ML20063K719
| ML20063K719 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 02/17/1994 |
| From: | Chris Miller Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20063K722 | List: |
| References | |
| NUDOCS 9403020055 | |
| Download: ML20063K719 (16) | |
Text
{{#Wiki_filter:. fa arc o [ . ?*, UNITED STATES f [th, Cf.j NUCLEAR REGULATORY COMMISSION '$ h 's WASHINGTON, D.C. 20555-0001 0 N e PHILADELPHIA ELECTRIC COMPANY DOCKET N0. 50-352 LIMERICK GENERATING STATION. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 68 License No. NPF-39 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment by Philadelphia Electric Company (the licensee) dated November 30, 1993, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; 8. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (1) 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; 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. 9403020055 940217 PDR ADOCK 05000352 p PDR
n - l 2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this-license amendment, y 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. L 68, 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. FOR THE NUCLEAR REGULATORY COMMISSION CAade /17hlbu Charles L. Miller, Director Project Directorate I-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications - l Date of Issuance: February 17, 1994 l l j
~ 9 ATTACHMENT TO LICENSE AMENDMENT N0.68 FACILITY OPERATING LICENSE NO. NPF-39 OOCKET 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 decument completeness.* Remove Insert 3/4 6-13 3/4 6-13* 3/4 6-14 3/4 6-14 8 3/4 6-3 B 3/4 6-3* B 3/4 6-3a B 3/4 6-3a l B 3/4 6-4 8 3/4 6-4 q l
<3 CONTAINMENT SYSTEMS LIMITING CONDITION FOR OPERATION (Continued) ACTION: (Continued) 3. With the suppression chamber average water temperature greater than 120 F, depressurize the reactor pressure vessel to less than 200 psig within 12 hours. With only one suppression chamber water level indicator OPERABLE and/or c. with less than eight suppression pool water temperature indicators, one in each of the eight locations OPERABLE, restore the inoperable indicator (s) to OPERABLE status within 7 days or verify suppression chamber water level and/or temperature to be within the limits at least once per 12 hours. d. With no suppression chamber water level indicators OPERABLE and/or with less than seven suppression pool water temperature indicators covering at least seven locations OPERABLE, restore at least one water level indicator and at least seven water temperature indicators to OPERABLE i status within 48 hours or be in at least HOT SHUTOOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. i With the drywell-to-suppression chamber bypass leakage in excess of e. the limit, restore the bypass leakage to within the limit prior to increasing reactor coolant temperature above 200*F. SURVEILLANCE REQUIREMENTS 4.6.2.1 The suppression chamber shall be demonstrated OPERABLE: a. By verifying the suppression chamber water volume to be within the limits at least once per 24 hours. b. At least once per 24 hours by verifying the suppression chamber l average water temperature to be less than or equal to 95 F, except: 1. At least once per 5 minutes during testing which adds heat to the suppression chamber, by verifying the suppression chamber average water temperature less than or equal to 105*F. l 2. At least once per hour when suppression chamber average water temperature is greater than or equal to 95 F, by verifying; a) Suppression chamber average water temperature to be less than or equal to 110 F, and b) THERMAL POWER to be less than or equal to 1% of RATED THERMAL POWER 12 hours after suppression chamber average water temperature.has exceeded 95*F for more than 24 hours. 3. At least once per 30 minutes following a scram with suppression chamber average water temperature greater than or equal to 95 F, by verifying suppression chamber average water temperature less than or equal to 120 F. LIMERICK - UNIT 1 3/4 6-13
l CONTAINMENT SYSTEMS i SURVEILLANCE RE0VIREMENTS (Continued) 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 hours, 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 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 5 120*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 consecutive tests fail to meet the specified limit, a test shall be performed at least every 24 months until two I consecutive tests meet the specified limit, at which time the test schedule may be resumed. I 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 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. Amendment No. 68 LIMERICK UNIT 1 3/4 6-14
