ML20039F864
| ML20039F864 | |
| Person / Time | |
|---|---|
| Site: | Palisades, Maine Yankee  |
| Issue date: | 12/31/1981 |
| From: | ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY |
| To: | |
| Shared Package | |
| ML13308A045 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.K.2.13, TASK-TM CEN-189-APP-C, NUDOCS 8201130487 | |
| Download: ML20039F864 (29) | |
Text
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EVALUATION OF PRESSURIZED THERMAL SH0CK EFFECTS DUE TO SMAL'. BREAK LOCA'S WITH LOSS OF FEEDWATER FOR THE MAINE YANKEE REACTOR VESSEL Prepared for s
- MAINE YANKEE ATOMIC POWER COMPANY NUCLEAR OWER S'.
TEMS DIVISION PI - ' POWER Bimi d SYSTEMS COMBUSTION ENGINEERING INC 8 2 0 3, n o ygy l
LEGAL NOTICE THIS REPORT WAS PREPARED AS AN ACCOUNT OF WORK SPONSORED BY COMBUSTION ENGINEERING, INC. NEITHER COMBUSTION ENGINEERING NOR ANY PERSON ACTING ON ITS BEHALF:
A.
MAKES ANY WARRANTY OR REPRESENTATION, EXPRESS OR IMPLIED INCLUDING THE WARRANTIES OF FITNESS FOR A PARTICULAR PURPOSE OR MERCHANTABILITY, WITH RESPECT TO THE ACCURACY, COMPLETENESS, OR USEFULNESS OF THE INFORMATION CONTAINED IN THIS REPORT, OR THAT THE USE OF ANY INFORMATION, APPARATUS, METHOD, OR PROCESS DISCLOSED IN THIS REPORT MAY NOT INFRINGE PRIVATELY OWNED RIGHTS;OR B. ASSUMES ANY LIABILITIES WITH RESPECT TO THE USE OF, OR FOR DAMAGES RESULTING FROM THE USE OF, ANY INFORMATION, APPARATUS, METHOD OR PROCESS DISCLOSED IN THIS REPORT.
k
/
ABSTRACT This Appendix to CEN-189 provides the plant-specific.
evaluation of pressurized thermal shock effects due to small ' break LOCA's with extended loss of feedwater for the Maine Yankee reactor vessel.
It'is ~ concluded that crack initiation w'ould not occur for the' transients considered for more than 32 effective full power-years, which is assumed
' to represent full plant life.
f
CEN-189-Appendix C TABLE OF CONTENTS SECTION
-TITLE PAGE ABSTRACT C1.
PURPOSE Cl C2.
SCOPE C1 C3.
INTRODUCTION Cl C4.
THERMAL HYDRAULIC ANALYSES Cl CS, FLUENCE DISTRI UTIONS C2 C6.
MATERIAL PROPERTIES C11 C7.
VESSEL INTEGRITY EVALUATIONS C17 C8.
CONCLUSIONS C24 i
e ti
Cl.0 PURPOSE This Appendix provides the plant-specific evaluation of pressurized thermal shock effects of the SB LOCA + LOFW transients presented in the main body of the CEN-189 report for the Maine Yankee reactor vessel.
C2.0 SCOPE The scope of this Appendix is limited to the evaluation of the SB LOCA +
LOFW transients presented in CEN-189, as applied to the Maine Yankee reactor vessel.
Other C-E NSSS reactor vessels are reported in separate Appendices.
C
3.0 INTRODUCTION
This Appendix to CEN-189 was prepared by C-E for Maine Yankee Atomic Power for their use in responding to Item II.K.2.13 of NUREG-0737 for the Maine Yankee rcactor vessel.
This Appendix is intended to be a companion to the CEN-189 report.
The transients evaluated in this Appendix are those reported in Chapter 4.0 of the main report. Chapter C5 of this Appendix reports the plant-specific fluence distributions developed as described in Chapter 5.0 of the main report. Chapter C6 reports the plant-specific material properties and change of properties due to irradiation, based on the methods of Chapter 6.0 of the report. Chapter C7 reports the results of comparing the fracture mechanics results of Chapter 7.0 of the report, to the material properties discussed in Chapter C6.
