Text

Proposed Tech Specs Allowing Plant to Operate at Uprated Power Level of 2558 Mwt
	ML20077S033
Person / Time
Site:	Hatch
Issue date:	01/13/1995
From:	; GEORGIA POWER CO.
To:	;
Shared Package
ML19311B658	List: HL-4724, Application for Amends to Licenses DPR-57 & NFP-5,proposing Changes to Ts,To Allow Plant to Operate at Uprated Power Level of 2558 Mwt.Also,Forwards Proprietary Power Uprate SAR for E I Hatch Plant Units 1 & 2; ML20077S033;
References
	NUDOCS 9501230239
	Download: ML20077S033 (49)
	v • d • e

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(2) Pursuant to the Act and 10 CFR Part 70. Georgia Power Company to receive, possess and use at any tine special nucicar material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Final Safety Analysis Report, as suppicmented and amended; (3) Pursuant to the Act and 10 CFR Parts 30, 40 and 70 Georgia Power Company to receive, i possess and use at any time any byproduct, source and special nuclear material as sealed neutron sources for reactor startup, sealed sources for reactor instrumentation and radiation monitoring equipment calibration, and as fission detectors in amounts as required; )

(4) Pursuant to the Act and 10 CFR Parts 30. 40 and 70. Georgia Power Ccmpany to receive, possess and use in amounts as recuired any byproduct, source or specia) nuclear material without rastriction to chemical or physical .

fom. for sample analysis or inst ument t O calibration or asst,ciated with rad',oactive apparatus or components; E

g (5) Pursuant to the Act and 10 CFR Parts 30 and 70, e Georgia Power Company to possess, but not .

separate, such byproduct and special nuclear g' .

materials as may be procuced by the operation ,

I of the facility. =

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C. This license shall be deemed to contain and is subject to the conditions specified in the following Consission regulations in 3 10 CFR Chapter 1: Part 20 Sectien 30.34 of P&rt 30. Section 40.41 of Part 40, Sections 50-54 and 50-59 of Part 50, and Section 70.32 of Part 70; is subject to all applicable provisions of the Act and to the rules, regulations, anc orders of the Cosnission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) Maximum Power t.evel The Georgia Power Company is authorized to l

operate the facility at steady state reactor l

core power levels not in excess of 34M megawatts thermal. (2ff8 h (2) Technical Specifications The Technical Specifications contained in appendices A and B, as revised through Amendment No. /55 are O hereby incorporated in the license. The 1scensee shall operate tne facility in accordance with the Tecnnical Specifications.

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I Definitions 1.1 ;

1.1 Definitions OPERABLE - OPERABILITY instrumentation, controls, normal or emergency electrical power, cooling and seal water, (continued) lubrication, and other auxiliary equipment that are required for the system, subsystem, division, component, or device to perform its specified :

safety function (s) are.also capable of performing -

their related support function (s). !

PHYSICS TESTS PHYSICS TESTS shall be those tests performed to l measure the fundamental nuclear characteristics of ;

the reacte core and related instrumentation. '

These tests are:

a. Described in Section 13.6, Startup and Power Test Program, of the FSAR; -

b. Authorized under the provisions of 10 CFR 50.59; or

c. Otherwise approved by the Nuclear Regulatory Commission.

O :

[

RATED THERMAL POWER RTP shall be a total reactor c eat transfer (RTP) rate to the reactor coolant of MWt.

2fff i REACTOR PROTECTION The RPS RESPONSE TIME shall be that time interval '

SYSTEM (RPS) RESPONSE from when the monitored parameter exceeds its RPS TIME trip setpoint at the channel sensor until !

de-energization of the scram pilot valve

solenoids. The response time may be measured by '

means of any series of sequential, overlapping, or total steps so that the entire response time is measured. ,

O (continued) -

HATCH UNIT 1 1.1-5 REVISION /U

l l

SLC System 3.1.7 IO V SURVEILLANCE REOUIREMENTS :

SURVEILLANCE FREQUENCY SR 3.1.7.5 -(continued) Once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after solution temperature is ,

restored within the Region A 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.

SR 3.1.7.7 Verify each pump develops a flow rate In accordance a gpm at a discharge pressure with the psig. Inservice 3 2 (V 120I Testing Program SR 3.1.7.8 Verify flow through one SLC subsystem from 18 months on a pump into reactor pressure vessel. STAGGERED TEST BASIS SR 3.1.7.9 Verify all heat traced piping between 18 months storage tank ano pump suction is unblocked.

E ,

Once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after pump suction piping temperature is restored within the Region A limits of Figure 3.1.7-2 (continued) ,

3.1-22 REVISION A HATCH UNIT 1

1 RPS Instrumentation ,

3.3.1.1 -i i

tante 3.3.1.1 1 (page 2 of 3) i seector erotection System instrumentation I

APPLICABLE CONDITIONS ,

MCSES OR REQUIRED REFERENCED-ctNER CHANNELS F RCpl

. SPECIFIED PER TRIP REOUIRED SURVEILLANCE ALLOWASLE FUNCT!CN CONelTIONS SYSTEM ACT!Du D.1 REQUIREMENTS VALUE Z. Average Power Range .

Monitors (continued)

c. Fined Neutron 1 2- F SR 3.3.1.1.1 s 1201 RTP. I Flum - nign SR 3.3.1.1.2 SR 3.3.1.1.8 4 SR 3.3.1.1.9 SR 3.3.1.1.10

SR 3.3.1.1.15 t

s. Downscale ! I r SR 3.3.1.1.5 2 6.2 atP Sc 3.3.1.1.8 SR 3.3.1.1.15

e. Inoo 1.2 2 C SR 3.3.1.1.8 NA ;

SR 3.3.1.1.9 SR 3.3.1.1.15 1005

3. aesctor vesset steam 1.2 2 C SR 3.3.1.1.1 s 1054 psig.

l!

Dome Pressure - High SR 3.3.1.1.9 ,

. SR 3.3.1.1.13 SR 3.3.1.1.15 p 1,2 2 C SR 3.3.1.1.1 e o inenes i

\ 6 Reactor vessel water '

Level - Low, Level 3 SR 3.3.1.1.9 SR 3.3.1.1.13 3 SR 3.3.1.1.15 i

! Main Steam Isotatic+ 1 8 F SR 3.3.1.1.9 1 10% closso valve - CL osure SR 3.3.1.1.13 l SR 3.3.1.1.15-

6. DryweL L Pressure - Nigh 1.2 2 C SR. 3.3.1.1.1 1 1.92 psig SR 3.3.1.1.9 SR 3.3.1.1.13 f SR 3.3.1.1.15 I

(continvec) ;

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REVISION A I HATCH UNIT 1 3.3-7

ATWS-RPT Instrumentation i

3.3,4.2 O SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.4.2.2 Perform CHANNEL FUNCTIONAL TEST. 92 days i

SR 3.3.4.2.3 Perform CHANNEL CALIBRATION. The 18 months Allowable Values shall be:

a. Reactor Vessel Water Level -- Low Low, Level 2: 2 -47 inches; and ;

b Re r Steam Dome Pressure -- High:

s psig. ;

ll135 SR 3.3.4.2.4 Perform LOGIC SYSTEM FUNCTIONAL TEST 18 months !

including breaker actuation.

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3.3-32 REVISION A HATCH UNIT 1 ,

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i LLS Instrumentation 3.3.6.3 r"'

fable 3.3.6.3 1 (page 1 of il Low Low set Instrumentation RE00!AED CMANNELs PER SURVEILLANCE ALLOWAgLE FUNCT!0u FUNCT10u REQUI AEMENis VALUE

1. Reactor steen Dome Pressure Migh 1 per LLs valve 54 3.3.6.3.1 5 1054 psig st 3.3.6.3.4 sa 3.3.6.3.5 sa 3.3.6.3.6 ( 0 SS

2. Low Low set Pressure setpoints 2 per LLs valve sa 3.3.6.3.1 Lows sa 3.3.6.3.4 Coen s 1005 osis

$t 3.3.6.3.5 Close s 857 psig sa 3.3.6.3.6 Medlun* Low Open 1 1020 psie close s 872 pois Medium-N igh t ;

open s 1035 pois Close s 887 psig Hight Coen s 1045 pois Close 1 897 pois r Teilpipe Pressure switch 2 per s/RV sa 3.3.6.3.2 2 80 psis one

3. s 100 psig 54 3.3.6.3.3 sa 3.3.6.3.5 54 3.3.6.3.6 ,

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i 3.3-64 REVISION A HATCH UNIT 1

Recirculation Loops Operating 3.4.1

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V es CPS AAflous eseT ALLOpuGO A06ecue M

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\ l p pc,- 7"b D IWC by g 4

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N CPG AATGOut ALLOuuBO As O g .

