ML20196D301
| ML20196D301 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 10/31/1988 |
| From: | Cross B, Iddings F SRI INTERNATIONAL |
| To: | |
| Shared Package | |
| ML100331022 | List: |
| References | |
| NUDOCS 8812090013 | |
| Download: ML20196D301 (79) | |
Text
.- --.
REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM FORINDIAN POINT UNIT NO. 2 ANALYSIS OF CAPSULE V FINAL REPORT SwRI Project No.17 2108 Prepared for Consolidated Edison Company of New York, Inc.
4 Ining Place NewYork,NewYork 10003 l
1 October 1933 IQh Luw: V2 %.
r- -k .M (
i = %
k- v ! V '; SOUTHWESTRESEARCHINSTITUTE i l s./ '-- ',JS AN ANTONIO HOUSTON
_ 1. <<i
[' [ R .
t a
{
l REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM FORINDIAN POINT UNIT NO. 2 ANALYSIS OF CAPSULE V FINAL REPORT SwRI Project No.17 2108 Prepared for Consol! dated Edison Company of New York, Inc.
4 Irving Place New York, New York 10003 I
October 1988 r
Wraten by Approved by be M E1 -
~
b F. A. Iddings B. T. Cross D. G. Cadena Director l l Mark Williams (Consultant) Department of NDE Science and R* search
- Southwest Research Institute 1
ABSTRACT Capsule V, the fourth vessel material surveillance capsule removed from the Indian Point Unit No. 2 nuclear power plant, has been tested, and the results have been evaluated. The analysis
, of the data (1) conarms the decrease in fluence rate from the low leakage core vs cycles prior 1
to cycle 6, and (2) indicates that the prest,ure vessel weld and plate rosterials will retain adequate shelf toughness throughout the 32 EFPY design life. time using either the new Regulatcy Guide 1.99, Revision 2 or the original Revision 1 requirements.
1 j
1 l
l 1
l i
l 1
i
{
l 1
( ii i
(
TABLE OF CONTENTS Etat LIST OF FIGURES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iv LIST OF TABLES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . v I.
SUMMARY
OF RESULTS AND CONCLUSIONS . . . . . . . . . . . . . . . . . I.1 II. BACKOROUND..,,............................ 11 1
- m. DESCRIPTION OF 2.1ATERIAL SURVErf1ANCE PROGRAM , . . . . . . . . III.1 IV. TESTING OF SPECIMENS FROM CAPSULE V. . . . . . . . . . . . . . . . . IV.1 A. Shipment, Opening, and Inspection of Capsvle . . . . . . . . . . . . . . . IV.1 B. Neutron Dosimetry . . . . . . . . . . . . . . . . . . . . . . . . . . . IV 3 C. Mechanical Property Tests . . . . . . . . . . . . . . . . . . . . . . IV 14 D. Chemical Analysis Results . . . . . . . . . . . . . . . . . . . . . . . IV.26 V. RESULTS OF ANALYSIS . . . . . . . . . . . . . . . . . . . . . . . . . V.1 VI. HEATUP AND COOLDOWN LIMIT CURVES FOR NORMAL OPERATION OF INDIAN POINT UNIT NO. 2 . . . . . . . . . . . . . . . . . . . . . . . . VI.1 l VII. REFERENCES .............................. VII.1 APPENDIX A Tensile Test Data Records i
f i
4 t
t 4
1 i
l t
1 ill
I1ST OF FIGURES Flavre f.agt m.1 Arrangement of Surveillance Capsules in the Pressure Vessel . . . . . . E2 m2 Indian Point Unit 2 Resetor Geometry . . . . . . . . . . . . . . . . . E3 m3 Vessel Material Surveillance Specimens . . . . . . . . . . . . . . . . E5 m4 Arrangement of Specimens in Capsule V. . . . . . . . . . . . . . . . E6 E5 Surveillance Capsule Geomety . . . . . . . . . . . . . . . . . . . . m.7 IV 1 Uranium and Neptunium Containers as Removed from Dosimeter Block . IV 4 IV.2 Radiation Responw ofIndian Point Unit 2 Shen Plate B2002 2. . . . . . IV.19 IV 3 Radiation Response of!ndian Point No. 2 Weld Metal . . . . . . . . . . IV 20 IV 4 Radiation Response ofIndian Point No. 2 Heat Affected Zone Mate rial . . . . . . . . . . . . . . . . . . . . . . . . . . . . . IV 21 IV 5 Radieth Response ofIndian Point No. 2 Correlation Monitor Mate rial . . . . . . . . . . . . . . . . . . . . . . . . . . , . . IV 22 V1 Effect of Neutron Fluence on RTNDT Shift, Indian Point Unit No. 2 . . . \ -4 V.2 Dependence of Cy Shelf Energy on Neutron Fluence, Indian Point Unit No. 2 . . . . . . . . . . . . . . . . . . . . . . V7 V3 Predicted Decrease in Shelf Energy u a Function of Copper Content an d Fl u t u ca . . . . . . . . . . . . . . . . . . . . . . . . . . . V9 1
VI.1 Indian Point Unit No. 2 Reactor Coolant Heat Up Umitations Applicable for Periods Up to 32 Effective Full Power Years 6Vith Criticality U mi t) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . VI 3 VI 2 indian %1nt Unit No. 2 Coolant Cooldown Umitations Applicable for Periods Up to 32 Effective Full Power Years . . . . . . . . . . . . . VI 4 ,
VI-3 Indian Point Unit No. 2 Resetor Coolant Heatup Umitations Applicable for Periods Up to 32 Effective Full Power Years GVith Leak Test U mi t) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . VI 5 I
l I
tv
LIST OF TABLES IAhlt East m1 Indian Point Unit No. 2 Reactor Vessel Surveillance Materials . . . . . . m4 m.2 Capsule V Neutron Flux Dosimeters. . . . . . . . . . . . . . . . . . m.8 IV 1 Summary of Reactor Operations . . . . . . . . . . . . . . . . . . . IV.5 IV.2 Results of Dlserete Ordinates Sn Transport Analysis . . . . . . . . . . IV.9 IV.3 Dosimeter Activities and Measured Fluence Rate in Capsule V. . . . . . IV.11 IV-4 Thermal Neutron Fluence Rate in Indian Point 2, Capsule V . . . . . . IV.13 IV.5 Charpyimpact Data With Photos of Fracture Faces. . . . . . . . . . . IV 15 IV.6 Charpy Impact Data With Photos of Frr..-ture Faces (Cont'd). . . . . . . IV.16 j IV.7 Charpylmpact Data With Photos o(Fracture Faces (Cont'd). . . . . . . IV 17
) IV 8 CharpyImpact Data With Photos of Fracture Faces (Cont'd). . . . . . . IV 18 IV.9 Summary of RTNDTShifts and Upper Shelf Energy Reduction (Cy)
- for Materials in Capsule V . . . . . . . . . . . .......... IV 23 IV.10 Tensile Test Data Reeords Results. . . . . . . . . . . . . . . . . . . IV.25 l IV.11 Sumtre ' ' mistry Values for Indian Point Unit No. 2 Materials . . .
IV.26 i
IV.12 Chemi: . < ',Irdian Point Unit 2 Materials Based on Reg C .R.v.2 . . . . . . . . . . . . . . . . . . . . . . IV.28 V.1 Adjusto l' .'gpTVaaes for Indian Point.2 . . . . . . . . . . . . . . V5 V.2 Compa.. . of haasured and Calculated RTNDT Values for !ndian PoinNr. Capsule V Materials. . . . . . . . . . . . . . . . . V.8 V.3 Reactor Vessel Surveillance Capsule Removal Schedule . . . . . . . . . V.10 l
! V.4 Comparison of End of Cycle 8 Fluence Values from Transport Calcula.
- tions and Capsule V Dosimetry Analysia . . . . . . . . . . . . . . V.11 i
y v,
~_ _ _ _ _ _ _ _ _ _ - _ _ _ - .
l J
l l
4 1 L
SUMMARY
OF RESULTS AND CONCLUSIONS j The analysis of the fourth material surveillance capeule removed from the Indian Point Unit No. 2 reactor pressure veneelled to the following , iclusions: ;
(1) Based on a calculated neutron spectral distribution, Cepeule V received a fast Suence of 5.3 x 1018 n/cm2(E > 1 MeV) atits radialcenterline. !
j, (2) The surveillance specimens of the core beltline plate materials experienced shifts in i i
t l RTNDT (50 ftlb. values) over the mage of 7FF (Plate B2002 2) to 23FF OVeld) as a f
{ result of fast neutron exposure up to the 1987 refueling outage.
a
\
the heatup and cooldown limitations throughout the design lifetime of the pressure vessel. [
- l I i i !
(4) From the previous capeule, Z, the estiinated maximum neutron cuence of 3.33 x 10 18* ;
-i j neutrons /cm2 (E > 1 MeV) was received by the veneel wall in 5.17 effective full power l years (EFPY) through Cycle 5, which is equal to a Suence rate of 6.44 x 1017* per EFPY. i At the end of Cycle 8 (8.6 EFPY) the neutron Auence was 4.45 x 1018 n/cm2giving 3.26
} r 1017 n/cm2per EFPY for Cycles 6 through 8. This calculated value for the decrease
.i i in Suence per EFPY agrees well with the experimental value for the decrease in Guence t l
rete; 1.e., 50.6% vs. 48.9%. The use of a low leakage core loading pattern beginning uith l i i Cycle 6 did signiacantly reduce the ouence rate on the pressure vesselwall. i t !
I :
1 i l f r
- Revised from Capsule Z report using the latest plant specific lead factors.
l i !
l I1 i
o
t (5) The Indian Point Unit No. 2 vessel weld metal located in the core beltline twion are prefected to retain suf5cient toughness to mest the curnat 50 ftlb Charpy upper snelf l
4 requirements of 10CFR60 Appendia G throughout the design life of the pr6sture vessel !
4 using either Radston 2 or Revision 1 requirements of Regulatory Guide 1.99. [
v i
(6) Based on Regulatory Guide 1,99, Rev. 2, tand curves, the prqlected maximum RTNDTf or
]
l the Indian Point Unit No. 2 veneel core bottline materials at the 1/4T and 3/4T positions !
i afte* 32 EFPY of operation are 266*F and 207'F, r+;h4. These values were used j i
I as the bases for cosaputing bestup and cooldown limit curves to be used for up ta 32 i i [
FJPY of operation.
]
I I i t f
i :
t
! l l
t f
f
$ e+ (
I -
i 2
- i 1
I i
I l
i 12 L t
1 1
1
II. BACKGROUND The allowable loadings on nuclear pressure vessels are determined by applybg the rules in Appendit G, "Fracture Tougbness Requirements,' of 10CFR50 (J) la the case of pressure.rueinmg compar.cnis made of ferritic materia's, the allowable loadings depend on the reference stress intensity fe_ett.,r (K I g) curve indexed to the refersace nil ductility temperature (RTNtyr) presented in Appendix G, 'Protecdon Against Non.Dactile Failure,' of Section III of the ASME Code @. Further, the matedals in the eeltline region of the reactor vessel must be monitored for radiation indueed changes in RTNDT per the requirements of Appendix H,
- Reactor Vessel Material Surveillance Program Requirementa,' of 10CFR50.
The RTNDT must be established for all matedals, including weld metal and heat a#ected zone (HAZ) material as well as base plates and forgings, which comprise the reactor coolant pressure boundary, it is well established that ferritic matedals undergo an increase in strength and hardness and a decrease in ductility and toughness when exposed to neutron Quences in excess of 10 17 neutrons per em2(E > 1 MeV) W. Also, it has been established that tramp elemente., particular!y copper i and nickel, affect the radiation embrittlement rer.ponse of ferritie := : *als (H). The relat;on-ship between increase in RTNDT and copper content is defined in Rigulatory Guide 1.99, Rev.1 and Rev. 2. Estimates of shifts in RTypt n ithis report are based on the May 19&S version of Revision 2 of Regulatory Guide 1.99 @.
In general, the only ferritic pressure boundary matedals in a nuclear plaut which are expected to recehe a fluence sufneient to afect RTNDT are those materials which are located in the core belt!ine region of the reactor pressure vesul. Therefore, materia surveillance programs Il 1
include specimens machined from the plate or forging material and weldments which are located La the core beltline region of high neutron flux density to provide the data rvquired to assess tha depee of neutron embrittlement. ASTM E 185 (2) desedbes the recommended practice for
, monitoring and evalusting the radiation induced changes occurdeg in the mechanical properties of pressure vessel beltline matedals.
Westinghouse has provided such a surveElance program for the Indian Point Unit No. 2 nuclear l power plant QQ). The encapsulated Cy specimens are located on the O.D. surface of the thermal shield where the fast neutron Dux density is 1.08 times that at the adjacent vessel wall surface ,
(at 4
- for Capsule V, see Table IV.2) Q1). Therefore, the increaws (shifts) in transition temperatures of the matedals in the pressure vessel are slightly less than the corresponding shifts observed in the surveillance specimers. However, because of azimuthal variations in neutron aux density, capsule Cuences may lead or lag 'io maximum vessel Guence in a corresponding exposure period. The capsules also contain several doalmeter materials for expedmentally determining the average neutron Gux density at each capsule location during the exposure period.
- The Indian Point Unit No. 2 material surveillance espsules also include tensile specimens as j l
recommended by ASTM E 185. At the present time, irradiated tensile properties are used only l l
to indiate that the matedals tested continue to meet the requirements of the appropdate l
material sped._f. cation, In addition, the matedal surveillance capsules contain wedge opening loading (WOL) fracture mechanics specime.a Current technology limits the testing of these specimens at temperatures well below the minimun. scrne temperature to obtain valid frarture mechanics data per ASTM E 399 QD, ' Standard Method of Test for Platie-Strain Fracture l
Toughness of Metallic Materials.' Currently, tha NRC suggests storing thee specimens until an seceptable testing procedure has been defined for determining the Jy fracture t..ughness q;p.
l Il2
This report desrEcs the results obtained from testing the contents of Capsule V. These data and those obtained previously from Capse'n T, Y, and Z @lO are analyzed to estimate the radiation induced chenges in the mechanicc'. properties of the pressure vessel at the end of Cycle 8 as well as predicting the changes expected to occur at selected times in the future operation of the Indian Point Unit No. 2 power plant. The future prejutions are based on the continued use of a low leabge core loading pattern, put in service at the start cf Cycle 6, which involves placing burnt assemblies at the periphery and minimal fresh assemblies instead of all fresh assemblies at the periphery n that the peak vessel wall neutron Dux is reduced by apprtaimately 45 to 50 percent.
I 11- 3
M. DESCRUTION OF MATERIAL SURVEILLANCE PROGRAM T
The Indian Point Unit No. 2 matedal surveillance program is described in detail in WCAP I
7323 (1Q), dated May 1969. Eight matadals surveillance caps ales (Ave Type I and thru Type II) i were placed in the ructor vessel between the thermal shiell and the vmeet wall before startup I
(see Figura m 1 and m 2). The verthm! center of each espoule is oppoelta the vertion! center of the core. The neutron flux density at each 4* capsult i eation slighdy ucads 1.00 timw the maximum Sux density on the vasel I.D. (II). However, the pak vessel exposure rate has been signtAcantly reduced since the introduction of a low leakage core loading pattern in Cycle 6. .
