ML20137K237

Amends 74 & 60 to Licenses NPF-4 & NPF-7,respectively, Revising Tech Specs to Update Pressure Temp Limit Curves to Be Applied During Heatup & Cooldown

ML20137K237
Person / Time
Site:	North Anna
Issue date:	01/15/1986
From:	Rubenstein L Office of Nuclear Reactor Regulation
To:
Shared Package
ML20137K241	List: ML20137K237 ML20137K250
References
NUDOCS 8601230526
Download: ML20137K237 (24)

Text

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VIRGINIA ELECTRIC AND POWER COMPANY OLD DOMINION ELECTRIC COOPERATIVE DOCKET NO. 50-338 NORTH ANNA POWER STATION, UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 74 License No. NPF-4 1.

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

A.

The application for amendment by Virginia Electric and Power Company, et al. (the licensee), dated April 30, 1985, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (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.

8601230526 860115 gDR ADOCM 05000338 PDR

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. 2.

Accordingly, the license is amended by changes to the Technical Speci-fications as indicated in the attachment to this license amendment, and paragraph 2.D.(2) of Facility Operating License No. NPF 4 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 74, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective within 30 days from date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Lb b

Lester S. Rubenstein, Director PWR Project Directorate #2 Division of PWR Licensing-A

Attachment:

Changes to the Technical Specifications Date of Issuance:

January 15, 1986 4

1 ATTACHMENT TO LICENSE AMENDMENT NO. 74 1

TO FACILITY OPERATING LICENSE NO. NPF-4 DOCKET NO. 50-338 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 document completeness.

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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 and 3.4-3 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:

a.

A maximum heatup of 60*F in any one hour period.

l b.

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

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 determina 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 operations 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 and500psig,respectively,withinthefolloNSg30 hours.

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, at the intervals required by 10 CFR 50, Appendix H.

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

NORTH ANNA - UNIT 1 3/4 4-26 Amendment No. J6,74

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Figure 3.4.3 North Anna Power Station I Reactor Coolant System Cooldown Limitations Valid up to 10 EFPY NORTH ANNA - UNIT 1 3/4 4-28 Amendment No. JS, 74

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REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION 3.4.9.3 At least one of the following overpressure protection systems shall be OPERABLE:

a.

Two power operated relief valves (PORVs) with a lift setting of:

1) less than or equal to 420 psig whenever any RCS cold leg temperature is less than or equal to 375'F, and 2) less than or equal to 350 psig whenever any RCS cold leg temperature is less than 185 F, or i

b.

A reactor coolant system vent of greater than or equal to 2.07 square inches, or c.

A maximum pressurizer water volume of 457 cubic feet with all RCS cold leg temperatures greater than or equal to 320'F.

APPLICABILITY: When the temperature of one or more of the RCS cold legs is less than or equal to 375'F, except when the reactor vessel head is removed.

ACTION:

a.

With one PORV inoperable, either restore the inoperable PORV to OPERABLE status within 7 days or depressurize and vent the RCS through 2.07 square inch vent (s) within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />; maintain the RCS in a vented condition until both PORVs have been restored to OPERABLE status.

b.

With both PORVs inoperable, depressurize and vent the RCS through a 2.07 ::qu:re inch vent (s) within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />; maintain the RCS in a vented condition until both PORVs have been restored to OPERABLE status.

c.

In the event either the PORVs or the RCS vent (s) are used to mitigate a RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to 5)ecification 6.9.2 within 30 days. The report shall describe tie circumstances initiating the transient, the effect of the PORVs or vent (s) on the transient and any corrective action necessary to prevent recurrence.

d.

The provisions of Specification 3.0.4 are not applicable.

NORTH ANNA - UNIT 1 3/4 4-31 Amendment No. H,74

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l 4.4.9.3.1 Each PORV shall be demonstrated OPERABLE by:

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Performance of a CHANNEL FUNCTIONAL TEST on the PORV actuation channel, but excluding valve operation, within 31 days prior to 1

entering a condition in which the PORV is required OPERABLE and at least once per 31 days thereafter when the PORV is required j

OPERABLE.

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b.

Performance of a CHANNEL CALIBRATION on the PORV actuation channel, at least once per 18 months.

Verifying the PORV keyswitch is in the Auto position and the c.

PORV isolation valve is open at least once per 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> when the l

l PORV is being used for overpressure protection.

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Testing pursuant to Specification 4.0.5.

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4.4.9.3.2 The RCS vent (s) shall be verified to be open at least once per l

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />

• when the vent (s) is being used for overpressure protection.

