ML20064B343
| ML20064B343 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 10/02/1990 |
| From: | Butler W Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20064B344 | List: |
| References | |
| NUDOCS 9010160202 | |
| Download: ML20064B343 (16) | |
Text
_ _ _ _ _ - _ _ _ _ _ _ _ -- ______--__- - -__ _ _____-___- _ ___ - __-_- _ - - ___ - - _
[f' UNITED STATES g'
g NUCLEAR REGULATORY COMMISSION p
wAssiwoTow. o. c. ro666
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PHIL ADELPHI A ELECTRIC COMPANY 00CKET NO. 50-357 LIMERICK GENERATING STATION, UNIT 1 AMENDMENT TO Fk,f1LITY OPERATING LICENSE Amendment No. 46 1.icense 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 October 11, 1989, as supplemented by letter dated April 9, 1990 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 provisions of the Act, and the rules and regulations of the Comission; 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 Comission's regulations; D.
The issuance of this emendment 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 sathfied.
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 Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 46 are hereby incorporated into this license. Philadelphia Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
9010160202 901002 ADOCK0500g{2 DR 1
l I
2 3.
This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/S/
Walter R. Butler, Director Project Directorate 1-2 Division of Reactor Projects 1/11
Attachment:
Changes to the Technical Specifications Date of Issuance: October 2, 1990 l
i 1
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l-d, PDI-2/P OGC PDI-2/D d) 1 fen RClark:
finh WButler 90
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3.
This license amendment is effective as of its date of issuance.
FOR THE NVCLEAR REGULATORY COMMISSION
().
Walter R. Butler, Director Project Directorate 1-2 Division of Reactor Projects 1/11 Attachment Charges to the Technical Specifications Date of Issuance: October 2, 1990 t
ATTACHMENT TO LICENSE AMENDMENT NO. 46 FACILITY OPERATING LICENSE NO. NPF-39 DOCKET NO. 50-352 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. Overleaf pages are provided to maintain document completeness.*
Remove Insert 3/4 6-43 3/4 6-43*
3/4 6-44 3/4 6-44 8 3/4 6-3 B 3/4 6-3*
B 3/4 6-4 B 3/4 6-4 i
i TABLE 3.6.3 1 PRIMARY CCNTAINMTNT ISOLATION VALVES N0'ATIGN l
i 90TES (Continued)'
21.
Automatic isolation signal causes TIP to retract; ball valve closes hen prece is fully retracteo.
i 22.
! solation carrier remains water filled or a water seal remains in the fire ocst*LCCA.
Isolation valve may be testec with water.
Isolation i
5 valve leakage is not inclucea in 0.60 La total Type B & C tests.
23.
Valve does not receive an isolation signal.
Valves will be open during Type A test.
Type C test not required.
24.
Both isolation signals required for valve closure.
25.
Deleted 26.
Valve stroke times Ifsted are maximum times verified by testing per Speci-fication 4.0.5 acceptance criteria.
The closure times for isolation valves in lines in which high energy line breaks could occur are identified with a i
single asterisk.
The closure times for isolation valves in lines which provide an open path from the containment to the environs are identified i
with a double asterisk.
27.
The reactor vessel head seal leak detection line (penetration 29A) excess flow check valve is not subject to OPERABILITY testing.
This valve will not be exposed to primary system pressure except under the unlikely con ditions of a seal failure where it could be partially pressurized to reactor pressure.
Any leakage path is restricted at the source; therefore, this valve need not be OPERA 8ILITY tested.
28.
(DELETED) 23.
Valve may be open during normal operation; capable of manual isolation from control room.
Position will be controlled procedurally.
30.
Valve normally open, closes on scram signal.
31.
Valve 41-1016 is an outboard isolation barrier for penetrations X 9A, 8 and X-44 Leakage through valve 41-1016 is included in the total for penetration X-44 only.
32.
Feedwater long path recirculation valves are sealed closed whenever the reactor is critical and reactor pressure is greater than 600 psig.
The valves are expected to be opened only in the following instances:
Flushing of the condensate and feedwater systems during plant startup.
a.
b.
Reactor pressure vessel hydrostatic testing, which is conducted follow-ing each refueling outage prior to commencing plant startup.
Therefore, valve stroke timing in accordance with Specification 4.0.5 is not required.
33.
