ML17158A236
| ML17158A236 | |
| Person / Time | |
|---|---|
| Site: | Susquehanna  |
| Issue date: | 04/15/1994 |
| From: | Chris Miller Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17158A237 | List: |
| References | |
| GL-88-01, GL-88-1, NUDOCS 9404250051 | |
| Download: ML17158A236 (27) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2$5554001 PENNSYLVANIA POWER
& LIGHT COMPANY ALLEGHENY ELECTRIC COOPERATIVE INC.
DOCKET NO. 50-387 SUS UEHANNA STEAM ELECTRIC STATION UNIT I AMENDMENT TO FACILITY OPERATING LICENS Amendment No. >34 License No. NPF-14 1.
The Nuclear Regulatory Commission (the Commission or the NRC) having found that:
A.
The application for the amendment filed by the Pennsylvania Power 8
Light Company, dated April 16,
- 1993, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Commission's 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 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 set forth in 10 CFR Chapter I; 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.
940425005l 940415 PDR ADOCK 05000387
..F'DR,
2.
Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in'he attachment to'his license amendment and paragraph 2.C.(2) of the Facility Operating License No. NPF-14 is hereby amended to read as follows:
(2)
Technical S ecifications and Environmental Protection Plan The Technical Specifira ions contained in Appendix A, as revised through Amendment No.
and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license.
PP&L shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of its date of issuance and is to be implemented within 90 days after its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
Attachment:
Changes to the Technical Specifications Date of Issuance:
Apr>l 15> 1994 goJm ffsiiA~
Charles L. Miller, Director Project Directorate I-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
ATTACHMENT TO LICENSE AMENDMENT NO. 134 FACILITY OPERATING LICENSE NO. NPF-14 DOCKET NO. 50-387 Replace the following pages of the Appendix A Technical Specifications with enclosed pages.
The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.
The overleaf page is provided to maintain document completeness.*
~RENOV 3/4 0-3 3/4 4-5 3/4 4-6 3/4 4-7 3/4 4-8 B 3/4 4-2 INSERT 3/4 0-3 3/4 4-5*
3/4 4-6 3/4 4-7 3/4 4-8 B 3/4 4-2
APPLICABILITY SURVEILLANCE REQUIREMENTS Continued ASME Boiler and Pressure Vessel Code and applicable Addenda terminology for inservice inspection and testing activities Weekly Monthly Quarterly or every 3 months Semiannually or every 6 months Every 9 months Yearly or annually Required frequencies for performing inservice inspection and testing activities At least once per 7 days At least once per 31 days At least once per 92 days At least once per 184 days At least once per 276 days At least once per 366 days c.
The provisions of Specification 4.0.2 are applicable to the above. required frequencies for performing inservice inspection and testing activities.
d.
Performance of the above inservice inspection and testing activities shall be in addition to other specified Surveillance Requirements.
e.
Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any Technical Specification.
f.
The Inservice Inspection Program for piping identified in NRC Generic Letter 88-01 shall be performed in accordance with the NRC Staff position on Schedule, Methods and Personnel, and sample expansions included in the Generic Letter.
SUSQUEHANNA - UNIT 1 3/4 0-3 Amendment No.
78, >3"
III ill
REACTOR COOLANT SYSTEM 3/4. 4. 2 SAFETY/RELIEF VALVES LINITING CONDITION FOR OPERATION 3.4.2 The safety valve function of at 1east 10 of the following reactor coolant system safety/relief va1ves shall be OPERABLE with the specified code safety va1ve function lift settings:" ""
l>~
2 safety-relfef valves 9 1146 4
safety-relief valves I 1175 4
safety-elief valves I 1185 3
safety-relief valves I 1195 3
safety-relief valves 0 1205 psig +lX psig
+1%
psig
+1%
psfg +1%
psig +1%
APPLICABILITY:
OPERATIONAL CONDITIONS 1, 2 and 3.
ACTION:
a.
N1th the safety va1ve funct1on of one or ears of ths above requtrec 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 COLO SHUTDO& within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
b.