• Except when the vent pPthway is provided with a valve which is locked, sealed, or otherwise secured in the open position, then verify these valves open at least once per 31 days.

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' F.I.GURE B 3/4 4.1 PAS. T IEUTRON PLUENCE (E >1ese) A8 A PUIeCTION OF PULL PoupER SERVICE LIFE 9

NORTH ANNA - UNIT 1 B 3/4 4-7 s

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REACTOR COOLANT SYSTEM BASES l

a The heatup analysis also covers the detemination of pressure-temperature limitations for the case in which the outer wall of the vessel becomes the controlling location.

The themal gradients established during heatup produce tensile stresses at the outer wall of the vessel.

These stresses are additive to the pressure induced tensile stresses which are already present. The thermal induced stresses at the outer wall of the vessel are tensile and are dependent on both the rate of heatup and the time along the heatup ramp; therefore, a lower bound curve similar to that described for the heatup of the inner wall cannot be defined.

Conse-quently, for the cases in which the outer Wall of the vessel becomes the stress controlling location, each heatup rate of interest must be analyzed on an individual basis.

2 The heatup limit curve, Figure 3.4-2, is a composite curve which was prepared by determining the most conservative case, with either the inside or outside wall controlling, for any heatup rate up to 60*F per hour. The cooldown limit curves of Figure 3.4-3 are composite curves which were prepared based upon the same type analysis with the exception that the controlling location is always the inside wall where the cooldown thermal gradients tend to produce tensile stresses while producing compressive stresses at the outside wall. The heatup and cooldown curves were prepared j

based upon the most limiting value of the predicted adjusted reference temperature at the end of 10 EFPY. The adjusted reference temperature was calculated using results from a capsule removed after the first cycle.

The results are documented in Babcock and Wilcox Reports BAW-1638, May 1981 and BAW-1872, April 1985.

The reactor vessel materials have been tested to detemine their initial RTunT. The results of these tests are'shown in Table B 3/4 4-1.

Reactor operation and resultant fast neutron (E>l Hev) irradiation will cause an increase in the RT Therefore, an adjusted reference temperature, based upon the fluEe and copper content of the material in question, can be predicted using Figures B 3/4.4-1 and B 3/4.4-2. The heatup and cooldown I

limit curves (Figure 3.4-2 and 3.4-3) include predicted adjustments for l

this shift in RT at the end of 10 EFPY, as well as ad;ustments for l

possible errors NTthe pressure and temperature sensing nstruments.

8 The actual shift in RT of the vessel material will be established periodicallyduringoperatiEbyremovingandevaluating,inaccordance i

with ASTM E185-70, reactor vessel material irradiation surveillance j

specimens installed near the inside wall of the reactor vessel in the 4

core area. Since the neutron spectra at the irradiation samples and i

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NORTH ANNA - UNIT 1 B 3/4 4-8 Amendment No. 74

REACTOR COOLANT SYSTEM BASES vessel inside radius are essentially identical, the measured transition shift for a sample can be applied with confidence to the adjacent section of the reactor vessel. The heatup and cooldown curves must be recalculated when the ART detemined from the surveillance capsule is different fromthecalb[atedART for the equivalent capsule radiation exposure.

NDT The pressure-temperature limit lines shown on Figure 3.4-2 for renctor criticality and for inservice leak and hydrostatic testing have been provided to assura compliance with the minimum temperature require-ments of Appendix G to 10 CFR 50.

The number of reactor vessel irradiation surveillance specimens and the frequencies for removing and testing these specimens are provided in Table 4.4-3 to assure compliance with the requirements of Appendix H to 10 CFR Part 50.

The limitations imposed on pressurizer heatup and cooldown and

)

spray water temperature differential are provided to assure that the pressurizer is operated within the design criteria assumed for the fatigue analysis performed in accordance with the ASME Code requirements.

The OPERABILITY of two PORVs or an RCS vent cpening of greater than l

2.07 square inches ensures that the RCS will be protected from pressure transients which could exceed the limits of Appendix G to 10 CFR Part 50 when one or more of the RCS cold legs are less than or equal to 375*F.

l Either PORY has adequate relieving capability to protect the RCS from overpressurization when the transient is limited to either (1) the start of an idle RCP with the secondary water temperature of the steam generator less than or equal to 50*F above the RCS cold leg temperatures or (2) the start of a charging pump and its injection into a water solid RCS.

When the temperature of the RCS cold legs is between 320*F and 375'F, overpressure protection can also be provided by a bubble in the pressurizer.

In such a case, a maximum pressurizer water volume of 457 cu. ft. has been 4

selected to provide at least 10 minutes for operator response in the event of a malfunction resulting in maximum flow from one charging pump.

