ML20065D660

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Heatup & Cooldown Limit Curves for Normal Operation Braidwood Unit 2 (Capsule U).
ML20065D660
Person / Time
Site: Braidwood  Constellation icon.png
Issue date: 02/28/1991
From: Meyer T
WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20065D639 List:
References
NUDOCS 9404070173
Download: ML20065D660 (25)
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HEATUP AND 000LOOWN L'IMIT CURVES-FOR. NORMAL'0PERATION. ,

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' -BRAIDWOOD UNIT 2-(CAPSULE U) 1

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February 1991. ,

n Work Performed Under Shop Order BMVP-106 g Prepared by Westinghouse ElectricLCorporation.-

for the Comonwealth!Edisori Company 4

Approved by:

bk T.A.IMeyer Manager ,

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Structural Reliability and. Life Optimization q l . l

WESTINGHOUSE ELECTRIC CORPORATION 2 e ..

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3' CRITERIA"FOR ALLOWABLE PRESSURE-TEMPERATURE. RELATIO y.

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HEATUPLANDjCOOLDOWN LIMIT CURVES ,

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' Braidwood Unit 2 Reactor Coolant' System Heatup:. 11' 1

M ' Limitations (Heat"up rate up to 100*F/hr): Applicable.-

for the First 16 EFPY (Without' Margins!For!!nstrumentatiioit .x

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'2 rates 'up'to 100*F/hr): Limitations Applicable. for' the- ,

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LIST OF TABLES 4

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EAS.t Table Ilth :U:r .

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': - 1 Braidwood Unit 2-Reactor Vessel Tough' ness; Table.

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-(Unitradiated);

J9; 2 Sumary' ofeAdjusted Reference)Temperatura y(ART).~atfl/4T

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HEATUP AND C00LDOWN LIMIT CUR'.

% i. FOR NORMAL OPEllATION 4

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

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. material fracture toughness properties and estimating the is designated as:the' higher of radiation-induced ARTNDT. RTHDT 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 60*F.

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

RTNDT 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 The. Nuclear' (such as copper and nickel) present in reactor vessel steels.

Regulatory Commission (NRC) has published a method for predicting radiation embrittlement in Regulatory Guide 1.99 Rev.-2 (Radiation Embrittlement of Reactor Vessel Materials)Ill. Regulatory Guide.l.99,-

Revision 2 is used for the calculation of RTNDT values at 1/4T andi3 locations (T is the thickness of the vessel at the' beltline' region).

2. FRACTURE TOUGHNESS PROPERTIES, The fracture-toughness properties of the.ferritic material,in'the reactor coolant pressure boundary are' determined in accordance"with the NRC. '

-Regulatory Standard Review Plani 23. The pre'-irradiation fracture-toughness' properties of the Braidwood. Unit 2 reactor vessel are presented in Table 1.

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3. CRITERIA FOR Au.0WABLE PRES 5URE-TEMPER,ATURE REL.10NSHIPS .

The ASME _ approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, Kg, for.

L the combined thermal and pressure stresses.at any time during heatup or cooldown cannot be greater than the reference stress intensity factor, KIR' for the metal temperature at that time. KIR is obtained from the reference -

The fracture toughness curve, defined in Appendix G to the ASME CodeI33.

K IR curve is given by the fol:owing equation:

(1)?

Ki g = 26.78 + 1.223 exp [0.0145 (T-RTNDT + 160)]

where K IR = reference stress intensity factor as a function of-the metal temperature T 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:

(2)'

C Kgg + KIT 5 KIR where Ki g - stress intensity factor caused by. membrane (pressure) stress K IT - stress intensity factor caused by the thermal: gradients -

K IR = function of temperature relative'to the RTNDT of the material

r. 2.0 for Level A and Level.B service' limits C

1.5 for hydrostatic and leak test conditions during whichLthe

" reactor core'is not critical t.

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At any time durins .ie.heatup or'cooldown transien i c 4. .

metal' temperature at the tip of the post'ulated flaw, the appropriate value.

