ML20034B937

Reactor Vessel Heatup & Cooldown Limit Curves for Normal Operation

ML20034B937
Person / Time
Site:	Beaver Valley
Issue date:	03/31/1989
From:	Meyer T, Ray N, Yanichko S WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML20034B933	List: ML20034B932 ML20034B934 ML20034B936 ML20034B937
References
NUDOCS 9005010189
Download: ML20034B937 (25)

Text

.

.i--,.,,'

^v Letter Report NT/SNART-BEAVER VALLEY UNIT 1 REACTOR VESSEL HEATUP AND=

C00LDOWN LIMIT CURVES FOR NORMAL OPERATION March 1989.

O Prepared by: 9.

M N. R. Ray Q

'~

i Verified by: / b

,t,. </ Mad -

5.E.ftnichko Approved by:

7O bA T. A. Meyer,' Manager Structural Materials Engineering-Prepared for Duquesne Light Company 4

Although information contained in this report is nonproprietary, no distribution shall be made outside Westinghouse or its licensees without the customer's approval.

i WESTINGHOUSE ELECTRIC' CORPORATION Nuclear and Advanced Technology Division P.O. Box 2728 Pittsburgh, Pennsylvania 15230-2728 f

387ts/032148:10 9005010189 900419 PlJR ADOCK 05000334 P

PDC 3

'o 1

TABLE OF CONTENTS-Section Title Page

1.0 INTRODUCTION

1 1

2.0 FRACTURE TOUGHNESS PROPERTIES 1-3.0 ADJUSTED REFERENCE TEMPERATURE 2

)

4.0 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS 3

5.0 HEATUP AND C00LDOWN. LIMIT CURVES 6 -.

6.0 REFERENCES

7 APPENDIX A-HEATUP AND COOLDOWN DATA i

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33 l

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LIST OF TABLES

%e Table Title Page 1$

1 Beaver Valley Unit 1 Reactor Vessel Fracture Toughness-9 4

f Properties 2

Calculation of Adjusted Reference Temperatures for Limiting 11 y

Beaver Valley Unit 1 Reactor Vessel Material - Intermediate Shell Plate - B6607-2

+

l 3.

.Q Ik i

Id LIST OF FIGURES E

k' Figure Title Page 1

Fluence Factor for Use in the Expression for ART 12 il i

NDT t

2 Beaver Valley Unit 1 Reactor Coolant System Heatup' 13 3

l Limitations Applicable for the First 9.5 EFPY j[

4 3

Beaver Valley Unit 1 Reactor Coolant System Cooldown 14 i

Limitations Applicable for the First 9.5 EFPY g

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1 HEATUP AND C00LDOWN-LIMIT CURVES FOR NORMAL OPERATION i

1,. 0 INTRODUCTION Heatup and cooldown limit curves are calculated using the most limiting value of RTNDT (reference nil-ductility temperature) for the reactor vessel. The f the material in the core region of the reactor vessel most limiting R.TNDT is determined by using the preservice reactor vessel material fracture tough-ness properties and estimating the radiation-induced ARTNDT.

NDT RT is designated as the higher of either the drop weight nil-ductility transition temperature (NDTT) or the temperature at which the material ~ exhibits at least 50 f t-lb of impact energy and 35-mil l'ateral expansion (normal to the major l

working direction) minus 60*F.

l RT increases as the material is exposed to fast-neutron radiation, j

NDT Therefore, to find the moet limiting RTNDT at any time period in the reactor's life, ART due to the radiation exposure associated with that-NDT time period must be added to the original unirradiated RT The extent of NDT.

the shift in RT is enhanced by certain chemical, elements (such as copper NDT and nickel) present in reactor vessel steels. Tne 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)Ill, i

2.0 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 Plan [3)

The pre-irradiation fracture-toughness properties for the materials in the Beaver Valley Unit I reactor vessel are presented in table 1.

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3.0- ADJUSTED REFERENCE TEMPERATURE From Regulatory Guide 1.99 Rev. 2 (1) the adjusted reference temperature (ART) 1 for each material in the beltline is given by the following expression:-

l l

ART = Initial RTNDT + ARTNDT + Margin.

(1) initial RT is the reference temperature for the unirradiated material as NDT l

defined in paragraph NB-2331 of Section III of the ASME Boiler and Pressure l

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.

