ML20082S833
| ML20082S833 | |
| 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: |
| References | |
| WCAP-12970, NUDOCS 9109170312 | |
| Download: ML20082S833 (28) | |
Text
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A P P R O'!C D.
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Letter No. N8553 o.i . : July 28, 1991 ftMElst( VALLEY AutH08 iffy SOEP (N) ST P A 1,w. 4 Q HEATUP AND COOLDOWN LIMIT CURVES
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'I= FOR NORMAL OPERATION b $ SEQUOYAH UNIT 1 I5h e 8
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ENU d June 1991 [
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Work Performed Under Shop Order TJAP-N004 I i
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Prepared by Westinghouse Electric Corporation for the Tennessee Valley Authority !
. i Approved by: b' %M T.A.Meyer,Ma[ager .
Structural Reliabil.ity & Plant Life Optimization f i
f WESTINGHOUSE ELECTRIC CORPORATION !
i Nuclear and Advanced Technology Division P.O. Box 355
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j l Pittsburgh, Pennsylvania 15230-0355 .;
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- 1991 Westinghouse Electric Corp.
u 91n9170312 910906 RIMS IT SLE-E o; ADOCK 05000327
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e s TABLE OF CONTENTS Section Title gg 1 INTRODUCTION 1 2 FRACTURE TOUGHNESS PROPERTIES 1 3 CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELATIONSHIPS 2 4 HEATUP AND C00LDOWN LIMIT CURVES 5 5 ADJUSTED REFERENCE TEMPERATURE 6 6 REFERENCES 15 S
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- _ _ _ _ _ _________--_____-_- _ -__ - __ - ---_- - ____- - -_---__- - -_ _ ~
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LIST OF ILLUSTRATIONS !
LLqute 111.le bag 1 Sequoyah Unit 1 Reactor Coolant System Heatup Limitations 11 (Heatup rates up to 20'F/hr) Applicable for the First 16 ;
EFPY (Without Margins for Instrumentation Errors) 2 Sequoyah Unit 1 Reactor Coolant System Heatup Limitations 12 (Heatup rates up to 40*T/hr) Applicable for the First 16 EFPY (Without Margins for Instrumentation Errors) 3 Sequoyah Unit 1 Reactor Coolant System Heatup Limitations 13 (Heatup rates up to 60*F/hr) Applicable for the First 16 ,
EFPY (Without Margins for instrumentation Errors) 4 Sequoyah Unit 1 Reactor Coolant System Cooldown (Cooldown 14 Rates up to 100*F/hr) Limitations Applicable for the First 16 EFPY (Without Margins for Instrumentation Errors)
LIST OF TABLES Igb_lg litle hgg 1 Sequoyah Unit 1 Reactor Vessel Toughness Table ,8 (Unirradiated) <
2 Calculation of Chemistry Factor Basad on Surveillance 9 Capsule Data l 3 Summary of Adjusted Reference Temperature (ART) at 1/4T 9 and 3/4T Location 4 Calculation of Adjusted Reference Temperature
- for 10 Limiting Sequoyah Unit 1 Reactor Vessel Mate, l- ,
Lower Shell forging 04 ii
l O 6 l
- 1. INTRODUCTIM l
Heatup and cooldown limit curves are calculated using the most limiting value l of QTNDT (reference nil-ductility temperature) for the reactor vessel. The ,
most limiting RTNDT of the material in the core region of the reactor vtssel j is determined by usi 9 the preservice reactor vessel material fracture toughness properties and estimating the radiation-induced ARTNDT* i RTNDT 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 ft-lb of impact energy and 35-mil lateral expansion (normal to the !
major working direction) minus 60*F. l RTNDT increases as the material is exposed to fast-neutron radiation. l 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 i time period must be added to the original unirradiated RTNDT. The extent of ;
the shift in RTNDT is enhanced by ceruin chemical elements (such as copper !
and nickel) present in reactor vess A steels. The Nuclear Regulatory '
Commission (NRC) has published a method for predicting radiatiun embrittlement [
in Regulatory Guide 1.99 Rev. 2 (Radiation Embrittlement of Reactor Ve sel I Materials)Ill. 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). !
