Text

Technical Specification Pages for Amendment Nos. 238 and 180
	ML032691352
Person / Time
Site:	Hatch
Issue date:	09/23/2003
From:	; NRC/NRR/DLPM
To:	;
References
	TAC MB7026, TAC MB7027
	Download: ML032691352 (78)
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for sample analysis or instrument calibration, or associated with radioactive apparatus or components; (6) Southern Nuclear, pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. This renewed license shall be deemed to contain and is subject to the conditions specified in the following Commission regulations in 10 CFR Chapter : Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Section 50.54 of Part 50, and Section 70.32 of Part 70; all applicable provisions of the Act and the rules, regulations, and orders of the Commission now or hereafter in effect; and the additional conditions specified or incorporated below:

(1) Maximum Power Level Southern Nuclear is authorized to operate the facility at steady state reactor core power levels not in excess of 2804 megawatts thermal.

(2) Technical Specifications The Technical Specifications (Appendix A) and the Environmental Protection Plan (Appendix B), as revised through Amendment No.

are hereby incorporated in the renewed license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

The Surveillance Requirement (SR) contained in the Technical Specifications and listed below, is not required to be performed immediately upon implementation of Amendment No. 195. The SR listed below shall be successfully demonstrated prior to the time and condition specified:

SR 3.8.1.18 shall be successfully demonstrated at its next regularly scheduled performance (3) Fire Protection Southern Nuclear shall implement and maintain in effect all provisions of the fire protection program, which is referenced in the Updated Final Safety Analysis Report for the facility, as contained in the updated Fire Hazards Analysis and Fire Protection Program for Edwin I. Hatch Nuclear Plant Units 1 and 2, which was originally submitted by letter dated July 22, 1986. Southern Nuclear may make changes to the fire protection program without prior Commission approval only if the changes Renewed License No. DPR-57 Amendment No. 238

Definitions 1.1 1;1 Definitions (continued)

MINIMUM The MCPR shall be the smallest critical power ratio (CPR) that CRITICAL POWER exists Inthe core for each class of fuel. The CPR Isthat power RATIO (MCPR) In the assembly that Is calculated by application of the appropriate correlation(s) to cause some point Inthe assembly to experience boiling transition, divided by the actual assembly operating power.

MODE A MODE shall correspond to any one Inclusive combination of mode switch position, average reactor coolant temperature, and reactor vessel head closure bolt tensioning specified In Table 1.1-1 with fuel in the reactor vessel.

OPERABLE - A system, subsystem, divisIon, component, or device shall be OPERABILITY OPERABLE or have OPERABILITY when It Is capable of performing Its specified safety function(s) and when all necessary attendant Instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for thd 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 measure the fundamental nuclear characteristics of the reactor 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.

RATED THERMAL RTP shall be a total reactor core heat transfer rate to the reactor POWER (RTP) coolant of 2804 MWt I REACTOR The RPS RESPONSE TIME shall be that time Interval from when the PROTECTION monitored parameter exceeds Its RPS trip setpoint at the channel SYSTEM (RPS) sensor until de-energization of the scram pilot valve solenoids. The RESPONSE TIME response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time Is measured.

(continued)

HATCH UNIT 1 1.1-4 Amendmt No. 238

SLs

2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 785 psig or core flow

< 10% rated core flow THERMAL POWER shall be s 24% .RTP. I 2.1.1.2 With the reactor steam dome pressure a 785 psig and core flow 2 10% rated core flow.

MCPR shall be 2 1.07 for two recirculation loop operation or 2. 1.09 for single recirculation loop operation.

2.1.1.3 Reactor vessel water level shall be greater than the top of active Irradiated fuel.

2.1.2 Reactor Coolant System (RCS) Pressure SL Reactor steam dome pressure shall be s 1325 psig.

2.2 SL Violations With any SL violation, the following actions shall be completed:

2.2.1 Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , notify the NRC Operations Center, Inaccordance with 10 CFR 50.72.

2.2.2 Within 2 hours0.0833 days 0.0119 weeks 0.00274 months :

2.2.2.1 Restore compliance with all SLs; and 2.2.2.2 Insert all nsertable control rods.

2.2.3 Within 24 hours1 days 0.143 weeks 0.0329 months , notify the plant manager, the corporate executive responsible for overall plant nuclear safety, and the offsite review committee.

(continued)

HATCH UNIT 1 2.0-1 it No. 238

APLHGR 3.2.1 3.2 POWER DISTRIBUTION LIMITS 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)

LCO 3.2.1 All APLH4GRs shall be less than or equal to the limits specified n the COLR.

APPLICABILITY: THERMAL POWER 2 24% RTP. I ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Any APLHGR not within A.1 Restore APLHGR(s) to 2 hours0.0833 days 0.0119 weeks 0.00274 months limits. within limits.

B. Required Action and B.1 Reduce THERMAL 4 hours0.167 days 0.0238 weeks 0.00548 months associated Completion POWER to < 24% RTP. I Time not met SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.1.1 Verify all APLHGRs are less than or equal to the Once withIn limits specified Inthe COLR. 12 hours0.5 days 0.0714 weeks 0.0164 months after 2 24% RTP I AND 24 hours1 days 0.143 weeks 0.0329 months thereafter HATCH UNIT 1 3.2-1 Amendmet No. 238

MCPR 3.2.2 32 POWER DISTRIBUTION LIMITS 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR)

LCO 3.2.2 All MCPRs shall be greater than or equal to the MCPR operating limits specified In the COLR.

APPLICABILITY: THERMAL POWER 2 24% RTP.

ACTIONS CONDITION -REQUIRED ACTION COMPLETION TIME A. Any MCPR not within limits. A1 Restore MCPR(s) to 2 hours0.0833 days 0.0119 weeks 0.00274 months within limits.

B. Required Action and B.1 Reduce THERMAL 4 hours0.167 days 0.0238 weeks 0.00548 months associated Completion POWER to < 24% RTP.

Time not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.2.1 Verify all MCPRs are greater than or equal to the Once within limits specified In the COLR. 12 hours0.5 days 0.0714 weeks 0.0164 months after 2 24% RTP AND 24 hours1 days 0.143 weeks 0.0329 months thereafter (continued)

HATCH UNIT 13 3.2-2 ,Amomt No. 23 8

RPS Instrumentation 3.3.1.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME C. One or more Functions with C.1 Restore RPS trip 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months RPS trip capability not capability.

maintained.

D. Required Action and D.1 Enter the Condition Immediately associated Completion referenced In Time of Condition A, B, Table 3.3.1.1-1 for the or C not met. channel.

E. As required by Required E.1 Reduce THERMAL 4 hours0.167 days 0.0238 weeks 0.00548 months Action D.1 and referenced POWER to < 27.6%

in Table 3.3.1.1-1. RTP.

F. As required by Required F.1 Be In MODE 2. 6 hours0.25 days 0.0357 weeks 0.00822 months Action D.1 and referenced InTable 3.3.1.1-1.

G. As required by Required G.1 Be In MODE 3. 12 hours0.5 days 0.0714 weeks 0.0164 months Action D.1 and referenced InTable 3.3.1.1-1.

H. As required by Required H.1 Initiate action to fully Immediately Action D.1 and referenced Insert all nsertable InTable 3.3.1.1-1. control rods Incore cells containing one or more fuel assemblies.

(continued)

II HATCH UNIT 1 3.3-2 Amendment No. 238

RPS Instrumentation 3.3.1.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME

1. As required by Required 1.1 Initiate alternate method 12 hours0.5 days 0.0714 weeks 0.0164 months Action D.1 and referenced to detect and suppress in Table 3.3.1.1 -1. thermal-hydraulic instability oscillations.

AND 1.2 Restore required 120 days channels to OPERABLE.

J. Required Action and J.1 Be InMODE 2. 4 hours0.167 days 0.0238 weeks 0.00548 months associated Completion Time of Condition I not met.

SURVEILLANCE REQUIREMENTS

- -- -~NOTES

1. Refer to Table 3.3.1.1-1 to determine which SRs apply for each RPS Function.

2. When a channel is placed Inan inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours0.25 days 0.0357 weeks 0.00822 months provided the associated Function maintains RPS trip capability.

SURVEILLANCE FREQUENCY SR 3.3.1.1.1 Perform CHANNEL CHECK. 12 hours0.5 days 0.0714 weeks 0.0164 months SR 3.3.1.1.2 -- NOTE-Not required to be performed until 12 hours0.5 days 0.0714 weeks 0.0164 months after THERMAL POWER 2 24% RTP. I Verify the absolute difference between the 7 days average power range monitor (APRM) channels and the calculated power is s 2% RTP while operating at 2 24% RTP. I (continued)

HATCH UNIT 1 3.3-3 Amedment No. 238

RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.1.11 Verify Turbine Stop Valve - Closure and 24 months Turbine Control Valve Fast Closure, Trip Oil Pressure - Low Functions are not bypassed when THERMAL POWER Is z 27.6% RTP. I SR 3.3.1.1.12 Perform CHANNEL FUNCTIONAL TEST. 24 months SR 3.3.1.1.13 NOTES

1. Neutron detectors are excluded.

2. For Function 1, not required to be performed when entering MODE 2 from MODE 1 until 12 hours0.5 days 0.0714 weeks 0.0164 months after entering MODE 2.

Perform CHANNEL CALIBRATION. 24 months SR 3.3.1.1.14 (Not used.)

SR 3.3.1.1.16 Perform LOGIC SYSTEM FUNCTIONAL TEST. 24 months SR 3.3.1.1.16 -NOTE Neutron detectors are excluded.

Verify the RPS RESPONSE TIME is within limits. 24 months on a STAGGERED TEST BASIS (continued)

HATCH UNIT 1 3.3-5 hamert No. 23 8

RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 1 of 3)

Reactor Protection System Instrumentaion APPUCABLE CONDmONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE

1. Intermediate Range Monlor

a. Neutron Flu -High 2 2(d) G SR 33.1.1.1. s 12U/125 SR .3.1.1.4 dvisions of fuji SR 3.3.1.1.6 scale SR 3.3.1.7 SR 3.3.1.1.13 SR 3.3.1.1.15 Isa) 2(d) H SR 3.3.1.1.1 S 12M125 SR 3.3.1.1.5 divislons of fI SR .3.1.1.13 scale SR 3.1.1.15

b. Inop 2 G SR .3.1.14 NA SR 3.31.1.15 5(a) 2(d) H SR 3..1.5 NA SR 3.3.1.1.15

2. Average Power Range Monitor

a. Neutron Fux - High 2 G SR 3.1.1.1 s 20%/ RTP (Setdown) SR 3.3.1.1.7 SR 3.3.1.1.8 SR 3.3.1.1.10 SR 3.3.1.1.13

b. SImlated Therral 1 3(c) F SR 3.3.1.1.1 0.57W +

Power - High SR 3.3.1.12 66.8% RTP SR 3..1.1.5 and SR 3.1.1.10 s115.5%

SR 3.1.1.13 RTP(b)

c. Neutron Flux - High 3(c) F SR 31.1.1 s 120 RP SR 3.3.1.12 SR 3.3.1.1.

SR 3.3.1.1.10 SR .3.1.1.13

d. Inop 1.2 3(c) G SR 3.1.1.10 NA (continued)

(a) With any control rod withdrawn from a core cel containing one or more fuel assemblies.

(b) 0.57W . 56.8% - 0.57 AW RTP when reset for single loop operation per LCO 3A.1, Redrculaffon Loops Operating.

(c) Each APRM channel provides Inputs to both tip systems.

(d) One channel hI each quadrant of the core must be OPERABLE whenever the IRMs are required to be OPERABLE. Both the RWM and a second iensed operator must verify compliance with the wilthdrawal sequence when less than three channels In any tip system are OPERABLE.

HATCH UNIT 3.3-7 Amendment No. 238

RPS Instrumentaton 3.3.1.1 Table 3.3.1.1-1 (page S of 3)

Reactor Protection System Instwnerteaon APPUCABLE CONDmONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDTONS SYSTEM ACTION D.1 REQUIREMENTS VALUE B. Turbine Stop Valve-Closure 227. °6%RTP 4 E SR 33.1.1.9 S10%dosed I SR 3..1.11 SR 3.3.1.1.13 SR 3.3.1.1.15 S. Turblne Control Valve Fast Closure, Trip OH Pressure -

2 27.6% RTP 2 E SR SR 3.3.1.1.9 3.1.1.1 Z 600 pslg I Low SR 3+/-1.1.13 SR 1.1.15 SR 3.3.1.1.16

10. Reactor Mode Swltch- 1.2 1 0 SR 3+/-1.1.12 NA Shutdown P8sl1on SR 3.3.1.1.15 5(a) 1 H SR 1.1.12 NA SR 3.3.1.1.15

11. ManulScram 1.2 1 G SR 3.3.1.1.5 NA SR 3.3.1.1.15 Eta) 1 H SR 3.1.1.1 NA SR 3.3.1.1.15 (a) With any control rod wtdrawn from a core cell contaling one or more fuel assembles.

HATCH UNIT 1 3.3-0 Ament No. 238

Feedwater and Main Turbine Trip High Water Level Instrumentation 3.3.2.2 3.3 INSTRUMENTATION 3.3.2.2 Feedwater and Main Turbine Trip High Water Level Instrumentation LCO 3.3.2.2 Three channels of feedwater and main turbine trip nstrumentation shall be OPERABLE.

APPLICABILITY: THERMAL POWER 2 24% RTP. I ACTIONS NRl t -

Separate Condition entry Is allowed for each channel.

CONDITION REQUIRED AC1iON COMPLETION liME A. One feedwater and main A.1 Place channel Intrip. 7 days turbine high water level trip channel Inoperable.

B. Two or more feedwater and B.1 Restore feedwater and 2 hours0.0833 days 0.0119 weeks 0.00274 months main turbine high water main turbine high water level trip channels level trip capability.

Inoperable.

C. Required Action and C.1 Reduce THERMAL 4 hours0.167 days 0.0238 weeks 0.00548 months associated Completion POWER to < 24% RTP.

Time not met I HATCH UNIT I 3.3-20 Amendmet No. 238

EOC-RPT Instrumentation 3.3.4.1 3.3 INSTRUMENTATION 3.3.4.1 End of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation LCO 3.3.4.1 a. Two channels per trip system for each EOC-RPT Instrumentation Function listed below shall be OPERABLE:

1. Turbine Stop Valve (TSV) - Closure; and

2. Turbine Control Valve (TCV) Fast Closure, Trip Oil Pressure - Low.

OR

b. LCO 3.2.2, MINIMUM CRITICAL POWER RATIO (MCPR), limits for inoperable EOC-RPT as specified Inthe COLR are made applicable.

APPLICABILITY: THERMAL POWER 2 27.6% RTP. I ACTIONS l.^YP I

------ -- 1 [Mt Separate Condition entry Is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more channels A.1 Restore channel to 72 hours3 days 0.429 weeks 0.0986 months Inoperable. OPERABLE status.

OR A.2 -- NOTE--

Not applicable If inoperable channel Is the result of an Inoperable breaker.

Place channel In trip. 72 hours3 days 0.429 weeks 0.0986 months (continued)

HATCH UNIT 1 3.3-27 Amndmnt No. 238

EOC-RPT Instrumentation 3.3.4.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. One or more Functions with B.1 Restore EOC-RPT trip 2 hours0.0833 days 0.0119 weeks 0.00274 months EOC-RPT trip capability not capability.

maintained.

OR B.2 Apply the MCPR limit 2 hours0.0833 days 0.0119 weeks 0.00274 months MCPR limit for Inoperable for Inoperable EOC-RPT not made EOC-RPT as specified applicable. Inthe COLR.

C. Required Action and CA Remove the associated 4 hours0.167 days 0.0238 weeks 0.00548 months associated Completion recirculation pump from Time not met. service.

OR C.2 Reduce THERMAL 4 hours0.167 days 0.0238 weeks 0.00548 months POWER to c 27.6% I RTP.

SURVEILLANCE REQUIREMENTS When a channel is placed In an Inoperable status solely for performance of required Surveillances, entry Into associated Conditions and Required Actions may be delayed for up to 6 hours0.25 days 0.0357 weeks 0.00822 months provided the associated Function maintains EOC-RPT trip capability.

SURVEILLANCE FREQUENCY SR 3.3.4.1.1 Perform CHANNEL FUNCTIONAL TEST. 92 days on an ALTERNATE TEST BASIS SR 3.3.4.1.2 Verify TSV - Closure and TCV Fast Closure, Trip 24 months Oil Pressure - Low Functions are not bypassed when THERMAL POWER is ? 27.6% RTP. I (continued)

HATCH UNIT 1 3.3-28 Amendment No. 238

Main Turbine Bypass System 3.7.7 3.7 PLANT SYSTEMS 3.7.7 Main Turbine Bypass System LCO 3.7.7 The Main Turbine Bypass System shall be OPERABLE.

OR LCO 3.2.2, MINIMUM CRITICAL POWER RATIO (MCPR)," limits for an Inoperable Main Turbine Bypass System, as specified in the COLR, are made applicable.

