NL-18-045, IP3 ILRT RAI Responses

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IP3 ILRT RAI Responses
ML18193A452
Person / Time
Site: Indian Point Entergy icon.png
Issue date: 07/03/2018
From:
Entergy Nuclear Operations
To:
Office of Nuclear Reactor Regulation
Shared Package
ML18193A465 List:
References
NL-18-045
Download: ML18193A452 (23)
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Text

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  • 61'l Ill 82'll IP3 ILRT RAI RESPONSES Prepared For

- E~te,gx Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 Buchanan, NY 10511-0249 Revision: 1 Project#: 1DEV21082 Project Name: IP3 ILRT RAI Responses Report#: 021082-RPT-02 Name and Date Oigita!ly signed by Donald E . Macl eod Preparer: ON : C"Ui>, Eadmacleod@jansenhu,ghes.com, Q:JE NSE N HUGHES . OU:Power Services Group.

Donald E. Macleod ~~~.:::.';.~:'-.;;::'.A Reason: l amthea uthor orthisdocumenl Coritactlnfo: 518-583-35 25 Date:-2018.06.2815:48:46.Q4'Cl0' Prepa rer:

Digitally signed by Harry Liao Date: 2018.06.28 15:23 :50-04'00' Reviewer:

Review Method Desig n Review [gJ Alternate Calculation D Approver:

Revision 1

021082-RPT-02

- Revis101

- Record Summary REVISION RECORD

SUMMARY

Revision Revision Summary 0 Initial Issue Revision to reference citations and page numbers.

Revision 1 Page i

021082-RPT -02 Table of Contents TABLE OF CONTENTS

1.0 RAl-01

External Events Screening .............................................................................. 1 1.1 Response .. .. .... ..... ...... .... ......... ... ......... ...... ........ .. .. .. ........................................... . 1 1.1.1 High Winds and Tornadoes (including Tornado and Hurricane Missiles) .. 1 1.1.2 External Flooding .......... .... ........ ....... .. ....... ... ...... ... .... ..... ...... .. .... .. ... .... ... .. 2 1.1 .3 Chemical, Transportation and Nearby Facility Accidents .......... ..... ..... ..... .4 1.1.4 Aircraft Hazards .... ....... ....... ..... ........... .... ...... .. ..... ... ..... .... .... ............ .... .... .7 1.1.5 lce ......... .... .... ....... ... .... ... .... .... .. ... .......... .... ... ...... .... .......... ... ... ... .. ..... .. .. .. .. 7 1.2 Summary Conclusion ... .... ..... ............. ........ ..... ... .. .... .......... .. .. .. ..... .... .... ... .. ..... ..... 7

2.0 RAl-02

Assumptions Used for Alternative Approach for External Events lmpact ... 8 2.1 Response .. ... .. .... ...... ..... ... ...... ...... ....... ...... ..... ..... ...... ..... ... .. .... ... .... .. ... ... ..... .. .... ..8 3.0 References ............... .. ...................................................... ............................................ 18 Revision 1 Page ii

021082-RPT 02 RAI f xte'n ii Events ~-rePn ng

1.0 RAl-01

EXTERNAL EVENTS SCREENING Electric Power Research Institute (EPRI) Technical Report No. 1009325, Revision 2-A (ADAMS Accession No. ML14024A045) states that "[w]here possible , the analysis should include a quantitative assessment of the contribution of external events (for example, fire and seismic) in the risk impact assessment for extended ILRT intervals. For example, where a licensee possesses a quantitative fi re analysis and that analysis is of sufficient quality and detail to assess the impact, the methods used to obtain the impact from internal events should be applied for the external event. " EPRI TR-1009325 , Revision 2-A further states that the "assessment can be taken from existing , previously submitted and approved analyses or another alternate method of assessing an order of magnitude estimate for contribution of the external event to the impact of the changed interval. "

In Section 5.7 of Attachment 1 to the LAR, the licensee performed an assessment of external event contribution. The licensee's analysis reflected the contribution from internal fire and seismic event. The licensee stated that high winds , external floods and "Other" external events were considered negligible in estimation of the external events impact on the ILRT extension application. This conclusion appears to be reached based on the IP3 Individual Plant Examination for External Events (IPEEE) analysis performed in 1997.

Consistent with the Regulatory Guide (RG) 1.174 guidance that the probabilistic risk assessment (PRA) scope, level of detail , and technical acceptability be based on the as-built and as-operated plant, and maintained to reflect the current operating experience at the plant, provide justification for the applicability of the IPEEE conclusions to the current plant and its environs, considering each of the external hazards screened from this application and taking into account any updated risk studies and insights. The analysis should include all hazard groups (i.e. , high winds , external flooding , transportation events, aircraft, industrial facilities, and other external hazards) 1.1 Response 1.1.1 High Winds and Tornadoes (including Tornado and Hurricane Missiles)

The major concern in a high-wind or tornado scenario are the wind loads imposed on the buildings/major structures and the potential for wind-generated missiles to disable systems or components necessary to shut down the plant or maintain the plant in a safe shutdown condition . IP3 wind and tornado loadings are evaluated under Section 16.2 of UFSAR [1 ]. All Class I buildings and structures were designed to withstand tornado winds corresponding to 300 mph ta ngential velocities , traverse velocities of 60 mph and a differential pressure drop of 3 psi in 3 seconds with no loss of function . In addition , all Class I buildings and structures were also designed to withstand various postulated tornado generated missiles, including a plank and a 4000-lb passenger car. Note that since the IP3 IPEEE [2] , RG 1.76 [3] was updated to lower the required design basis tornado wind speeds to 200 mph for the region in which IP3 is located.

IP3 IPEEE evaluated the probability and consequences of high winds and tornadoes for the as-built, as-operated plant. The IPEEE concluded that the risk of core damage associated with high winds and tornadoes was below the 1E-6 per year criterion for concern , and the impact of tornado-generated missiles on safety-related reinforced concrete structures at IP3 is of no significance.

Subsequent to the finalization of the IPEEE, additional analyses of high wind events were performed to support other industry initiatives. Hurricanes and tornadoes were "screened in" and assessed for IP3 FLEX based on screening criteria in NEI 12-06 [4] . All lP3 Phase 2 FLEX Revision 1 Page 1

021082 RPT 02 RAI lxter'lal E:ver t:-, Screening equipment is stored in structures with designs that are robust such that no one external event can reasonably fail the FLEX capability. Although none of the IP3 tanks have an overt protection from tornado missile damage, IP-RPT-13-00055 [5] shows that tank separation and relative location will ensure that either a Condensate Storage Tank (CST) or the City Water Tank (CWT) will survive a tornado event, and IP-RPT-15-00031 [30] concludes that the Refuel Water Storage Tank (RWST) is protected from a tornado driven missile hit by adjacent structures such as the Unit 3 Containment and Unit 1 Containment and thus will likely survive a tornado. In addition , IP-RPT-15-00030 [6] shows that IP3 water tanks , including the Primary Water Storage Tank (PWST), CWT, CST, RWST, and Fire Water Storage Tanks (FWSTs), can survive credible hurricane borne missiles . According to IP-CALC-14-00044 [7] and IP-CALC-14-00012 [8] , a single RWST can provide sufficient borated water to both units until an offsite borated water source can be established . With the CWT surviving , there is sufficient non-borated water sources to manage cool down of both units.