5
[
- 2. FRACTURE TOUGHNESS PROPERTIES , ,
i i.
The fracture-toughness properties of the ferritic material in the reactur coolant pressure boundary are determined in accordance with the NRC ";gulatory l Standard Review Plan [2]. The pre-irradiation fracture-toughness properties -
of the Secuoyah Unit I reactor vessel are presented in Table 1.
f I
l I
1 i
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- 3. CRITERIA FOR ALLOWABLE PRESSURE-TEMPERATURE RELA 110NSHIPS l
The ASME approach for calculating the allowable limit curves for various heatup and cooldown rates specifies that the total stress intensity factor, K; for j the combined thermal and pressure stresses at any time during heatup or t cooldown cannot be greater than the reference stress intensity factor, KIR' '
for the metal temperature at that time. KIR i s obtained from the reference fracture toughness curve, defined in Appendix G to the ASME Codel33. The i
KIR curve is given by the following equation:
l b
KIR = 25.78 4 1.223 exp (0.0145 (T-RTNDT + 160)) (1) [
where X1R = reference stress intensity factor as a function of the metal j temperature T and toe metal reference nil-ductility temperature l RTkDT !;
Therefore, the governing equation for the heatup-cooldown analysis is defined ;
in Appendix C cf the ASME CodeI33 as follows: !
CKg+ 7 KIT I XIR (2) where <
fi
, i Kjg = stress intensity factor caused by membrane (pressure) stress !
r i
X:T = stress intensity factor caused by the thermal gradients l l !
l K;g = function of temperature relative to the RTNDT of the material h i
! C - 2.0 for level A and Level 8 service limits l l
L C - 1.5 for hydrostatic and leak test conditions during which the ,
reactor core is not critical I t
l l .
2
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5
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y -- - - , , - - -
,.r-, . . ,. , , ---- --- . -- .w , _.. _ , , ,
At any time during the heatup or cooldown transient, KIR 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 factors are obtained and, from these, the allowable pressures are calculated.
For the calculation of the allowable pressure versus coolant temperature during 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 produce tensile stresses at the inside, which increase with increasing cooldown rates. Allowable pressure-temperature relations are generated for both steady-state and finite cooldown rate situations. From these relations, composite limit curves are constructed for each cooldown rate of interest.
The use of the composite curve in the cooldown analysis is necessary because contro; 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 vessel location is at a higher temperature than the fluid adjacent to the vessel 10. This condition, of course, is not true for the steady-state situation. It follows that, at any given reactor coolant temperature, the AT developed during cooldown results in a higher value of KIR 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 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 l
3
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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 l cooldown period. -
l Three separate calculations are required to determine the limit curves for f 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 f pressure. The metal temperature at the crack tip lags the coolant temperature; j therefore, the KIR for the 1/4 T crack during heatup is lower than the KIR f for the 1/4 T crack during steady-state conditions at the same coolant l temperi.ture. During heatup, especially at the end of the trabmnt, co:ditions !
i 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 I 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 !
t pressure-temperature limitations for the case in which a 1/4 T deep outside i surface flaw is assumed. Unlike the situation at the vessel inside surface, i r
the thermal gradients established at the outside surface during heatup pr.oduce l stresses which are tensile in nature and therefore tend to reinforce any
{
pressure stresses present. These thermal stresses are dependent en both the rate of heatup and the time (or coolant temperature) along the heatup ramp, f Since the thermal stresses at the outside are tensile and increase with 5 ,
increasing heatup rates, each heatup rate must be analyzed on an individual j basis. '
i 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 j i
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--- - - - - ----_--- _. - - - --- -_ O
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allewable pressure is taken to be the lesser of the three values taken from the curves under consideration. The use of the composite curve is necessary to set conservative heatup limitations because it is possible for conditions to exist ;
wherein, over the course of the heatup ramp, the controlling condition switches I from the inside to tne outside, and the pressure limit must at all times be '
based on analysis of the most critical criterion.
Finally, the 1983 Amendment to 10CFR50l43 has a rule which addresses the metal temperature of the closure head flange rd vessel flange regions. This rule states that the metal temperature of t'.e closure flange regions must exceed the material RTNDT by at least 120'F for normal operation when the ,
pressure exceeds 20 percent of the preservice hydrostatic test pressure.
