ML20128B477
| ML20128B477 | |
| Person / Time | |
|---|---|
| Site: | Zion File:ZionSolutions icon.png |
| Issue date: | 01/28/1993 |
| From: | Stimac S COMMONWEALTH EDISON CO. |
| To: | Murley T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9302030073 | |
| Download: ML20128B477 (58) | |
Text
7
'; Commorcoalth Edison 4 1400 Opus Place
' Downers Grove, Illinois 00515
' ~
January 28,1993 Dr. Thomas E. Murley, Director Olilco of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555 Attn' Document Control Desk
Subject:
Zion Station Units 1 and 2 Fracture Tou hness Requirements for Protection A inst Pressurized Thermal Sh k Events - 10 CFR 50.61 Response to NRC Roquest for AdditionalInformation NRO DocketNosJ0:295_and50 304 Referencos:
(a)
May 22,1992 lotter from S.F. Stimac to T.E. Murley (b)
December 2,1992 letter from C.Y. Shiraki to T.E. Murley Doar Dr. Murley:
Reference (a) provided Commonwealth Edison Campany's (CECO's) re assossment of end of life adjusted reference temperature, pressurized thormal shock (RT values for Zlon Units 1 and 2. Reference (b) transmitted the staff's request fofts)ditional Information (RAl) related to the reforonco (a) submittal. CECO's ad response to the RAI is enclosed.
The re assessment transmitted via reference (a) is cased upon weld metal speclile data which was obtained from the Babcock and Wilcox Owners Group (BAWOG) Mastor Integrated Reactor Vossol Surveillance Program. This surveillance pro 9 ram, developed in accordance with the requirements of 10 CFR 50, Appendix H, Section ll.C, was previously approved by NRC. Whilo much of the literature describing the program and data evalual on has been previously submitted to NRC by the BAWOG, not all of the reports have been submitted. Therefore, pertinent excerpts from the referenced reports have been included in the enclosed response to facilitato staff review.
Please direct any questions you may have to this office.
Vpry truly Yo f,,
p atephen F.
nac 010000 Nuclear Licensing Administrator Enclosure cc:
A. Bori Davis, Regional Administrator - Rlli C.Y. Shiraki, Project Manager - NRR J.D. Smith, Senior Resident inspector - Zion 930203o073 930gpg I
pR ADOCK 0500o295
/
L ZNLD/2478/1 PDR l
J
I o
COMMONHEALTH EDISON C0t@ANY RESPONSE TO NRC REQUEST FOR ADDITIONAL INFORMATION l
DATED DECEMBER 2, 1992 Re g elt E :
Provide the basis (mechanistic and statistical) for not including in your analysis unirradiated data with stress relief times greater than 35 hours4.050926e-4 days <br />0.00972 hours <br />5.787037e-5 weeks <br />1.33175e-5 months <br />.
MoongE: A B&W review of vessel fabrication records showed that the maximum time that any B&H-fabricated PHR reactor vessel, fabricated using HF-70, was stress relieved is 35 hours4.050926e-4 days <br />0.00972 hours <br />5.787037e-5 weeks <br />1.33175e-5 months <br />, hence data beyond 35 hours4.050926e-4 days <br />0.00972 hours <br />5.787037e-5 weeks <br />1.33175e-5 months <br /> was considered irrelevant.
In the case of Zion Units 1 and 2,-the stress relief times wore 23 hours2.662037e-4 days <br />0.00639 hours <br />3.80291e-5 weeks <br />8.7515e-6 months <br /> and 31 hours3.587963e-4 days <br />0.00861 hours <br />5.125661e-5 weeks <br />1.17955e-5 months <br /> respectively.
The rationale for this approach is discussed in I
detail in BAH-2100, and the appropriate paragraphs are extracted and presented as follows:
"LYAJyation of Initial PJggy_h All the initial value RTNDT data for Mn-Mo-N1/Linde 80 weld metals except those fabricated with weld wire 72105 are presented in Tables 14-1 and 14-These data are presented in Figure 14-1 and show that the general population of the Mn-Mo-Ni/Linde 80 initial RTNDT values lie betweer: -50 and +30F.
These data are for materials with a range of strest-relief times.
Typically, the toughness values of these materials exhibit a low sensitivity-to stress-relief times.
The initial RTNDT data for weld metals fabricated with weld wire 72105 are presented in Table 14-3.
These data are also identified as to source and stress-relief time.
The data are presented in Figure 14-2 along with the general population of the Mn-Ho-N1/Linde 80 initisi RTNDT values from Tables 14-1 and 14-2 and show that the weld metals fabricated with weld wire 72015 HF-70 and WF-209-1, have initial RTNDT values that.11e between Of and +120F and overlap the general Mn-Mo-N1/Linde 80 date in the temperature range of O' to 30F.
The range of the RTNDT data of the 72105-weld metals is nearly twice ihat of the' general population of Mn-Mo-Ni/Linde 80 weld metals.
This variability is attributable to the 72105 weld metals having a unique sensitivity to stress-relief time, ihis is supported by the data in Figure 12-2 which show the Charpy 50 ft-lb data from the results of Charpy evaluations performed to determine the effects of stress relief times.
(1)
ZNLD/2479/1
A statistical analysis was performed on the two data sets and selected subsets to determine the mean and standard deviation.
These data are summarized in Table 14-4.
A review of the data illustrates the influence of the weld metal 72105 data on both the mean initial value and in some instances standard deviation.
The influence of all the data for welds made with weld wire 72105 is pronounced on both the mean value and in some instances the standard deviation.
The mean value increases from -4.8F to
+7.0F ar.d the standard deviation nearly doubled from 19.7F to 32.2F.
However, by eliminating all the data for stress-relief times of 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> or greater, the influence of the data on the overall metn and standard deviation are negligible.
This latter position is acceptable because none of the reactor vessels containing weld metal WF-70 have stress-relief times greater than 35 hours4.050926e-4 days <br />0.00972 hours <br />5.787037e-5 weeks <br />1.33175e-5 months <br />.
However, to include all the weld metal 72105 data would unjustly penalize the other kald metal initial RTNDT temperatures especially as related to the increase in standard deviation, It is apparent from the above that there are two dt; tinct data populations with regard to Mn-Ho-Ni/Linde 80 weld metals.
The first is the total Mn-Mo-N1/Linde 80 less weld metal 72105 data, and the second is the 72105 weld metal. Based on this observation, it appears practical to identify an initial mean RTNDT value and standard deviation specifically for weld metal WF-70 based on the data obtained for weld wires fabricated with weld wire 72105.
The mean and standard deviation data base for the HF-70 and HF-209-1 weld metals are included in Table 14-4.
The various subsets were analyzed as based on the rationale described above for the selection of the data at stress-relief times of less than 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />.
Based on this criterion, the analyses of the data produced a mean initial RTNDT value of +17.4F and a corresponding standard deviatio=. of 20.0F, Actually, the data from the Oconee Unit 3 sursaillance program is inconsistent with similar data.
If the Oconee 3 data is deleted, the mean value would become +11.0F with a standard deviation of il.5F.
A background evaluation of the Oconee Unit 3 data did not provide any justifiable reason for deletion of the data.
Therefore, tha former set of values (i.e. which includes the '<onee 3 date) will be usen rather than the latter for licensing applications where initial RTNDT and standard deviation values are neeJed.
(2)
ZNLD/2479/2
The relationship of the WM 72105 data stress-relieved for less than 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> is shown in Figure 14-3 and is compared to the mean standard deviation for the Mn-Mo-N)/Linde 80 data less HM 72105 data.
If the Oconee Unit 3 data is deleted, the remainder of the less than 40 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> stress-relief data falls within the 2 standard deviation limits of the original Hn-Mo-Ni/Linde 80 data.
If the WM 72105 data base is expanded to include the 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> stress-relief point, the mean value increases to 19 '5 xi the standard deviation remains relatively constant.
Therefore, u s ; only the less than 40 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> stress-relief data does not cause any real bias in the selection of the chosen mean and standard deviation for the initial values for WF-70.
Re c ommen d e d _ Re g u l a tory _Eo sit.lon s InttlaLRIgI_hlue - The initial RTNOT value to be uset for weld metal WF-70 in those cases where no meaiured RTNDT exists is as follows:
Initial RTNDT - 18F Standard Deviation - 20F These values are based on all the data available from weld metals fabricated with weld wire 72105 and which received stress-relief of less than forty hours."
i (3)
ZNLD/2479/3
Table 14-1.
Mn Mo Ni/ Linde-80 Submerged Arc Weld Metal Procedure Qualification Drop Weight and Charpy Impact Data Used i22' to Determine initial Reference Temnerature Values Charov V Properties Weld Drop Wt.
- Temp, Energy, Lat. Exp.,
- RTuor, Number Tuor, F F
ft lbs mils F
Source
- 1
-40
+70 52,52,60 38,39,47
+10 PQ 3170 2
-20
+80 50,50,53 42,49,46
+20 PQ 2923 3
-20
+70 50,51,57 44,43,55
+10 PQ 3443 4
-30
+70 51,54,61 56,48,45
+10 PQ 3116 5
-30
+50 53,59,65 44,45,59
-10 PQ 3117 6
20
+40 65,68,68 35,36,40
-20 PQ 3299 7
40
+20 54,56,57 53,55,53
-40 WF-291 8
-40
+80 50,51,54 51,47,51
+20 WF-275 9
-30
+50 53,54,56 48,52,54
-10 WF-292 10
-40
+70 52,53,55 51,52,58
+10 WF-282-11
-40
+30 50,54,63 36,39,44
-30 WF 308 12
-30
+40 50,53,53 50,50,51
-20 WF-314 13
-40
+60 52,55,59 44,44,48 0
WF 351 14
-40
+70 50,55,55 40,40,41
+10 WF-635 15
-40
+40 52,54,67 45,49,54
-20 WF 307 16
-30
+30 53.55,57 40,40,43
-30 WF-324 17
-50
+40 50,53,59 50,41,43
-20 WF-696' 18
-30
+30 51,67,71 44,49,53
-30 WF-336 19
+10
+70 51,52,65 41,44,53
+10 SA 2050 20
+10
+70 50,52,65 41,44,53
+10 WF-353 21
-50
+40 54,62,72 43,52,58
-20 WF-645 22
-60
+50 50*'
39"
-10' WF 610
- These data were obtained from Weld Wire / Flux Qualification Test Results for Mn Ho-Ni filler wire and Linde-80 flux where PQ - Procedure Qualifi-cation tests and WF and SA - weld wire and flux lot qualifications.
