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

NEXTera ENERGY~

~AB ROOK October 27, 2016 10 CFR 50.90 SBK-L-16162 Docket No. 50-443 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Seabrook Station 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 Non-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:

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" Markup of Affected Seabrook Station Emergency Action Levels - Initiating Conditions, Threshold Values and Basis 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" to SBK-L-16162 Page 2of12

Background

By letter dated February 27, 2016 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML16068A128), 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.

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 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.

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. ML16166A191), 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.

to SBK-L-16162 Page 6of12 SA5 Note:

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.

RAl-Seabrook-5 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 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 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. ML16166A366) 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 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 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.

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."

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 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) a. Release of a toxic, corrosive, asphyxiant or flammable gas into any Table H 1.

rooms or areas.

AND

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 Basis:

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 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 .

(I) Dose rate greater than 15 mR/hr in ANY of the fo llowing areas :

Control Room RM6550 Central A larm 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.

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 safety-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 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:

• 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.

(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.

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 safety-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 safety-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

CU5 ECL: Notification of Unusual Event Initiating Condition: Loss of all onsite or offsite communications capabilities.

Operating Mode Applicability: 5, 6, Defueled Emergency Action Levels: (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:

Nuclear Alert System (NAS)

Backup NAS Al+Control RoomffSC plaffi telephones Cellular telephoAes OR (3) Loss of ALL of the following NRC communications methods:

Emergency Notification System (ENS)

Al+Control RoomffSC plaffi telephones FTS te lephones in the TSC Cellular telephoAes Basis:

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 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.

MU6 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 A-ltControl RoomffSC plaflt telephones Cellular telephones OR (3) Loss of ALL of the following NRC communications methods:

Emergency Notification System (ENS)

A-ltControl RoomffSC plaflt telephones FTS telephones in the TSC Cellular telephones Basis:

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 OROs 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 FGl GENERAL EMERGENCY 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 ot Applicab le A. Core Cooling (C) A. An automatic or A. Operation of a second A. A leaki ng or Not Applicable CSF - ORANGE manual SI actuation is charging pump in the RUPTURED SO is entry conditions met required by EITHER normal charg ing FAUL TED outside of CNOTE I} of the fo ll owi ng: mode is required by contain ment.

1. UN!SOLA BLE EITHER of the RCS leakage following:

OR 1. UNlSOLABLE RCS leakage

2. SO tube RUPTURE. OR

2. SO tube leakage.

OR B. RCS Integrity (P)

CSF - RE D entry conditions met w ith RCS press > 300 psig. CNOTE I}

2. Inadequate Heat Remova l 2. Inadequate Heat Removal 2. Inadequate Heat Removal A. Core Cooling (C) A. Core Cooling (C) Not App licabl e A. Heat Sink (H) CSF - Not Applicable A. Core Cooling (C) CSF CSF - RED entry CSF-ORA GE RED entry conditions - RED entry conditions conditions met. entry conditions met. met. CNOTE l} met for 15 minutes or CNOTE 1} CNOTE l} longer. (NOTE 1}

OR 8 . Heat Sink (H) CSF -

RED entry condi tions 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 licable A. Post LOCA Radiation Not Appl icab le Not App licable A. Post LOCA Radi ation Rad iation 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/gm Dose Equivalent! 131 as determined per Procedure CS0925 .0 I, Reactor Coolant Post Accident Sampling.

4. Containment Integrity or Bypass 4. Containment Integrity or Bypass 4. Containment Integrity or Bypass Not Appl icable Not Applicable Not Appli cabl e Not App licable A. Containment isolation A. Containment (Z) CSF -

is required RED entry cond iti ons AND met. (NOTE 1)

EITHER of the OR fo llowing: B. Containment H2

l. Containment concentration 2'. 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 operating per OR design for 15 B. Indications of RCS minutes or longer.

leakage outside of containment.