NORTH ANNI. - UNIT 1 B 3/4 4-11 Amendment No. 5,M/

• 74

a REACTOR COOLANT SYSTEM BASES a

3/4.4.10 STRUCTURAL INTEGRITY 3/4.4.10.1 ASME CODE CLASS 1, 2 AND 3 COMPONENTS The inspection programs for ASME Code Class 1, 2 and 3 Reactor Coolant System components ensure that the structural integrity of these components l

will be maintained at an acceptable level throughout the life of the plant.

l To the extent applicable, the inspection program for components is in compliance with Section XI of the ASME Boiler and Pressure Vessel Code.

NORTH ANNA - UNIT 1 B 3/4 4-12 Amendment No.16, 58

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UNITED STATES E

i NUCLEAR REGULATORY COMMISSION

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VIRGINIA ELECTRIC AND POWER COMPANY OLD DOMINION ELECTRIC COOPERATIVE DOCKET NO. 50-339 NORTH ANNA POWER STATION, UNIT NO. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 60 License No. NPF-7 1.

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

A.

The application for amendment by Virginia Electric and Power Company, et al. (the 'icensee), dated April 30, 1985, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission'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 Comission; 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; 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 Comission's regulations and all applicable requirements have been satisfied.

. I O

2.

Accordingly, the license is amended by changes to the Technical Speci-fications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-7 is hereby amended to read as follows:

(2) Technical Specifications t

The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 60, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective prior to restart after the forthcoming fifth (5) refueling outage.

FOR THE NUCLEAR REGULATORY COMMISSION

u. L L

Lester. Rubenstein, Director PWR Project Directorate #2 Division of PWR Licensing-A l

Attachment:

Changes to the Technical Specifications Date of Issuance: January 15, 1986 0

4

. _ _ _ _ _ _ =_ __ __.

d ATTACHMENT TO LICENSE AMENDMENT NO. 60 TO FACILITY OPERATING LICENSE NO. NPF-7 DOCKET NO. 50-339 i

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 document completeness.

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20 30 40 50 00 70 30 90 100 PERCENT OF RATED THERMAL POWER FIGURE 3.4-1 DOSE EQUIVALENT l-131 Primary Coolant Specific Activity Limit Versus Parcovet of RATED THERMAL POWER with the Primary Coolant Specific Activity >1.0 Ci/yem Dose Equivalent 1131 NORTH AhNA - UNIT 2 3/4 4-25 m

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I REACTOR COOLANT SYSTEM 3/4.4.9 PRESSURE / TEMPERATURE LIMITS REACTOR COOLANT SYSTEM LIMITING CONDITION FOR OPERATION l

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 and 3.4-3 during heatup, cooldown, criticality, and inservice leak and hydrostatic testing with:

A maximum heatup of 60*F in any one hour period.

a.

[

b.

A maximum cooldown of 100*F in any one hour period, A maximum temperature change of less than or equal to 10*F in any c.

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

APPLICA8ILITY: 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 i

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 operations or be in at least HDT STAN08Y within the next6hoursandreducetheRCSTl"I0 hours.and pressure to less than 200*F and 50 f

respectively, within the followin i

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$URVEILLANCE REQUIREMENTS i

4.4.9.1.1 The Reactor coolant System temperature and pressure shall be deter-eined to be within the limits at least once per 30 minutes during system heatup, cooldown, and inservice leak and hydrostatic testing operations.

l 4.4.9.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties, at the intervals required by 10 CFR 50, Appendix H.

The results of these examinations sha11 be used to update Figures 3.4-2 and 3.4-3.

NORTH ANNA - UNIT 2 3/4 4-26 Amendment No. 60

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i UNACCEPTABLE OPERATION f

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100 200 300 400 500 INDICATED TEMPERATURE (OF)

Figure 3.4.2 Reactor Coolant System Temperature-Pressure Heatup Limitations Valid up to 10 EFPY NORTH ANNA - UNIT 2 3/4 4-27 Amendment No. 60

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Figure 3.4.3 Reactor Coolant System Temperature-Pressure Cooldown Limitations I

Valid to 10 EFPY NORTH ANNA - UNIT 2 3/4 4-28 Amendment No. 60

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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 or cooldown of 200'F, in any one hour period, and b.

A maximum spray water temperature and pressurizer temperature differential of 320'F.

APPLICA8ILITY: At all times.

ACTION:

With the pressurizer temperature limits in excess of any of the above limits,

' restore the temperature to within the 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 />.

t I

SURVEILLANCE REQUIREMENTS I

4.4.9.2 The pressurizer temperatures shall be determined to be within the limits at least once per 30 minutes during system heatup or cooldown.

