ML20065E717

Proposed Tech Specs Re Starting Reactor Coolant Pump

ML20065E717
Person / Time
Site:	Vogtle
Issue date:	09/20/1990
From:	GEORGIA POWER CO.
To:
Shared Package
ML20065E716	List: ELV-01901, Application for Amends to Licenses NPF-68 & NPF-81,revising Tech Spec (TS) 3.4.1.3 Re Starting Reactor Coolant Pump & Providing Consistency Between Ts,Rcs Analysis Assumptions Demonstrating Compliance W/App G Limits & RHR Analysis ML20065E717
References
NUDOCS 9010020274
Download: ML20065E717 (7)

Text

c , ,

4' -

! 3.

{

l .

l ENCLOSURE 3 l V0GTLE ELECTRIC GENERATING PLANT ,

REVISION TO TECHNICAL SPECIFICATION 3.4.1.3 INSTRUCTIONS FOR INCORPORATION ,

The proposed amendment to Section 3.4.1.3 of the Vogtle Electric Generating Plant Technical Specifications would be incorporated as follows:

Remove Pace Insert Paae 3/4 4-3 and 3/4 4-4* 3/4 4-3 and 3/4 4-4*

B 3/4 4-1 and B 3/4 4-2* B 3/4 4-1 and B 3/4 4-2* i 8 3/4 4-15* and B 3/4 4-16 B 3/4 4-15* and B 3/t. 4-16 i

I l

h

• 0verleaf page containing no change 9010020274 900920 I

' {DR ADOCN 0500o424 Pl>C

e REACTOR COOLANT SYSTEM HOT SHUTDOWN LIMITING CONDITION FOR OPERATION 3.4.1.3 At least two of the loops / trains listed below shall be CPERABLE and '

at least one of these loops / trains shall be in operation:*

a. Reactor Coolant. Loop 1 and its associated steam generator and reactor  ;

coolant pump,** l Reactor Coolant Loop 2 and its associated steam ganerator and reactor  ;

b.

coolant pump,**

c. Reactor Coolant Loop 3 and its associated steam generator and reactor coolant pump,**

d. Reactor Coolant Loop 4 and its associated steam generator and reactor [

coolant pump,**

e. RHR train A, and ,

f. RHR train B. .

6 APPLICABILITY: MODE 4. ,

I ACTION:

a. With less than the above required loops / trains OPERABLE, immediately initiate corrective action to return the required loops / trains to  !

OPERABLE status as soon as possible; if the remaining OPCnABLE loop / train is an RHR train, be in COLD. SHUTDOWN within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. With no loop / train in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required loop / train to operation.

• All reactor coolant pumps and RHR pumps may be deenergized for up to I hour 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 10*F below saturation temperature. +

• • A reactor coolant pump shall not be started unless the secondary water temperature of each steam generator is less than,50'F above each of the Reactor Coolant System cold leg temperatures. WJ/ ee M ce,. lad /swy.'

M9%s,0i ydal n alnad b 75 f a/ a 6c5 i O

~ a2Lt g yso f ar / r Huir AuwAj -le 50

• I al a f c : L p k y gop V0GTLE UNITS - 1 & 2 3/4 4-3

3/4.4 REACTOR COOLANT SYSTEM BASES  ;

3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION . ,

The plant is designed to operate with all reactor coolant loops in operation and maintain DNBR above 1.30 during all normal operations and antici-l pated transients. In MODES 1 and 2 with one reactor coolant loop not in

, operation this specification requires that the plant be in at least HOT STANDBY '

l within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. ,

l In MODE 3, two reactor coolant loops provide sufficient heat removal l capability for removing core decay heat even in the event of a bank withdrawal l accident; however, a single reactor coolant loop provides suffia;ient heat i

removal capacity if a bank withdrawal accident can be preveKu, i.e. , by opening the Reactor Trip System breakers. '

In MODE 4, and in MODE 5 with reactor coolant loops filled, a single i reactor coolant loop or RHR train provides sufficient heat removal capability for removing decay heat; but single failure considerations require that at I least two trains / loops (either RHR or RCS) be OPERABLE.

i In MODE 5 with reactor coolant loops not filled, a single RHR train provides sufficient heat removal capability for removing decay heat; but single failure considerations, and the unavailability of the steam generators as a heat removing component, require that at least two RhR trains be OPERABLE. The

, locking closed of the required valves, except valves 1208-U4-176 and 1208-U4-177 for short periods of time to maintain chemistry control, in Mode 5 (with the loops not filled) precludes the possibility of uncontrolled boron dilution of the filled portion of the Reactor Coolant System. These actions prevent flow to the RCS of unborated water in excess of that analyzed. These limitations are consistent with the initial conditions assumed for the boron dilution accident in the safety analysis.

