NL-17-0766, License Amendment Request to Revise Technical Specification Section 5.5.12 for Permanent Extension of Type a and Type C Leak Rate Test Frequencies, Responses to NRC Second Set of Requests for Additional Information

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License Amendment Request to Revise Technical Specification Section 5.5.12 for Permanent Extension of Type a and Type C Leak Rate Test Frequencies, Responses to NRC Second Set of Requests for Additional Information
ML17152A413
Person / Time
Site: Hatch  Southern Nuclear icon.png
Issue date: 06/01/2017
From: Wheat J
Southern Nuclear Company
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-17-0766
Download: ML17152A413 (21)
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Text

Justin T. Wheat 40 ln,cmess Center Pdrkway

  • Southern Nuclear Nuclear Licensing Manager Po't Office Box 1295 Bmningham. AI. 1'i242 205 992 5998 tel 205 992 760 I fax jtwhc:tt @southcmco.com JUN 0 1 2017 Docket Nos.: 50-321 NL-17-0766 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{12 for Permanent Extension of Type A and Type C Leak Rate Test Frequencies 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 RAJ 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. Enclosed are SNC responses.

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

U. S. Nuclear Regulatory Commission NL-17-0766 Page 2 Mr. Justin T. Wheat states that he is the Nuclear Licensing Manager for SNC, is authorized to execute this oath on behalf of SNC and, to the best of his knowledge and belief, the facts set forth in this letter are true.

Respectfully submitted, C6).,..m;:;C~

Justin T. Wheat Nuclear Licensing Manager JTW/efb/lac Sworn and subscribed 12017.

~eX My commission expires: / () ; '( - do 1-1

Enclosure:

1. Responses to NRC Second Set of Requests for Additional Information cc: Regional Administrator, Region II NRR Project Manager- Hatch Senior Resident Inspector- Hatch RTYPE: CHA02.004

Edwin I. Hatch Nuclear Plant Units 1 and 2 License Amendment Request to Revise Technical Sp~cification Section 5.5.12 for Permanent Extension of Type A and Type C Leak Rate Test Frequencies Responses to NRC Second Set of Requests for Additional Information Enclosure 1 to NL-17-0766 SNC Response to Second Set of RAis By letter dated July 1, 2016 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML16188A268), as supplemented by letter dated August 24, 2016 (ADAMS Accession No. ML16238A477), Southern Nuclear Operating Company (SNC, the licensee) submitted a license amendment request (LAR) to revise the Edwin I. Hatch Nuclear Plant (HNP), Unit Nos. 1 and 2, Technical Specification (TS) 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 reviewed the LAR and supplement and issued a request for additional information (RAI) on December 5, 2016 (ADAMS Accession No. ML16330A128). SNC submitted its response to that RAI by letter dated February 10,2017 (ADAMS Accession No. ML17041A294). Based upon its review of that RAI resppnse, the NRC staff has determined that the following information is needed to complete its review of the LAR.

In the LAR, SNC stated that there is no meaningful change in core damage frequency (CDF) when considering the containment overpressure credit for net positive suction head (NPSH) for Unit 1 Emergency Core Cooling (ECCS) pumps, and therefore ~CDF was not quantitatively evaluated in the LAR. In RAI 6 of the December 6, 2016 letter, the NRC staff requested the licensee to provide additional justification to support the LAR conclusion of negligible impact on CDF. In its February 10,2017, response to RAis 6.a and 6.c, the licensee performed an updated risk evaluation and quantified the ~CDF due to loss of containment overpressure to 5.47E-07/year (RAI 6.a for sequences with containment heat removal), and 7.12E-07 /year (RAI 6.c for sequences without containment heat removal). It was not clear to the staff which value for ~CDF, or both, should be used in the quantification of total ~CDF.

a. Since the summation of the ~CDF values provided in the responses to RAI 6.a and 6.c of 1.26E-06/year could potentially result in a change to Large Early Release Frequency (LERF), the NRC staff requests an updated assessment of ~LERF due to loss of containment overpressure credit impacting Unit 1 ECCS NPSH, taking into account external hazards. Justify ~LERF due to loss of containment overpressure, including the contributions from external hazards to loss of containment overpressure risk.
b. The LAR stated that a pre-existing containment leak, resulting in the loss of adequate NPSH to the ECCS pumps would have the same result as containment failure from overpressure when containment heat removal is not available. The response to RAI 6.c provided an estimate of change in CDF from containment overpressure failure and stated that this change in CDF would not contribute to ~LERF. As indicated in the response to RAI 6.c the containment failure due to overpressure occurs in the 15-hour time frame. While this containment overpressure failure may not be categorized as LERF due to timing, the loss of NPSH to ECCS pumps caused by a potential pre-existing non-detected leak in the containment could happen much earlier than 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> and therefore it could be categorized as LERF.

Discuss and justify how all the accident scenarios with loss of containment heat removal are considered in the estimate of ~LERF from loss of containment overpressure, provided in response to item a. above.