APPLICABILITY: THERMAL POWER 2 24% RTP. I ACTIONS CONDITION . REQUIRED ACTION COMPLETION TIME A. Requirements of the LCO A.1 Satisfy the requirements 2 hours0.0833 days 0.0119 weeks 0.00274 months not met. of the LCO.

B.. Required Action and B.1 Reduce THERMAL 4 hours0.167 days 0.0238 weeks 0.00548 months associated Completion POWER to < 24% RTP. I Time not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.7.1 Verify one complete cycle of each main turbine 31 days bypass valve.

SR 3.7.7.2 Perform a system functional test. 24 months SR 3.7.7.3 Verify the TURBINE-BYPASS SYSTEM 24 months RESPONSE TIME Iswithin limits.

HATCH UNIT 1 3.7-18 A UAmer t No. 238

Reactor Core SLs B 2.1.1 BASES BACKGROUND to a structurally weaker form. This weaker form may lose Its Integrity, (continued) resulting In an uncontrolled release of activity to the reactor coolant.

The reactor vessel water level SL ensures that adequate core cooling capability is maintained during all MODES of reactor operation.

Establishment of Emergency Core Cooling System nitiation setpoints higher than this safety limit provides margin such that the safety limit will not be reached or exceeded.

APPLICABLE The fuel cladding must not sustain damage as a result of normal SAFETY ANALYSES operation and AOOs. The reactor core SLs are established to preclude violation of the fuel design criterion that a MCPR limit Is to be established, such that at least 99.9% of the fuel rods Inthe core would not be expected to experience the onset of transition boiling.

The Reactor Protection System setpoints [LCO 3.3.1.1, Reactor Protection System (RPS) Instrumentation, In combination with the other LCOs, are designed to prevent any anticipated combination of transient conditions for Reactor Coolant System water level, pressure, and THERMAL POWER level that would result Inreaching the MCPR SL.

2.1.1.1 Fuel Cladding Inteariv GE critical power correlations are applicable for all critical power calculations at pressures - 785 psig and core flows 2 100h of rated flow. For operation at low pressures or low flows, another basis Is used, as follows:

Since the pressure drop Inthe bypass region Is essentially all elevation head, the core pressure drop at low power and flows will always be > 4.5 psi. Analyses (Ref. 2) show that with a bundle flow of 28 x 108 b1/hr, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow with a 4.5 psi driving head will be > 28 x 10 lb/hr. Full scale ATLAS test data taken at pressures fron 14.7 psia to 800 psla Indicate that the fuel assembly critical power at this flow Is approximately 3.35 MWt With the design peaking factors, this corresponds to a THERMAL POWER

> 50% RTP. Thus, a THERMAL POWER limit of 24% RTP for reactor pressure < 785 psig Is conservative. I (continued)

HATCH UNIT I B 2.0-2 Amendmet No. 238

APLHGR B 3.2.1 B 3.2 POWER DISTRIBUTION LIMITS B 3.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)

BASES BACKGROUND The APLHGR is a measure of the average LHGR of all the fuel rods in a fuel assembly at any axial location. Limits on the APLHGR are specified to ensure that certain fuel design limits Identified in Reference I are not exceeded during anticipated operational occurrences (AOOs) and that the peak cladding temperature (PCT) during the postulated design basis loss of coolant accident (LOCA) does not exceed the limits specified In 10 CFR 50.46.

APPLICABLE The analytical methods and assumptions used In evaluating the fuel SAFETY ANALYSES design limits are presented In References 1 and 2. The analytical methods and assumptions used In evaluating Design Basis Accidents (DBAs), anticipated operational transients, and normal operation that determine the APLHGR limits are presented in References 1, 2,3, 4, 5, 6, and 7.

Fuel design evaluations are performed to demonstrate that the 1% limit on the fuel cladding plastic strain and certain other fuel design limits described In Reference 1 are not exceeded during AOOs for operation with LHGRs up to the operating mit LHGR. APLH1GR limits are equivalent to the LHGR limit for each fuel rod divided by the local peaking factor of the fuel assembly. APLiHIGR limits are developed as a function of exposure and the various operating core flow and power states to ensure adherence to fuel design limits during the limiting AOOs (Refs. 5, 6, and 7). Flow dependent APLHGR limits are determined (Ref. 7) using the three dimensional BWR simulator code (Ref. 8) to analyze slow flow runout transients. The flow dependent multiplier, MAPFACf, is dependent on the maximum core flow runout capability. The maximum runout flow Is dependent on the existing setting of the core flow limiter In the Recirculation Flow Control System.

Based on analyses of limiting plant transients (other than core flow increases) over a range of power and flow conditions, power dependent multipliers, MAPFACp, are also generated. Due to the sensitivity of the transient response to Initial core flow levels at power levels below those at which turbine stop valve closure and turbine control valve fast closure scram trips are bypassed, both high and low core flow MAPFACp limits are provided for operation at power levels between 24% RTP and the previously mentioned bypass power level.

(continued)

HATCH UNIT B 3.2-1 Amendment No. 238

APLHGR B 3.2.1 BASES (continued)

APPLICABILITY The APLHGR limits are primarily derived from fuel design evaluations and LOCA and transient analyses that are assumed to occur at high power levels. Design calculations (Ref. 7) and operating experience have shown that as power Is reduced, the margin to the required APLHGR limits Increases. This trend continues down to the power range of 5% to 15% RTP when entry Into MODE 2 occurs. When In MODE 2, the Intermediate range monitor scram function provides prompt scram initiation during any significant transient, thereby effectively removing any APLHGR limit compliance concern In MODE 2. Therefore, at THERMAL POWER levels ?4% 2 RTP, the I reactor is operating with substantial margin to the APLHGR limits; thus, this LCO Isnot required.

ACTIONS AM If any APLHGR exceeds the required limits, an assumption regarding an nitial condition of the DBA and transient analyses may not be met.

Therefore, prompt action should be taken to restore the APLHGR(s) to within the required limits such that the plant operates within analyzed conditions and within design limits of the fuel rods. The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time Is sufficient to restore the APLHGR(s) to within its limits and Is acceptable based on the low probability of a transient or DBA occurring simultaneously with the APLHGR out of specification.

If the APLHGR cannot be restored to within its required limits within the associated Completion Time, the plant must be brought to In a MODE or other specified condition Inwhich the LCO does not apply.

To achieve this status, THERMAL POWER must be reduced to c 24% RTP within 4 hours0.167 days 0.0238 weeks 0.00548 months . The allowed Completion Time Is I reasonable, based on operating experience, to reduce THERMAL POWER to < 24% RTP Inan orderly manner and without challenging I plant systems.

SURVEILLANCE SR 3.2.1.1 REQUIREMENTS APLHGRs are required to be Initially calculated within 12 hours0.5 days 0.0714 weeks 0.0164 months after THERMAL POWER Is2 24% RTP and then every 24 hours1 days 0.143 weeks 0.0329 months I thereafter. They are compared to the specified limits in the COLR to ensure that the reactor Is operating within the assumptions of the (continued)

HATCH UNIT 1 B 3.2-3 Amendment No. 238

APLHGR B 3.2.1 BASES SURVEILLANCE SR 3.2.1.1 (continued)

REQUIREMENTS safety analysis. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency Is based on both engineering judgment and recognition of the slowness of changes In power distribution during normal operation. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowance after THERMAL POWER 2 24% RTP is achieved is acceptable given the I large Inherent margin to operating limits at low power levels.

REFERENCES 1. NEDE-2401 1-P-A General Electric Standard Application for Reactor Fuel,' (revision specified Inthe COLR).

2. FSAR, Chapter 3.

3. FSAR, Chapter 6.

4. FSAR, Chapter 14.

5. NEDO-24205, ED.. Hatch Nuclear Plant Units 1 and 2 Single-Loop Operation, August 1989.

6. NEDO-24395, Load Line Limit Analysis,' October 1980.

7. NEDC-30474-P Average Power Range Monitor, Rod Block Monitor and Technical Specification Improvements (ARTS)

Program for E.l. Hatch Nuclear Plant, Units 1 and 2,'

December 1983.

8. NEDO-30130-A, 'Steady State Nuclear Methods," May 1985.

9. NEDO-24154, 'Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors,' October 1978.

10. NEDO-31376, E.I. Hatch Nuclear Plant SAFERIGESTAR-LOCA Analysis,' December 1986.

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

12. NEDC-33085-P, "SafetyAnalyis Report for Edwin I. Hatch Units 1 and 2 Thermal Power Optimization,' November 2002. I B 3.2-4 Amerxtnt No. 238 HATCH UNIT 1

MCPR B 3.2.2 BASES APPLICABLE benchmarked using the three dimensional BWR simulator code SAFETY ANALYSES (Ref. 9) to analyze slow flow runout transients. The operating limit Is (continued) dependent on the maximum core flow limiter setting In the Recirculation Flow Control System.

Power dependent MCPR limits (MCPRp) are determined mainly by the one dimensional transient code (Ref. 10). Due to the sensitivity of the transient response to nitial core flow levels at power levels below those at which the turbine stop valve closure and turbine control valve fast closure scrams are bypassed, high and low flow MCPRp operating limits are provided for operating between 24% RTP I and the previously mentioned bypass power level.

The MCPR satisfies Criterion 2 of the NRC Policy Statement (Ref. 11).

LCO The MCPR operating limits specified Inthe COLR are the result of the Design Basis Accident (DBA) and transient analysis. The operating limit MCPR Is determined by the larger of the MCPR and MCPR%

limits.

APPLICABILITY The MCPR operating limits are primarily derived from transient analyses that are assumed to occur at high power levels. Below 24% RTP, the reactor Is operating at a minimum recirculation pump speed and the moderator void ratio Is small. Surveillance of thermal limits below 24% RTP Is unnecessary due to the large Inherent margin that ensures that the MCPR SL Is not exceeded even if a I limiting transient occurs. Statistical analyses indicate that the nominal value of the initial MCPR expected at 24% RTP is > 3.5. Studies of the variation of limiting transient behavior have been performed over the range of power and flow conditions. These studies encompass the range of key actual plant parameter values Important to typically limiting transients. The results of these studies demonstrate that a margin Is expected between performance and the MCPR requirements, and that margins increase as power is reduced to 24% RTP. This trend is expected to continue to the 5% to 15% power I range when entry Into MODE 2 occurs. When In MODE 2, the Intermediate range monitor provides rapid scram Initiation for any significant power increase transient, which effectively eliminates any MCPR compliance concern. Therefore, at THERMAL POWER levels

< 24% RTP, the reactor Is operating with substantial margin to the I MCPR limits and this LCO Is not required.

(continued)

HATCH UNIT B 3.2-6 Amendment No. 238

MCPR B 3.2.2 BASES (continued)

ACTIONS A3 If any MCPR is outside the required limits, an assumption regarding an initial condition of the design basis transient analyses may not be met. Therefore, prompt action should be taken to restore the MCPR(s) to within the required limits such that the plant remains operating within analyzed conditions. The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time is normally sufficient to restore the MCPR(s) to within its limits and Is acceptable based on the low probability of a transient or DBA occurring simultaneously with the MCPR out of specification.

If the MCPR cannot be restored to within its required limits within the associated Completion Time, the plant must be brought to a MODE or other specified cordition n which the LCO does not apply. To achieve this status, THERMAL POWER must be reduced to c 24% RTP within 4 hours0.167 days 0.0238 weeks 0.00548 months . The allowed Completion Time s reasonable, based on operating experience, to reduce THERMAL POWER to < 24% RTP In an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.2.2.1 REQUIREMENTS The MCPR is required to be initially calculated within 12 hours0.5 days 0.0714 weeks 0.0164 months after THERMAL POWER is a 24% RTP and then every 24 hours1 days 0.143 weeks 0.0329 months thereafter. It is compared to the specified limits Inthe COLR to

-ensure that the reactor Isoperating within the assumptions of the safety analysis. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency Is based on both engineering judgment and recognition of the slowness of changes In power distribution during normal operation. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowance after THERMAL POWER k 24% RTP Is achieved Is acceptable given the large Inherent margin to operating limits at low power levels.

SR .2.2.2 Because the transient analysis takes credit for conservatism in the scram speed performance, it must be demonstrated that the specific scram speed distribution is consistent with that used in the transient analysis. SR 3.2.2.2 determines the value of a,which Is a measure of the actual scram speed distribution compared with the assumed distribution. The MCPR operating limit Is then determined based on an interpolation between the applicable limits for Option A (scram (continued)

HATCH UNIT B 32-7 Amendmt No. 238

RPS Instrumentation B 3.3.1.1 BASES APPLICABLE the specified Allowable Value, where appropriate. The setpolnt Is SAFETY ANALYSES calibrated consistent with applicable setpoint methodology LCO, and assumptions (nominal trip setpoint). Each channel must also respond APPLICABILITY within Its assumed response time, where appropriate.

(continued)

Allowable Values are specified for each RPS Function specified in the Table. Nominal trip setpoints are specified in the setpolnt calculations. The nominal setpoints are selected to ensure that the actual setpoints do not exceed the Allowable Value between successive CHANNEL CALIBRATIONS. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within Its Allowable Value, is acceptable. A channel Is Inoperable If ts actual trip setpoint is not within Its required Allowable Value.

Trip setpoints are those predetermined values of output at which an action should take place. The setpolnts are compared to the actual process parameter (e.g., reactor vessel water level), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit) changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis. The Allowable Values are derived from the analytic limits, corrected for calibration, process, and some of the Instrument errors.

The trip setpoints are then determined accounting for the remaining instrument errors (e.g., drift). The trip setpoints derived In this manner provide adequate protection because Instrumentation uncertainties, process effects, calibration tolerances, Instrument drift, and severe environmental effects (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for.

The OPERABILITY of scram pilot valves and associated solenoids, backup scram valves, and SDV valves, described Inthe Background section, are not addressed by this LCO.

The ndividual Functions are required to be OPERABLE In the MODES or other specified conditions specified In the Table, which may require an RPS trip to mitigate the consequences of a design basis accident or transient. To ensure a reliable scram function, a combination of Functions are required Ineach MODE to provide primary and diverse Initiation signals. The only MODES specified In Table 3.3.1.1-1 are MODES 1 (which encompasses 2 27.6% RTP) and 2, and MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies. No RPS Function Is required in MODES 3 and 4 since all control rods are fully Inserted and the Reactor Mode Switch Shutdown Position control rod withdrawal block (continued)

HATCH UNIT I B .3-3 Amendment No. 238

RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 2.a. Average Power Range Monitor Neutron Flux - High Setdown)

SAFETY ANALYSES (continued)

LCO, and APPLICABILITY abnormal operating transients In this power range. For most operation at low power levels, the Average Power Range Monitor Neutron Flux - High (Setdown) Function will provide a secondary scram to the ntermediate Range Monitor Neutron Flux - High Function because of the relative setpoints. With the IRMs at Range 9 or 10, it Is possible that the Average Power Range Monitor Neutron Flux - High (Setdown) Function will provide the primary trip signal for a corewide increase In power.

No specific safety analyses take direct credit for the Average Power Range Monitor Neutron Flux - High (Setdown) Function. However, this Function indirectly ensures that before the reactor mode switch Is placed in the run position, reactor power does not exceed 24%.RTP I (SL 2.1.1.1) when operating at low reactor pressure and low core flow.

Therefore, It indirectly prevents fuel damage during significant reactivity Increases with THERMAL POWER < 24% RTP. I The Allowable Value Is based on preventing significant Increases In power when THERMAL POWER Is < 24% RTP. I The Average Power Range Monitor Neutron Flux - High (Setdown)

Function must be OPERABLE during MODE 2 when control rods may be withdrawn since the potential for criticality exists.

In MODE 1, the Average Power Range Monitor Neutron Flux - High Function provides protection against reactivity transients and the RWM and rod block monitor protect against control rod withdrawal error events.

2.b. Average Power Range Monitor Simulated Thermal Power - High The Average Power Range Monitor Simulated Thermal Power - High Function monitors neutron flux to approximate the THERMAL POWER being transferred to the reactor coolant. The APRM neutron flux Is electronically filtered with a time constant representative of the fuel heat transfer dynamics to generate a signal proportional to the THERMAL POWER in the reactor. The trip level Is varied as a function of recirculation drive flow (i.e., at lower core flows, the setpoint Is reduced proportional to the reduction Inpower experienced as core flow Is reduced with a fixed control rod pattern) but Is clamped at an upper limit that Is always lower than the Average Power Range Monitor Neutron Flux - High Function Allowable Value.

(continued)

HATCH UNIT I B 3.3-7 Amendmnt No. 238

RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 8. Turbine Stop Valve - Closure (continued)

SAFETY ANALYSES, LCO, and reactor scram reduces the amount of energy required to be absorbed APPLICABILITY and, along with the actions of the End of Cycle Recirculation Pump Trip (EOC-RPT) System, ensures that the MCPR SL Is not exceeded.