- ' Determined from full Charpy impact Data curves (lower bound value),
l
't Table:14-2. Nozzle. Belt Dropout and Surveillance Program Mn Mo Ni/Linde 80 Submerged Arc 1 Weld Metal Drop Weight and Charpy impact Data Used-to 2ri.
Evaluate Initial Reference'Temnerature Values Charpy-Imonet Pronerties -
Drop Wt.
50 ft-lb 35 MLE-Weld ID Tuoy, F Temp...F Temp.,- F RTuoy, F-Source"' '
WF-25(9)"'
- 40 67 70
+ 7 *'
NBD WF 25(6)"'
- 10 93 125
+33*'
NBD WF-25(5)"'
- 20 47 5
-13 RVSP.-
SA-1036
- 70 59 13 1
NBD SA-1101 90 10
-7
-50 NBD WF 112 50 79 31
+19 RVSP SA-1526
- 40 40 25 NBD-WF-67
- 20 57 50
-3 NB0 SA-1585
- 50 53 10
-7 NBD WF-193
-100 B0 75
+20-RVSP WF-182-1
- 30 75 50
+15 RVSP WF 182-1
- 20 45 0
-15 RVSP~
"'These data were obtained from nozzle belt dropouts _ (NBD)- and reactor _
vessel mater.ial surveillance programs (RVSP).
"'These data have the RTuor controlled' hy the lateral expansion measurement which there is reason to believe is in error. Therefore.Lthe Charpy 50-ft-lb lower bound temperature data was'used to-set the RTuor value.
"' Denotes same weld wire / flux combination but from different reactor vessels.
Table 14-3.
Summary of Available Initial RTuor Data for Weld Metals Fabricated with Weld Wire 72105 and Linde-80 Flux
-Charpy S.R.-
Drop Wt.
50-ft-lb Source Material Time.-Hrs.
Tuoy, F;
_ Temp.,_F. -
lRTuor,:F*
Oconee 2-RYSP WF-209-1 33
-20
- + 67-
- ( 7:
' 116 Y 56 Oconee 3-RYSP WF-209-1 30
-20
+
Zion 1-RVSP WF-209-1 23
-70
+ 88
+ 28 Zion 2-RYSP WF-209-1 30
-70
+ 57-3 B&WOG-RVSP WF-70 48
-50
+118
+ 58l B&W-NBD WF-70 48
-50
' +134
+ 74 HSST-Series 3 WF 48
-50
+183
+123-Midland Beltline WF-70 23
-60
.+ 61-
-+1 Midland Beltline WF-70 23-
-60
+ 76
' +-16 Midland Beltline WF-70
'30
--60
+ 77
+ 17
~
Midland Beltline WF-70 40
-50
+ 96
+ 36-
~
Midland Beltline WF-76
- 50
-60
.+134
' : 74
+
- RTuor defined per. ASME Code NB-2331. In'all cases the RTuo, value.is controll.ed
~
by the Charpy 50 ft-lb lower bound temperature' data.
s k
w,.
e 9
5.
Table 18-4.
Summary of the Mean: Initial l RT, Values-and:
Standard Deviations-for Various. Groupings and-Subsets of Mn-Mo Ni/Linde-80 Wald Metal Data Mean. F. -Std. Dev.. F Mn-Mo Ni/Linde 80 Data Base:-
All Mn Mo-Ni/Linde 80 data
-l4.8 19.7
-less WM 72105 data All Mn-Mo-Ni/Linde 80 data
- 1.0 21.2 and WM 72105 (S.R..< 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />)
All Mn Mo Ni/Linde 80 data
+ 7.0 32.2 WF-70 and WF-209-1 Data Base:
All original WM 72105 data
+49.0 43.1 A1,1 original WM 72105 RVSP data
+29.2 27.7, 1
All current WM 72105 data
+28.8 28.2 All WM 72105-(< 40 hr. S.R. data)
+17.4 20.0 All WM 72105 (< 40 hr. S.R. data
+11.0 11.5 and less OC3 data
+
4 m..._-.._
ui i ni
1.
p 4
Figure-14-1 Distribution of. Initial RT or Values Obtained-:
r from Mn-Mo Ni/Linde 80 Submerged Arc Weld Metal--
Exclusive of Walds'Made with Wald-Wire 72105' 0_
y '.Totai-un Mo N. L noe 80 ' Data -
'O Less 72105 Weio W.ee Data
.Mean = +5F-Sen Dev-e 2'O. F i
5
,o 4
?
5 3
3 t
1 1 1
0, 0. 0 0. 0. '0. 0 - 0 0
0, 0,0,0
. t 00 50 0
'.50 100--
t 73:
1 45)
(a t t)
(*10)
.g 38i:
RT.,,F(C)
Figure 14-2 Comparison of the Distributions of Initial.RTuor Values from Mn Mo Ni/Linde 80 Submerged Arc Weld: Metal With the Initial' Values from Weld Metals Fabricated with Weld Wire -72105-
'O C--) Totai-Mn wo Natinae so Data 9
Less 72105 Wete Wire. Data -
+
a 20-F Mean. 5F Sto. Dev, WF.70 ano. WF.209.t(Wete W.f o 72105) Data Originat.oata.
.- 43 F Mean a 49F Sto Dev h[
Q WF.70 anc WF.209.ttweic We'e 6 6 4
72105) Data.C, wit:,nt Data -
1 Wean = 29F Ste. Dev =ad 8-
_o 3
-2 I
3 1
2 2
/
- cy i j
.ic i
i 3-0 h
~
II SY kh 0-0
-O 0 0 0 0.
0 500 50 0
50 100 6 73n
( 451 to t al -
4 10) --
38 FIT,..F(C)
E I
Figure 12-2 Relationship Of Initial Reference Temperature, Drop-Weight Data and Stress-Relief Time for WF-70 Beltline Weld Metal
+80 Note: RT,,ot based on Charpy lower bound data
+
~
M.. = Mean and standard deviation for Mn Mo Ni/Lince 80 data less WM 72105 cata
~
e WF 70(B)
+50 g
u.
+40 5
8 2.
[ +30
+20 g
+10 E
O M
5-i i
i
-10
-40 l
u.,
-50 E=~
-so g
70 i
I i
I 10 20 30 40 50 60 Stress Relief Time, Hours I
E g
awu = u:-
4 3
4 Figure 14 3 Effects of Stress-Relief Time on the Initial RTwo, Value of Weld Metals Fabricated with Weld Wire 72105 and Linde 80-Weld Flux compared to the Mean and Standard Deviation.for.
-Mn-Mo-Ni/Linde 80 Data Less WM 72105 Data
+90 i
a a
un uewnnw. no ou wee wu 13 05 "
ius.gon. u.'.s one'*sano.m o.ui.on w '
+B0 sal Wuf2105 a.m sitesi reu.f =40nt.
e WF 70 0 WF.209 1 88
+70
+60
' +50 1
y+0 g,,,
to C
30 o
+
+20
,y3 ic
+10 0
1 0
o M
10 0 0 20 30
- u. ' 40 5*
.e r
.50 3
60 2
70 0
0 I
I to 20 30 40 50 60 Stress Relief Time. Hours
.s-
Etquest #2:
Provide the basis'for concluding that the. increase in reference temperature for Zion weld metal may be determined from surveillance data irradiated in Babcock and Hilcox designed reactor vessels.
Resoonse #2: When all the available'trradiated RTNDT and' shift data for weld wire 72105 are presented (Figures 14-5 & 14-6), including irradiations from H and B&W vessels, alt data correlates well with Reg. Guide 1.99 Rev. 2 predictions. The available data therefore confirms that the integrated surveillance program approach is valid for B as well as B&W plants containing 72105 weld metal. The cross-referencing of tne surveillance data is allowed through the concept of an integrated surveillance program as described in 10CFR50, Appendix H, Section II.C.
The B&W Owners Group's integrated surveillance program which encompasses both B&W and Hestinghouse-designed reactor vessels is fully described in the NRC accepted report BAH-1543, Revision 3.
The procedures used to develop the trend and grouping of transition temperature shift data for weld metal HF-70 from both B&W and Westinghouse-designed vessel surveillance programs are those described in Regulatory Guide 1.99, Revision 2, Position 2.
The rationale for this approach is discussed in detail in BAH-2100, page 14-4 and the appropriate paragraphs are extracted and presented here as follows:
" Evaluation of Shift of RT g Qata Base - The surveillance data available for weld metals fabricated from-72105 constitute one of the largest sets of data for an individual weld metal-that exists in any data base. These fourteen individual data sets (fourteen separate irradiations) are presented in Section 11, Table 11-1, and repeated here in Table 14-5 for continuity, and are all the data available up to the time of this report.
All these data have been verified by a thorough review of all source documents and the fluence values are based on current re-evaluations of the capsule fluences.
Irradiated Data Behavior - The data were evaluated on the basis of the 30f t-lb temperature shif t and the n adiated 30 f t-Ib temperature.
These-data are plotted as a function of fluence and are presented in Figures 14-4 and 14-5.
For comparison, the Regulatory Guide 1.99, Revision 2 curves based on 0.35% copper and 0.59% nickel (CF=211) are included-in Figure 14-4 with the 30 ft-lb temperature shift data.
This comparison demonstrates the extreme conservatism of the Regulatory Guide methodology when compared to the actual 30 ft-lb temperature shift data.
The irradiated data is compared (11)
ZNLD/2479/4
to the prediction curves based on the Regulatory Guide (CF-184) and the initial mean RTNDT values and standard deviation, in Figure 14-5. -This-comparison also illustrates the conservatism of the Regulatory Guide methodology compared to actual data behavior and demon 3trates a more-realistic representation of the data behavior.-
The irradiated data were analyzed using the procedures described.in Regulatory Guide 1.99, Revision 2. Position 2.
The results of this' analysis based on tne 30 ft-lb temperature shift data are shown in Figure 14-6.
This-approach resulted in a chemistry factor of 174 as compared to a chemistry factor of 211 based on the typical HF-70 chemical composition of N1 - 0.59%
and Cu - 0.35%. A chemistry factor of 174 implies that the weld metal irradiation sensitivity is the equivalent of a copper content of 0.'25%
assuming Ni - 0.59%.
The procedure from Position 2 was applied to a data group consisting of the fourteen irradiated 30 ft-lb temperature data-. sets in contrast to the shift data for normal Position 2 analysis.
