ML20137A022
| ML20137A022 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 01/10/1986 |
| From: | Musolf D NORTHERN STATES POWER CO. |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| REF-GTECI-A-49, REF-GTECI-RV, TASK-A-49, TASK-OR TAC-59974, TAC-59975, NUDOCS 8601140113 | |
| Download: ML20137A022 (25) | |
Text
_ _ _ _
,--.e Northem States power Company 414 Nicollet Mall Minneapolis, Mnnesota 55401 Telephone (612) 330-5500 January 10, 1986 Director Office of Nuclear Reactor Regulation U S Nuclear Regulatory Commission Washington, DC 20555 PRAIRIE ISLAND NUCLEAR GENERATING PLANT DOCKET NOS. 50-282 LICENSE NOS. DPR-42 50-306 DPR-60 Assessment of Pressurized Thermal Shock Reference Temperature In Accordance with 10 CFR Part 50. Section 50.61 The purpose of this letter is to provide projectea values of pressurized thermal shock reference tem-perature at the inner vessel surface of reactor vessel beltline materials at the Prairie Island Nuclear Generating Plant.
This report satisfies the requirements of 10 CFR Part 50, Section 50.61(b)(1).
The following information is provided:
Figure 1 - Plot of Unit I reactor beltline material pressurized thermal shock reference temperature vs. effective full power years (EFPY).
Figure 2 - Plot of Unit 2 reactor beltline material pressurized thermal shock reference temperature vs. EFPY.
Table 1 - Unit 1 reactor beltline materials data and computation of pressurized thermal shock reference temperature at 60 EFPY.
Table 2 - Unit 2 reactor beltline materials data and computation of pressurized thermal shock reference temperature at 60 EFPY.
The screening criteria of 10 CFR Part 50, Section g
50.61 of 270 degrees F for forgings and 300 degrees 0
h0 F for circumferential weld materials is not ex-ceeded at 60 EFPY in either unit.
9) 8601140113 860110
^
PDR ADOCK 05000282 P
Northem States Fower Company Director of NRR January 10, 1986 Page 2 The operating licenses for Prairie Island Unit 1 and 2 currently expire on June 25, 2008.
This expiration date is based on a 40-year period following issuance of the construction permits for Prairie Island Units 1 and 2.
We will soon submit a license amendment request to change the expira-tion date of the operating license for Unit 1 to April 5, 2014 and the operating license for Unit 2 to October 29, 2014.
This will extend the license expiration dates to 40 years following issuance of the full power operating license for each unit.
We are also actively involved in life extension stud-ies for our' Prairie Island and Monticello Nuclear Generating Plants.
These studies may be used to support operating license extensions to 60 EFPY (approximatey 80 years of operation based on the current average design capacity factor of about 75%).
Using Figures 1 and 2, the margins to the pressurized thermal shock screening values can be read for all points in. plant life.
Specifically:
Date EFPY Unit 1 EFPY Unit 2 1-10-86 9.4 9.0 (current date) 6-25-08 25.9 (est) 26.4 (est)
(current operating license expiration)
The following information was used in computing the values presented in Figures 1 and 2 and Tables 1 and 2:
Vessel Beltline Materials Beltline materials consist of the inter shell forging, the lower shell forging, the nozzle shell to inter shell weld, and the inter shell to lower shell weld.
All welds are circumfer-ential welds.
Forging and weld metal c' emical a
composition and initial reference temperature are as reported to the NRC in Reference (1),
except that for conservatism, NRC worst case generic values were used for the weld metal.
Portions of Reference (1) are included in Attachment (1).
T Northem States Power Company Director of NRR January 10, 1986 Page 3 Vessel Fluence Vessel fast neutron flux used in calculating the pressurized thermal shock reference tempe-rature is as reported to the NRC in the dosi-metery section of summary technical reports of analysis of the second capsule from each unit removed as part of the reactor vessel radia-tion surveillance program (References 2 and 3).
Portions of these reports used in the calculations are provided in Attachment (2) and Attachment (3).
Core power distribution used by Westinghouse in these analyses are representative of time-averaged conditions derived from statistical studies of long term operation of Westinghouse two-loop plants.
These input distributions include rod-by-rod spatial variations for all peripheral fuel assemblies.
