ML20206U487
| ML20206U487 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 05/17/1999 |
| From: | Clifford J NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20206U493 | List: |
| References | |
| NUDOCS 9905250238 | |
| Download: ML20206U487 (6) | |
Text
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4 UNITED STATES p
g NUCLEAR REGULATORY COMMISSION
'f WASHINGTON, D.C. 20555Am
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PECO ENERGY COMPANY DOCKET NO. 50-353 UMERICK GENERATING STATION UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 08 License No. NPF-85
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1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by PECO Energy Company (the licensee) dated January 12,1999, as supplemented January 29 and March 10,1999, complies with the standards and requirements of the Atomic Energy Act of 1954, as
)
amended....
7 and the Commission's rules and regulations set forth in j
10 CFR Chaptw 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 activitics 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.
9905250238 990517 PDR ADOCK 05000353 P
i i 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-85 is hereby amended to read as follows:
Technical Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. og, are hereby incorporated into this license. PECO Energy Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
g 3.
This license amendment is effective as of its date of issuance and shall be implemented prior to startup following completion of the April 1999 refueling outage for Limerick Generating Station, Unit 2.
FOR THE NUCLEAR REGULATORY COMMISSION
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' Or ames W. Clifford, Chief, Section 2 j
Project Directorate 1 Division of Licensing Project Management Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: ity 17,1999 I
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1 ATTACHMENT TO LICENSE AMENDMENT NO. A FACILITY OPERATING LICENSE NO. NPF-85 DOCKET NO. 50-353 Replace the following pages of the Appendix A Technical Specifications with the attached 3
revised pages as indicated. The revised pages are identified by Amendment number and contain marginallines indicating the area of change.
Remove Insert 3/4 4-7 3/4 4-7 B 3/4 4-2 B 3/4 4-2 B 3/4 5-1 B 3/4 5-1
REACTOR COOLANT SYSTEM 3/4.4.2 SAFETY / RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.2 The safety valve function of at least 12 of the following reactor coolant system l
safety / relief valves shall be OPERABLE with the specified code safety valve function lift settings:*#
4 safety / relief valves @ 1170 ps g i3%
5 safety / relief valves @ 1180 ps g i3%
5 safety / relief valves @ 1190 ps g 13%
APPLICABILITY:
OPERATIONAL CONDITIONS 1, 2, and 3.
ACTION:
a.
With the safety valve function of one or more of the above required safety / relief valves inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
b.
With one or more safety / relief valves stuck open, provided that suppression pool average water temperature is less than 105*F, close the stuck open safety / relief valve
- if unable to close the stuck open valvejs within 2 minutes or if su) pre (s) ion pool average water temperature is 110 F)or ss greater, place tie reactor mode switch in the Shutdown position.
c.
With one or more safety / relief valve acoustic monitors inoperable, restore the inoperable acoustic monitors to OPERABLE status within 7 days or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
SURVEILLANCE RE0VIREMENTS l
4.4.2.1 The acoustic monitor for each safety / relief valve shall be demonstrated OPERABLE with the setpoint verified to be 0.20 of the full open noise level ## by performance of a:
a.
CHANNEL FUNCTIONAL TEST at least once per 92 days, and a b.
CHANNEL CALIBRATION at least once per 24 months **.
4.4.2.2 At least 1/2 of the safety relief valves shall be removed, set pressure tested and reinstalled or replaced with spares that have been previously set pressure tested and stored in accordance with manufacturer's recommendations at least once per 24 months, and they shall be rotated such that all 14 safety relief valves are removed, set pressure tested and reinstalled or replaced with spares that have been previously set pressure tested and stored in accordance with manufacturer's recommendations at least once per 54 months. All safety valves will be recertification tested to meet a i1% tolerance prior to returning the valves to service.
The lift setting pressure shall correspond to ambient conditions of the o
valves at nominal operating temperatures and pressures.
Co The provisions of Specification 4.0.4 are not applicable provided the Surveillance is performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure is adequate to perform the test.
Up to 2 inoperable valves may be replaced with spare OPERABLE valves with lower setpoints until the next refueling Initial setting shall be in accordance with the manufacturer's recommendation. Adjustment to the valve full open noise level shall be accomplished during the startup test program.
LIMERICK - UNIT 2 3/4 4-7
&mtmt tb. ?t, 33,34,51,93
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BASES RECIRCULATION SYSTEM (Continued)
Plant specific calculations can be performed to determine an applicable region for monitoring neutron flux noise levels.
In this case the degree of conservatism can be reduced since plant to plant variability would be eliminated.
In this case, adequate margin will be assured by monitoring the region wl.ich has a decay ratio greater than or equal to 0.8.
Neutron flux noise limits are also established to ensure early detection of limit cycle neutron flux oscillations.
