SBK-L-16162, Response to Request for Additional Information Regarding License Amendment Request 15-02, Adoption of Emergency Action Level Schemes Pursuant to NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactors

Response to Request for Additional Information Regarding License Amendment Request 15-02, Adoption of Emergency Action Level Schemes Pursuant to NEI 99-01, Revision 6, Development of Emergency Action Levels for Non-Passive Reactors

ML16302A414
Person / Time
Site:	Seabrook
Issue date:	10/27/2016
From:	McCartney E NextEra Energy Seabrook
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
SBK-L-16162
Download: ML16302A414 (66)

Text

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Seabrook Station NEXTera ROOK October 27, 2016 10 CFR 50.90 SBK-L-16162 Docket No. 50-443 Response to Request for Additional Information Regarding License Amendment Request 15-02, Adoption of Emergency Action Level Schemes Pursuant to NEI 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors"

References:

1. NextEra Energy Seabrook, LLC letter SBK-L-15120, "License Amendment Request 15-02, Adoption of Emergency Action Level Schemes Pursuant to NEI 99-01, Revision 6 'Development of Emergency Action Levels for Non-Passive Reactors"'

February 27, 2016(ML16068A128)

2. NRC letter "Seabrook Station, Unit No. 1 -Request for Additional Information Related to License Amendment Request to Adopt Emergency Action Level Schemes Pursuant to NEI 99-01, Revision 6 (CAC MF7439)," September 22, 2016(ML16230A533)

In Reference 1, NextEra Energy Seabrook, LLC (NextEra) submitted a license amendment request (LAR) to revise the current EAL scheme to one based upon the Nuclear Energy Institute (NEI) document NEI 99-01, Revision 6, "Development of Emergency Action Levels for Passive Reactors".

In Reference 2, the NRC staff determined that additional information is necessary to support the staff's continued technical review of the proposed EAL, scheme change. The enclosures to this letter provide the requested additional information.

The enclosed mark ups and clean pages supersede the corresponding pages in Reference

1. The changes to the LAR provided in this letter do not alter the conclusion in Reference 1 that the change does not involve a significant hazards consideration pursuant to 10 CFR 50.92, and there are no significant environmental impacts associated with this change. No new or revised commitments are included in this letter. NextEra Energy Seabrook, LLC, P.O. Box 300, Lafayette Road, Seabrook, NH 03874 U.S. Nuclear Regulatory Commission SBK-L-16162 I Page 2 Should you have any questions regarding this letter, please contact Mr. Kenneth Browne, Licensing Manager, at (603) 773-7932.

I declare under penalty of perjury that the foregoing is true and correct. Executed on October __n__, 2016. Sincerely, Eric McCartney Site Vice President NextEra Energy Seabrook, LLC

Enclosures:

Enclosure 1 Response to Request for Additional Information Regardinglicense Amendment Request 15-02, Adoption of Emergency Action Level Schemes Pursuant to NEI 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors" Enclosure 2 Markup of Affected Seabrook Station Emergency Action Levels -Initiating Conditions, Threshold Values and Basis Enclosure 3 Clean Copy of Seabrook Station Emergency Action Levels -Initiating Conditions, Threshold Values and Basis Enclosure 4 NEI 99-01, Rev. 6, Deviations and Differences, Seabrook Station Nuclear Power Plant -Unit 1 cc: NRC Region I Administrator NRC Project Manager NRC Senior Resident Inspector Director Homeland Security and Emergency Management New Hampshire Department of Safety Division of Homeland Security and Emergency Management Bureau of Emergency Management 33 Hazen Drive Concord, NH 03305 Mr. John Giarrusso, Jr., Nuclear Preparedness Manager The Commonwealth of Massachusetts Emergency Management Agency 400 Worcester Road Framingham, MA 01702-5399 Enclosure 1 to SBK-L-16162 Response to Request for Additional Information Regarding License Amendment Request 15-02, Adoption of Emergency Action Level Schemes Pursuant to NEI 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors" Enclosure 1 to SBK-L-16162 Page 2of12 Background By letter dated February 27, 2016 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 16068A128), NextEra Energy Seabrook, LLC (NextEra) submitted a license amendment request (LAR) to adopt the emergency action level schemes pursuant to Nuclear Energy Institute (NEI) 99-01, Revision 6, "Development of Emergency Action Levels for Non-Passive Reactors," at Seabrook Station, Unit No. 1. The NRC staff has determined that additional information provided below is necessary to complete the review. RAI-Seabrook-1 Section 2.7, "Classification of Short-Lived Events," does not contain the guidance provided in Section S.7 of NEI 99-01, Revision 6, which states, in part: If an event occurs that meets or exceeds an EAL, the associated ECL (emergency classification level) must-be declared regardless of its continued presence at the time of declaration.

Please explain why this key guidance from NEI 99-01, Revision 6, was omitted, or revise accordingly.

NextEra Response Section 2. 7 of the Seabrook Station technical basis is revised to incorporate the referenced guidance statement from NEI 99-01, Revision 6, Section S.7. RAI-Seabrook-2 The technical basis discussion for RA3 [AA3] in NEI 99-01, Revision 6, states: This IC (initiation condition) addresses elevated radiation levels in certain plant rooms/areas sufficient to preclude or impede personnel from performing actions necessary to maintain normal plant operation, or to perform a normal plant cooldown and shutdown.

The technical basis discussion for HAS [HAS] in NEI 99-01, Revision 6, states: This IC addresses an event involving a release of a hazardous gas that precludes or impedes access to equipment necessary to maintain normal plant operation, or required for a normal plant cooldown and shutdown.

Enclosure 1 to SBK-L-16162 Page 3of12 The proposed Table H1 includes "Equipment Vaults" as a plant room/area that require access to operate equipment as noted above. It is not clear to the NRG staff what required equipment is contained within the "Equipment Vaults," or if there are additional rooms/areas that are identified as "Equipment Vaults" that do not contain equipment, but require access to perform actions (e.g. operate equipment) necessary to maintain normal plant operation, or to perform a normal plant cooldown and shutdown.

For EAL RA3 [AA3] and HA5 [HA5], please address the following:

a. Please clarify what required equipment is contained in the "Equipment Vaults" identified in Table H1. Additionally, please provide justification for using the potentially vague room/area designation of "Equipment Vaults" as this designation could potentially impact a timely and accurate classification, or revise accordingly.

b. Table H1 indicates that access to the containment is required in Operating Modes 3 and 4. Please explain why access is required to the containment building for Mode 3 and 4, operations or revise accordingly.

This explanation should include: (1) a listing of the specific areas of the containment for which access is required in Operating Modes 3 and 4, and (2) what procedural requirements necessitate access for performing actions necessary to maintain normal plant operation or to perform normal plant cooldown and shutdown.

c. Table H1 indicates that access to the entire turbine building is required for Operating Modes 1, 2 and 3. Please explain why access is required to the entire turbine building for Operating Modes 1, 2 and 3 operations, or revise accordingly.

This explanation should include: (1) a listing of the specific areas of the turbine building for which access is required in Operating Modes 1, 2 and 3, and (2) what procedural requirements necessitate access for performing actions necessary to maintain normal plant operation or to perform normal plant cooldown and shutdown.

NextEra Response a. The term "Equipment Vaults" refers to the Residual Heat Removal (RHR)/Containment Building Spray (CBS) equipment vaults which are attached to the Primary Auxiliary Building.

Operation of RHR is required for transition from modes 3 to 4 and modes 4 to 5 respectively per Operations Procedure OS1000.04, Plant Cooldown from Hot Standby to Cold Shutdown.

Table H1 in EALs RA3 and HA5 is revised to replace the term "Equipment Vaults" with "RHR/CBS Equipment Vaults". b. Access is required to Containment levels 0 to -26 in modes 3 and 4 to put RHR in service per Operations Procedures OS1013.03 and OS1013.04, RHR Train A and RHR Train B Startup and Operation.

Prior to placing RHR in service for a scheduled Enclosure 1 to SBK-L-16162 Page 4of12 plant shutdown, ultrasonic testing (UT) at RHR piping sample points to verify water solid conditions is required to be conducted per surveillance procedure OX1456.02, ECCS Monthly System Verification.

Prerequisite 2.1.18 of OS1013.03 states, "If "A" RHR train is being placed in service as part of a scheduled plant shutdown, the "A" RHR lines have been verified water solid by performing ultrasonic testing on point RH-3 per PM ECCS-UT-PIPING.

Consideration should be given to test "B" train, point RH-5, at the same time." UT requires access to the loop 4 entry at level -26' to access the ladder to reach test point RH 5 at the -10' level downstream of RC-V-87 (RHR train B suction isolation valve). UT also requires access to the loop 1 entry at level -26' to access the ladder to reach test point RH 3 at the -1 O' level downstream of RC-V-22 (RHR train A suction isolation valve). c. Access is required to Turbine Building 21' and 50' levels in modes 1, 2 and 3 for alignment of feedwater with the Startup Feed Pump (SUFP) per Operations Procedures OS1000.03, Plant Shutdown from Minimum Load to Hot Standby, and OS1000.04, Plant Cooldown from Hot Standby to Cold Shutdown.

Procedure OS1035.02, Startup Feed Pump Operations, requires access to the SUFP area on the 21' level of the Turbine Building North when the condensate (CO) cleanup system is not in service. When placing CO cleaning in service per procedure ON1034.09, Condensate Cleanup System Operation, and realigning SUFP suction to CO cleaning, access is required to the 21' level of the Turbine Building East along the condensers and to the 50' level of the Turbine Building northeast area near the CO cleanup system filters. The use of the CO cleanup system is preferred for conservation of water inventory if the main condenser steam dumps are available for use, i.e., the main steam isolation valves are not closed and condenser vacuum is maintained.

Per OS1021.01, Steam Generator Slowdown System Operation, realignment of the flash tank vapor from FW-E-23C to the main condenser requires access to the 50' level of the Turbine Building.

Table H1 in EALs HA5 and RA3 is revised to specify the 21 ft elevation and the 50 ft elevation of the Turbine Building as areas requiring access in Operating Modes 1, 2 and 3. Further review of Table H1 resulted in identification of additional areas that are not required to access equipment as required by initiating conditions RA3 and HA5. These areas are non-essential switchgear room, steam and feedwater pipe chases, and -31 ft elevation of the Waste Process Building.

Accordingly, Table H1 is revised to delete these areas.

• RAI-Seabrook-3 For RU1 [AU1], EAL 1, the assessment criteria is based on one of the listed radiation monitors being greater than 2 times the offsite dose calculation manual (ODCM) limits. In addition to providing a list of site-specific monitors, the developer's guidance in NEI 99-01, Revision 6, states: Radiation monitor readings should reflect values that correspond to a radiological release exceeding 2 times a release control limit.

Enclosure 1 to SBK-L-16162 Page 5of12 Please explain how an assessment of this EAL can be performed in a timely and accurate manner, without including instrument values that represent 2 times the ODCM limits, or revise accordingly.

NextEra Response The ODCM values are conservatively based on an isotopic mix that will vary over time and may change depending on plant operations (e.g., a dry fuel storage campaign).

The alarms for the radiation monitors listed in EAL RU1, with the exception of the WRGM alarm, are set at levels that are well below the ODCM value. The WRGM alarm is set at the ODCM value. If a radiation monitor alarm comes in, the operating staff will enter the applicable AOP (OS1252.01, Process or Effluent High Radiation, or OS1252.02, Airborne High Radiation) and will direct the on-shift Chemistry Technician to validate the alarm using Chemistry procedure CS0905.10, Chemistry Response to ROMS or Waste Gas Oxygen Monitor Failure or Alarm. CS0905.10 contains instructions for the on-shift Chemistry Technician for obtaining and analyzing applicable samples in the event any of the monitors identified in RU1 is in alert or high alarm. Values that correspond to 2X the ODCM limit are identified in CS0905.10.

