ML17059A102
| ML17059A102 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 11/09/1993 |
| From: | Capra R Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17059A103 | List: |
| References | |
| NUDOCS 9311170327 | |
| Download: ML17059A102 (24) | |
Text
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-000I NIAGARA MOHAWK POWER CORPORATION DOCKET NO. 50-410 NINE MILE POINT NUCLEAR STATION UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.
51 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 May 21,
- 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 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 paragraph 2.C.(2) of Facility Operating License No.
NPF-69 is hereby amended to read as follows:
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(2)
Tec nical 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.
51 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 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:
November 9,
1993 Robert A. Capra, Director Project Directorate I-I Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO.
TO FACILITY OPERATING LICENSE NO. NPF-69 DOCKET NO. 50-410 Revise Appendix A as follows:
Remove Pa es 2-3 3/4 2-2 3/4 3-60 3/4 3-62 3/4 3-63 3/4 3-64 3/4 3-65 B3/4 2-1 6-22 Insert Pa es 2-3 3/4 2-2 3/4 3-60 3/4 3-62 3/4 3-63 3/4 3-64 3/4 3-65 B3/4 2-1 6-22
Zzm K
Im 0Oz I
z TABLE 2.2.1-1 FUNCTIONAL NIT 1.
Neutron Flux - High 2.
Average Power Range Monitor:
TRIP SETPOINT
<120/1 25 divisions of full scale ALL WABLEVAL E
<122/1 25 divisions of full scale REA R PR TECTI N
Y TEM IN TR MENTATION ET INT a.
Neutron Flux - Upscale, Setdown b.
Flow-Biased Simulated Thermal Power - Upscale 1)
Flow-Biased 2)
High-Flow-Clamped
<15% of RATED THERMALPOWER
<0.58 (W-b,W)" + 59%, with a maximum of <113.5% of RATED THERMALPOWER
<20% of RATED THERMALPOWER
<0.58 (W-hW)" + 62%, with a maximum of <115.5% of RATED THERMALPOWER c.
Fixed Neutron Flux - Upscale
<118% of RATED THERMALPOWER
<120% of RATED THERMALPOWER d.
Inoperative NA 7.
Drywell Pressure - High
<1.68 psig 3.
Reactor Vessel Steam Dome Pressure s1037 psig
- High 4.
Reactor Vessel Water Level-Low, a159.3 in. above instrumentzero*
Level 3 5.
Main Steam Line Isolation Valve -
s8% closed Closure 6.
Main Steam Line Radiation+ - High
<3.0 x full-power background a1057 psig a157.8 in. above instrument zero
<12% closed
<3.6 x full-power background
<1.88 psig z0 See Bases Figure B 3/4 3-1.
(a)
The Average Power Range Monitor Scram Function varies as a function of recirculation loop drive flow (W). hW is defined as the difference in indicated drive flow (in percent of drive flow which produces rated core flow) between two loop and single loop operation at the same core flow. bW=O for two loop operation. bW=5% for single loop operation.
(b)
See footnote (**)to Table 3.3.2-2 for trip setpoint during hydrogen addition test.
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ER Dl IB Tl N LIMIT WER RAN E M NIT R
ET NT LIMITIN NDITI N F
R PERATI N
3.2.2 The Average Power Range Monitor (APRM) flow-biased simulated thermal power-upscale scram trip setpoint (S) shall be established according to the relationship specified in the CORE OPERATING LIMITS REPORT.
0 to 25% of RATED THERMALPOWER.
QQTIQN; With the APRM flow-biased simulated thermal power-upscale scram trip setpoint less conservative than the value shown in the Allowable Value column for S, as above determined, initiate corrective action within 15 minutes and adjust S to be consistent with the Trip Setpoint value* within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> or reduce THERMALPOWER to less than 25% of RATED THERMALPOWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
RVEILLAN E RE IREMENT 4.2.2 The FRACTION OF RATED THERMALPOWER (FRTP) and the CORE MAXIMUM FRACTION OF LIMITINGPOWER DENSITY (CMFLPD) shall be determined, the value of T**
calculated, and the most recent actual APRM flow-biased simulated thermal power-upscale scram trip setpoint verified to be within the above limit or adjusted, as required:
I a.
At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, b.
Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMALPOWER increase of at least 15% of RATED THERMALPOWER, and c.
Initiallyand at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating with CMFLPD greater than or equal to FRTP.
d.
The provisions of Specification 4.0.4 are not applicable.
With CMFLPD greater than the FRTP rather than adjusting the APRM setpoints, the APRM gain may be adjusted so that APRM readings are greater than or equal to 100% times CMFLPD provided that the adjusted APRM reading does not exceed 100% of RATED THERMALPOWER and a notice of adjustment is posted on the reactor control panel.