Nth one or more safety/relfef valves stuck open, provfded that suppression pool average water temperature fs less than 105'F, close the stuck open relief valve(s); if unable to close the open valve(s) within 2 mfnutes or ff suppression pool water temperature is 105 F or greater, place the reactor mode switch fn the Shutdown posftfon.
c.
Nth one or more safety/relief valve acoustfc monitors inoperable, restore the fnoperable monitor(s) to OPERABLE status.within 7 days or be in at least HOT SHUTDOlA within the next I2 hours and fn COLD SHUTDOWN ~ithin the 1'ollowfng 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
SURVEILLANCE RE UIRBIENTS 4.4.2 The acoustfc Ionftor for each safety/relfef valve shall be demonstrated OPERABLE wfth the setpofnt veriffed to be 0.25 of the full open noise level by performance of a:
a.
CHANNEL FUNCTIONAL TEST at least once per 31 days, and a b.
Calibration fn accordance with procedures prepared fn conjunction with fts amufacturer 5 recomendatfons at liast once per l8 months.
gp h l1*
I dt ~I dM f
valves at nominal operatfng temperatures and pressures.
""Up to 2 inoperable valves may be replaced with spare OPERABLE valves with lower setpoints until the next refuelfng.
kfI'he provfsions of Specification 4.0.4 are not applfcable provided the
'urveillance fs performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steaa pressure is adequate to perform the test.
SUSQUEHANNA - UNIT 1 3/4 4-5 Aaen~nt No.36
REACTOR COOLANT SYSTEM 3 4.4.3 REACTOR COOLANT SYSTEM LEAKAGE LEAKAGE DETECTION SYSTEMS LIMITINGCONDITION FOR OPERATION II 3.4.3.1 At least the following reactor coolant system leakage detection systems shall be OPERABLE:
a.
Two drywell floor drain sump level channels, and b.
One primary containment atmosphere gaseous radioactivity monitoring system channel and one containment atmosphere particulate radioactivity monitoring system channel aligned to the drywell ~
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3.
ACTION:
a.
With one or both channels of the drywell floor drain sump level monitoring system inoperable, operation may continue for up to 30 days provided the drywell floor drain sump flow rate is monitored and determined by alternate means at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Otherwise, be in at least HOT SHUTDOWNwithinthe 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 />.
I b.
With both channels ofthe gaseous radioactivity monitoring system inoperable or with both channels of the particulate radioactivity moriitoring system
, inoperable, operation may continue for up to 30 days provided grab samples of the containment atmosphere are obtained and analyzed at least once per
.24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
If at least one channel of the affected monitoring system cannot be returned to OPERABLE status and aligned to the drywell within 30 days, or the grab samples are not obtained and analyzed as required, 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 REQUIREMENTS 4.4.3.1 The reactor coolant system leakage detection systems shall be demonstrated OPERABLE by a.
Primary containment atmosphere particulate and gaseous monitoring systems-performance of a CHANNEL CHECK at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, a
CHANNEL FUNCTIONAL TEST at least once per 31 days and a CHANNEL CALIBRATIONat least once per 18 months.
b.
Drywell floor drain sump level monitoring system-performance of a CHANNEL FUNCTIONALTEST at least once per 31 days and a CHANNELCALIBRATION at least once per 18 months.
SUSQUEHANNA - UNIT 1 3/4 4-6 Amendment No.
87, >>4
REACTOR COOLANT SYSTEIVI OPERATIONAL LEAKAGE LIIVIITINGCONDITION FOR OPERATION 3.4.3.2 Reactor coolant system leakage shall be limited to:
a.
b.
5 gpm UNIDENTIFIED LEAKAGE.
c.
25 gpm total leakage average over any 24-hour period.
d.
1 gpm leakage at a reactor coolant system pressure of 1000 k 10 psig from any reactor coolant system pressure isolation valve specified in Table 3.4.3.2-1.
e.