CS0905.10 directs the on-shift Chemistry Technician to notify the Shift Manager as soon as possible that the EAL may apply. The applicable AOP also directs the operating staff to evaluate the EALs if limits are exceeded as reported by the on-shift chemistry technician.

Therefore a monitor alarm provides ample time for the on-shift operating staff and chemistry personnel to verify a monitor level and compare it to the ODCM value. Procedures and processes are in place to validate a radiation monitor alarm and determine if the alarm value meets the EAL threshold in a timely manner. RAl-Seabrook-4 Please provide justification for not including power supply tables for EALs MA 1 [SA 1 ], MU1 [SU1] and CU2 [CU2], based on NRC staff resolution provided in Emergency Preparedness Frequently Asked Question (EPFAQ) No. 2015-15 (ADAMS Accession No. ML 16166A191), or revise accordingly.

NextEra Response Per EPFAQ No. 2015-15, a table of AC power sources that is included as a note in the current EAL SA5 will be added to NEI 99-01 Revision 6 EALs MA1and CU2. EPFAC No. 2015-15 says that a table of power sources is expected for MA1 [SA1] and CU2 only. MU1 concerns loss of offsite AC power sources only and therefore does not require the power supply table.

Enclosure 1 to SBK-L-16162 Page 6of12 SA5 Note: NOTE There are six power sources to consider:

RAl-Seabrook-5

• 345 kV offsite power Line 369

• 345 kV offsite power Line 363

• 345 kV offsite power Line 394

• Emergency Diesel Generator A

• Emergency Diesel Generator B

• SEPS. For SEPS to be considered available, both SEPS diesel generator sets must be functional.

For EALs MG8 [SG8], MG1 [SG1], MS1 [SS1] and CA2 [CA2], please provide justification for including a discussion related to a specific power source that could compel a decision-maker to make a declaration, even though mitigation strategies are effective, or revise accordingly.

NextEra Response EALs MG8, MG1, MS1, MA1 and CA2 include a discussion of the Supplemental Emergency Power System (SEPS) that can supply power to emergency buses E5 or E6 in the event of the loss of offsite power and the failure of both emergency diesel generators to start and load. Each of the referenced EALs contains a note that says "For a bus to be considered energized from SEPS, both SEPS diesel generator sets must be functional." The basis section for each of the EALs contains the statement "For power restoration from the SEPS, both SEPS diesel generator sets must be functional." SEPS Loading Calculation 9763-3-ED-00-02-F provides the basis for requiring both SEPS diesel generator engines to be functional in order to be credited for supplying power to an emergency bus. The calculation shows that the required load is greater than the capacity of one SEPS generator engine (2640 KW.) This calculation does not take into account the starting current required by various equipment.

RAl-Seabrook-6 For EALs CU5 [CU5] and MU6 [MU6], please address the following:

a. Criteria lists "all plant telephones" as an acceptable communication method. This could imply that an EAL would not have to be declared as long as there was at least Enclosure 1 to SBK-L-16162 Page 7of12 one functioning telephone on site. Please provide justification that supports the use of "all plant telephones," which addresses how this condition could be assessed in a timely and accurate manner. b. Criteria lists cellular telephones as an acceptable method of communication for offsite communications.

As stated in NEI 99-01, Revision 6, communication methods with the offsite response organizations and the NRG should be " ... described in the site Emergency Plan." Section 7 of the Seabrook Site Emergency Plan, which describes communication methods, does not include cellular phones. Please provide justification for listing cellular phones as a method of communication, or revise accordingly.

NextEra Response a. The term "All plant telephones" in EALs 2 and 3 of CU5 and MU6 is replaced by "Control Room/TSC Telephones".

b. Cellular telephones is deleted from the list of communications methods in EALs 2 and 3 of CU5 and MU6. RAl-Seabrook-7 For the fuel clad and reactor coolant system (RCS) fission product barriers, RED entry conditions for the heat sink critical safety function (CSF) are used as a threshold for a potential loss of the barrier. However, the NEI 99-01, Revision 6, guidance states: In accordance with EOPs (emergency operating plans), there may be unusual accident conditions during which operators intentionally reduce the heat removal capability of the steam generators; during these conditions, classification using threshold is not warranted.

This guidance is included in the barrier threshold basis discussions; however, it is not included in the relevant barrier thresholds.

Please explain why the NEI 99-01, Revision 6, guidance concerning making classifications for heat sink conditions when operators intentionally reduce heat removal capability, in accordance with EOPs, is not included in the fission product barrier thresholds, or revise accordingly.

NextEra Response Current Seabrook Station Emergency Response Procedure ER 1.1, Classification of Emergencies, contains a discussion of the proper use of critical safety function status trees (CSFSTs) for emergency classification (i.e., non-green CSFST must represent a true challenge to the CSF for emergency classification purposes).

This discussion is Enclosure 1 to SBK-L-16162 Page 8of12 retained in the revised ER 1.1 procedure that incorporates the NEI 99-01 Revision 6 EALs. A note is added to the Fission Product Barrier Table for EALs FG1, FS1 and FA1 that refers to the discussion of proper use of CSFSTs for emergency classifications.

RAl-Seabrook-8 Concerning EAL HG1 [HG1], NRC staff resolution to EPFAQ 2015-13 (ADAMS Accession No. ML 16166A366) was recently approved, which provides guidance that could be used, if deemed appropriate, to meet the intent of HG1 [HG1]. Please consider EPFAQ 2015-13 and revise EAL HG1 [HG1] if deemed appropriate, to reflect latest staff clarification of NEI 99-01, Revision 6 guidelines.

NextEra Response Per EPFAQ 2015-13, EAL HG1 is deleted. Because this is a deviation from NEI 99-01, Revision 6, the table of deviations and differences is revised to include the justification for deletion of EAL HG1. RAl-Seabrook-9 EAL HU4 [HU4] (2) in NEI 99-01, Revision 6, states: Receipt of a single fire alarm (i.e., no other indications of a FIRE). The NEI 99-01, Revision 6, technical basis for HU4 [HU4] (2) further states: A single fire alarm, absent other indication(s) of a FIRE, may be indicative of equipment failure or a spurious activation, and not an actual FIRE. For this reason, additional time is allowed to verify the validity of the alarm. The 30-minute period is a reasonable amount of time to determine if an actual FIRE exists; however, after that time, and absent information to the contrary, it is assumed that an actual FIRE is in progress.

The proposed HU4 [HU4] (2) includes an exception for the containment based on the following note: A containment fire alarm is considered valid upon receipt of an actuated alarm on CP-376, combined with any of the following:

• CP 376 panel -Multiple Zones Actuated

• Plant Equipment

-Spuriously Operating

• Containment Temperature

-Increasing

• Containment Particulate Radiation

-Increasing Please provide further justification for the apparent deviation from the NRG-endorsed guidance provided by NEI 99-01, Revision 6, for the receipt of a single fire alarm. For Enclosure 1 to SBK-L-16162 Page 9of12 example, this could potentially cause confusion with a declaration under EALs MA9 [SA9] and CA6 [CA6], where a containment fire causes spurious operation of equipment, e.g., is a rise in containment temperature or spurious operation of equipment to be considered as indications of degraded performance per MA9 [SA9] and CA6 [CA6]? NextEra Response Next Era proposes to make an exception in EAL HU4 (2) to exclude containment in modes 1 and 2 because accessing containment within 30 minutes to verify the status of a single alarm is a challenge, particularly in modes 1 and 2 when containment integrity is set and personnel safety concerns would preclude entry into certain areas of containment.

There are areas within containment where fire detectors are located that would be inaccessible during these modes due to elevated radiation levels. Based on prior experience, if containment were to be included in EAL HU4 (2) during modes 1 and 2, the potential would exist for an inordinate unneeded number of Notification of Unusual Event emergency classifications and subsequent retractions.

Seabrook Station's containment building contains 137 individual Pyrotronics detectors distributed over 9 zones. The first 4 zones provide detection for the O' elevation; the other 5 zones provide detection for the -26' elevation.

When a detector alarms, the zone alarm for the zone in which the detector is located will actuate on fire panel FP-CP-376.

137 individual fire detectors are an unusually large number that significantly increases the potential of a spurious alarm. The 137 fire detectors in the Seabrook Station containment building is approximately 4.5 times the average number of containment building fire detectors in other NextEra nuclear power plants. Actuation of more than one zone on FP-CP-376 is the most reliable indication of a valid fire detector alarm because of the volume of air flow throughout the containment building.

Due to construction of the intermediate floors and multiple openings in the floors it can be expected that smoke would migrate throughout containment.

There are six Containment Air Handling (CAH) cooling units located on the O' elevation of the containment building.

Five of the CAH cooling units are normally operating at any given time to cool the containment.

Each cooling unit discharges approximately 56,000 CFM into the common air distribution system. The units draw return air into each end of the unit. This constant flow of air (approximately 280,000 CFM) would draw any smoke towards the cooling units past the installed detectors thus affecting multiple zones. More than one zone actuated on FP-CP-376 is therefore the most reliable indication of a valid alarm and accurately meets the criteria of EAL HU4 (1). Verification of a single containment fire alarm that is likely to be spurious does not warrant the potential elevated exposure risks associated with an emergency entry of containment in modes 1 and 2. Therefore, Seabrook Station proposes to make EAL HU4(2) applicable to a single fire alarm in containment in Modes 3, 4, 5 and 6. , The note containing criteria for a valid containment fire alarm that would be applicable to HU4(1) during modes 1 and 2 is revised to read "A containment fire alarm is considered Enclosure 1 to SBK-L-16162 Page 10of12 valid upon receipt of multiple zones (more than 1) actuated on CP-376 panel." The alternate indications of spurious equipment actuation, increasing containment temperature, and increasing particulate radiation in containment are removed from the note to preclude potential confusion with the degraded safety equipment EALs CA6 and MA9. RAl-Seabrook-10 EAL HU4 [HU4] (4) in NEI 99-01, Revision 6, states: A FIRE within the plant or ISFSI (for plants with an ISFSI outside the plant Protected Area) PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish.

The proposed EAL HU4 [HU4] (4) does not include the independent spent fuel storage installation (ISFSI) (referred to as dry fuel storage facilityr Please explain why the dry fuel storage facility was not included for fires that require an offsite fire response to extinguish, or revise accordingly.

NextEra Response EAL HU4(4) is revised to identify the Dry Fuel Storage Facility.

RAl-Seabrook-11 For EALs MU5 [SU5], MA5 [SA5], and MS5 [SS5], a power level (<5%) was added to the EALs. The intent of NEI 99-01, Revision 6, is align the above EAL classifications with site-specific EOP criteria of a successful reactor shutdown, as the consistency between EALs and EOPs would benefit the decision-makers by providing consistent criteria.

The power level provided in the NEI 99-01, Revision 6, developer notes is an example that represents a typical EOP indication for a generic power plant. Please consider either using either the same EOP reactor shutdown criteria that the operators use in either the EOPs or operator training, or consider using wording similar to the guidance in NEI 99-01, Revision 6. NextEra Response The reference to "neutron flux <5%" is removed from EALs MU5, MA5 and MS5. The wording in NEI 99-01, Revision 6, for SU5, SA5 and SS5 is used for these EALs instead.

Enclosure 1 to SBK-L-16162 Page 11 of 12 RAl-Seabrook-12 For EAL MS5 [SS5], the second paragraph in the technical basis includes a discussion that classifications from MS5 [SS5] may be at a higher level than what would be determined by the fission product barrier recognition category.

Although this may be true for some licensees, the Seabrook fission product barrier recognition category for either core cooling or heat sink CSF red entry conditions met would result in a site area emergency based solely on the fission product barrier recognition category.

Please provide an explanation for including a discussion that does not appear to be specific to Seabrook, or revise accordingly.

