ML20082S836

Heatup & Cooldown Limit Curves for Normal Operation, Sequoyah Unit 2

ML20082S836
Person / Time
Site:	Sequoyah
Issue date:	06/30/1991
From:	Chicots J, Meyer T, Ray N WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20082S825	List: ML20082S822 ML20082S828 ML20082S833 ML20082S836
References
WCAP-12971, NUDOCS 9109170313
Download: ML20082S836 (34)

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HEATUP AND C00LDOWN LIMIT CURVES h

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'I Work Performed Under Shop Order TJAP-N004 i-Prepared by Westinghouse Electric Corporation for the Tennessee Valley Authority l

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Structtiral Reliability & Plant: Life Optimization 9

WESTINGHOUSE ELECTRIC CORPORATION i

Nuclear and Advanced Technology Division P.O. Box 355 Pittsburgh, Pennsylvania 15230-0355 e 1991 Westinghouse Electric Corp.

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TABLE OF CONTENTS l

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I Section litle Paag 1

INTRODUCTION 3

2 FRACTURE TOUGHNESS PROPERTIES 1

3 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS 2

4 HEATUP AND C00LDOWN LIMIT CllRVES 5

5 ADJUSTED REFERENCE TEMPERATURE 6

6 REFERENCES 15 i

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LIST OF ILLUSTRATIONS-l Ficure Title

' Pace 1

Sequoyah Unit 2 Reactor Coolant System Heatup Limitations

.11-1

'(Heatup rate up to 20*F/hr) Applicable for the First 16 f

EFPY (Without Margins for Instrumentation Errors) i t

2

-.Sequoyah Unit 2 Reactor Coolant System Heatup Limitations 12 (Heatup rate up to 40*F/hr) Applicable for the First 16

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EFPY (Without Margins for Instrumentation Errors) 3 Sequoyah Unit 2 Reactor Coolant System Heatup Limitations 13 (Heatup rate up to 60*F/hr) Applicable for the First 16 EFPY (Without Margins-for Instrumentation Errors) i i

4 Sequoyah Unit 2 Reactor Coolant System Cooldown (Cooldown 14

]j Rates up to 100'F/hr) Limitations Applicable fer the First 16 EFPY (Without Margins for Instrumentation Errors)

LIST OF TABLES i

Tablq Title P, igg l

1 Sequoyah Unit 2 Reactor Vessel Toughness Table

-8 (Unirradiated) 2 Calculation of Chemistry Factor Based on Surveillance 9

Capsule Data i

3 Summary of Adjusted Reference Temperature (ART) at 1/4T 9

l and 3/4T Location 4

Calculation of Adjusted Reference Temperatures for 10

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Limiting Sequoyah Unit 2 Reactor Vessel Material - Weld i

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1. INTRODUCTION Heatup and cooldown limit curves are calculated using the most limiting value of RTNDT (reference nil-ductility temperature) for the reactor vessel.

The most limiting RTNDT of the material in the core region of the reactor vessel is determined by using the preservice reactor vessel material fracture toughness properties and estimating the radiation-induced ARTNDT-RTNDT is designated as the higher of either the drop weight nil-ductility transition tempcrature (NDTT) or the temperature at which the material exhibits at least 50 ft-lb of impact energy and 35-mil lateral expansion (normal to the major working direction) minus 60'F.

RTNDT increases as the material is exposed to fast-neutron radiation.

Therefore, to find the most limiting RTNDT at any time period in the reactor's life, ARTNDT due to the radiation exposure associated with that time period must be added to the original unirradiated RTNDT.

The extent of the shift in RTNDT is enhanced by certain chemical elements (such as copper and nickel) present in reactor vessel steels.

The Nuclear Regulatory Commission (NRC) has published a method for predicting radiation embrittlement in Regulatory Guide 1.99 Rev. 2 (Radiation Embrittlement of Reactor Vessel Materials)ll).

Regulatory Guide 1.99, Revision 2 is used for the calculation of RTNDT values at 1/4T and 3/4T locations (T is the thickness of the vessel at the beltline region).

2.

FRACTURE TOUGHNESS PROPERTIES The fracture-toughness properties of the ferritic materici in the reactor coolant pressure boundary are determined in accordance with the NRC Regulatory Standard Review Plan [23 The pre-irradiation fracture-toughness properties of the Sequcyah Unit 2 reactor vessel are presented in Table 1.

1

6 3.

CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS The ASME approach for calculating the allowable 1fmit curves for various heatup and cooldown rates specifies that the total stress intensity factor, Kg, for the combined thermal and pressure stresses at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, KlR' for the metal temperature at that time.

KIR is obtained from the reference fracture toughness curve, defined in Appendix G to the ASME CodeI33 The Kyg curve is given by the following equation:

KIR - 26.78 + 1.223 exp [0.0145 (T-RTNDT + 160))

(1) where KIR = reference stress intensity factor as a function of the metal temperature 1 and the metal reference nil-ductility temperature RTNDT Therefore, the governing equation for the heatup-cooldown analysis is defined in Appendix G of the ASME CodeI33 as follows:

CKg+ KIT I KIR (2) i where Kyg = stress intensity factor caused by membrane (pressure) stress KIT = stress intensity factor caused by the thermal gradients KIR " function of temperature relative to the RTNDT of the material t

C

= 2.0 for Level A and Level 8 service limits C

- 1.5 for hydrostatic and leak test conditions during which the reactor core is not critical 2

At any time during the heatup or cooldown transient, XIR is determined by the metal temperature at the tip of the postulated flaw, the appropriate value for RTNDT, and the reference fracture toughness curve.' The thermal stresses resulting from the temperature gradients through the vessel wall are calculated and then the corresponding (thermal) stress intensity factors, KIT, for the reference flaw are computed.

From equation 2, the pressure stress intensity l

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

For the calculation of the allowable pressure versus coolant temperature during I

cooldown, the reference flaw of Appendix G to the ASME Code 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 i

produce tensile stresses at the inside, which increase with increasing cooldown rates.

Allowable pressure-temperature relations are generated for both r

steady-state and finite cooldown rate situations.

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

e The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on the 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 1/4 T vesrel location is at a higher temperature than the fluid adjacent to the ve.ssel ID.

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

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

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

The above procedures are needed because there is no direct control on temperature at the 1/4 T location and, therefore, allowable pressures may Y

unknowingly be violated if the rate of cooling is decreased at various 3

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intervals along a cooldown ramp.

The use of the composite curve eliminctes t

this problem and ensures conservative operation of the system for the entire cooldown period.

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Three separate calculations are required to determine the limit curves for j

finite heatup rates.

As is done in the cooldown analysis, allowable pressure-temperature relationships are developed for steady-state conditions as well as f

finite heatup rate conditions assuming the presence of a 1/4 T defect at the inside of the wall that alleviate the tensile stresses produced by internal i

pressure.

The metal temperature at the crack tip lags the coolant temperature; I

therefore, the KIR for the 1/4 T crack during heatup is lower than the K IR for the 1/4 T crack during steady-state conditions at the same coolant temperature. During heatup, especially at the end of the transient, conditions may exist so that the effects of compressive thermal stresses and lower KIR's do not offset each other, and the pressure-temperature curve based on steady-st

'onditions no longer represents a lower bound of all similar curves for finite heatup rates when the 1/4 T flaw is considered.

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

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The second portion of the heatup analysis concerns the calculation of the pressure-temperature limitations for the case in which a 1/4 T deep cutside i

surface flaw is assumed.

Unlike the situation at the vessel inside surface, the thermal gradients established at the outside surface during heatup produce strasses which are tensile in nature and therefore tend to reinforce any pressure stresses present. These thermal stresses are dependent on both the t

i rate of heatup and the time (or coolant temperature) along the heatup ramp.

Since the thermal stresses at the outside are tensile and increase with increasing heatup rates, each heatup rate must be analyzed on an individual basis.

I Following the generation of pressure-temperature curves for both the steady I

state and finite heatup rate situations, the final limit curves are produced by

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constructing a composite curve based on a point-by-point comparison of the steady-state and finite heatup rate data.

At any given temperature, the i

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.c allowable pressure is taken to be the lesser of th 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' wherein, over the course of the heatup ramp, the control-ling condition switches from the insi ' to the.outside, and the pressure limit must at all times be based on analy s of the most critical criterion.

Finally, the 1983 Amendment to 10CFR50I43 has a rule which addresses the metal temperature of the closure head flange and vessel flange regions.

This rule states that the metal temperature of the closure flange regions must exceed _the material RTNDT by at least 120'F for norm.11 operation when the pressure exceeds 20 percent of the preservice hydrostatic test. pressure.,

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i Table 1 indicates that the initial RTNDT of -13*F occurs in the closure head flange of Sequoy 5 Unit 2, so the minimum allov able temperature of :this region is 107'F.

