Text

Amend 21 to License NPF-5,changing Tech Specs to Establish Revised Safety & Operating Limits During Cycle 2 W/Reload 1 Fuel Inserted
	ML19351G318
Person / Time
Site:	Hatch
Issue date:	02/10/1981
From:	Reid R; Office of Nuclear Reactor Regulation
To:	;
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ML19351G319	List: ML19351G318; ML19351G322; ML19351G324;
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	NUDOCS 8102230502
	Download: ML19351G318 (26)
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GEORGIA POWER COMPANY OGLEFORPE POWER CORPORATION MUNICIPAL ELECTRIC _ AUTHORITY OF GEORGIA CITY OF DALTON, GEORGIA DOCKET NO. 50-366 EDWIN I. HATCH NUCLEAR PLANT, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 21 License No. NPF-5 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Georgia Power Company, et al., (the licensee) dated October 17, 1980, as supplemented January 30, 1981, complies with the standards and requirements of the Atomic Energy Act of 1954, as anended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

raio 22 3 0 boV

ATTACHMENT TO LICENSE AMENDMENT NO. 21 FACILITY OPERATING LICENSE NO. NPF-5 DOCKET NO. 50-366 Replace the following pages of the Appendix "A" Technical Specifications with the enclosed pages as indicated. The revised pages are identified by Amendment number and contain vertical lines indicating the area of change. The corresponding overleaf pages are also provided to maintain docunent completeness.

Pages 2-1 B2-1 B2-4 B2-9 3/4 2-1 3/4 2-4A (new) 3/4 2-6 3/4 2-7 3/4 2-7a (new) 3/4 2-7b (new) l B3/4 2-1 f

B3/4 2-3 5-1 4

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2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.-l SAFETY LIMITS THERMAL POWER (Low Pressure or Low Flow) 2.1.1 THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow.

APPLICABILITY:

CONDfTIONS 1 and 2.

ACTION:

With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

THERMAL POWER (High Pressure and High Flow) 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.07 with the reactor vessel steam dome pressure greater than 785 psig

(

and core flow greater than 10% of rated flow.

J APPLICABILITY: CONDITIONS 1 AND 2.

ACTION:

With MCPR less than 1.07 and the reactor vessel steam dome pressure l

greater than 785 psig and core flow greater than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

l I

REACTOR C0OLANT SYSTEM PRESSURE 2.1.3 The reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed 1325 psig.

APPLICABILITY:

CONDITIONS 1, 2, 3 and 4.

ACTION:

l i

With the reactor coolant system pressure, as measured in the reactor vessel steam doice, above 1325 psig, be in at least HOT SHUTDOWN with reactor coolant system pressure < 1325 psig with 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

l l

HATCH-UNIT 2 2-1 Amendment No.

21 r

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i SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.1 SAFETYLIMITS(Continuedl REACTOR VESSEL WATER LEVEL 2.14 The reactor vessel water level shall' be above the top of the-active. irradiated fuel..

A'PPLICABILITY:

CONDITIONS 3,,4 and 5 ACTION:

With the.reactorc vessel-water-level. at or below.the top of the active irradiated 4 fuel, manually initiate the low pressure ECCS to restore the reactor. vessel: water level, after depressurizing the reactor vessel, i f: required..

I h

HATCH,. UNIT 2:

2-2 e

a

l 2.1 SAFETY LIMITS BASES 2.0 The fuel cladding, reactor pressure vessel and primary system piping are the principal barriers to the release of radioactive materials to the environs. Safety Limits are established to protect the integrity of these barriers during normal plant operations and anticipated tran-sients. The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated.

Because fuel damage is not directly observable, a step-back approach is used U establish a Safety li'mit such that the MCPR is not less than 1.07. hw.

> 1.07 represents a conservative margin relative to the conditions required to maintain fuel cladding integrity. The fuel cladding is one of the physical barriers which separate the radioactive materials from the environs.

The integrity of this cladding barrier is related to its relative freedom from perforations or cracking. Although some corrosion or use related cracking may occur during the life of the cladding, fis-sion product migration from this source is incrementally cumulative and continuously measurable.

Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation signifi-cantly above design conditions and the Limiting Sa'fety System Settings.

While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a' threshold beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration. Therefore, the fuel cladding Safety Limit is defined with a margin to the conditions which would produce onset of transition boiling, MCPR of 1.0.