As a result of a study performed to identify potential tornado missile barrier vulnerabilities, the following two actions will be implemented to revise Section 4.0 of Attachment 3 of OAP-008 "Severe Weather Preparations" [1 O] .

  • Close tornado Door 07 located on the 73' elevation of the Unit 3 PAB, upon issuance of a tornado warning by the National Weather Center.
  • After tornado passage , inspect Service Water Vent pipe on Unit 3 PAB roof for damage.

If vent pipe is crimped shut, then open up pipe to allow for venting .

There have been no major changes to the buildings/major structures or location of important safety equipment within them since the IPEEE submittal in 1997. The only significant change is the addition of FLEX equipment and procedure additions or changes which provide the station with additional response capability to an event. Therefore , it is concluded that no new factors have been introduced at IP3 since the submittal of the IPEEE in 1997 that would result in an increase in the risk associated with high winds , tornadoes , tornado missiles and hurricane missiles.

1.1.2 External Flooding On March 12, 2012, the NRC issued a request for information associated with Near-Term Task Force (NTTF) Recommendation 2.1 for Flooding [11 ]. One of the Required Responses in that request 'directed licensees to submit a Flood Hazard Reevaluation Report (FHRR) using present-day methodologies and guidance. For IP3, the FHRR [12] was submitted on December 23, 2013. On December 9, 2014, FHRR Revision 2 [13] was submitted. IP3 also provided supplemental information in letters dated August 18, 2014 [14] , December 10, 2015 [15] , and April 5, 2016 [16] . After a review of the FHRR and supplemental information , the NRC staff concluded that the reevaluated flood hazard results for the following flood-causing mechanisms are not bounded by the current design basis (COB) of IP3.

  • Local intense precipitation (LIP) - the flood level is higher than the COB level ;
  • Flooding in streams and rivers - the Hudson River probable maximum flood height flood level exceeded the COB;
  • Dam breeches and failures - the flood level for a dam failure combined with Hudson River probable maximum flood (PMF) exceeded the COB;
  • Storm surge - the flood level for a combined event coincident with the probable maximum storm surge exceeded the COB.

For the above mechanisms, IPEC has performed a flooding mitigation strategies assessment (MSA) [17]. The MSA found limited impact to the site Flex Strategy from the Reevaluated Flood Revision 1 Page 2

021082-RPT 02 RAI E:..xtprrial E:..vents Screening Hazards and described the changes that the IPEC had implemented, or was in the process of implementing, to the existing FLEX Strategy such that they could be successfully implemented and deployed for the Reevaluated Flood Hazards. The modifications included raising manholes, sealing conduits, changes to procedures to implement temporary passive flood protection features, and raising the temporary flood protection features on two interior doors leading to the Unit 2 480 Volt switchgear room and exterior doors leading to both the auxiliary feedwater pump and control building from the IP2 and IP3 transformer yard during storm preparations. The NRC staff concluded that IPEC has demonstrated the mitigation strategies , if appropriately implemented , can reasonably be executed under the Reevaluated Flood Hazard conditions for beyond-design-basis external events [18] .

The Nuclear Energy Institute (NEI) prepared NEI 16-05 Rev. 1, "External Flooding Assessment Guidelines" [19]. The NRC endorsed NEI 16-05 Rev. 1 and recommended changes, which have been incorporated into NEI 16-05, Revision 1. NEI 16-05 indicates that each flood-causing mechanism not bounded by the Design Basis (DB) flood (using only stillwater and/or wind-wave run-up level) should be addressed by either a focused evaluation (FE) or an integrated assessment (IA) following one of the following five assessment paths :

  • Path 1: Demonstrate Flood Mechanism is Bounded Through Improved Realism;
  • Path 2: Demonstrate Effective Flood Protection;
  • Path 3: Demonstrate a Feasible Response to Local Intense Precipitation (LIP) ;
  • Path 4: Demonstrate Effective Mitigation;
  • Path 5: Scenario Based Approach .

IPEC performed a preliminary assessment of the four non-bounded flood-causing mechanisms and determined that three mechanisms (LIP, Dam Failure, and Streams and Rivers Stillwater) would screen as a Path 2 FE and the controlling storm surge combined event would trigger an IA. The assessment concludes:

  • The LIP event has effective flood protection for permanent plant equipment as discussed in [13] , [17], and [18];
  • The Dam Failure event has effective protection provided by site grade;
  • Streams and Rivers Stillwater with advance warning from a significant rainfall event or storm surge event is less than the temporary flood protection features currently implemented in abnormal operating procedures (AOPs);
  • The risk of the controlling storm surge combined event resulting in consequential impact to the Unit 2 & 3 Service Water Pumps is determined to be low based on the assessment of two scenarios: (1) a storm surge flood scenario that is below 17.7' NGVD29 (National Geodetic Vertical Datum of 1929) water surface elevation and (2) a storm surge flood scenario that is above the 17.7' NGVD29 water surface elevation. The expected frequency of the first scenario is below the 1E-04/year threshold of concern

[13]. The annual exceedance probability (AEP) for the storm surge combined event is on the order of 2E-05 for 17.7'. The NRC staff concluded that the storm surge events that exceed 17. 7' are rare and are unlikely to exceed the screening criteria of 1E-04/year

[20] . In addition, the NRC staff determined that (1) IP3 key safety functions are protected from flood mechanisms below 17.9', (2) key safety functions that are provided by structures, systems, or components east of the IP3 turbine building are protected to the storm surge reevaluated flood height of 18.9' for this area, and (3) IP3 modified FLEX strategy for the reevaluated storm surge, if implemented appropriately, could be effectively deployed to mitigate the storm surge event at 18.9'. Furthermore, the storm surge event should have sufficient warning time associated with it such that the plant is Revision 1 Page 3

021082 RPT 02 RAI 1:- I(' rr ii t:-ven's Sc eeri1nq placed in the best possible condition to withstand the event. This includes IP3 being shut down well in advance of a storm surge flood thus reducing the decay heat that needs to be removed from the reactor. Other preparations would be enacted in accordance with the licensee's severe weather procedures . That is , IP3 has an effective flood strategy as discussed in the SRM of COMSECY-15-0019 [21] per guidance in NEI 16-05 for a Path 5 IA.