NextEra Response The second paragraph of the technical basis for EAL MS5 is deleted. RAl-Seabrook-13 EAL MA1 [SA1] (1) in NEI 99-01, Revision 6, states: a. AC (alternating current) power capability to (site-specific emergency buses) is reduced to a single power source for 15 minutes or longer. AND b. Any additional single power source failure will result in a loss of all AC power to SAFETY SYSTEMS. For EAL MA 1 [SA 1 ], the condition that any additional single power source will result in a loss of all AC power to SAFETY SYSTEMS was removed from the proposed EALs as being redundant to the condition that AC power capability to both AC emergency buses E5 and E6 is reduced to a single power source for 15 minutes or longer. Although the conditions provided by NEI 99-01, Revision 6, both include the term power source, they are not redundant.

Please explain, in greater detail, why the condition, "Any additional single power source failure will result in a loss of all AC power to SAFETY SYSTEMS," was removed from the proposed EAL MA 1 [SA 1 ], or revise accordingly.

NextEra Response EAL MA 1 is revised to add the condition, "Any additional single power source failure will result in a loss of AC power to SAFETY SYSTEMS."

Enclosure 1 to SBK-L-16162 Page 12 of 12 RAl-Seabrook-14

• For EAL MA2 [SA2], please address the following:

a. As proposed, all core exit temperatures and all but one RCS temperatures would not require a classification.

Depending on the nature of the transient, an RCS temperature indication may or may not provide an accurate assessment of core conditions.

Please justify, including RCS temperature as an alternative to core exit temperatures or revise accordingly.

b. The Seabrook core cooling critical safety function status tree (CSFST) specifically uses reactor vessel level indication system (RVLIS) to assess the Core Cooling CSFST. However, the proposed EAL MA2 [SA2] uses pressurizer level. Depending on the nature of the transient, pressurizer level indication may or may not provide an accurate assessment of core conditions.

Please provide justification for not using RVLIS to determine RCS level for EAL MA2 [SA2], or revise accordingly.

NextEra Response a. EAL MA2 (1) a. is revised to delete RCS Temperature and utilize Core Exit Temperature as the indicated parameter.

b. EAL MA2 (1) a. is revised to delete Pressurizer Level and utilize RCS Level as the indicated parameter.

Enclosure 2 to SBK-L-16162 Markup of Affected Seabrook Station Emergency Action Levels -Initiating Conditions , Threshold Values and Basis 2.7 CLASSIFICATION OF SHORT-LlVED EVENTS Event-based ICs and EALs define a variety of specific occurrences that have potential or actual safety significance.

By their nature, some of these events ma y be short-lived and , thus , over before the emergency classification assessment can be completed.

If an event occurs that meets or exceeds an EAL, the associated ECL must be declared regardless of its continued presence at the time of declaration.

Examples of such events include a failure of the reactor protection s y stem to automatically scram/trip the reactor follo w ed by a successful manual scram/trip or an earthquake.

HAS ECL: Alert Initiating Condition:

Gaseous release impeding access to equipment necessary for normal plant operations , shutdown or cooldown.

Operating Mode Applicability:

All Emergency Action Levels: Note: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred , then no emergency classification is warranted. (1) Basis: a. AND Release of a toxic , corrosive , asphyxiant or flammable gas into any Table H 1. rooms or areas. b. Entry into the room or area is prohibited or IMPEDED. Table Hl Area Mode Primary Aux Building 25 ft elevation 1 , 2 , 3, 4 7 ft elevation

-26 ft elevation Turbine Building 21 ft elevation 1 , 2 , 3 50 ft elevation Switchgear Rooms Essential 1, 2 , 3, 4 NeA esseAtial Steam aAEl FeeewateF Pifle el=rnses +,-¥ Waste Process Building 25 ft elevation 1 , 2 , 3 -3 ft elevation 31 ft ele'f*atieA Containment 3, 4 RHR/CBS Equipment Vaults 3 , 4 IMPEDE: Entry into an area requires extraordinary measures to facilitate entry of personnel into the affected room/area by installing temporary shielding , requiring use of non-routine protective equipment, or requesting an extension in dose limits beyond normal administrative limits.

RA3 ECL: Alert Initiating Condition:

Radiation levels that IMPEDE acce s s to equipment necessary for normal plant operations , shutdown or cooldown.

Operating Mode Applicability:

All Emergency Action Levels: (1 or 2) Note: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted. (I) Dose rate greater than 15 mR/hr in ANY of the fo ll owing areas: Contro l Room RM6550 Central A l arm Station (CAS) by survey Secondary Alarm Station (SAS) by survey OR (2) An UNPLANNED event results in radiation levels that prohibit or IMPEDE access to any of the following plant rooms or areas: Table Hl Area Mode Primary Aux Building 25 ft elevation 1 , 2 , 3 , 4 7 ft elevation

-26 ft elevation Turbine Building 21 ft elevation 1, 2, 3 50 ft elevation Switchgear Rooms Essential 1, 2 , 3 , 4 l>foA esseAtial Steam aAe Feee*wateF Pipe ehases ;-¥ Waste Process Building 25 ft elevation 1 , 2 , 3 -3 ft elevation 31 ft: ele'f*atieA Containment 3 , 4 RHR/CBS Equipment Vaults 3 , 4 MA1 ECL: Alert Initiating Condition:

Loss of all but one AC power source to emergency buses for 15 minutes or longer. Operating Mode Applicability

1 , 2 , 3, 4 Emergency Action Levels: Notes:

• The STED/SED should declare the Alert promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.

• For a bus to be considered energized from SEPS, both SEPS diesel generator sets must be functional.

(1) a. AC power capability to BOTH AC emergency buses ES AND E6 is reduced to a single power source for 15 minutes or longer. Basis: AND b. An y additional single po w er source failure will result in loss of a ll AC power to SAFETY SYSTEMS. NOTE There are six power sources to consider:

• 345 kV offsite power Line 369

• 345 kV offsite power Line 363

• 345 kV offsite power Line 394

• Emergency Diesel Generator A

• Emergency Diesel Generator B

• SEPS. For SEPS to be considered a v ailable, both SEPS diesel generator sets must be functional.

SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, inc l uding the ECCS. Systems classified as safetyrelated. This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss ofall AC power to SAFETY SYSTEMS. In this CU2 ECL: Notification of Unusual Event Initiating Condition:

Loss of all but one AC power source to emergency buses for 15 minutes or longer. Operating Mode Applicability:

5, 6, Defueled Emergency Action Levels: Notes: (1)

• The STED/SED shou ld declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.

• For power restoration from the SEPS, both SEPS diesel generator sets must be functional.

a. AC power capability to Both AC emergency buses ES AND E6 is reduced to a s in gle power source for 15 minutes or l onger. AND b. Any additional s in g l e power source failure will result in l oss of all AC power to SAFETY SYSTEMS. NOTE There are six power sources to consider:

• 345 kV offsite power Line 369

• 345 kV offsite power Line 363

• 345 kV offsite power Line 394

• Emergency Diesel Generator A

• Emergency Diesel Generator B

• SEPS. For SEPS to be considered available, both SEPS diesel generator sets must be functional.

Basis: SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition , including the ECCS. Systems classified as related. This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this cond iti on, the so le AC p ower so urce may be powering one, or more than one, train of safetyrelated equipment.

When in the cold shutdown , refueling , or defueled mode , this condition i s not classified as an Alert because of the increa se d time available to restore another power source to service. Additional time is available due to the reduced core decay heat l oad , and the lower temperatures ECL: Notification of Unusual Event Initiating Condition:

Loss of all onsite or offsite communications capabilities.

Operat i ng Mode Applicability

5, 6 , Defueled Emergency Action Leve l s: (I or 2 or 3) (I) Loss of ALL of the following onsite communication methods: In-Plant (PBX) Telephones Gai-Tronics Plant Radio System OR (2) Loss of ALL of the following ORO communications methods: OR Nuclear Alert System (NAS) Backup NAS Al+Control RoomffSC plaffi telephones Cellular telephoAes (3) Loss of ALL of the following NRC communications methods: Basis: Emergency Notificat i on System (ENS) Al+Control RoomffSC plaffi te l ephones FTS te l ephones in the TSC Cellular telephoAes CU5 This IC addresses a significant loss of on-site or offsite communications capabilities.

While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to OROs and the NRC. This IC should be assessed only when extraordinary means are being utilized to make communications possible.

EAL #1 addresses a total loss of the communications methods used in support ofroutine plant operations. EAL #2 addresses a total l oss of the communications methods used to notify all OR Os of an emergency declaration.

The OROs referred to her e are Commonwealth of Massachusetts and State of New Hampshire. EAL #3 addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.

ECL: Notification of Unusual Event Initiating Condition:

Loss of all onsite or offs ite communications capabilities. Operating Mode Applicability:

1 , 2, 3, 4 Emergency Action Levels: (1 or 2 or 3) (1) Loss of ALL of the following onsite communication methods: In-Plant (PBX) Telephones Gai-Tronics Pl ant Radio System OR (2) Loss of ALL of the following ORO communications methods: OR Nuclear Alert System (NAS) Backup NAS A-lt Control RoomffSC plaflt telephones Cellular telephones (3) Loss of ALL of the following NRC communications methods: Basis: Emergency Notification System (ENS) A-lt Control RoomffSC plaflt telephones FTS telephones in the TSC Cellular telephones MU6 This IC addresses a significant lo ss of on-site or offsite communications capabilities.

While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to OROs and the NRC. This IC should be assessed only when extraordinary means are being utilized to make communications possible. EAL #1 addresses a total los s of the communications methods us ed in support of routine plant operations.

EAL #2 addresses a total loss of the communications methods used to notify all ORO s of an emergency declaration. The OROs referred to here are the Commonwealth of Massachusetts and State of New Hampshire.

EAL #3 addresses a total l oss of the communications methods used to notify the NRC of an emergency declaration.

1. Fission Product Barrier Table Thresholds for LOSS or POTENTIAL LOSS of Barrier s FGl GENERAL EMERGENCY FSl SITE AREA EMERGENCY FAlALERT L oss of a n y two barri e rs and Loss o r Potential Loss of the third barri e r. Fuel Clad Barrier LOSS POTENTIAL LOSS RCS or SG Tube L eakage 1. Loss or Pot e nti a l Loss of any two barri ers. Any Loss or a n y Pot e nti a l Loss of eit h e r the Fuel Clad or RCS bar r ier. RCS Barrier Containment Barrier LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS RCS or SG Tube Leakage 1. RCS or SG Tube Leakage ot App lic ab l e A. Co r e Coo lin g (C) A. An a u to m at i c or A. Operation of a second A. A l eaki ng o r Not A ppli cab l e CSF-ORANGE manual S I actuat i on is c h arg in g pump in the RUPTURED SO is e nt ry conditions met required b y EITHER n orma l c h a r g in g FAUL TE D outside of CNOTE I} of the fo ll owi n g: m ode is req uir e d by co n tain m e n t. 1. UN!SOLA BL E EITHER of the RCS l ea ka ge following:

OR 1. UNlSO L ABLE 2. SO tube RCS l eakage RUPTURE. OR 2. SO tube l eakage. OR B. RCS Inte g rity (P) CSF -RE D e nt ry conditio n s met w i th R CS pr ess > 300 psig. CNOTE I} 2. Inadequate Heat Remova l 2. Inadequate Heat Removal 2. Inadequate Heat Removal A. Core Cooling (C) A. Core Coo lin g (C) Not App li ca bl e A. H eat Sink (H) CSF -Not A pplic a bl e A. Core Coo lin g (C) CSF CSF -RED entry CSF-ORA GE RED entry conditions

-RED entry co nditi ons co ndit i ons met. entry conditions met. met. CNOTE l} m et for 15 minut es or CNOT E 1} CNOTE l} l o n ger. (NOTE 1} OR 8. Heat S ink (H) CSF -RED entry con di tio n s met. CNOTE ])

3. RCS Activity I Containment Radiation

3. RCS Activity I Containment Radiation

3. RCS Activity I Containment Radiation A. Post LOCA Not App licabl e A. Post LOCA Radiation Not Appl ic ab l e Not App licable A. Post LOCA Radi ation Rad i ation Monitors Monitors Monitors RM 6576A-l or RM RM 6576A-l or RM RM 6576A-l or RM 6576B-l 6576B-l 6576B-l 2'. 95 R/hr. 2'. 16 R/hr. 2'. 1 , 305 R/hr. . OR B. RCS activity> 300 uCi/g m Dose Equiva l ent! 131 as determined per Procedure CS0925.0 I , Reacto r Coo l ant Post Accident Sampling.