5. STED/SED Jud gment 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 ind icates 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 ofCSFSTs as EALs

HU4 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:

• 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 an actuated alarm multiple zones (more than 1) actuated on CP-376 panel., combined with any of the following:

o GP 376 panel Multiple Zones Actuated o Plant Equipment Spuriously Operating o Containment Temperature Increasing 0

(1) a. 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 OR (2) a. Receipt of a single fire alarm (i.e., no other indications of a FIRE).

AND

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.

MS5 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. An automatic or manual trip did not shutdown the reactor to neutron flux< 5%.

AND

b. All manual actions to shutdown the reactor have been unsuccessful.

AND

c. EITHER of the following conditions exist:

Core Coolin C CSF RED entr conditions met.

Heat Sink (H CSF RED entr conditions met.

Basis:

This IC addresses a fai lure 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 mitigation 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. An automatic or manual trip did not shutdown the reactor to neutron flmt < 5%.

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 fMCB 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.

MU5 ECL: Notification of Unusual Event Initiating Condition: Automatic or manual trip fails to shutdown the reactor to AeutroA flmt < 5%.

Operating Mode Applicability: 1 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 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 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 llowing 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 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 wi ll quickly fal l to a level w ithin 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 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 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 JC MAS . Depending upon the plant response, escalation is also possible via IC FAI .

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 been exceeded, or will likely be exceeded.

(1) a. 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 AND

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 Basis:

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 the ability to obtain SAFETY SYSTEM parameters from wi thin 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 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) a. Release of a toxic, corrosive, asphyxiant or flammable gas into any Table HI rooms or areas.

AND

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 Basis:

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.

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 safety-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:

• 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.

(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.

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 safety-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 safety-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

cus 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:

Emergency Notification System (ENS)

Control Room/TSC telephones FTS telephones in the TSC Basis:

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.

MU6 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:

Emergency Notification System (ENS)

Control Room/TSC telephones FTS telephones in the TSC Basis:

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 OROs 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 SOLAB LE RCS leakage

2. SGtube 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 operating per OR 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

HU4 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:

• 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.

(1) a. 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 OR (2) a. Receipt of a single fire alarm (i.e., no other indications of a FIRE).

AND

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) 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.

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, 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 been exceeded, or will likely be exceeded. *

(1) a. 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 AND

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 Basis:

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/ RADIOLOGICAL EFFLUENT ICS/EALS

( 1) Dose rate greater than 15 mR/hr in ANY of the fo ll owing areas: ( 1) Dose rate greater than 15 mR/hr in ANY of the fo llowing areas:

• Contro l Room I Control Room I

• Central Alarm Station I Central Alarm Station (CAS) by survey I (2)

• (other site-specific areas/rooms)

An UN PLANNED event results in radiation levels that prohibit or I Secondary Alarm Station (SAS) by survey I impede access to any of the fo llowing plant rooms or areas: (2) An UNPLANNED event results in radi ati on levels that prohi bit or (site-specific list of plant rooms or areas w ith entry-re lated mode IM PEDE access to any of the fo llowing pl ant rooms or areas:

applicabil ity identified) [_able Hl Area Mode r imary Aux Building 25 ft elevation I 2 3 4 7 ft elevation

- 26 ft elevati2!]

ifurbine Building 21 ft elevation I, 2, 3 50 ft elevation

~ential Switchgear Rooms I, 2, 3, 4 Waste Process Bui lding 25 ft elevatio I, 2, 3

-3 ft elevatio Containment 3, 4 RHR/CBS Eg!!irment Vaults 3,4 Difference /Justification Table Hl : Site specific info rmati on, see V7 - Table Hl Procedure References 12

COLD SHUTDOWN/ REFUELING SYSTEM MALFUNCTION ICS/EALS CU2: INITIATING CONDITIONS NEl 99-01Rev6 Seabrook Station Nuclear Power Plant Loss of all but one AC power source to emergency buses for 15 minutes or longer. Loss of all but one AC power source to emergency buses fo r 15 minutes or longer.