2 gpm increase in UNIDENTIFIED LEAKAGE within any 24-hour period in OPERATIONAL CONDITION 1
~
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3.
ACTION:
a.
With any PRESSURE BOUNDARY LEAKAGE, 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 any reactor coolant system leakage greater than the limits in b and/or c, above, reduce the leakage rate to within the limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 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 />.
c.
With any reactor coolant system pressure isolation valve leakage greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least one closed manual or deactivated automatic valves, 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 />.
d.
With one or more of the high/low pressure interface valve leakage pressure monitors shown in Table 3.4.3.2-1 inoperable, restore the inoperable monitor(s) to OPERABLE status within 7 days or verify the pressure to be less than the alarm pressure at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; restore the inoperable monitor(s) to OPERABLE status within 30 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 />.
e.
With any reactor coolant system UNIDENTIFIED LEAKAGE increase'greater than 2 gpm within any 24-hour period, in OPERATIONAL CONDITION 1 only, identify the source of leakage increase as not service sensitive. Type 304 or 316 austenitic stainless steel within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 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 />.
SUSQUEHANNA - UNIT 1 3/4 4-7 Amendment No. 88, >>4
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.3.2.1 The reactor coolant system leakage shall be demonstrated to be, within each of the above limits by:
a.
Monitoring the primary containment atmospheric particulate and gaseous radioactivity at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, and b.
Monitoring the drywell floor drain sump level 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.
Determining the total IDENTIFIED LEAKAGE at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
4.4.3.2.2 Each reactor coolant system pressure isolation valve specified in Table 3.4.3.2-1 shall be demonstrated OPERABLE by leak testing pursuant to Specification 4.0.5 and verifying the leakage of each valve to be within the specified limit:
a.
At least once per 18 months, and b.
Prior to returning the valve to service following maintenance, repair or replacement work on the valve which could affect its leakage rate.
The provisions of Specification 4.0.4 are not applicable for entry into OPERATIONAL CONDITION 3.
4.4.3.2.3 The high/low pressure interface valve leakage pressure monitors shall be demonstrated OPERABLE within the alarm setpoints per Table 3.4.3.2-1 by performance of a:
a.
CHANNEL FUNCTIONALTEST at least once per 31 days, and b.
CHANNEL CALIBRATIONat least once per 18 months.
SUSQUEHANNA - UNIT 1 3/4 4-8 Amendment No. ~>
>>4
3/4.4 REACTOR COOLANT SYSTEM BASES Continued 3/4.4.2 SAFETY/RELIEF VALVES The safety valve function of the safety/relief valves operate 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 10 OPERABLE safety-relief valves is required to limit reactor pressure to within ASME III allowable values for the worst case upset transient.
Demonstration of the safety/relief valve liftsettings willoccur only during shutdown and will be performed in accordance with the provisions of Specification 4.0.5.
3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE 3 4.4.3.1 LEAKAGE DETECTION SYSTEMS The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the reactor coolant pressure boundary.
3/4.4.3.2 OPERATIONAL LEAKAGE The allowable leakage rates for the reactor coolant system have been based on the predicted and experimentally observed behavior of cracks in pipes. The normally expected background
. leakage due to equipment design and the detection capability of the instrumentation for determining system leakage was also considered.
The evidence obtained from experiments suggests that for leakage somewhat greater than that specified for UNIDENTIFIEDLEAKAGE the probability is small that the imperfection or crack associated with such leakage would grow rapidly. The limitof unidentified leakage has been changed to reflect the requirements of Generic Letter 88-01.
However, in all cases, if the leakage rates exceed the values specified or the leakage is located and known to be PRESSURE BOUNDARY LEAKAGE, the reactor will be shutdown to allow further investigation and corrective action.
The Surveillance Requirements for RCS pressure isolation valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequence intersystem LOCA.
3 4.4.4 CHEIVIISTRY The water chemistry limits of the reactor coolant system are established to prevent damage to the reactor materials in contact with the coolant.
