ML20247H751

Amends 100 & 82 to Licenses NPF-9 & NPF-17,respectively, Updating Pressure & Temp Limits in Tech Spec 3/4.4.9 for Heatup & Cooldown of RCS

ML20247H751
Person / Time
Site:	McGuire, Mcguire
Issue date:	07/18/1989
From:	Matthews D Office of Nuclear Reactor Regulation
To:
Shared Package
ML20247H748	List: ML20247H746 ML20247H751 ML20247H761
References
NUDOCS 8907310120
Download: ML20247H751 (20)

Text

- - _ - _ - _ - -

A UNITED STATES f

g, NUCLEAR REGULATORY COMMISSION 5

j WASHINGTON. D. C. 20555

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DUKE POWER COMPAl4Y L

DOCKET NO. 50-369 McGUIRE NUCLEAR STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE L

Amendner.t No.100 License No. NPF-9 1.

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

A.

The application for amendment to the McGuire Nuclear Sthtion, Unit 1 (the facility) Facility Operating License No. NPF-9 filed by the Duke Power Company (the licensee) dated January 22, 1989, as supplemented May 17 and June 19, 1989, complies with the stendaros anc requirements of the Atomic Energy Act of 1954, as amenced (the Act) and the Com-ruission's rules and regulations as set forth in 10 CFR Chapter 1; B.

The facility will operate in conformity with the application, as onended, 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 Comission's regulations set forth in 10 CFR Chapter I; D.

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

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

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l 2.

Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachments to this license amendment, 3

and Paragraph 2.C.(2) of Facility Operating License No. NPF-9 is hereby l

amended to read as follows:

{

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment. No.100, are hereby incorporated into the license.

The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 14 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Original signed by:

David B. Matthews, Director Project Directorate 11-3 Division of Reactor Projects-1/II Office of Nuclear Reacte-Regulation

Attachment:

Technical Specification Changes Date of Issuance: July 18,1989 l

OFFICIAL RECORD COPY L

t

• See previous concurrence

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Accordingly, the license is hereby amended by page changes to the Technical Specifications as indicated in the attachments to this license amendment, and Paragraph 2.C.(2) of facility Operating License No. hPF-9 is hereby amended to read as follows:

(2) ghnicallpecifications N

The Technical ecifications contained in Appendix A, as revised through Anendnen No.

, are hereby incorporated into the license.

The licensee shal erate the f acility in accordance with the Technical Specifica s and the Environmental Protection Plan.

3.

This license amendment is effective as of its date of issuance.

OR Ti1E NUCLEAR REGULATORY COMMISSION David Bq Matthews, Director Project D{ rectorate 11-3 Division oT Reactor Projects-1/11 Office of N leur Reactor Regulation Attachnent:

Technical Specification Changes Date of Issuance:

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DUKE POWER COMPANY-DOCKET NO. 50-370

'McCUIRE NUCLEAR STATION, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 82 i

License No. NPF-17 1.

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

A.

The application for amendment to the McGuire Nuclear Station, Unit 2 (the facility) Facility Operating License No. hPF-17 filed by the Duke Power Company (the licensee) dated..anuary 22, 1989, as supplemented May 17 and June 19, 1989, complies with the standaros and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Com-mission's rules and regulations as set forth in 10 CFR Chapter I;-

B.

The facility will operate in conformity with the application, as I

amer.ded, the provisions of the Act, and the rules anc regulation:5 of the Comission; C.

There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health ~ ano safety of the public, and (ii) that such activities will be, conducted in compliance with the Commission's regulations set furth in 10 CFR Chapter I; D.

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

The issuance of this amendment is in accoroance with 10 CFR Part 51 of the Comission's regulations and cil applicable requirements have been satisfied.

4 l

. 2.

Accordingly, the license is hereby anended by page changes to the Technical Specifications as indicated in the attachments to this license amendnent, and Paragraph 2.C.(2) of facility Operating License No. NPF-17 is hereby amended to read as follows:

(2) Technict.2 Specifications

.The Technical Specifications contained in Appendix A, as revised through Amendment No. 82, are hereby incorporated into the license.

