ML19211D376
| ML19211D376 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 10/31/1979 |
| From: | SACRAMENTO MUNICIPAL UTILITY DISTRICT |
| To: | |
| Shared Package | |
| ML19211D374 | List: |
| References | |
| NUDOCS 8001180320 | |
| Download: ML19211D376 (4) | |
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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operations 3.1.2 PRESSURIZATION, HEATUP, AND C00LDOWN LIMITATIONS Specification 3.1.2.1 Inservice Leak and Hydrostatic Tests:
Pressure temperature limits for the first five EFP years of inservice leak and hydrostatic tests are given in Figure 3.1.2-3.
Heatup and cooldown rates shall be restricted according to the rates specified in Figure 3.1.2-3.
3.1.2.2 Heatup Cooldown:
For the first five EFP years of power opeations, the reactor coolant pressure and the system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with Figure 3.1.2-1 and Figure 3.1.2-2 respectively.
Heatup and cooldown rates shall not exceed the rates stated on the associated figure.
3.1.2.3 The secondary side of the steam generator shall not be pressurized above 200 psig if the temperature of the steam generator shell is below 130 F.
3.1.2.4 The pressurizer heatup and cooldown rates shall not exceed 100 F in any 1-hour period.
3.1.2.5 The spray shall not be used if the temperature difference between the pressurizer and spray fluid is greater than 4100F.
3.1.2.6 Prior to exceeding five effective full power ye'rs of operation, Figures 3.1.2-1,
-2, and -3 shall be updated for the next service period in accordance with 10 CFR 50, Appendix G, Section V.E.
The highest predicted adjusted reference temperature of all the beltline materials shall be used to determine the adjusted reference temperature at the end of the service period.
The basis for this prediction shall be submitted for NRC staff review in accordance with Specification 3.1.2.7.
3.1.2.7 The updated proposed technical specifications referred to in 3.1.2.6 shall be submitted for NRC review at least 90 days prior to the end of the service period. Appropriate additional NRC review time shall be allowed for proposed technical specifications submitted in accordance with 10 CFR 30, Appendix G, Section V.C.
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3.1.2.8 Emergency / Faulted Operation:
In the emergency / faulted condition when there is no forced or natural circulation in the reactor coolant system and there is high pressure 64 injection and/or makeup addition, the Reactor Coolant System temperatur Rev.1 and pressure shall be limited in accordance with the limit line shown on Figure 3.1.2-4.
Under the above energency/ faulted conditicns, Figure 3.1.2-2 will not apply.
3-3 Amendment No. 64 8001180
RANCHO SECO UNIT I TECHNICAL SPECIFICATIONS Limiting Conditions for Operation The maximum allowable pressure is taken to be the lowest pressure of the three calculated pressures.
The pressure limit is adjusted for the pressure differential between the point of system pressure measurement and the limiting component for all reactor coolant pump combinations.
The limit curves were prepared based upon the most 1imiting adjusted reference temperature of a1i the beIt1ine region materials at the end of the fifth effective full power year.
The actual shift in RT of the beltline region material will be established periodicallyduringopbtionbyremovingandevaluating, in accordance with Appendix H to 10 CFR 50, reactor vessel material irradiation surveillance specimens installed near the inside wall of the reactor vessel in the core area.
Because the neutron energy spectra at the specimen location and at the vessel inner wal1 location are essentially the same, the measured transition shift for a sample can be applied with confidence to the adjacent section of the reactor vessel. The limit curves must be recalculated when the ART determined from the surveillance capsule is different from the calculated a f r the equiva-DT lent capsule radiation exposure.
The unirradiated impact properties of the beltline region materials, required by Appendices G and H to 10 CFR 50, were determined for those materials for which sufficient amounts of material were available.
The adjusted reference temperatures are calculated by adding the radiation-induced ART nd the unirradiated RT ND NDT' The predicted ART recalculatedusingtherespec[iveneutron fluence and NDT copper and phosphorus contents in accordance with Reg. Guide 1.99 The assumed RT of the closure head region is 60 F and the outlet nozzle steel forgings is60NI The limitations imposed on pressurizer heatup and cooldown and spray water temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASP.E code requirements.
