ML20081H073
| ML20081H073 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 03/21/1995 |
| From: | Gundrum L NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20081H080 | List: |
| References | |
| NPR-03-A-196, NPR-3-A-196, NUDOCS 9503240086 | |
| Download: ML20081H073 (9) | |
Text
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UNITED STATES
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- ,I NUCLEAR REGULATORY COMMISSION f
WASHINGTON, D.C. 20666-0001
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TOLEDO EDISON COMPANY CENTERIOR SERVICE COMPANY AlfQ THE CLEVELAND ELECTRIC ILLUMINATING COMPANY DOCKET NO. 50-346 DAVIS-BESSE NUCLEAR POWER STATION. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.196 License No. NPF 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by the Toledo Edison Company, Centerior Service Company, and the Cleveland Electric Illuminating Company.
(the licensees) dated December 6, 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;
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B.
The facility will operate in conformity with the application, the i
provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (1) 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 i
satisfied.
i 2.
Accordingly,-the license is amended by changes to the Technical i
Specifications as indicated in the attachment to this license amendment, 3
and paragraph 2.C.(2) of Facility Operating License No. NPF-3 is hereby amended to read as follows:
l 9503240086 950321 P PDR ADOCK 050003467 l
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- (a) Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No.196, are hereby incorporated in the license.
The Toledo Edison Company 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 not later than 90 days after issuance.
FOR THE NUCLEAR REGULATORY COMMISSION U
l Linda L. Gundrum, Project Manager Project Directorate III-3 Division of Reactor Projects III/IV Office of Nuclear Reactor Regulation Att achmen'.: - Changes to the Technical l
Specifications l
Date of issuance:
March 21, 1995 l
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i ATTACHMENT TO LICENSE AMENDMENT NO.196
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FACILITY OPERATING LICENSE NO. NPF-3 DOCKET NO. 50-346 l
l Replace the following pages of the Appendix "A" Technical Specifications with the attached pages. The revised pages are identified by amendment number and i
contain vertical lines indicating the area of change.
Remove Insert 3/4 1-20 3/4 1-20 3/4 1-21 3/4 1-21 3/4 4-16 3/4 4-16 3/4 5-4 3/4 5-4 3/4 6-12 3/4 6-12 l
3/4 10-4 3/4 10-4 l
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F REACTIVITY CONTROL SYSTEMS G_ROUP HEIGHT - SAFETY AND REGULATING R0D GROUPS LIMITING CONDITION FOR OPERATIONS ACTION:
(Continued) c)
A power distribution map is obtained from the incore detectors and.F and F",, are verified to be within their limits within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />,, and d)
Either the THERMAL POWER level is reduced to 5 60% of the THERMAL POWER allowable for the reactor coolant pump combination within one hour and within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> the High Flux Trip Setpoint is reduced to s 70% of the THERMAL POWER allowable for the reactor coolant pump combination, or e)
The remainder of the rods in the group with the inoperable rod are aligned to within i 6.5% of the inoperable rod within one hour while maintaining the position of the rods with',n the limits provided in the CORE OPERATING LIMITS REPORT; the THERMAL POWER level shall be restricted pursuant to Specification 3.1.3.6 during subsequent operation.
SURVEILLANCE REQUIREMENTS 4.1.3.1.1 The position of each control rod shall be determined to be within the group average height limit by verifying the individual rod positions at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> except during time intervals when the asymmetric rod monitor is inoperable, then verify the individual rod position (s) of the rod (s), with the inoperable asymmetric rod monitor at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
4.1.3.1.2 Each control rod not fully inserted shall be determined to be OPERABLE by movement of at least 2% in any one direction at least once every 92 days.
l DAVIS-BESSE, UNIT 1 3/4 1-20 Amendment No.175, Jff, J$E, U8,196
REACTIVITY CONTROL SYSTEMS GPOUP HEIGHT - AXIAL POWER SHAPING ROD GROUP LIMITING CONDITION FOR OPERATION 3.1.3.2 All axial power shaping rods (APSR) shall be OPERABLE, unless fully withdrawn, and shall be positioned within i 6.5% (indicated position) of their group average height.
