ML20126E838

Amend 18 to License NPF-2,changing New Rod Bow Penalty Curve & Bases.Also Revises Heatup & Cooldown Curves & Bases

ML20126E838
Person / Time
Site:	Farley
Issue date:	02/13/1981
From:	Varga S Office of Nuclear Reactor Regulation
To:
Shared Package
ML20126E841	List: ML20126E838 ML20126E844 ML20126E848
References
NUDOCS 8103040067
Download: ML20126E838 (21)

Text

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ALABAMA POWER COMPANY DOCKET NO. 50-348 r

' JOSEPH M. FARLEY NUCLEAR PLANT, UNIT NO.1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No '18 License No. NPF-2 1.

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

A.

The applications for amendment by Alabama Power Company (the licensee) dated March 28,1980 (supplemented by letter dated January 5,1981) and January 9,1981 (supplemented by letter dated January 23, 1981), comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set

. orth in 10 CFR Chapter I; E.

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

There is reasonable assurance (1).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 nealth and safety of the public; 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'.

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Accordingly, the license is amended by changes to the Technical Specifications as indicated in thi a_tt'achment to this license amendment, and paragraph 2.C.(2) of Fa.cility Operating License No. NPF-2 is hereby amended to read as follows:

(2) Technical Specificatdons The Technical Specifications ' contained in Appendices A and B, as revised through Amendment No.18, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

,3. This license amendmeht is effective as of the date of its issuance.

THE CLC R REGUALTORY COMMISSION teven arga! h af f

' Operating Reactd s Branch #1 t

Division of Licen ng A :a :r:sr.::

Cr.ar ges :: the Technical 5:e :i fi cations,,

• suance: February 13, 1981

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ATTACHMENT TO LICENSE AMENDMENT AMENDMENT NO.18 TO FACILITY OPERATING LICENSE NO. NPF-1 DOCKET NO.30-348 Revise Appendix A as fpilows:

Remove Pager.

Insert Pages XI XI 3/4 2-10a 3/4 2-10a 3/4 4-25 3/4 4-25 3/4 4-26 3/4 4-26 3/4 4-27 3/4 4-27 3/4 4-28 3/4 4-28 B 2-2 B 2-2 B 3/4 2 5 B 3/4 2-5 B'3/4 4-6 B 3/4 4-6 B 3/4 4-7 8 3/4 4-7 8 3/4 4-8 B 3/4 4-8 B 3/4 4-9 8 3/4 4-9 B 3/4 4 B 3/4 4-10 B 3/4 4-11 B 3/4 4-11 B93/4 4-12 B 3/4 4-13 B 3/4 4-14' B 3/4 4-6a h

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INDEX l

+

BASES 4

SECTION PAGE i

3/4.3 INSTRUMENTATION 3/4.3.1 PROTECTIVE INSTRUMENTATION..............................

B 3/4 3-1 3/4.3.2 ENGINEERED SAFETY FEATURE INSTRUMENTATION...............

B 3/4 3-1 3/4.3.3 MONITORING INSTRUMENTATION..............................

B 3/4_3-1 4

1 3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 REACTOR COOLANT L00PS...................................

B 3/4 4-1 3/4.4.2 a n d 3/4. 4. 3 S AFETY '.'ALVES............................... B 3/4 4-1 3/4.4.4 PRESSURIZER.............................................

B 3/4 4-2 t

3/4,4.5 S TE AM G EN E RATO RS........................................ B 3/4 4-2 I

i 3/4.4.6 REACTOR COOLANT SYSTEM LEAKAGE.......................... B 3/4 4-4

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

B 3/4 4-5 3/4.4.8 SPECIFIC ACTIVITY.......................................

B 3/4 4-5 i

3/4.4.9 PRESSURE / TEMPERATURE LIMITS.............................

B 3/4 4-6 3/4.4.10 STRUCTURAL INTEGRITY...................................

B 3/4 4-14 FARLEY - UNIT 1 XI Amendment No.18

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REACTOR'C00LANT SYSTEM 3/4.4. 9 PRESSURE / TEMPERATURE LIMITS l

-REACTOR' COOLANT SYSTEM r

-LIMITING CONDITION FOR 0PERATION

3. 4. 9.1 The Reactor Coolant System'(except the pre'ssurizer) temperature ind pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2 and 3.4-3 during heatup, cooldown criticality, and inservice.

leak and hydrostatic testing with:

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

a.

b.

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

l c.