)
I i
Capsule V, a Type !! capsule, was removed during the 1987 refueling outage. The T)v II [
capsules ech wtain Charpy V notch, tensile, and WOL specimens machined from the thru f
, t SA533 Gr B, Cl 1 beltline shell plates plus charpy V notch specimens machined from a correla- l tion monitor heat of steel. The chemistdes and but trutments of the vnul surveillance j materials are summadzed tu Table m 1. All test specimens were machined from the test l 1
matedals at the quarter thickness (1/4T) location. The longitudinal base metal Cy specimens !
i were odented with their long axis parallel to the pdmary rolling direction and with V notchee I l
perpendicular to the major plate surfaces. Tensile specimens were machined with the longitu. j dinal axis parallel to the plate pdmary rolling direction. The WOL specimens were machined ,
i with the simulated crack perpendicular to the pdmary rolling direction and to the major plate l l
- surfaces. All mechanical test specimens (see Figure W 3) were taken at lust one plate thick- l l
ness from the quenched edges of the plate material. ,
i l
r m1 !
i I
L
t 1
?.(TrPeIl 173' t
. V ITYpe U)
/ '
ff/ 3 (TYPE El I
/
tl(TrPeIl 0'
.f. .
180' -
l
- : / ,
r W (TrPe 11 X (TrPe !)
/
/
//
L Y (Type U) l d' i 7 (T YP* I) l Thermal sueldlReasi Core Barret l
I l
r t
i Figure m.1. Arrangement of surveillance capsules in the pressure vessel !
t l
III.2 I i
i
i I
i i
e 4 !
l i
W/ 4 4C i
\
l A Ic4 "' se e , i l
V//// r,
/ w
[
I '36/g
/
'h
/
) ,
l j
W
- / t i l A I E I I T E1 7 1 -
T I
., / !/
/ 3 1
/
/ // l l/ /
j l l i
- I // i l
8
// l
\ /,' t I i 1,
// j l
1
/p _ _ _ _ .
i l
, t i
t Figure m 2. Indian Point Unit ? eactor geometry l (Reference ';)
D' 3
Table III 1 INDIAN POINT UNIT NO. 2 REACTOR VESSEL SURVEILLANCE hfATERLES (M)
Heat Treatment History Shell Plate hiaterial:
Heated to 15501600*F for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, water quenched.
Tempered at 1225'F for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, air cooled.
Stress relleved at 1150*F for 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />, furnace cooled to 600*F Weldment:
Stress relieved at 1150*F for 19,75 hours8.680556e-4 days <br />0.0208 hours <br />1.240079e-4 weeks <br />2.85375e-5 months <br />, furnace cooled to 600*F -
Correlation hfonitor-.
1650*F,4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, water quenehed to 300*F 1000*F,6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, air cooled.
Chemleal Comeosition (Percent) hinteris] C .Mn, P l ,fi _HL .M2. .fdl.
Plate B20021 0.20 1.28 v.010 0.019 0.25 0.53 0.46 0.25 Plate B2002 2 0.22 1.30 0.014 0.01S 0.22 0.46 0.50 0.14 Phte B2002 3 0.22 1.29 0.011 0.020 0.25 0.57 0.46 0.14
, Correlation hionitor 0.24 1.34 0.011 0.023 0.23 (a) 0.51 (a)
Weld hietal (a) (a) (a) (a) (a) (a) (a) (a)
(a) Not reported in WCAP 7323 (H).
i Capsule V contained 32 Charpy V. notched specimens,4 tensile specimens (2 from weld metal and 2 from plate), and 4 base plate WOL s,wimens. The spedmen numbering system and I
location within Capsule V is shown in Figures 111-4 and 1115.
III-4
l e
r 44 h
I i P
i g e N [
t a j W HItt i M
-a .
t l
la) Charpy V netch Impact Specimen
=' ;g 1 ..
.& N" g ,
H t' I
'+.
=_.e t
L3H
. - . . _I L itt 64N j I , M M i
$)J .
{ $- m-
,w. .. . .
<- i .i r. .. u. s p. . tm. .
- G-ED [
yy.au i , r ,i >
t r fiir 44 - ;;WI ' -- --
% t!T;' *= i
, a$.. .t\14 1 {
g$4' Y N. l ti ;
i., w. o. c,..a. t... m s,.. w..
I i
Figy ' 3, Venel mate 'al surveillance spedmens i
m-5 j 1
- - .- _ _._ _ _ _ _ _ _ . _ _ _ _ . ._ _ ._ ___ ______ __ _ ,,_.,_ -.._..._ m, _ - - _ _ _ . . ~ . . _ . _ _ . . - - .
._ c Ca re to Cu Co(CJ) Hi Co(Cd) Cu Co(Cd).
5g 35 as =3 g5 13 ' I" 33 I si is si s1 38 3 88 #1 31 II e i S2 ss s2 ts s LJ tl U U ili-ft 30 ei BR 579*F lionitor 590*F Honitter 579'T Honitor C, Tensile C, C, Cy C, Tena11e C,
_o_
-p_%9l H9 till Wil HIJ W12 HI3 W13 H14 W14 HI6 i
W16 W4 W3 - - - - - - L -
27 26 - -
R49' 2-41 RSI 24 3 RS2 244 R53 244 RS4 246 RS6 248 4
8 g 1 E C t ~
UOL I UOL WOL UOL C,
& I Hit W10 24 H15 WIS W3 W4 , Dosini- 211 212 - -
R50 242 eter R55 247 Block
\ l f I l l 1 1 h I
l l s I E I e
o To Vessel Botton ew 1 i Vessel Wall Side
+ : . #a 4 i : 9 ,o f
ru 2 - Plate B2002- 2 Il - IIAZ U - Ifeld Metal R - Correlation Honitor Plate co
-a e
o W
t>
19gure 1114. Arrangement of specimens in Capsule V
MOTE : RLL Dine 05to45 BR.E. 14 CE4TiMETtrt S 3 A8I ,
- 2. f.1 ,
- 2.41 ;,
l.L8 ;
_ l.23 i R40105 (CM)
O . 41 -+ :e A13. t 5 ep iwwww.
. s..
i ii1 e t > i to r is,g ,;,g
, 3 ',9, ,
FLV4 0 RES
- O - ala.)&
Flux WiR.1$ eN A)).6&
CSP 5uLE Ct ETER.
\ + C \ A ll.S )
C"n%j FWX WikE 6 -
- dc'kp a 11.1%
j N a. io. a f/ / / / SPR C,ER.
f e r i / e i t////,
,g
, og g 3
% N N NN NNNNNNNN ,'l' a g ' g bI 14 _ . .
THERMA). SM) ELD Figure III 5. Surveillance espsule geometry (Reference 17) dI 7
Capsule V also contained the following dosimeters for determining the neutron flux density:
Table III 2 CAPSULE V NEUTRON FLUX DOSIMETERS Tarnet Element Form Quantity t
Copper Bare wite Nickel Bare wire 1 Cobalt (in aluminum) Bare wire ,
Cobalt (in aluminum) Cd shielded wire 3 Uranium Oxide 1 Neptunium Oxide 1 In addition, ands were cut from 10 tested Charpy specimens to serve as iron dosimeters.
Three eutectic alloy thermal monitors had been inserted in holes in the steel spacers in Cap.
sule V. Two docated at tne top and bottom) were 2.5% Ag and 97.5% Pb with a melting point of 579'F. The other Oocated at the center of the capsule) was 1.75% Ag, 0.75% Sn, and 97.5% Pb having a melting point of 590*F.
III 8
IV. TESTING OF SPECIMENS FROM CAPSULE V t
The capsule shipment, capsule opealng, specimen testing, and reporting of results were carried out in accordance with the Project Plan for Indian Point Unit No. 2 Reactor Vessel Irradiation Suneillance Program. The SwRI Nuclear Projects Operating Procedures called out in this plan include:
(1) XIII MS 1041, ' Shipment of Westinghouse PWR Vessel Material Suneillance Capsule Using SwRI Cask and Equipment" (2) XI4fS 1011, ' Determination of Specific Activity and Analysis of Radiatica Detector Specimens *
(3) XI hiS-103-1, ' Conducting Tension Tests on afetallie Specimens *
(4) XI4fS-104 1, "Charpy Impact Tests on Metallic Specimens' (5) XIII41S 1031, ' Opening Radiation Suneillance Capsules and Handling and Storing Specimens
- i Copies of the above documents are on file at SwRI.
A- Shipment, Opening, and Inspection of Capsulo Southwest Research Institute utilized Nuclear Projects Operating Procedure XIII41S 1041, as incorporated in approved Consolidated Edison Co. procedures, for the shipment of Capsule V IV 1
._ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . - ]
to the SwRI laboratories. On March 30, 1988, SwRI personnel severed the capsule from its extension tube, sectioned the extension tube into several lengths, supervised the lovling of the capsule and extension tube materials into the shipping cask, and transported the cask to San Antonio, Texas. The capsule arrived at the SwRI Radiation I4boratory on April 5,1988, and unloading of the capsule commenced the next day.
The capsule was opened and the contents identified and stored in accordance with Proce-dure XIII.MS-1031. The long seam welds were milled off using a Bridgeport vertical milling machine. Before milling the long seam weld beads, transverse saw cuta were mado to remove the capsule ends. After the long seam welds had been milled off, the top half of the capsule shell was removed. The specimens and spacer blocks were carefully removed and placed in indexed receptacles identifying each capsule location. After the disassembly had been completed, each specimen was carefully checked to insure agreement with the identification and location as listed in WCAP 7323 QO. The following discrepancies were found and corrected:
Two Charples were both marked R 55 on one end and R 56 on the other end. The Charpy that was in the R-55 position was remarked properly on the other end and the R 56 Charpy was also remarked by crossing out the R 55 and remarking the end as R 56.
The thermal monitors and neutron dosimeter wires were removed from the holes in the spacers. The thermal monitors, contained in quart: vials, were examined. No evidence of melting was observed, thus indleating that the maximum temperature during exposurs of Capsule V did not exceed 579'F. All neutron dosimeters were in the positions called o' .t in WCAP 7323 and were correctly accounted for. However, the Neptunium container had an appearance that had not been encountered before. The Uranium and Neptunium containers are ly.2
-en in F!gwe IV 1. The deformed condition of the Neptunium container caused the loss of ast of the sample during opening.
i l
B. Neutron Dashnetry The dosimeter wires were weighed on a Mettler microbalance, and the Charpy slices were weighed on a Mettler digital balance. The gamma activities of the dosimeters were determined in accordance with Procedure XI MS 1011 using an IT-5400 multichannel analyzer and an intrinsic Ge enav;al detector system. The calibration of the equipment was accomplished with ,
54Mn, 60Co,and 137Cs radioactivity standards obtained from the U.S. Department of Commerce National Bureau of Standards. All activities were corrected to the time-of. removal (TOR) at reactor shutdown.
Infinitely dilute saturated activities (ASAT) were calculated for es:h of the dosimeters because A SAT si dimtly related to the product of the energy-dependent microscopic activation cross section and the neutron Dux density. The relationship between ATOR and ASAT is given by:
ATOR m=n At m I:
ASAT m=1 Pm(/1e' aim)\ e wbere: A = decay constant for the activation product, day *I; tm = decay time after operating period m, days; Tm = operating days; Pm = average fraction of full power during operating period.
The values of Tm and Pm up to the 1987 refueling shutdown for Indian Point Unit No. 2 are presented in Table IV 1. The calculation of the neutronic factors is described below.
IV 3
t 4 ff ,g s,
fj%
5 p- <
8' gg ,
! U NP l
Figure I%1. Uranium and Neptunium contafners as removed from dosimeter block I%4
~
TableIV 1
SUMMARY
OF REACTOR OPERATIONS INDIAN POINT UNIT NO,2 Operating Fraction of Operating Dates Days Shutdown Full Power Period Stars Stop G)m Days G)m 1 08/15/73 08/24/73 10 - 0.4377 08/25/73 08/25p3 - 1 -
2 08/26/73 09/07/73 13 - 0.4532 09/08/73 09/20/73 - 13 -
3 09/21/73 09/28/73 8 - 0.3161 09/29/73 09/30p3 - 2 -
4 10/01/73 10/12p3 12 - 0.3088 10/13/73 01/25/74 - 105 -
5 01/26/74 01/29/74 4 - 0.2412 01/30/74 03/21/74 - 51 -
6 03/22/74 04/18/74 28 - 0.5438 04/19/74 04/28/74 - 10 -
7 04/29/74 06/03/74 5 - 0.4062 06/04/74 06/04/74 - 1 -
8 06/06/74 06/10/74 6 - 0.4743 06/11/74 06/12/74 - 2 -
9 06/13/74 06/13/74 1 - 0.0730 06/14/74 06/20/74 - 7 -
10 06/21/74 06/14/74 25 - 0.6G53 06/15/74 06/16/74 - 2 -
11 06/17/74 07/22/74 36 - 0.7601 07/23/74 07/23/74 - 1 -
12 07/24/74 07/26/74 3 - 0.7503 07/27/74 08/06/74 - 10 -
13 08/06/74 00/06/74 32 - 0.C653 09/07/74 00/09/74 - 3 -
14 09/10/74 09/30/74 21 - 0.7420 10/01/74 10/11/74 - 11 -
15 10/12/M 11/09/74 20 - 0.8037 11/10,14 11/10/74 - 1 -
16 11/11/74 12/06/74 26 - 0.8300 12/07/74 12/07/74 - 1 -
17 12/08/74 01/01/75 25 - 0.8405 01/02/75 01/04/75 - t 3 -
18 01/05/75 01/05/75 1 - 0.5450 01/06/75 01/06/75 - 1 -
19 01/07/75 01/31/75 25 - 0.8810 02/01/75 02/02/75 - 2 -
20 02/03/75 02/28/75 26 - 0.9408 03/0105 04/03/75 - 34 -
21 04/04/75 05/02S5 29 - 0.7G32 05/03/75 05/03/75 - 1 -
22 05/04/75 07/28/75 86 - 0.0114 07/20/75 08/10/75 - 13 -
23 08/11/75 09/12/75 33 - 0.7108 00/13/75 09/13/75 - 1 -
24 00/14/75 10/16/75 33 - 0.7062 10/17/75 10/20/75 - 13 -
25 10/30/75 11/14/75 16 - 0.74G7 11/15/75 11/15/75 - 1 -
26 11/16/75 01/04/76 50 - 0.5427 01/05/76 01/05/76 - 1 -
27 01/06/76 01/20/76 24 - 0 8703 01/03/76 02/04/76 - 0 -
28 02/05/76 03/30/76 55 - 0.0122 03/31/76 09/26/76 - 180 -
20 03/27/76 CD/27/76 1 - 0.0030 03/28/76 03/28/76 - 1 -
30 09/20/76 10/29/76 31 - 0.8423 10/30/76 12/10/76 - 42 -
IV 5
I Table IV.1
SUMMARY
OF REACTOR OPERATIONS INDIAN POINT UNIT NO. 2 (CONT'D)
Operating Traction of Operating Dates Den Shutdown Full Power Period Start Stop a) m Den (Pm) 31 12/11/76 01/27/77 48 - 0.8306 01/28/77 01/29/77 - 2 -
32 01/30/77 02/01/77 3 - 0.7250 02/02/77 02/0607 - 4 -
33 02/06/77 03/11/77 34 - 0.3825 03/12/77 03/14/77 - 3 -
34 03/15/77 04/10/77 27 - 0.9242 04/11/77 06/13/77 - 33 -
35 06/14/77 07/02/77 50 - 0.8936 07/03/77 08/06/77 - 34 -
36 08/06/77 08/19/77 14 - 0.6372 08/20/77 08/21/77 - 2 ,
37 08/22/77 02/13nt 176 - 0.0022 ,
- 02/14/78 06/24/78 - 100 -
38 06/25ns 07/28n8 as - 0.8000 07/2908 07/30n8 - 2 39 07/31/78 09/15/78 47 - 0.9820 09/1608 10/05/7s - 20 -
40 10/06ns 11/23n8 49 - 0.93c0 11/2408 12/02n8 9 -
41 12/0308 06/15/79 196 - 0.0000 06/16/79 09/14/79 - 91 -
42 09/15/19 11/27/79 74 - 0.8120 11/28/79 11/29/79 - 2 -
43 11/30/79 12/02/79 3 - 0.1840 12/03/79 12/07/79 - 5 -
1 44 12/0800 01/11/80 35 - 0.8710 01/12n8 02/09/s0 - 29 -
45 02/10/80 02/14/80 5 - 0.4200 02/15/80 02/18/80 - 4 -
46 02/19/80 06/03/80 106 - 0.9310 06/04/80 06/11/80 - 8 -
47 06/12/80 08/10/80 60 - 0.0310 08/11/80 08/13/80 - 3 -
48 08/14/80 10/17/80 65 - 0.9400 10/18/80 05/21/81 - 216 -
49 06/22/81 07/10/81 50 - 0.7120 07/11/81 07/11/81 - 1 - ,
50 07/12/81 08/21/81 41 - 0.9C40 l 08/22/81 09/15/81 - '. 'S - t 51 09/16/81 10/05/81 20 . 0.0040 l 10/06/81 10/15/81 - til -
52 10/16/81 11/11/81 27 . 0.9710 l
12/12/81 11/22/81 - 11 53 11/23/81 04/02/82 131 - 0.9500 l 04/03/82 04/03/82 - 1 54 04/04/82 05/17/82 44 - 0.0230 f 08/18/42 05/23/82 - 0 -
55 05/24/82 08/12/82 81 - 0.0520 i I
03/13/82 08/14/82 - 2 -
56 08/15/82 00/02/82 19 - 0.7800 ;
5 - :
09/03/82 09/07/82 -
I 57 09/08/82 00/17/82 10 - 0.7080 09/18/82 01/01/83 - 100 -
58 01/02/83 01/05/83 4 - 0.3453 01/06/83 01/06/83 - 1 59 01/07/83 01/08/83 2 - 0.0355 2 -
01/00/83 01/10/83 -
60 01/11/83 01/31/83 21 - 0.7303 02/01/83 02/11/83 = 11 t
IV.6
Table IV.1 SI'MMARY OF REACTOR OPERATIONS IlTIAN POINT UNIT NO. 2 (CONT'D)
Operating Traction of Operating Detas Days Shutdown Full Power Period Start Step (Tm) Days (Pm) 61 02/12/83 02/13/83 2 - 0.0000 02/14/83 02/14/83 - 1 -
62 02/18/83 02/18/83 4 - 0.1025 02/19/83 *J2/19/83 - 1 -
63 02/20/83 08/27/83 189 - 0.9619 08/28/83 08/28/83 - 1 -
64 08/29/83 10/04/83 37 - 0.9672 10/06/83 10/28/83 - 21 -
66 10/26/83 01/08/84 72 - 0.9248 Ol/08/N 01/07/84 - 2 -
66 01/08/84 02/11/84 38 - 0.9228 02/12/84 02/26/84 - 15 - .-
67 02/27/84 06/01/84 96 - 0.9100 !
