ML20006F686

Proposed Tech Specs,Revising Bases Section 3/4.7.1.2 to Clarify Relationship Between Auxiliary Air Supply & Operability of Train of Auxiliary Feedwater

ML20006F686
Person / Time
Site:	Sequoyah
Issue date:	02/22/1990
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML20006F684	List: ML20006F683 ML20006F686
References
NUDOCS 9002280293
Download: ML20006F686 (8)

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PLANT SYSTEMS . -

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V = maximum' number of inoperable safety valves per steam line

.U = (

maximum number of inoperable safety valves per operating <

steam line.

109 = Power Range Neutron Flux-High Trip Setpoint for 4 locp operation.  ;

76? = Maximum percent of RATED THERMAL POWER permissible by P-8 Setpoint for 3 loop operation.

X. = Total relieving capacity of all safety valves per steam line in 1bs/ hour, 4.75 x 106 lbs/ hour at 1170 psig, i Y= Maximum relieving capacity of any one safety valve in -

, lbs/ hour, 950,000 lbs/ hour at 1170 psig. 4 . c

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3/4. 7.'1. 2 AUX 1LIARY FEEDWATER SYSTEM v: ~

The OPERABILITY of the auxiliary feedwater system ensures that the Reactor L

Coolant System can be cooled down to less than 350'F from normal operating l conditions in the event of a total loss of off-site power. .,

! ' . The steam driven auxiliary feedwater pump.is capable of delivering 880 [.

gpe (total feedwater flow) and each of the electric driven auxiliary feedwater ,

pumps are capable of delivering 440 gpm (total feedwater flow) to the entronce 1

of the steam generators at steam generator pressures of 1100 psia. At R119 1100 psia the.open steam generator safety valve (s) are capable of relievinti at least 11L of nominal steam flow. A total feedwater flow of 440 gpm at ' '

I pressures of.1100 psia is sufficient to ensure that adequate feedwater flow is available to remove decay heat' and reduce the Reactor Coolant System R119l temperature to less than 350'F where the Residual-Heat Removal System may te R placed into operation. The surveillance test values ensure that each pump will. provide at least 440 gpm plus pump recirculation flow against a steam R119 -

1: . generator pressure of 1100 psia.

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pA pM j 3/4.7.1.3 CONDENSATE STORAGE TANK The OPERABILITY of the condensate storage tank with the minimum water volume ensures that sufficient water is availaule to maintain the RCS at HCT 51ANDBf conditions for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> with steam discharge to the atmos'phere conctrrent with total loss of off site power. The contained water volume limit inclucen an allowance for water not usuable because of tank discharge line location or other physical characteristics.

.M SEQUOYAH - UNIT I B 3/4 7-2 Amendment No.115

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. May 11. 19.19

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,_ PLANT SYSTEMS.

BASES

.s SAFETY-VALUES (Continued)

~109 =

Power Range Neutron flux-High Trip Setpoint 'for 4 loop operai. ion 76 =

Maximum percent of RATED THERMAL POWER permissible by P 8 Setpoint for 3 loop operation.

• X =

Total- relieving' capacity of6 all safety valves per steam L line in Ibs/ hour, 4.75 x 10 lbs/hr at 1170_psig Y = Maximum relieving gapacity of any one safety valve in 1bs/ hour, 9.5 x 10 1bs/hr at 1170 psig.

3/4.7.1.2 AUXILIARYFEEDWAMRSYSTEM

- The OPERABILITY of the euxiliary feedwater system ensures that the Ritactor Coolant System can be cooled down to less than 350'F from norma.1 operatinil conditions in the'ev40t of a total loss of off-site power.

(total feedwater flow) and each of the electric driven auxiliarThe st pumps are capable of delivering 440 gpm (total feedwater flow) to ythe feedwate' ent'ance of the steam generators at steam generator pressures of 1100 psia. At '

10 1100 psia the open steam generator safet valve (s) are capable of relieviig at -

'least 11% of-nominal steam flow. A tota feedwater flow ~of-440 gpm at prissures

~ [ sof 1100 psia is sufficient to ensure that adequate feedwater flow is available lR i to remove decay heat and reduce the-Reactor Coolant System temperature to less

-than 350'F where the Residual Heat Removal' System may be placed into operntion.

