42E-Page l 58

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 59 of 82 07/year in Unit 4. Based on a 1 in 15 year test interval, from Tables 6-8a and 6-8b, the Class 3b LERF contribution (including corrosion) increases to 6.97E-07/year for Unit 3 and 7.17E-07/year in Unit 4. This results in a LERF of 5.58E-07/yr for Unit 3 and 5.74E-07/yr for Unit 4.

The total combined increase in LERF, CCFP, and change in dose rate due to the extension of the ILRT interval to 15 years is summarized below:

Table 6-10a: Unit 3 Acceptance Criteria Summary Contributor LERF (/yr) Person-rem/yr Dose (%) CCFP Internal Events 2.54E-09 8.27E-03 0.88% 0.87%

External Events 5.58E-07 1.82E+00 0.88% 0.87%

Total 5.61E-07 1.82E+00 0.88% 0.87%

Acceptance Criteria <1E-06 <1.0 person-rem/yr OR <1.0% <1.5%

Table 6-10b: Acceptance Criteria Summary Contributor LERF (/yr) person-rem/yr dose (%) CCFP Internal Events 2.55E-09 8.28E-03 0.81% 0.88%

External Events 5.74E-07 1.87E+00 0.81% 0.88%

Total 5.77E-07 1.88E+00 0.81% 0.88%

Acceptance Criteria <1E-06 <1.0 person-rem/yr OR <1.0% <1.5%

The 5.77E-07/yr increase in LERF (Unit 4) due to the combined internal and external events from extending the ILRT frequency from 3-per-10 years to 1-per-15 years falls within Region II between 1.0E-7 to 1.0E-6 per reactor year ("Small Change" in risk) of the RG 1.174 acceptance guidelines. Per RG 1.174, when the calculated increase in LERF due to the proposed plant change is in the "Small Change" range, the risk assessment must also reasonably show that the total LERF is less than 1.0E-05/yr. The baseline LERF for Unit 3 is 1.33E-06/yr and the baseline LERF for Unit 4 is 1.57E-06/yr. The LERF increase due to ILRT interval extension and total resulting LERF for each unit is summarized below:

Table 6-11a: U3 Total LERF Summary Internal Events EE/IF Sum Original LERF (/yr) 1.11E-08 1.31E-06 1.33E-06 3b LERF increase for 15 yr test interval (from 3 yr) (/yr) 2.54E-09 5.58E-07 5.61E-07 Total LERF (/yr) 1.36E-08 1.87E-06 1.89E-06 Table 6-11b: U4 Total LERF Summary Internal Events EE/IF Sum Original LERF (/yr) 1.11E-08 1.55E-06 1.57E-06 3b LERF increase for 15 yr test interval (from 3 yr) (/yr) 2.55E-09 5.74E-07 5.77E-07 Total LERF due to ILRT extension (/yr) 1.36E-08 2.13E-06 2.14E-06 Page l 59

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 60 of 82 The total increase in LERF due to the extension of the Integrated Leak Rate Testing interval to 15 years is shown to be between 1.0E-06/yr and 1.0E-07/yr and total resulting LERF is less than 1.0E-05/year for both Unit 3 and Unit 4, therefore the increase in LERF is considered small.

6.3.3 Release Class Distribution - Quantitative Evaluation It is assumed that the distribution of the internal flooding, internal fire, and external events (IF/EE) contributions to core damage frequency will be similar to that of internal events (Reference 8.18). The percent contribution of the total CDF to each accident class is provided in Table 4.2-3.

Sensitivity cases were developed to gain an understanding of the sensitivity of the results to the release class distribution. As described in Reference 8.18, the PTN Fire PRA has a lower proportion of CDF which contributes to LERF, indicating that a higher proportion of Class 1 (containment remains intact) and/or Class 7 (accidents induced by severe accident phenomena) scenarios exist in the CDF sequences caused by fire. Although the proportions of Class 2 and Class 8 sequences are already low for PTN, the release class distribution was adjusted to reduce the Class 2 and Class 8 proportions while increasing the Class 1 and Class 7 proportions. Lower proportions of Class 2/Class 8 releases increase the overall LERF and CCFP due to the ILRT extension to 15 years.

Due to the similarity of the Unit 3 and Unit 4 results this sensitivity is documented only for Unit

3. In the first sensitivity case, the proportions of both the Class 2 and Class 8 sequences were decreased by 50%, and the proportions of Class 1 and Class 7 sequences were increased accordingly. The second sensitivity assumes the Class 2 and Class 8 release class proportions are 0%, and the proportions of Class 1 / Class 7 sequences are increased accordingly. The second sensitivity case is conservative, as the PTN Fire PRA (Reference 8.47) identifies several Class 8 scenarios (containment bypass) as dominant risk contributors to both U3 and U4 LERF (Reference 8.47, Appendix F). All other calculations remain the same as those presented in Section 6.3.2. The calculations are shown in Tables 6-12 through 6-18.

Table 6-12: EPRI Release Class Distributions Sensitivity Percent of CDF EPRI Release Class Baseline Distribution Unit 3 Sensitivity 1 Unit 3 Sensitivity 2 1 41.67% 43.21% 44.75%

2 0.005% 0.002% 0%

7 52.16% 53.70% 55.25%

8 6.16% 3.08% 0%

Total 100% 100% 100%

Page l 60

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 61 of 82 Table 6-13a: Unit 3 EPRI Release Class Frequencies by Hazard (Distribution Sensitivity 1)

Frequency by Hazard Type (per year)

EPRI Release Distribution Class (% of CDF) Internal Internal Internal Seismic Other Events Flood Fire CDF 100% 2.91E-07 1.37E-08 6.22E-05 6.98E-07 1.00E-06 1 43.21% 1.26E-07 5.92E-09 2.69E-05 3.02E-07 4.32E-07 2 0.00% 6.94E-12 3.27E-13 1.48E-09 1.66E-11 2.38E-11 7 53.70% 1.56E-07 7.36E-09 3.34E-05 3.75E-07 5.37E-07 8 3.08% 8.97E-09 4.22E-10 1.92E-06 2.15E-08 3.08E-08 LERF - 1.11E-08 5.44E-11 1.25E-06 2.66E-08 3.81E-08 Table 6-13b: Unit 3 EPRI Release Class Frequencies by Hazard (Distribution Sensitivity 2)

Frequency by Hazard Type (per year)

