ML19308B535
| ML19308B535 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 07/31/1977 |
| From: | BABCOCK & WILCOX CO. |
| To: | |
| References | |
| BAW-1438, NUDOCS 8001090535 | |
| Download: ML19308B535 (75) | |
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7 BAW-1438 v
July 1977 DF*D *D'T o o . . m ANALYSIS OF CAPSULE OCIII-A FRCH DUKE POWER COF?ANY OLONEE NUCLEAR S D;iT0d, UNIT 3
- Reactor Vessel Materials Survei.' lance Program -
t NOTICE - i THE AT T ACHED Ftt ES ARE OFFICI AL RECORDS OF THE DI SiCN OF DOCUVE N T CONTHOL. T H E Y H AV E t*E E N CHARGED TO YOU FOft A LIMstED It'.1E PElilOD AND I
, MUST BE f*ETURNED TO THE HECORDS F ACILIT Y g l BRANCH 01#3 PL F ASE DO NOT SEND DOCUVENTS i j
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BAW-1438 July 1977 l
l ANALYSIS OF CAPSULE OCIII-A FROM I DUKE POWER COMPANY OCONEE NUCLEAR STATION, UNIT 3
-- Reactor Vessel Materials Surveillance Program --
l by A. L. Lowe, J r. , PE E. T. Chulick
- 11. S. Palme C. L. Whitcursh C. F. Zur11ppe l
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B&W Contract No. 595-7020-71 BABCOCK & WILCOX Power Generation Croup Nuclear Power Generation Division P. O. Box 1260 Lynchburg, Virginia 24505 m a i -
l CONTENTS l
Page l
l 1. -INTRODUCT!CN . . . . . . . . . . . . . . . . . . . . . . .... 1-1
- 2. BACKGROUND . . . . . . . . . . . . . . . . . . . . . . . .... 2-1
- 3. SURVEILLA24E PROGRAM DESCRIPTION . . . . . . . . . . . . .... 3-1
- 4. PHEIRRADI A!!ON TESTS . . . . . . . . . . . . . . . . . . .... 4-1 l
4.1. Ten.<ile Tests . . . . . . . . . . . . . . . . . . .... 4-1 j 4.2. Impact Tests . . . . . .. . .'.. . . . . . . . . . .... 4-1
- 5. POST!RRAD*ATION TESTS . . . . . . . . . . . . . . . . . .... 5-1 5.I. Thernal Monitors . . . . . . . . . . . . . . . . . .... 5-1 5.2. Chenical Analysis . . . . . . . . . . . . . . . . .... 5-1 5.1. Teasile Test Results . . . . . . . . . . . . . . . .... 5-1
- 5. 4 Charry V-Notch Impact Test Results . . . . . . . . .... 5-2 1
- 6. NEUTRON leSIMETRY . . . . . . . . . . . . . . . . . . . .... 6-1 l 6.1. Introduction . . . . . . . . . . . . . . . . . . . .... 6-1 6.2. AnaIy:Ical Approach . . . . . . . . . . . . . . . .... 6-2
- 6. 3. Results . . . . . . . . . . . . . . . . . . . . . .... 6-3
. 7. DISCUSSION OF CAPSULE RESULTS . . . . . . . . . . . . . .... 7-1 i
i 7.1. Preirradiation Property Data . . . . . . . . . . . .... 7-1 7.2. Irradiated Property Data . . . . . . . . . . . . . .... 7-1 7.2.1. Tensile Properties . . . . . . . . . . . . .... 7-1 7.2.2. Impact Properties . . . . . . . . . . . . .... 7-1
- 8. DFTERMINA! 0N OF RCPB PRESSURE-TEMPERATURE LIMITS . . . .... 8-1
- 9. SI'MMARY OF RESULTS . . . . . . . . . . . . . . . . . . . .... 9-1
- 10. SURVEILLANCE CAPSULE REMOVAL SCHEDULE . . . . . . . . . .... 10-1
- 11. CERTIFICA! ION . . . . . . . . . . . . . . . . . . . . . .... 21-1 I la. R Er EREN C Es . . . . . . . . . . . . . . . . . . . . . . . . . . . 12-i t
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-l. .: .r h t a r :. u s af Pre.< 4 u re- Te me rature 1.1-i t curves fe r t n . .ie. ~ J, , '. f c.i S t . N r c ", a Eighth rull h e- Year . . . . . . 8 *. {
A- L . s :r ei:' ance ' ." ra- ?'a t er f a l s ScIceeion Data f.,r (kenee i.. . A-3 c .' . . . : t z r i a '. +
- c. 7 e, i~ o : . ;.p re r ::uryc illance Opsules OCI I'-/ . ' !i-U. ar. r ;:1-E . . . . . . . . . . . . . . . . . A-4 i' A- ) . L leitals a12 peefrcas in L .cr surveillance rupsulet OCIII-T., DC : ! ! -!), and CCIII-F . . . . . . . . . . . . . . . . . A-4 ii- 1. Preirridia f an Tensi;e Propert les of Shell For::ing
."aterial. Heat ASK-M1 . . . . . . . . . . . . . . . . . . . . . B-2
!1- 2 . Preitr.td'ation Tennile ?raperties of Shell For. ting L t e ria l - :!A.' . Hea .ur-191 . . . . . . . . . . . . . . . . . . B-3 B- i . Preirradiatien Tensi'e Preperties ef Shell Ferring .
l L terial. W at AWs-; 2 . . . . . . . . . . . . . . . . . . . . . B *.
R-4 Preirradiat ion Tensi'.e Freperties of Shell For.in.:
".aterial. IL\2. Heat tJ.is- lo2 . . . . . . . . . . . . . . . . . . 5-5 6 i
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L List of Tables (Cont'd)
Table Pege
B-5. Preirradiation Tensile Properties of Weld Metal --
Longi tudinal, WF-209-1B . . . . ...... ... ... .... B-6 C-1. Preirradiation Charpy impact Dats for Shell Forging Material -
Longitudinal Orientation, Ileat ANK-191 . ...... ..... C-2 C-2. Preirradiation Charpy Impact Data for Shell Forging Material -
Transverse Orientation, Heat ANK-191 ...... ..... .. C-3 C-3. Preirradiation Charpy Impact Data for Shell Forging Nbterial -- -
HAZ, Longitudinal Orientation, Ileat ANK-191 . . . . ...... C-4 C-6. Preirradiation Charpy Impact 7 sta for Shell Forging Nbterial.
HAZ, Transverse Orientation, Heat ANK-191 . . . . . . . . . . . C-3 C-5. Prcirradiation Charpy Impact Data for Shell Forging Material - -
Longitudinal orientation, IIeat AWS-192 ....... .. ... C-6 C-6. Preirradiation Charpy Icpact Data for Shell Forging Nbterial -
Transverse Orientation, Heat AWS-192 ........... .. C-7 :
C-7. Preirradiation Charpy Impact uata for Shell Forging Material -- '
HAZ, Longitudinal Orientation, Heat AWS-192 . .. . .... .. C-8 .
I C-8. Preirradiation Charpy Impact Data for Shell Forging Nbterial - ,
HAZ, Transverse Orientation,Ileat AWS-192 . . . ........ C-9 '
C-9. Preirradiation Charpy Inpact Data for Weld Metal, WF-209-1B . . C-10 D-1. Detector Composition and Shielding . ...... ..... . D-2 D-2. Oconee 3. Cycle 1 Neutron Dosimeters ..... .. ..... . D-3 _
D-3. Dosir.eter Activation Cross Sections . ....... ..... . 3-8 m 1 l
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i List of Figures -
Figure = \
1 3-1. Reactor Vessel Cross Section Showing Surveillance Capsule Locations . . . . . . ... .... .. . .. .... . 3-4 -
5-1. )
Charpy Icpact Data From Irradiated Base Metal A, Longitudinal i orientation . . . . . . . . . .. ........ . .... .. 5-5 5-2. Charpy impact Data From Irradiated Base Metal A, Transverse Orientation . . . . . . . . . .. . ...... . .. .. ... 5-6 5-3. Charpy != pact Data From Base Metal A. HAZ, Longitudinal Orientation . . . . . . . . . .... .. .... .. ... .. 5-7 5-4 Charpy Icpact Data From Irradiated Base Metal B. Transverse Orientation . . . . . . . . . .. ... ..... .... ... 5-8 5-5. Charpy != pact Data From Irradiated k* eld Metal . . . ...... 5-9 2 6-1. Fas t Neut ron Fluence of Surveillance Capsule Center Compared to Various tecations Through Reactor Vessel Wall for First 10 EFPY . .. . . . . . . . . ... ... ... . . . ... .. 6-8 7-1. .trradiated Vs Unirradiated Charpy Icpact Properties of Base
>btal . Longitudinal Orientation . .. ... ... ..... .. 7-5 7- 2. Irradiated Vs l'nirradiated Charpy Icpact Properties of Base Metal, Transverse Orientation . .. . . .... .. ... ... 7-6 7-3. Irradiated Vs thirradiated Charpy Impact Properties of RAZ, Transverse Orientation . . . ... ... .... . . ... .. 7-7 l
-v- Babcock a.Wilcox
I List of Figures (Cont'd) i Figure Page -
7- 4 Irradiated Vs Unirradiated Charpy Irpact Properties of Ease
>btal, Transverse Orientation . . . . . . . . . . . . . . .. 7-8 7-5. Irradiateil Vs Unirradiated Charpy Irpact Pronerties of Weld !ktal . . . . . . . . . . . . . . . . . . . . . . . . .. 7-9 8-1. Predicted Fast Neutren Fluences at Various Locations Through Reactor Vessel Wall for First 10 EFPY (Oconee 3) . . . . . .. 8-5 8-2. Reactor Vessel Pressure-Temperature Limit Curves for Norral ,
Operat ion -- Heatup, Applicable for First Eight Ef fective f l
[ Full-Power Years (Oconee 3) . . . . . . . . . . . . . . . ... 8-6 j
( 8-3. Reacter Yeasel Pressure-Temperature Limit Curve f o r No r=al , l Operat ion - Cooldown, Applicable for First Eight Effective g l Full-Pever Yea rs (Oconee 3) . . . . . . . . . . . . . . .. . . 8-7 l M- 4 Reactor Vessel Pressure-Temperature Limit Curve for Inservice Leak and Ilydrostatic Tests. Applicable for First Eight Effective Full-Power Years (Ocence 3) . . . . . . . . . . . .. B- S A-1. Location and Identi fication of Materials Used in Fabrication of Fe.etor Pressure Vessel . . . . . . . . . . . . . . . ... A-5 I C-1. Cha rpy Impact Data Fron Unirradiated Base Metal A. l Lor.gi t udinal orientation . . . . . . . . . . . . . . . .... C-11 l
, C-2. Cha r;e lepact Data From t*nirradiated Base Metal A, !
Transverse Orientat ion . . . . . . . . . . . . . . . . . . .. C-12 C l. Charpv lepact Data Fro . t*nirradiated Base Metal A. !!AZ, Longitudinal orientation . . . . . . . . . . . . . . . . .. . C-13 C-4 Charpy Impact Data Fron Unirradiated Base Metal A. HAZ, Transverse Orientation . . . . . . . . . . . . . . . . . . .. C-14 C- 5. CS.i r;'y Impast flata Fron Unirradiated Base Metal S, Longit w11nal orie.itat ion . . . . . . . . . . . . . . . . . . . C-15 C-6 Charpv Irpact Data Frcm Unirradiated Base Metal B.
T rar.s ve rse Orientatien . . . . . . . . . . . . . . . . . .. . C-16 C-7. Cnarpy inpact Data from i*nitradiated Base P.etal is, ilAZ, l.ongi tudinal Orientation . . . . . . . . . . . . . . . .... C-17 C-M. Cnarpy impact Data From tinitradiated Base Sktal 3, IIAZ, T ransverse Orientat ion . . . . . . . . . . . . . . . . .. .. C-18 C-9. Charpy Impact . lata Fron Unirradiated Weld >ktal . . . . . ... C-19 1
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- 1. INTRODUCTION This report describes the results of the examination of the first capsule of Dube Power Company's Oconee Nucicar Station, Unit 3 (Oconee 3) reactor vessel
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surveillance program. The objective of the program is to monitor the effects of neutron irradiation on the tensile and impact properties of reactor pres-sure vessel materials under actual operating conditions. The surveillance program for Oconee 3 was designed and furnished by Babcock & Wilcox; it is described in BAW-10006A, Revision 3.I The program was planned to monitor the effects of neutron irradiation on the reactor vessel materials for the 40-year design life of the reactor pressure vessel.
The surveillance program for Oconee 3 was designed in accordance with E185-66 and has been updated to meet the intent of the first draft of ASTM E185-73.
The prog:am is not in compliance with Appendixes C ar.d H to 10 CFR 50 since the requirenents did not exist at the ti=e it was designed. Eccause of this difference, additional teste and evaluations were required to ensure teeting the requirements of these appendixes. The recommendations for the future operation of Oconce 3 included in this report do comply with these require-ments.
t 1-1 Babcock s. Wilcox
- 2. BACKGROUND The ability of the reactor pressure vessel to resist fracture is the primary factor in ensuring the safety of the primary system in light water cooled re-actors. The beltline region of the reactor vessel is the most critical re-gion of the vessel because it is exposed to neutron irradiation. The general
[ ef f ects of fast meutron irradiatien on the mechanical properties of such low-l i alloy ferritic steels as the SA503. Class 2 used in the fabrication of the Oconee 3 reactor vessel are well characterized and documented in the litera-ture. The low-alloy ferritic steel used in the beltline region of reactor vessels exhibit an increase in ultimate and yield strength properties with a corresponding decrease in ductility after irradiation. In reactor pressure vessel stects. the most sericus mechanical property change is the increase in te=perature for the transition f rom brittle to ductile fracture accompanied by a reduction in t he upper shelf impact strength.
