ML20236L500
| ML20236L500 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek  |
| Issue date: | 08/31/1987 |
| From: | Lippincott E, Meyer T, Yanichko S WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | |
| Shared Package | |
| ML20236L490 | List: |
| References | |
| WCAP-11553, NUDOCS 8711100376 | |
| Download: ML20236L500 (101) | |
Text
_- .
I 1 l
N WCAP-11553 1
WESTINGHOUSE CLASS 3 a
CUSTOMER DESIGNATED DISTRIBUTION o
.g o
ANALYSIS OF CAPSULE U FROM THE WOLF CREEK NUCLEAR OPERATING CORPORATION WOLF CREEK REACTOR' VESSEL RADIATION SURVEILLANCE PROGRAM S, E. Yanichko E. P. Lippincott
, L. Albertin J. C. Schaertz August 1987 APPROVED: ' b/ F t T. A. Meyer,' Manager Structural Materials and 3eliability. Technology l
Work performed under Shop Order No. KXEJ-106 Prepared by Westinghouse Electric Corporation for the Wolf Creek Nuclear Operating Corporation Although information e.ontained in this report is nonproprietary no distribution shall be made outside Westinghouse or its licensees f- without the customer's approvai r
l$
WESTINGHOUSE ELECTRIC CORPORATION l Power Systems Division l P. O. Box 2728 Pittsburgh, Pennsylvania 15230 t
8711100376 871104 PDR ADDCK 05G00482 P .PDR t_ m_ _ . _
- n. !,
T, 1
l PREFACE, 1
This report has'been t'echnically. reviewed'and verified. . , ;.;
+
I s i Reviewer ..,
Sections 1 through 5, 7 and 8 C. C.jH,einecke' d d M a l. I Geetion 8- S. L. Anderson . Id. ' fimd onand . I Appendix A A. C. Chan --
1' I
l i
i 1
G 1
e j -
1 Is
,9 i .
i l
LLL ,
-s
1
-(!
l
, l l
k 00NTENTS
- 1. SULVARY OF RESULTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1.
- 2. INTRODUCTION ..............................................-... 2-1
- 3. BACKCh0UND.................................................. . 3-1
- 4. DESCRIPTION OF PR0 CRAM...................................... . 4-1
- 5. TESTING OF SPECIMENS PROM CAPSULE U........................... 5-1 5.1 Overviev .......................... .................... 5-1 5.2 Charpy V-Notch Impact Test Results ..................... 5-3 5.3 Tension Test Results ................................... 5-4 5.4 Compact Tension Tests .................................. 5-5
- 6. RADIATION ANALYSIS AND NEUTRON DOSIMETRY ..................... 6-1 6.1 Introductica ..............,............................ 6-1 6.2 Discrete Ordinates Analysis ............................ 6-2 6.3 Radiometric Monitors ................................... 6-4 6.4 Neutrun Transport Analysis Results ..................... 6-12 6.5 Dosimetry Results ....................................... 6-13
- 7. SURVEILLANCE CAPSULE REMOYAI; SCHEDULE . . . . . . . . . . . . . . . . . . . . . . . 7 8. REFERENCES ................................................... 8-1 APPENDII A. HEATUP AND 000LDOWN LIMIT CURVE 8 FOR NORMAL OPERATION ............................................ . A-1 A.1 Introduction............................................. A-1 A.2 Fracture Toughness Properties........................... A-2 A.3 Criteria for Allowable Pressure-Temperatures -
Relationships........................................... A-2 A.4 Heatup and Cooldown Limit 0urves........................ A-6 A.5 Adjusted Reference Temperature.......................... A-7 Av
, ,, <. , j a., . w ; - ,, - .. .
s.
'e_,
^
- . . , L-b -' 2 ;/9 .
'. ),# / ' - ., .
- s. '
+.
n r. 1- , .~.,.;, . f g. _
s ,, - v s _ '. : .
4 GP ILLUST1gyIONS 4-1 Arrange 4-1 vessel.. ment ofcesurv illan Capsule U............e
. . . . . . . . . capsules in the 5-1 thermal monitodiagram Lara rs andshowing location. . . . . . . . . . . . . . . .re.a ctor s
vCharpy e V notch i do imeters..of sn ............ 4-4 N 5-2 ess l shell E Charpy V notch ves plate R2508-3 (lmpact properties or Wolf ................
f........ecim 5-3 sel shell impact longitudinal ori e 4-5 Cre k reactor vCharpy V notchplate rR2508-3 (tproperties entation).... for W 5-4 essel ew ld metalimpact pr ansv ers e olf Cr eek re 5-13 Charpy e V notch. . . . . . . .operties for Wolforientation) ...... 5-14 actor 5-5 w ld heat affectimpact eds zonproperties e
Creek reacto for W ............r..
Charpy reactor impact ves orientation.sel shellpecimen metal.........olf fr Cre 5-15 5-8 acture ........ek reactor
. . . s. . . . . . plate ces surfa R2508-3 ........... i 5-18 r Wolf (lon Charpy impact fo rea c tor ves orientation)sel shellpecimen
................ g tud
........inal fractur Creek 5-7 e rCharpy impact s. . . . . . . . . . ...........
plate R2508-3 5-17 (tr c
5-8 ea tor v essel weld es metalpecimen erse Creek fractur......
Charpy reactor v c impa t s .......... 5-18 5-9 Tensile pro ess l w ld heatpecimen e
e e ................ Creek fractur. . . . .. . . . .urfaces 5-10 plate R2508perties for W affected zonesurfaces5-19 for Wolf 3 (lon gitudinalolf Cr eek rtac t metal..... . .....
Creek plate R2508-3 (tTensilevesse properti 5-20 es erse ransv for WolfCr eek rea orientation)or........l shell orientation)ctor ves 5- 21
........sel shell
.......... 5-22
P_ag 5-23 vessel weld Wolf Creek reactor.................
Figure Tensile properties for...................lf l
Creek
.........reactor 5-24 y 5-11 meta 1............... specimeas from ....... wor 2508 3 (longitud 4
5-25 Fractured tensile ...................
Wolf Creek reactor ~ orienta 5-12 vessel shellspecimens.from plateorientation) verse . . . . . . 5-26 Fractured vessel shell pla tensilete R2508-3 (trans lf Creek reactor...............
5-27 5-13 specimens fromspecimens.......
Wo..................
c 5-14 Fra tured tensilevessel weld metal. 6-29 Typical stress-strainvessel surveillance.................
5-15 Flan view of a dual reactor..................of reactor 6-30 maxim 6-1 capsule..............l distributionwithin the.....................
u flux 6-31 6-2 Calculated azimutha(E >:1.0 MeV) ne tr reactor surveillance distributionwithin the MeV) neutron 6-32 Calculated radialMeV) neutron ..........
6-3 (. E > 1 0 of f sa t (E > 1.0 flux vessel................ expression A-16 axial variation reactor 4 the 6-4 Relativewithin flux theuse in Equation A-.....................
of A-17 ction A-1 Fluence for ART factor forET......................
limitations A-18 A-2 Past full power Neutron service life Fluence (E (EFPY) for tup coolant system hea7 EFPY...........
i s A-1 h first cooldown limitat ......
on Wolf Creek coolant reactorapplicable for t e A-3 first EFPY....................
system 7 Creekh reactor Wolf applicable to t e A-4 vi
i 1
'Y r '
( LIST OF ILLUSTRATIONS- )
i Figure Page 1 i
4-1 Arrangement of surveillance capsules in'the reactor i vesse1................................................... 4-4 )
4-1 Capsule U diagram showing location of specimens, thermal monitors acd dosimeters.......................... 4-5 5-1 Charpy Y-notch impact properties for Wolf Creek' reactor vessel shell plate R2508-3 (longitudinal orientation) . . . . 5-13 5-2 Charpy V-notch impact properties for Wolf Creek reactor vessel shell plate R2508-3 (transverse orientation) . . . . . . 14 5-3 Charpy V-notch impact properties'for Wolf Creek reactor !
vessel weld meta 1........................................ 5-15 l
. 5-4 Charpy V-notch impact properties for Wolf Creek reactor-weld heat affected zone meta 1............................ 5-18 5-5 Charpy impact specimen fracture surfaces for Wolf Creek reactor vessel shell plate R2508-3 (longitudinal orientation............................................. 5-17 5-6 Charpy impact specimen fracture surfaces for Wolf Creek-reactor vessel shell plate R2508-3 (transverse orientation)............................................ 5-18 - .
)
5-7 Charpy impact specimen fracture surfaces for Wolf Creek '
reactor vessel weld metal............................... 5-19 5-8 Charpy impact specimen fracture surfaces for Wolf Creek l reactor vessel weld heat affected zone metal............ 5-20 5-9 Tensile properties for Wolf Creek reactor vessel shell i' plate R2508-3 (longitudinal orientation) . . . . . . . . . . . . . -. . . 5-21 l 5-10 Tensile properties for Wolf Creek reactor vessel shell plate R2508-3 (transverse orientation).................. 5-22 .;
y t
t O
Figure Page
/ e ,
5-11 Tensile properties for Wolf Creek reactor vessel weld .
metal................................................... 5-23 5-12 Fractured tensile specimens from Wolf Creek reactor l vessel shell plate R2508-3 (longitudinal ,
ori e nt at io n) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5-24' 5-13 Fractured tensile specimens from Wolf Creek reactor vessel shell plate R2508-3 (transverse orientation...... 5-25 5-14 Fractured tensile specimens from Wolf Creek reactor vessel weld metal.........'.............................. 5-26 5-15 Typical stress-strain curve for tension specimens....... 5-27 6-1 Plan view of a dual reactor vessel surveillance capsule................................................. 6-29 6-2 Calculated azimuthal distribution of maximum f ast (E > 1.0 MeV) neutron flux within the reactor vessel-surveillance capsule geometry........................... 6-30 6-3 Calculated radial distribution of maximum fast -
(E > 1.0 MeV) neutron flux within the reactor vessel. . . . 6-31 6-4 Relative axial variation of f ast (E .) 1.0 MeV) neutron
- flux within the reactor vessel.......................... 6-32 A-1 Fluence factor for use in Equation A-4 the expression for ART A NDT***********".................................
, A-2 Fast Neutron Fluence (E > 1.0 MeV) as a function of f ull power service life (EFPY) f or Wolf Creek. . . . . . . . . . . A-17 A-3 Wolf Creek reactor coolant system heatup limitations applicable for the first 7 EFPY......................... A-18 A-4 Wolf Creek reactor coolant system cooldown limitations applicable to the first 7 EFPY.......................... A-19
.c
\
s .
\,.
': vi h,
1 l
j'+
..~
'l
= ;
'1
'l LIST OF TABLES i
l j
Table Pag, -
-1 4-1 Chemical Composition and Heat Treatment of the' Wolf
~
- Creek Reactor Yess.1 Surveillance Materials............. 4-3 5-1 Charpy Y-Notch Impact' Data for the Wolf Creek Shell
-PlgeR25g8-3 Irradiated.at550*FFluence3.39x l 10 n/cm (E > 1. 0 MeV) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
5-6 5-2 Charpy V-Notch Impact Data for the Wolf Creek Reactor Vessel Weld Meta1 1 pd hag Metal Irradiated at 550*F Fluence 3.39 x:10. n/cm . (E > 1. 0 MeV) . . . . . . . . . . . . . . . . . 5-7 il 5-3 InstrumentedCharpyImpactTestResultsforgolfCreek t
~S hell Plate R2508-3 Irradiated at 3.39 x 10i n '
(E'> 1.0 MeV)................................../cm-
......... 5-8' 1
5-4 Instrumented Charpy Impact Test Results for Wolf .!
~*
WeldMetalandHAZMetalIrradiatedat3.39x10 greek 1 'n I (E > 1.0 MeV)......................................./cm .... 5 l 5-5 Effect of 550*F Irradiation at 3.39'x 1018 ,7 , 2 (B > 1.0 MeV) on Notch Toughness Properties of Wolf Creek Reactor Vessel Materials.......................... 5 5-6 Cosparison of Wolf Creek 30 ft-lb Transition Temperature Results with Regulatory Guide 1.99 Revision 2.
Predictions............................................. 5-11 5-7 TensilePropertiesforWolfCreekReactorfessel Material Irradiated at 550*F to 3.39 x 10 1 n 2 (E > 1. 0 MeV) . . . . . . . . . . . . . . . . . .. .......... . . . . . . . . 5-12
. . . . . . /cm 6-1 SAILOR 47 Neutron Energy Group Structure................ 6 , 6-2 Nuclear Constants for Radiometric Monitors Contained in the Wolf Creek Surveillance Capsules.................... 6-16
. 6-3 Calculated Fast Neutron Exposure Parameters for the Peak Location of the Wolf Creek Reactor Vessel...'............ 6-17 6-4 Calculated Fast Neutron Exposure Parameters and the Lead Factors for the Wolf Creek Surveillance Capsules.......,. 6-18 vii
u Table P,ase
~
6-5 Calculated Neutron Energy Spectrum at the Center of Wolf Creek Surveillance Capsule'U....................... 6-19 6-6 Irre.diation History of Wolf Creek Surveillance f Capsule U............................................... 6-20 0
6-7 Measured Radiometric Monitor Activities and Reaction Rates for Wolf Creek Surveillance Capsule U.............. 6-21 6-8 Results of Neutron Dosimetry for Wolf Creek Surveillance Capsule U............................................... 6-23 i 6-9 FERRET-SAND II Results for Wolf Creek Surveillance 4 Capsule U............................................... 6-24 6-10 Comparison of Measured and Calculated Reaction Rates Used in the Analysis of Wolf Creek Capsule U............ 6-26 6-11 Integral Neutron Flux Results Derived from Adjusted i
Spectrum................................................ 6-27 I
6-12 Summary of Wolf Creek Fast (E > 1.0 MeV) Neutron Fluence Results Based Upon Surveillance Capsule U....... 6-28 .
A-1 Chemistry Factor for Welds, 'F.......................... A-11.
A-2 Chemistry Factor for Base Metal, 'F..................... A-13 A-3 Reactor Vessel Toughness Data (Unirradiated)............ A-15 1
l l 1 1
l I
viii
9
- 1.
SUMMARY
OF RESULTS The analysis of the reactor vessel material contained in surveillance Capsule U, the first. capsule to be removed from the Wolf Creek Nuclear Operating Corporation Wolf Creek reactor pressure vessel, led to the following conclusions:
- The capsule received an gerage2 fast neutron fluence (E > 1 MeV) of 3.39 x 10 n/cm
- Irradiatig of the reactor vessel lower shell plate R2508-3 to 3.39 x 10 n/cm* resulted in 30 and 50 ft-lb transition temperature increases of,30*F, for specimens oriented psrallel to the major working direction (longitudinal orientation) and increases of 25'F for specimens oriented normal to the major working direction (transverse orientation).
- Weld metal irradiated to 3.39 x 1018 n/cm2 resulted in both a 30 and 50 ft-lb transition temperature increase of 20*F.
- The average upper shelf energy (transverse orientation) of the plate R2508-3 increased from 93 to 95 ft-lb, and the limiting weld decrgsed 3f om 100 to 92 ft-lb after irradiation to 3.39 x 10 n/cm . Both materials exhibit a more than adequate upper shelf energy level for continued safe plant operation and are expected to maintain an upper shelf energy of no less than 50 ft-lb throughout the life of the vessel as required by 100FR50, Appendix G.
- The surveillance capsule test results do not indicate any significant changes in the RT E Yalues reactor vessel, and, therefore,Ta projectad for the low risk of vessel failure from pressurized thermal shock (PTS) events is postulated.
t
- 2. INTRODUCTION I' This report presents the results of the examination of Capsule U, the first capsule to be removed from the reactor in the continuing surveillance program which monitors the effects of neutron irradiation V on the Wolf Creek reactor pressure vessel materials under actual operating conditions.
The surveillance program for the Wolf Creek reactor pressure
. . vessel materials was designed and recommended by the Westinghouse Electric Corporation. A description of the surveillance program and the preirradiation mechanical properties of the reactor vessel materials are presented by L. R. Singer.(1) The surveillance program was planned to cover the 40-year design life of the reactor pressure vessel and was based on ASTM E-185-79, ' Standard Practice for Conducting Surveillance Tests for Light Water Nuclear Power Reactor Yessels'. Westinghouse Nuclear Energy Systems personnel were contracted for the preparation of procedures for removing the capsule from the reactor and its shipment to the Westinghouse Research and Development Center where the postirradiation mechanical tosting of the Charpy V-notch impact and tensile surveillance specimens were performed.
This report summarized the testing of and the postirradiation data obtained from surveillance Capsule U removed from the Wolf Creek reactor vessel and discusses the analysis of these data.
l
. I i
LA % 0 2-1
p '
,j, ,
c e
'3 . BACKGROUND The ability of the large steel pressure: vessel containing the reactor core and its primary coolant to' resist fracture constitutes:an>
important f actor in ensuring safety.in'the nuclear industry.= : The .
beltline region of the reactor pressure: vessel is the most critical:
region of"the vessel because it iw subjected to significant fast' neutron bombardment. The overall effects of; fast neutron irradiation on the mechanical properties of low alloy,iferritic. pressure. vessel steels such-as SA 533 Grade B Class 1 (base material'of the Wolf Creek reactor.
pressure vessel beltline) are well dociusented in the literature.
Generally, low alloy.ferritic materials show an-increase in hardness and tensile properties and a decrease in ductility and toughness under certain conditions of irradiation.
~A method for performing analyses to guard against' fast fracture in reactor pressure vessels have been presented in Protection Against Nonductile Failure", Appendix 0 to Section III of-the ASME' Boiler and:
Pressure Vessel Code. The method uses frteture mechanics concepts and is based on the reference nil-ductility temperature (RTgg.) . .
RT ET I", defined as the greater of either the' drop. weight nil- ]
cuctility transition temperature _ (NDTT per ASTM E-208) or the -l temperature 60'F less than the 50 f t-lb (and 35-mil lateral-expansion) . !
temperature as determined from Charpy specimens oriented normal j (transverse) to the major working direction of the material. The RT ET ]
of a given saterial is used to index that material to a reference stress j L intensity factor curve (KIR curve) which appears in Appendix G of the , j ASME Code. The K IR curve is a 1 wer bound of. dynamics, crack arrest',
and static fracture toughness results obtained from several heats of j
, pressure vessel steel.. When s'given material is indexed to the K j IR curve, allowable stress intensity factors can be obtained for this' H m
a
- r;
,(' , 1 o
- c. , ,.
i i a
! ,)1 j
- ( e I
material as a function of temperature. Allowable operating limits can # , .
then be determined using these allewable stress intensity factors. '
RT and, in turn, the operating limits of nuclear power plants-ET can be adjusted to account for the effects of radiatioh on the reacto,r .
vessel material properties. The radiation embrittlement changes in. ,
mechanical properties of a given reactor pressure vessel.st' eel"can.be o monitored by a reactor surveillance program such as;the Wolf Creek Reactor Vessel Radiation Surveillance Program,II) in which a surveillance capsule is periodically removed from the operating nuclear reactor and the encapsulated specimens arm tested. The' increase in the average Charpy V-notch 30 f t-lb temperature (ARTNDT) dhe to irradiation is added to the original RT ET to adjust the RT ET f r radiation embrittlement. This adjusted RTET (RTET 1"iDi"I + EIET) is used to index the material to thn K g enrve and, in turn, to set operr. ting ,
limits for the nuclear power plant which take into account the effects of irradiation on the reactor vessel materials.
f
, f f
d p'
/
I
('
1 e 1 3-2
i'! ,
, , my s r 3_
2 * ,~
y \tf q
. o (a.
- j. ,
p X.3 ;
O .
g q
.c.M ' gj 1 ,
=
x
'{.