.The licensee shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 14 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Original signed by:

David B. Matthews, Director Project Directorate 11-3 Division of Reactor Projects-1/Il l

Office of Nuclear Reactor Regulation

Attachment:

Technical Specification i

Changes l

Date of Issuance:

July 18, 1989 l

J l

l OFFICIAL RECORD COPY I

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Accordingly, the license is hereby arended by page changes to the Technical Specifications as indicated in the attachnients toLthis license amendment, Paragraph 2.C.(2) of-Facility Operating License No. NPF-17 is hereby a

ame ed to read as follows:

'(2)' Technical Specifications The'T hnica1' Specifications contained in--Appendix A, as revised through ndnent No.

., are hereby incorporated into the license.

The lice e shall' operate the facility in accordance with the Technical cificattons and the Environmental Protect 1on Plan.-

-3.

This license amendme t is effective as of its date of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION David B. Matthews, Director roject Directorate 11-3

' vision of Reactor Projects-I/II Of ce of Nuclear Reactor Regulation Attachnent:

Technical: Specification Changes Date of Issuance:

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l' ATTACHMENT TO LICENSE AMENDMENT NO.100 j

1 FACILITY OPERATING LICENSE NO. NPF-9 DOCKET NO. 50-369 AND T0 LICENSE AMENDMENT NO. 82 FACILITY OPERATING LICENSE NO. NPF-17 DOCKET NO. 50-370 Replace the following pages of_the Appendix "A" Technical Specifications with the enclosed pages. The revised pages are identified hy Amendment nunber and contain vertical lines indicating the areas of change.

The corresponding over-leaf pages are also provided to maintain document completeness.

Amended Page' Overleaf Page 3/4 4 3/4 31 3/4 4-32 3/4 4-33 l

3/4 4-34 1

3/4 4-35 3/4 4-36 B 3/4 4-7 B 3/4 4-8 l

B 3/4 4-13 8 3/4 4-14 8 3/4 4-16 B 3/4 4-15 l

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REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION 3.4.9.1 The Reactor Coolant System (except the pressurizer) temperature and pressure shall be limited in accordance with the limit lines shown on Figures 3.4-2, 3.4-3, 3.4-4, and 3.4-5 during heatup, cooldown, criticality, and l

inservice leak and hydrostatic testing with:

Maximum heatup rates as specified in Figures 3.4-2 and 3.4-3 l

a.

b.

Maximum cooldown rates as specified in Figures 3.4-4 and 3.4-5 l

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

1-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 to within the limit within 30 minutes; perform an engineering evaluation to determine the effects of the out of-limit condition on the structural integrity of the Reactor Coolant System; determine that the Reactor Coolant System 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 avg than 200 F and 500 psig, respectively, within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLANCE REQUIREMENTS 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be determined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations.

4.4.9.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and 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, 3.4-3, 3.4-4, and 3.4-5.

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1 McGUIRE - UNITS I and 2 3/4 4-30 Amendment No. 82 (Unit 2)

Amendment No.100 (Unit 1)

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FOR THE SERVI':E PERIOD UP TO 10 EFPY 0

0 50 100 150 200 250 300 350 400 450 500 INDICATED TEMPERATURE (OF)

CURVE APPLICABLE POR HEATUP MATERIAL BASIS RATES UP TO te*/HR FOR THE CONTROLLING MATERIAL WELO METAL SERVICE PERIOD L TO to EFPY COPPER CONTENT 4.30wt%

CONTAINE MARGIN FOR PossitLE PHOSPHORUS CONTENT-4.013wt%

INSTRUMENT ERRORE.