The operation of one reactor coolant pump (RCP) or one RHR pump provides adequate 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 capability of operator recognition and control, t The restrictions on starting an RCP with one or more RCS cold legs less than or equal to 350'F are provided to prevent RCS pressure transients, caused i by energy ad.ditions from the Secondary Coolant System, which could exceed the limits of Appendix G to 10 CFR Part 50. The RCS will be protected against overpressure transients and will not exceed the limits of Appendix G by restricting starting of the RCPs to when the secondary water temperature of each steam gengrator is less than 50'F above each of the RCS cold leg l temperatures, k AMO# // ff d y a. g q g d Ac p

) N 0 m y> q s a W + -/ $ / p.t cs m U f N DA)4A4 i144U W4 %fAcl -bo .jQ ff A Wtl l s puuutA N eA -L4LLb l wltm A HA I w. fur.

M M.e d .l@ h C 5 0 + t- y < b H < u l p u b & ' .

V0GTLE UNITS - 1 & 2 B 3/4 4-1 Amendment No. 28 (Unit 1)

Amendment No. 9 (Unit 2)

~

. ad & lh.pk R 4 % @uin 24 % ya .

I u,ab co4.J /4 L .gnup oZ6 3.4,/,3. d w n A L REACTOR C00LANi ,(STEM *"*d fb/ IA "A @ d'-M

• 4 4 ## O  %

.G h wluth aun Q ,.Au. a wd4 Q l

BASES J, i PRESSURE / TEMPERATURE LIMITS (Continued) j Although the pressurizer operates in temperature ranges above those for n which there is reason for concern of nonductile failure, operating limits are > i provided to assure compatibility of operation with the fatigue analysis l performed in accordance with the ASME Code requirements.  ;

COLD OVERPRESSURE PROTECTION SYSTEMS t The OPERABILITY of two PORVs, two RHR suction relief valves 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. Either PORV or either RHR suction relief valve has i adequate relieving capability to protect the RCS from overpressurization when

• i the transient is limited to, either: (1) the start of an idle RCP with the ,

secondary water temperature of the steam generator less than or equal to 50'F f  :

above the RCS cold leg temperatures, or (2) the start of all three charging pumps and subsequent injection into a water-solid RCS. ,  !

The Maximum Allowed PORV Setpoint for the Cold Overpressure Protection System (COPS) is der (ved by analysis which mod 91s the performance of the COPS assuming various mass input and heat input transients. Operation with a PORV Setpoint less than or equal to the maximum Setpoint ensures that the nominal 16 EFPY Appendix G reactor vessel NDT limits criteria will not be violated with consideration for a maximum pressure overshoot beyond the PORV setpoint which j can occur as a result of time delays in signal processing and valve opening, instrument uncertainties, and single failure. To ensure that mass and heat  :

input transients more severe than those assumed cannot occur, Technical Spec- '

ifications require lockout of all safety injection pumps while in MODES 4, 5, f '

and 6 with the reactor vessel head installed and disallow start of an RCP if -

secondary temperature is more than 50'F above primary temperature.- 4 W 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, Appendix H, and in accordance with the schedule in Table 16.3-3 of the VEGA FSAR.

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 1 life of the plant. These orograms 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 written relief has been granted by the Commission pursuant to 10 CFR 50.55a(g)(6)(1).' '

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

E i

l' REACTOR COOLANT SYSTEM HOT S;iUTDOWN LIMITING CONDITION FOR OPERATION 3.4.1.3 At least two of the loops / trains listed below shall be OPERABLE and at least one of these loops / trains shall be in operation:*

t i

a. Reactor Coolant Loop 1 and its associated steam generator and reactor coolant pump.**

b. Reactor Coolant Loop 2 and its associated steam generator and reactor coolant pump,**

c. Reactor Coolant Loop 3 and its associated steam generator and reactor coolant pump,"

d. Reactor Coolant Loop 4 and its associated steam generator and reactor coolant pump,**

e. RHR train A, and

f. RHR train B.

APPLICABILITY: HODE 4.

ACTION:

a. With less than the above required loops / trains OPERABLE, immediately initiate corrective action to return the required loops / trains to OPERABLE status as soon as possible; if the remaining OPERABLE loop / train is an RHR train, be in COLD SHUTDOWN within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

b. With no loop / train in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate corrective action to return the required loop / train to operation.

• All reactor coolant pumps and RHR pumps may be deenergized for up to-I hour 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 10*F below saturation temperature.

• • A reactor coolaat pump shall not be started unless the secondary water temperature of each steam generator is less than 50*F above each of the Reactor Coolant System cold leg temperatures. With no reactor coolant pump running, this value is reduced to 25'F at an RCS temperature of 350*F and varies linearly to 50*F at an RCS temperature of 200*F.

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

i; j ' ..