The~1rradiated Charpy 30 ft-lb temperature data in Figure 14-5 were analyzed using the methodology described in Regulatory Guide 1.99, Revision 2, Position 2.
A chemistry factor of 184 was calculated (see Figure 14-5).
Using the relationship that initial RTNDT plus irradiated shift equals the irradiated value, it follows in this case that the differences of_the two chemistry factors is 10.
If this is adjusted by multiplying by the fluence-factor, an RTNDT value approximately equivalent to the difference between 174 and 184 is.
obtained for the fluence of interest and is representative of the initial.
RTNDT value.
This relationship supports the selection of a value of IlF for the generic initial RTNDT value by indicating a 10F differential between the two values.
titthodology for Calculatina Shif t in RTg - The shift in RTNDT resulting from irradiation shall be in accordance with Regulatory Guide 1.99, Revision 2, Position 2.
Based on the available fourteen irradiated-data sets, the-chemistry factor for WF-70 weld metal is 174.
The margin shall.be calculated based on one standard deviation = 14F t i.e, '1/2 of Posliion 1 o3 - 28F)."
(12)
ZNLD/2479/5
l.
.D l
I Table 14-5. Mn-Mo-N1/Linde 80 Held Metal Data for Helds Fabricated Hith Held Hire No. 72105 (Data available through 5/31/91) 30 ft-lb Transition, Capsule Held
- Fluence, F. Temo. F Upoer-Shelf Enerav. ft-1b Plant Ident.
Metal n/cml_ Initial Irradiated Change Initial Irradiated Change E Survelliance Data i
Oconee Unit 2 C
HF-209-1A 1.02E+18
+4 49 45 67 51 16 22,23 Oconee Unit 2 A
3.37E+18-
+4 118 114 67
.47 20 22,24 Oconee Unit 2 E
1.21E+19
+4 183 179 b7 44 23 22,25 l
Oconee Unit 3 A
HF-209-1B 8.10E+17
+45 93 48 66 54 12 22,26 Oconee Unit 3 8
3.12E+18
+45 109 64 66 49 17 22.27 Oconee Unit 3 0
1.45E+19
+45 185 140 66 42 24' 28 Zion Unit 1 T
HF-209-1D 2.53E+18
+4 116 112.
64 56 8
22,29 Zion Unit 1 0
8.49E+18
+4 203 199 64 52 12 22,30 Zion Unit 1 X
1.26E+19
+4 203 199
'64 44
'20 22,31 Zion Unit 1 Y
1.56E+19
+4 209
' 205 64-44 20 32
[
i Zice Unit 2-
.U HF-209-1E 2.57E+18
-23 122 145 70 48 22 22,33 Zion Unit 2 T
8.04E+18
-23 168 191
.70-44 26 22,34
'Zlon Unit 2
~Y 1.48E+19
-23 208 231 70 50 20 35 Davis-Besse D1 HF-70(N) 6.63E+18
+45 180
- 135
'56
'43 13' 22,2.;
ZPHR/1161-19
._.a
,m
Figure'14-4 -Comparison of Regulatory Guide 1.99, Revision 2 " Estimate of Radiation Shift With Observed Shift for Surveillance' Caosule Data for Welds. Fabricated with Weld Wire 72105 -
405 32W f,4.uai.on Bas.o en sw Shisi of ter.oei.e CA.rpy 300 ao i-e :.*o.raiwe o.is e ** _J 272f
,, gaf (
" p
.I
{
_,.+2'W
- s.......c.
gg C
po,D "' '
gg g~
s",..-
a i
g p
3 it
.e
&a.-
~
p&
o e
g too ed,301 E
Y s'
~
te
/
4.
,y
~
i
/
'W"
. ocoa 'urwi 2 50 ffo g
a ocea urw s e B&W Owners Grouc ff_
I o Jen uaei i
/
/,d**
I a 2ea unei 2 0
SE.17 1E.18 4E+10 1E.19-4E.19^
Fluence (E 1MeV),nicm' Figure 14-5 Comparison of Regulatory Guide 1.99, Revision 2.
Position 2, Based on Irradiated Values With Observed Irradiated Data from Surveillance 1 Caosules' for Weld Fabricated with Weld Wire 72105 400 g
E v.audien Bes.o on weae.t.o Charov wi-e i.ko.r.i,. o...
300
. 2sss
- S.. -';
....a..
c4.- '-
,.3,6 j
g.g'1,'voy -
.i
,j a
?!
_ #p, W*pi',.-(e
- nil L
ioo t'l
[
s a
ed 1
got,a
, ' O**
o 2
da ~
- r!
1
. ocoa urm 2- -1,-
gp*'
[
a ocoa urui 3 i
3, e saw o.a.es Grouc a
j o 2ea van i a 2ea urus 2 t
t o
SE.t?
1E.te 4E.it 1E.19 4(*')
Fluence (E = 1MeV),ntem' 4
FigureLl4 6 Comparison of Regulatory Guide 1.99,~ Revision 2, Position 2.
Estimate of Radiation Shift With Observed Shift for Surveillance--
Caosule Data for' Welds Fabricated with Wald Wire 72105
- c. :
i,..... ear,u.o.. sn.iiei,...ieoca.rev 3ea
- 3.,...,,,. %.... o..
_. Ir.,
,,,o
,, y,,,,,, 7493_
- 2F
- s...... ne e
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Requelt_#3:
Plant Specific irradiation temperatures and spectra should be included in your evaluation of the effect of neutron irradiation on the increase in reference temperature.
Respone_ #J: Plant specific irradiation-temperatures are discussed in detall_in BAW-2166 which is the B&W Owners Group response to Generic Letter 92-01.__The applicable reactor coolant system temperatures as a function of power are reproduced here for the applicable reactor vessels.
" Material sensitivity to irradiation embrittlement is directly affected by irradiation temperature. Over the temperature range that most light-water cooled reactors operate, the irradiation embrittlement is inversely related to irradiation temperature.
However, since current generation' pressurized water cooled reactors operate over a the relatively narrow temperature range (i.e. 529-556F RV inlet temperature), the relative sensitivity of the belt line materials as a function of temperature is easily overshadowed by other parameters such as variations in material properties and Charpy impact testing techniques.
The development of Regulatory Guide 1.99, Revision 2, was based solely on surveillance data in the irradiation temperature range of-525 to 575F.
Normally, the Regulatory Guide 1.99, Rev. 2 data is applied directly in the evaluation of a reactor vessel on the assumption that the reactor vessel temperature was always within this temperature range.
However, as can be seen from a review of reactor coolant system temperature as a function of power, the inlet temperature can vary. This does not affect-the monitoring of irradiation embrittlement of the reactor vessel because.the surveillance capsules are located in the downcomer region of the reactor vessel and experience the same temperature history as the reactor vessel.
The reactor coolant system temperatures as a function of power for each plant included in this report are reviewed below.
These data were provided by each plant owner and are as stated in their respective FSAR's.
"4. L B&W-Designed 177-E M lants Figure 4-1 shows the reactor vessel outlet temperature (THot) and the reactor vessel inlet temperature (TCold) for the B&W 177-FA reactor vessels.
(17)
ZNLD/2479/6
4 This is representative of all 177-FA plants except-Davis-Besse.
!;Iese operating. limits are characterized by a constant system average-tempe.rature and an increase to the inlet temperature-(TCold);to 580F with a reduction in operating power.
These temperature characteristics result from the fact that t
initial approach to power is controlled by the water level-in the steam generator followed by a change in. operation to maintain the system average temperature constant. The increaso in inlet temperature may have the.effect of minimizing irradiation embrittlement for these plants.
1.2.
Davis-Stist Figure 4-2 shows the reactor vessel outlet temperature (THot) and th'e reactor vessel inlet temperature-(TCold) for the Davis-Besse reactor vessel.
The system behavior is similar to that of the other 177-FA plants with the' exception that the change from level control of system average temperature--is at approximately 287. power.
4.7.
Zlon Units _] and 2 Figure 4-7 shows the reactor vessel outlet temperature (THot) and the reactor vessel inlet temperature (TCold) as a function of power for the Zion Units 1 and 2 reactor vessels.
These operating limits are characterized by an increasing average temperature with increasing power levels.
The. inlet temperature decreases with increasing power and reaches a minimum at 1001.
power."
The differences in the operating-temperatures of-the B&H NSS. designs and the
~
Westinghouse NSS design for the Zion units shows an inlet temperature difference of approximately 26F; however, this temperature spread is within the allowable temperature range for the applicability of Regulatory Guide 1.99,-Rev. 2.
Therefore, it is not necessary to apply or to take special considerations of these temperature differences.
(18)
ZNLD/2479/7
Figure 4-1.
Reactor Coolant System Temperatures as a Function of Power for B&W 177-FA Plants Exceot Davis-Besse u.
I i
i 602F 600 T..
o
~
3 580F
,,,,,,,,,,,,,,,,,,,'A,j,'
,, _ 579F d
5 580 Nominalf,,,,
9 Te ie
=o 560 556F 5
540 o
532F { Hot Standby)
~
oo 520 "o3 500
- Level Control T.,,,... Control ceo I
I C
480 O
25 50 75 100 Nuclear System Power, Percent Figure 4-2.
Reactor Coolant System Temperatures as a Function of Power for Davis-Besse u.
I i
4 602F 600 7"
o 580F
'U.!' - - - -
5?-
580 Tcois o,
/
E o
560 H
556F 5
540 3
532F (Hot stancey) o o
520 o
~~
Level 500 Control Ta...... Control C
I l
I I
1 480 0
25 50 75 100 Nuclear System Power, Percent
Figure 4-7.
Reactor Coolant-System Temperatures;as a Function of Power for Westinahouse-Desianed Plants for Zion Units 1 and 2'
~
'620-4 N
600 589.4F ~
3, 4-T...
5' 580' ca.
E
! Taverase 559g,,
560 r
Tcois
~
E 547.oF
- 529.dF --
540-m (Hot Standby) oo o
520-w 1-h 500 m
ID 9
q 9 --
c-480-0 25 50- 100-Nuclear: System Power,- Percent-i t
I e
a 6
1 e
s
Likewise, the issue of the effects of neutron spectra was considered in the design of the Master Integrated Reactor Vessel Surveillance Program as_ described in BAW-1543, Revision 3 (approved _by NRC) and_ls extracted as follows:
"Itte._r_ththe neutron _ energy _snentrum is a function of-the' geometry and materials-of the reactor internals components.
As shown in' Tables 3-2 and 3-6, the materials of the reactors are-the seme, but the dimensions of the internals vary and will produce some variation in neutron spectra.