Valve also constitutes a Unit 2 Reactor Enclosure Secondary Containment Automatic Isolation Valve and a Refueling Area Secondary Containment Automatic Isolation Valve as shown in Table 3.6.5.2.1-1 and Table 3.6.5.2.2-1 respectively.
LIMERICK - UNIT 1 3/4 6-43 Amendment No. 23 JUN 141989
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l CONTAINMENT SYSTEMS 3/4.6.4 VACUUM RELIEF 1
SUPPRESSION CHAMBER - DRYWELL VACUUM BREAKERS LIMITING CONDITION FOR OPERATION l
l 3.6.4.1 Three pairs of suppression chamber - drywell vacuum breakers shall be OPERABLE and all suppression chamber - drywell vacuum breakers shall be closed.
l APPLICABILITY:
OPERATIONAL CONDITIONS 1, 2, and 3.
ACTION:
With one or more vacuum breakers in one of the three required pairs of a.
suppression chamber - drywell vacuum breaker pairs inoperable for opening but known to be closed, restore at least one inoperable pair of vacuum breakers to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 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 />.
b.
With one suppression chamber - drywell vacuum breaker open, verify the other vacuum breaker in the pair to be closed within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />; restore the open vacuum breaker to the closed position within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 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 />.
With one position indicator of any suppression chamber - drywell c.
1 vacuum breaker inoperable:
1.
Verify the other vacuum breaker in the pair to be closed within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and at least once per 15 days thereafter, or 2.
Verify the vacuum breaker (s) with the inoperable position indicator to be closed by conducting a test which demonstrates that the AP is maintained e.t greater than or equal to 0.7 psi for one hour without makeup within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and at least once per 15 days thereafter.
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 />.
l l
l l
LIMERICK - UNIT 1 3/4 6-44 Amendment No. 46
CONTAINMENT SYSTEMS BASES l
3/4.6.2.
DEPRESSURIZATION SYSTEMS l
The specifications of this section ensure that the primary containment pressure will not exceed the design pressure of 55 psig during primary system blowdown from full operating pressure.
The suppression chamber water provides the heat sink for the reactor coolant system energy release following a postulated rupture of the system.
)
The suppression chamber water volume must absorb the associated decay and structural sensible heat released during reactor coolant system blowdown from i
1040 psig.
Since all of the gases in the drywell are purged into the suppres-sion chamber air space during a loss-of-coolant accident, the pressure of the suppression chamber air space must not exceed 55 psig.
The design volume of the suppression chamber, water and air, was obtained by considering that the total volume of reactor coolant is discharged to the suppression chamber and that the drywell volume is purged to the suppression chamber.
i Using the minimum or maximum water volumes given in this specification, suppression pool pressure during the design basis accident is approximately 30 psig which is below the design pressure of 55 psig.
Maximum water volume 3
of 134,600 ft results in a downcomer submergence of 12'3" and the minimum volume of 122,120 fta results in a submergence approximately 2'3" less.
The majority of the Bodega tests were run with a submerged length of 4 feet and with complete condensation.
Thus, with respect to the downcomer submergence, this specification is adequate.
The maximum temperature at the end of the blowdown tested during the Humboldt Bay and Bodega Bay tests was 170'F and this is conservatively taken to be the limit for complete condensation of the reactor coolant, although condensation would occur for temperatures above 170'F.
Should it be necessary to make the suppression chamber inoperable, this shall only be done as specified in Specification 3.5.3.
Under full power operating conditions, blowdown through safety / relief 2
valves assuming an initial suppression chamber water temperature of 95'F results in a bulk water temperature of approximately 136'F immediately following blowdown which is below the 190*F bulk temperature limit used for complete condensation i
l via T quencher devices.
At this temperature and atmospheric pressure, the avail-able NPSH exceeds that required by both the RHR and core spray pumps, thus there I
l is no dependency on containment overpressure during the accident injection phase, if both RHR loops are used for containment cooling, there is no dependency on containment overpressure for post-LOCA operations.
Experimental data indicate that excessive steam condensing loads can be avoided if the peak local temperature of the suppression pool is maintained 3
below 200*F during any period of relief valve operation for T quencher devices.
Specifications have been placed on the envelope of reactor operating conditions so that the reactor can be depressurized in a timely manner to avoid the regime of potentially high suppression chamber loadings.