Resulting fluence values produce accurate long term neutron exposure levels for the pressure vessel.
Please contact us if you have any question related to the information we have provided.
l-J' L-,
David Musolf Manager - Nuclear Sup ort SerJr.as c:
Regional Administrator-III, NRc Resident Inspector, NRC NRR Project Manager, NRC G Charnoff Attachments
NorL5em States Power Company Director of NRR January 10, 1986 Page 4
References:
1.
Letter dated October 31, 1977 from L 0 Mayer, NSP, to D K Davis, Division of Operating Reactors, JSNRC, " Reactor Vessel Material Surveillance Program."
2.
Letter dated May 13, 1981 from L 0 Mayer, NSP, to Director of NRR, USNRC, " Summary Technical Report of Analysis of Capsule from Unit No. 2 Reactor Vessel Radiation Surveillance Program -
3.
Letter dated October 12, 1982 from D Musolf, NSP, to Director of NRR, USNRC, " Summary Technical Report of Analysis of Capsule from Unit No. 1 Reactor Vessel Radiation Surveillance Program -
TABLE 1
- 1 <-. ts.
N =r Neh?HkR4'E7A?Is"JE4fn chWlINv"_ FERTRtV"f.flNU AucEEin'otNElit:25*JLRur'uAtl i iEj V
^
,4n CALCULATED rOR 40 YEARE OF OPERATION r.
_____________SE"t2NgN{,,,,,,,,,, {N((,R{Ng{ g{Ng{,5Rgg,,Hfg,gU,,,,Wfg,N{,,,,{gUgNgg,,,g((((g{_
$j3k o.
NrEn suRLL rena:No i.
.E..
nEasunEn 4...E-.
7.s.E-.
..s=E.
i.ait..e m
LouEs enELL reno:No
_+...E...
nEasuaEn 7...E-.
....E-.
..s E.
i.e.E..a g
Noz-:N1En enttL uRLo ens,/sie.
cENEnso 3.1.E-.i i.s.E-ei m.ast.i, i..
E...
Noz-:NvEn sutto utLe 3..,/sie.
cENEnzo a.i.E-ei
....E-..
....E.3,
...it...
INTER-LoH WHELL NELO 1752/823.
CENERIo 1
4.E-.8 1.7.E-et
. 3 E.3, 3.73E..g iNrEn-Lou.nELL wEL.
3.../
3.
.ENE..
.E-.i
....E_..
E.
i.435...
TABLE 2 08k?Nk$4'EYAYEs"JEEIn#8bNlIN? _ FEXfRtV'f.flNU NucEEta'atNEX11 bl*1LRur'unti CALCULATED FOR 40 YEARS OF OPERATION
_____________E9"t2NEN{,,,,,,,,,, {Nj{,g{Ng{ g{Ng{,gggg,,Wfg,go,,,,Wf},N{,,,,{(UgNgg,,,g((((g{,
l INTER BHELL roRGING 4...E...
nEASuREo 7 5.E-.2 7 53E-ei
. 52E.3, 1 24E..t Loutn swEtt reno:No
_4...E...
nEasunto
....E-.
7...e-ci E.
i.33E...
NoZ-INTER SHELL HELD 375E/8243 GENERIC 3.4.E-.8 3 4.E-wa s.45E.1, 3.43E..E NOZ-INTER SHELL HELo 3 4,/1243 GENERIO 3.,.E-.1 1
3.E-.1 2.4SE.1, 1.73E..t l
INTER-Lou SHELL HELD 2728/3243 GENER80
....E-.2 1
3.E-.1
. 5 E.3, 1 24ES.E l
INTER-LoM BHELL HELO 3 4,/3343 GENERIO i.,.E-.1 1
3.E-.8
. 5:E.8, 3:34E..E G
l A -e
Dir:cter cf NRR J nuary 10, 1986 Figure 1 10 CFR PART 50, SECTION 50.61, SCREENING PRRIRIE ISLRND UNIT = 1 9
o.
R-LEGEN0 o - INTER SHELL FORGING 9
a - LOWER SHELL FORGING
+ - N0Z-INTER SHELL WELO 2269/1180 x - N0Z-INTER SHELL WELD 3049/1180 o
o - INTER-LOW SHELL WELD 1752/1230 d-v - INTER-LOW SHELL WELD 3049/1230 C.
g-9
- G-O9 tJ 3 -
=
g-u a9 G b~
q j
S E
~
~
e 8-C.
s~
o.