4. Containment Integrity or Bypas s 4. Containment Integrity or Bypass 4. Containment Integrity or Bypass Not A ppl i cable Not Ap plicabl e Not App li ca bl e Not App licable A. Contain ment i so lati on A. Co ntainm e nt (Z) CSF -is required RED entry cond iti ons AND met. (NOTE 1) EITHER of the OR fo llow ing: B. Containment H 2 l. Conta inm e nt concentration 2'. 6% integrity h as been OR l ost based on STED/SED C. I. Contain m ent judgment.

pre s sure > 1 8 psig OR AND 2. UNI SO LAB LE 2. Less than o n e full pathway from the train of contai nm e nt to Containme nt the enviro nment Building Spray exists. (CBS) is OR operating p er design for 15 B. Indications of R CS minutes or l onger. leakage outside of containment.

5. STED/SED Jud g ment 5. STED/SED Judgment 5. STED/SED Judgment A. ANY condition in A. ANY condition in the A. ANY condition in the A. ANY condition in the A. ANY condit ion in the A. ANY condition in the the opinion of the opi ni on of the opinion of the opinion of the opinion of the opinion of the ST E D/SED that STED/SED that STED/SED that STED/SED that STED/SED that STED/SED that ind i cates Loss of the indicates Potential indicates Loss of the indicates Potential indicates Loss of the indic ates Potential Loss Fue l Clad Barrier. Loss of the Fue l Clad RCS Barrier. Loss of the RCS Containment Barrier. of the Conta inment Barrier. Barrier. Barrier. NOTE 1: Refer to ER 1.1. Section 1.1 , Discussion concerning the proper use ofCSFSTs as EALs ECL: Notification of Unusual Event Initiating Condition:

FIRE potentially degrading the level of safety of the plant. Operating Mode Applicability:

All Emergency Action Levels: (1 or 2 or 3 or 4) Notes: (1)

• The STED/SED sho uld declare the Unusua l Event promptly upon determining that the applicable time ha s been exceeded , or will likely be exceeded.

-A containment fire alarm is considered valid upon receipt of an actuated alarm multiple zones (more than 1) actuated on CP-376 panel., combined with any of the following:

a. o GP 376 panel Multiple Zones Actuated o Plant Equipment Spuriously Operating o Containment Temperature Increasing 0 A FIRE is NOT extinguished within 15-minutes of ANY of the following FIRE detection indications:

Report from the field (i.e., visual observation)

Receipt of multiple (more than 1) fire alarms or indications Field verification of a single fire alarm AND b. The FIRE is located within ANY Table H2 plant rooms or areas: Table H2 Condensate Storage Tank Enclosure Fuel Storage Building Containment Primary Auxiliary Building Control Building Service Water Pump House Cooling Tower Steam and Feedwater Pipe Chases Diesel Generator Buildin g North Tank Farm Emergency Feedwater Pump House Startup Feedwater Pump Area RHR/CBS Equipme nt Vault OR (2) a. Receipt of a single fire alarm (i.e., no other indications of a FIRE). AND HU4 b. The FIRE is located within ANY of the Table H2 plant rooms or areas except Containment in Modes land 2 (see note above): AND c. The existence of a FIRE is not verified within 30-minutes of alarm receipt. OR (3) A FIRE within the plant PROTECTED AREA or Dry Fuel Storage Facility not extinguished within 60-minutes of the initial report , alarm or indication.

OR (4) A FIRE within the plant PROTECTED AREA-or Dry Fuel Storage Facility-that requires firefighting support b y an offsite fire response agency to extinguish.

ECL: Site Area Emergency Initiating Condition:

Inability to shutdown the reactor to neutron flm< < 5% causing a challenge to core cooling or RCS heat removal. Operating Mode Applicability

Emergency Action Levels: (1) a. AND b. AND c. Basis: An automatic or manual trip did not shutdown the reactor to neutron flux< 5%. All manual actions to shutdown the reactor have been unsuccessful.

EITHER of the following conditions exist: Core Coolin C CSF RED entr conditions met. Heat Sink (H CSF RED entr conditions met. MS5 This IC addresses a fai lur e of the RPS to initi ate or complete an automatic or manual reactor trip that results in a reactor shutdown , all subsequent operator actions to manually shutdown the reactor are unsuccessful , and continued power generation is challenging the capability to adequately remove heat from the core and/or the RCS. This condition will lead to fuel damage if additional mitigat i on actions are unsuccessful and thus warrants the declaration of a Site Area Emergency.

In some instances , the emergency classification resulting from this IC/EAL may be higher than that resulting from an assessment of the plant responses and symptoms against the Recognition Category F JCs/EALs.

This is appropriate in that the Recognition Category F ICs/EALs do not address the additional threat posed by a failure to shutdown the reactor. The inclusion of this IC and EAL ensures the timely declaration of a 8ite Area EmergenC)' in response to prolonged failure to shutdown the reactor. A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.

Escalation of the emergency classification level would be via IC RG 1 or FG 1.

MAS ECL: Alert Initiating Condition:

Automatic or manual trip fails to shutdown the reactor to neutron flm< < 5%, and subsequent manual actions taken at the Main Control Board are not successful in shutting down the reactor. Operating Mode Applicability:

Emergency Action Level: Note: A manual action is any operator action , or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies.

(1) a. AND b. Basis: An automatic or manual trip did not shutdown the reactor to neutron flmt < 5%. Manual actions taken at the MCB are not successful in shutting down the reactor. This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, and subsequent operator manual actions taken at the MCB to shutdown the reactor are also unsuccessful.

This condition represents an actual or potential substantial degradation of the level of safety of the plant. An emergency declaration is required even ifthe reactor is subsequently shutdown by an action taken away from the MCB since this event entails a significant failure of the RPS. A manual action at the f MCB is any operator action , or set of actions , which causes the control rods to be rapidly inserted into the core. This action does not include manually driving in control rods or implementation of boron injection strategies.

If this action(s) is unsuccessful , operators would immediately pursue additional manual actions at locations away from the MCB. Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room , are not considered to be "at the MCB". The plant response to the failure of an automatic or manual reactor trip will vary based upon several factors including the reactor power level prior to the event , availability of the condenser , performance of mitigation equipment and actions, other concurrent plant conditions , etc. If the failure to shutdown the reactor is prolonged enough to cause a challenge to the core cooling or RCS heat removal safety functions , the emergency classification level will escalate to a Site Area Emergency via IC MS5. Depending upon plant responses and symptoms, escalation is also possible via IC FSl. Absent the plant conditions needed to meet either IC MS5 or FSl, an Alert declaration is appropriate for this event. It is recognized that plant responses or symptom s may al s o require an Alert declaration in accordance with the Recognition Category F ICs; however , this IC and EAL are included to ensure a timely emergency declaration.

A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.

ECL: Notification of Unusual Event Initiating Condition:

Automatic or manual trip fails to shutdown the reactor to AeutroA flmt < 5%. Operating Mode Applicability:

1 MU5 Note: A manual action is any operator action , or set of actions , which causes the control rods to be rapidly in erted into the core, and does not include manuall y driving in control rods or implementation of boron injection strategies.

Emergency Action Levels: (1 or 2) (1) a. An automatic trip did not shutdown the reactor to AeutroA fltrn. < 5%. AND b. A subsequent manual action taken at the MCB is successful in shutting down the reactor. OR (2) a. A manual trip did not shutdown the reactor to AeutroA flux< 5%. AND b. EITHER of the following: 1. A subsequent manual action taken at the MCB is successfu l in shutting down the reactor. OR 2. A subsequent automatic trip is successful in shutting down the reactor. Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown , and either a subsequent operator manual action taken at the MCB or an automatic trip is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant. Fo ll owing the failure on an automatic reactor trip , operators will promptly initiate manual actions at the MCB to shutdown the reactor. If these manual actions are successful in shutting down the reactor, core heat generation wil l quickly fall to a level within the capabilities of the plant's decay heat removal systems. If an initial manual reactor trip is unsuccessful , operators will promptly take manual action at another location(s) on the MCB to shutdown the reactor. Depending upon several factors , the initial or subsequent effort to manually trip the reactor, or a concurrent plant condition, may le ad to the generation of an automatic reactor trip signal. If a subsequent manual or automatic trip is successful in shutting down the reactor , core heat generation wi ll quickly fal l to a level w ithin the capabilit i es of the plant's decay h eat removal systems. A manual action at the MCB is any operator action , or set of actions , which causes the contro l rods to be rapidly inserted int o the core. This action does not include manually driving in control rods or implementation of boron injection strategies.

Action s taken at back-panels or other locations within the Contro l Room , or any location outside the Control Room , are not considered to be " at the MCB". The plant response to the failure of an automatic or manual reactor trip will vary based upon several factors including the reactor power level prior to the event , availability of the condenser , performance of mitigation equipment and actions, other concurrent plant conditions , etc. If sub s equent operator manual actions taken at the MCB are also unsuccessful in shutt in g down the reactor , then the emergency c la ssification level will esca lat e to an Alert via JC MAS. Depending upon the plant response, escalation is also possible v ia IC FAI.

MA2 ECL: Alert Initiating Condition:

UNPLANNED lo ss of Control Room indications for 15 minutes or longer with a significant transient in progress.

Operating Mode Applicability:

1, 2, 3, 4 Emergency Action Levels: Note: The STED/SED should declare the Alert promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.

(1) Basis: a. AND An UNPLANNED event results in the inability to monitor one or more of the following parameters from within the Control Room for 15 minutes or longer. Reactor Power PFess1:1Fi:leF RCS Level RCS Pressure Core Exit eF RGS Temperature Level in at least two steam generators Steam Generator Emergency Feed Water Flow b. ANY of the following transient events in progress.

Automatic or manual run back greater than 25% thermal reactor power Electrical load rejection greater than 25% full electrical load Reactor trip SI actuation UNPLANNED:

A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.

The cause of the parameter change or event may be known or unknown. This IC addresses the difficulty associated w ith monitoring rapidly changing plant conditions during a transient without th e ability to obtain SAFETY SYSTEM parameters from wi thin the Control Ro om. During this condition, the margin to a potential fission product barrier challenge is reduced. It thus represents a potential substantial degradation in the level of safety of the plant. As used in this EAL, an " inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a l oss of all of the Control Room sources for the given parameter(s). For example, the reactor power level cannot be determined from any analog , digital and recorder source within the Control Room.

Enclosure 3 to SBK-L-16162 Clean Copy of Seabrook Station Emergency Action Levels -Initiating Conditions, Threshold Values and Basis 2.7 CLASSIFICATION OF SHORT-LIVED EVENTS Event-based ICs and EALs define a variety of specific occurrences that have potential or actual safety significance.

By their nature, some of these events may be short-lived and, thus, over before the emergency classification assessment can be completed.

If an event occurs that meets or exceeds an EAL, the associated ECL must be declared regardless of its continued presence at the time of declaration.

Examples of such events include a failure of the reactor protection system to automatically scram/trip the reactor followed by a successful manual scram/trip or an earthquake.

HAS ECL: Alert Initiating Condition:

Gaseous release impeding access to equipment necessary for normal plant operations, shutdown or cooldown.

Operating Mode Applicability:

All Emergency Action Levels: Note: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted.

