ML20207D650

Proposed Tech Specs,Adding Operability & Surveillance Requirements for Low Temp Overpressure Protection Sys

ML20207D650
Person / Time
Site:	Sequoyah
Issue date:	08/10/1988
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML20207D656	List: ML20207D650
References
NUDOCS 8808150411
Download: ML20207D650 (63)

Text

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3 ENCLOSURE 1 PROPOSED TECHNICAL SPECIFICATION CHANGE SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 (TVA-SQN-TS-88-03)

LIST OF AFFECTED PAGES Unit i VI XIII 3/4 1-7 3/4 1-8 3/4 1-9 3/4 1-10 3/4 4-2 3/4 4-23 3/4 4-28 (added) 3/4 4-29 (added) 3/4 4-30 (added)

B 3/4 1-2 B 3/4 1-3 r

B 3/4 4-1 B 3/4 4-7 B 3/4 4-14 5-6 Unit 2 VI XIII 3/4 1-7 3/4 1-8 3/4 1-9 3/4 1-10 3/4 4-3 3/4 4-28 3/4 4-34 (added) 3/4 4-35 (added) 3/4 4-36 (added)

B 3/4 1-2 B 3/4 1-3 B 3/4 4-1 B 3/4 4-8 B 3/4 4-14 B 3/4 4-15 (added) 5-6 8808150411 880810 ":

PDR ADOCK 05000327 p

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INDEX i,

-LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS L

PAGE

'SECTION

.3/4.4 REACTOR COOLANT SYSTEM

-l 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION S tar tup a nd Powe r Ope rati on...............................

3/4 4-1 l-1 Hot. Standby...............................................

3/4 4-la Shutdown..................................................

3/4 4-2 U

3/4.4.2 SAFETY VALVE S - SHUT D0WN.......... '..................,.....

3/4 4-3 3/4.4.3 SAFETY AND RELIEF VALVES - OPERATING Safety Valves - Operating......'......'.....................

3/4 4-4 Relief Valves - Operating...................,.............

3/4 4-da 3/4.4.4 PRESSURIZER...............................................

3/4 4-5

.T 3/4 4;5 STEAM GENERATORS..........................................

3/4 4-6

(

i 1

3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems.................................

3/4 4-13

. 0perational Leakage.......................................

3/4 4-14 1

3/4.4.7 CHEMISTRY.................................................

3/4 4-16 3/4 4-19 3/4.4.8 SPECIFIC ACTIVITY.................,........................

3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System....................................

3/4 4-23 Pressurizer...............................................

3/4 4-26 3/4.4.10 STRUCTURAL INTEGRITY i

ASME Code Class 1, 2 and 3 Components..................

3/4 4-27 7

3/4.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) t 3/4.5.1 ACCUMULATORS 9

Cold Leg Injection Accumula tors.......................

3/4 5-1 7

Upper Head injection Accumulators.....

3/4 5-3 5E000YAH - V'111 1 V[

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,PAGE SECTION 3/4.4.6 REACTOP COO LANT SYST EF L E AKAGE............................. B 3/4 4-3 3/4.4.7 CHEFISTRY..................................................

B 3/4 4 4 3/4.4.8 SPECIFIC ACTIVITY..........................................

B 3/4 4-5 3/4.4.9 PRESSUR E/TEMPERATUR E LIMITS................................ B 3/4 4-6 3/4.4.10 STRUCTURAL INTEGRITY......................................

8 3/4 4-14

,?l4 g ll OVCW86560& (20TCCitOA) MSTUN5 6 3/4 4 -l&

3/4.5 EFERGENCY CORE COOLING SYSTEuS (ECCS) 3/4.5.1 A C C UMJ LAT0 R S............................................... 8 3/4 5-1 3/4.5.2 and 3/4.5.3 ECCS SilRSYSTEFS................................

P 3/4 5-1 3/4.5.4 PORON INJECTION SYSTEF.....................................

B 3/4 5-2

[Y 3/4.5.5 REFUELIFG WATEP STOPAGE TANK (PWST)........................

R 3/4 5-3 3/4.6 CO*!TAINf'ENT SYSTEPS 3/4.6.1 PRIPARY C0NTAINFENT........................................

2 3/4 6-1 3/4.6.2 OEPRES SURIZATION AND COOLING SYSTEPS....................... B 3/4 6-3 3/4.6.3 CONTAINPENT ISOLATION VALVES...............................

R 3/4 6-3 3/4.6.4 COPSUST!BLE GAS CONTP0L....................................

B 3/4 6-3 3/4.6.5 ICE CONDENSER..............................................

8 3/4 6 4 3/4.6.6 YACUOM RELIEF VALVES.......................................

B 3/4 6-6 3/4.7 PLANT SYSTPS 3/4.7.1 TURBINE CYCLE..............................................

R 3/4 7-1 3/1.7.2 STEAM GENERATOR PRESSURE /TEPERATURE LIP!TATION.............

o 3/4 7-3 2 3/4 7-3 3/4.7.3 COMPONENT C00 lit!G WATER SYSTEw.............................

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SE000YAH - Uti!T 1 XIII l

,kEACTIVITY CONTROL SYSTEMS 3/4.1.2 CORATf0N SYSTEMS FLOW PATHS - SHUTDOWN LIMITING CON 0! TION FOR OPERATION 3.1.2.1 As a minimum, one of the following boron injection flow paths shall be OPERABLE:

a.

A flow path from the boric acid tank via a boric acid transfer pump and charging pump to the Reactor Coolant System if only the boric acid storage tank in Specification 3.1.2.Sa is OPERABLE, or b.

The flow path from the refueling water storage tank via a charging pump to the Reactor Coolant System if only the refueling water storage tank in Specification 3.1.2.5b is OPERABLE.

APPLICABILITY: MODES S and 6.

ACTION:

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44 4-ooe: Of th: 2:;: '!:u path: ^PMAElbuspend-+M-opeeMener+a W4*+et-I j

CP.E,CER"M'.5 r p;;i ti ; re::t!'ity change;.=

l SURVEILLANCE REQUIREMENTS 4.1.2.1 At least one of the above required finw paths shall be demonstrated OPERABLE:

a.

At least once per 7 days by verifying that the temperature of the heat traced portion of the flow path is greater than or equal to 145'F when a flow path from the boric acid tanks is used, b.

At least once per 31 days by verifying t' hat each valve (manual, power operated or automatic) in the flow path that is not locked, sealed, or othemise secured in position, is in its correct position.

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REACTIVITI CONTROL SYSTEMS FLOW PATHS - OPERATING LIMITING CONDITION FOR OPERATION z

At least two of the following three boron injection flow paths shall 3.1.2.2 be OPERABLE:

The flow path from the boric acid tanks via a boric acid transfer a.

pump and a charging pump to the Reactor Coolant System.

Two flow paths from the refueling water storage tank via charging b.

pumps to the Reactor Coolant System.

MODES 1, 2,h :nf '

3 APPLICABILITY:

1-dnd ACTION:

With only one of the above required boron injection flow paths to the Reactor Coolar.t System OPERABLE, restore at least two boron injection flow paths to the Reactor toolant System to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 1% delta k/k I

at 200*F within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore at least two flow paths to OPERABLE status within the next 7 days or be in 44MN> SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

TWT' SURVEILLANCE REOUIREMENTS At least two of the above required flow paths shall be demonstrated 4.1.2.2 OPERABLE:

At least once per 7 days by verifying that the temperature of the heat traced portion of the flow path from the boric acid tanks is a.

greater than or equal to 145'F when it is a required water source.

At least once per 31 days by verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed, b.

or otherwise secured in position, is in its correct position.

Atleastonceper18monthsduringshutdownbyverifyingthateach automatic valve in the flow path actuates to its correct position on c.

a safety injection test signal.

At least once per 18 months by verifying that the flow path required d.

by Specification 3.1.2.2a delivers at least 10 gpm to the Reactor Coolant System.

htAR 251982 3/4 1-8 Amencment No. 12 SEQUDYAH - UNIT 1

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REACTfVITY CONTROL SYSTEMS CHARGING PUMP - SHUTOOWN LIMITING CONDITION FOR OPERATION 3.1.2.3 One chargis.; pump in the boron injection flow path required by Sper:ification 3.1.2.1 shall be OPERABLE and capable of being powered from an OPERABLE shutdown board.

APPLICABILITY: MODES 5 and 6.

ACTICH:

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e 4.1.2.3 The abcve required charging pump shall be demonstrated CPERABLE by verifying, that on recirculation flow, the pump develops a discharce pressure of greater than or equal to 2400 psig when tested pursuant to Specification 4.0.5.

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V SEQUOYAH - UNIT 1 3/4 1-9

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4 REACTIVITY CONTROL SYSTEMS CHARGING PUMPS - OPERATING LIMITI"G CONDITION FOR OPERATION 3.1.2.4 At least two charging pumps shall be OPERABLE.

APPLICABILITY:

MODES-1, 2, 3 :rd '

ACTION:

da d' With only one charging pump OPERABLE, restore at least two charging pumps to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY and bnrated to a SHUTOOWN MARGIN equivalent to at least 1% delta k/k at 200*F within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore at least two charging pumps to OPERABLE status within the next 7 days or be in 44te SHUI 30WN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

H01-SURVEILLANCE REOUIREMENTS

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4.1.2.4 At least two charging pumps shall be demonstrated OPERABLE by verifying, that on recirculation flow, each pump develops a discharge pressure '

of greater than 2. equal to 2400 psig when tested pursuant to Specification 4.0.5.

,a SEQUDYAH - UNIT 1 3/4 1-10 Amencmer, No. 1 if]3 32 %

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,l REACTOR COOLANT SYSTEM

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5HUTDOWN

_ LIMITING CONDITION FOR OPERATION 3.4.1.3 a.

At least two of the reactor coolant and/or residual heat removal (RHR) loops listed below shall be OPERABLE:

1.

Reactor Coolant Loop A and its associated steam generator and reactor coolant pump,%

2.

Reactor Coolant Loop B and its associated steam generator and reactor coolant pump,A 3.

Reactor Coolant Loop C and its associated steam generator and reactor coolant pump,4s 4.

Reactor Coolant Loop D and its associated steam generator and reactor coolant pump,3-5.

Residual Heat Removal Loop A.

R16 6.

Residual Heat Removal Loop B.

,n b.

At least one of the above reactor coolant and/or RHR loops shall

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be in operation.**

APPLICABILITY:

MODE 4.

ACTION:

a.

With less than the above required loops OPERABLE, immediately initiate corrective action to return the reouired loops to OPERABLE status as soon as possible; be in COLD SHUTDOWN within 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />.

b.

With no reactor coulant or RHR loop in operation, suspend all opera-l tions involving o reduction in bcron concentration of the Reactor l

Coolant System and immediately initiate corrective action to return l

the required coolant loop to operation.

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^^All reactor coolant pumps and residual heat removal pumps may be de-energized 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.

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MAk20 1902 SEQUOYAH - UNIT 1 3/4 4-2 Amendment No. l'

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5-REACTOR COOLANT SYSTEM

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i 3/4.4.9 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM

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LIMITING CONDITION FOR OPERATION

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i 3.4.9.1 The Reactor Coolant System (except the pressuriz'er) temperature and pressure shall be limited in accordance with the limit lines shown on Figures.3.4-2 and 3.4-3 during heatup, cooldown, criticclity, and inservice leak and hydrostatic testing with:

a.

A maximum heatup of 100 F in any one hour period.

b.

A maximum cooldown of 100*F in any one hour period.

c.

