ML20059C217
| ML20059C217 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 10/31/1993 |
| From: | ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY, ASEA BROWN BOVERI, INC. |
| To: | |
| Shared Package | |
| ML20059C200 | List: |
| References | |
| -MECH-ER-13, -MECH-ER-13-R, 0-MECH-ER-013, 0-MECH-ER-013-R00, NUDOCS 9311010059 | |
| Download: ML20059C217 (19) | |
Text
-_ -__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .__ ___- _ ____
d EQUIVALENT MARGINS ASSESSMENT OF LOW CHARPY UPPER-SHELF ENERGY LONGITUDINAL WELDS FOR THE FORT CALHOUN STATION REACTOR VESSEL O-MECH-ER-013, REV. 00 October 1993 h$k kDOhK O O 285 P PDR j; -
TABLE OF CONTENTS Section Description Pane List of Figures 2 1.0 introduction 3 2.0 Evaluation Description 4 3.0 Results and Conclusions 6 4.0 Summary 8 l
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ADD Combustion Engineering Nuclear Sennees 0 MECH D1013, Fbv. 00 1
LIST OF FIGURES i
h Description Paoe 1 J-Integral vs. Crack Depth,100*F/hr CD, Pressure = 2750 psia l Axial Flaw Orientation, Level A, Weld Material, Flaw Growth 9 2 J-integral vs. Crack Depth,100*F/hr CD, Pressure = 2750 psia Axial Flaw Orientation, Level A, Weld Material, Flaw Stability 10 3 J-Integral vs. Time, Steam Line Break Axial Flaw Orientation, Level C, Weld Material 11 l
4 J-integral vs. Crack Depth, Steam Line Break ]
Axial Flaw Orientation, Level C, Weld Material, Flaw Growth 12 5 J-Integral vs. Time, Steam Line Break Axial Flaw Orientation, Level D, Weld Material 13 j I
6 J-Integral vs. Crack Depth, Steam Line Break j Axial Flaw Orientation, Level D, Weld Material, Flaw Growth 14 7 J-Integral vs. Time, Feedwater Line Break Axial Flaw Orientation, Level D, Weld Material 15 8 J-Integral vs. Crack Depth, Feedwater Line Break Axial Flaw Orientation, Level D, Weld Material, Flaw Growth 16 I l
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1.0 INTRODUCTION
ABB/CE, in sponsorship with the Combustion Engineering Owners Group (CEOG), has performed an equivalent margins analysis of low Charpy upper-shelf energy applicable to all CEOG reactor vessel beltline materials. The methods, results, and conclusions have been documented in report CEN-604, Revision 1 (Reference 1) which was previously transmitted to the NRC staff for information (Reference 2).
In summary, the CEOG evaluation addressed the selection of controlling transients, reactor vessel beltline material properties and geometry; along with the analytical evaluation method and assessment criteria from a generic viewpoint. The report is directly applicable to the Omaha Public Power District (OPPD) Fort Calhoun Station reactor vessel.
Recent discussions between OPPD and the NRC staff have identified a concern that the longitudinal beltline welds (Identification No.s 2-410 A, B, C and 3-410 A, B, C) may not, with certain conservative assumptions, exhibit the 10CFR50, Appendix G required 50 ft-lb (or greater) Charpy USE at end-of-life. As described in the generic equivalent margins analysis report, ABB/CE has projected the end-of-life Charpy USE values for the CEOG vessels using utility Generic Letter 92-01 response information and Regulatory Guide 1.99 Revision 2 (Reference 3) and concluded that CEOG vessel welds would not fall below 50 ft-Ib. Consequently, equivalent margins were not demonstrated for weld material.
To address this concern, OPPD contracted ABB/CE to demonstrate equivalent margins for ,
the Fort Calhoun Station beltline axial welds. This report provides a description of the evaluation and the results obtained.
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1 ADB Combustion Engineering Nuclear Servan O MECOEBo13, F*rv. 00 3 i
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2.0 EVALUATION DESCRIPTION The evaluation" details of the Fort Calhoun Station reactor vessel welds are provided predominantly by the CEOG equivalent margins ana!ysis (Reference 1). A brief discussion surrounding the evaluation is provided to identify parameters which remain unchanged and those which were modified. This evaluation of the beltline welds is considered to be a supplement to the original evaluation.
