ML20039F863
| ML20039F863 | |
| Person / Time | |
|---|---|
| Site: | Millstone, Palisades  |
| 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-E, TAC-46893, NUDOCS 8201130486 | |
| Download: ML20039F863 (34) | |
Text
- - - - - - - -
so-3
- l'Aiaixe EVALUA"10N OF PRESSURIZED THERMAL SH0CK E:FECTS DUE TO SMALL BREAK LOCA'S WITH LOSS OF FEEDWATER FOR THE Mll. STONE 2 REACTOR VESSE'.
Prepared for NORTHEAST UTILITIES POWER SERVICE COMPANY NUCLEAR WE SYSTEMS DIVISION Pi
' POWER Emaid SYSTEMS S 2 01 13 0 Wh COMBUSTION ENGINEERING INC
I LEGAL NOTICE THIS REPORT WAS PREPARED AS AN ACCOUNT OF WORK SPONSORED BY COM8USTION 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.
l
)
.._-i_
i ABSTRACT i
j This Appendix to CEN-189 provides the plant-specific i
evaluation of pressurized thermal shock effects due to small break LOCA's with extended loss of feedwater for the Millstone-2 reactor vessel.
It is concluded that; crack initiation would not occur for the transients considered for more than 32 effective full power years, which -is assumed to represent full plant life.
l i
I 1
I i
1
CEN-189 Appendix E TABLE OF CONTENTS SECTION TITLE PAGE ABSTRACT El.
PURPOSE El E2.
SCOPE El E3.
INTRODUCTION El E4.
THERMAL HYDRAULIC ANALYSES El E5, FLUENCE DISTRIBUTIONS ES E6.
MATERIAL PROPERTIES E13 E7.
VESSEL INTEGRITY EVALUATIONS E19 E8, CONCLUSIONS E29 i
ii
E1.0 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 Millstone 2 reactor vessel.
E2.0 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 Millstone 2 reactor vessel.
i Other C-E NSSS reactor vessels are reported in separate Appendices.
E3.0 INTRODUCTIOM This Appendix to CEN-189 was prepared by C-E for Northeast Utilities for their use in responding to Item II.K.2.13 of NUREG-0737 for the Millstone 2 reactor 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 E5 of this App.aix reports the plant-specific fluence distributions. developed as described in Chapter 5.0 of the main report.
Chapter E6 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 E7 reports the results of comparing the fracture mechanics results of Chapter 7.0 of the report, to the material properties discussed in Chapter E6.
E4.0 THERMAL HYDRAULIC 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 performing more detailed plant-specific thermal-hydraulic system analyses.
Removal of these available conser-vatisms by additional analyses was not performed due to the favorable conclusion achieved.
El
E5. Millstone Point-Unit 2 Fluence' Distribution Northeast Utilities provided the reactor energy output and detailed radial power distributions required to update the azimuthal fluence distribution from the time of the surveillance capsule removal to December 31, 1981.
Up to December 31, 1981 a cumulative energy generation of 93,905,659 flegawatt-hours (MwH) was quoted as shown in Table E5-1.
If full power is defined as 2700 Megawatts-thermal (Mwt) then this energy output corresponds to 3.97 Effective Full Power Years (EFPY)'. The results of the surveillance capsule analysis show a peak fast neutron fluence on the vessel wall of 18 (n/cm ) for an exposure of 3.0 EFPY at 2700 Mwt (ES-1). This 2
2.8 x 10 implies a rate of accumulation in the peak fluence of 0.933 x 1018 (n/cm2) per EFPY at 2700 Mwt. This value is an approximation because of the change in power levels during the third cycle, however, the change would only be on the order of 4%.
Using the above rate of accumulation in the _ peak fluence, the peak wall fluence as of December 31, 1981 (3.97 EFPY at 2700 Mwt) is 3.70 2
x 1018 (n/cm ),
TABLE ES-1 Millstone Point Unit 2 Energy Output 100%
Power Level Cycle EFPD (Mwt)
MW-HR 1
488.8 2560 30,031,872 2
297.33 2560 18,267,955 3
23.65 2560 1,453,056 3
300.78 2700 19,490,544 4
380.59 2700 24,662,232 (EFPD.= Effective Full Power Days)
A surmiary of the results obtained from the surveillance capsule analysis report is shown in Table E5-2.
TABLE E5-2 Effective Full Peak Fluence Accumu}ationRate Peak Wall Fluence Full Power Level Power Years 2
n/cm cer EFPY n/cm Mwt EFPY 2.8 x 10 2700 3.0 9.33 x 1037 18 3.70 x 1018 2700 3.97 9.33 x 1017
- E2
The azimuthal shape of the fluence distribution was obtained by updating the 00T-R9 azimuthal distribution corresponding to the surveillance capsule analysis (3.0 EFPY) to December 31,1981 (3.97 EFPY). The adjustment factors were calculated using the SHADRAC code as described in Section 5.2.2.