The thermal stresses t RT NDT, and_the reference fracture toughness curve.

resulting from the temperature gradients' through the vessel wall are and then the corresponding (thermal) stress intensity factors, KIT, for the reference flaw are computed.- . From equation 2, the pressure' stress inten f actors are obtained' and, from these, the allowable pressures are calcul For the calculation of the allowable pressure versus coolant temperature cooldown, the reference flaw of- Appendix G to the ASME Code is assu at the inside of the vessel wall. During cooldown,'the controlling locatio the flaw is always at the inside of the wall because the-thermal gradients produce tensile stresses at the inside, which-increase with increasin rates. Allowable pressure-temperature relations .are ~ generated for. both From these relations, steady-state and finite cooldown rate situations.

composite limit curves are constructed for each cooldown rate of interes The use of the composite curve in the cooldown analysis .is necessary beca control of the cooldown procedure is based on the measurement of. reactor coolant temperature,.whereas the limiting pressure is actually depend material temperature at the tip of the assumed flaw.

During cooldown, the 1/4 T vessel location is at a higher temperature This condition, of course, is not true for fluid adjacent to the vessel 10.

It follows that, at any. given reactor coolant the steady-state situation.

temperature, the AT developed during cooldown reruits in a higher v K IR at the 1/4 T location for finirte cooldown rates than for steady operation. Furthermore, if conditions exist'so that the increase in K IR exceeds K IT, the calculated allowable pressure during cooldown will .

greater than the steady-state value.

The above procedures are needed because there is no direct control on-temperature at the 1/4 7 location and, therefore, allowable. pressures may unknowingly be violated if the rate of cooling is decreased at various'.

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intervals along a ;oldown'. ramp. The use of the cs ,,osite curve eliminates

'this.problemandensuresconservativeopelationofthesystemfortheentirel cooldown period.

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

L finite heatup rates. As is done in the cooldown analysis,. allowable pressure-temperature relationships are developed for steady-state' conditions as well'as 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 pressure. The metal temperature at the crack tip lags the coolant temperature;.

therefore, the KIR for the 1/4 T crack during heatup is 16wer than the KIR h

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-state conditions 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.

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-outside surface flaw is assumed. Unlike the situation at the vessel inside _ surface,-

the thermal gradients established at the outside surfa'ce during'heatup produce 9 stresses which are tensile in nature and-therefore tend to reinforce any pressure stresses present. These ~ thermal stresses are dependent on both the' 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.

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

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 4

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" allowable . pressure. is taken to be the' lesser of ths .hree values taken from the -

The use of the composite curve is.necessary to' et curves under consideration.

[ conservative heatup limitations because it is possible for conditions to exist wherein, over the course of the heatup ramp, the controlling condition switches from the.inside tc the.outside, and the pressure limit must at all times be based on analysis of the most critical criterion. .

Finally, the 1983 Amendment to 10CFR50 I43 has a rule which addressee the This metal temperature of the closure head flange and vessel flange regions.

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

Table 1 indicates that the initial RTNDT of 20*F occurs in both the

' vessel flange and the closure head flange of Braidwood Unit 2, so the minimum-allowable' temperature of this region is 140*F. These limits are shown in Figures-1 and 2 whenever applicable.

4. HEATUP AND COOLDOWN LIMIT CURVES Limit curves for normal heatup and cooldown of the primary Reactor Coolant Figure 1 System have been calculated using the methods discussed in Section 3.

is the heatup curve up to 100*F/hr and applicable for the first 16 EFPY Figure 2 is the cooldown without margins for possible instrumentation errors.

curve up to 100*F/hr and applicable for the first 16 EFPY without margins for possible instrumentation error's.

Allowable combinations of temperature and pressure for specific temperature change rates-are below and to the-right of the limit lines shown in Figure and 2. This is in addition to other criteria which must be: met before reactor is made critical.

The leak limit curve shown in Figure 1 represents minimum temperature requirements.at'the 1cak test pressure specified by applicable 5

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by methods of X ,- ~ codes (2,3).- The t 4 test limit curve was determi.

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  • References'2 and 4.