S~
i l a
$~
o T,
9
'0. 0 5'. 0 lb.0 lb.0 2b.0 2b.0 3b.0 3b.0 4b.0 4b.0 Sb.0 5b.0 60.0 EFFECTIVE FULL POWER YERRS
Dir:ctor of NRR January 10, 1986 10 CFR PART 50, SECTION 50.61, SCREENING PRRIRIE ISLRNO UNIT = 2 o.
=
K-LEGEND o
-INTER SHELL FORGING
'o a - LOWER SHELL FORGING
+ - N0Z-INTER SHELL WELD 1752/1263 x - N0Z-INTER SHELL WELD 3049/1263 9
o - INTER-LOW SHELL WELD 2721/1263 H-v - INTER-LOW SHELL WELD 3049/1263 o
k~
-oti
[ k~
~
09 e B-g S9 6
G N~
0 g
gq e g_
~
o.
K~
9 8~
1 !
c g_
o.
N!
N 0.0 5.0 lb.0 1U.0 2b.0 2b.0 3b.0 35.0 4b.0 4$.0 Sb.0 55.0 60.0 EFFECTIVE FULL POWER YERRS w e
Director of NRR January 10, E PliAIRIEISLANDUNITNO.1 1986 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM 1.) The estimated maximum fluence (E > 1 Mev) at the inner surface of the i
reactor vessel as of March 31,1977 is 2.85 x 1018 n/cd.
2.) The effective full power years (EFPY) of operation accumulated as of March 31,1977 is 2.118 EFPY.
3.) Fabrication of the' reactor vessel was performed by Societe des Forges et Ateliers du Creusot (SFAC).
4.) a.) Sketch of the reactor vessel shewing materials 1,n the beltline region
~
is shown in Figure 1.
b.) Information on each of the welds in the beltline region is shown in Tables 1 through 4.
c.) Information.on each of the forgings in the beltline region is. shown in Tables 4 through 7.
- 5. ).
Information relative to the weld and forging material in the material surveillance program is shcwn in Tables 1 through 3 and 5 through 7.
e I
l i
i 1-1 l
f
\\
FIGURE 1 IDENTIFICATION AND LOCATION OF PRAIRIE ISLAND UNIT NO.1 REACTOR VESSEL BELTLINE REGION idELD AND FORGING MATERIAL
~
. M 1
s n
C W2 E
en
.y Forging C ue y
\\
Core.
- s W3
{
Forging 0
-2 e
g g
4 e
e op e
6 1-2
-. _ ~ _ _..
f
. TABLE]
IDENTIFICATION OF PRAIRIE ISLAND UNIT No. 1 REACTOR VESSEL BELTLINE REGION WELD METAL Weld Wire Flux Wald Welding Wald Post Weld Location Process Control No.
Type Heat No.
Type _
Lot No.-
Heat Treatment Submerged Arc PS-011 UM40 2269 UM89 1180 1022*F + 50*F-25 HR$ + -
i Nozzle Shell to i
Inter Shell
- HH40 3049 UM89 1180 1148*F _+ 45*F-20 HRS-FC i
circle Sean W2 I
Inter. Shell to Satunerged Arc PS-011 UM40 1752 UM89 1230 1022 + 50*F-25 HRS +
l
,__ Lower Shell
- UH40 3049 UM89 1230 1148'T + 45'F-20 HRS-FC
~
Circle Sean W3 i
Surveillance Weld - Same as the Inter. Shell to Lower Shell Circle Seam 1622*F-5 HRS + 1112*F
- 7 HRS-FC I
l
- Used only In Root Area of Wald t
I T
l TABLE 2 l,
CHEMICAL C0WOSITION OF VESSEL BELTLINE REGION WELD METAL l
Wald Wire Flux Weight Percent 1
- C_r, 31, Me, Cu, C,,o, '
r 1
P, S,
Sl_
l Type Heat No. fT,ype.
Lot No.
C_
Mn_ -
r..
UM40 2269 UH89 1180
.055 1.24..016
.012
. 43
.025
.15
.,44
.17
.023 l
i UM40 3049 1968 9 1180.