(1) Basis: a. AND Release of a toxic, corrosive, asphyxiant or flammable gas into any Table HI rooms or areas. b. Entry into the room or area is prohibited or IMPEDED. Table Hl Area Mode Primary Aux Building 25 ft elevation 1, 2, 3, 4 7 ft elevation

-26 ft elevation Turbine Building 21 ft elevation 1, 2, 3 50 ft elevation Switchgear Rooms 1, 2, 3, 4 Essential Waste Process Building 25 ft elevation 1, 2, 3 -3 ft elevation Containment 3, 4 RHR/CBS Equipment Vaults 3, 4 IMPEDE: Entry into an area requires extraordinary measures to facilitate entry of personnel into the affected room/area by installing temporary shielding, requiring use of non-routine protective equipment, or requesting an extension in dose limits beyond normal administrative limits.

RAJ ECL: Alert Initiating Condition:

Radiation levels that IMPEDE access to equipment necessary for normal plant operations, shutdown or cooldown.

Operating Mode Applicability:

All Emergency Action Levels: (1 or 2) Note: If the equipment in the listed room or area was already inoperable or out-of-service before the event occurred, then no emergency classification is warranted.

(1) Dose rate greater than 15 mR/hr in ANY of the following areas: Control Room RM6550 Central Alarm Station (CAS) by survey Secondary Alarm Station (SAS) by survey OR (2) An UNPLANNED event results in radiation levels that prohibit or IMPEDE access to any of the following plant rooms or areas: Table Hl Area Mode Primary Aux Building 25 ft elevation 1, 2, 3, 4 7 ft elevation

-26 ft elevation Turbine Building 21 ft elevation 1, 2, 3 50 ft elevation Switchgear Rooms 1, 2, 3, 4 Essential Waste Process Building 25 ft elevation 1, 2, 3 -3 ft elevation Containment 3, 4 RHR/CBS Equipment Vaults 3,4 MA1 ECL: Alert Initiating Condition:

Loss of all but one AC power source to emergency buses for 15 minutes or longer. Operating Mode Applicability:

1, 2, 3, 4 Emergency Action Levels: Notes:

• The STED/SED should declare the Alert promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.

• For a bus to be considered energized from SEPS, both SEPS diesel generator sets must be functional.

(1) a. AC power capability to BOTH AC emergency buses E5 AND E6 is reduced to a single power source for 15 minutes or longer. Basis: AND b. Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS. NOTE There are six power sources to consider:

• 345 kV offsite power Line 369

• 345 kV offsite power Line 363

• 345 kV offsite power Line 394

• Emergency Diesel Generator A

• Emergency Diesel Generator B

• SEPS. For SEPS to be considered available, both SEPS diesel generator sets must be functional.

SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. Systems classified as related. This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this CU2 ECL: Notification of Unusual Event Initiating Condition:

Loss of all but one AC power source to emergency buses for 15 minutes or longer. Operating Mode Applicability:

5, 6, Defueled Emergency Action Levels: Notes: (1)

• The STED/SED should declare the Unusual Event promptly upon determining that 15 minutes has been exceeded, or will likely be exceeded.

• For power restoration from the SEPS, both SEPS diesel generator sets must be functional.

a. AC power capability to Both AC emergency buses E5 AND E6 is reduced to a single power source for 15 minutes or longer. AND b. Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS. NOTE \ There are six power sources to consider:

• 345 kV offsite power Line 369

• 345 kV offsite power Line 363

• 345 kV offsite power Line 394

• Emergency Diesel Generator A

• Emergency Diesel Generator B

• SEPS. For SEPS to be considered available, both SEPS diesel generator sets must be functional.

Basis: SAFETY SYSTEM: A system required for safe plant operation, cooling down the plant and/or placing it in the cold shutdown condition, including the ECCS. Systems classified as related. This IC describes a significant degradation of offsite and onsite AC power sources such that any additional single failure would result in a loss of all AC power to SAFETY SYSTEMS. In this condition, the sole AC power source may be powering one, or more than one, train of related equipment.

When in the cold shutdown, refueling, or defueled mode, this condition is not classified as an Alert because of the increased time available to restore another power source to service. Additional time is available due to the reduced core decay heat load, and the lower temperatures ECL: Notification of Unusual Event Initiating Condition:

Loss of all onsite or offsite communications capabilities.

Operating Mode Applicability:

5, 6, Defueled Emergency Action Levels: (1 or 2 or 3) (1) Loss of ALL of the following onsite communication methods: In-Plant (PBX) Telephones Gai-Tronics Plant Radio System OR (2) Loss of ALL of the following ORO communications methods: Nuclear Alert System (NAS) Backup NAS Control Room/TSC telephones OR (3) Loss of ALL of the following NRC communications methods: Basis: Emergency Notification System (ENS) Control Room/TSC telephones FTS telephones in the TSC cus This IC addresses a significant loss of on-site or offsite communications capabilities.

While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to OROs and theNRC. This IC should be assessed only when extraordinary means are being utilized to make communications possible.

EAL #1 addresses a total loss of the communications methods used in support ofroutine plant operations.

EAL #2 addresses a total loss of the communications methods used to notify all OROs of an emergency declaration.

The OROs referred to here are Commonwealth of Massachusetts and State of New Hampshire.

EAL #3 addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.

ECL: Notification of Unusual Event Initiating Condition:

Loss of all onsite or offsite communications capabilities.

Operating Mode Applicability:

1, 2, 3, 4 Emergency Action Levels: (1 or 2 or 3) (1) Loss of ALL of the following onsite communication methods: In-Plant (PBX) Telephones Gai-Tronics Plant Radio System OR (2) Loss of ALL of the following ORO communications methods: Nuclear Alert System (NAS) Backup NAS Control Room/TSC telephones OR (3) Loss of ALL of the following NRC communications methods: Basis: Emergency Notification System (ENS) Control Room/TSC telephones FTS telephones in the TSC MU6 This IC addresses a significant loss of on-site or offsite communications capabilities.

While not a direct challenge to plant or personnel safety, this event warrants prompt notifications to OROs and the NRC. This IC should be assessed only when extraordinary means are being utilized to make communications possible.

EAL #1 addresses a total loss of the communications methods used in support of routine plant operations.

EAL #2 addresses a total loss of the communications methods used to notify all OR Os of an emergency declaration.

The OROs referred to here are the Commonwealth of Massachusetts and State of New Hampshire.

EAL #3 addresses a total loss of the communications methods used to notify the NRC of an emergency declaration.

Fission Product Barrier Table Thresholds for LOSS or POTENTIAL LOSS of Barriers FGlGENERALEMERGENCY FSl SITE AREA EMERGENCY FAlALERT Loss of any two barriers and Loss or Loss or Potential Loss of any two barriers.

Any Loss or any Potential Loss of either Potential Loss of the third barrier. the Fuel Clad or RCS barrier. Fuel Clad Barrier RCS Barrier Containment Barrier LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS LOSS POTENTIAL LOSS 1. RCS or SG Tube Leakage 1. RCS or SG Tube Leakage 1. RCS or SG Tube Leakage Not Applicable A. Core Cooling (C) A. An automatic or A. Operation of a second A. A leaking or Not Applicable CSF-ORANGE manual SI actuation is charging pump in the RUPTURED SG is entry conditions met required by EITHER normal charging FAUL TED outside of (NOTE 1) of the foJiowing:

mode is required by containment.

1. UNISOLABLE EITHER of the RCS leakage following:

OR 1. UNI SO LAB LE 2. SGtube RCS leakage RUPTURE. OR 2. SG tube leakage. OR B. RCS Integrity (P) CSF -RED entry conditions met with RCS press> 300 psig. (NOTE 1) 2. Inadequate Heat Removal 2. Inadequate Heat Removal 2. Inadequate Heat Removal A. Core Cooling (C) A. Core Cooling (C) Not Applicable A. Heat Sink (H) CSF -Not Applicable A. Core Cooling (C) CSF CSF -RED entry CSF-ORANGE RED entry conditions

-RED entry conditions conditions met. entry conditions met. met. (NOTE 1) met for 15 minutes or (NOTE 1) (NOTE 1) longer. (NOTE 1) J OR B. Heat Sink (H) CSF -RED entry conditions met. !NOTE 1)

3. RCS Activity I Containment Radiation

3. RCS Activity I Containment Radiation

3. RCS Activity I Containment Radiation A. PostLOCA Not Applicable A. Post LOCA Radiation Not Applicable Not Applicable A. Post LOCA Radiation Radiation Monitors Monitors Monitors RM 6576A-1 or RM RM 6576A-1 or RM RM 6576A-1 or RM 6576B-1 6576B-l 6576B-l 2': 95 R/hr. =:: 16 R/hr. ::>: 1,305 R/hr .. OR B. RCS activity>

300 uCi/gm Dose Equivalent I 131 as determined per Procedure CS0925.01, Reactor Coolant Post Accident Sampling.

4. Containment Integrity or Bypass 4. Containment Integrity or Bypass 4. Containment Integrity or Bypass Not Applicable Not Applicable Not Applicable Not Applicable A. Containment isolation A. Containment (Z) CSF -is required RED entry conditions AND met. (NOTE 1) EIT,HER of the OR following:

B. Containment H2 I. Containment concentration=::

6% integrity has been OR lost based on STED/SED c. I. Containment judgment.

pressure > 18 psig OR AND 2. UNI SO LAB LE 2. Less than one full pathway from the train of containment to Containment the environment Building Spray exists. (CBS) is OR operating per design for 15 B. Indications of RCS minutes or longer. leakage outside of containment.

5. STED/SED Judgment 5. STED/SED Judgment 5. STED/SED Judgment A. ANY condition in A. ANY condition in the A. ANY condition in the A. ANY condition in the A. ANY condition in the A. ANY condition in the the opinion of the opinion of the opinion of the opinion of the opinion of the opinion of the STED/SED that STED/SED that STED/SED that STED/SED that STED/SED that STED/SED that indicates Loss of the indicates Potential indicates Loss of the indicates Potential indicates Loss of the indicates Potential Loss Fuel Clad Barrier. Loss of the Fuel Clad RCS Barrier. Loss of the RCS Containment Barrier. of the Containment Barrier. Barrier. Barrier. NOTE 1: Refer to ER 1.1, Section 1.1, Discussion concerning the proper use of CSFSTs as EALs ECL: Notification of Unusual Event Initiating Condition:

FIRE potentially degrading the level of safety of the plant. Operating Mode Applicability:

All Emergency Action Levels: (1 or 2 or 3 or 4) Notes: (1) OR a.

• The STED/SED should declare the Unusual Event promptly upon determining that the applicable time has been exceeded, or will likely be exceeded.

• A containment fire alarm is considered valid upon receipt of multiple zones more than 1 actuated on CP-376 anel. A FIRE is NOT extinguished within 15-minutes of ANY of the following FIRE detection indications:

Report from the field (i.e., visual observation)

Receipt of multiple (more than 1) fire alarms or indications Field verification of a single fire alarm AND b. The FIRE is located within ANY Table H2 plant rooms or areas: Table H2 Condensate Storage Tank Enclosure Fuel Storage Building Containment Primary Auxiliary Building Control Building Service Water Pump House Cooling Tower Steam and Feedwater Pipe Chases Diesel Generator Building North Tank Farm Emergency Feedwater Pump House Startup Feedwater Pump Area RHR/CBS Equipment Vault (2) a. Receipt of a single fire alarm (i.e., no other indications of a FIRE). AND HU4 b. The FIRE is located within ANY of the Table H2 plant rooms or areas except Containment in Modes land 2 (see note above): AND c. The existence of a FIRE is not verified within 30-minutes of alarm receipt. OR (3) A FIRE within the plant PROTECTED AREA or Dry Fuel Storage Facility not extinguished within 60-minutes of the initial report, alarm or indication.