The detailed pin power distribution used in the DOT-R9 calculations were obtained from a pin power distribution calculated in cycle 3 which displayed an assembly power distribution that closely approximated the time averaged assembly power distribution from initial plant startup to the end of cycle 3.
The pin powers were then adjusted to correspond to the actual time averaged assembly power distribution. The resulting D0T-R9 azimuthal fast flux distribution was then adjusted to correspond to the time averaged power distribution as of December 31, 1981 (3.97 EFPY) using a nodalized form of the pin powers as shown in Figures E5-1 and E5-2.
The beginning, middle and end of cycle assembly power distributions for cycle 4 were supplied by Northeast Utilities. These assembly power distributions are shown in Figures E5-3, E5-4 and E5-5, respectively. The resulting azimuthal fluence distribution for December 31, 1981 is shown in Figure E5-6, where the 00 reference location is given in Figure E5-7.
One eighth core symmetry was assumed.
The axial and radial fluence distributions in the reactor vessel are obtained from DOT RZ calculations using the Millstone Point Unit 2 design as a reference plant. The axial shaoe of the fluence is shown in Figure ES-8.
The radial shapes of the fluence are given in Figure E5-9.
Refe rences:
E5-1.
Post Irradiation Evaluation of Reactor Vessel Surveillance Caosule W-97, Combustion Engineering, TR-N-MCM-008, (Draft).
E3
NILLSTOkJE Z.
Box 1
2-PohjER dlSTR160Tl00 SE6HEArTATicM socace Pc i.ii izi!)i!N.i /Z i3 19 is
!bl iN! !!.'7 !!!
llo 17
- F: : q. :.:..:]d :.:. )8 3
4 5 isiiii! !ilii.i 19 20 Zi ZZ Jo 31
- 32. 33 SF $9 40 #I 47 I
29'24 25 34 M
yg 43 qz; gg l
~
27 zir 64 37 45 qf, l
26 l
8 z3 l
0
'l9 50 51 SL 60 61 f
lo2 59 54 55 a m doEE. RAT
(
demEE M 56 57 58 65.
6G 14 53 71 72 q
ME DISTRlbOTIOL) oF PIIGS IM THE b]' AT 23 74 75 lbTER Roh) of SMAUEST SEbMEMS StGt AS SE6 MENT 3 1,z.,3,4 OF 60X 1 OR SEbhEMT319;30,22,24 Of BOX.3 B SPUT IMTo 3,4,4 AUD 3 PIMS foR D
15 FAcH SE6NEm i RESPEcWE.L4. TH DF PsMS In EACH of THE JOUMBER OTHER SE6MEMT3 OAM BE DETER FROM Tris Fl60RE.
FIGURE ES-1
MWM i_
M
\\,\\y\\
\\\\
i
\\\\.
\\\\\\
I i
l_
~
I I
I I
i 1
v CEWTEEL/NE
\\
E5
MILLSTONE UNIT 2 OCTANT 200 muJo/m t A 1.as a.qq V P
?
- 1. \\ %
t.2o oss5 0.to3
~
=
~
~).22.
o.sq o.ga 1.12.
OR8
- 1. ie l.o3 t.n 1.27 1.o2.
i.17 O.,I O.19 o.90 1.2)
\\.os OA4 0A9 1.09 0Rt 0.tolo
\\.\\\\
t.oJ t.2O o.t7
- t. \\ '1 1 74 j O.5(o L:
Beginning of Cycle 4 Assembly Power Distribution FIGURE E5-3
.E6
MILLSTONE UNIT 2 OCTANT l
DLUD (YTTN 1.11 035 RPD r
L
\\.\\S t.\\\\o oss3 0B
\\
\\.23-0.92 0.91
\\. \\2.
0.94-t I
\\.\\3
\\.03
- t. \\5
\\.21 0.R9
\\.\\ \\
l o.T3 N
o.89 oAs 1.s7 1.oy o.9y
\\.co G.97 c.93
[
lo c.17
\\.\\7 i.o?
i.19 c.9o
- 1. 2.z.
\\.19 r
I s
F r
Middle of Cycle 4 Asser.bly Power Distribution i
FIGURE E5-4 j
h l
E7 m
H
MILLSTONE UNIT 2 OCTANT
\\\\2S0 mloD/mTk 1.09 c.77
- PD
\\.l9 l.\\4 0.R5 0.( 45
- - ~ "~"
i.il O.43 0.92
- 1. I 2.