.055 1.32
.020
.011
.54
.035
.08
.44
.115
.020 i
UH40 1752 L3089 1230
.099 1.36
.016
.013
.36
.060
.17
.45
. 14 018 I
UM40 3049 UM89
'1230
.054 1.44 019 010
.42
.050
.08
.48,_
.115
,.019
.51
.13
<.001 l
Surveillance Wald
.052 1.30 017
.014-
.36
.015 e
1 i
1 1
l l
TABLE 3 ECHANICAL PROPERTIES OF VESSEL BELTLIE REGION WELD METAL Shelf Wald Wire Flux Energy RT TNOT at 10*F NOT Energy VS UTS Elong.
)
Type _, Neat No. Type Lot Nim.
'F ft-Ibs
- F ft-Ibs kst kst 5
5 66.1 80.6 29.8 i
UM40 2269 UM89 1180 0*
95,73.5,81.5.93.91 0* -
69.'4 84.0 32.0 UM40 3049 UM89 1180 0*
77.5,87.5,61.93.68 0*
UM40 1752 L9489 1230 0*
83.5,81.5,83.5,81.5,87.5' 0*
65.5 80.6 29.8 74.2 88.4 29.0 UM40 3049 1948 9 1230 0*
57,77.5,68,93,81.5 0*
2 Survet11ance Wald *
-13 75,50,52
-13 78.5 70.9, 86.3 26.6 69.8 l
-
- Estimated per NRC Standard Review Plan Section 5.3.2.
L TABLE 4 i
MAXIN#1 END-0F-LIFE FLUENCE AT VESSEL IlftER WALL LOCATIONS Fluence i
,je,2 4
I8 Nozzle She11 to Inter. Shell Seam W2 1.3 x.10 i
i II 4.3 x 10 Inter. Shell Forging C I8 Inter. Shell to Lower Shell Sean W3 4.3 x 10 I
4.3 x 10 '
Lower Shell For9tng D l
l e
eo.
l
~
TAKE 5 IDENIFICATION 0F PRAIRIE ISLAND UNIT No I REACTOR VESSEL BELTLINE REGION FORGIMS Heat Treatment Forging Material camponent No, Heat No.
Spec. No, Suppiter Austenttire Temper Stress Relief I:It:r. Shell C
21g18/38566 A508 CL3 SFAC 1652-1715*F-5 HR-WQ 1175-1238*F-5 HR-FC 1022*F-8 HR +
l 1652-1724*F-5 1/2 HR-WQ 1202-1238'F-5 HR-FC 1112*F-14 HR-FC i
i Lower Shell 0
21887/38530 A508 CL3 SFAC 1652-1715'F-5 HR-WQ 1175-1238'F-5 HR-FC 1022*F-13 HR +
l 1652-1715-5 HR-WQ 1202-1238*F-5 HR-FC 1112*F-7 HR-FC l
i l
Survaillance Material - Same as Inter. Shell Forging C l
I i
l TARE 6 l
CHEMICAL CONGSITION 0F VESSEL BELTLINE REGION FORGINGS Weight Percent Forging C !,
g!,
y, j
!!b.
C.,
H!!,
f, -
S.,
Sj, H.j, Mg, j
C
.17 1.41
.013
.005
.28
.72
.48
.06
.010 4.002
.28
.66
.53 07
.015
<.005 l
]
D
.17 1.43 -
.014
.007 l
Surveillance Material Forging C - Analysis as Reported Above by Fabricator l
I i
l-i
TABLE 7 ECHANICAL PROPERTIES OF VESSEL BELTLINE REGION FORGINGS 4
5pr Shelf ft-Ib T
- UTS Elong.
RA No.
- F
- F MD IMfD ksi ksi 1
1 C
14 14
.127 82.5*
67.5 86.5 28.5
' 72.6 '
D'
-4
-4 135 88*
68.8 87.'3 27.6 73.0 1
14 163 146 67.0 86.2
. 27.6 73.I'l C
h (SurveillanceTestData)
-4 134 D
- Estimated From Data in the Major Working Direction (MWD) Pet NRC Standard Review Plan Section 5.3.2 i
- i' l
'i j
j l
t l
i 1
\\
j t
0 l
T PRAIRIE ISLAND UNIT NO. 2 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM
' ~...___
1.) The estimated maximum fluence (E > 1 Mey? at-the inner surface of the reactor vessel as of March 31.1977 is 2.23 x loi8 n/cm.