A maximum temperature change of less than or equal to 5 F in any one hour period during inservice hydr,. static and leak testing operations above the heatup and cooldown limit cufves.

APPLICABILITY: At all times.

k ACTION:

f.

With any of the above limits exceeded, restore the temperature and/or pressure to.with;n the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural l

integrity of the Reactot oolant System; deter' that the Reactor Coolant l

System remains acceptable for continued operatn.

or be in at least HOT STAN0dY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce in RCS T,yg and pressure to less than 200 F and 500 psig, respectively, within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once nor 3C minutes during system heatup, cooldown, and inservice leak and hyd nstatic testing operations.-

4.4.9.1.2 The reactor vessel material irradiation surveillance specimens l

shall be removed and examineo, to determine change in material properties, at the intervals required by 10 CFR 50, Appendix H in accordance with the schedule in Table 4.4-5.

The results of these examinations shall be used to update M'

Figures 3.4-2 m4 3.4 p an d 4, y. 9, MAR 25 im

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SEQUOYAH - UNIT 1 3/4 4-23 Amendment No.12 g

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REACTOR COOL ANT SYSTE?t' Ler! TimMgNDes A OVERPRESIDRE DROTECTION SYS1Elj5, LIMITING CONDITION FOR ODFPATION

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Il 3.4.-9-G-At least one of the following Overpressure Protection Systems sha be OPERABLE:

Two p' ower operated relief valves (PCRVs) with' a nominal lif t a.

less than or equal to that shown in Figure 3.4-4, or b.

The Reactor Coolant System (RCS) depressurized with an RCS vent of greater than or equal to 3 square inches.

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b.

With both PORVs inoperable, depressurize and vent the RCS throug least a 3 souare inch vent within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.$

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an RC; aressure transient, a Special Report shall submined to the Commission pursuant to Spedffication 6.9.2 within 30 days.

The report shall cescribe the circumstances initiating the transient, the ef fect of the PORVs or RCS vent (s).on the transient and any corrective action necessary to prevent recurrence.

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The provisions of Specification 3.0.4 are not applicable.

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FIGUdE 3.4-4 PORV NOMINAL LIFT SETTINGS - APPLICABLE TO 9.2 EFPY

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V REACTOR COOLANT SYSTEM SURVEILLANCE REOUIREMENTS ll. l 4.4.N Each PORV shall be derenstrated OPERABLE b -

Performance of an ANALCO CHANNEL OPERATIONAL T T on the PORV a.

actuation channel, but excluding valve operatio, r " r. ::. :-

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. J and at least once per 21 days thereaf ter when the PCRV is required OPERABLE:

b.

Performance of a CHANNEL cal.IBRATION on the PORY actuation channel at least once per 18 months; and Verifying the PORV isulation valve is open at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> c.

when the PGFV is being used for overpressure p etection.

4.4.0.0.0 The ROS vent (s) shall be verified to be open at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />" when the vent (s) is being used for overpressure protection.

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-L*.~~0 W - UNIT 1 3f4 4 1

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REACTIVITY CONTROL SYSTEMS i.

BASES condition of all rods inserted (most positive MDC) to an all rods withdrawn condition and, a conversion for the rate of change of moderator density with temperature at RATED THERMAL POWER conditions.

This value of the MOC was then

~4 transformed into the limiting MTC value, -4.0 x 10 delta k/k/*F. The MTC value of -3.1 x 10'4 delta k/k/ F represents a conservative value (with correc-tions for burnup and soluble boron) at a core condition of 300 ppm equilibrium baron concentration and is obtained by making these corrections' to the limiting MDC value -4.0 x 10'4 delta k/k/ F.

The surveillance requirements for measurement of the MTC at the beginning and near the end of each fuel cycle are adequate to confirm that the MTC remains within its limits since this coefficient' changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup.

3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICALITY

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This specification ensures that the reactor will not be made critical with the Reactor Coolant System average temperature less than 541 F.

This limitati~on is required to ensure 1) the moderator temperature coefficient is within its analyzed temperature range, 2) the protective instrumentation is within its normal operating range, 3) the P-12 interlock is above its setpoint,

4) the pressurizer is capable of being in a OPERABLE status with a steam bubble, and 5) the reactor pressure vessel is above its minimum RT temperature.

NOT 3/4.1.2 BORATION SYSTEMS The boron injection system ensures that negative reactivity control is available during each mode of facility operation. The components required to perform this function include 1) borated water sources, 2) charging pumps, 3) seperate flow paths, 4) boric acid transfer pumps, 5) associated heat tracing systems, and 6) an emergency power supply from OPERABLE diesel generators.

With the RCS average temperature above F, a minimum of two separate I

and redundant boron injection systems are provided to ensure single functional l

capability in the event an assumed failure renders one of the systems inoperable.

l The boration capability of either flow path is sufficient to provide a SHUTDOWN l

MARGIN from expected operating conditions of 1.6% delta k/k af ter xenon decay and cooldown to 200*F.

The maximum expected boration capability requirement occurs at E0L from full power equilibrium xenon conditions and requires C

SEQUOYAH - UNIT 1 8 3/4 1-2

e REACTIVITY CONTROL SYSTEMS BASES 5408 gallons of 20,000 ppm borated water from the boric acid storage tanks or 64,160 gallons of 2000 ppm borated water from the refueling water storage tank.

M0 With the RCS temperature below 200 F, one injection system is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity change in the event the single injection system becomes inoperable.

The boron capability required below 200 F, is sufficient to provide a SHUTOOWN MARGIN of 1% delta k/k after xenon decay and cooldown from 200 F to 140 F.

This condition requires either 835 gallons of 20,000 ppm borated water from tne boric acid storage tanks or 9,690 gallons of 2000 ppm borated water from the refueling water storage tank.

The contained water volume limits include allowance for water not available T/

because of discharge line location and other physical characteristics.

The limits on contained water volume and boron concentration of the RWST BR also ensure a pH value of between 7.5 and 9.5 for the solution recirculated within containment after a LOCA.

This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corrosion on mechanical systems and components.

The OPERABILITY of one boron injection system during REFUELING ensures that this system is available for reactivity control while in MODE 6.

3/4.1.3 MOVABLE CONTROL ASSEMBLIES The specifications of this section ensure that (1) acceptable power distri-bution limits are maintained, (2) the minimum SHUTDOWN MARGIN is maintained, and (3) limit the potential effects of rod misalignment on associated accident analyses.

OPERABILITY of the control rod position indicators is required to determine control rod positions and thereby ensure compliance sith the control rod alignment and insertion limits.

l SEQUOYAH - UNIT 1 B 3/4 1-3 Revised 08/18/87 b

o i

3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 R ACTOR COOLANT LOOPS AND COOLANT CIRCULATION The pie.ni is designed to operate with all reactor coolant loops in opera-tion, and maintain DNBR above 1.30 during all normal operations and anticipated 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 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

In MODE 3, a single reactor coolant loop provides sufficient heat removal capabiliity for removing decay heat; however, single failure considerations require tnat two loops be OPERABLE.

In MODE 4, a single reactor coolant loop or residual heat removal (RHR) loop provides sufficient heat removal capability for removing decay heat; but R16 single failure considerations require that at least two loops be OPERABLE.

Thus, if the reactor coolant loops are not OPERABLE, this specification requir,es two RHR loops to be OPERABLE.

,. ~.

J R16 In MODE 5, single failure considerations require that two RHR loops be i

OPERABLE.

The operation of one Reactor Coolant Pump 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 rMm ~ O'h ~ R CP %ba bba>ubhNe iMl5 5 M b [ /95 W /8 M shimap k gN3W.% bg,m

& & psup a 4)v stuhd, Jk deam b Ade Mll aw'n0 M h tcsJ0tN hfMtW d4 dO M

W y

cha,gog pups n eapW a a,ned, This w,k w ll larven a

F a p. le y f u.J < esp +h de 3 uvt1, t, duiGdY N6fM'ed kaas 'enh d r*'*

'"5" f"'#

l se wnfu l.v i f i

3/4.4.2 and 3/4.4.3 SAFETY AND RELIEF VALVES The pressurizer code safety valves operate to prevent the RCS from being pressurized above its Safety Limit of 2735 psig.

Each safety valve is designed to relieve 420,000 lbs per nour of saturated steam at the valve set point.

The relief capacity of a single safety valve is adequate to relieve any over-j/, pressure condition which could occur during shutdown.

In the event that no MAR 25 G82 SEQUOYAH - UNIT 1 B 3/4 4-1 Amendment ho. 12 1

]

f REACTOR COOLANT SYSTEM RASES Heatup and cooldown limit curves are calculated usino the nost ifmitino value of the nil-ductility reference temperature, RTFDT, at the end of 9.2 effective full power years of service life. The 9.2 EFPY service at the 1/4T location life. period is chosen such that the limiting RTNDT in the core region is greater than the PTNOT of the limitino unirradiated assures that all naterial. The selection of such a limiting PTNDT components in the Peactor Coolant Systm will be operated conservatively in accordance with applicable Code recuirenents.

The reactor vessel materials have been tested to detemine their initial NOT; the results of these tests are shown in Table B 3/4.4-1.

Peactor PT operation and resultant fast neutron (E greater than 1 PEV) irradiation Therefore, an adjusted reference can cause an increase in the PTNOT.

temperature, based upon the fluence of the raterial in cuestion, can be predicted usino Figure R 3/4.4-1. The heatup and cooldokn limit curves of Figures 3.4-2 and 3.a-3 include credicted adjustnents for this shift in PT at the end of 9.2 EFPY, as well as adjustnents for oossible NDT errors in the pressure and temoerature sensino instrurents.

detemined in this manner nay be used until the Values of delta PTNOT results from the caterial surveillance procram, evaluated accordino to ASTP E185, are available. The first capsule will be removed at the end of the first core cycle. Succes!.ive capsules will be removed in accordance with the reouirements of ASTF ElPS-73 and i0 CFP 50, Appendix H.

The heatup and cooldown curves must be recalculated when the delta PT detemined fran the surveillance capsule exceeds the calculated NOT delta RT for the eouivalent caprule radiation exoosure.

NDT

& 0h h%

V gu. pwG ^W"b e

/

SEN10YAH - UPIT 1 P 3/4 a-7

REACTOR COOLANT SYSTEM

~

~

BASES E~

i-

' J-

' ~

T-

~~ ~

7

-..7 3/4.4.10 STRUCTURAL INTEGRITY. V...

.. ~....

- The inservice inspection and testing programs for ASME Code Class 1, 2 and 3 components ensure that the structural integrity and ooerational readiness of these canponents will be maintained at an acceptable level throuchout the life of the plant.

These programs are in accordance with Section XI of the ASFE Boiler and Pressure Vessel Code and applicable Addenda as reouired by 10 CFR Part 50.55a(g) except where specific written relief has been granted by the Canmission pursuant to 10 CFR Part 50.55a (g)(6)(i).

Components of the reactor coolant systen were designed prior to issuance of Section XI of the ASME Boiler and Pressure Vessel Code. These components will be tested to the extent practical within the limitations of the original

• plant design, oeometry, and materials of construction of the components.

3/4.e.11 OVERPRESSURE PROTECTION SYSTEM The operability of two PORVs or an RCS vent opening of at least three j[$gfl square inches 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 degrees F.

Either PORV has adequate relieving capability to protect the I

RCS from overpressurization when the transient is limited to either:

(1) the start of an idle RCP with a water-solid RCS and a secondary water I

temperature of the steam generator less than or equal to 50 degrees F above the RCS cold leg temperatures, or (2) the start of a charging pump and its injection into the RCS with letdown isolated.

