ML20128M870

,. =,

HATCH UNIT 1 3.6-18 Amendment No. 195

. - =.

SLC System 3.1.7 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.1.7.5 Verify the concentration of sodium 31 days i

i pentaborate in solution is within the Region A limits of Figure 3.1.7-1.

M-3 Once within.

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after i

water or sodium

{

pentaborate is r

ll added to solution M

i Once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after i

solution -

temperature is restored within 4

the Region A i

(

limits of '

Figure 3.1.7-2 SR 3.1.7.6 Verify each SLC subsystem manual and power 31 days operated valve in the flow path that is not locked, sealed, or otherwise secured in position is in the correct position, or can be aligned to the correct position.

4 i

I i

SR 3.1.7.7 Verify each pump develops a flow rate In accordance i

2 41.2 gpm at a discharge pressure with the a Hei psig.

Inservice l

p)

Testing Program a

SR 3.1.7.8 Verify flow through one SLC subsystem from 18 months on a pump into reactor pressure vessel.

STAGGERED TEST i

BASIS (continued) d HATCH UNIT 2 3.1-23 Amendment No. 138

'S/RVs 3.4.3 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.3 Safety / Relief Valves (S/RVs) to 8 LCO 3.4.3 Thesafetyfunctionogli S/RVs shall be OPERABLE.

n APPLICABILITY:

MODES 1, 2, and 3.

.s hr,

cy a

e ACTIONS

' N^W

~

T CONDITION REQUIRED ACTION COMPLET!0NtTIME m

n x

x h ','

naa S/RV ineper:ble.

Ari Restore-the-S/RV-to days

_a 4pERABL4-status.-

.s. s 4

M A

y, s,

S.

S gir Be in MODE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months

~'

aeeecistM Cer;1etion*

Tim ef-Condition-A-

/g%

8t

+"

-net-met-f4 +

Be in MODE 4.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months ..,g

-fE-M, '2.,

A, Two or more S/RVs j

1 HATCH UNIT 2 3.4-7 Amendment No. 135

1

[

S/RVs 3.4.3 SURVEILLANCE REQUIREMENTS

(

SURVEILLANCE FREQUENCY I

SR 3.4.3.1 Verify the safety function lift setpoints In accordance j

of the S/RVs are as follows:

with the Inservice Number of Setpoint Testing Program S1RL% Insto) -

f/

usb +3%

4 1120+33 6 4 130 -*-33. 9 -

F 3

1140+3 Following testing, lift settings shall be within

1%.

1 SR 3.4.3.2

NOTE---------------

Not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months i

after reactor steam pressure and flow are adequate to perform the test.

Verify each S/RV opens when manually 18 months actuated.

i i

i f

l

(

i HATCH UNIT 2 3.4-8 An,endment do.138

Qg d5lWiW;-

ECCS;-0perating &

" WR3;5.1't ^',

s n,y,

tie 3.5 EMERGENCYCORECOOLINGSYSTEMS(ECCS)ANDREACTORCOREISOLATIONCg0 LING.

(RCIC) SYSTEM m

wag.

< 7<,

~ c.

3.5.1 ECCS - Operating

- >;ig% Q u.-

y me

. &dik' t LC0 3.5.1 Each ECCS injection / spray subsystem and the Automatici'i

'Depressurization System (ADS) function of seven safety /By.r g g g, yig p a el'f ' ;c:

i

. relief valves.shall be OPERABLE.

+ sw

, gh, a., -sw;c%

s..,

1.,

A[;pjpMOW&

APPLICABILITY:

MODE 1, a J #C.I E ' '

H MODES 2 and 3, except high pressure %.,and ADS v olant'injecti GIPCI Ji reactor steam done pressure :s; 150. psig.

~.fyy%;g,,pg...,,.

y$4g o

, s.

n

.m. e,

ACTIONS

' NhhN 'b5

. e4.# # 0 rg$ 'arr

CONDITION.

6. REQUI.R.E. D ACT, ION.

COMPLE, TION 3 IME T

N

. m ww ump @6

~.

i

.m ; -

. ?g 3

p

.Q Wh 5&f lE; #PF) w A.

One low pressure ECCS A.1 Restore low pressure 7 daysf.

My p%he injection / spray

"#ECCS injection / spray subsystem inoperable.

, subsystem to OPERABLE co status.

2 '.

w W lr ? ~n~t ;! -

. ; p.a e _g r '

a ti

' / *a,, '

B.

Required Action and 8.1

'Be in MODE 3.

12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months syi d ;o associated Completion Time of Condition A E

W not met.

B.2 Be in MODE 4.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months ..

..r 4'.

C.

HPCI System C.1 Verify by I hour

l '

administrative means f'

RCIC System is OPERABLE.

..qQ3 E

C.2 Restore HPCI System 14 days to OPERABLE status.

(continued)

HATCH UNIT 2 3.5-1 Amendment No'. 135

.a i'. w L.s A

s d p% *

~

.pn.,.

-fl e 3

-' y

. '. d

\\ t

ECCS-Operating 3.5.1 i

ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME D.

HPCI System-D.1 Restore HPCI System 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months inoperable.

to OPERABLE status.

ANQ 9B One low pressure ECCS D.2 Restore low pressure 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months injection / spray ECCS injection / spray subsystem-is subsystem to OPERABLE inoperable. -

status.

N_

_r n-- anc a--

.- Erl -

Restore-ADS-valve-to-days-i=epersle.

OPERABLE-status.---

I i

\\

-F.- One-ADS-valve 5.1 --

Restore-ADS-valve to -

72 5:ure---

/

i

rdle.

ORERABLE-stat".

AME -

^ 0 ! w prester: ECC4- -F,2-Restore-low-pressure 72 hesis

- iafectic /:pr y ECCS_ injection /spr:y

d:y:t= ir.epereMe, s @ system-to-OPERABLE-

-:t:t":r

^

m

~

N L --Gr Two or more ADS valves @

Be in MODE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months inoperable.

AND DE

<-Gre-Reduce reactor steam 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Required Action and gg, z dome pressure to associated Complet n

s; 150 psig.

imeJfJ ndition r-E, Or not me.

c (continued)

HATCH UNIT 2 3.5-2 Amenument No. 135

ECCS-Oporating

-3.5.1 ACTIONS (continued)

CONDITION REQUIRED ACTION.

COMPLETION TIME-V.

/#.I Hir wo or more low Enter LCO 3.0.3.

Immediately.

pressure ECCS injection / spray-s

.v subsystems inoperable.

gg.

'{-

no HPCI System and r

"7%l,'

more ADSivalves -

N 1

+

.,M.

' ];$.

lf5 3 1.

a.a;fr '

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUEKds.

.u SR 3.5.1.1 Verify, for each ECCS injection / spray 31 days subsystem, the piping is filled with water.

1

.y...

from the pump discharge valve to the 42 injection valve.

4 (continued)

HATCH UNIT 2 3.5-3 Amendment No. 135

LLS Valves 3.6.1.6 e

i 3.6 CONTAINMENT SYSTEMS 3.6.1.6 Low-Low Set.(LLS) Valves

+-Hrt.e.oU LC0 3.6.1.6 The LLS function o four safety / relief valves shall be OPERABLE.

APPLICABILITY:

MODES 1, 2, and 3.

ACTIONS

]

CONDITION REQUIRED ACTION _

COMPLETION TIME O

LLS v lve 1

est e LLS valve opera e.

atus.

e A.I ed A,ct nfd Be in MODE 3.

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months qu f

difio,A [

as ciated mpTe o m 6fC A_NQ n

met.

/

h Be in MODE 4.

36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months f

A,E

( A. Two or more LLS valves inoperable.

HATCH UNIT 2 3.6-18 Amendment No. 135

i Edwin I. Hatch Nuclear Plant Request to Revise Technical Specifications:

l Safety / Relief Valve Setpoint Change Bases Changes 3

For Information Only 4

i

-i i

1 1

i 1

1 s

1 1

i HL-5230 E4-1

W SLC System B 3.1.7 BASES j

SURVEILLANCE SR 3.1.7.4 and SR 3.1.7.6 (continued)

REQUIREMENTS in the nonaccident position provided it can be aligned to the accident position from the control room, or locally by a i

dedicated operator at the valve control. This is acceptable j

since the SLC System is a manually initiated system. This Surveillance also does not apply to valves that are locked, i

sealed, or otherwise secured in position since they are i

verified to be in the correct position prior to locking, sealing, or securing. This verification of valve alignment does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves. The 31 day Frequency is based on engineering judgment and is consistent with the procedural controls governing valve operation that ensures correct 3

valve positions.

SR 3.1.7.5 This Surveillance requires an examination of the sodium pentaborate solution by using chemical analysis to ensure that the proper concentration of boron exists in the storage tank (within Region A limits of Figures 3.1.7-1 and 3.1.7-2).

SR 3.1.7.5 must be performed anytime sodium pentaborate or water is added to the storage tank solution to determine that the boron solution concentration is within the specified limits.

SR 3.1.7.5 must also be performed any time the temperature is restored to within the Region A limits of Figure 3.1.7-2, to ensure that no significant boron precipitation occurred. The 31 day Frequency of this Surveillance is appropriate because of the relatively slow variation of boron concentration between surveillances.

SR 3.1.7.7 Demonstrating that each SLC System pump develops a flow rate a 41.2 gpm at a discharge pressure a 1232 psig ensures that l

pump performance has not degraded during the fuel cycle.

This minimum pump flow rate requirement ensures that, when combined with the sodium pentaborate solution concentration requirements, the rate of negative reactivity insertion from the SLC System will adequately compensate for the positive (continued)

HATCH UNIT 1 B 3.1-44 00CR 96-31 9/6/96

i i

LLS Instrumentation B 3.3.6.3 BASES ACTIONS S_,.1 (continued) function if any S/RV tailpipe pressure switch instrument channel becomes inoperable. Therefore, it is acceptable for I

plant operation to continue with only one tailpipe pressure switch OPERABLE on each S/RV. However, this is only acceptable provided each LLS valve is maintaining initiation capability.

(Refer to Required Action A.1 and 0.1 Bases.)

Requirco Action B.1 requires restoration of the tailpipe pressure switches to OPERABLE status prior to entering MODE 2 or 3 from MODE 4 to ensure that all switches are OPERABLE at the beginning of a reactor startup (this is because the switches are not accessible during plant operation). The Required Actions do not allow placing the channel in trip since this action could result in a LLS valve actuation. As noted, LC0 3.0.4 is not applicable, thus allowing entry into MODE 1 from MODE 2 with inoperable channels. This allowance is needed since the channels only have to be repaired prior to entering MODE 2 or MODE 3 from MODE 4.

Yet, LC0 3.0.4 would preclude entry into MODE 1 from MODE 2 since the Required Action does not allow unlimited operations, fu.1 A failure of two pressure switch channels associated with one S/RV tailpipe could result in the loss of the LLS function (i.e., multiple actuations of the S/RV would go undetected by the LLS logic).

However, there is a total of 11 S/RVs. Therefore, it would be very unlikely that a single S/RV would be required to arm all the LLS logic.

Therefore, it is acceptable to allow 14 days to restore one pressure switch of the associated S/RV to OPERABLE status (Required Action C.1).

However, this allowable out of service time is only acceptable provided each LLS is maintaining initiation capability (Refer to Required Action A.1 and D.1 Bases).

If one inoperable tailpipe pressure switch cannot be restored to OPERABLE status within the allowable out of service time, Condition D must be entered and its Required Action taken.

The Required Actions do not allow placing the channels in trip since this action could result in a LLS valve actuation.

(continued) i 1

HATCH UNIT 1 B 3.3-191 DOCR 96-31 9/6/96 l

S/RVs B 3.4.3 l

B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.3 Safety / Relief Valves (S/RVs)

BASES BACKGROUND The ASME Boiler and Pressure Vessel Code requires the reactor pressure vessel be protected from overpressure during upset conditions by self-actuated safety valves. As part of the nuclear pressure relief system, the size and number of S/RVs are selected such that peak pressure in the nuclear system will not exceed the ASME Code limits for the reactor coolant pressure boundary (RCPB).

The S/RVs are located on the main steam lines between the reactor vessel and the first isolation valve within the drywell. The S/RVs can actuate by either of two modes:

the i

safety mode or the relief mode.

In the safety mode (or I

spring mode of operation), the spring loaded pilot valve opens when steam pressure at the valve inlet overcomes the spring force holding the pilot valve closed.

Opening the pilot valve allows a pressure differential to develop across the main valve piston and opens the main valve. This satisfies the Code requirement.

Each S/RV discharges steam through a discharge line to a point below the minimum water level in the suppression pool.

The S/RVs that provide the relief mode are the low-low set j

(LLS) valves and the Automatic Depressurization System (ADS) valves. The LLS requirements are specified in LC0 3.6.1.6,

" Low-Low Set (LLS) Valves," and the ADS requirements are i

specified in LCO 3.5.1, "ECCS - Operating."

i f

APPLICABLE The overpressure protection system must accommodate the most SAFETY ANALYSES severe pressurization transient.

Evaluations have determined that the most severe transient is the closure of all main steam isolation valves (MSIVs), followed by reactor scram on high neutron flux (i.e., failure of the direct scram associated with MSIV position) (Ref. 1).

For the purpose of the analyses, 10 of 11 S/RVs are assumed to l

operate in the safety mode. The analysis results demonstrate that the design S/RV capacity is capable of maintaining reactor pressure well below the ASME Code limit i

(continued)

HATCH UNIT 1 B 3.4-13 00CR 96-31 9/6/96 3

S/RVs B 3.4.3 BASES APPLICABLE of 110% of vessel design pressure (110% x 1250 psig -

SAFETY ANALYSES 1375 psig).

