ML20058G911
| ML20058G911 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 11/30/1993 |
| From: | PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC |
| To: | |
| Shared Package | |
| ML20058G902 | List: |
| References | |
| NUDOCS 9312100175 | |
| Download: ML20058G911 (7) | |
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ATTACHMENT 2 LIMERICK GENERATING STATION UNITS 1 AND 2 DOCKET N05. 50-352 50-353 LICENSE HOS. NPF-39 NPF-85 TECHNICAL SPECIFICATIONS CHANGE REQUEST NO. 93-07-0 AFFECTED PAGES UNIT I UNIT 2 3/4 6-14 3/4 6-14 B 3/4 6-3a B 3/4 6-3a B 3/4 6-4 B 3/4 6-4 ,
1 9312100175 931130 PDR ADOCK 05000352 P ^PDR
. CONTRINMENT SYSTEMS'
. i SdRVElllANCEREQUIREMENTS(Continued) !
. 1
- c. By verifying at least-two suppression chamber water level indicators >
and at least 8 suppression pool water temperature indicators in at least 8 locations, OPERABLE by performance of a: ;
- 1. CHANNEL CHECK at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,
- 2. CHANNEL FUNCTIONAL TEST at least once per 31 days, and
- 3. CHANNEL CALIBRATION at least once per 18 months, with the water level and temperature alarm setpoint for:
- 1. High water level 5 24'l "
- 2. High water temperature:
a) First setpoint 5 95'F b) Second setpoint s 105 F -
c) Third setpoint s llc"F ,
d) Fourth setpoint s 120*F
- d. Drywell-to-suppression chamber bypass leak tests shall be conducted at -
40 +/- 10 month intervals to coincide with the ILRT at an initial differential pressure of 4 psi and verifying that the A/(k calculated from the measured leakage is within the specified limit. If any drywell-to-suppression >
chamber bypass leak test fails to meet the specified limit, the test schedule for subsequent tests shall be reviewed and approved by the Commission. If two cor.secutive tests fail to meet the specified t limit, a test shall be performed at least every 24 months until two l ,
consecutive tests meet the specified limit, at which time the '
test schedule may be resumed. I
- c. By conducting a leakage test on the drywell-to-suppression chamber vacuum breakers at a differential pressure of at least 4.0 psi and verifying that the intal leakage area A//k contributed by all vacuum breakers is less :
than or equal to 24% of the specified limit and the leakage area for an individual set of vacuum breakers is less than or equal- to 12% of the ;
specified limit. The vacuum breaker leakage test shall be conducted during each refueling outage for which the drywell-to-suppression chamber bypass leak test in Specification 4.6.2.1.d is not conducted. '
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LIMERICK - UNIT 1 3/4 6-14 l
CONTAINMENT SYSTEMS l BASES l l
OfPRESSURIZATION SYSTEMS (Continued) ;
The drywell-to-suppression chamber bypass test at a differential pressure of at least 4.0 psi verifies the overall bypass leakage area for simulated LOCA conditions is less than the specified limit. For those outages where the drywell-to-suppression chamber ;
bypass leakage test in not conducted, the VB leakage test verifies that the VB leakage area is less than the bypass limit, with a 76% margin to the bypass limit to accommodate the i remaining potential leakage area through the passive structural components. Previous i drywell-to-suppression chamber bypass test data indicates that the bypass leakage through {
the passive structural components will be much less than the 76% margin. The VB leakage ;
limit, combined with the negligible passive structural leakage area, ensures that the .
drywell-to-suppression chamber bypass leakage limit is met for those outages for which the drywell-to-suppression chamber bypass test is not scheduled. i 3/4.6 32 PRIMARY CONTAINMENT ISOLATION VALVES !
l 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 relaase of radioactive material to the containment atmosphere or pressurization of the containment and is consistent with the requirements i 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-t ment will be consistent with the assumptions used in the analyses for a LOCA. i
}]_4 . 6 . 4 VACUUM RELIff i
Vacuum relief valves are provided to equalize the pressure between the i 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 ,
l not be inoperable in the open position since this would allow bypassing of the i 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 l 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 l position.
1
! Each vacuum breaker valve's position indication system is of great enough sensitivity to ensure that the maximum steam bypass leakage coefficient of A
7k - 0.05 ft' for the vacuum relief system (assuming one valve fully open) will not be exceeded.