EPRI Release Distribution Class (% of CDF) Internal Internal Internal Fire Seismic Other Events Flood CDF 100% 2.91E-07 1.37E-08 6.22E-05 6.98E-07 1.00E-06 1 44.75% 1.30E-07 6.13E-09 2.78E-05 3.12E-07 4.48E-07 2 0.00% 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 7 55.25% 1.61E-07 7.57E-09 3.44E-05 3.86E-07 5.53E-07 8 0.00% 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 LERF - 1.11E-08 5.44E-11 1.25E-06 2.66E-08 3.81E-08 Table 6-14a PTN Unit 3 Categorized Accident Classes and Frequencies, Sensitivity 1 EPRI Adjusted Frequency Using Scale Class Frequency (-/yr)

Class Factor of 1.0 (per yr) 1 2.76E-05 2.76E-05 2 1.52E-09 1.52E-09 7 3.43E-05 3.43E-05 8 1.97E-06 1.97E-06 Total 6.39E-05 6.39E-05 Frequency 3a =0.0092*(CDF-Class2-Class8) 5.70E-07 3b =0.0023*(CDF-Class2-Class8) 1.42E-07 Page l 61

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 62 of 82 Table 6-14b: PTN Unit 3 Categorized Accident Classes and Frequencies, Sensitivity 2 Adjusted Frequency Using EPRI Class Class Frequency (-/yr)

Scale Factor of 1.0 (per yr) 1 2.86E-05 2.86E-05 2 0.00E+00 0.00E+00 7 3.53E-05 3.53E-05 8 0.00E+00 0.00E+00 Total Frequency 6.39E-05 6.39E-05 3a =0.0092*(CDF-Class2-Class8) 5.88E-07 3b =0.0023*(CDF-Class2-Class8) 1.47E-07 Table 6-15a: Unit 3 Radionuclide Release Frequencies as a Function of Accident Class, Sensitivity 1 Accident Frequency (per Rx-yr)

Classes due to due to due to Description EPRI (Containment Corrosion Corrosion Corrosion Release Type) Methodology (3 in 10 yr) (1 in 10 yr) (1 in 15 yr)

No Containment 1 2.69E05 -5.51E-10 -3.16E-09 -7.31E-09 Failure Large Isolation 2 Failures (Failure to 1.52E-09 0.00E+00 0.00E+00 0.00E+00 Close)

Small Isolation 3a Failures (liner 5.70E-07 0.00E+00 0.00E+00 0.00E+00 breach)

Large Isolation 3b Failures (liner 1.42E-07 5.51E-10 3.16E-09 7.31E-09 breach)

Small Isolation 4 Failures (Failure to N/A 0.00E+00 0.00E+00 0.00E+00 seal -Type B)

Small Isolation 5 Failures (Failure to N/A 0.00E+00 0.00E+00 0.00E+00 sealType C)

Other Isolation 6 Failures (e.g., N/A 0.00E+00 0.00E+00 0.00E+00 dependent failures)

Failures Induced by 7 Phenomena (Early 3.43E-05 0.00E+00 0.00E+00 0.00E+00 and Late)

Bypass (Interfacing 8 System LOCA) 1.97E-06 0.00E+00 0.00E+00 0.00E+00 CDF All CET end states 6.39E-05 0.00E+00 0.00E+00 0.00E+00 Page l 62

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 63 of 82 Table 6-15b: Unit 3 Radionuclide Release Frequencies as a Function of Accident Class, Sensitivity 2 Accident Frequency (per Rx-yr)

Classes due to due to due to Description EPRI (Containment Corrosion Corrosion Corrosion Release Type) Methodology (3 in 10 yr) (1 in 10 yr) (1 in 15 yr)

No Containment 1 2.79E-05 -5.69E-10 -3.26E-09 -7.54E-09 Failure Large Isolation 2 Failures (Failure to 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Close)

Small Isolation 3a Failures (liner 5.88E-07 0.00E+00 0.00E+00 0.00E+00 breach)

Large Isolation 3b Failures (liner 1.47E-07 5.69E-10 3.26E-09 7.54E-09 breach)

Small Isolation 4 Failures (Failure to N/A 0.00E+00 0.00E+00 0.00E+00 seal -Type B)

Small Isolation 5 Failures (Failure to N/A 0.00E+00 0.00E+00 0.00E+00 sealType C)

Other Isolation 6 Failures (e.g., N/A 0.00E+00 0.00E+00 0.00E+00 dependent failures)

Failures Induced by 7 Phenomena (Early 3.53E-05 0.00E+00 0.00E+00 0.00E+00 and Late)

Bypass (Interfacing 8 System LOCA) 0.00E+00 0.00E+00 0.00E+00 0.00E+00 CDF All CET end states 6.39E-05 0.00E+00 0.00E+00 0.00E+00 Page l 63

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 64 of 82 Table 6-16a: PTN Unit 3 Annual Dose as a Function of Accident Class; Characteristic of Conditions for ILRT Required 3/10 Years (Sensitivity 1)

Accident EPRI Methodology Plus Classes EPRI Methodology Change Due to Person-Rem (50 Corrosion (Containment Description Corrosion Person-miles) Frequency Person-Rem/yr Frequency Person-Rem/yr Release Rem/yr Type) (per Rx-yr) (50 miles) (per Rx-yr) (50 miles)

No Containment 2.42E+04 2.69E-05 6.50E-01 2.69E-05 6.50E-01 -1.33E-05 1

Failure Large Isolation Failures 4.58E+06 1.52E-09 6.97E-03 1.52E-09 6.97E-03 0.00E+00 2

(Failure to Close)

Small Isolation Failures 2.42E+05 5.70E-07 1.38E-01 5.70E-07 1.38E-01 0.00E+00 3a (liner breach)

Large Isolation Failures 2.42E+06 1.42E-07 3.44E-01 1.43E-07 3.46E-01 1.33E-03 3b (liner breach)

Small Isolation Failures N/A N/A N/A N/A N/A N/A 4 (Failure to seal -Type B)

Small Isolation Failures N/A N/A N/A N/A N/A N/A 5 (Failure to sealType C)

Other Isolation Failures N/A N/A N/A N/A N/A N/A 6 (e.g., dependent failures)

Failures Induced by 5.42E+06 3.43E-05 1.86E+02 3.43E-05 1.86E+02 0.00E+00 7 Phenomena (Early and Late)

Bypass (Interfacing 6.38E+06 1.97E-06 1.26E+01 1.97E-06 1.26E+01 0.00E+00 8

System LOCA)

CDF All CET end states N/A 6.39E-05 2.00E+02 6.39E-05 2.00E+02 1.32E-03 Page l 64

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 65 of 82 Table 6-16b: PTN Unit 3 Annual Dose as a Function of Accident Class; Characteristic of Conditions for ILRT Required 3/10 Years (Sensitivity 2)