Turbine Stop Valve - Closure signals are Initiated from position switches located on each of the four TSVs. Two independent position switches are associated with each stop valve. One of the two switches provides Input to RPS trip system A; the other, to RPS trip system B. Thus, each RPS trip system receives an Input from four Turbine Stop Valve - Closure channels, each consisting of one position switch. The logic for the Turbine Stop Valve - Closure Function is such that three or more TSVs must be dosed to produce a scram. In addition, certain combinations of two valves closed will result In a half-scram. This Function must be enabled at THERMAL POWER a 27.6% RTP. This Is normally accomplished automatically by pressure switches sensing turbine first stage pressure; therefore, opening of the turbine bypass valves may affect this Function.

The Turbine Stop Valve - Closure Allowable Value is selected to be high enough to detect Imminent TSV closure, thereby reducing the severity of the subsequent pressure transient.

Eight channels of Turbine Stop Valve - Closure Function, with four channels In each trip system, are required to be OPERABLE to ensure that no single Instrument failure will preclude a scram from this Function If the TSVs should dose. This Function is required, consistent with analysis assumptions, whenever THERMAL POWER is 2:27.6% RTP. This Function is not required when THERMAL POWER is < 27.6% RTP since the Reactor Vessel Steam Dome Pressure - High and the Average Power Range Monitor Neutron Flux - High Functions are adequate to maintain the necessary safety margins.

9. Turbine Control Valve Fast Closure, Trio Oil Pressure - Low Fast closure of the TCVs results In the loss of a heat sink that produces reactor pressure, neutron flux, and heat flux transients that must be limited. Therefore, a reactor scram Is initiated on TCV fast closure Inanticipation of the transients that would result from the closure of these valves. The Turbine Control Valve Fast Closure, Trip Oil Pressure - Low Function is the primary scram signal for the generator load rejection event analyzed In Reference 2. For this event, the reactor scram reduces the amount of energy required to be (continued)

HATCH UNIT 1 B 3.3-16 Amndmnt No. 238

RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 9. Turbine Control Valve Fast Closure. TriD Oil Pressure - Low SAFETY ANALYSES, (continued)

LCO, and APPLICABILITY absorbed and, along with the actions of the EOC-RPT System, ensures that the MCPR SL Is not exceeded.

Turbine Control Valve Fast Closure, Trip Oil Pressure - Low signals are initiated by the electrohydraulic control (EHC) fluid pressure at each control valve. One pressure switch is associated with each control valve, and the signal from each switch Is assigned to a separate RPS logic channel. This Function must be enabled at THERMAL POWER 2 27.6% RTP. This Is normally accomplished automatically by pressure switches sensing turbine first stage pressure; therefore, opening of the turbine bypass valves may affect this Function.

The Turbine Control Valve Fast Closure, Trip Oil Pressure - Low Allowable Value Is selected high enough to detect Imminent TCV fast closure.

Four channels of Turbine Control Valve Fast Closure,,Trip ON Pressure - Low Function with two channels In each trip system arranged In a one-out-of-two logic are required to be OPERABLE to ensure that no single instrument failure will preclude a scram from this Function on a valid signal. This Function Is required, consistent with the analysis assumptions, whenever THERMAL POWER Is 2 27.6% RTP. This Function is not required when THERMAL POWER Is < 27.6% RTP, since the Reactor Vessel Steam Dome Pressure - High and the Average Power Range Monitor Neutron Flux - High Functions are adequate to maintain the necessary safety margins.

10. Reactor Mode Switch - Shutdown Position The Reactor Mode Switch - Shutdown Position Function provides signals, via the manual scram logic channels, directly to the scram pilot solenoid power circuits. These manual scram logic channels are redundant to the automatic protective nstrumentation channels and provide manual reactor trip capability. This Function was not specifically credited in the accident analysis, but It Is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.

The reactor mode switch is a single switch with two channels, each of which provides input into one of the RPS manual scram logic channels.

(continued)

HATCH UNIT I B 3.3-17 AAmenmnt No. 238

RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1..1 (continued)

REUIREMENTS between Instrument channels could be an ndication of excessive instrument drift Inone of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, It is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.

Agreement criteria are determined by the plant staff based on a combination of the channel Instrument uncertainties, Including indication and readability. Ifa channel s outside the criteria, it may be an Indication that the instrument has drifted outside Its limit.

The Frequency Is based upon operating experience that demonstrates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO.

SR 3.3.1.1.2 To ensure that the APRMs are accurately Indicating the true core average power, the APRMs are calibrated to the reactor power calculated from a heat balance. The Frequency of once per 7 days Is based on minor changes InLPRM senstivity, which could affect the APRM reading between performances of SR 3.3.1.1.8.

A restriction to satisfying this SR when < 24% RTP Is provided that requires the SR to be met only at . 24% RTP because It is dicult to accurately maintain APRM indication of core THERMAL POWER consistent with a heat balance when < 24% RTP. At low power levels, a high degree of accuracy Is unnecessary because of the large, inherent margin to thermal limits (MCPR and APLHGR). At 2 24% RTP, the Surveillance Is required to have been satisfactorily performed within the last 7 days, In accordance with SR 3.0.2. A Note is provided which allows an Increase InTHERMAL POWER above 24% If the 7 day Frequency Is not met per SR 3.0.2. In this event, the SR must be performed within 12 hours0.5 days 0.0714 weeks 0.0164 months after reaching or exceeding 24% RTP. Twelve hours Is based on operating experience and In consideration of providing a reasonable time In which to complete the SR.

(continued)

HATCH UNIT B 3.3-24 Amendment No. 238

RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.11 (continued)

REQUIREMENTS POWER is 2 27.6% RTP. This involves calibration of the bypass channels. Adequate margins for the Instrument setpoint methodologies are Incorporated Into the actual setpoint. Because main turbine bypass flow can affect this setpolnt nonconservatively (THERMAL POWER Is derived from turbine first stage pressure), the main turbine bypass valves must remain closed during the calibration at THERMAL POWER 2 27.6% RTP to ensure that the calibration Is valid.

Ifany bypass channel's setpoint is nonconservative (i.e., the Functions are bypassed at a 27,6% RTP, either due to open main turbine bypass valve(s) or other reasons), then the affected Turbine Stop Valve - Closure and Turbine Control Valve Fast Closure, Trip Oil Pressure - Low Functions are considered Inoperable. Alternatively, the bypass channel can be placed Inthe conservative condition (nonbypass). Ifplaced in the nonbypass condition (Turbine Stop Valve - Closure and Turbine Control Valve Fast Closure, Trip Oil Pressure - Low Functions are enabled), this SR Is met and the channel Is considered OPERABLE.

The 24 month Frequency Is based on a review of the surveillance test history, drift of the associated Instrumentation, and Reference 18.

SR 3.3.1.1.13 A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies that the channel responds to the measured parameter within the necessary range and accuracy.

CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations, consistent with the plant specific setpolnt methodology. For MSIV - Closure, SDV Water Level - High (Float Switch), and TSV - Closure Functions, this SR also includes a physical Inspection and actuation of the switches. For the APRM Sinulated Thermal Power - High Function, this SR also Includes calibrating the associated recirculation loop flow channel.

Note 1 states that neutron detectors are excluded from CHANNEL CALIBRATION because they are passive devices, with minimal drift, and because of the difficulty of simulating a meaningful signal.

Changes In neutron detector sensitivity are compensated for by performing the 7 day calorimetric calibration (SR 3.3.1.1.2) and the 1000 effective full power hours LPRM calibration against the TIPs (SR 3.3.1.1.8). A second Note Is provided that requires the IRM SRs (continued).

HATCH UNIT I B 3.3-28 Amerment No. 238

Feedwater and Main Turbine High Water Level Trip Instrumentation B 3.3.2.2 B 3.3 INSTRUMENTATION B.3.3.2.2 Feedwater and Main Turbine High Water Level Trip Instrumentation BASES BACKGROUND The feedwater and main turbine high water level trip Instrumentation is designed to detect a potential failure of the Feedwater Level Control System that causes excessive feedwater flow.

With excessive feedwater flow, the water level Inthe reactor vessel rises toward the high water level setpoint, causing the trip of the two feedwater pump turbines and the main turbine.

Reactor Vessel Water Level - High signals are provided by level sensors that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level Inthe reactor vessel (variable leg). Three channels of Reactor Vessel Water Level - High nstrumentation are provided as Input to a two-out-of-three Initiation logic that trips the two feedwater pump turbines and the main turbine. The channels Include electronic equipment (e.g., trip relays) that compare measured Input signals with pre-established setpoints. When the setpolnt is exceeded, the channel output relay actuates, which then outputs a main feedwater and turbine trip signal to the trip logic.

A trip of the feedwater pump turbines limits further Increase In reactor vessel water level by limiting further addition of feedwater to the reactor vessel. A trip of the main turbine and closure of the stop valves protects the turbine from damage due to water entering the turbine.

APPLICABLE The feedwater and main turbine high water level trip Instrumentation SAFETY ANALYSES Is assumed to be capable of providing a turbine trip in the design basis transient analysis for a feedwater controller failure, maximum demand event (Ref. 1). The high level trip Indirectly Initiates a reactor scram from the main turbine trip (above 27.6Y RTP) and trips the I feedwater pumps, thereby terminating the event. The reactor scram mitigates the reduction In MCPRF Feedwater and main turbine high water level trip instrumentation satisfies Criterion 3 of the NRC Policy Statement (Ref. 3).

(continued)

HATCH UNIT I B 3.3-53 Ament No. 238

Feedwater and Main Turbine High Water Level Trip Instrumentation B 3.3.2.2 BASES (continued)

ICO The LCO requires three channels of the Reactor Vessel Water Level - High instrumentation to be OPERABLE to ensure that no single Instrument failure will prevent the feedwater pump turbines and main turbine trip on a valid Reactor Vessel Water Level - High signal.

Two of the three channels are needed to provide trip signals In order for the feedwater and main turbine trips to occur. Each channel must have its setpoint set within the specified Allowable Value of SR 3.3.2.2.2. The Allowable Value is set to ensure that the thermal limits are not exceeded during the event. The setpoint Is calibrated to be consistent with the applicable setpoint methodology assumptions (nominal trip setpoint). Nominal trip setpolnts are specified Inthe setpoint calculations. The nominal setpoints are selected to ensure that the setpolnts do not exceed the Allowable Value between successive CHANNEL CALIBRATIONS. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within Its Allowable Value, is acceptable.

Trip setpoints are those predetermined values of output at which an action should take place. The setpolnts are compared to the actual process parameter (e.g., reactor vessel water level), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip relay) changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis. The Allowable Values are derived from the analytic limits, corrected for calibration, process, and some of the instrument errors. A channel Is Inoperable If Its actual trip setpoint Is not within Its required Allowable Value. The trip setpoints are then determined accounting for the remaining Instrument errors (e.g., drift). The trip setpoints derived Inthis manner provide adequate protection because nstrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe environmental effects (for channels that must function In harsh environments as defined by 10 CFR 50.49) are accounted for.

APPLICABILITY The feedwater and main turbine high water level trip instrumentation is required to be OPERABLE at 2 24% RTP to ensure that the specified acceptable fuel design limits are not violated during the feedwater controller failure, maximum demand event. As discussed In the Bases for LCO 3.2.1, Average Planar Linear Heat Generation Rate (APLHGR)," and LCO 3.2.2, MINIMUM CRITICAL POWER RATIO (MCPR)," sufficient margin to these limits exists below 24% RTP; therefore, these requirements are only necessary when operating at or above this power level.

(continued)

HATCH UNIT I B 33-54 Amendment No. 238

Feedwater and Main Turbine High Water Level Trip Instrumentation B3.3.2.2 BASES ACTIONS B.1 (continued) not maintained). Therefore, continued operation is only permitted for a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months period, during which feedwater and main turbine high water level trip capability must be restored. The trip capability Is considered maintained when sufficient channels are OPERABLE or Intrip such that the feedwater and main turbine high water level trip logic will generate a trip signal on a valid signal. This requires two channels to each be OPERABLE or In trip. f the required channels cannot be restored to OPERABLE status or placed in trip, Condition C must be entered and Its Required Action taken.

The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time Is sufficient for the operator to take corrective action, and takes into account the likelihood of an event requiring actuation of feedwater and main turbine high water level trip Instrumentation occurring during this period. It Is also consistent with the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time provided n LCO 3.2.2 for Required Action A.i, since this Instrumentation's purpose is to preclude a MCPR violation.

C.^i With the required channels not restored to OPERABLE status or placed Intrip, THERMAL POWER must be reduced to < 24% RTP I within 4 hours0.167 days 0.0238 weeks 0.00548 months . As discussed Inthe Applicability section of the Bases, operation below 24% RTP results In sufficient margin to the required I limits, and the feedwater and main turbine high water level trip Instrumentation is not required to protect fuel Integrity during the feedwater controller failure, maximum demand event. The allowed Completion Time of 4 hours0.167 days 0.0238 weeks 0.00548 months Is based on operating experience to reduce THERMAL POWER to < 24% RTP from full power conditions I In an orderly manner and without challenging plant systems.

SURVEILLANCE The Surveillances are modified by a Note to Indicate that when a REQUIREMENTS channel Is placed In an inoperable status solely for performance of required Surveillances, entry Into associated Conditions and Required Actions may be delayed for up to 6 hours0.25 days 0.0357 weeks 0.00822 months provided the associated Function maintains feedwater and main turbine high water level trip capability. Upon completion of the Surveillance, or expiration of the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note Is based on the reliability analysis (Ref. .2) assumption of the average time required to perform channel Surveillance. That analysis (continued)

HATCH UNIT I B 3.3-56

EOC-RPT Instrumentation B 3.3.4.1 BASES (continued)

APPLICABLE The TSV - Closure and the TCV Fast Closure, Trip Oil SAFETY ANALYSES, Pressure - Low Functions are designed to trip the recirculation LCO, and pumps Inthe event of a turbine trip or generator load rejection to APPLICABILITY mitigate the increase In neutron flux, heat flux, and reactor pressure, and to increase the margin to the MCPR SL. The analytical methods and assumptions used In evaluating the turbine trip and generator load rejection are summarized In References 2 and 3.

To mitigate pressurization transient effects, the EOC-RPT must trip the recirculation.pumps after nitiation of closure movement of either the TSVs or the TCVs. The combined effects of this trip and a scram reduce fiel bundle power more rapidly than a scram alone, resulting In an increased margin to the MCPR SL Aternatively, MCPR limits for an Inoperable EOC-RPT, as specified Inthe COLR, are sufficient to prevent violation of.the MCPR Safety Umit. The EOC-RPT function Is automatically disabled when turbine first stage pressure Is

< 27.6% RTP.

EOC-RPT instrumentation satisfies Criterion 3 of the NRC Policy Statement (Ref. 6)..

The OPERABILITY of the EOC-RPT Is dependent on the OPERABIUTY of the Individual nstrumentation channel Functions.

Each Function must have a required number of OPERABLE channels In each trip system, with their setpoints within the specified Allowable Value of SR 3.3.4.1.3. The setpoint is calibrated consistent with applicable setpoint methodology assumptions (nominal trip setpoint).

Channel OPERABILITY also Includes the associated EOC-RPT breakers. Each channel (including the associated EOC-RPT breakers) must also respond within Its assumed response time.

Allowable Values are specified for each EOC-RPT Function specified In the LCO. Nominal trip setpoInts are specified In the setpoint calculations. A channel Is Inoperable its actual trip setpoInt Is not within Its required Allowable Value. The nominal setpoints are selected to ensure that the setpoints do not exceed the -Allowable Value between successive CHANNEL CALIBRATIONS. Operation with a trip setpolnt less conservative than the nominal trip setpoint, but within Its Allowable Value, Is acceptable. Each Allowable Value specified is more conservative than the analytical limit assumed In the transient and accident analysis In order to account for Instrument uncertainties appropriate to the Function. Trip setpoints are those predetermined values of output at which an action should take place.

The setpoints are compared to the actual process parameter (e.g.,

TSV position), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip (continued)

HATCH UNIT I B 3.3-76 Ameent No. 238

EOC-RPT Instrumentation B 3.3.4.1 BASES APPLICABLE relay) changes state. The analytic limits are derived from the limiting SAFETY ANALYSES, values of the process parameters obtained from the safety analysis.

LCO, and The Allowable Values are derived from the analytic limits, corrected APPLICABILITY for calibration, process, and some of the Instrument errors. The trip (continued) setpolrits are then determined accounting for the remaining Instrument errors (e.g., drift). The trip setpoints derived In this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, Instrument drift, and severe environmental effects (for channels that must function Inharsh environments as defined by 10 CFR 50.49) are accounted for.

The specific Applicable Safety Analysis, LCO, and Applicability discussions are listed below on a Function by Function basis.

Alternatively, since this instrumentation protects against a MCPR SL violation, with the Instrumentation Inoperable, modifications to the MCPR limits (LCO 3.2.2) may be applied to allow this LCO to be met.

The MCPR penalty for the EOC-RPT Inoperable condition Is specified in the COLR.