- 4. DESCRIPTION 0FLPROGRAM d
.$ix. surveillance: capsules for monitoring the eff4 cts-of neutron: H v;
H erposure on the Wolf Creek.resctor preasure v'essel core region material. '
"j were inserted in'the-reactar> vessel prior to' initial' plant'startup., The. '
4 l
V six capsules were positioned;in the reactor vessel NtweenLthe' neutron '
shielding pads ead the vessel wall as ishown in Figure 4-1.1The vertital. ,, j eenter of the capsules is opposite the. vertical center of the' core.; i
- d Capsule'U was' removed after'l.08 effective full power; years of 7 l plant operation.- This capsule centained Charpy V-notch, tensilei and- j 1/2 T compact tensi'o n '(CT) specimens (Figure 4-2) from the lower'shell.-
plate R2508-3 and-submerged arc weld metal representative,of the. 5j intermediate:tolowershellbeltlineweldseamofthe' reactor l vessel ~and j
Charpy. 7-notch specimens from weld heat-affected isone (HAZ) material.' l
~
t Allheat-affectedrenespecimenswereobtainedfromwithin~theHAZo[ ;
piste R2508-3 of the representative weld.:
l The chemistry and heat, treatment of the Elf Creek surveillance 1 material'is presented in Table 4-1.
l All test' specimen were machined from the 1/4 thickness location 1 of the plato. Test specimens represent material taken at least one '
plate thickness from the quenched end of the plate. Base met'al Charpy d V-notch impact and. tension. specimens were. oriented with the longitudinal l axis of the specimen pt.rallel to the major working direction of tho' plate (longitudinal orientation) and also normal to the major working ~
direction (transverse orie2.tation).' Charpy V-notch and tensile specimensfromthe.weldmetalwereorientedwiththelongitudnalaxis~
- l. of the specimens' transverse to the welding direction. The CT specimens-in Capsule U were nachined such thst the simulated crack in the specimen
. would propagate normal and parallel to the: major working direction'for the plate specimen and parallel to the weld' direction.
I-4-1
...i__._m.__- ^
N CW 7
, 3r --
7%
. (U [
y' l(9 - l \rA { :,
, a;h
-3,
'3 .
.~
/r d [- ;y{ '
(
\
t
, Qu ep' u w
3; a.s a y, .4 .t z ,-
't 4 - "
- h. 'p
. Capsule U.contain$d dosI h ;[y ath M " "l- res of pure coppe $ , iron, (V nickel, ad e,1uminum-0.15% eMait '(cadnium-shielded-and unshielded) . In addition, cadmium' shielded dosimeters of neptunium (Np237) and uranium
'-- ~
(U238)1r' are contained-in' the Espule. . 4
Thermal monitors made from the two low-melting eutectic alloys:
Od se:. led,in Pyn\extLUeswerr.pacludedinthe~ capsule. The. composition < . . .. .. 3 55 of.the to alloy,.s . and 'their 'selting point's' are' as follows.
, e. s t t . , ,s-k
' 1, .- "-
., . t )
+ ,
. . < ,- 1 y 2.5% Agw 9 M W Pb Melting Point: '579'F- (304'0).
,EkT5ASn,971%Pb Melting Point: 590*F (310'C) 1 e ' '( % . i
< 1.75%. .,
,, 3 7 3
[ j. 4 f *y ' 'j.,
}
g' a :s-'
(,~ " ;4
- 2 jh,y by} age,h :tb y t _ $f ' thel'
,. u. m i arious mechanical specimens, dosimeters.
. v. c <. .
l7 a-R h M f1:s3 nil 6 N contain(d it: thsulelU'areshowninFigure4-2.
' c a :.
1 q, s . ,
' y :, \ 7 s\
- i I
's I k ) 'I i
% L,,.,j 91
's, ,f N r
- 1, g k '[ ' !
u .
n, y , i -
s; a
, s 5
\, ( i q ,* % }'
1 " , 1 b'1x k' s s
,,,s.
A 9
- f. s x5\, '
l
)
4
) w
, sq 6
. \\ '
p( 3 ;
-5.
i g
\\ j'g (-
- g9 4
s
(**
l \ \*
4 l. y g4 2 t
., 1 i ,
% q l, *, .
-3 y qt -
a
} ( )
? *
[t g 5 4 -
)' .' d 1 <. 4 t ,
b,/
- / l. ,[
c .
( '\ } ,
k.-
1 k g
' 1 (
l' ' 't
\
'4- 2 I
.; 4 4 .. 3
- y.- __ -
,, a o.y q.
+: > -
' Table 4-1.
^
CEEMICAL COMPOSITION AND HEAT TREATMENT.0F THE' WOLF CREEK REACTOR VLSSEL SURVEILLANCE MATERIALS =
Chemical Composition L(wt%)'
L Element' Lower Shell Plate R2508-3 Weld' Material (*)
0 0.20 0.11 Mn 1.45 1.'48 P. 0.008: 0.005- ;
S: 0.010 10.011- '
{
Si < 0. 20 . .0.48 l Ni- .0.82 O.09 Mo 0.55 :0.56 Cr- 0.05- 0.09 Cu- 0.07 0.04 Al 0.032 0.009'
- Co '0.014 0.010 l Pb <0.001 <0.001 l W <0.01 <0.01 Ti <0.01 <0.01 Zr
.<0.001 <0.001 1 V 0.003 .CL OO5 'l Sn 0.002 0.003-
!
- As 0.007 0.004 Cb < .01 (0.01 N
2 .0.007 'O.008 i B <0.001 <0;001- i Heat Treatment History' Material Temperature (*F) Time (Hr)--- Coolant L Lower Shell Austenitising 1575-1825 4 Water quenched ;
(Plate R2508-3) Tempered 1200-1250 4 Air cooled '
Stress Relief 1100-1200 8.5 Furnace cooled g Weld Metal Stress Relief 1100-1200 10.25 Furnace cooled
(*)This weldment was fabricated by. combustion Engineering, Inc., using
.. _3/18 inch Mil B-4 weld filler wire, best number'90148 and Linde 124:
flux, lot number 1081 and is identical to that used in. the actual fabrication of the reactor vessel intermediate to lower shell. girth-
. weld.
1 4-a b.
O. REACTCR VESSEL
,330 COFE BARREL.
(3.85) Z ,
NEUTRON PAD CAPSULE U (3.85)
I I V (3.65)
O se.s~~ 'se . s. L.
- 1 ._
270*- --- - _ _ _ g, 35* 2.5' i
) / y 50*_ 4a,7 (3.65) Y /-l l\
(3.85) X 3 W (3.85) 2 W' ieO*
PLAN VIEW Figure 4-1. Arrangement of surveillance capsules in the reactor vessel ,
4-4 l
- e-e 9 .
LEGEND: AL LOWER SHELL' PLATE'R2508 3 (LONGITUDINAL)
' AT - LOWER SHELL PLATE R2508 3 (TRANSVERSE)
AW WELD METAL -
'AH - HEAT AFFECTED-ZONE MATERIAL -
L l
CouPACT - CouracT CouMCT commCt I' AM TEN $tLE TENStoM TENSION 'CHAAPY CHAAPY CHARPT ' TEN $loN - . TENSION CHARPy i
AW3 AWil AH15 AW12 AMg g AWg AHg AWS AH$.
AWT AW4 AW3 AWf AW1 AW14 AH14 AWil AM1 t AW8 AHS AL4 AL3 AL2 AL1 AWS AH$
AW1 AW13 AH13 AW16 AH10 AW7 ANT AW4 AM j .
C. 3 ll l As .f s*. ce C.
lls 14 1 u',1 ._, : ' m moNston- A' 88% C* tCd3
~j ll ll i woNiron i1 I l1 f t iI p.
! sl,ll f Lt TO TOP OF YESSEL l
O
'M 0u3639 '
t
l
]
l MP - - .
(P = t OO$1M E T ER COMPACT - COMPACT -
'l -
CHACD2 SLOCK f tNetLE CHARPY CHARPY CHARPY CHARPY CHARPY TENSION - TENSION TENSILE xn cm as Arts uts Atil air Ars c -
Ars as Ars as Af* '
j t :
xs mer 465 42 Afla Aus ATil ni A tt AUl Afs ALS Art AL2 AT4 AU Af2 Afi AU j
- wi Asi . , ni Arts Aus Atto nio Art a:i Are 44 Ari si Ari j 1
t n N I
,. .i.. . . . . -
- i. u-ii'.C. C. ' ig lI ! u .is . C.
4i l:i 1 l; ,
j!I i ei e i l 1 18 ii ii.eIII i ewuw is a wus Al- i$'. Co iCdt F 8,7 g
l l M- 5*. Co 4Cd) lg si is i i Ie I :< 1 iI it i I II I )
g, !! !!i ,,
! .! H i CENT [Q REGION Of VES$tt i
. 1 i
Figure 4-2. Capsule U Diagram Showing Location of Specimens, Thermal Moniters, and Dosimeters i
j)
J TI 4-5 APERTURE g CARD g 7 /lIDD37h- O I j Amo AvaMahle On Aperture Card .
1 I
i l
~. -_ _ __
4 i
l 4
1
- 5. TESTING OF SPECIMENS FROM CAPSULE U i 5.1 OVERVIBW The post-irradiation mechanical. testing of the Charpy V-notch and tensile specimen was performed at the Westinghouse Research and Development Center with consultation by. Westinghouse Nuclear Energy Systems personnel. Testing was performed in-accordance with 10CFR50, l Appendices G and H,(2) ASTM Specification E185-82, and Westinghouse Procedure RMF 8402, Revision 0 as modified by RMF Procedures 8102 and 8103.
Upon receipt of the capsule at the laboratory, the specimens.and spacer blocks were carefully removed, inspected for identification number, and checked against the master list in WCAP-l'0015.(1) No discrepancies were found.
Examination of the two low-melting point 304*C (579'F) and 310'C (590*F) eutectic alloys indicated no melting of either type of thermal monitor. Based on this examination, the maximuu temperature to which the test specimens were exposed was.less than 304*C (579'F).
1 The Charpy impact tests were performed per ASTM Specification E23-82 and RMF Procedure 8103 on a Tinius-Olsen Model 74,358J machine.
The tup (striker) of the Charpy machine is instrumented with an Effects Technology Model 500 instrumentation system. With this system, load-time and energy-time signals can be recorded in addition to the standard measurement of Charpy energy p(E ) . From the load-time curve, the load of general yielding (PGY), the time to general yielding (tGY), the maximum load (P g ), and the time to maximum load (ty) can be determined.
Under some test conditions, a sharp drop in load indicative of fast fracture was observed. The load at which' fast fracture was initiated is
. identified as the f ast fracture load (P p), and the load at which f ast fracture terminated is identified as the arrest load- (PA)
- 5-1
l .
O l
The energy at maximum lead (Eg ) was determined by comparing the energy-time record and the load-time record. The energy at maximum' load.
is roughly equivalent to the. energy required to initiate a crack in the-specimen.. Therefore, the propagation energy for the crack (E p ) is the '
difference between the total energy to fracture (ED) and the energy at maximum load.
The yield stress (aY) is calculated from the three-point bend '
formula. The flow stress is calculated from the average of the yield and maximum loads, also using the three-point bend formula. j Percent shear was determined from post-fracture photographs
- v. sing the ratio-of-areas methods.in apliance with ASTM Specification A370-77. The lateral expansion was measured using a disl gage rig l similar to that shown in the came specification.
Tension tests were performed on a 20,000-pound Instron,. split-console test machine (Model 1115)- per ASTM Specification E8-83 and E21-79, and RMF Procedure 8102. All pull rods, grips, and pins were ;
made of Inconel'718 hardened to Re 45. The upper pull rod was. connected .j I
through a universal joint to improve. axiality of loading. The tests were conducted at a constant crosshead speed of 0.05 inches per minute -
throughout the test. 1-
~
Deflection measurements were made with a' linear variable.
displacement transducer (LVDT) extensometer The extensometer knife I edges tvere spring-loaded to the specimen and operated through specimen failure. The extensometer' gage length is 1.00 inch. The extensometer is rated as Class B-3'per ASTM E83-87.
Elevated test temperatures were obtained with a three-sone electric resistance split-tube furnace with a 9-inch hot zone. All l j tests were conducted in air.
Because of the difficulty in remotely attaching a thermocouple-directly to the specimen, the following procedure was used to monitor specimen temperature. Chromel-alumel thermoccuples were inserted in
- i. shallow holes in the center and each end of the gage section of a dummy.
- l. .
i 8-2 l
, - ._ .-___ _ _ - - a
l i
)
j
)
- specimen and in each grip. In the test configuration, with a slight' 1
' load onLthe specimen, a plot of. specimen temperature versus upper and lbwer grip and controller temperatures was developed over.the range-of <
room temperature to. 550*F (288'C) . : The' upper grip wt .used to control:
the furnace temperature. During the actual. testing ,ae grip.
temperatures were used to obtain desired specimen temperatures. 1 Experiments indicated that this' method is accurate to *2*P.
]
The yield: load, ultimate' load, fracture load,7 total. elongation, and uniform elongation'were determined directly from the' load-extension i curve. The yield strength, ultimate strength, and fracture strength l
were calculated using the original cross-sectional' area.' The. final'~ !
i diameter and' final gage length were determined.from post-fracture {
photographs. The fracture area used'to calculate'the fracture stress- l (true stress at fracture)' and percent reduction in area was computed-using the final diameter measurement. I 1
1 l .
5.2 CHARPY V-NOTCH IMPACT TBST RBSULTS The zesults of Charpy V-notch impact' tents performed on the l_
variousmaterialscontainedinCapsuleUirradiatedat3.39x'ip8 n/cm are presented in Tables 5-1 through 5-4 and Figures 5-1 through.5-4.- .
The transition temperature increases and upper shelf energy decreases- 1 s
for the Capsule U materials are summarised in Table 5-5. 1 Irradiation cf vessel lower shell plate R2508-3 material-(longitudinal orientation) specimens to 3.39 x 1018 ,7c,2 (Figure 5-1) resulted in a 30*F increase in 30 and'50 ft-lb. transition temperature and an upper shelf energy. decrease of 3 ft-lb.
Irradiation of vessel lower shell plate R2508-3' material (transverse orientation) sp'ecimens to 3.39 x 1018 37,,2 (Figure 5-2) .
resulted in both 30 and 50 ft-lb transition temperature increases of .
ll 25'F. The irradiated upper shelf energy experienced an increase of 2-f t-lb when compared to the unirradiated data'.
c: 1 5-3
^
__i l , , _ _ _ _ _ - - " " ' - ' - " " ' " " " ~
m Weld metal. irradiated to 3.39 x 1018 ,7,,2 (Figure J 5-3) . resulted.
-in both 30 and 50 ft-lb transition temperature increases'of 20*F and an upper shelf ~ energy decrease of 8 fti lh'.
Weld HAZ metal' irradiated'to 3.39'x 1018 ,je,2 (FigureL5-4) resulted in both 30 and 50 ft ilb transition' temperature increases of.-
65*F and an upper shelf energy decrease of 21 ft-lb..
The fractur's appearance of.each' irradiated Charpy specimen from-the various materials is shown in Figures 5-5.through 5-8 and sS w an.
increasingly ductile.or~ tougher appearance with increasing'. test.
temperature.
'A comparison'of the 30 ft-lb transition: temperature increases for the various Wolf Creek' surveillance' materials with predicted increases using the methods . of . NRO Regulatory Guide' l'.99, .' Revision' 2(3)..
is presented in Table'5-6. This comparison' indicates that the transition temperature increases resulting from. irradiation to 3.39 x 18 10 37,,2 are in good agreement with the Guide predictions.
5.8 TBNSION TBST RESULTS The results of tension tests performed o.n plate R2508-3 -
(longitudinal and transverse orientation) and the weld metal- irradiated to 3.39 x 1018 n/cm2 are shown in Table'5-7 and Figures 5-9, 5-10 'and 5-11. Plate R2508-3 test results are shown in Figures 5-9'and 5-10 and indicate that irradiation to 3.39 x 1018 ,j,,2 caused a less than'10'ksi increase in the 0.2 percent offset yield strength'and ultimate tens,ile strength. Weld metal tension test results shown in' Figure 5-11, show that the ultimate tensile strength and the 0.2 percent offset yield strength increased by less than 5 kai with irradiation. .The fractured tension specimens for the plate material are shown in Figures'5-12 and ;
5 *3, while the fractured specimens for the weld metal are shown.in.
Figure 5-14. A typical stress-strain' curve for the tension. tests.is shown in Figure 5-15. ,
s-4
l, 5.4 COMPACT TENSION TESTS Per the surveillance capsule testing prograa with'the Wolf Creek Nuclear Operating Corporation,.1/2 T-compact tension fracture mechanics.
apecimens will-not be: tested and will'.be stored at the hot cells-st the.
Westinghouse R&D Center.'
W 6
- 1 4
Y f.
5-5
p, , -
h ,
4'
< ].:
Table 5-1 CHARPY Y-NOTCH IMPACT DATA FOR THE WOLF. CREEK SHELL' PLATE R2508-3 F IATfD AT 550'F, FLUENCE 3.39 x 10 n/cm (E > 1.0 MeV)
. Temperature . Impact Energy Lateral Expansion . Shear SampleNo.,('yj, ('0) (ft-lb) g),- (mils). , ,(33), (%)
Longitudinal Orientation- .
AL13 (-46) 7.0__ ( 9.5) 9.0 (0.23) 0 AL14 (-32) - '14.0- ( 19.0); .
15.0 (0.38) 5, AL4- _0 (-18)- 17.0 (.23.0) 20.0 (0.51). 10 AL7.- 10 -(-12) 23.0 (.31.0)- 24.0 -(0.61). '10 AL1 '20 (- 7) .22.0_. ( 30.0) 26.5 (0.67) '15 AL10 25 (-31) 55.0. (.75.0) 45.5 (1.16) - 20 <
AL15 25 -(-31) 44.0 ( 60.0). 38.0 (1.00). 15-AL12 50 (10) 84.0 '(114.0) 64.5. . (1. 64) - 45 AL2 . 50 (10) 62.0=(.84.0). 50.5' (1.28)- 25 AL3 76 .( 24) 77.0 (104.5) 58.0 (1.47)- 45 AL9 100 -( 38) 125.0 (169.5) .74.5 (1.89). . 80 AL6 150 . (.66) 133.0 ;(180.5) 81.5 ,
(2.07) 90 AL8 225 (107) 151.0-(204.5) 79.0 .(2.01)- ' 100.-
AL5 325- (163) 148.0 (200.5) 81.0 (2.06). 100' AL11 400 (204) 136.0 (184.5)' 76.0 (1.93) 100 .
Transverse Orientation ATil -50 (-46) 7.0 ( 9.5 10.0 (0.25) 3 -
AT12 0 (-18) 23.0 .( 31.0 23.5- (0.60) 10 AT2 10 (-12) 24.0 (32.5 30.0; .(0.76) .10 -
AT15 10 (-12) 23.0-(31.0) 23.5 (0.60) 10 AT14 25 ( -4)- 32.0 (43.5) 30.0 (0.76)- '1G AT8 25 -4) 33.0 -(144.5) 32.5 ~(O.83): ,15 AT3 .40 4) 29.0 (39.5) '33.0 .(O.84)' 15.
AT13 50 10). 46.0 (62.5) 41.0 (1.04)' 25 AT1 50 (10)' 34.0 (:46.0)- 37.0 (0.94) 20 -
AT6 76 (24) 70.0 (95.0) 49.0 (1.24) 40 AT9 100 (.38) -69.0 (93.5) 58.0 (1.47) 50.