A maximum temperature change of.less than or equal to 10*F in any.

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one hour period during inservice hydrostatic and. leak testing operations above the heatup and cooldown limit curves.

APPLICABILITY:

At all times.

ACTION:

With any of the above limits exceeded, restore the temperature and/or pressure

.i to within the limit within 30 minutes; perform an engineering evaluation-or inspection to determine the effects,of the out-of-limit condition on the fracture toughness of the Reactor Pressure Vessel; determine that the Reactor.

• Pressure Vessel remains acceptable for. continued operation or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce the RCS T and pressure to less than 200*F and 500 psig, respectively, within the f8Tiowing 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.2.1,1 The Reactor Coilant System temperature and pressure shall be determined to be within the limits at least once per hour during system heatup, cooldown, and inservice leak and. hydrostatic testing operatinns.

4.4.9.1.2 The reactor vessel material irradiation surveillance specimens shall be removed aid examined, to determine changes in material properties, as required by 10 CFR 50, Appendix H in accordance with the schedule in Table 4.4-5.

The results of these examinations shall be used to update Figures,3.4-2 and 3.4-3.

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Amendment No.18 FARLEY-UNIT'l 3/4 4-25 j

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MATERIAL PROPERTY BASIS CONTROLLING MATERIAL: WE LD META (,

COPPER CONTENT: 0.24 WT%

j PHOSPHORUS CONTENT: 0.011 WT%

j 0

0F RTNDT_INITI AL:

AFTER 7 EFPY: 1/47,1850F RTNOT 3/4T. 1180F CURVE APPLICABLE FOR HEATUP RATES 0

UP TO 60 F/HR FOR THE SERVICE PERIOD UP TO 7 EFPY AND CONTAIlls ACCEPTABLE 3000 MARGINS OF 10 F AND 60 PSIG FOR REGION FOR 0

POSSIBLE INSTRUMENT ERRORS HYOROSTATIC TESTING OPERATIONS 3f LEAK TEST LIMIT w

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h 2000 UNACCEPTABLE c

OPERATION 8

4 HEATUP RATES ACCEPTABLE 0

E UP TO 60 F/HR OPERATION

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CRITICALITY LIMIT p

P BASED ON INSERVICE HYDROSTATIC TEST TEMPERATURE 0

(325 F) FOR THE SERVICE PERIOD UP TO 7 EFPY 0

0 100 200 300 400 500 INDICATED TEMPERATURE (OF)

Ficure 3,4-2 Farley Unit I. Reactor Coolant System Heatup Limitations Applicable For The First 7 EFPY FARLEY-UNIT 1 3/4 4-26 Amendment No.18 mm m

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MATERIAL PROPERTY GASIS CONTROLL;NG MATERIAL: WELD METAL COPPER CONTENT: 0.24 WT%

PHOSPHORUS CONTENT: 0.011 WT%

INITIAL: 0 F RTNOT NDT AFTER.7 EFPY: 1/4T,1850F RT 3/4T, 118of CURVE APPLICABLE FOR COOLDOWN 0

RATES UP TO 100 F/HR FOR THE SERVICE PERIOD UP TO' 7 EFPY AND CONTAINS 0

MARGINS OF 10 F AND 60 PSIG FOR POSSIBLE INSTRUMENT ERRORS 2000 o

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OPERATION COOLOOWN RATES OF/HR O N 20 --

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0 100 200 300 400 INDICATED TEMPERATURE (*F)

Ficure 3.4-3Farley Unit 1 Reactor Coolant System Csoldown Limitations Applicable For The First 7 EFPY FARLEY-UNIT 1 3/4 4-27 Amendment No.18 goo yc yg a

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U TABLE 4.4-5 d;

REACTOR VESSEL MTERIAL SURVEILLANCE PROGRAM-WITilDRAWAL SCllEDULE a

a VESSEL LEAD CAPSUL _E LOCATION FACTOR

'WITilDRAWAL TIME Y

3430 3.5 1st. Refueling Outage U

1070 3.5 3 EFPY.

