ML081820317

Technical Specifications, Implementation of Arts/Mella

ML081820317
Person / Time
Site:	Nine Mile Point
Issue date:	02/27/2008
From:	NRC/NRR/ADRO/DORL/LPLI-1
To:
david marshall NRR/DORL 415-1547
References
TAC MD5233
Download: ML081820317 (10)

Text

4-(1) Maximum Power Level Nine Mile Point Nuclear Station, LLC, Is authorized to operate the facility at reactor core power levels not in excess of 3467 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. 123 are hereby incorporated into this license. Nine Mile Point Nu.-lear Station, LLC shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3) Fuel Storage and HandlinQ (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 3.5.1,310, 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!Jr its supplements wherein the license condition h: discussed.

Renewed License No. NPF-69 Amr*mr_t N). 123

TABLE OF CONTENTS 1.0 USE AND APPLICATION 1.1 Definitions ....................................... 1.1-1 1.2 Lo gical Con n ecto rs .......................................................................... 1.2-1 1.3 C om pletion Times ........................................................................... - .3-I 1.4 Fre q u e ncy ......................................................................................... 1.4 -1 2_.0 SAFETY LIMITS (SI.s 2 .1 S Ls ................................................................................................... 2 .0 -1 2 .2 S L V iolatio n s ...................................................................................... 2 .0-1 3.0 LIMITING CONDITION FOR OPERATION (LCO) APPLICABILITY .......... 3.0-1 3.0 SURVEILLANCE REQUIREMENT (SR) APPLICABILITY ........................ 3.0-4 3.1 REACTIVITY CONTROL SYSTEMS 3.1.1 SHUTDOWN MARGIN (SDM) ......................... 3 .1.1-1 3.1.2 R eactivity Anom alies .................................................................... 3.1.2-1 3.1.3 Control Rod O PERABILITY .......................................................... 3.1.3-1 3.1.4 C ontrol R od Scram Tim es ............................................................ 3.1.4-1 3.1.5 Control Rod Scram Accum ulators ................................................. 3.1.5-1 3.1.6 Rod Pattern C ontrol ...................................................................... 3.1.6-1 3.1.7 Standby Liquid Control (SLC) System .................... 3.1.7-1 3.1.8 Scram Discharge Volume (SDV) Vent and D ra in Va lve s ............... ................. ................................. 3 .1.8-1 3.2 POWER DISTRIBUTION LIMITS 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (A P LH G R ) ................................. ...................................... 3 .2 .1-1 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR) ............................. 3.2.2-1 3.2.3 LINEAR HEAT GENERATION RATE (LHGR) .............. 3.2.3-1 3.3 INSTRUMENTATION 3.3.1.1 Reactor Protection System (RPS)

Instrum entation ..................................................... 3.3.1.1-1 3.3.1.2 Source Range Monitor (SRM) Instrumentation ............................. 3.3.1.2-1 3.3.2.1 Control Rod Block nstr m . .......................... ............... 3 2 1-1 3.3.2.2 Feedwater System and Main Turbine High Water Level Trip Instrumentation .................. ......3.3.2.2-1 3.3.3.1 Post Accident Monitoring (PAM) Instrumentation .......................... 3.3.3.1-1 3.3.3.2 Rem ote Shutdow n System ........................................................... 3.3.3.2-1 3.3.4.1 End of Cycle Recirculation Pump Trip (EOC-RPT)

Instrum e ntation ................................................................ 3.3.4 .1-1 3.3.4.2 Anticipated Transient Without Scram Recirculation Pump Trip (ATWVS-RPT) Instirumentation ............................. 3.3.4.2-1 3.3.5.1 Emergency Core Cooling System (ECCS)

Instrum entation .................................................. 3. 3.3.5.1-1 3.3.5.2 Reactor Core Isolation Cooling (RCIC) System Instrum entation ................................................................... 3 .3 .5.2-1 (continued)

NMP2 Amendment 94, 123

Definitions 1.1 1.1 Definitions LEAKAGE 2. LEAKAGE into the drywell atmosphere from (continued) sources that are both specifically located and known either not to interfere with the operation of leakage detection systems or not to be pressure boundary LEAKAGE;

b. Unidentified LEAKAGE All LEAKAGE into the drywell that is not identified LEAKAGE; and

c. Pressure Boundary LEAKAGE LEAKAGE through a nonisolable fault in a Reactor Coolant System (RCS) component body, pipe wall, or vessel wall.