These conditions represent a significant departure from the condition intended by design for planned operation.

2.1.1 THERMAL POWER (Low Pressure or Low Flow)

The use of the GEXL correlation is not valid for all critical power calculations at pressures below 785 psig or core flows less than 10% of rated flow. Therefore, the fuel cladding integrity Safety Limit is established by other means. This is done by establishing a limiting condition on core THERMAL POWER with the following basis. Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flows will always be grqater l

than 4.5 psi. Analyses show that with a bundle flow of 28 x 10J lbs/hr, bundle pressure drop is nearly independent of bundle power and has a l

value of 3.5 psi. Thus, the bundle flow with a 4.5 psi driving head 3

will be greater than 28 x 10 lbs/hr.

Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia 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 of more than 50% of l

RATED THERMAL POWER. Thus, a THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressure blow 785 psig is conservative.

i HATCH - UNIT 2 B 2-1 Amendment No. 21 L

=

SAFETY LIMITS BASES (Continued) 2.1.2 THERMAL POWER (High Pressure and High Flow)

The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated.

Since the parameters which result in fuel damage are not directly observable dur-ing reactor operation, the thermal and hydraulic conditions resulting in a departure from nucleate boiling have been used to mark the beginning of the region where fuel damage could occur. Although it is recognized that a departure from nucleate boiling would not necessarily result in damage to BWR fuel rods, the critical power at which boiling transition -

is calculated to occur has been adopted as a convenient limit.

However, the uncertainties in monitoring the core operating state and in the procedures used to calculate the critical power result in an uncertainty in the value of the critical power.

Therefore, the fuel cladding integrity Safety Limit is defined as the CPR in the limiting fuel assembly for.which more than 99.9% of the fuel rods in the core are expected to avoid boiling transition considering the power distribution within the core and all uncertainties.

Thermal Analysis Basis, GETAB(gtermined using the General Electric The Safety Limit MCPR is

. phich is a statistical model that combines all of the uncertainties in operating parameters and the pro-cedures used to calculate critical power. The probability of the occur-rence of boiling transition is determined using the General Electric Critical Quality (X) Boiling Length (L), GEXL correlation.

The GEXL correlation is valid over the range of conditions used in the tests of the data used to develop the correlation. These conditions are:

Pressure:

800 to 1400 psia 6

2 Mass Flow:

0.1 to 1.2510 lb/hr-ft Inlet Subcooling:

0 to 100 Btu /lb l

Local Peaking:

1.61 at a corner rod to 1.47 at an interior rod l

t l

(a) l General Electric BWR Thermal Analysis Bases (GETAB) Data, l

Correlation and Design application," NED0-10958 and NEDE-10958.

l HATCH - UNIT 2 B 2-2 l

l L

SAFETY LIMITS BASES (Continued) 2.1.2 THERMAL POWER (High Pressure and High Flow) (Continued)

Axial Peaking:

Shape Max / Avg.

Uniform 1.0 Outlet Peaked 1.60 Inlet Peaked 1.60 Double Peak 1.46 and 1.38 Cosine 1.39 Rod Array 64 Rods in an 8 x 8 array The required input to the statistical model are the uncertainties listed in Bases Table B 2.1.2-1, the nominal values of the core para-meters listed in Bases Table B 2.1.2-2, and the relative assembly power distribution shown in Bases Table B 2.1.2-3.

Bases Table B 2.1.2-4 shows the R-factor distributions that are input to the statistical model which is used to establish the Safety Limit MCPR.

The R-factor dis-tributions shown are taken near the beginning of the fuel cycle.

Thebgsfortheuncertaintiesinthecoreparametersaregivenin NED0-20340

, and the s for the uncertainty in the GEXL correlation is given in NED0-10958 The power distribution is based on a typical 764 assembly core in which the rod pattern was arbitrarily chosen to produce a skewed power distribution having the greatest number of i

assemblies at the highest power levels.

The worst distribution in Hatch - Unit 2 during any fuel cycle would not ce as severe as the dis-tribution used in the analysis. The pressure Safety Limits are arbi-trarily selected to be the lowest transient overpressures allowed by the applicable codes, ASME Boiler and Pressure Vessel Code,Section III, and USAS Piping Code, Section B31.1.

(a)

" General Electric BWR Thermal Analysis Bases (GETAB) Data, Correlation and design application," NED0-10958 and NEDE-10958.