X 2870 3.5 6 EFPY W

1100 2.9 11 EFPY M

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V 2900 2.9 20 EFPY 2

Z 3400 2.9 STBY r

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SAFETY LIMITS SASES N

and a ref erence cosine with a peak of 1.55 for axial power shak., cf 1.55 Tne curves are based on an enthalpy hot channel factor, F An -

g at reduced power based on allowance is included for an increase in FaH the expression:

?".. c 1.55 [1 + 0.2 (1-P)]

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.Where P is the fraction o.f RATED T, ERMAL POWER H

Tnese limitino heat flux conditions are ' higher than those calculated fer the range of all control rods. fully withdrawn to the maximum allowable

r.. :1 rod insertion assuming the axial powe'r imbalance is within the iirits Of.the f (61) function of the Overtemperature trip. When the

we* imbdiance is not within the tolerance, the ax:al power I

t?t

. :::an:e eect on the Overtemperature tT trips will redu:e the setpoints to :r n=e prote: tion consistent witn core safety limits.

No p 2 *.- EECTOR. COOLANT SYSTEM :RE55URE i:

e restriction of nis Safety. Limit prote::s the integrity of the

.51:::- 0 :ian: System from overpressuri ntion ah: thereby prevents th'e s'sats :' radionuclides :or.taine: in the rea:: r :::iant from reaching o

g : e ::

ainment a.m:s:here.

Tne reactor pressure vessel pressurizer and -he rea: tor coolant system piping and fittings.are designed to Section III of the ASME Code

.f:w Nuclear Power Plant @ich. permits.a maxim ^um transient pressure of 11 3 (2735 iisig) of design pressure.

' e Safety Limit of 2735 psig :s therefore c:nsisten with the

esi; criteria and associhted code requirements.

The entire Reactor Coolar.t System is hydrotested at 3107 psig,125%

l c' design pressure, to demons; rate integrity prier to* initial operation.

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FARLEY - UNIT 1 B 2 -2 Amendment.No. 5 8

PO',.TR DISTRIBUf10N LIMITS a

EASES.

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Abnormal perturbations'injhe radial power shame, such as.from a.

rod misalignment, effect rg more directly than Fp b.

Although rod movemen't has a direct influence upon limiting F to within its limit, such control is not readily available to limitFh,and I

c.

. Errors in'predi.ption for. control power shape ' detected during

'startup physics tests can -b'e compensated 'for in, F by g' estrict ing axial flux distributions. This compensation Eor F" is

less readily available.

Fuel rod bowing reduces the value of DNB ratio. Sufficent credit is available to offset this reduction. This credit comes from generic design margins totaling 9.1% and 3% margin in the difference between the 1.3 DNBR safety limit and the minimum DNBR -

e calculated for the Complete Loss of Flow event. The penalties applied ^to Fyto account for Rod Bow (Figure 3.2-3) as a function of burnup are consistent with those described in Mr.

John F. Stolz's (NRC) letter to T. M. Anderson (Westinghouse) dated April 5,1979, and WCAP-8691, Rev.1 (partial rad bow test data).

' The rrdial peaking factor, Fxy (:), is measured periodically to provide additional assurance that the hot channel, factor, F0 (2),' remains within' its -Ti=it.

The Fxy (c) limits were deter =ined from expected power control' maneuvers over the full range of burnup conditions in the core.

3/4.2.4 OUADRANT p0WER TILT RATIO The quardrant power til' ratic limit assures that the radial power distribution satisfies the design values used in the power capability analysis.

Radial power distribution measurements are made during start-up testing and periodically during power operation.

The limit of,1,02 at which corrective action 'is required provides '

DNE and linear heat generation rate protecticn with x-y plane power tilts.

A limiting tilt of 1.025 can be tolerated before the margin for uncertainty in Fg.is depleted.

The limit of 1.02 was selected to provide an allowance for the uncertainty associated with the. indicated power tilt.

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.The two hour time allowance for operation with a tilt condition

.creater than 1.02 but less than 1.09 is provided to allow identification 2nd correction of'a dropped or misaligned rod.

In the event such action does not correct the tilt, the margin for un:ertainty on F is reinstated by reducing the power by 3 percent for each percent of ti19 in ext.ess of 1.0.

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,R,EACTOR CCCLANT SYSTEM BASES i

I The ACTION statement permitting POWER OPERATION to continue for limited time periods with the primary coolant's specific activity > 1.0 pCi/ gram DOSE EQUIVALENT I-131, but within the allowable limit shown on Figure. 3.4-1, acconrnodates possible iodine spiking phenomenon which may occur following. changes in THERMAL POWER.

Operation with specific activity levels exceeding-1.0 uC1/ gram DOSE EQUIVALENT I-131 but within the limits shown on Figure 3.4-1 must be' restricted to no more than 10 percent of the unit's yearly operating time since the activity levels i

allowed by Figure 3.4-1 increase the 2' hour, thyroid dose at the site j

boundary by'a factor of up to 20 following a postulated steam generator j

tube rupture.

