~ Southern Nuclear

July 12, 2017

Docket Nos.: 50-321 50-366

U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001

J. J. Hutto Regulatory Affairs Director

Edwin I. Hatch Nuclear Plant Units 1 and 2 License Amendment Request to Revise

Technical Specification Section 5.5.12 for Permanent

40 lmemess Center Parkway Post Ofiice Box 1295 Birmingham, AL 35242 205 992 5872 tel 205 992 760 I fax

jjhutto@•muthemco.com

NL-17-1161

Extension of Type A and Type C Leak Rate Test Frequencies Supplemental Responses to NRC Second Set of Requests for Additional Information

Ladies and Gentlemen:

By letter dated July 1, 2016, as supplemented by letter dated August 24, 2016, Southern Nuclear Operating Company (SNC) submitted a license amendment request (LAR) to revise the Edwin I. Hatch Nuclear Plant (HNP), Units Nos. 1 and 2, Technical Specification 5.5.12, .. Primary Containment Leakage Rate Testing Program ... In part, the proposed changes would allow SNC to increase the existing Type A integrated leakage rate test interval for each unit from 10 years to 15 years, 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. The U.S. Nuclear Regulatory Commission (NRC) staff issued a request for additional information (RAI) on December 5, 2016, and SNC submitted its response to that request by letter dated February 10, 2017.

On April24, 2017, the NRC staff, upon review of SNC's February 10,2017 RAI response, determined that additional information was needed to complete its review, and issued a letter requesting that SNC respond to their second set of questions. SNC responded to that request on June 1, 2017. On June 23, 2017, the NRC staff requested clarifications of the June 1, 2017 response. Enclosed are supplemental responses to provide clarification.

This letter contains no NRC commitments. If you have any questions, please contact Ken McElroy at 205.992.7369.

I declare under penalty of perjury that the foregoing is true and correct. Executed on July 12, 2017.

J. J. Hutto Regulatory Affairs Director Southern Nuclear Operating Company

U. S. Nuclear Regulatory Commission NL-17-1161 Page2

Enclosure: 1. Supplemental Responses to NRC Second Set of Requests for Additional Information

cc: NRC Regional Administrator, Region II NRC NRR Project Manager- Hatch NRC Senior Resident Inspector- Hatch Georgia - State Director of Environmental Protection Division SNC Document Control RTYPE: CHA02.004

Edwin I. Hatch Nuclear Plant Units 1 and 2 License Amendment Request to Revise

Technical Specification Section 5.5.12 for Permanent Extension of Type A and Type C Leak Rate Test Frequencies

Supplemental Responses to NRC Second Set of Requests for Additional Information

Enclosure 1 (9 pages)

05080.000-MEM-13347 Hatch ILRT RAI Responses

NRC SUPPLEMENTAL INFORMATION REQUEST PART 1 AND 2

• Exclusion of scenarios with SORVs Justify the statement on page E1-8,. stating "any transients with SORVs would be subsumed by the MLOCA analysis" and that "NPSH is not considered to be lost for these scenarios during the first 24 hours. " Explain and justify assumptions of number of open SORVs and confirm that it covers the scenarios with all SRVs stuck open. If core damage occurs after 24 hours it should be taken into account in the estimate of delta CDF, or otherwise justify its exclusion.

• Exclusion of all other transient scenarios Page E1-12 states that the remaining scenarios from "General Transient, LOSP and Station Blackout Initiators" that are not SORVs scenarios "are bounded by the MSIV Closure scenario". Table 2 shows that the MSIV closure scenario would result in core damage. However Table 1 reports a zero delta CDF from these scenarios, with an explanation that because "the base model assumes core damage if containment heat removal fails", "loss of NPSH does not create additional CDF." Justify the zero increase in CDF for these scenarios, given the MAAP results. Also discuss all other transient type initiators, such as loss of main feedwater, loss of instrument air, loss of condenser, etc.

Explain the MAAP runs for the MSIV Closure scenarios and explain why core damage occurs much sooner when compared to the medium LOCA case.

Response to Supplemental Information Request Part 1 and 2

This response does not exclude SORV or Transient scenarios. The originaiiLRT RAI response to RAI 6a calculated a delta CDF increase of 5.47E-07/year. The delta CDF increase of 5.47E-07/yr will be further analyzed and refined. A delta CDF for SORVs (item 1 above) and delta CDF for all other transient scenarios (item 2 above) is also captured . The delta CDF calculated includes core damage scenarios longer than 24 hours.