Sensitivity analyses have demonstrated that (continued) 8 or 9 S/RVs operating in the pressure relief mode will maintain the reactor vessel below 1375 psig. This LC0 helps to ensure that the acceptance limit of 1375 psig is met during the Design Basis Event.

From an overpressure standpoint, the design basis events are bounded by the MSIV closure with flux scram event described above.

Reference 2 discusses additional events that are expected to actuate the S/RVs.

S/RVs satisfy Criterion 3 of the NRC Policy Statement (Ref. 4).

LC0 The S/RV safety function requires 10 of 11 S/RVs to be OPERABLE to satisfy the assumptions of the safety analysis (Refs. 1, 2, and 5), although margins to the ASME Vessel Overpressure Limit are substantial.

The requirements of this LC0 are applicable only to the capability of the S/RVs to mechanically open to relieve excess pressure when the lift setpoint is exceeded (safety function).

The S/RV setpoints are established to ensure that the ASME Code limit on peak reactor pressure is satisfied.

The ASME Code specifications require the lowest safety valve setpoint to be at or below vessel design pressure (1250 psig) and the highest safety valve to be set so that the total accumulated pressure does not exceed 110% of the design pressure for overpressurization conditions. The transient evaluations in the FSAR are based on these setpoints, but also include the additional uncertainties of 3% of the nominal setpoint drift to provide an added degree of conservatism.

Operation with fewer valves OPERABLE than specified, or with setpoints outside the ASME limits, could result in a more severe reactor response to a transient than predicted, i

possibly resulting in the ASME Code limit on reactor j

pressure being exceeded.

(continued) i HATCH UNIT 1 B 3.4-14 DOCR 96-31 9/6/96 i

. =..

4 S/RVs B 3.4.3 l

I BASES (continued)

{

APPLICABILITY In MODES 1, 2, and 3, 10 of 11 S/RVs must be OPERABLE, since l

considerable energy may be in the reactor core and the limiting design basis transients are assumed to occur in these MODES.

The S/RVs may be required to provide pressure relief to discharge energy from the core until such time that the Residual Heat Removal (RHR) System is capable of dissipating the core heat.

In MODE 4, decay heat is low enough for the RHR System to 4

provide adequate cooling, and reactor pressure is low enough that the overpressure limit i. unlikely to be approached by assumed operational transients er accidents.

In MODE 5, the reactor vessel head is unbolted or removed and the reactor 4

is at atmospheric pressure. Le S/RV function is not needed during these conditions.

ACTIONS A.1 and A.2 s

With 1 SR/V inoperable, no action is required, because an analysis demonstrated that the remaining 10 SR/Vs are l

capable of providing the necessary overpressure protection.

(See Reference 5.)

With two or more S/RVs inoperable, a transient may result in the violation of the ASME Code limit on reactor pressure.

The plant must be brought to a MODE in which the LC0 does not apply. To achieve this status, the plant must be brought to MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

(continued)

HATCH UNIT 1 B 3.4-15 DOCR 96-31 9/6/96

S/RVs B 3.4.3 i

BASES (continued) j i

SURVEILLANCE SR

3. 4 dd REQUIREMENTS This Surveillance requires that the S/RVs will open at the pressures assumed in the 'afety analysis of Reference 1.

The demonstration of the J/RV safety lift settings must be i

performed during shutdown, since this is a bench test, to be done in accordance with the Inservice Testing Program.

The j

lift setting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures. The S/RV setpoint is 3% for OPERABILITY; however, the valves are reset to 1% during the i

1 Surveillance to allow for drift.

Performance of this SR in accordance with the Inservice i

Testing Program requires an 18 month Frequency. The 18 month Frequency was selected because this Surveillance must be performed during shutdown conditions and is based on the time between refuelings.

SR 3.4.3.2 A manual actuation of each S/RV is performed to verify that, l

mechanically, the valve is functioning properly and no blockage exists in the valve discharge line.

This can be demonstrated by the response of the turbine control valves or bypass valves, by a change in the measured steam flow, or by any other method suitable to verify steam flow. Adequate i

reactor steam dome pressure must be available to perform this test to avoid damaging the valve. Also, adequate steam flow must be passing through the main turbine or turbine j

bypass valves to continue to control reactor pressure when the S/RVs divert steam flow upon opening.

Sufficient time is therefore allowed after the required pressure and flow are achieved to perform this test. Adequate pressure at which this test is to be performed is 920 psig (the pressure recommended by the valve manufacturer). Adequate steam flow is represented by at least 1.25 turbine bypass valves open, or total steam flow 2 lE6 lb/hr.

Plant startup is allowed prior to performing this test because valve OPERABILITY and the setpoints for overpressure protection are verified, per ASME Code requirements, prior to valve installation.

Therefore, this SR is modified by a Note that states the (continued)

HATCH UNIT 1 B 3.4-16 00CR 96-31 9/6/96 l

S/RVs B 3.4.3 BASES SURVEILLANCE SR 3.4.3.2 (continued)

Surveillance is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure and flow are adequate to perform the test. The 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowed for manual actuation after the required pressure is reached is sufficient to achieve stable conditions for testing and provides a reasonable time to complete the SR.

If a valve fails to actuate due only to the failure of the solenoid but is capable of opening on overpressure, the safety function of the S/RV is considered OPERABLE.

The 18 month Frequency was developed based on the S/RV tests required by the ASME Boiler and Pressure Vessel Code,Section XI (Ref. 3). Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency.

Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

REFERENCES 1.

FSAR, Appendix M.

2.

FSAR, Section 14.3.

3.

ASME, Boiler and Pressure Vessel Code,Section XI.

4.

NRC No.93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.

5.

NEDC-32041P, " Safety Review for Edwin I. Hatch Nuclear Power Plant Units 1 and 2 Updated Safety / Relief Valve Performance Requirements," April 1996.

HATCH UNIT 1 B 3.4-17 00CR 96-31 9/6/96

ECCS - Operating B 3.5.1 BASES I

BACKGROUND pumps without injecting water into the RPV.

These test (continued) lines also provide suppression pool cooling capability, as described in LCO 3.6.2.3, "RHR Suppression Pool Cooling."

Two LPCI inverters (one per subsystem) are designed to i

provide the power to various LPCI subsystem valves (e.g., inboard injection valves). This will ensure that a postulated worst case single active component failure, j

during a design basis loss of coolant accident (which includes loss of offsite power), would not result in the low pressure ECCS subsystems failing to meet their design function.

(Although an alternate power supply is available, the LPCI subsystem (s) may not be capable of meeting its design function if the alternate power supply (ies) is in service; i.e., the LPCI inverters are bypassed.)

j The HPCI System (Ref. 3) consists of a steam driven turbine pump unit, piping, and valves to provide steam to the turbine, as well as piping and valves to transfer water from the suction source to the core via the feedwater system line, where the coolant is distributed within the RPV through the feedwater sparger.

Suction piping for the system is provided from the CST and the suppression pool..

Pump suction for HPCI is normally aligned to the CST source to minimize injection of suppression pool water into the RPV. However, if the CST water supply is low, or if the suppression pool level is high, an automatic transfer to the suppression pool water source ensures a water supply for continuous operation of the HPCI System.

The steam supply to the HPCI turbine is piped from a main steam line upstream of the associated inboard main steam isolation valve.

The HPCI System is designed to provide core cooling for a wide range of reactor pressures (150 psig to 1185 psig).

l Upon receipt of an initiation signal, the HPCI turbine stop valve and turbine control valve open simultaneously and the turbine accelerates to a specified speed. As the HPCI flow increases, the turbine governor valve is automatically adjusted to maintain design flow.

Exhaust steam from the HPCI turbine is discharged to the suppression pool. A full flow test line is provided to route water from and to the CST to allow testing of the HPCI System during normal operation without injecting water into the RPV.

The ECCS pumps are provided with minimum flow bypass lines, which discharge to the suppression pool.

The valves in (continued)

HATCH UNIT 1 B 3.5-3 00CR 96-31 9/6/96

1 ECCS - Operating B 3.5.1 BASES APPLICABLE c.

Maximum hydrogen generation from a zirconium water SAFETY ANALYSES reaction is s 0.01 times the hypothetical amount that (continued) would be generated if all of the metal in the cladding surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react; d.

The core is maintained in a coolable geometry; and i

e.

Adequate long term cooling capability is maintained.

The limiting single failures are discussed in Reference 9.

The remaining OPERABLE ECCS subsystems provide the capability to adequately cool the core and prevent excessive fuel damage.

The ECCS satisfy Criteria 3 and 4 of the NRC Policy Statement (Ref. 12).

LC0 Each ECCS injection / spray subsystem and six of seven ADS l

valves are required to be OPERABLE.

The ECCS injection / spray subsystems are defined as the two CS subsystems, the two LPCI subsystems, and one HPCI System.

The low pressure ECCS injection / spray subsystems are defined as the two CS subsystems and the two LPCI subsystems.

With less than the required number of ECCS subsystems OPERABLE, the potential exists that during a limiting design basis LOCA concurrent with the worst case single failure, the limits specified in Reference 10 could be exceeded. All low pressure ECCS subsystems and ADS must therefore be OPERABLE to satisfy the single failure criterion required by Reference 10.

(Reference 9 takes no credit for HPCI.) HPCI must be OPERABLE due to risk consideration.

LPCI subsystems may be considered OPERABLE during alignment and operation for decay heat removal when below the actual RHR low pressure permissive pressure in MODE 3, if capable of being manually realigned (remote or local) to the LPCI mode and not otherwise inoperable.

At these low pressures and decay heat levels, a reduced complement of ECCS subsystems should provide the required core cooling, thereby allowing operation of RHR shutdown cooling when necessary.

(continued)

HATCH UNIT 1 B 3.5-5 DOCR 96-31 9/6/96

ECCS - Operating 8 3.5.1 BASES ACTIONS E.1 and E.2 (continued)

With one ADS valve inoperable, no action is required, because an analysis demonstrated that the remaining six ADS valves are capable of providing the ADS function, per Reference 13.

If any Required Action and associated Completion Time of Condition C or D is not met, or if two or more ADS valves l

are inoperable, the plant must be brought to a condition in which the LC0 does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reactor steam dome pressure reduced to s 150 psig within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

Entry into MODE 3 is not required if the reduction in reactor steam dome pressure to s 150 psig results in exiting the Applicability for the Condition, and the s 150 psig is achieved within the given 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

EJ l

When multiple ECCS subsystems are inoperable, as stated in Condition H, the plant is in a condition outside of the accident analyses. Therefore, LC0 3.0.3 must be entered immediately.

(continued)

HATCH UNIT 1 B 3.5-8 00CR 96-31 9/6/96

ECCS - Operating B 3.5.1 BASES (continued)

SURVEILLANCE SR 3.5.1.1 REQUIREMENTS The flow path piping has the potential to develop voids and pockets of entrained air.

Maintaining the pump discharge lines of the HPCI System, CS System, and LPCI subsystems full of water ensures that the ECCS will perform properly, injecting its full capac.ity into the RCS upon demand. This will also prevent a water hammer following an ECCS initiation signal. One acceptable method of ensuring that the lines are full is to vent at the high points.

In addition, when HPCI is aligned to the suppression pool (instead of the CST), one acceptable method is to monitor pump suction pressure.

The 31 day Frequency is based on the gradual nature of void buildup in the ECCS piping, the procedural controls governing system operation, and operating experience.

SR 3.5.1.2 Verifying the correct alignment for manual, power operated, and automatic valves in the ECCS flow paths provides assurance that the proper flow paths will exist for ECCS operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these l

were verified to be in the correct position prior to locking, sealing, or securing. A valve that receives an initiation signal is allowed to be in a nonaccident position provided the valve will automatically reposition in the proper stroke time. This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.

For the HPCI System, this SR also includes the steam flow path for the turbine and the flow controller l

position.

(continued)

HATCH UNIT 1 B 3.5-9 DOCR 96-31 9/6/96 l

ECCS - Operating B 3.5.1 BASES SURVEILLANCE SR 3.5.1.2 (continued)

REQUIREMENTS The 31 day Frequency of this SR was derived from the Inservice Testing Program requirements for performing valve testing at least once every 92 days.

The Frequency of 31 days is further justified because the valves are operated under procedural control and because improper valve position would only affect a single subsystem.

This Frequency has been shown to be acceptable through operating experience.-

This SR is modified by a Note that allows LPCI subsystems to be considered OPERABLE during alignment and operation for decay heat removal with reactor steam dome pressure less than the RHR low pressure permissive pressure in MODE 3, if capable of being manually realigned (remote or local) to the LPCI mode and not otherwise inoperable.

This allows operation in the RHR shutdown cooling mode during MODE 3, if necessary.

SR 3.5.1.3 Verification every 31 days that ADS air supply header pressure is a 90 psig ensures adequate air pressure for reliable ADS operation.

The accumulator on each ADS valve i

provides pneumatic pressure for valve actuation.

The design pneumatic supply pressure requirements for the accumulator are such that, following a failure of the pneumatic supply to the accumulator, at least two valve actuations can occur with the drywell at 70% of design pressure (Ref. 11). The ECCS safety analysis assumes only one actuation to achieve the depressurization required for operation of the low pressure ECCS. This minimum required pressure of 2 90 psig (for one actuation) is provided by the ADS instrument air supply. The 31 day Frequency takes into consideration administrative controls over operation of the air system and alarms for low air pressure.