Turbine StoD Valve - Closure Closure of the TSVs and a main turbine trip result In the loss of a heat sink and Increases reactor pressure, neutron flux, and heat flux that must be limited. Therefore, an RPT Is Initiated on a TSV - Closure signal before the TSVs are completely closed In anticipation of the effects that would result from closure of these valves. EOC-RPT decreases reactor power and aids the reactor scram In ensuring that the MCPR SL Is not exceeded during the worst case transient Closure of the TSVs Is determined by measuring the position of each valve. While there are two separate position switches associated with each stop valve, only the signal from one'swItch for each TSV Is used, with each of the four channels being assigned to a separate trip channel. The logic for the TSV - Closure Function Is such that two or more TSVs must be closed to produce an EOC-RPT. This Function must be enabled at THERMAL POWER 2 27.6% RTP. This Is normally accomplished automatically by pressure switches sensing turbine first stage pressure; therefore, opening of the turbine bypass valves may affect this Function. Four channels of TSV - Closure, with two channels In each trip system, are available and required to be OPERABLE to ensure that no single Instrument failure will preclude an EOC-RPT from this Function on a valid signal. The TSV - Closure Allowable Value Is selected to detect Imminent TSV closure.

(continued)

HATCH UNIT 1 B 3.3-77 Amermnt No. 238

EOC-RPT Instrumentation B 3.3.4.1 BASES APPLICABLE Turbine Stop Valve - Closure (continued)

SAFETY ANALYSISD LCO, and This protection Is required, consistent with the safety analysis APPLICABILITY assumptions, whenever THERMAL POWER Is k 27.6% RTP. Below 27.6% RTP, the Reactor Vessel Steam Dome Pressure - High and the Average Power Range Monitor (APRM) Neutron Flux - High Functions of the Reactor Protection System (RPS) are adequate to maintain the necessary margin to the MCPR SL Turbine Control Valve Fast Closure. Trio Oil Pressure - Low Fast closure of the TCVs during a generator load rejection results In the loss of a heat sink that produces reactor pressure, neutron flux, and heat flux transients that must be limited. Therefore, an RPT is Initiated on TCV Fast Closure, Trip Oil Pressure - Low in anticipation of the transients that would result from the closure of these valves.

The EOC-RPT decreases reactor power and aids the reactor scram In ensuring that the MCPR SL Is not exceeded during the worst case transient.

Fast closure of the TCVs Is determined by measuring the electrohydraulic control fluid pressure at each control valve. There Is one pressure switch associated with each control valve, and the signal from each switch Is assigned to a separate trip channel. The logic for the TCV Fast Closure, Trip Oil Pressure - Low Function Is such that two or more TCVs must be closed (pressure transmitter trips) to produce an EOC-RPT. This Function must be enabled at THERMAL POWER f 27.6% RTP. This Is normally accomplished automatically by pressure switches sensing turbine first stage pressure; therefore, opening of the turbine bypass valves may affect this Function. Four channels of TCV Fast Closure, Trip Oil Pressure - Low, with two channels Ineach trip system, are available and required to be OPERABLE to ensure that no single Instrument failure will preclude an EOC-RPT from this Function on a valid signal.

The TCV Fast Closure, Trip Oil Pressure - Low Allowable Value Is selected high enough to detect imminent TCV fast closure.

This protection Is required consistent with the safety analysis whenever THERMAL POWER Is 2 27.6% RTP. Below 27.6% RTP, the Reactor Vessel Steam Dome Pressure - High and the APRM Neutron Flux - High Functions of the RPS are adequate to maintain the necessary margin to the MCPR SL (continued)

HATCH UNIT I B .3-78 Amelnet No. 238

EOC-RPT Instrumentation B 3.3.4.1 BASES ACTIONS B.1 and B.2 (continued)

Required Actions B.1 and B.2 are Intended to ensure that appropriate actions are taken If multiple, Inoperable, untripped channels within the same Function result In the Function not maintaining EOC-RPT trip capability. A Function is considered to be maintaining EOC-RPT trip capability when sufficient channels are OPERABLE or In trip, such that the EOC-RPT System will generate a trip signal from the given Function on a valid signal and both recirculation pumps can be tripped. Alternately, Required Action B.2 requires the MCPR limit for inoperable EOC-RPT, as specified Inthe COLR, to be applied. This also restores the margin to MCPR assumed Inthe safety analysis.

The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time is sufficient time for the operator to take corrective action, and takes Into account the likelihood of an event requiring actuation of the EOC-RPT Instrumentation during this period. It is also consistent with the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time provided in LCO 3.2.2 for Required Action A.1, since this Instrumentation's purpose is to preclude a MCPR violation.

C.1 and C.2 With any Required Action and associated Completion Time not met, THERMAL POWER must be reduced to < 27.6% RTP within 4 hours0.167 days 0.0238 weeks 0.00548 months . I Alternately, the associated recirculation pump may be removed from service, since this performs the intended function of the instrumentation. The allowed Completion Time of 4 hours0.167 days 0.0238 weeks 0.00548 months is reasonable, based on operating experience, to reduce THERMAL POWER to c 27.6% RTP from full power conditions In an orderly I manner and without challenging plant systems.

SURVEILLANCE The Surveillances are modified by a Note to Indicate that when a REQUIREMENTS channel Is placed Inan Inoperable status solely for performance of required Surveillances, entry Into associated Conditions and Required Actions may be delayed for up to 6 hours0.25 days 0.0357 weeks 0.00822 months provided the associated Function maintains EOC-RPT trip capability. Upon completion of the Surveillance, or expiration of the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note Is based on the reliability analysis (Ref. 4) assumption of the average time required to perform channel Surveillance. That analysis demonstrated that the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months testing allowance does not significantly reduce the probability that the recirculation pumps will trip when necessary.

(continued)

HATCH UNIT 1 B 3.3 Amendmt No. 238

EOC-RPT Instrumentation

B3.3.4.1 BASES SURVEILLANCE SR 3.3.4.1.1 REQUIREMENTS (continued) A CHANNEL FUNCTIONAL TEST Is performed on each required channel to ensure that the entire channel will perform the Intended function. Any setpoint adjustment shall be consistent with the.

assumptions of the current plant specific setpoint methodology.

The 92 day on an ALTERNATE TEST BASIS Frequency Is based on a review of the surveillance test history and Reference 8.

SR 3.3.4.12 This SR ensures that an EOC-RPT Initiated from the TSV - Closure and TCV Fast Closure, Trip OH Pressure - Low Functions will not be Inadvertently bypassed when THERMAL POWER Is 2 27.p% RTP.

This Involves calibration of the bypass channels. Adequate margins for the Instrument setpoint methodologies are Incorporated Into the actual setpolnt. Because main turbine bypass flow can affect this setpolnt nonconservatively (THERMAL POWER Is derived from first stage pressure) the main turbine bypass valves must remain closed during the calibration at THERMAL POWER 2 27.6% RTP to ensure that the calibration Isvalid. If any bypass channel's setpoint Is nonconservative (i.e., the Functions are bypassed at k 27.6% RTP, either due to open maIn turbine bypass valves or other reasons), the affected TSV - Closure and TCV Fast Closure, Trip Oil Pressure - Low Functions are considered Inoperable. Alternatively, the bypass channel can be placed In the conservative condition (nonbypass). If placed in the nonbypass condition (Turbine Stop Valve - Closure and Turbine Control Valve Fast Closure, Trip Oil Pressure - Low Functions are enabled), this SR Is met with the channel considered OPERABLE.

The 24 month Frequency Is based on a review of the surveillance test history, drift of the associated Instrumentation, and Reference 7.

SR 3.3.4.1.3 CHANNEL CALIBRATION Is a complete check of the Instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy.

CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations, consistent with the plant specific setpoint methodology. For the TSV - Closure Function, this SR also Includes a physical Inspection and actuation of the switches.

(continued)

HATCH UNIT 1 B 3.3-81

Primary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE 1.b. Main Steam Line Pressure - Low SAFETY ANALYSES, LCO, and Low MSL pressure with the reactor at power Indicates that there may APPLICABILITY be a problem with the turbine pressure regulation, which could result (continued) in a low reactor vessel water level condition and the RPV cooling down more than 1000F/hour If the pressure loss Is allowed to continue. The Main Steam Une Pressure - Low Function Isdirectly assumed In the analysis of the pressure regulator failure (Ref. 2). For this event, the closure of the MSIVs ensures that the RPV temperature change limit (100 0F/hour) Is not reached. In addition, this Function supports actions to ensure that Safety Limit 2.1.1.1 Is not exceeded. (This Function closes the MSIVs prior to pressure decreasing below 785 psig, which results Ina scram due to MSIV closure, thus reducing reactor power to < 24% RTP.) I The MSL low pressure signals are Initiated from four switches that are connected to the MSL header. The switches are arranged such that, even though physically separated from each other, each switch Is able to detect low MSL pressure. Four channels of Main Steam Une Pressure - Low Function are available and are required to be OPERABLE to ensure that no single Instrument failure can preclude the solation function.

The Allowable Value was selected to be high enough to prevent excessive RPV depressurization.

The Main Steam Une Pressure - Low Function Is only required to be OPERABLE in MODE 1 since this Is when the assumed transient can occur (Ref. 2).

This Function Isolates the Group 1 valves.

1.c. Main Steam Line Flow - High Main Steam Line Flow - High Is provided to detect a break of the MSL and to Initiate closure of the MSIVs. If the steam were allowed to continue flowing out of the break, the reactor would depressurize and the core could uncover. If the RPV water level decreases too far, fuel damage could occur. Therefore, the Isolation s Initiated on high flow to prevent or minimize core damage. The Main Steam Line Fow -

High Function Is directly assumed In the analysis of the main steam line break (MSLB) (Ref. 2). The Isolation action, along with the scram function of the Reactor Protection System (RPS), ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46 and offsite doses do not exceed the 10 CFR 100 limits.

(continued)

HATCH UNIT I B 3.3-141 Ametmcnt No. 238

S/RVs B 3.4.3 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 SIRVs 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 satisfies the Code requirement.

Each SIRV discharges steam through a discharge line to a point below the minimum water level In the suppression pool. The StRVs 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 (iLS)

Valves," and the ADS requirements are specified In LCO 3.5.1,

ECCS - Operating.'

APPLICABLE The overpressure protection system must accorrmodate the SAFETY ANALYSES most 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 operate in the safety mode. The analysis results demonstrate that the design S/RV capacity is capable of maintaining reactor pressure below the ASME Code limit of 110% of vessel design pressure (110%h x 1250 psig = 1375 psig). This LCO helps to ensure that the acceptance limit of 1375 psig Is met during the Design Basis I

Event.

(continued)

HATCH UNIT BB 3.4-1 0 Amendwmt No. 238

Main Condenser Offgas B 3.7.6 B 3.7 PLANT SYSTEMS B 3.7.6 Main Condenser Offgas BASES BACKGROUND During unit operation, steam from the low pressure turbine is exhausted directly Into the condenser. Air and noncondensable gases are collected Inthe condenser, then exhausted through the steam jet air ejectors (SJAEs) to the Main Condenser Ofigas System. The offgas from the main condenser normally Includes radioactive gases.

The Main Condenser Ofpgas System has been Incorporated into the unit design to reduce the gaseous radwaste emission. This system uses a catalytic recombiner to recombine radiolytically dissociated hydrogen and oxygen. The gaseous mixture Is cooled by the offgas condenser; the water and condensables are stripped out by the offgas condenser and moisture separator. The radioactivity of the remaining gaseous mixture (i.e., the offgas recombiner effluent) is monitored downstream of the moisture separator prior to entering the holdup line.

APPLICABLE The main condenser offgas gross gamma activity rate is an SAFETY ANALYSES Initial condition of the Main Condenser Offgas System failure event, discussed In the FSAR, Section 9.4 and Appendix E (Ref. ). The analysis assumes a gross fallure In the Main Condenser Offgas System that results Inthe rupture of the Main Condenser Offgas System pressure boundary. The gross gamma activity rate s controlled to ensure that during the event the calculated offslte doses will be well within the Omits of 10 CFR 100 (Ref. 2).

The main condenser offgas limits satisfy Criterion 2 of the NRC Policy Statement (Ref. 3).

LCO To ensure compliance with the assumptions of the Main Condenser Offgas System failure event (Ref. 1), the-fission product release rate should be consistent with a noble gas release to the reactor coolant of 100 .1CVMWt-second after decay of 30 minutes. This LCO Is established consistent with this requirement (2436 MWt x 100 pCVMWt-second = 240 mCl/second). The 240 mCVsecond limit Is conservative for a rated core thermal power of 2804 MWt. I (continued)

HATCH UNIT 1 B 3.7-31

Main Turbine Bypass System B3.7.7 BASES (continued)

LCO The Main Turbine Bypass System is required to be OPERABLE to limit peak pressure In the main steam lines and maintain reactor pressure within acceptable limits during events that cause rapid pressurization, so that the Safety Limit MCPR Is not exceeded. With the Main Turbine Bypass System Inoperable, modifications to the MCPR limits LCO 3.2.2, MINIMUM CRITICAL POWER RATIO (MCPR)] may be applied to allow this LCO to be met. The MCPR limit for the Inoperable Main Turbine Bypass System Is specified in the COLR. An OPERABLE Main Turbine Bypass System requires the bypass valves to open In response to Increasing main steam line pressure. This response Is within the assumptions of the applicable analysis (Ref. 2).

APPLICABILITY The Main Turbine Bypass System Is required to be OPERABLE at a 24% RTP to ensure that the fuel cladding Integrity Safety Limit and I the cladding 1% plastic strain limit are not violated during the feedwater controller failure to maximum flow demand transient. As discussed Inthe Bases for LCO 3.2.1, AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR),u and LCO 3.2.2, sufficient margin to these limits exists at < 24% RTP. Therefore, these I requirements are only necessary when operating at or above this power level.

ACTIONS Ifthe Main Turbine Bypass System Is Inoperable (one or more bypass valves inoperable), or the MCPR limits for an Inoperable Main Turbine Bypass System, as specified In the COLR, are not applied, the assumptions of the design basis transient analysis may not be met.

Under such circumstances, prompt action should be taken to restore the Main Turbine Bypass System to OPERABLE status or adjust the MCPR limits accordingly. The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time Is reasonable, based on the time to complete the Required Action and the low probability of an event occurring during this period requiring the Main Turbine Bypass System.

B.1 Ifthe Main Turbine Bypass System cannot be restored to OPERABLE status or the MCPR limits for an Inoperable Main Turbine Bypass System are not applied, THERMAL POWER must be reduced to (continued)

HATCH UNIT I B 3.7-35 Famnmt No. 238

Main Turbine Bypass System B 3.7.7 BASES ACTIONS 3.1 (continued)

< 24% RTP. As discussed In the Applicability section, operation at

< 24% RTP results in sufficient margin to the required limits, and the Main Turbine Bypass System Is not required to protect fuel Integrity during the turbine generator load rejection transient. The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time Is reasonable, based on operating experience, to reach the required unit conditions from full power conditions n an orderly manner and without challenging unit systems.

SURVEILLANCE SR 3.7.7.1 REQUIREMENTS Cycling each main turbine bypass valve through one complete cycle of full travel demonstrates that the valves are mechanically OPERABLE and will function when required. The 31 day Frequency is based on engineering judgment, Is consistent with the procedural controls governing valve operation, and ensures correct valve positions. Operating experience has shown that these components usually pass the SR when performed at the 31 day Frequency.

Therefore, the Frequency Is acceptable from a reliability standpoint SR 3.7.7.2 The Main Turbine Bypass System Is required to actuate automatically to perform its design function. This SR demonstrates that, with the required system Initiation signals, the valves will actuate to their required position. The 24 month Frequency s based on the need to perform this Surveillance under the conditions that apply during a unit outage and because of the potential for an unplanned transient If the Surveillance were performed with the reactor at power. The 24 month Frequency is based on a review of the surveillance test history and Reference 5.

SR 3.7.7.3 This SR ensures that the TURBINE BYPASS SYSTEM RESPONSE TIME Isin compliance with the assumptions of the appropriate safety analysis. The response time limits are specified InTechnical Requirements Manual (Ref. 3). The 24 month Frequency Is based on the need to perform this Surveillance under the conditions that apply during a unit outage and because of the potential for an unplanned transient if the Surveillance were performed with the reactor at power.

(continued)

HATCH UNIT I B 3.7-36 Anendmt N. 238

(6) Southern Nuclear, pursuant to the Act and 10 CFR Parts 30 and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operation of the facility.

C. This renewed license shall be deemed to contain,.and is subject to, the conditions specified in the following Commission regulations in 10 CFR Chapter : Part 20, Section 30.34 of Part 30, Section 40.41 of Part 40, Section 50.54 of Part 50, and Section 70.32 of Part 70; all applicable provisions of the Act and the rules, regulations, and orders of the Commission now or hereafter in effect; and the additional conditions2 specified or incorporated below:

(1) Maximum Power Level Southern Nuclear is authorized to operate the facility at steady state reactor core power levels not in excess of 2,804 megawatts thermal, in accordance with the conditions specified herein.

(2) Technical Specifications The Technical Specifications (Appendix A) and the Environmental Protection Plan (Appendix B), as revised through Amendment No.

are hereby incorporated in the renewed license. Southern Nuclear shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3) Additional Conditions The matters specified in the following conditions shall be completed to the satisfaction of the Commission within the stated time periods following the issuance of the renewed license or within the operational restrictions indicated. The removal of these conditions shall be made by an amendment to the license supported by a favorable evaluation by the Commission.