IPEC was expected to complete the FE and IA by December 2018. In its letter to the NRC dated July 24, 2017 [22] , IPEC requested deferral of the FE and IA in light of its decision to cease operation for IP2 in 2020 and IP3 in 2021. In a letter dated October 4, 2017 [20] , the NRC granted the deferral request and concluded the following :

  • IPEC could develop an appropriate NEI 16-05, Revision 1, Path 5 IA for the storm surge flood mechanism (to demonstrate low flood event frequency and the availability of a feasible response strategy) ;
  • IP2 and IP3 key safety functions are protected from the non-bounded flood-causing mechanisms (i.e. LIP, Dam Failure , Streams and Rivers Stillwater, and Storm Surge),

and additional regulatory actions associated with the flood hazards, beyond those proposed as part of the mitigation strategies assessment, are likely not warranted for these hazards; The risk associated with these events is judged to be low based on the discussion above, the fact that the NRC has granted IPEC's FE and IA deferral request, and the following considerations (quoted from [20]) .

  • "IP2 and IP3 have achieved additional defense in depth for coping with an extended loss of alternating current power and loss of normal access to the ultimate heat sink due to external events, including those caused by seismic and flooding events, as a result of the licensee's compliance with Orders EA-12-049 [23] and EA-12-051 " [24] .
  • "The staff reviewed the licensee submittals which indicated that the impact to IP2 and IP3 from the Reevaluated Flooding Hazards is within IPEC's ability to cope with the hazards ."
  • "The information does not indicate any concern regarding adequate protection of public health and safety for either IP2 or IP3 dLJe to the reevaluated seismic and flooding hazard . Furthermore, based on the information developed to date, the staff did not identify any substantial safety improvements that would be cost-justified to implement during the period of deferral."

Based on the discussion above , it is concluded that the flood protection measures at IP3 are adequate for the non-bounded flood-causing mechanisms of LIP , Dam Failure , Streams and Rivers Stillwater, and Storm Surge.

1.1.3 Chemical, Transportation and Nearby Facility Accidents Toxic hazards The toxic hazards result from the release of toxic chemicals and asphyxiants such that control room habitability might be threatened.

As discussed in Section 1.3.2 of UFSAR, separate chlorine , anhydrous ammonia, and carbon dioxide diffusion type probes will detect the presence of these gases in the control room outside the air intake duct. Upon a toxic gas detection alarm , the operator will be able to place the control room air conditioning system in the 100% recirculation mode to stop the intake of outside Revision 1 Page 4

021082 RPT-02 RAI F xtt=Jrn2I E=vents Screen in~

air. An additional toxic gas detection system indicates the oxygen, chlorine, and anhydrous ammonia levels in the control room atmosphere.

IP3 IPEEE assessed the following to identify all potentially toxic chemicals and asphyxiants at or within five miles of the plant.

  • Military and industrial facilities within five miles of IP3 that store or use chemicals that could give rise to a toxic hazard at the IP3 site were identified;
  • Chemicals stored on-site at IP3 that could give rise to an explosion or airborne toxic hazard (by evaporation or combustion);
  • Possible transportation incidents in the vicinity of the plant involving the potential airborne release of toxic materials.

Shipments of toxic material by rail were eliminated as sources of concern because Conrail states that no hazardous chemicals are transported within five miles of the plant. Shipments of hazardous material by road , other than to or from local facilities, must use interstate highways.

Variou s potential toxic release incidents were analyzed. It was concluded that either the release poses no risk to control room habitability at IP3 or the predicted frequencies with which control room habitability at IP3 is affected are less than the frequency criterion of concern.

A control room habitability evaluation was performed in 2003 [31 ], 2008 [32] and 2013 [33] .

These evaluations presented a survey of stationary and mobile sources of hazardous chemical and asphyxiant releases that might pose a potential threat to control room habitability. Each evaluation consisted of a comprehensive survey of all sources of hazardous chemicals and asphyxiants at and within a 5-mile radius of each of the Indian Point Unit 2 and Unit 3 Central Control Rooms' makeup air intakes at IPEC, including chemicals deliveries to IPEC and shipments by water, rail, and ground within 5-mile radius of IPEC. The potential accidents involving hazardous chemicals identified in the comprehensive survey were then assessed against screening criteria devised to exempt low risk hazardous chemical releases from detailed evaluation . The control room habitability evaluations concluded that no additional mitigation was required to address releases of hazardous chemicals. The release of such chemicals will result in concentrations at the control room air intakes that are less than the immediate danger to life and health (IDLH) values for those chemicals or the frequency of release is predicted to be lower than the threshold for concern and the control room operators would in any case have adequate time to don protective gear.

  • Based on the information summarized above , it is concluded that core damage frequency associated with toxic hazards at IP3 is small , and would not impact the conclusions from the ILRT extension from the currently approved 15 years to 16 years ..

Explos ion hazards Potential explosion hazards at IP3 include explosions within plant structures, vapor cloud explosions (VCEs) and boiling liquid expanding vapor explosions (BLEVEs). IP3 IPEEE identified potential sources of these explosions within five miles of the plant and concluded that either explosions caused by hydrogen, propane, and carbon dioxide pose no risks to the plant or the explosion occurrence frequencies are below the frequency criterion for concern . Given that the cumulative core damage frequency from hydrogen explosions is approximately 1E-06/yr, installation of an excess flow valve at the outside hydrogen storage facility to stop flow in the event of a hydrogen line rupture was implemented as a mitigative measure to reduce the risk.

Two natural gas transmission pipelines owned and operated by the Algonquin Gas Transmission Company cross the IP3 plant site within 400 ft of safety related structures. The Revision 1 Page 5

02*os2 RPT 02 RAI ~xtcrr'll Events Streening two gas pipelines are discussed in Section 2.2.2 of UFSAR. With a very conservative approach ,

the IP3 IPEEE concluded that the rupture of a natural gas pipeline does not pose a major risk and the resulting contribution to the core damage frequency is less than 1E-6/year screening value. The NRC staffs evaluation of the IP3 IPEEE did not identify any concerns with the conclusion [29] .

The consequences of natural gas releases from the pipelines were re-evaluated in 2008 and 2015. The 2008 study [25] determined that the rupture of the pipelines and subsequent ignition of the methane released will not damage any safety related structure. The impact of pipe fragments on safety-related structures at the IPEC is bounded by the scenarios considered in the FSAR. Even in the unlikely event of a hypothetical vapor cloud explosion , structural damage to buildings other than the warehouse adjacent to the pipelines will not occur. A flammable vapor cloud fire that engulfs the plant is unlikely because the turbulent momentum with which the methane exits the pipe line will result in low methane concentrations close to the point of release.