E1-1 to NL-17-0766 SNC Response to Second Set of RAis

c. Provide an estimate of the total LlCDF, LlLERF, change in population dose, and change in the conditional containment failure probability from all contributors/hazards (internal events, fire, external events, loss of containment over pressure) and confirm that the acceptance criteria in Section 3.2.4.6 of the Safety Evaluation for Electric Power Research Institute (EPRI) Technical Report (TR) 1009325, Revision 2, are met for the application.

SNC Response SNC provides the following response to RAI 1:

Introduction A question is rai~ed as to whether the L\CDF of 5.47E-07/year calculatep in response to RAI 6a or 7 .12E-07/year calculated in response to RAI 6c, or both, should be used in the quantification of total L\CDF. These L\CDF contributions were calculated in different manners, providing two different L\CDF values. To avoid further confusion, the response to RAI a, band c will use a more refined analysis, not dependent on the previous responses.

An updated assessment of L\CDF and L\LERF due to loss of containment overpressure credit impacting Unit 1 ECCS NPSH, taking into account external hazards is performed and justified.

Part I of this document focuses on potential NPSH issues leading to a change in L\CDF. Part II reviews Class II containment failure endstates to determine if medium or late releases should be reclassified as large early releases leading to a change in L\LERF.

It is recognized that a loss of NPSH to ECCS pumps caused by a potential pre-existing non-detected leak in the containment could be postulated to occur earlier than 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> and therefore it could be categorized as LEAF. Modular Accident Analysis Program (MAAP) cases will be used to evaluate loss of NPSH and the timing to vessel breach and a large release. All accident scenarios with loss of containment heat removal are considered in the estimate of L\LERF from loss of containment overpressure.

An estimate of the total L\CDF, L\LERF, change in population dose, and change in the conditional containment failure probability from all contributors/hazards (internal events, fire, external events, loss of containment over pressure) is provided and it is confirmed that the acceptance criteria in Section 3.2.4.6 of the Safety Evaluation for EPRI Technical Report (TR) 1009325, Revision 2, are met for this application. This information is included in Part Ill RESULTS AND CONCLUSIONS.

A bounding assessment for combined Internal + External Events Contributions found the following:

L\CDF of 4.15E-08/year: This is less than <1 E-06/year placing this in the very small" region in Reg. Guide 1.174 Figure 4 Acceptance guidelines for core damage frequency.

L\LERF of 2.1 E-07/year: This is less than <1 E-06/year placing this in the "small" region in Reg.

Guide 1.174 Figure 5 Acceptance guidelines for large early release frequency. A calculation of total LEAF is required for a L\LERF in the "small region" Total LEAF of 2.7E-6/year: The calculated total LEAF meets the acceptance criteria of

<1 E-5/yr.

E1-2 to NL-17-0766 SNC Response to Second Set of RAis

!lPerson-rem/year of 3.43E-02: The calculated total is far below the acceptance criteria of ::;1.0 person-rem/year.

Change in the Conditional Containment Failure Probability (CCFP) is unchanged at 0.84%. The acceptance criteria of less than or equal to 1.5% is met Section 5.2.4 of EPRI Report 1018243 includes guidance on performing this risk assessment and provides the following examples of accident scenarios to be considered:

LOCA [loss-of-coolant accident] scenarios where the initial containment pressurization helps to satisfy the NPSH requirements for early injection in BWRs or PWR [pressurized-water reactor] sump recirculation Total loss of containment heat removal scenarios where gradual containment pressurization helps to satisfy the NPSH requirements for long term use of an injection system from a source inside of containment.

These considerations will be discussed for the Plant Hatch accident scenarios.

Recog~ized Conservatism in Methodology The EPRI methodology is conservative in that it assumes a significant leak of 1OOLa can exist without being detected. Plant Hatch is a BWR/4 with a Mark I containment. During power operation, the primary containment atmosphere is inerted with nitrogen to ensure that no external sources of oxygen are introduced into containment. The containment inerting system is used during the initial purging of the primary containment early in power operation and provides a supply of makeup nitrogen to maintain primary containment oxygen concentration within TS limits. As a result, the primary containment is maintained at a slightly positive pressure during power operation.

Primary containment pressure is continuously recorded and verified by TS surveillance on a frequency of every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> from the MCR. The fact that the containment is continuously pressurized by the containment inerting system, and is periodically monitored, provides assurance that gross containment leakage that may develop during power operation will be detected.

Additional Considerations Reactor Core Isolation Cooling (RCIC) Operation Emergency Operating Procedure (EOP) changes to credit long term RCIC operation have recently been incorporated into plant operations. These changes have yet to be incorporated in the Plant Hatch PRA model. A sensitivity analysis indicates that CDF may be reduced by greater than 10%. This would produce a measurable decrease in ILRT !lLERF calculations.

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

Procedure 31 EO-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 E1-3

Enclosure 1 to NL-17-0766 SNC Response to Second Set of RAis

  • Add authorization to defeat the RCIC high exhaust pressure isolation to RPV pressure control steps The above changes allow RCIC to operate beyond 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Fire Water Injection In addition, inventory make-up using Fire Protection Water firewater is in the PRA model but given a failure probability of 1.0. Recent changes to firewater hook-ups allow Operators to begin injecting firewater within 30 minutes as demonstrated by JPM LR-JP-36.23-07. Operators use a new valve installed at ground level at the CST enclosure to connect a short fire hose connection. Directions are contained in site procedures.