Definition of T is specified in the CORE OPERATING LIMITSREPORT.
NINE MILE POINT - UNIT 2 3/4 2-2 Amendment No.~ 51
NTR L R D BL K IN TR MENTATI N TRIPF N
I N
MINIMUM APPLICABLE OPERABLE CHANNELS OPERATIONAL ~au a.
Upscale b.
Inoperative c.
Downscale 1*
1*
1*
60 60 60 2.
R M
i a.
Detector Not Full In (b) b.
Upscale(c) c.
Inoperative(c) d.
Downscale(d) 3 2(f) 61 61 61 61 61 61 61 61 3.
In rm i
R Mni a.
Detector Not Full In b.
Upscale c.
Inoperative d.
Downscale(e) 4.
r m Di r
V m
Water Level - High, Float Switch 2
5.
R In R
ir I 'nFIw a.
Upscale b.
Inoperative c.
Comparator
'6.
R w'
a.
Shutdown Mode b.
Refuel IVlode 2,5 2,5 2,5 2,5 1, 2, 5**
3,4 5
61 61 61 61 62 62 62 62 62 62 NINE MILE POINT - UNIT 2 3/4 3-60 Amendment No.~ 51
,I
zzm K
Im-DOz I
Cz IPF N
1.
R a.
Upscale b.
Inoperative c.
Downscale D
NA
+5% of RATED THERMALPOWER AL WABL VAL E NA
>3% of RATED THERMALPOWER r
R M
i 3.
4.
a.
Detector Not Full In b.
Upscale c.
Inoperative d.
Downscale In rm i
R Moni a.
Detector Not Full In b.
Upscale c.
Inoperative d.
Downscale r
Di V
m NA
<1 x 10~ cps NA h3 cps**
NA a108/1 25 divisions of full scale NA
>5/1 25 divisions of full scale NA
~1.6 x 10~ cps NA z1.8 cps~*
NA a110/1 25 divisions of full scale NA a3/1 25 divisions of full scale Water Level - High, Float Switch a16.5 in.
a39.75 in.
z0
l
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Im 90z Cz hJ TRIP FUN I
N 5.
a.
Upscale b.
Inoperative c.
Comparator Reac or C n
em
-2 (Continued)
ALL WABLEVAL E s108% rated flow NA s10% flow deviation a111% rated flow NA
~11% flow deviation NTR L R D BLO K IN R MENTATION ET INT 6.
Reactor Mod Switch a.
Shutdown Mode b.
Refuel Mode NA NA NA NA Ptz0 Specified in the CORE OPERATING LIMITS REPORT For fuel loading and startup from refueling the count rate may be less than 3 cps if the following conditions are met: the signal to noise ratio is greater than or equal to 5, and the signal is greater than 1.3 cps.
z Ill K
IlTl 00z I
Cz hJ TRIP FUN I
N R
Bl k M ni r
TABLE4..
1 CHANNEL CHECK CHANNEL FUNCTIONAL ML OPERATIONAL CONDITIONS FOR WHICH CHANNEL SURVEILLANCE Mill NTR L R D BL K IN TR MENTATI N RVEILLAN E RE IREMENT a.
b.
C.
Upscale Inoperative Downscale NA NA NA S/U(b)(c), 0(c)
S/U(b)(c), 0(c)
S/U(b)(c), 0(c) 0 NA 0
1
- 1*
1*
2.
a.
b.
C.
d.
rce Ra M ni or Detector Not Full In Upscale Inoperative Downscale NA NA NA NA S/U(b), W S/U(b), W S/U(b), W S/U(b), W NA 0
NA 0
2,5 2,5 2,5 2,5 Ptz0 3.
4 Inerm i
R n e Moni a.
Detector Not Full In b.
Upscale c.
Inoperative d.
Downscale ra Di har V
I me Water Level - High, Float Switch NA NA NA NA NA S/U(b), W S/U(b), W S/U(b), W S/U(b), W 0
NA 0
NA 0
2,5 2,5 2,5 2,5 1,2 5**
zzm K
Im 00z I
Cz 5.
Reac or I n mR ir lai Fl w CHANNEL CHECK CHANNEL FUNCTIONAL OPE RATIONAI CONDITIONS FOR WHICH CHANNEL SURVEILLANCE
~*II Mill NTR L R D BL K IN MENTATION RVEILLAN E RE IREMENT 4)
Ol Ql 6.
a.
Upscale b.
Inoperative c.
Comparator Reactor Mode wi h
a.