06/02/84 10/20/84 - 141 - t 64 10/21/84 11/30/84 41 - 0.8706 12/01/84 12/01/84 - 1 -
69 12/02/84 12/19/84 18 - 0.9147 12/20/84 12/26/84 - 7 -
70 12/27/84 12/28/84 2 - 0.0060 12/28/84 12/31/84 - 3 -
71 01/01/88 09/20/88 263 - (19500 09/21/85 09/22/86 - 2 -
72 09/23/85 10/21/28 29 - 0.G813 10/22/85 10/23/85 - 2 -
73 10/24/85 01/13/86 82 - 0.9298 01/14/86 06/24/86 - 131 -
74 06/25/86 06/28/86 4 - 0.1083 ,
05/20/86 08/20/86 - 1 -
75 05/30/86 08/31/86 2 - 0.2835 06/01/86 06/06/86 - 6 ..
76 06/07/86 06/09/86 3 - 0.1020 06/10/86 06/10/86 - 1 -
77 06/11/86 10/20/86 132 - 0.0339 t 10/21/86 10/22/86 - 2 -
78 10/23/86 10/23/86 1 - 0.0710 -
10/24/86 10/26/86 - 3 -
(
79 10/27/86 11/06/86 11 - 0.9146 -
11/07/86 11/08/86 - 2 - l 80 11/09/86 11/15/84 7 - 0.7844 i 11/16/86 11/16/86 - 1 -
81 11/17/86 01/30/87 78 - 0.0393 01/31/77 02/06/97 - 7 - l 82 02/07/87 02/10/87 4 - 0.7053 ;
02/11/87 02/12/87 - 2 - >
83 02/13/87 06/27/87 135 - 0.0804 i 06/28/87 06/23/87 - 2 -
[
84 06/30/87 10/04/87 97 - 0 0310 l
l I
[
IV.7 (
s,- , . -, , -
Westinghouse performed a two-dimensional ordinates Su transport analysis to determine the neutron fluxes and energy spectrum within the reactor vessel and surveillance capsule of Indian Point Unit 2. This analysis was undertaken to calculate the spectrum averaged cross sections for the threshold and the fission detectors, the lead factors for use in relative neutron exposure of the pressure vessel to that of the surveillance capsule and iron atom displacement (DPA). t Westinghouse undertook two distinct calculations for the Indian Point Unit 2 reactor pressure vessel. First was a single computation in the conventional forward mode to obtain f relative neutron energy distdbutions throughout the reactor geomety as well as through the vessel wall. This transport calculation was carried out in R, i geometry using the DOT two- ,
dimensional discrets ordinates code and the SAILOR cross.section library. The SAILOR library i
is a 47 group ENDFB-IV based data set produced specifically for light water reactor appilcations.
In this calculation P 3anisotropic scattering and Sg order of angular quadrature was used. The reference forward calculations was normalized to a core mid. plane power density characteristic i r
I of operation at a thermal power level of 2758 MWt.
I The second calculation consisted of a series of adjoint analysis relating the fast neutron ,
I i
flux (E > 1.0 MeV) at surveillance capsule positions and several azimuthal locations on the i
i
- pressure vessel inner radius to neutron source distributions vithin the reactor core. All adjoint !.
i l
3 analyses were also carried out using an Sg order of angular quadrature and P3 anisotropic ,
scattering using the 47 group SAILOR Library as described above.
I l The core power distributions for each cycle used in fast neutron exposure evaluation were j taken from Indian Point Unit 2 nuclear design reports.
$ IV 8 f i .
4
.___.--,_m.___.--, ,,,n-
l The pertinent factors (1) calculated spectrum averaged reaction cross sections and (11) calculated cyde dependent Guence lead factors obtained from these transport calculaticus are summarized in Table IV 2. The calculated spectntm averaged reaction cross sections are employed in the analysis of fast stron monitors activity data for the prediction of fast neutron flux /Suence (E > 1.0 MeV) at surveillance capsule location and the calculated lead factors for the prediction of reactor vessel aux / fluence (E > 1.0 MeV) from the surveillance.
Neutron Cycla 5 lead factor results given in Table IV 2 are representative of a standard loading pattern cycle as Indian Point Unit 2 employed this loading pattern from Cycle 1 through i
Cyde 5. Cyde 8 results are for the low leakage loading pattern as the low leakage loading pattarn was implemented at indian Point Unit 2 starting from Cyde 6. f TableIV 2 RESULTS OF DISCRETE ORDINATES Sn TRANSPORT ANALYSIS (jl)
INDIAN POINT UNIT NO. 2 1
A. Calculated Spectrum Averaged Reaction Cross Sections (ae gr) for Analysis of Fast Neutron i Monitors (E > 1.0 MeV)
Reaction (barns) 4' 40*
54Fe(n,p)54Mn 0.0887 0.067 :
I 58Ni(np)58Co 0.116 0.0914
, 63Cu(n.o)60Co 0.00119 0.000694 238U(n,0137Cs 0.372 0.343 237Np(n,f)l37Cs 2.63 ,
2.84 l
B. Calculated Fluen:e Imad Factors (a) for Indian Point 2 Cycles 5 and 8 (
Cy.Gli :l* :1.Q* l 5 1.08 3.42 l 8 1.19 3.40 EOC Fluence at Surveillance Location j (a)L.F =
EOC Fluence at RPV O T Location IV 9
The primary result desired from the dosimeter analysis is the total neutron fluence (E >
1 MeV) which the surveillance specim: ns and pressure vessel have received. The average flux a Ic.ll poweris given by:
p=ASAT/NoI ASAT = Saturated activity (rste of decay = rate of production) in disintegration /sec or Bq where d = energy dependent neutron flux, n/cm2 ,,e 7 = spectrum averaged activation cross section, em2; and No = number of target atoms per mg.
The total neutron duence is then equal to the product of the average neutron flux and the equivalent reactor operating time at full power.
In Capsule V, the Correlation Monitor and B2002 2 shell plate Charpy specimens were located in the specimen layer nearest to the vessel wall and the weld metal, heat affected zone (HAZ) Charpy specimens were located in the specimen layer nearest to the core. Since there is a radial dependence of the fast neutron flux in the vessel, the neutron exposure received by the Correlation Monitor and B2002 2 shell plate Charpy specimens is expected to be lower than that received by the weld metal and HAZ Charpy specimens. The dosimetry program is capable of providing information on the radial dependence of the fast flux because the Charpy ends used for iron dosimetry were taken from both of the Charpy specimen 1systs (nearest to and farthest from the core).
Since Indian Point Unit No. 2 operated for 8.6 effective full power years (EFPY) up to the 1987 refueling outage, the calculated fluences for Capsule V and the vessel up to the 1987 refueling outage are as presented in Table IV 3. Thermal neutron dux (fluence rate) values from Capsule V are presented in Table IV4 IV 10
Table IV-3 DOSIMETER ACTIVITIES AND MEASURED FLUENCE RATE IN CAPSULE V Dosimeter ATOR ASAT Measured 4 (> 1 MeV)(a)
Position ID (Bq/Mg) (Bq/Mg) (n em 2 sec 1)
R= 211.18 (Core Side cf Charny Comnartment):
N1 1: 25.4 16860.0 2 08E10 Bottom Cu '/6.8 138.6 1.76E10 Top Cu 79.1 142.8 1.82E10 Bottom Fe W 9 670.2 842.4 1.52E10 Bottom Fe W 12 681.1 856.1 1.54E10 Bottom Fe H 12 717.7 902.1 1.63E10 Bottom Fe W 13 667.7 839.1 1.51E10 Top Fo H 16 751.3 944.1 1.70E10 Ave: 1.70E1021.9E9 R = 211.68:
238 U 239.1 1398.3 2.47E10 237 Np (9820) (5740) (1.J1E11)
NOTE: Np Results are not reliable because an inadequate sample was recovered (see comments in text)
R = 212.18 (Vessel Side of Charov Comoartment):
Bottom Fe 2 41 571.9 718.8 1.30E10 Bottom Fe 2-44 582.0 731.5 1.32E10 Boucat Fe R 52 615.6 773.8 1.39E10 Bottom Fe 2 45 565.8 711.2 1.28E10 Top Fe R 56 622.3 782.2 1.41E10 Aye: 1.34E10 6.0E8 (a) Measured 4 (> 1 MeV) = ASAT . (ATOR/h)
No 000 No 00 0 1
IV 11
i Table IV 3 (Cont'd)
DOSIMETER ACTIVITIES AND MEASURED FLUENCE RATE IN CAPSULE V Determination of Fluence Rate at Centerline of Surveillance Capsule V, Indian Point 2 Dosimeter Centerline Radial Dosimetic d (> 1 MeV) Gradient ((> 1 MeV)
Position ID n/cm2see Factor n/cm 2 ,,e 211.18 Ni 2.08E10 0.953 1.98E10 Cu (bottom) 1.76210 0.956 1.68E10 Cu (Top) 1.82E10 0.956 1.74E10 Fe W 9 1.52E10 0.951 1.45E10 Fe W.12 1.54E10 0.951 1.46E10 Fe H.12 1.63E10 0.951 1.55E10 FeW 13 1.51E10 0.951 1.44E10 Fe H 16 1.70E10 0.951 1.62E10 211.68 238g(a) 2.47E10 1.050 2.60E10 237Np(*) 1.37E11 1.049 1.44E11 212.18 Fe 2 41 1.30E10 1.152 1.50E10 Fe 2 44 1.32E10 1.152 1.52E10 Fe R 52 1.39E10 1.152 1.60E10 Fe 2 45 1.28E10 1.152 1.47E10 Fe R 56 1.41E10 1.152 1.62E10 Average (a) Fluence. Rate = 1.59E1011.5E9 at Center of Capsule V (a)238U and 237 Np results not included in average (Cs-137 halflife allows influence from high leakage cores in cycles 1 through 5) t Value is la from variation ofindividual values included in the average IV.12
Table IV-4 THERMAL NEUTRON FLUENCE RATE IN INDIAN POINT 2, CAPSULE V 59Co Bare 59Co Cd Covered Axial ATOR. ASATe("} ATOR, ASAT,(*) Thermal Flux Location Bq/Mg Bq/Mg Bq/Mg Bq/Mg n/cm2 ,
Top 3.22E6 5.81E6 1.37E6 2.47E6 f.C;E9 Middle 3.10E6 5.60E6 1.39E6 2.51E6 8.15E9 Bottom 3.49E6 6.30E6 1.28E6 2.31E6 1.05E10 Average 3.27E6 5.90E6 1.35E6 2.43E6 9.15E9 (a) 60Co saturation factor a h = 554; ASAT " ATOR/h The variations in the peak vessa.) Sux values (i9.4% from variations in individual values) determined frort the several dosimeter materials may be attributed to the uncertainties in measurements and calculations (in the calculated spectra and in the reaction cross sections).
Uranium dosimeter values are higher than others because the Cs-137 product half. life is 30.1 yr and retains some activity from the earlier higher leakage cores.
Neptunium dosimeter values are not dependable because insufficient material was recovered from the capsule. The aluminum shell containing the Neptunium was krittle and cracked open on the lathe while being opened. Most of the Neptunium oxide was not recoverable.
Aversging the results obtained from the Capsule V iron, copper, and nickel neutron dosim.
eters, the peak neutron flux incident on the center of Capsule V is calculated from Table IV 3 z to be 1.59 x 1010 n/cm2 sec, (E > 1 MeV). This is equivalent to 3.42 x 10 10 n/cm2 3,e (E > 1 MeV)(15).
IV 13
I I
C. Machanle=1 F.py Tests The irradiated Charpy V notch specimens were tested on a calibrated ** SATEC Model SI 1K 240 ft lb,16 ft/see impact machine in accordance with Procedure XI MS-1041. The test temperatures, selected to develop the ductile-brittle transition and upper shelf regions, were obtained using a liquid conditioning bath monitored with a Fluke Model 2168A digital thermom-eter. The Charpy V notch impact data obtained by SwRI on the spedmens contained in Capsule V are presented in Tables IV 5 through IV 8. The shifts in the Charpy V notch tranal.
tion temperatures determined for the three vessel plates and the correlation monitor are shown in Figures IV 2 through IV 5. The Capsule T (B), Capsule Y (H), and Capsuls Z (H) results, included in the figures for comparison, show that Capsule V is a low lead factor, low Gux capsule, as expected.
A summan of the shifts in RTNDT determined at the 46 ftlb level as specided in NUREG-0800 (H) and Appendix G to 10CFR50 q), and the reduction in Cy upper shelf energies for each material, is presented in Table IV 9.
s
- Inspected and calibrated using specimens and procedures obtained from the Army Materials and Mechanics Research Center.
IV.14
TABLE IV-5 CHARPY IMPACT DATA WITH PHOTOS OF FRACTURE FACES
"*
- 2' ""
WATER At . ( WE l.a )
SPECIMEN TE.wP ENERGY LATERAL FRACTURE PHOTOGMPH NO. 'F FT-L35 EXPANSION APPEARANCE 1X 74'F '
W. 9 24.0 .014 0 l
W 10 +130 26.5 .02) :o [_ -
I W ti +130 :.v . 3 .035 so 0 ~ ]f )n. .