Difference /Ju stification None THRESHOLDS NEC 99-01 Rev 6 Seabrook Station Nuclear Power Plant (1) AC power capab ility to (site-specific emergency buses) is reduced to a Note: For power restoration from the SEPS, both SEPS d iesel generator sets single power source for 15 minutes or longer. ust be functiona l.

AND ( 1) a. AC powe r capability to OT AC emergency buses 5 AND E6 is

b. Any addi tional single power source fa ilure will resul t in loss of all AC reduced to a sing le power source for 15 minutes or longer.

power to SAFETY SYSTEMS.

AND

b. A ny additiona l sing le power source fa ilure will res ul t in loss of all AC powe r 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 /Justification Added NOTE to clarify that both SEPS constitute a single power source. Added NOTE containing table of AC power sources per EPFAQ 2015-15.

26

COLD SHUTDOWN/ REFUELING SYSTEM MALFUNCTION ICS/EALS CUS: INITIATING CONDITIONS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant Loss of all onsite or offsite communications capabilities. Loss of all onsite or offsite communications capabilities.

Difference /Justification None THRESHOLDS NEJ 99-01 Rev 6 Seabrook Station Nuclear Power Plant

( 1) Loss of ALL of the following onsite communication methods: ( I) Loss of ALL of the following onsite communicat ion methods:

(site-specifi c li st of communications methods) I In-Plant (PBX) Tele_phones I (2) Loss of ALL of the following ORO communications methods: I Gai -Troni cs (s ite-specific li st of communications methods)

I (3) Loss of ALL of the following NRC communications methods:

I r 1ant Rad io System I (site-specific li st of communications methods) (2) Loss of ALL of the following ORO communications methods:

I !N uclear Alert System (i'JAfil I I ~ckup NAS I I Control Room/TSC telephones I I I (3) Loss of ALL of the following NRC communications methods:

I ~merge ncy Notification System (ENS) I I All plant telephones I I f TS telephones in the TSC I I I Diffe rence /Justification Prov ided site specific communications methods 29

INDEPENDENT SPENT FUEL STORAGE FACILITY {ISFSI) ICS/EALS EUl: I NITIATING CONDITIONS NEI 99-0 l Rev 6 Sea brook Station Nuclea r Power P la nt Damage to a loaded cask CONFINEMENT BOUNDARY. Damage to a loaded cask CONFINEMENT BOUNDARY.

Difference /Justification Non e THRESHOLDS NEI 99-0 l Rev 6 Seabrook Station Nuclea r Power P la nt

( 1) Damage to a loaded cask CONFlNEMENT BOUNDAR Y as ( I) Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation read ing greater than (2 ti mes the indicated by ANY of the fo llowing on-contact surface rad iation site-specific cask spec ific technical specification all owabl e rad iation readings greater than:

level) on the surface of the 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 exterion Difference /J ustification Added NOTE pull ed from the bas is all owing calcul ating surface dose from a di stance dose EUl. l: Site specific information, see V I 9 NU HOMS HSM Dose Rates Technical Specification 30

FISSION PRODUCT BARRIER ICS/EALS B. RCS cooldown rate greater than (site-specific pressurized thermal shock criteria/limits defined by site-specific indications).

Seabrook Station Nuclear Power Plant Not Applicab le A. Core Coo ling A. An automatic or A. Operation of a second A. A leaking or Not Applicable (C) CSF - ORANGE manual SI actuation is chargi ng pump in the RUPTURED SG is entry cond iti ons met required by EITHER of normal charging mode FAULTED outside of (NOTE I) the fo llowing: is required by EITHER containment.

• UNISOLABLE of the fo llowing :

RCS leakage I. UNISOLABLE OR RCS leakage

• SG tube OR RUPTURE.