Shutdown Mode b.
Refuel Mode NA NA NA NA NA S/U(b), 0 S/U(b), 0 S/U(b), 0 0
NA 0
NA NA 3,4 5
3 Q.
z0
I TRIB I
N LIMIT BA E
The specifications of this section assure that the peak cladding temperature following the postulated design basis loss-ofwoolant accident will not exceed the 2200'F limitspecified in 10 CFR 50.46.
4.2.1 AVERA E PLANAR LINEAR HEAT ENERATI N RATE The peak cladding temperature (PCT) following a postulated losswf-coolant accident is primadly a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is dependent only secondarily on the rod-to-rod power distribution within an assembly.
The peak clad temperature is calculated assuming an LHGR for the highest powered rod which is equal to or less than the design LHGR corrected for densification.
This LHGR times 1.02 is used in the heatup code along with the exposure-dependent steady-state gap conductance and rod-trod local peaking factor. The limiting value for APLHGR is specified in the CORE OPERATING LIMITS REPORT for twmecirculation-loop operation.
The calculational procedure used to establish the APLHGR specified in the CORE OPERATING LIMITSREPORT is based on a loss-of-coolant accident analysis.
The analysis was performed using General Electric (GE) calculational models which are consistent with the requirements of Appendix K to 10 CFR 50. A complete discussion of each code employed in the analysis is presented in Reference 1.
A list of the significant plant input parameters to the loss-ofwooiant accident analysis is presented in Bases Table B3.2.1-1.
For plant operations with single recirculation loop the MAPLHGR limits are specified in the CORE OPERATING LIMITS REPORT.
The constant factor is derived from LOCA analyses initiated from single loop operation to account for earlier boiling transition at the limiting fuel node compared to the standard LOCA evaluations.
4 2 PRM ET INT The fuel cladding integrity Safety Limits of Specification 2.1 were based on a power distribution which would yield the design LHGR at RATED THERMALPOWER.
The flow-biased simulated thermal powerwpscale scram setting of the APRM instruments must be adjusted to ensure that the MCPR does not become less than the fuel cladding integrity safety limit or that greater than or equal to 1% plastic strain does not occur in the degraded situation, The scram setpoint is I
adjusted in accordance with the formula in this specification when the combination of THERMAL POWER and CMFLPD indicates a higher peaked power distribution to ensure that an LHGR transient would not be increased in the degraded condition.
NINE MILE POINT - UNIT 2 B3/4 2-1 Amendment No.~
51
V
DMINISTRATIVECONTROLS EMIANNUALRADI A TIVE EFFLUENT RELEA E REPORT 6.9.1.8 (Continued)
The Semiannual Radioactive Effluent Release Reports shall include any changes made during the reporting period to the PROCESS CONTROL PROGRAM (PCP) and to the OFFSITE DOSE CALCULATIONMANUAL(ODCM), pursuant to Specifications 6.13 and 6.14, respectively, as well as any major change to liquid, gaseous, or solid radwaste treatment systems pursuant to Specification 6.15.
It shall also include a listing of new locations for dose calculations and/or environmental monitoring identified by the land use census pursuant to Specification 3.12.2.
The Semiannual Radioactive Effluent Release Reports shall also include the following:
an explanation of why the inoperability of liquid or gaseous effluent monitoring instrumentation was not corrected within the time specified in Specification 3.3.7.9 or 3.3.7.10, respectively, and a description of the events leading to liquid holdup tanks exceeding the limits of Specification 3.11.1.4.
CORE OPERATING LIMITS REPORT 6.9.1.9 Core operating limits shall be established prior to each reload cycle, or prior to any remaining portion of a reload cycle for the following:
1)
The AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) for Specification 3.2.1.
2)
The Average Power Range'Monitor (APRM) flow-biased simulated thermal power-upscale scram trip setpoint for Specification 3.2.2.
I 3)
The K, core flow adjustment factor for Specification 3.2.3.
4)
The MINIMUMCRITICAL POWER RATIO (MCPR) for Specification 3.2.3.
5)
The LINEAR HEAT GENERATION RATE (LHGR) for Specification 3.2.4.
6)
Control Rod Block Instrumentation Setpoint for the rod block monitor upscale trip setpoint and allowable value for Specification 3.3.6.
I and shall be documented in the CORE OPERATING LIMITS REPORT.
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 document.
1)
General Electric Company Analytical Model for Loss-of-Coolant Analysis in Accordance with 10 CFR 50, Appendix K, NEDE-20566 latest approved revision.
NINE MILE POINT - UNIT 2 6-22 Amendment No.gg 51
't