2.1 Transient Definition The transients analyzed to demonstrate equivalent margins of the reactor vessel beltline welds were consistent with those documented in the original evaluation.
Section 4.0 of the generic analysis (Reference 1) provides detailed transient profiles.
2.2 Reactor Vessel Geometry As described by the generic analysis (Reference 1, Section 5.0), CEOG vessels consist of four reactor vessel beltline geometries. The evaluation of the Fort Calhoun Station reactor vessel welds will only consider the nominal dimensions associated with the 140 inch inside diameter vessel. This geometry is directly applicable to the Fort Calhoun Station reactor vessel.
2.3 Material Properties t
The generic analysis (Reference 1) addressed plate, whereas this evaluation is for welds. The material properties necessary to perform an equivalent margins analysis for the weld material include the modulus of elasticity, yield strength, and the J-Integral Resistance Curve (J-R Curve). The modulus of elasticity for the weld is the same as for the plate. The yield strength for the plate is applied to the weld; this is conservative because the irradiated yield strength for the Fort Calhoun '
Station vessel welds will be higher based on the copper content relative to the plate materials (nominally 0.20 percent Cu for welds versus 0.10 percent Cu for plates). The J-R Curve for welds was obtained from Reference 4. The Charpy (CVN) model for Deformation-J was selected to define representative toughness properties for this analysis. This model was selected over the pre-irradisted Charpy USE (Cu-@t) or CVNp models, which use the vessel fluence as a variable.
In order to use the Cu-$t or CVNp models, the change in upper shelf energy due to irradiation is predicted using Regulatory Guide 1.99, Revision 2, and then the J-R curves are calculated by the model. It was concluded that the CVN model would provide a more direct approach by assuming the irradiated USE of interest for the material. Then Jo for various USE values could be compared to the Applied J to determine the lowest Charpy Upper Shelf which would show equivalent margins to Appendix G. This approach avoids possible errors of basing J-R Curve behevlor on estimated values of decreased USE due to irradiation. The weld CVN model (Reference 4) b as follows:
Jo = C1(Aa) 'exp[C3(Aa)]
where: InC1 = -4.12 + 1.48 In CVN - 0.00248 T + 0.153 in Bn C2 = 0.077 + 0.116 in C1 - 0.110 in Bn '
C3 = -0.0812 - 0.0092 In C1 - 0.0789 in Bn C4 = -0.483 Bn = 1.0 ABB Combusbon Engineering Nuclear Services OMECHERO13. Ikw. 00 4 ;
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2.4 Level A/B Service Loading Evaluation Methodology ASME Code Case N-512 (Reference 5) recommends evaluation of an axially oriented flaw when evaluating longitudinal vessel welds. The evaluation method and criteria for the axially oriented flaw is provided by Section 6.0 of Reference 1 i and has been app!!ed to evaluate the Fort Calhoun Station reactor vessel welds.
2.5 Level C/D Service Loading Evaluation Methodology The methodology for evaluating the Level C/D transients for the Fort Calhoun Station longitudinal welds is described predominantly by Section 7.0 of Reference
- 1. Again, it is only necessary to consider an axially oriented flaw.
The methodology was modified to include weld residual stresses. The weld residual stress had a peak magnitude of 8 ksi and was represented by a cosine distribution, tensile at the inner and outer surfaces and compressive at the midwall.
The linear elastic fracture mechanics evaluation methodology is described in detail by References 6 and 7. The J-Integral is estimated from the applied stress intensity factor as described in Reference 1.
Stresses induced due to the presence of the clad have also been incorporated in the methodology. Specifically, these stresses are due to the difference in thermal expansion coefficients of the stainless steel cladding and the weld metal. The stresses resulting from the difference in cladding and weld metal expansion coefficients is conservatively assumed to be negligible.