LIMERICK - UNIT 1 B 3/4 6-3 Amendment No. 33 OCT 3 01989
I CONTAINMENT SYSTEMS BASES DEPRESSURIZATION SYSTEMS (Continued)
Because of the large volume and thermal capacity of the suppression pool, the volume and temperature normally changes very slowly and monitoring these parameters daily is suf ficient to establish any temperature trends.
By requiring the suppression pool temperature to be frequently recorded during periods of significant heat addition, the temperature trends will be closely followed so that appropriate action can be taken.
In addition to the limits on temperature of the suppression chamber pool water, operating procedures define the action to be taken in the event a safety-relief valve inadvertently opens or sticks open.
As a minimum this action shall include: (1) use of all available means to close the valve, (2) initiate suppres-sion pool water cooling, (3) initiate reactor shutdown, and (4) if other safety-relief valves are used to depressurize the reactor, their discharge shall be separated from that of the stuck-open safety / relief valve to assure mixing and uniformity of energy insertion to the pool.
3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES The OPERABIt.ITY of the primary containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurization of the containment and is consistent with the requirements of GDC 54 through 57 of Appendix A of 10 CFR Part 50.
Containment isolation within the time limits specified for those isolation valves designed to close automatically ensures that the release of radioactive material to the environ-I ment will be consistent with the assumptions used in the analyses for a LOCA.
l 3/4.6.4 VACUUM RELIEF Vacuum relief valves are provided to equalize the pressure between the suppression chamber and drywell.
This system will maintain the structural integrity of the primary containment under conditions of large differential I
pressures.
The vacuum breakers between the suppression chamber and the drywell must not be inoperable in the open position since this would allow bypassing of the suppression pool in case of an accident.
Two pairs of valves are required to protect containment structural integrity.
There are four pairs of valves (three to provide minimum redundancy) so that operation may continue for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> with no more than two pairs of vacuum breakers inoperable in the closed position.
Each vacuum breaker valve's position indication system is of great enough sensitivity to ensure that the maximum steam bypass leakage coefficient of A
/E=0.05ft2 for the vacuum relief system (assuming one valve fully open) will not be exceeded.
LIMERICK - UNIT 1 B 3/4 6-4 Amendment No.46 l
'e UNITED STATES
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NUCLEAR REGULATORY COMMISSION
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PHIL ADELPHI A ELECTRIC COMPANY DOCKET NO. 50-353 LIMERICK GENERATING STATION, UNIT 2 AMENOMENT TO FACILITY OPERATING LICENSE Amendment No. 9 License No. NPF-85 1.
The Nuclear Pegulatory Comission (the Commission) has found that A.
The application for amendment by Philadelphia Electric Company (the i
licensee) dated October 11. 1989, as supplemented by letter dated April 9, 1990 complies with the standards and requirements of the i
Atomic Energy Act of 1954, as amended (the Act), and the Comission's
't rules and regulations set forth in 10 CFR Chapter 1; 4
B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; C.
There is reasonable assurance (i) that the activities authorized by this amendment an 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; j
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 l
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.
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 Specifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No. 9
. are hereby incorporated into this license. Philadelphia Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
V F
^
l i 3.
This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION
/S/
h Walter R. Butler, Director Project Directorate 1-2 Division of Reactor Projects 1/11
Attachment:
Changes to the Technical Specifications Date of Issuance: October 2, 1990 l
i i
skA PDI-2/PM OGC PDI-2/D I
Jin RClark:tr 4M WButler 90 of//6/90 of/tt /90 lO / \\/90 W
2 3.
This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION R
Walter R. Butler, Director Project Directorate 1-2 Division of Reactor Projects 1/11
Attachment:
Changes to the Technical Specifications Date of Issuance:
October 2, 1990 I
i i
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t ATTACHMENT TO LICENSE AMENDMENT NO. 9 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. Overleaf pages are provided to maintain document completeness.
Remove Insert 3/4 6-43 3/4 6-43*
3/4 6-44 3/4 6-44 B 3/4 6-3 B 3/4 6-3*
B 3/4 6-4 B 3/4 6-4
.i TABLE 3.6.3-1 PRIMARY CONTAINMENT ISOLATION VALVES NOTATION NOTES (Continued) 21.