The 2015 study [26] identified a potential equipment issue that was not addressed in the 2008 study [25] . The issue is related to elevated EOG air intake temperatures due to a prolonged elevated heat flux (on the order of Yi hour or more) in the general area impacted by the gas pipeline rupture. Although the two studies agree that the heat flux at the EOG structure will not exceed a level at which damage will occur, the 2008 study did not examine the EOG room air temperature. The potential jet fire effect on local intake ambient temperatures is currently yet to be determined and therefore a potential issue. The lack of damage to the dampers / motors is based on engineering judgment and not engineering evaluation . Based on the 2015 study, this is not a current operability concern since the potential for a pipeline rupture is judged to be sufficiently low for underground piping and alternate means to safely shut down the plant are available if IP3 EDGs are assumed unavailable due to pipeline rupture. Thus it is concluded that the prolonged elevated heat flux is a negligible contributor to risk.

The Algonquin Gas Transmission Company has installed a 42 inch gas pipeline after 2015 that crosses the Hudson south of the site property and turns north. It passes through the easternmost corner of the site and then crosses Broadway between the Switchyard and the GT 2 I 3 fuel oil storage tank. The 42 inch gas line joins the 30 inch and 26 inch gas pipelines north of the switchyard and east of the site. A safety evaluation [28] was performed for the proposed pipeline and the juncture with the 30 and 26-inch pipelines . Based on the proposed routing of the 42-inch pipeline further from safety related equipment at IPEC and accounting for the substantial design and installation enhancements agreed to by the Algonquin Gas Transmission Company, the evaluation concluded that the failure of these gas lines would not impair the safe operation of the plant and that there would be no significant reduction in the margin of safety because the effects are acceptable or can be excluded by probability.

An additional analysis [27] of a potential rupture of the underground portions of the existing pipeline was subsequently performed using the same methodology as for the 42-inch pipeline.

It was concluded that safe operation would not be impaired since the plant would be able to safely shutdown following any affects such as damage to overhead transmission lines and that the probability of ruptures were sufficiently remote that ruptures could be excluded based on probability. During operation of the 42 inch pipeline the 26 inch pipeline will normally be idled (at normal operating pressure but no gas flow) which further restricts probability.

The control room habitability evaluations performed in 2003 [31 ], 2008 [32] and 2013 [33] also considered the possibility and consequences of explosions and fires following chemical releases within the vicinity of the IPEC. The evaluations identified no chemicals that might pose an Revision 1 Page 6

021082 RPT 02 RAI External Events Screening explosion hazard that were not addressed or are not bounded by problems addressed in IP3 IPEEE and the gas pipeline analyses [25, 26, 27, 28] .

Based on the information summarized above, it is concluded that core damage frequency associated with explosive hazards at IP3 is small, and would not impact the conclusions from the ILRT extension from the currently approved 15 years to 16 years .

1.1.4 Aircraft Hazards IP3 IPEEE concluded that all SRP aircraft hazard screening criteria were satisfied and aircraft crashes on IP3 pose no significant risk of causing core damage. It is noted that the number of fatal crashes continues in a downward trend since the 1997 IP3 IPEEE. Plane crash data found at http://www.planecrashinfo.com/cause.htm , shows a more than 50% reduction in fatal crashes since 1997. A review of airports within 10 miles of IP3 found no new airports since 1997, and one airport within 10 miles of IP3 was closed in 2007. According to the information on the Westchester County Airport website and another independent website (see links below), there is a drop in yearly total aircraft operations at the Westchester County Airport.

  • https://a irport.westchestergov .com/about-us/
  • https ://www.lohud .com/story/news/i nvestigations/2016/04/06/westchester-ai rport--

crossroads/82521212/

Based on the discussion above, the conclusion in the IPEEE that aircraft hazards pose no significant risk of causing core damage is considered to be valid .

1.1.5 Ice IP3 IPEEE determined that as the design of the intake structure and pumps makes the probability of ice blockage remote: the service water pumps can obtain , water through four separate intakes; the service water pump suction is at 10 ft below mean sea level (and 6 ft below the hypothesized extreme low river level). According to Section 9.6.1 of the UFSAR (2017), there has been no change to the design of the service water intake structure and pumps and there is no significant risk to IP3 related to icing effects.

1.2 Summary Conclusion Based on the discussions above, the conclusions in the IPEEE indicating that the Fire and Seismic hazards are the most relevant to IP3 External Events risk are considered to remain valid , and the quantitative focus of the one-time ILRT frequency extension analysis on those hazards is considered to be appropriate.

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021082 RPT 02

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ASSUMPTIONS USED FOR ALTERNATIVE APPROACH FOR EXTERNAL EVENTS IMPACT Section 2.5.3 of RG 1.174, Revision 3, states, "[t]he impacts of using alternative assumptions or models may be addressed by performing appropriate sensitivity studies or by using qualitative arguments, based on an understanding of the contributors to the results and how they are impacted by the change in assumptions or models." In addition, Section 2.5.5 of RG 1.174 states, "[i]n general, the results of the sensitivity studies should confirm that the guidelines are still met even under the alternative assumptions (i.e. , change generally remains in the appropriate region) ."

In Section 5.7 of Attachment 1 to the LAR, the licensee performed an assessment of external event contribution . The licensee used two approaches : the multiplier approach, which applies a multiplier to the internal events results (the multiplier is derived from the ratio of external events core damage frequency (CDF) to the internal events CDF) and an "alternative" approach where each EPRI accident class frequency is re-examined . The licensee calculated an increase in population dose risk from changing the ILRT frequency from three in 10 years to once in 16 years as 3.67 person-rem/year or 1.00% (when using the multiplier approach) and 3.84 person-rem/year or 0.69% when using the alternative approach . The reported increase in total population dose is close to the acceptance criteria values of 1 person-rem/year or 1% provided in EPRI TR-1009325, Revision 2-A, and defined in Section 3.2.4.6 of the NRC safety evaluation for NEI 94-01, Revision 2 (ADAMS Accession No. ML081140105).

It appears that the reduction in the % population dose change obtained through the use of the "alternative" approach included in Section 5.7.5 of Attachment 1 to the LAR relies on increasing the total estimated population dose due to external hazards for the base case, corresponding to the three in 10-year frequency. Two assumptions appear to be key to this reduction in population dose change:

  • The frequency for the EPRI accident Class 7 sequences (accidents involving containment failure induced by severe accident phenomena) is increased by assuming that 50% of the CDF is due to late Class 7 sequences , as compared to the internal events value of 15%, with the justification that "the external events contributors are dominated by unrecoverable SBO-like scenarios".
  • The frequency o'f the EPRI accident Class 2 sequences (containment isolation failures) is increased by assuming that 0.1 % of the external events CDF is due to large containment isolation failures , as opposed to the internal events contribution of 0.03% ,

with the justification that "seismic and fire-initiated events would likely be more susceptible" to large containment isolation failures.

In accordance with the guidance in RG 1.17 4, provide a detailed justification for assuming the value of 50% for the external events contribution to the Class 7 late sequences and the value of 0.1% for the Class 2 sequences . The explanation should include the concepts used to identify the assumptions made for the contributors which ultimately resulted in dose risk change to less than 1.0%. Include a discussion of the conservatisms in the analysis and the risk significance of these conservatisms.

