ML15096A076

Issuance of Amendment Maximum Extended Load Line Limit Analysis Plus

ML15096A076
Person / Time
Site:	Nine Mile Point
Issue date:	09/02/2015
From:	Bhalchandra Vaidya Plant Licensing Branch 1
To:	Orphanos P Exelon Generation Co, Nine Mile Point
Vaidya B, NRR/DORL/LPL1-1, 415-3308
References
TAC MF3056
Download: ML15096A076 (21)

Text

OFFICIAL USE ONLY- PROPRIETARY INFORMATION UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 September 2, 2015 Mr. Peter Orphanos Vice President Nine Mile Point Exelon Generation Company, LLC Nine Mile Point Nuclear Station, LLC P. 0. Box 63 Lycoming, NY 13093

SUBJECT:

NINE MILE POINT NUCLEAR STATION, UNIT NO. 2 - ISSUANCE OF AMENDMENT RE: MAXIMUM EXTENDED LOAD LINE LIMIT ANALYSIS PLUS (TAC NO. MF3056)

Dear Mr. Orphanos:

The Commission has issued the enclosed Amendment No. 151 to Renewed Facility Operating License No. NPF-69 for the Nine Mile Point Nuclear Station, Unit No. 2 (NMP2). The amendment consists of changes to the Technical Specifications (TSs) in response to your application transmitted by letter dated November 1, 2013, as supplemented by letters dated January 21, February 14, February 25, March 10, May 14, June 13, October 10, December 11, 2014, and February 18, 2015.

Further, as a part of its application for the license transfer and conforming amendment of the Renewed Facility Operating License for NMP2, in the letter dated March 28, 2014, (Agencywide Documents Access and Management System Accession No. ML14087A274) Exelon Generation has stated that:

Prior to the license transfers, GENG made docketed submittals to the NRG that requested specific licensing actions, such as license amendment requests, relief requests, exemption requests, etc. Furthermore, in the application for the license transfers, Exelon stated that upon transfer of the licenses, Exelon would assume all current regulatory commitments made for these units. Accordingly, Exelon hereby adopts and endorses those docketed requests currently before the NRG for review and approval. Exelon requests that the NRG continue to process those pending actions on the schedules previously requested by GENG.

The amendment includes changes to the NMP2 TSs necessary to: (1) implement the Maximum Extended Load Line Limit Analysis Plus (MELLLA+) expanded operating domain; (2) change the stability solution to Detect and Suppress Solution - Confirmation Density (DSS-CD); (3) use the TRACG04 analysis code; (4) increase the isotopic enrichment of boron-10 in the sodium pentaborate solution utilized in the Standby Liquid Control System (SLCS); and (5) increase the Safety Limit Minimum Critical Power Ratio (SLMCPR) for two recirculation loops in operation.

NOTICE: Enclosure to this letter contains Sensitive Internal Information Upon separation from the Enclosure, this letter is DECONTROLLED.

OFFICIAL USE ONLY-PROPRIETARY INFORMATION

OFFICIAL USE ONLY-PROPRIETARY INFORMATION P. M. Orphanos A copy of the related Safety Evaluation is enclosed. A Notice of Issuance will be included in the Commission's next regular biweekly Federal Register notice.

Sincerely, Bhalchandra Vaidya, Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-41 O

Enclosures:

1. Amendment No. 151 to NPF-69

2. Safety Evaluation (Non-Proprietary)

3. Safety Evaluation (Proprietary) cc w/encls: Distribution via Listserv OFFICIAL USE ONLY-PROPRIETARY INFORMATION

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 NINE MILE POINT NUCLEAR STATION. LLC EXELON GENERATION COMPANY. LLC DOCKET NO. 50-410 NINE MILE POINT NUCLEAR STATION. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 151 Renewed License No. NPF-69

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

A The application for amendment by Nine Mile Point Nuclear Station, LLC (the licensee), dated November 1, 2013, as supplemented by letters dated January 21, February 14, February 25, March 10, May 14, June 13, October 10, December 11, 2014, and February 18, 2015, 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 1O 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:

(2) Technical Specifications 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. 151, are hereby in*corporated into this license.

Nine Mile Point Nuclear Station, LLC, 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 within 90 days.

FOR THE NUCLEAR REGULATORY COMMISSION Michael Dudek, Chief (A)

Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the License and Technical Specifications Date of lssuance:september 2, 201 5

ATTACHMENT TO LICENSE AMENDMENT NO. 151 TO 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 marginal lines indicating the areas of change.

