ML17058B899
| ML17058B899 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 08/11/1993 |
| From: | Capra R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17058B898 | List: |
| References | |
| NUDOCS 9309010183 | |
| Download: ML17058B899 (22) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 NIAGARA MOHAWK POWER CORPORATION DOCKET NO. 50-410 NINE M LE POINT NUCLEAR STATION UNIT 2 AMENDMENT TO FACILITY OPERATING LICENS Amendment No. 46 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 February 27,
- 1993, as supplemented June 18,
- 1993, 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 facility will operate in conformity wi'th the application, the provisions of the Act, and the rules and regulations of the Commission; C;
There is reasonable assur ance (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 paragraph 2.C.(2) of Facility Operating License No. NPF-69 is hereby amended to read as follows:
93090iOi83 9308ii j,i PDR ADQCK 050004i0.'(
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"A
(2) e c l 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.
46 are hereby incorpor ated 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 is effective as of the date of its issuance to be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION
Attachment:
Changes to the Technical Specifications Date of Issuance:
August 11, 1993 Robert A. Capra, Director Project Directorate I-I Division of Reactor Projects I/II Office of Nuclear Reactor Regulation
0
I T
CHM T TO ICENSE AMENDMENT AM NDMENT NO 46 TO FACILITY OPERATING LICENSE NO. NPF-69 DOCKET NO. 50-410 Revise Appendix A as follows:
Remove Pa es 1-2 82-1 82-2 82-3 82-4 83/4 2-2 83/4 2-3 83/4 2-4
~P 1-2 82-1 82-2 82-3 82-4 83/4 2-2 83/4 2-3 83/4 2-4
F
DEFINITI N HANNEL F N TIONALTEST 1s6 (Continued)
The CHANNEL FUNCTIONALTEST may be performed by any series of sequential, overlapping or total channel steps so that the entire channel is tested.
RE ALT RATI N
1.7 CORE ALTERATION shall be the addition, removal, relocation, or movement of fuel,
- sources, incore instruments or reactivity controls within the reactor pressure vessel with the vessel head removed and fuel in the vessel.
Normal movement of the SRMs, IRlVls, TIPs or special movable detectors is not considered a CORE ALTERATION. Suspension of CORE ALTERATIONS shall not preclude completion of the movement of a component to a safe conservative position.
RE MAXIM M FRA TI N OF LIMITINGPOWER DEN ITY 1.8 The CORE MAXIMUMFRACTION OF LIMITINGPOWER DENSITY (CMFLPD) shall be the highest value of the FLPD which exists in the core.
RITI AL P WER RATI 1.9 The CRITICAL POWER RATIO (CPR) shall be the ratio of that power in the assembly which is calculated by application of an approved critical power correlation to cause some point in the assembly to experience boiling transition, divided by the actual fuel assembly operating power.
D E E IVALENT l-1 1
1.10 DOSE EQUIVALENT 1-131 shall be that concentration of l-131, expressed in microcuries per gram, which alone would produce the same thyroid dose as the quantity and isotopic mixture of l-131, l-132, l-133, l-134, and l-135 actually present.
The thyroid dose conversion factors used for this calculation shall be those listed in Table III of TID-14844, "Calculation of Distance Factors for Power and Test Reactor Sites."
VERA Dl INTE RATI N ENER Y
1.11 P shall be the average, weighted in proportion to the concentration of each radionuclide in the reactor coolant at the time of sampling, of the sum of the average beta and gamma energies per disintegration; expressed in MeV, for isotopes, with half-lives greater than 15 minutes, making up at least 95% of the total non-iodine activity in the coolant.
EMER EN Y RE LIN Y TEM RE P
N E TIME 1.12 The EMERGENCY CORE COOLING SYSTEM (ECCS) RESPONSE TIME shall be that time interval from when the monitored parameter exceeds its ECCS actuation setpoint at the channel sensor until the ECCS equipment is capable of performing its safety function, i.e., the valves travel to their required positions, pump NINE MILE POINT - UNIT 2 1-2 Amendment No. 46
i1, 1
2.1 BA E
F R
AFETY LIMITS 2.1.