2. SG tube leakage.

OR B. CS Integri ty (P) CS

- RED entry conditions met with RCS 2 ress >

~ oo sig. (NOTE I)

Difference /Justification Fuel Clad Barrier Potential Loss LA: Site specific information , see V20 CSFST Core Cooling RCS Barrie r Potential Loss 1.B: Site spec ific information , see V2 l CSFST lntegrity 32

FISSION PRODUCT BARRIER ICS/EALS NEI 99-01 Rev 6

2. Inadequate Heat Remova l 2. I nadequate Heat Remova l 2. I nadeq ua te Heat Re mova l A. Core exit A. Core ex it Not Applicable A. Inadequate RCS heat Not Applicable A. I. (Site-specific thermocouple thermocouple removal capability via criteria for readings greater readings greater steam generators as entry into core than (site-specific than (site-specific indicated by (site- cooling temperature temperature specific indications). restoration value). value). procedure)

OR AND B. lnadequate RCS

2. Restoration heat removal procedure not capabi li ty via effective within steam generators 15 minutes.

as indicated by (site-specific indications).

Seabrook Sta tion Nuclear Power Plant A. Core Cooling (C) A. Core Cooling (C) Not Applicab le A. eat Sink (H) CSF - Not Applicab le A. Core Cooling (C)

CSF - RED entry CSF - ORANG ED entry conditions CSF - RED entry conditions met. entry cond itions met. (NOTE I) conditions met for (NOTE I) met.(NOTE I) 15 minutes or OR onger. (NOTE I)

8. eat Sink (H)

CSF - RED entry conditions met.

(NOTE I)

Differe nce /J ustification Fuel C lad Ba r r ier: Loss 2.A, Potentia l Loss 2.A and Containment Barrier Potentia l Loss 2.A: Site specific information, see V20 CSFST Core Cooling RCS Ba rrier: Potentia l Loss 2.A: Site specific information, see V22 CSFST Heat Sink 33

FISSION PRODUCT BARRIER ICS/EALS Not Applicable Not Applicabl e Not Applicable Not Appli cable A. Conta inment isolati on is A. Containment (Z) required CSF - RED entry AN D conditions met.

(NOTE I)

E ITHER of the following: OR

3. Contai nment B. Cnmt. hydrogen integri ty has been concentrati on ~ 6%

lost based on OR STED/SED C. I. Containmen ~

judgment.

pressure > 18 OR sig

4. UNISOLA BLE AND pathway from the

2. Less than one containment to the full train o environment exists.

Cnmt. Building OR S ray (CBS) is B. Indications of RCS operating per leakage outside of des ign for 15 containment. minutes or longer.

Difference /J ustification Conta inment Barrier: Potential Loss 4.A: Site specific in fo rmation, see V24 CSFST Containment Containment Barrier: Potential Loss 4.8: Site spec ifi c info rmation, see V 14 H2 concentration in con tainment Containment Barrier: Potential Loss 4.Cl : Site spec ific informati on, see V25 Conta inment Spray Setpoint NEI 99-01 R ev 6

5. Other Indications 5. Other Indications 5. Oth e r Indications A. (site-specific as A. (site-spec ifi c as A. (s ite-specific as A. (s ite-specific as A. (s ite-specific as A. (site-spec ific as applicabl e) appl icabl e) appli cabl e) applicable) app licabl e) applicable)

Seabrook Station Nuclear Power Plant 36

FISSION PRODUCT BARRIER ICS/EALS Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Not Applicable Difference /J ustification None NEI 99-01 Rev 6

6. Emergency Director Judgment 6. Emergency Director Judgment 6. Emergency Director Judgment 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 ion of the the opi nion of the opinion of the opin ion of the opinion of the the opinion of the Emergency Emergency Emergency D irector Emergency Director Emergency Director that Emergency Director Director that Director that that 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 Loss of the Fuel Barrier. Containment Barrier.

Barrier. Clad Barrier.