4 2.). The effective full power years (EFPY) of operation ac, cumulated as of March 31. 1977 is 1.661 EFPY.
3.) Fabrication of the reactor vessel was perfomed by Societe des Forges et Ateliers du Creusot (SFAC).
4.) a.) Sketch of the reactor vessel showing materials in the beltline region is shown in Figure 1.
b.) Infomation on each of the welds in the beltline region is. shown in Tables 1 through 4.
c.) Information on each of the forgings in the beltline region is shown in Tables 4 through 7.
5.) Infomation relative to the weld and forging material in the material surveillance program is shown in Tables 1 through 3 and 5 through 7.
l 9
,e l
e e
0 6
l-7
FIGURE 1
~
IDENTIFICATION AND LOCATION OF PRAIRIE ISLAND UNIT NO. 2 RCACTOR YESSEL BELTLINE REGION idELD AND FORGING MATERIAL w
2
=
3 i
ggt 1
- b2 2 i Fcrging C Core g
Forging 0
.2 4
k l
j I
i e
l 1-8
I
..id e
i
.. T TABLE 1 j
j IDENTIFICATION OF PRAIRIE ISLAND UNIT NO. 2 REACTOR VESSEL BELTLINE REGION WELD METAL l
4 Wald Wire Flux Wald WaldIng Wald Post Wald i
Location Process Control No Type.
Heat No.
- Type, Lot No.
Heat Treatment a
Nozzle Shell to Submerged Arc PS-011 UM40 1752 UM89 1263 1022*F + 50*F-25 luts +
Inter. Shell
- UM40 3049 UM89 1263 1148*F'T 45'F-20 HRS-FC I
~
Circle Sean W2 PS-011 UM40 2721 UM89 1263 1022*F + 50*F-25 HRS +
I Inter'. Shell to Submerged Arc Lower Shell
- UM40 3049 UM89 12,63 1148*F T5*F-20 HRS-FC '
^
Circle Seam W3 j
L Survalliance Weld - Same as the Inter. Shell to Lower Shell Circle Seam 1022*F-10 HRS + 1112*F
- 7 HRS-FC
- Used Only in Root Area of the Weld
)
1
'i' l
e TABLE 2 i
i VESSEL BELTLINE REGION CHEMICAL COMPOSITION l
Wald Wire
' Flux Weteht Percent
~~ Type Heat No. Type Lot No.-
C M
P
' S_,
Si,
. Cr, N.ji,
- jto_,
C[
Co, UM40 1752 UM89 1263
.060 1.39
.018
.014
.41
.050
.14
_.48
.14 020 l
j UM40 3049 L2489 1263
.062 1.33
.021
.010
.51
. 035
.13
.48
.19
.022 IM40 2721 L2089 1263 050 1.36
.016
.013
.42 030
.13
.44
.09
.033 Surveillance Wald 045 1.37
.019
.014
.47
.020
.072
.51
.082
.513 i
i l
s' I
i l
4 l
f
~
I
=_ _.
t TABLE 3 i
l MECHANICAL PROPERTIES OF VESSEL BELTLINE REGION WELD METAL Energy Shelf RT 7NOT at 10*F NDT Energy
-VS UTS Elon9
.RA
,TJE*. Heat No. Tyge, Let No.
- F ft-Ibs
- F ft-Ibs_
ksi ksi 5
5 68.3 84.0 31.0 UM40 1752 iM89 1263 0*.
