ML20086D116

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Proposed Tech Specs 3/4.4.4, Relief Valves
ML20086D116
Person / Time
Site: Vogtle  Southern Nuclear icon.png
Issue date: 11/18/1991
From:
GEORGIA POWER CO.
To:
Shared Package
ML20086D111 List:
References
NUDOCS 9111250260
Download: ML20086D116 (4)
v • d • e


Text

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i REACTOR COOLANT SYSTEM 3/4.4.4 RELIEF VALVEE LIMITING CONDITION FOR OPERATION s

3.4.4 Both power-operated relief valves (PORVs) and their associated block j valves shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3.

ACTION:

a. With one or both PORV(s) inoperable, because of excessive seat I leakage, within I hour either restore the PORV(s) to OPERABLE status or close the associated block valve (s) with power maintnined to the block valve (s); otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
b. With one or both PORV(s) inoperable due to causes other than exces- I sive seat leakage, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> either restore the PORV(s) to OPERABLE status or close the associated block valve (s) and remove power from the block valve, and
1. With only one PORY OPERABLE, restore at least a total of two PORVs to OPERABLE status within the following 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, or
2. With both PORVs inoperable, restore at least one PORV to OPERABLE I status within I hour or be in HOT STANDBY within the next "

6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

c. With one or both block valves inoperable, within I hour restore the block valve (s) to OPERABLE status or place its associated PORV(s) in manual control. Restore at least one block valve to OPERABLE status within the next hour if both block valves are inoperable; restore any remaining inoperable block valve to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; otherwise, be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />,
d. The provisions of Specification 3.0.4 are not applicable.

SURVEILLANCE RE0VIREMENTS 4.4.4.1 Each PORV shall be demonstrated OPERABLE at least once per 18 months by:

a. Operating the valve through one complete cycle of full travel, and
b. Performing a CHANNEL CALIBRATION.

l V0GTLE UNITS - 1 & 2 3/4 4-10 l g,6k^$5I2Elh4

M, -,%.~

, REACTOR COOLANT SYST@

l I

BASES 3/4.4.4 RELIEF VALVES I The power-operated relief valves (PORVs) and steam bubble function to relieve RCS pressure during all design transients up to and including the design step load decrease with steam dump. Operation of the PORVs minimizes the undesirable opening of the spring-loaded pressurizer Code safety valves.

Each PORY has a remotely operated block valve to provide a positive shutoff capability should a relief valve become inoperable. The PORVs are equipped with automatic actuation circuitry and manual control capability. Because no credit for automatic PORV operation is taken in the USAR analyses for MODE 1, 2 and 3 transients, the PORVs are considered OPERABLE in either the manual or automatic MODE. The automatic MODE is the preferred configuration, as this provides pressure relieving capability without reliance on operator action.

3/4.4.5 STEAM GENERATORS The Surveillance Requirements for inspection of the steam generator tuber ensure that the structural integrity of this portion of the RCS will be main-tained. The program for inservice inspection of steam generator tubes is based on a modification of Regulate.y Guide 1.83, Revision 1. Inservice inspection of steam generator tubing is essential in order to maintain surveil-lance of the conditions of the tubes in the event that there is evidence of mechanical damage or progressive degradation due to design, manufacturing errors, or inservice conditions that lead to corrosion. Inservice inspection of steam generator tubing also provides a means of characterizing the nature and cause of any tube degradation so that corrective measures can be taken.

The plant is expected to be operated in a manner such that the secondary coolant will be maintained within those chemistry limits found to result in negligible corrosion of the steam generator tubes. If the secondary coolant chemistry is not maintained within these limits, localized corrosion may likely result in stress corrosion cracking. The extent of cracking during plant operation would be limited by the limitation of steam generator tube leakage between the Reactor Coolant System and the Secondary Coolant System (primary-to-secondary leakage - 500 gallons per day per steam generator).

Cracks having a primary-to-secondary leakage less than this limit during operation will have an adequate margin of safety to withstand the loads imposed during normal operation and by postulated accidents. Operating plants have demonstrated that primary-to-secondary leakage of 500 gallons per day per steam generator can readily be detected by radiation monitors of steam generator blowdown. Leakage in excess of this limit will require plant shutdown and an unscheduled inspection, during which the leaking tubes will be located and plugged.

Wastage-type defects are unlikely with proper chemistry treatment of the secondary coolant. However, even if a defect should develop in service, it will be found during scheduled inservice steam generator tube examinations.

Plugging will be required for all tubes with imperfections exceeding the plugging limit of 40% of the tube nominal wall thickness. Steam generator tube inspections of operating plants have demonstrated the capability to reliably detect degradation that has penetrated 20% of the original tube wall thickness.