OR (4) A FIRE within the plant PROTECTED AREA or Dry Fuel Storage Facility that requires firefighting support by an offsite fire response agency to extinguish.

MSS ECL: Site Area Emergency Initiating Condition:

Inability to shutdown the reactor causing a challenge to core cooling or RCS heat removal. Operating Mode Applicability:

1 Emergency Action Levels: (1) Basis: a. An automatic or manual trip did not shutdown the reactor. AND b. All manual actions to shutdown the reactor have been unsuccessful.

AND c. EITHER of the following conditions exist: Core Coolin C CSP RED entr conditions met. Heat Sink H CSP RED entr conditions met. This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, all subsequent operator actions to manually shutdown the reactor are unsuccessful, and continued power generation is challenging the capability to adequately remove heat from the core and/or the RCS. This condition will lead to fuel damage if additional mitigation actions are unsuccessful and thus warrants the declaration of a Site Area Emergency.

A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.

Escalation of the emergency classification level would be via IC RGI or PG 1.

MAS ECL: Alert Initiating Condition:

Automatic or manual trip fails to shutdown the reactor, and subsequent manual actions taken at the Main Control Board are not successful in shutting down the reactor. Operating Mode Applicability:

1 Emergency Action Level: Note: A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies.

(1) a. An automatic or manual trip did not shutdown the reactor. AND b. Manual actions taken at the MCB are not successful in shutting down the reactor. Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, and subsequent operator manual actions taken at the MCB to shutdown the reactor are also unsuccessful.

This condition represents an actual or potential substantial degradation of the level of safety of the plant. An emergency declaration is required even ifthe reactor is subsequently shutdown by an action taken away from the MCB since this event entails a significant failure of the RPS. A manual action at the MCB is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core. This action does not include manually driving in control rods or implementation of boron injection strategies.

If this action(s) is unsuccessful, operators would immediately pursue additional manual actions at locations away from the MCB. Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the MCB". The plant response to the failure of an automatic or manual reactor trip will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If the failure to shutdown the reactor is prolonged enough to cause a challenge to the core cooling or RCS heat removal safety functions, the emergency classification level will escalate to a Site Area Emergency via IC MS5. Depending upon plant responses and symptoms, escalation is also possible via IC FSl. Absent the plant conditions needed to meet either IC MS5 or FSl, an Alert declaration is appropriate for this event. It is recognized that plant responses or symptoms may also require an Alert declaration in accordance with the Recognition Category F ICs; however, this IC and EAL are included to ensure a timely emergency declaration.

A reactor shutdown is determined in accordance with applicable Emergency Operating Procedure criteria.

MUS ECL: Notification of Unusual Event Initiating Condition:

Automatic or manual trip fails to shutdown the reactor Operating Mode Applicability:

1 Note: A manual action is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core, and does not include manually driving in control rods or implementation of boron injection strategies.

Emergency Action Levels: (1 or 2) (1) a. An automatic trip did not shutdown the reactor. AND b. A subsequent manual action taken at the MCB is successful in shutting down the reactor. OR (2) a. A manual trip did not shutdown the reactor. AND b. EITHER of the following:

1. A subsequent manual action taken at the MCB is successful in shutting down the reactor. OR 2. A subsequent automatic trip is successful in shutting down the reactor. Basis: This IC addresses a failure of the RPS to initiate or complete an automatic or manual reactor trip that results in a reactor shutdown, and either a subsequent operator manual action taken at the MCB or an automatic trip is successful in shutting down the reactor. This event is a precursor to a more significant condition and thus represents a potential degradation of the level of safety of the plant. Following the failure on an automatic reactor trip, operators will promptly initiate manual actions at the MCB to shutdown the reactor. If these manual actions are successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems. If an initial manual reactor trip is unsuccessful, operators will promptly take manual action at another location(s) on the MCB to shutdown the reactor. Depending upon several factors, the initial or subsequent effort to manually trip the reactor, or a concurrent plant condition, may lead to the generation of an automatic reactor trip signal. If a subsequent manual or automatic trip is successful in shutting down the reactor, core heat generation will quickly fall to a level within the capabilities of the plant's decay heat removal systems. A manual action at the MCB is any operator action, or set of actions, which causes the control rods to be rapidly inserted into the core. This action does not include manually driving in control rods or implementation of boron injection strategies.

Actions taken at back-panels or other locations within the Control Room, or any location outside the Control Room, are not considered to be "at the MCB". The plant response to the failure of an automatic or manual reactor trip will vary based upon several factors including the reactor power level prior to the event, availability of the condenser, performance of mitigation equipment and actions, other concurrent plant conditions, etc. If subsequent operator manual actions taken at the MCB are also unsuccessful in shutting down the reactor, then the emergency classification level will escalate to an Alert via IC MA5. Depending upon the plant response, escalation is also possible via IC FAl.

MA2 ECL: Alert Initiating Condition:

UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress.

Operating Mode Applicability:

1, 2, 3, 4 Emergency Action Levels: Note: The STED/SED should declare the Alert promptly upon determining that 15 minutes has (1) Basis: been exceeded, or will likely be exceeded.

• a. AND An UNPLANNED event results in the inability to monitor one or more of the following parameters from within the Control Room for 15 minutes or longer. Reactor Power RCS Level RCS Pressure Core Exit Temperature Level in at least two steam generators Steam Generator Emergency Feed Water Flow b. ANY of the following transient events in progress.

Automatic or manual runback greater than 25% thermal reactor power Electrical load rejection greater than 25% full electrical load Reactor trip SI actuation UNPLANNED:

A parameter change or an event that is not 1) the result of an intended evolution or 2) an expected plant response to a transient.

The cause of the parameter change or event may be known or unknown. This IC addresses the difficulty associated with monitoring rapidly changing plant conditions during a transient without the ability to obtain SAFETY SYSTEM parameters from within the Control Room. During this condition, the margin to a potential fission product barrier challenge is reduced. It thus represents a potential substantial degradation in the level of safety of the plant. As used in this EAL, an "inability to monitor" means that values for one or more of the listed parameters cannot be determined from within the Control Room. This situation would require a loss of all of the Control Room sources for the given parameter(s).

For example, the reactor power level cannot be determined from any analog, digital and recorder source within the Control Room.

Enclosure 4 to SBK-L-16162 NEI 99-01, Rev. 6, Deviations and Differences, Seabrook Station Nuclear Power Plant -Unit 1 ABNORMAL RAD LEVELS/ RAD I OLOGICAL EFFLUENT ICS/EALS (1) Dose rate g r eater t h a n 1 5 m R/hr i n ANY of t h e fo ll ow ing a r eas: (1) Dose ra t e greater than 1 5 m R/hr i n AN Y of t h e fo ll ow in g areas:

• Co n t ro l R oo m I Co n tro l R o o m I

• Ce n tra l A l ar m S t a ti o n I Ce n tra l A l a rm S tati o n (CAS) by s ur vey I * (ot h e r s it e-s p ec ific a r eas/roo m s) I Secondary A l arm Station (SAS) by survey I (2) A n UN P LANNE D eve nt r es ul ts in ra di a tion l eve l s that p ro h i bi t or imp e d e access to a ny o f th e fo ll ow ing p l a n t roo ms o r a r eas: (2) A n UNPLANNE D eve nt r es ult s in r a di a ti on l eve l s that p ro hi b i t or (s it e-s p ec ifi c li s t of p l ant ro o ms o r a re as w i t h e nt ry-r e l a t ed m o d e I M P E D E a ccess t o an y of th e fo llowin g pl a nt rooms o r a r eas: a pp l i c ab il i ty identifi e d) [_able Hl Area Mode r i mary Aux Buildin g 25 ft elevation I 2 3 4 7 ft elevation

-26 ft eleva ti2!] if urbine Building 21 ft elevation I , 2 , 3 50 ft elevation Switc h gear Rooms I , 2 , 3, 4 Waste Process Bui l ding 25 ft elevatio I , 2 , 3 -3 ft elevatio Containment 3 , 4 RHR/CBS Eg!!irment Vaults 3,4 Difference

/Ju s ti fi cation T able Hl: S it e spec i fic in fo rm a ti o n , see V 7 -Ta bl e Hl Proce dur e Refe r e n ces 1 2 COLD SHUTDOWN/

REFUELING SYSTEM MALFUNCTION ICS/EALS CU2: INITIATING CONDITIONS NEl 99-01R ev6 S eabrook S tation Nuclear Power Plant Loss of a ll bu t one AC powe r so ur ce to e m e r gency bu ses for 15 minu tes o r l o n ger. Loss of a ll bu t on e AC p owe r source to e m e r ge n cy bu ses for 1 5 m i nut es or l o n ger. Difference

/Ju s tification No n e THRESHOLDS NEC 99-01 Rev 6 Seabrook Station Nuclear Power Plant (1) AC p ower capab ili ty to (s it e-spec i fic e m e r ge n cy b u ses) i s r ed u ced to a Note: For power restoration from the SEPS, both SEPS d i esel generator sets sing l e power source for 15 m i nutes or l onger. ust be functiona l. AN D (1) a. AC po we r ca p a bili ty t o OT AC e m e r ge n cy buses 5 AND E6 i s b. A n y ad di tio n a l s in g l e p owe r so ur ce fa ilu re w ill r es ul t in l oss of a ll AC r e duced to a s in g l e pow e r so ur ce for 1 5 minut es or l o n ge r. powe r to SAFET Y S Y STE M S. A ND b. A n y a dditi o n a l s in g l e power so ur ce fa ilu re w i ll res ult in l oss of a ll AC po wer t o SAFET Y S Y STEMS. NOTE There are six power sources to consider: . 345 kV offsite power Line 369 . 345 kV offsite power Line 363 . 345 kV offsite power Line 394 . Emergency Diesel Generator A . Emerge ncy Die sel Generator B . SEPS. For SEPS to be considered available, both SEPS diesel generator sets must be functional.

Difference

/Ju s tification Added NO TE to clarify that bot h SEPS constitute a s i ng l e power source. Added NOTE containing table of AC power so urc es per EPFAQ 2015-15. 26

---

COLD SHUTDOWN/

REFUELING SYSTEM MALFUNCTION ICS/EALS CUS: INITIATING CONDITIONS NEI 9 9-01 Rev 6 Seabrook Station Nuclear Power Plant Loss of a ll onsite or offsite communications capabilities.

Loss of all onsite or offsite communications capabilitie

s. Difference

/Justification None THRESHOLDS NEJ 99-01 Rev 6 Seabrook Station Nuclear Power Plant (1) Loss of ALL of the following onsite communication m et hod s: (I) Loss of ALL of the following onsite communica t i o n method s: (site-spec ifi c li st of communications methods) I In-Plant (PBX) Te l e_phones I (2) Loss of ALL of the following ORO communications m et hod s: I Gai-Troni cs I (s ite-specifi c li s t of communications methods) I r 1ant R ad i o System I (3) Loss of ALL of the following NRC communications methods: (site-s p ec ifi c li st of communications method s) (2) Loss of ALL of the following ORO communicat i ons method s: I !N uclear Alert System (i'JAfil I I I I Co ntr o l Room/TSC telephon es I I I (3) Loss of ALL of the following NRC communicat i ons method s: I Notification System (ENS) I I A ll plant telephones I I f TS telephone s in the TS C I I I Diff e rence /Justification Prov ided site s pecific communications methods 29 INDEPENDENT SPENT FUEL STORAGE FACILITY {ISFSI) ICS/EALS EU l: I NITIATING CONDITION S NEI 9 9-0 l R ev 6 Sea b roo k S tation Nucle a r Po we r P l a nt D am age to a loade d cask CONFINEMENT BOUNDARY.