ABB Combustion Engineering Nuclear Somces O MEOiERo13. Hev. 00 5
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l 3.0 RESULTS AND CONCLUSIONS Level A/B Service Loads To assess the significance of the low Charpy USE weld material, the cooldown transient was analyzed using the recommendation provided by the ASME Code Case as described previously. The two criteria which need to be addressed are flaw growth and flaw stabliity.
The first criterion requires that the applied J-Integral be less than the resistance of the material (Jo) given 0.1 inch of flaw growth. Figure 1 graphically depicts the results obtained for the Fort Calhoun Station reactor vessel geometry. This figure provides the applied J-Integral for the axial flaw as a function of flaw depth and also depicts the weld material Jo for various levels of Charpy USE. The applied J-Integral given 0.1 inch of flaw -
growth was calculated to be 372.0 in-lb/in*. The material exhibiting the minimum Jo was determined to be weld metal with at least 33.2 ft-Ib Charpy USE.
The second criterion requires the flaw to bo stable under Level A/B service loadings.
Figure 2 depicts the applied J-Integral for an axial flaw utilizing the requisite factor of safety, along with material Jo values characterized by Charpy USE values. The criterion requires that the slope of the applied J-Integral be less than the slope of Jo at the point of intersection. Upon review of Figure 2, a weld material exhibiting at least 36 ft-Ib Charpy USE would be required to satisfy the criterion.
Level C Service Loadinos The results of the Level C transient evaluation are provided in Figures 3 and 4. Figure 3 depicts the applied J-Integral at 0.1 inch flaw extension for the axial flaw as a function of time. In addition, the weld material resistance to flaw extension (Jo) representative of various Charpy USE levels is also provided.
The first criterion is that the app!!ed J-Integral be less than Jo given 0.1 inch flaw extension under the defined loading conditions. Review of Figure 3 shows that the applied J-Integral due to an axial flaw is substantially below Jo of the material. Thus, the first criterion addressing growth has been met.
The second criterion is that the flaw be stable under Level C loading conditions. To address flaw stability, the time point which provides the smallest margin between the applied J-Integral and Jo was utilized to assess flaw stability. Figure 4 illustrated the applied J-Integral due to various flaw depths and the Jo associated with the weld material given the vessel temperature at this time point. The slope of the appiled J-Integralis less than the slope of Jo at the point of intersection. Consequently, flaw stability has been demonstrated for values of Charpy USE as low as 20 ft-lb.
Level D Service Loadinas The feedwater line break and steam line break were evaluated for the Fort Calhoun Station reactor vessel welds against Level D criteria. The specific criteria for Level D service loadings are (1) the flaw be stable given the potential of ductile growth and (2) the flaw depth not exceed 75 percent of the vessel thickness, with the remaining ligament being safe from tensile instabliity.
To address both these criteria, the more stringent criteria associated with Level A, B, and 'C service loadings can be demonstrated.
ABB Combustion Enginoonng Nuciew Sondcas CMAECHmo13. Ibv. 00 6
First, Figure 5 depicts the applied J-Integral for a postulated axial flaw due to the steam line break transient. The Jo of the weld material is also provided. Both identities are provided assuming 0.1 inch flaw growth and over the transient duration. As evident from the figure, flaw growth is not a concern since the applied J-Integral is below the material Jo.
Flaw stability was addressed at the time point within the transient duration with the smallest margin between the applied J-Integral and the material Jo. Review of Figure 6 ,
shows that the slope of the applied J-Integral is less than the material Jo. Hence, flaw stability is not a concern for the weld material.
Tensile instabliity is not a concern since the flaw does not extend under the imposed loading conditions.
The results associated with the feedwater llne break are provided by Figure 7 and 8. Again, utilizing the logic previously applied to the steam line break transient, flaw growth and flaw stability are not a concern for the weld material. Tensile instability is again moot due to the lack of crack extension.