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.t The surveillance test values' ensure that each pump will provide at least H0 gpa plus pump recirculation flow against a steam generator pressure of 1100 pita.

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-l 374.7.1.3 CONDENSATE STORAGE TANK l L

The OPERABILITY of the condensate storage tank with the minimum water volume ensures that sufficient water is available to maintain the RCS at 107 STANDBY conditions for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> with steam discharge to the atmosphere con:urrent .

with total loss of off site power. I

,-J The contained water volume limit inc1Jdes L

an allowance for water other physical characteristics. not usable because of tank discharge line location or 1

_3/4.7.1.4 ACTIVITY 1 p The limitations on secondary system specific activity ensure that the resultant off site radiation dose will be limited to a small fraction of 10 CFR Part 100 limits in the event of a steam line rupture. This dose also includes the effects of a coincident 1.0 GPM primary to secondary tube lesh in I

the steam generator of the affected steam line.

with the assumptions used in the accident. analyses.These values are consistent  !

SEQUOYAH - UNIT 2 'B 3/4 7-2 Amendment No. 10b May 11 1989 l

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Attachment Each motor-driven auxiliary feedwater pump (one Train A and one Train B) supplies flow paths to two steam generators. Each flow path contains an automatic air-operated level control valve (LCV). The LCVs have the same train designation as the associated pump and are provided trained air.

-The turbine-driven auxiliary feedwater pump supplies flow paths to all' four steam generators. Each of these flow paths coatains an automatic air-operated LCV, two of which are designated as Train A, receive A-train air, and provide flow to the same steam generators that are supplied by the B-train motor-driven auxiliary feedwater pump. The remaining two LCVs are designated as Train B, receive B-train air, and provide flow to the same steam generators that are supplied by the A-train motor-driven pump.

This design provides the required redundancy to ensure that at least two  ;

steam' generators receive the necessary flow assuming any single failure. f It can be seen from the description provided above that the loss of a :1 single train of air (A or B) will not prevent the auxiliary feedwater  !

system from performing its intended safety function and is no more severe than the loss of_a single auxillary feedwater pump. Therefore, the loss of a single train of auxiliary air only affects the capability of a single motor-driven auxiliary feedwater pump because the turbine-driven pump is i still capable of providing flow to two. steam generators that are separate-from the other motor-driven pump.

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ENCLOSURE 2 PROPOSED TECHNICAL SPECIFICATION CHANGE SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 (TVA-SON-TS-90-10)

DESCRIP110N AND JUS 11FICAT10N FOR REVISING BASES SEC110N 3/4.7.1.2 i

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[ ENCLOSURE 2 Description of Change l

Tennessee Valley Authority proposes to modify the Sequoyah Nuclear Plant I' (SQN) Units 1 and 2 technical specifications (ISs) Bases Section 3/4.7.1.2 to clearly define the auxiliary air _ requirements to meet the requirements i of Limiting Condition of Operation (LCO) 3.7.1.2.

Reason for Chance f-The auxiliary air system is common equipment to both units at SQN and L therefore supplies a trained backup / emergency air supply to equipment  ;

[ 1 required for operation of each unit. LCO 3.7.1.2 requires that at least '

4 three independent steam generator auxiliary feedwater (AFW) pumps and  :

associated flow paths be operable when the reactor is in Mode 1, 2, or 3.

l One interpretation of the LCO rettuires two AfH pumps to be declared inoperable when either train of aux 111ary air supply to the AfW level control valves is inoperable. The interpretation places both units at SQN in an action statement requiring shutdown within six hours (Action b of 4

LCO 3.7.1.2).