C4.0 THERMAL HYORAULIC ANALYSES The pressure-temperature transients used to perform the plant-specific vessel evaluation reported in this Appendix are those reported in Chapter 4.0 of CEN-189. As discussed in the body of the report, there are several plant parameter conservatisms included in the analyses to develop these transients due to the reference plant approach used which could be eliminated by perfoming more detailed plant-specific themal-hydraulic system analyses. Removal of these available conser-vatisms by additional analyses was not performed due to the favorable conclusion achieved.
Cl
C5. Maine Yankee Fluence Distribution Maine Yankee provided the fluence data used in the analysis of the Maine Yankee reactor vessel. The data as transmitted by Maine Yankee are shown in Tables C5-1 through C5-4 plus Figure C5-1.
Table C5-1 displays the cumulative energy generation and estimates the number of Effective Full Power Years (EFPY) at 2630 Megawatts-thermal (Mwt) to be 5.904 as of December 31, 1981.
The azimuthal flux distribution is given in Tabie C5-2.
The 00 reference point of Table C-2 is shown in Figure C5-1.
The conversion of the flux values in Table C5-2 to fluence as of December 31, 1981 was obtained by using the normalization factor of 1.51 defined by Maine Yankee and the 8
time in 5.904 EFPY (1.863 x10 seconds). The peak value of the fast neutron fluence was then determined to be 5.00 x 1018 n/cm2 as of December 31, 1981. The slight depression at -70 is due to the presence of the surveillance capsule in the model.
The axial flux distribution is shown in Table C-3.
The axial distribution at a depth equal to 0.0 was used. The radial flux distribution is given in Table C-4.
Continuous curves were fit to the tabular data for the azimuthal, axial and radial fluence distributions as shown in Figures C-2, C-3 and C-4, respectively. The uncertainty in the peak flux value was estimated as +10%, -35% by Maine Yankee.
The fluence distributions were applied as described in the following sections of this report.
C2 ham.-
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1 Table C5-1 Maine Yankee Cycles 1,1 A, 2-6 Effective Full Power Years of Operation to December 31, 1981
~
Cycle Cycle Cycle Cumul a tive Exposure Loa ding Energy EFPY Cycle (MWD /MTU)
(MTU)
(MWD) at 2630 MWt 1
10367 81.540 845,325 0.880
'83.110 373,995 1.269 1A 4500 2
17395 80.885 1,406,995 2.734 a
3 11075 83.065 919,945 3.692 4
10496 81.843 859,024 4.586**
5 10796 83.034 896,435 5.519 6
4500*
82.248 370,116 5.904 Total-5,671,335
- es e ima t ed to 12/31/81
- surveillance capsule 263 removed i
C3
or
/
TABLE C5-2 AZIMUTHAL DISTRIBUTION OF FAST NEUTRON FLUX (E>1.0 Mev) AT THE If4NER RADIUS OF THE MAINE YANKEE REACTOR PRESSURE VESSEL Azimuthal Neutron Flux I
Angle (Des.)
(n/cm Sec) 10 0.0 1.78 x 1010 1.74 x 10 1.0 10 2.0 1.69 x 10 10 3.0 1.67 x 10 10 4.0 1.72 x 10 10 5.0 1.73 x 10 10 6.0 1.72 x 10 10 9.0 1.56 x 10 10 12.0 1.30 x 10 10 l
15.0 1.05 x 10 9
18.0 8.96 x 10 9
21.0 8.88 x 10 9
24.0 9.33 x 10 -
9 l
9.74 x 10 27.0 9
9.86 x 10 30.0 9
33.0 9.50 x 10 9
8.53 x 10 36.0 9
39.0 7.90 x 10 9
42.0 7.48 x 10 9
45.0 7.37 x 10 Note: 1)0.0 is referenced to the perpendicular.to the core shroud.
- 2) Calculations were based on burnup averaged core power distributions for Cycles 1 through 6.
C4
OUTLET 180" NOZZLE
(--'-
m-Maine' Yankee
)-
Orientation Relative to Core Vessel Core Shroud and Capsule 263 INLET INLET N0ZZLE r-N0ZZLE CORE 9
f
/
9 SilROUD
\\
{
1
\\
/
/
\\
t
\\
Perpendicular
)
to Core Shroud h
0 260
~ ~ ~ ~ ~ ~ ~ ~ ~
90 l
{
.C O
~
2700
!__l Angular Location of p
Capsule 263 i
,e Capsule 263 -
(
Distance from core center to
)
5 center -line of the vessel g
g wall assembly tube (as-built)