PRA Modeling and Results

As noted in RAI 6a: " ... the current containment isolation failure logic can be increased by the Class 3b frequency at 15 years (i.e., 0.0023 * 5.0 = 0.0115) to estimate a bounding increase in CDF. With this increase applied to the loss of NPSH probability in the Hatch PRA model, the CDF increases from 7.57E-06/yr to 8.12E-06/yr representing an increase of 5.47E-07 /yr." The PRA modeling to obtain this NPSH will be explained. This will be followed by a refinement of the PRA modeling based on current plant procedures that allow RCIC to operate long term (i.e., well beyond 24 hours) without containment heat removal.

The current Hatch PRA model includes a term "NPSHLOSSPROB". Failure of this event coupled with loss of heat removal from the Main Condenser, Suppression Pool Cooling (SPC) and Shutdown Cooling (SOC) results in loss of injection from the Core Spray and RHR pumps.

Logic is shown below:

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05080.000-MEM-13347 Hatch ILRT RAI Responses

r..--------1..------. I I I I I I I L

LOSS OF NPSH FOR ECCS LP INJECTION

AFTER OPENING HARDENED VENT

FAILURE OF TORUS CLG PATHS, S/0 CLG PATHS &

MAIN CONDENSER

PROBABILilY FOR LOSS OF NPSH FOR LP ECCS

INJECTION AFTER OPENING HV

Show Parents

Parenls of event EMERGENCYVENT CSA. ~·:-~~CORE SPRAY TRAIN A FAILS CSB -CORE SPRAY TRAIN 8 FAILS LPCIA -LOOP A RHR LPCJ FAILS LPCIB -LOOP B RHR LPCI FAILS

The impact of Loss of NPSH to each group of initiating events is captured in the table below. This table shows heat removal systems credited in support of injection, as well as those injection systems not impacted by a 100 La leak and loss of SOC and SPC.

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05080. 000-M EM-1334 7 Hatch ILRT RAI Responses

Table 1 SUMMARY OF THE IMPACT OF 100 La LEAKAGE TO EVENT TREE LOGIC

Initiator

GENERAL TRANSIENT EVENT TREE

Heat Removal Systems

Credited with 100La

Leakage

(All general transients that do not result in a LOSP or A TWS)

General Transients without SORV Main . Reactor Trip Condenser . Turbine Trip (MC), Shutdown Cooling (SOC) . MSIV Closure and . Loss of Condenser Vacuum Suppression . Loss of Plant Service Water Pool Cooling

Loss of Feedwater (SPC) . . Loss of 4160 V Bus . Loss of 600 V Bus . Loss of Drywall Cooling . Loss of Startup Transformer . Loss of Main Control Room Air

Conditioning . Loss of DC Bus . Loss of Instrument Air . Flooding Initiators

Above General Transients with one SORVs Same as above

Above General Transients with two or more SORVs Shutdown Cooling (SOC) and Suppression Pool Cooling (SPC)

LOCA EVENT TREES

Small Break Loss of Coolant Accident (LOCA) Shutdown Cooling (SOC) and Suppression Pool Cooling (SPC)

Medium Break Loss of Coolant Accident (MLOCA) Shutdown A medium LOCA ranges from 0.01 tr to 0.1 tr for Cooling (SOC) water line breaks and from 0.03 tr to 0.2 tr for and steam line breaks. Suppression

Pool Cooling (SPC)

Injection Systems Credited with 100

La Leak & SDC/SPC Failure

Condensate (CD) RHR Service Water Injection (RHRSWINJ)

Same as above

RHR Service Water Injection (RHRSWINJ)

RHR Service Water Injection (RHRSWINJ)

RHR Service Water Injection (RHRSWINJ)

Event Trees and Comments

GTR . Primary Conversion System (PCS) (requires CD, FW and Main Condenser) is a success. . HPCI and RCIC are not credited for 24 long term success . . RHRSWINJ requires Torus Drywall Vent success.

Same as above

GTR . HPCI/RCIC not credited for depressurization.

SLOCA . Injection and Heat Removal logic is the same as GTR Event Tree w/o SORV except PCS is not credited for success.

MLOCA . Similar to SLOCA. HPCI is credited for depressurization; however, RCIC is not credited for depressurization.

