ML20069G332
| ML20069G332 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 09/08/1982 |
| From: | Vassallo D Office of Nuclear Reactor Regulation |
| To: | Power Authority of the State of New York |
| Shared Package | |
| ML20069G335 | List: |
| References | |
| DPR-59-A-070, TAC 48771 NUDOCS 8209280594 | |
| Download: ML20069G332 (9) | |
Text
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UNITED STATES y
- g NUCLEAR REGULATORY COMMISSION g
.y WASHINGTON. D. C. 20506
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POWER AUTHORITY OF THE STATE OF NEW YORK DOCKET NO. 50-333 JAMES A. FITZPATRICK NUCLEAR POWER PLANT AMENDMENT TO FACILITY OPERATING LICENSE Amndrent No. 70 License No. DPR-59 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The' application for amendment by the Pcwer Authority of the State of New York dated August 12, 1982 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 confomity with the application, the provisions of the Act, and the rules and regulations of l
the Comission; f
C.
There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health l
and safety of the public, and (11) that such activities will be conducted in compliance with the Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon defense and security or to the health and safety of the public; and E.
The issuance of this amendment is in accordance with 10 CPR Part 51 of the Comission's regulations and all applicable mquiremnts have been satisfied.
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2.
Accordingly, the license is amended by changes to the Techcical Specifications as indicated in the attachment to this license amendment and paragraph 2.C(2) of Facility Operating License No. DPR-59 is hereby amended to read as follows:
(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 70, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment became effective August 13, 1982.
FOR THE NUCLEAR REGULATORY COMMISSION N
Domenic B. Vassallo, Chief Operating Reactors Branch #2 Division of Licensing
Attachment:
Changes to the Technical Specif1 cations Date of Issuance: September 8, 1982
'8 AHACHMENT TO LICENSE AMENDMENT NO. 70 FACILITY OPERATING LICENSE NO. DPR-59 DOCKET NO. 50-333 Revise Appendix A Technical Specifications as follows:
Remove Insert 76c 76c 142a 142a 142b 143 143 143a 143b C
f 1
a i
9 s
c.
L W
4 NOTES FOR TABLE 3.2-f (CONTINUED)
In the event that all indications of this parameter is disabled and such indication cannot be "
2.
i restored in six (6) hours, an orderly shutdown shall be initiated and the reactor shall be in a flot Shutdown condition in six (6) hours and a Cold Shutdown condition in the following l
I eighteen (18) hours.
3.
Three (3) indicators from level instrument channel A, B, & C. Channel A or B are utilized for IIL h feedwater control, reactor water high and low level alarms, recirculation pump runback.
9 level trip of main turbine and feodwater pump turbine utilizes channel A, B,
& C.
4.
One (1) recorder utilized the same level instrument channel as selected for feedwater control.
5.
Three (3) indicators from reactor pressure instrument channel A.
D,
& C. Channel A or D are utilized (or feedwater control and reactor pressure high alarm.
for 6.
One (1) recorder. U,tilizes the same reactor pressure instrument channel as selected feedwater control.
The position of each of the 137 control rods is monitored by the Rod Position Information 7.
System. For control rods in which the position is unknown, refer to Paragraph 3.3.A.
l Heutron monitoring operability requirements are specified by Table 3.1-1 and Paragraph 3.3.B.4.
8.
9.
A minimum of 3 IRM or,2 APRM channels respectively must be operable (or tripped) in each safety system.
10.
Each Safety Relief Valve is equipped with two acoustical detectors of which one is in service and a backup thermocouple detector. In the event that a thermocouple is inoperable SRV l
p erformance shall be monitored daily with the associated acoustical detector.
I 11.
From and after th.e date that none of the acoustical detectors is operable but the thermocouple is operable, continued operation is permissible until the next outage in wi ich a primary t
containment. entry is made. Both acoustical detectors shall be made operable prior to restart.
12.
In the event that both primary and secondary indications of this parameter for any one valve are disabled and neither indication can be restored in forty-eight (48) hours, an orderly i
shutdown shall be initiated and the reactor shall be in a llot Shutdown condition in twelve (121 hours0.0014 days <br />0.0336 hours <br />2.000661e-4 weeks <br />4.60405e-5 months <br /> cnd in a Cold Shutdown within the next twenty-four (24) hours.
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13.
From and after the date that the minimum number of operable instrument channels is one less than the minimum number specified for each parameter, continued operation is permissable during the suc'ceeding 7 days unless the minimum number specified is made operable sooner.
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44,71,p[,70 76 c Amendment No.
l
r 3.6 (cont'd)
.1 Art 4PP 4.6(cont'd) i E.
Safet tand Safety /polief valves E.
Safe;ty and Safety / Relief valves 1.
During reactor power operating 1.