LINEAR HEAT GENERATION The LHGR shall be the heat generation rate per RATE (LHGR) unit length of fuel rod. It is the integral of the heat flux over the heat transfer area associated with the unit length.

LOGIC SYSTEM FUNCTIONAL A LOGIC SYSTEM FUNCTIONAL TEST shall be a test TEST of all required logic components (i.e., all required relays and contacts, trip units, solid state logic elements, etc.) of a logic circuit, from as close to the sensor as practicable up to, but not including, the actuated device, to verify OPERABILITY. The LOGIC SYSTEM FUNCTIONAL TEST may be performed by means of any series of sequential, overlapping, or total system steps so that the entire logic system is tested.

MINIMUM CRITICAL POWER The MCPR shall be the smallest critical power RATIO (MCPR) ratio (CPR) that exists in the core for each class of fuel. The CPR is that power in the assembly that is calculated by application of the appropriate correlation(s) to cause some point in the assembly to experience boiling transition, divided by the actual assembly operating power.

(continued)

NMP2 1 .1!-4 Amendment 94,123

SLC System 3,1,7 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.1.7.7 Verifyv each pump develops a flow rate in accordance

> 41.2 gpm at a discharge pressure with the

> 1325 psig. Inservice 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. AND 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 > 25 atom percent B-10. addition to SLC tank NMP2 3.1.7-3, Amendment 9 14, !!-, -1 7 ,123

RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.1.3 NOTE ..------------.--

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 > 25% RTP.

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

<2% RTP while operating at >_25% RTP.

SR 3.3.1.1.4 -------------- NOTE For Functions 1.a and 1.b, not required to be performed when entering MODE 2 from MODE 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.

Perform CHANNEL FUNCTIONAL 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. SRMs SR 3.3.1.1.6 Only required to be met during entry into MODE 2 from MODE 1.

Verify the IRM and APRM channels overlap. 7days (continued)

NMP2 3.3.1.1-4 Amendment 94, 92-,123

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 j ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION 0.1 REQUIREMENTS VALUE

1. Intermediate Range Monitors a, Neutron Flux - Upscale 2 3 H SR 3.3.1.1.1 5122/125 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 3.3.11.1.14 5(a) 3 SR 3.3.1.1.1 <_1221125 SR 3.3.1.1.4 divisions SR 3,3.1.1.13 of full SR 3.3.1.1.14 scale

b. Inop 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 <5.64W +

Thermal Power - Upscale channel SR 3.3.1.1.3 63.8% RTP" SR 3.3.1.1.7 and *115.5%

SR 3.3.1.1.10 RTP(b)

SR 3.3.1.1.13

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. Inop 1,2 3 per logic H SR 3.3.1.1.7 NA channel SR 3.3.1.1.i0

e. OPRM-Upscale 1 3 per logic F SR 3.3.1.1.2 As channel SR 3.3.1.1.7 specified SR 3.3.1.1.10 in the COLR SR 3.3.1.1.13 SR 3.3.1.1.16

f. 2-Out-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 (continued)

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

(b) Allowable Value is .58(W - 5%) + 62% RTP when reset for single loop operation per LCO 3.4.1, "Recirculation Loops Operating."

NMP2 3.3.1.1-8 Amendment 1--92323

Control Rod Block Instrumentation 3.3.2.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.2.1.2 -NOT-..-------- NO----------N Not required to be performed until 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after THERMAL POWER is*< 10% RTP in MODE 1.