0.9%
=-
i.13 t.03 i.is t.\\ 9 o.98 t.o9
~
o.34
".o3 oAs
).co o,q g OA5 08s
). ig 1
080 0.8)
O.S4
\\. \\ 9
\\.oto
- 1. \\ 9 o.99
- l. \\9
\\.\\(o h
End of Cycle 4 Assembly Powe.r Distribution FIGURE E5-5 EB
ll1 6
04
~ :
N i
53 N0 1
T 3
A I
RA V
E CN E E 5
A 2
U I
f'"
F L D 6
L 0 s.
F M S
A E
A Q A
E H L R
T ES U
S U
G E
M V
1 I
I 2
F l
A Z
A 2
E NO T
S 5
L S
1 L
\\
I M
1 h
1 5
m ll F
TSA F
M r I
LEW
- ~
- =
8 5
8 7
5 L
E 1
1
180 r ~ - - ~ ~ ' 'i
's/ Outlet NOIIIe i
/
's
(
i 3
Vessel
/
h 104 \\ m
- s.,
n 3
i
[- l}
l sNczIIe Nintet b
/
x j
core shroud
\\
'S
. //
,r' Core support Barrel
,/
ermal Vessel Vessel g
Shield 0
Reactor vessel 97 ~~~~ +
.:5 j/
core i j
/ "r
(,/{ap$$'
Vessel [
N ' Vessel "I i sule t
83o ',s /
,I 277 I \\'.
AsseinW
' Thermal Shield Sh~
t W
r /
',__ Vessel f s
j
,/
2se q\\.
k s
\\
\\
\\
I I
j
\\
i L '
~'
f N s
-,/
s
= xr #
q D -Core j'/
Reactor Support g
Barrel I
Vessel i
\\
/
\\
/
I l
I,,, - - -
- l U
O i
FIGURE Es-7
MILLSTONE 2AXIALFLUENCEVARIATION i
t i
AT YESSEL 03D INTEllFACE
'(
NBAtlIEDFASTFLl2 IEE i j N
\\
t 3
X IE-1
(
\\
\\
3 4
3 t-
\\
t IE-2
's t.
\\
4
\\
t 3
\\
4
\\
IE-3 t
\\.
4
\\
\\
3
\\
t
\\
IE-4 4
x
'n
(:
i 4
\\
3 i
g
\\
(
t 1;-5
's
\\
(;
\\
l g
3
\\
0
\\
C IE-5 C
I 25 Sa 75 108 125 12 175 298 225 22 715 328 325 352 c
C Distance From Core M.P. - CM Ell
HILLSTONE2RADIALFLUENCEVARIATION At VESSEL CLAD ININAE 10NLIIED FA51 FLUX IE2 -\\
\\ \\
UlR ElWI GE W.P.
L 5
y
\\,\\
s
' D!AL E!WI
\\\\ '-
s N
' s,
=198.12 G b
\\
2 3
s y
sN\\
N s,
N 3
\\
Ulu ElWI y s IE-1 s,y s,
=251.530
\\ \\
\\
\\ \\
\\
s s
. s
\\
1
\\
\\ \\
5 g \\
x x
\\
\\
2 IE-2 B 1 2 3 4 5 6 7 8 9 le 11 12 13 14 15 16 17 18 19 20 21 22 23 24 Distance From V.C. Interface - CM FIGURE 5-9
APPENDIX E MILLSTONE UNIT #2 E.6 MATERIAL PROPERTIES The methods used to develop and evaluate the materials for the Millstone Unit #2 reactor vessel are described in Section 6.0 in the main body of the report. The chemistry data (nickel, copper, and phosphorus content) and initial (pre irradiation) toughness properties of the reactor vessel shell course plates and welds are lum:aarized in Table E6-1.
In cases where the weld metal nickel content was not determined, it was conservatively estimated using information on the type of wire used (eg, high MnMo versus MnMcNi wire).
For the Millstone Unit #2 weldments, the weld in-spection records and welding certification reports for seam 1-203 indicated that the weld could be expected to contain high nickel (greater than 0.30 w/o) since it was f abricated with MnMcNi wire, so the nickel content was conserva-tively estimated to be 0.99 w/o as shown in Table E6-1.
For seams 2-203 and 3-203, the records indicated that the welds could be expected to contain low nickel (less than 0.30 w/o) since they were fabricated with high MnMo wire, so the nickel content was conservatively estimated to be 0.20 w/o as shown in the Table.
The toughness properties given in Table E6-1 are the drop weight NDTT (if determined) and the initial reference temperature, RTNDT.
For the plate materials, the RTNDT was determined using transversely oriented Charpy impact specimens or by converting longitudinal impact data using Branch Technical Position MTES 5-2*.
For the weld material, the RTNDT was estimated using 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 E6-2.
Each weld which exhibited an average Charpy energy of 79 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 welde i.e.,
" Fracture Toughness Requirements for Older Plants,"
U.S. Atomic Energy Commission, Regulatory Standard Review Plan.
l l
E6 L
r
-SSF or less.