Appendix G to 10 CFR 50. " Fracture Toughness Requirements." specifies minimum fracture toughness requit znents for the ferritic materials of the pressure-
! retaining co=penents of the reactor coolant pressure boundary (RCPB) of water-cooled po-tr reactors and provides specific guidelit.es for determining the pressure-tecperature limitations on operation of the RCFB. The toughness and operational requirements are specified to provide adequate safety nargins dur-ing any condition of normal operation. including anticipated operational occur-rences and syste= hydrostatic tests, to which the pressure houndary may be sub-jected over it s service lifetice. Although the rec,uirements of Appendix G to 10 CFR 50 beca:e effective on August 13. 1973, the. requirements are applicable l to all boiling and pressurized water-cooled nuclear power reactors, including thoss under construction or in operation on the effective date.
Apprndix H f., 10 CFR 50, " Reactor Yessel Materials Surveillance Program Re-quirenents, defines the material surveillance program esquired to monitor l chaages in the fracture toughness properties of ferritic materials in the rc-actor vessel beltline region of water-cooled reactors resulting from exposure 2-1 Babcock a.Wilcox I i
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to neut ren irr.diation and the t herm al envirennent. fra:ture t ou.:hne s s test data inre obt.s itwd t r o . r.a t e r i a l spee tr. ens sithdrawn perio.!ically fran ti.e re-actor ve%.l. Th.se d at a will ; cr=it detter.in.ition af the conditiens n: der which toe vv. s1 caa be nptrated with adequate sa:et y rar.; ins a.atnst i rr c t .:re
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t h roua.heu t it. acrvice life.
.\ nethed f or mrd ier, . spir at brittle fracture in reacter pressure vessels !~
described la ppenJ1x G to the 7 91E Boiler and Pressure Vessel Code. Sectio, Ill, thia r.thod utilites fracture necnantes concepts ar.d the reference nil-ductilitf t.
.pe r.it ure. RTND.. which is detined 1s the .tre ter of the drop weight n i l-h t i l i t :. traasition tsnperature (per .bMI E-205) or the teepera-ture t h '. t 1. e.0F 5elow that at which t he caterial exhibits 50 t t-1, and .55 l mits lateral e pan ton. the RT. o f a g iv n :.at e r ia l is used to index that 1
. iter ia l t o a rv f e renc.: stress intensit y f actor curve (K yg c u ne ) . @M ar a
- ica rs in .V ; caJ t x G a
- .U.'tE See t loa Ill. The K g curve is a lower bound of dynamic. at.it;c. and crack arre.t fracture toughness resalts obtained from i
8vveral heat. of ;>ressure vessel steel. '.Chen a given material is indexed to I the K;g curva. .i l l owah l s stress inten.sity factors can be obtained for thi . :u-t e r i a l a.. .i t ract ien on temperature. /dlowable operating li: nits can then be l' l
i a t e rnit.ed w thu allewable stress .aten..ity ' actors.
i . .e H,.,...b,_ a n J . In turn. the operatinr. lir.its of a nuclear power plant. can be l l su t u t e.: t.. .ia e rn t 'or ihe effeets of radiation on the propertlea of the re-
..c t o r 'e t . .. e ' r.a t e r i a l e . The railla t ion .e hri t t le cat anc the re.altant changes la r:cchan ical properties of a given pressure vessel stes-1 can be canitored by surveillante :.rwe r in in which a r;urveillance capsule cor.taining prepared speci-r .. n . o n the r :.e t or ve-;..el na t e r ial s is periodically re oved tres the operating I
nw l e.i r reac t ..r a n.1 t i.e specimens tested. The increase in the Garry V-notch W it-ib '. e . : erature, or the incre ue in the M nils of litteral expansion t em-
- era t ure. -
hi.ver results in the larger tenperature 4.i f t due to irradiation.
is .i!ded to the oritinal RTy.g to id j ust it for radiatica embrittlement. This adlusted is used to index the naterial to the Kg curve, which, in turn RT.,.
fs used to et operitlug i i n i t >. ior the nuclear power plant. TEcse new linits take into aceeunt the eticets of Irradiat ion on the reactor vessel materials.
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.,_4 Babcock a. Wilcox i
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- 3. SURVEILLANCE PROC?.A". DESCRIPTIC"'
The surveillance program for oconce 3 comprises six survei.*ance ecpsules de-signed to monitor the effects of neutron an.1 thermal environment on the mate-rials of the reactor pressure vessel core region. The capsules, which were inserted into the reactor vessel before initial plant startup, were positioned
( inside the reactor vessel between the ther=al shield and the vessel wall at the locations shown in Figure 3-1. The capsales, placed two in each holdcr tube, are positioned near the peak axia! and azi=uthat neutron flux. BA'a'- -
10C06A, Revision 3, includes a full description of capsule locations and design.1 Capsule OCIII-A was recoved during the first refuelicg shutdown of Oconee 3.
This capsule contained Charpy V-notch icpact and tensile specimens fabricated of SAind. Class 2 steel, weld metal, and correlation steel. The specimen con-tained in the capsule is described in Tahle 3-1, and the chenistry and heat treatment of the surveillance material are described in Table 3-2.
All test specinens were cachined f rom the 1/ ~-thickness location of the shell forgings. Charpy '.'-notch and tensile specimens f rom the vessel material were oriented with their longitudinal axes parallel to the principal working di-rection of the forging; the specimens were also oriented transverse to the principal working direction. Capsule OCIII-A contained dosime*er wires, de-scribed as follows:
Dositeter wire Shielding U-Al alloy Cd-Ag alley N P-Al alloy Cd-Ag alley Nickel Cd-Ag allcy 0.66% Co-Al alloy Cd 0.66% Co-Al alloy None Ye None 3-1 Babcock s. Wilcox
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Th2rmal monitors of low-melti g eutectic alloys were included in the capsule.
Tha eutectic alloys and their celting points are as follows:
Alloy Melting point. F 90% Pb, 5% Ag, 5% Sn 553 97.5Z Pb, 2.5% Ag 580 97.5% Pb, 1.5% Ag, 1.0% Sn 588 Lead 621 Cadmium 610 Table 3-1. Specitens in Surveillance Capsule OCIII-A i
No. of specicens Material description Tensile Charpy i
k' eld metal, kT-209-1B 3 12 lleat-affected zone (llAZ) lleat A - A'iK-191, longitud. 0 12 Baseline r:aterial j llea t A - M;K-191, longitud. 0 9 Ileat A - M;K-191, transverse 2 12 ,
Ilea t B - Ak'S-192, traasverse 0 9 Total per capsule 4 54
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Table 3-2. Chemistry and Haat Treatment of Surveillance Materials j l
Chemical Analysir He.t Heat Weld metal (b) kT-209-1B Element A'?K-lh **) AWS-192 C 0.24 0.22 0.067 Mn 0.72 0.58 1.56 P 0.014 0.011 0.020 S 0.012 0.015 0.005 S1 0.21 0.24 0.56 Ni 0.76 0.73 0.48 Mo 0.62 0.60 0.33 Cu 0.02 0.01 0.30 ifeat Trmtcent Time, Heat * *o . Temp, F b Cooli x;K-191 " 1620-1660 4O Water quench 1579-1610 4.0 Water quench 1230-1270 10.0 Water quench 1100-1150 40.0 Furnace-cooled AWS-192(* 1620-1 % 0 4.0 Water quench :
1570-1610 4.0 Water quench 1220-1260 10.0 Water quench 1100-1150 40.0 Furnace-co. .d
?-209-18 0'I 110C 50 30.0 Fu nace-cooled Per certificate of test.
}Per veld certification.
4 3-3 Babcock s. Wilcox V-
Figure 3-1. Reactor Vessel Cross Section Showing Surveillance Capsule locations X Surveillance Capsule Holder Tubes - Capsules CCIII-A, OCIll-D
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- 4. PREIRRADIATION TESTS ;
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L'airradiated material was evaluated for two purposes: (1) to establish a I baseline of data to which irradiated prcperties data ceuld be referenced, an3 (2) to determine those materials properties to the extent practical from avail- 4 able material, as required for compliance with Appendizes G and 11 to 10 CFR 50. j l
6.1. Tensile Testa l Tensile specieens were fabricated from the reactor vessel shell course forging and weld metal. The subsize specimens were 4.25 inches long with a reduced section 1.750 inches long by 0.357 inch in diameter. They were tested on a 20,000 *b-load-capacity universal test machine at a crosshead speed of 0.005 inch per minute. Test conditions were in accordance with the applicable re-quirements of ASTM A370-72. For each material type and/or condition, six specimens in groups of three were tested at both room temperature and 570F.
An LVDT-type clamp-end screw-on extensometer was used to determine the 0.2%
yield point.
The tension-compression load cell used had a certified accuracy of bet ter than ? 0.5% of full scale (10,000 lb). All test data for the pre-Irradiation tensile specimens are given in Appendix B.
4.2. Impact Tests Charry V-notch impact tests were conducted in accordance with the requirements of ASTM Standard Methods A370-72 and E23-72 on a remote controlled i= pact test- -
er cert ified to meet Watertown standards. Test specimens were of the Charpy V-co:ch type, which are 0.394 inch square and 2.165 inches long.
Prior to testing, specimens were temperature-conditioned in a combination re-sistance-heateJ/ carbon dioxide-cooled chamber, designed to cover the tempera-ture range from -85 to +550F. The specimen support arm, which is linked to the pneumatic transfer mechanism, is instrumented with a contacting thermo-couple allowing instantaneous specimen temperature determinations. Specimens were transferred frem the conditiening chamber to the test frame anvil and pre-cisely pretest-positioned with a fully automated, remotely controlled apparatus.
I l
i l
l t
l Transfer times were less than 3 seconde and repeat within 0.1 second. Once the specimen was positioned, the electrcnic interlock opened, and the pendulum was released from its preset drop height. After failing, the specimen, the hammer pendulum was slowed on its return stroke and raised back to its start poeition. Impact test data far the unirradiated baseline reference materials j are presented in Appendix C. Tables C-1 through C-9 contain the basis data which are plotted in Figures C-1 through C-9.
k l
- 5. POSTIRRADIATION TESTS 5.1. Thermal Monitors Surveillance capsule OCIII-A contained three temperature monitor holde- tubes,
- each containinA five fusible alloys with dif ferent melting points ranging from
{ 558 to 621F. .111 the thermal monitors at 558 and 580 had melted, while those at 588 and 610F showed slight celting, indicating a short time at temperature.
This may have occurred as part of hot functional testing during reactor start-up. The monitor at 621F remained in its original configuration as initially placed in the capsule. From these data it was concluded that the irradiated l
specicens had been exposed to a maximum temperature in the range of 580 to less than 610F daring the reactor vessel operating period. There appeared to be no significant temperature gradient along the capsule length.
5.2. Chenical_ Analysis Two broken inpact specimen halves taken at random from each of the two unir-radiated base metals (heats ANK-191 and AWS-192) and the weld metal (WF-209-IB) i were analyzed f or nickel, copper, phosphorus, and sulfur contents to verify original mill test report data. The analyses were perfor=ed using an emission spectrograph c31thrated with standards traceable to the National Bureau of Standards.
5.3. Tensile Test Results The results of the postirradiation tensile tests are presented in Table 5-2.
Tests were performed on specimens at both room temperature and 580F using the same test procedures and techniques used to test the unieradiated specimens (sect ion 4.1). In general, the ultimate yield strer.gth of the material in-
- l creased slightly with a corresponding slight decrease in ductility; both ef-l fects were the result of neutron radiation damage. The type of behavior ob-served and the degree to which the material properties changed is within the range of changes to be expected for the radiation environment to which the specimens were exposed.
T! e results of the preirradiation tensile tests are presented in Appendix B.
5.4. Charpy V-Notch Impact Test Results The test results from the irradiated Charpy V-notch specimens of the reactor vessel beltline caterial and the correlation monitor material are presented in Tables 5-3 and 5-4 and Figures 5-1 through 5-5. The test procedures and gechniques were the same as those used to test the unirradiated specimens (section 4.2). The data show that the material exhibited a sensitivity to Irradiation with!.n the values predicted from its chemical composition and the fluence to which it was exposed.
4 The results of the ; .elrradiation Charpy V-notch impact test are given in Ap-peadix C. j s:
k Table 5-1. Chemistry Data on L'nirradiated RVSP Material F
CVN- 1 specimen >bterial type /
"U *' "' **
No. heat No. Ni ? 0.01 Cu 0.02 Pt 0.003 S2 0.004 l
JJ 818 P,ase/ANK-191 0.75 0.05 0.012 0.007 l
KK O27 Base /AWS-192 0.74 0.06 0.013 0.015 JJ 091 Weld /WF-209-1B 0.58 0.29 0.017 0.010 i
Table 5-2. Tensile Properties of Ca sule OCIII-A Base and l'ald Metals Irradi_ated to 7.39 = 10 I nyt (E > 1 MeV) .