. AT5 150 (!66) 92.0 (124.5) 70.0 (1.78). -85 AT7. 225 (107) 98.01'(133.0) 74.0 1.88)_ 100
- AT10- 275 (135) 04.0 ~(128.0) 71.0 1.80) 100 AT4 325 -(163) 95.0 (129.0)~ 69.0 1.75) -100.
a
..j
=
1 i s-e l-i
i Table 5-2 CHARPY V-NOTCH IMPACT DATA FOR THE WOLF. CREEK REACTOR :
VESSEL WELD METAL AND H l l-1550*F, FLUENCE 3.39x10{gMETAgIRRADIATEDAT: ,n/cm (E > 1.0.MeV) i
. Temperature . Impact Energy Lateral Expansion Shear Samole No. (*F) (*C) L(ft-lbi.1J1. (mils)- (mm) (%)
Weld Metal- . ., ,
AW5 -100 (-73) 4.0 ( - 5. 57' 8.5' (0.22)1 ..2' AW15- -50 ' (-46) ' 17.0 L '( 23.0) 18.0 ,(0.46) 10 -
AW10 -50 (-46)L '29.0 (39.5) 24.0 (0.61) 15-AW1 (-40)' 22.0 ;( 30.0) l23.5' -(0.60)l 15 AW4 -40 (-40) 30.0 L(. 40.5) 30.0 -(0.76) 10.
AWB -20 (-29) . 25.0 : ( ' 34.0) : 24.5 (0.62)-. 20 AW14 -20 ' (-29) ' . 34.0. . ( 46.0) - 30.0 .(0.76)' 25 AW3 0- (-18): 51.0 (69.0) 47.0 (1.19) 45 1AW11 0 -(-18) 54.0- (73.0). : 41.01 (1.04) 45:
AW13 50 (10). - 72.0 ' ( 97.5)L 61.0 (1.55) 70 "
AW9 ~76 ( 24) 73.0 (99.0) 60.0 .(1.52) 70 AW2 150 (66)- 89.0- (120.5) '72.5- ' (1. 84) . 100 AW7 225- (107) 97.0' (131.5) 78.0 (1.98): 100 AW12 325 (163) ~92.0 (124.5) 73.0 (1.85). 100 AW6 375 (191) 89.0 .(120.5) 82.0- (2.08) :100 l
'HAZ Metal-AH15 -200 (-129) 2.0 ( 2.5) 7.0 (0.18) 0 AH9 -150 (-101) 14.0 '( 19.0) -
11.0 (0.28) 10' AH10 -100 (-73) 19.0- (26.0) 13.0 (0.33) 10 AH2 -75 (-59) 32.0 ( 43.5) 22.5L - (0. 57) . '15 AH7 -75 (-59) 38.0 (51.5) 22.5; (0.57) 15 AH6 -50 (-46) 33.0 (44.5). 26.0 (0.66); 20 AH8 -50 (-46) 99.0 (134.0) 64.0 (1.63) 50 AH4' -25 (-32) 130.0 -(176.5) 78.5 -(1.99) 65.
AH1 -25 (-32) 26.0 . ( 35.5) 23.5 (0.60) 30 AH3 -25 (-32) 50.0 ( 68.0) 40.0 (1.02)' 45L AH11 0 (-18) 84.0'-(114.0) 52.0' (1.32): 50
, AH5 50 (10) -143.0 ~ (194.0). 83.0' (2.11) 100' AH13 76 ( 24)- 132.0 (179.0)- '77.5 .(1.97) 100
'AH14 '150 (66) 134.0 (181.5)' 87.5 -(2.22) -100 AH12 250 (121). 150.0 (203.5). 83.5 (2.12) 100.
1 a-7
,,,s.. --. .
S
- - ' - - ~
= s) -
7 02735780
~
~
4898695661256 d si 218 8911991999987 - 7 l es 119 - 1 1- - 19196667 11 1 1 1' 1 erk :
it( .
YS t . ) -
L sds - 505500555 '
eap - 00 0 - 5000 -
roi 23 9 - - 8330 - - - - 242738990 - - - - '- - -
rLk - -
000O00002 -
A . (
00 0 - 0 1 2 l' -
~
~
e '
r
)
ud s 505 0 0000 . . 5005000005 5844892697
)
V e
t ap coi aLk (
452 333 1
4 -
6610 4352
- -. - - 3344334333 M r - - .
.~
F K0 .
E E1 -
R om) ? . - - - . . : l' C> t uc .
me 555 5 - 55050005 5555500500050 9330476725177 - 1 FE ei s mxp 373 - 11778227 L( 112 2 88777887 334443366666 O ia( .
Wmc TM R/
O n F m u - . . - 2 - -
8 )
g T0 md s i ap n050 o764 5
1 55055555 70409764 5 5' 05500550500 n6845802690998 L1 xoi i - - - o U aLk ( t333 4 44543333 i3344344334333 - 3 S x M a -
t -
E t a R9 n t 3 3 d e n
- T l) i e 5 S3 eec r - .
i -
E mi e O555 5 05000500 r5550000005000 e TT iYs 888 8 - - 87088558 O5788889887887 - 5 l A T # l 1 '
b T o( a e a CD t n s T AE i r PT d e MA - ) u .
v -
II dd s t055 5 00505505 s5005 00005050 D l ap i439 5 - 38585011 n5065905997853 YA eoi g - - - a PR ilk n332 3 32322222 r2333233222222 RR Y ( o .
- T - - : --
AI L H
C3 pA - - '
D8 srp o/ 106 155 - 92175971 9 97080648 022163689930' 354 867208072 6
E0 ePE 3267888 '
- - 112345 T5 i -
N2 g ER r )
M em 2 UE nu n RT EmAi -
TA i// 531 8 00080987 7312057615313 - 0 SL d xmb 468 05122117 2359895467576 - 3 NP eaEl 1 43433332 111111122222 I sM - . - - '
L i t
' L l ay f
(
E H m pA -
4 S rr/ 637574596071625 653368440461985 oad 513874597207199 589965377654856 NhE - 11114346600210 11122223557777 C -
11111 .~ ' -
L
)
yyb :
pgl 000000000000000 rr - 000000000000000 aet 747325424753186 733432946092845 h nf 11225468723543 22233234769999 CE( 11111 -
t p1 - '
smF 050005500600550 000055000600555 ee* 52 122255705220 5 1122455705272 TTf - - '
11234 - - - -
11223 er l e
_ pb 34 -
05 2 - ' 1 125 - 4 - 3' J
mm 1147I1121396851 1112813I1695714 au SN LLLLLLLLLLLLLLL TTTTTTTTTTTTTTT AAAAAAAAAAAAAAA -
AAAAAAAAAAAAAAA - - -
~
- e "
. c
. dsi -
^ - -
. ' - ~
l es 68973 68345261 209974071272989 erk 83145 - - 20029879 452243211200097 it( 1111 1111 .
111111111111 YS t )
sd s - -
eap - 500 - - 055 550505 roi - 624 600 - - - - - - - - - 557483 - - - -
rLk ( '
A 000 112 - '
010102
) e V r ) -
e ud s 55055 005 00555500550 t ap 64404 793 M coi - - - - - - -
80363063461 -
aLk 34455 433 -
0 r ( '
25445534424 F
1 K
E E>
R om)
C (E t uc me 1 Fm L c eis mxs 50550 80370 - - 00715568 00550550 550555550055555 550704630255265 O / ia( 2425 5$576666 W TM - 23454635866778 n -
g m F 10 u )
S mds 50000 55005000 000055005005005 T x iap 66415 - 81139653 905740734742520 L xoi - -
U 9 aLk ( 34455 - 44453334 S M 254455444444444 E 3 -
R 3 4 d l T l) a l S T eec t a-5 E A mie e50000 55550505 t e T D iYs M59990 888488781 e 555000550500050 l
T E T o(
u -
M 199999888888809 -
b d C T a A t l Z T A I e A P D M A )
W -
H I R dd s 00055 05050005 l ap 62646 82179537 000055555050050 YR eoi - - 359940653721394 P I ilk 24344 33332222 R L Y ( 143344333333322 -
A .
H A T
C E pA -
~
7 6' 0 4391 M o/ 06144 1225 556403510611124 D srp 14334 - - 513' E Z ePE - 1232' 4544 11223667376256 T i - '
4 1638778 N A g EH r )
M .D em 2 U nu n R N EmAi T A i// 21235 39075992 187463293151274 S L d xmb 29048 - - 51557440 123383343109424 NA I
eaEl zM -
212 22232223 11222312432333 T i t E l f M ay (
_ D m pA ~
6338667'. 9376139 L rr/ 274722415087117 oad E
W NhE C
333740713881841 12122244557777 115506900475670 11232724061002 -
- 1 11
)
yyb -
_ pgl rr -
aet 000000000000000 479205414239729 000000000000000 249283960043240 h nf 12232355778998
_ 11333925384335 CE(-
- 1 11 t p1 smF ee*
000000000060555 0554422 1 -
575227 1233 000550055500600 0507755222 211 5
TT1 - - - - - - - - - - - - - ' - - - -
er l e -
pb 50 4 13 2 5 0 1 mm 511148131192716 191276813415111 SN au WWWWWWWWWWWWWWW AAAAAAAAAAAAAAA HHHHHHHHHHHHHHH AAAAAAAAAAAAAAA -
?*
.M
~
)d y b e g lt rt - a 5 5 2 0 eati 4 9 9 4 e fd 1 Bs(e a r 1 -
el gt aI rr .
ab e rrh eoSd v s e Ablt Al a ai Fd a 8 3 0 1 4 9 0 8 1
.r 1 1 S .r -
S .i E
M s
U G 0 5 0
0 7
E03 ~
5 2 2 6 E d C e T t O b )' a 5
__ M IFi 0 5 5
- d 3 6 4 NS t( a -
OL f r -
A er
)I 0 rI YR 5 u eE t MT ead A gre
- 0. M aet rpa 1L emi 0 E v ed 0 0 5 1
>S ATa 4 1 1 S r - -
RE r .
(Y i n
2 R U eO cT 5 /C 0 5 5 5
- nA E E3 0 2 1 5 8 R e I d l 0K e b 1E t a E l n) a T xR i oFi 0 5 0 5 C mi*d 2 4 1 3 9 s( ar - -
- 3. LF 5 n 3 aer 3O prI W ex u T gEt' AF a ad O rl re N eaet OS v rpa IE A emi 0 5 5 0 TI t ed 1 2 2 9 AT aTa - - -
IR L r DE r AP . i RO n RR U IP F 5 0 5 0 E3 0 2 2 6 5
5 d e
F t O b) a IFi 0 5 0 0
_ T
- d 1 2 3 8 C t( a - -
_ E f r
_ F er
- F 0rI E 3 u t y
_ ead g gre r
_ aet rpa ~
e emi 0 0 0 5 n
_ ved 2 5 4 e
- 1 ATa r
- f
_ r l
_ i e
_ n h s
_ U
_ ) a
. 3l 3 l
- a -) e 8 n 8e l 0i 0s l s
- a 5d 5 r a l a
_ i 2 u 2 e t a e
_ r 1t i 1 v e t r e s M e c t eg en M n
_ a t n t a d I M ao ar l E )
lL 1T e A
- P( P( W B (
Y
- . - . ,, , , ,e
,g 3- .,
a p . .
, 'p.
.., 4 -
. f.:s
, ?:, 1 z
- h' .?
,' k .
i-[-'
, 1 .
, ,. 'i n, .>. r ;
+
<, *' 1 c .l .
'w 1 o. t
. :q ,
+ ,
1 h .-
,s. c
., o, .. [ s. ,)
7 ,*
' Tab 1E 5-6 1
-p t
i COMPARISON OF WOLF CREEK'30 FT-LB TRANSITION TEMPERATURE. 1 1
RESULTS WITH REGULATORY GUIDE l'99 REVISION ~2 PREDICTIONS i
~30'ft-lb Transition Tem 6. Shift :
F 10}gence R.G. 1.99 Rev. 2. Capuule U /
n/cm --
Material 2
(*F) (*F)
$) .<
. Plate 'R2508-3 (Trans.) ' 3.39 30 25 l Plate R2508-3 (Long.) 3.39 .30' 30 Weld Metal 3.39 23 20 '. , $
o 1
1
.q l
l 1
.j 1
l q
l l
e l'
l I
4
[
8-11 i
l l .!
~
n:
ie oa t r) -
cA5 ( 632 379 979 u 676 565 !366 d n~-- s ' -
ei -
R -
n o l
? -
~ i : -
1 lt . 347 383 835 aa) ~
975 O
t g% 662 954 on 222 '
122 ~
ToQ222 l
E n
~
L - - o mi '
A rt '555 958 8 5'. 8 IR )V oa) 132 932 Ee TM f g%
i n(
330 111 111 - 11 A no -
L M 0 Ul E
L 1 E>S eh rt ~
$ E u g) 000 101 121 E t ni V( _
ces 533 171 231 P2 m ark 555 656 666 O rt(
T c FS s C /n p e rs us) 396 632 700 7
K E 1 0 t ei crs 12n 049 134 5 E atk 693 374 599
'e Rx C F rS( 111 111 111 l 9 -
b F 3 e a L 3 r ) -
T O udp W C 000 000 500 t ai 766 080 010 R
O T cok aL( 222 323 333 r
F *F F
S 0 E 5 e I 5 th T at m g)
R T 655 565 677 E
P A i ni t es 510 150 553 O D l rk 888 8E8 899 RE T Ut(
S P A d E ID l eh L A it I
S R Y g) 210 011 944 NR %es ni 305 310 887 E I rk 665 566 677 T 2.t ( -
OS _
t p ),
smF 700 070 007 ee* 755 575 5 7 T T [, 15 5 1 5 er l e pb -
mm 123 123 123 au SN LLL TTT WWW AAA AAA AAA l
a 3 3 .
i - - .
r e8 . e8 s l e t0g t0n da t a5 n a5 a lt a l2 o l2 r ee M
PRL PRT WM .
- G r
ev . ' '
' curve 754333-A '
s 'l
- ( o C) >
1
, . -100' -50: 0. 50' :100 .'150' --200 250-i i i
' 2 :2' ' 3- 'I O t,- 100 -
1
, 3 80 - -
M - -
1 gg -
o 20 -
o -
j 0 ' -
I ' '
._ 100 i. I I 21 "
i 1 i 2. 5 -
(80
- 2. 0 60 -
o -
- 1. 5 lE
- d. E '
g 40 -
- o. -
LO -
-30F.
, g. 20 -- -
- 0. 5 -
0
' '2 ' ' ' '
0 200 280.
i i i i i i i i 180 -
240 160 2 ig Unirradiated
- a
- a
- r -
200 n
g 120 -
160
~
100
- Irradiated at 550 F ;
D 18 2 -
120 b 80 -
o8 3. 39 x 10 n/cth e
60 -
0 80 30 F b l==30 F '
40-20 0
I p' ' I I '
0
-200 -100 0 100 ' 200 -. 300 400- 500 Temperature ( F)
Figure 5-1. Charpy V-notch impact properties for Wolf Creek reactor vessel shell plate R2508-3-(longitudinal orientation) 5-13 1
L
l Curve 754331-A
( C)
-100 - 50 0 50 100 150 200 250 i i 1 I I i
i i2 21 '
l I 100 -
E 80 -
,q .,
bM
=
2 5 40 -
o 3 )
m _
l 0
i 2. 5 '
100 i i i i i i i g 2. 0 '
80 -
~ ,6 "" -
1.5liii 60 -
$ 40 - 2 -
- 1. 0 b
20 F z! 20 -
0.S
.i ' ' ' ' ' '
0 0-E i i i i i i i i 280 ,
180 - -
240 -
160 -
200 ., .
.o 140 -
l g 120 - -
160
)
h100 Unirradiated o- "
v jf 3
~
g 80 -
Irradiated at 550 F
" 3. 39 x 1018 n/cm2 80 60 - --
25*F 40 -
= 25 F -
40
' ' ' ' ' 0 ;
- 200 -100 0 100 200 300 400- 500 Temperature ( F) 1 Figure 5-2. Charpy V-notch impact properties for Wolf Creek reactor.
vessel' shell plate R2508-3 (transverse orientation) ..
5-14
curve 754332-A i
(oC)L 1
-100 .0 50- 100 150 200 ~250. ..
i i- i i '
2 2
100 - -
-h- .
- o ",
. s 80 -
e kes 60 l
2 .- l O
5 40 -
2 A -
m n- -
- 0 ' ' ' ' '
i L
l l
_M 100 i i i i i i i i 2.5 l o
\
80 --
.n -*
l 7
- 2. 0 i 60 -
o --
- 1. 5 E
- d. E a 40. -
-=
15 F-
- 1. 0 - ~
zi 20 -
- c. -
0.5 3 ' ' ' ' '
0' 0 !
200 -280 i i i i i i i i 180 -
240 160 -
_ 140 -
n g 120 -
160 Unirradiated o g;100 V j .o" ;
120 "
g 80 -
8 . Irradiated at 550 F 60 -
80 20 F 3. 39 x 1018 n/cm2 40 -
20 F -
40 20 -
o 0 I ' ' ' '
0
- 200 -100 0 100 200 300 400 500 Temperature- ( *F)
Figure 5-3. Charpy V-notch impact properties for Wolf Creek reactor'
. vessel weld metal 5-15
_________________________________________________________________:__.__ _____________-.____________1____.__
'(* C).
-200' -150 -100 -50 0: 50 100 150; 22- c 2.
100 - - cd : '-
g 80 -
!g .
m g&
.c -
y .
- d. .
0 .
100 > > > > , , 2.5 5 80 -
^
4 2,*S ^
2.0 e
- 60 - -
- 1. 5 e 55 F
$ 40 .
.- l. 0. 3 .
p 20 - -
0.5 0 0 200 , , , , . , , .
180 -
240 160 ~
Unirradiated -o
_ ig .
. 8 -
N0 .
- e .- .
di 120 -
160-g;100 -
q l b Irradiated at 550 F -
120 ~
c 80 ~
- 18 2
- 3. 39 x10 n/cm 60 -
o 65 F -
80 65* F o 40 -
o o e _
g 0
- 300 - 200 -100; O 100 -200 300 Temperature (*F) >
~
Figure 5-4. Charpy Y-notch impact properties for Wolf Creek reactor '
weld heat affected sone metal -
i 5-16 L
I
1 l
l
-- - ,. , y-,m ~ . - , . _, . . , . ry,, _,. , , , _, .
r- i c.
. T 1.
t y . s+;.- ,
x '. ; yy' ;
- k. J
-q ,
o 4
.is AL13 AL14 AL4 'L7 All
, .n. , ., , ..a,,.. .
t; AL10 AL15" " AL12. AL2 AL3' I AL9 AL6 AL8 AL5 AL11 Figure 5-5. Charpy impact specimen fracture surfaces for Wolf Creek reactor vessel shell plate R2508-3 (longitudinal orientation) i m
5-17 RM-14077
_w
1
, - - - , .. -~ - -
w-u x, n, c,m -- -
ATil AT12 AT2 AT15 -AT14 7 m p n w , ,n e xo:-;r m g:m y un, ..nv wmv,,n
' ^ '
AT8 AT3 AT13 'AT1 AT6 AT9 AT5 AT7 AT10 AT4 Figure 5-8. Charpy impact specimen fracture surfaces for Wolf Creek reactor vessel shell plate R2508-3 (transverse orientation) i j
e i
6-1s RM-14078 m_
, f ._,, , . . , , , , .
p 3 l:!
[:( ;
1 1
l y ::
j
.[ ;:
c .-
g '. . .. ..-:
m,, y. ,,,r m.y,m, ,
.1 m ,3,y ,, .< ,yn.,,,,, . ,, ,., ,,
- e. .
1 l l .
- l 1
4
- l .
- s. . . . .