Table 1 indicates that the initial RTNDT of -40'F occurs in the closure head flange of Sequoyah Unit 1, so the minimum allowable temperature of this region is 80*F. These limits are shown in Figures 1 and 2 whenever applicable.
- 4. HEATUP AND C00LDOWN LIMIT CURVES Limit curves for normal heatup and cooldown of the primary reactor pressure vessel have been calculated using the methods discussed in Section 3. Figures 1, 2 and 3 contaie the heatup curves for 20, 40 and 60*F/hr, respectively, i figure 4 contains the c,ooldown curves up to 100'F/hr. Figures 1 through 4 are applicable for the first 16 IFPY 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 the limit lines' shown in Figures 1 through 3. This is in addition to other criteria which must be met before the reactor is made critical.
The leak limit curve shown in Figures 1, 2 and 3 represents minimum temperature requirements at the leak test pressure specified by applicable codes [2'33 t
s l
b =
The leak test limit curvt was determined by methods of References 2 and 4.
The criticality limit curve shown in Figures 1, 2 and 3, specifies pressure-temperature limits for core operation to provide additional margin during actual power production as specified Reference 4. The pressure-temperature limits for core operation (except for low power physics 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 at least 40'F higher than the minimum pressure-temperature curve for heatup and cooldown calculated as described in Section 3. The maximum temperature for the inservice hydrostatic test for the .cquoyah Unit I reactor vessel is 327'F. A vertical line at 327'T 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.
Figures 1 through 4 define limits for ensuring prevention of nonductile failure for_the Sequoyah Unit I reactor vessel.
- 5. ADJUSTED REFERENCE TEMPERATURE From Regulatory Guide 1.99 Rev. 2 [1] the adjusted reference temperature (ART) for each material in the beltline is given by the following expression:
ART - Initial RTNDT + ARTNDT + Margin (3)
Initial RT ND7 is the reference temperature for the unirradiated material as defined in paragraph N2331 of Section 111 of the ASME Boiler and Pressure Vessel Code. If measured values of initial RT NDT f r 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 NDT is the mean value of the adjustment in reference temperature caused by irradiation and should be calculated as follows:
l l
6 1
l l
u -
ARTNDT - [CF)f(0.28-0.10 log f) (4)
To calculate ARTHDT 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.
I f(depth X)
- fsurface(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.
CF (*F) is the chemistry factor, obtained fram Reference 1. In addtion, the chemistry factor is also calculated using surveillance capsule data. A sample of this calculation is shown in Table 2.
All materials in the beltline region of Sequoyah Unit I 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 lower shell forging for heatup and cooldown curve 5 applicable up to 16 EFPY. A sample calculation for the RTNDT for 16 EFPY is shown in Table 4.
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I TABLE 1 SEQUOYAH UNIT 1 REACTOR VESSEL TOUGHNESS-TABLE (Unirradiated)
CU NI I-RTNDT (a)
Material Description (%) (%) ('F)
Closure Head Flange (b) -- --
-40 Vessel Flange (b) -- --
-49 Intermediate Shell Forging 0.15 0.86 40 Lower Shell Forging 0.13 0.76 73 Welds 0.33 0.17 -40
- a. The initial RTNDT (I) values for the forgings and welds are measured values.
- b. These items are to be used for considering flange recuirinents for heatup/cooldown curvesI43 f
4
=
l l
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TABLE 2 CALCULATION OF CHEMISTRY FACTOR BASED ON SURVElLLANCE CAPSULE DATA
_ Material Description _C aps ul e fluente FF -PRINpl FF*DRTNDT IfD 2 r Forging 04 (axial) T 0.274 0.647 60 38.825 0.419 ;
l.08 U 1.022 75 76.614 1.044 Forging 04 (tangential) T 0.274 0.647 70 45.296 0.419 U l.08 1.022 130 11E&l1B 1.044 293.532 2.924 i Chemistry Factor 293; 100.372 Weld Metal T 0.274 0.647 140 90.591 0.419 U l.08 1.022 160 163.441 1.044 254.034 1.462 Chemistry factor - 254;34,173,733 , ,
r TABLE 3 FUMMARY OF ADJUSTED REFERENCE TEMPERATURE (ART) AT 1/4T and 3/4T LOCATION 16 EFPY RINDT at Component 1/4T ('F) 3/4T ('F) ,
Intermediate Shell forging 195 162 Lower Shell Forging (195)* (166)*
Wolds (169) (120) s RTNDT numbers within ( ) are based on the chemistry factor calculated using capsule data.