Reducing T to <500*F prevents the release of activity should a a

steam generator dbe rupture since the saturation pressure of the primary coolant is below the lift pressure of the atmospheric steam relief valves.

The surveillance requirements provide adequate assurance that excessive specific activity levels in the primary coolant will be detected in suf#icien time to take corrective action.

Infomation obtained on' i

i: cine s:iking will be used to assess the parameters associated with spiking phenomena. A reduction in frequency of isotopic analyses folloviing power changes may be permissible if justified by the data obtainec.

j 3/4.4.9 PRE 55URE/TE!4PERATUF.E LIMITS i

The tamperature and pressure changes during heatup and c:oldown are limited to be consistent with the requirements given in the ASME Boiler and l

Pressure Vessel Code,Section III, Appendix G.

3 1)

The reactor coolant temperature and pressure and system heatup and cooldown rates (with the exception of the pressuri:er) shall be limited in-accordance with Figures 3.4-2 and 3.4-3.

a)

Allowable combinations of pressure and temperature for specific 1

temperature. change rates are below and to the right of 4'he limit lines shown.

Limit lines for cooldown rates between those presented may be obtained by interpolation.

^

b)

Figures 3.4-2 and 3.4-3 define limits to assure prevention of non'-

ductile failure only.

For normal operation,' other inherent plant characteristics, e.g., pump heat addition and pressuriter heater capacity, may limit the heatup and cooldown rates that can be achieved over certain pressure-temperature ranges.

i FARLEY-UNIT 1 B 3/4 4-6 Amendment No. 18

... ~.. -.

2)

These limit lines shall be ' calculated periodically using methods provided belew.

i 3)

The secondary side of the steam geherator must not be pressuri:ed above 200 psig if the temperature of the steam generator is below 70*F.

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"A 'JV.UfGT 1 B '3/4 4-6A Amendment No. 18

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BASES 4)

The pressurizer heatup and cooldown rates shall not exceed 100*F/ hr and 200*/hr respectively.

The spray shall not be used if the temperature difference between.the pressuri:er and the spray fluid is greater than 320*F.

5)

System and in-service leak and-hydrotests shall be performed at pressures in accordance with the requirements of ASl'.E Boiler and Pressure Vessel Code, S'ection XI.

The fracture toughness preperties of the ferritic =aterials in the reactor vessel are determined in accordance with ASTM E185-73, and in a::ordan:e with additional reactor vessel requirecents.

These properties are tnen evaluated in accordance witn Appendix G of the 1975 Summer Adder lca to Section III of the ASME Sciler and Pressure Vessel C de.

Heatup and cooldown limit curves are calculated using the most limiting value.of the nil-ductility reference temperature, RT.

, at the end of hva 7

effective full power years of service life. The 7 EFPY service life period is :hosen such that the limiting Ri at the 1/47 location gg7 in tne core region is greater than the RT of the limiting unirraciated gg7 caterial.

The selection of such a limiting RT assures that all NDT c:epenents in the Reactor C:olant System will be cperated ::nservatively in a::ordance with applicable Code requirements.

The reactor vessel materials have been tested to determine their initiai RTNDT; the results of these tests are shown in Table B 3/4.4-1.

Reactor i

]

cperation and resultant fast neutron (E greater than 1 MEV) irradiation

'can cause an increase in the RT Therefore, an adjusted reference NDT.

tch;erature, based upon the fluence and copper content of the material in cuestion, can be predicted using Figure B 3/4.4-1 and the recommendations of Regulatory Guide.1.99, Revision 1, " Effects of Residual Elements on Predicted Radiation Damage to Reactor Vessel Materials."

The heatup and cooldown limit curves of Figures 3.4-2 and 3.4-3 include predicted adjustments for this shift in RT at the end of 7 EFPY (as well as NDT adjustments for possible errors in the pressure and temperature

' sensing instruments).

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FAWLEY-UNIT 1 B 3/4 4-7 Amendment No.18

.REA:~:? ::'. ANT SYSTEM l

SA5E5 Values of ART determined in this manner may be used until the results ND7 fr:: the caterial surveillance program, evaluated according to ASTM E185,

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are available.

Capsules will be recoved in accordance with the req'uire-rents of ASTM E185-73 and 10 CFR 50, Appendix H.