Chloride limits are specified to prevent stress corrosion cracking of the stainless steel.
The effect of chloride is not as great when the oxygen concentration in the coolant is low, thus the 0.2 ppm limit on chlorides is permitted during POWER OPERATION.
During shutdown and refueling operations, the temperature necessary for stress corrosion to occur is not present so a 0.5 ppm concentration of chlorides is not considered harmful during these periods.
SUSQUEHANNA - UNIT 1 B 3/4 4-2 Amendment No. ~SR 134
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 2Q555-0001 PENNSYLVANIA POWER 5 LIGHT COMPANY ALLEGHENY ELECTRIC COOPERATIVE INC.
DOCKET NO. 50-388 SUS UEHANNA STEAM ELECTRIC STATION UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. >O4 License No.
NPF-22 1.
The Nuclear Regulatory Commission (the Commission or the NRC) having found that:
A.
The application 'for the amendment filed by the Pennsylvania Power Light Company, dated April 16,
- 1993, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Commission's 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 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.set forth in 10 CFR Chapter I; 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.
2.
Accordingly, the license is amended by changes to the Technical Specifica-tions as indicated in the attachment to this license amendment and paragraph 2.C.(2) of the Facility Operating License No.
NPF-22 is hereby amended to read as follows:
(2) Technical S ecifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No.
1o4
.and the Environmental Protection Plan contained in Appendix B, are hereby incorporated in the license.
PPKL shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of its date of issuance and is to be implemented within 90 days after its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
Attachment:
Changes to the Technical Specifications Date of Issuance:
April 15, 1994 Charles L. Miller, Director Project Directorate I-2 Division of Reactor Projects I/II Office of Nuclear Reactor Regulation
ATTACHMENT TO LICENSE AM NDMENT NO. 104 FACILITY OPERATING LICENSE NO. NPF-22 DOCKET NO. 50-388 Replace the following pages of the Appendix A Technical Specifications with enclosed pages.
The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.
The overleaf pages are provided to maintain document completeness.*
REMOVE 3/4 0-1 3/4 0-2 3/4 0-3 3/4 4-5 3/4 4-6 3/4 4-7 3/4 4-8 B 3/4 4-1 B 3/4 4-2 INSERT 3/4 O-l*
3/4 0-2 3/4 0-3 3/4 4-5*
3/4 4-6 3/4 4-7 3/4 4-8 B 3/4 4-1*
B 3/4 4-2
3 ¹:0 A LIC BIL LIMITINOC NDITION FOR OPERATION 3.0.1 Compliance with the Limiting Conditions for Operation contained in the succeeding Specifications is required during the OPERATIONAL CONDITIONS or other conditions specified therein; except that upon failure to meet the Limiting Conditions for Operation, the associated ACTION requirements shall be met.
3.0.2 Noncompliance with a Specification shall exist when the requirements of the Limiting Condition for Operation and associated ACTION requirements are not met within the specified time intervals.
If the Limiting Condition for Operation is restored prior to expiration of the specified time intervals, completion of the ACTION requirements is not required.
3.0.3 When a Limiting Condition for Operation is not met, except as provided in the associated ACTION requirements, within one hour action shall be initiated to place the unit in an OPERATIONALCONDITION in which the Specification does not apply by placing it, as applicable, in:
- 1. At least STARTUP within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />,
- 2. At least HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and
- 3. At least COLD SHUTDOWN within the subsequent 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Where corrective measures are completed that permit operation under the ACTION requirements, the ACTION may be taken in accordance with the specified time limits as measured from the time of failure to meet the Limiting Condition for Operation.
Exceptions to these requirements are stated in the individual Specifications.
This specification is not applicable in OPERATIONAL CONDITION 4 or 6.
3.0.4 Entry into an OPERATIONAL CONDITION or other specified condition shall not be made when the conditions forthe LImitingConditions for Operation are unmet and w
ified tim i
a Tl N
T
'ed cond'on I
i f rman et I 'io f ime fa i' r
h This provision shall not prevent passage through or to OPERATIONALCONDITIONS as required to comply with ACTION requirements.