L ABB Comboscon Engineenrig Nudear Services O MECHET4013, Fhrv. 00 7
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4.0
SUMMARY
In summary, equivalent margins have been demonstrated for the Fort Calhoun Station i reactor vessel welds under Level A/B service loadings.. Weld material exhibiting at least 36 ft-Ib Charpy USE will provide equivalent margins of safety required by 10CFR50 Appendix G. The Level C and D evaluations have demonstrated that weld materials exhibiting 20 ft-Ib Charpy USE adequately provide the requisite margins of safety required by 10CFR50 Appendix G.
In conclusion, using the guidance provide by ASME Code Case N-512, equivalent margins of safety to ASME Code, Section 111, Appendix G have been demonstrated for weld material as low as 36 ft-Ib Charpy USE for the Fort Calhoun Station reactor vessel.
s ABB Combustion Engineering Nuclear Services O MECRERo13, FW. 00 8
FIGURE 1 .
'J-INTEGRAL VS. CRACK DEPTH 100 F/HR CD, PRESSURE = 2750 psia AXIAL FLAW ORIENTATION, LEVEL A, WELD MATERIAL VESSEL SIZE: NOMINAL ID = 140 In, BELTLINE THICKNESS = 7.125 in FLAW GROWTH AXIAL FLAW O
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J-INTEGRAL VS. CRACK ' DEPTH
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J-INTEGRAL VS. TIME STEAM LINE BREAK AXIAL FLAW ORIENTATION, LEVEL C, WELD MATERIAL VESSEL SIZE: NOMINAL ID = 140 in, CLAD THICKNESS = 0.2188 in BELTLINE THICKNESS = 7.125 In 1500 - - - . .-- _ - _ . . ._ - . _ . . _.-.-. . .
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J-INTEGRAL VS. CRACK- DEPTH ~
STEAM LINE BREAK AXIAL FLAW ORIENTATION, LEVEL C, WELD MATERIAL VESSEL SIZE: NOMINAL ID = 140 in, BELTLINE THICKNESS = 7.125 in FLAW STABILITY 1000 _ - _ _
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FIGURE 5 '
J-INTEGRAL VS. TIME STEAM LINE BREAK AXIAL FLAW ORIENTATION, LEVEL D, WELD MATERIAL l VESSEL SIZE: NOMINAL ID = 140 in, CLAD THICKNESS = 0.2188 in i BELTLINE THICKNESS = 7.125 in i
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FIGURE 6 ..
J-INTEGRAL VS. CRACK DEPTH STEAM LINE BREAK AXIAL FLAW ORIENTATION, LEVEL D, WELD MATERIAL VESSEL SIZE: NOMINAL ID = 140 in. BELTLINE THICKNESS = 7.125 in FLAW STABILITY 1500 _. ._ _ . . _ - . _ . _
__+._. . _ . _ _ _ ._ _. . _ . . . . _ . _ .
1250 _. .. _ _ . . _ . - _ . . . _ . _ ..__ . . . _ . . .
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0 .25 .5 .75 1 1.25 1.5 1.75 2
. CRACK DEPTH, a (in) 5:
' FIGURE 7- .
e .J-INTEGRAL VS. TIME a FEEDWATER'.LINE BREAK -
AXlAL FLAW ORIENTATION, LEVEL D,. WELD MATERIAL VESSEL SIZE: NOMINAL' ID = 140 in, CLAD THICKNESS = 0.2188 in BELTLINE THICKNESS = 7.125 in ' ;
t:
1500 .-=
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-- ,_ . . , . -- ,=r. . - , . . , . . _ . . ...--..w..
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FIGURE 8 -
J-INTEGRAL VS. CRACK DEPTH
! FEEDWATER LINE BREAK AXIAL FLAW ORIENTATION, LEVEL D, WELD MATERIAL VESSEL SIZE: NOMINAL ID = 140 in, BELTLINE THICKNESS = 7.125 in-FLAW STABILITY ,
o 1000 _. n .- .- . --
.-. _. -.-. ~ 4-. - - . - + . - - - . ,----t- - - - . - - -.--
= - - - - " ~ " -- ~ ~ '- - - - - - - - - - ~
900 _ _. '-~------_ _ _ _ _ . _ _cm'-so - - . _.__
n.ee - e. . - - - . - . _...- ---..,%_..,4..,,._. .n.. ....._.,e-, -
4-.. -+---..e--.ea- _-.- --...-e- -.4%..e o
_. ._ . _-h--.-__-.-M_
- ~~ ~~~~ -- ~~ ~~ ~ -~ ~ ~~
n 800 - - .. _ . - - - .. .
a.. . . _, .. . _ _ . _ --- -
l
.. _ . . . _ ... __ _ _ . . . , , - , . -- .. n _ . , _
G,
_ . - - - -. - - . - . . ~ .