Differences in neutron spectra, however, are not unique'to integrated programs. A surveillance program for a single reactor must contend with the variation in spectrum through the reactor vessel wall and the difference in spectrum between the vessel and the survel!1ance capsule.
In the integrated program, the difference in spectra between the hestinghouse and B&W reactors is no larger than that already encountered in non-integrated programs.,The difference in spectra could be ignored if a perfect damage exposure index-existed.
Since this is not the case, it is desirable to hold the difference in spectra to a minimum and correlate damage in a specimen _with different points in the reactor vessel wall using the best available damage function.
The effect of spectra variations will continue'to be evaluated in the integrated program just at it would also continue to be evaluated in a non-integrated program.
As an example of the evaluation to be considered, the neutron energy spectrum at the reactor vessel for a typical Westinghouse system design is compared with a typical B&W system design and is shown in Figure 4-1."
(21)
ZNLD/2479/8
=
b b'
f J
Fioure 4-1 Soectral= Comoarison for Westinahouse and BW Plants
- :. _:._::.:..-.=-
E=-
~-
-fSpectral ComparisonLfor.== =-
~
Wand B&W Plants ih.-.
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N Table 3-2.
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lip,L Comarison of Plant Parameter 5 for the USW 177-FA RV5Ps (f,
e t
- W) LS.
A V.
i 1,
y
~'Y Davit-Passe Raa(% Scro M aatat weclear tryst al af ter g
Pla G r3 %tfCs
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_ On t hit I
_ tNit.3 _
fUtPet.th.lt l _ Ctow tMIL Z_ 9tl*tJhlt ). Isltad.9k1L1 t
1% lille 4
l Destge heat output (core).
7772 2772 2548 2544 2568 r%8 2563 iMg 4
MWL Q,h,
)
$3 j
Design overpower 1L 112 117 112
!!?
112 Ilf ill lif l 1 95 S'S -
i Systee pressere (nom), psla 7700 7200 7700 2200
??00 2200
??M 7700 8
U Nj-;
l Coolsat;flowrate. 106 gtsyh; 143.8; 387.000 143.8; 387.000 139.7: 375.000 139 6; 375.000 139.F; 375.000 139.7; 375.000 139 T 375.000 139.73 375.tkw ",
gpg _
' C. pip i.
,u Coolant temperatores f
!i
. msw j
stoel'eal!Inle t 558 558 555
!"6 556 556
' 554 554 Avg rise in vessel 49 49 47 46 47 47 47 47 Avg la,vesset 582 582 579 579 579 579 579 579 4
g3 No. of fueltassemblies 177 lif 177 477 171 IFF IFF IIF M*lh Typeafgfmel' assemblies Mark 8 (15:35) Mark 8 (15:15) Mark 8 (15:15)
Mark n (15:15) Park 8 (15:15)
Mark 8 (15:11) Meet 8 (15:15) MattI(15:151 e'
s Wf L
Core barrel'ID/00. In.
147/145 148/145 141/145 141/145 141/145 141/I45 3
141/145 148/141 Mrd i
ThermalishieldID/00.le.
147 I58 147/151 147/851 147/151 14r/151 147/151
/
j pep Core strectoral character-
- 147/851 147/111-istics ' g.,
~m. va,
l Cere'equiv diam. Is.
1;8 e 178.9 128.9 128.9 178.9
!!8.9 lit.1 Iti.9 g
Care active feel height.
143.2 141.8 141.8 141.8 141.8 141.8 141.5 142.3 In.ht e
kh I
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9'2 Table 3-2.
Comparison of Plant Parameters for the 85W 177-FA RVSPs (Cont'd) p fU l g-4e
- i: I
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~'vy.g kW L Cawls Sesse Beache N o Arkansas sheclear Crystal S6ver lhece till, Jg}
Plant Parneeters Unit i Jhell I
_One th t t i 141t 3 lyne.tynJLL 059 ate tMla Dar+te IMit 3. Island 3kdLL ' *4 ~
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Se xi...essei desi,a 4df'.
parameters
-Ja-(g.
g d (
prlacipal material SA508, Cl.2 SA533 Tp 8 Cl.I 5A533 Ip 8 Cl.1 5A533 fp 8 C.I 5A302 Gr8 Cl.IIU 5A508 C3.2 5AM8 Cl.2 SA302 Cr8W k3 Design pressure, psl8 2500 2500 2500 2500 2500 2500 2500 2500 t'
~
p Oi Design temperatere, f 650 650 650 658 450 650 650 650
'i, E*.
' ?3 Shell 10, la.
171 lit lll Ill.375 til til til 111 NQ Shell thickness,' In.
8.4g'I 8.44 8.44 8.44 8.44 8.44 8.44 8,44 I 4
,;Q 00 across non les, In.
4 261 2.49 2.49 249 241 241 249 241 p-d Overallcesseglesere 40* 8 7/8" 4G* 8 T/8*
40* 8 1/8*
40* 8 1/8*
40* 8 1/8*
40' 8 7/8*
48' 8 7/8*-
43* 8/18*
-Q3 head belght, ft in.
,4.,.
Core barrel-thermal shield Type 304 55 Type 304 55 Type 30t 55 Type 304 55 Type 304 55 type 304 55 Iype 304 55 lype 334 15 L-
-g.
4 4 principal matertal v.,
i 4
IN.w4 c
>4 k
I*I0ver cladding and lastrumentatlee norries.
M Wid
.Fg (b)As modified by Code Case 1333.
W
- J; L
<5 t
J vi (c)for Dawls-Besse Unit I this is the nosteal 00 across the inlet nortles. The a) across the outlet morales is 245 laches.
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Comoarison of Plant Parameters for the Westin3 ouse RVSPs h
i Table 3-6.
M i
~ 'l t
?'"
5 f
Point pe u h re at peach lasker Point larley Point l
R anp_
tMt I tin i t l _, _Un it J _ _t%JL 4...
$vrrYthtt.J $n athlU [Jeg AIQ g} q y PlantJara~ters 1570 M 15181.5 1515.5 2200 2200 2241/2546 2241/2546 3250/3391 32%C/3111 4
u l
taprated)
(upr ated)
(aprated)
(* prated)
Des tga l', eat et.trut (core). MWt
'i 5
IlO lie llo 110 lie lie llo llo 113 I
i 8-E
'r 2235/le85 2235/1985 2235 2235 2235 2235 2215 2235 t
Design cuerpo=*r.
l f
2235 Systes pressure ( m vnal). psig 101.*:214.000 110.2:271.900 110.7:271.900 114:364.500 135:344.50n i
toolant flow rate. 10 lb/h; 9pm 31.2;85.700 66,1:lRO 000 66.T:lf 0.000101.5:274.000 I
f 6
-l 546.S c. _
552.5 552.5 546.2 546.2 543 543 5y*.2 530.2 f
Coolant te geratures, f 7
Avg rise :n =esset 573.8 9..
5T.6 57.3 55.9 55.9 62.8 6L8 64 64 i'
54 574.2 574.4 574.4 562.2 562.2
.t
% inal inlet i
610.1 610.1 514.2
{
Avg la vessel s
121 7 121 128 157 157 IST 157 193 193 14:14 14:14 14:14 15 15 15:15 15:15 15:15 15:15 15ml5
[
l Mo. of fuel asseelies
'i
)
type of f uel a se+blies ofA/V.5 ct A otA/10rAR OI A/torMt St aMard Stem ard MP ora l
Of A/Y,.51 tore barrel 10700, in.
109/132.5 109/112.5 109/112.5 133.9/131.9 133.9/137.9 133.9/137.9 133.9/137.9 14f /152.5 148/112.5 l
Assembly design 115.3/!22.5 115.3/122.5 115.3/122.5 142.5/145.0 142.6/le.0 142.6/148.9 142.6/las.0 158.5/164 154.5/164 j
thermal -hf eld it'/CO. in.
f l
-(-
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96.9 -
96.1 96.I 189.2 819.2 119.7 119.2 132.2 132,2 i
(cre structural characteristles
,47 141.4 144 144 144 144 144 144 144 144 l
(cre act he f uel height. (n.
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{e-l Comparison of Piant Parameters for the Westinnhouse RVSPs (Cont'd)
II)
.i Table 3-6.
me t
f I
6-i i
p'c rotat tteeth reint Beach lerkey Psiat ischer relat r*
Plaat Parneglers R.L CLesA _ 3 alt 1
_U91L L _ Ma_it 3 J_tt 4 lern1hlU lyrn SaiLZ L19=. % L1 lie LRan I A-508 C12 5A 302 Cr8 A 508 Cl2 A-509012 A 508(12 A 533 Cr3 A-533 Cr8 A 533 Cr8 A-533 Cv5 N[
l Reactor vessel destge parameters (18 til til (Il -
T' i
Prfacfpal material
[
2485 2435 2485 2485 2485 2895 2495 2485 24:5 i
650 650 650 650 650 650 650 650 650 IJ2 132 132 155.5 155.5 157 157 173 gr3 f
Destge pressure, pstg i -
Oesign traperature,f 5.50 6.50 6.50 7.15 T.15 F.F5 7.75 p.44 g 44 Shell 10. la.
0D across Inlet /oottet mottles. (s.
230/219 230/219 230/211 174 lit 174 Its 262/258 ts2/258
+
t i
Shell thldness, la.
Overall vessel-closure beed helght 39' l.3*
37' 5*
37' 5" 42* 2*
42' 7" 40' 5*
40' 5*
43* 1.7*
43' 5.;*
'l; '
~
Core barrel thermal shleid principal A240 A240 A240 A240 A240 A240 A240 A240 A240 materfal Type 334 lype 304 Type 304 Type 304 Type 304 Type 304 type 304 type 304 Typ,3g4
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Re_ quest #3:
Provide the Charpy impact data and curves that were used to determine the increase in the 30 ft-lb transition temperature.
Response #4: The charpy impact data and curves are found in BAH-1803, Revision 1 and in surveillance capsule reports.
These data and_their-interpretation are.
included in BAH-2100 for weld _ metal HF-70. All Charpy data were reevaluated to provide a consistent approach on the determination of transition temperatures and upper shelf energies.
The appropriate paragraphs as extracted from BAH-1803, Rev.
I are presented as follows:
"4.4.
QAtA_lD_teIDIe_titi.OD 4.4.1. Charoy A tc_t Curves The Charpy impact data were obtained from capsule evaluation reports issued from 1969 to 1986.