A- ,'l)'
AW8. AW14 AW3 AW11 AW13 AW9 AW2 AW7 AW12 AW6 Figure 5-7. Charpy impact specimen fracture surfaces for Wolf Creek reactor vessel weld metal o
5-19 RM-14079
. . .a
- l 1
grw 7yswmegynyppea+rm engm~rm"N"" ~1 I . s
?
- .; , .A l
j . . .
- i. !
m m , ,, . . - - - ---s !
AH15 -AH9 AH10 .AH2 AH7 )
1
-4 7 -c - m . .
l AH6 AH8 AH4 AH1 AH3
. .m , c- , . , - . .
j,
. es .n,. . l
h :).
- 4~
EE}.
j 9 :
{ , :n AH11 AH5 AH13 AH14 AH12 Figure 5-8. Charpy impact specimen fracture surfaces for Wolf Creek reactor vessel weld heat affected zone metal 4
6-20 RM-14080-l
]
s curve 754329-A.
, (*C) 0 50 100 150
)
200-- 250 300 120 i i. .' I I I '- 800 1 110 -
100 - -
700 l J 90 -
Tensile Strength -
M01 7 _ -
500 m t 1
60 -
400- !
50 - - 3- l 0.'2 % Yield Strength 40 'I I T l' i i 300- 1 i Code i Open Points - Unirradiated 18 2
- Closed Points - Irradiated at 3. 9 x 10 n/cm 80 i i i
, i I
, e , Reduc 0 ion in Area r_
70 -
g _.
f% -
x s
~~
40 -
Total Elongation E
a 30 -
w -
w -
J
- ~
- -= ,
10 -
- - .1 Uniform Eloncation J 0 i i i i J
0 100 200 300 400 500 600 1 Temperature ( *F) )
L- i Figure 5-9. Tensile properties for Wolf Creek reactor. vessel shell 1 plate R2508-3 (longitudinal orientation)
{
- )
5-21 i l
- I
(. -I w_-_.-..- . .. . . ..
'. i i i t j3 . 'I' i
> 1 sf' curse 754330-A 1 *C)- e e 0 50 .100 150 200 250 -300. '
. j 120 i i -i l-i i 1 800 110 -
700:
100 -
j 90 -
6001 '
- s q Tensile Strength ,b -
n .
80
-2 500 m
60 2
K* m 2-400 ;
.50 -
d 2 % Yield:Strengt@3 ' -
I i i i- i i 300 >
40 '
Code ..
Open Points - Unirradiated 18. 2 Closed Points '- Irradiated at 3. 39 xf10 n/cm .
80 i i i- i i i i ..
70 -
e o Reduction in Area' -
R 60 24 .. s o-
? So - :-
$ 40 - -
30
- Total Elongation. -
3 -
m x ;- _2
- 7 ,
10 -
g , , .,. Uniform Elongation-0 100 200- 300 400 500' .600 l Temperature 1 *F)
' S Figure 5<-10. Tensile properties for Wolf Creek' reactor vessel'shell- -
L plate R2508-3 (transverse orientation); -
4 5-22
< c
..i.___..______ _ _ _ . _ . 2_'
7{
curve'754327-A
(?C) .
O~ 50 100- 150 200 250- 300 120 , , , , , .,- , 1 800 1 110 -
100 - -
700 q
=
- Tensile Strength j 3 90 F -
i- 600 g
- m 80 -
n- i 1
570 0 .2- 500 -
~
60 -
- 0. 2 % Yield Strength -
400 50 -
1 4g , , , , -, .,
300.-
Code Open Points - Unirradiated 18 2 Closed Points - Irradiated at 3. 39 x 10 n/cm 80 . , , , , , ,
. 70 -
. Reduction in Area -
. c o _.
60 -
R
_M -
$_. 40 t3 g 30 -
Total Elongation ~
2 20 -
a '
-f -
10
.- s -
n rm Bong @on 0 , , , ,
0 100 200 300 400 500 600 Temperature ( F)
Figure 5-11. Tensile properties for Wolf Creek reactor vessel weld
. metal 5-sa i
y ,a v% m
,-! y\.ry;ry;'s n t, - : .\ '- ,
< . '\ ' ;n N/i c 1 O .'. ;\
si j (v. . W ..
f ' '
m'eJ" si ' .' .
- q ' (.
[f jfl $ .
Ali}l. i: .A
^
w . ..
9, g
. hk u
(
'j yf
. W s
4 i.
s,
~ '
, [i ) ' '} ^
+
lyM 9r ' a.g m.-[k-l.
i j4 .
w@m$f! d,:t p@g A A
gm.. . ' ' pa 4.d
- fhk j;'
'i ,\ .
j ..~Qfdhllu.
. en n
%{l}fQlq
-Lm 3
' A .
. nm 6
nn
- 3 Specimen.All 7 77'F gi ,
y( q -
, g ,
a
&f { hj' f '. A 'Mf
~
i ,
' m *, _
% f' < ;w d:b fy%f w ? ,j% e ,
g,@%4 e f A .,
\ (, ,
- w; w
.g
,- 8
%
- A45
( ' %,m{ AIM / At . )' - SE- l p
b 4,
,"t g $& g, - ~ _
1 ! 5 ( --
, m n .?h; fh&y4k%i&$;Ikfk 51 f %q yj yy
~ v 7 g g yp" :Qrpl@6MS s *
% N ,. . .' sr-. . $ M650d$h $dkN$ U ds if!Ed$siE bG <S l s -s y e'a M
fs Specimen AL2 }150*F',
< t s% -
,, '. \
& ( ..
<z.
p'"y mn:/ p a m "%wgw v I e
'4 r)m'fiyg,yQp.ni.1"m%lqhye;f:;[fy%Q"'a is?f C q .. .
- EV M ;:- V M$ $
e
- 1 % W :,d W L.:
s s
, ma ww w' d n . ; n.t a p:f 9
M d' ~P MMd;.m@fd3Nid$y;:igk.gjg. { .2 s ,,,. ,
.;4fllQMM}if_jQQ/((QiWy ..
lh ~.&'mQl
~
l% .
g
- ,s -
- s .;:. y
- '...+p%}q
.g ?y j4 j .hlT) .W < : m[
T M ::s
'hQQjf p Q~ah W
j d,2': h
.; - ,J -$j~6 ., MQM+4 a I Op [!,d;<;; w., ,.ps g[O pm2 w,M.w, a&ig, ; nap.M .
! Specimen AL3 550*F; o ,
+
., j'Ifirn;e 5-12. Fractured tensile specimens from Wolf Orcek reactor-
, , "[ ji! vessel shell plate R2508-3 Longitudinal orientation)(
y '
g <)3
> tr ,
2
'ty , t ,I' )
i <
. yg t
{
! j 5-24 5
-/ ~! k t t .- g i 3
/k- aM 14081L ,
1\
! =,.
mm _1.m m_w_c_ 'L1 s. :0 2
e li J
ty- y !
, m QKy0 m uq!O?NfS!]aa*]s;e;'"$Of y!q:
.m .j 4 m y wfgc
/3777 7p pppgrj'up{q -* :Agj7~-
~
g iyQt;g ' Qw.;
f
- 'h', gly WWco 4 :ygS; l " 9; j0;g .p. g; pMh W);[fr,,l?'p , . ey A M.xsj e)gy:g g w xgM;y*g
, S n ? : - '-
a<
, tn
'7
,, I?
- . ywa ww .pg e p m<* 3 ,
1 x e
{ #?' w& ?: A :'f G e s a ;- '- I g[j?jk!k.{-&.33(h: jj 4 r g%ai ii%%9)h;)y d % W,Df}yh,$h.
. .l;=; y- k k $Q9qph 9 eh w' ~ a my Q,myL.Jp,m@p y%p: e.Y
)Q g.$>)Mdfd
%$979 my ON % M1, 'ig?/ 9 g ebgkS g;p% Ajj1epp;
.b t
. i 9IM' eye,Qn t:
f '
%e3M l OEhdL21}0{E: ddd n&
Specimen AT2 77'F 1 n wo w & emi ydM4(M. w n rm w Xi r.y]veQ Qh p gb ne A: y g
,Q
%"l%f@b;. * . Mbd T g;! Tk /y%Wp 412 iMM ,. 7 -j n'c a p% qpsny UK?k W W VnOW Q ; W; 9i ;
Agggggm%g., min q yL l a f p- :
.e -
g -
\
4 y.
;h.lk s %gjQg'9gg.4.c4.gv al.:~ 7.<gy $g, y yl::t;q ff p . m g g.,g g g n%,g.4gj l- p pw Q
~&
, 4 M-tN . Y-g .d4&J ,
e g@$R;; m -pNQ fQQ)g%gg' y%k;.h$j; w, .n
. g e ,<..~ .Q w.~ a nn m,w.. -,
Specimen AT3 150*F ' 7L
' #.. :y 4 @/. +\ b.s (I S y, 11(
- k. <q;
.i w;, y'; ..i ..,.3 : ! yw ,!S/MSt SY, w~ '
T']'d./
- i. .
, y; ;r,y^
.; s s ; :m 53n ; xv J dyla[p M.MrS"bddd$bUdb ^
4 1
, 9
' l I f '. k,.
M 7
'1
i ,
- w. g.t ,. >, .,
, . ,y..%!.lin.Q' 7 .g Specimen AT1 550*F Figure 5-13. Fractured tensile specimens from Wolf Creek reactor vessel shell plate R2508-3 (transverse orientation) t 5-25 RM-14082
'l f . , > r I a pg&:.J?
v; u ,1w$bdpsyWe 7 a
$fi 8gdC;THW;yFjdk%
3 "- P9W D $ $ n- 'i m .
;op -}
%mkNO
+
j.
.x O, MQ w
.g!,?ist A. rv : f@f$
;. aarww.; g. ..-c.r t
37 W c.<ht a %$ h c g ik +3p ' i . Mi.m g &p t ??E ,; l ,
;MGhM; olA d s'as%m VW f.fi.f_$k.mJ fifhO *pM u%(a$ff0f %
$iN'7re,wh0m,"g*
44M14! e: Jj MJn n,W a , < mw v. 34%gNOtk.c MW%#qm%QL%d%E,a%%kvi % Specimen AW2/ .0'F y ., w ref m y n c g ggp n g :p um g gr
$w;!s;: .%~. 91r w ) M h lzze & vs 4yy
~ >
^MyfWrannjk}fhltyA %gbf f.hc;' %.e
,y@g" f n
gpn 9% % , .
,/
g I a,
.l f !.?:l
,' .m
- g. n n yp.mg% p q np v;w; ~n;.m n,,
Mi
'y %?y$p.y ij ulg&g_ meg gn}%
cads $h:Yll@@IW ddac,$t !Ss M5MNM ' Specimen AW3 77'F . n g ,s c
- v. .
p:w7g . , a;2 j m?,;
' Qy y* ,, < - ) [f'. : ', 4
,f.
, p"J.q-g'qf.q.
{'t s-
?."' '
,. ':.n i i.
~r
, if . . , ,, m e H. /,s. y
).g',,
.,..._.,.p.
pik,
- 1.? A , ;/ ' t ( ;q,I 7
o. dddsd@,d.4'h"jp'Ejf.iQII% f@%@$ 7h s' : ,
'a;.
/ ,
a m y
.}, ,
n'
> ; la ' -f ;
63
~
4 <
,q., ,
s
,.,I, , ./ 0 $ , ,
. f, ' : %ew%tn Specimen AWi- 550*F Figure 5-14. Fractured tensile specimens from Wolf Creek reactor vessel weld metal 5-26 RM-14083
l
}
I
~1
~k
.i j
i Curve 754326-A' -l 120 , i 7 , i 100 80 --
-j r )
4 m . < a - e@ G' 1 40 - Specimen AL3 o 20 - ( 550*F) - I 0 I I i ! 0
~
0.05 0.10 0.15 0.20- 0.25 1 Strain, in/in J Figure 5-15. Typical stress-strain curve for tension specimens l 5-27
a i
=
i 4 4 l 1
- 6. RADIATION ANALYSIS AND NEUTRON DOSIMETRY ;
'l l
6.1 INTRODUCTION
l i Knowledge of the neutron environment within the reactor pressure " vessel and surveillance capsule geometry is required as an integral part of LWR reactor pressure vessel surveillance programs for two reasons. 3 First, in order to interpret the-neutron radiation-induced material l 1 property changes observed.in the test. specimens, the neutron environment' l (energy spectrum, flux, fluence) to which the . test specimens were . 5 exposed must be known. Second, in order to relate the changes observed j in the test specimens to the present and future condition of the reactor . I vessel, a relationship must be established between the neutron 'i environment at various positions within the reactor-vessel and that ' experienced by the test specimens.. The former requirement is normally met by employing a combination of rigorous analytical techniques.and .j
~
measurements obtained with passive neutron flux monitors contained in each of the surveillance capsules. The'latter information is derived l' solely from analysis. l This section describes a discrete ordinates 8, transport-analysis performed for the Wolf Creek reactor to determine the fast (E > 1.0 MeV) neutron flux and fluence as well as the neutron energy j spectra within the reactor vessel and surveillance capsules. The analysis data' were then used to develop lead factors for use in relating neutron exposure of the reactor vessel to that of the surveillance capsules. Based on the use of spectrum-averaged reaction cross sections derived from this calculation and the Wolf Creek power history, tho= analysis of the neutron dosimetry contained in Capsule U is presented. l e-1
, i 4
6.2 DISCRETB ORDINATES ANALYSIS A plan view of the Wolf Creek geometry at the core midf lane is shown in Figure 4-1. Six irradiation capsules attached to the neutron pad are included-in the reactor design to constitute the reactor vessel surveillance program. The capsules are located at 58.5, 61.0, 121.5,, 238.5, 241.0, and 301.5 degrees as shown in Figure 4-1. A plan. view of a dual surveillance capsule holder attached to the neutron pad is shown in Figure 8-1. The-stainless steel specimen containers are 1.182 by 1-inch and approximately 56 inches in_ height. The containers are positioned axially such that'the specimens are centered on the core midplane, thus spanning the central 5 feet of the-12-foot-high reactor core. From a neutron transport standpoint, the surveillance capsule structures are significant. They have.a marked effect on both the distribution of neutron flux and the neutron energy spectrum in the' water annulus between the neutron pad and the reactor vessel.- In' order to properly determine the neutron environment at the. test specimen . locations, the capsules themselves must be included in the analytical. model. This requires at least a two-dimensional calculation. . In the analysis of the neutron environment within the Wolf Creek reactor geometry, two sets of transport calculations,were carried out. The first, a single computation in the conventional forward mode, was used primarily to obtain spectrum-averaged reaction cross sections and radial gradient information for the pressure vessel. The second set of calculations consisted of a series of adjoint analyses relating the fast neutron (E > 1.0 MeV) flux at the surveillsnee capsule location and selected location on the reactor vessel inner wall to the power distributions in the reactor core. These adjoint importance functions, j when combined with cycle-specific core power distributions, yield plant-specific fast neutron exposure at the. surveillance capsule and j pressure vessel locations for each operating fuel cycle. Both the ' l forward and adjoint calculations used.an S8 angular quadrature.
?
6-2
,. g
- ~-
o - 4 wy
- f. , JL .
+
5; i ( ,
, f
# s Theforwardtransport:calculttionwas.carriediout.in;R,16:
geometry using'the DOT two' dimensional-discrete 1ordinat'se code (4) andI :i
'the SAILOR cross-section library. (5) LThe SAILOR library. is 'a 47 ' group, 1, ENDF-BIY based data set produced specifically for; light wat'er' reactor:
~
.i
'~
applications.,AnisotropicscatteringistreatedwithaPgexpansionof." the cross-sections. The energyJgroup structureLused in the analysis is listed .in T:,ble ' 6-1. The design basis' core power. distribution utilised in the forward' , analysis was derived from statistical-studies of long-term. operation of Westinghouse 4-loop plants. Inherent in the development of this design' i f basis' core power distribution is the use of'aa.out-in fuel management strategy; i.e., fresh fuel on the core periphery.- Furthermore, for the. l peripheral fuel assemblies, a 2a ' uncertainty derived from the- U l statistical' evaluation of. plant to plant and cycle to cycle' variations-in peripheral power was used.--'Since it is unlikely that a single-reactor would have'a power distribution at the nominal 42a level for a large number of fuel cycles,-the use of~this design' basis distribut' ion 1 is expected to yield somewhat conserve.tive results. This is.especially. J true in cases where low leakage fuel management'has been employed. l The adjoint analyses were also carried out using the P ' cross 3 q l section approximation from the SAILOR library. Adjoint source locations j l were chosen at the' center of each of the surveillance capsules as'well as at positions along the inner diameter of the-pressure v'essel. Again, j these calculations were run in'R, 8 geometry to provide power distribution importance functions for the' exposure parameters of
]
interest. Having the adjoint importance functions and appropriate core 3 power distributions, the response of interest is calculated'as a 1 L , , ,
. Lj R
R
- 0 *
- R' 0' E I(R,B,E) F(R,0,E)- dERdRd6 (6-1)' 1 j
e-s - H .
, 2: i * , i f, ( , _, f 1
. ; ( . .,
s , ,
, a-7 g .
~
. i
. p-
. ' ; \
--where: i Tgg; = Response -of.-interest.-(e.g.', f.(E > 1.0 MeV)) at q
' radius R:and'asinuthal angleLd.: j I(R,#,5) = Adjoint !importanceL function at radius R. an'd ;
< asinuthal' angle'8 for. neutron energy group E. ;
F(R,8,E) = Full power = fission density;st radius 3 an'd Y asinuthal angle 6'for' neutron' energy l group.E..
- The fission density distributions used. reflect the burnup-dependent-
- inyantory of. fissioning actinides, including U-235, U-238, Pu-239,..
Pu-240 and Pu-241.1
'. Core' power distrib'utions for use'.in the plant specific evaluation for- Wolf Creek,' cycle 'l were obtained' from WCAP-10483. (6)-
The data extracted .fron reference.6 represented cycle averaged relative; assembly powers. Therefore, the adjoint'results applicable'to capsule U-represent the neutron flux averaged'over cycle,1.which'when'aultiplied-by the cycle length yields the incremental fast neutron fluence for that cycle. .
'The transport methodology,.both forward and' adjoint, using th'e SAILOR cross-section library has been~ benchmarked against the-Dak Ridge. -
National' Laboratory - (ORNL) .Poolside Critical Assembly .(PCA) f acility ' as well as'against the Westinghouse power reactor surveillance capsule' data L l base.(7) The benchmarking studies indicate that the~use of'SAILDR cross-sections ~mnd generic design basis power distributions produce flux-levels that tend to be conservative by 7.!to_22%...-When plant specific power distributions are used with.the adjoint;importance1 functions,'the benchmarking studies show that fluence predictions'are within *15% of measured values at surveillance capsulejlocations. .The analysis is consistent with established-ASTM standards.(8-12) 6.3 RADIOMETRIC MONITORS The passive radiometric monitors included in the Wolf Creek
- surveillance program'are listed in Table 6-2. The first'five reactions' F ..
l-l.
~6-4. '
e -_ ._h hh.L.su EU- Wm-_411- " - ' o a 1--2.
i i i 1 i in Table 6-2 are used as fast neutron monitors to determine neutron- 4 fluence' (E > 1.0 MeV) . Exposure as measured by fluence (E > 1.0 MeV)'is the primary parameter currently used to correlate. measured material- ! l property changes. However,-the data provided by the five' monitors D sufficiently covers the neutron energy range so that other parameters proposed for damage evaluation (such as displacement per atom, .dpa) may -l be determined with an acceptable degree of uncertainty. In addition, bare and cadmium shielded cobalt-aluminum monitors are included in order' to provido a measure of the thermal and resonance region neutron. fluxes; l at the monitor location. These latter neutron energy regions are not d currently thought to make any significant contribution'to irradiation damage in steel, but may be used for evaluations'in the future. In addition, the thermal and resonance neutron measurements al3ow an ' I evaluation of burn-in and burn-out, and impurity effects on the fast ! 1 neutron reaction measurements. The relative locations of the various radiometric monitors j
. within the surveillance capsule are shown in Figure 4-2. The iron, nickel, copper, and cobalt-aluminum monitors,'in wire form, are placed j in holes drilled in spacers at several axial levels within the capsules.