S*# "Etfl' psi __ Elongation. T Test Specimen temp, Yield Ult. Uniform Total Red'n of ID No. F (YS) (UTS) (UE) (TE) area, %
Base Metal - lleat ANK-191, Transverse JJ 616 RT 63,015 87,490 11.77 29.93 68.4 619 580 57,210 86,360 11.56 28.71 67.4 I Weld Met al -- VF-209-1B ,,
JJ 006 RT 79,340 93,940 12.18 27.29 62.6 012 580 72,630 93,760 11.09 22.S6 60.3
- Babcock & \Milcox
Table 5-3. Charpy impact Data for Capsule CCIII-A Base Metal 1rradiated to 7.39 - 1017 nyt (E > 1 MeV)
Test Abs Lateral Shear Specimen teep, energy. expansion, fracture.
ID No. F ft-lb 103 in. %
3ase Metal - Heat _._ ANK-191, Loegitudinal JJ 806 361 154 69 l'00 807 239 150 72 100 835 180 131 68 75 814 139 124 71 65 815 84 110 74 20 828 65 52 45 4
- 819 35 43 37 3 838 41 60 43 2 847 19 36 26 O Base Metal - Heat ANK-191, Transverse JJ 613 360 124 73 100 619 280 136 71 100 632 200 130 69 100 644 139 116 76 70 647 87 92 66 45 668 64 S3 42 S 650 40 73 54 12 681 31 56 41 6 666 20 45 34 3 689 4 15 9 1 654 0 44 32 3 691 -20 8 2 O HA2 - Heat ANK-191, Transverse JJ 400 3
- 9 102 50 100 401 281 133 66 100 402 200 139 65 100 410 180 151 65 100 403 141 159 71 100 412 121 103 56 82 404 84 77 43 45 414 60 110 52 97 407 39 46 25 30 415 21 35 21 22 409 0.7 31 18 8
[ 420 -40 44 21 6 1
5-3 Babcock t.Wilcox
I Table 5-4. Charpy impact Data for Capsule OCI!1-A Base Metal Trradiated to 7.39 = 1017 nyt (E > 1 MeV)
Test Abs Lateral Shear Specimen temp, energy, expansion, fracture.
ID No. F ft-lb 103 in. !
Base Metal - lleat AWS-192, Transverse KK 612 359 ' 100 73 100 621 280 97 64 100 628 198 104 65 100 632 141 88 62 75 670 110 60 47 40 3 647 84 56 46 10 650 85 52 44 6 662 60 45 36 6 '
658 40 16 13 <1 +f Table 5-5. Charpy impact Data for Capsule OCIII-A Weld Metsi (VF-209-1B) Irradiated to I
7.39 < 1017 nvt (E > 1 MeV)
Test Abs Lateral Shear Specimen temp, energy, expansion, fracture,
__ID No. F ft-lb 103 in. 2 i JJ 014 359 62 51 100 015 279 56 47 100 017 199 54 47 99 041 169 54 45 99 027 140 46 32 94 046 124 50 30 90 028 109 53 40 97 071 95 26 16 45 029 84 34 28 25 085 60 25 20 15 036 39 37 26 28 096 21 28 21 14 5-4 Babcock & %Vilcox e
Fi gu re- 5-1. Charpy I:: pact Data Fro:n Irradiated Base Metal A. Longitudinal Orientation IT ,
. , , , . . ~ -
< 75 .
- e 1
- 50 .___._ __________ _ _ _ . - - - - - - - - - - - - _ . - .
=
d 25 - -
r, - - - - - - ' - - - - -
.G80 . . . . . i i . . . -
4 .
g .rfg . .
5 t ; e na ,oug . .
r - _ _ _ _ . . _ _ _ _ - . . _ _ _ _ _ _ _
7 e
t .020 - -
~~
r e e e e i e i . .
,,, i 2M . . . , , . . ,
3414 50 W RY pe, , T,,, __N A ,
l In (15 rul _32F
( 160 InISO 't-tD) 37F ,
l
- q. m t,,g3 155 ft-Ib _
- RI , -2 31' ( Be s t estimate) 140 .
. e
-& 120 e
I 3 1m - -
5' w
Y Sn . .
t.
t
m . e -
________ _t_______________________________.
40 - -
%rtenai. SA 309. C1 2 Qis tuvarion i.on gitud inal
'O -
Funact 7.39 = 1037 nvt HEAT Nwete ANI~191 p r r i i . , , , , ,
-80 -W D SO 80 120 160 2m 20 280 5n zo vn Test itsetsaruet, F 5-5 Babcock 8. Wilcox
i 1
Figure 5-2. Charpy Impset Data Froci Irradiated Base
, Metal A. Transverse Orientation IN . . . . .
1 l
75 -
'4 __ _ .- _ _ _ _ _ _ _ _ _ _ _ _ _ - .
$c - _.
.ll'
- l 5 -
5 J, 2s -
n .. . . . . . .
(
.D80 . . . . . .2 . . . .
i f e i .ith -
? .
?
.. .ty.o .
- )F
= . _- _ _ _ _ _. _ _ _ _ _ _ _ _ _ _. _ _ . . ~
g . ;
- i
- f .020 i . 1 i i
.5 , , . .
i e i , , e i 2M . . . . . . . . . . .
".AI4 S;l W PY l
NA I B*. . y act - .
~ 23F 3 (35 e l 2sr m . t. <so .r _
q.u,z g w 135 ft-Ib ~
186C R T,, -35F (Best estimate) ,
=
& IM -
N -
=
31%
{
a L 80 -
Q . {
t
- 6C e ~
teC
%ftsig SA 508 C1 2 l Onsturarson Transverse g D ~
. Fwract 7. 39 = 10 17 nvt ~ j Haar mma ANK-191 0= 40 80 20 2M 3M N .m 4.
jl
-60 -40 0 120 150 200 '
Test Tewcaarvet, F 5-6 Babcock 8. Wilcox i \
l l l
Figure 5-3. Charpy Irpact Data Fron Base Metal A.
HAZ. lone,itudinal Orientation try) , , , , ,
- 75 -
J at 50 _- -
O *
=
Y e -
m 25 -
e
. . f . . .' f .
g . .
.080 . . . . . . . . . . .
i e e e g .W .
g
- e
=
w, ,w . .
s . _ -_ .--- __ _.------- _ - _---__-
E ar e i* .020 -
e , e e , e i e . .
p i 2m , , , , , , , , , , ,
- CATA SUWRY N'\
If .I nct To (35 esp _ft2F
]g .Y g ISO tr-ts) W . .
(.yz (a,o 146 ft-lb e
, gar .c, -17F (Best estimate) , . I l
- I
. I E 120 1
- I
. e 1 l
.; IOC - -
a 80 -
O.
t
~ 60 -
40 - .
, Nitens SA_503. C1 2 Deirutation HAZ-Trans.
20 -
Funnes 7. 39 = 1017 nyt
- HEAT Mweca ANV-191 o
-80 -43 O 40 80 120 150 200 20 260 3M %9 @
itst itw taarves, F )
.. l
., _7 Babcock & Wilcox l
Fi gure 5-4. Charpy I: pact Data From Irradiated Base
- et al 5 , !*ansverse Crientation Im , , , , , , . . , .
- o. 15 -
1 e
4 yw ______________ _ _________________.
f .
m 25 -
9 f f e e a y a g i e e
- 5 5 5 5 3 I E 3 5 s r i
e g .ON) -
i
- O .0@ - . ht 5
E
.020 $
R L I f I f f f f f f e e u a a e a g : a g , a
">TA $;; WRY T NA len not .
T3 USrw 61rF 160 T(y (50 st-sa) 78F .
(.'fl (A,s) 101 ft-lb 140 - RT., +18F (best estimate) .
1
. 1 E 121 -
1 3 -
.7 110 , l 5
O 80 - .
- t t
- M - . .
40 - .
%ftniat SA 508. c1 2 .
Cattuvation Transyerse q d
20 - ~
Funnet 7.39 = 1017 nvt 1 HEAT Nuesta AFS-192 .
i . . , , , , , , , , .
o E0 -40 0 40 60 123 160 2M ?43 281 529 59 4% 1 itst itsetaatuat, F 5-s Babcock & Wilcox
1 i
I Figure 5-5. Charpy Icpact Data Fro:s Irradiated Weld Metal im . . . . .. -
". 75 Y
it y rA ---___--____ ____________________.
w I
3n -
- e -
e e e e e e n g a e e e a e
.NJ . . . . . . . . . . -
4 5 .%0 - -
5 .
e w .M -
e ,
5 - - - - - _ - _ - - - - - - - - - _ - _ - - _ - - - - _ _ _ - - - - - _ _ . i g , e * ,
a e t I 2
- .f20 l I 4
_"E f 9 f f I t 1 9 1 e e 2tV) . . . . . . . . . . .
DATA SUTWAY T NA l@ et -
J 120F Tn (35 at) 160 Ic, (50 st-u) 135F ,
(-USE(avs) 58 ft-1b
. it.0 - ti , +75F (Best estimate) .
L 120 -
1 2
- IT -
5 r
w M -
. i
- l t.o t - '
l
- w . ,
.
- i
~_-------- ---_, ____.__- _-___ _________ 1 40 -
e %rteras. Weld Metal e e bitsfatI0se
~
20 -
Fueset 7.19 = 10I7 evt" ht nween W-209-1B ,_,
o e i . . , , , , , , ..
~
-m -40 o na ao 120 no 200 2c 280 sa 5m 40n itst Ta m aavuot, F 5-9 Babcock & Wilcox 1
i L
~
- 6. NEUTRON DOSIMEIRY 6.1. Intreduction A'signfileast aspect of the surveillance program is to provide a correlation b:stween thd neutron fluence above 1 MeV and the radiation-icduced property change noted in the surveillance specicen. To permit such a cerrelation, act ivation detectors with reaction thresholds in the energy range of interest were plased in each sur* sillance capsule. The properties of interest for the detectors are given in lable 6-1.
Because of a long half-life (30 years) and an effective energy range of more than 0.5 v.e'.'. only the ceasurements of 137 0s production f rom fission reac'.lons i':!p (and 3 *U) are directly applicable to analytical determinations of In the-fant neutron (E e 1 MeV) fluence during cycle 1. The other dosimeter reac-tions are u>eful as corroborating data for shorter time intervals and/or higher energy :1uxes. Short-lived isotope activities are representative of reactor co:tditions oser t he latter portion of the irradiation period (fuel cycle) only, whereas reactions with a high threshold energy do not record a significant p., r t ni the total fast ilux.
The energy-dependent neutron flux is not directly available from activat ion detectors because the process provides the integrated effect of the neutron flux en the target naterial as a function of both irradiatien ti=c r;d neutron energy.. To obtain an accurate esticate of the average neutron flux inci*ent upon the de:ector, several parameters cost be known: the operating history of the reactor., the energy response of the given detector, ar.d the neutron spec-t rum at the detector location'. Of these parameters, the definition of the azutron spectrum is the most difficult to obtain; essentially, two means are availabic: . iterative unfolding of experimental foil data and analytical ce thod s. 'Because of a lack of sufficient threshold foil detectors satisfying both the threshold energy and half-life requirements necessary for a surveil-Irnce program, iterative unfolding could not be used. This leaves the speci-fication of the neutron spectrum to the analytical method.
6-1 Babcock a. Wilcox ,
I i
l 4
l 9
. i.
6.2. Analytical Approach Energy-dependent neutron fluxes seen by the detector .+re determined by a dis-crete ordinate solution of the Boltzmann transport equat ion. Specifically.
ANISNI , a one-dimensional coJc, and DOT", a two-dimenslenal code, were used to
- calculate the flux at the detector position. In both codes, the Oconee system g was modeled radially from the core out to the air gap outside the pressure ves-sel. The model included the core with a time-averaged radial pewer distribu-tien, core liner, barrel, thermal shield, pressure vessel, and water reglens.
Inclu' ling the internal components is necessary to account for the distortions of the required er. orgy spectrum by attenuation in these components. The ANISN code used the CnJK5 22-group neutron cross section set with an St. order of an-gular quadratu. and a P3 expansion of the scattering natrix. The problen was run alcng a radt ; across the core flats. Azimuthal variations ware obtained with a DOT r-theta calculation that modeled a one-eighth plan view of the core at its midplane and included a pin-by-pin, time-averaged power distribut ion.
The DOT calcualtion used S.:. quadrature and a Pt cross section set derived from c.s x.