(continued)

HATCH UNIT 1 B 3.5-10 DOCR 96-31 9/6/96 l

l

ECCS - Operating B 3.5.1 BASES SURVLILLANCE SR 3.5.1.12 (continued)

REQUIREMENTS The Frequency of 18 months is based on the need to perform the Surveillance under the conditions that apply just prior to or during a startup from a plant outage. Operating experience has shown that these components usually pass the SR when performed at the 18 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

REFERENCES 1.

FSAR, Section 6.4.3.

2.

FSAR, Section 6.4.4.

3.

FSAR, Seciion 6.4.1.

4.

FSAR, Section 6.4.2.

5.

FSAR, Section 14.4.3.

6.

FSAR, Section 14.4.5.

7.

8.

FSAR, Section 6.5.

9.

NEDC-31376P, "E.I. Hatch Nuclear Plant Units 1 and 2 SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis,"

December 1986, 10.

11. Memorandum from R.L. Baer (NRC) to V. Stello, Jr.

(NRC), " Recommended Interim Revisions to LCOs for ECCS Components," December 1, 1975.

12.

NRC No.93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.

13.

NEDC-32041P, " Safety Review for Edwin I. Hatch Nuclear Power Plant Units 1 and 2 Updated Safety / Relief Valve Performance Requirements," April 1996.

i HATCH UNIT 1 B 3.5-16 DOCR 96-31 9/6/96

4 RCIC System B 3.5.3 a

B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM B 3.5.3 RCIC System l

BASES 4

BACKGROUND The RCIC System is not'part of the ECCS; however, the RCIC System is included with the ECCS section because of-their similar functions.

The RCIC System is designed to operate either automatically or manually following reactor pressure vessel (RPV) isolation accompanied by a loss of coolant flow from the feedwater system to provide adequate core cooling and control of the RPV water level. Under these conditions, the High Pressure Coolant Injection (HPCI) and RCIC systems perform similar functions. The RCIC System design l

requirements ensure that the criteria of Reference 1 are satisfied.

l The RCIC System (Ref. 2) consists of a steam driven turbine pump unit, piping, and valves to provide steam to the i

turbine, as well as piping and valves to transfer water from the suction source to the core via the feedwater system line, where the coolant is distributed within the RPV i

through the feedwater sparger.

Suction piping is provided from the condensate storage tank (CST) and the suppression pool.

Pump suction is normally aligned to the CST to minimize injection of s:ppression pool water into the RPV.

However, if the CST water supply is low, or the suppression pool level is high, an automatic transfer to the suppression pool water source ensures a water supply for continuous operation of the RCIC System. The steam supply to the turbine is piped from a main steam line upstream of the associated inboard main steam line isolation valve.

The RCIC System is designed to provide core cooling for a wide range of reactor pressures (150 psig to 1185 psig).

l Upon receipt of an initiation signal, the RCIC turbine accelerates to a specified speed. As the RCIC flow increases, the turbine control valve is automatically adjusted to maintain design flow Exhaust steam from the RCIC turbine is discharged to the uppression pool. A full flow test line is provided to route water from and to the CST to allow testing of the RCIC System during normal operation without injecting water into the RPV.

(continued)

)

HATCH UNIT 1 B 3.5-23 DOCR 96-31 9/6/96

i LLS Valves B 3.6.1.6 BASES i

APPLICABLE.

assumption that simultaneous S/RV openings occur only on the SAFETY ANALYSEC initial actuation for DBAs.

Even though four S/RVs are (continued) designated for the LLS function, all four LLS S/RVs do not operate in any DBA analysis. Thus, operation with three of four LLS S/RVs OPERABLE is acceptable.

(See Reference 4.)

LLS valves satisfy Criterion 3 of the NRC Policy Statement

-l

~

(Ref. 3).

1 j

LC0 Three of four LLS valves are required to be OPERABLE to satisfy the assumptions of the safety analyses (Refs.1 and 4). The requirements of this LC0 are applicable to the i

mechanical and electrical / pneumatic capability f the LLS i

valves to function for controlling the opening ar.d closing of the S/RVs.

f l

J APPLICABILITY In MODES 1, 2, and 3, an event could cause pressurization of the reactor and opening of S/RVs.

In MODES 4 and 5, the j

probability and consequences of these events are reduced due L

to the pressure and temperature limitations in these MODES.

Therefore, maintaining the LLS valves OPERABLE is not

)

required in MODE 4 or 5.

ACTIONS A.1 and A.2 1

i l

With one LLS valve inoperable, no action is required,

]

because an analysis demonstrated that the remaining three LLS valves are capable of providing the necessary LLS function.

(See Reference 4.)

If two or more LLS valves are inoperable, the plant must be brought to a MODE in which the LC0 does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within M hours. The allowed Completion Times are reasonable, bassa on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

4 (continued)

HATCH UNIT 1 B 3.6-35 DOCR 96-31 9/6/96

I LLS Valves B 3.6.1.6 BASES SURVEILLANCE SR 3.6.1.6.1 REQUIREMENTS A manual actuation of each LLS valve is performed to verify that the valve and solenoids are functioning properly and no blockage exists in the valve discharge line. This can be demonstrated by the response of the turbine control or l

bypass valve, by a change in the measured steam flow, or by any other method that is r able to verify steam flow. -

Adequate reactor steam dont ressure must be available to

)

perform this test to avoid damaging the valve. Adequate pressure at which this test is to be performed is 2 920 psig 1

(the pressure recommended by the valve manufacturer). Also, adequate steam flow must be passing through the main turbine 4

or turbine bypass valves to continue to control reactor 2

pressure when the LLS valves divert steam flow upon opening.

Adequate steam flow is represented by at least 1.25 turbine bypass valves open, or total steam flow > IE6 lb/hr. The 18 month Frequency was based on the S/RV tests required by the ASME Boiler and Pressure Vessel Code,Section XI (Ref. 2). Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

Since steam pressure is required to perform the Surveillance, however, and steam may not be available during a unit outage, the Surveillance may be performed during the startup following a unit outage. Unit startup is allowed prior to performing the test because valve OPERABILITY and the setpoints for overpressure protection are verified by 4

ASME Section XI testing prior to valve installation. After adequate reactor steam pressure and flow are reached, 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is allowed to prepare for and perform the test.

Adequate pressure at which this test is to be performed is consistent with the pressure recommended by the valve i

manufacturer.

j l

a 4

(continued) i HATCH UNIT 1 B 3.6-36 DOCR 96-31 9/6/96 l

l

LLS Valves B 3.6.1.6 5

BASES SURVEILLANCE SR 3.6.1.6.2 REQUIREMENTS (continued)

The LLS designated S/RVs are required to actuate automatically upon receipt of specific initiation signals.

A system functional test is performed to verify that the mechanical portions (i.e., solenoids) of the LLS function operate as designed when initiated either by an actual or simulated automatic initiation signal. The LOGIC SYSTEM FUNCTIONAL TEST in SR 3.3.6.3.6 overlaps this SR to provide complete testing of the safety function.

The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potentia' for an unplanned transient if the i

l Surveillance were performed with the reactor at power.

~

Operating experience has shown these components usually pass the Surveillance when performed at the 18 month Frequency.

Therefore, the Frequency was concluded to be acceptable fror a reliability standpoint.

This SR is modified by a Note that excludts valve actuation.

This prevents a reactor pressure vessel pre',sure blowdown.

a REFERENCES 1.

FSAR, Section 4.11.

l 2.

ASME, Boiler and Pressure Vessel Code,Section XI.

3.

NRC No.93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.

4.

NEDC-32041P, " Safety Review for Edwin I. Hatch Nuclear Power Plant Units 1 and 2 Updated Safety / Relief Valve Performance Requirements," April 1996.

J

]

i i

4 HATCH UNIT 1 B 3.6-37 DOCR 96-31 9/6/96

SLC System B 3.1.7 BASES SURVEILLANCE SR 3.1.7.4 and SR 3.1.7.6 (continued)

REQUIREMENTS in the nonaccident position provided it can be aligned to the accident position from the control room, or locally by a dedicated operator at the valve control. This is acceptable since the SLC System is a manually initiated system. This Surveillance also does not apply to valves that are locked, sealed, or otherwise secured in position since they are verified to be in the correct position prior to locking, sealing, or securing. This verification of valve alignment does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position.

This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves. The 31 day Frequency is based on engineering judgment and is consistent with the procedural controls governing valve operation that ensures correct valve positions.

SR 3.1.7.5 This Surveillance requires an examination of the sodium pentaborate solution by using chemical analysis to ensure thtt the proper concentration of boron exists in the storage tank (within Region A 7 imits of Figures 3.1.7-1 and 3.1.7-2).

SR 3.1.7.5 oust be performed anytime sodium pentaborate or water is added to the storage tank solution l

to determine that the boron solution concentration is within I

I the specified limits.

SR 3.1.7.5 must also be performed any time the temperature is restored to within the Region A limits of Figure 3.?.7-2, to ensure that no significant boron precipitation occurred. The 31 day Frequency of this i

Surveillance is aprropriate because of the relatively slow variation of boron concentration between surveillances.

I SR 3.1.7.7 Demonstrating that each SLC System pump develops a flow rate 2 41.2 gpm at a discharge pressure 2 1232 psig ensures that l

pump performance has not degraded during the fuel cycle.

This minimum pump flow rate requirement ensures that, when combined with the sodium sentaborate solution concentration requirements, the rate of n eative reactivity insertion from the SLC System will adequately emoensate for the positwe (coi,tinued)

HATCH UNIT 2 B 3.1-44 DOCR 96-31 9/6/93

LLS Instrumentation B 3.3.6.3 BASES ACTIONS H.d (continued) function if any S/RV tailpipe pressure switch instrument channel becomes inoperable. Therefore, it is acceptable for plant operation to continue with only one tailpipe pressure switch OPERABLE on each S/RV. However, this is only acceptable provided each LLS valve is maintaining initiation capability.

(Refer to Required Action A.1 and 0.1 Bases.)

Required Action B.1 requires restoration of the tailpipe pressure switches to OPERABLE status prior to entering MODE 2 or 3 from MODE 4 to ensure that all switches are OPERABLE at the beginning of a reactor startup (this is because the switches are not accessible during plant operation). The Required Actions do not allow placing the channel in trip since this action could result in a LLS velve actuation. As noted, LC0 3.0.4 is not applicable, thus allowing entry into 40DE 1 from MODE 2 with inoperable channels. This allowan.t >s needed since the channels only have to be repaired prL to entering MODE 2 or MODE 3 from MODE 4.

Yet, LC0 3.0.4 would preclude entry into MODE 1 from MODE 2 since the Required Action does not allow unlimited operations.

fd A failure of two pressure switch channels associated with one S/RV tailpipe could result in the loss of the LLS function (i.e., multiple actuations of the S/RV would go undetected by the LLS logic). However, there is a total of 11 S/RVs. Therefore, it would be very unlikely that a single S/dV would be required to arm all the LLS logic.

Therefore, it is acceptable to allow 14 days to restore one pressure switch of the associated S/RV to OPERABLE status (Required Action C.1).

However, this allowable out of service time is only acceptable provided each LLS is maintaining initiation capability (Refer to Required Action A.1 and D.1 Bases).

If one inoperable tailpipe pressure switch cannot be restored to OPERABLE status within the allowable out of service time, Condition D must be entered and its Required Action taken.

The Required Actions do not allow placing the channels in trip since this action could result in a LLS valve actuation.

(continued) f HATCH UNIT 2 B 3.3-191 DOCR 96-31 9/6/96 d

-c

S/RVs B 3.4.3 8 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.3 Safety / Relief Valves (S/RVs)

BASES BACKGROUND The ASME Boiler and Pressure Vessel Code requires the reactor pressure vessel be protected from overpressure during upset conditions by self-actuated safety valves. As part of the nuclear pressure relief system, the size and number of S/RVs are selected such that peak pressure in the nuclear system will not exceed the ASME Code limits for the reactor coolant pressure boundary (RCPB).

The S/RVs are located on the main steam lines between the reactor vessel and the first isolation valve within the drywell. The S/RVs can actuate by either of two modes:

the safety mode or the relief mode.

In the safety mode (or spring mode of operation), the spring loaded pilot valve opens when steam pressure at the valve inlet overcomes the spring force holding the pilot valve closed. Opening the pilot valve allows a pressure differential to develop across the main valve piston and opens the main valve. This 1

satisfies the Code requirement.

Each S/RV discharges steam through a discharge line to a point below the minimum water level in the suppression pool.

The S/RVs that provide the relief mode are the low-low set (LLS) valves and the Automatic Depressurization System (ADS) valves.

The LLS requirements are specified in LC0 3.6.1.6,

" Low-Low Set (LLS) Valves," and the ADS requirements are specified in LC0 3.5.1, "ECCS - Operating."

APPLICABLE The overpressure protection system must accommodate the most SAFETY ANALYSES severe pressurization transient.

Evaluations have determined that the most severe transient is the closure of all main steam isolation valves (MSIVs), followed by reactor scram on high neutron flux (i.e., failure of the direct scram associated with MSIV position) (Ref.1).

For the purpose of the analyses,10 of 11 S/RVs are assumed to l

operate in the safety mode.

The analysis results demonstrate that the design S/RV capacity is capable of maintaining reactor pressure well below the ASME Code limit (continued)

HATCH UNIT 2 B 3.4-13 DOCR 96-31 9/6/96 1

l

S/RVs B 3.4.3 BASES i

APPLICABLE of 110% of vessel design pressure (110% x 1250 psig -

SAFETY ANALYSES 1375 psig).

Sensitivity analyses have demonstrated that (continued) 8 or 9 S/RVs operating in the pressure relief mode will maintain the reactor vessel below 1375 psig.