For those weld qualification test results exhibiting an average Charpy energy less than 79 ft-lb, the RT I
NDT 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 ft-lb or better exists was determined, and the RT was established at a temperature 60F below that value.
NDT A " map" of the cylindrical portion of the Millstone Unit #2 reactor vessel is given in Figure E6-1.
It shows the locations of the plates and welds listed in Table E6-1 and their corresponding values of initial RT
""9
'9' NDT NDT for the 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 E6-2 is a map of adjusted RT values f r important NDT locations at the inner surface of the Millstone Unit #2 vessel predicted for December 31, 1981.
The predictions are based on the best estimate neutron fluence, 0.370 x 10 n/cm (E>lMeV), (corresponding to 3 97 effective full power years at peak flux location on the inside surface of the reactor vessel), the initial RT and c Pper, P osphorus, and h
NDT nickel contents given in Table E6-1, and the normalizd neutron flux profiles given in Section E.5.
The values of adjusted RT nitial NDT l
P us predicted shift) are located in rectangles adjacent to the PTNDT plate and weld designations. The RT Value8 app y to the inner l
DT surface of the vessel in the region indicated by a circle.
The circled regions generally represent areas of peak neutron flux for_ a given weld seam or plate.
E6-2
REACT'dif VESSEL f1ATERI ALS Product Material Drop Weight Initial Chemical Content (%)
Form Identification NDTT ( F)
RTNDT ( F) liickel Coyper Phosphorus b
d P1 ate C-504-1 10 20 ^
0.53 0.13 0.011 D
Plate C-504-2 10 20 ^
0.56 0.13 0.009 b
Plate C-504-3
-10 10 0.56 0.13 0.012 Pla te C-505-1
-20 5^
0.64 0.13 0.006 a
P1 ate C-505-2
-10 25 0.64 0.13 0.003 a
Plate C-505-3
-10 D
0.65 0.13 0.007 c
c P1 ate C-506-1
-10 6
0.61 0.14 0.006 Plate C-506-2
-40
-30^
0.61 0.14 0.007 Plate C-506-3
-30 0^
0.70 0.13 0.005 Weld 1 -203 A,B,&C N/A
-45j 8
0.99 0.22 0.015 7
Weld 2-203 A,B,&C N/A
-50 0.20 0.12 0.013 d
7 Weld 3-203 A,B,&C N/A
-50 0.20 0.12 0.018 d
Weld 8-203 N/A
-60 0.18 0.30 0.013 c
c c
c c
Weld 9-203
-60
_55 0.06 0.30 0.015 L
t N/A Not Available Determined using Branch Technical Position MTEB 5-2 a
b Estimated based on average for Millstone Unit #2 plates having analyses reported c
Surveillance program data d
Estimated (see text and Table E6-2)
Estimated Ni content (high nickel type wire) e f
Estimated Ni content (low nickel type wire)
L _. _. _ _ _ _ _ _ _ _..
TABLE E6-2 f1ILLST0flE UNIT #2 REACTOR VESSEL WELD SEAf1 TOUGilNESS DATA c
Charpy Qualification Test Results Average Ener9y Estimated Weld Seam at 10 F (ft-lb) at 10 F (ft-lb)
RTNDT ( F) 1-203 A/C 62, 59, 60 60.3
-45 2-203 A/C 66, 75, 78 73.0
-50 3-203 A/C 66, 75, 78 73.0
-50 3-203 151, 121, 123 131.7
-80 101, 103, 107 105.3
-60 8
9-203 101, 108, 107 105.3
-55 110, 116, 107 111.0
-60 b
Surveillance Weld 77, 79, 30.5 73.3
-55 a Held seam 9-203 fabricated with same heat of. wire (10137) and lot of flux (Linde 0091 lot 3999) as surveillance weld, so same RTNDT b Actual RTf1DT based on drop weight and Charpy test data c Estimated using the method described in the text
-j
f!GJRE E6-1 HILLSTONE #2 REACTOR PRESSURE VESSEL MAP INITIAL RTNDT IN *T l
l l
OUTLET IflLET a
INLET GUNLET IflLET 0
U o
l'lLET GUTLET J\\
\\
)
\\
\\
'O
~
'y
/
E23 EIdl 80 g
C 504 2 C-504 3 C 5048 C 504 2 g
gg 6
6 203_ _
CI i
l m
e u uw cr o aza cc C 505t C 505 2 C 505 3 C 505 e E
____________(COCE_._
2203 ESG U
9 203 5
3203 8 Ela CED Ef23 EI3 C. 50 6 6 C 506 3 C 506 C
._______2__________ _ 506____.