None of the RCIC EOP or firewater injection procedure changes have been incorporated into the PRA or supporting HRA analysis. Therefore, quantitative assessments of a reduction in ILRT delta risk cannot be performed at this time. These procedure changes are qualitatively assessed as having a measurable and possibly significant reduction of the calculated risk I related to the ILRT interval extension.

PART 1- EVALUATION OF CHANGE TO CDF Evaluation of LOCA type Scenarios Large LOCA (LLOCA)

The earlier response to RAI 6b discussed MAAP cases that demonstrate that with a single train of containment heat removal operating the torus temperature remains well below 211 °F, a temperature level that can impact NPSH requirements for U1 components. Large LOCA scenarios are used as a bounding estimate due to the largest impact on torus temperature. The four MAAP calculations performed for the earlier 6b response demonstrate that for a range of containment leakage rates, the torus temperature remains below approximately 181 °F and NPSH is not lost if torus cooling is available.

For the purposes of this RAI response, an assumption is made that following a Large LOCA, NPSH will be lost should a large leak (1 OOLa) exist. The CDF increase is determined by inserting the following logic under gates for the low pressure pumps taking suction from the torus (Core Spray Pumps CSA and CSB and residual heat removal (RHR) Pumps RHRA, RHRB, RHRC and RHRD) and running the Hatch PRA model. The PRA modeling approach is very conservative in that RHR in Suppression Pool Cooling (SPC) is also not credited, even though this function is not affected by the loss of NPSH for vessel injection.

All CS AND RHR mAINS SUCTlOI'I FAll DUE TO CNT LEAK SUPPRE AND LLOCA FOR PU PROBABILITY FOR LOSS LARGE BREAK LOCA OFNPSH FROM INSIDE THE DRYWELL PRE-EXISTING CNT LEAK (EPRI CLASS 38)

E1-4 to NL-17-0766 SNC Response to Second Set of RAis The Basic Event LEAK-38-NPSH is added to the model and assigned a probability equal to the probability of an EPRI Class 3b leak size.

The calculated increase in risk is 3E-1 0/year as shown below:

Case CDF /year Sensitivity 7 .53402E-06 Baseline 7.53370E-06 Delta Risk 3.2E-10 The risk increase from not crediting Core Spray and RHR Pumps for injection or heat removal during a Large LOCA scenario is not significant. The RHR service water (RHRSW) system is credited for injection and the Torus/Drywall HCjlrdened Vent System is credited for containment heat removal. Credit for RHRSW injection antl the Hardened Vent System, combined with a low frequency Large LOCA initiating event lead to a small delta risk frequency of 3E-10/year.

MLOCA As a sensitivity, the same conservative assumption made for the Large LOCA scenario will be made for a Medium LOCA, NPSH will be lost should a large leak {1 OOLa) exist. A CDF increase is calculated by inserting the following logic under gates for the low pressure pumps taking suction from the Torus (Core Spray Pumps CSA and CSB and RHR Pumps RHRA, RHRB, RHRC and RHRD).

r All CS AND RHR TRAINS I I FAIL DUE TO CNT LEAK I I ANDMLOCA I I I I I I I CNT-LEAK-MLOCAI

_Q I

I I I I PROBABILITY FOR LOSS MEDIUM BREAK LOCA OFNPSH FROM INSIDE THE DRYWELL PRE-EXISTING CNT LEAK (EPRI CLASS 36) u I LEAK-36-NPSHI 2.30E-()3  %

I%MLOCAI 970E-051Y The calculated increase in risk is -4E-09 /year as shown below:

Case CDF /year Sensitivity 2 7.53739E-06 Baseline 7.53370E-06 Delta Risk 3.7E-09 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 E1-5 to NL-17-0766 SNC Response to Second Set of RAis removal. System redundancy, combined with a low frequency Medium LOCA initiating event result in a small delta risk frequency of -4E-09/year.

SLOCA A Small LOCA would be less severe than a Medium LOCA MAAP case. In the Medium LOCA 100 La case Low Pressure Core Spray (LPCS) is lost at 23 hours2.662037e-4 days <br />0.00639 hours <br />3.80291e-5 weeks <br />8.7515e-6 months <br /> due to NPSH issues. MAAP calculation results are shown in Table 2. Considering RPV leakage rates, LPCS would not be lost during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of a small LOCA.

Success Criteria includes Condensate/Condensate Booster Pump and the RHRSW Pumps for injection and the Power Conversion System (PCS), Torus and Drywell Ventilation System for Heat Removal. The PCS System includes the Main Condenser and necessary components to return coolant to the reactor pressure vessel. For information purposes only, a sensitivity was performed using the logic below.

r All CS AND RHR lRAINS I I FAIL DUE TO CNT LEAK I I ANDSLOCA I I I I I I ICNT-LEAK-SLOCAI I I

I _Cd_ I I

PROBABilrtY FOR LOSS SMAll BREAK LOCA OF NPSH FROM INSIDE THE DRYWELL PRE-EXI5nNG CNT LEAK (EPRI CLASS 3B) 0 if ILEAK-38-NPSH I I %SLOCAI 2.30E-03 4 90E-04/Y The calculated increase in risk is 1E-1 0 as shown below:

Case CDF /year Sensitivity 3 7.53384E-06 Baseline 7.53370E-06 Delta Risk 1.4E-1 0 The risk increase from not crediting Core Spray and RHR Pumps for injection or heat removal during a Small LOCA scenario is on the order of the risk increase associated with a LLOCA.