Automatic isolation signal causes TIP to retract.; ball valve closes when probe is fully retracted.
22.
Isolation barrier remains water filled or a water seal remains in the line post-LOCA.
Isolation valve may be tested with water.
Isolation i
valve leakage is not included in 0.60 La total Type 8 & C tests.
1 23.
Valve does not receive an isolation signal.
Valves will be open during Type A test.
Type C test not required, i
24.
Both isolation signals required for valve closure.
25.
Deleted 26.
Valve stroke times listed are maximum times verified by testing per Speci-fication 4.0.5 acceptance criteria.
The closure times for isolation valves in lines in which high-energy line breaks could occur are identified with a single asterisk.
The closure times for isolation valves in lines which provide an open path from the containment to the environs are identified with a double asterisk.
27.
The reactor vessel head seal leak detection line (penetration 29A) excess flow check valve is not subject to OPERABILITY testing.
This valve will not be exposed to primary systes pressure except under the unlikely con-ditions of a seal fnilure where it could be partially pressurized to reactor pressure.
Any leakage path is restricted at the source; therefore, this valve need not be OPERA 8ILITY tested.
l 28.
(DELETED) i 29.
Valve may be open during normal operation; capable of manual isolation from control room.
Position will be controlled procedurally.
30.
Valve normally open, closet, on scram signal.
l 31.
Vcive 41-2016 'is an outboard isolation barrier for penetrations X-9A, B and X-44.
Leakage through valve 41-2016 is included in the total for i
penetration X-44 only.
32.
Feedwater long path recirculation valves are sealed closed whenever the reactor is critical and reactor pressure is greater than 600 psig.
The valves are expected to be opened only in the following instances:
i a.
Flushing of the condensate and feedwater systems during plant startup.
l b.
Reactor pressure vessel hydrostatic testing, which is conducted follow-ing each refueling outage prior to commencing plant startup.
Therefore, valve stroke timing in accordance with Specification 4.0.5 is not required.
33.
Valve also constitutes a Unit 1 Reactor Enclosure Secondary Containment Automatic Isolation Valve and a Refueling ~ Area Secondary Containment Automatic Isolation Valve as shown in Table 3.6.5.2.1-1 and Table 3.6.5.2.2-1, respectively.
34.
Isolation signal causes recombiner to trip; valve closes when recombiner is not operating.
LIMERICK - UNIT 2 3/4 6-43 l
1 CONTAINMENT SYSTEMS 3/4.6.4 VACUUM RELIEF 1
SUPPRESSION CHAMBER - ORYWELL VACUUM BREAKERS i
i LIMITING CONDITION FOR OPERATION l
l 3.6.4.1 Three pairs of suppression chamber - drywell vacuum breakers shall be OPERABLE and all suppression chamber - drywell vacuum breakers shall be closed.
i APPLICABILITY:
OPERATIONAL CONDITIONS 1, 2, and 3.
ACTION:
With one or more vacuum breakers in one of the three required pairs of a.
suppression chamber - drywell vacuum breaker pairs inoperable for opening i
l.
but known to be closed, restore at least one inoperable pair of vacuum 3
breakers to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 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 1
following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
I b.
With one suppression chamber - drywell vacuum breaker open, verify the other. vacuum breaker in the pair to be closed within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />; restore the open vacuum breaker to the closed position within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 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 />.
With one position indicator of any suppression chamber - drywell c.
vacuum breaker inoperable:
i l
1.
Verify the other vacuum breaker in the pair to be closed within i
2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and at least once per 15 days thereafter, or 2.
Verify the vacuum breaker (s) with the inoperable position indicator to be closed by conducting a test which demonstrates that the AP is maintained at greater than or equal to 0.7 psi for one hour without makeup within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and at least once l
per 15 days thereafter.
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 />, l
LIMERICK - UNIT 2 3/4 6-44 Amendment No. 9 I
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CONTAINMENT SYSTEMS BASES 3/4.6.2.
OEPRESSURIZATION SYSTEMS The specifications of this section ensure that the primary containment pressure will not exceed the design pressure of 55 psig during primary system blowdown from full operating pressure.
The suppression chamber water provides the heat sink for the reactor coolant system energy release following a postulated rupture of the system.
The suppression chamber water volume must absorb the associated decay and structural sensible heat released during reactor coolant system blowdown from 1040 psig.