\\
/ OUTLET
\\
OUTLET
/
N0ZZLE Top of tube 85-29/64 in NOZZLE Bottom of tube 85-3/8 in L_._ _. i INLET 0
N0ZZLE FIGURE CS-1 r....--.,,,,
1 TABLE C5-3 t.
j
' RELATIVE AXIAL DISTRIBUTION OF FAST NEUTRON FLUX (E> 1.0 Mev) f WITHIN THE MAINE YANKEE REACTOR PRESSURE VESSEL Relative Flux r<eight Above Midplane (cm)
Depth = 0.0 Depth = 6.0 Depth = 15.0 Depth = 21.0 J
0.0 1.000 1.000 1.000 1.000 10.0 1.000 1.000 1.000 1.000 20.0 1.000 1.000 1.000 1.000 30.0 1.000 1.000 1.000 1.000 40.0 1.000 1.000 1.000 1.000 50.0 0.999 0.999 0.999 0.999 60.0 0.997 0.997 0.997 0.997 70,0 0.994 0.994 0.994 0.994 80.0 0.985 0.985 0.985 0.985 90.0 0.975 0.975 0.975 0.975 100.0 0.926 0.926 0.926 0.926 110.0 0.863 0.863 0.863 0.863 120.0 0.792 0.792 0.792 0.792 130.0 0.704
" 0.704 0.704 0.704 140.0 0.600 0.600 0.600 0.600 150.0 0.463 0.463 0.463 0.463 160.0 0.324 0.324 0.332 0.354 170.0 0.210 0.210 0.223 0.261 175.0 0.166 0.166 0.178 0.221 180.0 0.127 0.127 0.140 0.189 185.0 0.0960 0.0960 0.109 0.160 190.0 0.0700 0.0700 0.0831 0.136 195.0 0.0506 0.0510 0.0628 0.114 200.0 0.0350 0.0371 0.0468 0.0955 205.0 0.0243 0.0269 0.0346 0.0820 210.0 0.0158 0.0181 0.0251 0.0705 215.0 0.0100 0.0121 0.0184 0.0615 220.0 0.00615 0.00765 0.0132 0.0536 225.D 0.00368 0.00475 0.00960 0.0471 230.0 0.00221 0.00288 0.00695
- 0.0423 235.0 0.00129 0.00188 0.00561 0.0387
~4 240.0 7.55 x 10 0.00122 0.00461 0.0343
~4 8.11 x 10'4 0.00380 0.0300 245.0 4.58 x 10 2.86 x 10'4 5.46 x 10' O.00317 0.0274 250.0
-4 1.86 x 10-4 3.72 x 10 0.00266 3.0247 255.0
-4
-4 260.0 1.27 x 10 2.67 x 10 0.00226 0.0221
- o k6 U.00193 u.uzou 8.80 x 10 "
1.98 x
' " ^
TABLE C5-4 RELATIVE RADIAL DISTRIBUTION OF FAST NEUTRON FLUX (E>1.0 Mev. WITHIN THE MAINE YANKEE REACTOR PRESSURE VESSEL Depth Into Vessel (cm)
Relative Flux 0.00 1.000 1.00 0.934 2.00 0.850 3.00 0.754 4.00 0.668 5.00 0.587 6.00 0.513 7.00 0.448 8.00 0.389 9.00 0.337 10.00 0.291 11.00 0.253 12.00 0.220 13.00
-0.190 -
14.00 0.164 15.00 0.142 16.00 0.123 17.00 0.106 18.00 0.0908 19.00 0.0770 20.00 O.0650 21.00 0.0541 21.97 0.0455 1
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75 1R 125 19 175 292 225 22 25 E
Distance From Core M.P. - CM f(yREC5-3
MAINEYENKEERADIALFLUENCEVARIAll0N 10 R I E D FAST FL13 IES l
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Distance From U.C. Interface - CM FIGURE C5-4 O
C10
APPENDIX'C-MAINE YANKEE-i C.6 MATERIAL PROPERTIES The methods used to develop and evaluate the materials for the Maine Yankee _ reactor vessel are described in' Section 6.0 in-the main body of the report. The chemistry data (nickel,- copper, and phosphorus centent) and intial (pre-irradiation) toughness properties of the reactor vessel shell course plates and welds are-summarized in Table C6-1.