These limits are shown in F'a'~

l and 2 whenever applicable.

4.

HEATUP AND C00LDOWN LIMIT CURVES '

t Limit curves for normal heatup and cooldown of the primary reactor pressure vessel have been calculated using the methods discussed in Se'ction 3.

Figures 1, 2 and 3 contain the heatup curves for 20, 40 and 60*F/hr, respectively.

Figure 4 contains the cooldown curves up to 100*F/hr.

Figures 1 through 4 are applicable for the first 16 EFPY of operation.

No, margins were included in the development of heatup and cooldown curves to allow for possible instrumentation errors.

Allowable combinations of temperature and pressure for specific temperature change rates for heatup operation are below and to the right of th'e limit lines

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shown in Figures 1 th~ rough 3.

This is in addition to other criteria which must

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be met before the reactor is made-critical.

Theleaklimidcurveshowr.inFigures1,2and3representminimumtemperature requirements at the leak test pressure specified by applicable codes [2'33 f

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The leak test limit curve was determined by methods of References 2 and 4.

r The criticality limit curve shown in Figures 1, 2 Tnd 3, specifies i

pressure-temperature limits for core operation to provide additional margin during actual power production as specified in Reference 4.

The

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pressure-temperature limits for core operation (except for low power physics i

tests) are that the reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test, and f

at least 40*F higher than the minimum pressure-temperature curve for her. tup

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and cooldown calculated as described in Section 3.

The maximum temperature for f

the inservice hydrostatic test for the Sequoyah Unit 2 reactor vessel is

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274*F.

A vertical line at 274'F on the pressure-temperature curve, intersecting a curve 40*F higher than the pressure-temperature limit curve, constitutes the limit for core operation for the reactor vessel.

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J Figures I through 4 define limits for ensuring prevention of nonductile failure for the Sequoyah Unit 2 reactor vessel.

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5, ADJUSTED REFERENCE TEMPERATURE From Regulatory Guide 1,99 Rev. 2 [1] the adjusted reference temperature (ART)

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for each material in the beltline is given by the following expression:

ART = Initial RTNDT.4 ARTNDT + Margin (3) i I

Initial RT NDT is the reference temperature for the unirradiated material as defined in paragraph N2331 of Section III of the ASME Boiler and Pressure Vessel Code.

If measured values of initial RTNDT for the material in l

L question are not available, generic mean values for that class of material may l

l be used if there are sufficient test results to establish a mean and standard deviation for the class.

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ARTNDT is the mean value of the adjustment in reference temperature i

caused by irradiation and should be calculated as follows:

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ARTNDT = (CF]f(0.28-0.10 log f)

(4)

To calculat' ARTNDT at any depth (e.g., at 1/4T or 3/4T), the following f

e formula must first be used to attenuate the fluence at the specific depth.

i f(depth X)'* Isurface(e'*

)

(5) where x (in inches) is the depth into the vessel wall measured from the vessel clad / base metal interface.

The resultant fluence is then put into equation (4) to calculate ARTNDT at the specific depth.

t r

/ CF (*F) is the chemistry factor, obtained from Reference 1.

In addtion, the chemistry factor is also calculated using surveillance capsule data.

A sample of this calculation is shown in Table 2.

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5 All materials in the beltline region-of Sequoyah Unit 2 were considered for the limiting material. RTNDT at 1/4T and 3/4T are summarized in Table 3.

From Table 3, it can be seen that the limiting material is the weld for heatup'and t

cooldown curves applicable up to 16 EFPY.

A sample calculation for the RTNDT for 16 EFPY is shown in Table 4.

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1 TABLE 1 SEQUOYAH UNIT 2 REACTOR VESSEL T0VGHNES5' TABLE (Unirradiated)

CU NI l-RTNDT'(a)

Material Description

(%)

(%)

(*F)

Closure Head Flange (b)

-13 Vessel Flange (b)

-22 Intermediate Shell Forging 0.13 0.74 10 Lower Shell Forging 0.14 0.76

-22 Welds 0.13 0.11

-4 a.

The initial RTNDT (I) values for the forgings and welds are measured values.

b.