, . qvn,4o - . ._
4
_e 700 s
.. _ . .. - _ _ . _ _ . . . _, . - _. .._ .. _ __ . __.~.. .._
~~ - --
r 600 7 ^ ~~ ~-
c - - - - - ---
- F - - -- -- - -- ^- -
vn-so:: .==:
._. - .2-.- =:: :
"^ '
a 500 1 . f.'_
4 l .. . - .
3 .. - _. _ _ _ __ ._ ,
[ 400 ,-
( - - + - -- ~ - -
,3 4 "~
- lCVN=20 -- .
F - ~ ~ ~~
- ~~~~~~"~ -^ " ~~ ~ ~~ ~ "~ ~ ~ ~ ~
300 _" a_2 .-
z.; ._ . .
[m. _._
.,__._..u_n
. . __ _ .p _ __
1 , ~
~ " - ~ ~ ~ ~ ~ ~ ~ ~
200 i u- -.- * - + '-
._._ _ - .+ ;"
-+_ _ _ + ._4
_ _g.q _ . ._ 4._
_.- AXIAL FLAW
_ . . ._4__.. ._
_.:_ _ --..j q_.
~ ~ ~ ~ ~ ~ ~ ~ ~~* ~" " "
~
100 m4
/
4 Z '_..
_. y.
_.7_, _..._..
- g',
a"- ,. .n 4_. L_._
3 _ _y=1._
_ + _ - - m._ q .- p 0 - ' ' ' ~~' ' ' '- ' ' i ' ' ~ "m ~' ~ ' i ' ' ' 'T ' i'r . .
0 . 25 . 5 . 75 1 1.25 1.5 1.75 2 '
CRACK DEPTH, a (in)
G
, i i
% 7 -<9-%
-+.v.tr +~ =,p-w, wt.,e- -e-,-a - , - - w - tme . es-+--t 1r g* STW vrde +ow+v tr w ,i,e--'a w*-d w er e- e __.-__ta'u_= -
r.- wh----u.u- -.--.' .a
REFERENCES
- 1. " Final Evaluation of Low Upper Shelf Energy for Combustion Engineering Nuclear Steam Supply Systems Reactor Pressure Vessels," Report No. CEN-604, Revision 01, September 1993, ABB Combustion Engineering Nuclear Services
- 3. R. Burski to J. Richardson," Final Evaluation of Low Upper Shelf Energy," letter No. CEOG-93-479, September 27,1993
- 3. " Radiation Embrittlement of Reactor Vessel Materials," U.S. NRC Regulatory Guide 1.99, Revision 2, May 1988
- 4. E. D. Eason, J. E. Wright, and E. E. Nelson, "Multivariable Modeling of Pressure Vessel and Piping J-R Data," NUREG/CR-5729 (MCS 910401), Modeling and Computing Services, Newark, CA, May 1991
- 5. " Assessment of Reactor Vessels with Low Upper Shelf Charpy impact Energy Levels,"
ASME Section XI, Code Case N-512, approved date March 11,1993
- 6. " Low Temperature Overpressurization Transient Pressure-Temperature Limit for Determination of Low Temperature Overpressure Protection Setpoints," CE Report No.
CEN-381-P, Combustion Engineering, Inc., December 1988 (CE Proprietary) ;
- 7. " Low Temperature Overpressurization Protection Pressure-Temperature Limit for Methodology Response for USNRC Inquiry," CE Report No. CEN-381-P, Attachment 1-P, Combustion Engineering, Inc., August 1990 (CE Proprietary)
N38 Combuston Oviimeg Nucisar Serwces 044ECHER413. FA 00 17