The surveillance spe:iren withdrawal schedule is shown in Table 4.4-5.

The heatup and determined from the coolcewn curves must be recalculated when the ARTNOT f r the equivalent' survet11ance capsule exceeds the calculated ARTNDT caosule radiation exposure.

Allewable press,ure -temperature relationships for various heatup and cold:wn rates are calculated using methods derived frem Appendix G in Se:.icn III of the ASME Soiler and Pressure Vessel Code as required by A::endix G to 10 CFR Part 50 and theue metnods are discussed in detail in

P e fe.11: wing paragraphs, The general method for calculating heatup and cecidown limit curves is tissd up:n the principles f the li.. ear elastic fracture mechani:s (LEFM) ts:5. 1:;y.

In the calculation procedures a semi-elliptical surface dific. with a depth of one quarte.r f the wall thickness, T, and a length f 2/27 is assumed to exist a-the inside of the vessel wall as well as E-the Outside of the vessel wall.

The dimensions of this postulated c a:k, referred to in Appendix G of A5ME Section III as the rderence flaw, a piy exceed the current capabilities of inservice inspection techniques.

Therefore, the reactor operation linit curves developed for this reference cra:k are conservative and provide sufficient safety margins for protection a;a/nstnon-ductilefailure.

To assure that the radiation embrittlement effee.s are accounted for in the calculation of the limit curves, the

s. iimiting value of the nil ductility reference temperature, RTNDT, is used and this includes the radiation induced shift,.aRTNDT, e cresponding t; tha end of the period for which heatup and cooldown curves are generated t

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FARI.EY-ldIT I B 3/4 4-9 Amendment No. 18' 4

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Figure 3 3/4.4-1 Fast-Neutron Fluence (E > 1 Mev) as a Function of Full-Power Service Life FAP1EY-Vi;IT 1 B 3/4 4-10 Amendment No. 18

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t REACTOR COOLANT SYSTEM BASES The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, K, for the combined thermal _and pressure stresses'at any time during'heatup g

or cooldown cannot be greater than the reference stress intensity factor, KIR' for the metal temperature at that time.

K is obtained frem_the reference IR fracture toughness curve, defined in Appendix G to the ASME Code.

The KIR curve is given by the equation:

Kyg = 26.78 + 1.223 exp [0.0145(T-RTNOT + 160)]

0) where K is the ref.erence stress intensity factor as a function of the metal IR temperature T and the metal nil ductility reference temperature RT

Thus, NOT.

the governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME Code as follows:

t-CK73 + kit IR (2)

1. K Where, K,4 is the stress intensity factor caused by membrane (pressure) g stress.

kit.is the stress intensity factor caused by the thermal gradients.

K is provided by the code as a function of temperature relative IR l

to the RT f the material.

HDT C = 2.0 for level A and B service limits, and C = 1.5 for inservice hydrostatic and leak test operations.

At any time during the heautp or cooldown transient, K is determined by yg the metal temperature at the tip of the postulated flaw, the' appropriate value for RTNDT, and the reference fracture toug.hness curve.

The thermal stresses resulting from temperature gradients through the vessel wall are FARLEY-UNIT 1 B 3/4_4-11 Amendment No. 18 I

L.

REACTOR COOLANT SYSTEM BASES l

calculated and then the corresponding thermal stress intensity factor, KIT' l

for the reference flaw is computed.

From Equation (2) the pressure stress l

intensity factors are obtained and from these, the allowable pressures are calculated.

COOLDCWN For the calculation of the allowable pressure versus coolant temperature during cooldown, the Code reference flaw is assumed to exist at the inside of the vessel wall. During cooldown, the controlling location of the flaw is always at the inside of the wall because the thermal gradients produce tensile stresses at the inside, which increase with increasing cooldown rates.

Allowable pressure-teeperature relations are generated for both steady-state and finite ccoldown rate situations.

From these relations composite limit curves are constructed fer each cooldown rate of interest.

The use of the composite curve in the cooldown analysis is necessary be-cause control of the comidown procedure is based on measurement of reactor coolant temperature, whereas the limiting pressure is actually dependent on the material temperature at the tip of the assumed flaw.

During cooldown, the 1/4T vessel location is at a higher temperature than the fluid adjacent to the vessel ID.

This condition, of course, is not true for the steady-state situation.

It follows that at any given reactor coolant temperature, the delta T developed during cooldown results in a higher value of K at the 1/4T location for yg finite cooldown rates than for steady-state operation.