The maximum allowed PORV setpoint for the low temperature overpressure protection (LTOP) system is derived by analysis which models the perfor=ance of the LTOP system assuming various mass input and heat input transients.

Operation with the PORV setpoint less than or equal to the maxi =um s'etpoint ensures that Appendix G 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 processing and valve opening, instrument uncertainties, and single f ailure. To ensure that mass and heat input transients more severe than those assumed cannot occur, technical specifications require tagout or isolation of all l

but one centrifugal charging pump while in modes 4, 5, and 6 with the reactor vessel head installed and disallow restart of an RCP if a steam bubble does not exist in the pressurizer.

The LTOP system setpoints include a 50 degree F allowance for heat transport effects and a 27 degree F allowance for instrument accuracy.

An 800 psig pressure limit protects the PORV piping from the consequences of a possible water hammer caused by the rapid opening times associated with

• he PORVs.

l l

<EOUOYAH - U.'!JT I R 3/4 a ld l

l

~

1ABLE 5.7.1 d

C0!!PONENT CYCtlC OR 1RANSIENT LIMITS

')

2 CYCtlC OR DESIGN CYCLE CO !P0llEN1 IRANSitHI LlHII OR TRANSIENT c_C

]

Reactor Coolant System 200 heatup cycles at < 100*f/hr ifeatup cycle - I from < 200*f

'*9 and 200 cuoldown cycles at to > 550*F.

< 100*f/hr CooIdown cycle - I frca L 550*F to $ 200*F '9 200 pressurizer cooldown cycles Pressurizer cooldown cycle temperatures frem 1 50*f to 0

at $ 200*F/hr

_ 200*F.

> 15% of RATED IllERHAL POWER to 80 loss of load cycles, without immediate turbine or reactor trip.

0% of RAIED TilEHMAL POWER.

Lc,:; of of fsite A.C. electrical T

40 cycles of loss of offsite power tource supplying the onsite A.C. electrical power.

ESF Electrical System. -

Loss of caly one reactor 80 cycles of loss of flow in one coolant pump.

reactor coolant loop.

100% to 0% of RATED TilERHAL POWER.

400 reactor trip cycles.

Spray water temperature dif ferential g

12 spray actuation cycles.

> 320 F and 5 560*f.

t Pressurized to 2485 psig 50 leak tests Pressurized to 3105 psig k

5 hydrostatic pressure tests h

l 0 Lou %piNw e waie'-4?I?l CVWfMS**

Ylbtt's0lb *r,yY OCbJ' 2vZnis Pressurized to 1330 psig f.

5cconilary System 5 hydrostatic pressure tests o

,,, i..lo.ai c w. i

.. i. o. i,m,..o n.mn,

INDEX l

LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE RE0VIREMENTS b..f SECTION

_PAGE 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR CCOLANT LCOPS AND COOLANT CIRCULATION

~

.I S ta rtup and Powe r Ope ra tion...............................

3/4 4-1

'i Hot Standby..................'.............................

3/4,4-2

'{

Hot S'utdown..............................................

h I

3/4 4-3 Cold Shutdown.............................................,,'3/4 4-5. -

l 3/4.4.2 SAFETY VALVES - SHUT 00WN..................................

3/4 4-6:.

3/4.4.3 SAFETY AND RELIEF VALVES - OPERATING SafetyValves0perating...................................~3[44-7 Relief Valves 0perating..................................

3/4 4-8

~

3/4.4.4 PRESSURIZER...............................................

3/4 4-9

~

S iTAM G ENERA TOR S..................................

s 3/4.4.5 3/4 4-10 i '

f, 3/4.4.6 REACTOR COOLANT SYSTEM LEAXAGE Leakage De tec ti on Sys tems.................................

3/4 4-17

~ ~

Operational Leakage.......................................

3/4 4-18 3/4.4.7 CHEMISTRY.................................................

3/4 4-21 3/4.4.8 SPECIFIC ACTIVITY.........................................

3/4 4-24 3/4.4.9 PRESSURE / TEMPERATURE LIMITS Reactor Coolant System.............~t......................

3/4 4-28 Pressurizer...............................................

3/4 4-32 3/4.1.10 STRUCTURAL. INTEGRITY 1

ASME Code Class 1, 2 and 3 Components..............

3/4 4-33

, /l, ll Cylig(fesSUR6 f,COTECT/04 s MTtM4 3N 4d

(,

SEQUOYAH - UNIT 2 i

VI I

e

INDEX f'~~ff-wi, BASES

~~

SECTION PAGE,,

3/4.4.6 REACTOR COO LANT SYSTEM LEAKAGE.............................

8 3/4 4,4 3/4.4.7 CHEMISTRY..................................................

~8 3/4.4-5 '

3/4.4.8 SPECIFIC' ACTIVITY.............

........................... 8 3/4 4-5 3/4.4.9 PRESSURE / TEMPERATURE LIMITS................................

8 3/4 4-6 3/4. 4.' 10 - STRUCTURA L INT EGRITY......................................

8 3/4 4-14

,3,ly, fl DVenPREs.sv/E l'20TecTidal.s.nTeMA

$.3/4 4.-]'f' 3/4.5 EMERGENCY CORE COOLING SYSTEMS

~~

J 3/4.5.1 ACCUMULATORS.................................

8 3/4 5-1 3/4.5.2 and 3/4.5.3 ECCS SUBSYSTEMS.......................s.........83/45-1 3/4.5.4 BORON INJECTION SYSTEM.....................................

8 3/4 5-2 r '.,

3/4.5.5 REFUELING WATER STORAGE TANK...............~................

S 8 3/4 5-2 3/4.6 CONTAINNENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT........................................

8.3/4 6 3/4.6.2 OEPRESSURIZATION AND COOLING SYSTEMS.......................

8 3/4 6-3

~

3/4.6.3 CONTAINMENT ISOLATION VALVES...............................

8 3/4 6-3 3/4.6.4 COMBUSTIBLE GAS CONT 30L....................................

8 3/4 6-3 3/4.6.5 ICE C0HOENSER..............................................

8 3/4 6-4 3/4.6.6 VACUUM RELIEF VALVES.......................................

8 3/4 6-6 3/4.7 PLANT SYSTEMS 3/4.7.1 TURBINE CYCLE.........................................

B 3/4 7-1 3/4.7.2 STEAM GENERATOR PRESSUP.E/TEMERATURE LIMITATION............. 8 3/4 7-3 3/4.7.3 COMPONENT COOLING WATER SYSTEM...........................

8 3/4 7-3 f-,4 4%

SEQUOYAH - UNIT 2 XIII

(

-ww

- - - ~

a--

-r-

REACTIVITY CONTROL SYSTEMS 3/4.1.2 BORATION SYSTEMS

~

t FLOW PATH - SHUT 00WN LIMITING CONDITION FOR OPERATION 3.1.2.1 As a minimum, one of the following boron injection flow paths shall be OPERABLE:

A flow path from the boric acid tanks via a boric acid transfer pump a.

and a charging pump to the Reactor Coolant System if the boric acid storage tank in Specification 3.1.2.5a is OPERABLE, or The flow path from the refueling water storage tank via a charging b.

pump to the Reactor Coolant System if the refueling water storage tank in Specification 3.1.2.5b is OPERABLE.

APPLICABILITY: MODES 5 and 6.

ACTION:

~Ath acne c' the ec'f e '!c'1 path; OPEMBLE, ;u; pend all opeNt40ns-invo4vhuy 9

. ~ii; ty changen-

,rv,...m

(

SURVEILLANCE REOUIREMENTS 4.1.2.1 At least one of the above required flow paths snall oe demonstrated OPERABLE:

At least once per 7 days by verifying that the temperature of the a.

heat traced portion of the flow path is greater than or equal to 145*F when a flow patn from the boric acid tanks is used, b.

At least once per 31 days by verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

cff bm Ow pdl,.s cPro1BLE, susfncb MODE 'l -

W,lh ncne v

et II opa d.vas, n ecle.<w c'cAE BLTERAlicas er

. s. Y* u clan etn N res fere. o n e. flow l2d as nact,v.f[oss.esk.

sco,s a rrtuo6S 5 -

u).lL ncoe c F L <&c..<e fto pdLc et'reatc, supend

,.wct...,g aw Arrawcas er ps.La ^~0'h a"J b a tt cyc.d.c.,s c

owns.

SEQUOYAH - UNIT 2 3M 1-7

,' REACTIVITY CONTROL SYSTEMS i,

FLOW PATHS - OPERATING

\\

T LIMITING CONDITION FOR OPERATION 3.1.2.2 At least two of the following three boron injection flow paths shall be OPERABLE:

The flow path from the boric acid tanks via a boric acid transfer a.

pump and a charging pump to the Reactor Coolant System, b.

Two flow paths from the refua' ling water storage tank via charging pumps to the Reactor Coolant System.

APPLICABILITY: MODES 1, 2, and 4,.

ACTION:

[

With only one of the above required boron injection flow paths to the Reactor Coolant System OPERABLE, restore at least two baron injection flow paths to the Reactor Coolant System to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANOBY and borated to a SHUT 00bN MARGIN equivalent to at least 1%

delta k/k at 200 F within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore at least two flow paths to OPERABLE status within the next 7 days or be in M t0 SHUTOOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

Pt0T SURVEILLANCE REOUIREMENTS 4.1. 2. 2 At least two of the above required flow paths shall be demonstrated OPERA 8LE:

At least once per 7 days by verifying that the temperature of the a.

heat traced portion of the flow path from the boric acid tanks is greater than or equal to 145'F when it is a required water source.

]

i b.

At least once per 31 days by verifying that each valve (manual, power operated or automatic) in the flow path that is not locked, sealed, or otherwise secured in position, is in its correct position.

At least once per 18 months during shutdown by verifying that each c.

automatic valve in the flow path actuates to its correct position on a safety injection test signal, i

d.

At least once per 13 months by verifying that the flow path required by Specification 3.1.2.2a delivers at least 10 gpm to the Reactor Coolant System.

I SEQUOYAH - UNIT 2 3/4 1-8

(

~

.,;6 REACTIVITY CONTROL SYSTEMS CHARGING PUMP - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.1.2.3.One charging pump in the baron injection flow path required by Specification 3.1.2.1 shall be OPERABLE and capable of being powered from an OPERABLE shutdown board.

APPLICABILITY: MODES 5 and 6.

ACTION:

4,

'A tk ae e aag'ag po p OPERABLErsuspend--aM-operat4 ens-involeteg CCC:

'LTE"'*CNS cr p::iti c: :::ti ity :n=g :.

SURVEILLANCE REOUTREMENTS 4.1.2.3 The above required charging pump shall be demonstrated OPERABLE by verifying, that on recirculation flow, the pump develops a discharge pressure of' greater than or equal to 2400 psig when tested pursuant to Specifica-tion 4.0.5.

cbry p

CffpBt.E, s%

m/ all cfxrdl,w 5

1. 4 W.$

no cedAme.n n.c L.

cwhes

~

pcs.,k rwas o,

inwlv.,,U te.

ps o.as ra n -

ci, cy.ng p.g ns kre.

one and p-p opase, xye.,J J trd~ r wA y ak,g/? Ta2A Ticas or jovsM*w R~cE+*7 d"ff's "ws 6 -

• "I b is kis:,,

GRE It SEQUOYAH - UNIT 2 3/4 1-9

' REACTIVITY CONTROL SYSTEMS

~'

CHARGING PUMPS - OPERATING, LIMITING CONDITION FOR OPERATION J

j/. l. 2. 4 At least two charging pumps shall be OPERABLE.