Accident EPRI Methodology Plus Classes EPRI Methodology Change Due to Person-Rem (50 Corrosion (Containment Description Corrosion Person-miles) Frequency Person-Rem/yr Frequency Person-Rem/yr Release Rem/yr Type) (per Rx-yr) (50 miles) (per Rx-yr) (50 miles) 1 No Containment Failure 2.42E+04 2.79E-05 6.73E-01 2.79E-05 6.73E-01 -1.37E-05 Large Isolation Failures 2 4.58E+06 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 (Failure to Close)

Small Isolation Failures 3a 2.42E+05 5.88E-07 1.42E-01 5.88E-07 1.42E-01 0.00E+00 (liner breach)

Large Isolation Failures 3b 2.42E+06 1.47E-07 3.55E-01 1.48E-07 3.57E-01 1.37E-03 (liner breach)

Small Isolation Failures 4 (Failure to seal -Type N/A N/A N/A N/A N/A N/A B)

Small Isolation Failures 5 (Failure to sealType N/A N/A N/A N/A N/A N/A C)

Other Isolation Failures 6 (e.g., dependent N/A N/A N/A N/A N/A N/A failures)

Failures Induced by 7 Phenomena (Early and 5.42E+06 3.53E-05 1.91E+02 3.53E-05 1.91E+02 0.00E+00 Late)

Bypass (Interfacing 8 6.38E+06 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 System LOCA)

CDF All CET end states N/A 6.39E-05 1.93E+02 6.39E-05 1.93E+02 1.36E-03 Page l 65

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 66 of 82 Table 6-17a: PTN Unit 3 Annual Dose as a Function of Accident Class; Characteristic of Conditions for ILRT Required 1/15 Years (Sensitivity 1)

EPRI Methodology Plus Accident Classes EPRI Methodology Change Due to Person-Rem (50 Corrosion (Containment Description Corrosion miles) Frequency Person-Rem/yr Frequency Person-Rem/yr Release Type) Person-Rem/yr (per Rx-yr) (50 miles) (per Rx-yr) (50 miles) 1 No Containment Failure 2.42E+04 2.41E-05 5.81E-01 2.40E-05 5.81E-01 -1.77E-04 Large Isolation Failures 2 (Failure to Close) 4.58E+06 1.52E-09 6.97E-03 1.52E-09 6.97E-03 0.00E+00 Small Isolation Failures 3a (liner breach) 2.42E+05 2.85E-06 6.89E-01 2.85E-06 6.89E-01 0.00E+00 Large Isolation Failures 3b (liner breach) 2.42E+06 7.12E-07 1.72E+00 7.20E-07 1.74E+00 1.77E-02 Small Isolation Failures 4 (Failure to seal -Type B) N/A N/A N/A N/A N/A N/A Small Isolation Failures 5 (Failure to sealType N/A N/A N/A N/A N/A N/A C)

Other Isolation Failures 6 (e.g., dependent failures) N/A N/A N/A N/A N/A N/A Failures Induced by 7 Phenomena (Early and 5.42E+06 3.43E-05 1.86E+02 3.43E-05 1.86E+02 0.00E+00 Late)

Bypass (Interfacing 8 System LOCA) 6.38E+06 1.97E-06 1.26E+01 1.97E-06 1.26E+01 0.00E+00 CDF All CET end states N/A 6.39E-05 2.02E+02 6.39E-05 2.02E+02 1.75E-02 Page l 66

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 67 of 82 Table 6-17b: PTN Unit 3 Annual Dose as a Function of Accident Class; Characteristic of Conditions for ILRT Required 1/15 Years (Sensitivity 2)

EPRI Methodology Plus Accident Classes EPRI Methodology Change Due to Person-Rem (50 Corrosion (Containment Description Corrosion miles) Frequency Person-Rem/yr Frequency Person-Rem/yr Release Type) Person-Rem/yr (per Rx-yr) (50 miles) (per Rx-yr) (50 miles) 1 No Containment Failure 2.42E+04 2.49E-05 6.02E-01 2.49E-05 6.02E-01 -1.82E-04 Large Isolation Failures 2 (Failure to Close) 4.58E+06 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Small Isolation Failures 3a (liner breach) 2.42E+05 2.94E-06 7.11E-01 2.94E-06 7.11E-01 0.00E+00 Large Isolation Failures 3b (liner breach) 2.42E+06 7.35E-07 1.78E+00 7.43E-07 1.79E+00 1.82E-02 Small Isolation Failures 4 (Failure to seal -Type B) N/A N/A N/A N/A N/A N/A Small Isolation Failures 5 (Failure to sealType C) N/A N/A N/A N/A N/A N/A Other Isolation Failures 6 (e.g., dependent failures) N/A N/A N/A N/A N/A N/A Failures Induced by 7 Phenomena (Early and 5.42E+06 3.53E-05 1.91E+02 3.53E-05 1.91E+02 0.00E+00 Late)

Bypass (Interfacing 8 System LOCA) 6.38E+06 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 CDF All CET end states N/A 6.39E-05 1.94E+02 6.39E-05 1.94E+02 1.80E-02 Page l 67

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 68 of 82 Table 6-18a: Sensitivity 1 (including corrosion)

EPRI DOSE Base Case 3 in 10 Years Extend to 1 in 15 Years Class Person-Rem CDF/Yr Person- CDF/Yr Person-Rem/Yr Rem/Yr 1 2.42E+04 2.69E-05 6.50E-01 2.40E-05 5.81E-01 2 4.58E+06 1.52E-09 6.97E-03 1.52E-09 6.97E-03 3a 2.42E+05 5.70E-07 1.38E-01 2.85E-06 6.89E-01 3b 2.42E+06 1.43E-07 3.46E-01 7.20E-07 1.74E+00 7 5.42E+06 3.43E-05 1.86E+02 3.43E-05 1.86E+02 8 6.38E+06 1.97E-06 1.26E+01 1.97E-06 1.26E+01 Total N/A 6.39E-05 2.00E+02 6.39E-05 2.02E+02 ILRT Dose Rate from 4.83E-01 2.43E+00 3a and 3b Delta Total From 3 yr N/A 1.88E+00 Dose Rate change in dose From 3 yr N/A 0.94%

rate from base 3b Frequency (LERF) 1.43E-07 7.20E-07 Delta From 3 yr N/A 5.77E-07 LERF CCFP % 57.01% 57.91%

Delta CCFP From 3 yr N/A 0.90%

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Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 69 of 82 Table 6-18b: Sensitivity 2 (including corrosion)