BWR cores typically operate with neutron flux noise caused by random boiling and flow noise.
Typical neutron flux noise levels of 1-12% of rated power have been reported for the range of low to high recirculation loop (peak-to-peak)both single and dual flow during recirculation loop operation. Neutron flux noise levels which significantly bound these values are considered in the thermal / mechanical design of GE BWR fuel and are found to be of negligible consequence.
In addition stability testsatoperatingBWRshavedemonstratedthatwhenstabilityrelatedneutron flux limit cycle oscillations occur they result in peak-to-peak neutron flux limit cycles of 5-10 times the typical values. Therefore, actions taken to reduce neutron flux noise levels exceeding three (3) times the typical value are sufficient to ensure early detection of limit cycle neutron flux oscillations.
Typically, neutron flux noise levels show a gradual increase in absolute magnitude as core flow is increased (constant control rod pattern) with two reactor recirculation loops in operation.
Therefore, the baseline neutron flux noise level obtained at a specific core flow can be applied over a range of j
core flows. To maintain a reasonable variation between the low flow and high flow end of the flow range, the range over which a specific baseline is applied should not exceed 20% of rated core flow with two recirculation loops in i
operation. Data from tests and operating plants indicate that a range of 20%
of rated core flow will result in approximately a 50% increase in neutron flux noise level during operation with two recirculation loops.
Baseline data should be taken near the maximum rod line at which the majority of operation i
will occur. However, baseline data taken at lower rod lines will result in a conservative value since the neutron flux no(i.e. lower power) ise level is j
proportional to the power level at a given core flow.
3/4.4.2 SAFETY / RELIEF VALVES l
The safety valve function of the safety / relief valves operates to prevent the reactor coolant system from being pressurized above the Safety Limit of 1325 psig in accordance with the ASME Code. A total of 12 OPERABLE safety /
l relief valves is required to limit reactor pressure to within ASME III allow-able values for the worst case upset transient.
Demonstration of the safety / relief valve lift settings will occur only during shutdown.
The safety / relief valves will be removed and either set pressure tested or replaced with spares which have been previously set pres-sure tested and stored in accordance with manufacturers recommendations in the specified frequency.
LIMERICK - UNIT 2 B 3/4 4-2 Amtrnt % 98
3/4.5 EMERGENCY CORE COOLING SYSTEM BASES 3/4.5.1 and 3/4.5.2 ECCS - OPERATING and SHVTDOWN The core spray system is provided to assure that th(CSS), together with the LPCI mode of the RHR system, e core is adequately cooled following a loss-of-coolant accident and provides adequate core cooling capacity for all break sizes up to and including the double-ended reactor recirculation line break, and for smaller breaks following depressurization by the ADS.
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The CSS is a primary source of emergency core cooling after the reactor vessel is depressurized and a source for flooding of the core in case of accidental draining.
The surveillance requirements provide adequate assurance that the CSS will be OPERABLE when required. Although all active components are testable and full flow can be demonstrated by recirculation through a test loop durin reactor operation, a complete functional test requires reactor shutdown.g The pump discharge pip ng is maintained full to prevent water hammer damage to pip ng and to star cooling at the earliest moment The low pressure coolant injection (LPCI mode of the RHR system is provided to assure that the core is adequately) cooled following a loss-of-coolant accident.
Four subsystems, each with one pump, provide adequate core flooding for all break sizes up to and including the double-ended reactor J
recirculation line break, and for small breaks following depressurization by the ADS.
The surveillance requirements provide adequate assurance that the LPCI system will be OPERABLE when required. Although all active components are testable and full flow can be demonstrated by recirculation through a test loop during reactor operation, a complete functional test requires reactor shutdown.
The pump discharge piping is maintained full to prevent water hammer damage to piping and to start cooling at the earliest moment.
The high pressure coolant injection that the reactor core is adequately cooled (HPCI) system is provided to assure to limit fuel clad temperature in the event of a small break in the reactor coolant system and loss of coolant which does not result in rapid depressurization of the reactor vessel.
The HPCI system permits the reactor to be shut down while maintaining sufficient reactor vessel water level inventory until the vessel is depressurized.
The HCPI system continues to operate until reactor vessel pressure is below the i
pressure at which CSS operation or LPCI rrode of the RHR system operation i
maintains core cooling.
The capacit of the system is selected to provide the re uired core cooling.
The HPCI pump is designed to deliver greater than or equal to $600 gpm at reactor j
aressures between 1182 and 200 psig and is capable of delivering at least 5000 gpm l
3etween 1182 and 1205 psig.
Initially, water from the condensate storage tank is used instead of injecting water from the suppression pool into the reactor, but no credit is taken in the safety analyses for the condensate storage tank water.
LIHERICK - UNIT 2 B 3/4 5-1 kmhnt No, M, 93 J