The cadmium-shielded neptunium and urauium fission monitors are. accommodated within the dosimeter block located near the axial center of the capsule. All monitors are located radially at the center of the capsule and azimuthally within *0.23 degrees of the capsule center. l The use of passive monitors such as those'licted in Table 6-2 does not yield a direct measure of the energy-depeadent neutron flux level at the point of interest. Rather, the activation or fission process is a measure of the integrated effect that the time- and energy-dependent neutron flux has on the target material over the course of the irradiation period. An accurate assessment of the average neutron flux level incident on the various monitors may be derived from the activation measurements only if the irradiation parameters are well known. In particular, the following variables are.important: e-5
' 'a-'- - ' ' ' ' - ' -' ' '
p ,
- n. ?
4- 4 i 7 6 f
,4.'. , .
g<
- lj ,-
g
, e
- The operating' history of ths reactor-
~
- The. energy response of'the' monitor-y 1*: TheLneutron energy spectrum at;.the:nonitor location' SThe~ physical characteristics of the monitdr; .
The analysis of-the passive-monitors and the subsequent.
. derivation ~of.the'a7erage neutron !l f 'ux requ re two operat ons.' First,;
i i , the disintegration rateLof prSduct'nuclide per unit mass'of aonitor,mustD be-determined._Second,.in.cedertodefinea' suitable!spectrun-averagedI '
-reaction cross section, the. neutron energy lspectrus'at,the ecaitor:
location must be calculated. The' specific activity'of~each monitors is determined.unies., established ASTM procedures. (13-21)1 poggo,1,,. sample preparation'J the' , U activity of'each monitor is determined by:means:of.a lithium-drifted-
~
germanium, Ge(Li),; gaana ray. spectrometer. - The overall' standard: l y
-deviation of the measured dats is a fOsction of'the precision 1of:sampleu
~
weighing, the. uncertainty in counting, and the acceptable error'in J.. detector calibration.' For the: samples removed'from Wolf Cre6k,' the; overall 2a deviation'in the' measured data'is_ddt'erained toLbe plusor;. " minus 10 percent. The. neutron energy spectrum at the monithr location is determined analytically using the1aethod dsperibe'd !in paragr.aph 6-2 9 .. Having the-measured activity;of the monitors and the neutron 1 energy spectrum at the monitor locations of interest,sthe calculation'of'
.. i
'the neutron flux proceeds'as.follows. Thereactionproductlscactivity.'in. ,
the monitor is expressed as
'P y Ag = N,gFY 1gy p
at (E) p) dB p 1-e-A g t[ : e -Aggt -
; (6-2)'
E J=1 nax
- c. .
I s-o
. j
, . . . . . . . .. . _ . . i . . .
.<l ;
. where A-g = induced product: activity (dps 'per gram)..
,- L N,1 -= number of target element' atoms'per gram F{ = weight fraction of the target nuclide~in the: target material- I Y.g .= number'of' product atoms produced per reaction ]
ay(E) = energy dependent reaction cross section ') p(B) . = energy dependent neutron' flux at the monitor -location-with the reactor at full' (reference) power-P = average core power level during irradiation period'j- ' 3 -] P, = maximum or. reference core power level. q Ag = decay constant ofLthe product nuclide1 ' t) = length of irradiation period'j 1 t d = decay time following irradiation period j "n = total number.of irradiation periods '; a l This equation'may be simplified by defining R;, the reactions per second, with reaction i occurring at' full power. E is t sometimes- i referred to as the saturated activity or reaction rate. I J A I l" '
.(64)
N,gggfp FY P3 j=1 max
,1-e -A g
t) .e -Ag t) Substituting in Equation 6-2,. R g=, ag@) p @) dB (6-4) It should be noced that an assumption 'is made in the above equations thtt the ratio of p(E) to measured power.-(P )g does ~not vary with' time. This assumption is accurate for cases where' fresh fuel is loaded on the periphery and no major, fuel management changes-are made. In other cases, corrections to the shorter lived reaction products must be made and in' general, uncertainties in fluence derived from these products-must be increased.
.e-r '___/___._._.______ '
C5
=. I _1 , .
Two methods were used to derive the neutron flux based on Equation 6-4. The'first of these involves an' average based on each -
., reaction measurement had is described next. The second method involves a fit to the data using a least' squares tiochnique to minimize the uncertainty. The least squares method is described af terwards.
Because the neutron flux distributions are calculated using .
. multigroup transport. methods and because the main interest is in the f ast (E > 1.0 MeV) neutron flux, spectrum-averaged reaction cross sections are defined such that the integral term in Equation 6-4 is replaced by the following relation. ,
a(E) p(E) dB = i pf where M 47 a(E) p(E) dB 2a sK 2*1 F=# = 18 . p(E)dE 1p 8
*IMeV g=1
'. ', # 18 f= f p(E)dB = .2 p
*1MeV g=1 8
'. g = groups number from Table 6-1.
Thus, Equation 6-4 is rewritten , l 4 21=8ff
.i or, solving for the f ast (E > 1.0 MeV) neutron flux, ,
. E .
f=g- f (6-5) 0
, 6-8
i .
'a o
e
- . i The total fast. (E >- 1.0 MeV) neutron fluences is then given by i 1
. ;l i
n P. ] ff = ff 2 p' t j (6-6) ! 5=1 max where l
-i n P 2p max j=1 t) = total effective full ~ power seconds of l reactor _ operation up to the time of capsule i removal 1 An assessment of the potential for product nuclide burnout may be a.ule using the bare and cadmium shielded cobalt acasured activities and published data for the 2200 m/s absorption cross-section and the resonance integral. 'This is done by rewriting Equation'6-2 in terms of a monitor 2200 m/s neutron flux and a monitor resonance flux as follows:
R bare " f2200 f2200 + RI f,,, (6-7) Red * ,RI pres (6-8) l where 59 Rb m = bare Co reaction rate R ed
= cadmium shielded 59 Co reaction rate a 2200 = Published 2200 m/s absorption cross-section for 60 Co gy = published epicadmium dilute resonance integral for BOCo p2200 = m nitor 2200 m/s neutron flux to be. determined from measured activities p,,, = monitor resonance neutron flux to be determined from measured activities 7 - .,
8-9
- - _ _ . - _ - _ _ - - _ _ _ _ _ _ _ . - - . _ _ _ _ _ _ _____-___-_______a_'_
j j l
- Equations.'6-7 and 6-8 are solved for p2200 and p,,, using the average measured bare and cadmium shielded cobalt activities at the .
monitor loct, tion.
-The total loss rate of a product nuclide'say.then be ' expressed q as the sum of its radioactive decay rate and the neutron absorption rate--
in that nuclide while the. reactor is at power. The product nuclide ; neutron absorption rate may be estimated from the published: data for a 2200 and RI.and the monitor fluxes determined ~above. If the neutron-absorption rate is small when cospared to the decay rate then there is no concern regarding burnout. , Similar corrections can be determined for other reactions affecting the measured results. Of most significance are corrections 239 for 235U fission.in thel 238U dosimeter and Pu fission after capture of neutrons in U. The latter two step reaction only,becomes important for very long irradiations.
~
The least squares analysis technique is performed-with the FERRET (22) code. _ This code employs ~ a log normal least-squares' algorithm ,
, 3 which weights all the calculated fluxes, group cross' sections,'and measured values with assigned uncertainties-and correlations. ~All the .
quantities in the series of Equation 6-4 are simultaneously adjusted to give the minimum weighted least squares error. The lognormal approach: ! automatically accounts for the physical constraint of positive fluxes, j even with large assigned-uncertainties. { For the analysis of dosimetry data, the continuous quantities ] (i.e., fluxes and cross sections) were approximated in 53 groups. The i calculated 47 group flux-spectrum was expanded and contracted into the 53 group structure using the SAND II code. (23)- This procedure is carried out by first expanding the spectrum into the SAND II 620-group structure using a SPLINE interpolation procedure for interpolation.in regions where group boundaries do not coincide. The high end of the spectrum is extrapolated using a fission spectrum form. The 620-point . spectrum is then easily collapsed to the 53 groups. l s-to - _ = -
, u. }
.1 1
'1
;;d }
l 1 lThe cross sections were also. collapsed into the:53 energy-group c
*~
structure:using SAND II.with the calculated spectrum;(as-expande'd to 820-
. groups): ss the weighting function. .The cross' sections were.taken from .1 1
the ENDF/B-V dosimetry file. Uncertainty estimates and'53 x 53' 'J covariance matrices were constructed for each cross'section. ,
.l Correlations between cross' sections were neglected'due to data and code limitations,Lbut are expected to be unimportant.- )
Foreachset.ofdataorapriorivalues,theinverseofLthe , I- corresponding relative covariance matrix M is used as-a statistical 1 L weight. In some cases,as for the cross' sections, a multigroup covariance matrix is:used. For the fluxes, a simple'parameterized form .- is'used: I l Mgg, = i
+Rg R,p g gg, ,
, where NR 8pecifies an overall fractional normalization uncertainty {
(i.e., complete correlation) for the corresponding set of values. The
., fractional uncertainties Rgspecify additional random uncertainties for group g that are correlated with a correlation matrix.'.
l g 'l p g, = -(1 - 8) 6g , + 8 eXp . The first term specifies purely random uncertainties while the second term describes short-range correlations over a range 7 (S i specifies the strength of the latter term). l '. For the a priori calculated fluxes, a short-range correlation of 7 = 6 groups waly used. This choice implies that neighboring groups are strongly correlated when S is close to'1.. Strong long-range
;,- correlations' (or anticorrelations) are justified based on detailed I studies carried out at ORNL.(24) For'the. integral reaction rate q
, covariances, simple normalization and random uncertainties were combined' L s-11
\
i. as deduced from experimental uncertainties. Specific assignments of uncertainty vslues were made as follows: Flux normalization uncertainty 30%
. Flux group uncertainty ~ 30%. 03 > 0.1 MeV)
Short range correlation fraction 0 0.9 Reaction rate uncertainties See Table 6-7' The marnitude of the uncertainty assignments are based on general agreement of calculation and measurements for similar cases. Since the uncertainties assigned to the fluxes are in general larger than those for the measurements, the solution is little affected by the specific values assumed. 6.4 NBUTRON TRANSPORT ANALYSIS RBSULTS Results of the discrete ordinates transport calculations for the Wolf Creek reactor are summarized in this section. In Figure 6-2, the . calculated maximum f ast' (E > 1.0 MeV) neutron fluz levels at the radius of the surveillance capsule center, the reactor vessel inner radius, the
- reactor vessel 1/4 thickness location, and the reactor vessel 3/4 thickness location are presented as m' function of asimuthal angle. In Figure 6-3, the radial distribution of maximum f ast (E > 1.0 MeV) neutron flux through the thickness of the reactor vessel is shown. The relative axial variation of f ast neutron flux within the reactor vessel is given in Figure 6-4. Absolute axial variations of fast neutron flux may be obtained by multiplying the levels given in Figure 6-2 or 6-3 by the appropriate values from Figure 6-4. Table 6-3 provides the calculated fast neutron exposure parameters for the Wolf Creek reactor vessel.
Table 6-4 provides the calculated f ast neutron exposure parameters and updated lead f actors for all of the Wolf Creek
- 1 e-12
y - - 43 (' m 6 surveillance capsules. The. lead factor is defined-as'the ratio of thei fast (E > 1.0 MeV) neutron flux at .the dosimeter _ block location (capsule center)~ to.the maximum fast neutron flux st the resctor vessel innerz l radius. In order to derive neutron flux and fluence levels'from the.
. measured' disintegration rates, suitable spectrun-averaged reactionicross:
1
' sections are required. The calculated neutron energy spectrum at the'
' center of the Wolf Creek surveillance Capsule U is listed in Table 6-5.
6.5 DOSIMBTRY RBSULTS The irradiation history of the Wolf Creek reactor up:to the time-of removal of Capsule U is listed in' Table 6-6. These values were used^ l; to calculate the reaction rates according to Equation 6-3. .All the-measured activities'and reaction rate results are presented-in Table 6-7. Neutron fluxes derived using the spectrum averaged cross' sections.are listed in Table 6-8. A value for the' fast neutron: flux-'is - ]
~1 derived using the five fast neutron monitors.- Values for' resonance and; 'I 2200 m/s thermal fluxes are derived based on the bare and cadmium covered cobalt' measurements. The thermal and resonance. fluxes were used to estimate absorption rates in the ndioactive product nuclides. Burn- -1 out of these nuclides was found to have less than a 0.li effect for all'. :l 8
except the pg(,,p)58Co reaction where the correction is about 0.3%.- Results from the FERRET calculation are given in Tables.6-9c i through.6-11. Table 6-9 presents the a priori _and adjusted neutron spectra in the 53 group structure. Adjusted spectral uncertainties are also given. Table 6-10 presents the ratios of calculated to, measured- ; reaction rates.for all the reactions used both for the a priori spectrum! I and the adjusted spectrum. All the reaction rates' agree with ! calculation within the uncertainty. Derived values for integral fluxes- , ! and dpa are presented in Table 6-11. The uncertainty in the final-results are also given as derived by FERRET from the input assumed i
?
b
'6-18 .')
3 ,7 ,- - .
+ '
- n. ,,
7
.m ,
4
,i
(- , 1
, p; ' ,
1
; t $, . , . . ,
- uncertainties and correlations. .The uncertainties are..therefore.
.estimatesoftheactualuncertainty;inthefluxanMfluence'results,fand( r.-
may-be used:for guidance in setting' upper limit values.> 2xcellent; agreement was'obtained between the~two flux evaluations ; methods. This follows fror the consistency of the m priori, spectral'
^
t shape with the nereurements. 'Minot' differences between the two measured ~ flux' evaluations.are due partly to the'different weighting'given thel
~
rasction'ratesandthe.'diffegentgroup-struct'ures'used.:
' Based on the analysis;;an average value'of the two fluxL
~
2~ evaluation methods was chosent This value -Lis,- 1.00 'x 1011'(n/da-sec,. E.) 1.0 MeV) with an uncertainty of 8% as ~ evaluated'by' FERRET. This' ' valuemaybecompared..with.thefdesignbasiscalculatedfluxvalue'of: 11 2 1.20 x'10 - n/cm -sec i aEd a plant specific calculated flux:value for the Wolf Creek, Cycle'1 fuel' loading which is 9'.50 x-1010 ; 2 -sec. The. measured value is in excellent'agreementivith the latter value. A summary of measured and calculated current, fast (E >.-1.0 MeV).
. neutron exposures ~for.: Capsule U'is. presented in Table 6-12. The' - ,.
corresponding end-of-life ?(BOL) fluences. are miso presented? > Based ~on the data given in this analysis,- the bsstiestimate f astf(E f l.0 MeV); , neutron exposure of the Wolf Creek Capsule,U.is: -
# = 3.39 x 1018.,7 2 (E >'.1.0 MeV) at 1.08 EFPY '
4
...l:<
e-n
1 t i d Table 6-1 SAILOR 47 NEUTRON ENERUY GROUP STRUCTURE - Group Group Energy Lower Energy Energy Lower Energy ]
.)
Croup (MeV)' Group- (MeV) j 1 14.19(*)- 25 0.183 2 12.21 26 0.111-3 -10.00 27 ' O.0674 j
'4 8.61 28 0.0409 1 5 7.41 29 0.0318 i 6 6.07 30 - 0.0261 7 4.97 31 0.0242 8 3.68. 32 0.0219 9 3.01 33 0.0150 l 10 2.73 34 7.10 x 10 -3 11 2.47 3 35 3.36x:107 1.59 x 10 -3 12 2.37 36. ,
13 2.35- 37 4.54 x 10~4 2.14 x'10-4 14 2.23 38 15 1.92 39 ' 1.01 x 10-4 i 16 1.65 '40 . 3.73 x 10 -5 17 1.35 41 '1.07 x 10-5 18 1.00 42 5.04 x 10 -6 19 0.821 '43 1.86 x 10-6 20 0.743 44 8.76.x 10-I 21 0.608 45 4.14 x 10-7 22 0.498- 46' 1.00 x 10-7 23 0.369 47 0.00 24 0.298 (*)The upper energy of group.1 is 17.33 MeV. 1 6-15
f' Tab 1'e ' 6 ~ NUCLEAR CONSTANTS FOR RALIOETRIC MONITORS CONTAINED IN TE WOLF CREEK SURVEILLANCE CAPSULES Reaction Target- Fission Yield
.of: Weight Product Fraction LHalf-life (O Monitor Materials ' Interest Iron wire' F;54(,,p)y,54-
. 0.058 4 312.2 dyf <
Nicke1 wire NiS8' (n,p) Co58. 0.6827- 70.91'dy Copper wire CuS3 (n,a) Co 60 0.'6917- 5.272'yr 38 (n,f) Cs 137. - 1.0 J30.17 yr 6.0' Uranium-238(") in U38 0 - i Neptunium-237(*)'in Np0 Np237(,,fy g,137 1.0- 30.17 yr 6.5 2 Cobalt-aluminum (*) wire CoS9 (n,7) Co 60 0.0015- 5.270 yr 60 - 5.272 yr Cobalt-aluminum wire CoS9 (n,7) Co .O'.0015 (*)Denctes that the acaitor is cadmium-chielded. 9 j l 6-16
? ?. -
-) .
j l Table'6-3 , m - L-CALCULATED FAST NEUIRON EXPOSURE PAPAETERS FOR-TE-PEAK. " LOCATION OF TE WOLF CREEK REACTOR VESSEL Iron-
. Radial Location- FastNeugronFlux Displacement ,.j within the- . .
(n/cm -sec) .. , . Rat;e ' Reactor Vessel'
.(E > 1.0 MeV) -(E > 0.1'MeV) (dpa/sec)- !J 3.14 x 1L,10 ' 10 !