EPRI DOSE Base Case 3 in 10 Years Extend to 1 in 15 Years Class Person- CDF/Yr Person- CDF/Yr Person-Rem Rem/Yr Rem/Yr 1 2.42E+04 2.79E-05 6.73E-01 2.49E-05 6.02E-01 2 4.58E+06 0.00E+00 0.00E+00 0.00E+00 0.00E+00 3a 2.42E+05 5.88E-07 1.42E-01 2.94E-06 7.11E-01 3b 2.42E+06 1.48E-07 3.57E-01 7.43E-07 1.79E+00 7 5.42E+06 3.53E-05 1.91E+02 3.53E-05 1.91E+02 8 6.38E+06 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Total N/A 6.39E-05 1.93E+02 6.39E-05 1.94E+02 ILRT Dose Rate from 4.99E-01 2.51E+00 3a and 3b Delta Total From 3 yr N/A 1.94E+00 Dose Rate change in dose From 3 yr N/A 1.00%

rate from base 3b Frequency (LERF) 1.48E-07 7.43E-07 Delta From 3 yr N/A 5.95E-07 LERF CCFP % 55.48% 56.41%

Delta CCFP From 3 yr N/A 0.93%

The calculations performed in the tables above indicate that the reduction in Class 2 and Class 8 proportions in the release class distribution does increase the LERF due to the ILRT extension to 15 years. The delta LERF for sensitivity case 1 increases from 5.58E-07/yr to 5.77E-07/yr, while the delta LERF for sensitivity case 2 increases to 5.95E-07/yr. Similar to the conclusions formed in Section 6.3.2, this is considered an acceptable change since the LERF is between 1.0E-07/yr and 1.0E-06/yr and the total LERF remains less than 1.0E-05/year. The adjusted CCFPs are also acceptable in both sensitivity cases. The change in total dose rate is the limiting factor. If Class 2 and Class 8 proportions are reduced to 0%, the resulting change in dose rate is 1%. However, this case is extremely pessimistic and judged to be unrealistic since the dominant external event, fire, is shown to have several Class 8 scenarios as important contributors to risk at PTN (Reference 8.47, Appendix F).

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Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 70 of 82 Table 6-19: Distribution Class Sensitivity Summary LERF Increase Total LERF Case (/yr) (/yr) CCFP Total Dose Rate Baseline 5.58E-07 1.87E-06 0.87% 0.88%

Sensitivity 1 5.77E-07 1.89E-06 0.90% 0.94%

Sensitivity 2 5.95E-07 1.91E-06 0.93% 1.00%

Acceptance Criteria <1.0E-6 <1E-05 <1.5% <1.0%

The results indicate that even if conservative assumptions for the PTN release class distributions are used to maximize the LERF, CCFP, and change in population dose, the conclusions from the base analysis would not change.

Based on these sensitivity cases, the baseline quantitative external events and internal flooding CDF and LERF contributions are judged appropriate for use in the delta LERF impact of the ILRT Type A test interval extension.

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Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 71 of 82 7.0 Conclusions Permanently increasing the Type A ILRT interval to fifteen years is considered to be an insignificantly small change to the PTN risk profile. This conclusion is based on the following results from Section 5.0 and the sensitivity calculations presented in Section 6.0:

RG 1.174 (Reference 8.4) provides guidance for determining the risk impact of plant specific changes to the licensing basis based on changes to CDF and LERF. Since the ILRT does not impact CDF, the relevant criterion is LERF. The increase in LERF resulting from a change in the Type A ILRT test interval from three in ten years to one in fifteen years is conservatively estimated for Unit 3 as 2.54E-09/yr due to internal events contribution and 5.58E-07/yr due to internal flood and external events. The total combined impact for Unit 3 is 5.61E-07/yr.

For Unit 4, the impact is conservatively estimated as 2.55E-09/yr due to internal events contribution and 5.74E-07/yr due to internal flood and external events. The total combined impact for Unit 4 is 5.77E-07/yr.

RG 1.174 (Reference 8.4) states that changes in LERF less than 1.0E-07 per reactor year are considered very small and are acceptable without evaluation of total LERF. Regulatory Guide 1.174 (Reference 8.4) also states that when the calculated increase in LERF is in the small range of 1.0E-07 per reactor year to 1.0E-06 per reactor year, applications will be considered only if it can be reasonably shown that the total LERF is less than 1.0E-05 per reactor year. The impact due to an increase in the Type A ILRT interval to one in fifteen years is very small when considering only internal events. When including the impact from internal flood and external events, the change to LERF is in the small range for both PTN Unit 3 and Unit 4, therefore, the total LERF is evaluated. The resulting total LERF for Unit 3 is 1.89E-06/yr (from Table 6-11a). The resulting total LERF for Unit 4 is 2.14E-06/yr (from Table 6-11b). The total LERF for both Unit 3 and Unit 4 are below the RG 1.174 acceptance criteria for total LERF of 1.0E-05/yr and therefore this change satisfies both the incremental and absolute criteria with regard to the RG 1.174 LERF metric.

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, is 0.88% of the total dose rate for Unit 3 and 0.81% of the total dose rate for Unit 4 when considering internal events only. Including the effect of internal flood and external events, the resulting total change in plant risk is a 0.88% increase in population dose for Unit 3 and a 0.81% increase in population dose for Unit 4. EPRI Report No. 1009325, Revision 2-A states that a very small population dose is defined as an increase of 1.0 person-rem per year or 1% of the total population dose, whichever is less restrictive for the risk impact assessment of the extended ILRT intervals. This is consistent with the NRC Final Safety Evaluation for NEI 94-01 and EPRI Report No. 1009325 (Reference 8.29). Moreover, the risk impact when compared to other severe accident risks is negligible.

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Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 72 of 82 The increase in the conditional containment failure probability (CCFP) from the three in ten year interval to a permanent one time in fifteen year interval is 0.87% for Unit 3 and 0.88%

for Unit 4. EPRI Report No. 1009325, Revision 2-A states that increases in CCFP of 1.5 percentage points are considered very small. This is consistent with the NRC Final Safety Evaluation for NEI 94-01 and EPRI Report No. 1009325 (Reference 8.29). Therefore the increase in conditional containment failure probability is judged to be very small.

7.1.1 Previous Assessments The NRC in NUREG-1493 (Reference 8.6) has previously concluded that:

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

Given the insensitivity of risk to containment leakage rate and the small fraction of leakage paths detected solely by Type A testing, increasing the interval between integrated leakage rate tests is possible with minimal impact on public risk. Beyond testing the performance of containment penetrations, ILRTs also test the integrity of the containment structure.

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

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Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 73 of 82 8.0 References 8.1. Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J, NEI 94-01, Revision 3-A, July 2012.

8.2. Risk Impact Assessment of Revised Containment Leak Rate Testing Intervals, EPRI, Palo Alto, CA EPRI TR-104285, August 1994.

8.3. 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.

8.4. An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant Specific Changes to the Licensing Basis, Regulatory Guide 1.174, Revision 3, January 2018.

8.5. Response to Request for Additional Information Concerning the License Amendment Request for a One-Time Integrated Leakage Rate Test Extension, Letter from Mr. C. H.