Seab rook Station Nuclear Power Plant 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 ion 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 indicates Loss of the indicates Potential Loss indicates Loss of the indicates Potential the Fuel C lad Loss of the Fuel RCS Barrier. of the RCS Barrier. Containment Barrier. Loss of the Barrier. Clad Barrier. Containment Barrier.

NOTE I : Refer to ER 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 ACTION resulting in loss of physical control of the facility. DELETED 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 PROTECTED AREA as reported by the (site-specific security sh ift supervision).

AND

b. EITHER of the following has occurred :

1. ANY of the fo llowi ng safety functions cannot be controlled or mainta ined.

• Reactiv ity control

• Core cooling [PWR] I RPV water level [BWR]

• RCS heat removal OR

2. Damage to spent fuel has occurred or is IMMINENT.

Difference /Justificat ion HG I deleted 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. Release of a toxic, corrosive, asp hyxiant or flammab le gas ( 1) a. Release of a toxic, corrosive, asphyxiant or fl ammable gas into any of the fo llowing p lant rooms or areas: into any able HI rooms or areas.

(s ite-spec ific list of plant roo ms or areas w ith entry-re lated AND mode app licability identified) b. Entr into the room or area is rohibited or im eded.

AND able HI

b. Entry into the room or area is prohibited or impeded. Area ode Primary Aux Building 25 ft elevatio I, 2, 3, 4 7 ft elevatio

- 26 ft elevation ifurbine Building 21 ft elevatio 1, 2, 3 ssential Switchgear Rooms I, 2, 3, 4 Waste Process Building 25 ft elevatio I , 2, 3 3,4 3, 4 Difference /Justification HAS.lb: Site spec ific info rmation, see V7 - Table HJ Procedure References 45

HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS

( 1) a. A FIRE is NOT extinguished with in 15-minutes of ANY of NOTE: A containment fire alarm is considered valid upon recei t of multi le the fo llowing FIRE detection indications: ones more than I) actuated alarm on CP 376.

• Report fro m the fie ld (i.e., visual observation) ( 1) a. A FIRE is NOT extinguished with in 15-m inutes of ANY of the

• Recei pt of mul tiple (more than I) fire alarms or fo llowin FIRE detection ind ications:

indications Report from the fi eld (i .e., visual observation)

• Field verification of a single fire alarm Receipt of multi ple (more than 1) fire alarms or indications AND Field verification of a single fire alarm

b. The FIRE is located within ANY of the fo ll owing plant AN D rooms or areas: b The FIRE is located within ANY Table H2 olant rooms or areas:

(site-specific list of plant rooms or areas) ff able H2 (2) a. Receipt of a single fire alarm (i.e., no other indi cations of a Condensate Stora!!.e Tank Enclosure f uel Stora!!.e Buildin2!

Containmenv Primarv Auxi liarv Buildinlli FIRE) .

Control BuildinE!I Service Water Pump House AN D Cooling Tower Steam and Feedwater Pi

b. The FIRE is located within AN Y of the fo llowing plant Chases rooms or areas: Diesel Generator Buildin!!l North Tank Fam1 (site-specific list of pl ant rooms or areas) ~mergency Feedwater Pumo House Startuo Feedwater Pump Area AN D !Equipment Vault

c. The existence of a FIRE is not verified within 30-minutes of alarm receipt. (2) a. Receipt of a single fi re alarm (i.e., no other indications of a (3) A FIRE with in the plant or JSFSJ [for plants with an JSFSJ outside FIRE).

the plant Protected Area] PROTECTED AREA not extinguished AN D within 60-minutes of the initial report, alarm or indication. b. The FIRE is located within ANY of the ifable H2 plant rooms or (4) A FIRE withi n the pl ant or JSFSJ [for p lants with an JSFSJ outside areas except Contai nment in Modes 1 and 2. (see note above) the plant Protected Area] PROTECTED AREA that requires AN D firefi ghting support by an offsite fire response agency to extinguish. c. The existence of a FIRE is not verified within 30-minutes of alarm recei pt.

