ML20065J611
| ML20065J611 | |
| Person / Time | |
|---|---|
| Site: | Peach Bottom  |
| Issue date: | 04/07/1994 |
| From: | Chris Miller Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20065J613 | List: |
| References | |
| NUDOCS 9404180323 | |
| Download: ML20065J611 (9) | |
Text
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UNITED STATES l'
Mi NUCLEAR REGULATORY COMMISSION g.
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WASHINGTON, D.C. 20555-0001 PHILADELPHIA ELECTRIC COMPANY PUBLIC SERVICE ELECTRIC AND GAS COMPANY-DELMARVA POWER AND LIGHT COMPANY ATLANTIC CITY ELECTRIC COMPANY DOCKET NO. 50-277 PEACH BOTTOM ATOMIC POWER STATION. UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.188 License No. DPR-44 1.
The Nuclear Regulatory Comission (the Commission) has. found that:
A.
The application for amendment by Philadelphia Electric Company, et.
al. (the licensee) dated May 25, 1993, as supplemented on March 11, 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 Comm.ission's regulations; D.
The issuance of this amendment will not be inimical to the common defense and security or-to the health ~or 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. DPR-44 is hereby.
amended to read as follows:
9404180323 940407 PDR ADOCK 05000277 P
PDRf s..
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. (2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.188, are hereby incorporated in the license.
PECO shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION N
A Charles L. Miller, Director Project Directorate I-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: April 7, 1994 k
4
ATTACHMENT TO LICENSE AMEN 0 MENT N0.188 FACILITY OPERATING LICENSE NO. DPR-44 DOCKET NO. 50-277 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages.
The revised areas are indicated by marginal lines.
Remoyg Insert 39 39 49 49 50 50 71b 71b 226 226 229 229 i
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Ur.it 2 PBAPS-NOTES FOR TABLE 3.1.1 1.
There shall be two operable or tripped trip systems for each function.
If the minimum number of operable sensor channels for a trip system cannot be met, the affected trip system shall be placed in the safe (tripped) condition, or the appropriate actions listed below shall be taken.
A.
Initiate insertion of operable rods and complete insertion of all operable rods within four hours.
B.
Reduce power level to IRM range and place mode switch in the start up position within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
C.
Reduce turbine load and close main steam line isolation valves within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
D.
Reduce power to less than 30% rated.
2.
Permissible to bypass, in refuel and shutdown positions of the reactor mode switch.
3.
Deleted.
1 4.
Bypassed when reactor thermal power is less than 30% of rated as indicated by turbine first stage pressure.
5.
IRMs are bypassed when APRMs are onscale and the reactor mode switch is in the run position.
6.
The design permits closure of any two lines without a scram j
being initiated.
7.
When the reactor is suberitical and the reactor water temperature is less than 212 degrees F, only the following trip functions need to be operable.
A.
Mode switch in shutdown B.
Manual scram C.
High flux IRM D.
Scram discharge instrument volume high level-8.
Not required to be operable when primary containment integrity is not required.
'i 9.,
Not required to be operable when the reactor pressure vessel head is not bolted to the vessel. Amendment No. 33, 108, 121, 188
Unit 2 PBAPS 3.1 RASE 1 (Cont'd) the amount of water which must be accomodated during a
- scram, i
During normal operation the discharge volume is empty; however, should it fill with water, the water discharged to the piping from the reactor could not be accomodated which i
would reruit in slow scram times or partial control rod i
insertion. To preclude this occurrence, level switches have been provided in the instrument volume which alarm and scram the reactor when the volume of water reaches 50 gallons. As
]
indicated above, there is sufficient volume in the pipin5 to accomodate the scram without impairment of the scram times or amount of insertion of the control rods. This function shuts the reactor down while sufficient volume remains to accomodate the discharged water and precludes the situation in which a scram would be required but not be able to perform its function adequately.
A source range monitor-(SRM) system is also provided to supply additional neutron level information during start-up but has no scram functions (reference paragraph 7.5.4 FSAR).
Thus, the IRM and APRM are required in the " Refuel" and
" Start / Hot Standby" modes. In the power range the APRM system provides required protection (reference paragraph 7.5.7 FSAR). Thus the IRM System is not required in the "Run" mode. The APRM's cover only the power range. The IRMs and APRMs provide adequate coverage in the start-up i
and intermediate range.
The high reactor pressure, high drywell pressure, reactor low water level and scram discharge volume high level scrams are required for Startup and Run modes of plant operation. They are, therefore, required to be operational for these modes of reactor operation.
The requirement to have the scram functions indicated in Table 3.1.1 operable in the Refuel mode assures that shifting to the Refuel mode during reactor power operation does not diminish the protection provided by the reactor protection system.
The turbine condenser low vacuum scram is only required during power operation and must be bypassed to start up the unit. The main condenser low vacuum trip is bypassed except in the run position of the mode switch.
Turbine stop valve closure occurs at.10% of valve closure.
