Text

Issuance of Amendment No. 174 to Revise TS 3.5.1, ECCS-Operating, for One-Time Extension of High Pressure Core Spray Completion Time & Associated Surveillance Requirement (Emergency Circumstances)
	ML18342A015
Person / Time
Site:	Nine Mile Point
Issue date:	12/09/2018
From:	Marshall M; Plant Licensing Branch 1
To:	Bryan Hanson; Exelon Generation Co, Exelon Nuclear
	Marshall, M, NRR/DORL/LPL1
References
	EPID L-2018-LLA-0491
	Download: ML18342A015 (30)
	v • d • e

OFFICIAL USE ONLY PROPRIETARY INFORMATION UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 Mr. Bryan C. Hanson Senior Vice President Exelon Generation Company, LLC President and Chief Nuclear Officer Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555 December 9, 2018

SUBJECT:

NINE MILE POINT NUCLEAR STATION, UNIT 2 - ISSUANCE OF AMENDMENT NO. 174 TO REVISE TECHNICAL SPECIFICATION 3.5.1, "ECCS - OPERATING," FOR A ONE-TIME EXTENSION TO THE HIGH PRESSURE CORE SPRAY COMPLETION TIME AND ASSOCIATED SURVEILLANCE REQUIREMENTS (EMERGENCY CIRCUMSTANCES)

(EPID L-2018-LLA-0491)

Dear Mr. Hanson:

The U.S. Nuclear Regulatory Commission (the Commission) has issued the enclosed Amendment No. 17 4 to Renewed Facility Operating License No. NPF-69 for the Nine Mile Point Nuclear Station, Unit 2. The amendment consists of changes to the Technical Specifications in response to your application dated December 6, 2018 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML18340A142), as supplemented by letter dated December 7, 2018 (ADAMS Accession No. ML18341A343).

The amendment modifies the Nine Mile Point Nuclear Station, Unit 2, Technical Specification 3.5.1, "ECCS [Emergency Core Cooling System] - Operating," for a one-time extension of the high pressure core spray completion time. Specifically, the amendment revises the completion time for an inoperable high pressure core spray system from 14 days to 35 days.

Additionally, the amendment allows extending the completion of several surveillance requirements of equipment being protected during the replacement of the high pressure core spray diesel generator.

This license amendment is issued under emergency circumstances as provided in the provisions of paragraph 50.91 (a)(5) of Title 10 of the Code of Federal Regulations due to the time-critical nature of the amendment. In this instance, an emergency situation exists in that the amendment is needed to allow the licensee to avoid a plant shutdown.

NOTICE: Enclosure 3 to this letter contains Proprietary Information. Upon separation from Enclosure 3, this letter is DECONTROLLED.

OFFICIAL USE ONLY PROPRIETARY INFORMATION

B. Hanson OFFICIAL USE ONLY PROPRIETARY INFORMATION A copy of the related safety evaluation is also enclosed. The safety evaluation describes the emergency circumstances under which the amendment is issued and the final no significant hazards determination. Notice of Issuance addressing the final no significant hazards determination and opportunity for a hearing associated with the emergency circumstances will be included in the Commission's biweekly Federal Register notice.

Docket No. 50-410

Enclosures:

1. Amendment No. 174 to NPF-69

2. Safety Evaluation (Non-Proprietary)

3. Safety Evaluation (Proprietary) cc: Listserv Sincerely, Michael L. Marshall, Jr., Senior Project Manager Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation OFFICIAL USE ONLY PROPRllaTARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, O.C. 20555-0001 NINE MILE POINT NUCLEAR STATION, LLC LONG ISLAND LIGHTING COMPANY EXELON GENERATION COMPANY. LLC DOCKET NO. 50-410 NINE MILE POINT NUCLEAR STATION, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 17 4 Renewed License No. NPF-69

1.

The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Exelon Generation Company, LLC (Exelon, the licensee) dated December 6, 2018, as supplemented by letter dated December 7, 2018, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 1 O CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Renewed Facility Operating License No. NPF-69 is hereby amended to read as follows:

OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION (2)

Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, as revised through Amendment No. 17 4, are hereby incorporated into this license. Exelon Generation shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of the date of its issuance and shall be implemented prior to the expiration of the 14-day completion time of Required Action B.2 of Limiting Condition for Operation 3.5.1 in the Technical Specifications, but no later than December 10, 2018.

4.

The nine compensensatory actions listed in the application dated December 6, 2018, as supplemented by letter dated December 7, 2018, shall be implemented, to the extent practical, as described in the application.