2.1 Response As identified in RAl-02, the percent change in the population dose risk is at the threshold of the acceptance criterion for the proposed change. The purpose of the "alternative" approach included in Section 5.7.5 of Enclosure 1 to the LAR was to demonstrate that when the characteristics of the external events contributors are accounted for, rather than assuming they Revision 1 Page 8

021082-RP1 02 RAI Assumption"' used fo~ Alh rnat ve Apr_roach for Fxter'lal fver,ts lmoact behave in the same manner as the internal events contributors , the percent change in the population dose risk moves further below the acceptance criterion of 1 percent. As identified in RAl -02 , the key assumption used in that sensitivity assessment is that the risk profile of the external events scenarios is skewed more toward late containment failures and containment isolation failures than the internal events accident scenarios. In order to better justify this assumption , the IP3 IPEEE [2] was reviewed to identify supporting plant-specific information.

As documented in the LAR and in the response to RAl-01, the external hazards that are significant quantitative contributors to plant risk are limited to fire and seismic events ; therefore, the review of the external events hazards performed for this response has been restricted to those types of events.

While the IP3 IPEEE does include containment response information, the nature of the Level 2 risk assessments differed between the external events hazards that were analyzed. For example, a seismic containment event tree was developed and documented to an extent that it was possible to extract quantitative results from the analysis; however, for the fire analysis, a containment event tree was not provided an it was necessary to rely on the qualitative discussions in the IPEEE to determine how the fire frequency would be distributed among plant release categories .

EPRI Class 7, Failures Induced by Phenomena (Late) : The IP3 IPEEE includes a seismic containment event tree (Figure 3.1.6.1) that documents the release category and the corresponding probability of each of the sequence end states, which has been included in this response as Figure 2-1. The "late" containment failure sequences in that tree , which include long term containment overpressure failures and basemat melt-through failures, correspond to EPRI Accident Class 7, "Failures Induced by Phenomena (Late)". The frequency of the IP3 "late" seismic failures can be determined by multiplying the mean seismic CDF reported in the IPEEE of 4.4E-05 by the sequence probability and summing the result for each of the "late" sequences. The total "late" frequency of 2.71 E-05 represents 61 .6 percent of the seismic CDF, which is larger than the percent of CDF that was binned to EPRI Accident Class 7 in the "alternative" approach (50 percent).

The IP3 IPEEE Fire analysis, which was performed using the EPRI FIVE methodology, does not explicitly include a containment event tree quantification; however, Section 4.8.4 provides a qualitative summary of containment performance for Fire events. That section states that the IP3 IPE identified early containment failures are primarily the result of sequences in which the accident initiator causes containment bypass (interfacing systems loss of coolant accidents and steam generator tube rupture events), but the fire risk assessment concludes that there are no such dominant fire-induced initiators and that late containment failure is more probable than early containment failure. While this indicates that "late" failures are more likely than "early" failures for fire, it does not directly provide information describing the relative magnitudes of the "late" and "intact" scenarios. However, review of the largest contributors to the IP3 Fire CDF does indicate that most of these scenarios would likely lead to "late" containment failures:

  • Zone 14, 480-V Switchgear Room (62 .2% of fire CDF) : Over 77% of the contributors in this zone lead to loss of both 480-V switchgear 31 and 32. The Appendix R diesel backed 480-V bus 312 is potentially available , but this bus does not power the fan cooler units or the containment spray system such that containment heat removal would not be available without a recovery action . The recovery action, which is proceduralized ,

requires routing alternate, temporary power cables from the selected fan cooler units to the 312 bus and loading the fan cooler units onto the bus. There is currently not an explicit quantification of this operator action , but for post core damage conditions, credit for ex-MCR actions would likely be limited (e.g., in the range of 0.5). The credit taken for the action essentially determines how the CDF is distributed between the "intact" and the Revision 1 Page 9

021082 RPT 02 RA, As., 1mpt1ons uc:eo for Altern 1t1vt Appro,H,h for l=xternc1I E:-ve'1h lmpa~

"late" containment endstates; therefore, the impact of the HEP {human error probability) assigned to this action must be evaluated.

  • Zone 11, Cable Spreading Room ( 12.1% of fire CDF) : The dominant fires in this zone lead to a loss of power to the safe shutdown equipment, and in the case that the operators fail to align the Appendix R safe shutdown equipment, core damage occurs .

In these scenarios, containment heat removal would also not be available and late containment failure would be the dominant endstate.

  • Zone 14/37A (multi-compartment) , 480-V Switchgear Room/South Turbine Building (8%

of fire CDF) : The largest contributors for these scenarios are those that lead to a loss of both on-site and offsite power to the 480-V safeguard buses and safe shutdown equipment. As with the zone 14 fires , the containment heat removal systems themselves are not damaged, but supporting 480VAC power would need to be supplied via the action to manually route cables from the 312 bus to the fan cooler unit.

  • Zone 10, Diesel Generator Room 31 (3.8% of fire CDF): These scenarios are primarily SBO cases (after random LOSP) , which are dominated by "late" containment failures .
  • Zone 102A, Diesel Generator Room 33 (3.4% of fire CDF): These scenarios are similar to those in Zone 10 and would also primarily be binned as "late" containment failures.

Based on the discussion above, two cases have been developed to estimate the percent of the fire CDF that is binned to the "late" containment failure mode: "Fire Case A" in which an HEP of 0.9 is used for the manual alignment of 480V power to support the fan cooler units, and "Fire Case B" in which the HEP is assumed to be 0.0 (the operator never fails to recover containment heat removal):

  • Fire Case A: 100% of Zones 11 , 10, and 102A are binned to "late" containment failure, 90% of the contributors from Zones 14 for which both 480V AC buses have failed are binned to "late" containment failure , 90% of the Zone 14/37A contribution is binned to "late" containment fa ilure, and 0% of the "other" zones are binned to "late" containment failure .

o The percent contribution of the fire CDF to the "late" frequency is therefore :

62.2*0.77*0.9 + 12.1 *1.0 + 8.0*0.9 + 3.8*1.0 + 3.4*1.0 + 10.5*0 = 69 .6%

  • Fire Case B: 100% of Zones 11 , 10, and 102A are binned to "late" containment failure ,

0% of Zones 14 and 14/37A are binned to "late" containment failure , and 0% of the "other" zones are binned to "late" containment failure :

o The percent contribution of the fire CDF to the "late" frequency is therefore: 62 .2*0.0

+ 12.1 *1.0 + 8.0*0.0 + 3.8*1.0 + 3.4*1.0 + 10.5*0 = 19.3%

As presented above , Fire Case A approximates an upper bound of the percentage of fire scenarios which can lead to late containment failure , and Fire Case B approximates a lower bound of the percentage of fire scenarios which can lead to late containment failure .