CONTAINMENT SYSTEMS BASES 3l 4 3/4.6.2 DEPRESSURIZATION 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 rated conditions. Since all of the gases in the drywell are purged into the suppression 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 chan'ber. Using the minimum or maximum water volumes given in this specification, suppression pool pressure during the design basis accident is below the design pressure. Maximum water volume of 134,600 ft' results in a downcomer submergence of 12'3" and the minimum volume of 122,120 ft' 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 temperature 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 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 via T-quencher devices. At this temperature and atmospheric pressure, the available NPSH exceeds that required by both the RHR and core spray pumps, thus 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 containment overpressure for post-LOCA operations. l LIMERICK - UNIT 1 B 3/4 6-3 Amendment No. 33, 57, i
3/4.6.2 DEpRESSURIZATION SYSTEMS -(Cont.) One of the surveillance requirements for the suppression pool cooling (SPC) mode of the RHR system 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 portion 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 Specifications 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 cc: -itment to meet the requirements of GDC 46. Experimental data indicate that excessivt steam condensing loads can be avoided if thc 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. 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-1 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. 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 capabilities of the chamber. Potential sources 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 response to the postulated bypass leakage can be mitigated by' manually actuating tha - A/(ppression chamber sprays. An analysis was performed for a design bypass leakage araa o su k equal to 0.0500 ft* to verify that the operator has sufficient time to initiate me sprays prior.to exceeding the containment design pressure of 55 psig. The limit of M of~ the design' value of 0.0500 ft2 ensures that the design basis for the steam bypass a @. sis L_ is met. l l LIMERICK - UNIT 1 B 3/4 6-3a Amendment No. 57, 68
CONTAINMENT SYSTEMS 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 through the passive structural components. Previous dryweli-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 i limit, combined with the negligible passive structural leakage area, ensures that the i drywell-to-suppression chamber bypass leakage limit is met for those outages for which the drywell-to-suppression chamber bypass test is not scheduled. 3 4.6.3 PRIMARY CONTAINMENT ISOLATION VALVES J 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. 3J4.6.4 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. l 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 hours 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 7k = 0.05 ft' for the vacuum relief system (assuming one valve fully open) will not be exceeded. LIMERICK - UNIT 1 B 3/4 6-4 Amendment No. #6,68 j
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p *to (q ' ?g E Ut'ITED STATES .;i' NUCLEAR REGULATORY COMMISSION y t. ,/ WASHINGTON, D C. 20555-0001 PHILADELPHIA ELECTRIC COMPANY DOCKET NO. 50-353 LIMERICK GENERATING STATION. UNIT 2 AMENDMENT TO FACIllTY OPERATING LICENSE i Amendment No. 31 License No. NPF-85 1. The Nuclear Regulatory Commission (the Commission) has found that: A. The application for amendment by Philadelphia Electric Company (the licensee) dated November 30, 1993, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (1) 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; 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.
1 ! 1 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. 31, 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. FOR THE NUCLEAR REGULATORY COMMISSION Wl$ Charles L. Miller, Director Project Directorate I-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: February 17, 1994 t 1 w
i i ATTACHMENT TO LICENSE AMENDMENT N0. 31 4 FACILITY OPERATING. LICENSE NO. NPF-85 1 DOCKET NO. 50-353 l 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 i provided to maintain document completeness.* i Remove Insert 3/4 6-13 3/4 6-13* 3/4 6-14 3/4 6-14 i B 3/4 6-3 B 3/4 6-3* i B 3/4 6-3a B 3/4 6-3a i B 3/4 6-4 8 3/4 6-4 i i l i t l l t i J l l I
1 CONTAINMENT SYSTEMS t LIMITING CONDITION FOR OPERATION (Continued) ACTION: (Continued) 3. With the suppression chamber average water temperature greater than 120*F, depressurize the reactor pressure-vessel to less i than 200 psig within 12 hours, c. With only one suppression chamber water level. indicator OPERABLE and/or with less than eight suppression pool water temperature indicators, one in each of the eight locations OPERABLE, restore the inoperable indicator (s) to OPERABLE status within 7 days or verify suppression chamber water level and/or temperature to be within the limits at least once per 12 hours. .d. With no suppression chamber water level indicators OPERABLE and/or with less than seven suppression pool water temperature indicators covering at least seven locations OPERABLE, restore at least one water level indicator and at least seven water temperature indicators to OPERABLE status within 48 hours or be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUT 00WN within the following 24 hours. e. With the drywell-to-suppression chamber bypass leakage in