R27 APPLICABILITY:

MODES 1, 2, 3 ae4 ACTION:

M With only one charging pump OPERABLE, restore at least two charging pumps to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY and borated to a SHUTDOWN MARGIN equivalent to at least 1% delta k/k at 200 F within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; restore at least two charging pumps to OPERABLE status within the next 7 days or be in 69E9 SHUTDOWN within the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

tioT SURVEILLANCE REQUIREMENTS 4.1.2.4 At least two charging pumps shall be demonstrated OPERABLE by N

verif ing, that on recirculation flow, each pump develops a discharge pressure of greater than or equal to 2400 psig when tested pursuant to Specification 4.0.5.

i i

i beide

• t

With e centri al chargin ump inoperab the emerge y core cooling stem (EC may remai operable for additional '6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />

~

b nd tha identifie

'n the Acti statement.

's temporar, chan expire at 0848 or July 13, 1 August 23, 1984 s--

SEQUOYAH - UNIT 2 3/4 1-10 Amendment No. 27 i i

REACTOR COOLANT SYSTEM

6..

HOT SHUT 00WN y,

~

LIMITING C0NDITION'FOR OPERATION y 'j-3.4.1.3 a.

At'least two of the reactor coolant and/or Residual heat removal (RHR) loops listed below shall be OPERABLE:

1.~

. Reactor Coolant Loop A and its associated steam generator and reactor coolant pump,5 2.

Reactor Coolant loop B and its associated steam generator

'and reactor coolant pump, T.

3.

Reactor Coolant loop C and its associated steam generator and reactor coolant pump, t 4.

Reactor Coolant Loop G and i'.s associated steam generator and reactor ccolint ::t..ro. $

5.

Residual Heat Remo a! uoon A, 6.

' Residual Heat Removal loop 8.

2 b.

At least one of t7e abo e reactor coolant and/or RHR loops shall f, '

be in operatien."

. APPLICABILITY: MODE 4.

V.

ACTION:

a.

With less than the above required loops OPERABLE, immediately initiate corrective action to return the required loops to OPERABLE status as soon as possible; be in COLD SHUTOOWN within 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />, b.

With no reactor coolant or RHR loop in operation, suspend all operations involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate correct ~ive action to return the required coolant loop to cperation.

i 1

"Ali reactor coolant pumps and residual heat removal pumps :nay 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 permittee that.,ould cause dilution of the Reactor Ccolant System boron concentration, ano 2) core

~

' outlet temperature is maintained at least 10 F below saturation temperature.

L l}(gu\\w ss bwt* pJ.wf sldl no k be C'SIMIGl vn $eb h s hetm NlNt 9

hatig d0 p. e o v,'i 3 N, l

SEQUOYAH - UNIT 2 3/4 4-3 l

REACTOR COOLANT SYSTEM

/

~

(

- 3/4.4.9 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION

~

3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance.with the limit' lines shown on Figures 3.4-2 and 3.4-3 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:

a.

A maximum heatup of 100*F in any one hour period.

b.

A maximum cooldown of 100 F in any one hour period, c.

A maximum temcerature change of less than. or equal to 5 F in any one hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.

APPLICABILITY: At all times.

ACTION:

I i

'dith any of the above limits exceeded, restore the tempe ature and/or pressure to within the limit within 30 minutes; perform an engineering evaluation to determine the ef fects of the out-of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System remains acceptable for continued operation or be in at least HOT STANOBY within the' next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce the RCS T and pressure to less than 200 F and 500 psig, avg respectively, within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REOUIREMENTS 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system

'i n - ::c1down, and inservice leak and hydrostatic testing operations.

4.2.9.l.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, at the intervals required by 10 CFR 50, Appendix H in accordance with the schedule in Table 4.4-5.

The results of these examinations shall be used to update Figures 3.4-2 v.4 3.4-3, and 3,y cf, l

e SEQUOYAH - UNIT 2 3/4 4-28 C

r n.

e-

-~,

e

o REACTOR COOLANT SYSTE't-O Te~idrntut26

/\\ OVERPREf5DRE PROTECTICH sys Eljs, i

LIMITING CONDITION FOR ODFRATION Il 3.4.-9 At least one of the following Overpressure Protection Systems shall be OPERABLE:

Two p'ower operated relief valves (PORVs) with'a nominal lift setting a.

less than or equal to that shown.in Figure 3.4-4, or b.

The Reactor Coolant System (RCS) depressurized with an RCS vent of greater than or equal to 3 square inches.

APPLICABILITY: MODE 4 uk-a

• ha +e re-atu e < : y 3:! :23 k; n ;; :nc E ", MODE 5 and MODE 6 with the reactor vessel head on.

~

e em: *:

ACTION:

g,%g.

With one PORV inoperable, restore-the inoper bh POR" t a.

OPEnnE LE, s t a * d taahouare4nchvent 'ithiathiN7-cays-or--cencessur4ae and vent-the RCS -through-at-the next E heur;,

b.

With both PORVs inoperable, depressurize and vent the RCS through at least a 3 square inch vent within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.n M

d p'.

In the event either the PORVs or the RCS vent (s) are used to mitigate an RCS pressure transient, a Special Report shall be preoared and submitted to the Co:reaission pursuant to Specffication 6.9.2 within 30 days.

The report shall cescribe the circumstances initiating the transient, and any corrective action necessary to prevent recurrence.the.effect of t E, /

The provisions of Specification 3.0.4 are not applicable.

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V REACTOR COOLANT SYSTEM SURVEILLANCE REOUIREMENTS ll. l 4.4.6 Each PORY shall be der.onstrated OPERABLE b -

a.

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Perfomance of a CHANNEL CA1.IBRATION on the PORV actuation channel at least once per 18 months; and Verifying the PORV isolation valve is open at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

c.

when the PORV is being used for overpressure pmte: tion.

11. '2.

4.4.:.;.

The ROS vent (s) shall be verified to be open at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />" when the vent (s) is being used for overpressure protection.

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REACTIVITY CONTROL SYSTEMS m

BASES 3/4.1.1.3 MODERATOR TEMPERATURE COEFFICIENT (Continued) involved subtracting the incremental change in the MOC associated with a core condition of all rods inserted (most pos,itive MOC) to an all rods withdrawn condition and, a conversion for the rate of change of moderator density with temperature at RATED THERMAL POWER conditions.

This value of the MDC was then transformed into the limiting MTC value -4.0 x 10~4 delta k/k/*F. The MTC valve of -3.1 x 10 delta k/k/*F represents a conservative value (with corrections for burnup and soluble boron) at a core condition of 300 ppm equilibrium boron concentration and is obtained by making these corrections to the limiting MTC value of -4.0 x 10'4 k/k/'F.

The surveillance requirements for measurement of the HTC at the beginning and near the end of the fuel cycle are adequate to confirm that the hTC remains within its limits since this coefficient changes slowly due principally to the reduction in RCS boron concentration associated with fuel burnup.

3/4.1.1.4 MINIMUM TEMPERATURE FOR CRITICAtITY This specification ensures that the reactor will not be made critical with the Reactor Coalant System average temperature less than 541*F.

This limitation is required to ensure 1) the moderator temperature coefficient is within it analyzed temperature range, 2) the protective instrumentation is within its normal operating range, 3) the P-12 interlock is above its setpoint,

4) the pressurizer is capable of being in a OPERABLE status with a steam bubble, and 5) the reactor pressure vassel is above its minimum RT temperature.

NDT 3/4.1.2 00 RATION SYSTEMS The boren injection system ensures that negative reactivity control is available during each mode of facility operation.

The components required to perform this function include 1) borated water sources, 2) charginq pumps, 3) separate flow paths, 4) boric acid transfer pumps, 5) associated heat tracing systems, and 6) an emergency power supply from OPERA 8LE diesel generators.

4fD With the RCS average temperature above MO*F, a minimum of two separate and redundant boron injection system are provided to ensure single functional capability in the event an assumed failure renders one of the flow paths inoperable. The boration capability of either flow path is sufficient to s

SE0VOYAH - UNIT 2 8 3/4 1-2

REACTIVITY CONTHOL SYSTEMS BASES BORATION SYSTEMS (Continued) provide a SHUTDOWN PARGIN from expected operating conditions of 1.6% delta k/k after xenon decay and cooldown to 200 F.

The maximum expected boration capability requirement occurs at E0L from full power equilibrium xenon conditions and requires _5408 gallons of 20,000 ppm borated water from the boric acid storage tanks or 64,160 gallons of 2000 ppm borated water from the refueling water storage tank.

With the RCS temperature below F, one injection system is acceptable without single failure consideration on the basis of the stable reactivity condition of the reactor and the additional restrictions prohibiting CORE ALTERATIONS and positive reactivity changes in the event the single injection system becomes inoperable.

The boron capability required below 200*F is sufficient to provide a SHUTDOWN MARGIN of 1% delta k/k af ter xenon decay and cooldown from 200 F to 140 F.

This condition requires either 835 gallons of 20,000 ppm borated water from the boric acid storage tanks or 9,690 gallons of 2000 ppm borated water from the refueling water storage tank.

I

.The contained water wlume limits include allowance for water not available because of discharge line location and other physical characteristics.

BF The limits on contained water volume and boron concentration of the RWST also ensure a pH value of between 7.5 and 9.5 for the solution recirculated within containment after a LOCA.

This pH band minimizes the evolution of iodine and minimizes the effect of chloride and caustic stress corresion on mechanical systems and components.

The OPERABILITY of one boron injection cy:;te:r. during REFUELING ensures that this system is available for reactivity control while in MODE 6.

3/4.1.3 M0VABLE CONTROL ASSEMBLIES The specifications of this section ensure that (1) acceptable power distribution limits are maintained, (2) the minimum SHUT 00WN MARGIN is main-tained, and (3) limit the potential effects of rod misalignment on associated OPERABILITY of the control rod position indicators is accident analyses.

required to determine control rod positions and thereby ensure compliance with the control rod slignment and insertion limits.

SEQUOYAH - UNIT 2 B 3/4 1-3 Revisea 08/18/87

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 anticipated 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 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

In MODE 3, a single reactor coolant loop provides suf ficient heat removal capability for removing decay heat; however, single failure considerations require that two loops be OPERABLE.

In MODE 4, a single reactor coolant loop or residual heat reno /al (RHR) loop provides sufficient neat reraoval capability for removing decay heat; but single failure considerations require that at least two loops be OPERABLE.

Thus, if the reactor coolant loops are not OPERA 8LE, this specification requires two RiiR loops to be OPERABLE.

In MODE 5 single failure considerations require that two RHR loops be OPERABLE.

('

The operation of one Reactor Coolant Pump or one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity w

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, t2b hnc.&$

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SEQUOYAH - UNIT 2 B 3/4 4-1

REACTOR COOLANT SYSTEM 4 '

BASES PRESSURE / TEMPERATURE LIMITS (Continued)

Values of ART determined in this manner may be used until the results NDT from the material surveillance program, evaluated according to ASTM E185, are available.

The first capsule w.ill be removed at the end of the first core cycle.

Successive capsules will be removed in accordance with the requirements of ASTM E185-73 and 10 CFR 50, Appendix H.