RTwayiNITIAL 4'F 1/4T, 165.B'F NOTAFTER 10 EFPY 3/47,113*F RT FIGURE 3.4-2 McGUIRE UNIT 1, REACTOR COOLANT l

SYSTEM, HEATUP LIMITATIONS

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NRC RG 1.99 REV 2 APPLICA LE F R TH FI ficGUIRE - UNITS I and 2 3/4 4-31 men ent Amendment No. 6:

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CURVES APPLICASLE POR HEATUP MATERIAL SASIS RATES UP TO es*F/HR FOR THE CONTROLLING MATE RI AL-R EACTOR SERVICE PERIOD UP TO 8 EPPY VESSEL INTERMEDIATE SHELL OS ANO CONT AINS MARGINS OF COPPER CONTENT 0.18 10 F AND 00 PSIO FOR POSSISLE 8

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FIGURE 3.4-3 MCGUIRE UNIT 2 REACTOR COOLANT SYSTEM, HEATUk LIMITATIONS APPLICABLE FOR THE FIRST 8 EFPY McGUIRE - Units 1 and 2 3/4 4-32 Amendment No.100 (Unit 1)

Amendment No. 82 (Unit 2)

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SERVICE PERIOO LP TO 10 EPPY COPPER CONTENT-0.30wt%

CONTAINS MARGIN FOR POSSitLE PHOSPHORUS CONTENT-4.013wt%

8 IFISTRUMENT ERRORS.

RTNOTINITIAL-4*P 1/4T, 168.8 r RT APTER 10 EPPY 3/4Y,113*P NDT y

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FIGURE 3.4-4 McGUIRE UNIT 1, RE ACTOR COOLANT SYSTEM, COOLDOWN LIMITATIONS NRC RG 1.99 REV 2 APPLICABLE FOR THE FIRST 10 EFPY ticGUIRE - UNITS I and 2 3/4 4-33 Amendment No.100 Unit 1 l

Amendment No. 82 Unit 2 w________

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INSTRUMENT ERRORS.

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FIGURE 3.4-5 McGUIRE UNIT 2 REACTOR COOLANT SYSTEM, COOLD6WN LIMITATIONS APPLICABLE FOR THE FIRST 8 EFPY l1cGUIRE.- UNITS 1 and 2 3/4 4-34 Amendment No.100 (Unit 1)

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REACTOR COOLANT SYSTEM PRESSURIZER LIMITING CONDITION FOR OPERATION 3.4.9.2 The pressurizer temperature shall be limited to:

a.

A maximum heatup of 100 F in any 1-hour period, b.

A maximum cooldown of 200*F in any 1-hour period, and c.

A maximum spray water temperature differential of 320'F.

APPLICABILITY:

At all times.

ACTION:

With the pressurizer temperature limits in excess of any of the above limits, restore the temperature to within tne limits within 30 minutes; perform an engineering evaluation to determine the effects of the out-of-limit condition on the structural integrity of the pressurizer; determine that the pressurizer 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 pressurizer pressure to less than 500 psig within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

SURVEILLA'4CE REQUIREMENTS 4.4,9.2 The pressurizer temperatures shall be determined te be within the limits at least once per 30 minutes during system heatup or cooldown.

The spray water temperature differential shall be determined to be within the limit at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> during auxiliary spray operation.

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McGUIRE - UNITS 1 and 2 3/4 4-36 I

REACTOR COOLANT SYSTEM' BASES SPECIFIC ACTIVITY (Continued)

Reducing T to less than 500 F prevents.the release of activity should

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gyg a steam generator tube rupture since the saturation pressure of the reactor coolantLis below the lift pressure of the atmospheric steam relief valves.

The-Surveillance Requirements provide adequate assurance that excessive specific activity levels in the reactor coolant will be detected in sufficient time to'.take corrective ACTION.

Information obtained on iodine spiking will be used to assess the parameters associated with spiking phenomena.

A reduction in frequency of isotopic analyses following power changes may be permissible if justified by the data obtained.

3/4.4.9 PRESSURE / TEMPERATURE LIMITS The temperature and pressure changes during heatup and cooldown are limited to be consistent with the requirements given in the ASME Boiler and Pressure Vessel Code,Section III, Appendix G:

.1.

The reactor coolant temperature and pressure and system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with Figures 3.4-2, 3.4-3, 3.4-4 and 3.4-5 for the service l

period specified thereon:

a.

Allowable combinations of pressure and temperature for specific temperature change rates are below and to the right of the limit lines shown.

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

Figures 3.4-2, 3.4-3, 3.4-4 end 3.4-5 define limits to assure pre-l vention of non-ductile failure only.

For normal operation, other inherent plant characteristics, e.g., pump heat addition and pres-surizer heater capacity, may limit the heatup and cooldown rates that can be acitieved over certain prassure-temperature ranges.