The SLOCA initiating event is approximately two orders of magnitude higher than the Large LOCA frequency, however, additional defense in depth for both injection and heat removal is available.

The following steam LOCA type initiators are also bounded by the Medium LOCA MAAP cases and would not result in loss of NPSH in the first 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s:

  • Inadvertent Opening of Relief Valve (IORV)
  • All SRVs Stuck Open (ALOCA)
  • Stuck Open Relief Valve (SORV) -This scenario is treated in multiple trees. See ATWS discussion below for the impact of a SORV in an ATWS scenario. The E1-6 to NL-17-0766 SNC Response to Second Set of RAis General Transient, LOSP and Station Blackout SORV scenarios are bounded by the MLOCA scenario.
  • SORV1 - One SRV Fails to Reclose
  • SORV2 -Two SRVs Fail to Reclose

For AlWS sequences, a loss of Containment Accident Pressure (CAP) would have a negligible impact on CDF and LERF due to early containment failure from AlWS impacts. The Hatch PRA model groups initiators into 5 groupings and models these groupings with AlWS Initiators.

The total frequency of all initiators subject to a consequential AlWS is -1.2/year. These initiators are represented by AlWS "pseudo" initiators with a total initiating event frequency of 1.19/yr. This representation is equivalent to processing AlWS scenarios as consequential events following an initiating event. The probability of Reactor Protection System (RPS) failure in the Plant Hatch model is 2.56E-06. The probability of a pre-existing leak for an ILRT test interval of 3 per 10 years is 2.3E-03. Multiplying the initiator frequency input with the two probabilities yields 7.1 E-9/year. If Recirc Pump Trip logic fails or there are multiple Stuck Open Relief Valves, the sequence progresses to core damage and containment overpressure has no impact since low pressure injection is not credited. Also, if pressure relief (Reactor Pressure Control) fails, the sequence progresses to core damage and containment overpressure has no impact since low pressure injection is unavailable due to high RPV pressure.

If pressure relief succeeds, high pressure injection from the Feedwater or High Pressure Coolant Injection (HPCI) systems would have to fail in order to require depressurization and low pressure injection. If low pressure injection is required early due to a failure of the HPCI and Feedwater systems; the Condensate, Core Spray, and Low Pressure Coolant Injection (LPCI) systems can provide low pressure makeup. Note that the Condensate system does not depend on containment overpressure because it does not depend on the torus as a suction source.

Using a 0.1 failure probability for all high pressure and low pressure injection and multiplying the AlWS frequency and the chance of a pre-existing leak yields a frequency of 7.1 E-10/year.

Note, this equates to a core damage frequency at 1 per 15-year interval of 5

  • 7.1 E-1 0/year =

3.6E-9/year. This calculation does not credit overpressure that may exist and be adequate for continued ECCS pump injection in the AlWS scenario.

General Transient. Loss of offsite power (LOSP) and Station Blackout The SORVs and AlWS scenarios associated with General Transient, LOSP and Station Blackout initiators are discussed above. As the impact would be less severe than a MLOCA, there would be adequate time between core damage and RPV release to declare a General Emergency (GE) and evacuate personnel. The remaining General Transient, LOSP and Station Blackout scenarios are evaluated further:

General Transients have multiple heat removal systems available. These are the following:

Power Conversion System (PCS)( 1l RHR suppression pool cooling (1 RHR & 1 RHRSW pump)

RHR shutdown cooling (1l The PCS is not available for some General Transients, such as; Loss of Main Condenser (%LOCV), MSIV Closure (%MSIVC) and Loss of Instrument Air (&LOINSTAIR).

E1-7 to NL-17-0766 SNC Response to Second Set of RAis OR Drywell Spray OR Hardened Venting OR Shutdown Cooling PCS, RHR in SPC, or shutdown cooling are heat removal sources for most General Transient Events.

As discussed above any transients with SORVs would be subsumed by the MLOCA analysis.

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 />.

Reactor Pressure Vessel (RPV) Rupture RPV Rupture is assumed to lead to core damage and large early release. No credit is givlen for Low Pressure injection. Thereto~, loss of NPSH is not a consideration.

Summary of ~CDF Risk Increase From Loss of NPSH Due to a Large Pre-existing Leak (1 00 bill The results considering all accident scenarios are summarized below:

Table 1 Summary Table Accident Scenario i1CDF (3 per 10 i1CDF (1 per 15 Comments years) years)<1 >

Large LOCA 3.2E-10 /yr 1.6E-9 /yr A conservative assumption is made that all EGGS heat removal is lost in addition to EGGS low pressure injection when a 100 La leak is present.

Medium LOCA 3.7E-9 /yr 1.9E-8 /yr NPSH lost at 23 hours2.662037e-4 days <br />0.00639 hours <br />3.80291e-5 weeks <br />8.7515e-6 months <br />. Core Damage at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> with 100 La Leak size.