Remove Page Insert Page Page 4 Page4 Replace the following pages of 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 Insert Pages TS 2.0-1 TS 2.0-1 TS 3.1.7-3 TS 3.1.7-3 TS 3.3.1.1-2 TS 3.3.1.1-2 TS3.3.1.1-3 TS 3.3.1.1-3 TS 3.3.1.1-4 TS 3.3.1.1-4 (includes Rolled Over Changes)

TS 3.3.1.1-5 TS 3.3.1.1-5 (includes Rolled Over Changes)

TS 3.3.1.1-6 TS 3.3.1.1-6 TS3.3.1.1-8 TS 3.3.1.1-8 TS 3.3.1.1-9 TS 3.3.1.1-9 (includes Rolled Over Changes)

TS 3.3.1.1-10 TS 3.3.1.1-10 (includes Rolled Over Changes)

TS 3.4.1-1 TS3.4.1-1 TS 3.4.1-2 TS 3.4.1-2 TS 5.6-3 TS 5.6-3 TS 5.6-4 TS 5.6-4

(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 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. 151 are hereby incorporated into this license. Exelon Generation shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3) 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 NRC approval by October 31, 1989, a turbine system maintenance program based on the manufacturer's calculations of missile generation probabilities.

(Submitted by NMPC letter dated October 30, 1989 from C.D. Terry and approved by NRC letter dated March 15, 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, 146, 147, 148, 151

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 785 psig or core flow < 10% rated core flow:

THERMAL POWER shall be:::; 23% RTP.

2.1.1.2 With the reactor steam dome pressure ~ 785 psig and core flow ~ 10% rated core flow:

MCPR shall be ~ 1.09 for two recirculation loop operation or~ 1.09 for single recirculation loop operation.

2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel.

2.1.2 Reactor Coolant System Pressure SL Reactor steam dome pressure shall be :::; 1325 psig.

2.2 SL Violations With any SL violation, the following actions shall be completed within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:

2.2.1 Restore compliance with all SLs; and 2.2.2 Insert all insertable control rods.

NMP2 2.0-1 Amendment91, 105, 112, 140, 151

SLC System 3.1.7 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.1.7.7 Verify each pump develops a flow rate In accordance

?. 41.2 gpm at a discharge pressure with the

?. 1335 psig. lnservice Testing Program SR 3.1.7.8 Verify flow through one SLC subsystem 24 months on a from pump into reactor pressure vessel. STAGGERED TEST BASIS SR 3.1.7.9 Verify all heat traced piping between 24 months storage tank and pump suction valve is unblocked.

Once within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after piping temperature is restored to

?. 70°F SR 3.1.7.10 Verify sodium pentaborate enrichment Prior to is?. 92 atom percent B-10. addition to SLC tank NMP2 3.1.7-3 Amendment91, 111, 117, 123, 140,

+4d, 151

RPS Instrumentation 3.3.1.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. One or more Functions C.1 Restore RPS trip 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> with RPS trip capability.

capability not maintained.

D. Required Action and D.1 Enter the Condition Immediately associated Completion referenced in Time of Condition A, Table 3.3.1.1-1 for the B, or C not met. channel.

E. As required by E.1 Reduce THERMAL POWER to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Required Action D.1 < 26% RTP.

and referenced in Table 3.3.1.1-1.

F. As required by F.1 Initiate action to implement the Immediately Required Action D.1 Manual BSP Regions defined and referenced in in the COLA.

Table 3.3.1.1-1.

AND F.2 Implement the Automated BSP 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Scram Region using the modified APRM Simulated Thermal Power- High scram setpoints defined in the COLA.

AND F.3 Initiate action in accordance Immediately with Specification 5.6.8.

G. As required by G.1 Be in MODE 2. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Required Action D.1 and referenced in Table 3.3.1.1-1.

(continued)

NMP2 3.3.1.1-2 Amendment 91, 92, 140, 151

RPS Instrumentation 3.3.1.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME H. As required by H.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Required Action D.1 and referenced in Table 3.3.1.1-1.

I. As required by 1.1 Initiate action to Immediately Required Action D.1 fully insert all and referenced in insertable control Table 3.3.1.1-1. rods in core cells containing one or more fuel assemblies.

J. Required Action and J.1 Initiate action to implement Immediately associated Completion the Manual BSP Regions Time of Condition F not defined in the COLR.

met.