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M SS r de 3

came nO= m .AnDi F

Figure 3.4.1-1 (Page 1 of 1)

Power-Flow Operating Map with One Reactor i i

Coolant System Recirculation loop in Operation O

3.4-4 REVISION A HATCH UNIT 1

3/RVs 3.4.3 O l u

SURVEILLANCE REOUIREMENTS SURVEILLANCE FRE0VENCY I

SR 3.4.3.1 Verify the safety function lift setpoints In accordance of the S/RVs are as follows: with the Inservice i Number of Setpoint Testing Program S/RVs (osio) 35.3

~

4 6080=32.4l LO t 4 1090 = 32.7 \ iso i 33.4 3 (1100=33.0j 33,9 j Following testing, lift settings shall be within = l'A. ,

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 '

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

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

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3.4-8 REVISION A HATCH UNIT 1 ,

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Reactor Steam Dome Pressure 3.4.10 03.4 REACTOR COOLANT SvSTEM (RCS) 3.4.10 Reactor Steam Dome Pressure 1068 LC0 3.4.10 The reactor steam dome pressure shall be s psig.

APPLICABILITY: MODES 1 and 2.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Reactor steam dome A.1 Restore reactor steam 15 minutes pressure not within dome pressure to limit, within limit.

B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time aat met-O-

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.10.1 Verif reactor steam dome pressure is 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months s 0 0 psig.

LOTP O

HATCH UNIT 1 3.4-25 REVISION A

ECCS - Operating 3.5.1 m

SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY

~'

SR 3.5.1.6 -------------------NOTE--------------------

Only required to be performed prior to entering MODE 2 from MODE 3 or 4, when in MODE 4 > 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

f Verify each recirculation pump discharge 31 days valve cycles through one complete cycle of full travel or is de-energized in the closed position.

SR 3.5.1.7 Verify the following ECCS pumps develop the In accordance specified flow rate against a system head with the corresponding to the specified reactor Inservice pressure. Testing Program SYSTEM HEAD NO. CORRESPONDING 0F TO A REACTOR

(]) SYSTEM ELOW RATE PUMPS PRESSURE OF CS 2: 4250 gpm 1 2 113 psig i LPCI a 17,000 gpm 2 a 20 psig

-SR 3.5.1.8 -------------------NOTE-------------------- .

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.

105I8 Verify, with reactor pressure s 1020 psig 92 days -

and 2: 920 psig, the HPCI pump can evelop a flow rate a 4250 gpm against a system head corresponding to reactor pressure.

(continued) t O

HATCH UNIT 1 3.5-5 REVISION A

RCIC System 3.5.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.3.1 Verify the RCIC System piping is filled 31 days with water from the pump discharge valve to the injection valve.

SR 3.5.3.2 Verify each RCIC System manual, power 31 days i operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.

SR 3.5.3.3 ------------------NOTE--------------------

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.

n loS7 V Verify, with reactor pressure s psig 92 days and 2 920 psig, the RCIC pump can evelop a flow rate 2 400 gpm against a system head corresponding to reactor pressure.

SR 3.5.3.4 -------------------NOTE--------------------

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.

Verify, with reactor pressure s 165 psig, 18 months the RCIC pump can develop a flow rate 2 400 gpm against a system head corresponding to reactor pressure.

(continued)

O HATCH UNIT 1 3.5-12 REVISION A

SLC System B 3.1.7 (a') BASES SURVEILLANCE SR 3.1.7 1 UCI REQUIREMENTS (continued) Demonstrating that each SLC System pum develops a flow rat" 2 41.2 gpm at a discharge pressure 2 19 psig ensures that 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 reactivity effects encountered during power reduction, cooldown of the moderator, and xenon decay. This test confirms one point on the pump design curve and is indicative of overall performance. Such inservice inspections confirm component OPERABILITY, trend performance, and detect incipient failures by indicating abnormal performance. The Frequency of this Surveillance is in accordance with the Inservice Testing Program.

SR 3.1.7.8 and SR 3.1.7.9

. These Surveil tances ensure that there is a functioning flow path from the sodium pentaborate solution storage tank to O the RPV, including the firing of an explosive valve. The replacement charge for the explosive valve shl1 be from the same manufactured batch as the one fired or from another batch that has been certified by having one of that batch successfully fired. The pump and explosive valve tested should be alternated such that both complete flow paths are tested every 36 months at alternating 18 month intevals.

The Surveillance may be performed in separate steps to prevent injecting boron into the RPV. An acceptable method for verifying flow from the pump to the RPV is to pump demineralized water from a test tank through one SLC subsystem and into the RPV. 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 Frequency was concluded to be acceptable from a reliability standpoint.

p J

(continued)

HATCH UNIT 1 B 3.1-45 REVISION A

i Primary Containment Isolation Instrumer.tation B 2.3.6.1 ,

(q) BASES APPLICABLE 1.c. Main Steam line now - Hioh (continued)

SAFETY ANALYSES, LCO, and Flow - High Function for each unisolated MSL (two channels APPLICABILITY per trip system) are available and are required to be OPERABLE so that no single instrument failum will preclude detecting a break in any individual MSL.

The Allowable Value is chosen to ensure that offsite dose limits are not exceeded to the break. The Allowable n Value corresponds to :s psid, which is the parameter I A monitored on control roo instruments.

I t(s This Function isolates the Group 1 valves.

1.d. Condenser Vacuum - Low The Condenser Vacuum - Low Function is provided to prevent overpressurization of the main condenser in the event of a loss of the main condenser vacuum. Since the integrity of the condenser is an assumption in offsite dose calculations, the Condenser Vacuum - Low Function is assumed to be O OPERABLE and capable of initiating closure of the MSIVs.

V The closure of the MSIVs is initiated to prevent-the addition of steam that would lead to additional condenser pressurization and possible rupture of the diaphragm installed to protect the turbine exhaust hood, thereby preventing a potential radiation leakage path following an accident.

Condenser vacuum pressure signals are derived from four pressure transmitters that sense the pressure in the condenser. Four channels of Condenser Vacuum - Low Function are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.

The Allowable Value is chosen to prevent damage to the condenser due to pressurization, thereby ensuring its integrity for offsite dose analysis. As noted (footnote (a) to Table 3.3.6.1-1), the channels are not required to be OPERABLE in MODES 2 and 3 when all turbine stop valves (TSVs) are closed, since the potential for condenser overpressurization is minimized. Switches are provided to manually bypass the channels when all TSVs are closed.

This Function isolates the Group 1 valves.

O (continued)

B 3.3-153 REVISION D HATCH UNIT 1

Primary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE 3.a. 4.a. HPCI and RCIC Steam Line Flow - Hioh SAFETY ANALYSES, (continued)

- LCO, and APPLICABILITY recirculation and MSL breaks. However, these instruments prevent the RCIC or HPCI steam line breaks from becoming bounding.

The HPCI and RCIC Steam Line Flow - High signals are initiated from transmitters (two for HPCI and two for RCIC) that are connected to the system steam lines. Two channels .

of both HPCI and RCIC Steam Line Flow - High Functions are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.

The Allowable Values are chosen to be low enough to ensure that the trip occurs to prevent fuel damage and maintains gog gt? the MSLB event as he bounding event. The Allowable Valu t correspond to s 1 inches water column for HPCI and s 190 inches water column for RCIC, which are the parameters '

monitored on control room instruments.