Since these evaluations were performed at different test facilities, the data received different-interpretations which were influenced by the experience of the evaluator; to minimize _the effect of this, all the Charpy data were re-evaluated.
The Charpy data and re-evaluated curves are included in Appendix A.
Some of the reported' transition temperatures and C VSE values were found to be inconsistent with the reported Charpy data; y
other values were found to be different because of the interpretation of the data.
In those few cases where the data-were insufficient to define the complete upper-shelf region, our experience with these materials was-drawn upon to extend the data in a rational manner; these extended curves were drawn with a broken line.
4.4.2 Transition Tegnertt.urg To determine the effect of neutron radiation on the reference temperature, the shift in the Charpy energy curve tempeiature at 30 ft-lbs is measured, in agreement with the current requirements of 10CFR50, Appendix G.
This is a change from the original requirement of making the measurement at 50 ft-lbs.
(25)
ZNLD/2479/9
._=.._ -.
Whenever the. upper-shelf energy decreased to below the 50 f.t-lbs value, itI was impossible to measure a shift.
The new. requirement also states that the-shift measurements is to be based on the' average Charpy curve of the:
irradiated material relative to that of the unirradiated material.
The charpy shift data for the various Linde 80 weld metals include all the data from the B&H-designed reactor survelliance programs and all-the-data from the various pressurized water reactor surveillance programs that were established for reactor vessels fabricated by B&W.
The Charpy shift data for each weld metal arc prGsented in Tables 4-2 and 4-3.
The Charpy 30 ft-lb shift data for the B&W-designed NSS reactor v!stel surveillance programs are plotted as a function of fluence'in Figure 4-1.
The value: beside each plotted point is the copper content.
Similar data is shown in Figure 4-2 for non-B&W NSS reactor vessel surveillance programs.
These data are also plotted as a function of fluence along with their corresponding copper contents.
4.4J Upper-Shelf _ Energy The effect of neutron radiation on the fracture toughness of the weld metals, in the region where the material behavior is defined as plastic,:is measured by the decrease in C USE.
The fundamental difficulty with the measurement of y
this characteristic of the Charpy data is the lack of a standard definition of what constitutes the " upper-shelf",I The main reason for this is that Charpy curves exhibit varying characteristics (shapes) in the region of interest.
This can be seen by examining the curves in Appendix A.
For the purpose of this study, all C VSEs were defined by a_ combination of y
(1) an appreciation of the general characteristics of the behavior of the curves for a given weld metal, (2) comparison of the upper-shelf values in the temperature range of 150 to 200F above the temperature where 1001. ductile failure was observed, and (3) particular ovpertise gain M by working with and evaluation of these weld metals.
In nest cases, the values selected could be argued to be slightly above or below those chosen; however, this method provided consistency to the data Interpretation.
(26)
ZNLD/2479/10
r The decrease in C USE values is summarized in Tables 4-2 and 4-3'.
These y
tables include all the data from.the various pressurized-water reactor-surveillance programs for B&N-fabricated reactor vessels.
These data have-the same chemical composition as the' shift dhta.
The irradiated Charpy USE.
data for the B&Hadesigned NSS reactor vessel surveillance programs are-1 plotted as a. function of fluenco in Figure 4-3 and the corresponding non-B&W?
NSS data _is presented in Figure 4-4.
The copper contents of the data is
~
presented beside each data point."
I At the' time this work was in progress the ASTM-Committee Elo.02 had not completed formalizing-a definition of upper-shelf energy.
The_ process used in this report closely approximates the procedure recommended by ASTM.
(27)
ZNLD/2479/ll
-Renue11_t5:. Identify the neutron transport code, the-scattering. cross-sections-and quadrature 1 approximations used in determining the neutron fluences of the capsules.
Responie_f5: Neutron' fluence-analysis information, including-the neutron transport codes, scattering cross-sections and quadrature-approximations are found in the references identified in BAW-1803, Rev. 2, Tables 4-4 and 4-5 (see following page).
These data and their interpretation are included in BAW-2100 for-tLe weld metal HF-70.
The appropriate paragraphs as extracted from BAW-1803, Rev. 1, are presented as follows:
"1,1._ _Jeler tlen.sLElv e n c e_Valv e.s Fluence values for surveillance capsules are continually upgraded as t.etter-techniques are established for determining reector flux and fluences.
Surveillance capsule fluence values have been reviewed in detail as a part of a Nuclear Regulatory Commission (NRC) sponsored dosimetry improvement programs.
As a result of these reviews, revisions to the older fluence values have occurred.
The fluence values for this study were based on two considerations.
For B&W plants, fluence values were used as reported since these procedures were benchmarked as a part of a NRC sponsored program at Hanford Engineering Development Laboratory (HEDL); for some of the older-(non-B&W) plants, capsule fluence values were updated, when available*, to agree with the values calculated by HEDL. More-recent Westinghouse fluence values were used without change since their procedures were benchmarked against HEDL as part-of their dosimetry pcogram.
In the case of the latest available capsule data, the capsule fluence values were obtained from published capsule reports.
A listing of-the sources of the fluence data are presented in Tables 4-4 and 4-5."
Applies to Turkey Point 3, Capsule V.
L (28)
ZHLD/2479/12 l -
l i
Jable A-4 _50urtes._oL5urycillance.Dattfor_BLH-Des _lgned3Jants (Dtta available through 1/1/90)
Source for Source for P1&n11CAplulf._-
Charky. Data Bfit.rface*
J.httle_
Referencs*
Oconee Unit 2 - C BAH-1437 A-4 ORNL/NRC-EDB 18 Oconee Unit 2 - A BAH-1699 A-5 DAH-1699 A-5 l
Oconee Unit 2 - E BAH-2051 A-6 BAH-205)
A-6 Oconee Unit 3 - A BAH-1436 A-7 ORNL/NRC-EDB 18 Oconee Unit 3 - B BAH-1697 A-8 BAH-1697 A-8 Oconee Unit 3 - D BAH-2128/1 31
' BAH-2128/1 31 Davis-Besse
- Di BAH-1920P A-21 BAH-1920P A-21 1Abic 1.54_Soutces_oLSutyei11 anes _DA.ta_ lor-_He111nghaust:Dc31gngd P1ani5 (D6ta available through 1/1/90)
Source for Source for Plan. TIC 3Asule Charov Data Reference
- Llutact_._
Reference
- Zion Unit 1
-T BMI A-41 HCAP-10902 30
)
Zion Unit 1
-U HCAP-9890 A-42 HC%P-10902 30 Zion Unit 1
-X SHRI A-43 HCAP-10902 30 Zion Unit )
-Y BAH-2082 32 BAH-2082 32 Zion Unit 2
-U BMI A-41 HCAP-10902 30 Zion Unit 2
-T SHRI A-45 HCAP-10902 30 Zion Unit 2
-Y WCAP-12396 33 HCAP-12396 33 ORNL/NRC-EDB - NRC embrittlement data base-at ORNL, NEDL - Capsule fluence validation snne as part of LHRDIP.
.5
- Reference numbers prefixed with the letter A are located in Appendix B.
1 ZNLD/2479/13
Chemical Coccosition of Weld Metals in Data Base Used to Develop Correlation Models Table 4-1.
Chemical Composition. w/o Item Plant Weld ID C
.En_
P 5
Si Cr Ni Ho_
Cu I
Oconee Unit I WF-II2 0.08 1.47 0.016 0.015 0.54 0.07 0.59 0.40 0.32 WF-209-1A *I 0.11 1.55 0.022 0.010 0.65 0.09 0.58 0.39 0.36 I
2 Octnee Unit 2 WF-209-1B(a) 0.08 1.63 0.017 0.012 0.61 0.10 0.58 0.39 0.30 3
Oconee Unit 3 3
1HI Unit 1 WF-25 a.09 1.02 0.013 0.015 0.46 0.10 0.66 0.40 0.33 IN 0.08 1.65 0.021 0.013 1.00 0.07 0.10 0.45 0.41 5
Crystal River-3 WF-209-gI 6
ffl0 t' nit I WF-193A 0.09 J.4?
0.016 0.016 0.51 0.06 0.59 0.39 0.28 7
Rancho Seco WF-1938 *I 0.09 J.49 0.016 0.016 0.51 0.06 0.59 0.39 0.28 I
8 Davis-Besse WF-182-1 0.09 1.69 0.014 0.013 0.41 0.15 0.63 0.40 0.21 9
Pt. Beach Unit 1 5A-1263 0.09 1.47 0.019 0.024 0.49 0.13 0.57 0.39 0.22 WF-193C 'I 0.08 1.40 0.014 0.0!3 0.55 0.07 0.59 0.39 0.25 I
10 Pt. Beach Unit 2 11 R. E. Ginna SA-1036 0.08 1.41 0.012 0.016 0.59 0.09 0.56 6..i6 0.23 1
12 Turkey Pt. Unit 3 5A-1101 0.08 1.55 0.019 0.008 0.59 0.16 0.54 0.38 0.21 3
13 Turkey Pt. Unit 4 SA-1094 0.10 1.44 0.014 0.011 0.50 0.!4 0.60 0.36 0.30 0.09 1.51 0.020 0.013 0.68 0.06 0.57 0.39 0.35 WF-209-ID(I,I 34 Zion Unit 1 3) 0.08 1.51 0.017 0.013 0.68 0.Go 0.57 0.39 0.30 j
WF-209-1E 15 Zion Unit 2 16 Surry Unit I SA-1526 0.09 1.53 0.013 0.0D
'0.53 0.98 0.68 0.42 0.35 17 Kori Ur.it I WF-233 0.10 1.45 0.021 0.015 0.42 0.08 0.68 0.44 0.27 18 85W Owners Group WF-25 0.09 1.50 0.015 0.016 0.54 0.09 0.67 b.42 0.35-19 B&W Owners Group WF-67 0.08 1.55 0.021 0.016 0.58 0.10 0.60 0.40 0.22 20 B&W Owners Group SA-1585 0.08 1.45 0.GI6 0.0!S 0.51 0.09 0.59 0.38 0.21 21 B&W Owners Group WF-70 0.09 1.63 0.018 0.009 0.54 0.11 0.59 0.40 0.42 l
22 B&W Owners Group WF-ll2 0.08 1.47 0.016 0.015 0.54 0.07 0.59 0_4c 0.32 23 B&W Owners Group SA-Il35 0.08 1.45 0.011 3.013 0.49 0.08 0.59 0.38 0.27 l
D I')Same weld wire / flux combination used for more than one surveillance weld metal witti different processing treatment. Letter following identification signifies that welding parameters, or fabri-cation variables, may cause differences in properties between weidments with identical wire-flux gg sE combinations.