Cruse (Calvert Cliffs Nuclear Power Plant) to NRC Document Control Desk, Docket No. 50-317, March 27, 2002.

8.6. Performance-Based Containment Leak-Test Program, NUREG-1493, September 1995.

8.7. Evaluation of Severe Accident Risks: Surry Unit 1, Main Report NUREG/CR-4551, SAND86-1309, Volume 3, Revision 1, Part 1, December 1990.

8.8. Letter from R. J. Barrett (Entergy) to U.S. Nuclear Regulatory Commission, IPN-01-007, January 18, 2001.

8.9. United States Nuclear Regulatory Commission, Indian Point Nuclear Generating Unit No. 3 - Issuance of Amendment Re: Frequency of Performance-Based Leakage Rate Testing (TAC No. MB0178), April 17, 2001.

8.10. Impact of Containment Building Leakage on LWR Accident Risk, Oak Ridge National Laboratory, NUREG/CR-3539, ORNL/TM-8964, April 1984.

8.11. Reliability Analysis of Containment Isolation Systems, Pacific Northwest Laboratory, NUREG/CR-4220, PNL-5432, June 1985.

8.12. Technical Findings and Regulatory Analysis for Generic Safety Issue II.E.4.3 Containment Integrity Check, NUREG-1273, April 1988.

8.13. Review of Light Water Reactor Regulatory Requirements, Pacific Northwest Laboratory, NUREG/CR-4330, PNL-5809, Vol. 2, June 1986.

8.14. Shutdown Risk Impact Assessment for Extended Containment Leakage Testing Intervals Utilizing ORAM', EPRI, Palo Alto, CA TR-105189, Final Report, May 1995.

8.15. Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants, NUREG-1150, December 1990.

8.16. United States Nuclear Regulatory Commission, Reactor Safety Study, WASH-1400, October 1975.

8.17. Turkey Point Units 3 and 4, License Amendment Request 254, Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structure, Systems, and Components (SSCs) for Nuclear Power Reactors, Letter L-2017-156, December 22, 2017 (NRC ADAMS Accession No. ML17363A216).

8.18. S&L Evaluation 2019-01368, Rev. 0, Permanent ILRT Interval Extension Risk Assessment, May 2019.

8.19. Withdrawal of Application to Adopt 10 CFR 50.69, Risk-Informed Categorization and Treatment of Structures, System, and Components for Nuclear Power Reactors, January 19, 2018, (NRC ADAMS Accession No. ML18024A444).

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Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 74 of 82 8.20. Anthony R. Pietrangelo, One-time extensions of containment integrated leak rate test interval - additional information, NEI letter to Administrative Points of Contact, November 30, 2001.

8.21. Letter from J.A. Hutton (Exelon, Peach Bottom) to U.S. Nuclear Regulatory Commission, Docket No. 50-278, License No. DPR-56, LAR-01-00430, dated May 30, 2001.

8.22. Risk Assessment for Joseph M. Farley Nuclear Plant Regarding ILRT (Type A)

Extension Request, prepared for Southern Nuclear Operating Co. By ERIN Engineering and Research, P0293010002-1929-030602, March 2002.

8.23. Letter from D.E. Young (Florida Power, Crystal River) to U.S. Nuclear Regulatory Commission, 3F0401-11, dated April 25, 2001.

8.24. PTN-BFJR-14-009, Revision 0, Turkey Point Seismic CDF Estimate.

8.25. Risk Assessment for Vogtle Electric Generating Plant Regarding the ILRT (Type A)

Extension Request, prepared for Southern Nuclear Operating Co. by ERIN Engineering and Research, February 2003.

8.26. Risk Impact Assessment of Extended Integrated Leak Rate Testing Intervals, EPRI, Palo Alto, CA: 1009325 R2-A. (Also identified as EPRI TR-1018243.)

8.27. Letter from P.P. Sena III (FENOC) to Document Control Desk (NRC), dated June 18, 2009, Beaver Valley Power Station, Unit No. 1, Docket No. 50-334, License No. DPR-66, LER 2009-003-00, Containment Liner Through Wall Defect Due to Corrosion.

8.28. Letter from J.E. Pollock (AEP Indiana Michigan Power) to Document Control Desk (NRC), dated March 16, 2001, submitting LER 316/2000-001-01, Through-Liner Hole Discovered in Containment Liner.

8.29. Final Safety Evaluation For NEI Topical Report 94-01 Revision 2, Industry Guideline for Implementing Performance-Based Option of 10 CFR Part 50, Appendix J and EPRI Report No. 1009325, Revision 2, Risk Impact Assessment of Extended Integrated Leak Rate Test Intervals, June 25, 2008 (ADAMS Accession No. ML081140105).

8.30. Turkey Point Units 3 & 4, Applicants Environmental Report, Operating License Renewal Stage, Docket Nos. 50-250 and 50-251, Revision 1.

8.31. Turkey Point, Units 3 and 4, Docket Nos. 50-250 and 50-251 Proposed License Amendments, Integrated Leak Rate Testing Interval - One Time Extension, Letter L-201-177 (NRC ADAMS Accession No. ML013390251).

8.32. UFSAR 2018 (living document updated as of 04-06-2018), Turkey Point Units 3 & 4, Updated Final Safety Analysis Report.

8.33. Safety/Risk Assessment of Generic Issue (GI) 199, Implications of Updated Probabilistic Seismic Hazard Estimates in Central and Eastern United States on Existing Plants, August 2010, (ADAMS Accession Nos. ML100270639 and ML100270756).

8.34. EPRI NP-6395-D, Probabilistic Seismic Hazard Evaluations at Nuclear Plant Sites in the Central and Eastern United States: Resolution of the Charleston Earthquake Issue, April 1989.

8.35. U.S. Nuclear Regulatory Commission, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Regulatory Guide 1.200, Revision 2, 2009.

8.36. Turkey Point Nuclear Generating Unit Nos. 3 and 4, Issuance of Amendments Regarding Transition to a Risk-Informed Performance-Based Fire Protection Program in Page l 74

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 75 of 82 Accordance with Title 10 of the Code of Federal Regulations Section 50.48(c) (TAC Nos. ME8990 and ME8991), May 28, 2015, (ADAMS Accession No. ML15061A237).

8.37. PTN-BFJR-17-051, PTN Other External Events Screening, Rev. 0.

8.38. Florida Power & Light Company, Turkey Point Nuclear Plant Units 3 and 4 Subsequent License Renewal Application, April 2018, Revision 1 (ADAMS Accession No.ML18113A146).

8.39. Generic Letter 88-20, "Individual Plant Examination of External Events (IPEEE) for Severe Accident Vulnerabilities - 10 CFR 50.54(f), Supplement 4," USNRC, June 1991.

