ML20094M885
| ML20094M885 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 11/20/1995 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Philadelphia Electric Co |
| Shared Package | |
| ML20094M887 | List: |
| References | |
| GL-91-008, NPF-39-A-105, NPF-85-A-069 NUDOCS 9511270278 | |
| Download: ML20094M885 (24) | |
Text
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UNITED STATES y
NUCLEAR REGULATORY COMMISSION f
WASHINGTON, D.C. 20066 4 301
%,.....,o PHILADELPHIA ELECTRIC COMPANY DOCKET NO. 50-352 LIMERICK GENERATING STATION. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.105 License No. NPF-39 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Philadelphia Electric Company (the licensee) dated September 14, 1995, and supplemented by letter dated October 27, 1995, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I; 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 (i) that the activities authorized by j
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 Comission'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 Comission's regulations and all applicable requirements have been satisfied.
9511270278 951120 PDR ADOCK 05000352 P
. 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-39 is hereby amended to read as follows:
Technical Soecifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 105, are hereby incorporated into this license.
Philadelphia Electric Company 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, to'be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMMISSION J
F. Stolz, Direc r ject Directorate
-2 ivision of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance:
November 20, 1995
s ATTACHMENT TO LICENSE AMENDMENT NO.105 FACILITY OPERATING LICENSE NO. NPF-39 DOCKET NO. 50-352 i
Replace the following pages of the Appendix A Technical Specifications with the attached pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.
Remove Insert xiii xiii 1-6 1-6 1-7 1-7 3/4 6-48 3/4 6-48 3/4 6-49 3/4 6-49 3/4 6-50 3/4 6-50 3/4 6-51 3/4 6-51 3/4 6-51a 3/4 6-51a B 3/4 6-6 8 3/4 6-6 l
i l
=
INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS SECTION Egig CONTAINMENT SYSTEMS (Continued) 3/4.6.4 VACUUM RELIEF i
Suppression Chamber - Drywell Yacuum Breakers.........
3/4 6-44 j
3/4.6.5 SECONDARY CONTAINMENT Reactor Enclosure Secondary Containment Integrity.....
3'/4 6-46 l
Refueling Area Secondary Containment Integrity........
3/4 6-47 i
Reactor Enclosure Secondary Containment Automatic Isolation Va1ves......................................
3/4 6-48 Refueling Area Secondary Containment Automatic Isolation Valva......................................
3/4 6-50 I
Standby Gas Treatment System - Common System..........
3/4 6-52 Reactor Enclosure Reci rcul ation System................
3/4 6-55 3/4.6.6 PRIMARY CONTAINMENT ATMOSPHERE CONTROL Primary Containment Hydrogen Recombiner Systems.......
3/4 6-57 Drywell Hydrcgen Mixing System........................
3/4 6-58 Drywell and Suppression Chamber Oxygen Concentration..
3/4 6-59 3/4.7 PLANT SYSTEMS
.4/4.7.1 SERVICE WATER SYSTEMS Residual Heat Removal hrvice Water System - Common Systes................................................
3/4 7-I Emergency Service Water System - Common System........
3/4 7-3 Ultimate Heat Sink....................................
3/4 7-5 LIMERICK - UNIT 1 xiii Amendment No. 27,#,105
l DEFINITIONS i
PURGE - PURGING j
l 1.31 PURGE or PURGING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, 4
concentration or other operating condition, in such a manner that j
replacement air or gas.is required to purify the confinement.
RATED THERMAL POWER 1.32 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3293 MWt.
)
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REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY 1.33 REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY shall exist when:
a.
All reactor enclosure secondary containment penetrations required to be closed during accident conditions are either:
1.
Capable of being closed by an OPERABLE secondary containment automatic isolation system, or i
1 2.
Closed by at least one manual valve, blind flange, slide gate l
damper, or deactivated automatic valve secured in its closed position, except as provided by Specification 3.6.5.2.1.
l i
1 b.
All reactor enclosure secondary containment hatches and blowout panels j
are closed and sealed.
c.
The standby gas treatment system is in compliance with the requirements of Specification 3.6.5.3.
d.
The reactor enclosure recirculation system is in compliance with the requirements of Specification 3.6.5.4.
l e.
At least one door in each access to the reactor enclosure secondary containment is closed.
f.
The sealing mechanism associated with each reactor enclosure secondary containment penetration, e.g., welds, bellows, or 0-rings, is OPERABLE.
g.
The pressure within the reactor enclosure secondary containment is less than or equal to the value required by Specification 4.6.5.1.la.
