ML15334A453

R.E. Ginna Nuclear Power Plant- Interim Staff Response to Reevaluated Flood Hazards Submitted in Response to 10 CFR 50.54(f) Information Request-Flood Causing Mechanism Reevaluation

ML15334A453
Person / Time
Site:	Ginna
Issue date:	12/04/2015
From:	Tekia Govan Japan Lessons-Learned Division
To:	Bryan Hanson Exelon Generation Co
Govan, Tekia NRR/JLD 415-6197
References
TAC MF6098
Download: ML15334A453 (9)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Mr. Bryan C. Hanson President and Chief Nuclear Officer Exelon Generation Company, LLC 4300 Winfield Road Warrenville, IL 60555 December 4, 2015

SUBJECT:

R.E. GINNA NUCLEAR POWER PLANT-INTERIM STAFF RESPONSE TO REEVALUATED FLOOD HAZARDS SUBMITTED IN RESPONSE TO 10 CFR 50.54(f) INFORMATION REQUEST-FLOOD-CAUSING MECHANISM REEVALUATION (TAC NO MF6098)

Dear Mr. Hanson:

The purpose of this letter is to provide a summary of the U.S. Nuclear Regulatory Commission (NRC) staff's assessment of the re-evaluated flood-causing mechanisms described in the March 11, 2015 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML15072A009, flood hazard reevaluation report (FHRR) submitted by R. E. Ginna Nuclear Power Plant, LLC (the licensee) for R. E. Ginna Nuclear Power Plant (Ginna), as well as supplemental information resulting from requests for additional information and audits. By letter dated March 12, 2012, the NRC issued a request for information pursuant to Title 10 of the Code of Federal Regulations, Section 50.54(f) (hereafter referred to as the 50.54(f) letter) (ADAMS Accession No. ML12053A340).

The request was issued as part of implementing lessons-learned from the accident at the Fukushima Dai-ichi nuclear power plant. Enclosure 2 to the 50.54(f) letter requested licensees to re-evaluate flood-causing mechanisms using present-day methodologies and guidance.

Concurrently, with the reevaluation of flooding hazards, licensees were required to develop and implement mitigating strategies in accordance with NRC Order EA-12-049, "Requirements for Mitigation Strategies for Beyond-Design-Basis External Events" (ADAMS Accession No. ML12054A735).

On March 30, 2015, the Commission provided Staff Requirements Memoranda (SRM) (ADAMS Accession No. ML15089A236) to COMSECY-14-0037, "Integration of Mitigating Strategies for Beyond-Design-Basis External Events and the Reevaluation of Flooding Hazards," dated November 21, 2014 (ADAMS Accession No. ML14309A256), affirming that licensees need to address the reevaluated flooding hazards within their mitigating strategies for beyond-design-basis external events. The NRC staff has reviewed the information submitted by the licensee and has summarized the results of the review in the tables provided as an Enclosure to this letter. Table 1 provides the current design-basis flood hazard mechanisms.

Table 2 provides the reevaluated flood hazard mechanisms; however, the reevaluated flood hazard mechanisms bounded by the current design-basis (Table 1) are not included.

The NRC staff has concluded that the licensee's reevaluated flood hazards information, as summarized in the Enclosure, is suitable for the assessment of mitigating strategies developed in response to Order EA-12-049 (i.e., defines the mitigating strategies flood hazard information B. Hanson described in guidance documents currently being finalized by the industry and NRC staff) for Ginna. Further, the NRC staff has concluded that the licensee's reevaluated flood hazard information is a suitable input for other assessments associated with Near-Term Task Force Recommendation 2.1 "Flooding." The NRC staff plans to issue a staff assessment documenting the basis for these conclusions at a later time. In addition, Nuclear Energy Institute (NEI) guidance document NEI 12-06 "Diverse and Flexible Coping Strategies (FLEX) Implementation Guide" is currently being revised. This revision will include a methodology to perform a Mitigating Strategies Assessment (MSA) with respect to the reevaluated flood hazards. Once this methodology is endorsed by the NRC, flood event duration parameters and applicable flood associated effects should be considered as part of the Ginna MSA. The NRC staff will evaluate the flood event duration parameters (including warning time and period of inundation) and flood-related associated effects developed by the licensee during the NRC staff's review of the MSA. As stated above, Table 2 of the enclosure to this letter describes the reevaluated flood hazards that exceed the current design-basis.