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 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 Pre ure or L w Fl w 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 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'b/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'b/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.
NINE MILE POINT - UNIT 2 B2-1 Amendment No. 34'6
BA E
F R
AFETY LIMIT 2.1.2 THERMAL P WER Hi h Pre ur nd Hi h Fl w The fuel cladding integrity Safety Limit is set so that no fuel damage is calculated to occur if the 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 power.
Therefore, the fuel cladding integrity Safety Limit is defined as the CPR in the limiting fuel assembly for which more than 99.9/o 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 power correlation.
The critical power correlation is valid over the range of conditions used in the tests of the data used to develop the correlation.
The required input to the statistical model are the uncertainties listed in Bases Table B2.1.2-1 and the nominal values of the core parameters listed in Bases Table B2.1.2-2.
The bases for the uncertainties in the core parameters and the basis for the uncertainty in the critical power correlation are given in Reference 1.
The power distribution is based on a typical 764 assembly core in which the rod pattern was arbitrarily chosen to produce a skewed power distribution having the greatest number of assemblies at the highest power levels.
The worst distribution during any fuel cycle would not be as severe as the distribution used in the analysis.
~Rf risc 1.
General Electric Standard Application for Reactor Fuel, NEDE-24011-P-A (latest approved revision).
NINE MILE POINT - UNIT 2 82-2 Amendment No.
4'5
g5JANTITY Feedwater Flow BA E
TABLE 82.1.2-1 N ERTAINTIES ED IN THE DETERMINATION F THE F EL LADDING AFETY LIMIT'TANDARD DEVIATION F P INT 1.76 Feedwater Temperature Reactor Pressure Core Inlet Temperature Core Total Flow Channel Flow Area Friction Factor Multiplier Channel Friction Factor Multiplier TIP Readings R Factor Critical Power 0.76 0.5 0.2 2.5 3.0 10.0 5.0 8.7 1.6 3.0 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.
NINE MILE POINT - UNIT 2 B2-3 Amendment No.~
46
BASES TABLE B2.1.2-2 NOMINALVAL E F PARAMETERS" SED IN THE STATISTI AL ANALYSIS OF F EL LADDING INTEGRITY AFETY LIMIT PARAMETER THERMAL POWER Core Flow Dome Pressure Bundle Enrichment R-Factor:
0 - 10 GWD/ST 10 - 15 GWD/ST
> 15 GWD/ST VAilAE 3293 MW 102.5 Mlb/hr 1005 psig 3.0 Wt % U-235 0.91 5 0.954 0.954 The values in this table are for a representative plant.
NINE MILE POINT - UNIT 2 B2-4 Amendment No. 46
P WER Dl TRIB TI N LIMIT BA E
4.2.
MINIM M RITICAL P WER RATIO The required operating limit MCPRs at steady-state operating conditions as specified in Specification 3.2.3 are derived from the established fuel cladding integrity Safety Limit MCPR, and an analysis of abnormal operational transients.
For any abnormal operating transient analysis evaluation with the initial condition of the reactor being at the steady-state operating limit, it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient, assuming instrument trip setting given in Specification 2.2.
To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting transients have been analyzed to determine which result in the largest reduction in CRITICAL POWER RATIO (CPR).
The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease.
The evaluation of a given transient begins with the system initial parameters shown in USAR Tables 15.0-3 and A15.0-4 that are input to a GE-core dynamic behavior transient computer program.
The codes used to evaluate transients are discussed in Reference 2.
The principal result of this evaluation is the reduction in MCPR caused by the transients.
The purpose of the K, factor specified in the CORE OPERATING LIMITS REPORT is to define operating limits at other than rated core flow conditions.
At less than 100% of rated flow, the required MCPR is the product of operating limit MCPR and the K, factor.