AND J.2 Reduce operation to below 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> the BSP Boundary defined in the COLR.

AND J.3 -------------NOTE--------------

LCO 3.0.4 is not applicable

---

Restore required channel to 120 days OPERABLE.

K. Required Action and K.1 Reduce THERMAL POWER 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> asociated Completion Time to less than 18% RTP.

of Condition J not met.

NMP2 3.3.1.1-3 Amendment 91, 92, 151

RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS

---

N()TE -----------------------------------------------------------

1. Refer to Table 3.3.1.1-1 to determine which SRs apply for each RPS 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 <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> provided the associated Function maintains RPS trip capability.

SURVEILLANCE FREQUENCY SR 3.3.1.1.1 Perform CHANNEL CHECK. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> SR 3.3.1 .1.2 Perform CHANNEL CHECK. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> SR 3.3.1.1.3 --------------------------- N()TE ----------------------------

Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after THERMAL POWER~ 23% RTP.

Verify the absolute difference between 7 days the average power range monitor (APRM) channels and the calculated power

2% RTP while operating at ~ 23% RTP.

SR 3.3.1 .1.4 ~-------------------------- N()TE ----------------------------

For Functions 1.a and 1.b, not required to be performed when entering MODE 2 from M()DE 1 until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after entering M()DE 2.

Perform CHANNEL FUNCTl()NAL TEST. 7 days SR 3.3.1.1.5 Verify the source range monitor (SRM) and Prior to fully intermediate range monitor (IRM) channels withdrawing overlap. SR Ms (continued)

NMP2 3.3.1.1-4 Amendment 91, 92, 123, 140, 151

RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.1.6 --------------------------- N()TE ----------------------------

()nly required to be met during entry into M()DE 2 from M()DE 1.

---

Verify the IRM and APRM channels overlap. ?days SR 3.3.1.1.7 Calibrate the local power range monitors. 1000 effective full power hours SR 3.3.1.1.8 Perform CHANNEL FUNCTl()NAL TEST. 92 days SR 3.3.1.1.9 Calibrate the trip units. 92 days SR 3.3.1 .1.10 -------------------------- N()TES ---------------------------

1. For Function 2.a, not required to be performed when entering M()DE 2 from M()DE 1 until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after entering M()DE 2.

2. For Function 2.e, the CHANNEL FUNCTl()NAL TEST only requires toggling the appropriate outputs of the APRM.

---

Perform CHANNEL FUNCTl()NAL TEST. 184 days SR 3.3.1.1.11 Perform CHANNEL CALIBRATl()N. 18 months SR 3.3.1.1.12 Perform CHANNEL FUNCTl()NAL TEST. 24 months (continued)

NMP2 3.3.1.1-5 Amendment 91, 92, 151

RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.1.13 --------------------------- N()TES ---------------------------

1. Neutron detectors are excluded.

2. For Functions 1.a and 2.a, not required to be performed when entering M()DE 2 from M()DE 1 until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after entering MODE 2.

3. For Function 2.e, the CHANNEL CALIBRATl()N only requires a verification of OPRM-Upscale setpoints in the APRM by the review of the "Show Parameters" display.

Perform CHANNEL CALIBRATION. 24 months SR 3.3.1.1.14 Perform L()GIC SYSTEM FUNCTl()NAL TEST. 24 months SR 3.3.1.1.15 Verify Turbine Stop Valve - Closure, and 24 months Turbine Control Valve Fast Closure, Trip Oil Pressure - Low Functions are not bypassed when THERMAL POWER is~ 26% RTP.

SR 3.3.1.1.16 Deleted (continued)

NMP2 3.3.1.1-6 Amendment 91, 92, 140, 151 Correoted by letter of July 24, 2000

RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 1 of 3)

Reactor Protection System Instrumentation CONDITIONS APPLICABLE REQUIRED REFERENCED MODES OR OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE

1. Intermediate Range Monitors

a. Neutron Flux - Upscale 2 3 H SR 3.3.1.1.1 ~ 122/125 SR 3.3.1.1.4 divisions SR 3.3.1.1.5 of full SR 3.3.1.1.6' scale SR 3.3.1.1.13 SR 3.3.1.1.14 5(a) 3 SR 3.3.1.1.1 ~ 122/125 SR 3.3.1.1.4 divisions SR 3.3.1.1.13 of full SR 3.3.1.1.14 scale

b. lnop 2 3 H SR 3.3.1.1.4 NA SR 3.3.1.1.14 5(a) 3 SR 3.3.1.1.4 NA SR 3.3.1.1.14

2. Average Power Range Monitors

a. Neutron Flux - Upscale, 2 3 per logic H SR 3.3.1.1.2 ~20% RTP Setdown channel SR 3.3.1.1.6 SR 3.3.1.1.7 SR 3.3.1.1.10 SR 3.3.1.1.13

b. Flow Biased Simulated 3 per logic G SR 3.3.1.1.2 ~ 0.61W +

Thermal Power - Upscale channel SR 3.3.1.1.3 63.4% RTP SR 3.3.1.1.7 and~ 115.5%

SR 3.3.1.1.10 RTP(b)(e)

SR 3.3.1.1.13(c),(d)

c. Fixed Neutron 3 per logic G SR 3.3.1.1.2 ~ 120% RTP Flux - Upscale channel SR 3.3.1.1.3 SR 3.3.1.1.7 SR 3.3.1.1.10 SR 3.3.1.1.13

d. lnop 1,2 3 per logic H SR 3.3.1.1.7 NA channel SR 3.3.1.1.10 (continued)

(a) With any control rod withdrawn from a core cell containing one or more fuel assemblies.

(b) Allowable Value is .50(W - 5%) + 53.5% RTP when reset for single loop operation per LCO 3.4.1, "Recirculation Loops Operating."

(c) If the As-Found channel setpoint is outside its predefined As-Found tolerances, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

(d) The instrument channel setpoint shall be reset to a value within the As-Left tolerance around the nominal trip setpoint at the completion of the surveillance; otherwise, the channel shall be declared inoperable. Setpoints more conservative than the nominal trip setpoint are acceptable provided that the As-Found and As-Left tolerances apply to the actual setpoint implemented in the surveillance procedures to confirm channel performance. The nominal trip setpoint and the methodologies used to determine the As-Found and the As-Left tolerances are specified in the Bases associated with the specified function.

(e) With OPRM Upscale (function 2.e) inoperable, reset the APRM-STP High scram setpoint to the values defined by the COLR to imple,ment the automated BSP Scram Region in accordance with Action F.2 of this Specification.

NMP2 3.3.1.1-8 Amendment 91, 92, 123, 140, 151

RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 2 of 3)

Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE

2. Average Power Range Monitors (continued)

e. OPRM-Upscale ~18% RTP(f) 3 per logic F SR 3.3.1.1.2 NA channel SR 3.3.1.1.7 SR 3.3.1.1.10 SR 3.3.1.1.13

f. 2-0ut-Of-4 Voter 1,2 2 H SR 3.3.1.1.2 NA SR 3.3.1.1.10 SR 3.3.1.1.14 SR 3.3.1.1.17

3. Reactor Vessel Steam Dome 1,2 2 H SR 3.3.1.1.1 ~ 1072 psig Pressure - High SR 3.3.1.1.8 SR 3.3.1.1.9 SR 3.3.1.1.13 SR 3.3.1.1.14 SR 3.3.1.1.17

4. Reactor Vessel Water 1,2 2 H SR 3.3.1.1.1 <': 157.8 inches Level - Low, Level 3 SR 3.3.1.1.8 SR 3.3.1.1.9 SR 3.3.1.1.13 SR 3.3.1.1.14 SR 3.3.1.1.17

5. Main Steam Isolation 8 G SR 3.3.1.1.8 ~ 12% closed Valve - Closure SR 3.3.1.1.13 SR 3.3.1.1.14 SR 3.3.1.1.17

6. Drywell Pressure - High 1,2 2 H SR 3.3.1.1.1 ~ 1.88 psig SR 3.3.1.1.8 SR 3.3.1.1.9 SR 3.3.1.1.13 SR 3.3.1.1.14 (continued)

(f) Following DSS-CD implementation, DSS-CD is not required to be armed while in the DSS-CD Armed Region during the first reactor startup and during the first controlled shutdown that passes completely through the DSS-CD Armed Region. However, DSS-CD is considered OPERABLE and shall be marinated OPERABLE and capable of automatically arming for operation at recirculation drive flow rates above the DSS-CD Armed Region.