Figures 1 and' 2 define. limits for ensuring prevention of nonductile fail'ure for _.

the Braidwood Unit 2 reactor vessel.

5. ADJUSTED REFERENCE TEMPERATURE From Regulatory Guide 1.99 Rev. 2[Il the adjusted -reference temperature:-(ART) for each material in the beltline is given.by the following- expression:

ART = Initial RTNDT + ARTNDT + Margin ~(3) is the reference temperature for the unirradiated material as=

Initial RTNDT defined in paragraph NB-2331 of Section III of the ASME Boiler and Pressure-Vessel Code. If measured values of initial-RTNDT for the material in question are not available, generic mean values for that class of material may, be used'if there are sufficient test results to establish a'mean and standard.

deviation for the class.

is the mean value of the adjustment in reference temperature ARTNDT caused by irradiation and should be calculated as follows:

.(4)

ART NDT - (CF]f(0.28-0,10 log f)

To calculate ARTNDT at any depth (e.g., at;1/4T or 3/4T),'the following-formula must first be used to attenuate the fluence at the specific depth.

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f(depth X) " fsurface(e .24x) 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.

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},,  : materiall RT'NDT at! :1/4T fand' 3/4f are- summarizsd j in' Table. 2._ . From1Tabl~el2,: _

N M it can de"seen that the' limiting:materialLis.the'circumferentialsweldifor- , , ,

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heatup an'di c'ooldown:cdrves applicable:.up to!!6.EFPY, _ A': sample: calculation- forg

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BRAIDWOOD UNIT 2 REACTOR VESSEL TOUGHNESS TABLE (Unirrad (a) -

Heat No. Cu (%) Ni (%) ('F)

Comoonent

- - 20 Closure head flange (b) 2031-V-1

- - 20 Vessel flange (b) 124P455

.03 .71 Upper Shell 490963/

49C904 1-1

.06 .75

-30 Lower Shell 500102/

50C97 1-1

.04 .67 40 WF562 Circumferential Weld (c)

a. The initial RTNDT (1) values for the forgings a'nd weld .are measured :

values,

b. To be used for considering flange requirements for heatup/cooldown curvesI4}
c. These values from Weld Ce'rtification Test . Report WF562'are higher-tha'n'-

the values of 0.03% Cu and 0.67% Hi used in the' PTS submittal [5),

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TABLE 2

SUMMARY

OF ADJUSTED REFERENCE TEMPERATijRE (ART) AT 1/4T a 16'EFPY AT

'RTNDT r

1/4T (*F1 3/4T (*F1- ,

Comoonent 9 -2i p ' Upper Shell, 490963/49C904 1-1 40 21 Lower Shell, 500102/50C97 1-1 145* -115*

Circumferential Weld

  • These RT NDT numbers used to generate heatup and cooldown curves applicable up to'.16 EFPY_

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-CALCULATION OF ADJUSTED REFERENCE TEMPERATURES FOR LIMIT

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Reaulatory Guide1.99 " Revision-2:

f 16 EFPY

  • 1/4 T 3/4 T-Parameter c 54.00. J 54'. 00 Chemistry Factor, CF-(*F) 0.907 0.327~ [,

Fluence, f (10 19 n/cm2 )(a) 0.693 0.973 ,

Fluence Factor, ff ,

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ARTNDT - CF x ff'(*F) 40.0 '40.0' Initial RTN07,.I ('F) 4

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Margin, M (*F) (b)

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Revision 2 to Regulatory Guide 1.99 145' ?115 ' ,

Adjusted Reference Temperature, <

L ART - Initial;RTNDT + A'RTNDT +. Margin '

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l9 2 Fluence, f, is based upon.fsurf.(10 .n/cm ,.E>1LMev)i--1.51:at 16 ,

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EFPY I73'. The .Braidwood Unit 2 freactor vessel wall thickness itk8.

inches at the beltline region [6),  :

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standard deviation for the initial RTNDT; margin. term ((ag)lis;

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standard deviation for ART NDT (8A )yis128'F forLthe weld,p '

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MATERIAL. PROPERTY e4SI$. .. g CONTROLLING MATERIAL: CIRCUMFERENTIAL WELD

40*F INITIAL RTNOT:

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1/4T,145'F.