22G 260 O
,QO 360 H Z illUltf HL LOL H I I ON, Ot64t L5
FIGL"IT f6-2 MILLSTONE #2 REACTOR PHESSURE VESSEL MAP ADJUSTED RTNOT IN *F (12/31/81) 0.37 x 10gective Full Power Years 3.97 E n/cv Peak Surface fluence OIITLET IrlLET 2
INLET OLIT E LET IriLET U
I!1LET GUTLETE o
ru
/
C 504 2 C 504 3 C.5041 C 504 2 a.
203 E3]
g H28 E3Fi u
g's 201_.,_
s
^
s
.u 449 g
too y
C. 50 5 s [Eddl C 505 2 C.505 3 C SOS-4 z
EMI 83 g'
--qC0RE--
-a---
2 203
,.- M u,
. N.
n 1
9 263 gj u
v cc G_i) 3 203
/
2
'eo
/
OD C 506s@.67)
C 5 0 6-3 C 506 2 C 5064 220 760 '
liso 360 HZ I NU T HHL - LOC AI I ON,0EG8KES
E.7.0 Millstone 2 Vessel Integrity The fracture mechanics analysis is performed using the plant specific properties of the Millstone 2 vessel. The attenuation of the peak fluence value is considered in three dimensions (r, z, 9), 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 information is used in locatg the points in the vessel having the highest RT at riDT each of the three axial sections of interest:
- 1) middle of core, z 136.4 in.
=
2) top of core, z 66 in.
=
3) above-core, z 40.5 in,
=
where z is the axial distance below the centerline of the nozzle.
From the predicted RT values, the material toughness properties K and NDT IC K, are determined from the calculated temperatures for the SBLOCA g
+ LOFW transients using the method described in Section 7.3.3.
Critical crack depth diagrams are constructed from the applied K vs crack depth 7
curves 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 Millstone 2 vessel for the SBLOCA + LOFW transient is evaluated.
E.7.1 Summary of Physics and Materials Data Input to Fracture Mechanics Analysis A detailed survey was performed on the combined fluence and material properties maps of the Millstone 2 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 riDT the inner vessel radius were compared, and the location with the highest RT value was used in the fracture mechanics analysis.
NDT At the mid-core level, the location of. highest RT occurs in the plate fiDT material at an azimuthal angle of 93 degrees. The fluence factor at this location is 1.0 of the peak fluence in the vessel.
E19
=- _.
The materials data at this point are as follows:
PCT.
Ni
.64
=
PCT.
Cu
.13
=
PCT.
P
.008
=
25 %
Initial RT
=
t4DT 19 At the 12/31/81 level of 4.0 EFPY, and peak fluence of.373 x 10 2
19 n/cm (E> 1 MeV), this corresponds to a point fluence' of.369 x 10 2
n/cm and an adjusted surface RT value of 80%.
NDT At the top of core level the location of highest RT ccurs in the NDT plate material at an azimuthal angle of 93 degrees. The fluence factor at this location in the vessel is.36 of the peak fluence.
The materials data at this point are as follows:
PCT.
Ni
.64
=
PCT.
C9
.13
=
.008 PCT.
P
=
25 %
Initial RT
=
NDT 19 At the 12/31/81 level of 4.0 EFPY, and peak fluence of.373 x 10 2
19 n/cm (E >l MeV), this corresponds to a point fluence of.136 x 10 2
n/cm and an adjusted surface RT valueof58%.
NDT At the above-core level (about halfway between the top of core and the inlet nozzle), the location of highest RT occurs in the plate ilDT materials at an azimuthal angle of 267 degrees.
The fluence factor
- at this point is.005 of the peak fluence in the vessel. The materials data for this point are as follows:
.58 PCT.
Ni
=
PCT.
Cu
.13
=
.011 PCT.
P
=
20 %
Initial RT
=
NDT rise
19 At the 12/31/81 level of 4.0 EFPY, and peak fluence of.373 x 10 2
19 n/cm (E > 1 MeV), this corresponds to a point fluence of.002 x 10 2
n.cm and an adjusted surface RT value of 25 F.
l NDT This represents the materials infonnation available at the time of the analysis.
E.7.2 Results of Fracture Mechanics Analysis for SBLOCA + LOFW Open PORV's (Case 4) l The stress analysis for this casa is presented in Section 7.8.1 of the report.
The fracture mechanics analyses were performed using the Millstone 2 vessel properties and considering fluence levels up to the assumed end-of-life condition of 32 EFPY.
Figure E.7-1 shows the critical crack depth diagram for the mid-core level at 32 EFPY.
The calculated shifts in RT are relatively low, thus the initiation toughness level is not NDT exceeded for this transient. This is indicated by the lack of data on the critical crack depth diagram.
Therefore, no crack initiation would occur under these conditions at the mid-core level of the vessel.
Figure E.7-2 is the critical crack depth diagram for the top of core level of the vessel at 32 EFPY.