2.

These limit lines shall be calculated periodically using methods provided

below, 3.

The secondary side of the steam generator must not be pressurized above 200 psig if the temperature of the steam generator is below 70 F, 4.

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

The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 320 F, and 5.

System preservice hydrotests and inservice leak and hydrotests shall be I~

performed at pressures in accordance with the requirements of ASME Boiler and Pressure Vessel Code,Section XI.

l Amendment No.100(Unit 1)

McGUIRE - UNITS 1 and 2 B 3/4 4-7 Amendment No. 82(Unit 2) i

i m

REACTOR COOLANT SYSTEM BASES r

i

. PRESSURE / TEMPERATURE LIMITS (Continued)

The fracture toughness properties of the ferritic materials in the reactor vessel are determined in accordance with the NRC Standard Review Plan, ASTM E185-73, and in accordance with additional reactor vessel requirements.

These properties are then evaluated in accordance with Appendix G of the 1976 Summer Addenda to Section III of the ASME Boiler and Pressure Vessel Code and the calculation methods described in WCAP-7924-A, " Basis for Heatup and Cooldown Limit Curves, April 1975."

Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RTNDT, at the end of the effective full power years (EFPY) of service life identified on the applicable technical specification figure.

The 10 EFPY service life period is chosen such that the limiting RT at the 1/4T location in the core region is greater than 4

t NDT the RT f the limiting unirradiated material.

The selection of such a NDT limiting RT assures that all components in the Reactor Coolant System will NDT be operated conservatively in accordance with applicable Code requirements.

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

Reactor operation and resultant fast neutron (E greater than 1 MeV) irradiation can cause an increase in the RT Therefore, an adjusted reference temperature, NDT.

based upon the fluence, copper content, and phosphate content of the material in question, can be predicted using Figure B 3/4.4-1 and the largest value of ART F r Unit 1, the adjusted reference temperature has been computed by NDT.

Regulatory Guide 1.99, Revision 2.

For Unit 2, the adjusted reference temperature has been computed as discussed in WCAP-11029.

The heatup and cooldown limit curves of Figures 3.4-2, 3.4-3 3.4-4 and 3.4-5 include predicted adjustments for this shift in RT at the end of the identified service life.

NDT Adjustments for possible errors in the pressure and temperature sensing instruments are included when stated on the applicable figure.

Values of ART determined in this manner may be used until the results NDT

-from the material surveillance program, evaluated according to ASTM E185, are i

available.

Capsules will be removed in accordanta with the requirements of i

ASTM E185-73 and 10 CFR 50, Appendix H.

The surveillance specimen withdrawal schedule is shovr in Table 4.4-5.

The lead factor represents the relationship between the fast neutron flux density at the location of the capsule and the inner wall of the pressure vessel.

Therefore, the results obtained from the surveillance specimens can be used to predict the future radiation damage to the pressure vessel material by using the lead factor and the withdrawal time of the capsule.

The heatup and cooldown curves must be recalculated when the

(

LRT determined from the surveillance capsule exceeds the calculated ART NDT NDT for the equivalent capsule radiation exposure.

Allowable pressure-temperature relationships for various heatup and cool-down rates are calculated using methods derived from Appendix G in Section III of the ASME Boiler and Pressure Vessel Code as required by Appendix G to 10 CFR Part 50, and these methods are discussed in detail in WCAP-7924-A.

Amendment No.100(Unit 1)

McGUIRE - UNITS 1 and 2 B 3/4 4-8 Amendment No. 82(Unit 2)

Tnis page intentionally deleted.

i ficGUIRE - UNITS I and 2 B 3/4 4-13 Amendment No. 100 (Unit 1)

Arendment flo. 82 (Unit 2) i

A REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued)

The general method for calculating heatup and cooldown limit curves is based upon the principles of the linear elastic fracture mechanics (LEFM) technology.

In the calculation procedures a semi-elliptical surface defect l

with a depth of one quarter of the wall thickness, T, and a length of 3/2T i

is assumed to exist at the inside of the vessel wall as well as at the 1

outside of the vessel wall.