These items are to be used for considering flange requirements for heatup/cooldown curvesI43 8

F-e TABLE 2 CALCULATION OF CHEMISTRY FACTOR BASED ON SURVEILLANCE CAPSULE DATA Material Description Capsule Fluence FF DRTNDT FF*DRTNQI (FF)2 Forging 05 (axial)

T 0.220 0.592 25 14.811 0.351 V

0.643 0.876 62 54.327 0.768 Forging 05 (tangential)

T 0.220 0.592 60 35.546 0.351 V

0.643 0.876 93 81.490 0.768 186.173 2.238 Chemistry Factor = 18.

- 83.205 3

Weld Metal T

0.220 0.592 80 47.394 0.351 0

0.643 0.876 130 113,911 0.768 161.305 1.119 Chemistry Factor, 161 305 = 144.182 TABLE 3

SUMMARY

OF ADJUSTED REFERENCE TEMPERATURE (ART) AT 1/4T and 3/4T LOCATION 16 EFPY RTNDT at Component 1/4T ('F) 3/4T ('F4 Intermediate Shell Forging (95)

(73)

Lower Shell Forging 97 70 Welds (142)*

(104)*

RTNDT numbers within ( ) are based on the chemistry factor calculated using capsule data.

These RTNOT numbers are used to generate heatup and cooldown curves appIicable up to 16 EFPY.

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CALCULATION OF ADJUSTED REFERENCE TEMPERATURES FOR LIMITING SEQUOYAH. UNIT.2 REACTOR VESSEL MATERIALt-WELD ~

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'Reanlat.prv Guide 1.99 --Revision:2 16 EFPY-brameter 1/4 T-3/4 T t

Chemistry Factor,.CF (*F) 68-(144) 68 (144)

Fluence, - f (l(19 n/cm )(a)

.5206

.1889 2

Fluence: Factor, ff

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.556 ARTNDT = CF x-ff ('F) 56 (118).

38 (80)

Initial RTNDT' I ('f)-

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~4 Margin, M (*F) (b) 56 (28) 56 (28)

. Revision 2 to Regulatory Guide 1.99 Adjusted Reference Temperature, 108 '('142) 90 (104)_

ART -Initial RTNDT_+ ARTNDT + Margin I9 2

--(a) : Fluence,; f,.is based upon fsurf (10

.n/cm, E>l Mev) = 0.8644 tt 16-EFPY, _-The Sequoyah Unit.2 reactor vessel wall-thickness is 8.6]8 inches at'the beltline region.

. (b)- Margin. is. calculated as,- M

- 2 [ 072 + oA ]

5 The standard hviation j

for the' initial RTNDT margin term, oi, is assumed to be O'F'since the i

initial.RTNDT is a m6asured value.

The standard. deviation for ARTNDT term, L

oA, is 28'F for the weld, except that oA need not exceed 0.5 times the L

mean value of ~ARTNDT' U,

is 14'F for the weld (cut in half) when A

-surveillance data is used.

The numbers within ( ) are calculated using surveillance capsule data,

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6.

REFERENCES 1

Regulatory Guide 1.99, Revision 2, "Radiatior,~ Embrittlement of Reactor Vessel Materials," U.S. Nuclear Regulatory Commission, May,1988.

2

" Fracture Toughness Requirements," Branch Technical Position MTEB 5-2, Chapter 5.3.2 in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, NUREG-0800, 1981.

3 ASME soiter and Ptsnure Vessel Code,Section III, Division 1 -

Appendixes, " Rules for Construction of Nuclear Power Plant Components, Appendix G, Protection Against Nonductile Fgilure," pp. 558-563, 1986 Edition, American Society of Mechanical Engineers, New York,1986.

4 Code of Federal Regulations, 10CFR50, Appendix G, " Fracture Toughness Requirements," U.S. Nuclear Regulatory Commission, Washington, D.C.,

federal Register, Vol. 48 No.104, May 27,1983.

i 9

15

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r ENCLOSURE 5 PRESSURIZED THERMAL Sil0CK ASSESSMENT i

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a SEQUOYAll NUCLEAR PLANT UNITS 1 AND 2 - PRESSURIZED THERMAL SHOCK SCREENING CRITERIA 1.

PURPOSE The purpose of this calculation is to verify that Sequoyah

' Nuclear Plant Units 1 and 2 reactor pressuro vessels moot the pressurized thermal shock (PTS) scrooning critoria of 10 CFR 50.61 dated May 15, 1991.

2.

INTRODUCTION Pressurized thermal shock events are system transients in a pressurized water reactor (PWR) that can cause savoro overcooling followed by immediate reprosaurization.