This LC0 helps to ensure that the acceptance limit of 1375 psig is met during the Design Basis Event.

From an overpressure standpoint, the design basis events are i

bounded by the MSIV closure with flux scram event described above.

Reference 2 discusses additional events that are expected to actuate the S/RVs.

S/RVs satisfy Criterion 3 of the NRC Policy Statement (Ref. 4).

i i

l LC0 The S/RV safety function requires 10 of 11 S/RVs to be OPERABLE to satisfy the assumptions of the safety analysis (Refs. 1, 2, and 5), although margins to the ASME Vessel Overpressure Limit are substantial.

The requirements of this LCO are applicable only to _the capability of the S/RVs to mechanically open to relieve excess pressure when the j

lift setpoint-is exceeded (safety function).

The S/RV setpoints are established to ensure that the ASME Code limit on peak reactor pressure is satisfied.

The ASME Code specifications require the lowest safety valve setpoint to be at or below vessel design pressure (1250 psig) and the highest safety valve to be set so that the total accumulated i

pressure does not exceed 110% of the design pressure for overpressurization conditions. The transient evaluations in the FSAR are based on these setpoints, but also include the additional uncertainties of

3% of the nominal setpoint drift to provide an added degree of conservatism.

Operation with fewer valves OPERABLE than specified, or with setpoints outside the ASME limits, could result in a more severe reactor response to a transient than predicted, possibly resulting in the ARME Code limit on reactor pressure being exceeded.

i (continued)

HATCH UNIT 2 B 3.4-14 DOCR 96-31 9/6/96

S/RVs B 3.4.3 BASES (continued) l APPLICABILITY In MODES 1, 2, and 3, 10 of 11 S/RVs must be OPERABLE, since l

considerable energy may be in the reactor core and the limiting design basis transients are assumed to occur in these MODES. The S/RVs may be required to provide pressure j

relief to discharge energy from the core until such time that the Residual Heat Removal (RHR) System is capable of dissipating the core heat.

In MODE 4, decay heat is low enough for the RHR System to provide adequate cooling, and reactor pressure is low enough i

that the overpressure limit is unlikely to be approached by assumed operational transients or accidents.

In MODE 5, the reactor vessel head is unbolted or removed and the reactor is at atmospheric pressure. The S/RV function is not needed during these conditions.

ACTIONS A.1 and A.2 With 1 SR/V inoperable, no action is required, because an analysis demonstrated that the remaining 10 SR/Vs are capable of providing the necessary overpressure protection.

(See Reference 5.)

With two or more S/RVs inoperable, a transient may result in the violation of the ASME Code limit on reactor pressure.

The plant must be brought to a MODE in which the LC0 does not apply. To achieve this status, the plant must be brought to MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based cn operating experience, to reach required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

)

(continued)

HATCH UNIT 2' B 3.4-15 DOCR 96-31 9/6/96

S/RVs B 3.4.3 BASES. (continued)

SURVEILLANCE SR 3.4.3.1 REQUIREMENTS-This Surveillance requires that the S/RVs will open at the pressures assumed in the safety analysis of Reference 1.

The demonstration of the S/RV safety lift settings must be performed during shutdown, since this is a bench test, to be done in accordance with the Inservice Testing Program. The lift setting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures. The S/RV setpoint is 3% for OPERABILITY; however, the valves are reset to

1% during the Surveillance to allow for drift.

Performance of this SR in accordance with the Inservice Testing Program requires an 18 month Frequency.

The 18 month Frequency was selected because this Surveillance must be performed during shutdown conditions and is based on the time between refuelings.

SR 3.4.3 &

A manual actuation of each S/RV is performed to verify that, mechanically, the valve is functioning properly and no blockage exists in the valve discharge line. This can be demonstrated by the response of the turbine control valves or bypass valves, by a change in the measured steam flow, or by any other method suitable to verify steam flow. Adequate reactor steam dome pressure must be available to perform this test to avoid damaging the valve. Also, adequate steam flow must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure when the S/RVs divert steam flow upon opening.

Sufficient time is therefore allowed after the. required pressure and flow are achieved to perform this test. Adequate pressure at which this test is to be performed is 920 psig (the pressure recommended by the valve manufacturer).

Adequate steam flow is represented by at least 1.25 turbine bypass valves open, or total steam flow 21E6 lb/hr.

Plant startup is allowed prior to performing this test because valve OPERABILITY and the setpoints for overpressure protection are verified, per ASME Code requirements, prior to valve installation.

l Therefore, this SR is modified by a Note that states the (continued)

HATCH UNIT 2 B 3.4-16 DOCR 96-31 9/6/96 l

S/RVs B 3.4.3 BASES (continued)

SURVEILLANCE SR 3.4.3.2 (continued)

REQUIREMENTS Surveillance is not required to be performed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure and flow are adequate to perform the test. The 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowed for manual actuation after the required pressure is reached is sufficient to achieve stable conditions for testing and provides a reascnable time to complete the SR.

If a valve fails to actuate due only to the failure of the solenoid but is capable of opening on overpressure, the safety function of the S/RV is considered OPERABLE.

The 18 month Frequency was developed based on the S/RV tests required by the ASME Boiler and Pressure Vessel Code,Section XI (Ref. 3). Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency.

Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

REFERENCES 1.

FSAR, Supplement 5A.

2.

FSAR, Section 15, 3.

ASME, Boiler and Pressure Vessel Code,Section XI.

4.

NRC No.93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.

1 5.

NEDC-32041P, " Safety Review for Edwin I. Hatch Nuclear Power Plant Units 1 and 2 Updated Safety / Relief Valve Performance Requirements," April 1996.

HATCH UNIT 2 B 3.4-17 DOCR 96-31 9/6/96

i ECCS - Operating B 3.5.1 J

BASES i

BACKGROUND pumps without injecting water into the RPV.

These test (continued) lines also provide suppression pool cooling capability, as described in LC0 3.6.2.3, "RHR Suppression Pool Cooling."

J Two LPCI inverters (one per subsystem) are designed to provide the power to various LPCI subsystem valves (e.g., inboard injection valves).

This will ensure that a postulated worst case single active component failure, during a design basis loss of coolant accident (which i

includes loss of offsite power), would not result in the low i

pressure ECCS subsystems failing to meet their design function.

(Although an alternate power supply is available, the LPCI subsystem (s) may not be capable of meeting its design function if the alternate power supply (ies) is in 3

service; i.e., the LPCI inverters are bypassed.)

l The HPCI System (Ref. 3) consists of a steam driven turbine pump unit, piping, and valves to provide steam to the turbine, as well as piping and valves to transfer water from the suction source to the core via the feedwater system line, where the coolant is distributed within the RPV through the feedwater sparger.

Suction piping for the system is provided from the CST and the suppression pool.

4 Pump suction for HPCI is normally aligned to the CST source to minimize injection of suppression pool water into the RPV. However, if the CST water supply is low, or if the suppression pool level is high, an automatic transfer to the suppression pool water source ensures a water supply for l

continuous operation of the HPCI System. The steam supply j

to the HPCI turbine is piped from a main steam line upstream of the associated inboard main steam isolation valve.

The HPCI System is designed to provide core cooling for a wide range of reactor pressures (162 psid to 1200 psid, l

i vessel to pump suction). Upon recnipt of an initiation j

signal, the HPCI turbine stop valve and turbine control valve open simultaneously and the turbine accelerates to a specified speed. As the HPCI flow increases, the turbine governor valve is automatically adjusted to maintain design l

flow.

Exhaust steam from the HPCI turbine is discharged to 4

the suppression pool. A full flow test line is provided to route water from and to the CST to allow testing of the HPCI System during normal operation without injecting water into j

i the RPV.

i The ECCS pumps are provided with minimum flow bypass lines,

)

which discharge to the suppression pool. The valves in f

(continued) i HATCH UNIT 2 B 3.5-3 DOCR 96-31 9/6/96

~

ECCS - Operating l

B 3.5.1 i.

BASES APPLICABLE c.

Maximum hydrogen generation from a zirconium water i

SAFETY ANALYSES reaction is s 0.01 times the hypothetical amount that (continued) would be generated if all of the metal in the cladding j

surrounding the fuel, excluding the cladding 4

surrounding the plenum volume, were to react; d.

The core is maintained in a coolable geometry; and e.

Adequate long term cooling capability is maintained.

The limiting single failures are discussed in Reference 10.

The remaining OPERABLE ECCS subsystems provide the capability to adequately cool the core and. prevent excessive 4

fuel damage.

The ECCS satisfy Criteria 3 and 4 of the NRC Policy Statement (Ref. 13).

LC0 Each ECCS injection / spray subsystem and six of seven ADS l

valves are required to be OPERABLE.

The ECCS injection / spray subsystems are defined as the two CS subsystems, the two LPCI subsystems, and one HPCI System.

l The low pressure ECCS injection / spray subsystems are defined as the two CS subsystems and the two LPCI subsystems.

With less than the required number of ECCS subsystems OPERABLE, the potential exists that during a limiting design basis LOCA concurrent with the worst case single failure, the limits specified in Reference 11 could be exceeded.

All low pressure ECCS subsystems and ADS must therefore be OPERABLE to satisfy the single failure criterion required by Reference 11.

(Reference 10 takes no credit for HPCI.)

HPCI must be OPERABLE due to risk consideration.

LPCI subsystems may be considered OPERABLE during alignment and operation for decay heat removal when below the actual i

RHR low pressure permissive pressure in MODE 3, if capable i

of being manually realigned (remote or local) to the LPCI mode and not otherwise inoperable. At these low pressures and decay heat levels, a reduced complement of ECCS subsystems should provide the required core cooling, thereby j

allowing operation of RHR shutdown cooling when necessary.

I (continued) 1 HATCH UNIT 2 B 3.5-5 DOCR 96-31 9/6/96 a

ECCS - Operating B 3.5.1 BASES ACTIONS E.1 and E.2 (continued)

With one ADS valve inoperable, no action is required, because an analysis demonstrated that the remaining six ADS valves are capable of providing the ADS function, per Reference 16.

1 If any Required Action and associated Completion Time of Condition C or D is not met, or if two or more ADS valves l

are inoperable, the plant must be brought to a condition in which the LC0 does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reactor steam dome pressure reduced to s 150 psig within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

Entry into MODE 3 is not required if the reduction in reactor steam dome pressure to s 150 psig results in exiting the Applicability for the Condition, and the s 150 psig is achieved within the given 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

EJ l

When multiple ECCS subsystems are inoperable, as stated in Condition H, the plant is in a condition outside of the accident analyses. Therefore, LC0 3.0.3 must be entered immediately.

(continued)

HATCH UNIT 2 B 3.5-8 DOCR 96-31 9/6/96

k ECCS - Operating B 3.5.1 BASES (continued)

SURVEILLANCE SR 3.5.1.1 REQUIREMENTS The flow path piping has the potential to develop voids and pockets of entrained air. Maintaining the pump discharge lines of the HPCI System, CS System, and LPCI subsystems full of water ensures that the ECCS will perform properly, injecting its full capacity into the RCS upon demand. This will also prevent a water hammer following an ECCS initiation signal. One acceptable method of ensuring that the lines are full is to vent at the high points.

In addition, when HPCI is aligned to the suppression pool (instead of the CST), one acceptable method is to monitor 4

pump suction pressure. The 31 day Frequency is based on the gradual nature of void buildup in the ECCS piping, the procedural controls governing systen, operation, and operating experience.

SR 3.5.1.2 Verifying the correct alignment for manual, power operated, and automatic valves in the ECCS flow paths provides assurance that the proper flow paths will exist for ECCS i

operation.

This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these were verified to be in the correct position prior to locking, sealing, or securing. A valve that receives an initiation signal is allowed to be in a nonaccident position provided the valve will automatically reposition in the 3

proper stroke time.

This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position. This SR does not apply to valves that cannot be inadvertently misaligned, such as check valves.

For the HPCI System, this SR also includes the i

steam flow path for the turbine and the flow controller position.

d i

I 1

(continued) a HATCH UNIT 2 B 3.5-9 DOCR 96-31 9/6/96 l

i

i ECCS -- Operating B 3.5.1 BASES 1

i SURVEILLANCE SR 3.5.1.2 (continued)

REQUIREMENTS The 31 day Frequency of this SR was derived from the Inservice Testing Program requirements for performing valve testing at least once every 92 days. The Frequency of 31 days is further justified because the valves are operated under precedural control and because improper valve position would only affect a single subsystem.

This Frequency has been shown to be acceptable through operating experience.

This SR is modified by a Note that allows LPCI subsystems to be considered OPERABLE during alignment and operation for decay heat removal with reactor steam dome pressure less than the RHR low pressure permissive pressure in MODE 3, if capable of being manually realigned (remote or local) to the LPCI mode and not otherwise inoperable. This allows operation in the RHR shutdown cooling mode during MODE 3, if necessary.

SR 3.5.1.3 Verification every 31 days that ADS air supply header pressure is 2: 90 psig ensures adequate air pressure for reliable ADS operation. The accumulator on each ADS valve provides pneumatic pressure for valve actuation. The design pneumatic supply pressure requirements for the accumulator are such that, following a failure of the pneumatic supply to the accumulator, at least two valve actuations can occur with the drywell at 70% of design pressure (Ref.12). The ECCS safety analysis assumes only one actuation to achieve the depressurization required for operation of the low pressure ECCS. This minimum required pressure of 2: 90 psig for one actuation is provided by the ADS instrument air supply.

The 31 day Frequency takes into consideration administrative controls over operation of the air system and alarms for low air pressure.

1 (continued)

HATCH UNIT 2 B 3.5-10 DOCR 96-31 9/6/96 l

I ECCS - Operating B 3.5.1 BASES l

REFERENCES 6.