(3) A FIRE within the plant PROTECTED AREA or Dry Fuel Storage Facility not exti nguished with in 60-minutes of the initial report, alarm or indication.

(4) A FIRE withi n the plant PROTECTED AREA or Dry Fuel Storage Facility that requires firefi ghting support by an offsite fire response agency to extinguish.

Difference /Justification 52

HAZARDS AND OTHER CONDITIONS AFFECTING PLANT SAFETY ICS/EALS Added NOTE to clarify the contai nment fire alarm HU4.lb: Site specific information, see V28 Verification of Fire Areas HU4.2b: Containment is excepted in Modes I and 2 but is covered by the second note. EALl would be app licable. Entry into the Containment to perform verifications within 30 minutes in Modes 1 and 2 is a challenge.

53

SYSTEM MALFUNCTIONS MSS: INITIATING CONDITIONS NEI 99-01 Rev 6 Seabrook Station N uclear Power P lant Inabil ity to shut down the reactor causing a challenge to (core cooling [PWR] Inabil ity to shutdown the reactor causing a challenge to core cooling or RCS I RP V water leve l [BWR]) or RCS heat remova l. heat removal.

Difference /Justifica tion THRESHOLDS N EI 99-01Rev 6 Seabrook Station Nuclear Power Plant

( I) a. An automatic or manual (trip [PWR] I scram [BWR]) did not ( I) a. An automatic or manual tri p did not shutdown the reactor shutdown the reactor. AND

b. All manual action s to shutdown the reactor have been AND un successful.

AND

b. All man ual actions to shut down the reactor have been c. EITHER of the fo ll owing conditi ons ex ist:

un successfu l. I Core Coo ling (C) CSF - RED entry conditions met. I I ~eat Sink (H) - RED entry conditions met. I AND

c. EITHER of the fo llowing conditi ons ex ist:

• (S ite-specifi c indication of an inabi lity to adequately remove heat from the core)

• (S ite-spec ifi c ind ication of an inab ility to adequately remove heat fro m the RCS)

Diffe rence /Justifica tion MSS.lc: Site spec ific information, see V20 CSFST Core Cooli ng and V22 CSFST Heat Sink 58

SYSTEM MALFUNCTIONS MAl: INITIATING CONDITIONS NEJ 99-01 Rev 6 Seabrook Station Nuclear Power Plant Loss of all but one AC power source to emergency buses for 15 minutes or Loss of all but one AC power source to emergency buses for 15 minutes or longer. longer.

Difference /Justification None THRESHOLDS NEl 99-01 Rev 6 Seabrook Station Nuclear Power Plant

( 1) a. AC power capability to (s ite-specific emergency buses) is ( 1) a. AC power capability to OTH AC emergency buses 5 AN D E6 is reduced to a single power so urce for 15 minutes or longer. reduced to a single power source for 15 minutes or longer.

AND

b. Any additional sing le power source failure wi ll result in a AND loss 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 event resu lts 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 within the Contro l Control Room for 15 minutes or longer. Room for 15 minutes or longer.

Reactor Power fBWR parameter listl fPWR parameter listl RCS Level Reactor Power Reactor Power RCS Pressure RPV Water Level RCS Level Core Exit Temperature RPV Pressure RC S Pressure Levels in at least two steam generators Primary Conta inment In-Core/Core Exit Steam Generator Emergency Feed Water Flow Pressure Temperature AND Suppression Pool Leve l Levels in at least (site- b. ANY of the fo llowing transient events in progress.

specific number) steam Automatic or manual run back greater than 25%

generators thermal reactor power Suppression Pool Steam Generator Auxiliary Electrical load rejection greater than 25% full Temperature or Emergency Feed Water electrical load Flow I Reactor trip I AND I ECCS (SI) actuation I

b. ANY of the fo llowing transient events in progress.