3/4T,115'F 2500 E mm a

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Figure 1. Braidwood Unit 2 Reactor Coolant System Heatup Limitations:(Heat u rate up to 100*F/hr) Applicable for the First 16 EFPY (Without Margins For Instrumentation Errors) 11

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MATERIAL PROPERTY BASIS '$

CONTROLLING MATERIAL: CIRCUMFERENTIAL WELO 40*F INITIAL RTNDT:

ART AFTER 16 EFPY: 1/4T,-145'F 3/4T, ll5'F r:  :

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Figure 2. Braidwood Unit 2 Reactor Coolant System Cooldown (Cooldownl rates up.to 100*F/hr) Limitations Applicable for the First 16 EFPY' (Without Margins For Instrumentation Errors) 12

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wg * - ,- . 6. REFERENCES  :%

o

[1] Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor

' Vessel Materials.,1.S.

Nuclear Regulatory Commission,' May,J 1988.

" Fracture Toughness Requirements," Branch Technical Position MTEB S-2,

[2]

Chapter 5.3.2'in Standard Review ~ Plan for'the_ Review of Safety Analys Reports for Nuclear Power Plants : LWR Edition, NUREG-0800,1981.

ASME Boiler and Pressure Vessel Code,Section III,' Division 1 -

[3]

Appendixes, " Rules for Construction of_ Nuclear Power Plant Components, Appendix G, Protection Against Nonductile Failure," pp. 558-563,71986L Edition, American Society of Mechanical- Engineers,. New York,1986.

[4] Code of Federal Regulations, 10CFR50, Appendix G, " Fracture Toughn Requirements," U.S. Nuclear Regulatory Commission, Washington,' O.C.,

Federal Register, Vol. 48 No.104,' Hay 27,1983. T PTS Submittal for Braidwood Unit 1 Reactor Vessel. (Section:5.3, Braidwo

[5]

FSAR, Amendment 47, Commonwealth Edison Company, . April 1986).

MT-SHART-078(89), " Byron Unit 2 ond Braidwood Unit 2: Reactor Vessel He

[6]

and Cooldown Limit Curves for Normal Operation", N. K. Ray, February 198 WCAP-12845, " Analysis _ of Capsule U from the Commonwealth Edison Compa

[7]

Braidwood Unit 2 Reactor Vessel Radiation Surveillance Program",. E.- Te et. al . , February 1991.

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  1. .,.;*.. ATTACHMENT 1- ., .

DATA POINTS FOR HEATUP AND COOLDOWN CURVES (Without Margins for Instrumentation Errors) s a

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6.0 REFERENCES

~

3

1. Regulatory: Guide 1.99, " Radiation' Embrittlement 'of Reactor Vessel  ;

Materials," Revision 2, May .1988. ,

2. - NUREG-0800,' " Pressure-Temperature. Limits," Standard _ Review Plan, ", '

.Section 5.3.2. '

Letter from A. R; Checca (CECO)L o T.!E. Murley (USNRC),i

Subject:

~

t

.3. f men d nt to' Facility;

.Braidwood Station, Units1 1 and 2,'Applicati on or A me '

ij Operating License NPF-72 and NPF-77,-NRC Docket No. 50-456 and!50-457,f '

December 19, 1990._

4. - WCAP-11651, " Analysis ofl Capsule U'from:the Commonwealth Edison! Company >

Byron Unit 1 Reactor Vessel Radiation Surveillance Program," Novemberf 1987.

5. WCAP-12685, " Analysis of Caosule U'f rom the- Commonwealth = Edison CompanyJ .

Braidwood Unit 1, Reactor Vessel Radiation Surveillance' Program,". August! Tic

-1990. +..

y;

'6. . WCAP-12845, " Analysis- of Capsule U f rom tiie' Connonwealth Edison. Company , -

Braidwood Unit 2, Reactor Vessel Radiation Surveillance Program," March--

1991.

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