FSAR, Section 15.1.40.

(continued) 7.

FSAR, Section 15.1.33.

8.

9.

FSAR, Section 6.3.3.

10.

NEDC-31376P, "E.I. Hatch Nuc1 ear P1 ant Units 1 and 2 SAFER /GESTR-LOCA Loss-of-Coolant Analysis,"

December 1986.

11.

12. Memorandum from R.L. Baer (NRC) to V. Stello, Jr.

(NRC), " Recommended Interim Revisions to LCOs for ECCS Components," December 1, 1975.

13.

NRC No.93-102 " Final Policy Statement on Technical Specification Improvements," July 23, 1993.

14 '.

Technical Requirements Manual.

15.

NED0-32291, "Sy.eim Analyses for Elimination of Selected Respor e Time Testing Requirements,"

January 1994.

16.

NEDC-32041P, " Safety Review for Edwin I. Hatch Nuclear Power Plant Units 1 and 2 Updated Safety / Relief Valve Performance Requirements," April 1996.

j HATCH UNIT 2 B 3.5-16a DOCR 96-31 9/6/96

LLS Valves B 3.6.1.6 BASES APPLICABLE assumption that simultaneous S/RV openings occur only on the SAFETY ANALYSES initial actuation for DBAs.

Even though four S/RVs are (continued) designated for the LLS function, all four LLS S/RVs do not

=

operate in any DBA analysis. Thus, operation with three of four LLS S/RVs OPERABLE is acceptable.

(See Reference 4.)

4 LLS valves satisfy Criterion 3 of the NRC Policy Statement (Ref. 3).

f LC0 Three of four LLS valves are required to be OPERABLE to satisfy the assumptions of the safety analyses (Refs. I and 4).

The requirements of this LC0 are applicable to the mechanical and electrical / pneumatic capability of the LLS valves to function for controlling the opening and closing of the S/RVs.

l APPLICABILITY In MODES 1, 2, and 3, an event could cause pressurization of the reactor ant ?ening of S/RVs.

In MODES 4 and 5, the probability and 1 sequences of these events are reduced due to the pressure a..m temperature limitations in these MODES.

Therefore, maintaining the LLS valves OPERABLE is not 4

required in MODE 4 or 5.

ACTIONS A.1 and A.2 i

i With one LLS valve inoperable, no action is required, because an analysis demonstrated that the remaining three LLS valves are capable of providing the necessary LLS function.

(See Reference 4.)

If two or more LLS valves are inoperable, the plant must be brought to a MODE in which the LC0 does not apply. To j

achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and to MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

(continued)

HATCH VNIT 2 8 3.6-36 DOCR 96-31 9/6/96

LLS Valves B 3.6.1.6 BASES SURVEILLANCE SR 3.6.1.6.1 REQUIREMENTS A manual actuation of each LLS valve is performed to verify that the valve and solenoids are functioning properly and no blockage exists in the valve discharge line. This can be demonstrated by the response of the turbine control or bypass valve, by a change in the measured steam flow, or by any other method that is suitable to verify steam flow.

Adequate reactor steam dome pressure must be available to perform this test to avoid damaging the valve. Adequate pressure at which this test is to be performed is 2 920 psig (the pressure recommended by the valve manufacturer). Also, adequate steam flow must be passing through the main turbine or turbine bypass valves to continue to control reactor pressure when the LLS valves divert steam flow upon opening.

Adequate steam flow is represented by at least 1.25 turbine bypass valves open, or total steam flow 2 1E6 lb/hr. The 18 month Frequency was based on the S/RV tests required by the ASME Boiler and Pressure Vessel Code,Section XI (Ref. 2). Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency.

Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

Since steam pressure is required to perform the Surveillance, however, and steam may not be available during a unit outage, the Surveillance may be performed during the startup following a unit outage. Unit startup is allowed prior to performing the test because valve OPERABILITY and the setpoints for overpressure protection are verified by ASME Section XI testing prior to valve installation. After adequate reactor steam pressure and flow are reached, 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is allowed to prepare for and perform the test.

Adequate pressure at which this test is to be performed is consistent with the pressure recommended by the valve manufacturer.

(continued)

HATCH UNIT 2 8 3.6-37 DOCR 96-31 9/6/96 l

1

LLS Valves B 3.6.1.6 i

BASES SURVEILLANCE SR 3.6.1.6.2 REQUIREMENTS (continued)

The LLS designated S/RVs are required to actuate auwutically upon receipt of specific initiation signals.

A system functional test is performed to verify that the mechanical portions (i.e., solenoids) of the LLS function operate as designed when initiated either by an actual or simulated automatic initiation signal. The LOGIC SYSTEM FUNCTIONAL TEST in SR 3.3.6.3.6 overlaps this SR to provide i

complete testing of the safety function.

j The 18 month Frequency is based on the need to perform this-Surveillance under the conditions that apply during a plant outage and the potential for an unplanned transient if the Surveillance were performed with the reactor at power.

Operating experience has shown these components usually pass the Surveillance when performed at the 18 month Frequency.

Therefore, the Frequeacy was concluded to be acceptable from a reliability standpoint.

This SR is modified by a Note that excludes valve actuation.

This prevents a reactor pressure vessel pressure blowdown.

1 REFERENCES 1.

FSAR, Section 5.5.17.

2.

ASME, Boiler and Pressure Vessel Code,Section XI.

3.

NRC No.93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.

4.

NEDC-32041P, " Safety Review for Edwin I. Hatch Nuclear Power Plant Units 1 and 2 Updated Safety / Relief Valve Performance Requirements," April 1996.

HATCH UNIT 2 B 3.6-38 DOCR 96-31 9/6/96

Attachment to Enclosure 4 Marked Up Bases Pages l

l

SLC System 8 3.1.7 BASES i

SURVEILLANCE SR 3.1.7.4 and SR 3.1.7.6 (continued)

REQUIREMENTS in the nonaccident position provided it can be aligned to the accident position from the control room, or locally by a dedicated operator at the valve control. This is acceptable since the SLC System is a manually initiated system. This i

Surveillance also does not apply to valves that are locked, sealed, or otherwise secured in position since they-are-verified to be in the correct position prior to locking, i

sealing, or securing. This verification of valve alignment does not require any testing or valve manipulation; rather, it involves verification that those valves capable of being mispositioned are in the correct position. This SR does not

)

apply to valves that cannot be inadvertently misaligned, such as check valves. The 31 day Frequency is based on engineering judgment and is consistent with the procedural controls governing valve operation that ensures correct valve positions.

i SR 3.1.7.5 This Surveillance requires an examination of the sodium pentaborate solution by using chemical analysis to ensure that the proper concentration of boron exists in the storage tank (within Region A limits of Figures 3.1.7-1 and 3.1.7-2).

SR 3.1.7.5 must be performed anytime sodium pentaborate or water is added to the storage tank solution to determine that the boron solution concentration is within the specified limits. SR 3.1.7.5 must also be perfomed any time the temperature is restored to within the Region A limits of Figure 3.1.7-2, to ensure that no significant boron precipitation occurred. The 31 day Frequency of this Surveillance is appropriate because of the relatively slow variation of boron concentration between surveillances.

Z3L)

SR 3.1.7.7 Demonstrating that each SLC System pu ae.elops a flow rate k 41.2 gpm at a discharge pressure a psig ensures that l

pump performance has not degraded during the fuel cycle.

This minimum pump flow rate requirement ensures that, when combined with the sodium pentaborate solution concentration requirements, the rate of negative reactivity insertion from i

the SLC System will adequately compensate for the positive (continued)

HATCH UNIT 1 B 3.1-44 REVISION 6 i

LLS Instrumentation l

B 3.3.6.3 BASES l

ACTIONS L1 (continued) function if any S/RV tailpipe pressure switch instrument channel becomes inoperable. Therefore, it is acceptable for i

plant operation to continue with only one tailpipe pressure i

switch OPERABLE on each S/RV. However, this is on?y acceptable provided each LLS valve is maintaining Stiation capability.

(Refer to Required Action A.1 and D.1 bases.)

Required Action B.1 requires restoration of the tailpipe pressure switches to OPERABLE status prior to entering-MODE 2 or 3 from MODE 4 to ensure that all switches are OPERABLE at the beginning of a reactor startup (this is.

because the switches are not accessible during plant operation). The Required Actions do not allow placing the channel in trip since this action could result in a LLS valve actuation. As noted, LCO 3.0.4 is not applicable, t

thus allowing entry into MODE 1 from NODE 2 with inoperable channels. This allowance is needed since the channels only have to be repaired prior to entering MODE 2 or MODE 3 from l

MODE 4.

Yet, LCO 3.0.4 would preclude entry into MODE 1 I

from MODE 2 since the Required Action does not allow unlimited operations.

I bl i

l A failure of two pressure switch channels associated with one S/RV tailpipe could result in the loss of.the LLS

, 4 h ue o.s a.+ot 3 function (i.e., multiple actuations of the S/RV would go l

c En s/evs /al ewe' :-:::: =e. e rm e ev-t. :=e:: ef s/]RY:

d undetected by the LLS logic). However,$50S/RY:::

i i-iti:lly :;:: (:etreint: mee-et-seme-eettir: fr t t:tel Jf 11 S/"": in thr-cret::)J Therefore, it would be very unlikely that a single S/(V would be required to are all the LLS logic. Therefore, it is acceptable to allow 14 days to restore one pressure switch of the associated S/RV to OPERABLE status (Required Action C.1). However, this allowable out of service time is only acceptable provided each LLS is maintaining initiation capability (Refer to Required Action A.1 and D.1 Bases).

If one inoperable tailpipe pressure switch cannot be restored to OPERABLE status within the allowable out of service time, Condition D aust be entered and its Required Action taken. The Required Actions do not allow placing the channels in trip since this action could result in a LLS valve actuation.

(continued)

HATCH UNIT I B 3.3-191 REVISION 1

j S/RVs B 3.4.3 4

l B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.3 Safety / Relief Valves (S/RVs)

BASES 1

BACKGROUND The ASME Boiler and Pressure Vessel Code requires the reactor pressure vessel be protected from overpressure during upset conditions by self-actuated safety valves. As part of the nuclear pressure relief system, the size and number of S nuclear sys/RVs are selected such that peak pressure in the tem will not exceed the ASME Code limits for the reacter coolant pressure boundary (RCPB).

The S/RVs are located on the main steam lines between the reactor vessel and the first isolation valve within the drywell. The S/RVs can actuate by either of two modes: the 3

safety mode or the relief mode.

In the safety mode (or spring mode of operation the spring loaded pilot valve opens when steam pressure),at the valve inlet overcomes the i

J spring force holding the pilot valve closed. Opening the pilot valve allows a pressure differential to develop across the main valve piston and opens the main valve.

This

' satisfies the Code requirement.

Each S/RV discharges steam through a discharge line to a point below the minimum water level in the suppression pool.

The S/RVs that provide the relief mode are the low-low set (LLS) valves and the Automatic Depressurization System (ADS) valves.

The LLS requirements are specified in LCO 3.6.1.6,

~

" Low-Low Set (LLS) Valves," and the ADS requirements are specified in LCO 3.5.1, "ECCS - Operating."

i, APPLICABLE The overpressure protection system must accommodate the most SAFETY ANALYSES severe pressurization transient.

Evaluations have 4

determined that the most severe transient is the closure of all main steam isolation valves (MSIVs), followed by reactor i

scram on high neutron flux (i.e., failure of the direct scram associated with MSIV position)

Ref. 1). For the purpose of the analyses,11 S/RVs are(assumed to operate in i

the safety mode. The lysis results demonstrate that the a

j design S/RV capacity 1: capable of maintaining reactor pressure well below tho ASME Code limit of 110% of vessel 1

4 (continued) 1 HATCH UNIT 1 B 3.4-13 REVISION O

S/RVs B 3.4.3 i

BASES i

APPLICABLE design pressure (110% x 1250 psig = 1375 psig). Sensitivity SAFETY ANALYSES analyses have demonstrated that 8 or 9 S/RVs operating in (continued) the pressure relief mode will maintain the reactor vessel below 1375 psig. This LCO helps to ensure that the 4

acceptance limit of 1375 psig is met during the Design Basis Event.

i From an overpressure standpoint, the design basis events are bounded by the MSIV closure with flux scram event described above.

Reference 2 discusses additional events that are expected to actuate the S/RVs.

S/RVs satisfy Criterion 3 of the NRC Policy Statement (Ref. 4).

i rr

~

s i

wm Qatynazer v.

w n m oL s

4 LCO TheAsafety funcT

. re r 9 W C h (re / e vg M.0)

Hiiu 4 be

}

OPERABLE to satisf the assumptions of the safety analysis Refs.Ys

), although margins to the ASME Vessel 1

LCedT fpTe imit are substantial.

The requirements of this LCO are applicable only to the capability of the S/RVs to mechanically open to relieve excess pressure when the lift setpoint is exceeded (safety function).

The S/RV setpoints are established to ensure that the ASME Code limit on peak reactor pressure is satisfied. The ASME Code specifications require the lowest safety valve setpoint to be at or below vessel design pressure (1250 psig) and the highest safety valve to be set so that the total accumulated pressure does not exceed 110% of the design pressure for overpressurization conditions. The transient evaluations in the FSAR are based on these setpoints, but also include the additional uncertainties of 3% of the nominal setpoint drift to provide an added degree of conservatism.

Operation with fewer valves-CPERABLE than specified, or with setpoints outside the ASME limits, could result in a more severe reactor response to a transient than predicted, possibly resulting in the ASME Code limit on reactor pressure being exceeded.