ART is the mean value of the adjustment in reference temperature caused i

NDT by irradiation and should be calculated as follows:

ND7=(CF)f(0.28-0.10 log f) = (CF) (ff)

'(2)

ART The value, "f", used -in equation (2) is the, calculated value of the neutron fluence at the location in the vessel at the location of the postulated defect,n/cm2 (E > 1 MeV) divided by 1019. 'The fluence factor, "ff" is-shown in figure 1.

j 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, f(depth X)

• Isurface

• 4^)

(3)

I where x (in inches) is the depth into the vessel wall measured from the vessel inner (wetted) surface. The attenuated fluence is-then used in equation (2) to calculate ART at the specific depth.

4 NDT CF ('F) is the chemistry factor, obtained from reference 1 for the beltline region materials of the Beaver Valley Unit I reactor pressure vessel. The

+

limiting material was found to be the intermediate shell plate B6607-2 for Beaver Valley Unit 1 for 9.5 EFPY, The calculation of ART for this limiting material is shown in' table 2.

The ART values at 1/4T and 3/4T locations will

'be used to develop the reactor pressure vessel heatup and-cooldown curves as described in the following sections.

4.0 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS The ASME approach for calculating the allowable limit' curves for various heatup and cooldown rates specifies that the total stress intensity factor,

~

K, for the combined thermal and pressure stresses at~ any time during heatup g

or cooldown cannot be greater than the reference stress intensity factor, KIR, for the metal temperature at that time. K is obtained from the IR reference fracture toughness-curve, defined in Appendix-G to the ASME CodeI43 The KIR curve is given by the following ~ equation:

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

(4) where l

KIR = reference stress intensity factor as a function of the metal temperature T and the metal reference nil-ductility temperature'RT NDT Therefore, the governing equation for the heatup-cooldown analysis is defined k

I43 in Appendix G of the ASME Code as follows-CKIM.* EIT

• EIR (5)

• NOTE: RTNDT as used in the ASME Code [4] is in fact the adjusted reference 3

temperature (ART) as defined in NRC Regulatory Guide 1.99, Rev. 2 [1]

and calculated in section 3.0.

mn.mmio 3

where i

Kgg = stress intensity factor caused by membrane (pressure) stress t

KIT = stress intensity factor caused by the thermal gradients KIR = function of temperature relative to the RTNDT off the mater -1:

l s

l C

= 2.0 for Level A and Level B service limits I

C

= 1.5 for hydrostatic and leak test conditions during which the reactor L

core is not critical' l

is: determined by

[

l At any time during the heatup or cooldown transient, KIR 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 calcu-latedandthenthecorresponding(thermal).stressintensityfactors,K IT' for the reference flaw are computed. -From equation (5), the pressure stress intensity factors are rhtained and, from these, the allowable pressures are calculated.

1 For the calculation of the allowable pressure versus coolant temperature during cooldown, the reference flaw of Appendix G to the ASME Code (4) is assumed to exist at the inside of the vessel wall. - During cooldown, the l

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 i

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, i

b 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 assume,d flaw.

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During cooldown, the 1/4 T vessel location is at a higher temperature than the fluid adjacent to the vesse1~ inside surface.

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

It follows th:t, at any given reactor coolant temperature, the AT developed during cooldown results in a higher value of XIR at the 1/4 T location for finite cooldown rates than for steady-state operation. Furthermore, if conditions exist so that the increase i

in KIR exceeds KIT, the 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 1/4 T location and, 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 ~

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

Three separate calculations are required to determine the limit curves for 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 K f r the 1/4 T crack during heatup is lower IR than the K for the 1/4 T crack during steady-state conditions at the same IR coolant temperature. During I;eatup, especia11y'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, ar.d the pressure-temperature curve based on steady-state conditions no longer represents ~a lower bound of all l

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

l Therefore, both cases have to be analyzed in order to ensure that at any l

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 iri which a 1/4 T deep outside nn.am" "

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i surface flaw is assumed. Unlike tha situation at the vessel inside surface, j

the thermal gradients established at the o9tside surface during heatup produce i

(

stresses which are tensile in nature and therefore tend to reinforce any pressure stresses present. Theta 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 l

basis.