~

These Functions isolate the Group 3 and 4 valves, as Q appropriate.

3.b. 4.b. HPCI and RCIC Steam Sucoly Line Pressure - Low Low MSL pressure indicates that the pressure of the steam in the HPCI or RCIC turbine may be too low to continue i operation of the associated system's turbine. These isolations are for equipment protection and are not assumed in any transient or accident analysis in the FSAR. However, -

they also provide a diverse signal to indicate a possible system break. These instruments are included in Technical Specifications (TS) because of the potential for risk due to possible failure of the instruments preventing HPCI and RCIC initiations. Therefore, they meet criterion 4 of the NRC Policy Statement (Ref. 6).

The HPCI and RCIC Steam Supply Line Pressure - Low signals '

are initiated from transmitters (four_for HPCI and four for RCIC) that are connected to the system steam line. Four channels of both HPCI and RCIC Steam Supply Line Pressure - Low Functions are available and are required to J

(continued)

HATCH UNIT 1 B 3.3-158 REVISION

Recirculation Loops Operating B 3.4.1 O BASES LC0 and APRM Flow Biased Simulated Thermal Power - High setpoint (continued) (LC0 3.3.1.1) must be appliad to allow continued operation consistent with the assumptions of Reference 3. In

~

addition, core flow as a function of core thermal power must be in the " Operation Allowed Region" of Figure 3.4.1-1 to ensure core thermal-hydraulic oscillations do not occur.

APPLICABILITY In MODES 1 and 2, requirements for operation of the Reactor Coolant Recirculation System are necessary since there is considerable energy in the reactor core and the limiting design basis transients and accidents are assumed to occur.

In MODES 3, 4, and 5, the consequences of an accident are reduced and the coastdown characteristics of the recirculation loops are not important.

ACTIONS A.1 and B.1 ~lb A j

r DuS to thermal-hydraulic stability concerns, operation of C the plant with one recirculation loop is controlled by restricting the core flow to 2: 45% f rated core flow when THERMAL POWER is greater than the($_.g.$

0 rod line. This requirement is based on the recommendations contained in GE SIL-380, Revision 1 (Reference 4), which defines the region where the limit cycle oscillations are more likely to occur.

If the core flow as a function of core thermal power is in the " Operation Not Allowed Region" of Figure 3.4.1-1, prompt action should be initiated to restore the flow-power combinati6ii to within the Operation Allowed Region. The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time is based on the low probability of an accident occurring during this time period, on a reasonable time to complete the Required Action, and on frequent core monitoring by operators allowing core oscillations to be quickly detected. An immediate reactor scram is also required with no recirculation pumps in operation, since all forced circulation has been lost and the probability of thermal-hydraulic oscillations is greater.

(continued)

HATCH UNIT 1 B 3.4-4 REVISION A

Reactor Steam Dome Pressure B 3.4.10 B 3.4 CEACTOR COOLANT SYSTEM (RCS)

B 3.4.10 Reactor Steam Dome Pressure BASES BACKGROUND The reactor steam dome pressure is an assumed value in the determination of compliance with reactor pressure vessel overpressure protection criteria and is also an assumed initial condition of design basis accidents and transi,ents, toS7 APPLICABLE The reactor steam done pressure of s 02 psig is an '

SAFETY ANALYSES initial condition of the vessel overpressure protection analysis of Reference 1. This analysis assumes an initial .

maximum reactor steam dome pressure and evaluates the response of the pressure relief systen., primarily the safety / relief valves, during the limiting pressurization transient. The determination of compliance with the overpressure criteria is dependent on the initial reactor steam dome pressure; therefore, the limit on this pressure

_ ensures that the assumptions of the overpressure protection analysis are conserved. Reference 2 also assumes an initial O reactor steam dome pressure for the analysis of design basis accidents and transients used to determine the limits for fuel cladding integrity (see Bases for LC0 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)") and 1% cladding plastic strain i (see Bases for LCO 3.2.1, " AVERAGE PLANAR LINEAR HEAT ;

GENERATION RATE.(APLHGR)").

Reactor steam dome pressure satisfies the requirements of Criterion 2 of the NRC Policy Statement (Ref. 3).

toS1r LCO The ecified reactor steam dome pressure limit of s psig ensures the plant is operated within the assumptions of the overpressure protection analysis.

Operation above the limit may result in a response more severe than analyzed.

APPLICABILITY In MODES 1 and 2, the reactor steam dome pressure is required to be less than or equal to the limit. In these (continued)

HATCH UNIT 1 5 3.4-53 REVISION A

(

Reactor Steam Dome Pressure. I B 3.4.10 j BASES .

APPLICABILITY MODES, the reactor may be generating significant steam and (continued) events which may challenge the overpressure limits are <

possible.

In MODES 3, 4, and 5, the limit is not applicable because the reactor is shut down. In these MODES, the reactor pressure is well below the required limit, and no r anticipated events will challenge the overpressure limits.

i ACTIONS M With the reactor steam dome pressure greater than the limit, prompt action should be taken to reduce pressure to below the limit and return the reactor to operation within the bounds of the analyses. The 15 minute Completion Time is-reasonable considering the importance of maintaining the ;

pressure within limits. This Completion Time also ensures i that the probability of an accident occurring while pressure i is greater than the limit is minimized.

O u ,

If the reactor steam dome pressure cannot be restored to within the limit within the' associated Completion Time, the plant must be brought to a MODE in which the LCO 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 . The allowed Completion i Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is reasonable, based on operating :

experience, to reach MODE 3 from full power conditions in an ;

orderly manner and without challenging plant systems. ,

i SURVEILLANCE SR 3.4.10.1 REQUIREMENTS 1o57 !

Verification that reactor steam dome pressure is s 1020 psig l ensures that the initial conditions of the vessel :

overpressure protection analysis is met. Operating i experience has shown the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency to be sufficient ;

for identifying trends and verifying operation within safety analyses assumptions. .

i O (continued). 1 1

HATCH UNIT 1 B 3.4-54 REVISION A 1

l

_ _ . . _ ~ . . _.

4 ECCS - Ooerating B 3.5.1

/ "1 V BASES BACKGROUND from the suppression pool, to allow testing of the LPCI (continued) pumps without injecting water into the RPV. These test lines also provide suppression pool cooling capability, as described in LCO 3.6.2.3, "RHR Suppression Pool Cooling." r 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 includes loss of offsite power), would not result in the low pressure ECCS subsystems failing to meet their design function. (While an alternate power supply is available.

the low pressure ECCS subsystems may not be capable of j meeting their design function if the alternate power supply is in service.) 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 i line, where the coolant is. distributed within the RPV through the feedwater sparger. Suction piping for the O

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. ,

185 4 The HPCI System is designed to provide core cooling for a wide range of reactor pressures (150 psig to 2 psig).

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.

(continued)

B 3.5-3 REVISION A HATCH UNIT 1

i RCIC System B 3.5.3 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATIM 4

C0OLING (RCIC) SYSTEM B 't . 5. 3 RCIC System j i

BASES 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 requirements ensure that the criteria of Reference 1 are satisfied.

The RCIC System (Ref. 2) consists of a steam driven turbine pump unit, piping, and valves to provide steam to the l turbine, as well as piping and valves to transfer water from O the suction source to the core via the feedwater system line, where the coolant is distributed within the RPV through the fec0 water 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 suppression 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 i turbine is piped from a main steam line upstream of the _;

associated inboard main steam line isolation valve. 1 a a r4 ,

The RCIC System is designed to provide core ooling for a ,

wide range of reactor pressures (150 psig to psig). t 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 l CST to allow testing of the RCIC System during normal operation without injecting water into the RPV. 4 h

O < cont 4nuee>

B 3.5-23 REVISION A )

HATCH UNIT 1 i

Primary Containment B 3.6.1.1 ;

BASES (continued)

APPLICABLE The safr*.y design basis for the primary containment is that !

SAFETY ANALYSES it must withctand the pressures and temperatures of the .