The K, factors assure that the Safety Limit MCPR will not be violated.
The K, factors are calculated as described in Reference 2.
I At THERMAL POWER levels less than or equal to 25% of RATED THERMAL POWER, the reactor will be operating at minimum recirculation pump speed and the moderator void content will be very small.
For all designated control rod patterns which may be employed at this point, operating plant experience indicates that the resulting MCPR value is in excess of requirements by a considerable margin.
During initial startup testing of the plant, an MCPR evaluation will be made at 25% of RATED THERMAL POWER level with minimum recirculation pump speed.
The MCPR margin will thus be demonstrated such that future MCPR evaluation below this power level will be shown to be unnecessary.
The daily requirement for calculating MCPR when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes.
The requirement for calculating MCPR within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER ensures thermal limits are met after power distribution shifts, NINE MILE POINT - UNIT 2 B 3/4 2-2 Amendment No. PV 46
PARAMETER
~PI n:
BASES TABLE B.2.1-1 SIGNIFI ANT INP T PARAMETER T
THE LOS F- 0 LANT A IDENT ANALYSI VAL E 1.
Core THERMAL POWER.....~....~................,.........,..................
2.
Vessel Steam Output 3.
Vessel Steam Dome Pressure 4.
Design Basis Recirculation Line Break Area for:
3461 MWt""
which corresponds to 105% of rated steam flow 15.0 x 10'bm/hr which corresponds to 105% of rated steam flow 1055 psia a.
Large Breaks.....
4+2
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~ ~ ~ ~ ~ ~ ~ ~
3o1 IL b
Small Breaks...........~...........
~..... ~..,
0.09 ft FUEL TYPE Initial Core Reload 8x8 8x8 1 3.4 144 14 1.4 1.20 1.20 PEAK TECHNICAL INITIAL SPECIFICATION DESIGN MINIMUM LINEAR HEAT AXIAL CRITICAL FUEL BUNDLE GENERATION RATE PEAKING POWER GEOMETRY (I<W/ft)
FACTOR RATIO' more detailed listing of input of each model and its source is presented in Section II of Reference 1 and subsection 6.3.3 of the USAR.
This power level meets the Appendix K requirement of 102%.
The core heatup calculation assumes a bundle power consistent with operation of the highest powered rod at 102% of its Technical Specification LINEAR HEAT GENERATION RATE limit.
For single recirculation loop operation, loss of nucleate boiling is assumed at 0.1 second after LOCA regardless of initial MCPR.
NINE MILE POINT - UNIT 2 B 3/4 2-3 Amendment No.
<6
I g I
P WER DI TRIB TI N LIMIT BA E
4.2.
MINIM M CRITI AL P WER RATIO 3/4.2.3 (Continued) while still allotting time for the power distribution to stabilize.
The requirement for calculating MCPR after initially determining that a LIMITINGCONTROL ROD PATTERN exists ensures MCPR will be known following a change in THERMAL POWER or power
- shape, and therefore avoid operation while exceeding a thermal limit.
4.2.4 LINEAR HEAT ENERATION RATE This specification assures that the linear heat generation rate (LHGR) in any rod is less than the design linear heat generation rate even if fuel pellet densification is postulated.
The daily requirement for calculating LHGR when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER is sufficient, since power distribution shifts are very slow when there have not been significant power or control rod changes.
The requirement to calculate LHGR within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER ensures thermal limits are met after power distribution shifts while still allotting time for the power distribution to stabilize.
The requirement for calculating LHGR after initially determining a LIMITING CONTROL ROD PATTERN exists ensures that LHGR will be known following a change in THERMAL POWER or power shape that could place operation exceeding a thermal limit.
~Rf rien 1
~
General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix K, NEDE-20566, latest approved revision.
2.
General Electric Standard Application for Reactor Fuel, NEDE-24011-P-A, latest approved revision.
NINE MILE POINT - UNIT 2 B 3/4 2-4 Amendment No.~ 46