EPRI Class 2, Large Isolation Failures (Failure to Close): The ERPI Class 2 sequences are those comprised of large isolation failures, which can be caused by the failure of the support systems required to close the containment isolation valves . For fire and seismic initiators, there are potentially initiator induced failures not represented in the internal events model that would lead to the loss of these support systems such that it would not be possible to close the motor, solenoid, and air operated containment isolation valves . The IPEEE does not provide any quantitative information that would allow a plant-specific change to be made to the fraction of Revision 1 Page 10

021082 RP1 02 R.AI Asc;urnp*1ons uc;ed for Al*t rnat1ve Approal n fo f xte*ri ,I Fvents Impact CDF that would lead to containment isolation failures for seismic and fire events; therefore , the internal events model results have been retained (0.03% of CDF) for this updated analysis .

Updated Results: The "alternative" approach to account for external events contributions documented in Section 5.7.5 of Enclosure 1 to the LAR has been updated using the plant-specific information provided above in order to better demonstrate that the acceptance criteria for this LAR have been met:

  • The EPRI Class _7-CFL frequency has been recalculated as follows:

o Seismic: 2.65E-05

  • 0.616 = 1.63E-05 o Fire Case A: 2.55E-05
  • 0.696 = 1.77E-05 o Fire Case B: 2.55E-05
  • 0.193 = 4.92E-06
  • The EPRI Class_2 frequency has been recalculated as follows :

o 0.0003

  • 5.2E-05 = 1.56E-08 The updated results were used to generate new "base cases" (Tables 2-1 a and 2-1 b for Fire Cases "A" and "B" respectively) and the cases in which the ILRT frequency has been changed to 1 per 16 years (Tables 2-2a and 2-2b for Fire Cases "A" and "B respectively) . As shown in Table 2-3, the percent change in dose risk , including internal and external events contributors, is 0.57% for Fire Case "A" and 0.80% for Fire Case "B". The results of Fire Cases "A" and "B" are both below the acceptance criterion and they envelop the dose risk change of 0.69% that was estimated in Section 5. 7 .5 of Enclosure 1 of the LAR.

A potential conservatism in this assessment is the assumption that all of the "other" fire CDF contributors (i.e. those not in the fire zones explicitly discussed) do not contribute to the "late" containment release category. While some fraction of those events would result in "late" containment failure scenarios , the impact of explicitly including them in the dose risk calculations would be small, increase the margin to the acceptance criterion , and would not change the conclusions of the analysis. In addition, the results presented here and in Enclosure 1 of the LAR reflect the change in risk from a case in which the ILRT is performed three times in ten years to a case in which the ILRT is performed once per sixteen years. The change in risk between the currently accepted one-time ILRT frequency of one per fifteen years and the proposed frequency of one per sixteen years is much smaller.

Other approximations and assumptions used in the analysis may also have small impacts on the results ; however, the qualitative insight that the fire and seismic accident scenarios are more likely to lead to "late" containment failures than the internal events scenarios is considered to be the key issue because it implies that there would be more margin to the dose risk acceptance criterion when those contributors are incorporated into the analysis.

Revision 1 Page 11

021082-RPT-02 RAI Assumptions used for Alternative Approach for External Events Impact Table 2-1a Population Dose Risk As A Function Of Accident Class (IP3 Alternative External Events Base Case Sensitivity - Fire Case "A")1 Accident Class Description Frequency Dose Dose Risk (Containment (1/yr) (Person-Rem) (Person-Rel ease Type) Rem/yr)

Containment Intact 1.68E-05 4.41E+04 7.39E-01 2 Large Isolation Failures (Failure to 1.56E-08 5.08E+07 7.92E-01 Close) 3a Small Isolation Failures (liner 4.?0E-07 4.41 E+05 2.0?E-01 breach) 3b Large Isolation Failures (l iner 1.18E-07 4.41 E+06 5.18E-01 breach) 4 Small Isolation Failures (Failure to N/A N/A N/A seal -Type B) 5 Small Isolation Failures (Failure to N/A N/A N/A seal-Type C) 6 Other Isolation Failures (e.g., N/A N/A N/A dependent failures) 7-CFE Failures Induced by Phenomena 6.76E-07 5.08E+07 3.43E+01 (Early) 7-CFL Fai lures Induced by Phenomena 3.38E-05 1.63E+07 5.50E+02 (Late) 8-SGTR Containment Bypass (Steam 2.03E-07 5.08E+07 1.03E+01 Generator Tube Rupture) 8-ISLOCA Containment Bypass (Interfacing O.OOE+OO 5.08E+07 O.OOE+OO System LOCA)

CDF All CET End States (Including 5.20E-05 597 .3 Intact Case) 1 The results do not include the impacts of corrosion, which are demonstrated to be small in Enclosure 1, Section 6.1 of the LAR.

Revision 1 Page 12

021082 RPT 02 Table 2-1b Population Dose Risk As A Function Of Accident Class (IP3 Alternative External Events Base Case Sensitivity - Fire Case"B")2 Accident Class Description Frequency Dose Dose Risk (Containment (1/yr) (Person-Rem) (Person-Release Type) Rem/yr)

Containment Intact 2.93E-05 4.41 E+04 1.29E+OO 2 Large Isolation Failures (Failure to 1.56E-08 5.08E+07 7.92E-01 Close) 3a Small Isolation Failures (liner 4.?0E-07 4.41 E+05 2.0?E-01 breach) 3b Large Isolation Failures (liner 1.1?E-07 4.41 E+06 5.18E-01 breach) 4 Small Isolation Failures (Failure to N/A N/A N/A seal -Type B) 5 Small Isolation Failures (Failure to N/A N/A N/A seal-Type C) 6 Other Isolation Failures (e .g., N/A N/A N/A dependent failures) 7-CFE Failures Induced by Phenomena 6.76E-07 5.08E+07 3.43E+01 (Early) 7-CFL Failures Induced by Phenomena 2.12E-05 1.63E+07 3.46E+02 (Late) 8-SGTR Containment Bypass (Steam 2.03E-07 5.08E+07 1.03E+01 Generator Tube Rupture) 8-ISLOCA Containment Bypass (Interfacing O.OOE+OO 5.08E+07 O.OOE+OO System LOCA)

CDF All CET End States (Including 5.20E-05 393.8 Intact Case) 2 The results do not include the impacts of corrosion , which are demonstrated to be small in Enclosure 1, Section 6.1 of the LAR.