The heatup and cooldown curvesgmust be recalculated when the ART determined from the NOT surveillance capsule exceeds the calculated 4RT f r the equivalent NOT capsule radiation exposure, gg) }\\x hw fewfkbi A'Mf3WW )#

g-hd Allowable pressure -temperature relationships for various heatup and cooldown rates are calculated using methods derived from Appendix G in Section III of the ASME Boiler and Pressure Vessel Code as required by Appendix G to 10 CFR Part 50 and these methods are discussed in detail in WCAP-7924-A.

The general method for calculating heatup and cooldown limit curves is based upon the principles of the linear elastic fracture mechanics (LEFM) technology.

In the calculation procedures a semi elliptical surface defect with a depth of one quarter of the wall thickness, T, and a length of 3/2T is assumed to exist at the inside of the vessel wall as well as at the outside of the vessel wall.

The dimenstions of this postulated crack, referred to in Appendix G of ASME III as the reference flaw, amply exceed the current capabilities of inservice inspection techniques.

Therefore, the reactor operation limit curves developed for this reference crack are conservative and provide sufficient safety margins for protection against non-ductile failure.

To assure that the radiation emorittlement ef fects are accounted for in the calculation of the limit curves, the most limiting value of the nil ductility reference temperature, RTNOT' IS used and this includes the radiation induced shift, aRTNOT, corresponding to the end of the period for which heatup and cooldown curves are generated.

SEQUOYAH - UNIT 2 8 3/4 4-8 1

s, REACTOR COOLANT SYSTEM v

BASES PRESSURE / TEMPERATURE LIMITS (Continued)

Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves are produced as follows. A composite curve is constructed based on a point-by point comparison of the steady-state and fini.te heatup rate data. At any given temperature, the allowable pressure is taken to be the lesser of the three values taken from the curves under consideration.

The use of the composite curve is necessary to set conservative heatup limitations because.it is possible for conditions to exist such that over the course of the heatup ramp the controlling condition switches from the inside to the outside and the pressure limit must at all times be based on analysis of the most critical criterion.

Finally, the composite curves for the heatup rate data and the cooldown rate data are adjusted for possible errors in the pressure and temperature sensing instruments by the values indicated on the respective curves.

Although the pressurizer operates in temperature ranges above those for which there is reason for concern of non-ductile failure, operating limits are provided to assure compatibility of operation with the fatigue arialysis performed

)

in accordance with the ASME Code requirements.

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 Part 50.55a(g) except wh're specific written relief has been granted by the Commission pursuant to 10 CFR Part 50.55a (g) (6) (i).

Components of the rear. tor coolant system were designed prior to the issuance of Section XI of the ASME Boiler and Pressure vessel Code. These components will be tested to the extent practical within the limitations of the original plant design, gecmetry and materials of construction.

mm SEQUOYAH - UNIT 2 8 3/4 4-14

m tage sea.

4*

=

3/4.4.11 OVERPRESSURE ?ROTECTION SYSTEM i

The operability of two PORVs or an RCS vent opening of at least three square inches 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 degrees F.

Either PORV has adequate relieving ce-pability to protect the RC3 from overpressurization when the transient is limited to either:

(1) the start of an idle RCP with a water-solid RCS and a secondary water temperature of the steam generator less than or equal to 50 degrees F above the RCS cold leg temperatures, or (2) the start of c charging pump and its injection into the RCS with letdcwn icolated.

The maximum allowed PORV setpoint for the low temperature overpressure protection (LTOP) system is derived by analysis which mocels the performance of the LTOP system assuming various mass input and heat input transients.

Operation with the PORV setpoint less than or equal to the maximum setpoint ensures that Appendix G criteria will not be violated with consideration for a maximum pressure overshoot beyond the PORV se: point whien 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 specifications require tagout or isolt. tion of all but one centrifugal charging pump while in modes 4, 5, and 6 witn the reactor vessel head installed and disallow restart of an RC? If a steam bubble does not exist in the pressurizer.

The LTOP system setpoints include a 50 degree F allowance for heat transport effects and a 27 degree F allowance for instrument accuracy.

An 800 psig pressure limit prctects the PORV piping fror.i the consequences of a possible water hammer caused by the rapid coening times associated with the PORVs.

r B 3/4 4-15

. -~~--

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TABLE 5.7-1

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f Cor1PONLNT CYCLIC OR TRANSlfNT LIH115 5

CYCIIC OR DESIGN CYCLE O

COf tt'Oril til IRANSIf.NI LIMIi OR 1RANSIENT

[

Ih actor Coolant System 200 heatup cycles at < 100*f/hr lleatop cycle - T.y9 from < 200*F and 200 cooldown cycles at to > 550 f.

< 100*f/hr.

CooIdown cycle - T av from

> 550"f to < 200*f.

200 pressurizer cooldown cycles Pressurizer cooldown cycle at < 200"f/hr.

temperatures from > 650*F to

-< 200*r.

80 loss of load cycles, without

> 15% of RATED lifERHAL POWER to immedie'; turbine or reactor trip.

0% of RAIED IllERMAL POWER.

40 cycles of loss of of fsite Loss of offsite A.C. electrical A. C. electrical power.

power source supplying the onsite ESF Electrical System.

80 cycles of loss of flow in one loss of only one reactor reactor coolant loop.

coolant pump.

400 reactor trip cycles.

100% to 0% of RATED TiiERMAL POWER.

j[

12 spray actuation cycles.

Spray water temperature differential R2 g

> 320*F and < 560*F.

a f.

50 leak tests.

Pressurized to 2485 psig.

E 5 hydrostatic pressure tests.

Pressurized to 3105 psig.

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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-SQN-TS-88-03)

DESCRIPTION AND JUSTIFICATION FOR

' ADDITION OF OPERABILITY ANT) SURVEILLANCE REQUIREMENTS FOR LOW-TEMPERATURE OVERPRESSURE PROTECTION (LTOP) SYSTEM' S

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ENCLOSURE 2 Description of Change The proposed changes to the SQN technical specifications to incorporate-operability and surveillance requirements for the LTOP system consist of five parts.

First, the specific requirements for the LTOP system are added. Second, changes to the boration system requirements are necessary to conform with the administrative controls that support the LTOP system requirements.

Third, new requirements are added to minimize the potential for heat input overpressure transients caused by the restart of a reactor coolant pump in mode 4.

Fourth, changes to the reactor vessel specimen surveil *ance program are proposed. And fifth, cyclic limits are added for low-temperature, water-solid overpressure events.

Specification 3.4.11 is added to specify operability and surveillance requirements for the LTOP system.

Specifications 3.1.2.1, 3.1.2.2, 3.1.2.3, and 3.1.2.4 are modified to reflect the new administrative requirements that limit the number of operable centrifugal charging pumps in node 4 to minimize the potential for and severity of low-temperature overpressure transients.

Specification 3.4.1.3 is modified to reflect new administrative requirements that prohibit restart of a reactor coolant pump unless a steam bubble exists in the pressurizer to minimize the severity of any resulting pressure transient.

Specification 3.4.9.1 is modified to require that the LTOP setpoints be updated as a result of examination and analysis of reactor vessel material irradiation surveillance specimens. Table 5.7-1 is modified to include cyclic limits for low-temperature, water-solid overpressure events.

The bases to specifications 3/4.1.2, 3/4.4.1, 3/4.4.9, and 3/4.4.11 and the index are also modified to reflect LTOP system requirements.

Reason for Change As the result of an NRC inspection to support SQN unit 2 restart, NRC identified the fact that no technical specifications exist for the LTOP system. The LTOP system was required to be installed after licensing as a condition of the license (reference item 2.C(7) of DPR-79 and item 2.C(20] of DPR-77). However, no technical specifications were proposed 'oy TVA at that time, nor were any requested by NRC.

The inspection team noted that NRC Branch Technical Position RSB 5-2 recommends specific technical specifications.

TVA agreed to submit proposed technical specifications for the LTOP system.

LTOP System Description Administrative procedures have been developed to aid the or ator in controlling reactor coolant system (RCS) pressure during low-temperature operation.

However, to provide a backup to the opccator and to minimize the frequency and severity of RCS overpressurization, an automatic system is provided to mitigate the pressure excursion within the allowable pressure limits.

The LTOP system is described in the Final Safety Analysis Report (FSAR), sections 5.2.2.4 and 7.6.7.

The LTOP system

. control circuitry is shown on FSAR figures 5.1-5 (coordinates A9-A12) and'

.1-6 (coordinates C2-C4).

The LTOP system functional logic is shown on FSAR figure 7.6.7-1.

Analyses have shown that one power-operated relief valve (PORV) is sufficient to mitigate the pressure excursions produced by anticipated mass and heat input transients.

However, redundant protection against such an overpressurization event is provided through the use of two PORVs.

The mitigation system is required only during low-temperature operation and was designed to'oe automatically enabled and actuated.

It should be noted that TVA has implemented a temporary alteration to disable the LIOP system when RCS temperature exceeds 350 degrees Fahrenheit (F).

This alteration was mandated by the corrective action to Licensee Event Report (LER) 8S020.

In May 1988, it was determined that the LT0r system could actuate during a postulated main steam line break (MSLB) accident.

During an MSLB, the RCS temperature in the affected loop could decrease below 350 degrees F, thereby automatically arming one train of the LTOP system.

If coincident single failure of a second wide-range temperature channel occurred in another RCS loop, the LTOP system could actuate and cause one or both PORVs on the pressurizer to open.

Opening the pressurizer PORVs would exacerbate the decreasing RCS pressure transient associated with the MSLB and increase the potential for departure from nucleate boiling (DNB) to occur in the core.

This event has not been analyzed as part of the SQN design basis.

The automatic arming feature of this system was designed to ensure the LTOP system would be functional any time the RCS temperature decreased below 350 degrees F.

However, this design did not con:ider the potential consequences of an inadvertent actuation of the LTOP system as a result of a credible design basis event.

TVA is preparing a design change to install a selector switch that will delete the automatic arming of the LTOP system.

As a result of this design change, the operators will be required to manually arm the system when the RCS temperature decreases below 350 degrees F and manually disarm the system when the RCS temperature is greater than 350 degrees F.

Two pressurizer PORVs are supplied with actuation logic to ensure that a redundant and independent ECS pressaac control backup feature is provided during low-temperature operations.

Th.s system provides the capability for additional RCS inventory letdown to relieve tha pressure transient produced by both mass and heat input overpressure events, thereby maintaining RCS pressure within allowable limits.

The basic function of the system logic is to continuously monitor RCS temperature and pressure conditions whenever plant operation is at low temperatures.

An auctioneered system temperature will be continuously converted to an allowable pressure and then compared with the actual RCS pressure.

The system logic will first annunciate a main control board alarm when the measured pressure approaches the allowable setpoint,

, indicating that a pressure transient is occurring.

On a further increase

~

in measured pressure, an actuation signal is transmitted to the PORVs when required to mitigate the pressure transient.

Pressure Transient Analyses Transient analyses were performed to determine the maximum pressure for both postulated mass input and heat input events.

These analyses are described in FSAR section 5.2.2.4.2.

The most severe mass input transient would occur if the letdown flow centrols failed to the zero flow condition while the charging flow

-controls failed to the full flow condition.

This failure mode would result in the maximum charging / letdown flow mismatch event but does not result in the isolation of the RHR system relief valves from the RCS.

Therefore, ad RHR system relief valve would mitigate this transient and prevent an overpressure condition in either the RHR system or the RCS.

4 The most likely way that a charging / letdown flow mismatch would occur is for the 1HR system (and relief valves) to be inadvertently isolated from the RCS oy spurious closure of the RHR-system inlet isolation valves.

Such a spurious closure is credible due to the presence of the automatic closi g signal.

