ML17056A905
| ML17056A905 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 06/19/1990 |
| From: | Capra R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17056A906 | List: |
| References | |
| NPF-69-A-016 NUDOCS 9006270051 | |
| Download: ML17056A905 (14) | |
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UNITED STATES NUCLEAR R EG ULATORY COMMISSION WASHINGTON, D. C. 20555 NIAGARA-MOHAWK POWER-CORPORATION DOCKET-NO. 50-410
NINE.MILE POINT-NUCLEAR STATION.UIIIT.2 AMENDMENT-TO-FACII.ITY OPERATING. LICENSE Amendment No. 36 License No. NPF-69 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Niagara Mohawk Power Corporation (the licensee) dated December 8, 1989, 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 10 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 paIagraph 2.C.(2) of Facility Operating License No. NPF-69 is hereby amended to read as follows:
9006270051 9006l9 PDR ADOCK 050004i0 P
(2)
Technical S ecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix 8, both of which are attached
- hereto, as revised through Amendment No. 16 are hereby incorporated into this license.
Niagara Mohawk Power Corporation shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment will become effective prior to startup following the first refueling outage.
FOR THE NUCLEAR REGULATORY COMMISSION Robert A. Capra, Director Project Directorate I-I Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
June 19, 1990
ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO.
16 TO FACILITY OPEBATING LICENSE NO.
NPF 69 DOCKET NO. 50-410 Revise Appendix A as follows:
Remove Pa es 2-1 82-1 82-3 3/4 4-1 Insert Pa es 2-1 82-1 82-3 3/4 4-1
1
2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS THERMAL POWER Low Pressure or Low Flow
- 2. 1.1 THERMAL POWER shall not exceed 25X of RATED THERMAL POWER with the reactor vessel steam dome pressure less than 785 psig or core flow less than 10X of rated flow.
.APPLICABILITY:
OPERATIONAL CONDITIONS 1 and 2.
ACTION:
With THERMAL POWER exceeding 25K of RATED THERMAL POWER and the reactor vessel steam dome pressure less than 785 psig or core flow less than 10K of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the require-ments of Specification 6.7.
THERMAL POWER Hi h Pressure and Hi h Flow 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.07 with two recirculation loop operation and shall not be less than 1.08 with single recirculation loop operation with the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10K of rated flow.
APPLICABILITY:
OPERATIONAL CONDITIONS 1 and 2.
ACTION:
With MCPR less than 1.07, with two recirculation loop operation or less than 1.08 with single loop operation, the reactor vessel steam dome pressure greater than 785 psig, and core flow greater than 10'f rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.
REACTOR COOLANT SYSTEM PRESSURE
- 2. 1.3 The reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed 1325 psig.
APPLICABILITY:
OPERATIONAL CONDITIONS I, 2, 3, and 4.
ACTION:
With the reactor coolant system pressure as measured in the reactor vessel steam dome above 1325 psig, be in at least HOT SHUTDOWN with reacto~ coolant system pressure less than or equal to 1325 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.
- 2. 1. 4 The reactor vessel water level shall be above the top of the active irradiated fuel.
NINE MILE POINT - UNIT 2 2-1 Amendment No.
16
k f,
.2.
1.0 INTRODUCTION
The fuel cladding, reactor pressure
- vessel, and primary system piping are the principal barriers to the release of radioactive materials to the environs.
Safety Limits are established to protect the integrity of'hese barriers during normal plant operations and anticipated transients.
The fuel cladding integrity Safety Limit is set so that no fuel damage is calculated to occur if the limit is not violated.
Because fuel damage is not directly observable, a
step-back approach is used to establish a Safety Limit so that the MCPR is not less than 1.07 for two recirculation loop operation and 1.08 for single recirculation loop operation.
MCPR greater than 1.07 for two recirculation loop operation and 1.08 for single recirculation loop operation represents a
conservative margin relative to the conditions required to maintain fuel cladding integrity.
The fuel cladding 'is one of the physical barriers that separate the radioactive materials from the environs.
The integrity of this cladding barrier is related to its relative freedom from perforations or cracking.
Although some corrosion or use-related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable.