Small LOCA, IORV, -0 -0 MLOCA calc bounds these scenarios. NPSH ALOCA, SORV would not be lost within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the start of the event.

ATWS 7.1E-10 /yr 3.6E-9 /yr Based on a bounding calculation described above.

The bounding calculation does not credit overpressure that may exist during this scenario.

General Transient, LOSP -0 -0 The base model assumes core damage if and Station Blackout containment heat removal fails. Therefore, loss of NPSH does not create additional CDF.

Reactor Pressure Vessel -0 -0 RPV Rupture is assumed to lead to core damage.

Rupture No credit is given for Low Pressure injection.

Therefore, loss of NPSH is not a consideration.

Total 4.7E-9 /yr 2.4E-8 /yr 11> Per EPRI methodology 1 per 15 year ILRT LlCDF is 5 x 3 per 10 year ILRT LlCDF.

E1-8 to NL-17-0766 SNC Response to Second Set of RAis PART II- EVALUATION OF CLASS II SCENARIOS This section explicitly addresses the RAI b statement "Discuss and justify how all the accident scenarios with loss of containment heat removal are considered in the estimate of !lLERF from loss of containment overpressure ...." Note that discussion in this section will overlap the discussion regarding !lCDF in the previous section. MAAP results previously provided will again be referenced.

As noted in RAI b, " ... the containment failure due to overpressure occurs in the 15-hour time frame. While this containment overpressure failure may not be categorized as LEAF due to timing, the loss of NPSH to ECCS pumps caused by a potential pre-existing non-detected leak in the containment could happen much earlier than 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> and therefore it could be categorized as LEAF".

Timing related in Class II events affected by loss of NPSH is analyzed further using MAAP results shown in Table12. The overall question to be answered is whether or not Class II failures should be reclassified LEAF scenarios. I E1-9 to NL-17-0766 SNC Response to Second Set of RAis Table 2 MAAP Results11 )

Time to Large Containment LPCS Lost Time To Core Vessel Release to LLOCACase Leak Size Initial Conditions12)

At Time Damage Failure - Env.

Equivalence (Csl > 10%)

HA050516D1 1 La 16.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> 17.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> 26.8 hrs. LPCS, No SPC 20.4 hrs.

HA050516D2 100 La 5.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> 7.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> 14.5 hrs. LPCS, No SPC 15.3 hrs.

HA050516D3 200 La 4.25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> 5.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 12.3 hrs. LPCS, No SPC 13.0 hrs.

HA050516D4 400 La 3.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 4.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> 12.1 hrs. LPCS, No SPC 12.6 hrs.

Containment Time to Large LPCS Lost Time To Core Vessel MLOCACase Leak Size Initial Conditions Release to Env.

At Time Damage Failure Equivalence (Csl > 10%)

HA050516D1 MLOCA 1 La N/A N/A N/A LPCS, No SPC N/A HA050516D2MLOCA 100 La 23.0 hrs. 24.0 hrs. 34.7 hrs. LPCS, No SPC 35.8 hrs.

HA050516D3MLOCA 200 La 17.2hrs. 18.2 hrs. 30.0 hrs. LPCS, No SPC 30.5 hrs.

HA050516D4MLOCA 400 La 9.9 hrs. 10.9 hrs. 19.3 hrs. LPCS, No SPC 19.9 hrs.

Time to Large Containment LPCS Lost Time To Core Vessel Release to MSIV Closure Case Leak Size Initial Conditions At Time Damage Failure Env.

Equivalence (Csl > 10%)

HA050516D1T 1 La 14.5 hrs. 20.7 hrs. 26.3 hrs. LPCS, No SPC 38+ hrs.

HA050516D2T 100 La 6.3 hrs. 11.0 hrs. 19.1 hrs. LPCS, No SPC 38+ hrs.

HA050516D3T 200 La 4.3 hrs. 8.6 hrs. 15.8 hrs. LPCS, No SPC 38+ hrs.

HA050516D4T 400 La 3.3 hrs. 7.1 hrs. 14.8 hrs. LPCS, No SPC 38+ hrs.

Notes to above MAAP Runs (a) All MAAP case durations are 38 hours4.398148e-4 days <br />0.0106 hours <br />6.283069e-5 weeks <br />1.4459e-5 months <br />.

(b) Core spray is used consistent with the MAAP cases presented in the original LAR.

E1-10 to NL-17-0766 SNC Response to Second Set of RAis Large LOCA Additional MAAP Cases for Class II Containment Scenario MAAP has the capability to determine when a loss of NPSH occurs. MAAP case results where one LPCS Pump is providing injection and there is no containment cooling are shown in Table 2 and summarized below:

Leakage Size LPCS Lost At Time To Core Time to Large Release LLOCA Case (Equivalence) (Csl > 10%)

Ti me Damage HA050516D2 100 La 5.9 hrs. 7.0 hrs. 15.3 hrs.