• Automatic or manual run back greater than 25%

therma l reactor power

• Electrical load rejection greater than 25% full electrical load

• Reactor scram [B WR] I trip [PWR]

• ECCS (SI) actuation

• Thermal power osci llation s greater than 10% [BWR]

Difference /Justification Non e 62

SYSTEM MALFUNCTIONS MAS: INITIATING CONDITIONS NEr 99-01 Rev 6 Seabrook Station Nuclear Power Plant Automatic or manua l (trip [PWR] I scram [BWR]) fails to sh ut down the Automatic or manual trip fails to shutdown the reactor and subsequent reactor, and subsequent manual actions taken at the reactor control conso les manual actions taken at the Main Control Board are not successful in shutting are not successful in shutting down the reactor. down the reactor.

Difference /Justification None THRESHOLDS NEI 99-01Rev6 Seabrook Station Nuclear Power Plant (I) a. An automatic or manual (trip [PWR] I scram [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 actions taken at the reactor control conso les are not successful in shutting down the reactor.

D ifference /Justification 63

SYSTEM MALFUNCTIONS MU2: INITIATING CONDITIONS NEl 99-01 Rev 6 Seabrook Station Nuclear Power Plant UNPLANNED loss of Contro l Room indications for 15 minutes or longer. UNPLANNED loss of Control Room indications for 15 minutes or longer.

Difference /Justification 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. An 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 longer.

Reactor Power

[BWR parameter listl [PWR parameter 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 Steam Generator Emergency Feed Water Flow Primary Containment In-Core/Core Exit Pressure Temperature Suppression Pool Level Levels in at least (site-specific number) steam generators Suppression Pool Steam Generator Temperature Auxiliary or Emergency Feed Water Flow Difference /Justification None 66

SYSTEM MALFUNCTIONS MUS: INITIATING CONDITIONS NEl 99-01 Rev 6 Seabrook Station Nuclear Power Plant Automatic 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 shutdown (I) a. An automatic trip did not shutdown the reactor the reactor. AND AND b. A subsequent manual action taken at the MCB is successful 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 subsequent manual action taken at the CB is successful

1. A subsequent manual action taken at the reactor control in shutting down the reactor.

conso les is successful in sh utting down the reactor. OR OR 2. A subsequent automatic trip is successful in shutting down

2. A subsequent automatic (trip [PWR] I scram [BWR]) is the reactor.

successful in shutting down the reactor.

Diffe rence /Justification 69

SYSTEM MALFUNCTIONS MU6: INITIATING CONDITIONS NEI 99-0 1 Rev 6 Sea brook Station Nuclear Power Plant Loss of all onsite or offsite communications capabil ities. Loss of all ons ite or offsite commun ications capab ili ties.

Difference /Justification Non e THRESHOLDS NEI 99-01 Rev 6 Seabrook Station Nuclear Power Plant

( 1) Loss of ALL of the fo llowi ng onsite comm unication methods: ( I) Loss of ALL of the fo llowi ng onsite communication methods:

(site-spec ific Iist of commun ications methods) I [ n Plant (PBX) Telt'.j?hones I (2) Loss of ALL of the fo llow ing ORO co mm unications methods: I Gai Tronics (site-spec ific list of commun ications methods)

I (3) Loss of ALL of the fo llowing NRC com munications methods: I Plant Radio Systell] I (site-spec ific list of commun ications methods) (2) Loss of ALL of the fo llowing ORO communi cations methods:

I Nuclear Alert System (NAS) I I 1Backup NAS I I Control Room/ TSC plant telephones I I I (3) Loss of ALL of the fo ll owi ng NRC communications methods:

I ~mergency Notification System @NS) I I All plant telephones I I f TS telephones in the TSC I I I Difference /J ustification Provided site spec ifi c comm unicati ons methods 70 L