In order to complete its response to the information requested by Enclosure 2 to the 50.54(f) letter, the licensee is expected to submit an integrated assessment or a focused evaluation, as appropriate, to address these reevaluated flood hazards, as described in the NRC letter, "Coordination of Request for Information Regarding Flooding Hazard Reevaluation and Mitigating Strategies for Beyond-Design-Basis External Events" (ADAMS Accession No. ML15174A257).

This letter describes the changes in the NRC's approach to the flood hazard reevaluations that were approved by the Commission in its SRM to COMSECY-15-0019, "Closure Plan for the Reevaluation of Flooding Hazards for Operating Nuclear Power Plants" (ADAMS Accession No. ML15209A682).

B. Hanson If you have any questions, please contact me at (301) 415-6197 or e-mail at Tekia.Govan@nrc.gov.

Docket No. 50-244

Enclosure:

Summary of Results of Flooding Hazard Re-Evaluation Report cc w/encl: Distribution via Listserv Sincerely, Tekia Govan, Project Manager Hazards Management Branch Japan Lessons-Learned Division Office of Nuclear Reactor Regulation ENCLOSURE:

SUMMARY

TABLES OF REEVALUATED FLOOD HAZARD LEVELS

R.E. Ginna Nuclear Power Plant Table 1. Current Design Basis Flood Hazards for Use in the MSA Mechanism Local Intense Precipitation Screen House Streams and Rivers PMF -Reactor Building PMF -Auxiliary Building PMF -Turbine Building PMF -Control Building PMF -All-Volatile-Treatment Building PMF -Standby Auxiliary Feedwater Pump Building PMF -Screen House PMF -Diesel Generator Building Failure of Dams and Onsite Water Control/Storage Structures Stillwater Elevation 254.5 ft NGVD29 272.0 ft NGVD29 273.8 ft NGVD29 256.6 ft NGVD29 272.0 ft NGVD29 272.0 ft NGVD29 273.0 ft NGVD29 256.6 ft NGVD29 i i I Waves/ Run up Minimal i Not I , applicable Not applicable Not applicable Not applicable Not applicable

!I Not applicable Not applicable 256.6 ft Not NGVD29 applicable i I Not included , Not included in DB 1 in DB Design Basis Hazard Elevation 254.5 ft NGVD29 272.0 ft NGVD29 273.8 ft NGVD29 256.6 ft NGVD29 272.0 ft NGVD29 272.0 ft NGVD29 273.0 ft NGVD29 256.6 ft NGVD29 256.6 ft NGVD29 Not included in DB Reference FHRR Section 1.3 FHRR Table 2.2-6 FHRR Table 2.2-6 FHRR Table 2.2-6 FHRR Table 2.2-6 FHRR Table 2.2-6 FHRR Table 2.2-6 FHRR Table 2.2-6 FHRR Table 2.2-6 FHRR Section 3.1.3 R.E. Ginna Nuclear Power Plant Table 1. Current Design Basis Flood Hazards for Use in the MSA --------**--------------------

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Mechanism Stillwater Waves/ Design Basis Reference Elevation Run up Hazard Elevation Storm Surge I Tropical Storm and Associated 253.3 ft 7.7 ft 260.9 ft FHRR Section 1.3 phenomena NGVD29 NGVD29 I Seiche Not included Not included Not included FHRR Section 3.1.5 in DB in DB in DB I I Tsunami I Not included Not included Not included FHRR Section 3.1.6 I in DB I in DB in DB I I I I Ice-Induced Flooding ! No Impact No Impact No Impact FHRR Section 3.1.7 I on the Site i on the Site on the Site Identified I Identified Identified Channel Migrations/Diversions Not included Not included Not included FHRR Section 3.1.8 in DB in DB in DB Note 1: Reported values are rounded to the nearest one-tenth of a foot.