FOR THE NUCLEAR REGULATORY COMMISSION

<C)~~--

Attachment:

Changes to the Renewed Facility Operating License and Technical Specifications Date of Issuance: December 9, 2018 Jlmes G. Danna, Chief Plant Licensing Branch I Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation OFFICIAL USE ONLY PROPRIETARY INFORMATION

OFFICIAL USE ONLY PROPRIETARY INFORMATION ATTACHMENT TO LICENSE AMENDMENT NO. 174 NINE MILE POINT NUCLEAR STATION, UNIT 2 RENEWED FACILITY OPERATING LICENSE NO. NPF-69 DOCKET NO. 50-410 Replace the following page of the Renewed Facility Operating License with the attached revised page. The revised page is identified by amendment number and contains a marginal line indicating the area of change.

Remove Page 4

Insert Page 4

Replace the following pages of the Appendix A, Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines indicating the areas of change.

Remove Pages 3.5.1-1 3.3.5.1-8 3.5.1-4 3.5.3-2 3.8.1-6 3.8.1-7 3.8.1-14 3.8.1-16 3.8.3-3 Insert Pages 3.5.1-1 3.3.5.1-8 3.5.1-4 3.5.3-2 3.8.1-6 3.8.1-7 3.8.1-14 3.8.1-16 3.8.3-3

(1) Maximum Power Level Exelon Generation is authorized to operate the facility at reactor core power levels not in excess of 3988 megawatts thermal (100 percent rated power) in accordance with the conditions specified herein.

(2)

Technical Specifications and Environmental Protection Plan (3)

The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, both of which are attached hereto, as revised through Amendment No. 174, are hereby incorporated into this license. Exelon Generation shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

Fuel Storage and Handling (Section 9.1.SSER 4)*

a.

Fuel assemblies, when stored in their shipping containers, shall be stacked no more than three containers high.

b.

When not in the reactor vessel, no more than three fuel assemblies shall be allowed outside of their shipping containers or storage racks in the New Fuel Vault or Spent Fuel Storage Facility.

c.

The above three fuel assemblies shall maintain a minimum edge-to-edge spacing of twelve ( 12) inches from the shipping container array and approved storage rack locations.

d.

The New Fuel Storage Vault shall have no more than ten fresh fuel assemblies uncovered at any one time.

(4)

Turbine System Maintenance Program (Section 3.5.1.3.10 SER)

The operating licensee shall submit for NRG approval by October 31, 1989, a turbine system maintenance program based on the manufacturer's calculations of missile generation probabilities.

(Submitted by NMPG letter dated October 30, 1989 from G.D. Terry and approved by NRG letter dated March 16, 1990 from Robert Martin to Mr. Lawrence Burkhardt, Ill).

The parenthetical notation following the title of many license conditions denotes the section of the Safety Evaluation Report (SER) and/or its supplements wherein the license condition is discussed.

Renewed License No. NPF-69 Amendment 117through 140,141,143,144,145,146,147,150,151,152,154,156,157,158,159,160,161, 163,164,165,166,167,168,169,170,172,174

ECCS - Operating 3.5.1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), RPV WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.5.1 ECCS - Operating LCO 3.5.1 Each ECCS injection/spray subsystem and the Automatic Depressurization System (ADS) function of six safety/relief valves shall be OPERABLE.

APPLICABILITY:

MODE 1, ACTIONS MODES 2 and 3, except ADS valves are not required to be OPERABLE with reactor steam dome pressure s 150 psig.

NOTE-------------------------------------------------------

LCO 3.0.4.b is not applicable to HPCS.

CONDITION REQUIRED ACTION COMPLETION TIME A.

One low pressure ECCS A.1 Restore low pressure 7 days injection/spray ECCS injection/spray subsystem inoperable.

subsystem to OPERABLE status.

B.

High Pressure Core B.1 Verify by administrative Immediately Spray (HPCS) System means RCIC System is inoperable.

OPERABLE when RCIC is required to be OPERABLE.

AND 8.2 Restore HPCS System 14 days*

to OPERABLE status.

{continued)

A one-time change to this Completion Time from 14 days to 35 days due to the HPCS DG replacement has been approved via emergency license amendment request This Completion Time expires on 12/31/2018 at 0100.

NMP2 3.5.1-1 Amendment 94,-4, OQ,~. 17 4

ECCS Instrumentation 3.3.5.1 SURVEILLANCE REQUIREMENTS

NOTES---------------------------------------------------------

1.

Refer to Table 3.3.5.1-1 to determine which SRs apply for each ECCS Function.

2.

When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Function or the redundant Function maintains ECCS initiation capability.

SURVEILLANCE FREQUENCY SR 3.3.5.1.1 Perform CHANNEL CHECK.

In accordance with the Surveillance Frequency Control Program SR 3.3.5.1.2 Perform CHANNEL FUNCTIONAL TEST.

In accordance with the Surveillance Frequency Control Program*

SR 3.3.5.1.3 Calibrate the trip unit.