Perform CHANNEL FUNCTIONAL TEST.- 92 days SR 3.3.2.1.3 Perform CHANNEL FUNCTIONAL TEST. 184 days SR 3.3.2.1.4 ----- - ----------- NOTE -------------

Neutron detectors are excluded.

Verify the RBM: 24 months

a. Low Power Range - Upscale Function is not bypassed when APRM Simulated Thermal Power is >- 28% and < 63% RTP and a peripheral control rod is not selected.

b. Intermediate Power Range - Upscale Function is not bypassed when APRM Simulated Thermal Power is ->63% and

< 83% RTP and a peripheral control rod is not selected.

c. High Power Range - Upscale Function is not bypassed when APRM Simulated Thermal Power is > 8%- RTP and a perip,'hie'"*aol,,,

rod is not selected.

SR 3.3.2.1.5 Verify the RWM is not bypassed when 24 months THERMAL POWER is < 10% RTP.

SR 3.3.2.1.6 -------------------- NOTE---------

Not required to be performed until 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> after reactor mode switch is in the shutdown position.

Perform CHANNEL FUNCTIONAL TEST. 24 months (continued)

NMP2 3.3.2.1-4 Amendment 94,123

Control Rod Block Instrumentation 3.3.2.1 Table 3.3.2.1-1 (page 1 of 1)

Control Rod Block Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS CHANNELS REQUIREMENTS VALUE 1 Rod Block Monitor

a. Low Power Range- Upscale (a) 2 SR 3.3.2.1.3 (h)

SR 3.312.1.4 SR 3.3.2.1.70)

b. Intermediate Power Range- (b) 2 SR 3.3.2.1.3 (h)

Upscale SR 3.3.2.1.4 SR 3.3.2.1.7(i)

c. High Power Range- Upscale (c)(d) 2 SR 3.3.2.1.3 (h)

SR 3.3.2.1.4 SR 3.3.2.1.7(1)

d. Inop (d)(e) 2 SR 3.3.2.1.3 NA

2. Rod Worth Minimizer 1(0,2(0 1 SR SR 3.3.2.1.1 3.3.2.1.2 NA SR 3.3.2.1.5 SR 3.3.2.1.8

3. Reactor Mode Switch - Shutdown (g) 2 SR 3.3.2.1.6 NA Position (a) APRM Simulated Thermal Power is Ž 28% and < 63% RTP and MCPR < limit specified in the COLR and no peripheral control rod selected.

(b) APRM Simulated Thermal Power is > 63% and < 83% RTP and MCPR < limit specified in the COLR and no peripheral control rod selected.

(c) APRM Simulated Thermal Power is > 83% and <90% RTP and MCPR < limit specified in the COLR and no peripheral control rod selected.

(d) APRM Simulated Thermal Power is Ž 90% RTP and MCPR < limit- specified in the COLR and no peripheral control rod selected.

(e) APRM Simulated Thermal Power is > 28% RTP and < 90% RTP and MCPR < limit specified in the COLR and no peripheral control rod is selected.

With THERMAL POWERs5 10% RTP, except du ring the reactor shutdown process if the coupling of each withdrawn control (f) rod has been confirmed.

(g) Reactor mode switch in the shutdown position.

"(h) Allowable Value specified in the COLR.

(i) If the as-found channel setpoint is not the nominal trip setpoint (NTSP), the channel is inoperable. The NTSP is specified in the COLR. The methodology used to determine the NTSP is specified in the Bases.

NMP2 3.3.2.1-6 Amendment 91,-2-1---, 123

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, OR One recirculation loop shall be in operation with the following limits 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 94-,*92-,13

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. Reactor Protection System Instrumentation Setpoint for the OPRM - Upscale Function Allowable Value 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-2401 1-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 SDM, 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)

NMVP2 5:6-3 Amendment 91, 92,101 123