W-12 -2:0 53.0 .oes 53 f, W-13 +:50 52.3 .c:a 95 W- u -300 7o.o .uu o5
' *o +325 72.5 .uos 45 W-;5 -350 7*.0 .0,7 tco .
I i t
I l
i IV 15
TABLE IV-6 CHARPY IMPACT DATA WITH PHOTOS OF FRACTURE FACES i
F r : e : ' . 2. 17-2106-001
'" l " 2' I*
'A7E9lAL - B :002-2
~
SPECIMEN TEMP ENERGY LATEML FMCTURE PHOTOGMPH NO. 'F FT-LB5 EXPANSICN APPEAUNCE 1x ,
1-i 74 r 17.5 .0te 5 j 2 42 +120 50.0 .042 15
) -
.- }
2-ed +150 60.5 .246 20 b _I _ _v . . _
2-aa +130 93.0 .059 e0 2 43 +220 111.0 .050 90 f'i
<!' ; .. g 2-45 + 60 109.5 .078 100 2 a6 +300 110.0 .075 100 s
s s *J l '
2*EI *130 10e.0 .ce ;co f ,
l
\
e- >
N l I i i
, i I
i :
i l l l
1 l i l
t IV 16 l
1 l
l 1
t _
TABLE IV 7 CHARPY IMPACT DATA %TTH PHOTOS OF FRACTURB FACES F. ; e : .;. 17 .'iv$ 00i
a!~1* L . (Reference) -'
! i 2PECIMEN TEMP ENERGY LATERAL FRACTURE PHOTOGRAPH NO. 'F FT-LBS EXPANSION APPEARANCE ,1_X l
l 4:i o ' ..
! n-a9 74*y i : ). 5 .014 5 JL.i ' . ' 'j
! I -
R-50 +130 32.0 .021 20 , , . , , , ,
l.1,. >2., .e>, >e 3 p.
,.s1
..me ,e., .e . ,,
g_p._.
A-52 +22u e2.0 .uss 15
<-,a 2:a , . , .m ;-
y
..,. .,2,
,.; .~ ,cc y .
R - 5 .* +35U 7..Q .ce' 'CO .s.
6 .
J'?
sm' , -
I
- i
, , 1 a
1
(
l j
IV 17 I
TABI.E IV 8 CHARPY IMPACT DATA WITH PHOTOS OF FRACTURE FACES
.=. : ..:. . ; . iu s-w a
- a t . J '2 n e 2. i4 55 u;. q.;( , ( pg ) __
i ,
SPECittEN TEMP ENERGY l LATERAL FRACTURE PHOTOGRAPH NO. 'F FT-LBS I EXPANSION l
APPEARANCE LX h-11 0 30.5 .003 25 H-10 +30 55.0 .052 90
{
~
H9 53.5 .0a0 50 l,n )q \ .< '
H-12 +110 $3.5 .047 $0 H-13 -l50 93.0 .05) so
- 200 '*)." 1(;0
- d-la
. Gee AL, . I H-lo +250 73.9 .067 .o .
L"_ _
i -
H 15 *250 ;22.5 .977 'G0 l
'w' -
l l 1 I
- l ' ,
\ _ 6 I
l i ,
l I l l l '
l
! l l i I . ,
l , I I I ,
1 IV 18
rLATE B2002-2
. i i . . i i i 160 - -
O BASEUNE
- CAPSULE T ,
4 CAPSULE Z
+ CAPSULE V O n 120- -
M
& +
1 0
+ +
30 -
0 ;
85'F
- y. 77 ft.lb. o= ,
0 m ,
w
( ,
46 ft.lb. g 0
80'F
, oOO O -
1 t 1 f f f f
-200 --100 0 100 200 300 400 4
TEMPERATURE, deg F.
J l
i i i i e i i
' 100 - _
O BA3EUNE q % CAPSULE T j
= $ CAPSULE Z 0
- [ 75 -
+ CAPSULE V + +
+
5 0
/
l 2 0 . I 4 Z ,
i 0-50 -
M 0 077'y *r W _j J 0 4
<C 05 mil ,
m 25 -
, a ;
', w e t > 0 0 i o l
- 5. .
l 0 i O -
)
i a i i , ;__ , I ;
( -200 -100 0 100 200 300 400 TEMPERATURE, deg F.
Figure IV 2. Radiation Response ofIndian Point Unit 2 Shell Plate B2002 2 i !
IV 19
i i
WELD METAL 160 -
O BASEUNE A CAPSULE Y i + CAPSULE V I
g 120 - i l
- 9 i
O l i
80 r J p i +
' ^ l C 4
% A w 239' a Z ': -
50 ft.lb. t 40 L -
Oo =
204' !
] o 30 ft.lb. ;
O O i 0- - -
f ? t i I f 1
-200 - 100 0 100 200 300 400 TEMPERATURE (deg F) f 4
j
, = _ . y _ _ .-. 7. . . , , ,
i 100 -
J l 0 BASEUNE n ,
i i g A CAPSULE Y c
- i :: + CAPSULE V E
i v 75 - -
Z O o
l Di k
c.
50 -
^
1 * -
X A ^
I W O
_ 210'
) a .= ,
2
<( 00 m 25 -
J' l
$ o 35 mil l
d O O ^
i 0 -
J l r , , , ,
-200 -100 0 100 200 300 400
! TEMPERATURE (deg F)
Figure IV 3. Radiation Response ofIndian Point Unit No. 2 Weld Metal IV 20
HAZ MATERIAL I
4 i i i e i i 160 - -
O BASEUNE j A CAPSULE Y
+ CAPSULE V m 120 - + -
m
- S O 1 0
+
0 + 0 80 -
a o A A
+ -
>- 77 ft.lb. 190'F E A W
Z 46 ft.lb.-
C Of/o a 162*F W ~
- 40 -
g 0
0 - -
t t t t t t 1 ;
-200 -100 0 100 200 300 400 TEMPERATURE (deg F) ;
e i 3 i I i i i 100 - -
O BASEUNE Q A CAPSULE Y 0 '
- + CAPSULE V o E o A o +
t v 75 -
g Z o + A 9 o O M A Z # A + 1 4 50 - -
l I m o 0 r
- y l LJ o 164'F A '
O J
l 4 0 & l m 25 -
35 mil 1 L i
w I r
& O O l
{
L 0 -
! ' I f i ? f
-200 -100 0 100 200 300 400 i TEMPERATURE (deg F) ;
j Figure IV 4. Radiation Response ofIndian Point No. 2 Heat Affected Zone ';faterial IV 21 ,
[
CORRELATION MONITOR 6 I i i 4 I g 100 -
O B4EUNE
- CAPSULE T
$ CAPSULE Z
~ + CAPSULE V 7 O CAPSULE Y
~
$ 00 J. o .
50 -
9** -
y 46 ft.lb.
104*F
$ 0 la ,
z W o 0 0 e++* a 25 - ,
O
.a +
s a 0 - -
f I t f i f f
-200 -100 0 100 200 300 400 TEMPERATURE, deg F.
i i i i i i i i 100 - -
O BASEUNE g $ CAPSULE T
- . $ CAPSULE Z 75 - + W SULE V _
O CAPSULE Y Z
- O +* g*
z n.
50 -
+ -
d - 88'F #+
a (d +
- 35 mil w
25 -
O a O
l 5 . ;
a 0 - -
l t f f f f
-200 100 0 100 200 300 400 ,
4 TEMPERATURE, deg F.
Figure IV.S. Radiation Response ofIndian Point No. 2 Correlation hionitor hiaterial IV.22
Table IV.9
SUMMARY
OF RTNIyf SHIFTS AND UPPER SHELF ENERGY REDUCTION (Cy)
FOR MATERIAIEIN CAPSULE V A. Sa==ary of Muence and Measured ARTNIyr Values for Test E-;+-! - :la Capsule V Fluence Type of Neutron Measured ARTNDT PF)
Matedal em 2 50 Ft Lbs 30 Ft Lbs 35 mils * ;
Weld 5.59E18 239 104 230 ,
77 Ft Lbs 40.'t Lbs l
Plate B2002 2 4.57E18 85 80 97 !
HAZ 5.59E18 190 162 184 Correlation Monitor 4.57E18 NA" 104 108 B. Decreasein Upper Shelf EnerEy(Cy )
Initial Shelf Capsule V"' Cy Material Ftlb _f,tdb Jtdh % Decrease il I
B2002 2 117 111 6 5 Weld Metal 118 75 43 36 l HAZ 100 93 i(nil) 2 f
Correlation Monitor 118 70 43 41 l
'35 mil + 20*F included in table.
"The upper shelf energy for this capsule was below 77 ft lbs. j
"* Average of 3 Charpy meuurements at as 100% ductile failure. !
l i
l r
[Y*23 [
i
Table IV 9 (Cont'd)
SUMMARY
OF RTNDTSHIFTS AND UPPER SHELF ENERGY REDUCTION (Cy)
FOR MATERIAIEIN CAPSULE V Charpy Impact Data for Deeresse la Upper Shelf Energy j
Shell Plate B2002 2 Weld Metal .-
Samnle EkLh % Ductilitv* Samole EtJA % Ductility 2 45 109.5 100 W 14 76.0 95 2 46 116.0 100 W 16 72.5 95 2 47 1Qfj 100 W 15 264 100 Ave." 111.0 Ave." 75.0 I
Heat Affect *dlang correlation Monitor Samole ,% Duetility* Ekkh % Ductility Et-Lh SADHdt H 14 93.5 100 R 53 67.5 100 H 16 78.0 40 R 54 70.5 100 H 15 1216 100 R 55 224 100 l Ave." 98.0 Ave." 70.0 i ,
i
- Fracture Appearance Ave." Average of 3 highest values with = 100% ductility Tensile tests were carried out in accordane) with Procedure XI MS 1031 using a 22 kip capacity MTS Model 810 Material Test System equipped with an Instron Catalogue No. G 51 13A I 2 inch strain gage extensometer and Hewlett Packard Model 7004B X Y autographic recording r
equipment. Tensile tests on the plate material and the weld metal were run at room tempers-i l l ture at a strain rate of 0.005 in/in/ min. through the 0.2% offset yield strength using i
servo control and ramp generator. The results, along with the room terrperature tensile data j
l l reported by Westinghouse on the unirradiated materials QQ), are presented in Table IV 10. The !
load strain records are included in Appendix B.
i .
I I IV 24
_ -_ _ _ - - _ _ _ _ _ . _ _ _ = _ - _ _ . - - .- _. .
Table IV-10 TENSILETEST DATA RECORDS Cagnale V DATA (a)
Fracture Undona Total Reduction Tcd Spc. Tear.p 0.2%S LTTS Fracture Stress Elongation Elongation in A-ca Material EL. {.D. _fksi) (bi) Load (N Sail _ ._ (5t_. i f%) (%)
Irradiated (a) l Itse 2-2 76 653 8(J 2940 157.9 24.6 25.5 62.4 l H2ilil2-2 2-7 550 66.4 90.4 31'JO 250.4 17.9 17.4 74D Wcld W-3 76 92.7 106.9 3460 188.2 21A 22.0 61.6 W-4 550 825 100.2 3460 174 3 19.6 20.7 58.2 Unisradiated(b) l N ELac -
Rooma 62.4 833 (c) (c) 27.1 (c) ~10.0 l B3112-2 -
! - Roon 663 903 (c) (c) (c) 28.2 69.6 l
(40 533 782 (c) (c) (c) 22.7 64.4 1 -
600 54.7 81.4 (c) (c) (c) 24.7 64.4 1
WcLI -
Roose 63 80.7 (c) (c) (c) 285 73.9 Ronan 65A 81A (c) (c) (c) 26.9 713 600 56.6 793 (c) (c) (c) 24.4 62A
( 410 56.6 ~n .2 (c) (c) (c) 24A 66.9
('llhence - 539 x 1018 m/can 2 ,y > g ge, (le)wap 7333 (c) Data ad regwnted in WCAP 7323
l I Testing of the WOh specimens was deferred at the request of Consolidated Edison Company.
The specimens are in starsge at the SwRI radiation laboratory.
D. Chesnical Ans@ sis Resuha Check analyses for copper and nickel content d the tan broken Charpy V notch specimens used for iron dosimetry and the three ter**d tensile specimens were run using ASTM Method E 322 (12). The results listed in Table IV.11 and IV 12 were obtained. For compic.iness, the list includes chemistry data from prior analyses of these and other survu w ,Luples af reactor vessel materials.
TABLE IV.11
SUMMARY
OF CHEMISTRY VALUES FOR INDIAN PO T UNIT NO. 2 MATERIALS i biaterial Souree of Dar qu,.W!% bl.%'1 .
Plate B2002.1 WCAP 7323 (.25)* (.58)*
Capsule Z: Cy Specimen 13s .22 .62 i Capsule-Z: Cy Specimo 138 .19 .71 CapsuleZ: Tensile Specimen 15 (.29)* .61 Capsule T: Cy Specimen 12 .17 -
C:Nule T: Cy Specimen 1-3 .15 -
Capsle T: Tensile Specimen 11 21 -
. re:sge
.19 .65 Plate B2002 2 WCAP 7323 (.14)* (.46)* ,
6.psule V: CySpecimen 2 44 .17 .46 i Capsule-V: Cy Specimen 2-44 .15 .41 Capsule V: Tensile Specimen 2 6 (.06)* (.27)*
Capsule V: Tensile Specirnen 2 7 (.08)* .42 Capsule Z: Cy Spas %en 2-33 .19 .47 l t
Ca,w ale Z: Cy Specimen 2 36 .17 .46 Capsule-Z: Cy Specimen 2 40 .20 .50 Capsule-Z: Tensile Specimen 2 5 .15 .52 Capsule T: Cy Specimen 2 2 .18 -
Capsuo T: Cy Specimen 2 3 .17 -
(.4psule T: Tensile Specimen 21 .13 -
1 Average .17 .46 IV.26
TABLE IV 11(Cont'd) 4
SUMMARY
OF CHEMISTRY VALUES FOR INDIAN POINT UNIT NO. 2 MATERIAIS Material Source of Data fdLW3 ELES Plata B2002-3 WCAP 7324 (.14)* (.57)*
Capsule Z: Cy Specimen 3-33 .30 .64 Capsule-Z: Cy Specimen 3 38 .27 .59 Capsule Z: Tensile Specimen 3 5 .23 .58 Capsule Y: Cy Speelmen 3-41 .21 -
Capsule Y: Cy Specimen 3-45 .22 -
Capsule Y: Tensile Specimen 3-6 ( 11)* -
Capsule Y: Tensile Specimen 3 7 (.10)* -
Capsule T: Cy Specimsa 3 2 .27 -
Capsule T: Cy Specimen 3-3 .23 -
Capsule T: Tensile Specimen 31 (.09)* -
Average .25 .60 ILh3 Capsule V:Cy Specimen H 16 .08 1.2 .
Capsule-V: Cy Specimen H 12 .06 1.2 Capsule Y: Cy Specimen H 21 .15 -
Capsule Y: Cy Specimen H 23 .20 -
Average .12 1.2 Eald Capsule-V: Cy Spechnen W 13 .23 1.02 Capsule V: Cy Specimen W.12 .20 1.06 Capsule V: Tenalle Specimen W 3 .20 (.69)*
j' Capsule V:Cy Tensile Specimen W 4 (.12)* 1.00 Capsule Y: Cy Specimen W 17 .19 -
Capsule Y: Cy Specimen W.19 .22 -
Capsule Y:Teasile Specimen W 5 .18 -
Capsule Y: Tensile Specimen W-6 .20 -
Average .20 1.03 I Correlation Monitor Capsule-V: Cy Specimen R 56 .20 .18 Capsule V: C,, Specimen R-52 .18 .27 Capsule Z: Cy Spaelmen R 33 .35 .28 Capsule Z: Cy Spedmen R 36 .31 .27 Capsule Z: Cy Specimen R-40 .21 .21 Capsule Y: Cy Spelmen R-60 .17 -
Capsule Y: Cy Specimen R 62 .19 -
4 Capsule T: Cy Specimen R 2 .25 -
Average .23 .24 3
- ' Values in parentheses discarded because of peessive deviation or were WCAP values.