The case most representative of the ILRT evaluation, but still conservative, is the 100 La leakage case shown above. As mentioned earlier, it is expected that gross leakage (such as a 100 La leak) would be detected by on-going monitoring. The larger leakage sizes in Table 2 are provi~ed for information. The Plant Hatch evacuation time i ~ Level 2 analysis is 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and 50 minutes or - 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. As can be seen above, core damage occurs -7 hours after the start of the event in the most representative case. Based on the Emergency Action Levels (EALs), the Loss of RCS Barrier and Loss of Fuel Cladding Barrier would occur due to low RPV water level.

With core damage occurring, no source of injection, and containment conditions trending towards failure, the General Emergency would be declared 30 minutes after core damage based on Emergency Director (ED) discretion at 7.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. The magnitude of the release based on Cs-1 content would not reach the "Large" state until after 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br />, therefore a General Emergency would be declared more than 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> before a large release occurring at 15.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.

Note, the bounding assumptions of loss of NPSH were applied in the CDF increase calculations described in the previous section. Further conservatism is unwarranted. Therefore, Large LOCA Class II scenarios with loss of Low Pressure injection remain as non-LEAF releases.

Medium LOCA The Large LOCA assessment above bounds the Medium LOCA scenario. However, the assessment is further analyzed for information purposes:

Medium LOCA Case HA050516D2MLOCA with a leak size equivalent to 100 La, is the appropriate for analysis for Medium LOCA Class II L scenarios. MAAP case results where one LPCS Pump is providing injection and no RHR pump providing containment cooling are shown in Table 2 and summarized below:

Leakage Size LPCS Lost At Time To Core Time to Large MLOCA Case (Equivalence) Release (Csl > 10%)

Time Damage HA050516D2MLOCA 100 La 23.0 hrs. 24.0 hrs. 35.8 hrs.

As shown above, Core Damage occurs at 24.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> leading to a Large Release for the most representative leakage size (1 00 La). Based on the EALs, the Loss of RCS Barrier and Loss of Fuel Cladding Barrier would occur due to low RPV water level. With core damage occurring, no source of injection, and containment conditions trending towards failure, the General Emergency would be declared 30 minutes after core damage based on ED discretion at 24.5 E1-11 to NL-17-0766 SNC Response to Second Set of RAis hours. The magnitude of the release based on Cs-1 content would not reach the "Large" state until after 35 hours4.050926e-4 days <br />0.00972 hours <br />5.787037e-5 weeks <br />1.33175e-5 months <br />. Therefore, a General Emergency would be declared more than 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> before a large release occurring at 35 hours4.050926e-4 days <br />0.00972 hours <br />5.787037e-5 weeks <br />1.33175e-5 months <br />.

Note, the bounding assumptions of a late General Emergency were applied in the CDF increase calculations performed earlier. Further conservatism appears unwarranted. Therefore, Medium LOCA Class II scenarios with loss of Low Pressure injection remain as non-LEAF releases.

Small LOCA.IORV. ALOCA. SORV These scenarios are bounded by the Medium LOCA scenario described above.

ATWS All ATWS scenarios are considered "early releases", including releases that are caused by containment failure. Therefore, there is no increase in LEAF caused by a late release being reclassified as an early release.

General Transient. LOSP and Station Blackout The SORVs and ATWS scenarios associated with General Transient, LOSP and Station Blackout initiators are discussed above. The remaining General Transient, LOSP and Station Blackout scenarios are bounded by a MSIV Closure scenario. MSIV Closure Case HA050516D2T with a leak size equivalent to 100 La, is the appropriate for analysis for these class II scenarios.

MAAP results where one LPCS Pump is providing injection and there is no containment cooling is shown in the Table below:

MSIV Closure Leakage Size LPCS Lost At Time To Core Time to Large Release Case (Equivalence) Time Damage (Csl > 10%)<1>

HA05051602T 100 La 6.3 hrs. 11.0 hrs. 38+ hrs.

1

<> For the MSIV closure cases the Csl release does not exceed 10% during the 38 hours4.398148e-4 days <br />0.0106 hours <br />6.283069e-5 weeks <br />1.4459e-5 months <br /> run.

As shown above, Core Damage occurs at 11.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> for the most representative leakage size (1 00 La). A large release does not occur until significantly later in the accident scenario at 38+

hours. Based on the EALs, the Loss of RCS Barrier and Loss of Fuel Cladding Barrier would occur due to low RPV water level. With core damage occurring, no source of injection, and containment conditions trending towards failure, the General Emergency would be declared 30 minutes after core damage based on ED discretion at 11.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. The magnitude of the release based on Cs-1 content would not reach the "Large" state until after 38 hours4.398148e-4 days <br />0.0106 hours <br />6.283069e-5 weeks <br />1.4459e-5 months <br /> therefore a General Emergency would be declared more than 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> before a large release occurring at 38 hours4.398148e-4 days <br />0.0106 hours <br />6.283069e-5 weeks <br />1.4459e-5 months <br />.

Reactor Pressure Vessel Rupture RPV Rupture is assumed to lead to core damage. No credit is given for Low Pressure injection.