, Au_s,tificationforChang s

As presently stated in Bases Section 3/4.7.1.2, a total feedwater flow to the steam generators of 440 gallons per minute at a pressure of 1,100 pounds per squarc inch absolute (psla) is sufficient to ensure that adequate feedwater flow is available to remove decay heat and reduce the reactor coolant system temperature to less than 350 degrees fahrenheit where the residual heat removal system may be placed into operation. The AfH system provides complete redundancy in pump capacity and water supply for all cases for which the system is required. 'Under all credible l accident conditions, at least one AfW pump is available to supply each steam generator not affected by the accident with its required feedwater.

p Only two steam generators are required to be usable for any credible accident condition. Redundant electrical power and air suppites ensure reliable system initiation and operation. The electric motor-driven pumps are powered by offsite or onsite sources; the turbine-driven pump takes L

steam from either of two. main steam lines upstream of isolation valves.

Each motor-driven AFW pump (one Train A and one Train B) supplies two steam generators-through its associated flow paths and air-operated level control valves. The turbine-driven AfW pump is capable of delivering i required flow to each of the four steam generators through separate flow h paths. Two of the flow paths and associated air-operated level control valves are Train A and the remaining two are Train B. The loss of a

-single train of auxiliary air would result in the loss of the flow paths y associated with the same train of motor-driven AFW pump and the two flow paths of the same train from the turbine-driven AfW pump. The flow paths associated with the opposite train of auxiliary air (two from the motor-driven and two from the turbine-driven pumps) would be unaffected.

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' Loss of, auxiliary air does not affect the operability of either the I'

- motor-driven pumps or the turbine-driven pump but only the associated .

level control valves. It should be noted that the A-train LCVs associated with the turbine-driven AFW pump supply flow to the same two steam '

generators as the B-train motor-driven AFW pump,-and the B-train paths y supplied by the turbine-driven pump supply the same two generators as the i i' A-train motor-driven pump. Therefore the AFW system would be able to .

. ' supply the required flow to the steam generators assuming any single-failure..  ;

L Based on the discussion provided'above, loss of a single train of. .

L auxillary (emergency) air to the AFW system is no more severe than the {

loss of a single AFW pump and therefore should have a consistent action i 4 , time. .

a dcag re u es not involve an unreviewed environmental question because operation of SQN Units 1 and 2 in accordance with this change would not:

1. Result in a significant increase in any adverse environmental impact previously evaluated in the final Environmental Statement (FES) as modified by.the Staffi s-testimony to the Atomic Safety and Licensing.

Board, supplements to the FES, environmental impact appraisals, or decisions of the Atomic Safety and Licensing Board.

2. Result-_in a significant change in effluents or power levels.

3. Result in matters not previously reviewed in the licensing basis for SQN'that may have.a significant environmental impact.

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.ENCt.0SURE 3  ;

l . PROPOSED TECHNICAL SPECIFICATION CHANGE- .

l SEQUOYAH NUCLEAR PLANTJUNITS 1 AND 2c .j DOCKET N05. 50-327 AND'50-328 'l 9 1 Q

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ENCLOSURE 3 Significant Hazards Evaluation TVA has evaluated the p"' ased TS change and has determined that it does not represent a signif ha7ards consideration based on criteria established in 10 CFR 5. . Operation of SQN in accordance with the proposed amendment will not.

(1) Involve a significant increase in the probability or consequences of an accident previously evaluated.

LC0 3.7.1.2 ensures that the reactor coolant system can be cooled down to less than 350 degrees Fahrenheit from normal operating conditions under all credible accident conditions. The bases change does not alter this requirement. The change merely clarifies the relationship between aux 111ary air supply and the operability of a train of AfW. Because the design requirement for the AFW system (i.e., minimum water delivery capability) remains unchanged, the bases change will not increase the probability or consequences of an accident previously evaluated.

(2) Create the possibility of a new or different kind of accident from any previously analyzed.

The bases change is made to clarify the relationship between auxiliary air supply and the operabillty of a train of AFN. There are no changes to design, method of operation, or testing.

Therefore, the possibility of a new or different kind of accident from those previously analyzed is not created.

(3) Involve a significant reduction in a margin of safety.

As described above, the change made to Bases Section 3/4.7.1.2. 1s made to clarify the relationship between auxiliary air supply and the operability of a train of AfW. Because the design requiremeats of the AfH system remain unchanged and surveillance testing will ensure thi.t these requirements continue to be met, there is no reduction in the margin of safety.

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