Fuel cladding perforations,
- however, can result from thermal stresses that occur from reactor operation significantly above design conditions and the Limiting Safety System Settings.
Although fission product migration from cladding perforation is just as measurable as that from use-related
- cracking, the thermally 'caused cladding perforations signal a threshold beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration.
Therefore, the fuel cladding Safety Limit is defined with a margin to the conditions that would produce onset of transition boiling, MCPR of -1.0.
These conditions represent a significant departure from the condition intended by design for planned operation.
2.1.1 THERMAL POWER Low Pressure or Low Flow The use of GEXL correlation is not valid for all critical power calculations at pressures below 785 psig or core flows less than 10'/. of rated flow.
Therefore, the fuel cladding integrity Safety Limit is established by other means.
This is done by establishing a limiting condition on core THERMAL POWER with the following basis.
Since the pressure drop in the bypass region is essentially.all elevation
- head, the core pressure drop at low power and flows will always be greater than 4.5 psi.
Analyses show that with a bundle flow of 28 x
10~ lb/hr, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 ps'hus, the bundle flow with a 4.5-psi driving head will be greater than 28 x
103 lb/hr.
Full-scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MNt.
With the design peaking'factors, this corresponds to a THERMAL POWER of more than 50'4 of RATED THERMAL POWER.
- Thus, a
THERMAL POWER limit of 25'/. of RATED THERMAL POWER for reactor pressu're below 785 psig is conservative.
NINE MILE POINT - UNIT 2 82-1 Amendment No.
16
BASES TABLE B2.1.2-1
~ UNCERTAINTIES USED IN THE DETERMINATION OF THE FUEL CLADDING SAFETY LIMIT*
UANTITY Feedwater Flow Feedwater, Temperature Reactor Pressure Core Inlet Temperature Core Total Flow Two Recirculation Loop Operation Single Recirculation Loop Operation Channel Flow Area Friction Factor Multiplier Channel Friction Factor Multiplier TIP Readings Two Recirculation Loop Operation Single Recirculation Loop Operation R Factor Critical Power STANDARD DEVIATION (1. OF POINT) 1.76 0.76 0.5 0.2 2.5 6.0 3.0 10.0 5.0 8.7 9.1 1.6 3.6 The uncertainty analysis used to establish the corewide Safety Limit MCPR is based on the assumption of quadrant power symmetry for the reactor core.
The values herein apply to both two recirculation loop operation and single recirculation loop operation, except as noted.
NINE MILE POINT UNIT 2 B2-3 Amendment No.
16
3/4. 4 REACTOR COOLANT SYSTEM 3/4. 4. 1 RECIRCULATION SYSTEM RECIRCULATION LOOPS LIMITING CONDITIONS FOR OPERATION 3.4. l. 1 Two reactor coolant system recirculation loops shall be in operation with:
a.
Total core flow greater than or equal to 45K of rated core flow, or b.
THERMAL POWER within the unrestricted zone of Figure 3.4. 1. 1-1.
APPLICABILITY:
OPERATIONAL CONDITIONS 1" and 2".
ACTION:
ao With one reactor coolant system recirculation loop not in operation:
1.
Within four hours:
a)
Place the recirculation flow control system in the Loop Manual (Position Control) mode, and b)
Reduce THERMAL POWER to < 70K of RATED THERMAL POWER, and, c)
Increase the MINIMUM CRITICAL POWER RATIO (MCPR) Safety Limit by 0.01 to 1.08 per Specification 2.1.2,
- and, d)
Reduce the Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limit to a value of 0. 81 times the two recirculation loop operation limit per Specification 3.2.1,
- and, e)
Reduce the Average Power Range Monitor (APRM) Scram and Rod Block and Rod Block Monitor Trip Setpoints and Allowable Yalues to those applicable for single recirculation loop operation per Specifications
- 2. 2.1, 3.2. 2 and 3. 3.6.
f)
Reduce the volumetric drive flow rate of the operating recirculation
= loop to < 41,800"" gpm.
~ See Special Test Exception 3.10.4.
"" This value represents the. volumetric recirculation loop drive flow which produces 100'ore flow at lOOX THERMAL POWER.
NINE MILE POINT" UNIT 2 3/4 4-1 Amendment No.
16