[

7 -

liEACTOR COOLANT SYSTEM l 0 ys ,

BASES -l y

PESSURE/ TEMPERATURE LIMITS (Continued) ,

Although the pressurizer operates in temperature ranges above those for ,.

which there is reason for concern of nonductile failure,. operating limits are provided '.o assure compatibility of operation with the fatigue analysis .-'

performet in accordance with the ASME Code requirements.

, COLD OVlitPRESSURE PROTECTION SYSTEMS The OPERABILITY of two PORVs, two RHR su. tion relief valves or an RCS vent i capable of relieving at least 670 gpm water flow at 470 psig ensures that the  ;

RCS will be protected f rom 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. Either PORV or either RHR suction relief valve has -

adequate relieving capability to 9rotect the RCS from overpressurization when i the transient is limited to either: (1) the start of an idle RCP with the >

secondary water temperature of the steam generator less than or equal to 50*F '

above the RCS cold leg temperatures, or (2) the start of all three charging pumps and subsequent injection into a water-solid 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 input and heat input transients. Operation with a PORV Setpoint less than or equal to the maximum Setpoint ensures that the nominal  ;

16 EFPY Appendix G reactor vessel NOT limits criteria will not be violated with- - !

consideration for a maximum pressure overshoct beyond the PORV setpoint which k can occur as a result of time delays in signal processing and valve = opening, instrument uncertainties, and single failure. Tc ensure that mass and heat input transients more severe than those assumed ,annot occur. Technical Spec- .

ifications require lockout of all safety-injW on pumps while in MODES 4, 5, and 6 with the reactor vessel head installed su disallow start of an RCP if secondary temperature-is more than 50*F abovt primary temperature. Additional temperature limitations are placed on the starting of a reactor coolant pump 7 These limitations assure that the RHR system remains in Specification.3.4.1.3.

within its ASME design limits when the RHR relief valves are used to prevent RCS overpressurization.

7 The Maximum Allowed PORV Setpoint for the COPS will be updated based on the' l results of examinations of reactor vessel material irradiation surveillance specimens performed as required by 10 CFR Part 50, Appendix H, and in accordance with the schedule in Table 16.3-3 of the VEGP FSAR.

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

'l;fe 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 written relief has been granted by the Commission purusar,t to 10 CFR 50.55a(g)(6)(i).

V0GTLE UNITS - 1 & 2 83/44-16

n. 4 - ,4 ,

.- 3'/4 . 4 REACTOR COOLANT SYSTEM. l

. BASES 3 /4 .' 4 .1 - REACTOR COOLANT LOOPS AND COOLANT CIRCULATION

.The plant is designed to operate with all reactor coolant loops in operation and maintain DNBR above 1.30 during all normal operations and antici-pated transients, in MODES 1 and 2 with one reactor coolant loop not in operation ~ this specification requires that the plant be in at least HOT STANDBY ,

within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

In MODE 3, two reactor coolant loops provide sufficient heat removal capability for removing core decay heat even in the event of a bank withdrawal  :

accident; however, a single reactor coolant loop provides suf ficient heat removal capacity if a bank withdrawal accident can be prevented, i.e., by '

opening the Reactor Trip System breakers, in H0DE 4, and in MODE 5 with reactor coolant loops filled, a single reactor coolant loop or RHR train provides sufficient heat removal capability '

for removing decay heat; but single failure considerations require that at least two trains / loops (either RHR or RCS) be OPERABLE.

i In MODE 5 with reactor coolant loops not filled, a single RHR train provides sufficient heat removal capability for removing decay heat; but single failure considerations, and the unavailability of the steam generators as a <

heat removing component,-require that at least two RHR trains be OPERABLE. The locking closed of the required valves, except valves.1208-U4-176 and "

1208-U4-177 for short periods of time to maintain chemistry control, in MODE 5 (with the loops not filled) precludes the possibility of uncontrolled boron dilution of the filled portion of the Reactor Coolant System. These actions i prevent flow to the.RCS of unborated water in excess of that analyzed. These limitations are consistent with the initial conditions assumed for the boron

! dilution accident in the safety analysis.

The operation of one reactor coolant pump (RCP) or one RHR pump provides adequate 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 capability of operator recognition and control.

The restrictions on starting on RCP with one or more RCS cold legs less than or equal to 350*F are provided to prevent RCS pressure transients, caused L by energy additions from the Secondary Coolant System, which could exceed the l limits of Appendix G to 10 CFR Part 50.' The RCS will be protected against l overpressure transients and will not exceed the limits of Appendix G by restricting starting of the RCPs to when the secondary water temperature of each steam generator is less than 50*F above each of the RCS cold leg  !

i temperatures. In MODE 4 the starting of an RCP, when no other RCP is operating, i l: is restricted to a range of temperatures that are consistent with analysis -j b assumptions used to demonstrate that the RHR design pressure is not exceeded >

when RHR relief valves are used for RCS overpressure protection. .

I  !

i V0GTLE. UNITS - 1 & 2 B 3/4 4-1 i

L i