Exceptions to these requirements are stated in the individual Specifications.
Compliance with this Specification for the inoperable "S" SRV acoustic monitor is not required for the period beginning January 21, 1994, until the next unit shutdown of sufficient duration to allow for containment entry, not to exceed the sixth refueling and inspection outage.
SUSQUEHANNA - UNIT 2 3/4 0-1 Amendment No. P3.
100 Jhi'J 3 i j-"ii)
APPLICABILITY SURVEILLANCE REQUIREMENTS 4.0.1 Surveillance Requirements shall be met during the OPERATIONAL CONDITIONS or other conditions specified for individual Limiting Conditions for Operation unless otherwise stated in an individual Surveillance Requirement.
4.0.2 Each Surveillance Requirement shall be performed within the specified surveillance interval with a maximum allowable extension not to exceed 25 percent of the specified surveillance interval.
4.0.3 Failure to perform a Surveillance Requirement within the allowed surveillance interval, defined b S ecification 4.0.2 shall constitute noncompliance with the OPERABILITYrequirements for a Limiting Condition for Operation.
The time limits of the ACTION re uirements are a
licable at the time it is identified that a
Surveillance Re uirement has not been erformed.
The ACTION re uirements ma bedela edforu to24hoursto ermitthecom letionofthesurveillance whenthe allowable outa e time limits of the ACTION re uirements are less than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
Surveillance Requirements do not have to be performed on inoperable equipment.
4.0.4 Entry into an OPERATIONAL CONDITION or other specified applicable condition shall not be made unless the Surveillance Requirement(s) associated with the Limiting Condition for Operation have been performed within the applicable surveillance interval or as otherwise specified.
This rovision shall not revent assa e throu h or to OPERATIONAL CONDITIONS as re uired to com I
with ACTION re uirements.
4.0.5 Surveillance Requirements for inservice inspection and testing of ASME Code Class 1, 2, 5. 3 components shall be applicable as follows:
- a. Inservice inspection of ASME Code Class 1, 2.and 3 components and inservice testing of ASME Code Class 1, 2, and 3 pumps and valves shall be performed in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50, Section 50.55a(g), except where specific written relief has been granted by the Commission pursuant to 10 CFR 50, Section 50.55a(g)(6)(i).
- b. Surveillance intervals specified in Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda for the inservice inspection and testing activities required by the ASME Boiler and Pressure Vessel Code and applicable Addenda shall be applicable as follows in these Technical Specifications:
Compliance with this Specification for the inoperable "S" SRV acoustic monitor is not required for the period beginning January 21, 1994, until the next unit shutdown of sufficient duration to allow for containment entry, not to exceed the sixth refueling and Inspection outage."
SUSQUEHANNA - UNIT 2 3/4 0-2 Amendment No.
APPLICABILITY SURVEILLANCE REQUIREMENTS Continued ASME Boiler and Pressure Vessel Code and applicable Addenda terminology for inservice inspection and testing activities Weekly Monthly Quarterly or every 3 months Semiannually or every 6 months Every 9 months Yearly or annually Required frequencies for performing inservice inspection and testing activities At least once per 7 days At least once per 31 days At least once per 92 days At least once per 184 days At least once per 276 days At least once per 366 days c.
The provisions of Specification 4.0.2 are applicable to the above required frequencies for performing inservice inspection and testing activities.
d.
Performance of the above inservice inspection and testing activities shall be in addition to other specified Surveillance Requirements.
e.
Nothing in the ASME Boiler and Pressure Vessel Code shall be construed to supersede the requirements of any Technical Specification.
f.
The Inservice Inspection Program for piping identified in NRC Generic Letter 88-01 shall be performed in accordance with the NRC Staff position on Schedule, Methods and Personnel, and sample expansions included in the Generic Letter.