Because this spurious valve closure event causes an RCS pressure transient by stopping the letdown flow and negating the pressure control and concurrently isolating the RHR system relief valves from the RCS, it is considered a "design bases" transient for the LTOP system. The analysis considered the operation of a single charging pump with injection flow rates ranging from 40 gallons per minute to 500 gallons per minute; the lower charging rates being associated with high-pressurizer-level operations.

Any additional mass injection, which could potentially produce higher injection rates than the overpressure mitigation system is capable of mitigating, is prevented trom occurring by locking out the appropriate valves or pumps.

If the RHR system is inadvertently isolated from the RCS by closure of the isolation valves (as by a spurious operation of the auto-close interlock) while the plant is water solid and in mode 4, technical specification 3.4.1.3 requires that a reactor coolant pump be restarted if an RHR loop is not returned to service promptly.

During the potential delay period, a temperature asymmetry in the reactor coolant loops (because of the continued input of cold seal injection water) could develop and not be apparent to the operator.

Then when the reactor coolant pump is restarted, a pressure transient could occur if the RCS is water solid.

This particular heat input transient is therefore considered a "design bases" transient for the LTOP system because the RHR system relief valves would not be available to mitigate the transient because the RHR system isolation valves are closed.

Based on the above discussion, most of the mass input and heat i.1put transients will be mitigated by relief valves in the RHR system.

However, for the remote cases that occur when the RHR system has become isolated l

frcm the RCS (most likely caused by the spurious closure of the RSR system inlet valves by the auto-close interlock), the LTOP rystem may be l

o

l

. called upon to mitigate certain increasing pressure transients.

Specifically, the LTOP design bases transients are defined as:

(1) the mass input transient caused by a charging / letdown flow mismatch after the termination of letdown flow and (2) the heat input transient caused by the restart of a reactor coolant pump when the RCS is water solid and the RHR system is not open to the RCS (i.e., the isolation valves are closed, and i

a temperature asymmetry exists within the RCS because of the continued injection of cold seal injection water).

The mass input analysis considered constant mass injection flow rates between 40 gallons per minute and 500 gallons per minute.

Four parameters were developed for each flow rate over a wide range of PORV pressure setpoints.

These four parameters are:

1.

Setpoint pressure overshoot (aP over).

2.

Peak RCS pressure (P max).

3.

Setpoint pressure undershoot (oP under).

4.

Minimum RCS pressure (P min).

The heat input analysis considered inadvertent startup of an RCS pump with temperature asymmetry between the RCS and steam generator (heat source).

The RCS was assumed to be in a water-solid condition.

The same four parameters as those used in the mass input analysis were developed for a range of steam generator temperatures (120 degrees F to 350 degrees F) and PORV pressure setpoints.

These four parameters were used to develop graphical algorithms for both the mass input and heat input cases. A range of permissible setpoints was then selected for each PORV.

Each of the two valves has different pressure setpoints versus RCS temperature such that only one valve will open and mitigate the transient (the second valve operates only if the first valve fails to open on command).

Both heat input and mass input analyses took into account the single failure of one PORV; therefore, only one PORV was assumed to be available for pressure relief.

The evaluation of the transient results concludes that the allowable limits will not be exceeded and therefore will not constitute an impairment to vessel integrity and plant safety.

The LTOP system setpoints have been developed to ensure that the requirements of 10 CFR 50, Appendix G, are met for the postulated heat input and mass input design basis transients.

In addition, the LTOP system setpoints have been developed to ensure that actuation of the LTOP system will occur only at pressures low enough to preclude the need for an evaluation of the system after each water-solid system actuation. An 800-pound-per-square-inch-gauge (psig) pressure limit was imposed to protect the PORV piping from the consequences cf a possible water hammer, given the rapid opening times characteristic of the solenoid-operated valves.

f,

I 1

Table 1 to this enclosure contains the individual PORV lift setpoints developed for SQN. These setpoints ensure that the existing plant design will not require reanalysis for at least ten challenges to the LTOP system I

over the life of the plant.

The large difference between the Appendix G curve and the setpoints at higher temperatures is also due to heat transport effects. The design basis for the LTOP system considers a maximum difference'of 77 degrees F between the RCS coolant and the vessel (50 degrees F due to heat transport plus 27 degrees F instrument accuracy). Therefore, if the measured RCS temperature is 275 degrees F, the LTOP system must consider an Appendix G limit corresponding to 198 degrees F when setpoints are being developed.

I By comparison, it can be seen that the unit 1 Appendix G curve is more restrictive than that for unit 2.

Application of the setpoints developed for unit 1 to unit 2 is, therefore, conservative with respect to unit 2.

Justification for Changes Reactivity Control Systems Specifications 3.1.2.1, 3.1.2.2, 3.1.2.3, and 3.1.2.4 are modified to reflect the fact that only one centrifugal charging pump will be allowed to be capable of operating in mode 4.

This change is necessary to implement actions to minimize the potential for and severity of mass input, low-temperature overpressure events.

As a result of these actions, the requirements for redundant boration capability in mode 4 cannot always be met.

Consequently, the reactivity control system requirements are changed to only require redundant capability in modes 1, 2, and 3.

The action statements for specifications 3.1.2.2 and 3.1.2.4 are also modified to reflect the change in applicability. These changes are consistent with the standard technical specification because the arming temperature for the SQN LTOP system is 350 degrees F.

These changes also conform to the mass input pressure transient analysis that was performed using the charging flow of a single centrifugal charging pump.

The changes to technical specifications 3.1.2.1, 3.1.2.2, 3.1.2.3, and 3.1.2.4 involve a reduction in the redundancy requirements for boration capability in mode 4.

However, this reduction is necessary to incorporate requirements that minimize the potential for and severity of mass input, low-temperature overpressure events by limiting the number of centrifugal charging pumps that can automatically inject into the RCS.

NRC has acknowledged the greater importance of LTOP system requirements and has structured the standard technical specification requirements accordingly.

TVA is adopting the standard requirements, that presumably prov;de a greater margin of safety.

An additional change is made to unit 2 specification 3.1.2.4 to delete an outdated footnote.

The change is administrative in nature and is being made because other changes are proposed on the same page.

. RCS - Hot Shutdown Specification 3.4.1.3 is modified to add new limitations on the restart of a reactor coolant pump. A footnote is added to require that reactor coolant pump restart only be allowed when a steam bubble exists in the pressurizer.

This change will minimize the severity of any potential heat l

input, low-temperature overpressure event because the design basis heat input pressure transient analysis was performed for a water-solid system.

1 This change is consistent with the reactor coolant pump restart restrictions listed in FSAR sectica 5.2.2.4.4 and accepted by NRC in response to FSAR question 5.25.

The change is similar to the standard technical specifications and, in effect, functionally equivalent in miniaizing the severity of any heat input, low-temperature overpressure event.

RCS - Pressure / Temperature Limits Surveillance requirement 4.4.9.1.2 is modified to require that the LTCP system setpoints are updated on the same schedule as the heatup and cooldown pressure and temperature li.aits.

The LTOP system setpoints are based, in part, on the reactor vessel material properties.

The change to surveillance requirement 4.4.9.1.2 will ensure that the LTOP system setpoints are updated whenever changes in the reactor vessel material properties are detected by examination of irradiation, surveillance specimens. As a result, the LTOP system will always ensure that the requirements of 10 CFR 50, Appendix G, will be met during temperature operation.

RCS - LTOP System Specification 3.4.11 is added to identify operability and surveillance requirements for the LTOP system. The specification requires two operable PORVs with setpoints derived frcm the heat input and mass input transient analyses.

The setpoints are a function of temperature and are shown on the new technical specification figure 3.4-4 Up to a temperature of approximately 280 degrees F, the LTOP setpoint is based on the requirements of 10 CFR 50, Appendix G, for the design basis heat input and mass input transients.

Between 280 degrees F and 380 degrees F, the setpoint is based on a more restrictive TVA limit imposed to protect the PORV piping from the consequences of a possible water hammer caused by l

rapid opening times characteristic of the solenoid-operated valves. Only the higher setpoints for the PORVs are shown in the technical specifications.

Operation of either or both PORV at these setpoints will ensure that the requirements of 10 CFR 50, Appendix G, are met.

The actual setpoints used at SQN will have one PORV setpoint curve set at less than the technical specification limit to provide staggered valve operation.

The technical specification is applicable during modes 4, 5, and 6 when the reactor vessel heaa is on.

This requirement is consistent with the design base analyses and the LTOP arming temperature of 350 degrees F.

-7 The proposed changes are consistent with the design basis low-temperature overpressure transient analyses.

The changes are also consistent with the standard technical specifications with five exceptions.

First, the surveillance requirements to verify that both safety-injection pumps and one centrifugal charging pump are incapable of automatically injecting into the RCS have been included with the LTOP system requirements. The standard technical specifications have these same types of requirements in the specifications for the boration system and emergency core cooling system.

The TVA proposal is considered a human factors improvement over the standard technical specifications because the surveillance requirements and limiting conditions for operation for the LTOP system are functionally grouped.

Second, an action statament is added to address the condition for when the pumps required to be incapable of automatically injecting into the RCS are actually capable of automatic injection.

The standard technical specifications have no corresponding action statement because the action statements associated with the boration system and emergency core cooling system do not address LTOP requirements. The TVA proposal is considered a human factors improvement over the standard technical specifications because specific and correct remedial actions are specified for potential plant conditions that degrade the protection provided by the LTOP system.

Third, options are provided for demonstrating that the required pumps are incapable of automatic injection into the RCS.

In addition to the action specified in the standard technical specifications for the pump motor circuit breakers, IVA has proposed two additional methods:

pump controls in the pull-to-lock position and system isolation with motor-operated valves.

The first alternative is the preferred method because it renders the pump incapable of automatic actuation but keeps it available to the operator for manual action in an emergency situation. The second

+

alternative is necessary to permit pump testing, troubleshooting, and maintenance activities that cannot be performed without the pump running.

Both alternatives satisfy the intent of the requirements by preventing automatic injection into the RCS.

The proposed alternatives are l

consistent with the administrative controls listed in FSAR section 5.2.2.4.4 It is recognized that reactor coolant pump seal flow nust be maintained at all times.

In order to preserve seal flow while switching charging pumps for maintenance or test activities, the preferred method is l

to isolate normal charging and injection pathways, start the second j

charging pump, and allow it to stabilize before stopping the first pump.

Once the system has stabilized, the first pump may then be stopped and l

placed in pull-to-lock and the normal charging pathway restored.

This l

process will ensure continuous reactor coolant pump seal flow.

The TVA l

oroposal is considered an operati.3nal improvement over the standard i

because it provides additional capabilities for emergency situations and permits necessary plant activities and testing to occur without increasing the potential for and severity of the cass input, low-temperature overpressure events.

t

. Fourth, an alternate limiting condition for operation and action statements are provided for the situation when the LTOP system or the PORV(s) is not operable. These alternate actions put the RCS in a condition that is not vulnerable to low-temperature overpressure events if plant conditions and systems cannot be restored to the specified conditions within the allotted times.

These alternate actions, which are summarized in the bases to the standard technical specifications, provide operational flexibility as well as a viable alternative to placing the plant in a transient condition and introducing a cooldown.

Fifth, surveillance requirements are proposed to test the PORVs and to verify that other plant conditions exist when required to support low-temperature overpressure protection. A special test provision has been incorporated to allow entry into mode 4 before requiring performance I

of an Analog Channel Operational Test.