In accordance with the Surveillance Frequency Control Program SR 3.3.5.1.4 Perform CHANNEL CALIBRATION.

In accordance with the Surveillance Frequency Control Program SR 3.3.5.1.5 Perform CHANNEL CALIBRATION.

In accordance with the Surveillance Frequency Control Program' SR 3.3.5.1.6 Perform LOGIC SYSTEM FUNCTIONAL TEST.

In accordance with the Surveillance Frequency Control Prooram*

'Following return to OPERABILITY of the HPCS System, the past due Surveillances will be completed by January 11, 2019.

NMP2 3.3.5.1-8 Amendment B~~;-498, 174

SURVEILLANCE REQUIREMENTS SR 3.5.1.1 SR 3.5.1.2 SR 3.5.1.3 SURVEILLANCE Verify, for each ECCS injection/spray subsystem, locations susceptible to gas accumulation are sufficiently filled with water.

NOTE-----------

Not required to be met for system vent paths opened under administrative control.

Verify each ECCS injection/spray subsystem manual, power operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.

Verify:

a.

For each ADS nitrogen receiver discharge header, the pressure is 2: 160 psig; and

b.

For each ADS nitrogen receiver tank, the pressure is 2: 334 psig.

ECCS - Operating 3.5.1 FREQUENCY In accordance with the Surveillance Frequency Control Program*

In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program

( continued)

Following return to OPERABILITY of the HPCS System, the past due Surveillances will be completed by January 11, 2019.

NMP2 3.5.1-4 Amendment ~a0~~;-4-70,, 17 4

SURVEILLANCE REQUIREMENTS SR 3.5.3.1 SR 3.5.3.2 SR 3.5.3.3 SR 3.5.3.4 SURVEILLANCE Verify the RCIC System locations susceptible to gas accumulations are sufficiently filled with water.

NOTE-------------------------------

Not required to be met for system vent flow paths opened under administrative control.

Verify each RCIC System manual, power operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.

NOTE---------------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure and flow are adequate to perform the test.

Verify, with reactor pressure s 1035 psig and~ 935 psig, the RCIC pump can develop a flow rate ~ 600 gpm against a system head corresponding to reactor pressure.

NOTE -----------------------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure and flow are adequate to perform the test.

Verify, with reactor pressure s 165 psig, the RCIC pump can develop a flow rate

~ 600 gpm against a system head corresponding to reactor pressure.

RCIC System 3.5.3 FREQUENCY In accordance with the Surveillance Frequency Control Program*

In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program (continued)

Following return to OPERABIL TY of the HPCS System, the past due Surveillances will be completed by January 11, 2019.

NMP2 3.5.3-2 Amendment 94,4.W,4~. 174

SURVEILLANCE REQUIREMENTS continued SURVEILLANCE AC Sources - Operating 3.8.1 FREQUENCY SR 3.8.1.2

*-------------------NO TE -----------------------------

All DG starts may be preceded by an engine prelube period and followed by a warmup period prior to loading.

Verify each required DG starts from standby conditions and achieves:

a.

In s 10 seconds, voltage ~ 3950 V for Division 1 and 2 DGs and ~ 3820 V for Division 3 DG, and frequency~ 58.8 Hz for Division 1 and 2 DGs and~ 58.0 Hz for Division 3 DG; and

b.

Steady state voltage ~ 3950 V and s 4370 V and frequency ~ 58.8 Hz and s61.2Hz.

Following return to OPERABILITY of the HPCS System, the past due Surveillances will be completed by January 18, 2019.

NMP2 3.8.1-6 In accordance with the Surveillance Frequency Control Program*

continued Amendment 94;-4~. 174

SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE AC Sources - Operating 3.8.1 FREQUENCY SR 3.8.1.3

NOTES ---------------------------

SR 3.8.1.4 SR 3.8.1.5 SR 3.8.1.6

1.

DG loadings may include gradual loading as recommended by the manufacturer.

2.

Momentary transients outside the load range do not invalidate this test.

3.

This SuNeillance shall be conducted on only one DG at a time.

4.

This SR shall be preceded by, and immediately follow, without shutdown, a successful performance of SR 3.8.1.2.

Verify each required DG is synchronized and loaded and operates for~ 60 minutes at a load ~ 3960 kW and ::; 4400 kW for Division 1 and 2 DGs, and ~ 2340 kW and s; 2600 kW for Division 3 DG.

Verify each required day tank contains

~ 403 gal of fuel oil for Division 1 and 2 DGs and ~ 282 gal for Division 3 DG.

Check for and remove accumulated water from each required day tank.

Verify each required fuel oil transfer subsystem operates to automatically transfer fuel oil from the storage tank to the day tank.