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l LIMERICK - UNIT 1 B 3/4 6-4
A 14.5.2'DEPRESSURINT10N SYSTEMS' (Cont.) )J One of the surveillance requirements for the suppression pool cooling (SPC) mode'of the.RHR s'ystem is to' demonstrate that each RHR pump develops a flow rate 210,000 gpm while operating in the SPC mode with flow through the heat exchanger and its associated closed bypass valve, ensuring that pump performance
-has not degraded during the cycle and that the flow path is operable. This test ,
confirms one point on the pump design curve and is indicative of overall
. performance. Such inservice inspections confirm component operability, trend '
performance and detect incipient failures by indicating abnormal performance. The
'RHR heat exchanger bypass valve is used for-adjusting flow through the heat ,
exchanger, and is not designed to be a tight shut-off valve. With the bypass valve closed, a potion of the total flow still travels through the bypass, which can affect overall heat transfer. However, no heat transfer performance :
requirement of the heat exchanger is intended by the current Technical Specification surveillance requirement. This is confirmed by the lack of any flow requirement for the RHRSW system in Technical Specification Section 3/4.7.1.
Verifying an RHR flowrate through the heat exchanger does not demonstrate heat ,
removal capability in the absence of a requirement for RHRSW flow. LGS does "
perform heat transfer testing of the RHR heat exchangers as part of its response to Generic letter 89-13, which verified the commitment to meet the requirements of GDC 46.
-l Experimental data indicate that excessive steam condensing loads can be ,
avoided 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 r that the reactor can ne depressurized in a timely manner to avoid the regime of i potentially high suppression chamber loadings.
Because of the large volume and thermal capacity of the ' suppression pool,. l' 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 i that appropriate action cm be taken. ;
In addition to the limits on temperature of the suppression chamber pool l water, operating procedures define the action to be taken in the event a safety- 1 relief valve inadvertently opens or sticks open. As a minimum this action shall i 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. :
During a LOCA, potential leak paths between the drywell and suppression chamber i airspace could result in excessive containment pressures, since the steam flow into !
the airspace would bypass the heat sink capabilities of the chamber. Potential sources i of bypass leakage are the suppression chamber-to-drywell vacuum breakers (VBs), .
penetrations in the diaphragm floor, and cracks in the diaphragm floor and/or liner plate.and- !
downcomers located in the suppression chamber airspace. The containment pressure '
rssponse to the postulated bypass leakage can be mitigated by manually actuating the ;
sup A/(pression k equal tochamber sprays.
0.0500 ft2 An analysis to verify that the was performed operator for a design has sufficient time bypass leakage to. initiate the area of ;
sprays prior to exceeding the containment design pressure of 55 psig. The limit of 10% of the design value of 0.0500 ft2 ensures that the design basis for the steam bypass analysis j is met. ;
LIMERICK UNIT 1 B 3/4 6-3a
^ ~
- ~
C0sTAINMENT SYSTEMS SURifEllLANCE REQUIREMENTS (Continued) !
l
- c. By verifying at least two suppression chamber water level. indicators ,
and at least 8 suppression pool water temperature indicators -in at ;
least 8 locations, OPERABLE by. performance of a: -
.i
- 1. CHANNEL CHECK at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, ,
- 2. CHANNEL FUNCTIONAL TEST at least once per 31 days, and
]t
- 3. CHANNEL CALIBRATION at least once per 18 months, with the water level and temperature alarm setpoint for:
- 1. High water level s 24'1 "
- 2. High water temperature: ,
a) First setpoint s 95'F b) Second setpoint s 105 F c) Third setpoint s 110 F d) Fourth setpoint s 120*F
- d. Drywell-to-suppression chamber bypass leak ' tests shall .be conducted at ,
40 +/- 10 month intervals to coincide with the ILRT at an initial differential ;
pressure of 4 psi and verifying that the A/(k calculated from the measured -l leakage is within the specified limit. If any drywell-to-suppression ;
chamber bypass leak test fails to meet the specified limit, the test schedule for subsequent tests shall be reviewed and approved by the Commission. If two consecutive tests fail to meet the specified limit, a test shall be performed at least every 24 months until two j !
consecutive tests meet the specified limit, at which time the test schedule may be resumed. l- !