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05080. 000-MEM-1334 7 Hatch ILRT RAI Responses

Table 1 SUMMARY OF THE IMPACT OF 100 La LEAKAGE TO EVENT TREE LOGIC

Heat Removal Systems Injection Systems

Credited with Credited with 100 100 La La Leak& Event Trees and

Initiator Leakage SDC/SPC Failure Comments

Large Break Loss of Coolant Accident (LLOCA) Shutdown RHR Service Water LLOCA (Large LOCAs are defined as any water line breaks Cooling (SOC) Injection . Similar to greater than 0.1 tr and steam line breaks greater and (RHRSWINJ) MLOCA, than 0.2 tr for piping connected to the reactor Suppression however vessel inside the primary containment.) Pool Cooling depressurization

(SPC) is assumed successful.

LOCAs Outside Containment (ULOCA) N/A N/A ULOCA Feedwater, Main Steam, HPCI, RCIC, RWCU line . A failure to breaks. isolate the line

break is assumed to result in CD and LER events. A large pre-existing leak would not impact CDF or LERF.

Interfacing Systems LOCA (ISLOCA) N/A N/A ILOCA . Same as U LOCA . Represented by a single event that leads to CD and LER events.

Excessive LOCA (RPVRUPTURE) N/A N/A RPV RUPTURE . RPV Rupture is assumed to result in CD and LER events. A large pre-existing leak would not impact CDF or LERF.

All SRVs Open (ALOCA) Shutdown RHR Service Water ALOCA Cooling (SOC) Injection . Similar to Large and (RHRSWINJ) LOCA except Suppression vapor Pool Cooling suppression is (SPC) not required

since the SRVs discharge to the suppression pool.

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05080. 000-MEM-1334 7 Hatch ILRT RAI Responses

Table 1 SUMMARY OF THE IMPACT OF 100 La LEAKAGE TO EVENT TREE LOGIC

Initiator

Heat Removal Systems

Credited with 100 La

Leakage

Injection Systems Credited with 100

La Leak& SDC/SPC Failure

Event Trees and Comments

ATWS EVENT TREES (100 La Leak does not impact the ATWS event tree logic)

Failure to Scram Transient Events no SORVs.

Failure to Scram Transient Events one SORV.

Failure to Scram Transient Events two or more SORVs.

LOSS OF OFFSITE POWER EVENTS

Loss of Off-Site Power

SBO EVENT TREE

LOSP with No AC Power (Diesels Generators Fail)

MC with 2" Emerg. Vent or SPC with Drywell Sprays

SPC

N/A

Shutdown Cooling (SDC) and Suppression Pool Cooling (SPC)

Power Conversion AlWS System with CD, . CS or RHR LPI

Power Conversion AlWS System with RHR . LPI

N/A ATSW .

RHR Service Water LOSP Injection (RHRSWINJ)

SDC and SPC RHR Service Water SBO Injection (RHRSWINJ)

Requires Standby Liquid Control System success.

Requires Standby Liquid Control System success.

Leads to Core Damage and Large Early Release.

Similar to General Transient Event Tree except PCS (CD, FWand Main Condenser) is not credited.

The CDF increase from increasing the probability of isolation failure is the result of loss of low pressure injection (CS and RHR LPI, see logic preceding the table above) coupled with loss of injection from outside the containment. This loss of low pressure injection occurs when containment heat removal is failed (either SOC or SPC, depending on the event tree logic) and containment isolation failure.

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05080.000-MEM-13347 Hatch ILRT RAI Responses

A review of the delta CDF cutsets found that 82% of the increase in CDF were from Sequences GT _6, GT _7 and GT _8. The injection nodes are shown below:

ROC DE LO QR Class Name

OK GT_1

OK GT_2

#OR CD GT_3

#LO CD GT_4

OK

#QR co

#LO co

CD

GT_9

#QR CD GT_10

RCIC-1 #LO co GT_11

#DE CD GT 12

As shown above, RCIC is credited for short term success leading to RPV depressurization. Low pressure injection is required following RCIC success. If RCIC could be credited for long term success, this injection success would lead to Gates GT _6, GT _7 and GT _8 no longer being applicable, as depressurization and low pressure injection would not be required. RCIC injection source is the Condensate Storage Tank (CST) and therefore, NPSH is not a concern.

Plant specific Emergency Operating Procedures (EOPs) and Severe Accident Management Guidelines (SAMGs) have been revised based on the generic BWROG Emergency Procedure Guidelines/Severe Accident Guidelines (EPGs/SAGs) Revision 3 issued in 2013.