Following return to OPERABILITY of the HPCS System, the past due Surveillances will be completed by January 18, 2019.

NMP2 3.8.1-7 In accordance with the SuNeillance Frequency Control Program*

In accordance with the SuNeillance Frequency Control Program*

In accordance with the SuNeillance Frequency Control Program' In accordance with the SuNeillance Frequency Control Program*

(continued)

Amendment 94,-4~. 174

SURVEILLANCE REQUIREMENTS continued SURVEILLANCE AC Sources - Operating 3.8.1 FREQUENCY SR 3.8.1.13

NOTES ---------------------------

1.

This Surveillance shall be performed within 5 minutes of shutting down the DG after the DG has operated 2'. 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months loaded ~ 3960 kW for Division 1 and 2 DGs, and~ 2340 kW for Division 3 DG.

Momentary transients below the load limit do not invalidate this test.

2.

All DG starts may be preceded by an engine prelube period.

Verify each required DG starts and achieves:

a.

In :5 10 seconds, voltage ~ 3950 V for Division 1 and 2 DGs and ~ 3820 V for Division 3 DG, and frequency 2'. 58.8 Hz for Division 1 and 2 DGs and~ 58.0 Hz for Division 3 DG; and

b.

Steady state voltage ~ 3950 V and

5 4370 V and frequency~ 58.8 Hz and :-::: 61.2 Hz.

Following return to OPERABILITY of the HPCS System, the past due Surveillances will be completed by January 18, 2019.

NMP2 3.8.1-14 In accordance with the Surveillance Frequency Control Program*

continued Amendment 94.,4S2, 174

AC Sources - Operating 3.8.1 SURVEILLANCE REQUIREMENTS continued SR 3.8.1.16 SURVEILLANCE

NOTE----------------------------

This Surveillance shall not normally be performed in MODE 1, 2, or 3. However, this Surveillance may be performed to reestablish OPERABILITY provided an assessment determines the safety of the plant is maintained or enhanced.

Credit may be taken for unplanned events that satisfy this SR.

Verify interval between each sequenced load block, for the Division 1 and 2 DGs only, is ~ 90% of the design interval for each automatic load sequence time delay relay.

FREQUENCY In accordance with the Surveillance Frequency Control Program"'

continued

Following return to OPERABILITY of the HPCS System, the past due Surveillances will be completed by January 11, 2019.

NMP2 3.8.1-16 Amendment Q-t,1~,-1ea, 174

Diesel Fuel Oil, Lube Oil, and Starting Air 3.8.3 SURVEILLANCE REQUIREMENTS SR 3.8.3.1 SR 3.8.3.2 SR 3.8.3.3 SR 3.8.3.4 SR 3.8.3.5 SURVEILLANCE Verify each fuel oil storage tank contains:

a.

~ 50,000 gal of fuel for Division 1 DG and Division 2 DG; and

b.

~ 35,342 gal of fuel for Division 3DG.

Verify lube oil inventory is:

a.

~ 99 gal for Division 1 DG and Division 2 DG; and

b.

~ 168 gal for Division 3 DG.

Verify fuel oil properties of new and stored fuel oil are tested in accordance with, and maintained within the limits of, the Diesel Fuel Oil Testing Program.

Verify each DG air start receiver pressure is:

a.

~ 225 psig for Division 1 DG and Division 2 DG; and

b.

~ 190 psig for Division 3 DG.

Check for and remove accumulated water from each fuel oil storage tank.

FREQUENCY In accordance with the Surveillance Frequency Control Program*

In accordance with the Surveillance Frequency Control Program In accordance with the Diesel Fuel Oil Testing Program In accordance with the Surveillance Frequency Control Program In accordance with the Surveillance Frequency Control Program'

Following return to OPERABILITY of HPCS System, the past due Surveillances will be completed by January 18, 2019.

NMP2 3.8.3-3 Amendment~2. 174

ENCLOSURE 2 SAFETY EVALUATION RELATED TO AMENDMENT N0.174 NINE MILE POINT NUCLEAR STATION, UNIT 2 Proprietary information pursuant to Section 2.390 of Title 10 of the Code of Federal Regulations has been redacted from this document.

Redacted information is identified by blank space enclosed within double brackets ((( ))).

1.0 OFFICl.",L USE ONLY PROPRIETARY INFORMATION UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555--0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 174 TO RENEWED FACILITY OPERATING LICENSE NO. NPF-69 NINE MILE POINT NUCLEAR STATION. LLC LONG ISLAND LIGHTING COMPANY EXELON GENERATION COMPANY. LLC.