(a) Fire Protection Southern Nuclear shall implement and maintain in effect all provisions of the fire protection program, which Is referenced in the the Updated Final Safety Analysis Report for the facility, as contained 2The original licensee authorized to possess, use, and operate the facility was Georgia Power Company (GPC). Consequently, certain historical references to GPC remain in certain license conditions.

Renewed License No. NPF-5 Amendment No. 180

Definitions 1.1 1.1 Definitions (continued)

PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor core and related Instrumentation. These tests are:

a. Described InChapter 14, Initial Tests and Operation, of the FSAR;

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

c. Otherwise approved by the Nuclear Regulatory Commission.

RATED THERMAL RTP shall be a total reactor core heat transfer rate to the reactor POWER (RTP) coolant of 2804 MWL I REACTOR The RPS RESPONSE TIME shall be that time interval from when the PROTECTION monitored parameter exceeds ts RPS trip setpoint at the channel SYSTEM (RPS) sensor until de-energization of the scram pilot valve solenoids. The RESPONSE TIME 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 SDM shall be the amount of reactivity by which the reactor Is subcritical MARGIN (SDM) or would be subcritical assuming that

a. The reactor is xenon free;

b. The moderator temperature is 680F; 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.

STAGGERED A STAGGERED TEST BASIS shall consist of the testing of one of TEST BASIS the systems, subsystems, channels, or other designated components during the Interval specified by the Surveillance Frequency; so that all systems, subsystems, channels, or other designated components are tested during n Surveillance Frequency Intervals, where n Is the total number of systems, subsystems, channels, or other designated components in the associated function.

THERMAL POWER THERMAL POWER shall be the total reabtor core heat transfer rate to the reactor coolant.

(continued) 1.1-5 Amendment No. 180 UNIT 2 HATCHI UNIT HATCH 2 1.1-5 Anendment No. 180

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 785 pslg or core flow c 10%h rated core flow.

THERMAL POWER shall be z 24% RTP. I 2.1.12 With the reactor steam dome pressure 2 785 psig and core flow 10% rated core flow:

MCPR shall be k 1.08 for two recirculation loop operation or 2 1.10 for single recirculation loop operation.

2.1.1.3 Reactor vessel water level shall be greater than the top of active Irradiated fuel.

2.1.2 Reactor Coolant System RCS} Pressure SL Reactor steam dome pressure shall be 5 1325 psig.

2.2 SL Violations With any Sl violation, the following actions shall be completed:

2.2.1 Within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , notify the NRC Operations Center, In accordance with 10 CFR 50.72.

2.2.2 Within 2 hours0.0833 days 0.0119 weeks 0.00274 months :

2.2.2.1 Restore compliance with all Sls; and 2.2.2.2 Insert all Insertable control rods.

2.2.3 Withn 24 hours1 days 0.143 weeks 0.0329 months , notify the plant manager, the corporate executive responsible for overall plant nuclear safety, and the offslte review committee.

(continued)

H-ATCH UNIT 2 2.0-1 Amendment No. 180

APLHGR 3.2.1 32 POWER DISTRIBUTION LIMITS 32.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)

LCO 32.1 All APLHGRs shall be less than or equal to the limits spealfied n the COLR.

APPUCABIUTY: THERMAL POWER 2 24% RTP. I ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Any APLHGR not within A.1 Restore APLHGR(s) to 2 hours0.0833 days 0.0119 weeks 0.00274 months limits. within limits.

B. Required Action and B.1 Reduce THERMAL 4 hours0.167 days 0.0238 weeks 0.00548 months associated Completion POWER to < 24% RTP.

Time not met. I SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.1.1 Verify all APLHGRs are less than or equal to the Once within limits specified in the COLR. 12 hours0.5 days 0.0714 weeks 0.0164 months after 2 24% RTP I AND 24 hours1 days 0.143 weeks 0.0329 months thereafter HATCH UNIT 2 3.2-1 Amendment No. 180

MCPR 3.2.2 3.2 POWER DISTRIBUTION LIMITS 3.2.2 MINIMUM CRITICAL POWER RATIO (MCPR)

LCO 3.2.2 All MCPRs shall be greater than or equal to the MCPR operating imits specified In the COLR.

APPLICABILITY: THERMAL POWER 2 24% RTP. I ACTIONS - _

CONDITION REQUIRED ACTION COMPLETION TIME A. Any MCPR not within limits. A.1 Restore MCPR(s) to 2 hours0.0833 days 0.0119 weeks 0.00274 months within limits.

B. Required Action and B.1 Reduce THERMAL 4 hours0.167 days 0.0238 weeks 0.00548 months associated Completion POWER to c 24% RTP. I Time not met.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.2.1 Verify all MCPRs are greater than or equal to the Once within limits specified In the COLR. 12 hours0.5 days 0.0714 weeks 0.0164 months after 2 24% RTP I AND 24 hours1 days 0.143 weeks 0.0329 months thereafter (continued)

HATCH UNIT 2 3.2'2 Amendment No. 180

RPS Instrumentation 3.3.1.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TiME C. One or more Functions with C.1 Restore RPS trip 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months RPS trip capability not capability.

maintained.

D. Required Action and D.1 Enter the Condition Immediately associated Completion referenced In Time of Condition A, B, Table 3.3.1.1-1 for the or C not met channel.

E. As required by Required E.1 Reduce THERMAL 4 hours0.167 days 0.0238 weeks 0.00548 months Action 1.1 and referenced POWER to < 27.6% I In Table 3.3.1.1-1. RTP.

F. As required by Required F.1 Be in MODE 2. 6 hours0.25 days 0.0357 weeks 0.00822 months Action D.1 and referenced In Table 3.3.1.1-1.

G. As required by Required G.1 Be In MODE 3. 12 hours0.5 days 0.0714 weeks 0.0164 months Action D.1 and referenced In Table 3.3.1.1-1.

H. As required by Required H.1 InItiate action to fully Immediately Action D.1 and referenced Insert all Insertable In Table 3.3.1.1-1. control rods Incore cells containing one or more fuel assemblies.

(continued)

HATCH UNIT 2 3.3-2 Amendment No. 180

RPS Instrumentation 3.3.1.1 ACTIONS (continued) .

CONDITION REQUIRED ACTiON COMPLETION TIME

1. As required by Required 1.1 Initiate alternate method 12 hours0.5 days 0.0714 weeks 0.0164 months Action D.A and referenced to detect and suppress In Table 3.3.1.1-1. thernal-hydraulic Instabilrty oscillations.

AND 1.2 Restore required 120 days channels to OPERABLE.

J. Required Action and J.1 Be InMODE 2. 4 hours0.167 days 0.0238 weeks 0.00548 months associated Completion Time of Condition I not met.

SURVEILLANCE REQUIREMENTS NOTES~ ~ ~ ~

1. Refer to Table 3.3.1.1-1 to determine which SRs apply for each RPS Function.

2. When a channel Is placed in an Inoperable status solely for.performance of required Surveillances, entry Into associated Conditions and Required Actions may be delayed for up to 6 hours0.25 days 0.0357 weeks 0.00822 months provided the associated Function maintains RPS trip capability.

SURVEILLANCE FREQUENCY SR 3.3.1.1.1 Perform CHANNEL CHECK 12 hours0.5 days 0.0714 weeks 0.0164 months SR 3.3.1.12 NNTE _ .

Not required to be performed until 12 hours0.5 days 0.0714 weeks 0.0164 months after THERMAL POWER Z 24% RTP. I Verify the absolute difference between the 7 days average power range monitor (APRM) channels and the calculated power Is 5 2% RTP while operating at 2 24% RTP. I (continued)

HATCH UNIT 2 3.3-3 Amendment No. 180

RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.1.11 Verify Turbine Stop Valve - Closure and 24 months Turbine Control Valve Fast Closure, Trip Oil Pressure - Low Functions are not bypassed when THERMAL POWER Is 2 27.6% RTP. I SR 3.3.1.1.12 Perform CHANNEL FUNCTIONAL TEST. 24 months SR 3.3.1.1.13 NOTES

1. Neutron detectors are excluded.

2. For Function 1, not required to be performed when entering MODE 2 from MODE 1 until 12 hours0.5 days 0.0714 weeks 0.0164 months after entering MODE 2.

Perform CHANNEL CALIBRATION. 24 months SR 3.3.1.1.14 (Not used.)

SR 3.3.1.1.15 Perform LOGIC SYSTEM FUNCTIONACTEST. 24 months SR 3.3.1.1.16 NOTES

1. Neutron detectors are excluded.

2. (Not used.)

3. For Function 5, One equals 4 channels for the purpose of determining the STAGGERED TEST BASIS Frequency.

Verify the RPS RESPONSE TIME is within limits. 24 months on a STAGGERED TEST BASIS (continued)

HATCH UNIT 2 3.3-5 Anendment No. 180

RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 1 of 3)

Reactor Protection System Instbumentaton APPUICABLE CONDlIlONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDONS SYSTEM ACTION 0.1 REQUIREMENTS VALUE

1. tntermediate Range Mortor

a. Neutron Flu - High 2 2(d) G SR &3.1.1.1 s 1201M25 SR 3.3.1.1.4 cdivisions of ftul SR 3.1.1.6 scale SR 33.1.1.7 SR 3.31.1.13 SR 3..1.1.16 6(a) 2(d) H SR S.3.1.1.1 S 12OM25 SR 33.1.1.5 dlivisons of MI SR 3+/-1.1.13 scale SR 33.1.1.15 h Inop 2 2(d) G SR 3.1.1A NA SR 1.1.16 6Sa) 2(d) H SR 31.1.5 NA SR 3.1.1.15 2 Average Power Range Monhor

a. Neubn Fla- High 2 SC) G SR 3M1.1.1 5 2h RTP (Setdown) SR 3.1.1.7 SR 3..1.1B SR 3.3.1.1.10 SR 3S.1.1.13

b. Simulated Thermal 1 3(c) F SR 33.1.1.1 5057W Power - High SR .31.12 66.8% RTP I SR 3.1.1 and s 115.5%

SR 3..1.1.10 RTp@b)

SR .31.1.13

c. Neutron Fklu- HKgh 3(c) F SR 31.1.1 s 120% RTP SR 3+/-1.1.2 SR 3L1.1.6 SR 3.3.1.1.10 SR &.1.1.13

d. Inop 1.2 3(c) G SR 3.3.1.1.10 NA (coninued)

(a) With any control rod withdrawn rom a core cell contalning one or more fel assemblies.

(b) 0.7W+ 56.8% - 0.57 AW RTP when reset for single Ioop operation per LCO SA.1, Rerciulation Loops Operating. I (c) Each APRM dcannel provides Inputs to both trp systens.

(d) One channel In each quadrant of Ihe core must be OPERABLE whenever the IRMs are required to be OPERABLE. Both the RWM and a second icensed operator must verily complance with the withdrawal sequence when less than three channels in anw trip system are OPERABLE.

HATCH UNIT 2 3.3-7 Amendment No. 180

RPS Instrumentation 3.3.1.1 Table 3.31.11 (page Sot 3)

Reactor Protection System Instrurntation APPUCABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE' FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE S. Tbine Stop Valve -Closure X27.6A RTP 4 E SR SR

.1.110%

331.1.11 dosed I1 SR 82.1.1.13 SR 1.1.16 SR 1.1.16

9. TurbineControl Valve Fast a 27.6% RTP 2 E SR 3.31.1.9 t6 s1g Closure, Tilp 0 Pressure - SR 1.1.11 Low SR 33.1.1.13 SR 33.1.1.15 SR 1.1.16

10. ReactorModeSwitch- 1.2 2 G SR 3+/-1.1.12 NA Shutdown Position SR 3+/-1.1.15 S(a) 2 H SR 3.3.1.1.12 NA SR &3.1.1.15

11. ManualScram 1,2 2 G SR 3.3.1.1.5 NA SR L31.1.15 2 H SR 3.3.1.1.5 NA SR 3.3.1.1.15 (a) Wth any contr d wihawn mn a core ce containing one or more l assernblies.

HATCH UNIT 2 3.3-9 Amendment No. 180

Feedwater and Main Turbine Trip High Water Level Instrumentation 3.3.2.2 3.3 INSTRUMENTATION 3.3.2.2 Feedwater and Main Turbine Trip High Water Level Instrumentation LCO 3.3.2.2 Three channels of feedwater and min turbine trip nstrumentation shall be OPERABLE.

APPLICABILITY: THERMAL POWER 2 24% RTP. I ACTIONS

- ---- I'VI-B~~~~f~L Separate Condition entry Is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One feedwater and main A.1 Place channel In trib. 7 days turbine high water level trip channel noperable.

B. Two or more feedwater and B.1 Restore feedwater and 2 hours0.0833 days 0.0119 weeks 0.00274 months main turbine high water main turbine high water level trip channels level trip capability.

Inoperable.

C. Required Action and C.1 Reduce THERMAL 4 hours0.167 days 0.0238 weeks 0.00548 months associated Completion POWER to c 24% RTP.

lime not met I HATCH UNIT 2 3.3-20 Amendment No. 180

EOC-RPT Instrumentation 3.3.4.1 3.3 INSTRUMENTATION 3.3.4.1 End of Cycle Recirculation Pump Trip (EOC-RPT) Instrumentation LCO 3.3.4.1 a. Two channels per trip system for each EOC-RPT Instrumentation Function listed below shall be OPERABLE:

1. Turbine Stop Valve (TSV) - Closure; and

2. Turbine Control Valve (TCV) Fast Closure, Trip Oil Pressure - Low.

OR

b. LCO 3.2.2, MINIMUM CRITICAL POWER RATIO (MCPR),, limits for Inoperable EOC-RPT as specified Inthe COLR are made applicable.

APPLICABILITY: THERMAL POWER 2 27.6% RTP. I ACTIONS

-I~~~~~~~~~~~~~~ iL~~~~~~1W

& ~Tt I I t.

Separate Condition entry Is allowed for each channel.

CONDITION REOUIRED ACTION COMPLETION TIME A. One or more channels A.1 Restore channel to 72 hours3 days 0.429 weeks 0.0986 months Inoperable. OPERABLE status.

OR A.2 -N NOTE Not applicable If Inoperable channel Is the result of an Inoperable breaker.

Place channel Intrip. 72 hours3 days 0.429 weeks 0.0986 months (continued)

HATCH UNIT 2 3.3-27 Amendment No. 180

EOC-RPT Instrumentation 3.3.4.1 ACTIONS (continued)

CONDITION REQUIRED ACTION COMPLETION TIME B. One or more Functions with B.1 Restore EOC-RPT trip 2 hours0.0833 days 0.0119 weeks 0.00274 months EOG-RPT trip capability not capability.

maintained.

AND B.2 Apply the MCPR limit 2 hours0.0833 days 0.0119 weeks 0.00274 months MCPR limit for Inoperable for noperable EOC-RPT not made EOC-RPT as specified applicable. Inthe COLR.

C. Required Action and C.1 Remove the associated 4 hours0.167 days 0.0238 weeks 0.00548 months associated Completion recirculation pump from Time not met, service.

OR C.2 Reduce THERMAL 4 hours0.167 days 0.0238 weeks 0.00548 months POWER to c 27.6% I RTP.

SURVEILLANCE REQUIREMENTS IL rai I When a channel Is placed Inan Inoperable status solely for performance of required Surveillances, entry Into associated Conditions and Required Actions may be delayed for up to 6 hours0.25 days 0.0357 weeks 0.00822 months provided the associated Function maintains EOC-RPT trip capability.

SURVEILLANCE FREQUENCY SR 3.3.4.1.1 Perform CHANNEL FUNCTIONAL TEST. 92 days on an ALTERNATE TEST BASIS SR 3.3.4.12 Verify TSV - Closure and TCV Fast Closure, Trip 24 months Oil Pressure - Low Functions are not bypassed when THERMAL POWER Is 2 27.6% RTP. I (continued)

HATCH UNIT 2 3.3-28 Amendment No. 180

Main Turbine Bypass System 3.7.7 3.7 PLANT SYSTEMS 3.7.7 Main Turbine Bypass System LCO 3.7.7 The Main Turbine Bypass System shall be OPERABLE.

OR LCO 3.2.2, MINIMUM CRmCAL POWER RATIO (MCPR)," limits for an inoperable Main Turbine Bypass System, as specified Inthe COLR, are made applicable.

APPLICABILITY: THERMAL POWER a 24% RTP. I ACTIONS CONDiTION REQUIRED ACTION COMPLETION TIME A. Requirements of the LCO A.1 Satisfy the requirements 2 hours0.0833 days 0.0119 weeks 0.00274 months not met of the LCO.

B. Required Action and B.1 Reduce THERMAL 4 hours0.167 days 0.0238 weeks 0.00548 months associated Completion POWER to < 24% RTP. I Time not met SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.7.1 Verify one complete cycle of each main turbine 31 days bypass valve.

SR 3.7.7.2 Perform a system functional test 24 months SR 3.7.7.3 Verify the TURBINE BYPASS SYSTEM 24 months RESPONSE TIME Iswithin limits.