.. limiting DBA without exceeding the design leakage rate.

The DBA that postulates the maximum release of radioactive !

material within primary containment is a LOCA. In the ;

analysis of this accident, it is assumed that primary ,

containment is OPERA 8LE such that release of fission .

products to the environment is controlled by the rate of primary containment leakage. ;

Analytical methods and assumptions involving the primary ,

containment are presented in References I and 2. The safety analyses assume a nonsechanistic fission product release following a DBA, which forms the basis for determination of offsite doses. The fission product release is, in turn, ;

based on an assumed leakage rate from the primary containment. OPERABILITY of the primary containment ensures ;

that the leakage rate assumed in the safety analyses is not ,

exceeded.

~ The maximum allowable leakage rate for the primary 1 containment (L ) is 1.2% by weight of the containment air Q per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months al the maximum peak containment pressure (P,)

of sig (Ref. 1). l l

Primary containment satisfies Criterion 3 of the NRC Policy Statement (Ref. 4). l i

LCO Primary containment OPERABILITY is maintained by limiting leakage to less than L except prior to the first startup :

after performing a req , uired 10 CFR 50, Appendix J, leakage l test. At this time, the combined Type B and C leakage must {

be < 0.6 L , and the overall Type A leakage must be < 0.75 1 L,. Comp 1fancewiththisLCOwillensureaprimary !

containment configuration, including equipment hatches, that ;

is structurally sound and that will limit leakage to those !

leakage rates assumed in the safety analyses.

Individual leakage rates specified for 'the primary containment air lock are addressed in LCO 3.6.1.2.

i 1

(continued)

HATCH UNIT I B 3.6-2 REVISIONfpl3 p

i

Primary Containment Air Lock B 3.6.1.2 r

b] BASES BACKGROUND containment leakage rate to within limits in the event of a (continued) DBA. Not maintaining air lock integrity or leak tightness

_. may result in a leakage rate in excess of that assumed in the unit safety analysis.

APPLICABLE The DBA that postulates the maximum release of radioactive SAFETY ANALYSES material within primary containment is a LOCA. In the analysis of this accident, it is assumed that primary containment is OPERABLE, such that release of fission products to the environment is controlled by the rate of primary containment leakage. The primary containment is designed with a maximum allowable leakage rate (L ) of 1.2%

byweightofthecontainmentairper24hoursatIhe M' calculated maximum peak containment pressure (P ) of

. psig (Ref. 2). This allowable leakage rate forms the l s for the acceptance criteria imposed on the SRs associated with the air lock.

Primary containment air lock OPERABILITY is also required to minimize the amount of fission product gases that may escape primary containment through the air lock and contaminate and pressurize the secondary containment.

The primary containment air lock satisfies Criterion 3 of the NRC Policy Statement (Ref. 4).

LC0 As part of primary containment, the air lock's safety function is related to control of containment leakage rates following a DBA. Thus, the air lock's structural integrity and leak tightness are essential to the successful mitigation of such an event.

The primary containment air lock is required to be OPERABLE.

For the air lock to be considered OPERABLE, the air lock interlock mechanism must be OPERABLE, the air lock must be in compliance with the Type B air lock leakage test, and both air lock doors must be OPERABLE. The interlock allows only one air lock door to be opened at a time. This provision ensures that a gross breach of primary containment does not exist when primary containment is required to be (continued)

HATCH UNIT 1 B 3.6-7 REVISION A

Drywell Pressure B 3.6.1.4

( B 3.6 CONTAINMENT SYSTEMS B 3.6.1.4 Drywell Pressure BASES BACKGROUND The drywell pressure is limited during normal operations to preserve the initial conditions assumed in the accident analysis for a Design Basis Accident (DBA) or loss of coolant accident (LOCA).

APPLICABLE Primary containment performance is evaluated for the entire SAFETY ANALYSES spectrum of break sizes for postulated LOCAs (Ref. 1).

Among the inputs to the DBA is the initial primary containment internal pressure (Ref.1). Analyses assume an initial drywell pressure of 1.75 psig. This limitation l ensures that the safety analysis remains valid by maintaining the expected initial conditions and ensures that the peak LOCA drywell internal pressure does not exceed the maximum allowable of 62 psig.

The maximum calculated drywell pressure occurs during the reactor blowdown phase of the DBA, which assumes an instantaneous recirculation line break. The calculated peak drywell pressure for this limiting event is O.6 psig l (Ref. 1). L. g Drywell pressure satisfies Criterion 2 of the NRC Policy Statement (Ref. 2).

LC0 In the event of a DBA, with an initial drywell pressure s 1.75 psig, the resultant peak drywell accident pressure l will be maintained below the drywell design pressure.

APPLICABILITY In MODES 1, 2, and 3, a DBA could cause a release of radioactive material to primary containment.

In MODES 4 O (continued)

HATCH UNIT 1 B 3.6-28 REVISION /

Main Condenser Offgas B 3.7.6 O BASES LCO withthisrequirement(2436MWtx100ki/MWt-second-(continued) 240 mci /second). The 240 m Ci/ second limit is cence+vt

. tor a ra4ca core therm.) puu o f 2 sTt Mu4 APPLICABILITY The LCO is applicable when steam is being exhausted to the main condenser and the resulting noncondensibles are being processed via the Main Condenser Offgas System. This occurs during MODE 1, and during MODES 2 and 3 with any main steam line not isolated and the SJAE in operation. In MODES 4 and 5, steam is not being exhausted to the main condenser and the requirements are not applicable.

ACTIONS a.d If the offgas radioactivity rate limit is exceeded, 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is allowed to restore the gross gamma activity rate to within the limit. The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time is reasonable, based on engineering judgment, the time required

. to complete the Required Action, the large margins associated with permissible dose and exposure limits, and O the low probability of a Main Condenser Offgas System rupture.

B.I. B.2. B.3.1. and B.3.2 If the gross gamma activity rate is not restored to within the limits in the associated Completion Time, all main steam lines or the SJAE must be isolated. This isolates the Main Condenser Offgas System from the source of the radioactive steam. The main steam lines are considered isolated if at i least one main steam isolation valve in each main steam line is closed, and at least one main steam line drain valve in the drain line is closed. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Completion Time is reasonable, based on operating experience, to perform the actions from full power conditions in an orderly manner and without challenging unit systems.

An alternative to Required Actions B.1 and B.2 is to place the unit in a MODE in which the LC0 does not apply. To achieve this status, the unit must be placed in at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in MODE 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The (continued)

HATCH UNIT 1 B 3.7-33 REVISION A

t Inservice Leak and Hydrostatic Testing Operation B 3.10.1 h- B 3,10 SPECIAL OPERATIONS B 3.10.1 Inservice Leak and Hydrostatic Testing Operation j l

BASES I BACKGROUND- The purpose of this Special Operations LCO is to allow l certain reactor coolant pressure tests to be performed in l MODE 4 when the metallurgical characteristics of the reactor >

pressure vessel (RPV) require the pressure testing at !

temperatures > 212*F (normally corresponding to MODE 3). l I

System hydrostatic testing and system leakage (same as i inservice leakage tests) pressure tests required by Section XI of the American Society of Mechanical Engineers i (ASME) Boiler and Pressure Vessel Code (Ref.1) are performed prior to the reactor going critical after a refueling outage. Inservice system leakage tests are :

performed at the end'of each refueling outage with the ,

system set for normal power operation. Some parts of the Class 1 boundary are not pressurized during these system !

tests. System hydrostatic tests are required once per l interval and include all the Class 1 boundary unless the i O test is broken into smaller portions. Recirculation pump operation and a water solid RPV (except for an air bubble for pressure control) are used to achieve the necessary ;

temperatures and pressures required for these tests. The '

minimum temperatures (at the required pressures) allowed for ,

these tests are determined from the RPV pressure and temperature (P/T) limits required by LCO 3.4.9, " Reactor Coolant System (RCS) Pressure and Temperature (P/T) Limits."