Fluxes caler'ated with this DOT model must be adjusted to account for lach of Ps crosa section detail in calculations et anisotropic scattering, a pertur-bation caused by the presence of the capsule, and the ar.ial power distribution. g The first two items are both energy- z.nd radial location-dependent. A P /P; 3 correct ion f actor was obtained by comparing two ANISM o.ae-dicensional model calculat ions .in which only the order of scattering was varf ed. The capsule perturbat ion factor was ob tained f rom a comparison of two DOT x-y nodel calcu-lations, one with a capsule explicitly modeled -- SS 304 cladding, aluminu=
fliier region, and carbon steel specimens and the cther with water in those regions. The effect of axial power distribution was d-termined from burnun calculations as a function of axial location for the outer rows of fuel as-semblies. The net. result f rom these parametric studies was a flux adjustment factor K(C), which should be applicable to the appropriate dosineters in all 177-luel assembly surveillance programs in which the capsules are located at a radius of 211 en f rom the core center and 11 degrees from a sajor axia s' axial relative power distributions are similar. (See Table 6-2.)
e
. Tha calculation described above provides the neutren flux as a function of ensrgy at the detector position. These calculated data are used in the 6-2 Babcock s. Wilcox
i l '
following equations to obtain the calculated activities used for comparison with the experitental values. The basic equation 6 for the activity D (in kC1/g) is as follows:
" ~
i*j "~ 1( ~*j) (~
Dg CN 1
=gi- 3.7
- 10* i E
n }
j=1 j( ~"
where C = normalizing constant, ratio of measured to calculated flux, N = Avagadro's number, A g = atomic weight of target materici 1, f = either weight fraction of target isotope in nth material g
or fission yield of desired isotope,
)"(E) = group-averaged cross sections for material n (listed in Table D-3),
- (E) = group-averaged fluxes calculated by DOT analysis.
F) * (raction of full power during jtb time in;erval, t ,
i = decay constant of ith materimi,
- = interval of power history, T = sun of total irradiation time, i.e., residual time in reac-tor, and wait time between reactor shutdown and counting,
= cumulative time from reactor startup to end of jth time 1
period, i.e., r = r t j j 9;; k.
l l
The normalizing tonstant C can be obtained by equaring the right s!de of equa-tien 6-1 te the neasured activity. With C specified, the neutron fluence greater than 1 Mi? can be calculated f rom 15 MeV 11
- (E ' l.0 MeV) = C [ :- (E) j*I 3
[ Ft (6-2)
E=1 where it is the number of irradiation time intervals; the other values are defined above.
6.3. Results l
Calculated act _vities are compared to measurements of the dosimeters in Table !
6-3. The 127Cs data show that fast flu:: (E > 1 MeV) is somewhat underpredicted (10 to 15%) by the analytical model described herein (if one assumes that the calculated flux spectrut is correct). Such agreement is probably within the
l i
l tancertainty 11=its of this analysi 6 A conservatice nurmalizaticn f actor of 1.1 was selected based on the similarity of results to previou+ analyses and f 2J comparison to the other isotopie data (range 0.83 to 1.23). Be:5 % and - Co ,
I showed an overprediction of flux with Eel ?!cV (17 and 'C), wnich ladicates that the .alcul.sted spectrt.m could be sonewhat harder (skewed teward higher energies) than measure.l. Although not report ed herein, addit ienal isotopic data ( Cc. - ' Ro. " Zr) were obtained to corroborate the selution of a 1.1 norr:.nl izat ion f act or.
The results of this analysis are consistent with previous 177-fuel assembly sirveillance erecic.en analyses.- Future capsule data should add confidence to ,
the analyt ical prc,cedure and possibly clarify variat ions in the normalization
- l constant.
ha-cd on a norc.alization tonstant of 1.1, an average fast flux for cycle I was 9
calei. lated at tne capsule location and at the inside surface ot the prsssure
- v. -cl w l ! . Ihe data (rable 6-5) were converted to fast fluer.cc values of
. H 10- 4m at the capsule center and 4.14 - 10l ' n /en- at the pressure "c %. 1 al: ?..r 1.i days of cycle I at the full power rating c: 2 364 Wt.
- l. !u.m e al. nl..t ed f or t he pressure vesrel wall (inside .4urt ace) refer to the t 1 n ; na r1-ax. .nich c.sv be located at .i different azimuthal and axial position l i
ban t ue surveillance capsule. In this analysis the maximum fluence at the g l
- = r e a.
- e .c m : etcurred at an sizinuthat position of S* :ren a n lor axis (cap- ;
vale 1.wated : !!') and about 90 cm above the lower act ive fuel line. This is i functjen e- p,.er distribution in the core. The effect of extending the flux '
.r . i n ,:e acwn to 0.1 MeV was to approxic:ately double the fluence at the capsule and the pre ure se.ssel. Since the same normalization factor (1.1) is used addi-tieaa' unc.rtainty is :ntroduced in this result because none of the dosineter react:ons . ire cit eet 1se over this ent ire energy range.
-d on t he nurveillance sanple analysis for cycle 1 and predicted core leak- ,
ar.d fuel burnups, for future cycles, pressure vessel fluence was predicted up to 10 FFPY (ef f ect ive full-pow,;r years). These data are listed in Table b-6 and should 5e more definitive than the generic design valu o in reference h
- 9. Figure 6-1 gives a coc;iarison of the fast neutron fluence of the surveil-lance cap 8ule center relative to variou.S locations through the reactor vessel .
wall during t irst 10 EFPY.
i i i
l l
l
l 1
l ,
Table 6-1. Surveillance Capsule Detectors-Threshold Isotope Detector reaction energy. MeV half-life i
- Fe (n.p) !'Mn >2.5 303 days
- ~Ni(n.p)!~Co >2.3 71. 3 days
~~U(n f)!37Cs 21.1 30 ytars I 7 Nptn f)I'7Cs >0.5 30 years
- U(n.f)I2?Ru >l.1 39.5 days
' 5p(n,f)l0'Ru 20.5 39.5 days i
Table _6-1._ Flux AJjustment Factor Energy Axial power Capsule r_sf ict "eV factor 3 perturb'n K(E)
.1 1.17 1.22 1. 51 1.90
"' S 1.11 1.23 1.26 1.81
^! " 1.17 1.23 1.14 1.71
- 2. 3 1.17 1.25 1.00 1.46 2.5 1.17 1.25 0.99 1.45 Table 6-3. Dosimeter Activations A B neasured calculated C = A/B activity.(a) activity, norm.il iza t ion R*d
- ion __aC_1/n p C i /_g__ constant
'Te(n.psS'Mn 363 437 0.33
Ni(n.p)t+Co 719 787 0.91
'~r(n.f)l*7Cs 1.05 0.98 1.07 l 'hPl=.f)II7Cs 5.42 4.n$ 1,g7 "C(n.f)" 'Ru 38.7 34.4 1.12
-'*7 Np(n.f)!Ru 180 146 1.23 (a) Average of four dosimeter wires from Table D-2.
i l
Table _6-4 Normalized Flux Spectra, Ee 1 MeV In water Energy range, near pressure 2 2 '- U MeV vessel w.ill f i ss ion _
l l 12.2-15.0 0.0016 0.0002 10.0-12.2 0.0064 0.0013 8.18-10.0 0.018 0.0052 I 6.36-S.18 0.050 0.021 4.96-6.36 0.092 0.051 4.06-4.96 0.078 0.052 3.01-4.06 0.113 0.159 2.16-3.01 0.122 0.132 f 2.35-2,46 0. 0.19 0.034 0.178
, 1.83-2.35 O.152 f l.11-1.8) 0.27% 0.32 1 1.0 -1.11 0.0457 0.044 e k
1.00 1.0 to i
5 l
l Table 6-5. Neutren Fluence h i
Fast f linx. Cvele I g n/cm*'-s (345 EFP3) y I_'
- f. J.l_uy n e ez,E, 1 MeV C.ipsule center 2.48+10 7.19+17 Pressure vessel wall (:nx) 1.39+10 4.14+17 Fluence. E > 0.1 McV 1.40+15 Capsule center 4.68&l0 f Pressurr vessel wall (max) 2.84+10 8.47+17 I
i
r l
Table 6-6. Predicted Fast Neutron Flue g in Pressure Vessel for 10 F.FPY I.oc a t inn in pressure vessel Inside Ot.t s id e wall T/4 3/4T wall 4
Avg fast flux, 1.7+10 9.4+9 2.3+9 9.7+S n/cm -s Fast aluence, n /en- 5.3+18 1.0+1S 7.3+17 3.1+17 i
i I
\
l O.)These data ,are based on the hypothesis that pressure vessel ;
I f!u.1ce is ; ropor t ional to calculated average core leakage !
ituves for cycles 1 ani 2. Subsequent fuel cycles were as- l o ice.1 to be the sa :e .n s those predicted for econce 2.
l I
i l
l l
- - o.s....a. . m::
i Figure 6-1. Fast Neutron Fluence of Surveillance Capsule Center Compared to Various 1.ocations Through Reactor Vesset k'all for First 10 EFPY 10 9.5 = 1018 nyt i
l 8 _
i i
/
N !
- ~
- b. l l _ 1 M
N* 5.3 = 10 1 " nvt 3 I
.: V ,
1 1 /c*
=
l i i 6
7
- ce E 3.0 - 1018 e/(S nyt l
l' 9 >b*
l 5goo 2 -
pc8 il 7.3 = 1017 nvt
, 3/4T Location 3.1 x 1017 nvt I
,,g n 4 . s. e 0
l .
4 2
- 6 6 10 Time. EFPY ,
,o Babcock s. Wilcox 8
- - _ . - _ , . 7 _., .
1 l
l l
l l
- 7. DISCt'SSION OF CAPSUI.E RESI'LTS
_7._1. Preirradiation Property Data A review of the unirradiated properties of the reactor vessel cere belt region indleared no significant deviatica f ra expected properties except in the case o f t !.c upper shelf properties of the weld metal. E.ased on the predicted end-o f- te rvlee peak neut ron fluence value at the 1/4T vessel wall location and the copper co:ttent of this weld, it is predicted that the end-of-service Charpy upper shelf energy (L'SE) will be belw 50 f t-lb. This veld was selected for it. elusion in the surveillance program in accordance with the criteria in effect at the tirw the program was designed for Oconee 3. The applicable selection c r iterion wa s based on t he unitradiated properties only.
'.: E - _ I rygtigt,=,d, J,r. pne r t v tu t a
.I_ A d L.I G .i_Ie. Prenert_ ifs Ta51s 7- 1 c o.mp .i r e a irradiated and t.nirradiated tensile properties. At both room tenperat are .ud SciOP. the ulti= ate and yield strength changes as a result of irradiatica and the corresponding changes in du:tility are nagligible.
Ibere appears to be some strengthening, as indicated by increases in ultimate and vield stren,;th and similar decreases in ductility properties. All changes observed are so s=all as to be considered within experimental error. In either case, the srall change in tensile properties is insignificant relative to the analysis of t he reactor vessel caterials at this p ?rted in service life.
7.2.2. i=,1ct Pros rties The hehavior of t he Charpy V-notch i= pact data is :ure significant to the cal-culation of the reactor system's operating limitations. Table 7-2 co pares the observed changes in irradiated Charpy impact properties with the predicted changes as shevn in Figures 7-1 through 7-5.
~
The 50-f t-lb t ransition temperature shif t for the weld retal and one of the two base metals was smaller than the shift that would be pred'eted according 7-1 Babcock 8.Wilcox
E3 Pagulatory Culde 1.99. A similar comparison of the shif t of the other base cecs! shows poor agreement. The latter poor comparison nay be attributed to the spread in the data of the unirradiated material co-bined with a minimum cf d ta points to establish the irradiated curve. If the lower bound of the unirrcdiated data curve as used for reference, the shift is close to that predicted using RG 1.99. L'nder these corditions, the comparisen indicates th t the estimating curves in RC 1.99 for medium-copper naterials and at low flu 2nce levels are reasonably accurate for predict 2ng the 50-f t-lb transition tomperature shifts. The estinating curves for high-cepper material at low fluence levels are not in good agreement with the observed data hacause of the abnornal shif ts thet develop in low-USE material as the Charpy curve upper shelf approaches the SG-ft-Ib design linit.
The increcse in the 35-mil lateral expansion transition temperature is com-parest with the shift in RT curve data in a manner sin!!ar to the comparison NDI made for t he 50- it-lb t ransition temperature shif t. These data show a behav-f or similar to t ha t observed from the comparison of tie observed and predicted 50- f t-th t ra. wit ion data. Again, tie significant difference is the larger 8hift ex:.ibited by one of the base metals. This large shif t appears to be re t.it ed t o t he scat ter of the data.
lhe data for the decrease in Charpv USE with irradiation showed a good compar-ison fer both tsase metals, having a medium copper content. "Ihe weld ectal data conpare very well with t he predicted value in view of the lack of data for high-topper-centent weldcents at low fluence values that will be used to develop the estimating curves.
Th.s shi f tw shown are not in complete agreement with thase predicted f rom Reg-ulatory Guide 1.99 .it the fluence level of this capsule. This indicates that ttie estinating curves have greater inaccuracies at tha very los neutron fluence lesels (;l -
10l ' n/cm2 ). This inaccuracy is probably a result of the limited data at the low fluence values and of the fact taat the najority of the data used to define the curves in RG 1.99 are based on the shift at 30 ft-lb as compar3d to t!.e current requirement of 50 f t-Ib. For most materials the shif ts neatured at 50 f t-lb/35 MLE are expected to be higher than those ocasured at 30 ft-lb. The significance of the shif ts at 50 f t-lb and/or 35 MLE is not
'w;11 understood at present, especially for materials having USEs that approach ths 50 it-lb level and/or the 35 MLE level. Materials with this characteristic
, chould be evaluated'at transition energy levels lower than 50 tt-lb.
7-2 Babcock & Wilcox ,
. . - =- . - --. _
i The design cur.es for predicting the shif t at 50 f t-1b/35 23.E will p'robably -
ba modified as data become available: until that t i r.e . the design curves fcr ET shif t as given in RG 1.99 are consid< red adequate for Predicting the RT predicting the kT shif t of thrvie r.it rials for which data are not available er.d will continue to be used to establish the pressure-temperature operational i: limitatlens for the irradiated portions of t w reactor v.*ssel.