(b) General Electric " Process Computer Performance Evaluation Accuracy" NE00-20340 and Admendment 1, NEDO-20340-1 dated June 1974 and December 1974, respectively.

HATCH - UNIT 2 B 2-3

Bases Table B 2.1.2-1 UNCERTAINTIES USED IN THE DETERMINATION OF THE FUEL CLADDING SAFETY LIMIT

Standard Deviation Quanti ty

(% of Point)

Feedwater Flow l.76 Feedwater Temperature.

0.76 Reactor Pressure 0.5 Core Inlet Temperature 0.2 Core Total Flow 2.5 Channel Flow Area 3.0 Friction Factor Multiplier 10.0 Channel Friction Factor Mul tiplier 5.0 TIP Readings 8.7 R Factor 1.6 Critical Power 3.6

The uncertainty analysis used to establish the core wide Safety Limit MCPR is based on the assumption of quadrant power symmetry for the reactor core.

HATCH - UNIT 2 B 2 Amendment No. 21

2.2 LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS The Reactor Protection System Instrumentation Setpoints specified in Table 2.2.1-1 are the values at which the reactor trips are set for each parameter. The Trip Setpoints have been selected to ensure that the reactor core and reactor co.ol, ant system are prevented from exceeding their Safety Limits. Operation with a trip set less conservative than its Trip Setpoint, but within its specified Allowable Value, is acceptable on the basis that each Allowable Value is equal to or less than the drif t allowance assumed for each trip in the safety analyses.

1.

Intermediate Range Monitor, Neutron Flux The IRM system consists of 8 chambers, 4 in each of the reactor trip systems. The IRM is a 5 decade 10 range instrument. The trip setpoint of 120 divisions of scale is active in each of the 10 ranges. Thus, as the IRM is ranged up to accommodate the increase in po'wer level, the trip setpoint is also ranged up. The IRM instruments provide for overlap with both the APRM and SRM systems.

The most significant source of reactivity changes during the power-increase are due to control rod withdrawal.

In order to ensure that the IRM provides the required protection, a range of rod withdrawal accidents have been analyzed, Section 7.5 of the FSAR. The most severe case involves an initial condition in which the reactor is just subcritical and the IRM's are not yet on scale. Additional conservatism was taken in this analysis by assuming the IRM channel closest to the rod being withdrawn is bypassed. The results of this analysis show that the reactor is shutdown and peak power is limited to 1%

of RATED THERMAL POWER, thJs maintaining MCPR above 1.07.

Based on this l

analysis, the IRM provides protection against local control rod errors and continuous withdrawal of control rods in sequence and provides backup protection for the APRM.

l 2.

Average Power Range Manitor For operation at low pressure and low flow during STARTUP, the APRM scram setting of 15/125 divisions of full scale neutron flux provides adequate thermal margin between the setpoint and the Safety Limits.

The margin accommodates the anticipated maneuvers associated with power plant startup.

Effects of increasing pressure at zero or low void content are minor and cold water from sources available during startup 1s not much colder than that already in the system.

Temperature coefficients are small and control rod patterns are constrained by the RSCS and RWM.

Amendment No. X,'21 o

HATCH - UNIT 2 B 2-9 l

2.2 LIMITING SAFETY SYSTEM SETTIN35 BASES (Continued)

REACTOR PROTECTION SfSTEM INSTRUMENTAT10tl SETFOINTS (Continued)

Average Power Range Monitor (Continued)

Of all the possible sources of reactivity input, uniform control. rod witt.-

Be:ause drawal is the most probable cause of significant power increase.

the flux distribution associated with uniform rod withdrawals does n:t involve high local peaks and because several rods must be moved to cnangs -

power by a significant amount, the rate of power rise is very slow.

Gen-In ar.

erally the helt flux is in near equilibrium with the fission rate.

assumed uniform rod withdrawal approach to the trip level, the ra te of power rise is not more than Si of RATED THERMAL POWER per minute anc the APRM system would be more than adequate to assure shutdown before tre power could exceed the Safety Limit.

The 15% neutron flux trip remains active until the mode switch is placed in the Run position.

The APRM flux scram trip in the Run mode consists of a t'ow referentsd simulated thernal power scram setpoint and a fixed neutron flux scra. se;-

The APRM flow ref erenced neutron flux signal is passed throgh a poit!.

filtering network with a time constant which is representative of tre fuel dynamic s.