Damage to a l oaded cask CONFINEMENT B OUNDARY. Diff ere n ce /J u s tifica t ion N on e THRESHOLDS NEI 9 9-0 l R ev 6 S eabrook S tation Nucle a r Power P l a nt (1) Damage to a l oaded cask CONFlNEMENT B OUNDA R Y as (I) Damage to a l oaded cask CONFINEMENT BOUNDARY as i ndica t ed by a n o n-co nt act r a d iat i on r ead in g greater than (2 ti mes th e indi cated b y A N Y of t h e fo ll owi ng o n-contact surface rad i at i o n site-spec ifi c cask spec ifi c t ec h n i ca l s p ecifica ti o n a ll owa bl e rad i at i o n rea din gs greater th a n: level) o n t h e surface of t h e spent fue l cask. 1600 mrem/hr at the front bird screen 4 mrem/hr at the door centerline 4 mrem/hr at the end shield wall exterio n Diff e r e nc e /J u s tific at ion A dd e d NOTE pull e d from t h e b as is a ll ow in g ca l c ul a ting s u rface d ose fro m a di s t a n ce dose EU l.l: Site spe ci fic i nfo rm a tio n, see V I 9 NU HOMS HSM Dose Rates Technical Specificat i on 30 FISSION PRODUCT BARRIER ICS/EALS B. RC S cooldown rate greater than (site-s pecific pres s urized th e rm a l shock criteria/limits defin ed by s it e-s pecific indic atio n s). Seabrook Station Nuclear Power Plant Not Applicab l e A. Core Coo lin g A. An automatic or A. Operation of a second A. A leakin g or Not App li cab l e (C) CSF -ORANGE manual SI actuation i s char gi n g pump in the RUP TURE D SG is e n try cond iti o n s m et required by EITHER of norm al charging mode FAULTED outside of (NOTE I) the fo llo wing: i s required by EITHER containme nt.

• UNISOLABLE of the fo llowin g: RCS l eakage I. UNISOLABLE OR RCS l ea k age

• SG tube RUPTURE. OR 2. SG tube leak age. OR B. CS In teg ri ty (P) CS -R ED e n try conditions m et wit h R CS 2 r ess > sig. (NOTE I) Difference

/J ustification Fuel Clad Barrier Potential Loss LA: Site s p ecific information , see V20 CSFST Co r e Cooling RCS Barrie r Potential Loss 1.B: Site spec i fic information , see V2 l CSFST lnte g rity 32 FISSION PRODUCT BARRIER ICS/EA LS NEI 99-01 Rev 6 2. Ina d e q uate Heat R e m ova l 2. I n a d equate Heat Remova l 2. I n adeq u a t e Heat Re m ova l A. Co re ex i t A. Core ex it Not Applicable A. Inadequate RCS heat Not Applicable A. I. (Site-specific thermocouple thermocouple removal capability via criteria for readi n gs greater readings greater steam ge nerators as entry into core than (site-specific than (site-specific indicated by (site-cooling tempera t ure temperature s pecific indication s). restoration value). value). procedure) OR AND B. lnadequate RCS 2. Restoration heat remova l procedure not capabi li ty v i a effective within steam generators 15 minute s. as indicated by (site-specific indications).

Se abrook S t a tion Nuclear Pow e r Plant A. Core C o olin g (C) A. Core Coo lin g (C) Not Applicab l e A. eat Sink (H) CSF -Not Applicab l e A. Core Cooling (C) CSF -RED entry CSF-ORANG ED entry conditions CSF -RED entry conditions met. entry cond ition s m et. (NOTE I) conditions met for (NOTE I) me t.(N OTE I) 15 minutes or OR onger. (NOTE I) 8. eat S ink (H) CSF -RED entry conditions me t. (NOTE I) Di ffere nc e /J u stification Fue l C l a d Ba r r i er: Loss 2.A, Potentia l Loss 2.A and Co n tainment Barr i er Pote n t i a l Loss 2.A: Site specific information , see V20 CSFST Core Cooling R CS Ba rri er: Potent i a l L o ss 2.A: Site specific informatio n , see V22 CSFST Heat Sink 33 FISSION PRODUCT BARRIER ICS/EALS N o t A pplicabl e Not A ppli ca bl e N o t A pplic ab l e Not A ppli ca bl e A. C on ta inm e nt i s ol a ti on i s A. Containment (Z) r equire d CSF -RED entry AN D conditions met. E I THE R of t h e (NOTE I) followi n g: OR 3. Co n tai nm e n t B. Cn m t. h y dro gen i n teg ri ty h as bee n co n ce n tra ti o n 6% l ost b ase d on OR STE D/SE D C. I. judg m e nt. pressure > 18 OR s i g 4. UN I S O LA B LE AND p a thw ay from t h e 2. Less than on e co nt a inm e nt to t h e full train o e n v ironm e n t exists. Cnmt. Building OR S ray (CBS) is B. Indi cat i o n s o f R CS o p era tin g p er l eakage out s ide o f d es i g n for 1 5 co n ta inm e n t. minu tes or l o n ge r. Difference

/J u s tification Conta i nment Barri e r: Potential Loss 4.A: S i te s p e cific in fo rm a ti o n , see V 24 CSFS T Co n ta inm e nt Containment Barri e r: Potential Los s 4.8: S i te s p ec ifi c in fo rm at i o n , see V 1 4 H 2 co nc e ntr at ion in c on ta i n m e nt Containment Barri e r: Potential Lo ss 4.Cl: S i te s p ec ifi c in for m ati on , see V 25 C on ta inm e n t S pra y Se tp oint NEI 99-01 R ev 6 5. Other Indication s 5. Other Indications

5. Oth e r Indication s A. (site-s p ec i fic as A. (site-spec ifi c as A. (s i te-specific as A. (s i te-specific as A. (s i te-spec i fic as A. (site-s p ec i fic as a pplic a bl e) a ppl ica bl e) a ppli ca bl e) a ppli ca bl e) app l ica bl e) app li cab l e) Seabrook Station Nuclear Power Plant 36 FISSION PRODUCT BARRIER ICS/EALS Not Applicable Not Applicable Not Applicable Not Applicab l e Not Applicable Not Applicable Difference

/J u st ification None NEI 99-01 Rev 6 6. Emergency Director Judgment 6. Eme rgenc y Director Judgment 6. Emergency Director Judgment A. ANY co n dition in A. ANY conditio n in A. ANY condition in the A. ANY condition in the A. ANY condition in the A. ANY condition in the opin i on of the the opi nion of th e opinion of t he opin i on of the opinion of the the opinion of the Emerge nc y Emerge ncy E m ergency D i rector Emerge nc y Director Emerge nc y Director tha t Emerge nc y Director Director that Director that t h at indicates Loss of that indicates Potential indicates Loss of the that indicates indicates Loss of indicates Potential the RCS Barrier. Loss of the RCS Containment Barrier. Potential Loss of the the Fuel Clad Lo ss of the Fuel Barrier. Containment Barrier. Barrier. Clad Barrier. Seab r ook Statio n Nuclear Power P l ant A. ANY condition in A. ANY condition in A. ANY condition in the A. ANY condition in the A. ANY condition in the A. ANY condition in the opin i on of the the opinion of the opinion of the opinion of the opinion of the the opinion of the STED/SED that STED/SED that STED/SED that STED/SED that STED/SED that STED/SED that indicates Loss of indicates Potential indicat es Loss of the indic ates Potential Loss indicates Loss of the indicates Potential the Fue l C l ad Loss of the Fuel RCS Barrier. of the RCS Barrier. Containment Barrier. Loss of the Barrier. Clad Barrier. Containment Barrier. NOTE I: Refer to E R I. I Section 1.1 , Discussion concerning the proper use of CSFSTs as EALs Difference

/Justification None 37 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS HGl: INITIATING CONDITIONS NEl 99-01Rev6 Seabrook Station Nuclear Power Plant HOSTlLE ACTIO N resulting in lo ss of physical control of the facility.

DEL ETE D Difference

/Justification None THRESHOLDS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant (1) a. A HOSTILE ACTION is occurring or has occurred within DELETED the PROTECTE D AREA as reported by the (si te-specific security sh ift supervision).

AND b. EITHER of the following h as occurred: 1. ANY of the fo llo wi n g safety functions cannot be co ntrolled or m a int a in ed.

• Re act i v i ty control

• Core cooling [PWR] I RPV water lev el [BWR]

• RCS h eat removal OR 2. Damage to spe nt fuel has occurred or i s IMMINENT.

Difference

/J ustifica t ion HG I del eted per staff resolution to EPF AQ 2015-13. HG 1 conditions are bounded by initiating conditions RA2, RS2, RG2, RS I , RG 1 , HS 1 , HS6, HS7 and HG2. 38 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS (1) a. b. Re l ease of a tox i c , corros i ve , asp h yxiant or flammab l e gas into a n y of t h e fo ll ow i ng p l a n t ro o m s or areas: (s it e-spec i fic li st o f pl a n t roo ms o r areas w i t h e ntr y-re l a t e d m ode app li ca bili ty i d e n tifie d) AND E n try into t h e room or area i s p ro h i b ited or imp e d ed. Diff e r e nce /Ju s tifica t ion HAS.lb: Site s p ec ifi c in fo rm at ion , see V 7 -Ta b le HJ Proced u re R efere n ces (1) 45 a. Release of a tox i c , corros i ve , asphyx i an t or fl ammab l e gas in to a n y able HI rooms or areas. A ND b. E n tr in to t h e room o r a r ea i s ro hibi ted o r i m eded. Area Primary Aux Building 25 ft elevatio 7 ft elevatio -26 ft elevation if urbine Building 21 ft elevatio able HI ssential Switchgear Rooms Waste Process Building 25 ft elevatio ode I , 2 , 3 , 4 1 , 2 , 3 I, 2 , 3 , 4 I, 2 , 3 3 ,4 3, 4 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS (1) (2) (3) (4) a. b. a. A F IR E is N O T ext in g ui s h e d w ith in 1 5-minut es of ANY of t h e fo ll ow ing F IRE d e t ect ion indi ca t io n s:

• R eport fro m th e fie ld (i.e., vi s u a l obse rv a ti o n)

