ML17059C072
| ML17059C072 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 06/04/1998 |
| From: | Bajwa S NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML17059C073 | List: |
| References | |
| NUDOCS 9806100411 | |
| Download: ML17059C072 (20) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 NIAGARAMOHAWKPOWER CORPORATION DOCKET NO. 50-410 NINE MILE POINT NUCLEAR STATION UNIT2 AMENDMENTTO FACILITYOPERATING LICENSE Amendment No.82 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 15, 1997, as supplemented by letter dated April24, 1998, 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 1; B.
The facilitywilloperate 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 willbe conducted in compliance with the Commission's regulations; D.
The issuance of this amendment willnot 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 paragraph 2.C.(2) of Facility Operating License No. NPF-69 is hereby amended to read as follows:
9806i004ii 980604 PDR ADOCK 050004i0 P
C I
(2)
Technical S ecifications 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. 82 are hereby incorporated into this license.
Niagara Mohawk Power Corporation shall operate the facilityin accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of the date of its issuance to be implemented before startup of the Unit 2 reactor to begin fuel operating cycle 7.
FOR THE NUCLEAR REGULATORY COMMISSION S. Singh Bajwa, Director Project Directorate I-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
3une 4, 1998
ATTACHMENTTO LICENSE AMENDMENT AMENDMENTNO. ~2 TO FACILITYOPERATING LICENSE NO. NPF-69 DOCKET NO. 50-410 Revise Appendix A as follows:
Ill 2-1 3/4 4-1 82-1 82-2 82-3 82-4
~Insert Pa e
iii 2-1 3/4 4-1 82-1 82-2 82-3 82-4
INDEX DEFINITI N
'PA E
1A8 VENTILATIONEXHAUST TREATMENT SYSTEM..........................
1-9
.49 VENTING 1
.. 1-9 1.60 CORE OPERATING LIMITSREPORT....................................
1-9 Table 1.1 Surveillance Frequency Notations '............... ~................
1-10 Table 1.2 Operational Conditions........................................
1-11 AFETY LIMIT AND LIMITIN AFETY TEM SETTIN S
2.1 AFETY LIMITS THERMAL POWER, Low Pressure or Low Flow............................
2-1 THERMAL POWER, High Pressure and High Flow... ~......................
2-1 Reactor Coolant System Pressure............. ~.......................
2-1 Reactor Vessel Water Level....... ~.................. ~....,..... ~... 2-1 2.2 LIMITIN AFETY YSTEM SETTIN Reactor Protection System Instrumentation Setpoints..............
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1
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~ 2 2 Table 2.2.1-1 Reactor Protection System Instrumentation Setpoints.................
1 AFETY LIMIT I'troduction.......,......................................
B2 1
THERMAL POWER, Low Pressure or Low Flow...........................
B2-1 THERMAL POWER, High Pressure and High Flow.........................
B2-2 Bases Table B2.1.2-1 Deleted.........................................
B2-3 Bases Table B2.1.2-2 Deleted...........;...............................
B2-4 NINE MILE POINT - UNIT 2 Amendment No. 682
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2.0 SAFETY LIMIT AND LIMITING AFETY SYSTEM SETTINGS 2.1 SAFETY LIMITS THERMAL POWER 'Low Pr sure or Low Flow 2.1.1 THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of. rated flow.
APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.
ACTION'ith THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel steam dome pressure less than 785 psig or core flow less than 10%'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 requirements of Specification 6.7.
THERMAL POWER Hi h Pressure and Hi h Flow 2.1.2 The MINIMUMCRITICAL POWER RATIO (MCPR)" shall not be less thorn 1.09 with two recirculation loop operation and shall not be less than 1.10 with single recirculation loop operation with the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow.
d ACTION:
With MCPR" less than 1.09, with two recirculation loop operation or less than,1.10 with I
single loop operation, the reactor vessel steam dome pressure greater than 785 psig, and core flow greater than 10% 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 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 1, 2, 3, and 4.
