ML20206T470
| ML20206T470 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 06/30/1986 |
| From: | TOLEDO EDISON CO. |
| To: | |
| Shared Package | |
| ML20206T455 | List: |
| References | |
| 1268, TAC-61937, NUDOCS 8607080143 | |
| Download: ML20206T470 (6) | |
Text
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I3/4.4. REACTOR COOLANT SYSTEM i
3/4.4.1. COOLANT LOOPS AND COOLANT CIRCULATION y STARTUP AND POWER OPERATION FOR NOMMTION ONW i
LIMITING CONDITION FOR OPERATION 3'. 4.1.1 Both reactor coolant loops and both reactor coolant pumps in each' loop shall be in operation.
APPLICABILITY: M00ES 1 and 2*.
ACTION:
With one reactor coolant pump not in operation, STARTUP and POWER OPERA-a.
TION may be initiated and may proceed provided THERMAL POWER is re-stricted to less than 79.7% of RATED THERMAL POWER and within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> l
the setpoints for the following trips have been reduced to the values specified in Specification 2.2.1 for operation with three reactor cool-ant pumps operating:
1.
High Flux 2.
Flux-aFlux-Fl ow SURVEILLANCE REOUIREMENTS 4.1.1.1 Th'e above required reactor coolant loops shall be verified to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
4.4.1.2 The reactor protective instrumentation channe'Is specified in the applicable ACTIDN statement above shall be verified to have had their trip setpoints changed to the values specified in Specification 2.2.1 for the ao-plicable nuster of reactor coolant pumps operating either:
Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after switching to a different pump comeination if the a.
switch is made while operating, or b.
Prior to reactor criticality if the switch is made while shutdown.
- See Special Test Except1.on 3.1G.3.
Amendment Mo 75. 33. 38, D, DAVIS-8 ESSE, UNIT 1 3/4 4.1 80 8607080143 860630 DR ADOCK 05000346 PDR 9 -
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' 3/4.6 RIACTOR COOZ, ANT SYSTEM 5HUTDOWN AND HOT STANDBY LIMITINC CONDITION FOR OPERATION
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3.4.1.2 a.
At least two of the coolant loops listed below shall be OPERAB M:
1.
Reactor' Coolant Loop 1 and its associated stasa" generator, 2.
Reactor Coolant 14op 2 and,its associated steam generator, 3.
Decay Heat Renoval Imop 1,*
4.
Decay Best Removal 14op 2.*
b.
At least one of the above coolant loops shall be in operation.**
c.
Not more than one decay heat removal pump any be operated with the sole suction path through DE-11 and DE-12 unless the control power has been removed from the DE-11 and DE-12 valve operator, or manual valves DE-21 and DE-23 are openad.
I d.
The provisions of Specificatsons 3.0.3 and 3.0.4 are not applicable.
AFFLICABILITY: HDDES 3, 4 and 3 ACTIGE:
a.
With less than the above required coolant loope CPIIABIZ, immediately initiate corrective actima to return the required coolant loops to OPERABI.E status as soon as possible, or be in COLD SEUTDOWN within 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.
l b.
With none of the above required coolmat loops in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant Sy' stem and iissediately initista corrective action to totura the required coolant loop to operation.
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- The normal or emergency power source may be inoperable in MODE 5.
This j
loop may not be selected in MODE 3 unless the primary side temperature and pressure are within the decay heat removal system's design conditions.
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- The decay heat removal pumps may be de-energized for up to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> provided (1) no operations are permitted that would cause dilution of the reactor coolant system boron concentration, and (2).. core outlet temperature is maintained at least y below saturation temper tura(
l 40 F m;uw&cd DAVIS-BESSE UNIT 1 3/4 4-2 Amendment No. 4, 8, 88, 88, 92
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3/4.4 REACTOR COOLANT SYSTEM FOR INFORMATION ONLY SURVEILIANCE REQUIRIMENTS I
The required decay heat removal loop (s) shall be determined h
4. 4.1.,2.1 OPERABLE per Specification 4.0'.5.
4.4.1.2.2 The required steam generator (s) shall be determined OPERABLE by verifying secondary side level to be greater than or equal to (a) 18 inches above the lower tube sheet once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> if an associated reactor coolant puny is operating, or, (b) 35 inches above the lower tube sheet once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> if no reactor coolant pumps are operating.
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4.4.1.2.3 at least one coolant loop shall be verified to be in l
operation and circulating reactor coolant at least once I
per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
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i DAVIS-BESSE UNIT 1 3/4 4-2a l
Ame.ndme_rt Ab.ee 4
s FOR INFORMATION ONLY v4.4 RExcToR coauNT SYSm SURVEILLANCE REQUIREMETTS l4.4.1.2.1 The required decay heat renoval loop (s) shall be determined oPEPJulLE per Specification 4.0.5.