Similarly, the diagram for the level above the core is given in Figure E.7-3.
Both of these diagrams indicate that there are no potential regions for crack initiation at these locations in the vessel for this transient.
i 1
E.7.3 Results of Fracture Mechanics Analysis for SBLOCA + LOFW Restoration of Feedwater (Case 5)
The stress analysis for this transient is presented in Section 7.8.2 of the report.
Fracture mechanics analyses were performed using the Millstone 2 vessel properties with fluence levelsup to the assumed end-of-life condition of 32 EFPY.
Critical crack depth diagrams were constructed as shown in Figure E.7-4 for the mid-core level of the vessel at 32 EFPY. The calculated values of K for this transient do not exceed g
the initiation toughness level under these conditions.
This indicates that i
crack initiation would not occur at the mid-core level of the vessel under E21
k m'
y -_
v
- y.
.,c,...
y-g;
~
~
.,g L
T v
this combination of loading and veise riatorial properties.
The critical crack depth diagram for tha tcp of co e level.of the
{
vessel at 32 EFPY is given n Figure E.7-5.
The p?ot of crit'1' cal crack depths for the level of the vessel above the core'is shown in ~
h Figure E.7-6.
The fact th5t no crack initiation regions' exist in these E
figures indicates that no crack initiation would occur in these' areas of the vessel throughout the life of the plant for this transient loading.
candition.
c-
~
r
- -E.7.4' Conclusion
~
These results demonstrate that the integrity of the Millstone 2 vessel would
/be assured throughout the assusted life o' the plant for SBLOCA + LOF4 transient with recovery ci feedwater, and' for the SBu0CA1' l.0R4 transient where PORV's are opened.
s 7
K-
~
s
.h.'-
s
(.
s'
-r em 4
i
_s x
1 q
N
,s
\\
~
s
-N
.N g.
t 1
s g
M
\\_
i g
x
=E E22
h
)
h
)
]
e, s
O 44o
+'. s '. N o s '.si.*s s'.s<>s5s*
- ss s<
s'.s'.*".ss s4s'.%*..*,s'a sd' s%,s' O
,., o a
g 3
O*
E 8
8.
0 8
e.
I*
8 8M g
.O t 4 s
!M t@
i GP p
b 0
1
(
8 t
R.
g 3
0 t.na E
O E
89 U
3 g
. 4 te.
g O
W p
O
.nf e.
O J
4
?
o O
3 E
8 4
8 t
4
+
m e
143 9
Z W<
M e
e=o i
3, U.
s.
r
~
w o
2 e
o m
w w
2 o
e s
Z s.=
amu J
8 J
x
. o.
o
.,o
,e, f
.s a
i I
i, i.
1
': s s s s. s,. s,.. s s,. s. s,. s,. s s, s. s, s,!..
I 1
i y i
i sss.s,.ss.ssss ssss<
. 3j i
1 i
i i
i t
i 1
t l
t==
c e
o o,
,' 4 o
e o 1 o
s e S,
o
-e o 2: *l LJ
)o e.
o.
e.
o.
g o.
I 4
e o
o.
e
[
+
e
- {
. i o}
O
,W J
J J
J i
J a
.aa J -
J
.J j
- p e }
i l
l w o
o e
=
l e
o f
e o
o e;
rg O l
, o o
a6 o
o e
o a
o o
o 4
j
=
t s) e o1 o
or e
o o
o o
==
1.
'a 2o l
=*
'.P SI h
4
.fi e
9 N
==
oea. o.3 I
e o
e o
e o
e e
e e
et c o
i l
1 i
~ -
i
=9=5 i
<ss e.- s a. a s
ua<un o a s. - :
l I
CU m
i 4.6 Z,- oc i> e e-2 a
l g
Il 2== Q s=
i r
[
[
+
t..'
U12 i
l 1
l
- 1 I
sus l
.Z 5
m
+
e M
3 O
O C
C C
O O
O O
O
^
n.
.w
.,s..
,s m.
sns.v u.w w.m s.uu.,ws,.-,s<,.,s-,ss. sv p.
.e i
b P
b b
+
I I
l l
I l
N.
l I
I I naJ
.l.
I I
P I
t
, N.