NINE MILE POINT NUCLEAR STATION. UNIT 2 DOCKET NO. 50-410 INTRODUCTION By letter dated December 6, 2018 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML18340A142), as supplemented by letter dated December 7, 2018 (ADAMS) Accession No. ML18341A343), pursuant to Title 10 of the Code of Federal Regulations (10 CFR) Section 50.90, Exelon Generation Company, LLC (Exelon, the licensee) submitted a license amendment request (LAR, application) for the Nine Mile Point Nuclear Station, Unit 2 (Nine Mile Point 2 or NMP Unit 2). The proposed amendment would allow for a one-time change to extend the completion time for an inoperable high pressure core spray (HPCS) in Section 3.5.1, "ECCS [Emergency Core Cooling System] - Operating," of the Nine Mile Point 2 Technical Specifications (TSs). Specifically, the application proposed to revise the completion time for an inoperable HPCS system from 14 days to 35 days. Additionally, the amendment would allow extending the completion of several surveillance requirements (SRs) of components or systems that are being protected during the replacement of the high pressure core spray diesel generator (DG).

The licensee is proposing the changes to the Nine Mile Point 2 TSs because Exelon needs to replace the HPCS DG at Nine Mile Point 2. In its application dated December 6, 2018, the licensee stated, in part:

On November 26, 2018, at 0100, NMP Unit 2 declared the High-Pressure Core Spray (HPCS) system inoperable and entered a 14 day Limiting Condition for Operation, under TS 3.5.1, to perform eighteen-year preventive maintenance of the HPCS Diesel Generator (DG) which included power pack inspections, cleaning lube oil coolers, fuel injector replacement, and additional eighteen-year OFFICIAL USE ONLY PROPRIETARY INFORMATION

2.0 2.1 OFFICIAL USE ONLY PROPRIETARY INFORMATION maintenance. A note in TS LCO [Limiting Condition for Operation] 3.8.1 allows taking exception to the Applicability requirements for HPCS Division 3 power sources, provided the HPCS System is declared inoperable (Reference TS LCO 3.8.1 Bases).

At 0518 on December 2, 2018, during post maintenance testing of the HPCS DG, the Main Control Room was notified by an Operator to shut down the HPCS DG.

There was sparking (from metal to metal contact) and smoke visible along with other noted damage to the HPCS DG. The DG was safely shutdown from the Main Control Room. The initial inspection of the HPCS DG found lower casing covers for cylinders 3 and 13 were off, a piston had come loose inside the diesel crankcase and that one piston connecting rod was broken and another piston connecting rod was bent. Additional signs of internal damage were also noted.

Upon further investigation it was determined that based on the extent of the damage to the HPCS DG, it requires replacement, which is expected to take 21 days to complete, including post-maintenance testing.

REGULA TORY EVALUATION Description of System Nine Mile Point 2 employs a nuclear steam supply system consisting of a single-cycle, forced circulating General Electric (GE) Boiling Water Reactor. The rated core thermal power level is 3,988 megawatts thermal (MWt). The primary containment is a GE Mark II type structure consisting of the drywell; the pressure suppression chamber, which stores a large volume of water; and the drywell floor, which separates the drywell and suppression chamber.

The alternating current distribution system consists of three divisional load groups: Divisions 1, 2, and 3, with each division powered by an independent Class 1 E 4.16 kilovolt (kV) bus. In the event of a loss of the normal onsite source of power in Divisions 1 and 2, the 4.16 kV buses each have one separate and independent source of offsite power. The Division 3 4.16 kV (HPCS) bus can be supplied by either source of offsite power. In the event of a loss of offsite power (LOOP), each of the three buses is energized from its own independent DG: Division 1, Division 2, and the HPCS DG Division 3 generators, respectively.

During a LOOP, Divisions 1 and 2 are independent redundant divisions and supply all nuclear safety-related loads except the HPCS system. The HPCS system and related equipment are supplied by the dedicated HPCS DG.

The HPCS system provides and maintains an adequate coolant inventory inside the reactor pressure vessel to limit fuel cladding temperatures in the event of breaks in the reactor coolant pressure boundary. The system sprays water into the reactor vessel over a wide range of operating pressures. The HPCS provides reactor vessel coolant inventory makeup in the event of a small break loss-of-coolant accident (LOCA) that does not immediately depressurize the reactor vessel and helps to depressurize the reactor vessel. This system also provides spray cooling for long-term core cooling after a LOCA. The HPCS system also serves as a backup to the reactor core isolation cooling (RCIC) system to provide makeup water in the event of a loss of feedwater flow transient. The system is initiated by either high pressure in the drywell or low water level in the vessel. It operates independently of all other systems over the entire range of pressure differences from greater than normal operating pressure to zero. The HPCS cooling OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION decreases vessel pressure to enable the low-pressure cooling systems to function. The HPCS system pump motor is powered by an onsite DG if offsite power is not available.