Surveillance specimen WCAP values not used since chemical analyses were available.
! IV 27 1
i
(
TableIV 12 CHEMISTRY FACTORS FOR INDIAN POINT 2 MATERIAIS BASED ON REG. GUIDE 1.99, REV,2 Reg. Guide 1.99, Rev. 2 Material W% Cu 3(iN1 Chemlet_rv Factor (*F)
~
Plate B20021 .19 .65 151 Plate B2002 2 .17 .46 115 Plate B2002 3. .25 .60 176 Sutveillance HAZ .12 1.2 86 Surveillance Weld Mat. .20 1.03 226 Correlation Monitor .23 .24 130 1
i l
1 IV.23
i V. RESULTS OF ANALYSIS The analysis of data obtained from surveillance program specimens has the following goals:
f I' l i- (1) Esimate the period of time over which the properties of the vessel beltline materials [
will meet the fracture toughness requirements of Appendix G of 10CFR50. This requires a projection of the measured reduction in Cy upper shelf energy to the vessel wall using i knowledge of the energy and spatial distribution of the neutron flux and the dependence [
i of Cy upper shelf energy on the neutron fluence. -
(2) Develop heatup and cooldown curves to describe the operational limitations for selected i periods of time. This requires a predoction of the measured shift in RTNDT to the vessel
{
q wall using knowledge of the dependence of the shift in RTNDT on the neutron fluence l and the energy and spatial distribution of the neutron flux.
I
! The energy and spatial distribution of the neutron flux for Indian Point Unit No. 2 was calcu.
lated for Capsule V with a discrete ordinates transport by the Power Systems Division of i
Westinghouse Electric Corporation (12). Results from this analysis establish the means for the i
interpretation of surveillance capsule dosimetry and for the subsequent projection of neutron [
exposure results to the pressure vessel wall. Furthermore, the results of the evaluations are i
appropriate for absolute comparison with measurement.
A method for estimating the increase in RTNDT as a function of neutron fluence and chemistry l is given in Regulatory Guide 1.99, Revisions 1 and 2 @. However, the Guide also permits !
interpolation betwwn credible surveillance data and chemistry factors and extrapolation by extending the response curves parallel to the guide trend curves. These surveillance capsule results agree well with the Regulatory Guide 1.99, Revision 1, trend curves (see Figure V 1).
V1 [
_---e
Revision 1 information is provided for compadson with earlier capsule data to show the effect of the low Gux leakage core loading which produewi e 48.9% reduction in fast neutron Dux (2
> 1 MeV) for cydes 6 through 8 as compared to the flux for the $rst 5 cydes. Revision 2 results from Capsule V are also induded in this section.
The weld metal has now become the controlling matedal in place cf the U2002 3 plate as can be seen in Table V.1. A long. term projection of vessel RTNDT bas been mad. ' rom Cycle 8 and beyond using a low leakage core loading pattern which significarAly reduces the pressure vessel cuence rate from that produced by the Dasigh Dr. sic Cao (H) (see Table V.1).
A method for estimating the adjusted RTNDT and the reduction in Cy upper shelf energy as a function cf neutron fluence is also given in Regulatory Guide 1.99, Revision 1 @, The results from Capsules T, Y, Z, and V are compared in Figures V 1 and V 2 which are adapted from the Regulatory Guide 1.99, Revision 1, Figures 1 and 2. The shelf energy responses of the pressure vessel sutveillance materials from all four capsules are reasonably consistent and fah below the predictive trend curves of Regulatory Guide 1.99, Revision 1, for nominal weld chemistries of 0.20% Cu and 1.03% Ni. A long. term projection of the degradation in upper shelf energy has been made using the end of. life fluences is given in Table V 2. The projected 32 EFPY 1/4T fluence is less than that received by Capsule Z and the shelf energy decreases of the Capsule V vessel specimens were all lower than earlier capsule specimens as shown in Figure V 2. Extrapolation to 1.2 x 10 19 n/cm2for 32 EFPY predicta that all Indian Point Unit 2 matedals will be below upper limit values for either RTNDT or decrease in shelf energy.
Similar results are obtained using Revision 2 of Regulatory Guide 1.99 for Capsule V matedals in Figure V 3. Extrapolation to 32 EFPY fluence of 1.2 x 10 19 n/cm2on Figure V 3 gives predicted values of greater than 50 ft. Ib. for shelf energies for weld ruetal (controlling materials) as well as plate and HAZ materials.
V.2
h The current Indian Point Unit No. 2 reactor vessel surveillance program removal schedule conforms to ASTM E 185-79 (2) and is summarized in Table V 3. There are four capsules .
remaining in the vessel, of which three are standbys.
Table V 4 provides a comparison of End of Cycle 8 (EOC8) Suence values from transport calculations w. h Capsule V dosimetry analysis and a compadson of prqlected Suence rates with transport calculations for Cycle 9. These compadsons, comparisons calculated with experimental values, show excellent agreement. EOC8 values differ by only two percent and the Cuence rates for Cycle 9 differ by only about 10 percent.
t 1
]
The Qux derived from Capsule V,1.59E10tl.5E9 compared with the transport calculation .'or the j same case agrees within the measurement uncertainties as shown in Table V 4.
i l.
I a
j l
)
i
)
V3
600 . , p i ,
- l. l { q
- - - i q _ . .
7
,.j ,. _
g '
_ _ _ ._ _ ._ t. l l ~
i ~
400 -
'-1
, j ,
hh h
~ ~
' ~~
Reg. Guide 1.99 kI : -~
f ry
~
2
_ l Upper Limit ,
f --
E' E i : i. o
,r' --,,-
,i'
,p,g,,ii.'
. l V ' ',
u' '
_'(_
_ _ _ , _ , i i , ,, .
{
~
200 - -
[ _~ ',,, , ,
lb ,
!,i,'
_ f1------ ; ,,
, =
- jl;
.u -
,e g _ _
,l g , - - - s rg _ ._ _ _
,- y -
n _ _ -
i r -
_m.
- - ,i - - -
y g
100 -
3,
,,,,,u'" p* t -
n j '.~
$*9 p- h '
~ ~ ~
b 3 y g c,
, n -
Code:
. j ,i
,e '
Capsules T, Y, and Z Capsule V pat"b-
= '
.s
~
.f 'f
' 4 ;
60 a
._: /
g .
./
h 0 .*1 ate B2002-1 d
1i E i .E :.
E* A Plate B2002-2 @
0.25% Cu 0.15% Cu 1.
li 40 -
0.012% P 0.012% P O Plate 82002-3 g 4 Weld Metal @
U
" f.. e HAZ Material @
I Correlation Monitor @
'"1 _EE .. . y W
g
~ ~ ~ ~
~~
- : l \\.
20 1 E I h '
i 2 x 1017 1018 1019 6 x 1019 Neutron Fluence, n/cm2 (E > 1 MeV)
Figure V-1. Effect of acuaron fluence cc RTNDT shift, Indian Point Unit No.2 (Regulatory Guide 19), Redsion 1)
Table V.1 ADJUSTED RTNDT VALUES FOR INDIAN POLVT 2 ART Initial Fluence (*) ART Time Eterial !antion RTNDT (> 1 mV) DPA Rev.1 Nr. 2 (AdNsted Rev.1 RTRN)PTS (')
EOC8 B2002-3 OT 34*F (5E18 (SE tt 154 137 188 205
[8.6 EFPY1 (Plate) 3/4T 34 2.3E18 2.0E18 110 117 144 185 3/4T 34 (7E17 L1E18 50 77 84 145 HAZ OT 0*F 4.5Z18 (5E18 154 125 154 181 1/4T 0 2.3E18 2.9E18 110 107 110 161 3/4T 0 4.7E17 L1E18 50 71 50 107 Welda) OT 0*F 15E18 (5Z18 172 176 172 232 1/4T C 2.3E18 2.9E18 123 150 123 206 3/4T 0 4.7E17 L1E18 56 99 56 149 15 EFPY B2002 3 OT 34*F 6.5E18 6.5E18 185 155 219 223 (PMe) 1/4T 34 14E18 (2E18 134 134 168 202 3/4T 34 6.9E17 1.6E18 60 92 94 160 HAZ OT 0* F 6.5E18 &5E18 185 142 185 108 1/4T 0 3.4E18 (2E18 134 122 134 178 3/47 0 &DE17 1.6E18 60 82 60 123 WeldO) OT 0* F &5E18 6.5E18 207 109 207 255 1/4T 0 14E18 4.2E18 150 172 150 228 3/4T 0 &9 Elf LSE18 67 117 67 173 20 EFPY B2002 3 OT 34'T 8.2E18 8.2E18 208 166 242 234 (Plate) 1/4T 34 4.3E18 5.3E18 151 145 185 213 3/47 34 8.5E17 2.0E18 67 102 101 170 HAZ OT 0* F 82E18 8.2E18 208 152 208 268 1/4T 0 (3E18 5.3E18 151 133 151 183 0 8SE17 2.0E18 67 94 67 141 3/4T WeldO) OT O'T 8.2E18 8.2E18 232 213 232 263 1/4T 0 (3E18 5.3E18 168 186 168 242 0 8.5 Elf 2.0E18 75 131 75 187e.
3/4T 32 EFPY B2002 3 OT 34'T 1.2E19 L2E19 253 185 287 253 250 (Plate) 34 6.4E18 7.8E18 153 164 217 232 224 1/4T 34 1.3E18 10E18 82 118 116 186 174 3/4T HAE OT 0* F L2E19 1.2E19 253 163 253 225 1/4T 0 6. 4E18 7.8E18 183 150 183 206 0 1.3E18 3.0E18 82 109 82 164 3/4T Welda) OT O'T L2E19 L2E19 282 237 282 203 212 1/4T 0 6. 4E18 7.BE18 204 210 204 266 186 0 L3E18 10E18 91 151 91 207 138 3/4T (a) The fluence value8 shown at 1/4T and 3/4T beations are based on attenuation factors for l MeV) and used for Rev.1 calculations only. The cctual udT and 3/4T /7uence used p(>Rev.
sn 2 results were based on DPA attenuations, consesvatively estimated to be 0.65 and 0.25 respectively (see Table V 2). Thus based on this approach the fluence at 1/4T and 3/4f locations is equal to the 0-T fluence multiplied by DPA attentation factors.
(b) Composition of weld No. 9-042 assumed to correspond to the surveillance data 0.20 percent Cu and 1.03 percent NL, for Rev. 2 analysis.
(c) RTPTS values based on PTS rule in 10CFR50.61 V5
i l
Table V 1(Cont'd)
RELATIVE RADIAL VARIATION OF DISPLACEMENT PER ATOM (DPA) AND I FLUX (E > 1 MeV) ATTENUATION WITHIN RPV AT LOCATION OF MAXIMUM INCIDENT FLUX l Reladve Reladve ;
i Radius Fluz DPA [
,,,[gg},,, Attannaden Attanuadon I ;
220.27(1) 1.00 1.00 l 220.64 0.977 0.943 l 221.66 0.885 0.915 222.99 0.756 0.820 224.31 0.637 0.730 0.534 0.647 225.63 225.75f * ) 0.526 0.640 j 226.95 0.443 0.573
- 228.28 0.367 0.507 ;
229.60 0.303 0.449 i 230.92 0.250 0.397 232.25 0.206 0.349 l
- 233.57 0.169 0.307 [
l 234.89 0.138 0.269 8 1 0.113 0.233 -
236.22 236.70(b ) 0.105 0.221 237.54 0.0912 0.201 r
}' 238.86 0.0736 0.170 i l
240.19 0.0584 0.141 !
1 241.51 0.0454 0.113 (
i 242.17(2) 0.0422 0.106 1
3 L
! I
' NOTES: (1) Base Metal!nner Radius !
Base Metal Outer Radius i
- (2)
(a) 1/4TImadon j (b) 3/4T Locadca ~
1 i
l h
- L 3
I f
! I 1
l 1
) V6 i
}
60 MM I '
[. f $ h
~
! I
+ .. O[
s .
l 3 h.(;h
,,,, ,,,,,. d. ,
p , -
~~~ ~ *"
d!
~ ~ ~ -
~~ "
l ! p, -"' ",1h., ii' {i; 'h 40 + to w ~i a- l ld!-- - -$ ---
of "
s-i -",,- !
,[ y 7--- ,,,
q
';yj 7
",,...' li h : l .- .: , -
W k'l ;
Reg. Guide 1.99 - 1 3,,,- - i.
I y ",,,,
,"j 7 j ,
Upper 1.1mit
T T ,, d' g
,"t '
, ,, m '"" ,
g l "' " i j 'I l j_ -
.i -%j '-
E
)) ;
? 3 4'
g
",,,,,, o""i
,4
~'
- l '
j l q i : -
,,,- - o l" :
9D
- 20 - -
@j 0.25% Cu 97 ;, ,,,,{
g hl!
l l
j l ,
], ! d ~
I " ,
Z7
~ ~'
t i 0.15% Cu - -
1 l m 3 q;l .
m i _-
0 3 'ifnq -- - - -
l ! il i- h 10 - -
, 1,[- -t j
r ,
< r i I!ll q'
-3 t
"4 ilij 3 ~ ~ '
=-
~T~ I l
L " '
f '.
lft t~ " h "M - -
l jl m li , L ~T '
c : r :( .
~' Capsule V
[ h" HI
'~
+
j . _
~ ~
ll Code:
6 - " -
i ,
~
- : O Plate B2002-1 f
~
$ .., [ l 2
1 ,,,, ; a Plate B2002-2 @
- F.
8 .
l
= -
t O Plate B2002-3 f
i 4 - '
l d 1
+ Weld Metal @
h e IIAZ Material
,,L f 5 7 llH @
. 'j
~ ' '
,,j.. r
- i 1; _ _
1 1_
, Ul 2
i Wi d., ,l - 11 I l o m. L h I ! pq W 2 x 1017 1018 1019 6x10I9 Neutron Iluence, n/cm2 (E > 1 MeV) l I%ure V-2. Dependcace of C, shelf energy on neutron fluence, Indian I*oint Unit No.2 (Regulatory Gokle I.W. Redsion 1)
i Table V 2 COMPARISON OF MEASURED AND CALCULATED RTNDT VALUES FOR INDIAN POINT 2 CAPSULE V MATERIAIA he G==tA 199 ! Material Measured (*) Rev.It ') Rev. 2 Rev.2 + Margin j i Plate B2002 2 78 90 90 124 l Weld 204 196 189 245 t IIAZ 166 170 136 191 ! 72 (c) 106(e) Cornlation Moalter 90 90 102 136 ; i (*) 30 Ft Lbs or 46 Ft Lbs Value, as appropriata, Figures IV.2,3,4, and 5; Table IV 9 !
- 0) Obtalaed kom Fig. V 1 at 5.59 x 1018 n/cm2for weld and HAZ or 4.57 x 1018 n/cm2 go, Plate 2002 2 and Correlation Monitor, (c) Based on Weld and Base Plate Correlation, respectively V8
~-m , ..
67.** +* 5 - - : ;T ;
'N.
i.
. !. Da OO
- -- =: ,g; C w l r
= = -. Z,m 0 o .
I-T*1 ?" . o :
- 5 ******* - N c
- . m e- - m ::: r .s c- -
p:r-p:::,:::- ".tri t p- 3 O T t --; . @ T;32 : .+..
' ::.;;2 :-
4.:=w-t.. +-4 ===.
. 3:=: . t* . 'ti. - 0 N t E
. - .. . C. t 0
** g
- tm N
- r T.