Internal stressos caused by rapid cooling of the reactor vossol insido rurface combined with the pressure stresses increase the potential for fracturo if an initiating flaw is present in 3ow toughness material.

This material may exist in the reactor vossol boltlino, adjacent to the caro, where neutron radiation gradually embrittles the material during operation of the plant.

The degran of embrittlement depends upon the chemical composition of the stool, especially the copper and nickel contents.

The pressurized thermal shock rule (PTS rulo) outablishes a scrooning cr!.torion which limits the amount of embrittlement of a reactor boltline beyond which the plant cannot operate without justification based on a plant specific analysis.

The toughness of reactor vessel materials is characterized by a reference temperature for nil ductility transition (RTHDT), which is determined by destructive tests of material specimens.

RTHDT is designated as the hignor of either the drop weight nil-ductility transition temperature (NDTT) or the temperature at which the material exhibits at least 50 ft-lb of impact energy and 35-mil lateral expansion (normal to the major working direction) minus 600P.

I t

The pressurized thermal shock rule, 10 CPR 50.61, adopted May 15, 1991, changes the procedure for establishing the limiting level of embrittlement.

This screening critorion is given in PTS to distinguish it from other terms of RTNDT, but called RT proceduras for calculating RTHDT*

2.

REFERENCE TEMPERATURE FOR PRES 8URIZED THERMAL SHOCK 0

The pressurized thermal shock (PTS) screening criterion is 270 F 0

for plates, forgings, and axial weld materials, or 300 F for circumferential wold materials.

e

?

SEQUOYAll NUCLEAR PIAllT UNITS 1 AllP 2 - PRESSURIZED TilERMAL SilOCK SCREENING CRITERIA From 10 C FR 50.61 dated May 15, 1991, RT is datormined for eachmaterialinthereactorvossolboltk,$3no fromt i

RTPTS I

+

M

+ ARTPTS (E40' 1)

=

L whoro I = tho initial reference temperatureunirradiated material as defined in(R of the of Section III of the ASME Doller and Prosauro Vossol Codo.

M = the margin to be added to covered uncertaintion in i

the values of initial RTNDT, coppor and nickol

[

contents, fluenco, and calculational p5 " d"r "-

equation 1, M is 66 F for wolds and 48 F for baso metals if generic values of I are used, and M is 56 F 0

for wolds and 34 F for bano metal if measured values of I are used.

t A AT the mean value of tho adjustmont in referenco PTS = temperature caused by irradiation and is calculated l

os follows:

l t

A RTPTS = (CF) f (0.28 - 0.10 log f)

(Eqn. 2)

I whuro CF(OF) = the chemistry factor, a function of copper and nlckel content, given in tables 1 and 2 of 10 CFR 50,61.

f = the best estimgge of t}eutron fluence, in units of 10 n/cm* (E greater then 1 MeV), at the clad-base-metal i

interface at the location where the matorial in question recoivos the highest fluence for the porlod of service in question.

f f a fluence f actor = f(0.28 - 0.10 log f) i 3.

FRACTURE TOUGilNESS PROPERTIES The fracturn-toughness proporties of the bolt 11no materials woro determined in accordance with the NRC Regulatory Standard Review i

Plan (Ref. 2).

The pro-irradiation fracture-toughness proporties i

of Sequoyah Units 1 and 2 are given in Tables 3 and 4.

Those f

data are the same as presented in references 3 and 4 on heatup and cooldown limit curves.

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SE?UOYAll HUCLEAlt PLANT UNITS 1 AllD 2 - PRESSUltIZED THERMAL Ull0CK f

SCREEN 1HG CRITERIA i

4. CALCUIATTON8 L

TABLE 1" SRQQDXMLU_Ilitd e

Scrooning i

Material A RT( PTS) d Initial Critorion h

Ihulcriution E'

ICrxtu RIpra Mat 91n RTgg Limit

(

Forging 05 116 157 40 34 231 270 Forging 04 95 128 73 34 235

270, (100) 133 73

[17) 223 270 Wolda 158 213

~40 56 229 300 i

(173) 230

-40

[28]

218 300 a.

All unita in O F.

b.

Chemistry Factor obtair.ed from tables 1 and 2 of 10 CFit 50.61 c.

Values in ( ? were datormined from credible survoillance data i

as reported un haatup and cooldown limit calculationn (Hof.

[

3, page 9).

d.