I 1
I I
Table 7-1. Copparison of Tensile Tc_st Result _s '
Room t erm t es t SF0F test inirr Irrad L'nirr irrad 1 B .,
- e 5'e t .i ! -- A*.F-! 91. T ransve rr.c riuence. Itsi n/c5 (. I ".eV) 0 7.39 0 7.39 ttt. t e:w ile strength. ;if 95.4 65.0 8 7. 5, , 86.4 D.27 vi.!J utren.th. Ist 63.1 63.0 35.7 57.2 i !_en . .i: s on ' ' 3n.4 29.9 28.6 28.7
- s. t6.8 A6.4 66.5 67.4 E._ ! .' ,e t. Q , '4 F- DN- 1 B 1 : i.c:a c . les n/* r (' l' reV) 0 7.39 0 7.39
, !t. ' en:< i ;e s t ren.:t h ke t
_ 90.5 93.9 37.8 93.7
's . .M . t.> ! 3 st rength, k si - 7 '. 9 79.3 67.4 72.6 i 1 an t;a t isn . ' 23,1 27.3 21.4 22.8 PA. 0 62.9 62.6 52.1 60.3 l
l i
! 7-3 . Babcock a. Wilcox
- - 1
Table 7-2. Observed Vs Predicted Changes in Irradiated Charpy Impact Properties meerial Observed Predicted (#'
I crease in 50-ft-lb trans temp, F 3ase material (ANK-191)
Longitudinal 12 16 Transverse 13 16 Ease naterial (AWS-192) 41 17 Heat-affected zone (ANK-191) (11)(* 16 ,
- i21d metal (WF-209-1B) 50 79 Increase in 35-MLE t rans temp, F h
' Basa material ( ANK-191) longitudinal 14 16 .
Transverse 18 16 l Saco material (AWS-192) 48 17 HA: (ANK-191) (25)I* 16 g Bald wtal (WF-209-1B) 39 79 De.rease in Charpy USE, ft-lb Case material (ANK-191) longitudinal 25 18 Transverse 13 14 Base ruterial (AWF-192) 8 11 ILC (ANK-191) (+54)ICI 9 Weld rm tal (WF-209-18) 8 20
"'These values predicted per Regulatory Guide 1.99, Revision 1.
Based on the assumptien that M1.E as well as 50-f t-15 transition tempera- l ture is used to contro? the shif t in RTNDT*
Scatter in both unirradiated and irradiated data precludes performing valid corparisons. f 1
i I. 7-4 Babcock s.Wilcox
i i
Fi gure 7-1. Irradiated Vs Unirradiated Charpy Iepact Properties of Base Metal, Longitudinal Orientation tro , , , , , , ,
3 -
.: Unirradiated 1
yw ____________ __
3 7.39 1087 nyt
.A 25 -
c , , , , , , , ,
.f41 . . . . . . . . . . .
.e.
E- '*. _ l'nitradiated 7.39 = 10I7 nyt -
7 2
., ,y
- T = 14F q
- E di"
/ .r)? -
2.
~
g , i i s , , e i e . .
2rr , , , , , , , , , , ,
,3,
- AUSE = 25 ft-lbq -
t F -
n -
. 140 -
r
[ IM -
Unirradiated i 7.39 = 10I7 nyt j 1rr -
a 2 sn -
L 1
%~
9 AT = 12F ~
40 -
%ftstat SA $08. c1 2 -
20 - Osit varsonLoneitudinal_
Fun cg See above b tihmeta _ANK-191 ;
e , , , , , , .
-E3 -*3 0 40 . 80 120 150 2n0 20 281 y.3 520 WM fast itsetaarvet, F
- _5 Babcock r. Wilcox 4
I 1
l l
Tigure 7-2. Irradiated '.'s Unirradiated Charpy Iepact Properties of Base .%tal, Transverse Orientation IT . . , , . .
I 75 -
y Unirradiated 7.39 , 1017 nyt i*
yse ._ __----_-- - _ - - - - _ _ - - - - - - _ . _ - - - - - - -
b l 2$ .
6 . g>
.CEE . . . . . . . . . .
Unirradiated 7.39 - 1017 nyt g .4c -
N Z
$ ,or.c t.T = 18F 3 _ _ _ _ _ _9 _ _____-.___._____________._________- .
E at I -
- .f2:
I f f f f
- f f f f 9 e 2M . . . . . . . . . . .
~
1 **
1 Ig- -
~l l
.s IM -
& IM
[ Unirraidated _
7.39 a 1017 nyt
- IT t,
Y -
w se -
O.
- t E ~
AT = 13F
~
9 tC - -
%rta ntSA 508. C1 2 l
1 Omstatario. Transverse 20 -
FustuCg _See above
~
MEATNate AS.K 10' e , e r , , i f , ,
! g i
-so -40 o 40 so im Iso 2m 24) 280 5n :o en Test Tcmearses, F 74 Babcock 8. Wilcox
Figure 7- 3. Irradiated Vs U.urradiated Charpy I= pact Properties of HAZ, Transverse Orienca:icn In0 . . s c ,
Unirradiated ,
5 2
W sc _ _ ._ _. __ _ _______ -_____--.
7.39 = 1037 nyt 3 25 - -
o
.080 , , , , , , . . . . .
i h .vc- - -
= ,
I t.T = 25F P ' \ nitradiated U
", .30 g ___---__l_ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ . _
- s
=
7.39 = 1017 g .rp; - nyt -
~
3 e r e e . . , . .
g i i 2M . . . . . . . . . . .
y . .
160 -
t
. l@ - -
r
~
E 12C -
- 7.39 a 10g7 nyt (t.USE = $4 f t-lb) 3 100 - -
? -
.Y 80 -
L. Unirtadiated -
O.
t T = 11F s
- s,c - 4 .
40 - -
%traig SA 508 C1 2 e
De:Enuts:n EAZ. Trans.
20 - -
FwEnct See above HeatitssEt ANE*19I 1 e p e e . e , , , .
-80 -Q 0 4 80 120 160 200 24 27) 32') %3 m Test Tceta4tvet F 7-7 Babcock & Wilcox
Figure ~-4 Irradiated *.'s Unirradiated 2.arpy I= pact Properties of Base Metal, Tra sverse crie ration Iro - , , , ,
/ -
" 75 -
Unirradiated 3
0 - - - - - - - - - - - - - - - -
g 50 - - - - - - - - - - - - - - - - - - - - - -
5 j 3 ,
7.39
- 10 17 rrat 0 ,-
e 3 g g g . I J B .
3 Unirradiated k
i .ru
[ b 7.39 = 1.17 nyt w . 040 -
- T=~BFl'- "l e E [
t ar i .02G -
1 a
- f I f f f f f 1 1 f NC . . . . . . . . . i 280 16C I . 14 -
, i l ,-- t,US2 = 1 f:-Ib I -
-5 LM s _
J %) -
Unirradiated 5
5 0 80 -
'- 7.36 = *LP uvt O.
1
- g - I.T = 41F - - -
f f y -
gyge g SA 508, c1 2 Maturation Transverse
! 20 - ru ,r.c, See above
%at Nu,me AWS-192 9 I f f f f f f f f 0
- t.0 -4 3 c.o so 120 33 2nc 2c m sn m n itsr Tieturm, F 7_; Babcock & Wilcox
i I '
- l 1
l 1 1
Figure 7-5. Irradiated Vs Unirradiated Charpy Impact Properties of k' eld Metal IM . . . ,
i l
Unitradiated j 75 .
- 7.39 = 1037 nyt i
1 3 1 0
% ~ ------- -- - - _ - - _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _
m Y
e 25 -
0
- ' s s a . . , , ,
, .CV) , *
- a a a . . , , ,
l
- E .% -
~
y Unitradiated 2 AT = 39F --
.oe l- - l g
y .020 p
7.39 = 1017 avt .
.rfn ' i e , , , , ,
2m * * * . . . . . , , ,
10 - .
If4 -
. l@ -
l E llo -
2 3
- IT -
?
w r
80 -
{ O SE = 8 ft-lb
-*~
af = 50F -
f .
I .
__q________________
w-Unitradiated 7.39 = 1037 %'teeg b' eld Metal 29 . Oniturattae
% g _See above
~~
litat hek'F-209-1B 0 e .
0 0 4 W 120 160 7b 23 h) 3h ,]) g Test Twenarweg. F
d
- 8. DETERMINATION OF RCPB PRESSURE-TEMPERATURE LIMITS The pressure-temperature limits of the reactor coolant pressure boundary (RCPB) of Oconee 3 have been established in accordance with the requirements of 10 CFR
- 59. Appendix G. The methods and criteria employed to establish operating pres-sure and temperature limits are described in topical report BAW-LOO 46.8 The objective of these limits is to prevent nonductile failure during any normal operating condition. including anticipated operational occurrences and system hydrostatic tests. The loading conditions of interest include the following:
- 1. Normal operations, including heatup and cooldown.
- 2. Inservice leak and hydrostatic tests.
- 3. Reactor core operation.
1 The major components of the RCPB have been analyzed in accordance with 10 CFR l 50 Appendix G. The closure head region, the reactor vessel outlet nozzle. l and the beltline region have been identified as the only regions of the reac-tor vessel, anil consequently of the RCPB, that regulate the pressure-tempera-ture limits. Since the closure head region is significantly stressed at rel-atively low temperatures (due to mechanical loads resulting from bolt preload),
this region largely controls the pressure-temperature limits of the first sev-eral service periods. The reactor vessel outlet nozzle also affects the pres-sure-temperature limit curves of the first several service periods. This is due to the high local stresses at the inside corner of the nozzle, which can be two to three tines the membrane stresses of the shell. After the first several vears of neut ron radiation exposure, the RT t i elt ne region materi-NDT l als will be high enough that the beltline region of the reactor vessel will I start to control the pressure-temperat ure limits of the RCPB. For the service period for which the limit curves are established, the maximum allowable pres-cure as a function of fluid temperature is obtained through a point-by-point conparison of the limits imposed by the closure head region, the outlet noz-zie, and the beltline region. The maximum allowable pressure is taken to be the lowest of the three calculated pressu en.
r
?
\ y. . w . . . .; ; . .-: ,y.+,s > ,,- p g.v a m ..,, _ m .g m m ,; 9 .,; x y p :w. osy,;_;,ip y : .
Tho eighth full-power year was selected because it is estic:ated that the sec-ond surveillance capsule will be withdrawn at the end of the refueling cycle, whrn the fluence corresponds to approximately the ninth full-power year. The time dif ference between the withdrawal of the first and seccad surwillance c1psules provides adequate time for re-establishing the operating pressure and temperature limits for the period of operation between the second and third surveillance capsule withdrawals.
The linit curves for econee 3 are based on the predicted values ef the ad-justed reference temperatures of all the beltline region materials at the end of the eighth full-power year. De unirradiated icipact properties were deter-mined for the surveillance beltline region materials in accordance with 10 CFR 50. Appendixes G and H. For the other beltline region and RCPS materials, h the unirradiated impact properties were estimated using the creth@ described in BAW-10046P.a ne unirradiated impact properties and residual elements of g the heltline region materials are listed in Table A-1. The adjusted reference temperatures are calculated by adding the predicted radiation-in.du, ed 7.RT NDT %
. tad t he unirradiated RT The predicted ART is calculated using the re-NDT- NDT spective neutron iluence and copper and phosphorus contents. We design ,
curv. . ot Regulatory Guide 1.99* vere used to predict the radiation-induced $
'<1 valtws as a f unction of the cuterial's copper and phosphorus content
~
and neut ron fluence. Figure 8-1 illustrates the calculated peak neutron flu- l ence at several locations through t he reactor vessel belt line region wall as j a function at exposure time. The support ing information for Figure 8-1 is de-scribed in RAW-10100. The neutron fluence values of Figure 8-1 are the pre-dict d f luences, which have been demonst rated (section 6) to be conservative.
Ihe peutron fluences and adjusted RT DI. values f the bel, line region materi-als at t he er.d o f t he e i gh t h f ul l-powe r y ea r a re listed in Tab le 8-1. %e j neutron f loences and ad justed RT value. are given for the 1/47 and 3/4T 2T vessel wall locations (T = wall thickness). De assumed RTNDT
- C I " * "
head region and t he out let nozzle steel forgings is 60F, in accordance with
! BAW-10046p.
l l
Re vis ion 1 January 1976.
Figure 8-2 shows the reactor vessel pressure-temperature limit curves for normal heatup. This figure also shows the core criticality ILaits as required by 10 CFR 50, Appendix C. Figures 8-3 and 8-4 show the vessel pressure-temperature limit curve for normal cooldown and for heatup during inservice Icak and hydrostatic tests, respectively. All pressure-temperature limit curves are applicable up to the ninth ef fective full-power year. Protection against nonductile failure is ensured by maintaining the coolant pressure be-low the upper limits of the pressure-temperature limit curves. The accept-4ble pressure and temperature combinations for reactor vessel operation are below and to the right of the limit curve. The reactor is not permitted to go critical intil the pressure-temperature combinations are to the right of the criticality limit cu rve . To establish the pressure-temperature limits for protect ion against nonductile failure of the RCPB, the limits presented in Figures M-2 through 8-4 must be adjusted by the pressure differential between the point of system pressure measurement and the pressure on the reactor ves-s.l controlling the limit curves. This is necessary because the reactor ves-w! is the nost limiting component of the RLPB.
i l
l t
6 5
e
1 1
l i
i Jw7 I . ~ ,
e . **I a a ~ = = 4 (a ."
e * .4
- r e e e n o
= g
- c 4 e a
- l e I [ U#Dwl ~ " "
3 ; y 4g';t
. ~ .. . . .
i Eme m
. .e . t. *e
-s < m u e, = -e, .. e
. * .e h '
,1 * ,y - '
i o
to.e 4
y, 0
, ..a ,. e, e.e ~ ~ ~ ~ ~.