IAtypical weld metal.
.tI a
o L-
b
[
BBm5 E
a m
25 m a a
a a
e asasan.as I
j 4
Table 4-2.__Linde 60 Weld Metal Data frn 5&W-Desicned-Reactor-Vesse! Surveillance Programs f
(Data available through I/1/9G) i 30 ft-Ib Traisition. F. Tere. F Uerer Shelf Enercy. ft-Ib l
Capsule Weld Fluence.
E Initial Irradiated. Change intilal Irradiated Change i
Plant ident.
Metal n/cm i
Oconee Unit 1 E
WF-ll2 1.50E+18
-5 73 78 64 55 9
- Oconee Unit I A
8.95E+18
-5 185 191 64 52 12 Oconee Unit !
C 9.85E*18
-5 ISO ~
185 64 52 12 Oconee Unit 2 C
WF-209-IA 1.02E*18
+4 49 45 67 51 16 Oconee Unit 2 A
3.37E518
+4 118 lit 67 al 20 Oconee Unit 2 E
1.21E+19
+4 183 179 67 44 23 Oconee Unit 3 A
WT-209-1B 8.10E+17
+45 93 45 E6 54 12 I
Oconee Unit 3 8
3.12E+18
+45 109 64 66 49 17 L
IMI Unit 1 E
WF-25 1.07E*I8
-56 68 124 81 64 17
~
,i IMI Unit I C
8.66E+18
-55 147 233 81 50 31 i,
AND Unit I E
WF-193A 7.f!7E+17
+5 110 105 73 58 15 AND Unit !
A 1.0X*19
+5 155 151 73 45 28 l
Ano Unit I C
I.4GE+19
+5 190 185 73 47 26 i
Rancho Seco 8
WF-1938 3.99E+I8
-14 85 99 69 51 17 4
Rancho Seco D
6.ECE*IS
-14 138 152 ES 53 15 Rancho Seco F
I.42E+19
-14 152 166 68 48 20 Davls-Besse F
WF-182-1 1.16E*18
-11
!!S 127 10 64 6
Davis.8 esse 8
5.92E*18 114 125 70 57 13 Davis-Besse A
1.29E*19
-11 164 175 70 62 8
Crystal River 01 WF-25 7.79E+18
-20 194 214 73 48 25
.k Crystal River Cl WF-67 6 09E+18
+13 173 160.
72 57 IS 4lll:
Erystal River Cl SA-1585 5.10E+18
-27 121 148 77 55 22 Et
$E Dawls-Besse DI WF-10 6.63tel8
+45 180 135 56 43 13
~ '
g3 Davls-8 esse DI hT-I!2 8.21E*18
-52 152 204 79 50 29 Davls-Besse DI SA-II35 1.03E*19
-39 103 142 76 55 21 g
.t f
t
~
j..
. Table 4-3.
Einde 80 Weld Metal Data from Westirchouse-Desinned-Reactor-Vessel Survelliance Programs (Data available through I/1/90) 30 ft-Ib Transition. F. Temo. F Uccer Shelf Enen. r. ft-Ib Capsule Weld
- Fluenge, Plant ident.
79etal n/cmd initial Irradiated Change lattial Irradined thange Pt. Beich Unit I R
54-1263 2.10E419
-45 120 165 65 53 12 Ft. Beech Unit 1 5
7.58E+18
-45 120 165 65 54 II
(
Pt. Beach Unit 1 T
2.llE+19 '
-45 130 175-65 55 10 Pt. Beach Unit i V
6.20E+18
-45 65 110 65 54 11 Pt. Beach Unit 2 R
WF-193C 2.15E*19 0
235 235 66 48 18 Pt. Beach Unit 2 i
8.36E*18
~0 150 150 66 56 10 Pt. Beach Unit 2 Y
6.14E-18 0
165 165 66 42 ~
24 R. E. Ginna R
SA-1036 1.02E419
-25 140 165 80 51 29 R. t. Cinna i
1.78E+19
-25 125 150 83 53 27 R. E. Ginna Y
6.53E418
-25 115 140 80 53 27' Iurtry Pt. Unit 3
'T SA-II0l 7.0lE+18
+12 116 164 -
66 62 4
Turkey Pt. Unit 3 Y
1.23E419
+12 190 178 66 48 18 ivrkey Pt. Unit 4 i
5A-1994 -
7.54E-618 0
224 224 68 43 25 Zion Unit.I I
WF-209 2.53E+I8
+4 116 112 64 56 8
Zion Unit l' U
8.49E*18
+4 203 199 64 52 12 i
Zion Unit !
I 1.26t+19
+4 203 199 64 44 20 Zion Unit 2 U
WF-209-IE 2.57E+18
-23 122 145 70 48-22-Zion Unit
- T 8.04E*18
-23 168 191 TO 44 26 Sorry Unit I T
54-1526 2.86E+18
-15 152-167 70 53 17 D
Kori Unit I Y
WF-233 i
Kort Unit I
_i 4.67E*18.
-6 185 191 66 48 Is I.08E419
-6
'181 187 66 42.
24 Kori Unit I 1.21E+19
-6 216 222 66 47 19 1
1 1
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i APPLNDIX A e
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9 8
ZPHR/1161-18
~
Data Set A-5.
Oconee Unit-2: Unirradiated Charpy Impact Data for Reactor vessel Surveillance Program Weld Metal WF-209-1 (See Reference A-41 i
Test Absorbed lateral Shear Specimen Temp.,
Energy.
Expagston, Fracture.
No.
F ft-lb 10 in.
1 i
E.
EE014 361 67.0 53.5 100
}n g
EE032 361 72.0 52.5 100
.EE035 360 65.0 49.0
.100 EE019 202 63.0 49.5 100 s
EE036 201 67.0 45.0 100 g,,
EE016 198 64.0 46.0 100 I.
EE020 121 68.0 47.0 98 5..
s EE0li 120 66.0 47.0 100 i
EE006
~120 66.0 44.0 100 m
EE031 81 60.0 42.5 75.
- m..,,,,
EE015 80 60.0
'42.0 80 e,.n...uv
~
EE012 80 55.0 32.0 60
'," [ " *' 7 -
^
EE024 0
27.0'
.25.5 25' EE022 0-31.0 28.5 20 g ~"
g g
m EE006 0
25.0 23.0
- 15 I"
g%
i" 59 e.ee se,e.e.e ee na e.m w,,, WI JWB 9 ee 4-W emu
,g
e 2, Ca Charpy Impact Data for Weld Metal WF-209-1. Irradiated to OconeeUnjgn/ca'psuleC:(E > 1 MeV) (See Reference A-4)
Data Set A-6.
1.02 x 10 Test Absorbed Lateral Shear
'~
. Specimen Temp.,
- Ener9y, Expagston,
- Fracture, No.
F ft-lb 10 in.
j EE009 321 51.0
,37.5 100
)e EE028 231 54.0 39.5 99
~
- EE040 180 51.0 35.0 96 EE004 139 50.0 32.0 97 j..
_j..
EE034 115 40.0 25.5 75 EE025-86 38.0 25.0 55 a
- EE017 60 32.0 24.0-
'35 I
EE001 40 28.0 20.0
'25 I
in ses, __.
ee t,...._ =. :.
f t M anet
- '3W
~
o "f,EWr ep. E W
i, n.....et__
.I" t.-sts e_11'1'"
se g
.zjw e
sg,
_g.
,e
,.. i.er..".r..'
~
- -m i E
1 i
a a
s
__i_ _ _
Data Set A-7.
OconeeUnlg-2,CapsuleA: Charpy Impact Data for Weld Metal WF-209-I, Irradiated to i'
l-3.37 x 10 n/ca' (E > 1 MeV) (See Reference A-5)
I' Test Absorbed Lateral Shear
='
Specimen Temp.,
- Energy, Expagston, Fracture.
No.
F ft-lb 10 in.
il-
~
EE013 38 17.0' 18.5 10 l-j
=
EE003 75 19.0 28.5 20 EE021 110 32.0 33.0 65 EE039 140 33.0 34.5 90 f..
EE029' 193 42.0 48.0 100
},,,
EE002 226 46.0 44.0 100 j
i 5
EE005 284 46.0 46.0 100 s o.:
~
EE033 408 44.0 44.0 100 s
IN e
a s
- saia seisme -
.n.... vu t,, ein n-sss. 8,_
e, e m..
- w
.i. * 'c,en.
_t t":s+.
4
.. ~. - - - -
er ij ee s
5y.
' ~ ~
m
_l.
w
,,, i. n. =. i. -
a 4
e a
ir 2, Ca Charpy Impact Data for Weld Metal WF-209-1, Irradiated to OconeeUnjgn/ca'psuleE:(E > 1 MeV) (See Reference A-6)
Data Set A-8.
1.21 x 10 Lateral Shear Test' Absorbed specimen Temp.,
'Encrgy, Expagston, Fracture.
No.
F ft-lb 10 in.
i ii EE007 110 18.0 15.0
.40 ln EE023'
.160 26.0 27.0 70 EE030 200 33.0 26.0 100
}.
EE038 260 41.0 31.0 80 EE037 330 42.0 38.0 90
_[,,
EE018' 400 46.0 41.0 300 j
^
EE027 450 42.0 46.0 100 5
sew EE026.
550 45.0 43.0 90 I
sw
- men sesse -
e,en... E _
r, en n see s.a _
w e
.m n
. q.
,,,,, u ri n i,, _ "
- o ha s
aE" g
e w..=+=
s.
- .2 se".A.#
. w w.*
a e
a e
W p
9 99 m
M 6
eene essee.m.
r
.w.
m
-m-..
Data set A-9.
Oconee Unit-3: Unirradiated Charpy Impact Data for Reactor Vessel Surveillance Program Weld Metal WF-209-IB (See Reference A-71 Test Absorbed lateral Shear v,
Specimen Temp.,
Energy.
Expagston, Fracture.
No.
F ft-lb 10 in.
j s
JJ063 359 57.0 49.0 100 g3 JJ083 357
'58.0 48.0 100 JJ089 357 65.0 53.0 100 JJ091 201 57.0 46.6 100 s,,,
JJ053 201 63.0 50.0 100 j,,
4 j,,
8 JJ084 199 78.0 57.0 100 JJO93 140 60.0 46.0 100 5s.=
JJ060 140 69.0 52.0 100 i
l JJ065 140 61.0 49.0 100
=
JJO92 110 64.0 46.0 100 JJ088 110 58.0 46.0 100
,,, m...
