NL-17-1161, 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

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

ML17194B078
Person / Time
Site:	Hatch
Issue date:	07/12/2017
From:	Hutto J Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-17-1161
Download: ML17194B078 (12)

Text

~ Southern Nuclear J. J. Hutto Regulatory Affairs Director 40 lmemess Center Parkway Post Ofiice Box 1295 Birmingham, AL 35242 205 992 5872 tel 205 992 760 I fax jjhutto@*muthemco.com July 12, 2017 Docket Nos.: 50-321 NL-17-1161 50-366 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 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 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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. " 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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) 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..------. Show Parents LOSS OF NPSH FOR I ECCS LP INJECTION I AFTER OPENING Parenls of event EMERGENCYVENT I HARDENED VENT CSA. ~*:-~~ CORE SPRAY TRAIN A FAILS CSB -CORE SPRAY TRAIN 8 FAILS I LPCIA -LOOP A RHR LPCJ FAILS I LPCIB -LOOP B RHR LPCI FAILS I

I L

FAILURE OF TORUS CLG PROBABILilY FOR LOSS PATHS, S/0 CLG PATHS & OF NPSH FOR LP ECCS MAIN CONDENSER INJECTION AFTER OPENING HV 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 Heat Removal Systems Injection Systems Credited with Credited with 100 100La La Leak & Event Trees and Initiator Leakage SDC/SPC Failure Comments GENERAL TRANSIENT EVENT TREE (All general transients that do not result in a LOSP or A TWS)

General Transients without SORV Reactor Trip Main Condenser Condensate (CD)

RHR Service Water GTR

. Primary (MC), Shutdown Injection Conversion Turbine Trip Cooling (SOC) (RHRSWINJ)

System (PCS)

MSIV Closure and (requires CD, FW

. Loss of Condenser Vacuum Suppression and Main Pool Cooling Condenser) is a Loss of Plant Service Water (SPC) success.

Loss of Feedwater HPCI and RCIC Loss of 4160 V Bus are not credited

. Loss of 600 V Bus for 24 long term

. Loss of Drywall Cooling success .

. Loss of Startup Transformer Loss of Main Control Room Air RHRSWINJ requires Torus

. Conditioning Drywall Vent success .

Loss of DC Bus

. Loss of Instrument Air Flooding Initiators Above General Transients with one SORVs Same as above Same as above Same as above Above General Transients with two or more SORVs Shutdown Cooling (SOC) and RHR Service Water Injection (RHRSWINJ)

GTR

. HPCI/RCIC not credited for Suppression depressurization.

Pool Cooling (SPC)

LOCA EVENT TREES Small Break Loss of Coolant Accident (LOCA) Shutdown Cooling (SOC) and RHR Service Water Injection (RHRSWINJ)

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

Medium Break Loss of Coolant Accident (MLOCA)

A medium LOCA ranges from 0.01 water line breaks and from 0.03 tr tr to 0.1 to 0.2 tr tr for for Shutdown Cooling (SOC) and RHR Service Water Injection (RHRSWINJ)

MLOCA Similar to SLOCA. HPCI is steam line breaks. Suppression credited for Pool Cooling depressurization; (SPC) 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)

(Large LOCAs are defined as any water line breaks greater than 0.1 tr and steam line breaks greater Shutdown Cooling (SOC) and RHR Service Water Injection (RHRSWINJ)

LLOCA Similar to 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)

Feedwater, Main Steam , HPCI, RCIC, RWCU line breaks.

N/A N/A ULOCA A failure to 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 Cooling (SOC) and RHR Service Water Injection (RHRSWINJ)

ALOCA Similar to Large 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-13347 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 ATWS EVENT TREES (100 La Leak does not impact the ATWS event tree logic)

Failure to Scram Transient Events no SORVs. MC with 2" Emerg. Vent or SPC with Power Conversion System with CD, CS or RHR LPI AlWS

. Requires Standby Liquid Drywell Sprays Control System success.

Failure to Scram Transient Events one SORV. SPC Power Conversion System with RHR LPI AlWS

. Requires Standby Liquid Control System success.

Failure to Scram Transient Events two or more SORVs .

N/A N/A ATSW

. Leads to Core Damage and Large Early Release.

LOSS OF OFFSITE POWER EVENTS Loss of Off-Site Power Shutdown RHR Service Water LOSP Cooling (SDC) Injection Similar to and (RHRSWINJ) General Suppression Transient Event Pool Cooling Tree except PCS (SPC) (CD, FWand Main Condenser) is not credited.

SBO EVENT TREE LOSP with No AC Power (Diesels Generators Fail) SDC and SPC RHR Service Water SBO Injection (RHRSWINJ)

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

1. OR CD GT_3

1. LO CD GT_4 OK

1. QR co

1. LO co CD GT_9

1. QR CD GT_10 RCIC-1 #LO co GT_11

1. 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 Page6

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 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in General Transient scenarios. The increase in CDF is estimated to be 2. 74E-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 (3-per-10 yr ILRT) {1-per-1 0 year ILRT) ILRT) LERF lncrease 11 ,

2 Internal Events Contribution( l 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 1

( l 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 .

Frequency Internal Events LERF 1.12E-06 External Events LERF (Internal events LERF x 1.21) 1.36E-06 Internal Events LERF due to ILRT (at 15 years) 2.82E-07 External Events LERF due to ILRT (at 15 years) (Internal 3.42E-07 Events LERF

• 1.21)

Total LERF (Internal + External) 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) 1

( J 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 1o/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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