8.40. NEI Letter to USNRC, "Final Revision of Appendix X to NEI 05-04/07-12/12-16, Close-Out of Facts and Observations (F&Os)," February 21, 2017 (ADAMS Accession No. ML17086A451).

8.41. USNRC Letter to Mr. Greg Krueger (NEI), "U.S. Nuclear Regulatory Commission Acceptance on Nuclear Energy Institute Appendix X to Guidance 05-04, 7-12, and 12-13, Close Out of Facts and Observations (F&Os)," May 3, 2017 (ADAMS Accession No. ML17079A427).

8.42. Containment Integrated Leak Rate Test Reports 8.42.1. Turkey Point Plant Procedure 3/4-OSP-051.16, Unit 3 Report, November 2004.

8.42.2. Turkey Point Plant Procedure 3/4-OSP-051.16, Unit 4 Report, May 2005.

8.42.3. Turkey Point Plant Procedure 3/4-OSP-051.16, Unit 3 Report, July 2012.

8.42.4. Turkey Point Plant Procedure 3/4-OSP-051.16, Unit 4 Report, November 2013.

8.42.5. Florida Power and Light Company Turkey Point Units 3 and 4 Operating Procedure 13100.1,Integrated Leakage Rate Test, November 1992.

8.42.6. Florida Power and Light Company Turkey Point Units 3 and 4 Operating Procedure 13100.1,Integrated Leakage Rate Test, October 1991.

8.43. USNRC Letter to Mr. J.A. Stall (FPL Energy), Turkey Point Units 3 and 4 - Issuance of Amendments Regarding One-Time Extension of the Integrated Leak Rate Testing Interval (TAC NOS. MB3249 AND MB3250), January 29, 2002 (ADAMS Accession No. ML020300286) 8.44. PTN-BFJR-00-001, PTN PRA Model Update, Rev. 11.

8.45. PTN-BFJR-099-10, Level 2 Analysis for Turkey Point Units 3 and 4, Rev. 1.

8.46. PTN-BFJR-11-009, PTN Internal Flooding Analysis, Rev. 2.

8.47. PTN-BFJR-16-057, Turkey Point Nuclear Plant FPRA Summary Report NUREG/CR-6850 Task 16, Rev. 1.

8.48. NRC Regulatory Issue Summary 2007-06, Regulatory Guide 1.200 Implementation, March 22, 2007 (ADAMS Accession No. ML07650428).

8.49. PTN-BFJR-18-060, Turkey Point Internal Events PRA Model Results, Rev. 0, October 2018.

8.50. PTN-BFJR-19-006, Turkey Point IE, IF, Fire Finding Closure Report - Enercon, Rev. 1.

8.51. PTN-BFJR-19-002, Turkey Point F&O Closures, Rev. 0.

8.52. NUREG/CR-6338 Resolution of the Direct Containment Heating Issue for All Westinghouse Plants with Large Dry Containments or Sub-atmospheric Containments.

January 1996 (NRC ADAMS Accession No.ML081920672).

8.53. Turkey Point Nuclear Plant, Units 3 and 4, Technical Specification 6.8.4.h.

8.54. Turkey Point Nuclear Plant Units 3 and 4, Applicants Environmental Report, Subsequent License Renewal Stage, January 2018 (NRC ADAMS Accession No. ML18113A145).

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Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 76 of 82 8.55. Application for Renewed Operating Licenses, Turkey Point Units 3 & 4, September 11, 2000.

8.56. Turkey Point Units 3 and 4 - Issuance of Amendments Regarding One-Time Extension of the Integrated Leak Rate Testing Interval (Tac. Nos. MB3249 and MB3250) (NRC ADAMS Accession No. ML020300286).

8.57. Turkey Point Nuclear Generating Unit Nos. 3 and 4 - Issuance of Amendment Regarding Transition License Conditions for Reactor Coolant Pump Seals (EPID L-2018-LLA-0280), March 27, 2019, (NRC ADAMS Accession No. ML19064A903).

8.58. Turkey Point, Units 3 and 4, Docket Nos. 50-250 and 50-251, License Amendment Request 265, Revise NFPA 805 License Condition for Reactor Coolant Pump Seals, Letter No. L-2018-170, October 17, 2018, (NRC ADAMS Accession No. ML18292A842).

8.59. Response to Sixth Request for Additional Information Regarding License Amendment Request 236, Revision to the Technical Specifications to Adopt Risk Informed Completion Times TSTF-505, Revision 1, "Provide Risk-Informed Extended Completion Times - RITSTF Initiative 4b, Letter L-2018-118, June 12, 2018 (NRC ADAMS Accession No. ML18179A162).

8.60. Turkey Point Nuclear Generating Unit Nos. 3 and 4 - Issuance of Amendments Regarding Adoption of Risk-Informed Completion Times in Technical Specifications (CAC Nos. MF5455 and MF5456), December 3, 2018 (NRC ADAMS Accession No. ML18270A429).

8.61. License Amendment Request No. 216, Transition to 10 CFR 50.48(c) - NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition), Letter L-2012-092 and Supplemental Information Regarding License Amendment Request No. 216 Transition to 10 CFR 50.48(c) - NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition), Letter L-2012-354, (NRC ADAMS Accession No. ML12278A106).

8.62. FPL letter L-2014-303 to NRC, "Turkey Point Nuclear Generating Units 3 and 4, Docket Nos. 50-250 and 50-251, Response to Request for Additional Information Regarding License Amendment Request No. 216, Transition to 10 CFR 50.48(c) - NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition)," dated November 5, 2014.

8.63. PWROG-19008-P, Independent Assessment of Facts & Observations Closure of the Turkey Point Probabilistic Risk Assessment, PA-RMSC-1673, Rev. 0-A, June,2019.

8.64. PTN-BFJR-19-022, Turkey Point One-Top All-Events PRA Model, Rev. 0.

8.65. PTN-BFJR-00-001, Turkey Point PRA Model Update, Rev. 12.

8.66. PTN-BFJR-11-009, Turkey Point Internal Flooding Analysis, Rev. 3.

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Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 77 of 82 Appendix A1 PRA Acceptability for the PTN ILRT Interval Extension Risk Impact Assessment Introduction The Internal Events, Internal Flood, and Fire PRA models described below have been peer reviewed. The PRA models credited in this request are the PRA models used in the NFPA 805 application with maintenance updates applied (References 8.44). Capability Category (CC) II of the NRC-endorsed ASME/ANS PRA Standard is the target capability level for the NFPA 805 application. The acceptability of the PRA model was reviewed by the NRC for the license amendment request (LAR) to adopt a risk-informed, performance-based fire protection program in accordance with 10 CFR 50.48(c) and NFPA 805 (2001 Edition) (Reference 8.36) and found to be technically acceptable. The impact of recent NRC-approved LARs to adopt a risk-informed Technical Specification Completion Time (RICT) program and to revise reactor coolant pump (RCP) seal PRA credit is also discussed in this appendix.