87.5.81.5.79.5 0*
69.4 ~ 85.0 27.6 i
IM40 3049 IM89 1263 0*
45.43.48 0*
69.4 84.0 28.4 IM40 2721 UM89 126'3 0*
79.5.79.5.77.5.81.5.81.5 '
0*
Surveillance Weld
-31
-31 103 66.5 80.2-27.4 73.8
~
CEs'timated Per NRC Standard Review Plan Section 5.3.2 j
1 5
i h
TABLE 4 MAXIMUM END-0F-LIFE FLUENCE AT VESSEL IltlER WALL LOCATIONS i
I Fluence 1
n/cm2 II f
Nozzle Shell to Inter. Shell Sean W2 1.3 x 10 II Inter. Shell Forging C 4.3 x IQ l
Inter. Shell to Lower Shell Seam W3 4.3 x 10 '
I8 4.3 x 10 j
Lower Shell Forging D l
1 1
j A
e l
i i
~
t TAK E 5 IDENTIFICATION OF PRAIRIE ISLAND UNIT No. 2 REACTOR VESSEL BELTLINE REGION' FORGINGS Heat Treatiment l
Forging Material j
Camponent
& Heat No. Spec. No. Supplier Austenttire Temper Stress Ralfef i
Inter. Shell-C 2282g A508 CL3 SFAC 1650-1745*F-5 HR-WQ 1170-1260*F-5 int-FC 1022*F-12 HR +
j 1565-1655'F-5 HR-WQ 1200-1260*F-5 HR-FC 1130*F-14-1/4 HR-FC Lower Shell D
22642 A508 CL3 SFAC 1652-1715*F-5 HR-WQ 1175-1238*F-5 HR-FC 1022*F-11-1/2 HR +
~
1652-1724*F-5-1/2 HR-WA
' 1202-1238*F-5 HR-FC 1112*F-7 HR-FC i
Surveillance Material same as Lower She11 Forging D i
i TABLE 6 l
CHEMICAL COMPOSITION 0F VESSEL BELTLINE REGION FORGINGS I
h Wefght Percent Forging No.
C Mn P
S' St Ni Mo Cu Co V
i C
.170 1.32
.010
.013 *
.270
.73
.480.
<.005
.075 030 i
8
.175 1.22
.011
.013
.285
.70
.445'
.085
.026
<.005 l
Survei11ance Material Forging D - Analysis as Reportsd Above by Fabr1cator t
i l
i l
l s:
l
(.
l i
b
9 r
TABLE 7 ECHANICAL PROPERTIES OF VESSEL BELTLINE REGION FORGINGS U>per 3
S ulf i
ft-lbs T
U Forging NOT NDT VS UTS Elong.
RA No,
- F
- F DGID IMfD ksi ksi 1
1 i
C
-4 10*
134.0 87*
60.4 80.6 31 73.4 i
l D
-13 10*
121.5 79*
65.8 87.8 28 70.5, i
112 C
-4
-6 150.5 108 66.7 86.4 27.3 71.6 D
1
- Estimated From Data in the Major Working Direction (NWD) Per NRC Standard Review Plan Section 5'.3.2.
}
I 7
i G
4 i
l l
i
-e 3
I t
r t
4 l
l
~
Director of NRR UNIT 2 -- WCAP-9877 Jcnuary 10, 1966 3,
Attachment (2) 10 "
8
~
147.64
~
171.77 IR I
2 1/4T h
~
81 1/2T 10 180.02
{ 10 3
g 8
g S
g*
3/4T 3
184.15 3
4 OR g
NO g
PRESSURE VESSEL 10' I
I I
I i
i i
i M
170 172 174 173 178 18d 182 134 Igg RA0108 (cm)
Figure 6-4.
Calculated Radial Distnbution of Maximum Fast Neutron Flux (E > 1.0 Movl Within the Pressure Vesset 1
2-1
- - ~..
18.318 24 100 5
2 10-1 E.,.
w 5
g E
ll3w 2
ua2 2
>-4 awa 10-2 5
CORE MIDPLANE 2
TO VESSEL CLOSURE HEAD 10-3 I
-300
-200
-100 0
100 200 ~
300 DISTANCE FROM CORE MIDPt.ANE (CM)
Figure 6-5.
Relative Axial Variation of Fast Neutron Flux (E > 1.0 Mev) Within the Pressure Vessei i
2-2 I
e-<
... - ~ -.
n.-
l
~
n, i
1 TA8LE A4 4
i HEACTOR VESSEL TOUGHNESS DATA (UNIRRADIATEDI l
1 Teaneseese *I I
50 ft Ibl35 soils I
ces F
'BeDTT Leeeest Espenniese peDT Aseeses Teemoverse *I RI i
e--r----*
Moserief Type 11 0 (10
(*F)
Teser (*f?