R.E. Ginna Nuclear Power Plant Table 2. Reevaluated Flood Hazards for Flood-Causing Mechanisms for Use in the MSA I Mechanism Stillwater Waves/ Reevaluated Reference I Elevation Run up Hazard I I Elevation I I Local Intense Precipitation

! ! LIP 6-h event: Reactor 270.9 ft Minimal 270.9 ft FHRR Table 2.1-1 Containment I NGVD29 NGVD29 I I LIP 6-h event: Auxiliary Building 270.7 ft Minimal 270.7 ft FHRR Table 2.1-1 NGVD29 NGVD29 I ' I I LIP 6-h event: Turbine Building 255.8 ft Minimal 255.8 ft FHRR Table 2.1-1 I NGVD29 NGVD29 I I ' LIP 6-h event: Control Building i 270.9 ft Minimal 270.9 ft FHRR Table 2.1-1 i NGVD29 NGVD29 ! I LIP 6-h event: All-Volatile-i 270.8 ft Minimal 270.8 ft FHRR Table 2.1-1 Treatment Building NGVD29 NGVD29 I I LIP 6-h event: Standby Auxiliary 270.2 ft Minimal 270.2 ft I FHRR Table 2.2-1 Feedwater Pump Building NGVD29 NGVD29 LIP 6-h event: Proposed Standby 270.5 ft Minimal 270.5 ft FHRR Table 2.1-1 Auxiliary Feedwater Pump NGVD29 NGVD29 Building Annex LIP 6-h event: Screen House 255.8 ft Minimal 255.8 ft FHRR Table 2.1-1 NGVD29 NGVD29 LIP 6-h event: Diesel Generator 255.8 ft Minimal 255.8 ft FHRR Table 2.1-1 Building NGVD29 NGVD29 Streams and Rivers ' Reactor Building (stillwater:

PMF; 272.4 ft Not 272.4 ft FHRR Table 2.2-6 total water level: PMF, H.1, H.4.2) NGVD29 applicable NGVD29 FHRR Sections 2.9.3 and 3.1.9 Turbine Building (stillwater:

PMF; 258.2 ft Not 258.2 ft FHRR Sections 2.9.3, 3.1.9 and total water level: H.4.2) NGVD29 applicable NGVD29 FHRR Enclosure 1 Table 4 ' I I R.E. Ginna Nuclear Power Plant Table 2. Reevaluated Flood Hazards for Flood-Causing Mechanisms for Use in the MSA Mechanism Control Building (stillwater:

PMF; total water level: PMF, H.1, H.4.2) Standby Auxiliary Feedwater Pump Building (stillwater:

PMF; total water level: H.1) Proposed Standby Auxiliary Feedwater Pump Building Annex (stillwater:

PMF; total water level: H.1) j Screen House (stillwater:

PMF; total water level: H.4.2) Diesel Generator Building (stillwater:

PMF; total water level: H.4.2) Stillwater Elevation 272.4 ft NGVD29 272.8 ft NGVD29 273.5 ft NGVD29 258.2 ft NGVD29 258.4 ft NGVD29 Waves/ Run up Reevaluated Hazard Elevation Not I 272.4 ft applicable 1 NGVD29 0.9 ft 0.9 ft I 273.7 ft NGVD29 274.4ft NGVD29 Not 258.2 ft applicable NGVD29 Not applicable 258.4 ft NGVD29 Reference FHRR Table 2.2-6 FHRR Section 2.9.3 and 3.1.9 FHRR Table 2.2-6 FHRR Enclosure 1 Tables 2 and 3 FHRR Enclosure 1 Tables 2 and 3 FHRR Sections 2.9.3, 3.1.9 and FHRR Enclosure 1 Table 4 FHRR Sections 2.9.3, 3.1.9 and FHRR Enclosure 1 Table 4 Note 1: The licensee is expected to develop flood event duration parameters and applicable flood associated effects to conduct the MSA. The staff will evaluate the flood event duration parameters (including warning time and period of inundation) and flood associated effects during its review of the MSA. Note 2: Reevaluated hazard mechanisms bounded by the current design basis (see Table 1) are not included in this table Note 3: Reported values are rounded to the nearest one-tenth of a foot.

B. Hanson If you have any questions, please contact me at (301) 415-6197 or e-mail at Tekia.Govan@nrc.gov.

Docket No. 50-244

Enclosure:

Sincerely, IRA! Tekia Govan, Project Manager Hazards Management Branch Japan Lessons-Learned Division Office of Nuclear Reactor Regulation Summary of Results of Flooding Hazard Re-Evaluation Report cc w/encl: Distribution via Listserv DISTRIBUTION:

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