HATCH UNIT 2 3.7-18 Amendment No. 180

Reactor Core SLs B 2.1.1 BASES BACKGROUND to a structurally weaker form. This weaker form may lose Its Integrity, (continued) resulting In an uncontrolled release of activity to the reactor coolant.

The reactor vessel water level SL ensures that adequate core cooling capability Is maintained during all MODES of reactor operation.

Establishment of Emergency Core Cooling System Initiation setpoints higher than this safety limIt provides margin such that the safety limit will not be reached or exceeded.

APPLICABLE The fuel cladding must not sustain damage as a result of normal SAFETY ANALYSES operation and AOOs. The reactor-core SLs are established to' preclude violation of the fuel design criterion that a MCPR limit s to be established, such that at least 99.9%/l of the fuel rods In the core would not be expected to experience the onset of transition boiling.

The Reactor Protection System setpolnts [LCO 3.3.1.1, Reactor Protection System (RPS) nstrumentations, Incombination with the other LCOs, are designed to prevent any anticipated combination of transient conditions for Reactor Coolant System water level, pressure, and THERMAL POWER level that would result In reaching the MCPR Safety Umit.

2.1.1.1 Fuel Claddinc Interity GE critical power correlations are applicable for all critical power calculations at pressures 2 785 psig and core flows k 10% of rated flow. For operation at low pressures or low flows, another basis Is used, as follows:

Since the pressure drop Inthe bypass region is essentially all elevation head, the core pressure drop at low power and flows will always be > 4.5 psi. Analyses (Ref. 2) show that with a bundle flow of 28 x 10' b1hr, bundle pressure drop is nearly Independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow with a 4.5 psi driving head will be > 28 x 10' lb/hr. Full scale ATLAS test data taken at pressures from 14.7 psla to 800 psla Indicate that the fuel assembly critical power at this flow Is approximately 3.35 MWt. With the design peaking factors, this corresponds to a THERMAL POWER

> 50°%RTP. Thus, a THERMAL POWER limit of 24% RTP for reactor pressure < 785 psig Is conservative. I (continued)

HATCH UNIT 2 B 2.0-2 Amendment No. 180

APLHGR B 32.1 B 32 POWER DISTRIBUTION UMITS B 32.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR)

BASES BACKGROUND The APLHGR Is a measure of the average LHGR of all the fuel rods in a fuel assembly at any axial location. imits on the APLHGR are specified to ensure that certain fuel design limits Identified In Reference 1 are not exceeded during anticipated operational occurrences (AOOs) and that the peak cladding temperature (PCT) during the postulated design basis loss of coolant accident (LOCA) does not exceed the limits specified n 10 CFR 50.46.

APPLICABLE The analytical methods and assumptions used Inevaluating the fuel SAFETY ANALYSES design limits are presented In References 1 and 2. The analytical methods and assumptions used in evaluating Design Basis Accidents (DBAs), anticipated operational transients, and normal operation that determine the APLHGR limits are presented In References 1, 2,3,4, 5,6, and 7.

Fuel design evaluations are performed to demonstrate that the 1%limit on the fuel cladding plastic strain and certain other fuel design limits described In Reference are not exceeded during AOOs for operation with LHGRs up to the operating mit UHGR. APUiGR limits are equivalent to the LHGR limit for each fuel rod divided by the local peaking factor of the fuel assembly. APLHGR limits are developed as a function of exposure and the various operating core flow and power states to ensure adherence to fuel design limits during the limiting AOOs (Refs. 5, 6, and 7). Flow dependent API-HGR limits are determined (Ref. 7) using the three dimensional BWR simulator code (Ref. 8) to analyze slow flow runout transients. The flow dependent multiplier, MAPFACn, Is dependent on the maximum core flow runout capability. The maximum runout flow Is dependent on the existing setting of the core flow limiter In the Recirculation Flow Control System.

Based on analyses of limiting plant transients (other than core flow Increases) over a range of power and flow conditions, power dependent multipliers, MAPFACp, are also generated. Due to the sensitivity of the transient response to Initial core flow levels at power levels below those at which turbine stop valve closure and turbine control valve fast closure scram trips are bypassed, both high and low core flow MAPFACp limits are provided for operation at power levels between 24% RTP and the previously mentioned bypass power level. I (continued)

HATCH UNIT 2 B 3.2-1 Amendment No. 180

APLHGR B 32.1 BASES (continued)

APPLICABILITY The APLHGR limits are primarily derived from fuel design evaluations and LOCA and transient analyses that are assumed to occur at high power levels. Design calculations (Ref. 7) and operating experience have shown that as power is reduced, the margin to the required APLHGR limits Increases. This trend continues down to the power range of 5% to 15% RTP when entry nto MODE 2 occurs. When In MODE 2, the intermediate range monitor scram function provides prompt scram initiation during any significant transient, thereby effectively removing any APLHGR limit compliance concern in MODE 2. Therefore, at THERMAL POWER levels:5 24% RTP, the I reactor is operating with substantial margin to the APLHGR limits; thus, this LCO is not required.

ACTIONS AM If any APLHGR exceeds the required limits, an assumption regarding an Initial condition of the DBA and transient analyses may not be met.

Therefore, prompt action should be taken to restore the APLHGR(s) to within the required limits such that the plant operates within analyzed conditions and within design limits of the fuel rods. The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time is sufficient to restore the APLHGR(s) to within Its limits and is acceptable based on the low probability of a transient or DBA occurring simultaneously with the APLHGR out of specification.

If the APLHGR cannot be restored to within Its required limits within the associated Completion Time, the plant must be brought to Ina MODE or other specified condition Inwhich the LCO does not apply.

To achieve this status, THERMAL POWER must be reduced to

< 24% RTP within 4 hours0.167 days 0.0238 weeks 0.00548 months . The allowed Completion Time Is reasonable, based on operating experience, to reduce THERMAL POWER to < 24% RTP Inan orderly manner and without challenging I;

plant systems.

SURVEILLANCE SR 3.2.1.1 REQUIREMENTS APLHGRs are required to be initially calculated within 12 hours0.5 days 0.0714 weeks 0.0164 months after THERMAL POWER Is2 24% RTP and then every 24 hours1 days 0.143 weeks 0.0329 months I thereafter. They are compared to the specified limits In the COLR to ensure that the reactor is operating within the assumptions of the (continued)

HATCH UNIT 2 B 32-3 Amendment No. 180

APLHGR B 32.1 BASES SURVEILLANCE SR 32.1.1 (continued)

REQUIREMENTS safety analysis. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency Is based on both engineering judgment and recognition of the slowness of changes In power distribution during normal operation. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowance after THERMAL POWER 2 24% RTP Isachieved Is acceptable given the I large Inherent margin to operating limits at low power levels.

REFERENCES 1. NEDE-2401 1-P-A General Electric Standard Application for Reactor Fuel," (revision specified in the COLR).

2. FSAR, Chapter 4.

3. FSAR, Chapter 6.

4. FSAR, Chapter 15.

6. NEDO-24205, 6E.I. Hatch Nuclear Plant Units 1 and 2 Single-Loop Operation,' August 1989.

6. NEDO-24395, Load Line Umit Analysis," October 1980.

7. NEDC-30474-P Average Power Range Monitor, Rod Block Monitor and Technical Specification Improvements (ARTS)

Program for E.l. Hatch Nuclear Plant, Units 1 and 2,8 December 1983.

8. NEDO-30130-A, Steady State Nuclear Methods," May 1985.

9. NEDO-24154, Qualification of the One-Dimensional Core Transient Model for Boiling Water Reactors, October 1978.

10. NEDO-31376, E.I. Hatch Nuclear Plant SAFER/GESTAR-LOCA Analysis," December 1986.

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

12. NEDC-33085-P, Safety Analysis Report for Edwin I. Hatch Units 1 and 2 Thermal Power Optimization," November 2002. I HATCH UNIT 2 B3.2-4 Amendment No. 180

MCPR B 3.2.2 BASES

APPLICABLE benchmarked using the three dimensional BWR simulator code SAFETY ANALYSES (Ref. 9) to analyze slow flow runout transients. The operating limit is (continued) dependent on the maximum core flow limiter setting Inthe Recirculation flow Control System.

Power dependent MCPR limits (MCPRp) are determined mainly by the one dimensional transient code (Ref. 10). Due to the sensitivity of the transient response to initial core flow levels at power levels below those at which the turbine stop valve dosure and turbine control valve fast closure scrams are bypassed, high and low flow MCPRI operating limits are provided for operating between 24% RTP I and the previously mentioned bypass power level.

The MCPR satisfies Criterion 2 of the NRC Policy Statement (Ref. 11).

LCO The MCPR operating limits specified in the COLR are the result of the Design Basis Accident (DBA) and transient analysis. The operating limit MCPR Is determined by the larger of the MCPRf and MCPR, limits.

APPLICABILITY The MCPR operating limits are primanly derived from transient analyses that are assumed to occur at high power levels. Below 24% RTP, the reactor Is operating at a minimum recirculation pump speed and the moderator void ratio Is small. Surveillance of thermal I.

limits below 24% RTP is unnecessary due to the large inherent margin that ensures that the MCPR SL Is not exceeded even if a limiting transient occurs. Statistical analyses indicate that the nominal value of the Initial MCPR expected at 24% RTP is > 3.S. Studies of the variation of limting transient behavior have been performed over the range of power and flow conditions. These studies encompass the range of key actual plant parameter values important to typically limiting transients. The results of these studies demonstrate that a margin Isexpected between performance and the MCPR requirements, and that margins increase as power Is reduced to 24% RTP. This trend is expected to continue to the 5% to 15% power I range when entry into MODE 2 occurs. When In MODE 2, the intermediate range monitor provides rapid scram Initiation for any significant power Increase transient, which effectively eliminates any MCPR compliance concem. Therefore, at THERMAL POWER levels

< 24% RTP, the reactor is operating with substantial margin to the I MCPR limits and this LCO is not required (continued)

HATCH UNIT 2 B 3.2-6 Amendment No. 180

MCPR B 3.2.2 BASES (continued)

ACTIONS M If any MCPR Is outside the required limits, an assumption regarding an initial condition of the design basis transient analyses may not be met. Therefore, prompt action should be taken to restore the MCPR(s) to within the required limits such that the plant remains operating within analyzed conditions. The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time is normally sufficient to restore the MCPR(s) to within Its limits and Is acceptable based on the low probability of a transient or DBA occurring simultaneously with the MCPR out of specification.

If the MCPR cannot be restored to within Its required limits within the associated Completion Time, the plant must be brought to a MODE or

-other specified condition in which the LCO does not apply. To achieve this status, THERMAL POWER must be reduced to < 24% RTP within 4 hours0.167 days 0.0238 weeks 0.00548 months . The allowed Completion Time Is reasonable, based on operating experience, to reduce THERMAL POWER to < 24% RTP In an orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.2.2.1 REQUIREMENTS The MCPR Is required to be Initially calculated within 12 hours0.5 days 0.0714 weeks 0.0164 months after THERMAL POWER is k 24% RTP and then every 24 hours1 days 0.143 weeks 0.0329 months thereafter. It Is compared to the specified limits In the COLR to ensure that the reactor Is operating within the assumptions of the safety analysis. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency is based on both engineering judgment and recognition of the slowness of changes In power distribution during normal operation. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowance after THERMAL POWER a 24% RTP is achieved Is acceptable given the large Inherent margin to operating limits at low power levels.

SR 3.2.2.2 Because the transient analysis takes credit for conservatism Inthe scram speed performance, It must be demonstrated that the specific scram speed distribution is consistent with that used Inthe transient analysis. SR 3.2.2.2 determines the value of a,which Is a measure of the actual scram speed distribution compared with the assumed distribution. The MCPR operating limit Is then determined based on an interpolation between the applicable limits for Option A (scram (continued)

IHATCH UNIT 2 B 3.2-7 Amendment No. 180

RPS Instrumentation B 3.3.1.1 BASES APPLICABLE the specified Allowable Value, where appropriate. The setpoint is SAFETY ANALYSES, calibrated consistent with applicable setpoint methodology LCO, and assumptions (nominal trip setpoint). Each channel must also respond APPU CABILITY within its assumed response time, where. appropriate.

(continued)

Allowable Values are specified for each RPS Function specified Inthe Table. Nominal trip setpoints are specified n the setpoint calculations. The nominal setpoints are selected to ensure that the actual setpoints do not exceed the Allowable Value between successive CHANNEL CALIBRATIONS. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within Its Allowable Value, is acceptable. A channel Is Inoperable If its actual trip setpoint Is not within its required Allowable Value.

Trip setpolnts are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., reactor vessel water level), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip unit) changes state. The analytic limits are derived from the limiting values of the process parameters obtained from the safety analysis. The Allowable Values are derived from the analytic limits, corrected for calibration, process, and some of the Instrument errors.

The trip setpoints are then determined accounting for the remaining Instrument errors (e.g., drift). The trip setpoints derived In this manner provide adequate protection because Instrumentation uncertainties, process effects, calibration tolerances, Instrument drift, and severe environmental effects (for channels that must function in harsh environments as defined by 10 CFR 50 49) are accounted for.

The OPERABILITY of scram pilot valves and associated solenoids, backup scram valves, and SDV valves, described in the Background section, are not addressed by this LCO.

The individual Functions are required to be OPERABLE In the MODES or other specified conditions specified Inthe Table, which may require an RPS trip to mitigate the consequences of a design basis accident or transient. To ensure a reliable scram function, a combination of Functions are required In each MODE to provide primary and diverse nitiation signals. The only MODES specified In Table 3.3.1.1-1 are MODES 1 (which encompasses z 27.6% RTP) and 2, and MODE 5 with any control rod withdrawn from a core cell containing one or more fuel assemblies. No RPS Function s required In MODES 3 and 4 since all control rods are fully Inserted and the Reactor Mode Switch Shutdown Position control rod withdrawal block (continued)

HATCH UNIT 2 B 3.3-3 Amendment No. 180

RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 2.a. Average Power Range Monitor Neutron Flux- High (Setdown)

SAFETY ANALYSES, (continued)

LCO, and APPLICABIUTY abnormal operating transients In this power range. For most operation at low power levels, the Average Power Range Monitor Neutron Flux - High (Setdown) Function will provide a secondary scram to the Intermediate Range Monitor Neutron Flux - High Function because of the relative setpoints. With the IRMs at Range 9 or 10, Itis possible that the Average Power Range Monitor Neutron Flux - High (Setdown) Function will provide the primary trip signal for a oorewide Increase In power.

No specific safety analyses take direct credit for the Average Power Range Monitor Neutron Flux - High (Setdown) Function. However, this Function Indirectly ensures that before the reactor mode switch Is placed Inthe run position, reactor power does not exceed 24% RTP I (SL 2.1.1.1) when operating at low reactor pressure and low core flow.

Therefore, it Indirectly prevents fuel damage during significant reactivity Increases with THERMAL POWER < 24% RTP. I The Allowable Value Is based on preventing significant Increases. In power when THERMAL POWER Is < 24% RTP. I The Average Power Range Monitor Neutron Flux - High (Setdown)

Function must be OPERABLE during MODE 2 when control rods may be withdrawn since the potential for criticality exists.

In MODE , the Average Power Range Monitor Neutron Flux - High Function provides protection against reactivity transients and the RWM and rod block monitor protect against control rod withdrawal error events.

2.b. Average Power Range Monitor Simulated Thermal Power - High The Average Power Range Monitor Simulated Thermal Power - High Function monitors neutron flux to approximate the THERMAL POWER being transferred to the reactor coolant The APRM neutron flux is electronically filtered with a time constant representative of the fuel heat transfer dynamics to generate a signal proportional to the THERMAL POWER Inthe reactor. The trip level Is varied as a funcrtion of recirculation drive flow (i.e., at lower core flows, the setpoint s reduced proportional to the reduction In power experienced as core flow is reduced with a fixed control rod pattern) but Is damped at an upper limit that Is always lower than the Average Power Range Monitor Neutron Flux - High Function Allowable Value.

(continued)

HATCH UNIT 2 B 3.3-7 Amendment No. 180

RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 8. Turbine Stop Valve - Closure (continued)

SAFETY ANALYSES, LCO, and reactor scram reduces the amount of energy required to be absorbed APPLICABILITY and, along with the actions of the End of Cycle Recirculation Pump Trip (EOC-RPT) System, ensures that the MCPR SL is not exceeded.

Turbine Stop Valve - Closure signals are Initiated from position switches located on each of the four TSVs. Two Independent position switches are associated with each stop valve. One of the two switches provides Input to RPS trip system A; the other, to RPS trip system B. Thus, each RPS trip system receives an Input from four Turbine Stop Valve - Closure channels, each consisting of one position switch. The logic for the Turbine Stop Valve - Closure Function Is such that three or more TSVs must be closed to produce a scram. In addition, certain combinations of two valves closed will result In a half-scram. This Function must be enabled at THERMAL POWER k 27.6% RTP. This Is normally accomplished automatically by pressure switches sensing turbine first stage pressure; therefore, opening of the turbine bypass valves may affect this Function.