The dimensions of this postulated crack, referred to in Appendix G of ASME Section III as the reference flaw, amply exceed the current capabilities of inservice inspection techniques.

Therefore, the reactor operation limit curves developed for this reference crack are conservative and provide sufficient safety margins for protection against non-ductile failure.

To assure that the radiation embrittlement effects are accounted for in the calculation of the limit curves, the most limiting value of the nil-ductility reference temperature, RTNDT, is used and this includes the radiation-induced shift, ARTNDT, corresponding to the end of -

the period for which heatup and coeldown curves are generated.

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 oh,cooldowncannotbegreaterthanthereferencestressintensityfactor,K yg, for the metal temperature at that time.

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

The K curve is g hen by the equation:

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

(1)

Where:

K is the reference stress intensity factor as a function of the metal temperatub$Tandthemetalnil-ductilityreferencetemperatureRT

Thus, NDT.

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

CKyg + Kyg < KIR

(?)

Where:

K is the stress intensity factor caused by membrane (pressure) yg

stress, K

is the stress intensity factor caused by the thermal gradients, yg K

is provided by the code as a function of temperature relative tbRthe RT of the material, NDT 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 themetaltemperatureatthetipofthepostulatedflaw,thkRappropriate value for RTNDT, and the reference fracture toughness curve.

The thermal stresses McGUIRE - UNITS 1 and 2 B 3/4 4-14

4 REACTOR COOLANT SYSTEM BASES PRESSURE / TEMPERATURE LIMITS (Continued) resulting from temperature gradients through the vessel wall are calculated and then the corresponding thermal stress intensity factor, K for the reference flaw is computed.

FromEquation(2)thepressuresNe,ssintensity factors are obtained and, from these, the allowable pressures are ca:culated.

C00LDOWN For the calculation of the allowable pressure versus coolant temperature during cooldown, the Code reference flaw is assumed to exist at the inside of i

the vessel well.

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-temperature relations are generated for both steady-state and finite cooldown rate situations.

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

The use of the composite curve in the cooldown analysis is necessary because control of the cooldown 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 IR location for finite cooldown rates than for steady-state operation.

Further-more, if conditions exist such that the increase in K exceeds Kgg, the yp calculated 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.

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HEATUP Three separate calculations are required to determi e the limit curves n

for finite heatup rates.

As is done in the cooldown analysis, allowable

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pressure-temperature relationships are developed for steady-state conditions as well as finite 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 stresses produced by internal pressure.

The metal temperature at the crack tip lags the coolant temperature; therefore, the K for the 1/4T crack during heatup is lower than the K frthe1/4Tcrackdhingsteady-state IR conditions at the same coolant temperature.

During heatup, especially at the McGUIRE - UNITS 1 and 2 B 3/4 4-15

~

1 REACTOR COOLANT SYSTEM l

BASES PRESSURE / TEMPERATURE LIMITS (Continued) 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 rates IR do not offset each other and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for L

finite heatup rates when the 1/4T flaw is considered.

Thereforo, both cases have to be analyzed in order to assure that at any coolant temperature the lower value of the allowable pressure calculated for steady-', tate and finite heatup rates is obtained.

The second portion of the heatup analysis concerns 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 be 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 follows.

A composite curve is constructed baseo on a point-by-point comparison of the steady-state and finite heatup rate data.

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

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 of the heatup ramp the controlling condition switches from the inside to the outside and the pressure limit must at all times be based on analysis of the most critical criterion.

Finally, the composite curves in technical specifications for the heatup rate data and the cooldown rate data may be adjusted for possible er' ors in the pressure and temperature sensing instruments by the values indicated on the respective curves.

Where technical specification curves have not been adjusted, such adjustments are made by plant procedures.

Although the pressurizer operates in temperature ranges above those for which there is reason for concern of non-ductile failure, operating limits are provided to assure compatibility of operation with the fatigue analysis performed in accordance with the ASME Code requirements.

The OPERABILITY of two PORVs or an RCS vent opening of at least 4.5 square inches ensures that the xCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of Amendment No. 82 (Unit 2)

McGUIRE - U' LITS 1 and 2 B 3/4 4-16 Amendment No.100(Unit 1)