This was necessary to preclude the removal of common instrumentation from service that is required for both the reactor vessel level indicator system (RVLIS) and the LTOP system.

In addition, portions of the Analog Channel Operational Test disable the automatic arming signal of the redundant PORV, which makes both PORVs inoperable, thereby introducing an 8-hour action requirement. A footnote has been added to the 8-hour action requirement to allow a more reasonable timeframe for performing this test.

In summary, the proposed change to incorporate the LTOP system within the SQN technical specifications will ensure that the proper controls exist for overpressurization protection while operating at low RCS temperatures.

The new LTOP specification and the provisions contained in the other revised specifications (i.e., reactivity control and reactor coolant system) remain consistent with standard technical specifications and the design basis.

TVA's proposed enhancements to add (1) a surveillance requirement to verify pumps are incapable of automatic injection into the RCS when their automatic operation cannot be permitted, (2) an action statement to address conditions when the LTOP system is inoperable or the automatic injection is not appropriately disabled. (3) additional methods for rendering pumps incapable of automatic injection, and (4) a special test provision to allow performance of an Analog Channel Operational Test provide improvements over the standard technical specifications and allow for operational flexibility and human factor considerations.

i t

ENCLOSURE 2 TABLE 1 SQN UNITS 1 AND 2 COLD OVERPRESSURE MITIGATION SYSTEM (COMS) PORV SETPOINTS Indicated PORV Lift Pressurizer PORV 1 RCS Temp ('F)

Setpoint (psig) 70.0 450.0 130.0 450.0 190.0 440.0 230.0 485.0 250.0 550.0 270.0 630.0 285.0 710.0 380.0 710.0 460.0 2,350.0 Indicated PORV Lift Pressurizer PORV 2 RCS Temp ('F)

Setpoint (psig) 70.0 470.0 190.0 470.0 210.0 490.0 230.0 520.0 250.0 590.0 270.0 670.0 285.0 755.0 380.0 755.0 450.0 2,350.0 COMS arming temperature = 350 'F NOTE:

The use of unit 1 setpoints is appropriate for unit 2 because the setpoints are conservative relative to the unit 2 Appendix G limits.

4

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s a

. ENCLOSURE ~3 y

PROPOSED TECHNICAL SPECIFICATION CHANGE.

SEQUOYAH-NUCLEAR PLANT UNITS l'AND 2 DOCKET NOS. 50-327 AND 50-328 (TVA-SQN-TS-88-03):

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS

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.i ENCLOSURE 3 Significant Hazards Evaluation l

l

'TVA has evaluated the proposed technical specification change and has determined that it does not represent a significant hazards consideration based on criteria established in 10 CFR 50.92(c).

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.

The low-temperature overpressure events had been previously evaluated in the FSAR. The addition of

-operability and surveillance requirements for the LTOP system is administrative in nature because it incorporates existing requirements into the technical specifications.

Conforming changes to technical specifications 3.1.2.1, 3.1.2.2, 3.1.2.3, 3.1.2.4, 3.4.1.3, 3.4.9.1, and 5.7 are necessary to fully integrate the LTOP system requirements. The addition of a special test condition for surveillance requirement 4.4.ll.l.(a) ensures that instrumentation common to both the RVLIS and LTOP systems remains in service as required.

The deletion of an outdated footnote to unit 2 specification 3.5.2 is purely editorial in nature. No new hardware or operating practices are introduced by these changes.

Consequently, the probability or consequences of accidents previously evaluated are unchanged.

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

The low-temperature overpressure events had been previously evaluated in the FSAR. The addition of operability and surveillance requirements for the LTOP system and conforming changes to other technical specifications is administrative in i

nature.

The deletion of an outdated footnote to unit 2 specification.

l 3.5.2 is purely editorial in nature.

No new hardware changes are l

introduced by these changes. The addition of a special test condition for surveillance requirement 4.4.11.1.(a) ensures that instrumentation common to both the RVLIS and LTOP systems remains in service as required. Consequently, the possibility of a new or different kind of accident from any previously analyzed is not created.

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

The addition of operability and surveillance requirements for the LTOP system will increase the margin of safety by affording a higher level of administrative controls over LTOP system operability.

The conforming changes to technical specifications 3.4.1.3, 3.4.9.1, and 5.7 will increase the margin of safety by also affording a higher level of administrative controls over related LTCP requirements.

The changes to technical specifications 3.1.2.1, 3.1.2.2, 3.1.2.3, and 3.1.2.4 involve a reduction in the redundancy requirements for boration capability in mode 4.

However, this reduction is necessary to incorporate requirements that minimize the potential for and severity of mass input, low-temperature overpressure events by limiting the

-2_

f number of centrifugal charging pumps that can automatically inject into the RCS. The addition of a special test condition for surveillance requirement 4.4.ll.l.(a) ensures that instrumentation common to both the RVLIS and LTOP systems remains in service as required. NRC has acknowledged the greater importance of LTOP system requirements and has structured the standard technical specification requirements accordingly. TVA is adopting the standard requirements, and some additional requirements, that presumably provide.a greater margin of safety. The deletion of an outdated footnote is administrative in nature and has no effect on the margin of safety.

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5 ENCLOSURE 4 PROPOSED TECHNICAL SPECIFICATION CHANGE SEQUOYAH :,'UCLEAR PLANT UNITS l AND 2 DOCKET NOS. 50-327 AND 50-328 (TVA-SQN-TS-88-03)

FSAR CHANGES FOR THE LTOP SYSTEM

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SNP A detailed functional description of the process equipment g*

a s sociated with the high pressure trip is provided in, Reference i

4.

The upper limit of overpressure protection is based upon the positive surge of the reactor coolant produced as a result of turbine, trip u n d e r. full load, assuming the core continues to produce full pow e r.

The self-actuated safety valves are sized on the basis of steam flow from the pressurizer to accommodate this surge at a setpoint of'2500 psia and a total accumulation of 3 percent.

Note that no credit is taken for the relief capability provided by the power operated relief valves during this surge.

The RCS design and operating pressure toge ther with the safety pow e r relief and pressurizer spray valve setpoints and the protection system'setpoint pressures are listed in Table 5.2.2-2.

System component s, who se design pressure and temperature are less than the RCS design limits are provided with overpressure protection devices and redundant isolation means.

System discharge from overpressure protection devices is collected in the pressurizer relief tank in the RCS.

Isolation valves are provided at all connections to the RCS.

5.2.2.4 RCS Pressure Control During Low Tempera _ture Operat_lon Administrative procedures have been developed to aid the operator

{

in controlling RCS pressure during l ow temperature operation.

H ow ev e r, to provide a back-up to the operator and to minimize the frequency of RCS overpressurization, an automatic system is provided to mitigate the pressure e xcursion wi thin the allowable pressure limits.

Analysis have shown that one PORY is sufficient to mitigate the pressure excursions produced by anticipated mass and heat input i transients.

How ev e r, redundant protection against such over-pressurization event is provided through use of two PORYs.

The mitigation system is required only during low temperature water solid operation and is' automatically enabled and actuated.

5.2.2.4.1 System Operation Tw o pressurizer power operated relief valves are both supplied I

with actuation logic to ensure that redundant and independent RCS pressure control back-up feature is provided for the operator during low temperature operations.

This system provides the capability for additional RCS inventory letdown to relieve the pressure transient produced by tha overpressure events thereby maintaining RCS pressure, within a ll ow a bl e limits.

Refer to Sections 5.5.7, 5.5.10, 5.5.13, 7.6.7, and 9.3.4 for additional information on RCS pressure and inventory control during other modes of operstion.

5.2-42

{

_----_--------------------J

c SNP-3 2

i The hasie function of the sy st em logic is to continuously monitor RCS temperature and pressure conditions whenever pl ant operation is at low temperatures.

An auctioneered sy st em temperature w il l be co n tin uo u s ly converted to an all owa bl e pressure and th e n compa red to the actual R.C S pressure.

This sy s t em l ogi c w ill first annunciate a m ain. board al a rm whe nev er th e me a s ure d pressure approaches within a pr ede te rm ine d amount of the allowable pressure, th e r e by indica tin g a pressure transient is occurring.

On a f ur th e r' i nc r e a se in m ea sured pre s sur e, an a ctua tion s i g nal is transmitted to the power operated relief valves when required t o mitiga te the pressure transient.

5.2.2.4.2 Evaluation of Low Temocrature Overoressure Transients

.P r e s s u r e Transient Analysis ASME Section III, Appendix G, establishes guideline s and upper limits for RCS Pressure primarily for low temperature conditions.

i

(<350*F).

The mi tiga tion system discussed above sa ti sfie s th e se conditions as discussed in the following p a r a gr a ph s.

Tra n sie n t a naly se s were pe rf orme d to determine the mazimum pressure for the po st ul a ted m a s s input and heat input events.

Th e ma s s input pressure transient whi ch w oul d oc cur mo s t f reque ntly during the co ur se of normal pl ant opera tion would involve letdown i sola tion with the charging pumps d el iv e rin g an input less than or eq ual to 120 gpm.

However, the mass input a n a ly si s was pe rf orme d a s s umin g letdown i sola tion wi th one Fkt ch a rging p ump operating in the configuration pr o du ci ng n a zimum pessNb d el iv e ry ratesg Thi s more unlikely eve n t and more severe eco nf i gur a ti on was cho se n to provide additional sy st em fl e xi bil i ty for pressure control.

Mass inj e c tion eve nt s, which co ul d potentially produce higher inj ection rates than the overpressure mi tiga tion sy s t em is ca pa bl e of mi ti ga tin g, are pr eve nte d from occurring by locking out the r eq ui red v al ve s and pumps.

The 2

a dmi ni s t ra t ive controls are i n c l u d e_d in the plant procedures.

~fo k RCS sg(se.< freswe- (Wy 40 Scogym). )

The a c a7 input - transtent a naly si s is performed over the entire RCS sh ut dow n t em pe ra t ur e range.

This a na ly si s assumes an i na dv er t e nt start of a reactor coolant pump with a 5 0* F m i sm a t ch between the RCS and the temperature of the hotter secondary side g e n e r a t or sj wit ( tkc RC$ in 4 wah/ Colid comllfA, of the steam Doth heat input and mass input a naly se s took into account the sin gl e f a ilure of one PORV th er ef or e, the only one Power Ope r a t ed Rel ief V al ve (PORV) was a s s um e d to be av ail abl e for pressure relief.

The ev al ua tio n of the tran sie n t results concludes that the all ow abl e limits w il l not be exceeded and l

therefore w il l not co n st i tut e an i mpa i rm e n t to vessel integrity and plant safety.

I l

l 5.2-43

SNP-2 5.2.2.4.3 Operatina Basis Earthquake (OBE) Evaluation A fl ui d sy s t em s evalua tion has been pe rf o rae d considering the po t e nt i al for overpressure transients f oll ow in g an Ope ra ting Dej6f

~

e most 1 ely f a 11 u r4 o f th old pgR,VS Me Basis Ba r thq na ke (OBE).J s

r pit.

ON 0 opWd.h The pow e r-ope r a t e d relief val ve s h av e been designed in accordance yg}yGS.

with the ASME code to provide integrity required for the reactor coolant pressure bo un da ry.

They have been analyzed for accident l oa ds and for l oa ds impo se d by se i sm ic events a nd h ave been sh ow n t o m ain t ain th ei r int e g ri ty.

Th er ef or e, the PORY s w ill be av ail abl e to provide pressure relief f oll ow ing an OB E.