• Recei pt o f m ul tiple (m o re t h an I) fire a l ar m s or i n dica ti o n s

• F i e ld ve rifi ca ti on of a s in g l e fire a l a rm AN D T h e FIRE i s l oca t ed wit hin ANY of the fo ll ow in g pl an t roo ms o r a r eas: (s it e-spec ific li s t of p l ant roo ms o r a r eas) Receipt of a s in g l e fire a l ar m (i.e., n o ot h e r indi cat i o n s of a F IR E). AN D b. T h e FIRE i s l oca t ed wit hin AN Y of the fo ll ow ing p l a n t rooms o r a r eas: (s it e-s p ec ific l is t of pl ant r o oms o r areas) AN D c. T h e existe nc e of a FIRE is n o t ve r ified w i t hin 30-mi nut es o f a l a r m rece ipt. A F I RE w ith i n t h e pl a nt or JSFSJ [f o r p l a nts w i th a n J SFS J outs i de the p l a n t P rotected A r ea] P ROTECTE D AREA n o t ext in g ui s h e d wit hin 60-mi n ut es of th e ini t i a l re p o rt , al ar m or in dicat i o n. A F IR E w ithin t h e pl a nt or JSFSJ [for p l a nt s wit h a n J SFS J outside t h e p l a n t P rotected Area] PRO TEC T E D AREA t h at r e quir es fire fi g ht ing s upp o rt by a n offsite fir e r es pon se age n cy t o ext in g ui s h. Differ e n c e /Ju s t ifi ca t ion NOTE: A containment fire alarm is considered valid upon recei t of multi le ones more than I) actuated a larm on CP 376. (1) a. A F I RE is NOT ex tin g u is h ed w ith i n 15-m i nut es o f ANY of the fo ll ow in F I RE d etectio n ind i ca ti o n s: (2) R e p ort from th e fi e ld (i.e., v i s u a l o b servatio n) R eceipt of mul ti ple (m ore tha n 1) fir e a l a r ms o r in d i cations F i e ld verifica ti o n of a s i ng l e fire a l arm AN D b T he F I RE i s l ocate d wit h i n AN Y T a bl e H2 o l a n t roo m s or areas: ff able H2 Con d e n sa t e Stora!!.e Tank Enclosure f u el Stora!!.e Bu ildin 2! Con t ainmen v Prim arv Auxi li arv Bu il din lli Co ntrol Buildin E!I Service Water P ump Hou se Coo lin g Tower Ste a m and Fe ed water Pi Chases D i ese l Generator Buildin!!l No rth Tank Fa m1 Feedwa ter Pumo H o u se Startuo Fee d wa t e r Pump Area !E quipment Vault a. R ece ip t of a s in g l e fi re a l ar m (i.e., n o othe r in d i cat i o n s of a FI RE). AN D b. T h e FIRE is l oca t e d w i t h i n ANY o f th e if ab le H 2 pl a nt roo ms o r a r eas exce p t Co nt ai nm e nt i n Mo d es 1 a nd 2. (see not e above) AN D c. T h e existe n ce of a F I RE is n ot ve rifi e d wit hin 30-minu tes of a l a rm recei p t. (3) A F I RE within t h e pl a n t PR OTECTE D A RE A o r Dry F uel Storage Fac ility n o t exti n g u is h e d wi th in 60-minut es of t he i nit ial r e p o rt , a l a rm or indi ca ti o n. (4) A F I RE w i thi n th e pl a nt PR OTECTE D AREA o r Dry F u el Storage Fac ili ty t h at r e quir es fir e fi g h t ing s upport by a n offs i t e fir e res p o n se age n cy to ext in g ui s h. 52 HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS Added NOTE to clari fy the contai nm ent fire alarm HU4.lb: Site specific informatio n , see V28 Verification of Fire Areas HU4.2b: Contai nm ent is excepte d in Modes I and 2 but is covered by the seco nd n ote. EALl would be app li cable. Entry into the Containme n t to perform verifications within 30 minutes in Modes 1 and 2 is a challenge.

53 SYSTEM MALFUNCTIONS MSS: INITIATING CONDITIONS NEI 99-0 1 Rev 6 Se abrook S t at ion N uclear Power P l a nt In a b il ity to sh u t d own th e reacto r ca u s i ng a c h a ll e n ge to (co r e coolin g [PW R] In ab il ity to s hutd own t h e r eactor ca u si ng a c h a ll enge to c or e coo l ing o r R CS I RP V water l eve l [BW R]) o r R CS h ea t remova l. h eat remova l. Diffe r ence /Ju s tifica t ion THRESHOLDS N EI 9 9-01R ev 6 Se abrook Station Nuclear Power Plant (I) a. An a u to m atic or m a nu a l (trip [PWR] I scram [BWR]) di d n ot (I) a. A n a u tomat ic or m a n ua l tri p di d n ot s hu t d ow n t h e reactor s h utdow n the reacto r. AND b. A ll m a nu a l act ion s to s hutd o wn th e r eactor have b een AN D un s u ccess ful. AND b. A ll m an u a l act i o n s to s hu t dow n th e r eactor h ave b ee n c. E ITHER of t h e fo ll owi ng c onditi o n s ex i st: un s u ccessfu l. I Core Coo lin g (C) CSF -RED entry condit ion s met. I AND I Sink (H) -RED entry conditions met. I c. E ITH E R of t h e fo llo w in g co nditi o n s ex i st: * (S i te-s p ec ifi c in dication of a n in abi l ity t o a d e qu ate l y r e m ove h eat from the co r e) * (S i te-s p ec ifi c ind icatio n o f a n i n ab ility to a d e qu ately r e m ove h ea t fro m t h e R CS) Diffe r ence /Justifica t ion M S S.lc: Site s p ec ifi c in for m a ti o n , see V20 CSFST Core Coo li ng and V22 CSFST Heat S in k 5 8 SYSTEM MALFUNCTIONS MAl: INITIATING CONDITIONS NEJ 99-01 Rev 6 Seabrook Station Nuclear Power Plant Los s of all but one AC power source to emergency buse s for 15 minutes or Loss of all but one AC power source to emergency buses for 15 minutes or lon ger. lon ger. Difference

/Justification None THRESHOLDS NEl 99-01 Rev 6 Seabrook Station Nuclear Power Plant (1) a. AC power capability to (s it e-specific emergency buses) i s ( 1) a. AC power capability to OTH AC emergency buses 5 AN D E6 is reduced to a single power so urc e for 15 minutes or lon ger. reduced to a s ingl e power so ur ce for 15 minutes or lon ger. AND b. An y additiona l sing l e power so ur ce failure wi ll re s ult in a AND lo ss of all AC power to SAFETY SYSTEMS. b. Any additional single power source failure will result in loss of all AC power to SAFETY SYSTEMS. NOTE There are six power sources to consider: . 345 kV offsite power Line 369 . 345 kV offsite power Line 363 . 345 kV offsite power Line 394 . Emergency Diesel Generator A . Emergency Diesel Generator B . SEPS. For SEPS to be considered available , both SEPS diesel generator sets must be functional.

Difference

/J ustification 60 SYSTEM MALFUNCTIONS (1) a. An UNPLANNED event results in the inability to monitor (I) a. An UNPLANNED eve nt r esu lt s in the inability to monitor one one or more of the following parameters from within the or more of the fo ll owing parameters from wit hin the Contro l Co ntrol Room for 1 5 minutes or lon ger. Ro om for 15 minut es or l onger. Reactor Power f BWR parameter listl f PWR parameter listl RCS Level Reactor Power R eactor Power RCS Pressure RPV Water Level R CS Level Core Exit Temperature RPV Pressure RC S Pr ess ur e Levels in at l east two stea m ge n erators Prim ary Co nt a inm ent In-Core/Core Exit Steam Generator Emerge nc y Feed Water Flow Pr essu r e Temperature AND Suppressio n Pool Leve l Levels in at l east (site-b. ANY of the fo llo wing transient events i n progress.

s pecific number) steam A utomatic or manual run b ack greate r t h a n 25% ge n erators thermal reactor power S uppr essio n Pool Stea m Generator Auxiliary Electr ic a l lo ad rejection g r eater than 25% full Temperat ur e or Emergency Feed Wate r electrical l oad Flow I Reactor trip I AND I ECCS (SI) actuatio n I b. ANY of the fo llo w in g transie nt eve nt s in progress.

• Automat i c or manual run back greater than 25% therma l reactor power

• E l ectrica l load r eject ion greate r than 25% full e l ectr ical l oad

• Reactor scram [B W R] I trip [PWR]

• ECCS (S I) actuation

• Thermal pow e r osci ll at ion s greater than 10% [BWR] Diff e rence /Justific a tion Non e 62 SYSTEM MALFUNCTIONS MAS: INITIATING CONDITIONS NEr 99-01 Rev 6 Seabrook Station Nuclear Power Plant A utom atic or m anua l (trip [PWR] I scram [BWR]) fails to sh ut down the Automatic or manual trip fails to shutdown the reactor and s ub se quent reactor , and subsequent manual actions taken at the reactor control conso l es manual act ion s taken at the Main Control Board are not successful in shutting are not successful in shutting down the reactor. down the reactor. Difference

/J ustification None THRESHOLDS NEI 99-01Rev6 Seabrook Station Nuclear Power Plant (I) a. An automatic or manual (trip [PWR] I scra m [BWR]) did (I) a. An automatic or manual trip did not shutdown the reactor not shutdown the reactor. AND b. Manual actions taken at the CB are not successful in shutting AND down the reactor. b. Manual actio n s taken at the reactor control conso l es are not s uc cessful in s huttin g down the reactor. D i fference /Justification 63 SYSTEM MALFUNCTIONS MU2: INITIATING CONDITIONS NEl 99-01 Rev 6 Seabrook Station Nuclear Power Plant UNPLANNED lo ss of Contro l Room indications for 15 minutes or longer. UNPLANNED loss of Control Room indications for 15 minutes or longer. D i fference /J ustification None THRESHOLDS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant (I) a. An UNPLANNED event results in the inability to monitor (I) a. A n UNPLANNED event results in the inability to monitor one or one or more of the following parameters from within the more of the following parameters from within the Control Room for 15 Control Room for 15 minutes or longer. minutes or lon g er. Reactor Power [BWR param e t e r listl [PWR param e t e r listl RCS Level Reactor Power Reactor Power RCS Pressure Core Exit Temperature RPV Water Level RCS Leve l Level in at least two steam generators RPV Pressure RCS Pressure Stea m Generator Emergency Feed Water Flow Primary Co nt ainment In-Core/Core Exit Pressure Temperature Suppress ion Pool Level Levels in at lea st (site-specific number) steam generators Suppre s sion Pool Steam Generator Temperature Auxiliary or Emergency Feed Water F lo w Diff e rence /J ustific a tion Non e 66 SYSTEM MALFUNCTIONS MUS: INITIATING CONDITIONS NEl 99-01 Rev 6 Seabroo k S tation Nuclear Power Plant A ut omatic or manual (trip [PWR] I scram [BWR]) fails to shutdown the Automatic or manual trip fails to shutdown the reactor .. reactor. Difference

/Justification THRESHOLDS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant (I) a. An automatic (trip [PWR] I scram [BWR]) did not shutdow n (I) a. An automatic trip did not shutdown the reactor the r eacto r. AND AND b. A subsequent manual action taken at the M CB is su ccessful in b. A subsequent manual action taken at the reactor control shutting down the reactor. consoles is successfu l in shutting down the reactor. OR (2) a. A manual trip ([PWR] I scram [BWR]) did not sh utdo wn the (2) a. A manual trip did not shutdown the reactor reactor. AND AND b. EITHER of the following:

b. EITHER of the following: 1. A s ub se qu ent manual action taken a t the CB is s u ccessful 1. A subseque nt manual actio n taken at the reactor control in sh uttin g down the reactor. conso l es is successful in sh utt ing down the reactor. OR OR 2. A subsequent automatic trip is successful in shutting down 2. A s ubsequent a ut omatic (trip [PWR] I scram [BWR]) is the reactor. successful in s huttin g down the reactor. Di ffe rence /J ustific at ion 69 SYSTEM MALFUNCTIONS MU6: INITIATING CONDITIONS NEI 9 9-0 1 R ev 6 Sea brook S t a tion N uclear Po we r Pl a nt Loss of all onsite or offsite co mmuni cations capa bil ities. Loss of a ll o n s i te or offsite commun i cations capab ili ties. Diff e r e nce /Justifi c ation Non e THRESHOLDS NEI 9 9-01 Re v 6 S eabrook S tation N uclear Power Pl a nt (1) Loss of ALL of the fo ll owi n g ons i te co mm unication m et h ods: (I) Loss of ALL of t h e fo ll owi n g o n s i te co mmu nicatio n met h ods: (site-s p ec i fic I i st of co mmun icatio n s m et h ods) I [n Plant (PBX) Telt'.j?hones I (2) Loss of ALL of t h e fo ll ow in g ORO co mm u n icat i o n s m et h ods: I Gai Tronics I (site-spec i fic l ist of co mmun icatio n s m et h ods) I Pla nt Radio Sy s te ll] I (3) Loss of A L L of t h e fo ll ow in g NRC com mun icatio n s m et h ods: (si t e-s p ec i fic li st of co mmun icatio n s m e thod s) (2) Loss of A LL of t h e fo ll owi n g O R O co mmuni cat i o n s met h o d s: I Nuclear Alert System (NAS) I I 1Backup NAS I I Control Room/TSC plant telephones I I I (3) Loss of A LL of t h e fo ll owi n g NRC communica t io n s me t ho d s: I Notification System @NS) I I All plant telephones I I f TS telephones in the TSC I I I Diff e r e nc e /J u s tific a tion Pro v id e d site s p ec ifi c comm unic a ti o n s m et h o d s 70 L