This provides a flow referenced signal that approximates the average heat flux or thermal power that is deve'.oped in the core during transient or steady-state conditicos.

The APRM flow referenced simulated thermal power scram trip setting at full recirculation flow is adjustable up to 113.5% of RATED THERMAL F0 DER.

This reduced flow referenced trip setpoint will restit in an earlier scram during slow thermal transients, such as the loss of 100 F feedwater heating event, than would result with the 118% fixed neutron flux scram tri;.

Tne lower flow. ef erenced scram setpoint therefore decreases the severi:y, LCPR, of a slow thermal transient and allows lower operating limits if such a transient is the limiting abnormal operational transient during a certain exposure interval in the fuel cycle.

The APRM fixed neutron flux signal does not incorporate the time corstant, but responds directly to instantaneous neutron flux.

This scram se: point scrams the reactor during fast power increase transients if credit is nct taken for a direct (position) scram, and also serves to scram the reactor if credit is not taken for the flow referenced simulated thermal power scram.

The APRM setpoints were selected to provide adequate margin for the Safety Limits and yet allow operating margin that reduces the possibility cf unnecessary shutdown. The flow referenced trip setpoint or APRM pain l

must be adjusted by the spccified fonnula in Specification 3.2.2 in order-I to maintain these margins when the CMFLPD exceeds the FRTP.

HATCH - UNIT 2 B 2-10 Amendment No. 14

' 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) for each type of fuel as,a function of AVERAGE PLANAR EXPOSURE shall not exceed the limits shown in Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, or 3.2.1-4.

l APPLICABILITY: CONDITION 1, when THERMAL POWER > 25% of RATED THERMAL POWER.

ACTION:

With an APLHGR exceeding the limits of Figure 3.2.1-1, 3.2.1-2, 3.2.1-3 or 3.2.1-4, initiate corrective action within 15 minutes and continue corrective action so that APLHGR is within the limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 25% of RATED THE'RMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 4.2.1 All APLHGRs shall be verified to be equal to or less than the applicable limit determined from Figure 3.2.1-1, 3.2.1-2, 3.2.1-3, or 3.2.1-4: l a.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , i

l b.

Whenever THERMAL POWER has been increased by at least 15% of RATED THERMAL POWER and steady state operating conditions have been established, and Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is I

c.

l operating with a LIMITING CONTROL R0D PATTERN for APLHGR.

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l HATCH - UNIT 2 3/4 2-1 Amendment No. 21

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POWER DISTRIBUT10'4 LIMITS 3/4.2.2-APRM SETP0lflT5 Ll!ilTING C0tiDIT10*! FOR OPERATION The APRM flow referenced simulated thennal power scram trip set-3.2.2 point (5) and control rod block trip setpoint (SRB) shall be established

according to the following relationships:

l 5

(0.66W + 51%)

I gg 1 (0.66W + 42;)

s 5 a nd S e a re in percent of RATED THERMAL POWER, and where:

W = Loop' recirculation flow in percent of rated flow.

p APPLICABILITY:

C01;DIT10N 1, when THERMAL POWER 2 25% of RATED THERMAL POUER.

/CTION:

L:ith 5 or S exceeding the allowable valuc. initiate corrective action within 15 minutes and continue corrective action so that 5 and 5 p

are N to l

within the reouired limits

within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduca THERMAL POW less t' an 25'. of RATED THERMAL PCWEO. within the nei.t 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLAt:CE REC'lREMENTS The CMTLT? shall be determined and the APRM flow referenced 4.2.2 simulated ther:na' power scram and control red block trip setpoints or LPRM readings adjusted, as required:

a.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , Whenever THERMAL POWER has been increased by at least 155 of b.

RAT ED THERMAL POWER and steady state operating conditions have been established, and Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor'is l

c.

operatino with a CMFLPD > FRTP.

Pith CORE MAX!M.:M FRACTION OF L]M:ilNG POWER DEt 51TY (CMFLPD) greater E

than the fraction of RATED THERMAL' P0'/ER (FRTP), R up to 95% of RATED THERMAL POWER, rather than adjusting the APRM setpoints, the. APRM gain may be adjusted such that APRM readinos art greater than or ecual to 100i times CHFLPD, provided that 'the adjusted APRM reading does not exceed 100L of RATED THERF%L -POWER and the required gain adjustment increment does not exceed 101 of RATED THERMAL POWER.