Procedure 31EO-EOP-010-2 RC RPV Control (Non-ATWS) Version 10 Revision 3 includes directions to:

• Limit RPV depressurization to preserve RCIC operation if needed for core cooling • Add authorization for defeating the high RPV water level RCIC isolation • Add authorization to defeat HPCI and RCIC high area temperature isolations to RPV

pressure control steps • Add authorization to defeat the RCIC high exhaust pressure isolation to RPV pressure

control steps

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05080. 000-MEM-13347 Hatch ILRT RAI Responses

RCIC long term usage is supported by current operating procedures. Long term make-up can be supplied by FLEX equipment to allow RCIC running past -60 hours.

It is estimated that RCIC long-term usage would reduce delta CDF by >50o/o. Previous calculations found a CDF change from 7.57E-06/yr to 8.12E-06/yr representing an increase of 5.47E-07 /yr with a conditional probability of containment isolation failure. A conservative increase in delta CDF of%* 5.47E-07 /yr = 2.74E-07 /yr is estimated. This increase includes SORV and all transient scenarios.

MAAP Run Explanation

For the MAAP cases where only LPCS is available without SPC available, the MSIV closure cases show core damage sooner. This is due to the loss of NPSH for LPCS injection from the pool sooner due to a lack of containment overpressure. With LPCS as the only credited injection source, core damage occurs shortly after NPSH is lost. The Hatch PRA model does not credit LPCS for preventing core damage with a pre-existing leak unless SPC is available. Note: the MAAP runs are no longer used to justify excluding a delta CDF for MSIV closure scenarios.

NRC SUPPLEMENTAL INFORMATION REQUEST PART 3

• Medium and Large LOCA estimates On page E 1-5 explain the credit for Torus/Drywell Hardened Vent System for containment heat removal. If the hardened vent system is credited for assuring sufficient NPSH to the ECCS pumps, include justification for this credit.

Response to Supplemental Information Request Part 3

Page E1-5 states: Consistent with the LLOCA approach for Medium LOCA the RHRSW system is credited for injection and the Torus/Drywall Hardened Vent System is credited for containment heat removal.

To clarify this statement: The Torus/Drywall Hardened Vent System is credited only with RHRSW system success. Core Spray and RHR pumps are failed in the sensitivity run if a pre-existing leak exists. This scenario leads to Core Damage if injection sources outside containment are unavailable.

NRC SUPPLEMENTAL INFORMATION REQUEST PART 4

A delta LERF resulting from loss of NPSH can be, as a first order, equated to delta CDF. If the justifications provided above for delta CDF support the current delta CDF estimate of 4.15E-8/year, no further explanation regarding delta LERF is necessary. If however the delta CDF is re-estimated and results in a significantly larger value, any credit for reducing delta LERF due to loss of NPSH below the value for delta CDF will have to be sufficiently justified.

Response to Supplemental Information Request Part 4

The delta CDF has been re-estimated and is significantly higher than in the previous RAI response. Some credit has been applied for the use of RCIC past 24 hours in General Transient scenarios. The increase in CDF is estimated to be 2. 7 4E-07 /yr from the baseline risk to 1 per 15 year ILRT interval. The delta LERF will be equated to delta CDF.

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05080. 000-MEM-1334 7 Hatch ILRT RAI Responses

Calculation of LlLERF and Overall Total LERF

Section 1: Relevant RAIBb Information

As discussed in RAI Response 8a, the seismic CDF estimate is lower per Gl-199 than that used in the ILRT submittal. Using the lower seismic CDF value of 6.6E-7/yr, the calculations are repeated as follows:

• External Events Multiplier= External Events CDF/Internal Events CDF • External Events CDF =Seismic CDF +Internal Fire CDF +High Winds CDF +External

Flood CDF • External Events CDF = 6.6E-7/yr + 7.5E-6/yr + 1.0E-6/yr + 1.0E-8/yr = 9.17E-6/yr • Internal Events CDF = 7.57E-6/yr • External Events Multiplier= 9.17E-6/yr /7.57E-6/yr = 1.21

Updated Tables

Table 5.8-2 (Updated Internal Events CDF and Seismic CDF)

HATCH CLASS 3b (LERF) AS A FUNCTION OF ILRT FREQUENCY FOR INTERNAL AND EXTERNAL EVENTS (INCLUDING AGE ADJUSTED STEEL CORROSION LIKELIHOOD)