REACTOR PROTECTION SYSTEM RESPONSE TIME 1.34 REACTOR PROTECTION SYSTEM RESPONSE TIME shall be the time interval frc.a when the monitored parameter exceeds its trip setpoint at the channai sensor until de-energization of the scram pilot valve solenoids. The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.
B WlELING FLOOR SECONDARY CONTAINMENT INTEGRITY 1.35 REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY shall exist when:
a.
All refueling floor secondary containment penetrations required to be closed during accident conditions are either:
LIMERICK - UNIT 1 1-6 Amendment No. 33,6f,105 w
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DEFINITIONS REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY (Continued) 1.
Capable of being closed by an OPERABLE secondary containment automatic isolation system, or 2.
Closed by at least one manual valve, blind-flange, slide gate damper, or deactivated automatic valve secured in its closed position, except as provided by Specification 3.6.5.2.2.
l b.
All refueling floor secondary containment hatches and, blowout panels are closed and sealed.
c.
The standby gas treatment system is in compliance with the requirements of specification 3.6.5.3.
d.
At least one door in each access to the refueling floor secondary containment is closed.
e.
The sealing mechanism associated with en-h refueling floor secondary containment penetration, e.g., welds, beliows, or 0-rings', is OPERABLE.
i f.
The pressure within the refueling floor secondary containment is le s than or equal to the value required by Specification 4.6.5.1.2a.
REPORTABLE EVENT 1.36 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50.
R00 DENSITY 1.37 R00 DENSITY shall be the number of control rod notches inserted as a fraction of the total number of control rod notches.
All rods fully inserted is equivalent to 100% R00 DENSITY.
SHUTDOWN MARGIN 1.38 SHUTDOWN MARGIN shall be the amount of reactivity by which the reactor is suberitical or would be subcritical assuming all control rods are fully inserted except for the single control rod of highest reactivity worth which is assumed to be fully withdrawn and the reactor is in the shutdown condition; cold, i.e. 68'F; and xenon free.
SITE BOUNDARY 1.39 The SITE BOUNDARY shall be that line as defined in Figure 5.1.3-la.
1.40 (Deleted)
SOURCE CHECK 1.41 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.
LIMERICK - UNIT 1 1-7 Amendment No. (8,66,105
CONTAINMENT SYSTEMS REACTOR ENCLOSURE SECONDARY CONTAINMENT AUTOMATIC ISOLATION VALVES 1
LIMITING CONDITION FOR OPERATION 3.6.5.2.1 The reactor enclosure secondary containment ventilation system auto-matic isolation valves shall be OPERABLE.
l APPLICABILITY:
OPERATIONAL CONDITIONS 1, 2, and 3.
ACTION:
With one or more of the reactor secondary containment ventilation-system automatic isolation valves inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is open and within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> either:
Restore the inoperable valves to OPERABLE status, or b.
Isolate each affected penetration by use of at least one daactivated valve secured in the isolation position, or c.
Isolate each affected penetration by use of at least one closed manual valve, blind flange or slide gate damper.
Otherwise, in OPERATIONAL CONDITION 1, 2, or 3, 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 4.6.5.2.1 Each reactor enclosure secondary containment ventilation system automatic isolation valve shall be ' demonstrated OPERABLE:
l a.
Prior to returning the valve to service after maintenance, rep' air or replacement work is performed on the valve or its associated actuator, control or power circuit by cycling the valve through at least one complete cycle of full travel and verifying the specified isolation time.
b.
At least once per 24 months by verifying that on a containment isolation test signal each isolation valve actuates to its isolation position.
c.
By verifying the isolation time to be within its limit at least once per 92 days.
LIMERICK - UNIT 1 3/4 6-48 Amendment No. 23,33,71,105
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I I
1 1
1 i
- 1 4
i f
THE INFORMATION FROM THIS TECHNICAL SPECIFICATIONS SECTION HAS BEEN RELOCATED TO THE UFSAR.
LIMERICK - UNIT 1 3/4 6-49 Amendment No. 23,105
1 CONTAINMENT SYSTEMS REFUELING AREA SECONDARY CONTAINMENT AUTOMATIC ISOLATION VALVES LIMITING CONDITION FOR OPERATION 3.6.5.2.2 The refueling area secondary containment ventilation system automatic isolation valves shall be OPERABLE.
l APPLICABILITY:
OPERATIONAL CONDITION *.
ACTION:
With one or more of the refueling area secondary containment ventilation system automatic isolation valves inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is open and within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> either:
a.