Amendment ilo. 14 HATCH - UNIT 2 3/4 2-5 m

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POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO LIMITING CONDITION FOR OPERATION 3.2.3 The MINIMUM CRITICAL POWER RATIO (MCPR), as a function of average scram time and core flow,- shall be equal to or greater than shown in Figure 3.2.3-1 or Figure 3.2.3-2 multiplied by the Kf shown in Figure 3.2.3-3, where:

0 or (Tave - TB), whichever is greater, r =

n B

A=

1.096 sec (notch 36) tion 3.1.3.3 scram time limit to Specifica

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t 4

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t =

g I

i i=1 n

= E N#T*

T

1 ave i=1 n

Nj i=1 number of surveillance tests performed to date in cycle, n =

Nj = number of active control rods measured in the ith surveillance

test, rg = average scram time to notch 36 of all rods measured in the ith surveillance test, and total number of active rods measured in 4.1.3.2.a.

N

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g APPLICABILITY:

CONDITION 1, when THERMAL POWER > 25% RATED THERFML POWER r

ACTION:

,With MCPR less than the applicable.limi t determined from _ Figure. 3.2.3-1 or Figure

'3.2.3-2 initiate corrective action within 15 minutes and continue corrective action so that MCPR is equal to or greater than the apolicable limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERF%L POWER to less than 25% of RATED THERFML POWER within the i

l-next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE RE0VIREMENTS

'4. 2. 3 The MCPR limit at rated flow shall be determined for each type of fuel (8X8R and P8X8R) from Figures 3.2.3-1 and 3.2.3-2 using:

1.0 prior to :he init al scram time measurements for the i

a.

t =

cycle performed in accordance wi th.Specifica tion 4.1.3.2.a. or Amendment No. 21 HATCH-UNIT 2 3/4 2-6

1 3/4.2.3 MINIMUM CRITICAL POWER RATIO SURVEILLANCE REQUIREMENTS (Continued) as defined in Specification 3.2.3; the determination of the b.

Tlimit must be completed within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of the conclusion of each scram time surveillance test required by Specification 1

MCPR shall be determined to be equal to or greater than the applicable limit:

a.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b.

Whenever THERMAL POWER has been increased by at least 15% of RATED THERMAL POWER and steady state operating corditions have been established, and Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is c.

operating with a LIMITING CONTROL R0D PATTERN for MCPR.

1 HATCH-UNIT 2 3/4 2-7 Amendment No. 21 4

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i MCPR LIMIT FOR P8X8R FUEL AT RATED FLOW HATCH. UNIT 2 FIGURE 3.2.3-2 3/4 2-7a Amendment No. 21 l

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N Amendment No. 21 MTCH - IMIT 2 3/4 2-7 b l

l 1

POWER DISTRIBUTION LIMITS 3/4.2.4 LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.4 All LINEAR HEAT GENERATION RATES (LHGRs) shall not exceed 13.4 Kw/ft.

APPLICABILITY:

CONDITION 1, when THERMAL POWER > 25% of RATED THERMAL POWER ACTION:

With the LHGR of any fuel rod exceeding the limit, initiate corrective action within 15 minutes and continue corrective action so that the LHGR is within the limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less then 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 4.2.4 LHGRs shall be determined to be equal to or less than the limit; a.

At least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , b.

When THERMAL POWER has been increased by at least 15% of RATED THERMAL POWER and steady state operating conditions haYe been l

established, and i

c.

Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is i

~

l operating on a LIMITING CONTROL R0D PATTERN for LHGR.

1 I

l l

HATCH - UNIT 2 3/4 2-8 l

l l

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p'3/4.2 POWER DISTRTBUTION LIMfTS

)

BASES The specifications of this section assure that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the 22000F limit specified in the Final Acceptance Criteria (FAC) issued in June 1971 considering the postulated effects of fuel pellet densification.

3/4.2.1 AVERAGE PEAf1AR LINEAR HEAT GENERATION RATE This specification assures that the peak cladding temperature follow-ing the postulated design basis loss-of-coolant accident will not exceed l

the limit specified in 10 CFR 50, Appendix K.

l The peak cladding temperature (PCT) following a postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is dependent only secondarily on the rod to rod power distribution within an assembly. The peak clad temperature is calculated assuming an LHGR for the highest powered rod which is equal to or less than the design LHGR corrected for densification.