NMP2 3.3.1.1-9 Amendment 91, 92, 140, 151

RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 3 of 3)

Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE

7. Scram Discharge Volume Water Level - High

a. Transmitter/Trip Unit 1,2 2 H SR 3.3.1.1.1  ::;; 49.5 SR 3.3.1.1.8 inches SR 3.3.1.1.9 SR 3.3.1.1.11 SR 3.3.1.1.14 5(a) 2 SR 3.3.1.1.1  ::;; 49.5 SR 3.3.1.1.8 inches SR 3.3.1.1.9 SR 3.3.1.1.11 SR 3.3.1.1.14

b. Float Switch 1,2 2 H SR 3.3.1.1.8  ::;; 49.5 SR 3.3.1.1.13 inches SR 3.3.1.1.14 5(a) 2 SR 3.3.1.1.8  ::;; 49.5 SR 3.3.1.1.13 inches SR 3.3.1.1.14

8. Turbine Stop  ;;:: 26% RTP 4 E SR 3.3.1.1.8. S:_7% Cl9S$d Valve - Closure SR 3.3.1.1.13 SR 3.3.1.1.14 SR 3.3.1.1.15 SR 3.3.1.1.17

9. Turbine Control Valve  ;;:: 26% RTP 2 E SR 3.3.1.1.8  ;;:: 465 psig Fast Closure, Trip Oil SR 3.3.1.1.13 Pressure - Low SR 3.3.1.1.14 SR 3.3.1.1.15 SR 3.3.1.1.17

10. Reactor Mode 1,2 2 H SR 3.3.1.1.12 NA Switch - Shutdown Position SR 3.3.1.1.14 5(a) 2 SR 3.3.1.1.12 NA SR 3.3.1.1.14

11. Manual Scram 1,2 4 H SR 3.3.1.1.4 NA SR 3.3.1.1.14 5(a) 4 SR 3.3.1.1.4 NA SR 3.3.1.1.14 (a) With any control rod withdrawn from a core cell containing one or more fuel assemblies.

NMP2 3.3.1.1-10 Amendment 91, 92, 14 0, 151

Recirculation Loops Operating 3.4.1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 Recirculation Loops Operating LCO 3.4.1 Two recirculation loops with matched flows shall be in operation, One recirculation loop shall be in operation provided the plant is not operating in the MELLLA or MELLLA+ domain defined in the COLR and provided the following limits are applied when the associated LCO is applicable:

a. LCO 3.2.1, "AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)," single loop operation limits specified in the COLR;

b. LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR)," single loop operation limits specified in the COLR; and

c. LCO 3.3.1.1, "Reactor Protection System (RPS)

Instrumentation," Function 2.b (Average Power Range Monitors Flow Biased Simulated Thermal Power- Upscale},

Allowable Value of Table 3.3.1.1-1 is reset for single loop operation.

APPLICABILITY: MODES 1 and 2.

NMP2 3.4.1-1 Amendment 91, 92, 123, 151

Recirculation Loops Operating 3.4.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. No recirculation loops A.1 Be in MODE 2. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> in operation.

AND A.2 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> B. Recirculation loop B.1 Declare the 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> flow mismatch not recirculation loop within limits. with lower flow to be "not in operation."

AND B.2 Prohibit operation in the 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> MELLLA domain or MELLLA+ domain defined in the COLR.

C. Requirements of the C.1 Satisfy the 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> LCO not met for requirements of the reasons other than LCO.

Conditions A and B.

D. Required Action and D.1 Be in MODE 3. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> associated Completion Time of Condition C not met.

NMP2 3.4.1-2 Amendment~' 151

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

1. The APLHGR for Specification 3.2.1.

2. The MCPR for Specification 3.2.2.

3. The LHGR for Specification 3.2.3.

4. The Manual Backup Stability Protection (BSP) Scram Region (Region I), the Manual BSP Controlled Entry Region (Region 11), the modified APRM Simulated Thermal Power - High setpoints used in the OPRM (Function 2.e), Automated BSP Scram Region, and the BSP Boundary for Specification 3.3.1.1.

5. The Allowable Values, NTSPs, and MCPR conditions for the Rod Block Monitor - Upscale Functions for Specification 3.3.2.1.

b. The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1. NEDE-24011-P-A-US, "General Electric Standard Application for Reactor Fuel," U.S. Supplement, (NRC approved version specified in the COLR).