_4 =
l"["[ k JJ077 109 47.0' 37.0 96 JJ064 80 61.0 40.0 98
- * ' ' ' ' ' " " ~ ' "
u g"
JJO95 80 38.0 32.0 95 i
=
JJ081 79 48.0 35.0 85 i*
s JJO90 40 28.0 28.0 10 i*
JJ069 40 33.0 29.0 15 l=
JJ075 40 21.0 20.0.
12
=
JJ061
- 40 17.0 16.0 5
C C ~ ~' ' * **
JJ050
. 16.0 14.0 2
JJ057
- 39 19.0 18.0 6
1
I I
Da'taSetA-10..OconeeUng-3,CapsuleA: Charpy Impact Data for Weld Metal WF-209-1, Irradiated to 8.10 x 10 n/ca' (E > I MeV) (See Reference A-7)
Test Absorbed Lateral Shear
-Specimen Temp.,
- Energy, Expagston, Fracture.
No.
F ft-lb 10 in.
i
. i 5
JJ014_
359 62.0 51.0 100
{e JJ015 279-56.0 47.0 100 JJ017 199 54.0 47.0 99 JJ041 169 54.0 45.0 99 j..
JJ027 140 46.0 32.0 94 j..
s JJ046 124 50.0 30.0 90 JJ028 109 53.0 40.0 97
-f,,
JJ071 95 26.0 16.0 45 I
JJ029 84 34.0 28.0 25
~ " " " " " ~ ~
JJ085 60 25.0 20.0' 15 JJ036 39 37.0 26.0 28 v m== -=*
= -,,m,,
.mr JJO96 21 28.0 21.0 14
,,,,,,,,,,gy__
4.w.
P M-*.
4 n.,,
W
=
%eveean SPgg egyget.S 4.de 99
%g WW -J8. 9 gre g
een a no esen y
-ww--
q-
-pr-r w-v-'-
-e,-
m
.A a2 m.
m m
Charpy lapact Data for Weld Metal WF-209-1, Irradiated to Oconee Unlg-3, Cy(sule 8:E > 1 MeV) (See Reference A-8)
Data Set A-II.
3.12 x 10 n/ca'
" Test Absorbed Lateral Shear
- Specimen
-Tew.,
- Energy, Expagston, Fracture.
~'
No.
F ft-lb 10 in.
1
(
E.
2 JJ032 0
12.0 12.5 15
)r JJ048 2
19.0 14.0 5
JJ002 38 15.0 14.0 5
JJ019 75 24.0 23.5 20 5.
JJ006 100-30.0 29.0 85 l.
- JJ031 146 34.0 34.0 70 j,,
JJ007' 175 46.0 44.5-91 5
JJ020 '
194 51.0 51.0 100
}-
I-JJ013 230 45.0 44.5 100
" ' * * " " ~
JJ047-280' 44.0.
.46.5 100
==
JJ003 440 52.0 50.5 100
,,.. n..
~
JJ026 580 54.0 58.5 100 "Z1
,g=
c.-m w e._
8
~ - -
5 5~
49 W
Je We64 en e4 6eep 99 stesse,me *.!***=4.='
s.
e
..m.
~
A J..
e
~
Data Set A-33.
Davis-Besse, Capsule 01: Unirradiated Charpy impact Data for Reactor Vessel Survelliance Procram Weld Metal WF-70 (See Reference A-21)
Test Absorbed lateral Shear Specimen Temp.,
- Energy, Expags'on,
- Fracture, No.
F ft-Ib 10 in.
i.
PR283 40 27.0 24.0 10 je PR319 75 35.0 30J 40 PR249 118 50.0 43.0 100 PR748 156 58.0 52.0 100 j.
PR320 222 55.0 53.0 100 j..,
__ j PR318 378 55.0 55.0 100 j,,
PR282 549 71.0 69.0 100 l,,
HR291*
O 21.0 18.0 5
3 HR295*
40 30.0 18.0 30 sw
- " " " " " ~
HR293*-
100 47.0 44.0 80
-ser HR2%*
100 35.0 24.0 55 s,.. n.. -* *
' f, tw,,eeat18F HR304*
100 46.0 31.0 85
,,,,,,,, n HRZ97*
150 42.0 28.0 83
) * **"' 7 "
~~
HR302*
150 52.0 36.0 100
~
HR294*
200 58.0 51.0 100
+
a HR298*
200 54.0 41.0 100 i
HR301*
275 55.0 42.0 100 j=
HR292*
350 55.0 55.0 100
=
HR299*
350 55.0 42.0 100 m
w s
,<.u s
=-*
e.
e.
m *-re
- HSSI program data
=
r.
Swe..
Capsule 01: Charpy Impact Data for Weld Metal WF-70, Irradiated to Davis-Besgg,/ca'(E>1MeV)(SeeReferenceA-21)
Data Set A-34.
6.63 x 10 n
Test Absorbed lateral Shear r
~
~
Specimen Temp.,
- Energy, Expagston,
- Fracture, No.
F ft-lb 10 in.
j
=
PR279 40 12.0 12.0 20
}.-
PR278 68 16.0 12.0 10 PR268 104 20.0 16.0 30 ea j
PR287 152 28.0 25.0 50 PRZ85 200 33.0 28.0 50 j..
PRZ84 225 29.0 25.0 40 PR267 275 33.0 32.0 100 g,,
PR263 300 47.0 39.0 100 PR273 350 47.0 47.0 100 in
- "" """ ~
PR286 400 38.0 34.0 100
==
e.,,
ea-PR272 500 40.0 37.0 100 g i n... -r* "
g rw
.A
- g... = =
.;, 8B '(q w 4 9 f 9 1D, e7 B t-
.1-E1-j.
W N
welf8tr tt.49ese99 e
- 6.*hw' F h Me WI:M...
ses e
ese e
ius
Data Set A-57.
Zion Unit-1: Unirradiated Charpy Impact Data for Reactor Vessel Surveillance Program I
Weld Metal WF-209-1 (See Reference A-401 i
i
- ~
Test Absorbed lateral Shear Specimen Temp.,
- Energy, Expagston,
- Fracture, z
Mo.
F ft-lb 10-i ri.
(
j_-
-100 6.0 7.0 5
},
-100 6.0 7.0 9
i
-100 5.0 7.0 5
- 50 16.0 17.0 17
- 50 17.0 17.0 21 s
- 50 9.0 11.0-12 i**
e
- 20 25.0 26.0 37-g**
~
- 20 18.0 20.0 32
- 20 23.0 24.0 37 i.
10 3!.0 34.0' 43 5
10 37.0 37.0 47 E'*
10 27.0 31.0 43 80 52.0 56.0 91 80 51.0 54.0 90
_..,0, w _
80 46.0 52.0 90 110 59.0 64.0 95
.,* k.
-m 110 65.0 70.0 100
....+
210 65.0 67.0 100
_ " ' 'b ' m a *=' "
210
- 63.0 66.0 100 210 62.0
~66.0 100 i,, "-
~
5 300 61.0 67.0 100 300 61.0 72.0 100
}
300 68.0 73.0 100 l,
- i O
e.
3=
sete eartet.eI.ae to
= -.
l area ep. W J99.9 est ame se, W
e r
--w
,e2 m-
=
l l
i Data Set A-58.
ZionUnitQ. Cap'suleT: Charpy Impact Data for Weld Metal WF-209-1, Irradiated to 2.53 x 10 n/ca (E > 1 MeV) (See Reference A-41) i' Test Absorbed
. Lateral Shear
'~ i
, ~
Specimen Temp.,
- Ener9y, Expagston, Fracture.
1 No.
F.
ft-lb 10 in.
i y.
W35 0
6.0 5.5 5
In W39
' 72 17.0 20.5 25 W34 100 29.0 28.0 45
==
1 W38 131 35.0 35.5 35 g==
4 W36 165 41.0 42.0 75
{..
W37 200 51.0 52.5 95 f.
4 1
l W33 350 55.0 55.0 100 l,,
I W40 350 57.0 54.5 100 i
ese t, _
f,sfletet
- EM t, ese nw.amm e,, em nw.Jer_
.h* 't,-M W.1!!1.%
g, er.,,
.I 1.
) w i
Je
% sete asetet t 4*ee ##
j.
a e,a. tm. n*-1.s anee se 88 M-8
~
b-
.b
.E E.
S
~
,,________________._._.___..m
i.
Capsule U: Charpy Impact Data for Weld F ".a1 WF-209-1. Irradiated to Zion Unitg,n/ca" (E > 1 MeV) (See Reference A-42)
Data Set A-59.
8.49 x 10
~
Test Absorbed Lateral Shear Specimen Temp.,
Energy.
Expagston, Fracture.
No.
F ft-lb 10 in.
1 i
E,.
W47 75 10.0 8.0 14
)e W45 150 22.0 18.0 41 W41 200 23.0 18.0 48 W48 225 40.0 38.0 95 j..
W44 250 42.0 42.0 98
{..
W42 300 47.0 43.0 100 j,,,
W43 350 52.0 47.0 100 l,,,.
W46 400 48.0 47.0 100 3
in SH. WWWW - -
W 't o,.
t, tM anet i f,, tW eeet ;}N_ _
9,, t 2 es oo.*ME.
[
't,W taans 3_[b)h5 M.,
~
f s=
l i
g p
o eege moet.s ema se 2 M e I8'
. mae ar gg g,,g het %>.
W gM I e
l e
n 9e e
SW m-W eWD
- g@ tamuuegeget, 9 a
.4
'(
a Data set A-60.
ZionUnitg.CapsuleX: Charpy Impact Data for Weld Metal WF-209-1, Irradiated to 2
j 1.26 x 10 n/cm (E > 1 MeV) (See Reference A-43)
^
^
Test' Absorbed lateral Shear Specimen-Temp.,
- Energy, Expagston.
- Fracture, m
j No.
F ft-lb 10 in.
no 4
l W25 75 11.0 10.0 5
}r i-W28 120 16.0 16.0 10 i
W26 160 22.0 22.0 20
.a
~
W32 180 23.0 23.0 25
~
3..
W27 210 35.0 35.0 50 j..
j,,,, _
W31 250 38.0 40.0 50 j.
W29 300 43.0 35.0 100 l,,
W30 350 45.0 47.0 100
'3 sw
~ 948 W ~
te t,,
1,, EM antB
- E j.