This LHGR times 1.02 is used in the heatup code along with the exposure dependent steady state gap conductance and rod-to-rod local peaking factor. The Technical Specification APLHGR is this LHGR of the highest powered rod divided by its local peaking factor.

The limiting value for APLHGR 1s shown in Figures 3.2.1-1, 3.2.1-2, 3.2.1-3 l

I and 3.2.1-4.

The calculational procedure used to establish the APLHGR shown on Figures 3.2.1-1, 3.2.1-2, 3.2.1-3 and 3.2.1-4 is based on a loss-of-coolant l

accident analysis. The analysis was perfonned using General Electric (GE) calculational models which are consistent with the requirements of Appendix K to 10 CFR 50. A complete discussion of each code employed in the analysis is presented in Reference 1.

Differences in this analysis compared to previous analyses performed with Reference 1 are:

(1) the analysis assumes a fuel assembly planar power consistent with 102% of the MAPLHGR shown in Figures 3.2.1-1, 3.2.1-2, 3.2.1-3 and 3.2.1-4; (2) fission product decay is computed l

assuming an energy release rate of 200 MEV/ fission; (3) pool boiling is assumed after nucleate boiling is lost during the flow stagnation period; and (4) the effects of core spray entrainment and counter-current flow limitation as described in Reference 2, are included in the reflooding calculations.

I A list of the significant plant input parameters to the loss-of-coolant accident analysis is presented in Bases Table B 3.2.1-1.

HATCH - UNIT 2 B 3/4 2 1 Amendment No. 21 1

Bases Table B 3.2.1-1 SIGNIFICANT INPUT PARAMETERS TO THE LOSS-0F-COOLANT ACCIDENT ANALYSIS 1

FOR HATCH-UNIT 2 Plant Parameters:

Core Thernal Power............... 2531 Mwt which corresponds to 105% of license core power

6 Vessel Steam Output..............

10.96 x 10 lbm/h which corresponds t) 105% of rated steam flow Vessel Steam Dome Pressure.......

1055 psia Design Basis Recirculation Line Break Area For:

2 a.

Large Breaks............ 4.0, 2.4, 2.0, 2.1 and 1.0 f t 2

b.

Small Breaks............ 1.0, 0.9, 0.4 and 0.07 ft Fuel Parameters:

PEAK TECHNICAL INITIAL SPECIFICATION DESIGN MINIMUM LINEAR HEAT AXIAL CRITICAL FUEL BUNDLE GENERATION RATE PEAKING POWER FUEL TYPE GE0 METRY (kw/ft)

FACTOR RATIO Initial Core 8x8 13.4 1.4 1.18 A more detailed list of input to each model and its source is presented in Section II of Reference 1 and subsection 6.3.3 of the FSAR.

The core

This power level meets the Appendix K requirement of 102%.

heatup calculation assumes a bundle power consistent with operation of the highest powered rod at 102% of its Technical Specification linear heat generation rate limit.

4 HATCH - UNIT 2 B 3/4 2-2

POWER DISTRIBUTION LIf1ITS BASES 3/4.2.2 APRM SETPOINTS The fuel cladding integrity Safety Limits of Specification 2.1 were based on a power distribution which would yield the design LHGR at RATED THERMAL POWER.

The scram setting and rod block functions of the APRM instru-ments or APRM readings cust be adjusted to ensure that the MCPR does not become less than 1.0 in the degraded situation. The scram settings and rod j

block settings or APRM readings are adjusted in accordance with the formula in this specification when the combination of THERMAL POWER and CMFLPD indicates a higher peaked power distribution to ensure that an LHGR transient would not be increased in the degraded condition.

i 3/4.2.3 MINIMUM CRITICAL POWER RATIO The required oper?.ing' limit MCPRs at steady state operating conditions as specified in Specification 3.2.3 are derived from the established fuel cladding integrity Safety Limit MCPR of 1.07, and an analysis of abnornal l

operational transients.

For any abnormal operating transient analysis evaluation with the initial condition of t.le reactor being at the steady state operating limit, it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrument trip setting as given in Specification 2.2.1.