E1-12 to NL-17-0766 SNC Response to Second Set of RAis Class II LERF Bounding Assessment As described above, even with a large containment leak Class II sequences can still be classified as non-LERF. However, a bounding assessment considers EP personnel do not declare aGE. Based on the long timeframes available to declare aGE, general HRA methods would show a HEP of less than 1E-3. Conservatively a value of 1E-3 is used to bound the potential failure to declare a General Emergency early. This bounding assessment is the following:

Class IIA frequency+ Class Ill frequency= 3.39E-6/yr + 4.11 E-1 0/yr Class II Total frequency= 3.39E-6/yr Probability EP personnel fail to declare a General Emergency early = 1E-3 Frequency of a General Emergency not declared prior to a Class II = 3.39E-6

  • 1E-3 llLERF = 3.39E-9/yr E1-13 to NL-17-0766 SNC Response to Second Set of RAis PART Ill-RESULTS AND CONCLUSIONS Results Summary An estimate of the totalllCDF, llLERF, change in population dose, and change in the conditional containment failure probability from all contributors/hazards (internal events, fire, external events, loss of containment over pressure) is provided and compared to acceptance criteria in Section 3.2.4.6 of the Safety Evaluation for EPRI TR 1009325, Revision 2, are met for the application.

Calculation of llCDF llCDF acceptance criteria are typically not addressed explicitly in the EPRI Guidance as ILRT changes do not significantly impact CDF. Acceptance criteria from Reg. Guide 1.174 will be used.

The bounding analysis above notes a potential increase in CDF associated with loss of NPSH.

The llCDF from loss of NPSH assumed ECCS systems are lost for both injection and heat removal in Large and Medium LOqA scenarios. A calculation for llCDF for ATWS scenarips is also performed. The EPRI guidance for a bounding analysis is to assume injection is considered lost with a 100 La leak. It is also noted that the Plant Hatch bounding analysis also assumed loss of heat removal. This bounding assessment led to a CDF increase of 2.4E-8/year when changing ILRT frequency from 3 per 10 years to 1 per 15 years. An internal events CDF to external events CDF multiplier of 1.21 is used for this calculation. The 1.21 multiplier is discussed in more detail in the llLERF calculation discussion below.

CDF CDF Contribution Contribution (3-per-1 0 year CDF Contribution (1-per-15 year ILRT) (1-per-10 year ILRT) ILRT) CDF lncrease< 1>

Internal Events Contribution< 2> 4.70E-09 1.57E-08 2.35E-08 1.88E-08 External Events Contribution 5.69E-09 1.89E-08 2.84E-08 2.27E-08 (Internal Events x 1.21)

Combined (Internal + External) 1.04E-08 3.46E-08 5.19E-08 4.15E-OB 1

( l Associated with the change from the 3-per-1 0 year frequency to the proposed 1-per-15 year frequency.

(2 l Values from Table 1 of this report.

Reg. Guide 1.174 identifies a llCDF less than 1E-06 /year as being in the very small" region.

Therefore, one need not consider the impact to total CDF.

Calculation of llLERF and Overall Total LEAF In response to RAI 8b, the external event multiplier was recalculated using the current CDF and seismic event CDF information rather than the IPEEE internal events CDF and older seismic information. This analysis is considered the best estimate and will be updated to demonstrate the impact of the loss of over pressurization caused by a per-existing large Leak.

Section 1 presents the calculation tables from the response to round 1 RAI 8b. Section 2 updates the tables showing the impact of loss of over pressurization.

E1-14 to NL-17-0766 SNC Response to Second Set of RAis Section 1: Relevant RAJ Bb Information As discussed in RAI Response Sa, 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 = 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 I 7.57E-6/yr = 1.21 Table 5.8-2 (Updated Internal Events CDF and Seismic CDF)

HATCH CLASS 3b (LEAF) AS A FUNCTION OF ILRT FREQUENCY I FOR INTERNAL AND EXTERNAL EVENTS (INCLUDING AGE ADJUSTED STEEL CORROSION LIKELIHOOD) 3b 3b Frequency 3b Frequency Frequency (1-per-15 year (3-per-1 0 yr ILRT) (1-per-10 year ILRT) ILRT) LEAF lncrease< 1l Internal Events Contribution< 2l 1.58E-08 5.29E-08 7.97E-08 6.39E-08 External Events Contribution 1.91E-08 6.40E-08 9.64E-08 7.73E-08 (Internal Events x 1.21)

Combined (Internal + External) 3.49E-08 1.17E-07 1.76E-07 1.41 E-07 1

<l Associated with the change from the 3-per-1 0 year frequency to the proposed 1-per-15 year frequency.

<2l Values from Table 5.7-1 of the ILRT Risk Assessment The results of using the higher external events multiplier based on the current internal events CDF and lower seismic CDF shows a LEAF increase of 1.41 E-7/yr which is in the "small" impact for deltas of <1 E-6/yr. Using the same external events multiplier the total LEAF is calculated to confirm that total LEAF is < 1E-5/yr as directed by RG 1.174, as follows.

Frequency Internal Events LEAF 1.12E-06/yr External Events LEAF (Internal events LEAF x 1.21) 1.36E-06/yr Total LEAF (Internal + External) 3.26E-06/yr Section 2- Updated Tables.