These limits are conservatively based on the fracture i toughness of the reactor vessel, taking into account anticipated vessel neutron fluence. The hydos c test requires increasing pressure to approximately 0 pstg.-II33 The system leak test requires increasing pressure to i approximately sig. i lo 35-With increased reactor vessel fluence over time, the minimum '

allowable vessel temperature increases at a given pressure. !

Periodic updates to the RCS P/T limit curves are performed ;

as necessary, based upon the results of analyses of -

irradiated surveillance specimens removed from the vessel.

i O (coatiaued)

HATCH UNIT 1 B 3.10-1 REVISION D

\

. 4 .

(1) r,astne Fower Level Georgia Fower Company is author 12eo to operate the f acility at steady state reactor core power levels not in excess of f436,regewatts thermal in accordance with the conditions e specified herein and in Attachment 2 to this license.

2M Attachment 2 is an integral part of this license.

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 4(a, are hereby incer-

~

porated in the license. The licenste shall operate the I facility in accordance with the Tecnnical Specifications.

\

(3) Additional Conditions The matters specified in the following conditions shall be completed to the satisfaction of the Corraission within the stated tine periods following the issuance of the license or within the operational restrictions indicated. The !

removal of these conditions shall be made by an amenoment to the license supported by a f avorable evaluation by the Comission.

( .; r _ :' " ' : - - -- - )

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Definitions 1.1 1.1 Definitions PHYSICS TESTS b. Authorized under the provisions of (continued) 10 CFR 50.59; or

c. Otherwise approved by the Nuclear Regulatory Commission.

RATED THERMAL POWER RTP shall be a total reactor core heat transfer MWt.

(RTP) rate to the reactor coolant of Q4325'ST REACTOR PROTECTION The RPS RESPONSE TIME shall be that time interval SYSTEM (RFS) RESPONSE from when the monitored parameter exceeds its RPS O

d TIME trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

SHUTDOWN MARGIN (SDM)

SDM shall be the amount of reactivity by which the reactor is subcritical or would be subcritical assurr.ing that:

a. The reactor is xenon free;

b. The moderator temperature is 68*F; and

c. All control rods are fully inserted except for

& the single control rod of highest reactivity

- worth, which is assumed to be fully withdrawn. ,

With control rods not capable of being fully inserted, the reactivity worth of these '

control rods must be accounted for in the determination of SDM.

(continued) 1.1-6 REVISION HATCH UNIT 2 i

SLC System 3.1.7 O

'd SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.7.5 (continued) Once within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after solution temperature is restored within the Region A 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.

O SR 3.1.7.7 Verify each pump develops a flow rate 2 41.2 gpm at a discharge pressure a psig.

In accordance with the Inservice Testing Program SR 3.1.7.8 Verify flow through one SLC subsystem from 18 months on'a pump into reactor pre:,sure vessel. STAGGERED TEST BASIS (continued)

O !

1 HATCH UNIT 2 3.1-22 REVISION A

1 i

RPS Instrumentation 3.3.1.1

,r %

d* Table 3.3.1.1 1 (page 2 of 3)

Reactor Protection system Instrunentation f F

APPLICABLE CONDi?!ONs MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM ,

SPECIFIED PER TRIP REQUIRED SURVE!LLANCE ALLOWASLE FUNCTION CONDITIONS sisTEM ACTION D.1 REQUIREMEN1s VALUE

2. Average Power Reige ,

Monitors (continued)

c. Flaed Neutron 1 2 F SR 3.3.1.1.1 sW 9 Flua . Nish sa 3.3.1.1.2 sa 3.3.1.1.8 sa 3.3.1.1.9 sa 3.3.1.1.10 l sa 3.3.1.1.15 sa 3.3.1.1.16

d. Dowwcat e 1 2 7 st 3.3.1.1.5 t 4.2% RTP SR 3.3.1.1.8 st 3.3.1.1.15

c. Inoo 1,2 i G st 3.3.1.1.8 NA

$R 3.3.1.1.9 st 3.3.1.1.15

3. Reactor vesset steen 1,2 2 G st 3.3.1.1.1 s 05 pais Dome Pressure - utyn SR 3.3.1.1.9 SR 3.3.1.1.13 SR 3.3.1.1.15 fobs sa 3.3.1.1.16

4. Reactor vessel Water 1,2 2 G st 3.3.1.1.1 t 0 inches Levet - Low, Level 3 st 3.3.1.1.9 v st 3.3.1.1.13 SR 3.3.1.1.15 i SR 3.3.1.1.16 ,

5. Main steam isolation 1 8 F SR 3.3.1.1.9 s 10% closed volve - Closure SR 3.3.1.1.13 st 3.3.1.1.15 st 3.3.1.1.16

6. Drywl! Pressu.'s - Nish 1,2 2 G SR 3.3.1.1.1 s 1.92 pois SR 3.3.1.1.9 ;

sa 3.3.1.1.13 ;

SR 3.3.1.1.15 (continued)

HATCH UNIT 2 3.3-8 REVISION A

ATWS-RPT Instrumentation 3.3.4.2 rn

() SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.4.2.2 Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.4.2.3 Perform CHANNEL CALIBRATION. The 18 months Allowable Values shall be:

a. Reactor Vessel Water Level - Low Low, level 2: 2 -47 inches; and

b. Re tor Steam Dome Pressure - High:

s psig.

Il76 SR 3.3.4.2.4 Perform LOGIC SYSTEM FUNCTIONAL TEST 18 months including breaker actuation. l O l O

HATCH UNIT 2 3.3-33 REVISION A l

l

LLS Instrumentation 3.3.6.3 Table 3.3.6.3-1 (page 1 of 1)

Low Low set Instr e entation REQUIRED CHAhhELS PER SURVE!LLI.NCE ALLOWASLE FUNCT!0m FUNCTION REQUIREMENTS VALUE

1. Reactor Steen Dome Pressure Migh 1 per LLS valve SR 3.3.6.3.1 s 1054 psig SR 3.3.6.3.4 sa 3.3.6.3.5 1R 3.3.6.3.6 1095

2. Low Low Set etessure Setpoints 2 per LLS valve SR 3.3.6.3.1 Low:

54 3.3.6.3.4 Open s 1010 psis st 3.3.6.3.5 Close s 860 psis st 3.3.6.3.6 Nedlue Low Open s 1025 psis close s 875 pois Mediun Migh:

Open 5 1040 psig Close s 890 pste High:

Open s 1050 peig ,

close s 900 pais

3. Taltpipe Pressure switch 2 per $/tv sa 3.3.6.3.2 2 80 psig and SR 3.3.6.3.3 s 100 pais sa 3.3.6.3.5 SR 3.3.6.3.6 t

e HATCH UNIT 2 3.3-66 REVISION A l

Recirculation Loops Operating 3.4.1 o i .

t i

\

30 et SPSAAfl0Bs 8007 aumee usan M=

a

/

i- #

I -

/T l o w e r- rk ;,s I n e_

'O I u b1 VI

T aumen g ,

28eaces i M

u i

m a m .

samt nar a nAram l

l Figure 3.4.1-1 (Page 1 of 1)

Power-Flow Operating Map with One Reactor '

Coolant System Recirculation loop in Operation 4

l 0 1 1

3.4-4 REVISION A HATCH UNIT 2 i

_--- _ _ _ _ __ _ _ _ ,~ _. - - . ---

S/RVs 3.4.3 -

[D V SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY i

SR 3.4.3.1 Verify the safety function lift setpoints In accordance of the S/RVs are as follows: with the Inservice Number of Setpoint Testing Program S/RVs (osia) .

4 7090t32.7J l120 i 33,(,

4 1100 t 33.0 ;- g i3o i 33,q 3 <ll10_t33.j3 y9g i

Following testing, lift settings shall be within i 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 after reactor steam pressure and flow are adequate to perform the test.

Verify each S/RV opens when manually 18 months I

actuated.

i i

l O

HATCH UNIT 2 3.4-8 REVISION A

n . .- . .. .- - .