To assure that the fuel cladding integrity Safety Limits are not exceeded during any anticipated abnormal operational transient, the most limiting transients have been analyzed to determine which results in the largest reduction in CRITICAL POWER RATIO (CPR). The type of transientr. evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease.

The limiting transient which determines the required steady state MCPR limit is the load rejection trip with failure of the turbine bypass.

This transient yields the largest A CPR. When added to the Safety Limit MCPR of 1.07 the required minimum operating limit MCPR of Specification l

3.2.3 is obtained.

i HATCH - UNIT 2 B 3/4 2-3 Amendment No J4, 21 3

+

l

POMER DISTRIBUT10'! LIMITS l

BASES MINIMJM CRITICAL POWER RATIO (Continued)

The evaluation 'of a given transient begins with the system initial paraireters shown in FSAR Table 15.1-6 that are input to a GE-core dynarric behavior transient computer program described in NE

{

Also, the void reactivity coefficients that were input to the transient calculational procedure are based on a new method of calculation temed NEY which provides a better agreement between the calculated and plant The outputs of this program along with instrument power distributions.

the initial MCPR form the input for further analyses of the themally l

limiting bundle with the sing channel transient thermal hydraulic SCAT code described in NED0-20566 The principal result of this evalt.ation l

is the reductier. in MCPR caused by the transient.

The purpcse of the X, factor is to define operating limits at cther At less than 100% of rated flow the re:Jired than rated flow conditioni.

Specif-MCPR is the product of the operating limit MCPR and the K, factor.

factor provides the required thermal margih to protect ically, the K The most limiting transient ini-iated agair.st a flow increase transient.from less than rated flow condit caused by a motor-generator speed control failure.

factors For operation in the automatic flow control mode, the Kf

" assure that the operating limit MCPR of Specification 3.2.3 will no-te flow.-

violated should the cost limiting transient occur at less than rate:

factors assure that the Safety In the manual flow control mode, the Kf tirit MCFR vill not be violated should the most limiting transient cccur at less than rated flow.

factor values shown in Figure 3.2.3-1 were developed ger.erically The K The K and are applicable to all BWR/2, BWR/3 and BWR/4 reactors.

f f

facters were derived using the flow control line corresponding to R*TED THERPAL POWER at rated core flow.

For the manual flow control mode, the 'K factors were calculated su:h that for the maximum flow rate, as limited by the pump scoop tube set point, and the corresponding THERMAL POWER along the rated flow control-lire, the limiting bundle's relative power, was adjusted 'until the MCPR was s'.ightly Using this relativa bundle power, the MCPRt were above the Safety Limit.

calculated at different points along the rated flow control line correspond-ing to different core flows..The ratio of the MCPR calculated at a given point of core flow, divided by the operating limit MCPR.' detemines the K.

f 8 3/4,2-4 HATCH - UNIT 2

l

,a 5.0 DESIGN FEATURES 5.1 SITE EXCLUSION AREA 5.1.1 The exclusion area shall be as shown in Figure 5.1.1-1.

LOW POPULATION ZONE 5.1.2 The low population zone coincides with the exclusion area and is also shown in Figure 5.1.1-1.

5.2 CONTAINMENT CONFIGURATION 5.2.1 The primary containment is a steel structure composed of a series I of vertical right cylinders and truncated cones which form a drywell.

This drywell is attached to a suppression chamber through a series of vents. The suppression chamber is a steel pressure vessel in the shape of a torus. The primary containment has a total minimum free air volume of 255,978 cubic feet.

DESIGN TEMPERATURE AND PRESSURE 5.2.2 The primary containment is designed and shall be maintained for:

a.

Maximum design internal pressure 56 psig, b.

Maximum allowable internal pressure 62 psig.

c.

Maximum internal temperature 340 F.

d.

Maximum external pressure 2 psig.

5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The rea: tor core shall contain 560 fuel assemblies with each fuel assembly containing 62 fuel rods and 2 water rods clad with Zircaloy -4.

Each fuel rod shall have a nominal active fuel length of 150 inches and contain a maximum total weight percent of 3341 grams uranium. The initial core loading shall have a maximum average enrichment of 1.87 weight percent U-235.

Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum average enrichment of 2.90 weight percent U-235.

l HATCH - UNIT 2 5-1 Amendment No. 21 m

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EXCLUSION AREA AND LOW POPULATION ZONE FIGURE 5.1.1-1 l_

HATCH.. UNIT 2 5-2 c

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