The spray water temperature differential shalf 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.

NORTH ANNA - UNIT 2 3/4 4-29

REACTOR COOLANT SYSTEM OVERPRESSURE PROTECTION SYSTEMS LIMITING CONDITION FOR OPERATION

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3.4.9.3 At least one.of the following overpressure protection systems shall i

be OPERABLE:

i i

Two power operated relief valves (PORVs) with a lift setting of:

1) a.

less than or equal to 520 psig whenever any RCS cold leg temperature is less than or equal to 340*F, and 2) less than or equal to 375 psig whenever any RCS cold leg temperature is less than 190*F, or b.

A reactor coolant system vent of greater than or equal to 2.07 l'

square inches, or i

A maximum pressurizer water volume of 457 cubic feet with all RCS c.

cold leg temperatures greater than or equal to 320'F.

APPLICABILITY:

When the temperature of one or more of the RCS cold legs is less than or equal to 340*F, except when the reactor vessel head is removed.

j ACTION:

3 With one PORV inoperable, either restore the inoperable PORV to o

a.

1 OPERABLE status within 7 days or depressurize and vent the RCS i

through 2.07 square inch vent (s) within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />; maintain, the RCS in a vented condition until both PORVs have been restored to OPERABLE status, b.

With both PORVs inoperable, depressurize and vent the RCS through a 2.07 square inch vent (s) within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />; maintain the RCS in a j

vented condition until both PORVs have been restored to OPERABLE j

status, l

In the event either the PORVs or the RCS vent (s) are used to mitigate c.

4 a RCS pressure transient, a Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within i

30 days.

The report shall describe the circumstances initiating the transient, the effect of the PORVs or vent (s) on the transient and any corrective action necessary to prevent recurrence.

d.

The provisions of Specification 3.0.4 are not applicable.

NORTH ANNA - UNIT 2 3/4 4-30 Amendment No. 60

REACTOR COOLANT SYSTEM BASES j

Heatup and cooldown limit curves are calculated using the most limiting value of the nil-ductility reference temperature, RTNDT, at the end of 10 effective full power years of service life. The 10 EFPY service life period is chosen such that the limiting RT at the 1/4T location in the NDT core region is greater than the RT of the limiting unirradiated material.

NDT f

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

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

Reactor i

operation and resultant fast neutron (E greater than 1 MEV) irradiation can cause an increase in the RT NDT. Therefore, an adjusted refer;ence tempera-ture, based upon the fluence and copper content of the material in question, can be predicted using Figures B 3/4.4-1 and B 3/4.4-2. The adjusted reference temperature was calculated using results from a capsule removed after the first core cycle. The results are documented in Babcock and Wilcox Reports BAW-1794, October 1983 and BAW-1872, April 1985. 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 10 EFPY, as well as I

NDT adjustments for possible errors in the pressure and temperature sensing instruments.

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 available. The first capsule was removed at the end of the first I

core cycle. Successive capsules will be removed in accordance with the requirements of ASTM E185-73 and the latest revision of 10 CFR 50, l

Appendix H.

The heatup and cooldown curves must be recalculated when the ART detennined from the surveillance capsule exceeds the calculated NDT ART for the equivalent capsule radiation exposure.

NDT

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4 NORTH ANNA - UNIT 2 B 3/4 4-11 Amendment No. 60

4 REACTOR C00LA4T SYSTEM 8ASES l

Allowable pressure -temperature relationships for various heatup and cooldown 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.

A 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 j

defect with a depth of one quarter of the wall thickness, T, and a length of 3/2T is assumed to exist at the inside of the vessel wall as well as at the outside of the vessel wall.

The dimensions of this postulated u

j crack, referred to in Appendix G of ASME III as the reference fiaw, amply exceed the current capabilities of inservice inspection techniques.

l Therefore, the reactor operation limit curves developed for this reference I]

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, RT NDT' j

is used and this includes the radiation induced shift, ARTNDT, corresponding l

to the end of the period for which heatup and cooldown curves are generated.

l The ASME approach for calculating the allowable limit curves for various 1

heatup and cooldown rates specifies that the total stress intensity factcr, 1

K, for the combined thermal and p h sure 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 from the reference IR l

fracture toughness curve, defined in Appendix G to the ASME Code. The K IR j

curve is given by the equation:

Kgg = 26.78 + 1.223 exp [0.0145(T-RTNOT + 1601)

(1)

NORTH ANNA - UNIT 2 8 3/4 4-12 i

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