5.2.2.4.4 Administrative Procedures Although the sy s t em de s c ri be d in Seetion 5.2.2.4.1 is in s t alle d to mitiga te the pressure excursion to address the allowable pressure l imi t s, a dmi ni s t ra t iv e procedure s are re comme n de d for minimiz ing the po t e n ti al for any transient that co ul d actuate the overpressure relief sy st em.

The following discussion h i ghl igh t s these procedural co nt rol s, listed in hierarchy of th ei r f un c t i on in mi tiga ting RCS c ol d ov erpre s s u riza tion transients.

Of primary importance is the ba sic me thod of ope ra tion of t h' e

(..

pl a n t.

No rm al plant opera tin g pro ce dure s ma zimiz e the use of a pre s suriz er cushion (steam bubble) during periods of low pressure and low tempe ra tur e operation.

This cush ion w ill dam pe n the plants response to po t e n ti al transient ge ne ra tin g inputs, providing easier pressure control with the sl ow e r response rates.

An adequate cushion subs t a nt i ally reduces the se v e ri ty of some potential pressure tra n sie nt s such as reactor coolant pump induced heat input and slows the rate of pressure rise for o th e r s.

In conj un c tion with the previously dis cus se d al a rm s, this provides r ea sonabl e assurance that most po t e nt i al transients can be t e rmi na t ed by o pe ra t or a e tion before the overpressure relief sy st em actuates.

How ev e r, for th o se modes of ope r a tion when water solid operation may still be possible, procedures f ur th e r hi g hli gh t precautions that m inimiz e the po t e nt i al for de v el opin g an 2

ov erpre s s u riz a tion t ran sie n t.

Th e following specific recommendations are made:

?

5.2-44

SNP-2 s

1.

Do not isolate th e r e s i d'u a l heat removal inl e t lines from the reactor coolant loop uniess the charging pumps are stopped.

This precaution is to assure there is a relief pa th from the reactor coolant loop to the re sid ual heat removal s u c tio n line relief valves when the RCS is at low pressure (less than 500 psi) and i s wa t er solid.

2.

Whenever the pl an t is water solid and the rea ct or coolant pressure is bein g m aint ained by the low pressure letdown co n t rol valve, le tdow n flow must bypass the no rm al l e t dow n orifices, the valve in the by pa s s line should be in the full open po si tion.

During this mode of operation, all three l e t dow n orifices must al s o remain open.

3.

If all reactor coolant pumps have st oppe d for more than 5 min ute s during plant heatup, and the reactor coolant tempe ra ture is greater th an the charging and seal inj ection g /

water t em p e ra tur e, do not a t t em pt to restart a pump unless a jg steam bubble is formed in the pressurizer.

This pr e ca ut i on will minim iz e the pressure transient whon the pump is started and the cold vator pr ev iously inj ect ed by th e charging pumps is circulated through the warmer reactor coolant ceuponents.

The steam bubbl e w ill accommodato the resultant expansion a: the cold wa te r is ra pi dly w a rme d.

4.

If all reactor coolant pumps are stopped and the RCS is being cooled down by th e re sidual heat exchangers, nonuniform

\\ /

t em pe r a tur e distribution may occur in the reactor coolant

/\\

loops.

Do not a t t em p t to restart a reactor coolant p um p unl e s : a steam bu bbl e is formed in the pr e s sur iz er.

5.

During plant c o ol dow n, all s t e am generators should be connected t o th e steam header to a s sure a uniform cool down of the reactor coolant loops.

6 At least one react or coolant pump must remain in service until the reactor coolant temperature is reduced to 160'F.

These special precautions back-up the no rm al operational mode of ma simiz in g periods of steam bubble o pe ra tion so that cold overpressure transient prevention is continued during periods of t ra n si tional o pera tion s.

The spocific pl ant operating instructions for pl a n t co ol do w n and the surveillance in s truc tion s for ECCS te s tin g include a dm ini s t r a tiv e co nt rol s and proce dur e s to preclude overpressure t ra n sie nt s.

These in clu de :

s.4-43

4'

~

dar-3 c

L..

To pravant. inadvertent ECCS actuation during plant heatup

['

and cooldown, procedures require blocking portions of the safety injection signal actuation logic. The pressurizer low pressure and level actuation signal is blocked when RCS pressure is.

reducted to 1900 psig.

The high steem flowcoincident with'1c,w steam line pressure or lo-lo T vg actuation signal is blocked a

when T is approximately 540*F.

avg 2.

Thc Upper Head Injection (UHI) accumulator isolation valves are closed and gagged and power removed from the gag motor when RCS pressure is_ reduced to 1900 psig.

3.

The Safety Injection System (SIS) low pressure accumulator isolation valves are closed and tagged with power removed when RCS pressure is reduced below 1000 psig and RCS temperature is less than 425'F.

4.

The Safety Injection pumps are locked out when the RCS N

temperature is reduced to less than 350*F.

5.

The non-operating centrifugal charging pump will be locked out when the RCS temperature is reduced to less than tSi'.F.

350 6.

The procedures for periodic ECCS pump performance testing will be to test the pumps during normal power operation or at i

hot shutdown conditions. This precludes any potential for developing a cold overpressurization transient.

Should cold shutdown testing of the pumps be desired, the test will be done when the vessel is open to atmosphere again precluding overpressurizatbn potential.

If cold shutdown testing with the vessel closed is necessary, the procedures will require charging pump discharge valve

})/

closure and RHRS alignment to both isolate potential ECCS pump k

input and to provide backup benefit of the RHRS relief valves.

7.

"S" signal circuitry testing (safety injection actuation C,

testing), if done during cold shutdown, requires RHRS alignment and non-operating charging pump power lockout to preclude developing cold overpressurization transients.

The above procedural recommendations covering normal operations with a steam bubble. transitional operations where potentially water solid, followed by specific testing operations provide in-depth cold overpressure preventions or reductions, a,ugmenting the installed overpressure relief system.

5.2-46 i

.,"r t '1' SNP-5 operator. The design modification includes easy access, clear protective covers attached to the c.ain control board panel over each respective control room switch except FCV-63-1.

The operator would be required to open this protective cover before he operates the control switch.

~For FCV-63-1, 67, 80, 98 and 118 operating instructions specify the removal of valve actuator power during normal operation.

After removal of power, redundant valve position indication is provided to the operator in accordance with Branch Technical Position (BTP) ICSB 18.

7.6.7 Interlocks for RCS Pressure Control During Low Temeerature Operation The basic function of the RCS overpressure mitigation system during low temperature operation is discussed in Section 5.2.2.4.

As noted in Section 5.2.2.4, this pressure control system includes automatic actuation logic for the two Pressurizer Power Operated Relief Valves (PORVs).

The function of this actuation logic is to continuously monitor RCS temperature and pressure conditions, with the actuation logic only unblocked when plant operation is at a temperature below the arming setpoint.

The monitored system temperature signals are processed to generate the reference pressare limit program which ic compared to the actual measured system pressure. This comparison will provide an actuation signal to an actuation device which will cause the PORV to automatically open if necessary to prevent pressure conditions from exceeding allowable limits. See Figure 7.6.7-1 for the block diagram showing the interlocks for RCS pressure control during low temperature optration.

As shown on this figure, the station variables required for this interlock are channelized as follows:

1.

Protection Set I Wide Range RCS Temperature from dot Legs and Cold Legs.

a.

2.

Protection Set II Wide Range RCS Temperature from Hot Legs and Cold Legs.

a.

3.

Protection Set III Wide Range RCS System Pressure (PT-68-68, W no PT-405).

a.

4 Non-Olvisional Wide Range RCS system pressure (PT-68-64, no vendor ID) a.

The wide range temperature signals, as inputs to the Protection Sets I and II, continuously =enitor RCS temperature conditions 7 A.3

• c SNP-S

(

whenever plant operation is at a temperature below the arming y

setpoint.

In Protection Set I, the existing RCS hot leg and cold leg J

wide range temperature channels will supply through an isolation device continuous analog input to an auctioneering device, which is located in the Process' Rack of Control Rack Group 3.

The lowest reading will be selected and input to a function generator which. calculates the reference pressure limit program considering the plant's allowable pressure and temperature limits. Also available from Protection Set III is the wide range RCS system pressure signal which is sent through an isolation device to Control Rack Group 3.

The reference pressure from the function generator is. compared to the actual RCS system pressure monitored by the wide. range pressure channel. The error signal derived from the difference between the reference pressure and the actual measured pressure will first annunciate a main board alarm whenever the actual measured pressure approaches, within a predetennined amount, the reference pressure.

On a further, increase in measured pressure, the error signal will generate an annur.ciated actuation signal. The actuation signal available from Control Rack Group 3 will control the train A PORV (PCV 340A) whenever a temperature-dependent permissive from Control Group 2 is present. The temperature-dependent permissive to the FORVs actuation device effectively disarms (blocks) the actuation signal at temperature greater than the range of conce rn.

Above this permissive interlock, the normal pressure protection system (as discussed in Section 5.2) ensures that the

{

system pressure temperature limitations are not exceeded.

This permissive prevents unnecessary system actuation at normal RCS operating conditions as a result of a failure in the process sensors.

The monitored generating station variables that generate the actuation signal for the Train B PORV (PCV-68-334) are processed in a similar manner.

In the case of the train B PORV, the reference temperature is generated in Control Rack Group 2 from the lowest auctioneered wide range hot leg and cold leg temperature.

The auctioneering device derives its inputs from the RCS wide range temperature in Protection Set II and the actual measured pressure 5

signal is available from PT-68-64 Therefore, the generating station variables used for the train B PORV are derived from a Protection Set that is independent of the Sets from which generating station variables used for the train A PORV are derived.

The error signal derivation itself used for the actuation signals is available from the Control Group.

Upon receipt of the actuation signal, the actuation d:"ice will automatically cause the PORV to open. Upon sufficient hC9 inventory letdown, the operating RCS pressure will decrease, clearing the actuation signal. Removal of this signal causes the PORV to close.

f 7.6-10

SNP-3 7.6.7.1 Analysis of Irterlocks Many critoria prosanted in IEEE 279.1971 and IEEE 338-1971 standards do not apply to the interlocks for RCS pressuro control during low temperature operation because.ae interlocks do not perform a protective function but rather provide automatic proscuro control at low temperatures as a backup to the oporator.

However, although IEEE-279 critoria do not apply, some advantages of the dependability and bonofits of an IEEE-Std-279 design havo occurred by including the pressure and temperature signal oloments as noted above in the Protoction Sots and by organizing the control of the two PORVs into dual channels wherevor practical.

Either of the two PORVs can accomplish the RCS pressure control function.

The design of the low temperaturo interlocks for RCS pressure control is such that pertinent features include:

1.

No credible failur9 at the output of the protection set racks, after the output loaves the racks to interface with the interlocks, will provent the associated protection system channel from performing its protective function because such outputs that leave the racks go through an isolation dovice as shown in Figure 7.6.7-1.

2.

Testing capability for elements of the interlocks within (not external to) the overpressure mitigation system is consistent with the testing principles and methods discussed in Section 7.2.2.2.3.

3 3.

A loss of of fsite power will not defeat the provisions for an electrical power sourco for the interlocks because those provisions are through onsito power which is described in Sections 8.3 and 7.6.1.

7.6.8 Liquid Level Monitoring Systems Two types of level measuromont systems used insido containmont are described below along with the particular application:

1.

An open coluen reference leg is used for steam generator (SC) level measurement.

The instrument is connected to the SC liquid by a condensato chamber at the upper tap. The 11guld in the reference log will be at ossentially ambient temporature.

7.6-11

o I

s 9

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