excess of the limit, restore the bypass leakage to within the limit prior to increasing reactor coolant temperature above 200*F. SURVEILLANCE REQUIREMENTS 4.6.2.1 The suppression chamber shall be demonstrated OPERABLE: a. By verifying the suppression chamber water volume to be within the e limits at least once per 24 hours. b. At least once per 24 hours by verifying the suppression chamber average water temperature to be less than or equal to 95*F, except: 1. At least once per 5 minutes during testing which adds heat to the suppression chamber, by verifying the suppression chamber average water temperature less than or equal to 105'F. 2. At least once per hour when suppression chamber average water temperature is greater than or equal to 95'F, by verifying: a) Suppression chamber average water temperature to be less j than or equal to 110'F, and b) THERMAL POWER to be less than or equal to 1% of RATED THERMAL POWER 12 hours after suppression chamber average water temperature has exceeded 95'F for more than 24 hours. 3. At least once per 30 minutes following a scram with suppression chamber average water temperature greater than or equal to 95*F, j by verifying suppression chamber average water temperature less than or equal to 120*F. LIHERICK - UNIT 2 3/4 6-13 I y
^ CONTAINMENT SYSTEMS 9 SURVEllLANCE RE0VIREMENTS (Continued) i c. By verifying at least two suppression chamber water level indicators 4 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 or,ce per 24 hours, 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 s 241 " 2. High water temperature: a) First setpoint 5 95'F i 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 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 l consecutive tests meet the specified limit, at which time the test schedule may be resumed. l 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 12x 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. LIMERICK - UNIT 2 3/4 6-14 Amendment No. 31
4 CONTAINMENT SYSTEMS BASES 3/4.6.2 DEPRESSURIZATION 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 suppression 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 1s below the design pressure of 55 psig. Maximum water volume of 3 134,600 ft rgsultsinadowncomersubmergenceof12'3"andtheminimumvolume of 122.120 ft results in a submergence approximately 2'3" less. The majority of the Bodega tests were run with a submerged length of 4 feat and with complete condensation. Thus, with respect to the downcomer submergence, this specification is adequate. Themaximumtemperatureatgheendoftheblowdowntestedduringthe Humboldt Bay and Bodega Bay tests was 170 F and this is conservatively taken tobethelimitforcompletecondensationoftheregctorcoolant,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. Underfullpoweroperatingconditions,blowdownthroughsafgty/reliefvalves assuminganinitialsuppressionchamberwatgrtemperatureof95Fresultsina bulk water temperature of approximately 136 F immediately following blowdown 0 which is below the 190 F bulk temperature limit used for complete condensation via T-quencher devices. At this temperature and atmospheric pressure, the available NPSH exceeds that required by both the RHR and core spray pumps, thus 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 containment overpressure for post-LOCA operations. LIMERICK - UNIT 2 B 3/4 6-3 Amendment No. 23 NOV 5 1932
L3/4.6.2 DEPRESSURIZATION SYSTEMS (Cont.) One of the surveillance requirements for the suppression pool cooling (SPC) mode of the RHR system 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 faiiures by indicating abnormal performance. The e 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 portion of the total flow still travels through the bypass, which can affect overall heat transfer. However, no heat transfer performance i 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 Specifications 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. 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. S)ecifications have been placed on the envelope of reactor operating conditions so t1at the reactor can be 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, the volume and temperature normally changes very slowly and monitoring these i 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 j that appropriate action can 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-j relief valve inadvertently opens or sticks open. As a minimum this action shall l 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 ca3 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 ccntainment pressure response to the postulated bypass leakage can be mitigated by manually actuating the suppressicn chamber sprays. An analysis was performed for a design bypass leakage area of A//k qual to 0.0500 ft' to verify that the operator has sufficient time to initiate the sprays prior. l to exceeding the containment design pressure of 55 psig. The limit of 10% of the @ ;qn I value of 0.0500 ft2 ensures that the design basis for the steam bypass analysis is v l LIMERICK - UNIT 2 8 3/4 6-3a Amendment No. 23, 31}}