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SOM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

(continued)

NMP2 5.6-3 Amendment 91, 92, 105, 123, 151

Reporting Requirements 5.6 5.6 Reporting Requirements (continued) 5.6.6 Post Accident Monitoring (PAM) Instrumentation Report When a report is required by Condition B or F of LCO 3.3.3.1 ,

"Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

5.6.7 Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)

a. RCS pressure and temperature limits for heatup, cooldown, low temperature operation, criticality, and system leakage and hydrostatic testing as well as heatup and cooldown rates shall be established and documented in the PTLR for the following:

1. Limiting Condition for Operation 3.4.11, "RCS Pressure and Temperature (P!T) Limits."

2. Surveillance Requirements 3.4.11.1 through 3.4.11 .9

b. The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1. N EDC-33178P-A, Revision 1, "General Electric Methodology for Development of Reactor Pressure Vessel Pressure-Temperature Curves," dated June 2009. The licensee will calculate the fluence for determining the adjusted reference temperature using either; (1) values determined using an NRG-approved, RG 1.190-adherent method, or (2) a fluence estimate, which the licensee has verified as conservative, using an NRG-approved, RG 1.190-adherent method.

c. The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto.

5.6.8 OPRM Report When a report is required by Required Action F.3 of TS 3.3.1.1, "RPS Instrumentation," a report shall be submitted within the following 90 days.

The report shall outline the preplanned means to provide backup stability protection, the cause of the inoperability, and the plans and schedule for restoring the required instrumentation channels to operable status.

NMP2 5.6-4 Amendment 91, 92, 145, 151

OFFICIAL USE ONLY PROPRIETARY INFORMATION P. M. Orphanos A copy of the related Safety Evaluation is enclosed. A Notice of Issuance will be included in the Commission's next regular biweekly Federal Register notice.

Sincerely, IRA/

Bhalchandra Vaidya, Project Manager Plant Licensing Branch 1-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-410

Enclosures:

1. Amendment No. 151 to NPF-69

2. Safety Evaluation (Non-Proprietary)

3. Safety Evaluation (Proprietary) cc w/encls: Distribution via Listserv DISTRIBUTION:

PUBLIC RidsNrrDraAphb RidsNrrDraArcb RidsNrrDeEicb RidsNmssFlb RidsNrrDssScvb RidsDssSnpb RidsNrrDssSrxb RidsNrrStsb RidsNrrDraAfpb LPLl-1 R/F RidsNrrDeEeeb RidsNrrDeEmcb RidsNrrDeEpnb RidsNrrDorlLPL 1-1 RidsRgn1 MailCenter RidsAcrsAcnw_MailCTR RidsNrrDssSbpb RidsNrrLAKGoldstein RidsNrrDeEsgb RidsNrrPMNineMilePoint DSchroder, Region 1 I. Dozier, NRR/ARCB R. Stattel, NRR/EICB N. Karipineni, NRR/SCVB M. Panicker, NRR/SNPB A. Guzzetta, NRR/SRXB D. Saenz, NRR/SRXB M. Honcheric, NRR/STSB R. Pederson. NRR/ARCB G. Thomas, NRR/SRXB I. Tseng, NRR/EMCB J. Tsao, NRR/EPNB D. Frumkin, NRR/AFPB ADAMS Accession Nos.: Package: ML15230A487 Transmittal Letter & Amendment: ML15096A076 Non-proprietary SE: ML152238144 Proprietary SE Enclosure 3: ML15195A257 (*) SE transmitted by memo **)By email OFFICE DORULPL 1-1\PM DORULPL 1-1\LA DSS/SRXB/BC DSS/SNPB/BC DSS/SCVB/BC DE/El CB/BC NAME BVaidva KGoldstein CJackson(*) JDean(*)w/SRXB RDennia(*) JThoro(*)

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I DATE 8/10/2015 1/14/2015 5/04/15, 6/04/15 8/13/2015 6/18/2015 6/22/2015 NMS~~I OFFICE DSS/SBPB/BC DRA/APLA/BC DRA/AFPB/BC DE/ESGB/BC DE/EVIB/BC NAME GCasto(**) SRosenberg(*) AKlein(**) GKulesa(**) JMcHale(**) IVIVCllltJ Not Applicable DATE 12/17/14, 6/09/15 8/13/2015 7/27/2015 6/15/2015 8/11/2015 8/14/2015 OGC(NLO OFFICE DORULPL 1-1/BC(A) DORULPL 1-1/PM w/comments)

NAME Jlindell MDudek BVaidva DATE 08/11/2015 9/01/2015 9/02/2015 OFFICIAL USE ONLY PROPRIETARY INFORMATION