As stated in the NRC Final Safety Evaluation for NEI 94-01, Revision 2 and EPRI Report No. 1009325, Revision 2 (Reference 8.29), CC I of the ASME PRA Standard shall be applied as the standard for assessing PRA quality for ILRT extension applications, as approximate values of CDF and LERF and their distribution among release categories are sufficient to support the evaluation of changes to ILRT frequencies. The NRC Safety Evaluation also states the assessment of external events can be taken from existing, previously submitted and approved analyses or other alternate method of assessing an order of magnitude estimate for contribution of the external event to the impact of the changed interval. Therefore, the ILRT interval extension risk assessment is allowed to use the existing Internal Events, Internal Flooding, and Fire PRA models and external hazard evaluations described below.

Additionally, note that no mitigating strategies (i.e.,FLEX) are incorporated in the baseline or other external event PRA model used in this evaluation.

Internal Events and Internal Flooding The PTN results for the internal events and flooding hazard are from the plant-specific PRA models which meet the ASME PRA Standard at Category II or higher (Reference 8.44, 8.46).

The internal events and internal flooding PRA models used in this analysis have been updated (Reference 8.64). The internal events CDF and LERF values decreased. The internal flooding CDF value increased slightly from 1.37E-08 to 2.06E-08 and from 1.27E-08 to 1.46E-08 for U3 and U4 respectively. The U3 LERF result also increased slightly from 5.44E-11 to 5.77E-11, while the U4 LERF decreased from 1.44E-10 to 1.16E-10 [Ref. 8.66]. Since the flooding contribution to the external event CDF and LERF is small (<1%), the increase in the U3 internal flooding CDF will not affect the ILRT extension risk results significantly. Moreover, consideration of flooding risk sensitivities is not required to be credited in the ILRT interval extension per EPRI Report 1009325. A decrease in the remaining internal events and internal flooding CDF and LERF results ensures that the ILRT interval extension risk results in this analysis are conservative as the risk is evaluated based on scaling the total CDF and LERF results.

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Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 78 of 82 As discussed above, it is noted that modifications required to adopt NFPA 805 were incorporated in the PRA model used in this analysis (Reference 8.44).

Fire Hazards The PTN Units 3 and 4 results for fire hazards are from the peer reviewed plant-specific fire PRA model credited in the NFPA 805 LAR submittal (Reference 8.61). The Fire PRA model was developed consistent with NUREG/CR-6850 and utilizes methods previously accepted by the NRC (Reference 8.47, 8.61). Most NFPA 805 related plant modifications have been completed at this time (Reference 8.62, 8.57). The modifications required for full compliance with the NFPA 805 transition are modeled in the PRA as part of the application to adopt a risk-informed, performance-based fire protection program in accordance with 10 CFR 50.48(c)

(Reference 8.47, Section 4.1). The NextEra risk management process ensures that the PRA model used in this application reflects the as-built and as-operated plant for each of the PTN units. It is noted that the plant PRA models with the NFPA 805 modifications included best represents the plant conditions during the period of ILRT interval extension as the modifications will have been completed (Reference 8.57, Safety Evaluation Section 2.3).

Seismic Hazards The PTN seismic CDF results are from a plant specific seismic CDF estimate developed in PTN-BFJR-14-009 (Reference 8.24). This calculation does not develop a full seismic PRA in compliance with the ASME PRA Standard. The ILRT interval extension guidance provided in EPRI Report 1009325 (Reference 8.29) requires a qualitative assessment of the contribution of seismic events (Section 4.2.7). Quantitative assessments of the seismic risk change due to the ILRT interval extension may be evaluated based on alternate methods to assess the order of magnitude of the seismic contribution to risk (Appendix H, Section 3.0). Therefore, the estimated seismic CDF utilized in this evaluation is considered acceptable.

Other External Hazards All other external hazards were screened from applicability to PTN, Units 3 and 4, per a plant-specific evaluation (Reference 8.37). The methods used are consistent with the screening and assessment processes identified in the supporting requirements of Part 6 of the ASME/ANS PRA Standard, as endorsed by NRC Regulatory Guide 1.200, Rev. 2. Attachment 4 to the LAR to adopt 10 CFR 50.69 (Reference 8.17) provides a summary of the other external hazards screening results as well. Attachment 5 to the LAR to adopt 10 CFR 50.69 (Reference 8.17) provides a summary of the progressive screening approach for external hazards.

Other LARs There are two recently approved LARs for PTN which will result in changes to the PRA models.

License Amendment Request 265: Revise NFPA 805 License Condition for Reactor Coolant Pump (RCP) Seals The PRA models credited in this ILRT interval extension submittal do not include the changes to the modeling of RCP seal leakage described in LAR 265. LAR 265 describes revisions to the RCP seal modeling to use the guidance from the NRC approved document WCAP-16175-P-A Page l 78

Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 79 of 82 instead of the guidance from the Flowserve RCP Seal Topical Report (Reference 8.58). This change results in the Human Error Probability (HEP) values increasing as a result of the decrease in time available to trip the RCPs from 60 minutes, assumed in the Flowserve Topical Report, to 20 minutes, as specified in WCAP-16175-P-A. Additionally, the probability of seal failure for the RCPs has been adjusted.

As discussed in the NRC Safety Evaluation associated with LAR 265, the overall impact in the change in methodology for modeling RCP seal leakage was determined to have a small impact on risk results. The model update which incorporates the modification to the modeling of the Flowserve RCP seals (Reference 8.65) results in a small decrease in overall CDF and LERF results. Therefore, the effect on the risk results for this submittal would not be significant because the ILRT extension delta risk results are based on a scaling approach using the total CDF/LERF risk values. Thus, changes associated with LAR 265 will not impact the conclusions of this ILRT interval extension risk assessment.

License Amendment Request 236: Revision to the Technical Specifications to Adopt Risk Informed Completion Times TSTF-505 The PRA models credited in this ILRT interval extension submittal do not incorporate the six implementation items listed in LAR 236 (Reference 8.59). These items are to be incorporated into the applicable PRA models prior to the time that the RICT Program is implemented (Reference 8.60). The PRA models that have been updated for calculations of RICTs have lower CDF and LERF results than those used in this ILRT interval extension risk assessment (Reference 8.64). A decrease in the internal events and external events CDF and LERF results ensures that the ILRT interval extension risk results in this analysis are conservative, as the risk is evaluated based on scaling the total CDF and LERF results.