SUSQUEHANNA - UNIT 2 3/4 0-3 Amendment No. ~~~ >04
I
3 4,4 V
VE LIIVImNGC DITI N FOR OPERATION t
3.4,2 The safety valve function of at least 10 of the following reactor coolant system safety/relief valves shall be OPERABLE with the specified code safety valve function liftsettings:
2 safety-relief valves @ 1146 psig ~1%
4 safety-relief valves 5 1175 psig a1%
4 safety-relief valves @ 1185 psig a1%
3 safety-relief valves 5 1196 psig a1%
3 safety-relief valves 5 1205 psig a1%
ACT~I Fl 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 106'F, dose the stuck open relief valve(s); if unable to close the open valve(s) within 2 minutes or ifsuppression pool water temperature is 1064F or greater, place the reactor mode switch in the Shutdown position.
c.>>>>>> With one or more safety/relief valve acoustic monitors inoperable, restore the inoperable monitor(s) to OPERABLE itatus 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 />..
SU V E
I 4A.2 The acoustic monitor for each safety/relief valve shall be demonstrated OPERABLE with the setpoint verified to be 0.26 of the fullopen noise level>> by performance of a:
a.
CHANNEL FUNCTIONALTEST at least once per 31 days, and a b.
Calibration in accordance with procedures prepared in conjunction with its manufacturer's recommendations at least once per 18 months.
The Nt setting presswe shall correspond to ambient conditions of tha valves at nominal operating temperatures and pressures.
0 ~
Up to 2 Inoperable valves may be replaced with spare OPERABLE valves with lower setpointe 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.
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.
¹ Compliance with these requirements for the S
SRV acoustic monitor is not required for the period beginning January 21, 1994, until the next unit shutdown of sufficient duration to allow for containment entry, not to exceed the sixth refueling end inspection outage.
SUSQUEHANNA - UNIT 2 3/4 4-5 JQ ~,',-mendment.No. 100
REACTOR COOLANT SYSTEM 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE 1
LEAKAGE DETECTION SYSTEMS LIMITINGCONDITION FOR OPERATION 3.4.3.1 At least the following reactor coolant system leakage detection systems shall be OPERABLE:
ri~
a.
Two drywell floor drain sump level channels, and b.
One primary containment atmosphere gaseous radioactivity monitoring system channel and one containment atmosphere particulate radioactivity monitoring system channel aligned to the drywell.
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3.
ACTION:
ae With one or both channels of the drywell floor drain sump level monitoring system inoperable, operation may continue for up to 30 days provided the drywell floor drain sump flow rate is monitored and,determined by alternate means at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
Otherwise, 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 />.
b.
With both channels ofthe gaseous radioactivity monitoring system inoperable or with both channels of the particulate radioactivity monitoring system inoperable, operation may continue for up to 30 days provided grab samples of the containment atmosphere are obtained and analyzed at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
If at least one channel of the affected monitoring system cannot be returned to OPERABLE status and aligned to the drywell within. 30 days, or the grab samples are not obtained and analyzed as required, 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 REQUIREMENTS 4.4.3.1 The reactor coolant system leakage detection systems shall be demonstrated OPERABLE by:
a.
Primary containment atmosphere particulate and gaseous monitoring systems-performance of a CHANNEL CHECK at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, a
CHANNEL FUNCTIONAL TEST at least once per 31 days and a CHANNEL CALIBRATIONat least once per 18 months.
b.
Drywell floordrain sump level monitoring system-performance of a CHANNEL FUNCTIONALTEST at least once per 31 days and a CHANNELCALIBRATION at least once per 18 months.
SUSQUEHANNA - UNIT 2 3/4 4-6 Amendment No.
S8, >04
REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITINGCONDITION FOR OPERATION 3.4.3.2 Reactor coolant system leakage shall be limited to:
a.
b.
5 gpm UNIDENTIFIED LEAKAGE.
c.
25 gpm total leakage average over any 24-hour period.
d.
1 gpm leakage at a reactor coolant system pressure of 1000 2 10 psig from any reactor coolant system pressure isolation valve specified in Table 3.4.3.2-1
~
e.