The existing 3b Frequency from Table 5.8-2 above is added to f1CDF from loss of NPSH and f1LERF from Class II reassigned to LEAF. For example, the 3 per 10 year ILRT input is:

IE Contribution from Table 5.8-2 1.58E-08/yr f1CDF from Table 1 4.7E-09/yr Class II frequency assigned to LEAF 3.4E-09/yr Total 3 per 10 year 3b frequency = 2.4E-08/yr E1-15 to NL-17-0766 SNC Response to Second Set of RAis The calculation will use 2.4E-08 /year as an input to update Tables in Section 1 and reproduced below.

Table 5.8-2 (Includes RAI3b Increase)

(Updated Internal Events CDF and Seismic CDF)

HATCH CLASS 3b (LEAF) 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-1 0 yr ILRT) (1 -per-10 year ILRT) ILRT) LEAF lncreasel1>

Internal Events Contribution1 2 l 2.40E-08 7.99E-08 1.20E-07 9.60E-08 External Events Contribution 2.90E-08 9.67E-08 1.45E-07 1.16E-07 (Internal Events x 1.21)

Combined (Internal ij External) 5.30E-08 1.77E-07 2.s5F-o1 2.12E-07 11l Associated with the change from the 3-per-1 0 year frequency to the proposed 1-per-15 year frequency.

12l Values from Table 5.7-1 of the ILRT Risk Assessment The results of using the higher external events multiplier based on the current internal events CDF and lower seismic CDF shows a LEAF increase of 2.1 E-7/yr which is in the "small" impact for deltas of <1 E-6/yr. Using the same external events multiplier the total LEAF is calculated to confirm that total LEAF 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) 9.60E-08 External Events LERF due to ILRT (at 15 years) (Internal 1.16E-07 Events LERF

  • 1.21)

Total LERF (Internal+ External) 2.69E-06 The calculated total LEAF of 2.7E-6/yr meets the acceptance criteria of <1 E-5/yr.

E1-16

Enclosure 1 to NL-17-0766 SNC Response to Second Set of RAis Change in Population Dose The change in population dose is reflected in a change in 3b frequency. A smallllLERF 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 P-All 3b Freq. P-REM/yr 3b Freq. REM/yr 3b Freq. P-REM/yr P-REM/yr lntervalsl 1> (3/10) (3/10). (1/1 0)) (1/1 0). (1/15) (1/15). Increase Internal Events 1.15E+05 2.40E-08 2.76E-03 7.99E-08 9.19E-03 1.20E-07 1.38E-02 1.10E-02 Contribution12 >

External Events 1.15E+05 2.90E-08 3.34E-03 9.67E-08 1.11 E-02 1.45E-07 1.67E-02 1.34E-02 Contribution (Internal Events x 1.21) I I Combined (Internal 1.15E+05 5.30E-08 6.10E-03 1.77E-07 2.03E-02 2.65E-07 3.05E-02 2.44E-02

+External)

!l) 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.

12> These values are from Table 5.8-2 (Includes RAI NPSH t.LERF 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 + 2.44E-02/yr llperson-rem/year = 3.43E-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% of the total population dose, whichever is less restrictive. The llperson-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 is 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.

E1-17 to NL-17-0766 SNC Response to Second Set of RAis Conclusions An updated assessment of 6LERF due to loss of containment overpressure credit impacting Unit 1 ECCS NPSH, taking into account external hazards, was performed. 6LERF due to loss of containment overpressure, including the contributions from external hazards to loss of containment overpressure risk, was justified through the use of bounding calculations. An increase of CDF associated with LOCA scenarios where the initial containment pressurization helps to satisfy the NPSH requirements for early injection was performed by defeating both ECCS low pressure injection and heat removal functions for Large and Medium LOCAs. The EPRI guidance for a bounding analysis is to assume injection is considered lost with a 100 La leak. Also, assuming the heat removal function is lost is an additional conservatism. ATWS scenarios were assumed to be impacted by a loss of NPSH as described in PART I.

Overpressure was conservatively assumed inadequate for the ATWS scenario.

The LAR stated that a pre-existing containment leak, resulting in the loss of adequate NPSH to the ECCS pumps would have the same result as containment failure from overpressure when containment heat removal is not available. Twelve additional MAAP runs were completed and all the acci~ent scenarios with loss of containment heat removal w4re evaluated to determine if an increase in 6LERF from loss of containment overpressure is justified. Rationale for continuing to call the releases a late release was provided. In addition, a bounding assessment was performed to address uncertainty in the declaration of a GE.

An estimate of the total 6CDF, 6LERF, change in population dose, and change in the conditional containment failure probability from all contributors/hazards (internal events, fire, external events, loss of containment over pressure) was performed. It was confirmed that the acceptance criteria in Section 3.2.4.6 of the Safety Evaluation for EPRI TR 1009325, Revision 2, are met for this application.

As noted in the introduction section, recent plant operational changes related to long term use of RCIC and Fire Water injection, are not yet incorporated into the Plant Hatch PRA. These changes will lead to a reduction in CDF likely off-setting the increase of 6CDF and 6LERF related to loss of ECCS injection from loss of NPSH due to a pre-existing leak.

Bounding calculations and other considerations continue to support a conclusion that increasing the ILRT interval to 15 years is considered to represent an insignificant change in risk for Plant Hatch.

E1-18

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