Below 30% of rated reactor thermal power the scram signal due to turbine stop valve closure is bypassed because the flux and pressure scrams are adequate to protect the reactor. Amendment No. 117, 188
9 PBAPS Unit 2 3.1 BASES (Cont'd.)
Turbine control valves fast closure initiates a scram based on pressure switches sensing Electro-Hydraulic Control (EHC) system oil pressure.
The switches are located be-tween fast closure solenoids and the disc dump valves, and are set relative (500<P<850 psig) to the normal EHC oil pressure of 1600 psig gauge that, based on the small system volume, they can rapidly detect valve closure or loss of hy-draulic pressure.
This scram signal is also bypassed when reactor thermal power is less than 30% of rated as indicated by turbine first stage pressure.
The requirement that the IRM's be inserted in the core when the APRH's read 2.5 indicated on the scale in the Startup and Refuel modes assures that there is proper overlap in the neutron monitoring system functions and thus, that adequate coverage is provided for all ranges of reactor operation. Amendment No.188
TABLE 3.2.B (CONTINUED)
Unit 2 INSTRUMENTATION THAT INITIATES OR CONTROLS THE CORE AND CONTAINMENT COOLING SYSTEMS Minimum No.
Number of of Operable Instrument Instrument Trip Function Trip Level Setting Channels Remarks Channels Per Provided by Trip System (1)
Design 2 per 4 kV Emergency Trans-87% (iS%) of
- 1. Trips emergency Bus former Degraded Rated Voltage.
transformer feed voltage (Inverse Tests at 2940 to 4 kV - emer-time - voltage).
volts in 30 seconds gency bus.
e (CV-6)
( 10%)
- 2. Fast transfer permissive.
a-8 l
8 R
8e w
b 4
FDAPS Unit i
l LIMITING CONDITIONS FOR OPERATION SURVEILLANCE REOUIREMENTS 3.10.A Refueling Interlocks 4.10.A.2 (Cont'd)
{
3.
The fuel grapple hoist load fully withdrawn and all switch shall be set at s 1000 other operable rods 1bs.
fully inserted. Alternatively if the remaining control rods 4.
If the frame-mounted are fully inserted and have auxiliary hoist or the monorail-their directional control mounted auxiliary hoist valves electrically dis-is to be used armed, it is sufficient for handling fuel with to demonstrate that j
the head off the reactor the core is subcritical vessel, the load limit with a margin of at j
switch on the hoist to least 0.25% A k at any be used shall be set at time during the main-5 400 lbs, tenance.
A control i
rod on which maintenance 5.
A maximum of two nonadjacent is being performed control rods may be withdrawn shall be considered 4
from the core for the purpose inoperable.
of performing control rod and/or control rod drive maintenance, provided the following conditions are satisfied:
a.
The reactor mode switch i
shall be locked in the
" refuel" position.
The refueling interlock which prevents more than one control rod from being withdrawn may be bypassed for one of the control rods on which maintenance is being performed.
All other refueling interlocks shall be operable.
b.
A sufficient number of control rods shall be operable so that the core can be made suberitical with the strongest operable control rod fully withdrawn and all other operable control rods fully inserted, or all
-226-Amendment No. 69,188
PBAPS Unit 3 1
3.10 HAEEE A.
Egiyelina Interlocks The refueling interlocks are designed to back up procedural core reactivity controls during refueling operations.
The interlocks prevent an inadvertent criticality during refuel-l ing operations when the reactivity potential of the core is being altered.
To minimize the possibility of loading fuel into a cell con-taining no control rod, it is required that all control rods are fully inserted when fuel is being loaded-into the reactor This requirement assures that during refueling the core.
refueling interlocks, as designed, will prevent inadvertent criticality.
The refueling interlocks reinforce operational procedures that prohibit taking the reactor critical under certain situations encountered during refueling operations by restricting the movement of control rods and the operation of refueling equipment.
The refueling interlocks include circuitry which senses the condition of the refueling equipment and the control rods.
Depending on the sensed condition, interlocks are actuated which prevent the movement of the refueling equipment or withdrawal of ' control rods (rod block).
Circuitry is provided which senses the following conditions:
1.
All rods inserted.
2.
Refueling platform positioned near or over the core.
3.
Refueling platform hoists are fuel-loaded (fuel grapple, frame mounted hoist, monorail mounted hoist).
4.
Fuel grapple not full up.
5.
Deleted.
6.
One rod withdrawn.
When the mode switch is in the "RefuelH position, interlocks prevent the refueling platform from being moved over the core if a control rod is withdrawn and fuel is on a hoist.
Like-wise, if the refueling platform is over the core with fuel on a hoist, control rod motion is blocked by'the interlocks.
When the mode switch is in-the-refuel position, only one con -
trol rod can be withdrawn.
The refueling interlocks, in com-bination with core nuclear design and refueling procedures, limit the probability of an inadvertent criticality..The nuclear characteristics of.the core assure that the reactor j
-229-Amendment No.188 i
.