Unit 1 fluence at 32 EPPY iggtaken as 2 timon the 16 EFPY i

surface fluenco.

f n/cm

= 1.94 at 16 pagoSOfr(10 EFPYf=E>lMeV) i EFPY (Ref.

3, ff(32 1.35.

l TABLE 20-groquoyah__ Unit 2 Screening f

Material ART (PTS) d Initial Cr!torion i

b E,c gigg HTgg Marg 1B RTgg Limit D.cscription Forging 05 95 108 10 34 152 270

[83) 95 10 (17) 122 270 Forging 04 104 119

-22 34 131 270 l

Wolda 68 76

-4 56 128 300 3

(144) 164

-4 (28) 188 300 i

a.

All unita in OF.

b.

Chemistry Factor obtained from tables 1 and 2 of 10 CFR 50.61 c.

Valuca in [ l were determined from credible surveillanco data as reported in hoatup and cooldown limit calculations (Ref.

4, page 9).

d.

Unit 2 fluence at 32 EFPY in taken an 2 timos the 16 EFPY 19 2

surface fluence, f

(10 n/cm, E>l MeV) = 0.0644 at 16 EFPY (Ref.

4, pageSOfg ff(32 EFPY) 1.14.

=

h

-3 b

.r-

,-w c - e

.r,,,,w....,

w-

,. v

,w,~

, - +. -, - -. _ _,, -

-r

v. -

-.m, r.,

• 4 4

SI:QUOYAll NUCLEAll PLANT UNITS 1 A!1D 2 11tESSUll1 ZED Till:ltMAL SilOCK SCitEENING CIU TEll1 A 5.

CONCLUDIOND For each of the materials in the vennel boltline regions of both Sequo 'ah Units 1 and 2, the calculated value of the itTyk,g in 1000 than he scroonithj criteria limitn.

No further action u

required.

6.

REFERENCED 1.

10 Cril 50. 61, "Prac+.uro Toughneun Requiremonta for Protection Against Pronourir.ed Thermal Shock Eventu," Fodoral Itogintor, Vol. 56 No. 94, May 15, 1991.

2.

"Fracturo Toughnenn Requirements," liranch Technical Position HTEli 5-2, Chapter 5.3.2 in Sinndard_lley.luwEnn for the Review of Safety Analynin Hoporto for lluclear Power Planta, IRR Edition, NUREG-0800, 1981.

3.

WCAP-12970, "llcatup and Cooldown Limit Curvou for Normal Operation," Sequoyah Nuclear Plant Unit 1, Juno 1991.

4.

WCAP-12971, "lleatup and Cooldown Limit Curven for Normal Operation," Snquoyah Nuclear Plant Unit 2, June 1991.

5.

Codo of redoral Regulationa, 10 CPR 50, Appendix G, "Fracturo Toughnoon Requiremonta," U.S.

Nuclear Itogulatory Comminnion, Washington D C.,

redoral Hogister, Vol. 48 No. 104, May 27, 1983.

6.

Regulatory Guido 1.99, Hovinion 2, " Radiation Embrittlement of Henctor Vonnel Materials," U.S. Nuclear Hogulatory Comminaion, May, 1988.

Prepared by ad

• 1 [, \\ J/i[4i Hoviewed },[ #( [Id fy\\ _6/ /_

Approved by _ M

\\ 8-/8/

K3 i

1 4

b \\O SI:QUOYAll 110CLEAll PIAliT UNITS 1 AlJD 2 - PRESSUl(IZED Till:RMAL Sil0CK SCHEl:llING CRITERI A TABLE 3 SI:QUDYAll UNIT 1 REACTOR VESSEL TOUGilNESS TAllLE (Unirradiated)

CU NI Initial ItTHDT (a)

Material Description (t)

(t)

( F)

Intermediato Shell 0.15 0.86

+40 (Forging 05)

Lower Shell 0.13 0.76

+73 (Forging 04)

Weldo 0.33 0.17

-40 a.

The initial RTHDT valuen for the forgings are meanured values.

TAllLI: 4 SEQUOYAll UllIT 2 IlEACTOR VESSEL TOUGilllESS TAllLE ( Uni rrad i a tet'.)

CU HI Initial RTHDT (a)

Material Description (t)

(t)

(OF)

Intermediate Shell 0.13 0.74 410 (Forging 05)

LcWer Shell 0.14 0.76

-22 (Forging 04)

Weldo 0.13 0.11 4

a.

The initial ItTHDT values for the forgingn are meanured values.

_ S.

L