2.2 Description of Proposed Change The licensee is proposing a one-time change to the completion time for Required Action B.2, "Restore HPCS System to OPERABLE status," in TS 3.5.1, "ECCS - Operating," of the Nine Mile Point 2 by adding a footnote to extend the completion time from 14 days to 35 days. The footnote would state:

A one-time change to this Completion Time from 14 days to 35 days due to the HPCS DG replacement has been approved via emergency license amendment request. This Completion Time expires on 12/31/2018 at 0100.

Additionally, the licensee is proposing to defer the completion of the following SRs associated with components or systems that are being protected during the replacement of the HPCS DG.

The completion of the following SRs may be extended to January 11, 2019:

SR 3.3.5.1.2 - CHANNEL FUNCTIONAL TEST for TS 3.3.5.1-.1 Functions 2.e, 2.f, 2.g, and 2.h SR 3.3.5.1.5 - CHANNEL CALIBRATION TS 3.3.5.1-.1 Functions 2.e, 2.f, 2.g, and 2.h SR 3.3.5.1.6 - LOGIC SYSTEM FUNCTIONAL TEST TS 3.3.5.1-.1 Functions 2.e, 2.f, 2.g, and 2.h SR 3.8.1.16 - Automatic Load Timer Relays for Division 2 Diesel SR 3.5.1.1 - LPCS [Low Pressure Core Spray] Gas Accumulation Monitoring SR 3.5.3.1 - RCIC Gas Accumulation Monitoring The completion of the following SRs may be extended to January 18, 2019:

SR 3.8.1.2 - Diesel Voltage and Frequency (Standby Start)

SR 3.8.1.3 - Diesel Synchronization and Loading SR 3.8.1.4 - Diesel Day Tank Fuel Oil SR 3.8.1.5 - Day Tank Water Accumulation SR 3.8.1.6 - Fuel Oil Transfer System SR 3.8.1.13 - Diesel Voltage and Frequency (Hot Restart)

SR 3.8.3.1 - Fuel Oil Storage Tank SR 3.8.3.3 - Diesel Fuel Oil Properties SR 3.8.3.5 - Fuel Oil Storage Tank Water Accumulation 2.3 Description of Regulatory Requirements The requirements in 10 CFR 50.46, "Acceptance criteria for emergency core cooling systems for light-water nuclear power reactors," in part, establish standards for the calculation of ECCS accident performance and acceptance criteria for that calculated performance.

The requirements in 10 CFR Part 50, Appendix K, "ECCS Evaluation Models," in part, establish required and acceptable features of evaluation models for heat removal by the ECCS after the blowdown phase of a loss-of-coolant accident.

OFFICIAL USE ONLY PROPRIETARY INFORMATION 0FFICIAb USE ONbY PROPRIETARY INFORMATION Appendix A, "General Design Criteria for Nuclear Power Plants" (GDC), to 10 CFR Part 50, applicable to this LAR are discussed below:

GDC 29, "Protection Against Anticipated Operational Occurrences," insofar as it requires protection and reactivity control systems be designed to assure an extremely high probability of accomplishing their safety functions in the event of anticipated operational occurrences.

GDC 34, "Residual Heat Removal," insofar as it requires to remove decay heat and other residual heat from the reactor core at a rate such that specified acceptable fuel design limits and the design conditions of the reactor coolant pressure boundary are not exceeded.

GDC 35, "Emergency Core Cooling," insofar as it requires that a system to provide abundant emergency core cooling be provided to transfer heat from the reactor core following any LOCA.

GDC 38, "Containment Heat Removal," insofar as it requires that a containment heat removal system be provided, and that its function shall be to rapidly reduce the containment pressure and temperature following a LOCA and maintain them at acceptably low levels.

GDC 50, "Containment Design Basis," insofar as it requires that the containment and its associated heat removal systems be designed so that the containment structure can accommodate, without exceeding the design leakage rate and with sufficient margin, the calculated temperature and pressure conditions resulting from any LOCA.

The requirements in 10 CFR 50.36, "Technical specifications," state, in part, that SRs are related to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and LCOs will be met.

3.0 TECHNICAL EVALUATION

3.1 One-Time Change to HPCS Completion Time LOCA Analysis To support the one-time extension of the completion time of TS 3.5.1, REQUIRED ACTION 8.2, "Restore HPCS System to OPERABLE status" from 14 days to 35 days, the licensee performed a supplemental LOCA analysis using the currently used and NRG-approved SAFER/PRIME-LOCA application methodology to demonstrate that the ECCS acceptance criteria are met. The analysis is performed for the most limiting scenario documented in the current analysis.