(continued)

HATCH UNIT 1 B 3.4-14 REVISION 0

S/RVs l

B 3.4.3 BASES (continued) em APPLICABILITY Ia MODES 1, 2, and 3 must be OPERABLE, since considerable energy m,ay be in the reactor core and the limiting design basis transients are assumed to occur in these MODES. The S/RVs may be required to provide pressure relief to discharge energy from the core until such time that the Residual Heat Removal (RHR) System is capable of dissipating the cora heat.

In MODE 4, decay heat is low enough for the RHR System to provide adequate cooling, and reactor pressure is low enough that the overpressure limit is unlikely to be approached by assumed operational transients or accidents.

In MODE 5, the reactor vessel head is unbolted or removed and the reactor is at atmospheric pressure.

The S/RV function is not needed during these conditions.

~e m

ACTIONS L1 With the afety functio of one S/RV inoperabl the remaining ERABLC S/RVs e capable providi the necessary ov pressure pro ction. H ver, the verall liability o the pressure elief syst is reduc beca e a itional fail es in the r ining OPE LE S/RVs ould res t in failur to adequate 1 relieve p sure dur a

limit g event.

this reaso continued peration permitt for a lim ed time on1 e 14 day ompletion me to resto the inope ble S/RV O

LE sta as is base on the reli capabilit of the r

ning S/R'

, the low obability o an event quiring S/RV tuation, nd a reas able time t complete t t

equir Action.

WA 1 52/ V L opera.bu-.,

ho (LClion i*, cefNb,

.,=--.x

~

hecusAbadyY f4wc'oe m re 5l N s

WittF = r: + %n en: S '""

perable, a transient may result pm; 3af;;y fEnttion-PothMtd ASME Code limit on reactor

[ cite s4rAhd

g*

I FecinoperaMeT/RV-tannotT ~ ~

cataN 10 M2N' torM te OPERABLE-status-within-the-associated-Completion-are capaw 8 prvvid Mccosag,owrpreswe%

Ti= cf 5;; ired Actton-Ad, Or " ^h tn :r =r: S[""; is in;;qsMedike s&fety f; action-of-e, iiTut'must lie br6iigEt I

e v t e c b c ) g( g o a MODE in wnich the LCO does not apply. To achieve this

(

status, the plant must be brought to MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (continued) i HATCH UNIT 1 B 3.4-15 REVISION 0

S/RVs 4

1 B 3.4.3 BASES i

SURVEILLANCE SR 3.4.3.2 (continued)

REQUIREMENTS recommended by the valve manufacturer). Adequate steam flow is represented by at least 1.25 turbine bypass valves open, or total steam flow 11E6 lb/hr. Plant startup is allowed prior to performing this test because valve OPERABILITY and the setpoints for overpressure protection are verified, per ASME Code requirements, prior to valve installation.

Therefore, this SR is modified by a Note that states the Surveillance is not required to be psrformed until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after reactor steam pressure and flow are adequate to perfom the test. The 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowed for manual actuation after the required pressure is reached is sufficient to achieve stable conditions for testing and provides a reasonable time to complete the SR.

If a valve fails to actuate due only to che failure of the solenoid but is capable of opening on overpressure, the safety function of the S/RV is considered OPERABLE.

The 18 month Frequency was developed based on the S/RV tests required by the ASME Boiler and Pressure Vessel Code,Section XI (Ref. 3). Operating experience has shown that these components usually pass the Surveillance when performed at the 18 month Frequency. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

~

REFERENCES 1.

FSAR, Appendix M.

2.

FSAR, Section 14.3.

3.

ASME, Boiler and Pressure Vessel Code,Section XI.

4.

NRC No.93-102, " Final Policy Statener.' on Technical Specification Improvements," July 23, 1993.

f

5. uene. - 32.c>4i P, soch Eev;ew fc,<- Edwla T. Ha4ch

)

hJu d u v T>o m e p ta,a Ltatts t cnu 2 Ltpda M So.feh/6f VaWe. Pcrhe%m Repics.w h "

f e

y A pril 19%.

HATCH UNIT 1 B 3.4-17 REVISION 0

i ECCS - Operating B 3.5.1 i

BASES

\\

BACKGROUND pumps without injecting water into the RPV. These test

{

(continued) lines also provide suppression pool cooling capability, as i

described in LCO 3.6.2.3, "RHR Suppression Pool Cooling."

i

\\

j Two LPCI inverters (one per subsystem) are designed to i

provide the power to various LPCI subsystem valves (e.g., inboard injection valves). This will ensure that a postulated worst case single active component failure, 4

during a design basis loss of coolant accident (which j

includes loss of offsite power), would not result in the low i

i pressure ECCS subsystems failing to meet their design function.

(Although an alternate power supply is available, 1

the LPCI subsystem (s) may not be capable of meeting its j

design function if the alternate power supply (ies) is in j

service; i.e., the LPCI inverters are bypassed.)

1 i

The HPCI System (Ref. 3) consists of a steam driven turbine pump unit, piping, and valves to provide steam to the i

turbine, as well as piping-and valves to transfer water from the suction source to the core via the feedwater system line, where the coolant is distributed within the RPV through the feedwater sparger. Suction piping for the i

system is provided from the CST and the suppression pool.

Pump suction for HPCI is normally aligned to the CST source i

to minimize injection of suppression pool water into the RPV. However, if the CST water supply is low, or if the suppression pool level is high, an automatic transfer to the suppression pool water source ensures a water supply for i

continuous operation of the HPCI System. The steam supply i

to the HPCI turbine is piped from a main steam line upstream of the associated inboard main steam isolatio ve.

HE )

The HPC1 System is designed to provide core cooling for a 1

wide range of reactor pressures (150 psig to psig).

l Upon receipt of an initiation signal, the HPCI turbine stop valve and turbine control valve open simultaneously and the 4

turbine accelerates to a specified speed. As the HPCI flow increases, the turbine governor valve is automatically adjusted to maintain design flow.

Exhaust steam from the i

HPCI turbine is discharged to the suppression pool. A full j

flow test line is provided to route water from and to the l

i CST to allow testing of the HPCI System during normal j

operation without injecting water into the RPV.

i j

The ECCS pumps are provided with minimum flow bypass lines, which discharge to the suppression pool. The valves in (continued)

HATCH UNIT 1 B 3.5-3 REVISION 6 t

1 1

1

ECCS - Operating B 3.5.1 i

BASES i

i APPLICABLE c.

Maximum hydrogen generation from a zirconium water l

SAFETY ANALYSES reaction is s 0.01 times the hypothetical amount that (continued) would be generated if all of the metal in the cladding i

surrounding the fuel, excluding the cladding i

surrounding the plenum volume, were to react; d.

The core is maintained in a coolable geometry; and e.

Adequate long tem cooling capability is maintained.

I The limiting single failures are discussed in Reference 9.

i The remaining OPERABLE ECCS subsystems provide the capability to adequately cool the core and prevent excessive 4

fuel damage.

l The ECCS satisfy Criteria 3 and 4 of the NRC Policy l

Statement (Ref. 12).

mW LC0 Each ECCS injectios/ spray subsystem an von ADS valves are required to be OPERABLE. The ECCS inj ion / spray subsystems are defined as the two CS subsystems, the two LPCI subsystems, and one HPCI System. The low pressure ECCS injection / spray subsystems are defined as the two CS subsystems and the two LPCI subsystems.

With less than the required number of ECCS subsystems OPERABLE, the potential exists that during a limiting design basis LOCA concurrent with the worst case single failure, the limits specified in Reference 10 could be exceeded. All low pressure ECCS subsystems nd ADS must therefore be OPERABLE to satisfy the single failure criterion required by.

Reference 10.

(Reference 9 takes no credit for HPCI.) HPCI must be OPERABLE due to risk consideration.

LPCI subsystems may be considered OPERABLE during alignment and operation for decay heat removal when below the actual IU5t low pressure permissive pressure in MODE 3, if capable of being manually realigned (remote or local) to the LPCI mode and not otherwise incperable. At these low pressures and decay heat levels, a reduced complement of ECCS subsystems should provide the required core cooling, thereby allowing operation of RHR shutdown cooling when necessary.

(continued)

HATCH UNIT 1 B 3.5-5 REVISION 0

i ECCS - Operating B 3.5.1 j!

i BASES 3

ACTIONS E.d LCO req /

/

(continued)

T uires even ADS valves to OPERABLEinorderdo providedhe ADS'functi,on. Jteference'9containstheresults 2

of ap/analyff's thaVevaluated the,effect of one, ADS val' e v

bejffg out ef se fee. fe'r this analysis, oper'ation of only f

f ix ADS valve ill provide the re lure jn)fabilityof th' quired depressur f

i Howev

,o rall re e ADS 1;(reduced, because as gle the OPERABLE' ADS valves couJd result in j

educ} n in,d6 pressurization capabji'ity. Therefore, /

)

enly alle6ed f pera fon is,ime is, based jm,sr a limited time / The 14 dat erencVll and has beeh found to,4bility st'udy cited /n f

C etion T a reli R

be acceptable thr,ough i

perat og expa l,ence.

E.1 and F.2 If a one 1 pressu/eECCSin'ection/sp y subsys is i

in erable n additi'on to o inoperabi ADS valve adequate re coo) ng is en'sured S injection)(ITY of H % and the he OPERAB q

remainJflglowpressureE' liability,Ms r/ spray subsystem.

How e active;all ECCS educedebecause a r, over l

si ith a des

(

compon t failure toncurren uired ECC,S,)qn ba is LOCA'could r uit in t minimum equipmenf not be og availa e.

Since th a hig ressure syste (ADS) a a low ssure su tem are

operable, i

mor restrictt e Comp on Time 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months s requir to 9

f alve to APERABLE p tus. Th Comple Time i p6ased on tore either the 1 pressure CS subs em or th y S a reliability study cited i eferen 1 and h g been f d

l to b /'cceptabM through eratin perienc a

1 E, \\ Q E 2-

\\

,,m WLA one. Ab5 wdve.

or i (h^oP" N " M c^

i If any Requi eD

Action and associated Completion Time of rag' ed, because.c W

5 Condition C ivalvesareinopYralde,iisnotmet,oriftwoormoreADS the plant must be brought to a d.

condition in which the LCO does not apply. To achieve this i

dew e Mu b.taa A status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and reactor steam done pressure reduced to i

" *."9

p s 150 psig within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . Entry into MODE 3 is not i

Ud Nes 4.a ca.FAW' required if the reduction in reactor steam dome pressure to j

or provWigh AM s 150 psig results in exiting the Applicability for the r

j g

Condition, and the s 150 psig is achieved within the given L(G rL& 0

y (continued) 4 HATCH UNIT 1 B 3.5-8 REVISION 1

ECCS - Operating B 3.5.1 BASES

.i uct E.i.

l ACTIONS I e..d 0.2 (continued) 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed Completion Times are reasonable, l

3 j

based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

FJ O

When multiple ECCS subsystems are inoperable, as stated in Condition H, the plant is in a condition outside of the accident analyses. Therefore, LCO 3.0.3 must be entered immedL tely.

i SURVEILLANCE SR 3.5.1.1 REQUIREMENTS The flow path piping has the potential to develop voids and pockets of entrained air. Maintaining the pump discharge lines of the HPCI System, CS System, and LPCI subsystems full of water ensures that the ECCS will perform properly, injecting its full capacity into the RCS upon demand. This will also prevent a water hammer following an ECCS initiation signal. One acceptable method of ensuring that the lines are full is to vent at the high points.

In addition, when HPCI is aligned to the suppression pool pump suction pressur)e.(instead of the CST, one acceptable me

)

The 31 day Frequency is based on the gradual nature of void buildup in the ECCS piping, the procedural controls governing system operation, and operating experience.

i l

SR 3.5.1.2 Verifying the correct alignment for manual, power operated, and automatic valves in the ECCS flow paths provides assurance that the proper flow paths will exist for ECCS operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position since these were verified to be in the correct position prior to locking, sealing, or securing. A valve that receives an initiation signal is allowed to be in a nonaccident position provided the valve will automatically reposition in the (continued)

HATCH UNIT 1 B 3.5-9 REVISION 1 i

ECCS - Operating 8 3.5.1 BASES SURVEILLANCE SR 3.5.1.12 (continued)

REQUIREMENTS The Frequency of 18 months is based on the need to perform the Surveillance under the conditions that apply just prior to or during a startup from a plant outage. Operating experience has shown that these components usually pass the SR when performed at the 18 month Frequency, which is based on the refueling cycle. Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

REFERENCES 1.

FSAR, Section 6.4.3.

2.

FSAR, Section 6.4.4.

3.

FSAR, Section 6.4.1.

4.

FSAR, Section 6 x 2.

5.

FSAR, Section 14.4.3.

6.

FSAR, Section 14.4.5.

7.

8.

FSAR, Section 6.5.

9.

NEDC-31376P, "E.I. Hatch Nuclear Plant Units I and 2

~

SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis,"

l December 1986.

10.

11. Memorandum from R.L. Baer (NRC) to V. Stello, Jr.

(NRC), " Recommended Interim Revisions to LCOs for ECCS Components," December 1, 1975.

12. NRC No.93-102, " Final Policy Statement on Technical S ecificagrovements,"_ July 23 1991 m

I3' s cac -3 zuiT, "sa766y w,w L Fa,J-L Haack huae WWct hn f, u n L4 s I and 2 u pcice4WI CQdrh4y / ElitC s

Vgtu.Fub, % ec2epdn w,ds",

AprM 1996 C

HATCH UNIT 1 B 3.5-16 REVISION 1

RCIC System B 3.5.3 4

B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM B 3.5.3 RCIC. System-

)

BASES i

BACKGROUND The RCIC System-is not part of the ECCS; however, the RCIC System is included with the ECCS section because of their j

similar functions.