> t 3 a. 4.. =. I. at. 4
, =. m. a.
c.
y f }w
- a- + + e + e e 6
=R !
U b* e
,.4 msi
.. s -
fl . a
.e.
.e a
= ,
3
% * *. 4 4 4 l w L. P Ee.;,. *e
- e. ee ~ e. .,. .
m u A % h. .. .
o ;r , .
M L; i c. h .e. - -
- e- * -r.
e u.
'4' e, s
=. c. c. a. .c. =. .
- t. g **. ".
f, - o e e o e = y
] E y 1s. .
c .'
L; , .
H-
.s .
u
. 4 e.
4
- e. . ..
a *41 : g -
e o c- e e c:
s- .; g t hM 3
- 7. ."
[s ? Ib r
e.
- g r cr 6
w r,,,
y .
a
)
wC O .* - , , . .
=
- ; *- d a-c.= e
- O *. "
- J .
c: '.. .,
l 6, i 1.
~3 r . J. . . e e .
~~ < . e
,. e e .
s e. e .e.
u r:,
6k
'.O. 's . , {1B-*
.a
. ~.
2 b . . . - . +
s e u,. c. .&. -3 .
c' . $. .s
'* e, a e
6 ,
. s . .
.f t . . e r
oc s
6, g
3 e, a t
l
.y .,._ e~, ~, , . . ~ . - , . . ~ . .,
a
- M A -e a r E '. E 2 3 e,
l
- y
- -
- a. .,
- * ' ' S.6 4 4 . . . .
- J sn 3 3 3 3 l
- i U *.
so 4 e.
.e
- '. 8 e
e,
=
e o
e=
e e ,
.I, e
as > * *
. l f.
E
- e. et 4 3 3 e.
A B l
i
= * * . . .
l l
l
Fi tpare d-1.
Predicted fsst Neutron Fluences at Various I.ocations Through licactor Veur.el L'all f or Tirret 10 f.FPY (Oconee 3) 6.0 5.6 ~
5.3 = 1038 nvt 5.2 -
- 4.8 -
p 4.4 -
- 4.0 -
a 3.6 -
w 37 ..
3.0 = 101 " nvt a e N 2.b - 'ig f o
s#gM
,- 2.4 -
E N 2.0 -
o 6
.g to'a0o E 1.6 - \
u 5
y
. 1.2 -
7.3 = 1037 nvt 3 0.H -
3/4T Location 0.4 - 3.1 = 1017 nyt
. out3.ide Suriace n i i f I I f j 0 1 2 3 4 5 6 7 8 9 10 lime. EFPY I
1
Figiere M-2. He.ictor % ssel Pressuse-letper.iture 1.imit Curves for Normal Operation -
IIc.itup. Applic.thic for Fl st right Effect tve Full-Power Ye.irie (Oconee ))
2200 -
Beltline Region 1/4T 147 Beltline Region 3/4I 87 Closure ficad Region 60 2000 -
Outlet Nozzle 60 D H 1800 -
Pressure. Temp.
y g Point pet F 1600 - A 480 70 w B 625 188 E C 625 273 1400 D 2000 298 k
n.
E 2250 308
- F 625 273 Applicable for co e 1200 6 Heatup Rates 4 .3 .
0 3 1 2250 348 UP to 100F/h 7* 1000 -
eo
" Criticality 800 -
o I.lini t B C U ,
i y 600 - y A
400 -
rh.-....,...,......-..p.r.i....i.i.......i.
g
. ..a s iii. . i v. .: sw. u.n . ...<.i. n ii. . ... 4 a' '
- 8 ' ** 'a n
cr a
. '. .8
. a. l, a, l'. ' ..*.
. ..O
... ., '.' .'..H i .' .' '..':'i h. p i .. '. r o Pala
. . " .. ,,' < e .e 6 200 -
...: ,... . . .s u as a . u .n i... .e aar **
- u i '+i4 8 Q ... c a ..t.sv ror ,.... m . i .i ai ,, roe.
" e f i e I t i 1 i i t i I f a kn 40 80 120 160 200 240 280 320 340 Q Reactor Vessel Coolant Temperature. F
~
- ,.an m .r e. o.- w +
Figure H-l. Re . art ir W ssel l'ressure-Teriperatu re I.f cit t Curve for Norral oper.it li n Cisiilsiown. Applicable for First Eight fi feet Ive full-Power Years (Oconee l) 2400 A=sunn il RINirT, F E
2200 -
Beltline degion 1/4T 147 Beltilne Reglim 3/4T 87 Closure Head Hegion 60
, 2000 -
Outlet Norrie 60 1800 -- Pressure. Temp.
y Point g et F E A 250 70 '
. 1600 -
B 625 136
- C 625 273 E D 970 210
$ 1400 -
E 2250 296 u
CD Ae e
M 6' 5 1200 -
o Applicable for f
0 Cooldown Rates 1000 up to 100F/h e
e 6
800 -
p O
B J 2 600 -
C
=
[
a;r 400 -
p,, ,,,,,,,.,,,,.,,,,.,,,,.,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,,
..e . .. . e i s ..e i *. i .
- at.e, ......
i., ,,4 n .,,, .,; i, ,,.,pe....., 4 ,.e . ...:..,,,,. 6,,,,,,ei,
,i ,
O A ,,,,,...........ea.....,,.i,,...,
1 i
E- 200 -
. . - s i,ii < *,. . ., eu s i s .i '
, . ,- e,si. . .* , i n e ..o..i..........
,.........p..
= l I I I I I I I I I I 0 I I Q 40 60 J/O 160 200 240 280 3; 0 M
Renetur Vessel Coolant Iceperature. F
s
'Y.
I Figure 8-4. Reactor Vessel Pressure-Tecperature Li=it Curve for Inservice Leak and Hydrostatic Tests, Applicable for First Eight Effective Full-Pc.cer L ars (Oconee 3) 2600 Assumed RTg.F E 2400
- Beltline Region 1/4T 147 Beltline Region 3/47 87 l Closure Head Region 60 2200 l - Outlet Nozzle 60 l 2000 -
Pressure. Temp, Point pst F l 5 1800 -
A 210 70 9_ B 625 156
. C 625 245
- D 1520 258 1600 -
3 E 2500 309 f* D i 1400 - lp 5
}
12 % -
Applicable for Heatup and Cooldown Rates up T to 100F/h (150F in -
? 10 % -
any 30-min. period) 5 f_ h% -
6x - C B
l 4d - l
,. me...o................ . . . . - . ..,,i.
eM te tw e . e- * .' '.e :sett c .e w - N
.=..;...e ve .
s.
,. . s : .- . . . . . . . . . . . . . . . . . . . ., (
e..te* ar. . ,,aw. .reo..e .md the . . ..r. . * + = *..e .
2' P( "" *e r . I i s *e t ' . Isost %f..: ..
e ne . ee s 4.+
w , a , ,e . . . . . e m.e s., *- s e t
, - .,,w
, f f f f I f f ' ' '
? f 30 lou 1O lou 220 i%s 350 Re.sc tor Vessel Coc]. ant lemperature. '
l l
33 Babcock & Wilcox
~
e
- 9. SUMw.ARY OF RESUI.TS The analysis of the reactor vessel naterial contained in the first surveillance capsule (OC111-A) removed from 'the Oconee 3 pressure vessel led to the follow-ing conclusions:
- 1. The capsule received an average fast fluence of 7.29 1017 n/cm (E > 1 MeV). The predicted fast fluence for the reactor vessel 1/4T location at the end of the.first fuel cycle is 2.28 = 10 17 (E - 1 MeV) .
- 2. The fast fluence of 7.39 = 1017 n/cm2 (E > 1 MeV) increased the RTND'i the capsule reactor vessel core region shell materials to a maximum of SOF.
- 3. Based on a ratio of 1.6 between the fast flux at the "urveillance capsule locatien to that at the vessel wall and an 50* load factor. the projected fast tluence that the Oconee 3 reactor pressure vessel vill receive in 40 ca lend..r years' operation is 1.69 1019 nicm2 (E > 1 MeV).
4.
We increase in RT,,J7 for the shell forginr material was less than that predicted by the currently used design curves of J.RT DT *##*"" fl"*"##
except in the case vbere the scatter in the unirradiated data did not per-mit accurate deter =ination of the shif t. Furthermore. inaccuracies in the prediction curves resulting from lack of irradiation data for low-copper materials at low fluences may be a factor in less-than-accurate predictions.
The current techniques for predicting irridiation-induced changes are con-servative.
- 5. The increase in the RT f r the weld metal was less than that predicted NDT by the currently used design curves of LRT.DT because of the inaccuracies
' in the method of measuring the shif t in transition terperature at lev flu-eneas. The current techniques used to predict irradiation-induced changes are conservative. '
- 6. The current techniques used for predicting the change in Charpy i= pact upper shelf >r..perties due to irradiation are conservative.
9-1 Babcock t iVilcox
~,
- . , -6
- 7. .' ne analysis of the' neutron dosimeters' derenstrated that the analy;* al techniques used to predict the neutron flux and fluence were accurate.
8 .' . he thernal monitors indicated that the capsule design was satisfactory
~ for. mintaining. the specimens within the desired temperature range.
?
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9-2 Babcock a.Wilcox
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- 10. SURVEILIECE CAPSULE REMOVAL SCHEDL*LE Based on the postirradiation test results of capsule OCIII-A, the following schedule is recon: mended for exar.ination of the remaining capsules in the Ocenee 3 reactor. vessel surveillance program:
Evaluation schedule Est capsule Mi d EFP). Est date(b)
Capsule fluence, data ID(8) n/cm 2 Surface 1/4T available OCiII-B 3.1 - '1015 5- 9 1979 OC I I I-C 1.2 10l ' 18 33 1984
<rtII-D 2.2 = 10P' 32 58 1986
, OCIII-E Standby -- -- -
Oc! ! !- F Standby .-- -- --
(a) Capsules contain weld metal specimens.
00 These dates do not represent the earliest dates that data will be available- for the ruterials that control the operat ing limitations. Similar materials are in-cluded as part of the BW Integrated Reactor Surveil-lance. Program, which will provide necessary data on a tirely basis. .The earliest date that these data will -
be available is 1980.
i L
10-1 Babcock s. Wilcox
- 11. CERTIFICATIO5 The specimens were tested, and the data obtair.ed fro:: sconee Nuclear Sta-ion, Unit .3 surveillance capsule OCIII-A were evalua:ed using accepted techniques and established standard methods and procedures is accordance with the require-ments of 10 CFR 50, Apper.dixes G and H.
. /- ' , .
)? s @ E.k, .$ t, / Da : te h
~. L. ,
Proje:tcve. Jr. , p>L2 Tec.%ical nager This report has been r+ viewed for technical cceten: and accuracy.
K. E. M>cre 7- W"17 Date Technical Staf f 11-1 Babcock s. Wilcox
+ e 9 6 w
4
- 12. REFERENCES i
I -G.
J. Snyder and G. S. Cart f Eeactor Vessel Material Surveillance Pro-gran, BAW-10006A, Rev. 3, Babcock & Wilcox Lynchburg, Virginia, January 1975.
1 # A.
L. Lowe, Jr., et al. , Analysis of Capsule CJ1-E From Duke Power Company Oconee L' nit
,s 1 Reactor Vessel Materials Surveillance Program, BAW-1436, hab oc k & Wilcox, Lynchburg Virginia (to be published).
3 User's Manual for ANISN, a (me-Dimensional Discrete Ordicates Transport Code Wit h Anisot ropic Scat tering, K-1693 (RSIC-CCC-82), Union Carbide Corp.,
Nuclear Livision, March 1967.
User's .HLnual for the D0T-11W Discrete Ordinates Transport Computer Code, WANI-N-1982, December 1969.
CASK - 4'J-Croup Coupled Neutron and Gamma-Ray Cross Section Data, RSIC-DLC_-Jj, v.adiation Shtelding Information Center.
Draf:
!a w Standard E482-00, " Recommended Practice for !kutron Dosimetry f or Reactor Pressure Vessel Surveillance," October 10, 1974.
7 W.
L. Z15p, Review of Activatica .'kthods for the Determination of East Neu-t ron fget ra, Reactor Centrum Nederland, Petten, May 1965.
' II . S . Palme and H. W.
Behnke, Pethods of Co=pliance With Fracture Toughness and Operat ional Requirements of Appendix C to 10 CFR 50, 3AW-10046?, Bab-cock & VIIcox, I.ynchburg, Virginia, October 1975.
- 11. S. Palme, C. S. Carter, and C.