Restore the inoperable valves to OPERABLE status, or b.
Isolate each affected penetration by use of at least one deactivated valve secured in the isolation position, or c.
Isolate each affected penetration by use of at least one closed manual valve, blind flange or slide gate damper.
3 Otherwise, in OPERATIONAL CONDITION *, suspend handling of irradiated fuel in the refueling area secondary containment, CORE ALTERATIONS and operations with a potential for draining the reactor vessel. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS I
4.6.5.2.2 Each refuelin area secondary containment ventilation system auto-matic isolation valve sha 1 be demonstrated OPERABLE:
l
]
a.
Prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit by cycling the valve through at least one complete cycle of full travel and verifying the specified isolation time.
b.
At least once per 24 months by verifying that on a containment isolation test signal each isolation valve actuates to its isolation position.
c.
By verifying the isolation time to be within its limit at least once per 92 days.
- Required when (1) irradiated fuel is being handled in the refueling area secondary containment, or (2) during CORE ALTERATIONS, or (3) during opera-tions with a potential for draining the reactor vessel with the vessel head removed and fuel in the vessel.
I LIMERICK - UNIT 1 3/4 6-50 Amendment No. 6,49,7f,105
P f
I l
i l
THE INFORMATION FROM THIS TECHNICAL SPECIFICATIONS SECTION HAS BEEN RELOCATED TO THE UFSAR.
LIMERICK - UNIT 1 3/4 6-51 Amendment No. 23,105 1
_ _. _.. ~... _. _. _ _ _.. _ -.
1 f
?
THE INFORMATION FROM THIS TECHNICAL SPECIFICATICHS SECTION HAS BEEN RELOCATED TO THE UFSAR.
LIMERICK - UNIT 1 3/4 6-51a Amendment No. 23,105
CONTAINMENT SYSTEMS BASES
{
3/4.6.5 SECONDARY CONTAllMENT (Continued)
(
The field tests for bypass leakage across the SGTS charcoal adsorber and j
HEPA filter banks are performed at a flow rate of 3000 10% cfm. This flow l
rate corresponds to the maximum overall three zone inleakage rate of 3264 cfs.
The SGTS filter train pressure drop is a function of air flow rate and i
filter conditions.
Surveillance testing is performed using either the SGTS or drywell purge fans to provide operating convenience.
4 d
Each reactor enclosure secondary containment zone and refueling area i
secondary containment zone is tested independently to verify the design leak l
tightness. A design leak tightness of 1250 cfm or less for each reactor enclosure and 764 cfm or less for the refueling area at a 0.25 inch of vacuum water gage will ensure that containment integrity is maintained at an acceptable level if all zones are connected to the SGTS at the same time.
The Reactor Enclosure Secondary Containment Automatic Isolation Valves and Refueling Area Secondary Containment Automatic Isolation Valves can be found in the UFSAR.
The post-LOCA offsite dose analysis assumes a reactor enclosure secondary containment post-draw down leakage rate of 1250 cfm and certain post-accident X/Q values. While the post-accident X/Q values represent a statistical inter-pretation of historical meteorological data, the highest ground level wind speed which can be associated with these values is 7 mph (Pasquill-Gifford stability Class G for a ground level release). Therefore, the surveillance requirement assures that the reactor enclosure secondary containment is verified under meteorological conditions consistent with the assumptions utilized in the design basis analysis. Reactor Enclosure Secondary Containment leakage tests that are successfully performed at wind speeds in excess of 7 mph would also satisfy the leak rate surveillance requirements, since it shows compliance with more conservative test conditions.
3/4.6.6 PRIMARY CONTAINMENT ATMOSPHERE CONTROL The OPERABILITY of the systems required for the detection and control of hydrogen combustible mixtures of hydrogen and oxygen ensures that these systems will be available to maintain the hydrogen concentration within the primary containment below the lower flammability limit during post-LOCA conditions.
The primary containment hydrogen recombiner is provided to maintain the oxygen concentration below the lower flammability limit. The combustible gas analyzer is provided to continuously monitor, both during nomal operations and post-LOCA, the hydrogen and oxygen concentrations in the primary containment. The primary containment atmospheric mixing system is provided to ensure adequate mixing of the containment atmosphere to prevent localized accumulations of hydrogen and oxygen from exceeding the lower flammability limit. The hydrogen control system is consistent with the recommendations of Regulatory Guide 1.7, " Control of Combustible Gas Concentrations in Containment Following a LOCA," March 1971.
