ML20070K969
| ML20070K969 | |
| Person / Time | |
|---|---|
| Site: | Farley  (NPF-02-A-109, NPF-2-A-109) |
| Issue date: | 07/22/1994 |
| From: | Matthews D Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20070K973 | List: |
| References | |
| NUDOCS 9407280115 | |
| Download: ML20070K969 (6) | |
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E UNITED STATES NUCLEAR REGULATORY COMMISSION "g ~,,,/
WASHINGTON, D.C. 20555 6 1 SOUTHERN NUCLEAR OPERATING COMPANY. INC.
DOCKET NO. 50-348 JOSEPH M. FARLEY NUCLEAR PLANT. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 109 License No. NPF-2 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Southern Nuclear Operating Company, Inc. (Southern Nuclear), dated June 17, 1994, 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 license 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-2 is hereby amended to read as follows:
i 9407280115 940722 PDR ADOCK 05000348 P
PDR l
. (2)
Technical Soecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 109, are hereby incorporated into
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the license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 30 days of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION 1'
G L
David B. Matthews, Director Project Directorate II-l Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
July 22, 1994 m
ATTACHMENT TO LICENSE AMENDMENT NO.109 FACILITY OPERATING LICENSE NO. NPF-2 DOCKET NO. 50-348 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages as indicated.
The revised areas are indicated by marginal lines.
Remove Paaet insert Paaes B 2-2 B 2-2 3/4 2-8 3/4 2-8 B3/4 2-4 B3/4 2-4 e
e.
Safety Limits s
Bases The curves of Figures 2.1-1 and 2.1-2 are based on the most limiting result using an enthalpy hot channel factor, FN AH, of 1.70 for l
VANTAGE 5 fuel and an FNAH of 1.55 for LOPAR fuel and a reference cosine with a peak of 1.55 for axial power shape. An allowance la included for an increase in FNAH at reduced power based on the expression:
N l
F ag = 1.70 [1 + 0.3 (1 - P)] for VANTAGE 5 fuel and FN H = 1.55 [1 + 0.3 (1 - P)) for LOPAR fuel where P is the fraction of RATED THERMAL POWER.
These limiting heat flux conditions are higher chan those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the f t
(delta 4; function of the Overtemperature trip. When the axial power imbalance is not within the tolerance, the axial power imbalance effect on the Overtemperature delta T trips will reduce the setpoints to provide protection consistent with core safety limits.
2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.
The reactor pressure vessel, pressurizer and the reactor coolant system piping and fittings are designed to Section III of the ASME Code for Nuclear Power Plant which permits a maximum transient pressure of 110% (2735 psig) of design pressure.
The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated code requirements.
The entire Reactor Coolant System is hydrotested at 3107 psig, 125% of design pressure, to demonstrate integrity prior to initial operation.
EARLEY - UNIT 1 B 2-2 AMENDMENT NO. 37,73,87,92,109
POWER DISTRIBUTION LIMITS 3/4.2.3 NUCLEAR ENTHALPY HOT CHANNEL FACTOR -
LIMITING CONDITION FOR OPERATION 3.2.3 FNAH shall be limited by the following relationships FN l
AH s 1.70 [1 + 0.3 (1 - P)] for VANTAGE 5 fuel and FNAH 5 1.55 [1 + 0.3 (1 - P)] for LOPAR fuel where P=
THERMAL POWER RATED THERMAL POWER APPLICABILITY:
MODE 1 ACTION:
With FNAH exceeding its limits a.
Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> either:
1.
Restore FN H to within the above limits and demonstrate through in-core mapping that FNAH is within its limit within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of exceeding the limit, or 2.
Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux - High Trip setpoints to C 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, and l
b.
Demonstrate through in-core mapping, if not previously performed per a.1 above, that FNAH is within its limit within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after exceeding the limit or reduca THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, and I
Identify and correct the cause of the out of limit condition prior c.
to increasing THERMAL POWER above the reduced limit required by a or b, above; subsequent POWER OPERATION may proceed provided that FNAH 1s demonstrated through in-core mapping to be within its limit at a nominal 50% of RATED THERMAL POWER prior to exceeding this THERMAL POWER, at a nominal 75% of RATED THERMAL POWER prior to exceeding this THERMAL POWER and within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after attaining 95% or greater RATED THERMAL POWER.
i FARLEY - UNIT 1 3/4 2-8 AMENDMENT NO.
26,37,64,92,109
2 s
POWER DISTRIBUTION LIMITS BASES FNAH will be maintained within its limits provided conditions a. though d.
above are maintained. The relaxation of FNAH as a function of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits.
When an Fg measurement is taken, an allowance for both experimental error and manufacturing tolerance must be made. An allowance of 5% is appropriate for a full core map taken with the incore detector flux mapping system and a 3%
allowance is appropriate for manufacturing tolerance.
When FNAH is measured, experimental error must be allowed for and 4% is the appropriate allowance for a full core map taken with the incore detection system. The specified limit for FNAH contains an 8% allowance for uncertainties.
The 8% allowance is based on the following considerations:
Abnormal perturbations in the radial power shape, such as from rod a.
misalignment, affect FN H more directly than Fg, b.
Although rod movement has a direct influence upon limiting Fg to within l
its limit, such control is not readily available to limit FN AH, and i
Errors in prediction for control power shape detected during startup c.
t physics tests can be compensated for in Fg by restricting axial flux dis t ribution This compensation for FN dH is less readily available.
N If F ag exceeds its limit, the unit will be allowed 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to restore NF ag to within its limits. This restoration may, for example, involve realigning any misaligned rods or reducing power enough to bring F AH within its N
power dependent limit. When the FNAH limit is exceeded, the DNBR limit is not
{
likely violated in steady state operation, because events that could J
significantly perturb the FNAH value, e.g., static control rod misalignment, are considered in the safety analyses.
However, the DNBR limit may be violated if a N
DNB limiting event occurs while F ag is above its limit.
The increased allowed action time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> provides an acceptable time to restore FNAH to within its lindts without allowing the plant to remain in an unacceptable condition for an extended period of time.
I once corrective action has been taken, e.g., realignment of misaligned rods or reduction of power, an incore flux map must be obtained and the measured value of FN AH verified not to exceed the allowed limit.
Twenty additional hours are provided to perform this task above the four hours allowed by Action Statement 3/4.2.3.a.
The completion time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is acceptable because of the low probability of having a DNB limiting event within this 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period and, in the etent that power is reduced, an increase in DNB margin is obtained at lower power levels. Additionally, operating experience has indicated that this completion time is sufficient to obtain the incore flux map, perform the required calculations, and evaluate FN g, FARLEY - UNIT 1 B 3/4 2-4 AMENDMENT NO. 26,64,92,109
.