Revision 1 Page 13

021082-RPT -02 RA Ac,c;umot1ons uc;ea for Altwn<it1ve Approach for E=xternal E=verts Impact Table 2-2a Population Dose Risk As A Function Of Accident Class (IP3 Alternative External Events Evaluation Characteristic of Conditions For 1 in 16 Year ILRT Frequency Sensitivity - Fire Case "A") 3 Accident Class Description Frequency Dose Dose Risk (Containment (1/yr) (Person-Rem) (Person-Release Type) Rem/yr)

Containment Intact 1.42E-05 4.41E+04 6.27E-01 2 Large Isolation Failures (Failure to 1.56E-08 5.08E+07 7.92E-01 Close) 3a Small Isolation Failures (liner 2.51 E-06 4.41E+05 1.11 E+OO breach) 3b Large Isolation Failures (liner 6.27E-07 4.41E+06 2.76E+OO breach) 4 Small Isolation Failures (Failure to N/A N/A N/A seal -Type B) 5 Small Isolation Failures (Failure to N/A N/A N/A seal-Type C) 6 Other Isolation Failures (e .g., N/A N/A N/A dependent failures) 7-CFE Failures Induced by Phenomena 6.76E-07 5.08E+07 3.43E+01 (Early) 7-CFL Failures Induced by Phenomena 3.38E-05 1.63E+07 5.50E+02 (Late) 8-SGTR Containment Bypass (Steam 2.03E-07 5.08E+07 1.03E+01 Generator Tube Rupture) 8-ISLOCA Containment Bypass (Interfacing O.OOE+OO 5.08E+07 O.OOE+OO System LOCA)

CDF All CET End States (Including 5.20E-05 600 .3 Intact Case) 3 The results do not include the impacts of corrosion , which are demonstrated to be small in Enclosure 1, Section 6.1 of the LAR.

Revision 1 Page 14

021082-RPT-02 RAI Assumptions used for Alternat,w Approach for fxtern1I Events Impact Table 2-2a Population Dose Risk as a Function of Accident Class (IP3 Alternative External Events Evaluation Characteristic of Conditions For 1 in 16 Year ILRT Frequency Sensitivity - Fire Case "8")4 Accident Class Description Frequency Dose Dose Risk (Containment (1/yr) (Person-Rem) (Person-Release Type) Rem/yr)

Containment Intact 2.67E-05 4.41E+04 1.18E+OO 2 Large Isolation Failures (Failure to 1.56E-08 5.08E+07 7.92E-01 Close) 3a Small Isolation Fai lures (liner 2.51 E-06 4.41E+05 1.11 E+OO breach) 3b Large Isolation Failures (liner 6.27E-07 4.41E+06 2.76E+OO breach) 4 Small Isolation Failures (Failure to N/A N/A N/A seal -Type B) 5 Small Isolation Failures (Failure to N/A N/A N/A seal-Type C) 6 Other Isolation Failures (e.g., N/A N/A N/A dependent failures) 7-CFE Failures Induced by Phenomena 6.76E-07 5.08E+07 3.43E+01 (Early) 7-CFL Failures Induced by Phenomena 2.12E-05 1.63E+07 3.46E+02 (Late) 8-SGTR Containment Bypass (Steam 2.03E-07 5.08E+07 1.03E+01 Generator Tube Rupture) 8-ISLOCA Containment Bypass (Interfacing O.OOE+OO 5.08E+07 O.OOE+OO System LOCA)

CDF All CET End States (Including 5.20E-05 396.8 Intact Case) 4 The results do not include the impacts of corros ion , which are demonstrated to be small in Enclosure 1, Section 6 .1 of the LAR.

Revision 1 Page 15

021082-RPT 02 RAI Assur'pt*orc:; .Jsed fer Ater*1at ve Approach for External Events Impact Table 2-3 Comparison to Acceptance Criteria Including Alternative External Events Evaluation Contribution for IP3 Sensitivity - Fire Case "A" and Fire Case "B" 5 Contributor Person-Re Person-Rem/yr (Fire Case A) (Fire Case B)

IP3 Internal EventSJ-in-10 81 .15 81 .15 IP3 External EventSJ-in-10 597.3 393.8 IP3 Totab-in-10 678.4 474.95 IP3 Internal Events1 -in-1 6 81 .96 81 .96 IP3 External Events 1-in-16 600.3 396.8 IP3 Totah -in-16 682.3 478.76 Delta 3.90/yr 3.81/yr (0 .80%)

(0 .57%)

Acceptance Criterion <1 .0 person- <1.0 person-rem/yr or rem/yr or

<1 .0% <1.0%

The external events results do not include the impacts of corrosion, which are demonstrated to be small in Enclosure 1, Section 6.1 of the LAR.

Revision 1 Page 16

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Revision 1 Page 17

021082 RP1 0 J Re'ererices

3.0 REFERENCES

1. Indian Point Energy Center, Indian Point Energy Center Unit 3 Ultimate Final Safety Analysis Report (UFSAR) (Rev . 07), 2017.
2. Indian Point Unit 3 Nuclear Power Plant, Individual Plant Examination of External Events (IPEEE), 1997.
3. U.S. Nuclear Regulatory Commission. Regulatory Guide 1.76, Design-Basis Tornado and Tornado Missiles for Nuclear Power Plants (Rev. 1), 2007.
4. Nuclear Energy Institute , Diverse and Flexible Coping Strategies (FLEX) Implementation Guide (Rev. 4), December 2012.
5. Indian Point Energy Center, IP-RPT-13-00055 (Rev. 0), Indian Point Energy Center Unit 3 Tornado Missile Impact on Water Storage Tanks.
6. Indian Point Energy Center, IP-RPT-15-00030 (Rev. 0), Effects of Hurricane Missiles on FLEX Credited Water Sources.
7. Indian Point Energy Center, IPCALC-14-00044, (Rev. 2), IP2 RCS Inventory Evaluation for FLEX, August 23, 2017.
8. Indian Point Energy Center, IPCALC-14-00012 (Rev . 1 &(2, Indian Point Energy Center Unit 3 RCS Inventory Evaluation for FLEX (EC 70064)
9. IP-CALC-13-00064, Rev. 0, FLEX Event Evaluation of Turbine Driven Auxiliary Feed Pump Room Heat-Up
10. Indian Point Energy Center, CR-IP3-2018-01279 CA-1, OAP-008 "Severe Weather Preparations".

11 . U.S. Nuclear Regulatory Commission, letter from Eric J. Leeds, Director, Office of Nuclear Reactor Regulation and Michael R. Johnson, Director, Office of New Reactors , to All Power Reactor Licensees and Holders of Construction Permits in Active or Deferred Status, "Request for Information Pursuant to Title 1 0 of the Code of Federal Regulations 50 .54(f) Regarding the Recommendations 2. 1, 2.3 , and 9.3, of the Near. Term Task Force Review of Insights from the Fukushima Dai-ichi Accident," March 12, 2012, ADAMS Accession No. ML12053A340.