L. Whitmarsh, Reactor Vessel Material Sur-i veillance Program - Compliance With 10 CFR 50, Appendix H, for Oconee-Class Reactors, BAW-10100A, Babcock & Wilcox, Lynchburg, Virginia, February 1975.
10 H. S. Palme, Radiation Embrittlement Sensitivity of Reactor Pressure Vessel Steel, 3AW-10056A., Babcock & Wilcox, Lynchburg, Virginia, August 1973.
12-1 Bebcock & Wilcox l
m
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e APPD; DIX A l
Reactor Vessel Surveillance Program -
Background Data and Information I
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- 1. Material Selection Data The data used to select the materials for the specimens in the surveillance program, in accordance uith E-185-66, are shown in Table A-1. The locatione of these materials within the reactor vessel are shown in Figures A-1 and A-2.
i
- 2. Definition of Beltline Region l
Thz beltline region of Oconee Nuclear Station, Unit 3 was defined in accord-ance with the data given in BAW-10100A.9
- 3. Capsule Identification
('
The capsules used in the Oconee 3 surveillance program are identified below
] i l
by identification number, type, and location. t e
's~
Capsule Cross Reference Data .
E S
ID No._ Type Location OCIII-A A Upper OCIll-B B Lower OCI I I-C A Upper OCIII-D B Lower OCII I-E A Upper j
! OCIll-F B Lower
,4 . Specimens per Survei_llance Capsule ,
See tables A-2 and A-3.
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Table A-2. Sterials and Specirens in Upper Survei1 lance Capsules 0CI1I-A, OCIII-g and OCIII-E No. of specimens Mterial description Tensile Charpv Weld cetal, WF-209-1B 2 12 Ileat-affected zone (llAZ) liest A - ASK-191, longitud. 0 12 i
Baseline material lleat A - ANK-191, longitud. 0 9 Ileat A - ASK-191, transverse 2 12 j I
lleat B - AWS-192, t ransverse 1 9 Total per capsule 4 54 i
h Table A-3. Nterials and Speciw ns in lower Surveillance Capsules OCIII-B, [t OC111-D, and OCI I_1-F I
No. of speciewns h terial_ description Tensile Charpy l l
Weld met al WF-209, longitud. 2 12 l Weld HAZ lleat A - ANK-191, longitud. 'D 12 Ileat B - AWS-192, longitud. 0 6 Baseline material lleat A - ANK-191, longitud. 'O O Heat A - ASK-191, transverse 2 12 Ilea t B - AWS-192, longitud. U , O I Heat B - AWS-192, transverse 0 6 Correlation itSST plate 02 0 6 Total per capsule 4 54 1
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Figure A-1. Location and Identification of Materials 1* sed l in Fabrication of Reactor Pressure Vessel l
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RD.% 6SO (Lower Nozzle Belt)
I 4+ WF-200 f AWS-192 (Upper Shell)
I LT-67 (75: ID)
WF-70 (25 OD)
ANK- 191 (lever Shell) 4 % WF-169 417525-1 (Dutchc:an)
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APPENDIX B Preirradiation Tensile Data l
l 1
B-1 Babcock & Wilenu l l
i
Table B-1. Preirradiation Tensile Properties of Shell Forging Material. Heat ANK-191 Test . Red'n trenEt , Psi __ Elongation, .
of area, Specimen temp,
- No. F Yield Ult. 1*n i f . Total Longitudinal JJ-802 KT 59,040 84,320 13.9 30.0 70.2 803- RT 60,180 84,310 13.4 31.4 72.2 29.3 71.2 l 804 RT 58,110 83.300 13.5 1kan KT 59,110 83,980 13.6 30.23 71.2 Std dev'n 850 478 0.22 0.87 0.82 JJ-801 580 53,690 83.290 13.76 27.9 68.7 {
l 805 580 54,280 -84,240 12.77 27.1 68.8 I 806 580 55,380 84,330 12.71 27.1 63.7 l Mean SSO 54,450' 83.450 13.08 27.37 68.73 k Std dev'n 25 700 470 0.48 0.33 0.05 ,
1 Transverse $
JJ-603 RT 70,070 85,060 13.4 30.7 66.9 604 RT 60,940 86,520 14.8 31.1 63.4 5 608 RT- 5s. 310 84,730 13.3 29.3 65.1 M Mean RT 63,110 S5,440 13.83 30.37 66.S Std dev'n 5,040 780 0.66 0.77 0.55 g
,JJ-607- 580 56,200 85.300 12.26 28.6 66.1 610 580 56,220 85,580 13.82 28.6 61.1 g 611 580 54,600 84,150 9.6 28.6 67.5 y Mean 580 55,670 85,010 11.89 28.6 66.57 Std dev'n 25 760 620 1.74 0 0.66 I
i I
l l
i a,
Tabic 3-2. Preirradiation Tensile Properties of Shell Forging .%terial - HAZ, I! ear ANK-191 Test- Red'n Specimen Strength, psi Elongation, -.
temp, of araa, No. __F__ Yield Ult. L'n i f . Total *
!.on gi t udinal JJ-501 RT 59,460 84,840 13.1 29.3 70.6 504 RT 59,310 84,440 13.6 29.3 70.3 505 RT 58,130 83,170 13.5 32.1 71.3 Mean RT 58,970 84,150 13.4 30.23 70.73 Std dev'n 595 -712 0.22 1.32 0.42 J1-502 580 54,880 83,570 12.79 25.7 67.7 503 580 53,920 82,780 13.24 28.6 67.2 507 -580 58,010 84,010 12.72 39.0 67.9
'le a n 580 55,600 63,450 12.92 28.1 69.6 Jt d dev'n 75 1,750 509 0.23 1.79 0.29 T ran 4ve rse
- JJ-301 KT 58,730 84,850 10.4 22.1 64.9
!O2 RT 59,270 84,300 9.8 22.1 64.9 306 RT 59,040 84,520 7. 8 21.4 61.4
.ean RT -59,010 84,560 9.33 21.87 63.7 Std dev'n 220 226 1.11 0.33 1.65
.fJ-30 3 550 58,100 82,150 7.98 20.7 66.6
~ 304 580 57.100 82,350 7.98 20.0 63.9 305 580 58,100 82.150 8.38 20.7 64.9 Mean 580 57,770 82,220 8.11 20.46 65.13 Std dev'n h5 470 .94 0.19 0.33 1.11 1
B-3 Babcock a.Wilcox
Table B-3. Preirradiation Tensile Properties of Shell Forging Material, Heat Ak'S-192 ,__
Test Red'n Streng . Psi _, Elongation, Specimen temp, of area, No. F Yield Ult. Unif. Total %
Longitudinal KK-802 RT 62,020 85,980 11.5 28.6 67.3 803 RT 59,240 86,030 12.4 27.9 69.0 805 RT 57,440 81,640 13.4 28.6 70.3 Mean RT 59.570 84,550 12.43 28.3o 68.87 !
Std dev'n 1,880 2.060 0.78 0.11 1.51 580 12.39 27.1 63.7 KK- 801 -
804 580 59,210 55,230 35,630 82,840 12.92 27.1 66.6 l 806 580 53,360 81,780 13.71 28.6 62.1 Mean Std dev'n 580 35 55.930 2,440 83,420 1,620 13.01 0.54 27.6 0.71 64.13 1.86 f
Transvery;e KK-601 RT 61,780 85,970 12.3 27.1 64.0 603 RT 57,780 83,070 13.4 26.4 62.5 607 RT 55,060 80,180 13.4 29.2 65.8
.'k an RT 58,210 83.070 13.03 27.6 64.1 Std dev'n 2,760 2.360 0.52 1.24 1.35 KK-602 580 55,370 82,:.00 11.97 25.7 60.8 604 580 57,490 84,490 14.86 23.6 56.6 ,
605 580 56.070 83.010 11.82 25.0 58.4 I Sk.m 580 56,310 83,270 12.88 24.77 58.6 Std dev'n 15 882 912 1.40 0.S7 1,72 l
g.4 Babcock s. Wilcox
~,
b Table h-4 ~ l'reirradiation Tensile Properties of Shell Forging Material, HAZ, Heat AWS-192 Test Specimen teg. Strength, ps_1
_ Elongation. % Red'n 9g ,
__ No. F Yleid. Ult. I'n i f. Total i 1.on gi t_ud inal KK-502 RT 56,950 50i 80,980 11.6 30.0 69.9 RT 55,820 504 80,430 14.6 31.4 70.6 kT 54,400 83,620 13.3 27.9 66.1 Sun RT 55,720 81,680 13.16 29.77 69.53 Std dev'n -1,041 3,392 1.23 1.44 1.05 KK- 501 580 55,540 83,060 505 12.99 28.6 66.0 SFO 52.650 79,990 S t s, ' 13.41 29.3 66.9 580 57,88') 79,860 11.84 27.1 -63.3
&an 580 55.360 So,970 Std dev'n 12.75 28.33 67.07
.5 2,140 1,480 0.66 0.92 0.95 Tranw..rse EK-301 RT 55,840 87,910 lo.' 10.6 24.3 66.0 RT 57,040 80,860 305 10.8 24.3 66.0 RT 56.870 50,960 9.5 24.3 66.9 Nan RT 56,580 l SLJ dev'n 83,240 10.3 24.3 66.3 530 3,300 0.57 0 l 0.043 i i
KK-TO1- 530 58,610 82,900 8.48 20.0 56.9 10 ~. Sho 55,530 79,980 38t .
8.69 20.7 59.5 '
-580 57,810 84,210 ;
9.58 20.7 55.7 l Nan 580 57,320 82,370 l Std dev'n t5 8.98 20.47 57.37 1,310 1,775 0.45 0.33 1.59 a-5 Babcock a. Wilcox
\
. i . . . .. .. . -
Table'B-5. . Preirradiation Ten-f le Properties of k' eld Me t a l, -_ _I.on gi t ud i nal , L'F-209- 1B Test 1 Re d ' n Strength, psi Elongation, % gg Speciewn tenp, ,
- No. F Yield til t . L*n i f. Total JJ-002 RT- 74,630- 90,460 12.5 29.3 63.2 004 KT' 73.540 .89,110 13.1 27.1 63.6 018 RT '
76,720 91,910 11.9 27.9 62.0 Nan RT 74,960 90,490 12.5 23.1 62.9 Std dev'n 1,320 1,140 0.49 0.91 0.68 JJ-001 5:st) '67,540 87,700 10.83 20.7 52.9 011 580 66,460' 88,300 11.54 22.1 53.0 013 580 6S,210 87,620 10.33 21.4 50.3 g Wan 530 67.410 87.870 10.9 21.4 52.07 Std dev'n * 's 712 303 0.50 0.57 1.25 3
3 s
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+- B-6 Babcock & WilCOX t * ~ -w - ~
APPD.T!X C Preirrattiat Ion Charpy Ir. pact Data c-I Babcock s. Wilcox
Table C-1. Preitradiation Charpy lepact Data for Shell Forging .tterial - Longitudinal Orienta-tion _, !! eat ASK-191 Test Absorbed Latera l Shear energy, expansion, fracture, Specimen teep, No. F ft-lb 10-3 in. t JJ 831 360 192 72 100 833 360 166 72 100 ^
816 357 181 72 100 180 65 100 JJ 860 283 859 279 178 71 100 853 278 164 77 100 JJ 818 201 187 68 100 802 200 161 70 100 g 846 200 165 70 100 JJ 822 140 142 73 85 829 140 136 77 70 75 f
845 140 140 69 JJ 830 PD 124 75 50 848 80 93 67 25 825 80 127 75 60 JJ 821 41 88 68 20 41 88 69 6 842 832 40 58 50 3 JJ 851 30 72 60 5 i H57 29 52 44 3 850 29 62 51 8 24 17 <1 JJ 854 16 H52 16 21 20 <1 856 16 30 26 <1 JJ 804 0 38 . 30- 1 817 0 35 30 2 820 0 31 25 1 JJ 849 -60 4 2 O M55 -60 2 1 0 858 -60 6 1 0 s
c-2 Babcock & Wilcox
, . Table C-2. Preirradiatfor Charpy In: pact Data for Shell Forging Material - Transverse Orientation.
Heat A';K-191 Test Absorbed . Lateral Shear Specimen' ten:p , energy, expansion, fracture, No. F f t -Ib 10-3 in. %
JJ 651 362 172 75 100 640 361 144 66 100 l 621 357 128 72 100 JJ 701 284 136 73 100 705 283 140 73 100
.. 700 277 135 71 100 JJ 635 200 150 71 100 g 625 200 153 67 100
! -637, 200 141 73 100 JJ 704 140 130 74 88
~
703 140 126 74 85 702 140 118 71 70 JJ 676 80 123 71 55 611 79 130 78 65 657 79 129 74 60
.:J 670-- 41- 77 62 5 685 41 71 58 4 612 41 56 66 10 JJ 708 25 ' 68 55 5 706' 24 74 62 10 707 24 57 46 8 675 21 58 48 6 643 20 47 41 4
~JJ 622 0.0 35 26 1 692 0.9 56 44 2 680 0.5 18 15 0 JJ 698 -39 3 2 0 697 -40 8 8 0 699' -41 14 12 0
- l' '
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C-3 Babcock a. Wilc0x
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Table C-3. Preirradi.ition Charry lepact Data for Shell Forging Pitorial - EG, Longitudinal Orien-tation. Heat ANK-191 Test Absorbed Lateral Shear Specimen teep, energy, expansion, fracture.