(:t: g. -- + ii..: ' . * .* '. *' *
- g. Q w C ,
- y o
- =. ;
. :K [. .
.a.T.;'*.;* ! 1. N w 9
0 ** h 1,..
.. ... .g ..'**T.1' e
. o go q % 4, g Hpg .. .. .
....% E at ., ,o L.} .:. =ow =
2 . -m. - o
,d q. c.. . 1 -
. w
=
- 5. a t- -
m
;. .- : . .t- =- o- e o o
- 5. N
,s ,_
# 3, - ; ,:,r .- . ,
...- .. .,. a = v
. g.
1
=.p .:.. p., ; ,.
. 8. v a. .
,., g.
.. c,
- o. . r. .tr-9e '!
<) 4 e g n 8s a.
- w. f. m LE uk1 1
-. i i . ~
.nt
.. m u
.riq ji T .d,i i tilo - 1 i !
I
~"
_ * '"I "'
"o 8."
W 1.@bE Rd'M'- k. @N d7*'J fl 'W I i,h - " ' *g* _+i p u g y ; F.T. f.p.g..w
. a .,= . ._"-t. _A. .. .a:ura:
t.
,. y,5
- p. u_ p. . _r.j gg. n_Em :.t. p. .p+_ ._h* .
= p-w.r-
.c_.q e a
- . =g : ..,: .
,o
., u,,
d=
=r = n W:u :.d d.e . %
r : ....
- a. 91 , . :. .. .. _. ,. e _ ,:::ya . .
- 3 ,,
C
- t. 4.c::m:dyy;+---***:=
;;;i :
m F-+:_T. -
- .-c:: r.< :,,. efh::.Ldn:-::d:*r:g
'r =: : ::.
rt *--
- .:=
g a .O .
. . t . 7i
.. wm ,s : .. _ . .
a
=
g
..... .. .. . . . a. .
. p._.. . _b_,\.=
. . .. ~.
*1. .
.. .. . w
*n GW R@c s. M--WWr. i. W n.L&v==hueh
~
N.
. . . . . . ru..... R I. r.e.. M. d .d.,r.-
. .. . _- ..4.:.
= ._ f .= . -". _ _- . . ..
m_
.mtr.. .L .
..nr-e.
.EL.= . = . .
.==.t_ :_:.
.c a.
hos
; ., a
_r d
.- 3
[.g. .- pl.tdg m=.- n - e.
.d:.:=. :.= c: :t. .E .:m =:=d ==
t r..c
-..o - . --
w
.# e u a
.. .s
..1
_. _; .1. . . _ ..
...4_
z ,b
- a.. ._
_ m n .s ,- n w
. a.
.., s .a :
... , .. .w
+ u-q.y.g. %-.
p_ ..w.. 4.. %
.. , .J., t.
m, - g u. g 4
} .. .
. ..p ..i4 6...
... %. . g y@
Ar . .?_ "* g
.p. . . .. \. } ,
.}, +1 '
i _jl.q 4
.y@
- s. ,i... s .
4 .r .. .,
.. s
. . -w
- i. g
.. . .. . . .. . . ... .,3 6. . . .. , .+ . %g.
, k. .- . . , .
i.
,i.
i. j.))iil J+- _
..,i. q i.
. ., I.
. j- i :i i i
.q3. ,... ...i t.9 ,
10 o m "p g5
-W N m i 1 j l 4k.t* W. u tr e it i.N +. c $-
w W W W Grq TFW l 3 @@ @ $ e $ W.-.: ,7. ;. - *- !E gJ i c:: .-j p g- W ccc d d 6 , M (1)1 T 7. .aQ3 $* I a:. .' w:: A
- . A" .,;;.
; n
.~. ;
...:-...A.,.
1;it :: I .: F. -
'. .=.1 7 y
@ P
= :.. .::: o :4:3.:.=: T.,*O*. ft. l. : *::.j n:- - t: d
**4.q="2".
;.*.*-- ,: .* *. U
- J TC N. *":J ;*: k
. . g W g*C. p
. . . . . -- ~ ". .
Om O'-' a( W N e v= *.* .*..*.
-t 1, . . -
.P.I..,..,
@2OQd d d d . , .;,; ,:
; . 2.
r;t ~f M
.t: .m. 2. . ;
; = -t" *t ' - * *
, , . . ,* n:-.
T T :.[
': ';; ' F* *. 9.
_1 . . . .
~~-6..... " * - - - - , . _ .*4 - -~ . _. .:'.: *.4 T* !. L~m.,.* ".' T.-. .
.. *"1. . :;.1__ .
.._.".":.-. b . .- . . . . -
. *: ::: L. .1 :; *
. . L .*
..;.4..l-'".,..*-'.'
- g. _F.:.7_3 : . t ".".T. . .=_
. ... .. T.. ,< ,,.;d' t. ,;f..-*,, ,n.' ., , ;,
- r. 4'n; l . .. q.... %._..
. .1 ,.p.....
o
.4.... .
..t.. 6. i. a_
e o o o o o co e er m N X ! Q@ T M N
- N 3 I
tussaad AS4eu3 pays ut steeJoa0 i i V.9 l r
i Table V 3 i REACTOR VESSEL SURVEILLANCE CAPSULE REMOVAL SCHEDULE (21) INDIAN POINT UNIT NO,2 L Cansula Ident. WOL Removej Equivalent Vessel , & Qdt klatsial Time Fluence - 1 T Three Plates 1.08 EFPY8) 3.4 EFPY at LD. 2 Y Weld & B2002 3 2.34 EFPY 6) 11 EFPY at 1.D. f 3 Z Three Plates 5.17 EFPpe) 29 EFPY at I.D. 4 V Weld & B2002 2 8.6 EFPdd) 8.92 EFPY at !.D. i (*) Removed after core cycle 1. (b) Removed aAer core cycle 3. f (*) Removed after core cycle 5. (d) Removed after core cycle 8. f r I I Note: Fihh capsule is scheduled for removal at the. end of Cycle 16. { t The remalr.ing capsules within the reactor vessel are:
- Q.lg WOL M2tedal S Weld & B20031 ,
l U Three Plates t W Three Plates i X Three Plates l J l I j L i i V 10
i Table V4 COMPARISON OF END OF CYCLE 8 FLUENCE VALUES FROM TRANSPORT . CALCULATIONS AND CAPSULE V DOSIMETRY ANALYSIS Transport Dosimetry Locadon Calculg) (n/cm don Resultg) (n/cm C/E* ; 4' S. C. 5.19E18 5.30E18 0.98 i 40'S.C. 1.48E19 1.51E18 0.98 RPVO T 4.35E18 4.45E18 0.98 COMPARISON OF PROJECTED FLUENCE RATES WITH TRANSPORT CALCULATIONS YOR CYCLE 9 Transport Dosimetry" Calegtion Resultg Location (n/cm sec) (n/cm see ) C/E' ; 4' S. C. 1.75E10 1.57E10 1.11 40'S.C. 3.77E10 3.42E10 1.10 , . RPVO T 1.13E10 1.03E10 1.10 l l I 4 *C/E is calculated / experimental. i j ,
"Capsule V values und as the ' projected
- dosimetry results.
i f l i ! 1 . t l I
, I 4
l I l
! v.11 .
l l ~~ .. , _ l
VL HEATUP AND COOLDOWN LIMTP CURVES FOR NORMAL OPERATION , OF INDIAN POINT UNIT NO. 2 i Indian Point Unit No. 2 is a 2758 Mwt pressudzed water reactor operated by Consolidated !
- Edisor. Company. The unit has been provided with a reactor vessel matedal surveillance program as required by 10CFR50, Appendix H.
The fourth suneillance capsule (Capsule V) was removed dudng the 1987 refueling outage. This capsule was tested by Southwest Research Institute, the results being described in the ; earlier sections of this report. In summary, these resulta show a marked decrease in fluence as , compared to three capsules (Capsules T, Y, and Z) but now indicate that the weld matedal will l I , control tha value of RTNDT over the plant design lifetime. I i The adjusted RTNDT (Regulatory Guide 1.99, Rev. 2, May 1088) after 32 effective full power years (EFPY) of operation is predicted to be 266*F at the 1/4T and 207'F at the 3/4T vessel } { walllocations, as controlled by weld matedal. Tae Unit No. 2 heatup aid cooldown limit cunes for up to 32 EFPY of operation have been computed on the basis of the above values of l adjusted RTNDT using Code procedures Q) and the following pressure vessel constants: i A ! VaselInr.er Radius, rj = 86.50 in. l 1 Vessel Outer Radius, ro a 95.23 in. l Operating Pressure, Po = 2235 psig i Initial Temperature, To = 70*F
- Final Temperature, Tt - 550*F
! r Effective Coolant Flow Rate, Q = 6 136.3 x 10 lbm/hr i Effective Flow Area, A = 26.719 ft2 , i Effective Hydraul!c Diameter, D = 15.051 in. l VI 1
Heatup curves were computed for heatup rates of 2&F/hr,4rF/hr,6FF/hr and 10FF/hr.
, The Unit No. 2 heatup, cooldown, and leak f ast curves for up to 32 EFPY are given in Figu2w VI 1, VI 2, and VI 3.
1 I, 4 h } { l i I l 1 j I h. 4 l I l i I t l, i ! l i' i I,
~
i
\ \
r l { YI+2 , t i t
2600 NI i i i l I i l 1 l } l -
,. 3; a r :il e m .
.t:- i: pmt n. . ,
i 2400 L- A 1., e 1 MATERIAL PROPERTY BASIS J 4 2 4ik L o '
; F A J u :!if - m
, g. ... a.. : .a -
2200 - Weld Metal Cu - 0.20% - - - -- T - r- 9 -~9 +--
+--
i
++
- t- .jp .j., p' 1 , t. - "
.n p.
Initial RT , 0*F ttrp- :- -i 4: 1? 1 . . .
'r :1 .
? a .n S M di @
~ M[
~ h n . ni- Hj. j di.
2000
- u -=
-p= ,i
;h.-ll .:,
At 32 Effective Full N-- -
~. 4-+-
4 u ..
+#' 'r .'.t !n.Hi. k.! .HI
+ .+++= - + -
U. . ,' a.' ,4 4'
& .L t.. .
E. Power Years: ' , ..i. IUUU "
' e,
~5'
*' i? ;F 1, .
0 W : ,! .i
% RT ,at 1/4T - 266*F
~ ~ ' ~~ ~ ~~ ^ ~ ~ " ~
T j"' '~ y" ~ a RTu at 3/4T 207*F i ~M 'if y ]'4 yy{
~ 1600 - - - - - -- -
-- =
s y y{ lij 'i[ [ yf n i - l- 7 1
- s ! i i' 1 I
+
.d n
- f -iH ;h. '
b :ii {U u 1400 - - - . , - --- --- --- -- -- -- l' n' A pl ;
.l ; .jp a ; i
,L v ~+- r f ,: m oy p
m g.l .4. . . l .'
- o. 1200 !
i if I!l p
i 1 ;- ' a'!- p
- j -- -- j-- '
-l ; - - - - - - -
ri a j ' Ibrgins for Instrument Error: " a 1000 # !;
-30 psig and +10 Deg F
/
-7,nht Oc"~ t n (.
~ ~-
u ) r w
+
c 800
. f I
l l ( -/
/
r -
. , .... , ~.tl.
- p
.. c: --
c 's l l ~ . .r
" ..ni . ,y v. i .; Tigi-~ '.
Il
/ //r -
' 1 600 rm l
j b ' I l r-id IP
'jk "
i Ti
+
400
/
Criticality Limit
/"'
4o 9
. 4 --
" <I ._h/ i
+a -
.ij! !!,.
l lt .i
!!H th,, .n,,t
, , , I i ; j ..
IIcatup Rates up ..t
.i.
m
.r M .n : !,
200 to 100 Deg F/hr
, 4. . . _ . ..u ._- . .i h.. ..
- d. .. .
) , , i -
- !. 74 .,.. 'l.. J . I.; g :r .g: :j
.. :.l. .g :
.lt g' j l:. ,!
+ :h.
O -
- * ' l ! 7 , ';nih hi 100 150 200 250 300 350 400 ,$0 Indicated Temperatare, Deg F Figurc VI.I. Indian Point Unit No. 2 Reactos Coolant IIcatup Umitations Applicabic for Periods Up to 32 Eficoive Full Pimer Years (With Critica'ity Umit)
2600 '
. I i ! l i l1 '
l l ,
; ! i 2400 l MATERIAL. r@PERTY BASIS l '
l i i
- l. l .
! Weld Metal Cu -
0.207. ' 2200 ! l
- laitial RT , - 0*F i ,
2000 i At 32 Effective Full - l -- - - - - - d- -- -- Power Years: 3 2800 266*F g , j RT ,at 1/4T RT , at 3/4T
- 207*F l 4
+
l
.[ . .. . .,.. ' '
l ,
- 1600 i I,_
o I i i -. , ,
; I I
- f a w 6 ! i p i <
{ d o 1400 ' t i .- j . I ; i. 0 . _ _ _.
$ i '
i L 8 3
! } ! , i ' ,i
- p. 1200 i '
; I' 3 I nirgins for Inst.rument Error: , ,
i 3p
-30 psig and +10 Deg F I I i l l 1000 -
', l o l ' , ,
j i *
, ; } l .g_ _ ._ .. _
O I - I L i ve ' 1 l 1
?
! ' i I e 800 I i I '
j Cooldown Rates, i e i i 600 9 ! '
, , l .
l _ i t - ' 400 20- i I
> 4
; , g } j I- -- - -- -
200 60: ._ i i ! - l ! i ; , _p ;
- l. f t ..; .] ._ ._ . . . .. ..
100F ! I i i
! 4 , i i l, ,
'=
0 ' a I - ' 100 150 200 250 300 350 400 450 Indicated Temperature, Deg F l Tigure VI-2. Indian Point Unit No. 2 Coolant Cool &mn 1. imitations AM for Perio6 Up to I 32 Effective Full Power Years
2600 , i i I ! l # 1 l l l l l .I,p .-r. - 1 1 1, -
'h 2400 i MATERIAL PROPERTY BASIS ! 1:
't
.I.
' i .t. . _ . _ _ . .
.. t .. .. .i.
p i Leak Test Limit l Weld Metal Cu - 0.20% i
- f .
r' F J ~ 2200 ?. Initial RT.
~
- 0*F
~ ~
. I Y'
~
-l:l,.. r. -
.. m
- ps " . ~ . .Y 2000 .
At 32 Effective full
.t.. .
i, .,
..i. ,,
e n m.h.'. . . . a. ..a 3-
+ LI- 2 4L C -. -
Power Years: ' ,, a.
.. , i a
e ' n' m i i
. p >
D' l
- i. ,
w 1800 , I i RTe at 1/4T - 266*f ! i. _ . ' _ ., ' . __'.i .ri L 'F k n ' 4 RT. at 3/4T - 207*f , -
+-- .-
O. l , h 1 j { L.* . 1600 -L - - -I. $E L ?- o ' i
- 1. +-
u l ',.' h! l
- s i , ap .i.. ...
o 1400 I ' , b' l, U 7tp - - - -7 m ; j
- - .- - - - -- - - ---.+-
o I , w i , t m - ~
- c. 1200 i j t
+- - - --
- - - - - - - ~ - - ---
, H p . ;- J.
+ ~~
fa .o I 4, o -
; . ..p .' , ',
j tj 1000 - p
~j-tiargins for Instrument Error: 7" 7
~
' ~[j-~ ~ ~ ~ ~ ~
' ~ ' ~
T ~ ~~
" 7 *
- u -30 psig and +10 Deg F - " " ' ' -
e o 800 ' -- c j , j
+
s , q.
- l .
e 7,j.-- .g ..7. .... 600 l j. t j-J. -' 400 l l i E ' -
- - r g -T .