,* 'I,49 99 W L t, t E 0948. ?
e e
r,
=
- * -i f
,"ame i -
3 e*
r g
w j
ewee hee 4-4 lose go
- egraaf
' M * *l' etee =>. W M-t b
E
$s
.an
=
t
~
Data Set A-61.
Zion Ur.it-2: Unirradiated Charpy Impact Data for Reactor Vessel Survelliance Program Weld Metal WF-209-1 (See Reference A-441
~
o -
i Test-Absorbed Lateral Shear Specimen Temp.,
Energy.
Expagston, Fracture.
No.
F ft-lb 10 in.
i i,,,
-100 10.0 7.0 17 jr g*z 11.0 8.0 13
-100
~17.0 14.0 23
- 60
- 60 20.0 16.0 33
- 60 14.0 12.0 23 s
- 25 35.0 32.0 58 i'"
- 25 25.0 23.0 47 j.,
- 25 32.0 27.0 42 10 37.0 33.0 55 i * *-
8 10 40.0 38.0 61 5
10 41.0 40.0 71 E * *'
40 47.0 43.0 65 40 51.0 48.0 71 40 43.0 40.0 57 w.=
75 60.0 58.0 95
.w 75 62.0 57.0 97-e,..n...
75 56.0 53.0 97
%.= a**.m -
'a ' * " *' A" -
210 69.0 69.0 100 e
1 e" '"'"""3"
210 69.0 70.0 190
" " ' ~ ~ ~ ' ~ ~
___ {
210:
62.0 61.0 100 i=
300 68.0 65.0 100 j
300 72.0 68.0 100 j,,
300 70.0 71.0' 100
.I -
_.. ~.
- ~
a w
a 4
4 u
i.i i
.,.i......
._.........s
Capsule U: Charpy lupact Data for Weld Metal WF-209-1, Irradiated to Zion Unitgg,n/ca' (E > 1 MeV) (See Reference A-41)
Data Set A-62.
2.57 x 10
~
/'
Test Absorbed Lateral Shear Specimen Temp.,
Ener9y.
Expagston,
- Fracture, n
i.
No.
F ft-lb 10 in.
y M5 50' 16.0 12.5 10 l^
7 W47 73 -
20.0 20.5 20 W42 125 31.0 29.5 65'
'a-s W41 200 45.0 41.5 99 p=
W44 285 48.0 51.0 100 je=
W43 345 50.0 49.0 100 g.
W46 350 46.0 71.0 100 l,,
W48 400 50.0 52.5 100 ew
- i. - e., _ _
,,,,..,E_
i,, is.. :5 e,...
-st'_
~
.j*
t, est s.=s te st-ana g
_ =.
~
8' 5,
hw S1*
- a
=
=
19 e=H mesog g 3.e, se t.u.
=.i..r w Sp. 88 % ' 9
' ".=
E E
E a.
O 9*** Baump. gen,e. s
Data Set A-63.
ZionUnitg. Cap'suleT: Charpy Impact Data for Weld Metal WF-209-1 Irradiated to 8.04 x 10 n/ca (E > 1 MeV) (See Reference A-45) i
~ '
Test Absorbed Lateral Shear Specimen Temp.,
- Energy, Expagston,
- Fracture, a
No.
F ft-lb 10 in 1,, _.
~
W34 71 12.0 13.0 15 W33 135 22.0 18.0 35 W35 185 40.0 38.0 80 2
W39 235 36.0 34.0 80 i'"
}**
W38 300 43.0 43.0 100 1
W37 350' 44.0 47.0 100 t-W36 400 42.0 45.0 100
-1.=
s i
a e
a e
e a
II8
=
e 5
5
- 'f.,,.-
I,E15amet _.*. M.
t,eis eesee C f, 9 9 ** ##. *.*_
.g* ' t, WE teuse M1-108
~
f
_EE e _ _. -
I-e j.
9 go a
h MO %9d-$ % OO se s
e.e,s* y w.f.J e
a an.
- I'e '
E.
E.
e a
-a I
y r
+'o t--:-
--- -=
Irradiated to 1.45 x 10 n/cm* py Impact Data for Weld Metal WF-209-1, Oconee Unit-3, Capsule D: Char
' BAW-2128/1, P 5-5.
(E > 1 MeV) (See Reference 31) s
~
~
~
Test
- Absorbed lateral shear Specimen Temp.,
Energy.
Expagston,
- Fracture, n
r No.
-F ft-lb 10 in i
. j
.w 18.
20
}"
JJ 037 70 18.0 JJ 011 125 16.5 16 10 a
i I
JJ 045 160 -
30.0 27 40 ea 1'
JJ 058 175 27.0 31 70 gem f
/
j JJ 051' 185 22.0 19 30 le.
[
- i. e e,_
i
=
JJ 052 200 41.0 38 90 sem i
JJ 044 250 49.5 47 100 JJ 025 300 39.0 36 100 3
T JJ 021 350 39.5 41 100
-- t in I
JJ 035 400 40.0 40 100
- m. r, JJ 012 450 47.0
~ 47 100
% < m==* 1 = -
l g tis n san.an,.
JJ 005 550.
39.0 45 100.
gin -.w
,3= c.- a w n-ia' 4
i e.e.
I jg.y#
l s.'
l-l a
1 3
t e
n,
. am Am e A s se gg gm 9.M.19#./ eel I'
- g. A W -2= -1 I
b b
, ens
.se e
33 i
..w.
,e y-
.c.,
~
,y4.,
m
,,,,.,,. ~,, _,,.., _.
,s--
~.
Charpy(impact Octa for Weld Metal WF-209-l.
BAW-2082, P 5-6.
Zion Unit-1, Capsule Y:
Irradiated to 1.56 x 10 n/cm E > I MeV) (See Reference 32)
Test Absorbed lateral Shear i
Specimen Temp.,
- Energy, Expa3sion,
- Fracture, n
No.
F ft-lb 10-in.
i a.
W53 70 7.5 0.005 5
}
2-W55 140 20.0 0.018 40 e
W52 175 25.5 0.022 40
. =a
[-..
W54 225 30.0 0.028 70 W49 290 43.0 0.041 100
[..
W5J 350 43.5 0.046 100 j,,,
W56 400 44.5 0.042 100 5
gem W51 550 46.5 0.046 100 3
na O
f, (fl en.si W f,s3 % 3 8.9.
- W I, f M fN9
- 88 *(qgg 44 ft-ite i
?*
3 a
Iw en
=
)$
sole **tet g,,,
T.W10"*1M. -
33 gm
- #*I east En.
-m e
m m
m m
)o APPENDIX D Refererttes Cited in Replies #4 anM A-4.
/. L. Lowe, Jr., el_11., Analysis of Capsule 0;II-C from Duke Power Company Oconee Nuclear Station, Unit 2, SAH-lail, Babcock & Hilcox, Lynchburg, Virginia, May 1977.
A-5.
A. L. Lowe, Jr., el_A1., Analysis of Capsule OC11-A from Duke Power Company Oconee Nuclear Station, Unit 2. DAH-1599, Babcock & Hilcox, Lynchburg, Virginia December 1981.
A-6.
A. L. Lowe, Jr., Et_al., Analysis of Capsule OCII-E from Duke Powe-Company Oconee Nuclear Station, Unit 2, BAR:2051, Babcock & Hilcox.
Lynchburg, Virginia, May 1977.
A-7.
A. L. Lowe, Jr., at_al., Analysis of Capsule OCIII-A from Duke Power Comp 6ny Oconee Nuclear Station, Unit 3. BAH:1138, Babecck & Hilcox, Lynchburg, Virginia, July 1977.
A-8.
A. L. Lowe, Jr., et_al., Analysis of Capsule OCIII-B f ron' Duke Power Company Oconee Nuclear Station, Unit 3, BAH-1697, Babc>ch & Hilcox, Lynchburg, Virginia October 1981.
A-21.
A. L. Lowe, Jr., g1_al., Analysis of Capsule DB1-LG1, labcock &
Wilcox Owners Group Integrated Reactor Vessel Materials Surv0111ance Program.
bah-1920E, Babcock & Hilcox, Lynchburg, Virginia, October
- 1986, a-41.
J. S. Perrin, gi_al., Zion Nuclear Plant Reactor Pressure Vesse?
Surveillance Program: Unit No. 1 Capsule T, and Un1+ No. 2 Capsulo V, BCL:585:4, Battelle Columbus Laboratories, Columbus, Ohio, March 25, 1978.
A-42.
S. E. Yanichko, e.t_al., Analysis of Capsule U f rem the Conmonwealth Edison Company Zion Nuclear Plant Unit 1 Reactor iessel Radiation Surveillance Program, MCAE-9&SO, Hestinghouse Ele:tric Corporation, Pittsburgh, Pennsylvania, March 1981.
ZNLD/2479/14
)so ',
A-43.
E. B. Norris, Reactor Vessel Material. Surveillance Program for Zion Unit No. 1 Analysis of Capsule X, Southwest Research Institute, San Antonio, Texas, March 1984.
A-45.
E. B. Norris, Reactor Vessel material Surveillance Program for Zion Unit No. 2, Analysis of Capsule T Final Report, Southwest Research Institute, San Antonio, Texas, July 6, 1983.
18.
F. H. Stallmann, The Pressure Vessel Steel fmbrittlement Data Base (EDB), NUREGIR-4316 mVolumes 1 and 2 Prepared for U.S. Nuclear-Regulatory Commission by Oak _ Ridge National Laboratory, Oak Ridge, Tennessee, June 1988.
30.
S. L. Anderson, Plant Specific Neutron Fluence Evaluation for Zion Units 1 and 2, NCAP-10902, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, August 1985.
31.
A. L. Lowe Jr., fLt_Al., Analysis of Capsule OCIII-D, Duke Power Company, Ocorce Nuclear Station Unit-3,-DAH-2128. Rev.1, B&W Nuclear Service Company, Lynchburgh, Virginia, May 1992.
32.
A. L. Lowe, Jr., gL Al., Analysis of Capsule Y from the Commonwealth Edison Company Zion Unit 1 Reactor Vessel Radiation Surveillance Program, BAH-2082, B&W Nuclear Service Company, Lynchburg, Virginia,-
March 1990.
33.
E. Terek, et al., Analysis of Capsule Y from the Commonwealth Edison Company Zion Unit 2 Reactor Vessel Radiation Surveillance Program, HCAP-12396, Westinghouse Electric Corporation, Pittsburgh, Pennsylvania, September 1989.
2NLD/2479/15