, Inner Surface 7.61 x 10
' 4.99 x 10-11 (R = 86.500 inches) 1/4 Thickness 1.73 x 10 10 6.70 x 10 10 - 3,25 x 10-11 (R = 88.673 inches)-
i 1/2 Thickness 7.82 x 10 9 4.72 x 10 10 1.92'x 10 -11 (R = 90.846 inches)L > 3/4 Thickness 3.27 x 109 2.98 x'10 10 1.09 x 10-11 (R = 93.020 inches) Outer Surface 1.35'x 10 9 1.50 x'10 10 5.24 x 10 -12
.(R = 95,193 inches)
(*)The peak is located at 26.5 degrees azimuthally and on the. core midplane. The peak exposure occurs:in octants with a 12.5* neutron. pad and is calculated using design basis power distributions. [ .; [ l 6-17
Table 6-4 CALCULATED FAST NEUTRON EXPOSURE PARAMETFJS AND LEAD FACTORS FOR THE WOLF CREEK SURVEILLANCE CAPSULES Iron Asimuthg) Fast Negtron Flux Displacement Capsule. Location (n/cm -sec) Rate Lead I.D. (Degrees) (E > 1.0 WeV) (E>0.1WeV),-- '(dpa/see) ' Factor (b) U 58.5 1.20 x 10 11 5.51'x 10 11 2.27 x 10 -10 3.85 V 81 1.13 x 10 11 5.14 x 10 11 2.12 x 10-10 3.65-10 W 121.5 1.20 x 10 11 5.51 x 10 11 2.27 x 10 3.85 X 238.5 1.20 x 10 11 5.51 x 10 11 2.27 x 10 -10 3.85 Y 241 1.13 x 10 11 '5.41 x 10 11 2.12 x 10 10- 3.65 301.5 1.20 x 10 11 :5.51 x 10 11 2.27 x 10 -10 Z 3.85 (*)The radina of the surveillance capsule center is 81.625 inches. (b)The lead factor is the ratio of the fast (E >~ 1.0 MeV) neutron - flux at the center of the' surveillance capsule to that at the peak location on the re:.etor. vessel inner sur.f ace. 4 j
)
.l U-18
l 1 1 q Table 6-5
~
CALCULATED (") NEUTRON ENERGY SPECTRUM AT THE CENTER OF WOLF CREEK SURVEILLANCE CAPSULE..U ,
\
' Energy Group
-NeutrgaFlux Energy Neutrga Flux (n/cm_-see) . Group .(n/cm -see) 7 1 ~2.25 x 10 . 25. 6.77.x 10 10 3 2' 8.25 x 10 7 26 7.03 x 10 10 8
l 3 2.85 x 10 27 5.66'x 10 10 l 4 5.17 x 10 8 28 3.97 x'10 10 5 8.55 x 10 8 29 1.20x1040 ; 1.88 x 10 0 6 '30 6.44 x 10 0 7 2.57 x 10 9 31 1.74 x 10 10 ] 8 5.20 x 10 9 32 1.11 x 10 10 j 9 4.73 x 109 33 '2.01 x-1010. .] 10 3.96 x 10 9' 34 2.96 x 10 10 11 4.75 x 10 9 35 5.01 x 10 10 j 2.37 x 100 4.51 x 10 10 12 36 L 13 7.30 x 10 8 37 6.30 x 10 10 ' 14 3.67 x 10 9 38 '3.43 x 10 10 15 9.o4 x 10 9 39 3,81 x 10 10 1.33 x 10 10 1-16 10 40 5.15 x 10 17 2.07 x 10 10 41 6.08 x 10 10 18 4.61 x 10 10 42 3.37 x 10 10-19 3.52 x'10 10 43 3.85 x 10 10 20 1.67 x 10 10 44 2.39 x 10 10 21 6.07 x 10 10 45 1.86 x 10 10 22 4.75 x 10 10 46 2.63 x 10 10 l 23 5.94 x 10 10 47 3.27 x 10 10 24 5.75 x 10 10 (*) Design basis power distribution. 6-19 L-____ ,
p. 4
-4.
a
~ Table 626 ,
^
IRRADIATION HISTORY OF WOLF. CREEK SURVEILLANCE CAPSULE U. / P .P Irradiation Time Decay Time-3 aax P.3j P,,,
- Month '-Year.- (Wt) (Wt) gg,y,) (p,y,)
6-' 1985 531 '3411' O.156 , 28- 582'
- 31. 5511 7' 1985. 1379- 3411'- 0.404 8 1985 2209 3411s 0.648 31 520 9 1986 .2851; '3411' O.836 30 -- 490 10 1985 2805 =3411 -0.822 Sli -459 11 1988 3287 3411 0. 964 .~ '30 .429-12L 1985 3184 .3411' O.933: 31. 398 1 1986 3334 3411 0.977- 31' - 367.'
2 1986. 2985_ 3411 ,0.875 28 339 3 1986 3378- 3411 0.990. 31 -308-4- 1986 1296- -3411" 0.380 30 278 5 1986 3147 3411 0.923 31' -247 6 '1986 2319' 3411 0.680 30 .217 , . 7- 1986 '2971 3411 0.871 31 186. 8 1986 3279 3411 'O.961 31' .155. , -9 1986 3343- 3411 0.980 30 125-10 1986 ~3388 3411 0.993: 15 110 (1) Decay time is referenced to 2/2/87. (2) Total irradiation time. is 3.39 x 107 effective full power seconds (EFPS) .or 1.08 effective full power years -(EFPY). (3)Pg is the. average core power during the irradiatio~n period.
. l l
q 6-20
u a i i i c -) J
-1 Table 6-7 -j ~
MEASURED RADIOWETRIO MONITOR ACTIVITIES AND REACTION RATES'
-FOR WOLF CREEK SURVEILLANCE CAPSULE U Monitor and Activity Reaction Rate Uncertainty (*) l Axial Location dis /see-ag reactions /see-atom %
238 ggo,f)1370s middle 1.43 x 102 3.81 x 10-14' l correchd() 1.22 x 102 3.31 x 10~14 15
- l. 237No(n.f)1370s
! middle 1.24 x 103 3.11 x 10 -13 10 5y,go,,)54,. g 1.510 x 10 3 top 5.60x10}5 middle 1.500 x 10 3 5.57 x 10 -15 botton 1.800 x 10 3 6.68 x 10-15 average - 5.59 x 10-15 -5: 58gggo,,)58Co top 1.64 x 10 4 7.74 x 10-15 middle 1.61 x 104 7.60 x 10-15 botton 1.76 x 104 8.30 x 10 -15
-average -
7.88 x 10-15 8 f l 630u(n,a)60Co top 4.44 x 10 1 5.38 x 10-1 ! middle 4.40 x 10 1 5.33 x 10 ~17 botton 4.75 x 10 1 5.76 x 10 -17 average - 5.49 x 10 -17 5 l . [. l 6-21 m_ _ _ .. . .. .. . . . . . , , .
e ' Table 6-7 (cont'd.) ' Monitor.'and .. Activity Reaction Rate Uncertainty (") Axial Location- dis /see-ma reactions /see-stom % 59Co(n,y60Co(Cd) top 5.27 x 103 2.75 x 10 -12 middle 5.14 x 10 3 2.68 x'10-12 bottom 4.89 x;10 3 2.55 x 10 average - 2.66 x 10-12 10
.59Co (n , M60Co(bare)'
top 1.04 x 10 4 5.33 x 10 -12 middle 1.00 x 10 4 5.12 x 10-12 botton 1.01 x 10 4 5.18 x 10-12 average - 5.21 x'10 10 (*) Uncertainties are based timated absolute radiometric counting uncertainties except (1) 33 uncertainty is increased due to , uncertainties in corrections .for other fissions in the monitor and rariations in other surveillance capsule results, ~(2) Ni uncertaintyis.ggeressedduetotime.historI0ncertaintiesbecause. u . , of the shorter Co half-life, and (3) the Co reactions.have-increased uncertainty due to greater spatial. Variations of the-low energy neutron flux. (b) Corrected by 15% for 235 U fissir.n and other effects. 1. i
=
l 6-22
f ., 1 m
] 1
,a l
Table 6-8 .'
':. ;)
.RESULTS OF NEUTRON DOSIMETRY FOR WOLF CREEK SURVEILLANCE CAPSULE U !
2 Reaction ' Reaction' Rate a (barns)(*): ~ Flux (n/cm -see) FAST NEUTRONS. (E ) 1.0 MeV) . 54 p ,p)54Mn 5.95'x 10-15 5.83 x 107 (b) 2 10.21 x.10 10-q 58 63 Ni(n,p) 8Ni. 7.88 x'10-15 7.90 x 10-2(b)[ 9.97 x.10 10 0u(n,a)60Co 5.49 x 10~1 -7.00 x 10~4(N '7s81 x.1010 - l 238g (,,f)1370s 3.31 x 10-14 3.2'O'x 10-1(b) .10.34 x.10 10L 237 Np(n,f)1370s 3.11 x 10 -13 3.30(b) 9.42xj0 10-Average 9.55k10 10. RESONANCE & THERMAL NEUIRONS 59 Co(n,7)60Co (Cd) 2.66 x 10 -12 7.55. >; 23 (c) 3.52 x 10 10(e) 59 Co(n,7) OCo (bare) 5.21 x 10-12 Bare Minus Cd 2.55 x 10-12 3.72 x 10-23(d).. -6.85 x 10 10(f) _ i F 1 ! (") Cross sections are from ENDF/B-V do.simetry file (b) Cross section for f ast neutrons (E > .1.0 MeV) . (")Rosonance integral i o } (d) Thermal (2200 m/s) cross section (*) Resonance flux (I)2200 m/r flux l I
'l 1
, i
)
l \. - l > l l l 1 1 i I 6-23 L t' )
e p, , i > - t i
+
Table.6-9 o; . # 2 . FERRET-SAND II:RESULTS FOR WOLF CREEK SURVEILLANCE CAPSULE ti; Energy A Priogi~ Flux
- TAdjustgd Flux % Uncertain.
Grouo' (MeV)- (n/es -seci- (n/cm -sec) -(l'Std)
- 1. 1.733E+01' 1.052E+074 . 7.771E+061- 22 2- .1.492E+01; .2.381E+07 1.766E+07: 120~
3 1.350E+011 9.153E+07 :6.871E+07s 17 4 .1.162E+01 L2.018E+08 1.546E+08 14-5 1.000E+01- 4.356E+08- .3.444E+081. 12' 6 8.607E+00l 7.193E+08 5.G79E+08: 11. 7 '7.4088+00 1.586E+09' - 1.390E+09 10 8 6.065E+00 2.165E+09 2.021B+09 10 9 -4.966E+00. 4.364E+09 '4.316E+09 , 10
'10 3.679E+00 5.576E+09 5.690B+09 :10 11 2.865E+00 : 1.136E+10' 1.181B+10 11-12- 2.231E+00 1.536E310 1;F96E+10. 12 13 '1.738E+00 2.141E+10 '2.206E+10 13 14 "1.353E+00 2.406E+10 ,2.445E+10 14' 15 1.108E+00 .4.432E+10- 4.402E+10 15
- 16. 8.208E-01 .5.121E+10 4.970E+10- :17 17 -6.393E-01 .5.3873+10' '5.106E+10- l19 -
18 4.979E-01 3.959E+10. 3.675E+10 :21: > 19 3.877E-01 5.658E+10 f5.151E+10. - 23 ' 20: 3.020E-01 .5.898E+10 5.292E+10 .25 21 1.832E-01 5.908E+10- 5.251E+10 27 .. 22 1.1115-01 :4.769E+10- 4.219E+10- 29 23 6.738E-02 3.337E+10 2.951E+10' 30 < 24- 4.087E-02 1.895E+10 11.681E+10- 31 25 2.554E-02
- 2.494E+10 .2.222E+10- 32-26 1.989E-02 1.230E+10- 1.102E+10' :33L 27 1.503E-02 1.561EA10- 1.'408E410. 33' 28 9.119E-03 2.249E+10' 2.036E+10 60L 29 5.531E-03 2.9178+10. 2.352E+10 60 30 3.355E-03 9.070E+09 8.282E+09- :60.
31 2.8393-03 8.637E+09 7.910E+09 59 32 2.404E-03 8.299E+09 7.619E+09' 59
'33 2.035E-03 2.335E+10 2.146E+10 58 34 1.234E-03 2.150E+10 1.980E+10 56 l 35 7.485E-04 1.990E+10 1.836E+10 53;
'l 36 4.E40E-04 1.892E+10 1.749E+10. 51- R' 37 2.754E-04 2.033E+10 1.880E+10- 48 38 - 1.670E-05 2.152E+10 2.013E+10 15' ,
39 1.'013E-05 2.177E+10 2.026E+10: 47J ~ k k 6 o _
v .,> ' c
.r 1-
}
,. 3 i.:,2 .
. , 7 i .
t . s i
-
- I
) ' .
' \*(. p[: ',
a N. - d h)\; li ., \
/ / (, [ 1 n
(. a
. Table . 6-9. (' cont'd.)
' ~
Energy.- I'riogi~ Flux * 'idjEstgd Phx- % Uncertain' ' Grou pi I,NeV) (n/em-seck (n/cm -sec)' (1 Std)
.L l40-,E 7 .1 144E-05 2.168E+10 . 2,010E+10. . 49: -
411/ , T 727E-05 . ' 2.10E+10-' 1.963B+10 " :52 42 j"2.260E-05 2.025E+10- 0 1.902B+10 54 .
'4't 1'.Si1E-05 ;' 1 $53E+10 1. 845E+10', 56
'44 (8.315E-06i9d A52E+10 1.761E+10 57, 45l l1.632E+10' 58
' 46 % y 3.059B-06 5.N3E-06 ? '.[1.702E+10 1'543E+10 1.492E+10 58.
47 1.855E-06' '1'.378E+10 1.344E+10.. 59 48' .1,125E-06 Y.103E+10 l'.173E+10 ;59 l' 49 6.826E-07 9.884E+09 '1.101E+10 '103-50 4.140E 7.966E+09 9.944E+09 '99 ' 51 2.511E-07 ' 7.406bO9' 1.054E+10. 93-52' 1.523E-07' 8.6988+09' .1.397B+10- :85- , 4 53 9.237E-08 2.552E+10 5.792E+10 34
}
a t. , h
*The a priori flux SS taken from a generic 4 loop plant. .,
l calculatgn'andnormalisedtogiven'pproximatelythe.
- 9 correct Fe(n,p) r'waction rate for the Yolf. Creek capsule. I The ' thermal flux, which is eiverely depressed at' the center -
of tg capril' is e separately uljusted to approximately fit l the Co(nd) re.netion.L ,, l l t j c 5 '< l .s
', t. ,
4 l v, I 1 j L l
- b
'd 6-25
. ; .l l . < !
.:_ - u L L - i'
- n. - - - - -
t f j Table 6-10c COMPARISON OF E ASURED AND CALCULATED. REACTION RATES.
- USED IN THE ANALYSIS OF WOLF CREEK CAPSULE U, Reaction Rate- . .' . Ratio.Cale/ Mess-Reaction Measured: . Uncertainty 00 A Priori .,Adiusted:
b (n,p) ,
~
5.67_x 10-15 I5 1.00- 0.97 58 ' '8~' -0.99: Ni(n,p) 8.17 x 10 ,1.02
.63 0u(n,a)' ~II' 1.03
'5.66 x'10 5 1.30L 238U (n,f) -14 0.99 1.00' 3.48 x 10 - 15i 237 1.09 :1.04 Np(n;f) ' 2.77.'x 10-13 10.
'59 10. I'.07 1.01 Co(n,7)Cd :2.73 x 10-12 5.21 x 10 -12
~
59 0.98 Co(n,7) Bare 10l 0.79 e ea G 6-36 l
Table 8-11 INTEGRAL NEUTRON FLUI RESULTS DERIVED FROM ADJUSTED SPECTRUM-Adjustgd alue.: Parameter- (n/cm -sec) Uncertainty (la)- Neutron Flux (E S l'.0 MeV) -. 1.04 x 10 11' 8% Neutron Flux (E > 0.1.MeV) 4.36 x 10 11 '15% Neutron Flux - (B <: 0.414 MeV) 9.24'x 10 10 -26% Total Neutron Flux 1.03 x.10 12 . 15%' Iron' Displacement (dps/s) 1.94 x'10-10 11%' Neutron Fluence (E > 1.0 MeV)- 3.52 x'10 18 8% Iron Displacements (dpa) .8.58 x 10 -3 11% q i I l' I I l 6-27 .- l
y ' 'i Table 6-12 .
SUMMARY
OF WOLF CREEK FASTi(E >- 1.0 MeV) NEUTRON ~ i
, -FLUENCE RESULTS BASED UPON SURVEILLANCE CAPSULE U.
-End-of-Life Current Fastf(E ).1.0 MeVJ Fast (E. > 1.0 MeV . -
Neutron Fluence (*) Neutron Fluence (b i
.(n/cm):
2 , (,7,,2); Location- Measured Calculated Measured Calculated 18 3.22 x 10 18 Capsule U' 3.39 x 10 , ,. Vessel IR 8.80 x 10 17 '8.36 x 10 17 '2.61 x 10 18' 3.12 x 10 18 17
'4.60 x 10 17 1!44!x.10 18 ~ 1.72 x'10 19l
~
Vessel 1/4T' 4.84 x 10 8.69 x.10 10' 2.71.x 10 18 :3.24 x 10 18 Vessel 3/4T 9.15 x 10 16 ..' (*) Current fluences are based on operation.at 3411 Mwt for.1.08 EFPY ' and use a plant specific calculated fluence. (b)EOL fluences based on operation at 3411 Mwt for 32 EFPY and use ' design basis calculated fluences. 4 I 6-28
..L:
n, i-1 1 l'. I i
)
l (TYPICAL)
'C - 61.C0-58.5 0' /
- \ l r ^ "R r
- 81.625 IN.
\
1
////// 4 9 N
k_ $h ~%% -k_ l Figure 6-1. Plan view of a dual reactor vessel surveillance capsule-6-29
l- , [ curve 753074-A 20.0 , ., , , i. m 10.0 -
- 8. 0
- 6. 0 - Surveillance Capsule' Geometry -
e '4. 0 - Reactor VesselIR X E j ~2.0 - .- 5' w_
.0 - ~
1/4 T 1.ocation' .
- 0. 8 A
$0.6 - -
.E
- ~
3/4T Location 0.2 -
'0.1 i i I i-l 0 '20 40 60 80 ~100 L
Azimuthal Angle (Degreti) Figure 6-2. Calculated asinuthal' distribution of maximum fast { (E > 1.0 MeV) neutron flux within the reactor . vessel-surveillance capsule geometry . a-so 4 , , i
; ,<: 1. ;
1 T Curve 753073-A 100-1 -1 1. .I 80 - - 60 ~
^~
219.71: 40 - -- 1 i R
,2 ' 225.23 X 1 m . '
id 20 - IR- - 1 "p - i
- a. ' 230.75
-E f 1/4T
_ 10 - - -
%~8 g
236,27-. -
~
^ 1/2T
~
w , p 4 - - C - 241.79 :j 1 q 3/4 T ! 2 . - Oh 1 1 1 1 1 200 210 220 230 240 250' Radius (cm) i
-Figure 6-3. Calculated radial distribution of maximum fast- , (E > 1.0 MeV) neutron flux within the reactor vessel 6-81'
p .- l- .l l l . . I ) t 11756$
)
1,000 0.700 - 0.500 - 0.200. - 0,100 - 5 0.070 - d - 5 'O.050-
=
H S 2 5 0.020 - G e . 0.010 - 0.007 - 0.005 -
~
CORE MIDPLANE
**00I ~
. TO VESSEL
' CLOSURE HEAD 0.001 I ! ! I l L i
! 300 200 -100 0 100 200 300 OlSTANCE FROM CORE MIDPLANE (em) Figure 6-4. Relative axial variation of f ast (E > 1.0 MeV) ! neutron flux within the reactor vessel , l 1 6-82 . l
4
- 7. SURVEILLANCE CAPSULE REMOVAL SCHEDULE- a The following removal schedule'meeta ASTM E185-82 'and is-recemoended for future capsules to be removed from the Wolf Creek reactor vessel:
l l Capsule Estimated-Location Lead Flu'nge e capsule -(der.) Factor Removal Time (*) (n/cm ) U 58.5 3.85- 18 1.08 (Removed). 3.39.x 10 Y 241 3.65 5 1.78 x 10 19(b) V 61 3.65 9 3.20 x 10 19 (*) .. I. 238.5 3.85 15 5.03 x;10 19
. W 121.5 3.85 Standby- -
Z 301.5 3.85 Standby - l 1 (*) Effective full power years from plant startup. (b) Approximate fluence at 1/4 thickness reactor vessel wall at end of life. (") Approximate fluence at reactor vessel inner wall at end of life. 1 != o i { I
.j 7-1 q l
P
, y. '
3,
,7~ '
7 , 4 i.,4 .,,o , , 1
.k/ j 11 r 6'
'r
'8.' REFERENCES'
- 1. L.'R.LSinger,.'KansaGasandElectrhcCompanyWolf. Creek
' Generation Station Unit No.t1~ Reactor Yessel Radiation- '
Surveillance Program'.,.WCAO-10015,' June.1982.