-- -a-i q.- - }.-. - . -- .L.
4' ' - w i .I---.
.- .C '
l ; ,i; 300 ' I 8 i i ' I l . 200 . i. . Ileatup Rates up i iT O iif, "" 7 $ " j 9~ 7.. }. . '.'" y Q
~ ~'~
4 , .,
. to 100 Deg F/hr...........,f 7 a.
........7.. .,
0 --
' . . .'{ l l l [ l l l U ! ' p' .ti P'h 100 150 200 250 300 350 400 450 Indicated Temperat.ure, Deg F ligure VI-3. Indian Point Unit No. 2 Reactor Coolant IIcatup Limitations Applicable for Perkxis Ilp to 32 Effective f~ut! Power Years (With Leak Test Ilmit)
VII. REFERENCES l
- 1. Title 10, Code of Federal Regulations, Part 50, 'Ueensing of Production and Utilization Facilities.'
- 2. ASME Boller and Pressure Vessel Code, Section III, ' Nuclear Power Plant Components.'
- 3. ASTM E 208 81, ' Standard Method for Conducting Drop Weight Teet to Determine N1.Duetility Transition Temperature of Ferd:ic dteels,' 1982 Annual Book of ASTM Standards.
- 4. Steele, L E., and Serpan, C. Z., Jr., ' Analysis of Reactor Vessel Radiation E5ects Suneil.
lance Programs,' ASTM STP 481, December 1970 1
- 5. Steele, L E., ' Neutron Irradiation Embdttlement of Reactor Pressure Vessel Steels,' i International Atomic Energy Agency, Technical Reports Sedes No.163,1975.
f
- 6. ASME Boiler and Pressure Vessel Code, Section XI, ' Rules for Insenice Inspection of Nuclear Power Plant Components,' 1974 Edition.
- 7. Randall, P. N., 'NRC Perspective of Safety and Ueensing Issues Regsrding Reactor Veuel l
Steel Embdttlement . Cdteda for Trend Curve Development,' presented at the American Nuclear Society Annual:.!eeting, Detroit, Michigan, June 14,1983. ( i , S. Regulatory Guide 1.99, Revision 1, Omce of Standards Development, U.S. Nuc%r Regula. i tory Commission, Apdl 1977 and Regulatory Guide 1.99, Revision 2. May 1953, i l r VII.1 i
- 9. ASTM E 185 79, "Standard Recomn.enied Pantice for Surveillance Tests for Nuclear Reactor Vessels,' 1981 Annual Book of ASTM Standards.
- 10. ' Indian Point Unit No. 2 Reactor Vessel Radiation Surveillance Program,' WCAP 7323, May 1969.
- 11. ASTM E 399 81, ' Standard Method of Test for Plane-Strain Fracture Toughness of Metallie Materials,' 1932 Annual Book of ASTM Standards.
- 12. ASTM E 813-81,
- Standard Test Method for Jge, a Measure of Fracture Toughness,' 1982 Annual Book of ASTM Standards,
- 13. Norris, E. B., 'Resetor Vessel Matenal Surveillance Program for Indian Point Unit No. 2; Ane]ysis of Capsule Y,' SwRI Report 02-5212, November 16,1980.
- 14. Norris, E. B., ' Reactor Vessel Material Surveillance Program for Indian Point Unit No. 2; i Analysis of Capsule T,* Final Report 02 4531, June 30,1977, and Supplement to Floal
! Report 02 4531 December 19SO.
- 15. Norris, E. B., ' Reactor Vessel Material Surveillance Program for Indian Point Unit No. 2; Analysis of Capsule Z,' SwRI Report 06 7379, April 19S4.
i
- 16. Anderson, S. L, ' Analysis of Neutron Flux Levels and Surveillance Capsule Lead Factors i
for the Indian Point Unit No. 2 Resetor Using a lau Leakage Core,' July 1952. i VII2
- 17. Anderson, S. L., ' Plant Speci8c Fut Neutron Exposure Evaluation of the Indian Point Unit 2 Reactor Pressure Vesed and Surveillance Capeules Fuel Cycles 1 through 9,*
Westinghouse Electric Corp., Power Systems Division, July 1988,
- 18. U.S. NRC Standard Review Plan, NUREG.0400, Section 5.3.2, Pressure Temperature Ilmits, Revision 1, July 1981.
- 19. ASTM E 322, "Standard Method for Spectrochemical Analysts of Low Alloy 8 teens and Cast Irons Using an X. ray Fluorescence Spectrometer.'
- 20. latter dated March 29, 1978, from W. J. Cahill. Jr. (Consolidated Edison) to R. W. Reid (NRC),
- Indian Point Unit No. 2 Reactor Vel Material Surveillance Program.'
- 21. Indian Point Unit No. 2 Technical Specifications.
V!!.3
i APPENDIX A TENSILE TEST DATA RECORDS M ^ _=gi etSpedmeas After Testing f;:'-- _- W3 W-4 24 27
.s
's
l l \ W T,.IL I 2 < 5 ,' F 5 8 0'F R RT Photograph of tensile specimens after testing l _-.,nr
Southwest Research Institute Department of Materials Sciences TENSILE TEST OATA SHEET Scecimen No. d.3 Project No. / 7c.2/0# Test Temperature ,W Mschine Ident. Y Strain Rate .m4'Y.rx)/.uk Date of Test t// r /AG Initial Diameter .2.97 Final Diameter .. /#3 Initial Area .p7W Final Ares . 0 / cLa Initial Gage Lengtn / cs Final Gage Lengtn /,;po Specimen Temperature: Maximum Lead 37,2o Too T.C. / 0.2". Offset Loac Wyo Middle T.C. 7l 7pp Fracture Load J V/, o BottemT.C.],s / Eleng. to Max. Lead ,u 71 Idd k h O*k % U.T.S. = Maximum Lead / Initial Area =
/S4JgA ,
J.2". Y.S. = 0.2" Offset load / Initial Area = 9 J,7/ 2 Feature Stress = Fracture Load / Final Area = / 99,## % R. A. = 100 (Init. Area-Final Area)/Init. Area = //, 4 ) i Total Elong. = 100 (Final G.L.-Init. G.L.)/Init. 3.L. = _2 J 0 ,_, i Uniform Eicng. = 100 (Eleng. to Max. Load)/Init. 3.L. = J 0 <r 4 Test Ferfor ed Oy:, N W A/4/ /# Amr# Calculatiens Perfor ed ey: M -, 21/,, s b ;0atei g//p/95 Calculatiens Checked ey: / b c<$ 4 OM Adate) 6 S//3Y
M:I .V L Y,1 . ? '"
' 'J. *1
- 04) N 09 #%
k7 __ r L' - p
.f .l
. f w .
wj
- ~ . ~ - - -
. .m.-.
_-+ ,
-.[.--
3 f 6 F.. . i ' _ 4- - W-..4 - i l
._-J l
1 I b, -,. - ~ _.. a
.. . - . . ,-, . - . . 5- ..-4 s
.. I 4. -
_ . . . ~ . - . - . . - - . . . , . . _ . ~
- M i
. , , . . a- - . . - ~ . , , . . .. .-
s > , I
-_ . . . . . .~ . . . . _. -
r p _ ._ -4 - .-.- .-. E- , -- 9 6 1 1
.y. . .- -w .-1.-.4.. . . - . . . . - . - . _ _ . - - - . .. .. . _
L t 5 4 8 g n .
, . -~_.e--.-~ ,
.3 1 .
_ __a
+ . . . ,_ . r ,
. h,, f...
,s ,- + . . - . .,.- ..
t
-.4. . 7 - . . .-
s h 7 __L
?
{ J
.n : - l _2 g _ v .. _ ._..
I i
._ rf. . J' - . . I-- . - . .
f j
-._.$_. .j _. *
.-,.-__.,.y'.
j* r i . t , , t I , i . ! t i L t i t ! -I. .4
~. . . .- ' - .- -- - . !
1 g ,
)
g ,
-.,- . . d J 4 - - . - -~ .. .---4-..4 ...--_4--.! % >
h , 1 L } S' d } 6 s _ ,.. . . _,'-,.~....-7 .
.-_---,._+.-n i h
- . - - - . . . . r--.
*/f* '
; i l 4 .,} l e 4
9 , e ? t i t . . i ! w~-4-_'- _. .I . . . . !. y.-,_.4-..
- w . , .
....._3
,D > (* 1 t % ! t.
a ,- , N
'4 .
l [ f j
. , - . . .. _ ... , _ . , _ .m
-. p --.._i g .-
_...9 - - - - - - ._ mQ
- e h h 1
_.4. 7 - _ i4. _. i a._ -t .g _ u __ q, . . _q 9_. r-- 4 7
- . _1 '
_ ___ _ m f I
. . - - .l ~ [
'f.
' _f--f 6
, l Y I i' e
.L_ -
- k. l
_-_L._ .. . , .~ .. -_.. . _ . .. .. . .. , . . . . .
-- . . _ ~ . ,
I.-.,g s - - - . . . . .
. ., . . - I.
..e--.-ea.e -._e-e4. - - . ...we - , _ . .- - e- -. - . . . .
1 . . . . - - f _ _ _ _ - - _ . _ _ ._ - ._.b. ! t_-.. . _ . . , I n ( _ es=-r- mr .e - _.-s. . - . - .ee . . . . - . = - . . . - - . -, ., ._.. l l
. _ . . . . . . .. - m..~~._-- , - - . -._ .. _. . , , , _
m e- .
-*g.- 61.M-.w*..-t I
I l . . _ . . 1 1 d _- . ._ 6 l % ;,- 9 i 3 , e s l
- s-N- >
3 i l '
# N- Ds s I
k J*., '% e e
'eM M -
. . a
.*b **
e, m o O
a ..g g g ., Pe]
,,,,,g , ,g,q ,,gg, ,
g y,43
,, . ,$(),,)
b y. '%& _ _ . - - - - - . . . . - - = ,
.m....-
r 1 7-. . .. _ . . . --._'. < s l-
. .. . .. ~ . - . . ~ . . ~
i i .
. -.-.__q__...,.__ . . _ ._.i I l l l [' l
... - .-. . . .. - .. . _ _ _ ;w - _ -- .--._.u .
e --- r ,-6 . .. .- ' ' , w. --. . . ,. _-- .- .e i i
- a I 9 f
L
~A~~ Y I
'4-+-. .- $
. -. . _ _ _ . . . _ . . . . ____ _ _ . _ _ _ _ . _ _ - _ . . . -.. . vc. <
e 4*=MN-* '. 6 28.'.'M a _ w.. ' MWN suh6e$ .e -W 9 . d
=.4.e. ..hr e s ,4 * - .
seere .m.. m ., og %- . . . . . ,,..,.e- _. .e M464 6 $ M N b' -M8 Me 4hgu m6M4W me e.m -e* wm mw e.m..hp- - M .
- e. s .. .. m 6
_ .. - . .._....a fe. I y . #-. . . . , . l .-- ta -y q . ,e . .... . 4 _q 8
.f .-4 ,-, . . *.. . se - +4=..
_ . - , . - , , _ . , ,~ . . - e- 4 es.s4 * -.- 1 4
. - . . . ..-. .ee -- .svg-e.,ww-... .%-, ... . .. . . em . . - we 4 %
. . ,.. - . -= .,_ -
l
.~.
S -emy+$t. I eem.~.e e -- new.i.. ..e me.g. .as.o.ne- ..g.e. er e - .-= S hh j ' PD. y
., ~.1 i
- y ,. . .. . e, T..._
1 i
% ._ _ .. ~- _ - .. - . - , _ .-, . __m 4 . . .
1
.e.= . +.-
-4. = _ -.t e_* . m . e a
.* 4
_ . _ _ . . . . . . . . , , . . _ m . .. - O p ._.___ . . _ . ._.- . _. _ _ ._ ___.._ _ _. __ i p MJm.m
" ** .'. ' * ** ** 5 N
- O" yk.-. .. . - .. .. .
9 i I e f i l u . : i i 4 (
- 1* \ s %
.D hO 4 g I
o . *: N N t *
# 4
% * '. j O
a I k
'e
/.* .Q-w
*. y
Southwest Research Institute Department of Materials Sciences TENS!LE TEST CATA SHEET . Specimen No. WY Project No. / 7-J/o 7 Test Temperature fS'd Y Machine Ident. Y Strain Rate .evs*NNAwia Cate of Test 4//4/rs Initial Diameter ,39% Final Diar.eter _ / _f9 Initial Area . o v7 ar ~ Final Area _ , c e n W. Initial Gage Lengtn /. o Final Gage'Lengtn /, u 7 Specimen Temperature: Maximum Load e/ 7e c Top T.C. s- 3 3 *r 0.2% offset Loac 31.4o Middle T.C. W/4 Fractere Lead _ ? W. 0 4ttom T.C. fr3 -7 Elong, to Max. Load .nfus U.T.S. = Maximum Load / Initial Area = /OrJr0 0.2*. Y.S. = 0.2". Offset Load / Initial Area = #2 $~/B Feature Stress = Fracture Load / Final Area = / 7 9' M J t R. A. = 100 (Init. Area-Final Area)/Init. Area = f f 32 i Total Elong. = 100 (Final G.L..! nit. G.L.)/Init. G.L.
- 4 0.76 i L'niform Eleng. = 100 (Elong, to Max. Lead)/Init. 3.L. = / 'T &
Test Perfor.ed by: THA G A fs ) /2 A rn*ff' Calculations Perfor ee by: d dd M --40ata) 44f.r Calculations Checked tytt << I /d< W 8/(Data), 6/7//s
- wee .
.*s
...a ..a t( ec . c. m. .. - .t. . 4 7 ,. u. 3
* +
/
%,:t
_ 43 4 .-
) -; _ _ . . . _. _ _. .. .j_ ,_, _._ 4 : :- .
I -
~ . .. - . ..- -._. -.. .-
I t l . j .3 J g
%..d- ._ .-
t , .; - *-
.. . ... , E
._ f ~ * *^ -"
..q
,-. ._, ..- . _ . . . . . . . . _ . . _ . . . . . . _ . . . _ _ . _ _ _ . . .___. f
.. ~ . . .. . .. - - ~N
-- t
_ _ -.. d-. - -.. _ , . _ . . - - - .~~. -- . . ..._ .. . s -.. _ . ..
..-....e - .- . -. . _.. -..
l I
..-.7.. ....,.
I- _! _ - __ 4
, _. . .._. .._.. . 4 . ._ ~
f;-.. g'-. . ! f
. ..... . . .- . .... ,_ _ .__ ... . .w.
-4 4 f a
...l d 6
.*. k I . ' ,
_._.i _ . . . .$ .. . . . - . 4.. .. . i
. - . - * * . = - 9.g e .e. w .wwh-. .e. w. m. 4- . ump e.=.-
i j -- .
. Ys ;
. _a f
i- l.+. - -. . . . . .- _.en--.e-._--... ...4 . . 8 f I.-_..I . . . _ _ . . k. 4 ._.
} ..
fp , . . i . .
/. !
l .~1 t g ,..-.-6.....4... . 1 f. . e._.. ..L..
, J f J i . f. f ; .s hl -
l o ~-.
- i. . . . . ;- . ..-. y' .. . ,. - '
p*J._.* g
. ..~.-t. ,, .. ep_s. -
~ ._ _ _ . -
. . g ,.
6N._6 M8' D6& 6 .*"a.*W."h. - M."6
- 8"* d m.m.,, e--ew . -
m...M= .4.h* 4* -- .6 J8.'-*d'm+.* .
.__.4 ..,W...mme. .g . .e. e- . - -M."M**
. ,..* _ ..... 1
,,..,,ha. .g.. p .a. e -= 9 - -- . * - w.
p _ . , , - . . .be - ,
. + = - - .--e- . . . . .
l I l l l e - . m t' W 5- . l O (DkB% t:r T
- uN5)a Q% S' a -
- y. -
s** b s n 1
... . . . . - . . - - - . .. .6..