I

.I Reactor Steam Dome Pressure 3.4.10

~

3 4' REACTOR COOLANT SYSTEM-(RCS) j 3.4.10 Reactor Steam Dome Pressure l

I ON j

__ LCO 3.4.10 The re' actor steam dome pressure shall be s psig. l t

APPLICABILITY:- MODES I and 2. i ACTIONS !

CONDITION REQUIRED ACTION COMPLETION TIME A. Reactor steam dome A.1 Restore reactor steam 15 minutes !

pressure not within dome pressure to !

limit. within limit. 4 B. Required Action and B.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 1 associated Completion Time not met.-

i SURVEILLANCE REQUIREMENTS i e

SURVEILLANCE FREQUENCY l SR 3.4.10.1 Veri reactor steam dome pressure is. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months !

s psig. !

los P !

O ,

HATCH UNIT 2 3.4-2S REVISION A

ECCS -- Operating I 3.5.1

(). SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY

. . . . SR 3.5.1.6 -------------------NOTE--------------------

Only required to be performed prior to entering MODE 2 from MODE 3 or 4, when in MODE 4 > 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .

Verify each recirculation pump discharge 31 days valve cycles through one complete cycle of full travel or is de-onergized in the closed position.

SR 3.5.1.7 Verify the following ECCS pumps develop the In accordance specified flow rate against a system head with the corresponding to.the specified reactor Inservice pressure. Testing Program SYSTEM HEAD N0. CORRESPONDING

~

0F TO A REACTOR SYSTEM FLOW RATE PUMPS PRESSURE OF 0 CS 2: 4250 gpm 1 2: 113 psig LPCI 2: 17,000 gpm 2 2: 20 psig SR 3.5.1.8 -------------------NOTE--------------------

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.

to$17 Verify, with reactor pressure s 020 psig 92 days and 2: 920 psig, the HPCI pump can evelop a flow rate 2: 4250 gpm against a system head corresponding to reactor pressure.

(continued)

O HATCH UNIT 2 3.5-5 REVISION'A !

RCIC System 3.5.3

(]) SURVEILLANCE REQUIREMENTS i SURVEILLANCE FREQUENCY ,

SR 3.5.3.1 Verify the RCIC System piping is filled 31 days with water from the pump discharge valve to the injection valve.

SR 3.5.3.2 Verify each RCIC System manual, power 31 days operated, and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position.

SR 3.5.3.3 -------------------NOTE--------------------

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.

tofP

,/~'1 Verify, with reactor pressure s 02 psig 92 days v and 2: 920 psig, the RCIC pump can evelop a flow rate a 400 gpm against a system head corresponding to reactor pressure. '

SR 3.5.3.4 -------------------NOTE--------------------

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.

Verify, with reactor pressure s 165 psig, 18 months the RCIC pump can develop a flow rate 2: 400 gpm against a system head .

corresponding to reactor pressure.

(continued) !

O .

HATCH UNIT 2 3.5-12 REVISION A

l SLC Systea B 3.1.7 h BASES SURVEILLANCE SR 3.1.7.7 pg REQUIREMENTS (continued) Demonstrating that each SLC System pum develops a flow rate 2: 41.2 gpm at a discharge pressure 190 psig ensures that 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 l requirements, the rate of negative reactivity insertion from the SLC System will adequately compensate for the positive reactivity effects encountered during power reduction, cooldown of the moderator, and xenon decay. This test confirms one point on the pump design curve and is indicative of overall performance. Such inservice inspections confirm component OPERABILITY, trend performance, and detect incipient failures by indicating '

abnormal performance. The Frequency of this Surveillance is in accordance with the Inservice Testing Program. ,

SR 3.1.7.8 and SR 3.1.7.9

~

These Surveillances ensure that there is a functioning flow path from the sodium pentaborate solution storage tank to

(] the RPV, including the firing of an explosive valve. The v replacement charge for the explosive valve shall be from the same manufactured batch as the one fired or from another batch that has been certified by having one of that batch successfully fired. The pump and explosive valve tested should be alternated such that both complete flow paths are tested every 36 months at alternating 18 month intervals.

The Surveillance may be performed in separate steps to prevent injecting boron into the RPV. An acceptable method for verifying flow from the pump to the RPV is to pump demineralized water from a test tank through one SLC subsystem and into the RPV. 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 Frequency was concluded to be acceptable from a reliability standpoint.

(continued)

HATCH UNIT 2 B 3.1-45 REVISION A

Primary Containment Isolation Instrumentation B 3.3.6.1 O BASES APPLICABLE . 1.c. Main Steam Line Flow-Hich. (continued) l SAFETY ANALYSES, LCO, and per trip system) are available and are required to be APPLICABILITY. OPERABLE so that no single instrument failure will preclude )

detecting a break in any individual MSL.

The Allowable Value is chosen to ensure that offsite dose !

limits are not exceeded due to the break. The Allowable l psid, which is the parameter.

' Value corresponds monitored to s @ instruments.

on control room .

l+5 This function isolates the Group 1 valves.

1.d. Condenser Vacuum-low The Condenser Vacuum-Low Function is provided to prevent l overpressurization of the main condenser in the event of'a :

loss of the main condenser vacuum. Since the integrity of .

the condenser is an assumption in offsite dose calculations, !

the Condenser Vacuum-Low Function is assumed to be OPERABLE !

and capable of initiating closure of the MSIVs.

O The closure of the MSIVs is initiated to prevent the addition of steam that would lead to additional condenser pressurization and possible rupture of the diaphragm installed to protect the turbine exhaust hood,- thereby !

preventing a potential radiation leakage path following an accident. .t Condenser vacuum pressure signals are derived from four pressure transmitters that sense the pressure in the condenser. Four channels of Condenser Vacuum-Low Function are available and are required to be OPERABLE to ensure that l no single instrument failure can preclude the isolatiou function.

The Allowable Value is chosen to prevent damage to the

condenser due to pressurization, thereby ensuring its integrity for offsite dose analysis. As noted (footnote (a) i to Table 3.3.6.1-1), the channels are not required to be l OPERABLE in MODES 2 and 3 when all turbine stop valves (TSVs) are closed, since the potential for condenser f overpressurization is minimized. Switches are provided to .

manually bypass the channels when all TSVs are closed. l

'This function isolates the Group 1 valves.

l (continued) l HATCH UNIT 2 B 3.3-153 REVISION A l l

. t

Primary Containment Isolation Instrumentation B 3.3.6.1 Q BASES APPLICABLE 3.e. 4.a. HPCI and RCIC Steam Line Flow - Hiah SAFETY ANALYSES, (continued)

. LCO, and APPLICABILITY recirculation and MSL breaks. However, these instruments prevent the RCIC or HPCI steam line breaks from becoming bounding.

The HPCI and RCIC Steam Line Flow- High signals are initiated from transmitters (two for HPCI ::ad two for RCIC) that are connected to the system steam lines. Two channels of both HPCI and RCIC Steam Line Flow - High Functions are available and are required to be OPERABLE to ensure that no single instrument failure can preclude the isolation function.

The Allowable Values are chosen to be low enough to ensure that the trip occurs to prevent fuel damage and maintains gg5 2 p. the MSLB evcnt as the bounding event. The Allowable Values i

correspond to s@for RCIC, which are the parametersinches inches water column monitored on control room instruments. *

~

These Functions isolate the Group 3 and 4 valves, as O PProPriate-3.b. 4.b. HPCI and RCIC Steam Supply Line Pressure - Low Low MSL pressure indicates that the pressure of the steam in the HPCI or RCIC turbine may be too low to continue operation of the associated system's turbine. These isolations are for equipment protection and are not assumed in any transient or accident analysis in the FSAR. However, they also provide a diverse signal to indicate a possible ;

system break. These instruments are included in Technical l Specifications (TS) because of the potential for risk due to i possible failure of the instruments preventing HPCI and RCIC :

initiations. Therefore, they meet Criterion 4 of the NRC j Policy Statement (Ref. 7). l l

The HPCI and RCIC Steam Supply Line Pressure - Low signals j are initiated from transmitters (four for HPCI and four for i RCIC) that are connected to the system steam line. Four channels of both HPCI and RCIC Steam Supply Line Pressure - Low Functions are available and are required to (continuad)

HATCH UNIT 2 B 3.3-158 REVISION

Recirculation Loops Operating 8 3.4.1

,y

() BASES LCO and APRM Flow Biased Simulated Thermal Power - High setpoint (continued) (LC0 3.3.1.1) must be applied to allow continued operation consistent with the assumptions of Reference 3. In addition, core flow as a function of core thermal power must be in the " Operation Allowed Region" of Figure 3.4.1-1 to ensure core thermal-hydraulic oscillations do not occur.