Furthermore, if conditions exist such that the increase in K exceeds kit, the calculated IR allowable pressure during cooldown will be greater than the steady-state value.

The above procedures are needed because there is no direct control on temperature at the 1/4T location; therefore, allowable pressures may unknowingly be violated if the rate of cooling is decreased at various intervals along a cooldown ramp.

The use of the composite curve eliminates this problem and assures conservative operation of the system for the entire cooldown period.

FARLEY-UNIT 1 B 3/4 4-12 Amendment No. 18

l REACTOR COOLANT-SYSTEM BASES HEATUp

]

Three separate calculations are required to determine the limit curves I

for. finite heatup rates.

As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as f'inite heatup rate conditions assuming'the presence of a 1/4T defect at the inside of the vessel wall.

The thermal gradients during heatup produce compressive stresses at the-inside of the wall that alleviate the tensile st esses produced by internal pressure.

The metal tamperature at the crack tip lags the coolant temperature; therefore, the K f r the 1/47 crack IR during heatup is lower than the K for the 1/4T crack during steady-state yg conditions at the same coolant temperature.

During heatup, especia11; at the end of the transient, conditions may exist such that the effects of compressive thermal stresses and different K

's for steady-state and finite heatup ratas IR do not offset each other and the pressure-tamperature curve based on steady-stata conditions no longer represents a icwer bound of'all similar curves for-finite heatup rates when the 1/47 flaw is c:nsidered.

Therefore, both cases

. have tc be analy:ed in order to assure that at any coolant temperature the lower value of the allowabie pressure calculated for steady-state and finite heatup rates is obtained.

The second portion of the heatup analysis conceins' the calculation of

' pressure-temperature limitations for the case in which a 1/4T deep outside surface flaw is assumed.

Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce stresses which are tensile in nature and thus tend to reinforce any pressure stresses *present.

These thermal stresses, of course, are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp.

Furthermore, since the thermal stresses, at the outside are tensile and increase with increasing heatup rate, 'a lower bound curve cannot bc defined.

Rather, each heatup rate of interest must be analyzed on'an individual basis.

Following the generation of pressure-temperature curves for both the steady state and finite heatup rate situations, the final limit curves are produced as fo11cvs.

A composite curve is constructed based on a point-by-point comp:rison of the steady-state and finite heatup rate data.

At any given tomoerature, the allowable pressure is taken to be the lesser of the thres values taken from the curves under consideration.

FARLEY-UNIT 1-B 3/4 4-13 0 * * ]D D

I

-_-__ __whB2U

REACTOR CCOLANT SYSTEM BASES The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist such that over the course cf the heatus ramp the contro11ing' condition switches from the inside to the cutside' and the pressure limit =ust at all times be, based:on analysis of'the most critical criterion.

Finally, the composite curves for the heatup rate data and the cooldewn rate data are adjusted for possible errors in the pressure and temperature.

sensing instruments by the values indicated on the respective curves.

Although the pressuri:er operates in temperature ranges above those for.

which there is reason for concern of non-ductile failure, operating limits are provided to assure ccmpatibility of operation with the fatigue analysis performed in acc:rdance with the ASME Code requirements.

f The OPERASILITY of two RHR relief valves or an RCS vent opening of greater

^

than or equal to 2.85 square inches ensures that the RCS will be protected frc: pressure transients which could exceed the li=its of Appendix G to 10 CFR part 50 wh'en ene er c:re of the RCS cold legs are less than or equal to 310*F.

Either RHR relief valve has adequate relieving ca: ability to protect the RCS fr cver:ressuri:ation when the transient is limited to either (1) the star:

of an it's RC7 with the sec:ndary water te :erature of the steam generat:r less thar.-or squal to 50*? above the RC3 c:1d leg tamperatures or (2).the i

start of 3 charging pumps and their injection into a water solid RCS, 3/4.".*C STRUCTL'?AL INTEGRITY The inservice inspection and testin; programs for A5ME Code Class I, 2 and 2 :: ;onents ensure that the structural integrity ano operational readiness of these c:=penents will be maintained at an acceptable level throughout the life of the plant.

These programs are in accordance with Section XI of the ASME Eciler and Pressure Vessel Code and applicable Addenda as required by 10 CFR Part 50.55a(g) except where specific written relief has been granted by the C ::ission pursuant to 10 CFR Part 50.55a (g) (5) (i).

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ee FARLEY-UNIT 1 B 3/4 4 Amendment No. 18

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