- e. By conducting a leakage test on the drywell-to-suppression chamber vacuum breakers at a differential pressure of at least 4.0 psi and verifying that the total leakage area A//k contributed by all vacuum breakers is less than or equal to 24% of the specified limit and the leakage >
area for an individual set of vacuum breakers is less than or equal to 12% of ,
the specified limit. The vacuum breaker leakage test hall be conducted during i cach refueling outage for which the drywell-to-suppression chamber bypass leak ;
test in Specification 4.6.2.1.d is not conducted.
s f
LIMERICK - UNIT 2 3/4 6-14 7 4 r - ,. - ., . -, - - , - . .
3/I.6_.2'DEPRESSUR_lZATION SYSTEMS (Cont.)
One' of the surveillance requirements for the suppression pool cooling (SPC) imode of'the RHR-system is to demonstrate that each RHR aump develoas a flow rate 210,000 gpm while operating in the SPC mode with flow tirough the Teat exchanger and its associated closed bypass valve, ensuring that pump performance 1 has not degraded during the cycle and that the flow path is operable. This test confirms one point on the pump design curve and is indicative of overall
. performance. Such inservice inspections confirm component operability,. trend performance and detect incipient failures by indicating abnormal aerformance. The RHR heat exchanger bypass valve is used for adjusting flow throug1 the heat exchanger, and-is not designed to be a tight shut-off valve. With the bypass ,
valve closed, a potion of the total flow still travels through the bypass, which ;
can affect overall heat transfer. However, no heat transfer "erformance requirement of the heat exchanger is intended by the currer ; chnical '
Specification surveillance requirement. This is confirmed )y ,~ 0 lack of any flow -
requirement for the RHRSW system in Technical Specification E.cion 3/4.7.1.
Verifying an RHR flowrate through the heat exchanger does not demonstrate heat removal capability in the absence of a requirement for RHRSW flow. LGS does perform heat transfer testing of the RHR heat exchangers as part of its response to Generic letter 89-13, which verified the commitment to meet the requirements of GDC 46.
Experimental data indicate that excessive steam condensing loads can be avoided if the peak local temperature of the suppression pool is maintained below
'200*F ouring any period of relief valve operation for T-quencher devices.
Saecifications have been placed on the envelope of reactor operating conditions so 11at the reactor can be depressurized in a timely manner to avoid the regime of potentially high suppression chamber loadings. ,
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 i 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 valvo 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. '
During a LOCA, potential leak paths between the drywell and suppression chamber airspace could result in excessive containment pressures, since the steam flow into the airspace would bypass the heat sink ca) abilities of the chamber. Potential sources of bypass _ leakage are the suppression chamaer-to-drywell vacuum breakers (VBs),
- penetrations in the diaphragm floor, and cracks in the diaphragm floor and/or liner ' plate and ,
downcomers located in the suppression chamber airspace. -The containment aressure response '
to the postulated bypass leakage can be mitigated by manually actuating tie suppression chamber sprays. An analysis was performed for a design bypass leakage area-of A//k equal .
to 0.0500 ft* to verify that the operator has sufficient time to initiate the s arays prior to exceeding the containment design pressure of 55 psig. The limit of 10% of tie design -
value of 0.0500 ft' ensures that the design basis for the steam bypass analysis is met.
LIMERICK - UNIT 2 B 3/4 6-3a =i t
, - - e
!C6NTAll1 MENT SYSTEMS t
BASES DEPRESSURIZATION SYSTEMS (Continued)
The drywell-to-suppression chamber bypass test at a differential pressure of '!
at least 4.0 psi verifies the overall bypass leakage area for simulated LOCA ,
conditions is less than the specified limit. For those outages where the !
drywell-to-suppression chamber bypass leakage test in not conducted, the VB leakage test verifies that the VB leakage area is less than the bypass limit, with a ,
76% margin to the bypass limit to accommodate the remaining potential leakage area i through the passive structural components. Previous drywell-to-suppression chamber ,
bypass test data indicates that the bypass leakage through the passive structural -
components will be much less than the 76% margin. The VB leakage limit, combined ,
with the negligible passive structural leakage area, ensures that the drywell-to- i suppression chamber bypass leakage limit is met for those outages for which the drywell-to-suppression chamber bypass test is not scheduled.
P 3/4.6.3 PRIMARY CONTAINMENT [ SOLATION 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 t 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. l 3/4.6.4 VALUUM 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 l 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 o)eration 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 areakers 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 7k - 0.05 ft' ,
for the vacuum relief system (assuming one valve fully open) will not be exceeded.
l l
l LIMERICK - UNIT 2 B 3/4 6-4 !
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