2 gpm increase in UNIDENTIFIED LEAKAGE within any 24-hour period in OPERATIONAL CONDITION 1.
APPLICABILITY: OPERATIONAL CONDITIONS 1, 2 and 3.
ACTION:
a.
With any PRESSURE BOUNDARY LEAKAGE, 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 any reactor coolant system leakage greater than the limits in b and/or c, above, reduce the leakage rate to within the limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> 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'ithin the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
c.
With any reactor'coolant system pressure isolation valve leakage greater than the above limit, isolate the high pressure portion of the affected system from the low pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least one closed manual or deactivated automatic valve, 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 />.
d.
With one or more of the high/low pressure interface valve leakage pressure monitors shown in Table 3.4.3.2-1 inoperable, restore the inoperable monitor(s) to OPERABLE status within 7 days or verifythe pressure to be less than the alarm pressure at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />; restore the inoperable monitor(s) to OPERABLE status within 30 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 />.
e.
With any reactor coolant system UNIDENTIFIED LEAKAGE increase greater than 2 gpm within any 24-hour period, in OPERATIONAL CONDITION 1 only, identify the source of leakage increase as not service sensitive Type 304 or 316 austenitic stainless steel within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 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 />.
SUSQUEHANNA - UNIT 2 3/4 4-7 Amendment No.
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.3.2.1 The reactor coolant system leakage shall be demonstrated to be within each of the above limits by:
a.
Monitoring the primary containment atmospheric particulate and gaseous radioactivity at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, and d
b.
Monitoring the drywell floor drain sump level 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.
Determining the total IDENTIFIED LEAKAGE at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
4.4.3.2.2 Each r'eactor coolant system pressure isolation valve specified in Table 3.4.3.2-1 shall be demonstrated OPERABLE by leak testing pursuant to Specification 4.0.5 and verifying the leakage of each valve to be within the specified limit:
a.
At least once per 18 months, and b.
Prior to returning the valve to service following maintenance, repair or replacement work'on the valve which could affect its leakage rate.
The provisions of Specification 4.0.4 are not applicable for entry into OPERATIONAL CONDITION 3.
4.4.3.2.3 The high/low pressure interface valve leakage pressure monitors shall be demonstrated OPERABLE within the alarm setpoints per Table 3.4.3.2-1 by performance of a:
a.
CHANNEL FUNCTIONALTEST at least once per 31 days, and b.
CHANNEL CALIBRATIONat least once per 18 months.
SUSQUEHANNA - UNIT 2
,3/4 4-8 Amendment No. 8~8
>04
Operation with one reactor recirculation loop inoperable has been evaluated and found acceptable, provided that the unit is operated in accordance with Specification 3.4.1.1.2.
LOCA analyses for two loop operating conditions, which result in Peak Cladding Temperatures (PCTs) below 2200'F, bound single loop operating conditions.
Single loop operation LOCA analyses using two-loop MAPLHGR limits result in lower PCTs. Therefore, the use of two-loop MAPLHGR limits during single loop operation assures that the PCT during a LOCA event remains below 2200'F.
The MINIMUMCRITICALPOWER RATIO (MCPR) limits for single loop operation assure that the Safety Limit MCPR is not exceeded for any Anticipated Operational Occurrence (AOO).
In addition, the MCPR limits for single-loop operation protect against the effects of the Recirculation Pump Seizure Accident. That is, for operation in single-loop with an operating MCPR limit a 1.30, the radiological consequences of a pump seizure accident from single-loop operating conditions are but a small fraction of 10CFR100 guidelines.
For single loop operation, the RBM and APRM setpoints are adjusted by a 8.5% decrease in recirculation drive flow to account for the active loop drive flow that bypasses the core-and goes up through the inactive loop jet pumps.
Surveillance on the pump speed of the operating recirculation loop is imposed to exclude the possibility of excessive reactor vessel internals vibration.