At least one half of all conditions anil prior to start up safety / relief valves shall be bench from a cold condition, or whenever checked or replaced.elth bench checked reactor coolant pressure is great er valves once each operating cycle. The t.han atmosphere and temperature safety / relief valve settings greater than 212"r, shall be set as required in Specification the sa fety pw>de of al l 2.2.H.
All valves shall be tested every safety / relief valves shall be two operatlay cycles.
ororable, exc. apt as specifiel by Specification 3.6.E.2 The Automatic Depressurization System valves shall he operable as required by Specification 3.5.D.
o Amen <went no. vi, yrt X, 70 142a This page is effective after the October 1982 outage.
3
3 3.6 (cont'd)
JAFNPP 4.6 (c nt'd) i E. Saf ety and Saf ety/ Relief : Valves E.
Safety and Safety / Relief Valves 1.
At least one half of all
- l. During reactor power operating safety / relief valves shall conditions and prior to startup be bench checked or re-from a cold condition, or whenever l
reactor coolant pressure-is placed with bench checked valves once each operating greater ensn atmosphere and temperatura greater than 212 F, cycle.
The safety / relief l
the safety mode of all safety / relief valve settings shall be set valves shall be operable, except as required in Specification 2.2.B.
All valves shall be i
as specified by Specification tested every two operating 3. 6. E. 2.-
The Automatic Depressurization System valves shall be operable as required cycles.
by Specification 3.5.D.
- 2. Reactor operation may continue with one safety / relief valve inoperable.
From and after the date that two safety / relief valves are made or found inoperable, continued reactor operation i
is permissible only during the succeeding 7 days,unless one valve is made operable.
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l i
142b This page is effective until Amendment No. 70' the October 1982 outage.
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-- r (cont'd)
JPP 4.6 (cont'd) 2.
a.
From and after the date 2.
At least one safety / relief that the safety valve function of one safety /
valve shall be disassembled relief valve is made or and inspected once/ operating cycle.
found to be inoperable, continued operation is 3.
permissible only during The integrity of the safety /
the succeeding 30 days relief valve bellows shall be unless such valve is continuously monitored.
sooner made operable.
The bellows monitoring a.
pressure switches shall i
b.
From and after the time,
be removed and hench that the safety valve j
function on two safety /
checked once/ operating relief valves is made or cycle. Modified safety /
found to be inoperable, relief valves with two-stage continued reactor operation
~
assemblies do not have a is permissible only during bellows arrangement and are, therefore, not subject the succeeding 7 days unless such valves are sooner made to this requirement.
4.
The integrity of the nitrogen 3.
If Specification 3.63.1 and system and components which 3.63.2 are not met, the reactor provide manual and ADS actuation shall be plate:t in a cold condition of the safety / relief valves shall within 24 hr.
be demonstrated at least once every 3 months.
- 4. Low power physics #:es+ fng and reactor operator t;-d r ing shall be permitted with inoperable l
components as specified in j
Item 8.2 above, provided that reactor coolant temperature is s 212*F and the reactor vessel is vented or the reactor vessel head is removed.
Amende^"t No. JHr. 70 143 This page is effective after the October 1982 outage.
o m,
..,no
JAFNPP 4.6 (c nt'd) 3.6 (cont'd) 2.
At least one safety / relief f
3.
If Specif ication 3. 6.E. l and 3. 6. E. 2 valve shall be disassembled are not met the reactor shall be end inspected once/ operating placed in a cold condition within cycle.
24 hr.
3.
Deleted 4.
Low power physics testing and reactor operator training shall be permitted 4.
The integrity of the nitrogen with inoperable components as specified system and components which in 3.6.E.2, and provided that reactor provide manual and ADS coolant temperature isq{_2120F and actuation of the safety / relief the reactor vessel is vented or the valves shall be demonstrated vessel head is removed.
at least once every 3 months.
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Amendment No. 70 143a This page is effective until the October 1982 outage.
JAFNPP 3.6 (cont'd)
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If, for a period of longer than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the temperature of any safety / relief discharge pipe is more than 400F above its steady state valu e, or the acoustical monitor reading of any safety / relief valve discharge pipe is more than 3 times greater than its steady state value, the following actions shall be taken:
a report shall be issued in accordance a.
with 6.9.A.4.1 which addresses the actions that have been taken or a schedule of actions to be taken.
b.
an engineering evaluation shall be per-formed 3ustifying continued operation
'l for the corresponding increase in tem-perature or acoustical monitor reading.
the affected safety / relief valve shall c.
be removed at the next cold shutdown of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or more, tested in the as-found condition, and recalibrated as necessary prior to reinstallation, d.
NRC approval of.the engineering evaluation specified in 3.6.E.6.b above shall be'obtained prior to continuing power operation for more than 90 days after the initial discovery of the 40 F increase in temperature or the factor of 3 increase in acoustical monitor reading.
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The steady state values of temperature and acoustical monitor readings shall be as measured after 5 days of steady state power operation.
This page is effective until the Amendment No. 70 143b October 1982 outage.
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