Since all of the gases in the drywell are purged into the suppres-sion chamber air space during a loss-of-coolant accident, the pressure of the suppression chamber air space must not exceed 55 psig.
The design volume of the suppression chamber, water and air, was obtained by considering that the total volume of reactor coolant is discharged to the suppression chamber and that the drywell volume is purged to the suppression chamber.
Using the minimum or maximum water volumes given in this specification, suppression pool pressure during the design basis accident is approximately 30 psig which is below the design pressure of 55 psig.
Maximum water volume 3
of 134,600 ft results in a downcomer submergence of 12'3" and the minimum volume of 122,120 ft3 results,in a submergence approximately 2'3" less.
The majority of the Bodega ttsts were run with a submerged length of 4 feet and with complete condensation.
Thus, with respect to the downcomer submergence, this specification is adequate.
The maximum temperature at the end of the blowdown tested during the Humboldt Bay and Bodega Bay tests was 170'F and this is conservatively taken to be the limit for complete condensation of the reactor coolant, although condensation would occur for temperatures above 170*F.
Should it be necessary to make the suppression chamber inoperable, this shall only be done as specified in Specification 3.5.3.
Under full power operating c'enditions, blowdown through safety /reliet i
valves assuming an initial suppression chamber water temperature of 95'F esults i
in a bulk water temperature of approximately 136*F immediately following ') lowdown which is below the 190'F bulk temperature limit used for complete condensation via T quencher devices.
At this temperature and atmospheric pressure, t'ie avail-able NPSH exceeds that required by both the RHR and core spray pumps, tr.us there is no dependency on containment overpressure during the accident injection phase.
If both RHR loops are used for containment cooling, there is no dependency on L
containment overpressure for post-LOCA operations.
Experimental data indicate that excessive steam condensing loads can be avolded if the peak local temperature of the suppression pool is maintained below 200'F during any period of relief valve operation for T quencher devices.
Specifications have been placed on the envelope of reactor operating conditions so that the reactor can be depressurized in a timely manner to avoid the regime of potentially high suppression chamber loadings.
LIMERICK - UNIT 2 B 3/4 6-3
CONTAINMENT SYSTEMS BASES DEPRESSURIZATION SYSTEMS (Continued)
Because of the large volume and thermal capacity of the suppression pool, the volume and temperature normally changes very slowly and monitoring these parameters daily is sufficient to establish any temperature trends.
By requiring the suppression pool temperature to be frequently recorded during periods of significant heat addition, the temperature trends will be closely followed so that appropriate action can be taken.
In addition to the limits on temperature of the suppression chamber pool water, operating procedures define the action to be taken in the event a safety-relief valve inadvertently opens or sticks open.
As a minimum this action shall include: (1) use of all available means to close the valve, (2) initiate suppres-sion pool water cooling, (3) initiate reactor shutdown, and (4) if other safety-relief valves are used to depressurize the reactor, their discharge shall be separated from that of the stuck-open safety / relief valve to assure mixing and uniformity of energy insertion to the pool.
3/4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES The OPERABILITY of the primary containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurization of the containment and is consistent with the requirements of GDC 54 through 57 of Appendix A of 10 CFR Part 50.
Containment isolation within the time limits specified for those isolation valves designed to close automatically ensures that the release of radioactive material to the environ-ment will be consistent with the assumptions used in the analyses for a LOCA.
3/4.6.4 VACUUM RELIEF Vacuum relief valves are provided to equalize the pressure between the suppression chamber and drywell.
This system will maintain the structural integrity of the primary containment under conditions of large differential pressures.
The vacuum breakers between the suppression chamber and the drywell must not be inoperable in the open position since this would allow bypassing of the suppression pool in case of an accident.
Two pairs of valves are required to protect containment structural integrity.
There are four pairs of valves (three to provide minimum redundancy) so that operation may continue for up to 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> with no more than two pairs of vacuum breakers inoperable in the closed position.
Each vacuum breaker valve's position indication system is of great enough sensitivity to ensure that the maximum steam bypass leakage coefficient of A,
4 = 0.05 ft2 i
for the vacuum relief system (assuming one valve fully open) will not be exceeded.
LIMERICK - UNIT 2 B 3/4 6 4 Amendment No. 9