The limitations during energency/ faulted operation when all reactor coolant flow and all feedwater flow is lost to the OTSG's are established to take into consideration that HPl gives false cold leg temperatures.
This transi-64 ent is controlied by Figure 3.1.2-4 and the vessel beltline temperature is Rev.1 calculated using incore thermocouples and subtracting 150 F for conserva-tism. When the coolant flow or feedwater flow is re-established, a four hour transition period will be allowed to progress f rom Figure 3.1.2-4 to Figure 3.1.2-2.
1765 308 3-0 Amendment No. 64 Rev. I
RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Limiting Conditions for Operation 3.4.2.2 Both auxiliary feed pumps are operable.
One auxiliary feed pump may be out of service for maintenance for a period of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
3.4.2.3 An auxiliary feedwater flow path is operable, otherwise the reactor 64 shall not remain critical.
Rev.I Bases.
The feedwater system and the turbine bypass system are normally used for decay heat removal and cooldown above 280 F.
Feedwater makeup is supplied by operation of a condensate pump and main feedwater pump.
In the event of complete loss of electrical power, feedwater is supplied by a turbine driven auxiliary feedwater pump which takes suction from the condensate storage tank.
Steam relief would be through the system's atmospheric relief valves, if neither main feed pump is available, feedwater can be supplied to the steam generators by an auxiliary feedwater pump and steam relief would be through the turbine bypass system to the condenser.
In order to heat the reactor coolant system above 280 F the maximum steam removal capabil i ty required is 4-1/2 percent of rated power. This is the maximum decay heat rate at 30 seconds after a reactor trip. The requirement for two steam system safety valves per steam generator provides a steam relief capability of over 10 percent per steam generator (1,341,938 lb/h).
In addi. tion, two turbine bypass valves to the condenser or two atmospheric dump valves will provide the necessary capacity.
The 250,000 gallons of water in the condensate storage tank is the amount needed for cooling wcter to the steam genenators {7J a p-iod in excess of one day follow-Ing a complete loss of all unit ac power.
The mi in 13,329,163 lb/hr. 2)um relief capacity of seventeen steam system safety valves is This is sufficient capacity to pr(o ect the steam system under the 3
design overpower condition of 112 percent.
REFERENCES (1) FSAR paragraph 14.1.2.8.4 1765 309 (2)
FSAR paragraph 10.3.4 (3)
FSAR Appendix 3A, Answer to Question 3A.5 3-24 Amendment No. 64 Rev.1
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RANCHO SECO UNIT 1 TECHNICAL SPECIFICATIONS Surveillance Standards 4.8 AUXILIARY FEEDWATER PUMP PERIODIC TESTING Applicability Applies to the periodic testing of the turbine and motor driven auxiliary feed-water pumps.
Obj ec t iv,e To verify that the auxiliary feedwater pump and associated valves are operable.
Specification 64 4.8.1 At least every 92 days on a staggered test basis at a time when the average Rev.i reactor coolant system temperature is 3305 F, the turbine / motor driven and
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motor driven auxiliary feedwater pumps shall be operated on recirculation to the condenser to verify proper operation.
The 92 day test frequency requirement shall be brought curzent w!*n*-0 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> af ter the average reactor coolant system temperature is 1305 F.
Acceptable performance will be indicatad if the pump starts and operates for fifteen minutes at the design f l o,, af 780 gpm.
This flow will be verified using tank level decrease and pump dif ferential aressure.
4.8.2 All valves, including those that are locked, scaled, or otherwis 64 secured in position, are to be inspected monthly to veri'?y they ar Rev.1 in the proper position.
Bases The quarterly test frequency will be sufficient to verify that.tu: turbine /metor driven and motor driven auxiliary feedwater pumps are operable.
Verification of
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correct operation will be made both from the control room instrmmentation and direct visual observation of the pumps.
REFERENCE 1765 310 4-39 Amendment No. 64 Rev. I