LIMERICA - UNIT 1 B 3/4 6-6 Amendment No. 8,105
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20086-0001
%.....)'
l PHILADELPHIA ELECTRIC COMPANY i
DOCKET NO. 50-353 q
LIMERICK GENERATING STATION. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE i-Amendment No. 69 l
License No. NPF-85 l
l l
1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Philadelphia Electric Company (the licensee) dated September 14, 1995, and supplemented by letter dated October 27, 1995, complies with the standards and requirements of the Atomic Energy Act of.1954, as amended (the Act), and the l
l Comission's rules and regulations set forth in 10 CFR Chapter I; 1
l B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; l
C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conduct <;d without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; i
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 CFR Part 51 l
of the Comission's regulations and all applicable requirements have l
been satisfied.
l-l 9
7 7+-
. 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 Soecifications The Technical Specifications _ contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 69, are hereby incorporated into this license.
Philadelphia Electric Company 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, to'be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COP 911SSION Y F. Stolz, Dirg: tor oject DirectoraM I-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Ch?.: :?3 to tS lethnical Specifications Date of Issuance:
November 20, 1995
ATTACHMENT TO LICENSE AMENDMENT NO. 69 FACILITY OPERATING LICENSE NO. NPF-85 DOCKET NO. 50-353 Replace the following pages of the Appendix A Technical Specifications with the attached pages. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change.
Remove Insert xiii xiii 1-6 1-6 1-7 1-7 3/4 6-48 3/4 6-48 3/4 6-49 3/4 6-49 3/4 6-50 3/4 6-50 3/4 6-51 3/4 6-51 3/4 6-51a 3/4 6-51a B 3/4 6-6 B 3/4 6-6
4 l
INDEX l
LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS l
SECTION PAlig I
CONTAINMENT SYSTEMS (Continued) 3/4.6.4 VACUUM RELIEF l
Suppression Chamber - Drywell Vacuum Breakers.........
3/4 6-44 3/4.6.5 SECONDARY CONTAINMENT Reactor Enclosure Secondary Containment Integrity.....
3/4 6-46 Refueling Area Secondary Containment Integrity........
3/4 6-47 i
l Reactor Enclosure Secondary Containment Automatic Isolation Va1ves......................................
3/4 6-48 I
Refueling Area Secondary Containment Automatic I s ol at i on Va1 ve s......................................
3/4 6-50 i
I Standby Gas Treatment System - Common System..........
3/4 6-52 Reactor Enclosure Recirculation System................
3/4 6-55 3/4.6.6 PRIMARY CONTAINMENT ATMOSPHERE CONTROL Primary Containment Hydrogen Recombiner Systems.......
3/4 6-57 Drywell Hydrogen Mixing System........................
3/4 6-58 l
Drywell and Suppression Chamber Oxygen Concentration..
3/4 6-59 1
l i
3/4.7 PLANT SYSTEMS I
3/4.7.1 SERVICE WATER SYSTEMS Residual Heat Removal Service Water System - Common System................................................
3/4 7-1 Emergency Service Water System - Common System........
3/4 7-3 Ultimate Heat Sink....................................
3/4 7-5 1
LIMERICK - UNIT 2 xiii Amendment No. 69 I
DEFINITIONS PURGE - PURGING 1.31 PURGE or PURGING shall be the controlled process of discharging air or gas from a confinement to maintain temperature, pressure, humidity, concentration or other operating condition, in such a manner that
(
replacement air or gas is required to purify the confinement.
RATED THERMAL POWER 1.32 RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 3458 MWt.
1 REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY 1.33 REACTOR ENCLOSURE SECONDARY CONTAINMENT INTEGRITY shall exist when:
a.
All reactor enclosure secondary containment penetrations required to be closed during accident conditions are either:
1.
Capable of being closed by an OPERABLE secondary containment automatic isolation system, or 2.
Closed by at least one manual valve, blind flange, slide gate damper or deactivated automatic valve secured in its closed position, except as provided by Specification 3.6.5.2.1.
l b.
All reactor enclosure secondary containment hatches and blowout panels i are closed and sealed.
c.
The standby gas treatment system is in compliance with the requirements of Specification 3.6.5.3.
d.
The reactor enclosure recirculation system is in compliance with the requirements of Specification 3.6.5.4.
e.
At least one door in each access to the reactor enclosure secondary containment is closed.
f.
The sealing mechanism associated with each reactor enclosure secondary containment penetration, e.g., welds, bellows, or 0-rings, is OPERABLE.
g.