V0GTLE UNITS - 1 & 2 8 3/4 4-3 l

REACTOR COOLANT SYSTEB COLD OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.9.3 At least one of the following groups of Cold Overpressure Protection Devices shall be OPERABLE when the reactor coolant system (RCS) is not depressurized thrcugh a vent path capable of relieving at least 670 gpm water flow at 470 psig.

a. Two power-operated relief valves (PORVs) with lift settings which vary with RCS temperature and which do not exceed the limits established in Figure 3.4-4a (Unit 1), Figure 3.4-4b (Unit 2), or
b. Two residual heat removal (RHR) suction relief valves (SRVs) each with a setpoint of 450 psig 3%, or
c. One RHR SRV and one PORV with setpoints as described above.

APPLICABILITY: MODES 4, 5; and MODE 6 with the reactor vessel head on.

ACTION:

a. In MODE 4, with only one PORV or one RHR SRV OPERABLE, restore one additional valve to OPERABLE status within the next 7 days or depressurize and vent the RCS, as specified in 3.4.9.3 above, within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />,
b. In MODES 5 and 6, with only one PORV or one RHR SRV OPERABLE, restore one additional valve to OPERABLE status within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or depressurize and vent the RCS, as specified in 3.4.9.3.

above, within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

c. In MODES 4, 5 and 6, with none of the PORVs or RHR SRVs OPERABLE, depressurize and vent the RCS, as specified in 3.4.9.3 above, l within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
d. In the event that the PORVs and/or RHR SRVs or the RCS vent (s) are used to mitigate an RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.8.2 within 30 days. The report shall describe the circumstances initiating the transient; the effect of the PORVs, the RHR SRVs, or RCS vent (s) on the transient; and any corrective action necessary to prevent recurrence,
e. The provisions of Specification 3.0.4 are not applicable.

V0GTLE UNITS - 1 & 2 3/4 4-34

BEACTOR COOLANT SYSTEM I

- BASES PRESSURE / TEMPERATURE LIMITS (Continued)

Although the pressurizer operates in temperature ranges above those for which there is reason for concern cf nonductile failure operating limits are provided atibility of operation with the fatigue analysis performed in )

toassurecomphtheASMECoderequirements.

accordance wi.

COLD OVERPf M RE PROTECTION SYSTEMS The OPERABillTY of two PORVs two RHR suction relief valves, a PORV and RHR SRV, or an RCS vent capable of relieving at least 670 gpm water flow at 470 psig ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 350 F. The PORVs have adequate relieving capability to p(rotect theofRCS

1) the start from an idle overpressurization RCP with the secondary when the transient water temperature of is thelimited steam to either:

2 generator the start of less all than threeorcharging equal to 50 F above pumps andthe RCS cold leg subsequent temperatures, injection or (lid into a water-so )

RCS. The RHR SRVs have adequate relieving capability to protect the RCS from (1 the start of an overpressurizaton when the transient is limited to either: idle RCP with the secondary to generator less than or equal to 25 F at an RCS temperature of 350 F and varies linearly to 50 F pumps at an RCS the start of all three charging and temperature of 200 Finto subsequent injection or less, or (2)d RCS, a water-soli A combination of a PORV and a RHR SRV with the above notes also provides overpressure protection for the RCS.

The Maximum Allowed PORV Setpoint for the Cold Overpressure Protection System (COPS) is derived by analysis which models the performance of the COPS assuming various mass irout and heat input transients. Operation with a PORV Setpoint less than or equal to the maximum Setpoint ensures that the nominal 13 EFPY for Unit I and 16 EFPY for Unit 2 Appendix G reactor vessel NDT limits l criteria will not be violated with consideration for a maximum pressure overshoot beyond the PORV setpoint which can occur as a result of time delays in signal and single failure. To processing ensure that and mass valve andopening, instrument heat input transientsuncertainties, more severehant those assumed cannot occur Technical Specifications require lockout of all safety injection pumps while,in MODES 4 5, and 6 with the reactor vessel head installed and disallow startofanRCPIfsecondarytem[eratureismorethan50'Faboveprimary temperature. Additional tempera ure limitations are placed on the starting of a Reactor Coolant Pump in Specification 3.4.1.3. These limitations assure that the RHR system remains within its ASME design limits when the RHR relief valves are used to prevent RCS overpressurization The Maximum Allowed PORV Setpoint for the COPS will be updated based on the results of examinations of reactor vessel material irradiation surveillance specimens performed as required by 10 CFR Part 50 with the schedule in Table 16.3-3 of the VEGP FSAk, Appendix H, and in accordance 3/4.4.10 STRUCTURAL INTEGRITY The inservice inspection and testing programs for ASME Code Class 1, 2, and 3 components ensure that the structural integrity and operational readiness of these components will be maintained at an acceptable level throughout the life of the plant. These programs are in accordance with Section XI of the ASME Boiler and Pressure Vessel Code and applicable Addenda as required by 10 CFR 50.55a(g) except where specific writter. relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(i).

V0GTLE UNITS - 1 & 2 B 3/4 4-16

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