There are two limiting points in the break spectrum: (1) the large break, double-ended-guillotine (DEG) recirculation pump suction line break, and (2) the limiting small break of 0.07 foot (ft)2 area of the recirculation pump suction line under the HPCS DG failure scenario. The licensee analyzed ((

)) For both the nominal and 10 CFR OFFICIAL USE ONbY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION Part 50 Appendix K, conditions, the limiting case in the current analysis is the small break LOCA in recirculation pump suction line in the presence of a LOOP and failure of the HPCS DG. In the supplemental analysis, the licensee considered failure of Division 1 emergency diesel generator (EDG) or Division 2 EDG, in addition to the unavailability of HPCS DG. Table 1 below provides the analysis, conditions, and results, as provided in GE Hitachi Nuclear Energy Report 005N0847, Revision 1, December 2018, "Supplemental NMP2 LOCA Evaluation for HPCS DG OOS [Out-of-Service]" (ADAMS Package Accession No. ML18341A342).

Table 1: Supplemental Small and Large Break LOCA Equipment Failure Assumptions and Results Parameters Supplemental Analysis Small Break Large Break and Results Failure Case 1 Failure Case 2 Failure Case 1 Failure Case 2 Break Area 0.07 ft2 0.07 ft2 DEG DEG Equipment HPCS DG and HPCS DG and HPCS DG and HPCS DG and Failure Division 2 EDG Division 1 EDG Division 2 EDG Division 1 EDG ECCS 1 LPCS 1 LPCS Equipment 1 LPCI 2 LPCI 1 LPCI 2 LPCI Available Maximum PCT with

((

11

((

11

((

11

((

11 Nominal Inputs Maximum PCT with

((

11

((

11

((

11

((

))

Appendix K Inputs Abbreviations:

LPCI - Low Pressure Coolant Injection LPCS - Low Pressure Core Spray PCT - Peak Claddinq Temperature From the results in Table 1 above, it is concluded that the most limiting peak cladding temperature (PCT) in the supplemental analysis is ((

)) for an Appendix K small break LOCA, which is less than the 10 CFR 50.46 acceptance criteria of 2200 degrees Fahrenheit (OF).

In the current analysis, the maximum PCTs for small break LOCA were determined to be ((

11 The maximum PCTs for the large break LOCA were determined to be ((

11 In the supplemental analysis, the maximum local oxidation is less than 1 percent for nominal results, and less than 4 percent for 10 CFR Part 50, Appendix K, results and is, therefore, bounded by the 10 CFR 50.46 limit of 17 percent. The core wide oxidation is less than 0.1 percent for both nominal and Appendix K results, which is bounded by the 10 CFR 50.46 limit of 1 percent.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION The results of the licensee's supplemental LOCA analysis show that the ECCS will perform its function, meeting the 10 CFR 50.46 acceptance criteria of 2200 °F PCT and 17 percent maximum oxidation acceptance criteria for all normal operating conditions, and with the HPCS DG inoperable. The supplemental analysis also showed that the subsequent core heatup for the long-term evaluation has peak cladding temperature significantly less than the 10 CFR 50.46 limit of 2200 °F, satisfying the long-term core cooling requirement.

The NRC staff finds the LOCA analysis results acceptable because there is significant margin to the 10 CFR 50.46 acceptance criteria of 2200 °F, even after considering the conservative licensing basis 1 O CFR Part 50, Appendix K, requirements, along with multiple concurrent failures in the ECCS.

Transient Events The licensee stated that the following transient events described in the Nine Mile Point Updated Final Safety Analysis Report, Chapter 15, credit either the RCIC system or HPCS injection to maintain long-term reactor water level:

Loss of instrument air Feedwater controller failure - maximum demand Pressure regulator failure - open Generator load rejection Turbine trip Loss of feedwater flow Trip one recirculation pump Trip both recirculation pumps Fast closure of one main recirculation valve Fast closure of both recirculation valves Recirculation pump seizure During these events, the RCIC system or HPCS are powered by offsite power. The operation of HPCS DG is not a mitigating action for these events. The equipment failure that causes these events is considered to be the single failure, and no additional failure or LOOP is assumed. The NRC staff finds it acceptable that the long-term reactor water level is maintained during these events and is unaffected by the loss of the HPCS DG, because neither LOOP or any additional single failures are considered during these events.

Failure of Residual Heat Removal Shutdown Cooling The safety function of shutdown cooling is accomplished by equipment powered from Divisions 1 and 2 only. The failure of residual heat removal shutdown cooling may use RCIC and/or HPCS for shutdown cooling using power from Divisions 1 or 2. The current evaluation shows that availability or failure of the HPCS DG equipment does not affect the shutdown cooling and can be accomplished by power from either Division 1 or 2. The NRC staff finds it acceptable that with the failure of residual heat removal shutdown cooling mode, the shutdown cooling by the RCIC system or HPCS is unaffected by the loss of HPCS DG.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIA.b USE ONbY PROPRIETARY INFORMATION Increase in Reactor Coolant Inventory The licensee stated that an inadvertent manual HPCS startup event would increase the reactor coolant inventory. This event assumes the HPCS to be normally powered, and this event is unaffected by loss of the HPCS DG. The NRC staff finds the licensee's evaluation acceptable.