The pressure within the reactor enclosure secondary containment is less than or equal to the value required by Specification 4.6.5.1.la.
REACTOR PROTECTION SYSTEM RESPONSE TIME 1.34 REACTOR PROTECTION SYSTEM RESPONSE TIME shall be the time interval from when the monitored parameter exceeds its trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids. The response time may be measured by any series of sequential, overlapping or total steps such that the entire response time is measured.
REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY 1.35 REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY shall exist when:
a.
All refueling floor secondary containment penetrations required to be closed during accident conditions are either:
LIMERICK - UNIT 2 1-6 Amendment No. 48,51,69
DEFINITIONS
+
REFUELING FLOOR SECONDARY CONTAINMENT INTEGRITY (Continued) 1.
Capable of being closed by an OPERABLE secondary containment automatic isolation system, or 2.
- Closed by at least one manual valve, blind flange, slide gate damper or deactivated automatic valve secured in its closed position, except as provided by Specification 3.6.5.2.2.
b.
All refueling floor secondary containment hatches and blowout panels are closed and sealed.
c.
The standby gas treatment system is in compliance with the requirements of Specification 3.6.5.3.
d.
At least one door in each access to the refueling floor secondary containment is closed.
e.
The sealing mechanism associated with each refueling floor secondary containment penetration, e.g., welds, bellows, or 0-rings, is OPERABLE.
f.
The pressure within the refueling floor secondary containment is less than or equal to the value required by Specification 4.6.5.1.2a.
}
REPORTABLE EVENT 1.36 A REPORTABLE EVENT shall be any of those conditions specified in Section 50.73 to 10 CFR Part 50.
ROD DENSITY 1.37 ROD DENSITY shall be the number of control rod notches inserted as a fraction l
of the total number of control rod notches. All rods fully inserted is equivalent to 100% R00 DENSITY.
4 SHUTDOWN MARGIN 1.38 SHUTDOW MARGIN shall be the amount of reactivity by which the reactor is subcritical or would be subcritical assuming all control rods are fully.
inserted except for the single control rod of highest reactivity worth which is assumed to be fully withdrawn and the reactor is in the shutdown condition; cold, i.e. 68'F; and xenon free.
SITE BOUNDARY 1.39 The SITE BOUNDARY shall be that line as defined in Figure 5.1.3-la.
1 1.40 (Deleted)
SOURCE CHECK 1.41 A SOURCE CHECK shall be the qualitative assessment of channel response when the channel sensor is exposed to a radioactive source.
LIMERICK - UNIT 2 1-7 Amendment No, ff,48,69
CONTAINMENT SYSTEMS REACTOR ENCLOSURE SECONDARY CONTAINMENT AUTOMATIC ISOLATION VALVES LIMITING CONDITION FOR OPERATION l
3.6.5.2.1 The reactor enclosure secondary containment ventilation system auto-catic isolation valves shall be OPERABLE.
l j
APPLICABILITY:
OPERATIONAL CONDITIONS 1, 2, and 3.
ACTION:
With one or more of the reactor secondary containment ventilation system automatic isolation valves inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is open and within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> either:
a.
Restore the inoperable valves to OPERABLE status, or b.
Isolate each affected penetration by use of at least one deactivated valve secured in the isolation position, or c.
Isolate each affected penetration by use of at least one closed manual valve, blind flange or slide gate damper.
Otherwise, in OPERATIONAL CONDITION 1, 2, or 3, 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 4.6.5.2.1 Each reactor enclosure secondary containment ventilation system automatic isolation valve shall be demonstrated OPERABLE:
l a.
Prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit by cycling the valve through at least one complete cycle of full travel and verifying the specified isolation time.
l b.
At least once per 24 months by verifying that on a containment i
isolation test signal each isolation valve actuates to its isolation I
position.
c.
By verifying the isolation time to be within its limit at least once per 92 days.
l LIMERICK - UNIT 2 3/4 6-48 Amendment No. H,69 l
1
I 4
l l
THE INFORMATION FROM THIS TECHNICAL SPECIFICATIONS SECTION HAS BEEN RELOCATED TO THE UFSAR.
l i
LINERICK - UNIT 2 3/4 6-49 Amendment No. 69
,s.
CONTAINMENT SYSTENS REFUELING ARE?. SECONDARY CONTAINMENT AUTOMATIC ISOLATION VALVES LINITING C0fEITION FOR OPERATION 1
3.6.5.2.2 The refueling area secondary containment ventilation system automatic isolation valves shall be OPERABLE.
l APPLICABILITY:
OPERATIONAL CONDITION *.