APPLICABILITY In MODES 1 and 2, requirements for operation of the Reactor Coolant Recirculation System are necessary since there is considerable energy in the reactor core and the limiting design basis transients and accidents are assumed to occur.

In MODES 3, 4, and 5, the consequences of an accident are reduced and the coastdown characteristics of the recirculation loops are not important.

ACTIONS A.1 and B.1 g e/,

~

n Due to thermal-hydraulic stability con erns, operation of

(~) the plant with one recirculation loop 's controlled by restricting the core flow to 2: 45% of, rated core flow when THERMAL POWER is greater than the@ rod line. This requirement is based on the recommendations contained in GE SIL-380, Revision 1 (Reference 4), which defines the region where the lihlit Cycle oscillations are more likely to occur.

If the core flow as a function of core thermal power is in the " Operation Not Allowed Region" of Figure 3.4.1-1, prompt action should be initiated to restore the flow-power combination to within the Operation Allowed Region. The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time is based on the low probability of an accident occurring during this time period, on a reasonable time to complete the Required Action, and on frequent core monitoring by operators allowirig core oscillations to be quickly detected. An immediate resetor scram is also required with no recirculation pumps in operation, since all forced circulation has been lost and the probability of thermal-hydraulic oscillations is greater.

l (continued)

HATCH UNIT 2 B 3.4-4 REVISION A

1 Reactor Steam Dome Pressure B 3.4.10 i

O B 3.4 REACTOR C00 TANT SYSTEM (RCS) l B 3.4.10 Reactor Steam Dome Pressure ,

BASES BACKGROUND The reactor steam dome pressure is an assumed value in the determination of compliance with reactor pressure vessel overpressure protection criteria and is also an assumed initial condition of design basis accidents and transients.

APPLICABLE The reactor steam dome pressure of s h g is an SAFETY ANALYSES initial condition of the vessel overpressure protection analysis of Reference 1. This analysis assumes an initial maximum reactor steam dome pressure and evaluates the response of the pressure relief system, primarily the safety / relief valves, during the limiting pressurization '

transient. The determination of compliance with the overpressure criteria is dependent on the initial reactor steam dome pressure; therefore, the limit on this pressure ensures that the assumptions of the overpressure protection analysis are conserved. Reference 2 also assumes an initial O reactor steam dome pressure for the analysis of design basis accidents and transients used to determine the limits for fuel cladding integrity (see Bases for LC0 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)") and 1% cladding plastic strain (see Bases for LC0 3.2.1, " AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)").

Reactor steam dome pressure satisfies the requirements of Criterion 2 of the NRC Policy Statement (Ref. 3).

LC0 Th ied reactor steam dome pressure limit of s sig ensures the plant is operated within the assumptions of the overpressure protection analysis.

Operation above the limit may result in a response more severe than analyzed.

APPLICABILITY In MODES 1 and 2, the reactor steam dome pressure is required to be less than or equal to the limit. In these (continued)

HATCH UNIT 2 B 3.4-53 REVISION A

. - - . . . .- . _ - - . ~~

Reactor Steam Dome Pressure B 3.4.10 ,

OL BASES.

APPLICABILITY MODES, the reactor may be generating significant steam and (continued) events which may' challenge the overpressure limits are i possible.

In MODES 3, 4, and 5, the limit is not applicable because the reactor is shut down. In these MODES, the reactor pressure is well below the required limit, and no anticipated events will challenge the overpressure limits.

ACTIONS M With the reactor steam dome pressure greater than the limit, prompt action should be taken to reduce pressure to below the limit and return the reactor to operation within the ,

bounds of the analyses. The 15 minute Completion Time is -

reasonable considering the importance of maintaining the .

pressure within limits. This Completion Time also ensures !

that the probability of an accident occurring while pressure .

is greater than the limit is minimized.

~

O u :

If the reactor steam dome pressure cannot be restored to within the limit within the associated Completion Time, the plant must be brought to a MODE in which the LCO 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 . The allowed Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is reasonable, based on operating I experience, to reach MODE 3 from full power conditions in an i orderly manner and without challenging plant systems. )

i SURVEILLANCE SR 3.4.10.1 REQUIREMENTS IO N ;

Verification that reactor steam dome pressure is :s 02 psig ensures that the initial conditions of the vessel overpressure protection analysis is met. Operating experience has shown the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency to be sufficient for identifying trends and verifying operation within safety analyses assumptions.

l O (coatinuee) I l

. HATCH UNIT 2 B 3.4-54 REVISION A

- . _ 1

ECCS - Operating B 3.5.1 BASES BACKGROUND from the suppression pool, to allow testing of the LPCI (continued) pumps without injecting water into the RPV. These test lines also provide suppression pool cooling capability, as described in LCO 3.6.2.3, "RHR Suppression Pool CJoling."

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 includes loss of offsite power), would not result in the low pressure ECCS subsystems failing to meet their design function. (While an alternate power supply is available, the low pressure ECCS subsystems may not be capable of meeting their design function if the alternate power supply is in service.)

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 A system is provided from the CST and the suppression pool.

V 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. d The HPCI System is designed to provide core for cooling a [Il wide range of reactor pressures (162 psid to Q3ypsid, 3

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

the suppression pool. A full flow test line is provided to i route water from and to the CST to allow testing of the HPCI ;

System during normal operation without injecting water into

~

the RPV. ,

(continued)

HATCH UNIT 2 B 3.5-3 REVISION A l

l

RCIC System B 3.5.3 B 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS) AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEli 8 3.5.3 RCIC System BASES BACXGROUND 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 requirements ensure that the criteria of Reference 1 are satisfied.

The kCIC System (Ref. 2) consists of a steam driven turbine p)

(~

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 is provided from the condensate storage tank (CST) and the suppression pool. Pump suction is normally aligned to the CST to minimize injection of suppression 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.

16Y The RCIC System is designed to provide core cooling for a wide range of reactor pressures (150 psig to sig). i Upon receipt of an initiation signal, the RCIC turbine accelerate; 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 frcm and to the CST to allow testing of the RCIC System during normal operation without injecting water into the RPV.

f

( (continued)

HATCH UNIT 2 B 3.5-23 REVISION A l

1

Primary Containment B 3.6.1.1

(] BASES (continued)

APPLICABLE The safety design basis for the primary containment is that SAFETY ANALYSES it must withstand the pressures and temperatures of the limiting DBA without exceeding the design leakage rate.

The DBA that postulates the maximum release of radioactive material within primary containment is a LOCA. In the analysis of this accident, it is assumed that primary containment is OPERABLE such that release of fission preducts to the environment is controlled by the rate of primary containment leakage.

Analytical methous and assumptions involving the primary containment are presented in References 1 and 2. The safety analyses assume a nonmechanistic fission product release following a DBA, which forms the basis for determination of offsite doses. The fission product release is, in turn, based on an assumed leakage rate from the primary containment. OPERABILITY of the primary containment ensures that the leakage rate assumed in the safety analyses is not exceeded.

. The maximum allowable leakage rate for the primary containment (L ) is 1.2% by weight of the containment air O d per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months al. the maximum peak containment pressure (P*)

psig (Ref. 1). l Primary containment satisfies Criterion 3 of the NRC Policy Statement (Ref. 4).