1 APPLICABILITY: MODES 1* and 2*.
ACTION:
With a maximum of cne APSR inoperable or misaligned from its group average height by more than t 6.5t (indicated position), operation may continue provided that within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:
a.
The APSR group is positioned such that the misaligned rod is restored to within limits for the group average height, or b.
It is determined that the imbalance limits of Specification 3.2.1 are satisfied and movement of the APSR group is prevented while the rod remains inoperable or misaligned.
SURVEILLANCE REQUIREMENTS 4.1.3.2.1 The position of each APSR rod shall be determined to be within the group average height limit by verifying the individual rod positions at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> except during time intervals when the asymmetric rod monitor is inoperable, then verify the individual rod position (s) of the rod (s), with the inoperable asymmetric rod monitor at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.
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- See Special Test Exceptions 3.10.1 and 3.10.2 DAVIS-BESSE, UNIT 1 3/4 1-21 Amendment No. (6f,196
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SURVEILLANCE REQUIREMENTS l
4 4.4.6.2.1 Reactor Coolant System leakages shall be demonstrated to be within I
each of the above l'imits by:
a.
Monitoring the containment atmosphere gaseous or particulate radioactivity at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
i b.
Monitoring the containment sump level and flow indication at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
c.
Measurement of the CONTROLLED LEAKAGE from the reactor coolant pump seals to the makeup system when the Reactor Coolant System I
pressure is 2185 i 20 psig at least once per 31 days.
d.
Performance of a Reactor Coolant System water inventory balance at i
least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> during steady state operation.
4.4.6.2.2 Each Reactor Coolant System Pressure Isolation Valve specified in Table 3.4-2 shall be individually demonstrated OPERABLE by verifying leakage testing (or the equivalent) to be within its limit prior to entering MODE 2:
l a.
After each refueling outage, b.
Whenever the plant has been in COLD SHUTDOWN for 7 days, or more, l
and if leakage testing has not been performed in the previous 9 months, and c.
Prior to returning the valve to service following maintenance, i
repair or replacement work on the valve.
d.
The provisions of Specification 4.0.4 are not applicable for entry into MODES 3 or 4.
4.4.6.2.3 Whenever the integrity of a pressure isolation valve listed in Table 3.4-2 cannot be demonstrated, determine and record the integrity of the high pressure flowpath on a daily basis.
Integrity shall be determined by performing either a leakapa test of the remaining pressure isolation valve, or a combined leakage test of the remaining pressure isolation valve in a series with the closed motor operated containment isolation valve.
In addition, record the position of the closed motor-operated containment isolation valve located in the high pressure piping on a daily basis.
t 9/##t ##f## #/20/81 DAVIS-BESSE, UNIT 1 3/4 4 5
Amendment No.54,115, 187, 196 I
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SURVEILLANCE RE0VIREMENTS (continued) b.
At least once'per 18 months, or prior to operation after ECCS piping has been drained by verifying that the ECCS piping is full of water by venting the ECCS pump casings and discharge piping high points.
By a visual inspection which verifies that no loose debris (rags,be c.
trash, clothing, etc.
is present in the containment which could transported to the con)tainment emergency sump and cause restriction of the pump suction during LOCA conditions. This visual inspection shall be perforfbed:
i 1.
For all accessible areas of the containment prior to establishing CONTAINMENT INTEGRITY, and i
2.
For' all areas of containment affected by an entry, at least once daily while work is ongoing and again during the final exit after completion of work (containment closecut) when CONTAINMENT INTEGRITY is established.
d.
At least once per 18 months by:
1 1.