The RCIC System is designed to operate either automatically i

or manually following reactor pressure vessel (RPV)

{

isolation accompanied by a loss of coolant flow from:the feedwater system to provide adequate core cooling and i

control of the RPV water level. Under these conditions, the i

High Pressure Coolant Injection (HPCI) and RCIC systems perform similar functions. The RCIC System design-requirements ensure that the criteria of Reference I are

{

satisfied.

i The RCIC Systes. (Ref. 2) consists of a steam driven turbine pus) unit, piping, and valves to provide steam to the tur)ine, as well as piping and valves to transfer water-from-i the suction source to the core via the feedwater system; line, where the coolant is distributed within the RPV through the feedwater sparger. Suction piping is provided i

from the condensate storage tank (CST) and the suppression i

pool.

Pump suction is normally aligned to the CST to j

minimize injection of suppression pool water into the RPV.

However, if the CST water supply is low, or the suppression i

pool level is high, an automatic transfer to the suppression pool water source ensures a water supply for continuous operation of the RCIC System. The steam supply to the turbine is piped from a main steam line upstream of the associated inboard main steam line isolation valv

!)3 C The RCIC System is designed to provide core col n or a wide range of reactor pressures (150 psig to H54 psig).

l Upon receipt of an initiation signal, the RCIC turbine accelerates to a specified speed. As the RCIC flow increases, the turbine control valve is automatically adjusted to maintain design flow.

Exhaust steam from the RCIC turbine is discharged to the suppression pool. A full flow test line is provided to route water from and to the CST to allow testing of the RCIC System during normal operation without injecting water into the RPV.

(continued)

HATCH UNIT 1 B 3.5-23 REVISION 6

LLS Valves B 3.6.1.6 BASES 1

APPLICABLE assumption that simultaneous S/RV openings occ)lt only on the 4

SAFETY ANALYSES iniAlayctuationforDBAs. Even though four

;ifiQ all four LLS S/RVs do not operate @in__any_DSAS/RVs a (continued (o@PEeifs! % ope ca.hcy vdilt uru-ct6 eta e/EV3 b e--L, t 1.6 an undhong Ac is 4.ccap4-dhu.- Du. Eef. bf l

LLS valves satisfy Criterion 3 of the NRC Policy Statement (Ref. 3).

I f

Three. c8 Four LLS valves are required to be OPERABL to Tsfy the LCO assumptions of the safety analyses (Ref.1. The i

requirements of this LCO are applicable to the mechanical j

and electrical / pneumatic capability of the LLS valves to l

function for controlling the opening and closing of the S/RVs.

1 APPLICABILITY In MODES 1, 2, and 3, an event could cause pressurization of the reactor and opening of S In MODES 4 and 5, the probability and consequences /RVs.

of these events are reduced due to the pressure and temperature limitations in these MODES.

i Therefore, maintaining the LLS valves OPERABLE is not required in MODE 4 or 5.

ACTIONS With e LL alve operabl he r ining RABL LS val s ar adequ toperf6rathe signed uncti H

ver the ov all rel tbility i reduc Th 14 d lit /

omp1 ion Ti takes to accou the r dundan capa aff ed by he resa ing LLS lves a the prgabi y

A.) a.nd A. 2 of an eve in whic the resa ing LL valve a)fli uld inadequa'.

wMk one LLS.

VC.ltic Lno(vr6k; no gckn L5 repruh j "a m d M ^ % ^' j M B A If two or more LLS valves are inoperab1MS--theX

' ' ^

M (155erehlTELSTiive c;nnst be rest:r;!-to-OPERABLE-statuD

f. N N L " D * -

Wiin-the rewired-Completion ~ Timer,"~thi pTant must be t

LLG valves m-b To capctW S provM%

rought to a PODE in which theTCO does not apply.

% necesa.q LL5 b 6sce rek reaco 2). _;

(continued)

HATCH UNIT 1 B 3.6-35 REVISION 1

LLS Valves B 3.6.1.6 i

BASES 1'

SURVEILLANCE SR 3.6.1.6.1 (continued)

REQUIREMENTS Adequate pressure at which this test is to be perfomed is consistent with the pressure recommended by the valve manufacturer.

SR 3.6.1.6.2 The LLS designated S/RVs are required to actuate automatically upon receipt of specific initiWon signals.

-A system functional test is performed to verify that the mechanical portions (i.e., solenoids) of the LLS function operate as designed when initiated either by an actual or simulated automatic initiation signal. The LOGIC SYSTEM FUNCTIONAL TEST in SR 3.3.6.3.6 overlaps this SR to provide l

complete testing of the safety function.

The 18 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and the potential for an unflanned transient if the Surveillance were perfomed with the reactor at power.

Operating experience has shown these components usually pass the Surveillance when performed at the 18 month Frequency.

Therefore, the Frequency was concluded to be acceptable from a reliability standpoint.

This SR is modified by a Note that excludes valve actuation.

This prevents a reactor pressure vessel pressure blowdown.

REFERENCES 1.

FSAR, Section 4.11.

2.

ASME, Boiler and Pressure Vessel Code,Section XI.

3.

NRC No.93-102, " Final Rolicy Statement on Technical Specification Improvements," July 23, 1993.

( d NIEht-32c)4 % "safeAy i24.vkw O Ed w La I.

HMck bl ac kct e h e e ~A+ uds l AJ E LL dc&A Sd4 /ReLaC Vg.)ve perke muco P

2cpremeb/ AycLL i996.

HATCH UNIT 1 B 3.6-37 REVISION 1 l

1 1

4 SLC Systeo B 3.1.7 4

4 BASES SURVEILLANCE SR 3.1.7.4 and SR 3.1.7.6'(continued) i REQUIREMENTS i

in the nonaccident position provided it can be aligned to the accident position from the control room, or locally by a i

dedicated operator at the valve control. This is acceptable since the SLC System is a manually initiated system.' This Surveillance also does not apply to valves that are locked, i

sealed, or otherwise secured in position since they'are i

verified to be in the correct position prior to locking, j

sealing, or securing. This verification of valve alignment does not require any testing or valve manipulation;urather, l

it involves verification that those valves capable of being mispositioned are in the correct position. This SR does not

{

apply to valves that cannot be inadvertently misaligned, such as check valves. The 31 day Frequency is based on engineering judgment and is consistent with the procedural controls governing valve operation that ensures correct valve positions.

SR 3.1.7.5 This Surveillance requires an examination of the sodium pentaborate solution by using chemical analysis to ensure that the proper concentration of boron exists in the storage tank (within Region A limits of Figures 3.1.7-1 and 3.1.7-2).

SR 3.1.7.5 must be performed anytime sodium pentaborate o" water is added to the storage tank solution to determine that the boron solution concentration is within the specified limits. SR 3.1.7.5 must also be performed any time the temperature is restored to within the Region A limits of Figure 3.1.7-2, to ensure that no significant boron precipitation occurred. The 31 day Frequency of this Surveillance is appropriate because of the relatively e ton variation of boron concentration between surveillancet.

SR 3.1.7.7 p

Demonstrating that each SLC System pum develops a flow rate a 41.2 gpm at a discharge pressure a psig ensures that l

pump performance has not degraded during the fuel cycle.

This minimum pump flow rate requirement ensures that, when combined with the sodium pentaborate solution concentration requirements, the rate of negative reactivity insertion from the SLC System will adequately compensate for the positive (continued)

HATCH UNIT 2 B 3.1-44 REVISION 5

/

5 LLS Instrumentation B 3'.3.6.3 ij v

l BASES W

sl I

ACTIONS L 1 (continued) i l

g w.-

function if any S/RV tailpipe pressure switch instrument "

)

channel becomes inoperable.

Therefore, it is acceptab 4 for plant operation to continue with only one tailpipe pressure"",7 l

switch OPERABLE on each S/RV. However, this is only M tW g i acceptable provided each LLS valve is maintaining initiation l

capability.

(Refer to Required Action A.1 and D.1 Ba ' y.

4 l

'RequiredActionB.I'[equiresrestorationofthetN j

pressure switches to OPERABLE status prior to entert

. Mj W l

MODE 2 or 3 from MODE 4 to ensure that all switcLes[are p

. OPERABLE at the beginning of a reactor startup (thisVis W l

because the switches are not accessible during plant @#b l

operation). The Required Actions do not allow placingstbec channel in trip since this action could result in aiLLS! a valve actuation. As noted, LCO 3.0.4 is not applicableh.'"

i thus, allowing entry into MODE 1 from MODE 2 with inopirable channels. This allowance is needed since the channels'enly have to be repaired prior to entering MODE 2 or MODE:!3).from-l MODE 4.

Yet, LCO 3.0.4 would preclude entry into M00Eik -

from MODE 2 since the Required Action does not allowy:$r unlimited operations.

"4 l

l

, +-

,; W i'

L1

~

My J

%=

i A failure of two pressure switch channels associated with.

{

one S/RV tailpipe could result.a the loss of the LLS T function (i.e., multiple actuations of the S/RV would go.

undetected by the LLS_ logic)ing-an-even$the-S/RVs-are a s

However, W

orgaritiidln-groups-andrdir trgroups-of-S/RVs-4hd ts.s dU

-i ni t 1 ally-open-(setpoin t s-are-a t-s ame-set t i ngs -for-a-tot al

-of-1-1-SfRVs in three iirs,@s)-; Therefore, it would be-very i

S gh.

unlikely that a single S/RV would be required to are all the LLS logic. Therefore it is acceptable to allow 14 days to i

restore one pressure s, witch of the associated S/RV to-OPERABLE status (Required Action C.1). However, this i

allowable out of service time is only acceptable provided j

aach LLS is maintaining initiation capability (Refer to-i Required Action A.1 and D.1 Bases).

If one inoperable <

tailpipe pressure switch cannot be restored to OPERABLE' status within the allowable out of service time, Condition D aust be entered and its Required Action taken. The Required Actions do hot allow placing the channels in trip since this

}

action could result in a LLS valve actuation.

j (continued) l HATCH UNIT 2 B 3.3-191 REVISION 1

s S/RVs B 3.4.3 i

)

B 3.4 REACTOR COOLANT SYSTEM (RCS)

B 3.4.3 Safety / Relief Valves (S/RVs)

BASES 4

i BACKGROUND The ASME Boiler and Pressure Vessel Code requires the-reactor pressure vessel be protected from overpressure-during upset conditions by self-actuated safety valves. As part of the nuclear pressure relief system, the size and number of S/RVs are selected such that peak pressure in the-nuclear system will not exceed the ASME Code limits for the reactor coolant pressure boundary (RCPB).

The S/RVs are located on the main steam lines between thei reactor vessel and the first isolation valve within ther drywell. The S/RVs can actuate by either of two modes:.the safety mode or the relief mode.

In the safety mode (ord spring mode of operation), the spring loaded pilot valve i

opens when steam pressure at the valve inlet overcomes the-

{

spring force holding the pilot valve-closed. Opening the pilot valve allows a pressure differential to develop 'across the main valve piston and orsns the main valve.

This 1

satisfies the Code requirement.

i Each S/RV discharges steam through a discharge line to a point below the minimum water level in the suppressian pool.

The S/RVs that previde the relief mode are the low-low set.

(LLS) valves and the Automatic Depressurization System (ADS) valves.

The LLS requirements are specified in LCO 3.6.1.6,

" Low-Low Set (LLS) Valves," and the ADS requirements are specified in LCO 3.5.1, "ECCS - Operating."

l APPLICABLE The overpressure protection system must accommodate the most SAFETY ANALYSES severe pressurization transient.

Evaluations have determined that the most severe transient is the closure of all main steam isolation valves (MSIVs), followed by reactor scram on high neutron flux (i.e., failure of the direct' scram associated with MSIV position) (Ref. 1).

For the purpose of the analysas 11 S/RVs are assumed to operate in the safety mode. The a lysis results demonstrate that the design S/RV capacity i capable of maintaining reactor pressure well below the ASME Code limit of 110% of vessel (continued)

HATCH UNIT 2 B 3.4-13 REVISION O

i S/RVs

]

B 3.4.3 BASES I

APPLICABLE design pressure (110% x 1250 psig - 1375 psig).

Sensitivity i

SAFETY ANALYSES analyses have demonstrated that 8 or 9 S/RVs operating in

{

(continued) the pressure relief moJe will maintain the reactor vessel below 1375 psig. This LC0 helps to ensure that the acceptance limit of 1375 psig is met during the Design Basis Event.

j From an overpressure standpoint, the design basis events are i

bounded by the MSIV closure with flux scram event described i

above.

Reference 2 discusses additional events that are expected to actuate the S/RVs.

S/RVs satisfy Criterion 3 of the NRC Policy Statement (Ref. 4).

.A i

[ frt4'eb/geG4 Vc luc' y.;fc3 go 4 g 3/g,vg l

LCO Th'dAnf4tyNtW%c, NWrequ red te be OPERABLE to satisfy the assumptions of the safety analysis (Refs_._1 -d 2), although margins to the ASME Vessel nd 6 overpreiMreLimitaresubstantial. The requirements of tt.!t LCO are applicable only to the capability of the S/RVs to w,echanically open to relieve excess pressure when the 1

lift setpoint is exceeded (safety function).

i i

The S/RV setpoints are established to ensure that the ASME l

Code limit on peak reactor pressure is satisfied. The ASME i

Code specificttions require the lowest safety valve setpoint to be at or below vessel design pressure (1250 psig) and the j

highest safety valve to be set so that the total accumultted pressure does not exceed 110% of the design pressure for overpressurization conditions. The transient evaluations in the FSAR are based on these setpoints, but also include the i

additional uncertainties of

3% of the nominal setpoint drift to provide an added degree of conservatism.