3b 3b Frequency 3b Frequency Frequency (1-per-15 year

LERF lncrease11, (3-per-10 yr ILRT) {1-per-1 0 year ILRT) ILRT)

Internal Events Contribution(2l 7.06E-08 2.35E-07 3.53E-07 2.82E-07

External Events Contribution 8.54E-08 2.84E-07 4.27E-07 3.42E-07 (Internal Events x 1.21)

Combined (Internal + External) 1.56E-07 5.20E-07 7.80E-07 6.24E-07

(1l Associated with the change from the 3-per-10 year frequency to the proposed 1-per-15 year frequency.

(2

) Values from Table 5.7-1 of the ILRT Risk Assessment are added to the delta CDF from NPSH. For example, the 3b Frequency for 1 per 15 years is 2.74E-07 /yr from NPSH and 7.97E-08 /yr from Table 5.7-1.

The results of using the higher external events multiplier based on the current internal events CDF and lower seismic CDF shows a LERF increase of 6.2E-7/yr which is in the "small" impact for deltas of <1 E-6/yr. Using the same external events multiplier, the total LERF is calculated to confirm that total LERF is <1 E-5/yr as directed by RG 1.174, as follows.

Internal Events LERF

External Events LERF (Internal events LERF x 1.21)

Internal Events LERF due to ILRT (at 15 years)

External Events LERF due to ILRT (at 15 years) (Internal Events LERF * 1.21)

Total LERF (Internal + External)

Frequency

1.12E-06

1.36E-06

2.82E-07

3.42E-07

3.10E-06

The calculated total LERF of 3.1 E-6 /yr. meets the acceptance criteria of <1 E-5/yr.

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05080.000-MEM-13347 Hatch ILRT RAI Responses

Change in Population Dose

The change in population dose is reflected in a change in 3b frequency. A small f1LERF of 3.39E-9/year comes from Class II being reclassified as an EPRI Class 3b contributor resulting in a small reduction in EPRI Class 7 contribution. For simplicity, this reduction will not be credited. The assumed change in the ILRT LAR provides the following information. Note, person-rem is based on information from Table 4.2-5 of the original LAR request.

Per-Rem All 3b Freq. P-REM/yr 3b Freq. P-REM/yr 3b Freq. P-REM/yr P-REM/yr lntervals<1

> (3/10) (3/10). (1/10) (1110). (1/15) (1/15). Increase

Internal Events 1.15E+05 7.06E-08 8.12E-03 2.35E-07 2.70E-02 3.53E-07 4.06E-02 3.25E-02 Contribution(2J

External Events Contribution 1.15E+05 8.54E-08 9.82E-03 2.84E-07 3.27E-02 4.27E-07 4.91E-02 3.93E-02 (Internal Events x 1.21)

Combined 1.15E+05 1.56E-07 1.79E-02 5.20E-07 5.98E-02 7.80E-07 8.97E-02 7.18E-02 (Internal + External)

(1J Per Rem, All intervals is calculated as 100 * Per-Rem associated with 1 La release (1 .15E+03) in Table 4.2-5

column 2030 Population Assigned Dose.

(ZJ These values are from Table 5.8-2 (Includes RAI NPSH llLERF Increase) above.

Total person-rem/year for type 1 testing as reported in the original LAR is 9.90E-03 person-rem/yr. This would be reduced if the latest external event multiplier was used. For simplicity, the increase in person-rem/year calculated above is added to 9.90E-03 person-rem/yr.

Total increase in person-rem/yr = 9.9E-03/yr. + 7.2E-02/yr.

f1person-rem/year = 8.2E-02.

The EPRI acceptance criteria is s1.0 person-rem/year or <1.0% person-rem/year. Specifically, 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 o/o of the total population dose, whichever is less restrictive. The f1person-rem/year calculated above is far below the acceptance criteria of S1.0 person-rem/year. Therefore, the acceptance criteria is met.

Change in the Conditional Containment Failure Probability (CCFP)

The original LAR submittal calculated an increase in conditional containment probability of 0.84%. EPRI acceptance criteria includes a small increase in CCFP defined as a value marginally greater than that accepted in a 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 point. The containment overpressure loss of NPSH does not impact the CCFP. Therefore there is no additional change in CCFP and acceptance criteria of less than or equal to 1.5% is met.

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