Verifying that the interlocks:
a)
Close DH-11 and DH-12 and deer,ergize the pressurizer heaters, if either DH-11 or DH-12 is open and a simulated reactor coolant system ressure which is greater than the trip setpoint (<438 psi )is not required if the valve is is applied. The interlock to close DH-11 and/or DH-1 closed and 480 V AC power is disconnected from its motor operators.
b)
Prevent the opening of DH-11 and DH-12 when a simulated or actual reactor coolant system pressure which is greater than the trip setpoint (<438 psig) is applied.
2.
a)
A visual inspection of the containment emergency sump which verifies that the subsystem suction inlets are not restricted by debris and that the sump components (trash racks, screens, etc.
show no evidence of structural distress or corrosion).
b)
Verifying that on a Borated Water Storage Tank (BWST)
Low-Low Level interlock trip with the motor operators for the BWST outlet isolation valves and the containment emergency sump recirculation valves energized, the BWST Outlet Valve HV-DH7A (HV-DH78) automatically close in 575 t
seconds after the operator manually pushes the control switch to open the Containment Emergency Sump Valve HV-DH9A (HV-DH98) which should be verified to open in 175 seconds.
l 3.
Verifying a total leak rate 120 gallons per hour for the LPI system at:
a)
Normal operating pressure or hydrostatic test pressure of 1 l
150 psig for those parts of the system downstream of the pump suction isolation valve, and b)
> 45 psig for the piping from the containment emergency sump Tsolation valve to the pump suction isolation valve.
DAVIS-BESSE, UNIT 1 3/4 5-4 AmendmentNo.g25,RS,Q,77,U),M,
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CONTAINMENT SYSTEMS SURVEILLANCE RE0VIREMENTS (Continued) t c.
At least once per 18 months by verifying a total leak rate 5 20 gallons per hour for the system at:
l.
Normal aperating pressure or a hydrostatic test pressure
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of 1150 psig for those parts of the system downstream of the pump suction isolation valve, and 2.
2 45 psig for the piping from the containment emergency sump isolation valve to the pump suction isolation valve.
i d.
At least once per 10 years by performing an air or smoke flow test through l
each spray header and verifying each spray nozzle is unobstructed.
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DAVIS-BESSE, UNIT 1 3/4 6-12 Amendment No. 196
SPECIAL TEST EXCEPTIONS SHUTDOWN MARGIN LIMITING CONDITION FOR OPERATION 3.10.4 The SHUTDOWN MARGIN requirement of Specification 3.1.1.1 may be suspended for measurement of control rod worth and shutdown margin provided:
a.
Reactivity equivalent to at least the highest estimated control rod worth is, available for trip insertion from OPERABLE control rod (s),
and 4
b.
All axial power shaping rods are withdrawn to at least 35% (indicated position) and OPERABLE.
APPLICABILITY: MODE 2.
l ACTION:
a.
With any safety or regulating control rod not fully inserted and with less than the above reactivity equivalent available for trip insertion or the axial power shaping rods not within their withdrawal limits, immediately initiate and continue boration at > 25 gpm of 7875 ppm boric acid solution or its equivalent until thi SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.
b.
With all safety or regulating control rods fully inserted and the reactor subtritical by less than the above reactivity equivalent, immediately initiate and continue boration at > 25 gpm of 7875 ppm boric acid solution or its equivalent until thi SHUTDOWN MARGIN required by Specification 3.1.1.1 is restored.
SURVEILLANCE REQUIREMENTS 4.10.4.1 The position of each safety, regulating, and axial power shaping rod either partially or fully withdrawn shall be determined at least once per 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
4.10.4.2 Each safety or regulating control rod not fully inserted shall be demonstrated capable of full insertion when tripped from at least the 50%
withdrawn position within 7 days prior to reducing the SHUTDOWN MARGIN to less l
than the limits of Specification 3.1.1.1.
DAVIS-BESSE, UNIT 1 3/4 10-4 Amendment No, fsf, 196