In cases where the chemistry exceeded the Regulatory Guide 1.99 prediction limits (0.35% and 0.012% P),
. those upper limit values were used in the reference temperature shift calcu-
- lations.
lIn cases where the weld metal nickel content was not determined, it was :
conservatively estimated using information on the type of wire -(eg, high MnMo versus MnMoNi-wiro) or the weld process- (inclusion of Ni-200 wire-during weld deposition). For the Maine Yankee weldments, the weld inspection records and welding certification reports indicated that all the welds could be expected
~
to contain high nickel (greater than 0.30 w/o), so the nickel content was conservatively estimated to be 0.99 w/o as indicated in Table.C6-1.
The toughness properties given in Table C6-1 are the drop weight NDIT
]
. (if determined) and the initial reference' temperature, RTNDT. For the plate materials,'the RTNDT was determined using transversely oriented Charpy impact 4
specimens or by converting longitudinal impact data using Branch Technical Position MTEB 5-2*.
For the weld materials, the RTNDT was estimated using i
the-weld qualification. test results benchmarked to the surveillance weld for
. the vessel as discussed in Section 6.0 and described below.
The individual weld qualification test results. (three Charpy impact specimens tested at. +10F) are listed in Table C6-2.
Each weld which 4
exhibited an average Charpy energy of 49 ft-lb or' greater (the average Charpy energy for the surveillance weld at 10F) was considered to be at least as tough as the surveillance weld; i.e., that weld seam ET.DT w"*~
-30F or less. For those weld. qualification test results exhibiting an
- " Fracture Toughness Requirements for Older Plants, U.S. Atomic Energy Commission, Regulatory Standard Review Plan.
C$i~*
average Charpy energy less than 49 f t-lb, the RTNDT "**
I ""
amount equivalent to the temperature difference between the average Charpy energy transition curve for the surveillance weld and the average Charpy energy for the vessel weld test results. In effect, the temperature at which 50 f t-lb or better exists was determined, and the RT as establisb.ed at a temperature 60F below tb.at value.
J NDT A 9 nap" of the cylindrical portion of the Maine Yankee reactor vessel is given in Figure C6-1.
It shows the locations cf the plates and welds listed in Table C6-1 and their corresponding values of initial RTNDT (F) located within a rectangle on the Figure.
RT NDT vertical weld seams (designated 1-203, 2-203, and 3-203) are shown at a single seam but apply to all three vertical seams in a given shell course.
Included in the Figure are the locations of the inlet and outlet nozzles, the core midplane, and the extremities of the active core.
Figure C6-2 is a map of adjusted FT val es f r important NDT locations at the inner surface of the Maine Yankee vessel predicted 13r December 31, 1981. The predictions are based on the best estimate neutron fluence, 0.499 x 10 n/cm ( E>lMeV), (corresponding to 5.904 effective full power years at peak flux location on the inside surface of the reacter vessel), the initial RT a
e pper, phosphorus, and NDT nickel contents given in Table C6-1, and the normalizd neutron flux profiles given in Section C.S.
The values of adjusted RT NDT RT,g7 plus predicted shift) are located in rectangles adjacent to the plate and weld designations. The FT
- ** EE I NDT surface of the vessel in the region indicated by a circle. The circled 4
regions generally represent areas of peak neutron flux for a given weld seam or plate.
TABLE C6-1 MAINE YANKEE TEACTTIR VESSEL MATERIALS Product Material Drop Weight Initial Chemical Content (%)
Form Identification NDTT ( F)
RTNDT ( F)
Plate D-8405-1
-10 Da 0.51 0.17 0.010 b
Plate.