Peer Review Process The PRA models described above have been assessed against ASME/ANS RA-Sa-2009 and RG 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Revision 2 (Reference 8.35) consistent with NRC RIS 2007-06 (Reference 8.48).

The models were subject to an industry peer review as documented in PWROG-19008-P (Reference 8.63):

2002 full-scope peer review of the Turkey Point Internal Events PRA model.

2011 focused-scope peer review of the Turkey Point Internal Flooding PRA model.

2013 focused-scope peer review of the Turkey Point PRA model with respect to specific supporting requirements such as CCF, ISLOCA, and LERF analysis Technical Elements.

2010 full-scope peer review of the Turkey Point Fire PRA model.

2012 focused-scope peer review (FSS, HRA, and PRM) of the Turkey Point Fire PRA model.

Based on the full-scope peer reviews and subsequent focused-scope peer reviews (Reference 8.63), 49 unique F&Os were reviewed in the closure assessment:

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Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 80 of 82 Internal Events: 9 findings Internal Flood: 8 findings Fire: 32 findings Findings were reviewed and closed using the process documented in Appendix X to NEI 05-04, NEI 07-12 and NEI 12-13, Close-out of Facts and Observations (Reference 8.63) as accepted by NRC in the staff memorandum dated May 3, 2017 (Reference 8.41). The results of this review have been documented and are available for NRC audit. As a result of the independent F&O closure, all 49 F&Os have been judged to be closed or superseded by new F&Os. PRA upgrades to close the original F&Os required a focused-scope peer review of the SRs associated with the upgrade. Following the focused-scope peer review, new F&Os were assigned to address the remaining issues. There are three open F&O findings and one new suggestion as a result of this effort. These findings are dispositioned with respect to the ILRT interval extension risk assessment in the table below.

PRA model upgrades made to close F&Os are incorporated into the latest model revisions. The updated models have lower total CDF and LERF results compared to those of the model revisions on which this ILRT extension risk assessment is based (References 8.64 and 8.65). A decrease in the internal events and external events CDF and LERF results ensures that the ILRT interval extension risk results in this analysis are conservative, as the risk is evaluated based on scaling the total CDF and LERF results. Note that the intent of this risk assessment is to assess the risk associated with any additional containment liner degradation which may occur as a result of the ILRT extension. Because the EPRI methodology is a bounding approach to estimating delta LERF based on scaling CDF (i.e., Class 3b frequency is based on CDF after subtracting the frequency of containment isolation failure (Class 2) and containment bypass (Class 8) scenarios) then only F&Os with a significant impact on CDF will affect the delta LERF for this application.

F&Os that have a significant impact on total LERF also need to be evaluated for this application because the total calculated increase in LERF is in the small range of 1.0E-07 per reactor year to 1.0E-06 per reactor year.

The information in this appendix demonstrates that the PRA is of sufficient quality and level of detail to support this submittal, and has been subjected to a peer review process assessed against a standard or set of acceptance criteria that is endorsed by the NRC. Therefore, the use of the PRA models listed in this analysis is considered acceptable for use in this application.

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Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 81 of 82 Disposition and Resolution of Open Peer Review Findings and Self-Assessment Open Items Supporting Disposition for ILRT Interval Extension Finding Number Description of Finding Requirement(s) Application Uncertainty is not required to be considered per The uncertainty and sensitivity calculation(s) need to be guidance in EPRI Report 1009325. This F&O UNC-A1-01 UN-A1 performed for the latest FPRA does not affect the risk evaluation of the ILRT model.

risk extension.

This F&O is a suggestion only and not a finding.

This is a documentation recommendation which Update the discussion in PTN-BFJR-16-034 to identify does not affect the ILRT extension risk the conservatisms in that the calculation. The use of conservative inputs, as FSS-C1-01 FSS-C1 generic treatments and combined ignition source discussed in the F&O, has a conservative effect references are based on HRR data on the FPRA results which are used as input to defined in NUREG/CR-6850 and not NUREG-2178. the ILRT risk assessment. Therefore, this is an additional conservatism in the ILRT extension risk evaluation.

In reviewing the calculation of the NSP development for Note that Supporting Requirement FSS-C1 was the hot gas layer scenario in determined to meet Capability Category II PAU 068, it was identified that the equation is incorrect despite this F&O. Capability Category I is and not appropriately accounting for the transient sufficient to meet the requirements of the ILRT frequency. The use of this equation and its inputs should extension risk application.

be reviewed for all scenarios of this type to ensure that FSS-C1-02 FSS-C1 the applicable inputs are being used correctly in the Additionally, this F&O affects a limited set of overall calculation. This is a concern for those PAUs that transient scenarios that involve full room model full room burnout transient scenarios. The burnup. The FPRA results used for this ILRT transient frequency is only included in the hot gas layer extension are judged to be sufficiently scenario. This is currently being tracked in the PTN conservative that this F&O would have a model change log under PTN-0 negligible effect on the ILRT extension risk 005. results.

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Turkey Point Nuclear Plant L-2020-003 Docket Nos. 50-250 and 50-251 Enclosure 3 Page 82 of 82 Supporting Disposition for ILRT Interval Extension Finding Number Description of Finding Requirement(s) Application Note that Supporting Requirement FSS-C1 was determined to meet Capability Category II The NSP calculation for TGO is calculated improperly for despite this F&O. Capability Category I is scenario 078-J-PTB. The NSP calculation uses a motor sufficient to meet the requirements of the ILRT as the HRR. The ignition source is oil for this source. extension risk application.

The NSP for this scenario needs to be updated to FSS-C1-03 FSS-C1 appropriately account for the ignition source type and Additionally, this F&O only affects the limited set characteristics. Other TGO fires should be reviewed for of scenarios that credit suppression of turbine this same incorrect calculation. This is currently being generator oil fire scenarios. The Fire PRA tracked in the PTN model change log under PTN 19- results used for this ILRT extension are judged 004. to be sufficiently conservative that this F&O would have a negligible effect on the ILRT extension risk results.

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L-2020-003, License Amendment Request 268, Request to Extend Containment Leakage Rate Test Frequency

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3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

5.0 ENVIRONMENTAL CONSIDERATION

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3.5.1 Methodology

4.0 REGULATORY EVALUATION

5.0 ENVIRONMENTAL CONSIDERATION

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L-2020-003, License Amendment Request 268, Request to Extend Containment Leakage Rate Test Frequency

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SUMMARY

3.0 TECHNICAL EVALUATION

4.0 REGULATORY EVALUATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

SUMMARY

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3.5.1 Methodology

4.0 REGULATORY EVALUATION

5.0 ENVIRONMENTAL CONSIDERATION

6.0 REFERENCES

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