12. Indian Point Energy Center, "Entergy's Required Response for NTTF Recommendation 2.1: Flooding - Hazard Reevaluation Report. Indian Point Unit Numbers 2 and 3. Docket Nos. 50-247 and 50-286 . License Nos. DPR-26 and 64," December 23 , 2013 , ADAMS Accession No. ML13364A006. [Original FHRR submittal.]
13. Indian Point Energy Center, "Entergy Fleet Fukushima Program Flood Hazard Reevaluation Report for Indian Point Energy Center (IPEC) Units 2 and 3. Docket Nos.

50-247 and 50-286," Document No.: 51-9195289-002 , December 9, 2014, ADAMS Accession No. ML14357A052 [ADAMS package containing revision to original FHRR submittal. The FHRR was prepared by AREVA and dated May 2, 2014) . The report is indexed in ADAMS in three parts : ADAMS Accession Nos. ML14356A634, ML14356A635, and ML14356A636. The cover/transmittal letter is ML14356A633.]

14. Indian Point Energy Center, "Revised FL0-2D Analysis to Address the Current LIP Regarding the Flooding Aspects of Recommendations 2. 1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident. Indian Point Unit Numbers 2 and 3. Docket Nos. 50-247 and 50-286. License Nos. DPR-26 and 64 ," August 18, 2014, ADAMS Accession No. ML16116A060.

Revision 1 Page 18

021082-RPT-02

15. Indian Point Energy Center, "Entergy Submittal of Revision 1 to 'Flooding Hazard Re-evaluation - Combined Effects Floods-Coastal Processes' in Response to NRC Request for Information Pursuant to 10 CFR 50.54(f) Regarding the Flooding Aspects of Recommendations 2.3 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident. Indian Point Unit Numbers 2 and 3. Docket Nos. 50-247 and 50-286. License Nos. DPR-26 and 64," December 10, 2015 , ADAMS Accession No .

ML15351A068, non-public. The cover/transmittal letter is at ADAMS Accession No . ML15351A071 , non-public.

16. Indian Point Energy Center, "Entergy Supplement to Basis for Performance of the Mitigating Strategies Assessment with the Flood Hazard Information and Report for Recommendations 2.1 of the Near-Term Task Force Review of Insights from the Fukushima Dai-ichi Accident," April 5, 2016, ADAMS Accession No. ML16104A041.
17. Indian Point Energy Center, "Mitigation Strategies Assessment (MSA) for Flooding Submittal for Indian Point Units 2 and 3 (CAC Nos . MF-3313 and MF-3314) Docket Nos.

50-247 and 50-286, License Nos. DPR-24 and DPR-64," October 27, 2016 , ADAMS Accession No. ML16305A331.

18. U.S . Nuclear Regulatory Commission, "Ind ian Point Nuclear Generating Unit Nos. 2 And 3 - Flood Hazard Mitigation Strategies Assessment (CAC Nos. MF7935 and MF7936) ,"

April 10, 2017 (ADAMS Accession No. ML17059C227).

19. Nuclear Energy Institute, External Flooding Assessment Guidelines (Rev. 1) (NEI 16-
05) , June, 2016.
20. U.S. Nuclear Regulatory Commission, "Indian Point Nuclear Generating Unit Nos. 2 And 3 - NRC Response to Request for Deferral of Actions Related to Beyond-Design-Basis Seismic and Flooding Hazard Reevaluations", October 4, 2017 (ADAMS Access No. ML17222A239).
21. U.S. Nuclear Regulatory Commission , "Mitigating Strategies and Flood Hazard Reevaluation Action Plan," Commission Paper COMSECY-15-0019, June 30, 2015, ADAMS Accession Nos. ML15153A104 (Package, two documents, ML15153A105, "Closure Plan for the Reevaluation of Flooding Hazards for Operating Nuclear Power Plants" [cover letter] and ML15153A110, "Enclosure 1 - Mitigating Strategies and Flooding Hazard Re-Evaluation Action Plan".*

22 . Indian Point Energy Center, "Request for Deferral of Actions Related to Beyond -Design-Basis External Events Flooding Actions - Commitment Changes Docket Nos. 50-247 and 50-286, License Nos. DPR-26 and DPR-64," July 24, 2017, ADAMS Accession No. ML17209A740. 23 . U.S. Nuclear Regulatory Commission, letter from Eric J. Leeds , Director, Office of Nuclear Reactor Regulation and Michael R. Johnson, Director, Office of New Reactors , to All Power Reactor Licensees and Holders of Construction Permits in Active or Deferred Status, "Issuance of Order to Modify Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events ," Order EA-12-049, March 12, 2012 , ADAMS Accession No. ML12054A736.

24. U.S. Nuclear Regulatory Commission, letter from Eric J. Leeds, Director, Office of Nuclear Reactor Regulation and Michael R. Johnson, Director, Office of New Reactors, to All Power Reactor Licensees and Holders of Construction Permits in Active or Deferred Status, "Issuance of Order to Modify Licenses with Regard to Reliable Spent Revision 1 Page 19

021082-RPT-02 ReferPnces Fuel Pool Instrumentation," Order EA-12-051, March 12, 2012, ADAMS Accession No. ML12056A044.

25. Indian Point Energy Center, IP-RPT-08-00032, "Consequences of Fire and Explosion Following the Release of Natural Gas from Pipelines Adjacent to Indian Point," David Allen , Risk Research Group, 2008.

26 . Indian Point Energy Center, CR-IP3-2015-5090, "Vulnerability of the IP3 EDGs to High Air Intake Temperatures/ Damper Damage ," 2015.

27. Indian Point Energy Center, IP-RPT-15-00048, "Consequences of a Postulated Fire or Explosion Following the Release of Natural Gas from the Existing 26" and 30" Pipelines Near IPEC," David Allen , Risk Research Group, October, 2015.
28. Indian Point Energy Center, Entergy letter NL-14-106, 10 C.F.R. 50.59 Safety Evaluation and Supporting Analyses Prepared in Response to the Algonquin Incremental Market Natural Gas Project Indian Point Nuclear Generating Unit Nos. 2 & 3 (ML14253A339) ..
29. U.S . Nuclear Regulatory Commission , Letter to M Kansler regarding "Review of Individual Plant Examination of External Events (TACNO. M83632),"' February 15, 2001.
30. Indian Point Energy Center, IP-RPT-15-00031 (Rev. 0) , FLEX Strategy Tornado Protection-lP2, May 3, 2016.
31. Indian Point Energy Center, IP3-RPT-CRHV-03379 (Rev. 0) , "Units 2 and 3, Control Room Habitability Evaluation, Final Report," The Risk Research Group, February 20 ,

2003.

32. Indian Point Energy Center, IP3-RPT-CRHV-03379 (Rev. 1), "Units 2 and 3, Control Room Habitability Evaluation , Final Report," The Risk Research Group, December 15, 2008.
33. Indian Point Energy Center, IPEC-RPT-CRH-2013, "Units 2 and 3, Control Room Habitability Evaluation , Final Report ," The Risk Research Group, April 17, 2014.

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