4.4.1.2.2 The required steam generator (s) shall be determined OPERABLE by verifying secondary side level to be greater than or equal to (a) 18 inches above the lower tube sheet once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> if an associated reactor coolant pump is operating, or. (b) 35 inches above the lower tube sheet once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> if no reactor coolant pumps are operating.
4. 4.1.2. 3 At least one coolant loop shall be verified to be in operation and circulating reactor coolant at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
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DAVIS-BESSE UNIT 1 3/4 4-2a O
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,a 3/4.4 REACTOR COOLANT SYSTEM FOR INFORMATION ONLY BASES 3/4.4.1 REACTOR COOLANT LOOPS The plant is designed to operate with both reactor coolant loops in operation, and maintain DNBR above 1.30 during all normal operations and anticipated transients.
With one reactor coolant pump not in operation.in one loop, THERMAL POWER is restricted by the Nuclear Overpower Based on RCS Flow and AXIAL POWER IMBALANCE, ensuring that the DNBR will be maintained above 1.30 at the maximum possicle THERMAL POWER for the number of reactor coolant pumps in operation or the local quality at the point of minimum DNBR equal to 22%, whichever is more restrictive.
In MODE 3 when RCS pressure or temperature is higher than the decay heat removal system's design condition (i.e. 330 psig and 350*F), a single reactor coolant loop provides sufficient heat removal capability.
The remainder of MODE 3 as well as in MODES 4 and 5 either a single reactor coolant loop or a DHR loop will be sufficient for decay heat removal; but single failure considerations require that at least two loops be OPERABLE.
Thus, if the reactor coolant loops are not OPERABLE, this specification
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requires two DHR loops to be OPERABLE.
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Natural circulation flow or the operation of one DHR pump provides adequate i
flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the Reactor Coolant System.
The reactivity change rate associated with boron reduction will, therefore, be within the capacity of operator recognition and control, s
3/4.4.2 and 3/4.4.3 SAFETY VALVES The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2750 psig.
Each safety valve is designed to relieve 336,000 lbs per hour of saturated steam at the valve's setpoint.
The relief capacity of a single safety valve is adequate to relieve any i
overpressure condition which could occur during shutdown.
In the event that no safety valves are OPERABLE, an operating DHR loop, connected to the RCS, provides overpressure relief capability and will prevent RCS overpressurization.
During operation, all pressurizer code safety valves must be OPERABLE to prevent the RCS from being pressurized above its safety limit of 2750 psig.
The combined relief capacity of all of these valves is greater than the maximum surge rate resulting from any transient.
The relief capacity of the decay heat removal system relief valve is adequate to relieve any overpressure condition which could occur during shutdown.
In the event that this relief valve is not OPERABLE, reactor coolant system pressure, pressurizer level and make up water inventory is limited and the capability of the high pressure injection system to inject water into the reactor coolant s system pressure ystem-is disabled to ensure operation within reactor coolant temperature limits.
Demonstration of the safety valves' lift settings will occur only during shutdown and will be performed in accordance with the provisions of Section XI of the ASME Boiler and Pressure Code.
gjDAVIS-BESSEUNIT1 8 3/4 4-1 Amendment No. Jr, Jr, F, 92
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FOR INFORMATION ONLY REACTCH COOLMT 575 TEM 3ASES The pressurizer code sa fety valves must be set such that the peak Reactor Coolant System pressure does not exceed 110% of design system pressure (2500 psig) or, 2750 psig.
The control rod group withdrawal accident will result in the most limiting high pressure in the RCS.
The analysis assumes RPS high pressure trip at 2200 psig and the code sa fety valves open at 2500 psig.
The tolerance on the Rps instrument accuracy is 30 psi and, it is +3% for the code safety valva settinrs.
The pressurizer electromatic relief valve was assumed not to open for this transient.
The resulting system peak pressure was calculated to be 2716 psig.
There fore, the code safety valve setpoint is conserva~-
tively set at < 2525 ps.ig which is the maximum pressure of 2500 psig
+1% for tolerance.
The pressurizer electromatic relief valve should be set such that i't will open be fore the code safety valves are opened.
However, it should not open on any anticipated transients.
Loss of Feedwater (LOFW) was identified as the limiting anticipated transient for RCS pressure.- The analysis assumes RPS high pressure trip at 2300 psig; with 30 psi for instrument errors, the resulting peak RCS pressure is calculated to te 2380 psi g.
This includes a 50 psig pressure overshoot on a LOFW transient.
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i DAVIS-BESSE, UNIT 1 B 3/4 4/la Amendment No 22, 60 U