D-8405-2 0
40a 0.54 0.17 0.010 b
P1 ate D-8405-3 0
20 a 0.58 0.17 0.011 Plate D-8406-1
-10 c
-10 c 0.59 0.15 0.01 3 Plate 0-8406-2
-20 Oa 0.56 0.17 0.009 Plate D-8406-3
-30 Da 0.62 0.12 0.010 Plate D-8407-1
-20
-20 a 0.62 0.24 0.008 Plate D-8407-2
-20 2a
- 0.62 0.23 0.007 Plate D-8407-3
-20 Oa 0.65 0.13 0.007 8
Weld 1-203 A,B,&C N/A
-50d 0.93 0,07 0,009 I
Weld 2-203 A,B,&C N/A
-50d 0.99 0.35 9 0.0129 I
Weld 3-203 A,B,&C N/A
-40 d 0.99 0.22 0.015 f
Weld 8-203 N/A
-25d 0.99 0.359 0.0129 c
c h
h Weld 9-203
-30
-30 c 0.78 0.35 0.012 O
w N/A Not Available a
Determined using Branch Technical Positinn MTEB 5-2 b
Estimated based on average for Maine Yankee plates havinq reported analyses c
Surveillance program data d
Estimated (see text and Table C6-2) e Upper bound for coated electrodes (see Table 6-3, Main Report) f Estimated Ni content (high Nickel type wire or weld process) 9 Regulatory Guide 1.99 upper bound prediction limit h
RG 1.99 upper bound, actual values 0.36 w/o Cu, 0.015 w/o P
lABLE C6-2 MAINE YANKEE REACTOR VESSEL UELO SEAM TOUGHNESS DATA d
Charpy Qualification Test Results Average Ener9y Estimated Weld Seam at 10 F (ft-lb) at 10 F (ft-lb)
RTNDT ( F)
I 1-203 A/C 113, 123, 140 125.3
< -50 2-203 A/C 50, 60, 72 61.0
-50 l
l 3-203 A/C 62, 47, 62 57.0
-40 l
79, 79 82 30.0
-50 l
3-203 35, 50, 43 44.3
-25 d
9-203 79, 63, 64 70.3
-30 nI:
151, 121, 123 1 31.7
< -50 b
Surveillance Weld 50.4,52.6,42.7 43.6
-30C a Same as surveillance weld, so same RTNDT b Surveillance weld fabricated with same heat of wire (IP 3571) and lot of flux (Linde 1092 lot 3953) as weld 9-203 c Actual RTNDT based on drop weight and Charpy test results d Estimated using the nothod described in the text
't
FIGURE C6-1 MAINE VANPIE RCACIOR FRESSURE VESSEL MAP INITIAL RTNOT IN 'F 2
R S
o
_lNLET GUILET IRLET GUT lET I*LET OUTLET IllLET "
1 I
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C. 7.0.
Maine Yankee Vessel Integrity The fracture mechanics analysis is performed using the plant specific properties of the Maine Yankee vessel. The attenuation of the peak fluence value is considered in three dimensions (r, z, 0), and the superposition of the fluence profile and the weld geometry map is used in calculating the predicted RT value at all points in the vessel NDT as a function of Effective Full Power Years (EFPY). This infomation is used in locating the points in the vessel having the' highest RTNDT at e?ch of the three axial sections of interest:
- 1) middic of core, z 136. in.
=
- 67. in.
- 2) top of core, z
=
41 in,
- 3) above-core, z
=
where z is the axial distance below the centerline of the nozzle. From the predicted RT values, the material toughness properties K and K NDT IC Ia are determined from the calculated temperatures for the SBLOCA + LOFW transients using the method described in Section 7.6.
Critical crack vs crack depth curves depth diagrams are constructed from the applied Ky and the calculated material toughness curves. By performing the same fracture mechanics analysis a number of times for increasing plant life (EFPY)' the integrity of the Maine Yankee vessel for the SBLOCA + LOFW transient is evaluated.
C.7.1 Sumary of Physics and Paterials Data Input to Fracture Mechanics Analysis
)
A detailed survey was performed on the combined fluence and material properties maps of the Maine Yankee vessel to determine the most critical locations in terms of radiation embrittlement. The properties are considered independently at the three. axial sections. At each section, the combination of fluence and materials data were evaluated for a large number of points around the circumference. The adjusted RT values at NDT the inner vessel radius were compared, and the location with the highest RT value was used in the fracture mechanics analysis.
NDI At the mid-core level, the location of highest RT occurs in the NDT weld material at an azimuthal angle of 270 degrees. The fluence factor
. at this location is.86 of the peak fluence in the vessel.
C17
_ _. _.. _. _ _ _. -., _ _, ~ _ - _. _ -. - _... - _.
. ~ - _
The materials data at this point are as follows:
.99 PCT.
Ni
=
.35 PCT.
Cu
=
.012 PCT.
P
=
-20%
Initial RT
=
fiDT 19 At the 12/31/81 level of 6.0 EFPY, and peak fluence of.507 x 10 19 n/cm2 (E > 1 MeV), this corresponds to a point fluence of.437 x 10 0
,n/cm and an adjusted surface RT value of 198 F.
fiDT occurs in the At the top of core leve, the location of highest RTNDT weld material at an azimuthal angle of 270 degrees. The fluence factor at this location in the vessel is.17 of the peak fluence. The materials data at this point are as follows:
1.06 PCT.
Ni
=
.35 PCT.
Cu
=
.012 PCT.