Operation with fewer valves-0PERABLE than specified, or with setpoints outside the ASME limits, could result in a more 2

severe reactor response to a transient than predicted, possibly resulting in the ASME Code limit on reactor pressure being Exceeded.

l i

i l

(continued)

HATCH UNIT 2 B 3.4-14 REVISION 0

l

}

S/RVs

{

B 3.4.3 j

BASES (continued)

A i

APPLICABILITY In MODES 1, 2, and 3, aii S/RVs must be OPERABLE, since considerable energy may be in the reactor core and the 3

H aiting design basis transients are assumed to occur-in i

these MODES. The S/RVs may be required to provide pressure reliaf to discharge energy from the core until such time, j

that the Residual Heat Removal (RHR) System is capable of-dissipating the core heat.

In MODE 4, decay heat is low enough for the RHR System to j

provide adequate cooling, and reactor-pressure is low >enough that the overpressure limit is unlikely to be approached 1by j

assumed operational transients or accidents.

In MODE 5,sthe-reactor vessel head is unbolted or removed and the reactore 4

i is at atmospheric pressure. The S/RV function ir, not needed during these conditions.

I I

ACTIONS M

\\

m W

t safe fu io one Vi parab.)

th!

esa ing 0 RABL S/ Vs are c ble f proyfding,the<

ne ssary verp ssu protec on. systeer/, thefoverallf

\\

oweve i

r iabi yo the ressure elle is reduced because dditi al.ilur in th resa ing OP RABLE /RVs codld

\\

resu in allu to ade atel relie pre ure durf'ng a 11 tin even For t s re on, c tinue operat on is i

rei ed fo a limi d ti only Th 14 d Comple on T to stor the in erable /RV q

ERABL status s ba d on e rel of cap lity the resa ng S/R, th ow p abil y of event qui ng j

i S/

ctuat n, a a re onabl time t comple th j

R ired ion.

A.t o. d A L

(=d B&-

H2O ML f

sa/a Q,au p ca2*~ & 4 n w ce " re i

Withb& e th

-- - S ' A inoperable, a transient may result

( (4f the :;afety functica of the 1;;;;;;:b1:in the M: n

" F " # "

"H +'

)

S/RV-eannet be-

] xasured-to-OPERABt.E status-within-the-associated-Completion-eeasd M D' O fd*N Of Re;" ired Act-ion-Arir-or

-the-safety funct_iorro Ti:::

rw'et (twej r_,re S/R" b incier:bl he plant must be brought are ca%pablo 'E

~

to a MODE in which the LCO does not apply. To achieve this p r+vcd

'N "1 status, the plant must be brought to MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months c ur ce w '

c&

"'**'R

]

,Yc V

(continued)

HATCH UNIT 2 B 3.4-15 REVISION 0

l S/RVs B'3.4.3 i

l l

BASES i

SURVEILLANCE SR 3.4.3.2 (continued)

Me REQUIREMENTS i

recommended by the valve manufacturer). Adequate steam flow 4 is represented by at least 1.25 turbine bypass valves.open i

or total steam flow a: IE6 lb/hr. Plant startup is allowedi e

prior to performing this test because valve.0PERA81LITYFand n e i

i the setpoints for overpressure protection are verifi ASME Code requirements, prior to valve installation ed,:perc

.D J Therefore,- this SR is modified by a Note that statessthe Y v i

Surveillancecis not required to be: performed:until 12, hours; T after reactor steam pressure and flowiare' adequate:to@M 4 perform the test. The 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowed for manual actuation after the required pressure is reached:is sufficient;H O '

to M

}

achieve stable conditions for testing'and provides a reasonable time to complete the SR.

If a valve fails toj.

s actuate due only to the failure of the solenoid buttis FL.

capable of opening on overpressure,- thessafety functionrof j

the S/RV is considered OPERABLE.

"b W'~. 'y

}'

.v The 18 month Frequency was developed-based on the S/RVith k I

required by the ASME Boiler and Pressure Vessel Code W.$

Section XI (Ref. 3). Operating experience has shown thatt j

these components usually pass the Surveillance when z

~:

i perforised at the 18 month Frequency. Therefore, ther.e j

Frequency was concluded to be acceptable from a reliability standpoint.

~

i 1

l REFERENCES 1.

FSAR, Supplement SA.

a 2.

FSAR, Section 15.

3.

ASME, Boiler and Pressure Vessel Code, Sectio 9 XI.

4.

NRC No.93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.

L.n.

me -szow, " safe % ecuu L casou x. Hua Pad l

Muttuu Pc wer-PLLsd-LLdh, i Ad 2 LA 64.[eAq/Lud Valve Perfer%cc aphe.m.sts->"

AprCL r9%

HATCH UNIT 2 B 3.4-17 REVISION 0

i ECCS - C g rating B 3.5.1 i

BASES BACKGROUND pumps without injecting water into the RPV. These test (continued) lines also provide suppression pool cooling capability, as described in LCO 3.6.2.3, "RHR Suppression Pool Cooling."

Two LPCI inverters (one per subsystem) are designed to i

provide the power to various LPCI subsystem valves i

(e.g.

inboard injection valves).

This will ensure that a i

postulatedworstcasesingleactivecomponentfailure, during a design basis loss of coolant accident (which includes loss of offsite power), would not result in the low pressure ECCS subsystems failing to meet their design function.

the low pres (While an alternate power supply is available;.

sure ECCS subsystems may not be capable off' meeting their design function if the alternate power supply is in service.)

The HPCI System (Ref. 3 consists of a steam driven turbine pus) unit, piping, and alves to pro Me steam to the i

tur)ine, as well as piping and valves i.o transfer water,from the suction source to the core via the feedwater systeu line, where the coolant is distributed within the RPV i

through the feedwater sparger. Suction piping for the system is provided from the CST and the suppression pool.

Pump suction for HPCI is normally aligned to the CST source i

to minimize injection of suppression pool water into the RPV. However, if the CST water supply is low, or if the suppression pool level is high, an automatic transfer to the suppression pool water source ensures a water supply for continuous operation of the HPCI System. The steam supply to the HPCI turbine is piped from a main steam line upstream of the associated inboard main steam isolation valve The HPCI System is designed to provide core foo iig for a wide range of reactor pressures (162 psid to M49 psid, l

vessel to pump suction). Upon receipt of an initiation signal, the HPCI turbine stop valve and turbine control valve open simultaneously and the turbine accelerates to a specified speed. As the HPCI flow increases, the turbine governor valve is automatically adjusted to maintain design flow.

Exhaust steam from the HPCI turbine is discharged to the suppression pool. A full flow test line is provided to route water from and to the CST to allow testing of the HPCI System during normal operation without injecting water into the RPV.

The ECCS pumps are provided with minimum flow bypass lines, which discharge to the suppression pool.

The valves in (continued)

HATCH UNIT 2 B

5-3 REVISION 5

4 l

ECCS - Operating i

B 3.5.;

i i

BASES e

I i

APPLICA8LE c.

Maximum hydrogen generation from a zirconium water i

SAFETY ANALYSES reaction is s 0.01 times the hypothetical amount that i

(continued) would be generated if all of the metal in the cladding surrounding the fuel, excluding the cladding surrounding the plenum volume, were to react; j

d.

The core is maintained in a coolable geometry; and I

e.

Adequate long tem cooling capability is maintained.

j The limiting single failures are discussed in Reference 10.

The remaining UPERA8tE ECCS subsystems provide the capability to adequately cool the core and prevent excessive fuel damage.

The ECCS satisfy Criteria 3 and 4 of the NRC Policy j

Statement (Ref.13).

j EachECCSinjection/spraysubsystemanhv#

LCO en/ ADS valves are required to be OPERA 8LE. The ECCS injectTiii/ spray l

subsystems are defined as the two CS subsystems, the two j

LPCI subsystems, and one HPCI System. The low pressure ECCS i

injection / spray subsystems are defined as the two CS subsystems and the two LPCI subsystems.

With less than the required number of ECCS subsystems OPERABLE, the potential exists that during a limiting design basis LOCA concurrent with the worst case single fai ure, the limits specified in Reference 11 could be exceeded. All low pressure ECCS subsystems and ADS must therefore be OPERABLE to satisfy the single failure criterion required by Reference 11.

(Reference 10 takes no credit for HPCI.)

HPCI must be OPERABLE due to risk consideration.

LPCI subsystems may be considered OPERABLE during alignment and operation for decay heat removal when below the actual RHR low pressure pemissive pressure in MODE 3, if capable of being manually realigned (remote or local) to the LPCI mode and not othemise inoperable. At these low pressures and decay heat levels, a reduced complement of ECCS subsystems should provide the required core cooling, thereby allowing operation of RHR shutdown cooling when necessary.

(continued) 1 HATCH UNIT 2 B 3.5-5 REVISION 0

ECCS - Operating j

B 3.5.1 1

BASES

'[ /

ACTIONS (continued)

/

i Th LC0 q re save S va s

tMV0PERA8LE i orderk i

ovi t

func on.

erenIce JC co dains! e res its h

i of ye s tha evalu ed thf effect f on ADS va e-beIng a 1

t d se ca.

r th ap61ys o

ationt only ix 5 alves ill p vide he v' equi

-d ressu ation. f ve ; ove 1 re abil y thefADS i reduc ;-because si le--f ure the E

LE ADS'va es c resulfin i

a uct in pres i

ion apab ity, ereforg; rati

-is y al for lim ed t r

4 day /

The )d:in /

)

omp1 ion me:1 as on rel ili study ite j

-f ~ o Ref nce

~ a as en und o be ceptab th rati expe en i

j F.

and

.2, y

a n

ou ess EC injecyon/spr sub stem t 4

ino ra esin it n;t one inpperabi Sv ve, ate c ec li is e u by the OPERA 8 TV o HPCI' the ing~ ow p ssu ECCS jecti /spr subsy ear f

ver,- ver E S reli 111ty s red ed-use:a'

(

l s gle etiv copponent ilure oncu nt wi a de v

ast DCA oulf resul in th ini requi ECC equ not ing ailabl. Si e both high res re i

s tem S and a ow pre ure syst are i per e, a re-str etive emplet n Ti of 72 ours^

red to res re ther-e low ress ECCS ubsystem-or he ADS v ve OPE LE st us.

is I etio dime s bas on re abili y stud cited n Refe nce 1 and as be found to ac ptable hroug operat ex ten j

E.\\

a d E 2-

/

~

l Qtg m Am vale

-4<1-e ^ 6 O f.p ru a % u M M ( If any Requi -

Actinynd associated Completion Time of

" I'["'h Condition Cp C n.,is not met, or if two or more-ADS 1

\\ valves are noperable, the plant must be brought to a ha Tcondition in which the LCO does not apply. To achieve this dw awa4clhr %. status, the plant must be brought to at least MODE 3 within M*W M 4Ds

/12 hours and reactor steam done pressure reduced to s 150 psig within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . Entry into MODE 3 is not Va w are eA W required if the reduction in reactor steam dome pressure to t

of vrvedgh Aos s 150 psig results in exiting the Applicability for the 44, Condition, and the s 150 psig is achieved within the given ReSe<t w (6-(continued)

(

HATCH UNIT 2 B 3.5-8 REVISION 1

ECCS - Operating 8 3'.5.1 BASES m

/ G..I C J C. 2-ACTIONS (11 rd 0.2, (continued) 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The allowed Completion Times are reasonable,-

l based on operating experience, to reach the reouired. plant conditions from full power conditions in an orterly~ manner and without challenging plant systems.

F.I a a

.... l d When multiple ECCS subsystems.are' inoperable, as stated 61n Condition H, the plant.is in a condition outside of ther accident analyses. Therefore, LCO 1-0.3 must be entered) immediately.

i e ;:t SURVEILLANCE ~

SR 3.5.1.1 REQUIREMENTS The flow path piping has the potential to develop voldssand pockets.of entrained air. Maintaining the pump discharge lines of the HPCI System,:CS System, and LPCI subsystemsq full of water ensures that the ECCS will perform properl,.

injecting its full capacity.into the.RCS upon~ demand s is will also prevent a water hammer following an ECCS initiation signal. One-acceptable method of ensuring-that the lines are full is to vent at the high points.

In.

addition, when-HPCI is aligned to the suppression pool ~,~

(instead of the CST), one acceptable method is to monitor pump suction pressure. The 31 day Frequency is based on the gradual nature of void buildup in the ECCS piping, the procedural controls governing system operation, and operating experience.

SR 3.5.1.2 Verifying the correct alignment for manual, power operated; and automatic valves in the ECCS flow paths provides assurance that the proper flow paths will exist for ECCS operation. This SR does not apply to valves that are -

locked, sealed, or otherwise secured in position since these were verified to be in the correct position prior to locking, sealing, or securing. A valve that receives an initiation signal is allowed to be in a nonaccident position provided the valve will automatically reposition in the -

(continued)

HATCH UNIT 2 8 3.5-9 REVISION 1

ECCS - Oporating B:3.5.1

.w

)w$,

BASES REFERENCES 6.

FSAR, Section 15.1.40.

.k (continued) n 7.

FSAR, Section 15.1.33.

8.

N:i

.g" 9.

FSAR, Section 6.3.3.

s'

10. NEDC-31376P,"E.I.HatchNuclearPlantUnits1anU2 SAFER /GESTR-LOCA Loss-of-Coolant Analysis,"

December 1986.

" %, f 1

11.