.a 8
O I n.aJ l
.i o
i i
es=
I i
I teJ l
,I z
,I
i 4 j J'l 1 8't , me mm ==*8') b l i e as @== a e-e e . i e o e i e e g l t eN e a I I i naJ,E 3 4 f, UZ447g Q laJ 1 B-Ih 4,, :I u u f l 1 w2 ea Z DE>c i i l a us l i i . I. i == g g 4 i l I D O O O O C O O O O O O rei 4 e.. s,.,...,...,....ss. ,....s,. s. s,. s,.... s,.s., s,. s,. ...,,,,. ~ 00 t 9 =e 8 6 8 e 0 9 0 0 9 e 9 6 0 4 e e 0 8 t t 0 8 Sl2 t t4l4 l ee 0 0@ 0 I p 8 0 0 8 8 e 9 8 1 9 0 6 + 9 + t 8 8 0 0 8 8 0 ft 4 8e W t 6e E 8 8W C 8 8 4p W 3 +w e fe .nJ 5 64 4 9 O c 0 3 9 9 4 0 0 8 0 8 8 0 6 8 d. + 0e 0 la# M 8 2 W 9 8 9 4 N. 8 9 ee W 8 6 ef t f 3 tO E gg l t 9 e N 0 = 0 0 SW 0 le 2 b e ee "3 9 @ Inf G e $4 n.J Z m 8 0 L 9 8 1 e 8 W 3 9 e* .J 8 0 .a 9 l
- =
== O 8 K + 1 0 e 3 8 0 0 0 8 9 I f98 4 0e 8 9 e 9 8 Jf 9 8e I 99 e 9 fW 9 9 8 9 8 4, 8 9 8 el e 9 e> 0 0 0 8 4 0 + e 9 2 8 9 8 I h, 9 t ul I 0 ) 0 t l 8 4 1' 8 0 0 si i. i t I < s s s.i j 3 s,. s,.. s. s s s. s, s,..,. s, s..ss,s..s,.s.s,ss.ss ,.sesss t .I .i 1 I Li i o o o 4, O O W e o e o e + e e e 8 Q a .nJ 3 W J t we wJ w W .J J 1ta i sC l a,s o o 9 o 4, a 3 o o a c6 e o o o 3 l e f o e o e 3 o e O - e o o o o == wi == e l L si. 4 J e M N -o >= i 3 e M.= **3 o. e. ano==o.le e o e e + p e [ i rv. i e e. a e. s ai > 3 lL W OE4u1: 3 J CL > 1:
- 4%3 tlCu 3
- l'.
oO gie-. e-2 .i z== u *= u CL 3 i h e-lL .Z. ^ ^ O O O G S 9 7 J 8 O O o C C l 444o m *'m%9m%< N',%%. eS %',s4 %%.N'.*O e*, %9. % 4, %*, N96 e45%'m %'>s%%' m 84 3 e 04 3 0 '> 0 8 9 0 0 8 0 8 g 6 9 8 e 8 0 8 8 6 6 0 g 6 Fl e 9O g 94 !W 3 9@ g 84 g S 4 8 0 8 8 0 0 8 0 5 e 6 S 3 8 tad 0 0 E g 8 O 3 4M U S 0o e 0 4 4 !W w e 8@ J e 0e 3 3 9 Q g 9 a. I e 0 1 1 0 8 J 0 0 8 3 8 0 9 es e = l 8 't laJ e 9 0 2 >a 0 8 2> f 8 U 3 DM S l 2 9 b4 K = g e4 N g e 0W w O a 0 8o 2 3 e
- O e
fU M d 9 E9 W o a e o O
- /
t C K 2 0 0 W g 8 9 J 0 8 .J J g 0 t= i o e 9 Z 9 .U. 0 t e t 3 9 2 D 0 W 0 0M I Oo 0 0 4 I t 0W 0 eo + e
- O t
Io 0 0 f4 0 6 0 0 3 8 i 3 0 el 0 0 PP e 6 9 3 0 I t 6 0 0 3-0 0 3. + 0 0 W I 8 j.i 1 e s 8 q r -l t i t i i !,, N,, s. s. %,.. % s.. %. s,i. s..,. s,,,, < % % s %. s s. s. %,. N s. %,.. %.,, % w s.i ,3 g I g g I I l .4 e i O j le o' l s e o e, o o e. i J ol i o CD o o J o. 3 o. D. o. - o o. e e t J 3 e .,J a .J l 4 .b -h o o a 4 J J .J
- >.e l
O o o e 10 l I. i o o o j e o c o o o ( 3 3 o O e i M O ,j o o e a o, o o o o .4. m. z o .o. =
- P.
a. N.