LCO Primary containment OPERABILITY is maintained by limiting except prior to the first startup leakage to less a after performing than requL , ired 10 CFR 50, Appendix J, leakage test. At this time, the combined Type B and C leakage must be < 0.6 L , and the overall Type A leakage must be < 0.75 L,. ComplfancewiththisLCOwillensureaprimary containment configuration, including equipment hatches, that is structurally sound and that will limit leakage to those leakage rates assumed in the safety analyses.

Individual leakage rates specified for the primary containment air lock are addressed in LC0 3.6.1.2.

I i

s) (continued)

HATCH UNIT 2 B 3.6-2 REVIS'ONf(

Primary Containment Air Lock (

8 3.6.1.2 l

~

BASES BACKGROUND containment leakage rate to within limits in the event of a (continued) DBA. Not maintaining air lock integrity or leak tightness l

. may result in a leakage rate in excess of that assumed in i the unit safety analysis. l l

APPLICABLE The DBA that postulates the maximum release of radioactive SAFETY ANALYSES material within primary containment is a LOCA. In the >

analysis of this accident, it is assumed that primary -

containment is OPERABLE, such that release of fission products to the environment is controlled by the rate of primary containment leakage. The primary containment is ,

designed with a maximum allowable leakage rate (L ) of 1.2% i byweightofthecontainmentairper24hoursatIhe i calculated maximum peak containment pressure (P,) of 95:5' . psig (Ref. 2). This allowable leakage rate forms the l i s for the acceptance criteria imposed on the SRs associated with the air lock.

Primary containment air lock OPERABILITY is also required to

~

minimize the amount of fission product gases that may escape ;

primary containment through the air lock and contaminate and ' '

Q pressurize the secondary containment.

The primary containment air lock satisfies Criterion 3 of the NRC Policy Statement (Ref. 4). ,

LC0 As part of primary containment, the air lock's safety '

function is related to control of containment leakage rates following a DBA. Thus, the air lock's structural integrity and leak tightness are essential to the successful mitigation of such an event.

The primary containment air lock is required to be OPERABLE.

For the air lock to be considered OPERABLE, the air lock interlock mechanism must be OPERABLE, the air lock must be in compliance with the Type B air lock leakage test, and -

both air lock doors must be OPERABLE. The interlock allows -

only one air lock door to be opened at a time. This provision ensures that a gross breach of primary containment does not exist when primary containment is required to be (continued)

HATCH UNIT 2 8 3.6-7 REVISION g

Drywell Pressure B 3.6.1.4 8 3.6 CONTAINMENT SYSTEMS B 3.6.1.4 Drywell Pressure BASES BACKGROUND The drywell pressure is limited during normal operations to preserve the initial conditions assumed in the accident analysis for a Design Basis Accident-(DBA) or loss of I coolant accident (LOCA). ;

APPLICABLE Primary containment performance is evaluated for the entire SAFETY ANALYSES spectrum of break sizes for postulated LOCAs (Ref.1).

Among the inputs to the DBA is the initial primary containment internal pressure (Ref.1). Analyses assume an initial drywell pressure of 1.75 psig. This limitation l ensures that the safety analysis remains valid by maintaining the expected initial conditions and ensures that the peak LOCA drywell internal pressure does not exceed the maximum allowable of 62 psig.

pI' The maximum calculated drywell pressure occurs during the reactor blowdown phase of the DBA, which assumes an 5d instantaneous recirculation line break. The alculated peak drywell pressure for this limiting event is 48.7 psig l (Ref. 1). yg g-Drywell pressure satisfies Criterion 2 of the NRC Policy Statement (Ref. 2).

LCO In the event of a DBA, with an initial drywell pressure s 1.75 psig, the resultant peak drywell accident pressure l will be maintained below the drywell design pressure.

APPLICABILITY In MODES 1, 2, and 3, a DBA could cause a release of radioactive material to primary containment.

In MODES 4 (continued)

HATCH UNIT 2 8 3.6-28 REVISION [A

l Main Condenser Offgas l B 3.7.6 ;

/7 BASES V

LC0 wit 5thisrequirement(2436MWtx100kCi/MWt-second-(continued) 240 mCsecond). The 240 met /sema lWit is cwervM b a Mr A con + hem.t p a ar . S 1ssP mu f ,

APPLICABILITY The LC0 is applicable when steam is being exhausted to the main c.ondenser and the resulting noncondensibles are being processed via the Main Condenser Offgas System. This occurs during MODE 1, and during MODES 2 and 3 with any main steam line not isolated and the SJAE in operation. In MODES 4 and 5, steam is not being exhausted to the main condenser and the requirements are not applicable.

ACTIONS Ad If the offgas radioactivity rate limit is exceeded, 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is allowed to restore the gross' gamma activity rate to within the limit. The 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time is reasonable, based on engineering judgment, the time required to complete the Required Action, the large margins associtted with permissible dose and exposure limits, and the low probability of a Main Condenser Offgas System O rupture.

B.l. B.2. B.3.1. and B.3.2

~

If the gross gamma activity rate is not restored to within the limits in the associated Completion Time, all main steam lines or the SJAE must be isolated. This isolates the Main Condenser Offgas System from the source of the radioactive steam. The main steam lines are considereo isolated if at least one main steam isolation valve in each main steam line is closed, and at least one main steam line drain valve in the drain line is closed. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Completion Time is reasonable, based on operating experience, to perform the actions from full power conditions in an orderly manner and without challenging unit systems.

An alternative to Required Actions B.1 and B.2 is to place the unit in a MODE in which the LC0 does not apply. To !'

achieve this status, the unit must be placed in at least MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months and in MODE 4 wit.iin 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The (continued)

HATCH UNIT 2 B 3.7-33 REVISION A l

L. - -_ . _ _ . _.

, . .. . -. - . ~ .-. . - - - -- . . . . .

1 Inservice Leak and Hydrostatic Testing Operation B 3.10.1 )

. B 3.10 SPECIAL OPERATIONS

-B 3.10.1 Inservice Leak and Hydrostatic Testing Operation BASES BACKGROUND The purpose of this Special Operations LCO is to allow '

certain reactor coolant pressure tests to be performed in MODE 4 when the metallurgical characteristics of the reactor J pressure vessel (RPV) require the pressure testing at temperatures > 212'F (normally corresponding to MODE 3). ,

System hydrostatic testing and system leakage (same as inservice leakage tests) pressure tests required by -

Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Ref.1) are performed prior to the reactor going critical after a i refueling outage. Inservice system leakage tests are '

performed at the end of each refueling outage with the system set for normal power operation. Some aarts of the Class I boundary are not pressurized during t1ese system tests. System hydrostatic tests are required once per '

interval and include all the Class 1-boundary unless the -

test is broken into smaller portions. Recirculation pump '

(N operation and a water solid RPV (except for an air bubble :

for pressure control) are used to achieve the necessary temperatures.and pressures required for these tests. The :

minimum temperatures (at the required pressures) allowed for t these tests are determined from the RPV pressure and temperature (P/T) limits required by LCO 3.4.9, " Reactor iA :

Coolant System (RCS) Pressure and Temperature l These limits are conservatively based on the fr(P/T) actureLimits." ,

toughness of the reactor vessel,'taking into account >

anticipated vessel neutron fluence. The hydros ic test psig. 'llPl requires increasing pressure to approximately The system le approximately test requires increasing pressure to psig.

d;

[

lo3r !

With increased reactor vessel fluence over time, the minimum-allowable vessel temperature increases at a given pressure.

Periodic updates to the RCS P/T limit curves are performed !

as necessary, based upon the results of analyses of ;

irradiated surveillance specimens removed from the vessel.

3 O

APPLICABLE Allowing the reactor to be considered in MODE 4 during ;

SAFETY ANALYSES hydrostatic or leak-testing, when the reactor coolant i temperature is > 212*F, effectively provides an exception to t MODE 3 requirements, including OPERABILITY of primary

O nontinued)

HATCH UNIT 2 B 3.10-1 REVISION D

.9 Enclosure 4 Edwin I. Hatch Plant Units 1 and 2 GE Report NEDC-32405P Power Uprate Safety Analysis Report O

O l

t I

ML20077S033

Text

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