Following the generation of pressure-temperature curves for both the steady-state and finite heatup rate situations, the final limit curves tre pr.oduced I

by constructing a composite curve based on a point-by-point comparison of the l

steady-state and finite heatup rate data.

At any given temperature, the i

i 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 i

to set 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'to the outside, and the pressure limit must at all times be based on analysis of the most critical criterion.

N Finally, the 1983 Amendment to 10CFR50 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 nust exceed the material RT by at least 120'F for normal operation when the NDT pressure exceeds 20 percent of the preservice hydrostatic test pressure.

5.0 HEATVP AND C00LDOWN LIMIT CURVES Limit curves for normal heatup and cooldown of the primary Reactor Coolant System have been calculated using the methods described in section 4.0, and the Westinghouse procedure of reference 6.

l l

Allowable combinations of temperature and pressure for specific temperature change rates are below and to the right of the limit lines shown in figures 2 m m.,m ie 6

P e

e r

,,i.-,

.ey,

... - -. + -... - - -2

.h l

t l

and 3 for 9.5 EFPY. This is in addition to other criteria which must be met I'

before the reactor is made critical.

f i

i The leak limit curve shown in figure 2 represents the minimum temperature r'equirements at the leak test pressure specified by applicable codesI3'43 The Teak test limit curves were determined by the methods of references 3 and 5.

Finally, table 1 indicates that the limiting flange RTNDT of 60'F occurs in the closure head and vessel flange so the minimum allowable temperature of i

this region is 180'T per reference 5.

These limits are shown in figure 3.

l l

l Figures 2 and 3 define the limits for ensuring prevention of nonductile failure for the Beaver Valley Unit 1 Primary Reactor Coolant System.

6.0 REFERENCES

1.

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

2.

Response to V. S. Nuclear Regulatory Comission Generic Letter 88-11 for the Beaver Valley Unit 1. MT-SMART-209(88), November 1988, f

3.

" Fracture Toughness Requirements," Branch Technical Position MTEB 5-2, i

Chapter 5.3.2 in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition,' NUREG-0800,1981.

4.

ASME Boiler and Pressure Vessel Code, Section Ill, Division 1 -

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

5.

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

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

MN4321W 10 y

5

-r 4

6.

' Procedure for Developing Heatup and Cooldown Curves," Westinghouse Electric Corporation, Generation Technology Systems Division Procedure GTSD-A-1.12 (Rev. 0), July 13,1988.

7.

? Beaver Valley Unit 1 Reactor Vessel Fluence and RTPTS evaluations" WCAP-ID829 Revision 1.

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TAOLE 1 BEAVER VALLEY UNIT 1 REACTOR VESSEL TOUGHNESS PROPERTIES (UNIRRAoIATED)

Component Heat No.

Code No.

meterial i Cu M1

'Not "Tuoi Upper snet r Enerar (Ft-1tt)

(%)

(%)

(*F)