The Turbine Stop Valve - Closure Allowable Value Is selected to be high enough to detect Imminent TSV closure, thereby reducing the severity of the subsequent pressure transient Eight channels of Turbine Stop Valve - Closure Function, with four channels In each trip system, are required to be OPERABLE to ensure that no single Instrument failure will preclude a scram from this Function If the TSVs should close. This Function Is required, consistent with analysis assumptions, whenever THERMAL POWER Is 27.6% RTP. This Function Is not required when THERMAL POWER Is < 27.6%/ RTP since the Reactor Vessel Steam Dome Pressure - High and the Average Power Range Monitor Neutron Flux - High Functions are adequate to maintain the necessary safety margins.

9. Turbine Control Valve Fast Closure. Trin Oil Pressure - Low Fast closure of the TCVs results Inthe loss of a heat sink that produces reactor pressure, neutron flux, and heat flux transients that must be limited. Therefore, a reactor scram Is Initiated on TCV fast closure In anticipation of the transients that would result from the closure of these valves. The Turbine Control Valve Fast Closure, Trip Oil Pressure - Low Function Isthe primary scram signal for the generator load rejection event analyzed In-Reference 2. For this event, the reactor scram reduces the amount of energy required to be (continued)

HATCH UNIT 2 B .3-16 Amendment No. 180

RPS Instrumentation B 3.3.1.1 BASES APPLICABLE 9. Turbine Control Valve Fast Closure, Trip Oil Pressure - Low SAFETY ANALYSES, (continued)

LCO, and APPLICABILITY absorbed and, along with the actions of the EOC-RPT System, ensures that the MCPR SL is not exceeded.

Turbine Control Valve Fast Closure, Trip Oil Pressure - Low signals are initiated by the electrohydraulic control (EHC) fluid pressure at each control valve. One pressure switch Is associated with each control valve, and the signal from each switch Is assigned to a separate RPS logic channel. This Function must be enabled at THERMAL POWER z 27.6% RTP. This Is normally accomplished automatically by pressure switches sensing turbine first stage pressure; therefore, opening of the turbine bypass valves may affect this Function.

The Turbine Control Valve Fast Closure, Trip Oil Pressure - Low Allowable Value Is selected high enough to detect Imminent TCV fast closure.

Four channels of Turbine Control Valve Fast Closure, Trip Oil Pressure - Low Functiori with two channels In each trip system arranged In a one-out-of-two logic are required to be OPERABLE to ensure that no single Instrument failure will preclude a scram from this Function on a valid signal. This Function Is required, consistent with the analysis assumptions, whenever THERMAL POWER Is 2 27.6% RTP. This Function Is not required when THERMAL POWER Is < 27.6% RTP, since the Reactor Vessel Steam Dome Pressure - High and the Average Power Range Monitor Neutron Flux - High Functions are adequate to maintain the necessary safety margins.

10. Reactor Mode Switch - Shutdown Position The Reactor Mode Switch - Shutdown Position Function provides signals, via the manual scram logic channels, to each of the four RPS logic channels, which are redundant to the automatic protective Instrumentation channels and provide manual reactor trip capability.

This Function was not specifically credited In the accident analysis, but It Is retained for the overall redundancy and diversity of the RPS as required by the NRC approved licensing basis.

The reactor mode switch is a single switch with four channels, each of which provides Input Into one of the RPS logic channels.

(continued)

HATCH UNIT 2 B 3.3-1 7 HAmendment No. 180

RPS Instrumentation B 3.3.1.1 BASES SURVILLANCE SR 3.3.1.1.1 (continued)

REQUIREMENTS between Instrument channels could be an indication of excessive instrument drift In one of the channels or something even more serious. A CHANNEL CHECK will detect gross channel failure; thus, It Is key to verifying the instrumentation continues to operate properly between each CHANNEL CALIBRATION.

Agreement criteria are determined by the plant staff based on a combination of the channel Instrument uncertainties, ncluding Indication and readability. If a channel Is outside the criteria. It may be an Indication that the Instrument has drifted outside ts limit.

The Frequency Is based upon operating experience that demonstrates channel failure Is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays associated with the channels required by the LCO.

-SR 3.3.1.1.2 To ensure that the APRMs are accurately indicating the true core average power, the APRMs are calibrated to the reactor power calculated from a heat balance. The Frequency of once per 7 days Is based on minor changes InLPRM sensitivity, which could affect the APRM reading between performances of SR 3.3.1.1.8.

A restriction to satisfying this SR when c 24% RTP Is provided thatl requires the SR to be met only at 2 24% RTP because It is difficult to accurately maintain APRM Indication of core THERMAL POWER consistent with a heat balance when < 24% RTP. At low power levels, a high degree of accuracy Is unnecessary because of the large, Inherent margin to thermal limits (MCPR and APLHGR). At 2 24% RTP, the Surveillance Is required to have been satisfactorily performed within the last 7 days, In accordance with SR 3.0.2. A Note Is provided which allows an Increase InTHERMAL POWER above 24% If the 7 day Frequency Is not met per SR 3.0.2. In this event, the SR must be performed within 12 hours0.5 days 0.0714 weeks 0.0164 months after reaching or exceeding 24% RTP. Twelve hours Is based on operating experience and In consideration of providing a reasonable time In which to complete the SR.

(continued)

HATCH UNIT 2 B 3.3-24 Amendment No. 180

RPS Instrumentation B 3.3.1.1 BASES SURVEILLANCE SR 3.3.1.1.11 (continued)

REQUIREMENTS POWER Is 2 27.6%6 RTP. This Involves calibration of the bypass channels. Adequate margins for the Instrument setpolnt methodologies are Incorporated Into the actual setpoint. Because main turbine bypass flow can affect this setpoint nonconservatively (THERMAL POWER is derived from turbine first stage pressure), the main turbine bypass valves must remain dosed during the calibration at THERMAL POWER a 27.6% RTP to ensure that the calibration Is valid.

If any bypass channers setpoint Is nonconservative (i.e., the Functions are bypassed at 2 27.6% RTP, either due to open main turbine bypass valve(s) or other reasons), then the affected Turbine Stop Vaive - Closure and Turbine Control Valve Fast Closure, Trip Oil Pressure - Low Functions are considered Inoperable. Alternatively, the bypass channel can be placed Inthe conservative condition (nonbypass). If placed in the nonbypass condition (Turbine Stop Valve - Closure and Turbine Control Valve Fast Closure, Trip Oil Pressure - Low Functions are enabled), this SR Is met and the channel Is considered OPERABLE.

The 24 month Frequency Is based on a review of the surveillance test history, drift of the associated instrumentation, and Reference 20.

SR 33.1.1.13 A CHANNEL CALIBRATION Is a complete check of the Instrument loop and the sensor. This test verifies that the channel responds to the measured parameter within the necessary range and accuracy.

CHANNEL CALIBRATION leaves the channel adjusted to account for Instrument drifts between successive calibrations, consistent with the plant specific setpoint methodology. For MSIV - Closure, SDV Water Level - High (Float Switch), and TSV - Closure Functions, this SR also Includes a physical Inspection and actuation of the switches. For the APRM Simulated Thermal Power - High Function, this SR also Includes calibrating the associated recirculation loop flow channel.

Note 1 states that neutron detectors are excluded from CHANNEL CALIBRATION because they are passive devices, with minimal drift and because of the difficulty of simulating a meaningful signal.

Changes In neutron detector sensitivity are compensated for by performing the 7 day calorimetric calibration (SR 3.3.1.1.2) and the 1000 effective full power hours LPRM calibration against the TIPs (SR 3.3.1.1.8). A second Note is provided that requires the IRM SRs (continued)

HATCH UNIT 2 B 3.3-28 Amendment No. 180

Feedwater and Main Turbine High Water Level Trip Instrumentation B 3.3.2.2 B 3.3 INSTRUMENTATION B 3.3.2.2 Feedwater and Main Turbine High Water Level Trip Instrumentation BASES BACKGROUND The feedwater and main turbine high water level trip nstrumentation is designed to detect a potential failure of the Feedwater Level Control System that causes excessive feedwater flow.

With excessive feedwater flow, the water level Inthe reactor vessel rises toward the high water level setpoint, causing the trip of the two feedwater pump turbines and the main turbine.

Reactor Vessel Water Level - High signals are provided by level sensors that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level in the reactor vesseL.(variable leg). Three channels of Reactor Vessel Water Level - High nstrumentation are provided as input to a two-outf-three initiation logic that trips the two feedwater pump turbines and the main turbine. The channels include electronic equipment (e.g., trip relays) that compare measured input signals with pre-established setlioints. When the setpoint Is exceeded, the channel output relay actuates, which then outputs a main feedwater and turbine trip signal to the trip logic.

A trip of the feedwater pump turbines limits further Increase Inreactor vessel water level by limiting further addition of feedwater to the reactor vessel. A trip of the main turbine and closure of the stop valves protects the turbine from damage due to water entering the turbine.

APPLICABLE The feedwater and main turbine high water level trip instrumentation SAFETY ANALYSES Is assumed to be capable of providing a turbine trip Inthe design basis transient analysis for a feedwater controller failure, maximum demand event (Ref. 1). The high level trip Indirectly Initiates a reactor scram from the main turbine trip (above 27.6% RTP) and trips the I feedwater pumps, thereby terminating the event The reactor scram mitigates the reduction InMCPR.

Feedwater and main turbine high water level trip Instrumentation satisfies Criterion 3 of the NRC Policy Statement (Ref. 3).

(continued)

HATCH UNIT 2 B 3.3-53 Amendment No. 180

Feedwater and Main Turbine High Water Level Trip Instrumentation B 3.3.2.2 BASES (continued)

LCO The LCO requires three channels of the Reactor Vessel Water Level - High nstrumentation to be OPERABLE to ensure that no single Instrument failure will prevent the feedwater pump turbines and main turbine trip on a valid Reactor Vessel Water Level - High signal.

Two of the three channels are needed to provide trip signals In order for the feedwater and main turbine trips to occur. Each channel must have its setpoint set within the specified Allowable Value of SR 3.3.2.2.2. The Allowable Value Is set to ensure that the thermal limits are not exceeded during the event. The setpoint Is calibrated to be consistent with the applicable setpoint methodology assumptions (nominal trip setpoint). Nominal trip setpoints are specified In the stpoint calculations. The nominal setpoints are selected to ensure that the setpolnts do not exceed the Allowable Value between successive CHANNEL CALIBRATIONS. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within Its Allowable Value, Is acceptable.

Trip setpoints are those predetermined values of output at which an action should take place. The setpoints are compared to the actual process parameter (e.g., reactor vessel water level), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip relay) changes state. The analytic limits are derived from the limitir values of the process parameters obtained from the safety analysis. The Allowable Values are derived from the analytic limits, corrected for calibration, process, and some of the Instrument errors. A channel is Inoperable if Its actual trip setpoint Is not within Its required Allowable Value. The trip setpolnts are then determined accounting for the remaining instrument errors (e.g., drift). The trip setpoints derived In this manner provide adequate protection because instrumentation uncertainties, process effects, calibration tolerances, Instrument drift, and severe environmental effects (for channels that must function In harsh environments as defined by 10 CFR 50.49) are accounted for.

APPLICABILflY The feedwater and main turbine high water level trip nstrumentation Is required to be OPERABLE at 2 24% RTP to ensure that the specified acceptable fuel design limits are not violated during the feedwater controller failure, maximum demand event. As discussed In the Bases for LCO 3.2.1, Average Planar Linear Heat Generation Rate (APLHGR)," and LCO 3.2.2, "MINIMUM CRITICAL POWER RATIO (MCPR),o sufficient margin to these limits exists below 24% RTP; therefore, these requirements are only necessary when operating at or above this power level.

(continued)

HATCH UNIT 2 B 3.3-54 Amendment No. 180

Feedwater and Main Turbine High Water Level Trip Instrumentation B 3.3.2.2 BASES ACTIONS B.1 (continued) not maintained). Therefore, continued operation Is only permitted for a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months period, during which feedwater and main turbine high water level trip capability must be restored. The trip capability Is considered maintained when sufficient channels are OPERABLE or in trip such that the feedwater and main turbine high water level trip logic will generate a trip signal on a valid signal. This requires two channels to each be OPERABLE or In trip. If the required channels cannot be restored to OPERABLE status or placed In trip, Condition C must be entered and Its Required Action taken.

The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time Is sufficient for the operator to take corrective action, and takes Into account the likelihood of an event requiring actuation of feedwater and main turbine high water level trip Instrumentation occurring during this period. It Is also consistent with the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time provided In LCO 3.2.2 for Required Action A.1, since this nstrumentation's purpose Is to preclude a MCPR violation.

0.1 With the required channels not restored to OPERABLE status or placed intrip, THERMAL POWER must be reduced to < 24% RTP I within 4 hours0.167 days 0.0238 weeks 0.00548 months . As discussed In the Applicability section of the Bases, operation below 24% RTP results In sufficient margin to the required I

limits, and the feedwater and main turbine high water level trip Instrumentation Is not required to protect fuel Integrity during the feedwater controller failure, maximum demand event. The allowed Completion Time of 4 hours0.167 days 0.0238 weeks 0.00548 months Is based on operating experience to reduce THERMAL POWER to < 24% RTP from full power conditions I In an orderly manner and without challenging plant systems.

SURVEILLANCE The Surveillances are modified by a Note to Indicate that when a REQUIREMENTS channel Is placed In an inoperable status solely for performance of required Surveillances, entry Into associated Conditions and Required Actions may be delayed for up to 6 hours0.25 days 0.0357 weeks 0.00822 months provided the associated Function maintains feedwater and main turbine high water level trip capability. Upon completion of the Surveillance, or expiration of the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note Is based on the reliability analysis (Ref. 2) assumption of the average time required to perform channel Surveillance. That analysis (continued)

HATCH UNIT 2 B 3.3-56 Amendment No. 180

EOC-RPT Instrumentation B 3.3.4.1 BASES (continued)

APPLICABLE The TSV - Closure and the TCV Fast Closure, Trip Oil SAFETY ANALYSES, Pressure - Low Functions are designed to trip the recirculation LCO, and pumps In the event of a turbine trip or generator load rejection to APPLICABILITY mitigate the Increase Inneutron flux, heat flux, and reactor pressure, and to Increase the margin to the MCPR SL. The analytical methods and assumptions used Inevaluating the turbine trip and generator load rejection are summarized in References 2 and 3.

To mitigate pressurization transient effects, the EOC-RPT must trip the recirculation pumps after initiation of closure movement of either the TSVs or the TCVs. The combined effects of this trip and a scram reduce fuel bundle power more rapidly than a scram alone, resulting In an increased margin to the MCPR SL. Altematively, MCPR limits for an inoperable EOC-RPT, as specified Inthe COLR, are sufficient to prevent violation of the MCPR Safety imit The EOC-RPT function is automatically disabled when turbine first stage pressure Is

< 27.6% RTP.

EOC-RPT instrumentation satisfies Criterion 3 of the NRC Policy Statement (Ref. 6).

The OPERABILTY of the EOC-RPT Is dependent on the OPERABILITY of the Individual Instrumentation channel Functions.

Each Function must have a required number of OPERABLE channels In each trip system, with their setpoints within the specified Allowable

,Value of SR 3.3.4.1.3. The setpoint Is calibrated consistent with applicable setpoint methodology assumptions (nominal trip setpoint).

Channel OPERABILITY also Includes the associated EOC-RPT breakers. Each channel (including the associated EOC-RPT breakers) must also respond within Its assumed response time.

Allowable Values are specified for each EOC-RPT Function specified in the LCO. Nominal trip setpoints are specified In the setpoint calculations. A channel Is inoperable If Its actual trip setpolnt is not within Its required Allowable Value. The nomial setspoints are selected to ensure that the setpoints do not exceed the Allowable Value between successive CHANNEL CALIBRATIONS. Operation with a trip setpoint less conservative than the nominal trip setpoint, but within its Allowable Value, Is acceptable. Each Allowable Value specified is more conservative than the analytical limit assumed In the transient and accident analysis in order to account for instrument uncertainties appropriate to the Function. Trip setpoints are those predetermined values of output at which an action should take place.

The setpoints are compared to the actual process parameter (e.g.,

TSV position), and when the measured output value of the process parameter exceeds the setpoint, the associated device (e.g., trip (continued)

HATCH UNIT 2

B 3.3-76 Amendment No. 180

EOC-RPT Instrumentation B 3.3.4.1 BASES APPLICABLE relay) changes state. The analytic limits are derived from the limiting SAFETY ANALYSES, values of the process parameters obtained from the safety analysis.

LCO, and The Allowable Values are derived from the analytic limits, corrected APPLICABILITY for calibration, process, and some of the nstrument errors. The trip (continued) setpoints are then determined accounting for the remaining instrument errors (e.g., drift). The trip setpoints derived Inthis manner-provide adequate protection because nstrumentation uncertainties, process effects, calibration tolerances, instrument drift, and severe environmental effects (for channels that must function in harsh environments as defined by 10 CFR 50.49) are accounted for.