- 2. . Code'of.Pederal' Regulations,:10CFR50, Appendix G, ' Fracture- .
' Sughness: Requirements'.;and Appendix H,o' Reactor Vescel Material- ,
' surveillance Program. Requirements",! U.S.' Nuclear Regulatory Commission, Washington, D.C.. ' '
- 3. Regulatory;Cuide~1.99, Proposed Revision 2, ' Radiation Damage.to.'
.ReactorLVessel. Materials',U.S.: Nuclear:Reguletoryl Commission',,
February, 1986.
- 4. R..G. Soltess, R. K. Disney, J. Jedruch,:and S.3L.' Zieglar, . >
" Nuclear Rocket Shielding Methods, Modification, Updating'and Input Data Preparation. Vol. 5--Two'-Dimensional Discrete.
Ordicstes Transport Technique", WANL-PR(LL)-034', Vol 5, ' August : 1970.
- 5. 'O'RNL-RSIC Data Library Collection DLC-76, SAILOR coupled Self = .
Shielded, 47 Neutron, 20 Gamma-Ray,'P3, Cross Section, Library for Light Water Reactors'; S. P. C. Cook and E. M. Spier, 'The Nuclear Design >and Core Physics. Characteristics of the Wolf ~ Creek Generating Station' Unit 1, Cycle 18, WCAP-10483,. February, 1984.
- 7. Benchmark Testing'of. Westinghouse Neutron transport' Analysis ~
Methodology (to be publish ~ed). f~ 8. ASTM Designation E482-82, ' Standard Guide for Application of~ Neutron.Tran.rport Methods for Reactor Vessel. Surveillance"., in ASTM Standarde...Section 12, American Sociuty for Testing.and * ' Materials, Philadelphia,.PA,:1984.
, 1
- 9. ASTM Designation E560-77, " Standard Recommended Practice for: q Extrapolating Reactor _Yessel' Surveillance Dosimetry Results',-la ;
ASTM Standards, Section;12, American Society'for Testing and -
-1 Materials, Philadelphia, PA, 1984.
g ,1
-t i
1
i
- 10. ASTM Designation E693-79, " Standard Practice for Characterizing Neutron Exposures in Ferritic Steels in Terms of Displacements per Atom (dpa)", in ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia, PA, 1984.
- 11. ASTM Designation E706-81a, " Standard Master Matrix for Light-Water l Reactor Pressure Yessel Surveillance Standards", in ASTM' Standards,'Section 12, American Society for Testing.and Materials,.
Philadelphia, PA, 1984.
- 12. ASTM Designation E853-84, " Standard Practice.for Analysis and-q Interpretation of Light-Water Reactor Surveillance'Results', in :
ASTM Standards, Section 12, American Society for Testing and Materials, Philadelphia,.PA,1 1984.- '
\
- 13. ASTM Designation E261-77, " Standard Method for Determining Neutron j Flux, Fluence, and Spectra by Radioactivation Techniques", in' ASTM :j Standards, Section 12, American Society for Testing'and Materials, Philadelphia., PA, 1984.
14
- 14. ASTM Designation E262-77, ' Standard Method for Measuring Thermal j Neutron Flux by Radioactivation Techniques", in ASTM Standards, 3 Saetion 12, American Society for Testing and Materials,' l Philadelphia, PA,'1984.
]
- 15. ASTM Designation E263-82, ~" Standard Method for Determining Fast-Neutron Flux Density by Radioactivation of Iron", in ASTM j Standards, Section 12, American Society for Testing and Materials, e Philadelphia, PA,.1984.
j I
- 16. ASTil Designation E264-82, ' Standard Ethod for Determining Fast- )
Neutron Flux Dens!ty by Radioactivation of Nickel', in ASTM A Standards, Section 12, American Society for Testing and Materials., j Philadelphia, PA, 1984. t 1
- 17. ASTM Designation E481-78, ' Standard Method for Mearnring Neutron- :
Flux Density by Radioactivation of Cobalt and Silver', in ASTM i Standards, Section 12, American Society for Testing and Materials, i Philadelphia, PA, 1984. l .
- 18. ASTM Designation E523-82, " Standard Method for Determining Fast-l Neutron Flux Density by Radioactivation of Copper', in ASTM j l Standards, Section 12, American Society for Testing and Materials,
! Philadelphia, PA, 1984. l 1 l-l s~s '
- 19. . ASTM Designation E704-84, ' Standard Method for Measuring Reaction Rates by Radioactivation of Uranium-238', in ASTM Standards
- Section 12,.American Society for, Testing and Materials, Philadelphia, PA,-1984.
- 20. ASTM Designation E705-79,." Standard Method for Measuring Fast-Neutron Flux Density by Radioactivation of Neptunium-237',;in ASTML Standards,-Section 12, American. Society for Testing and~ Materials, Philadelphia, PA, 1984.
21.' ASTM Designation E1005-84, " Standard Method for Application and Analysis- of Radiometric Monitors for' Reactor Vessel Surveillance', in ASTM Standards, Section'12, American' Society for Testing and ) M4terials, Philadelphia, PA,.1984.
- 22. F. A. Schmittroth, FERRET Data Asalysis Cor_e, HEDL-TME 79-40,-
Hanford Engineering Development Laboratory, Richland, WA, September 1979.
- 23. W. N. McElroy, S. Berg and T.' Crocket, A Computer-Automated Iterative Vethod of Neutron Fivx Spectra Determined by Foil.
Activation, AFWL-TR-67-41, Vol. I-IV, Air Force Weapons. Laboratory, Kirkland AFB, NM, July 1987.
- 24. EPRI-NP-2188, " Development and Demonstration of an Advanced
- Methodology for LWR Dosimetry Applications','R. E.- Maerker, et al., 1981.
j I l
~l s-a i
s
i l
. 3 i
l I APPENDIX A i i HEATUP AND COOLDOWN LIMIT CURVES i FOR NORMAL OPERATION '! I A.1 INTRODUCTION Eeatup and cooldown limit curves are calculated using the roost limiting value of RTET (reference nil-d":tility temperature) . The most limiting RTET f the material in the core region of the reactor vessel is determined by using the preservice reactor vessel material properties and estimating the radiation-induced ART ET. RT ET is designated as the ! higher of either the drop weight nil-ductility transition temperature i (NDTT) or the temperature at which the material exhibits at least 50 ft- 1 lb of impact energy and 35-mil lateral expansion (normal to the major working direction) minus 60*F. RT ET increases as the material is exposed to fast-neutron radiation. Therefore, to find the most limiting RTET *D ""7 DI** period in the reactor's life, ARTg7 due to the radiation exposure associated with that time period must be added to the original unirradiated RTET. The extent cf the shif t in RT ET is enhanced by certain chemical elements (such as copper, nickel, and phosphorus) present in reactor vessel steels. Westinghouse, other NSSS vendors, the U.S. Nuclear Regulatory Commission and others have developed trend curves for predicting adjustment of RTNDT as a functi n cf fluence and copper, nickel and/or phosphorus content. The Nuclear Regulatory Commission (URC) trend curve is published in Regulatory Guide 1.99 (Effects of Residual Elements on Predicting Radiation Damage to Reactor Vessel Materials).1 Regulstory Guide 1.99 was originally published in July 1975 with a Revision 1 being issued in April 1977. Currently, a A-1
be< - 4 , y y w< g bl e
, , t ,
s
.g.
n ., , t i"> 6 y .
+
'Is J
;\
l'\ f,. Revis12n'2.to Rsgulatory Guide'1.99 is uSder consideration.within the;
- NRC. ; - The Ichemistry f actor; 'CF" (*F)'1 a fune tion of > copper and nicke?, '
content' identified!in' Regulatory, Guide;1.99; Revision [22'1, ,gy,, g, .
'* Table A-1: for weldk and Table 'A-2) for . base ' metal ? (plates t and Lforgings)?. -
InterpolSionlispermitted. ThevalueJ?.f",,given'inFigureA-1isthel> dculated'valueofthe1o neutron' fluence-atthe'locatiYn'ofihterest a . . (inner'surfe.ce," 1/4T, ' or' 3/4T): in the vessel at the11ocation of 'thez , postulated defect, n/cm > 1 MeV) divided by 105- The. fluence - # factor'is determined from Figure lA-l', '
,GiventheLeopper?andnickel)contentsofthemost! limiting 2 material, the!iradiation-induced 3RTgcan' be estimated from Tables' A-lj .
and A-2 and Figure A-1. The maximum fastineutron fluenee (Ef1 Mei):at- . r r -, .
. the. inner surf ace, the' 1/4Tl (wall . thickness), and 3/4T).(wall i th'ickness) .
f vessel locati6ns.is given as r function'.of' full-power service life in-Figure A-2 :for the vessel' core region.
/
A.2 FRACTUR8 TOUCHNBSS PROPERTIBS , The preirr6distion fracture-toughness properties of the Wolf Creek reactor vessel r,aterials are presented in Table A-3.' The. ' fracture-toughness' properties 'of the ferritic :natorial in the reactor ' coolant' pressure boundary are determined in'accordance with the NRC~ ; Regulatory Standard Review Plan.3 The postirradiation fracture s toughness properties of. the reactor vessel beltline materialTere
.obtained directly from.the Wolf Creek Vessel Material Surveillance Program. These results show that the transition temperature shift is.
~
less than'the shift as predicted by Rev.' 2 of Reg. Guide 1.99, thus: validating the conservatism of Reg.-Guide 1.99:Rev. 22 for generating'. . l the heating;and cooldown curves. ,
'i .
A.3 . CRITERIA FOR AI;LOWABLE PRESdVRB-TEMPERATURE RELATION 8fIPS The ASME approach for calculati'ng the allowable' limit curves.for .j various heatup and cooldown rates specifies,that the total etress c j intensity factor, K ,7for the combined thermal'and pressure stresses at - L 1 l l s > l ! .A-2 ' 3
,g
- , T .. .c. l
.. . I5 i . r
, ,c, j ,
g i ' i t . ,k ;-
. l i.
l-[' .gl ,
,x r ,
9 p- m' 1 W
'anyitimeduringbestup,orcooldowncannotibe'greaterthan'thereference'(
p'
. stress intensity l factor, KIR, f r the metal. temperature at'that time.
KIR.is obtained'from the reference fracture toughness curve'cdefined in~
.' App $ ndix G to the! ASE Code. 4 . Thi K IR curve is given by theffollowing.. <
j equation. o
]
Kg = 26 78 + :1.223. exp [0.0145 '(T-RT@T + 160)]; (A-1).
,1 where.
1
'l
.1 KIR = reference stEess. intensity factor.as a fusetion of the' metal' temperature T'and'the metal reference nil-ductility temperature RT '
ET - 4 q S Therefore, the governing f.quation for'the-heatup-ecoldown j analysis is ' defined in Appendix G.of the ASE Code 4 as follows.' ' I J f CKg+ KIT UEIR (A-2)~
, a shere Kg = stress ' intensity factor cau'se'd by membrane (pressurp) stress ' ]
KIT = stress intensity _ factor caused by the thermal gradients
]p.
a Kg = function of temperature relative to the RTET . f the saterial - ; a C = 2.0 for Level A and Level B service limits- d s
, i j
C =.1.5 for hydrostatic and leak test. conditions d& ring which the-
'1 reacter' core is not crisical 1
' .a j %.
E i s . n . 1 A-s :
- {
t
' ' 'E' h6 l,
3 .
,I
)-
q. ,e At'any time dEring.the heatup or cooldown transient,'Kyg is determised by.the. metal temperature at.the tip of the postulated flaw,'-- E the appropriate value for.RTg , and the reference fracture-toughness- , curve. The thermal-stresses resulting from temperature gradients , through the vessel wall are calculated and then the corresponding: 9 (thermal) stress intensity f actors, _7 K j,J for;the reference flaw are computed. From Equation 1-2, the preusureistress-intensity 1 factors are obtained and, from these,1the'allpwable pressures are' calc 11ated. , For the calculation of the allowable pressure versus coolant temperature'during cooldown, the rqference flaw'of Appendix.G to:the' ASE Code'is assumed to exist at the'.inside of tho' vessel wall._ ;During cooldown, the controlling location'of the! flaw is'always at the:inside of'the wall because the thermal' gradients producettencile; stresses at'- the inside,-which increase with increasing cooldown' races.1 Allowable: pressure-temperature relations are generated for both steady-state andi finite cooldown rate situdions. From these relations, composite limit-curves are constructed for each cooldown rate of interest'.' , , The use of the composite curve in the cooldown analysis is necessary because control of the cooldown procedure is based on - measurement of reactor coolant temperature,'whereas the limiting. pressure is actually dependent on the material temperature at the tlp~ofi the assumed flaw, o During cooldown, the 1/4 7 vessel location is at,a higher i temperature than the fluid adjacent to the'yessel'ID. This condition,1 of course,.is not true for the steady-state situation. It follows that,
~
E at any given reactor coolant temperature, thd AT developed during l o cooldown results in.a higher value of KIR ,at the 1/4 T, location for_ finite cooldown rates'than for steady-state operation.' Furthermore, iff conditions exist'so that-the increase in XIR *****d' KIT, the calculated '
.a llowable pressure during cooldown will be greater than' the steady-state
, q
, value. ,
l The above procedures are needed because there is no direct l
)
control on temperature at the 1/4 T-location and, therefore,: allowable - '
]
1 l \,
' A.4
o ,
. i 1 , ,
.i' . , , .. , .
i, ,
i
-i
- J l
pressures may unknowingly be violated if.the rate of cooling is I decreased at various intervals along a cooldown ramp. The use of the composite curve eliminates this problem and' ensures conservative operation of the system for the entire cooldown period. Three separato calculations are required to determine the limit I curves for finite h'estup rates; As is done in the cooldown analysis, i allowable pressure-temperature relationships are develop'ed for steady- ! state conditons as well'as finite heatup rate conditions assuming.the L presence of a 1/4 T defect at the inside' of-the vessel wall. The. l thermal gradients during heatup produce compressive stresses at the { inside of the wall'that alleviate the tensile stresses produced by 1 internal pressure. The metal temperature at the crack tip lags the coolant temperature; therefore, the K f r the 1/4 T crack during-IR heatapislowerthantheKjR f r the 1/4 T crack during steady-state conditions at the same coolant temperature. .During.heatup, especially l at the end of the transient, conditions may exist so that the effects of
- l. compressive thermal stresses and lower K 's do not offset each other, IR and the pressure-temperature curve based on steady-state conditions no longer represents a lower bound of all similar curves for~ finite heatup rates when the 1/4 T flaw is considered. Therefore, both cases have to l be analyzed in order to ensure that at any coolant temperature the lower I
value of the allowable pressure calculated for steady-state and f.inite heatup rates is obtained. ! The second portion of the heatup analysis concerns the calculation of pressure-temperature limitations for the case in which a 1/4 T deep outside surface flaw is assumed. Unlike the situation at the vessel inside sufface, the thermal gradients established at the'outside j surface during heatup produce stresses,which are tensile in nature and therefore tend to reinforce any pressure stresses present. These l' thermal stresses are dependent on both the rate of heatup and the time (or coolant temperature) along the heatup ramp. Since the thermal-stresses at the outside are tensile and increase with increasing heatup rates, each heatup rate must be analysed on an individual basis. l l A-5, lt t
-__-w--~~
l
~
- m. ,
, D. < ' - ,
7, , J
,( ' - '
J j -l ] b , x, .
.- 1
'[, . . , .
s
$l i Following the:. generation of pressure-temperature curves for.both i ithe' steady-state'and finit'e heatup rate ~ situations,'the; final; limit. '-
,- , curvesarepr,oduced5bycconstructing'scompositecurvebasedonapoint-
;, u .. .
.by-pointcompariEonofthe-steady-stateiand' finite'hestdpratedata.!At' arygiven. temperature;:theallbwablepressureis..taken..tobethelesser' j
of the thsie values'taken fromthi curves under consideration. The;use j i of the composite curve.is necessary.to < . set' conservative heatup' , , ,
' 'i limitations'because'it. !1s possible for conditions to exis't wherein,('over' ..
1 i 4 the course of the heatup ramp, the controlling condition [switchesTfrom
-the-inside to the'outside'and the: pressure limit must at all times b'e' * >
.: . . . .. H based on' analysis of th'e most critical criterione Thenp composite-curvesforthe'heatuprate'dataandtheCcooldownratedabaare/ adjusted' ]
~
for possible errors in;the' pressure' sad temperature sensing ins $ruments R by 'the values indi' at'ed c ' on' Figures' A-3.and'A-C:
- 5 Finally,thenew10CFR50.AppendixG. rule'which/addressesthe; 4 metal temperature of the'cissure head!flangefand vessel flange rigions' ,
is considered. This 10CFR50 rule states that)the metsl' temperature ofl . . the closure' flange regions.must exceed the material RTg by at'least ; 120*Ffornormaloperationwhen(thepressureexceedAl20percentofLthe' . preservice hydrostatic test' pressure (621 psig for Wolf; Creek).. Table.: A-3 indicates shat the limiting RTg .'of 40*F occurs 'n'the i vessc1' . f flange of Wolf Creek', and the minimum allowable , temperature of this , region is 160*F st pressures greater.than 621 paig. These limits.are; , shown by the indentation on Figures.A-4. F A.4 HBATUP AND-C00LDOWN LIMIT CVRVBS Limit curves for normal'heatup and cooldown of the primary ReactorCoolantSystemha'vebeencalculate1using-theme $ hods / discussed-in Section4A-3. The derivation of the limit. curves.is presented in.the- . l NRC Regulatory Standard Review Plan.3 Transition temperature shifts' occurring in the pressure vessel' -
- 1 n
materials.due to radiation exposure have been'obtained directly fres the. R 1 l- reactor pressure vessel surveillance program. -
;j l 1 1
h A-6 q l g
._ _]
1 Allowable combinations of temperature and pressure for specific ( o temperature change rates are below and to the right of the limit lines ; shown in Figure A-3 and A-4. This is in addition to other criteria ! which must be met before the reactor is made critical. The leak limit curve shown in Figure A-3 represents minimum j temperature requirements at the leak test pressure specified by applicable codes. '4 The leak test limit curve was determined by j methods of References 3 and 5. l 1 Figures A-3 and A-4 define limits for ensuring prevention of i nonductile failure. Figure A-3 pertains to heatup rates up to 60*F/hr as well as 100*F/hr. A.5 ADJUSTED R3FERENCE TEMPERATURE From Regulatory Guide 1.99 Rev. 22 the adjusted reference temperz.ture (ART) for each material in the beltline is given by the following expression: ART = Initial RTET + ARTET + Margin (A-3) Initial RT ET is the reference temperature for the unirradiated material as defined in Paragraph NB-2331 of Section III of the ASME Boiler and Pressure Vessel Code.4 If measured values of Initial RT ET for the material in question are not available, generic mean values for that class of material may be used if there are sufficient test results l to establish a mean and standard deviation for the class. ART ET is the mean value of the adjustment in reference temperature caused by the irradiation and should be calculated as i l follows: l ART ET surfact; = (CF]f(0.7 0.10 log f) (A-4) 1
^
I f.- 7 l
, , , n 4
, , ;g w
#. o
' i
'y l
, 4
.]
.ti ,
i s
,,i f Lj'.
-NE i ,
- 1 rf b.