ACTION:
With one or more of the refueling area secondary containment ventilation system automatic isolation valves inoperable, maintain at least one isolation valve OPERABLE in each affected penetration that is open and within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> either:
a.
Restore the inoperable valves to OPERABLE status, or b.
Isolate each affected penetration by use of at least one deactivated valve secured in the isolation position, or c.
Isolate each affected penetration by use of at least one closed manual valve, blind flange or slide gate damper.
Otherwise, in OPERATIONAL CONDITION *, suspend handling of irradiated i
fuel in the refueling area secondary containment, CORE ALTERATIONS and i
operations with a potential for draining the reactor vessel. The provisions of Specification 3.0.3 are not applicable.
SURVEILLANCE REQUIREMENTS 4.6.5.2.2 Each refueling area secondary containment ventilation system auto-matic isolation valve shall be demonstrated OPERABLE:
l a.
Prior to returning the valve to service after maintenance, repair or replacement work is performed on the valve or its associated actuator, control or power circuit by cycling the valve through at least one complete cycle of full travel and verifying the specified isolation time.
- b. At least once per 24 months by verifying that on a containment isolation test signal each isolation valve actuates to its isolation position.
- c. By verifying the isolation time to be within its limit at least once per 92 days.
- Required when (1) irradiated fuel is being handled in the refueling area secondary containment, or (2) during CORE ALTERATIONS, or (3) during operations with a potential for draining the reactor vessel with the vessel head removed and fuel in the vessel.
LIMERICK - UNIT 2 3/4 6-50 Amendment No. H,69
m I
THE INFORMATION FROM THIS TECHNICAL SPECIFICATIONS SECTION HAS BEEN RELOCATED TO THE UFSAR.
LIMERICK - UNIT 2 3/4 6-51 Amendment No. 69
1 l
l l
1 1
THE INFORMATION FROM THIS TECHNICAL SPECIFICATIONS SECTION HAS BEEN RELOCATED TO THE UFSAR.
LIMERICK - UNIT 2 3/4 6-514 Amendment No. 69
s.
CONTAINMENT SYSTEMS BASES f
SECONDARY CONTAIMENT (Continued) i The SGTS fans are sized for three zones and therefore, when aligned to a single zone or two zones, will have excess capacity to more quickly drawdown the affected zones. There is no maximum flow limit to individual zones or pairs of zones and the air balance and drawdown time are verified when all three zones are connected to the SGTS.
The three zone air balance verification and drawdown test will be done after any major system alteration, which is any modificatia which will have an effect on the SGTS flowrate such that the ability of the WTS to drawdown the reactor enclosure to greater than or equal to 0.25 inch of vaam water gage in less than or equal to 126 seconds could be affected.
The field tests for bypass leakage across the SGTS charcoal adsorber and HEPA filter banks are performed at a flow rate of 3000 t 10% cfm. This flow rate corresponds to the maximum overall three zone inleakage rate of 3264 cfa.
1 The SGTS filter train pressure drop is a function of air flow rate and filter conditions.
Surveillance testing.is performed using either the SGTS or drywell purge fans to provide operating convenience.
j Each reactor enclosure secondary containment zone and refueling area i
secondary containment zone is tested independently to verify the design leak
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4 tightness. A design leak tightness of 1250 cfm or less for each reactor enclosure and 764 cfm or less for the refueling area at a 0.25 inch of vacuum water gage will ensure that containment integrity is maintained at an acceptable level if all zones are connected to the SGTS at the same time.
The Reactor Enclosure Secondary Containment Automatic Isolation Valves and Refueling Area Secondary Containment Automatic Isolation Valves can be found in the UFSAR.
The post-LOCA offsite dose analysis assumes a reactor enclosure secondary containment post-draw down leakage rate of 1250 cfm and certain post-accident X/Q values. While the post-accident X/Q values represent a statistical inter-pretation of historical meteorological data, the highest ground level wind speed which can be associated with these values is 7 mph (Pasquill-Gifford stability Class G for a ground level release). Therefore, the surveillance requirement assures that the reactor enclosure secondary containment is verified under meteorological conditions consistent with the assumptions. utilized in the design basis analysis. Reactor Enclosure Secondary Containment leakage tests that are successfully performed at wind speeds in excess of 7 mph would also satisfy the leak rate surveillance requirements, since it shows compliance with more conservative test conditions.
LIMERICK - UNIT 2 B 3/4 6-6 Amendment No. 51, 69
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