Surveillance on differential
.temperatures below the threshold limits of THERMAL POWER or recirculation loop flow mitigates undue thermal stress on vessel nozzles, recirculation pumps and the vessel bottom head during extended operation in the single loop mode.
The threshold limits are those values which will sweep up the cold water from the vessel bottom head.
Specifications have been provided to prevent, detect, and mitigate core thermal hydraulic instability events.
These specifications are prescribed in accordance with NRC Bulletin 88-07, Supplement 1, "Power Oscillations in Boiling Water Reactors (BWRs)," dated December 30, 1988.
LPRM upscale alarms are required to detect reactor core thermal hydraulic instability events.
The criteria for determining which LPRIVI upscale alarms are required is based on assignment of these alarm to designated core zones.
These core zones consist of the level A, 8 and C alarms in 4 or 5 adjacent LPRM strings.
The number and location of LPRM strings in each zone assure that with 50% or more of the associated LPRM upscale alarms OPERABLE sufficient monitoring capability is available to detect core wide and regional oscillations.
Operating plant instability data is used to determine the specific LPRM strings assigned to
, each zone.
The core zones and required LPRM upscale alarms in each zone are specified in appropriate procedures.
An inoperable jet pump is not, in itself, a sufficient reason to declare a recirculation loop inoperable, but it does in case of a design basis accident, increase the blowdown area and reduce the capability of reflooding the core; thus, the requirement for shutdown of the facility with a jet pump inoperable.
Jet pump failure can be detected by monitoring jet pump performance on a prescribed schedule for significant degradation.
SUSQUEHANNA - UNIT 2 B 3/4 4-1 Amendment No. ~
O~T8 8 1992
REACTOR COOLANT SYSTEM BASES Recirculation pump speed mismatch limits are in compliance with the ECCS LOCA analysis design criteria for two loop operation.
The limits will ensure an adequate core flow coastdown from either recirculation loop following a LOCA. In the case where the mismatch limits cannot be maintained during the loop operation, continued operation is permitted in the single loop mode.
In order to prevent undue stress on the vessel nozzles and bottom head region, the recirculation loop temperatures shall be within 50'F of each other prior to startup of an idle loop. The loop temperature must also be within 50'F of the reactor pressure vessel coolant temperature to prevent thermal shock to the recirculation pump and recirculation nozzles.
Since the coolant in the bottom of the vessel is at a lower temperature than the coolant in the upper"regions of the core, undue stress on the vessel would result if the temperature difference was greater than 145'F.
3/4 4.2 SAFETY RELIEF VALVES The safety valve function of the safety/relief valves operate 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 10 OPERABLE safety;relief valves is required to limit reactor pressure to within ASME III allowable values for the worst case upset transient.
Demonstration of the safety/relief valve liftsettings willoccur only during shutdown and will be performed in accordance with the provisions of Specification 4.0.5..
3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE 3/4.4.3.,1 LEAKAGE.DETECTION SYSTEMS The RCS leakage detection systems required by this specification are provided to monitor and detect leakage from the reactor coolant pressure boundary.
3/4 4.3.2 OPERATIONAL LEAKAGE The allowable leakage rates for the reactor coolant system have been based on the predicted and experimentally observed behavior of cracks in pipes. The normally expected background leakage due to equipment design and the detection capability of the instrumentation for determining system leakage was also considered.
The evidence obtained from experiments suggests that for leakage somewhat greater than that specified for UNIDENTIFIEDLEAKAGE the probability is small that the imperfection or crack associated with such leakage would grow rapidly. The limitof unidentified leakage has been changed to reflect the requirements of Generic Letter 88-01.
However, in all cases, if the leakage rates exceed the values specified or the leakage is located and known to be PRESSURE BOUNDARY LEAKAGE, the reactor will be shutdown to allow further investigation and corrective action.
The Surveillance Requirements for RCS pressure isolation valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequence intersystem LOCA.
SUSQUEHANNA - UNIT 2 B 3/4 4-2 Amendment No. ~H, >04