?Q , f..i . e. G '. 4?III!
-%,&~. /
6d ,
* ;l
. . - . . -----..- q
._4. . - ,
_. .._. . - - - - -- - ~ ~ - - - . - I _ . . . _ . __ I
.- , _. . , _ . . ... . . _.. - - . ...-d By -em.4.e. -em.mw. s+66-- m me l . . .M.. .. - . ..-%%---_ -- - ~ - - * - -
l l . 3- !_ _. _ .t._.x ._J...-__-- _ L . . . _ . .._ .. ... - - .~. --._. - . - . --. . -- - - . - - - --- - -- -- l i I
-e.p ..wem e amme.e w e.e g e- g e e.e. &-* - * * * " " -
- e mme--.w*a- ..ms
. . . . . - .. _. . . .. . -- - -. ..- - - . . . - -- - - -n-------- - ---- - - - -
em-m - * . - =u=.'e=.'s* - .* N6
.4._ .. -,.+.eeew -$r =W-5 *d-N.'Om-m-N* D-M MdUu. er i ._ __:. . _ .
...._ . _ . ._ .-4.- - - . -.. ---.-- - --- - ---- ~~~ - - . - - - -
' 1 i I i
.,_,.w .- e e.- * * * - " "
_ . , . m. s.mme.-m ge gbp+eems.N - P *e g-eee 6=+ - + - . * * -P ^ * *- "*P - m -M
} ~ ..-. jf-_ al? [ f *$ , - -- - - ---- *.-- ~- - --.--~~+ -~
b T3 . _ , _ _ . ... - . . - -*=
---------}- . -- ----
..- - e -
t & 3 ....n..n. .-- --- - - - - . = - --
- - -_ - ~ *- ~- - - ~ ~ ' ~
L - _ g >, . _ .
= = = . . , _ _ . . . _ .--- - --- --.- -- - - - . -
l 9 l *U l __ me- d *
, p..ma 1
M l 1
^
4 ..- s 4.a. e
. e.--+ ~ wam-I 1
h ! % us. ! m y, ht yn. ucoN i N n'.s\ : s.
- 'e, 2.
& s N
I.
f Southwest Research Institute Department of Materials Sciences TENSILE TEST DAiA SHEET , Specimen No. ,2-d Project No. / 7- W8 Test Temperature /2 T Machine Ident. M Strain Rate .er'E/m4 Date of Test 4// r'/W
/
Initial Diameter ,2f/ Final Diameter , /gY Initial Area ,g 4 9 74 Final Area , c/ gg 42 Initial Gage Lengtn /. o Final Gage Lengtn /. .t f r ipecimen Temperature: Maximum Load va70 Top T.C. / 0.2% Offsee Loco 52go Middle T.C. a '/ 74"# Fracture load aWO Bottom T.C. /L' Elong. to Max. Loao ~ . ,yr 78 N At YM ' U.T.S. = Maximum Load /!nitial Area = 9%, AIR,_ 0.2". Y.S. = 0.2". Offset Load / Initial Area =_ 4 5~, 30 >' Feature Stress = Fracture Load / Final Area = /eZ#97
" R. A. = 100 (Init. Area-Final Area)/Init. Area
= A J, 753_,.
i Total Elong. = 100 (Final G.L.-Init. G.L.)/Init. G.L. = #5' f' l 5 Uniform Elong. = 100 (Elong, to Max. Load)/Init. G.L. = ;2 (G4 Test Perfomed by: r*~MRor,0 /A AnVOL, Calculations Perfomed by: MsMd/x - (Data) 6//9/99 calculations checked by:.[ecM4 6-[wf (cate) 6/.7 _ff
* ; e i +1 e -s e
. .S af.s e% g g,eage .3 a.Cas e e
7 14,.) .
/
*e. {./
s~-% 8 , j 1* d__ .. 3
.- x_ ;- . . -' .t . .
t..__.. _.- ___ _ ._ .,.
- n. . , Q ,
. .u -
t ._w . ..
+
. = _ . _ . . . . . . .
l- i ( - j
._- l
-l 9
' - * ~ -
_-- ......__._p-_u-1.
.e . . _.. .m. me._. -. ,. .
. '. . f t - --[.-~ *g- ...t+. * : -t I
1 J
. _ .. .. - _ . ?
f
. .- . . . . . - -. .. .. . . . .. . _ . . ~ . . . - - ~..- - .. . _.
1 e - . ... .
. _ . a .
l _ . _ . . - . --.._.- ;. , - . . . ., ._. ._ .. . . . . - - . . . . . . .,_ ..._ . . __. ..._ ) e
>-. _~ _ , ,
I
._-.[ . - . . _ . . ..
.y _. .
_ . . . ._., .._., ._.....t . . . _ _ . _ _ . . . . . . . _ " .*
- 1 4
. . . . q. . a --. - .
- . . . _ . _....__ . ._ _a__. .
Q .
- i ,
- )
-4
_ . .C. . . _. - _ . . - _ _ . . . . . _ . . _ _ _ . y . ... .
* - e. .- . e.._e -_ e .. . _ . . . _ . . - - . . . . . p.:--.._--me . .. e.-...
. . . . - . .s*
l
,.e. . e_ . - . ...-.y
' .e.o e * .=_e. -e . g, .. ,.e.a.e - .e. .e o. w ..
l
. y . _. . . ) ..
. - . .- .m . = = >
-...e.-..e ,. . . ... .w- * * + . . . ..e.
. _ _ _ _ _ . . . _ .. . __ L ..._._..__.p_. . . _ . _ . . , _ . .. . . . -_ ____ J
"*4* *h_. 'B-* * * '
- 4. **-.
.p.e.--- * . - -
-.=e . ." *.6.e. . ee=-
, . ne. .,.
- . ._ . - . . -es e. e4
_ * * * . * * * - . * - -- . - *--^ *_W
- _ . * .
.-...,e- g.-_-.
eg
.g..
.>--6.W.a. m.-- p + we -h* . + + .
, w.y., , , , ..e w .o +.m . -_ee->w E
,6 B =,
[ s O 9 y* N I f w hAl, w N, . N4* =g A
*l D '
?w t
s. A 8 me gik l 9 %
vei .i.*. . *L.# " * '
.*" *1 1221
% {s+ s ,
- .- - . . s v i .
, t
- t. 3 t.
b
. ? !.e.a m i
j
.m.-w..~ . eo. m ms m ~
4,
._.,..-.i_._.....-.._.-3-.. .--. . . . - -. .
l
- _ . 1
, . - .r .. 7..a-
- ._._...._._7_. . - - --
. . _ _ . .. -. .... ._-. . - l 1 t j . !"
-. . _ . - 1..
- i : t--- _ u t. -. +: .--.-
t
.. - t . . . .
. . . . . . ~ .
I
= _ - - * . --
---.*-a.-.--*-*-*---==*- - - * * ""O r- --
. _. . . . ._. - . ~ _ . .
... ! . _ . _ . . I 4_._.. . . _ - _ .
.[ 4 f
. i
.I._,
8 6 1 j _ _ 4t .. g g .- _..___.__.p__...._.. _~
.a
' '}- .
'.. I
-. f ' d !
3 a
.. t.p _ p ; i
.. . .:I _ ;...,1 _. ;__ . .: ._._ ..;, . . _ . . 9_ . 4 ;
; [ t
, t ' 1 .t i .r- i '
t t'_... 3... .
; ; ;- , j
...,....J___.i,
- t. . _ . i i . ._.. _ _,
. - L
.__.,-f~*
l
, . . . p. '
j i l e
. .' I I I'
' I
. .a . . . H - _ . _ . .
.c _ . . . - - . . - . _
e j I j .
, j. j. .
.M. . _+_
.'......_,..m.
h l f .. l I 4 ..*_ Q .- .. .. . t
. . __ f t .. - .--
. d. ' I
. -)
1 4._. - 4
.._..b--..--.-.
=
I t d
. _ . . k _1 a _ ,
._... -q
'.._.g ~
.+ .
i [ p
- . l. .
-3 5.- _ . -. . - . .. .
_ . g. . _ .2. ._=a.---,.._3_*.-
.%* O '
b
- ; .a --i . . 7, . _ _
---4-.--.-. - - - -
- . . _ - -_ l .._j -
i 3 i 1 , I t i **
..p.____, . ..
1
' *. _ j _ . - .i. . . . ' . _ . - - --
t ( l 1 6
.M, . . _. -. P ^ __ .
3 . - _ _ . . _ . . . _ . . _ _ . . . _ _ . -_. y o e. ao . m i Nt _ . ... - . _ -_. .-. I * ... . . _ . _ . . - . . _ - . . . -. --- - - - -- { N.
,..e g
~ . _ . _ _ . . . ~ . _ . . - . . . . _ . . - . - .
l . l 1 A ak ._ ._._ _ v
$9 1 $
~
7-y . . _ . u
+A. ~
N, kyasM h N% ><
- xg N% **b a'a h cp W .N to.
1
l Southwest Research Ins .tute Depart. ment of Materials Scie.nces TENSILE TEST DATA SHEET , Specimen No. ot-7 Project No. / 7 _a /o e Test Temcerature 7C0 # Machine Ident. Y Strain Rate .co 6 YA> /s/ Cate of Test C //l /79 Initial Diameter , .;2 Y 9 Final Diameter /J 7 Initial Area ,04167 Final Area ,o/a66 Initial Gage Lengtn /,o Final Gage Lengtn /,/ 7 y Specimen Temperature: Maximum Load e/ t'oo Top T.C. Cf/ */ 0.2*. Offset Loaa 3230 Middle T.C. A/M Fracture Load 3/ 70 30ttom T.C. fry *A Elong. to Max. Loac .e 7 %# (b b a Ah U.T.S. = Maximun load / Initial Area = 9494 [ 0.2", Y.S. = 0.2". Offset Load / Initial Area = 4 t ,,Mf Frature 5 tress = Fracture Load / Final Area = A fd237 f
= 73 f f i R. A. = 100 (Init. Area-Final Area)/Init. Area i Total Elong. = 100 (Final G.L.-Init. G.L.)/Iait. G.L. = f 7, +'d i Unifom Elong. = 100 (Elong. to Max. Load)/Init. G.L. = / 7, 76 />
Test Perfor ed by: WAcerAj /p' A 7)XC// Calculations Perfor ed by: I f,_ [/ d.v b fcate) 4//v/TP Calculations Checked by: hc< ,6 8/ - [tbate) 4,/2//9f
/ ,
j
f.*5 hl .L. 3% *.:,; .;.' '" 4 47 .m p>
%4 y// w / &' f
_. __ j.; . f ..
.! t. JJ . - - - J
._. _ . . . ._._ _ . . . . . _ . . _ .. p - -
- . , .j -
p p.u-. ;.- !- -- !
- .j 8
t i i
. . . , j_____...___p_._____...__ .
.._q t
-. - . . _.-4' .'
i i-i i
.___ ,_.... % .t
, I j !. ;
i j l ..! l- I j t . .
.I., . .
__3
\ . .
. i 0
l i + ., , j
. . i }
I .
~; 1 u._. '. l
__....a.._; i _ . .t - ' ' -- '. '
.._L.._-,_..
i
. . . .._ t_._
I
!. l 1
- . -_-_ q t - I ' i i I i l I ! i f r
- t
-.7----+----. rj
, . l- -
_._-._....a
.___.__.j. .!
4 ! e 4 I - .I - N .,._4.- t-- n ; m.2 r. .[ [ ._i_..._.;..__
/i 4_2
.,i k. - . -t *-
1- ' j ' -
. } . j. -' 1 l , - j
_ L-- _.l___.....
- 1. : i __._, 4i
~{ . _ L._: _ . _.2..j . .y.-- 1
._ ] - ).
] ~ .1_ {_. ._ ...i u _. __ .
i I . g- , l , ,
! i
- s. , - ' .
t l ' -, _ _ _ . _ . . _ L ._ -.i _. _ .t.._. (
.._ \. .,;_ _l_. , -. '__-l l I
l 3 I" t 'f ^ ' o 3
-- ~ I
' ~I I- #
! l !
~ - -
i -
,e o f-p
*#> . - ** a t . l # -.t
~h-*- '
.}, , , ,g .
a, 4, _ . _ . . _ . . _ _ . . . .
'# =T
- l l - ! i
. t Cc
- e
. , , i . r c r .
i
# 4. .
g-
. .t . j .i g g .
._f
. =
p: . l: ! y y - I
. ' j
.._.l . h . ;,._ }. .
.. j .. . ..
3h g
*"/ ? '
- l. l
- l
.L- , . g' {: >
j l - } I' !.-l. l. . I . l _3 _. ; _- l . . . 7
,_ . 4. . -t.
1 j.
.}. .~ _. ..
-.-.! .1 l
'i
'. l x' 'j4 [ s 1 _ , .. .. ! 2 __ j . a._;._,-..
- - , i .
f
,, __.. .__i .~ .._w. -. .. . . _ . . . .
. f .
. _ _ _ .. .~. . ._ . _ . _ . . . . . , _ . . _ . .
k 4 . .
-e._.. _-.. ._- .__.._-._4._ . _ _w . _. - . . - ----... . -- .
I
' p.
.e 4
., . e- .-a w
--[ i l
\ _. _ .. -
. . rhmee e. . . _ _ - em-
, -.e , e.e- e eew+..N+
1WQ
.. ._. . t
*o
- s. -
.I 4 .,
A e ee -
- s, ,. -
- r. M
#8I .'s .' i. . . . . "i".
- 471 43 pp,Y J'4 lP bW y l ) , .
6 , . _C _ _ ._ . ._ . ; - : .. 4
**=he. . e m.
"88"'"'-***m*+ +eep +e +w mm,,., ,_.
'* ~ ~ ' ' " * -- . . . - . . . .
* " ' * " ' " * * ***O __ . . _ , , ,
a*w-4e_m..
+ '" **" 6 %_mm.,
t !
- - - * * * +-- - .. _ .a AQ, .
** h
- ~"-- -* -------- ..- _- _. .. __ ..,, .
"-""^"****"WWW=m*. ,es, l
1
* ' ' @ .me , , , , _
*.'**-G.huee. 4g
*e* -Nme _
*M N""'"*-*'** * *= . . . . . . , .
* '"* **"-"". MM-w.w._me.i p.,, ,
* ' *8*- mea.. , , _
.e
- e
~ ~ - - - . - - + - . . . . _ . _ . _ . . . _. . . _ , , _ , __ , , _ _ _ _ _ _ , {
- . - - ._ i i
* ' " * * - - -- -.= e-_. ._- ,% _ , , _
_D
- ;- ------ .. . L ___
s
. . 4,
* ~ ~ - -+--- - - - - . . . _ _ _ _
-- --- l - - . a-.- _ . . - ;. . , , ;
' r
- ---- .-- a . _ _. .._. ______ _ . . . _ _ _ _ _ , , _,,_,_
- UP# ;7 7 {*
- ----- - . - . . _ _ _ _ . ... . . _. _ . . . _ _ . j .-
#A .
* ** w-eas.-. ..
J -
-%.._%., m, , ,
I
. I--.~-._ . . .
1,; -- -- -- .. -..__..._-_ _ _y_,,_,,_, 3 e A.. * * ' ' * * - * * * * ==*_% . .~ . . .. ,, , ,, ,_
..9
% - . = e. ee, 94
' ' ~ * " * * * * ' '~^~ -.
f, m
*~'* ' * * * * * * " ' ' " - * - ' = - = = . . . . - . . , . _,
+*Mhe4we..
~~~ -* . ._n ..
" = = * - .eee.g. ._ .
. O e==+. _ g e
e, ( h \A. 1 ) .sb$, gaN
) A
.p ,
)hA J
",:P ,
A
,J E, M g ',
. e*}}