Loss of Alternating Current Power This loss of alternating current power event utilizes HPCS and RCIC powered by Divisions 1 or 2 for reactor water level control and is unaffected by the loss of HPCS DG. The NRC staff finds it acceptable that the reactor water level control can be maintained during the loss of HPCS DG.

Anticipated Transient Without Scram The licensee stated that during the anticipated transient without scram event, the HPCS pump is placed in pull-to-lock condition to prevent injection inside the core shroud. The NRC staff finds it acceptable that the HPCS inoperable will have no impact on the anticipated transient without scram analysis.

Primary Containment Pressure Response The current primary containment peak pressure of 42.08 pounds per square inch gauge (psig) is based on initial lower bound drywell temperature and relative humidity, and high initial suppression pool temperature of 90 °F. The containment design pressure is 45 psig. The peak pressure occurs in the second peak during the blowdown phase of the recirculation suction line break LOCA. The second peak is approximately 0.6 pounds per square inch (psi) greater than the first peak. The licensee evaluated the peak pressure sensitivity to ECCS lineups and flows and determined that it varies less than 1 psi, depending on the assumed initiation time and ECCS single failure assumptions. The licensee stated that the peak pressure with the assumed minimum ECCS with only two low pressure coolant injection (LPCI) systems and HPCS inoperable remains within the bounds of the current analysis, and no significant increase in the second peak is predicted. The NRC staff finds the licensee's evaluation acceptable and the reduced ECCS flow due to HPCS inoperable does not affect the second peak.

Primary Containment Temperature Response The primary containment peak temperature is defined by the small steamline break, which is analyzed assuming reactor water level is maintained with feedwater and the control rod drive system only. HPCS, RCIC, LPCS, and LPCI are not credited. Therefore, the NRC staff finds that HPCS inoperable does not impact the containment temperature response.

Equipment Qualification The licensee stated that the equipment qualification pressure and temperature profiles are defined conservatively and bound the accident peak pressure and temperature such that the assumed reduced ECCS flows do not impact equipment qualification profiles. The NRC staff finds that HPCS inoperable does not impact the equipment qualification.

OFFICIAL USE ONLY PROPRIETARY INFORMATION OFFICIAL USE ONLY PROPRIETARY INFORMATION ECCS Pump Net Positive Suction Head In the current analysis, the ECCS pump available net positive suction head is based on a conservatively assumed suppression pool temperature of 212 °F, without crediting containment accident pressure. The calculated peak suppression pool temperature for the design-basis LOCA is less than 210 °F. The reduced ECCS flows that would result from HPCS inoperable reduce the pump heat assumed added to the suppression pool and, therefore, the actual peak suppression pool temperature is reduced, which increases the available net positive suction head. Therefore, the NRC staff finds that HPCS inoperable does not negatively impact the ECCS pump net position suction head.

Conclusion The NRC staff concludes that a one-time extension of the completion time of TS 3.5.1, REQUIRED ACTION 8.2, "Restore HPCS System to OPERABLE status," from 14 days to 35 days is acceptable. The one-time extension is acceptable because the LOCA analysis described in the application demonstrates that the ECCS acceptance criteria are met, and the analysis was performed using established standards as required by 10 CFR 50.46 and acceptable features of evaluation models required by 10 CFR Part 50, Appendix K. The NRC staff also concludes that Nine Mile Point 2 will continue to meet the requirements of 10 CFR Part 50, Appendix A, GDC 29, 34, 35, 38, and 50, during the extension period, because the design of the plant will not be changed.

3.2 Extension of SRs Completion The licensee has requested to extend the following SRs:

Table 2: Surveillance Rec uirements Surveillance Surveillance Current 125%

Complete Maximum Requirement Division Frequency Scheduled Grace No Later Extension Date Date Than SR 3.3.5.1.2 92 days 11/29/2018 12/21/2018 01/11/2019 21 days SR 3.3.5.1.5 92 days 11/29/2018 12/21/2018 01/11/2019 21 days SR 3.3.5.1.6 92 days 11/29/2018 12/21/2018 01/11/2019 21 days SR 3.8.1.16 92 days 11/29/2018 12/21/2018 01/11/2019 21 days SR 3.5.1.1 31 days 12/20/2018 12/27/2018 01/11/2019 15 days SR 3.5.3.1 31 days 12/07/2018 12/14/2018 01/11/2019 28 days SR 3.8.1.2 1