(*F) usso Musso Closure Head C6213-18 86610 A5338 C1. 1

.15

.54

-40 0*

121 Dome Closure Head A5518-2 98611 A5338 C1. 1

.14

.60

-20

-20*

131 S'9-Closure Head ZV3758 96601-1 A508 C1. 2

.08

.65 60*

60*

>100 Flange Vessel Flange ZV3661 96603 A508 C1. 2 12

.71 60*

60*

166 Inlet Nozzle 9-5443 56008-1 A508 C1. 2 10

.54 60*

60*

82.5 i

Inlet Mozzle 9-5460 96608-2 A506 C1. 2

.10

.83 60*

60*

94 Inlet Nozzle 9-5712 96606-3 A508 C1. 2

.06

.80 60*

60*

97 L

Outlet Nozzle 9-5415 86603-1 A506 C1. 2

.79 60*

60*

97 to Outlet Nozzle 9-5415 96603-2 A508 C1. 2

.79 60*

60*

112.5 Outlet Nozzle 9-5444 86003-3 A508 C3. 2

.09 78 60*

60*

103 Upper Shell 123v339 96604 A508 C1. 2

.67 40 40*

155 Inter. Shell C4381-2 96607-2 A5336 C1. 1

.14

.62

-10 73 123 82.5 Inter.'Shell C4381-1 86607-1 A5338 C1. 1 14

.62

-10 43 128.5 90 i

Lower Shell C6317-1 86903-1 A5338 C1. 1

.20

.54

-50 27 134 80 Lower Shell C6293-2 87203-2 A5338 C1. 1

.14

.57

-20 20 129.5 83.5 Trans. Ring 123v223 86402 A508 C1. 2 30 30*

143 Bottom Hood C4423-3 86618 A5338 C1. 1 13

.55

-30

-29*

124 Seg.

Bottom Head C4462-1 86619 A533B C1. 1

.13

.55

-50

-33*

125.5 Inter. Shell (W1re 305424 and SAW

.28'**.63

-56**

Long. Weld Linde 1092 Flum) 3675s-032189:10

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TABLE 1 (cont.)

BEAVER VALLEY UNIT 1 REACTOR VESSEL TOUGHMESS DATA (UNIRRADIATED)

~

Component Heet No.

Code No.

Meterla1 Type Cu N1 NOT T800T Upper Shelf Enectar (Ft-Ib)

(%)

(%)

(*F)

(*F)

Impo puussp Inter. to Lower (Wire 90136 and SAW

.29'".07

-56**

'r Shell G1rth Linde 0091 Flux)

Lower She11 (Wire 305414 and SAW

. 34* ".61

-56**

Weld (Linde 1092 F1UK)

Weld HAZ

-40

-40 136.5 I

Estimated per MUREG-0000 MRC Standard Review Plan Branch Technical Postttori NTES 5-2 Generic IIean Value Estleeted per 10 CFR 50.61 PTS Rule mfd - Mejor Working Otroction peefD - Norsaal to mejor Working Direction s

• " From Reference 7.

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I TABLE 2 CALCULATION OF ADJUSTED REFERENCE TEMPERATURES FOR LIMITING l

BEAVER VALLEY UNIT 1 REACTOR VESSEL MATERIAL -

INTERMEDIATESHELLPLATEB6607-2[2]

i

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Regulatory Guide 1.99 - Revision 2 I

9.5 EFPY Parameter 1/4 I

1. 4 T t

100.5 100.5 ChemistryFactor,CF(*F))

18 n/cm)(a 0.81 0.32

[

2 Fluence,f(10 Fluence Factor, ff 0.94 0.69 l

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • s****************************

ARTNDT=CFxff('F) 95 69 Initial RTNDT,I('F) 73 73 Margin,M('F) 34 34

• • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • • 2.***********************

Revision 2 to Regulatory Guide 1.99 AdjustedR'eferenceTemperature, 202 176 f

ART = Initial RTNDT + ARTNDT + N""9i" 18 2

l (a) Fluence, f, is based upon fsurf(10 n/cm, E>l Nev) = 1.30 at 9.5 EFPY at inner surface. The Beaver Valley Unit I reactor vessel wall thickness j

is 7.875 inches at the beltline region.

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CONTROLLING MATERIAL:

INTERNEDIATE SHELL PLATE B6607-2 RT AFTER 9.5 EFPY:

1/4T 202'F NDT 3/4T,176'F CURVES APPLICABLE FOR HEATUP RATES UP TO 60'F/HR FOR THE SERVICE PERIOD UP TO 9.5 EFPY. DOESNOTCONTAINMARGINFORPdSSIBLEINSTRUNENTERRORS.

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_ the Service 500 Period Up Te 9.6 EFPY 250 Acceptable a - --

opgg 50

$60 sul 200 50 300 35 " 'dd " ase 300 essicatto truptaatung (etc.r)

Figure 2.

Beaver Valley Unit 1 Reactor Coolant System Heatup Limitations Applicable for the First 9.5 EFPY

-i i.

33

'l

'ATERIAL PROPERTY BA515 N

CONTROLLING MATERIAL:

INTERMEDIATE SHELL PLATE B 6607-2 RT AFTER 9.5 EFPY:

1/4T. 202'F NOT 3/4T. 176'F i

CURVES APPLICABLE FOR COOLDOWN RATES UP TO 100 *F/HR FOR.THE SERVICE PERIOD UP TO 9.5 EFPY.

DOES NOT CONTAIN MARGIN FOR POSSIBLE INSTRUMENT ERRORS.

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Beaver Valley Unit 1 Reactor Coolant System Cooldown Limitations Applicable for the First 9.5 EFPY i

14

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