~ACTI N:
With the reactor coolant system pressure as measured in the reactor vessel steam dome above 1325 psig, be in at least HOT SHUTDOWN with reactor 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.
REACTOR VESSEL WATER LEVEL 2.1.4 The reactor vessel water level shall be above the top of the active irradiated fuel.
" MCPR values are applicable to Cycle 7 operation only.
d NINE MILE POINT - UNIT 2 2-1 Amendment No. 7k 82
3 4.4 REACTOR COOLANT YSTEM 3 4.4.1 RECIRCULATION SYSTEM RECIRCULATION LOOPS LIMITINGCONDITION F
R PERATION 3.4.1.1 Two reactor coolant system recirculation loops shall be in operation with:
a.
Total core flow greater than or equal to 45% 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:
a.
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 ~70% of RATED THERMAL POWER, and, c)
Increase the MINIMUMCRITICAL POWER RATIO (MCPR)""" Safety Limit by 0.01 to 1.10 per Specification 2.1.2, and, d)
Reduce the Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) 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 Values 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 a 41,800** gpm.
See Special Test Exception 3.10.4.
This value represents the volumetric recirculation loop drive flow which produces 100%
core flow at 100% THERMAL POWER.
I MCPR values a>e applicable to Cycle 7 operation only.
NINE MILE POINT - UNIT 2 3/4 4-1 Amendment No. kk, 6 82
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2.1 BASES FOR SAFETY LIIVIIT 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 these barriers during normal plant operations and anticipated transients. The fuel cladding integrity Safety Limitis 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 1.09 for two recirculation loop operation and 1.10 for single recirculation loop operation.
MCPR" greater than 1.09 for two recirculation loop operation and 1.10 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 Pr s ure r L w Flow The use of critical power correlations is not valid for all critical power calculations performed at reduced 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 atl 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 Ib/hr, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi ~ Thus, the bundle flow with a 4.5 psi driving head will be greater than 28 x 10 Ib/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 MWt. With the design peaking factors, this corresponds to a THERMAL POWER of more than 50% of RATED THERMAL POWER.
Thus, a THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig is conservative.
MCPR values are applicable to Cycle 7 operation only.
NINE MILE POINT - UNIT 2 B2-1 Amendment No. 7k, kk 82
1 BASES FOR AFETY LI 21 T ERM LP WER HihPr n
Hi h' The fuel cladding integrity Safety Limit is set so that no'fuel damage is calculated to occur ifthe limit is not violated.
Since the parameters that result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions resulting in a departure from nucleate boiling have been used to mark the beginning of the region in
, which fuel damage could occur. Although it is recognized that a departure from nucleate boiling would not necessarily result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit.
However, the uncertainties in monitoring the core operating state and in the procedures used to calculate the critical power result in an uncertainty in the value of the critical
'ower. Therefore, the fuel cladding integrity Safety Limitis defined as the CPR in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling transition considering the power distribution within the core and all" uncertainties.
The Safety Limit MCPR is determined using a statistical model that combines all of the uncertainties in operating parameters and the procedures used to calculate critical power.
The probability of the occurrence of boiling transition is determined using an approved critical po~er correlation.
The critical power correlation is valid over the range of conditions used in the tests of the data used to develop the correlation.
Details of the fuel cladding integrity Safety Limit calculation are given in Reference
- 1. Reference 1 also includes a tabulation of the uncertainties used in the determination of the Safety Limit MCPR. The plant specific values of the parameters used in the Safety Limit MCPR statistical analysis are found in the cycle specific analysis.
~Rf rien 1.
General Electric Standard Application for Reactor Fuel, NEDE-24011-P-A (latest approved revision).
NINE MILE POINT - UNIT 2 B2-2 Amendment No. 4 82
This Page is Not Used NINE MILE POINT - UNIT 2 B2-3 Amendment No. 6, 4k 82
This Page is Not Used NINE MILE POINT - UNIT 2 B2-4 Amendment No. 4k, 6 Sp
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