These RTNDT numbers are used to generate heatup and cooldown curves applicable up to 16 EFPY.
9 I
6 .
TABLE 4 CALCULATION OF ADJUSTED REFERENCE TEMPERATURES FOR LlHITING SEQUOYAH UNIT 1 REACTOR VESSEL MAllRIAL --LOWER SHELL FORGING Reaulatory Guide 1.99 - Revision 2 16 EFPY Parameter 1/4 T 3/4 T Chemistry Factor, CF (*F) 95 ;'.00) 95 (100)
Fluence, f (10 19 n/cm2 )(a) 1.168 .4239 Fluence Factor, ff 1.043 .762 ARTNDT - CF x ff ('F) 99 (105) 72 (76)
Initial RTNDT, I (*F) 73 73 Margin, H ('F) (b) 34 (17) 34 (17)
Revision 2 to Regulatory Guide 1.99 '
Adjusted Reference Temperature, 206 (195) 179 (106)
ART - Initial RTNDT + ARTNDT + Margin
- c..**************... ****.......*****....*****.......**. .****....**....
(a) Fluence, f, is based upon fsurf (10 I9 n/cm2 , E>l Mev) - 1.94 at 16 _
EFPY. The Sequoyah reactor vessel wall thickness is 8.608 inches at the beltline region.
(b) Margin is calculated as, M = 2 [ 012 + oA 3 5 The standard deviation -
for the initial RTNOT margin term, 01, is assumed to be O'T since the initial RTNDT is a measured value. The standard deviation for ARTNDT term, O,A is 17'F for the plate, except that oA need not exceed 0.5 times the mean value of ARTNDT* 8, 4 is 8.5'F for the plate (cut in half) when surveillance data is used.
The numbers within ( ) are calculated using surveillance capsule data, i
10
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MATERIAL PROPERTY BASIS LIMITING ART AFTER 16 EFPY: 1/4T, 195'F -
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0 50 100 150 200 250 300 350 400 45u 500( ,
INDIC*TED TEMPERATURE (DEC.F)
Figu e 2. Sequoyah Unit 1 Reactor Coolant System Heatup Limitations (Heat up rates up to 40*F/hr) Applicable for the First 16 EFPY (Without Margins for Instrumentativn Errors) i 12
o -
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Figure 3.
Sequoyah Unit 1 Reactor Coolant System Heatup Limitations { Heat up rates up to 60*F/hr) Applicable for the First 16 EFPY ',Without ;
Margins for Instrumentation Errors) 13
MATERIAL PROPERTY BASIS LIMITING ART AFTER 16 EFPY: 1/4T, 195'F 3/4T, 166*F 2500 ,,m g,_ ; ii i i i ; ,,i; ; ; iii; ; i ; ; ; ;;;
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Figure 4.
Sequoyah Unit 1 Reactor Coolant System Cocidown (Cooldown rates up to 100*F/hr) Limitations Applicable for the First 16 EFPY (Without Margins for Instrumentation Errors) '
14
H. -
o- ,
- 6. REFERENCES 1 Regulatory Guide 1.99, Revision 2. " Radiation-Embrittlement of Reactor Vessel Materials," U S, Nuclear Regulatory Commission, May,1988.
2 " Fracture Toughness Requirements," Branch Technical Position HTED 5-2, Chapter 5,3.2_ in Standard Reviey_ Elta for the Review of Safety Analysis Reports for Nuclear Power Plants, LWR Edition, NUREG-0800,1981, 3 ASME Boiler and Pressure Vessel Code, Section 111, 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, 4 CoJe 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.
4 15
ATTACHMENT I DATA POINTS FOR HEATUP AND C00LDOWN CURVES (Witiiout Margins for Instrumentation Errors)
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