.Tc' calculate ART ET st.any depth (e.g.,JatL1/4T or 3/4T),ftihe~ ,
following attenuation ~ formula was used:. , .-j . (A~ " ARTET = [ARTETI'"#I'**3
- wherex,(in' inches)Listhi~depthinto;the'. vessel'wallmeasured;fromthe' <
vessel inner (wetted) surface.1 . s CF l'F) Lis the chemistryjfactor,Ta functi'on of copper andiniEkhil -- content. CF. is given in Table" A-I for: welds and.'in Table A-2 for base- ' metal'(plates and forgings). Linearfinterpolation'isjpermitted;iIn Tables Al l and A-2.' weight-perce't n copper' and'8 weight-percenti nickel *,
- are the.best-estimate values for thi material,'whibh;will normally he' >
the mean of the measured values l for a plate er forging.or-for weld r samples made with the weld wire heat number that matches the critical? vessel weld. The calculated endLof life fluence for 32 effective full power-years (EFPY) is 3.12 x 10 18 n/cm 2 ' at the vessel'inside radius. To: .: - cospute the fluence (f) for 7 EFPY, 'this value i's multiplied by (7/32) . < 2 L Thisyieldsafluencef'=0.6825x10In/cm ' 1 and,the resulting - flueace f actor, is 0.893. Based on Table A-3, the beltline material with the highest! initial RTET (40*F) as well asLthe highest % Cu and'% N1:is the: lower shell plate R2508-3. Based on Table ~A-2, the. chemistry factor)(CF)'for this plate is 44. - From Equation (A-4),c the1 ART 2 ET surface is. (39.3*F). Reg. Guide 1.99 Rev. 2 provides a formula and rules for-the margin:: 2 Margin'= 2 , a y v ,A ' i(A-0) l Since a measured value of initial RTET is used,1the temperature .j a y_is taken as sero, and the temperature a 'Is taken at 17'F. As a-3 ,
.p]
u i i i 1 A-s
.v .
w .' , i -(
}
result the margin is 1 2l (0)2 [17)2 = 34'F l
' Substituting the obtained values for ART ET surface and' margin, along with the' initial RTET f'40'F into Equation' (A-3) gives the ART for the - {
inside-surfae6: .i i ART = 40 4 39.3 + 34 = 113.3*F. I I ITsing the vessel thickness of. 8.625 inches at the beltline, Equation'A-5 is used to y h d ate.the ART ET at the 1/4T and 3/4T locations.- These are 34.0*F and 25.5'F, respectively. The ART at any location is'given by the initial'RTET, Plus the shif t (ARTET) at that location, plus the margin. Thus_at the 1/4T and 3/4T locations, ART is equal to 108.0'F and 99.5*F respectively af ter 7 l
- EFPY.
4 A-9 d ._m
m p c .
, -n g , . ,
V
.\ *w -[ l: .
c a h l I ., REFERENCES- 2
- 1. Regulatory Guide 1.99,. Revision l', " Effects of-Residual ::lements-on Predicted Radiation Damage to Reactor' Vessel Materials','U.S.
Nuclear Regulatory Commission,-April'1977. 2.. Regulatory Guide 1.99, Revision 2, " Radiation Damage to Reactor
. Vessel Materials', U.S. Nuclear Regulatory Commission, February 1986.
' 3. - " Fracture. Toughness Requirements", Branch TechAical Fosition'MTEB 5-2,^ Chapter 5.3.2 in Standard Review Plan for the Review of:
Safety Analysis. Reports for Nuclear Power Plants, LWR Edition,: NUREG-0800,.1981.
- 4. ASME Boiler,and Pressure Vessel Code, Section III,. Division ~1 -
Appendices, " Rules for Constructionxof Nuclear Vessels, ' Appendix G, ' Protection' Against Nonductile Failure', pp. 559-564, 1983 Edition, American Society of Mechanical ~ Engineers, New York,.1983. *- 1
- 5. Code of Federal Regulations, 100FR50, Appendix 0, " Fracture Toughness Requirements,8-U.S. Nuclear Regulatory Commission, Washington, D.C., Amended May 17, 1983 (48 Federal-Register 24010).
- 6. Pressure-Temperature Limits,' Chapter 5.3.2 in Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power E ts, LWR Edition, NUREG-0800, 1981.
i b. A-lo
- - - iMJ A a 3_' _s -_&oeib. A. _
[n Table A-1 3 CHEMISTRY FACTOR FOR WELDS,'*F, Copper, Nickel. Wt-% 1 Wt-% 0 0.20 0.40 0.60 -0.80 1.00 1.20-0- 20 20 20 20 20 20 ' 20 ,. 0.01 20 20 20 20' 20 '20. ~20-0.02 21- 26 27' 27 27 27 27 0.03 22 35 .41 41 41' 41 141' O.04' 24 43 54 54 54 '54 .54 i
.{
1 0.05. 26 49' 67 68 68- 68. 68) ! 0.06 29 52 77 82 82 82 82 0.07 32 55- 85 95 95 95 95. 0.08 36 58 90 106 '108 108 108 0.09 40 61 94 115 122 122 122 '$ j 0.10 44 65- 97 122 133 135 135 n 0.11 49 68 101 130 144 : .148 ' 148 - 0.12 52 72 103 135 153 161 161 , 0.13 58 76 106 139 162 172 176 ] 0.14 61 79 109 142 168 182 188 0.15 66 84 112 146 175 191 200 0.16 70 88 115 149 178- 199 211 0.17 75 92 119 151 184 207 221 0.18 79 95 122 154 187 214- 230 j 0.19 83 100 128 157 191 .220 238.
.j 4
1 J J A-11
..J
lE Table'A-1 (cont'd.) .
-Copper, Nickel. Wt-%-
Wt-% 0 0.20. 0.40 0.60 0.80 1.00 1.20 0.20 -88 '104 129. '160 194. 223 245 0.21 92 108 133 164 .197 229 252 0.22 97 112 137 167: 200 232 257 0.23 101 117- 140 169. '203 ~236 263, 0.24 105 121 144 .173 206- g239 268. 0.25 110. 126- ,148 176' 209, 243 272 0.28 113 130 151. 180 212 246 276. a0.27 119 134 '155 .184 21C .249L 280 0.28 122 :138 -160 187 218 251 284 0.29 128 142 164 191 222 -254 287 0.30 131 146 167 194 225 257 290'
~
O.31 136 151 172 198 228 260 293 0.32 140 155 175 202 231 263. 296 ' 0.33 144 160 180 205 234 266 299 O.34 149 164 184 209 238 269 302
- . 0.35 153 168 '187 212 241 '272 305' O.36 158 172 191 216 245 275 '308 0.37 162 177 196 220 248 ,278- 311 0.38 166 182- 200 223 250 281 314
, 0.39 171 185 203 227 254 285 317 0.40 175- 189 207 231 257 288 320 I
1 h A-12 o a ed I <
,m
' i
. i:'
'} : t e
. Table A-2 I
' CHEMISTRY FACTOR FOR BASE METAL, *F Copper, "
Nickel' Wt-5
. .I Wt-% 0 0.20. 0.40, 0.60 0.80 1.00 1.20. !
0 20 20 20- -20 . ' 20 20 20 0.01 20. 20 20 20 -20 20- '20: !
'O 02
. '20 20 20 20 '20 20 20; )
0.03 201 20 20 20. 20 20- 20 l- 0.04 22- '26 26 '26 26 26. .26 I 0.05 25 31 31 31 31- 31 - 31 ' O.06' 28 37 '37- 37 37 37- 37- , O.07- 31 43 44 - 44 44 44 44-l ' O.08 34 48 51 51 51 .51 '51 ] 0.09 37 53 58 58 -58 58 58 j t 0.10 41 58 65 65 67 67 67 l
. 0.11 45 62 72 74 77 77 - 77 ' f 0.12 49 67 79 83 86 86 86 t
0.13 53 71 85' 91 -96 96 96 - 0.14 57 75 91 100 '105 106 106 j O.15 '61 80 99 .110 115 '117 -117 - 0.61 65 84 104- 118 123 = 125 -125 ,1 1 0.17 69 88 110 127 132 135 135 ,l 0.18 73 92 -115 134 141 144 144 0.19 78 97 120 142 150. 154 154-r 0.20 82 102 125' 149 159 164 165
. 0.21 86 107- 129- '155 167' 172 174 0.22 91 112 134 161' 176 : 181 184 0.23 95 117 138 -167 184 ' 190 194-0.24 100 121 143 172 191 . 199 204
-1
.1 A-la l
)
+
a , , 5
~
Table A-2;(cont'd.)' .. Copper, Nicke1~. Wt'-%
' Vt-% 0 0.20 0.40. 0.60 .0.80 1.00 1.20 0.25 104. :12A , 148 - 176 199 208- 2141 0.26 109 130 151 1E0 205 . 216 231 0.27 114 134 155 184 211' ' 225 230 0.28 119 138. 160 '187- 216 233 239
- 0.29 124 142 164 191- 221 . 241 .248 0.30 128 1146 l6'7' 194 '225 249 '257 0.31- 134L 151 172 198' 228 255 266 0.32 139 155 175 202 ;231 260' 274' O.33 144 160 180 205 : 234 264- 282-0.34 149 164 184 -209 238 268 290 0.35 153 168 187' .212 241 ' 272 '298 0.36 158 173 191 216. 245 275~ ,303 0.37 162 177 196 220 248 ' 278 308 ,
0.38 168 182 200 223 250 281 313 0.39 171 185 203- 227 254 . 285 317 0.40 175 189 207 231 257 288' 320 l-1 1 ( A-14
^
k- _ . _ _ _ _ _ . . _ _ - _ _ _ _ _ _ _ _ . _ . _ _ _ _ - - - _ _ _ _ _ _ __ _ . _ _ . _ _ . _ _ _ ,
--___.-__.______---______-__._____-______.-__z.-______. _ - _ _ _ _
t f b ' i - l l - e
- 634877- -
h Dt
- - - - - - - - - - - - - - - - - 546122 -
111111 S Wf xn' o r
'M(
e 1 - p p ) - U
- b -
- . - - ' ~~
~
7'
- l D - 29927766696481377 113043552231123223808955 5706240 8 1 r Wt 9 7 v Mf 111111111111111111 1 11 -
1 A N( 1 - F
- 000000000000000000000000 22222231 - : 12 222~ 14365 0
5 0 8 f - - - - ^- - - ' 1 -
~
), f
) _ - - .
1
- 1
.? -
T : D E F
- 00000000000000000'0000000 422222231 445233412865 0
5 0 8 T f - - L A '
=
r- ~ ~. . I l
~
~
466020' 226515945643742444 D 5 A R iQ 666788586676666666366660 0 ] R N (. 0O0000O00O00000000000000 0-I -
; . ^ .
N U (
. : ~ ' -
1 ;H - : 1 093209006977018877988G' - 96 7 A 101110110000110000000100 0 T A q 000000000000000000.~000000 0 e 3
- D Pf 000000000000000000O00000 l
0 A J
~
S 22 5 4' 4445967124 11 l e S E N uq 11 11 -
- - - -- 0 0 0 0 0 0,0 0 0.:110 5
0 ' b H Cf 00 00 - 000000000000 O a C . T U 11 l
~ : -
O ~ - - 1111111111 1 T 2222222222' J. LL L E l o aN CC . . . . . . ' . . . .LLLLLLLLLLL CCCCCCCCCCC
. S i LLLLLLLLLL
,,CCCCCCCCCC . , , ~ , , c , :i = ,~ , , ,~ '
S r . BB 2 E ec B B B B B B' BPBBB V t e ap 33888888888833333333333 33000000000033333333333W W - R MS 55555555555555C55555555A AAAAAAAAAAAAAAAAAAAA1AAS A S O -
. i T . .
C . J : o
~ -
A E N 11111234123412312312311 R - - - - - - - - - - - - -- - - - - - - '- - - - - e 654122223333444555884786 6 d 110000000000000000000110 55555555555500000055555 , 1 o 22222222222222 222222222 C RRRRRRRRRRRRRRDRRRRRRR2C _ E 3 l l l e e l e sg h eun mra s uera msma
- ssh ool rmeere DTF oowses w e eeea . ; -
TDo , d d g e e e e llllllllll - ld od maanllll ssszllllllllldd lll t eeeasssssssseeeeeelll aad e e n HHHl ssssoooohhhhhh ( eeeeenwow - e n eee rrrl FooooiNNNSSSSSShhhHHa.t NNNN tttt eee . ..SSS nm .g .hA ~ tH Z
. o p uuuetttt eeeelll rrrrrroornrr m sssseeeel1ll o o o sllll t t ssseeeeeett t t sssttt wwwtttlt eoeidgl o lll ennnnuuuueoe nnooooon n e C CCCVIIIIDDOONMNhIILLLBBI .
K - I W .
>A*. ~
1 I
7 .. . . ,.. , ,.
.c u.
..}
e h.
, .. )
.- .l W i.
i ! n y I , s
- as -
% <a . ,
u:
.. . ... " o.
e m. e a h' .o,4
- m. ,
= 1 m ,
.- , '4 '(-
e- -. .,
, .a-y e o vf .. ..
A 3,
- ,o*
c - - n' . . ., 1 . a , n .o,4-a 3 4
~
6-
. w' . -
a. m N !. . h e a a e- k m O P n- p
.o
.a
.. v.
d v u m n G I .M \ M t i-- l v 4 ' en
.g.
M et . 9 [9 q. .
- 9. .
4 em eresre F ****d N
. A- 16 , . ,
e.a.-._.--_------._ -
--..-._..-.--.~._'__-1... _-_.'.a.__.----_ '
..L_ '
..,L..
E l , ;[g"m f " p 1j] [' ,. , ' f- , < , ) ( ! d} , . ' . .,j.'.s
,J ,.k/'
, '., . N " t'( , ' '
, f k k c. .
..'"'J., Tl w ys .t
- o g, j i ,
p' < ' y"E[N %-{[ N ; r 4 7 7
- y a' f . ,
, y >
h ,
. .? /
~
t [, ( : 3 .,.
**i 3.-' 'i^ .-
f p'i i, , '.: , q 'g,., ': , , it
-e-1
,p 5, '.[fj.1 . , i 1 i
/ '
;s L. ' . , ,, Ty( v, 9 . j j' ui f.p.
< t -y;
.1'0%
w: - : ( 1 ;. L e.. -
.I f) .
'v f
y' '. ; !*:. 9' i
.c m' '
^
+
4 y;
- r
(.;g
# g)
,4 i
y, i 1 3., ' i W Mg', j .' ( (d, '"a f f , l.
,e %, (c -
r g L , , ' ' f .,.
's
_ :- ,tft ,.a , '. i ; 1
- y. , .{ ' tn ;f..;. , ;, .;
t.
.! j j . ,
i
' lj .; i i
',- j .l
, <- y. .
p., ,
., .~- i c , ,r., t x< c r. .,! y -l
=* '
.- -2
, )' .
;c A O s .3 , 4nM. JMER( SURFACE --
V ' g ! !v ' [ >
-Qff 'n '
.f 0 r-t .- 1
[ y - ,-
; 4 0. c . , . c -
) r-..t.- ..<. ne ! .
4
," - w r ~ 1/4 TH h e
? *, ,
- a. r .
10, ' ,
* / 3f' W
/ g-p.
u s 4
.gg
)
,_ , , .g
-- ,; . , . n.
N : ,. . , - . ..
.y u, , ,:7; ,< .
e u , 2
. . , i. l t< g/ 1j > o 2
l' ' L.[ - 3/4 T' %
* .'> . i
- T 8
l:v .. s 'j
+ - " ,
<. e f. 4 2,. i> .- \ i w- 1 * #
- p 1
). .j
~
47 s u' m,Q . E. ) t< .
,j.c -o .-r . ,
w 1+
',,,.- .,,1 ;
D h' * ,' D 't . - ,; - N . u,. ' ", , 8 ' '
*- - l l
g . 10' 8 _ , , .t . mq w 2 W 9-z . 3- . 4 i e, f .
,,[ r'
_ ;G - l 4 4 r . w- '3 .
,t c t..
y ,
=,
. .e
.s
)
.a .- ,> .
y . t t 3
- l l
l
, , ;l '
- i
, a 4 ,
1
. 4
- f. - .y 3 a l'1 L3 i 10 0
1-- w
- i { f 't _ _ _f 4..'
5 .10 15 . -20 25 '30 '32 35 SERVICE LIFE: '(EFFECTIVEFULLPOWERYEARS)/ '.! y [ .
- 1 j 4
Pimeq 4 c. Fast neutron fhence (E > 51. MeVj na aL! unction of full'.' g power sevice life -(EFPY)i for Wolf Creek' g
'i 1 .
. q, 1'
e 3 f
,{
- . d' . _>
p, 3 , i9 ; i b' A-17 - 1-n
'ge J
' . ' .- ', b.3
- . . . . . . . . , . . _ . . ,-.m. ....-...
Material Pennarty Ramis Controlling Hataric1: RV Lower Sh:11 g Copper Content: 0.07 Weight 5-Nickel Content: 0.[2 Weight 1 Initial RT 40 7 T Afteh:EFPY RT j 1/4T: 108.0 F l 3/4T: 99.47% l 10 F and 60 PSIG Margins Incorporation for Possible Instrumentation Errors. J 2500 , e i a r I J r t
/ R I C I n i r. d 2250 i I ( 'I f
/ / , I ,
ii , / / l / e i i / f I / / 64 i / > I I 2000 ~ c l l I i I i f I / / 1 / i / LEGEE i / ' 1750 f
/
, f j A - Leak Test Curve B - Heatup RLtes j
/ / [ / ,
j r ; , up to 607/HR / / g 1500 C - Heatup Rakes ' f
/ '
f
/
! I
/
G
- n. --
up to 1007/HR D - Criticality Limit r
/ / i
/ /
/
up to 60 4/HR g 1250 __ E - Criticality Limit / f f
/
f f o up to 1007/HR ' / > / E -- F - Criticality Limit ' ' ' W -- /' j j Based on Inservice [ 1000 --
/
f
/ /
/ f /
a Hydrostatic Test'F) Temperature (267
/
w e.
/' ,
' /
/
'/ / /
750 ' U
/ , / '
h
/
l l/ 500 l s 3 250 j ,
,i c _ ,i i i p l
i e i i '
! i ! I I
) I i
I i f i t 6! l 0 O 50 100 150 200 250 300 350 400 450 500 I INDICATED TEMPERATURE (DEG.F) { Figure A-3. Yolf Creek reactor coolant system heatup limitations . applicable for the first 7 EFPY I A-18
Materini Pmnerty Rnsis Controlling Material: RV Lower Shell Copper Content: 0.07 Weight % Nickel Content: Initial RT 40[70. 2 Weight 5
- RT ET AfteN: EFPY 1/4T: 108.0 0F 3/4T: 99.47 F
^
10 F and 60 PSIG Margins Incorporation for Possible Instrumentation Errors. 2500 , , , i . ,
/
I
/
2250 l ,' '
, i
/
r i
/
i
/
2000 ' j l t 1750 i j
/
! I iI > /
g 1500 l j /' G ' ii , / t
- c. Unacceptable
./
Region w a: 1250 I i
/
/
3 I i
/
\
/
w I
, [ 1000 l ,
/
o I I ll / 8 d - Cooldown Rate /HR 0
,j p Acceptable Region '
u 750 , , , "p,, _J. . e i A 3 I me '
/J'b'/
500 t 0 - -w-- 6
* ! 20 -w A ,
i1 40 'V u , i i ' 60 ' ,
, i 250 ,100 ; i l l i
!'i I i i , ij l ,
ll
! ! ! ! ! l I ' I ! !
il ,,I i i i
'i l l t i1 ! i, ) i ; 6 ) , j "~
O' i ii1 1 1 'I i e! ! ! < 4 11 i ; O 50 100 150 200 250 300 350 400 450 500 INDICATED TEMPERATURE (DEG.F) Figure A-4. Wolf Creek reactor coolant system cooldown limitations . applicable to the first 7 EFPY A-19 '}}