(*F)
Upper Shelf (ft Ibl i
84*I I
ICI E
6 62 Oouse Head Dniene A533 Ge. 3,0,1 8'7 *I I
i
' lei *I
-35
-31 Head Flange A600 c.3 I
seI*I IS *I
-22 Vessel Flange A500 c. 3
-22 97 *I y.
I injectiose Noaales A600 C. 3
-22
-Il4 'I
-22 89 *I I
I 60 *I
-10
-13 Balet asial Outies Nosite A508 0. 3 SE *I I
41 *I
-13 I
Uppee Shell A600 0.3
-13 Inser. Shelllbl A608 O. 3 0.075 0.080
-4 56
-4 112 j
Lowee Shen lbl A500 0. 3 0 006 0.081
-13 64
-8 108 r
i 74 *I I
1 Trans. Ring A508 0. 3 10 50 le seI*I l
Bostoni Head, A533 Ge. B O.I
-13 54
-4 4
Weidnient Wiid 0 002 0.019
-38
-8
-31 103 I
1
-38
-:ll 117 I
HAZ HAZ
- s. W-wienerei n.smet se etw sensees weehene eines seen i
In. Senest en arseas sees==su des 4 semoush some sueueissaen.e sweepesse
- a. Estenosed essmeis "Phews-Tenenmeemse Ussess." Sessmue 5 3 2 of Seandseef Areses Man. SfuREG-7WDelF.
1975 feesus W.e.d sesse
n.
Director of NRR JCnuary 10, 1986 Attachment (3)
UNIT 1 - WCAP-10102 10"l i
s 3
h 6
I a
l 4
l s
1/4 T 2
1 i
i s
1 1/27 3
x' 2a 10*
m h8 l
e 3
6 U4 7 9
Z g
4 l
~
CR
~
2 PRESSURE VESSEL
, H:0 A
1 s'
I I
I 5-l 10' 160 162 164 166 168 170 172 174 176 178 180 182 184 186 188 RADIUS (cm)
Figure 6-4.
Calculated Radlai Distribution of Maximum Fast Neutron Flux (E > 1.0 Mov) within the Pressure Vessel e
3-1
e 0
8 8
4 2
~
10*'
3 u
3 8
2 Q 4 3
1
[2 a
,E
~
10~'
4 8
1 4
W3 5
m 8
2 0
l To VESSEL t o.,
CLOSURE HEAD l
l l
,l l
-300
-200
-100 0
100 200 300 DISTANCE FROM COAE MlOPLANE (cm) f Mgure 8-8.
Relative Aximi Variation of Fast Neutron Mux (E > 1.0 Mov) within the Pressure Vesset l
3-2 l
O e
i 1
I l
TABLE A-1 REACTOR VESSEL TOUGHNESS DATA (UNIRRADIATED) i l
NMWD 50 ft Ib/35 mits NMWD j
Lateral Espansion Upper Shelf RT l
Material Cu P
NDTT Temperature NDT Energy Component Type
(%)
(%)
(* F)
(* F)
("F)
(ft Ib) 4 *l 75 *3 Closure Head Dome A533 Gr. B, Cl.1
-4 64 *3 1
3 3
y Head Flange "
A508 Cl. 3
-4 12 *3
-4 *l 84 *3 1
1 3
1 1
1 l
Vessel Flange A508 Cl. 3
-4 41 *l
-4*l 77.5*l l
Injection Nozzles A508 Cl. 3
-22
-114*l
-22'l 97 *l 1
8 1
S *3 92 *l l
1 inlet and Outlet Nozzle A508 Cl. 3
+5 39 *3 3
Upper Shell A508 Cl. 3
-4 39*3
-4 *3 85 *l 8
1 3
Inter. Shell A508 Cl. 3 0.06 0.013
+14 14 14 143 Lower Shell A508 Cl. 3 0.07 0.014
-4 45
-4 134 Trans. Ring A508 Cl. 3
+5 63 *3
'S*l 79 *l 1
l f
1 8
3 Bottom Head A533 Gr. B Cl.1
-4 57 *l
-3 'l 68.5 *l l
Inter. to Lower Shell j
Girth Weld Sub-Arc Weld 0.13 0.017 0
10 0
78.5 O*l
<-100 0
211 l
- a. Estimated using the NRC Standard Review Plan.
i 1
1 j
i