- 4. :;
.4, e. j -t N. l
- =. I 3
o2 me-oc i l i i i ' ad *
- l l
l l l 1 oz<ux sws-: <%2 a a. a icu.= L l l J ~.o z n l
- ' 2 >= u e.
l >%O 2. 3 j 4 n o ra u- [26 m e = ~ m n p n p p q ,i v v ent D b o e".N5%< %9, %95 e'N*,%di %9,N5,e9,%9,%< %%,%% e',N9,%4, %*,%%, e%,%*,NeN% %%, eo 0 8o 9 0 =0 0 e < 0 0 t / 0 t I e 0 6 0 4 0 e e 0 0 t t t t 0 0 0 0N 0 Io l 4 8 e t 04 4 4o e o e 0 1 eo I i ta j t t 4 i 8 l 6 ? I e 0 t e 8 8 8 0 I l 0 I e e 0 0 0 0 t 0 0 0 0 0 ml W 4 04 3 4 0 e O 8 0 m3 O e 8o e eo m 8 8 o O 9 04 9 0 4 8 O 9 0 ~ 0 0 0 4 8 5 0 0 0 0 e e = 0 0 W {i 8 2 t 9 W 9 8 4 W 2> 0 0 =e U 8 9 tN 0 8 eo Z t t 4 N y 11 3 0J = 0 W 8 00 2 C O e ee i e to W L 2 0 04 W Z D U 0 0 0 CT. 4 0 0 Z >= E 8 0 W LA. Q t 8 =* .J t 0 0 a J 0 0 = E 4 4 4 o 8 8 0 0 9 0 3 8 8 y 8 6 el 9 0 o 0 04 8 4 L4 t 8o e> ee '8 0o 0 e rg i C g 0 8 4> 0 9 0 9 el 4 9 r> i 9 8 l 8 8 i e o 8 0 3' 0 t c, t I h, 0 0 61 9 9 0 t t t 3 8 0 tb 8 8
- ,%9,%<
N*,N% e% %',%4 N%,%*,4 C' 4 .NNN<>N'.Name 5%'.N<>N%%'.e %%'.N< N*.NS.* I tlk b o en a en e e> e e o e' e -> w, e W
- O 3;
o, o e o e o et ,o o o e o e + e e e e 4 e W idJ W W W LI W LS W Ld W t> kJ o. e.r) o o o epQ o e o o o Ci O o 4 o e o o o o o o o O w ma O O o. e. .a. n, N. nmo. ~ e ~ e l >N 9 e l s E 3 4 8 8 e i I\\M kdI >" ) b N O$IOII hNkN)I O L'OU 2 2
- O
.t e-2 wi e-i m== U== O Q. 2 6..*=Q. .Z e C i C C-- C C C, Co G Q G G S g __._______x ~ _ _ es C e nm) l .-,ssso s, s,.,,s, s+s*,s, e,s*,s< s*,s*,. *, s*,s< s, s*,. *, s*. s<s*.s* eO 0 0O 0 ** I C < 0 0 0 9 0 0 0 0 0 4 0 0 0 l 0 0 4 4 6 i 0 0 61 0 0 :% 0 0
- l 0) 9 0:
!.s O e - e i 9 e 8 g l gg a ?e l 0. s I-e O I I i 0 0 I O 0' 8 0 0 0 ? e 9 0 I I e e 0 0 1 0 8 0 01 f 84 neJ t 0R C 0 8W C 8 8O U eO e O 9O tea 0 4 6 O 8 0 4 3 2 0 0 4 0 0 kl 0 0 0 K t 5 == s a e =. no e j 0 8 taJ O 8 0 2 W EA. 4 6 4
- 1 I
6 =* U 8 9 9N 8 1 0 eO I s 0.i N O, 8 0w == O to 4 8O Z eO j a e I 9 9O W T I 84 W 2 t u a e o l 4 0 4 Z E I 4 W 8 8 f I C .J 0 0 6- == 4 0 0 E e u o == 0 8 9 0 == 8 6 2 0 8 Q 8 8 el 0 $O 9 0* 6 0W 8 0O e O e 8 9O 8 el 0 9 i 0 t ab 8 8 8 0 el 8 9> 0 a 0 0 3 0 0 e e 8 8 3' 0 Q O h I 8 kl 9 8 4 0 0 6 3 0 8 4> 0 0 N*,N'.e*,N*,N< N*,N*,e*,N',% < %%,N*.O i ,N'.N<>N'.N*.'.N'.% <,N'.N'.e*.N *.% < C ul6 en o e e> e el e O e o e o a> 0 db o e eup 2. o. e + e + 4 + e t e nnJ W h68 I60 hI leJ 4d led 44 le7 aI LaJ o o e o ei O O o o en
- Q o
e o e o e\\ o e o e o 4 k.l .O., O o o e o .e e, == o. e. ..o 8 o. ~. ,o. n, ~ ~ >NO i ee k i.- #> om<us: oWa-ao < *,3 t>C@ 3Z. ( E Ec8e
- -c.,
i. .z I l 1 C C C G G G G G G G O O O O O 9-Gl E
8.0 CONCLUSION
S This Appendix to CEN-189 provides the results of analytical evaluations of pressurized thermal shock effects on the Millstone-2 reactor vessel for cases of a SBLOCA + LOR 4, 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 system 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 each 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 + LOFW transients considered, for more than 32 effective full power years of operation, which is assumed to represent full plant life.
nse