P
=
-20%
Initial Ri
=
ttDT 19 At the 12/31/81 level of 6.0 EFPY, and peak fluence of.507 x 10 19 2 (E > l MeV), this corresponds to a point fluence of.087 x 10 n.cm 2
n/cm and ad adjusted surface RT valueof77%.
f4DT At the above-core level (about halfway between the top cf core and the inlet nozzle), the location of highest RT occurs in the plate material iiDT at an azimuthal angle of 170 degrees. The fluence factor at this point is.0007 of the peak fluence in the vessel. The materials data for this point are as fol?cws:
.58 PCT.
Ni
=
.17 PCT.
Cu
=
.011 PCT.
P
=
20 %
Initial RT
=
!iDT C18
}
19 At the 12/31/81 level of 6.0 EFPY, and peak fluence of.507 x 10 6
n/cm2 (E > 1 MeV), this corresponds to a point fluence of.335 x 10 a/cm and an adjusted surface RT valueof23%.
t4DT This represents the materials infonnation available at the time of values were subse@endy the analysis. Lower initial weld metal RTNDT justified by additional testing. The use of the present values therefore provides a conservative evaluation of vessel integrity.
C.7.2 Results of Fracture Mechanics Analysis for SBLOCA + LOFW Restoration of Feedwater (Case 6)
The stress analysis for the Maine Yankee SBLOCA + LOP 4 transient'is.
presented in Section 7.8.3 of the report. The fracture mechanics analyses were performed for this case using the Maine Yankee vessel proper-ties and predicted fluence levels up to the assumed end-of-life condition of 32 EFPY. The critical crack depth diagram at the mid-core level of the-vessel for 32 EFPY is given in Figure C.7-1.
An arrest region is indicated for this transient. However, the initiation toughness level is not exceeded for this case, therefore, no crack initiation would be expected to occur. The upper shelf toughness line indicates the flaw depths for which K; =- 200 ksiN. This represents the upper limit of applicability for linear elastic fracture mechanics. A ductile failure mechanism would be expected for crack sizes above this limit. The fact than no initiation region is apparent within the rance of applicability of LEFM indicates that brittle fracture would not occur for this corbination of transient loading conditions and end-of-life material properties.
The warm-prestress time is indicated on the critical crack depth diagram.
Warm-prestressing would occur beyond 42 min tes in the transient as seen from the' peak in the X vs time plot given in Figure 7.20 of the g
report.
C19
The center critical crack depth diagram for the top of core level.
at 32 EFPY is given in Figure C.7-2.
Similarly, the diagram for the above the core level of the vessel at 32 EFPY is shown in Figure C.7-3.
Both of these figures indicate that the initiation toughness level is not exceeded.at these locations in the vessel throughout the expected plant life for this transient loading condition.
C.7.3 Conclusion These results demonstrate that the integrity of the Maine Yankee vessel would be assured throughout the assumed life of the plant for the Maine Yankee SBLOCA + LOFW transient.
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-C23 e
8.0 -CONCLUSIONS This Appendix to CEh-189 provides the results of analytical evaluations of pressurized thermal shock effects on tha Maine Ynkee reactor vessel for cases of a SBLOCA + LOFW, in response to the requirements of Item II.K.2.13 of NUREG-0737. Two different scenarios were chosen for eval-uation based on remedial actions to prevent. inadequate core cooling:
1.
SBLOCA + LOFW + PORV's opened after 10 minutes 2.
SBLOCA + LOFW + Aux. FW reinstated after 30 minutes Thermal-hydraulic systen, transient calculations were performed on a reference-plant basis, as reported in CEN-189 with the parameter variations over the range representing all operating plants, Four different cases were analyzed for each of the two different scenarios defined above, for a total of eight cases. The most challenging of the two different scenarios was analyzed using linear elastic fracture mechanics methods to determine the critical crack tip stress intensity values for comparison to plant specific materials properties at various times in plant life. The effect of the warm prestress phenomenon is identified where applicable for each transient, and credited where appropriate.
In this Appendix, the results of plant specific neutron fluence pro-file calculations are superimposed on plant specific material proper-ties to define vessel capability versus plant life. The results of the generic LEFM analyses were evaluated using the plant specific material properties.
It is concluded that crack initiation would not occur due to the SBLOCA + LOPd transients considered, for more than 32 effective full power years of operation, which is assumed to represent full plant life.
C24
COMBUSTION ENGINEERING, INC.
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