The specific Applicable Safety Analysis, LCO, and Applicability discussions are listed below on a Function by Function basis.

Altematively, since this nstrumentation protects against a MCPR SL violation, with the instrumentation inoperable, modifications to the MCPR limits (LCO 322) may be applied to allow this LCO to be met.

The MCPR penalty for the EOC-RPT Inoperable condition Is specified Inthe COLR.

Turbine Stop Valve - Closure Closure of the TSVs and a main turbine trip result In the loss of a heat sink and ncreases reactor pressure, neutron flux, and heat flux that must be limited. Therefore, an RPT Is initiated on a TSV - Closure signal before the TSVs are completely closed In anticipation of the effects that would result from closure of these valves. EC-RPT decreases reactor power and aids the reactor scram in ensuring that the MCPR SL Is not exceeded during the worst case transient Closure of the TSVs Is determined by measuring the position of each valve. While there are two separate position switches associated with each stop valve, only the signal from one switch for each TSV Is used, with each of the four channels being assigned to a separate trip channel. The logic for the TSV - Closure Function is such that two or more TSVs must be closed to produce an EOC-RPT. This Function must be enabled at THERMAL POWER 2 27.6% RTP. This is normally I accomplished automatically by pressure switches sensing turbine first stage pressure; therefore, opening of the turbine bypass valves may affect this Function. Four channels of TSV - Closure, with two channels in each trip system, are available and required to be OPERABLE to ensure that no single Instrument failure will preclude an EOC-RPT from this Function on a valid signal. The TSV - Closure Allowable Value Is selected to detect Imminent TSV closure.

(continued)

HATCH UNIT 2 B 3.3-77 Amendment No. 180

EOC-RPT Instrumentation B 3.3.4.1 BASES APPLICABLE Turbine Stop Valve - Closure (continued)

SAFETY ANALYSES, LCO, and This protection Is required, consistent with the safety analysis APPLICABILITY assumptions, whenever THERMAL POWER Is 2 27.6% RTP. Below 27.6% RTP, the Reactor Vessel Steam Dome Pressure - High and the Average Power Range Monitor (APRM) Neutron Flux - High Functions of the Reactor Protection System (RPS) are adequate to maintain the necessary margin to the MCPR Safety Umit.

Turbine Control Valve Fast Closure. TrIo Oil Pressure - Low Fast closure of the TCVs during a generator load rejection results In the loss of a heat sink that produces reactor pressure, neutron flux, and heat flux transients that must be limited. Therefore, an RPT Is Initiated on TCV Fast Closure, Trip Oil Pressure - Low in anticipation of the transients that would result from the closure of these valves.

The EOC-RPT decreases reactor power and aids the reactor scram In ensuring that the MCPR SL Is not exceeded during the worst case transient Fast closure of the TCVs Is determined by measuring the electrohydraulic control fluid pressure at each control valve. There Is one pressure switch associated with each control valve, and the signal from each switch Is assigned to a separate trip channel. The logic for the TCV Fast Closure, Trip Oil Pressure - Low Function Is such that two or more TCVs must be dosed (pressure transmitter trips) to produce an EOC-RPT. This Function must be enabled at THERMAL POWER 2 27.6% RTP. This Is normally accomplished automatically by pressure switches sensing turbine first stage pressure; therefore, opening of the turbine bypass valves may affect this Function. Four channels of TCV Fast Closure, Trip Oil Pressure - Low, with two channels In each trip system, are available and required to be OPERABLE to ensure that no.single Instrument failure will preclude an EOC-RPT from this Function on a valid signal.

The TCV Fast Closure, Trip Oil Pressure - Low Allowable Value Is selected high enough to detect imminent TCV fast closure.

This protection Is required consistent with the safety analysis whenever THERMAL POWER is 2 27.6% RTP. Below 27.6% RTP, the Reactor Vessel Steam Dome Pressure - High and the APRM Neutron Flux - High Functions of the RPS are adequate to maintain the necessary margin to the MCPR SL (continued)

HATCH UNIT 2 B 3.378 Amendment NUo. 180

EOC-RPT Instrumentation B 3.3.4.1 BASES ACTIONS B.1 and B.2 (continued)

Required Actions B.1 and B.2 are intended to ensure that appropriate actions are taken Ifmultiple, noperable, untripped channels within the same Function result Inthe Function not maintaining EOC-RPT trip capability. A Function Is considered to be maintaining EOC-RPT trip capability when sufficient channels are OPERABLE or In trip, such that the EOC-RPT System will generate a trip signal from the given Function on a valid signal and both recirculation pumps can be tripped. Alternately, Required Action B.2 requires the MCPR limit for inoperable EOC-RPT, as specified Inthe COLR, to be applied. This also restores the margin to MCPR assumed In the safety analysis.

The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion ime Is sufficient time for the operator to take corrective action, and takes into account the likelihood of an event requiring actuation of the EOC-RPT instrumentation during this period. It Is also consistent with the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time provided In LCO 3.2.2 for Required Action A.1, since this Instrumentation's purpose s to preclude a MCPR violation.

C.1 and C.2 With any Required Action and associated Completion Time not met, THERMAL POWER must be reduced to < 27.6% RTP within 4 hours0.167 days 0.0238 weeks 0.00548 months . I Alternately, the associated recirculation pump may be removed from service, since this perfonrs the intended function of the Instrumentation. The allowed Completion Time of 4 hours0.167 days 0.0238 weeks 0.00548 months is reasonable, based on operating experience, to reduce THERMAL POWER to < 27.6% RTP from full power conditions In an orderly I manner and without challenging plant systems.

SURVEILLANCE The Surveillances are modified by a Note to Indicate that when a REQUIREMENTS channel Is placed in an Inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours0.25 days 0.0357 weeks 0.00822 months provided the associated Function maintains EOC-RPT trip capability. Upon completion of the Surveillance, or expiration of the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months allowance, the channel must be returned to OPERABLE status or the applicable Condition entered and Required Actions taken. This Note Is based on the reliability analysis (Ref. 4) assumption of the average time required to perform channel Surveillance. That analysis demonstrated that the 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months testing allowance does not significantly reduce the probability that the recirculation pumps will trip when necessary.

(continued)

HATCH UNIT 2 B 3.3-80 Amendment No. 180

EOC-RPT Instrumentation B 3.3.4.1 BASES SUREVILLANCE SR 3.3.4.1.1 REQUIREMENTS (continued) A CHANNEL FUNCTIONAL TEST is performed on each required channel to ensure that the entire channel will perform the intended function. Any setpoint adjustment shall be consistent with the assumptions of the current plant specific setpoint methodology.

The 92 day on an ALTERNATE TEST BASIS Frequency is based on a review of the surveillance test history and Reference 8.

SR 3.3.4.1.2 This SR ensures that an EOC-RPT initiated from the TSV - Closure and TCV Fast Closure, Trip Oil Pressure - Low Functions will not be inadvertently bypassed when THERMAL POWER is 2 27.6% RTP.

This Involves calibration of the bypass channels. Adequate margins for the Instrument setpoint methodologies are incorporated Into the actual setpoint. Because main turbine bypass flow can affect this setpolnt nonconservatively (THERMAL POWER Is derived from first stage pressure) the main turbine bypass valves must remain closed during the calibration at THERMAL POWER ; 27.6% RTP to ensure that the calibration Is valid. If any bypass channel's setpoint Is nonconservative (i.e., the Functions are bypassed at 2 27.6Yo RTP, either due to open main turbine bypass valves or other reasons), the affected TSV - Closure and TCV Fast Closure, Trip Oil Pressure - Low Functions are considered inoperable. Alternatively, the bypass channel can be placed Inthe conservative condition (nonbypass). If placed Inthe nonbypass condition (Turbine Stop Valve - Closure and Turbine Control Valve Fast Closure, Trip Oil Pressure - Low Functions are enabled), this SR Is met with the channel considered OPERABLE.

The 24 month Frequency is based on a review of the surveillance test history, drift of the associated Instrumentation, and Reference 7.

SR 3.3.4.1.3 CHANNEL CALIBRATION is a complete check of the Instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy.

CHANNEL CALIBRATION leaves the channel adjusted to account for instrument drifts between successive calibrations, consistent with the plant specific setpoint methodology. For the TSV - Closure Function, this SR also includes a physical nspection and actuation of the switches.

(continued)

HATCH UNIT 2 B 3.3-81 Amendment No. 180

Primary Containment Isolation Instrumentation B 3.3.6.1 BASES APPLICABLE 1.b. Main Steam Line Pressure - Low SAFETY ANALYSES, LCO, and Low MSL pressure with the reactor at power ndicates that there may APPLICABILITY be a problem with the turbine pressure regulation, which could result (continued) In a low reactor vessel water level condition and the RPV cooling down more than 10r0 F/hr if the pressure loss Is allowed to continue. The Main Steam Line Pressure - Low Function Is directly assumed in the analysis of the pressure regulator failure (Ref. 2). For this event, the closure of the MSIVs ensures that the RPV temperature change limit (l00F/hr) Is not reached. In addition, this Function supports actions to ensure that Safety Umit 2.1.1.1 Is not exceeded. (This Function doses the MSIVs prior to pressure decreasing below 785 psig, which results Ina scram due to MSIV closure, thus reducing reactor power to < 24% RTP.) I The MSL low pressure signals are Initiated from four switches that are connected to the MSL header. The witches are arranged such that, even though physically separated from each other, each switch Is able to detect low MSL pressure. Four channels of Main Steam Une Pressure - 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 was selected to be high enough to prevent excessive RPV depressurization.

The Main Steam Line Pressure - Low Function is only required to be OPERABLE n MODE 1 since this Is when the assumed transient can occur (Ref. 2).

This Function Isolates the Group I valves.

1.c. Main Steam Line Flow - Hi-h Main Steam Une Flow - High Is provided to detect a break of the MSL and to Initiate closure of the MSIVs. If the steam were allowed to continue flowing out of the break, the reactor would depressurize and the core could uncover. If the RPV water level decreases too far, fuel damage could occur. Therefore, the Isolation Is Initiated on high flow to prevent or minimize core damage. The Main Steam Une Flow -

High Function Is directly assumed In the analysis of the main steam line break (MSLB) (Ref. 2). The Isolation action, along with the scram function of the Reactor Protection System (RPS), ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46 and offslte doses do not exceed the 10 CFR 100 limits.

(continued)

HATCH UNIT 2 B .3-141 Amendment No. 180

S/RVs B 3.4.3 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 SIRVs 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 SIRVs are located on the main steam lines between the reactor vessel and the first isolation valve within the drywell. The SIRVs 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 pfot 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 SIRV discharges steam through a discharge line to a point below the minimum water level Inthe suppression pool. The SIRVs 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 (LS)

Valves," and the ADS requirements are specified in LCO 3.5.1, 8ECCS - Operating.

APPLICABLE The overpressure protection system must accommodate the SAFETY ANALYSIS most 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 operate Inthe safety mode. The analysis results demonstrate that the design SIRV capacity is capable of maintaining reactor pressure below the ASME Code limit of 110% of vessel design pressure (1 10% x 1250 psig = 1375 psig). This LCO helps to ensure that the acceptance limit of 1375 psig is met during the Design Basis I

Event.

(continued)

HATCH UNIT 2 B 3.4-1 0 Amendment No. 180

Main Condenser Offgas B 3.7.6 B 3.7 PLANT SYSTEMS B 3.7.6 Main Condenser Offgas BASES BACKGROUND During unit operation, steam from the low pressure turbine Is exhausted directly into the condenser. Air and noncondensable gases are collected in the condenser, then exhausted through the steam jet air ejectors (SJAEs) to the Main Condenser Offgas System. The ofgas from the main condenser normally Includes radioactive gases.

The Main Condenser Offgas System has been incorporated into the unit design to reduce the gaseous radwaste emission. This system uses a catalytic recombiner to recombine radiolytically dissociated hydrogen and oxygen. The gaseous mixture is cooled by the offgas condenser; the water and condensables are stripped out by the offgas condenser and moisture separator. The radioactivity of the remaining gaseous mixture (i.e., the offgas recombiner effluent) is monitored downstream of the moisture separator prior to entering the holdup line.

APPLICABLE The main condenser ofigas gross gamma activity rate is an SAFETY ANALYSES initial condition of the Main Condenser Offgas System failure event, discussed in the FSAR, Sections 11.3 and 15.1.35 (Ref. 1). The analysis assumes a gross failure In the Main Condenser Ofgas System that results in the rupture of the Main Condenser Offgas System pressure boundary. The gross gamma activity rate Is controlled to ensure that, during the event, the calculated offsite doses will be well within the limits of 10 CFR 100 (Ref. 2).

The main condenser offgas limits satisfy Criterion 2 of the NRC Policy Statement (Ref. 3).

LCO To ensure compliance with the assumptions of the Main Condenser Offgas System failure event (Ref. 1), the fission product release rate should be consistent with a noble gas release to the reactor coolant of 100 pCVMWt-second after decay of 30 minutes. This LCO is established consistent with this requirement (2436 MWt x 100 pCVMWt-second = 240 mClsecond). The 240 mCVsecond limit is-conservative for a rated core thermal power of 2804 MWt. I (continued)

HATCH UNIT 2 B 3.7-31 Amendment No. 180

Main Turbine Bypass System B 3.7.7 BASES (continued) 1CO The Main Turbine Bypass System is required to be OPERABLE to limit peak pressure Inthe main steam lines and maintain reactor pressure within acceptable limits during events that cause rapid pressurization, so that the Safety Limit MCPR is not exceeded. With the Main Turbine Bypass System inoperable, modifications to the MCPR limits [LCO 3.2.2, MINIMUM CRITICAL POWER RATIO (MCPR)J may be applied to allow this LCO to be met. The MCPR limit for the inoperable Main Turbine Bypass System is specified Inthe COLR. An OPERABLE Main Turbine Bypass System requires the bypass valves to open Inresponse to Increasing main steam line pressure. This response is within the assumptions of the applicable analysis (Ref. 2).

APPLICABILITY The Main Turbine Bypass System Is required to be OPERABLE at 2 24% RTP to ensure that the fuel cladding Integrity Safety Umit and I the cladding 1%plastic strain limit are not violated during the feedwater controller failure to maximum flow demand transient As discussed Inthe Bases for LCO 3.2.1, AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR),u and LCO 3.2.2, sufficient margin to these limits exists at < 24% RTP. Therefore, these I requirements are only necessary when operating at or above this power level.

ACTIONS A.1 If the Main Turbine Bypass System Is Inoperable (one or more bypass valves inoperable), or the MCPR limits for an inoperable Main Turbine Bypass System, as specified In the COLR, are not applied, the assumptions of the design basis transient analysis may not be met Under such circumstances, prompt action should be taken to restore the Main Turbine Bypass System to OPERABLE status or adjust the MCPR limits accordingly. The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time Is reasonable, based on the time to complete the Required Action and the low probability of an event occurring during this period requiring the Main Turbine Bypass System.

B.1 If the Main Turbine Bypass System cannot be restored to OPERABLE status or the MCPR limits for an Inoperable Main Turbine Bypass System are not applied, THERMAL POWER must be reduced to (continued)

HATCH UNIT 2 B 3.7-35 Amendment No. 180

Main Turbine Bypass System B 3.7.7 BASES ACTIONS B.1 (continued)

24% RTP. As discussed in the Applicability section, operation at c 24% RTP results In sufficient margin to the required limits, and the Main Turbine Bypass System is not required to protect fuel ntegrity during the turbine generator load rejection transient. The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time Is reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without challenging unit systems.

SURVEILLANCE SR 3.7.7.1 REQUIREMENTS Cycling each main turbine bypass valve through one complete cycle of full travel demonstrates that the valves are mechanically OPERABLE and will function when required. The 31 day Frequency is based on engineering judgment, is consistent with the procedural controls governing valve operation, and ensures correct valve positions. Operating experience has shown that these components usually pass the SR when performed at the 31 day Frequency.

Therefore, the Frequency is acceptable from a reliability standpoint SR 3.7.7.2 The Main Turbine Bypass System Is required to actuate automatically to perform its design function. This SR demonstrates that, with the required system Initiation signals, the valves will actuate to their required position. The 24 month Frequency Is based on the need to perform this Surveillance under the conditions that apply during a unit outage and because of the potential for an unplanned transient If the Surveillance were performed with the reactor at power. The 24 month Frequency is based on a review of the surveillance test history and Reference 5.

SR 3.7.7.3 This SR ensures that the TURBINE BYPASS SYSTEM RESPONSE TIME is in compliance with the assumptions of the appropriate safety analysis. The response time limits are specified in Technical Requirements Manual (Ref. 3). The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a unit outage and because of the potential for an unplanned transient If the Surveillance were performed with the reactor at power.

(continued)

HATCH UNIT 2 B 3.7-36 Anendment No. 180

v t e Letter Sequence Acceptance Review
TAC:MB7026, (Approved, Closed) TAC:MB7027, (Approved, Closed)
Initiation Acceptance Administration Withholding Request Acceptance, Withholding Request Acceptance Questions Answers Results Approval Other: ML032240287
MONTHYEAR2003-08-22 [Table View]

ML032691352

Text

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