No. F_ ft-!b 10-3 in. % I JJ 504 362 153 76 100 511 360 74 53 100 g 514 360 66 55 100 JJ 527 281 157 79 100 528 280 212 -
72 100 l JJ 510 200 75 51 100 502 200 70 54 100 505 199 103 52 100 l JJ 516 140 12e 62 97 JJ 517 121 186 71 100 518 129 146 68 100 523 120 1%4 76 100 JJ 506 SO 62 43 95 509 80 5' 29 45 513 80 76 50 75 JJ 501 40 47 30 35 l 507 '. n 106 56 50 f 512 40 41 60 25 l l
1 J.* 503 0 15 26 12 508 0 105 66 IS 515 0 34 24 23 JJ 519 -40 75 45 30 521 -40 78 50 35 526 -40 33 26 25 JJ 520 -60 to 30 5 525 -60 57 33 6 524 -61 71 44 . 12 I
i
I
- Table C-4. Preirradiation Charpy Impact Data for Shell Forging Material, ilAZ Transverse Orienta-tion,Ileat ANK-101 _ , , ,
Test Absorbed Lateral Shear Specimen temp. energy, expansion, fracture.
No. F ft-Ib 10~' in. I 1
l JJ 313 360 86 53 100 315 359 106 63 100 335 358 73 54 100 JJ 328 201 102 54 100 323 201 92 59 100 308 200 83 49 100 JJ 322 60- 76 42 85 l 80 98 100 319 51 344 80 74 45 60 157 40 66 44 85 1 12 40 57 33 45 139 40 55 30 55 101 20 46 29 70 10 4 20 36 27 30 l
JJ All 0 35 24 35 0 33 20 21 12 4 117 0 47 33 35 l
l JJ 424 -59 27 20 2 l 420 -60 56 44 4 l 421 -61 67 35 6 l
f
, - . . 4, ._ _ , ,; - . . .c. , .. . .. , .
T.tSle t - 5. Preirradiation Charpy Icpact Data for Shell Forging Fsterial - Longitudinal Orienta-tion, Heat AH-192 Test Absorbed lateral Shear
. Specimen temp. energy, expansion, f rae t ta re ,
W. F ft-lb IC~ f in. I KK dli 360 1 39 77 100 i' l 1 1% 134 70 100 hol 133 137 65 100 KK h2d 2x0 167 67 100
-!4 280 twl 75 100 327 279 162 67 100 KK 704 201 144 70 100
-a7 201 144 70 100
- I '. 200 146 71 100 KK V> 1 39 136 63 100 e2l 1 19 166 66 100 slo 1 39 162 74 100 KK . sol no 128 51 65 805 80 133 Is 55
- n '. '
.O 112 73 40
- 9. wl l 44 42 73 35
- ^12 40 81 65 8 306 40 oO 65 14 KK s21 16 107 77 20 81H 16 104 77 30 l 622 15 97 74 25 i KK 802 0 67 55 6 ,
M09 0 41 33 +1 1 316 0 o2 51 i KK 817 -41 40 30 1 j 824 -41 30 25 1 826 -41 56 44 5
Table C-6. Preirradiation Charpy Icpact D.ata for ShcIl Forging Nterial -- Transverse Ori *ntation.
Heat Na'S- l_92 Te s t Abscrbed Lateral Shear Specinen temp. energy. expansion, fracture.
No. .F J -lh 10-5 in. '
KK 665 ~ 360 99 77 100 641 160 101 70 100 6 34 359 109 72 100 KA 976 '82 111 71 100 677 2.% 0 10n 71 100 t> 74 .'79 104 70 100 KK 619 200 lli 71 100 a27 199 103 68 100 n6'* ! 9 115 72 100
.3.*'1 140 !!9 73 190 ai* 140 92 71 90 97* 1 19 113 67 100
%4 I 39 114 71 100 t's a22 81 64 51 8
- '> ! *O 83 62 25
.'4-30 99 71 35
- *. 7 5 eal . .O 65 40
- . 7 i e,0 80 64 35 l
%;, 60 64 54 25
.li- e,0 50 44 15 s"*' *> 1
- 40 $1 43 6 ajo 40 40 36 3 l nW *0
. 19 50 12 l KF 6 5.' O 17 30 <1 eli O 5 .1 42 4 6ns , 0 21 18 <1 KK %81 -59 2 1 0 ns: -60 in 1 26 too -60 3 2 0 l
i
{
l Table C-7. Preirradiation Charpy Impact Data for Shell Forging .%terial ttAZ, Longitudinal Orien-tatlon, Heat Ak'S-192 __
Test Absorbed 1.a t e ral Shear Specinen temp, energy, expansion, fracture, No. F ft-lb 10 ' in. 7 ,
( .__
KK 503 361 133 75 100 507 160 128 73 100 + ,
510 358 158 75 100 KK 511 201 163 72 100 508 200 111 54 100 g 504 199 86 56 100 KK 519 122 146 70 100 l 528 121 144 72 100 524 121 122 62 100
(
KK 509 60 130 71 100 516 80 156 66 100 512 79 126 58 100 KK 521 40 120 79 45 521 39 104 66 70 522 38 118 59 SS KK 106 0 103 67 30 511 0 40 59 40 ,
.50 30 514 0 70 l KK 526 -30 59 39 li 527 -30 50 36 6 517 - 31 98 61 40 l KK 520 -49 25 20 4 l 518 -49 34 26 2 !
525 -50 84 59 30 l KK 505 -78 18 13 2 l
501 -79 26 15 2 )
515 -80 55 33 4 ;
i
l Table C-8 Preirradiation Charpy Iepact Data for Shell ,
Forging Material -- HA2, Transverse Orienta-t ion, ficat N45-192 i
i Test Absorbed Late ral Shear Specimen temp. energy. expansion, fracture, No. F ft-lb 10-3 in. ___ T KK 325 361 135 65 100 136 360 153 77 100 318 358 128 63 100 KK 343 278 118 76 100 339 278 125 76 100 l 344 277 132 76 100 1 i KK 333 199 65 37 100 i 310 199 122 66 -
100 313 197 78 53 100 KK 346 139 131 77 100 337 139 132 67 100 308 139 121 71 100 KK 312 80 91 46 94 i
307 79 76 41 85 335 79 63 36 98 KK 30S 40 14] 67 100 316 40 72 42 35 327 40 66 40 45 KK 341 20 93 70 40 340 20 104 77 65 345 19 82 64 30 KK 317 0.5 106 64 45 324 0.1 43 31 18 326 0.1 43 32 40 KK 309 -40 30 22 20 330 -39 80 51 28 i
KK 342 -39 27 21 2 347 -40 52 35 5
! 338 -40 38 28 3 l
l
4 Table C-9. Preirradiation Charpy Impact Data for Weld Metal WF-209-1B Test Absorbed Lateral Shear Sp'cimen e temp, energy, expansion, fracture, No. F ft-lb 10-3 in. _
JJ 063 359 57 49 100 083 357 58 48 100 089 357 65 53 100 JJ 091 201 57 46 100
- G53 201 63 50 100 f 084 199 78 57 100 }
JJ 093 140 60 46 100 060 140 69 52 100 065 140 61 49 100 JJ 092 110 64 46 100 088 110 58 46 100 g 077 109 47 37 96 r JJ 064 80 61 40 98 095 80 38 32 95 g OSI 79 48 35 85 .
JJ 090 40 28 28 10 069 40 33 24 15 075 40 21 20 12
- JJ 061 -40 17 16 5 030 -40 16 14 2 057 -39 19 18 6
- c'
, --, .. , , .n
1 1
1 1
l 1
Fi gu re C- 1. Charpy != pact Data Free Unirradiated Base
.% t a l A. Longitudinal crientation
- i s y y 3 e a - - 1 e
.5 . -
. e e
2 j x _ _________ ____ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
v T -
e -
, 2%
l l
p &
e a e e f a f 1 o a 5 ,3 a
, 5 8 e le t '
,; N- .
m t * '
g D: -
e -
3 5 9 2 I
. OV -
1 3 - - -
__ y __ - - _ _ _ _ _-_- - -._ - __ _ _ _ _ _ _ _ _ ___-
I r e e I
- .G~ - d -
3 e -
l ;
g f f I t f f f
- I e t
- a s s s a a 5 a s a i
EAIA SL w ra .
T +20F e
- ( nor , -
T,s'5*ul+18F/+35F
- e
- p.- . 7,150 er-ta) t23F/+37F . _
l l
(+USE (avs) 180 ft-lb _
, l< - RT,3, +20F 3 -
- ' l e ,
- I e I
. e i E 17 - -
1
- i l
h17 - -
l a e Lv - -
7,
< e t
y - ! -
t l
_ ._ _____ ap __________________________________
~
1 t.C -
- 1 e
e
% teint _SA SCS. C1 2
, Or:twtation LonAltudinal h - e .
Faws %r. e
%r Nuota ANK-191 p t t # # t t i f f I t
-D 40 0 % SO 120 150 2T N 20 5:9 !% vn itsr Tawtaarver, F
\
l
Ficure C-2. Charpy Impact Data Fron Unirradia:ed Base Metal A. Transverse Orier. tat 1.'n im , , , , , , ._ .
.e 15 -
. e ll
- 2 r,c .-__ __ _ -..___ - - ._ - -- - _ _ _ - _ _ - - .
5 Y -
w 25 -
e a e e e e a a g e
.2 i . i q . . . . . .
e
. g -
e e . $
g .YC - * -
Z
.9se } h E
f.______________._--.__________--______-
= 1
- .r2; - 1 f
2 , , , . , , , , . .
W . . . . . . . . . .
!ATA St,W77 y +20F .
1r ser - -
r, as ,o + sri +17r . I Ig totsa et.a) +12F/+23F _
~
148 ft-lb
- q.eg . can:
14 -G % , +20F ,
- " S
- a 5 123 I
3 - -
.; I E J. an - -
t O. e e l t
~ E -
- / .
/
-- -__ y._--_.---------___-._-------_.-----
M -
Sitaias. SA 508. c1 2 Distnratie,, Transverse 26' -
e Foutwg None
~
+
HEAT Nsett ANE-191 D
-80 -40 0 5c' 80 120 160 2m 2o 2n m a v.n itst T vteatunt, F c-12 Babcock 8. Wilcox
l l
Fi,;ure C-3. Charpy Imp.ut Data Fra L'nirradiated S ne Stal A HAZ, Lo:igitt. inal Orient it 15 -
er
- a a i ,a e e a w i e = w
- N .
e .
_~,
f3 f] .- = _. - _ _ _ _ - - . _ _ . _ - _ . _ _ .
s- e 3 o e
. 5Y .
e e -
e e e
q w' A R A e a i i s a e i 3 3 5 5 g i *4 a w a g
, e e
. 4 8 4 .w; -
e 5- e e I I
~* .%C 7 * * *
- . e
.c' -
, 5 i ;
a .
r4 1 l' f f f f f a f e a F > < < i . . . . . . .
. - _ ' 4 ? A_SO9.'l.A.RY
- a . I=or --
I, (35 eu e Not determined t.r fa(50 er-tal Net determined
- e ,
(.G ta,0 _Not de t e rm.ined e l e
. :< . RI ocr Not determined- _
l 5 13 -
i .-
L
- l l
- r - * . ,
i . . l l C 1
- P .
I r., e
- e e, o ,
j t
- y . e
_ l l
e
___---____m___________________________.__.
e ,
j 8.C , .
I g irasat SA 508s CL 2 3 ,
Omstavartoo LonRitadinal Fumact None -
nur w oe. ANK.191 o
' + i . . , , , , , ,
-m -o o e so in 160 rn .'o m 5m su <n Test Tasessarues, F c-13 Babcock & Wilcox l
1 i
I I
l l
! Figure C-4 Charpv lepact Data Fror inirradiated Base "c:al A. HAZ. Transverse Orientation 1 11 , , ,
A ..3 e e e
ee 15 - .
J 1
j So . _ - _ . __.*
e j I ;e f y . -
t e M -
g e a e e a e a e a e a 1
- s I 5 s a 3 s e a i
- e 1
E .%) -
I
- e e e
\
~ .9.t -
s* -
\
E --o---~,--- r'-
,5 . /
.tDC -* -
E eg i 1 9 9 9 t t t i e e W i - i . . . . . . . .
1 1 -
Saia suwn _
rnor +20F j Ir _
l l
rag35 m, +37F/+55F
- g i n650 sf-ts
- +32F/+38F ,
(-USEfavs) _92 it-lb
, . 1.0 - RT , +20F _
i :
5 129 I
i l E.
- 1T -
, e -
y e -
- y
- e w 80 - -
O. I e t e e
~ 60 -
1
---_---- p--------__--_--_-----__--_________.
o ep 40 - -
e %rtain SA 508, el 2
- th:tstatice Transverse 20 . -_
Furtnet None NEAT Nuesta ANK-191 0 ' '
-60 40 0 W E0 120 160 200 243 280 320 WJ @
Test Ttwtearweg, F c-14 Babcock & Wilcox
Figure C-5. Char;. ::np.ic t D.a:a Fr. r Uni rrad iat. d ',is Sta: L. l.on,:i t : *.ina. ir i en t at i.'r.
p
= . i e i s
, '3 '$ '
')
w, .
.: e 1
ay _ _ . _ _ _ _ _ - _ _ _ _ . - . _ _ _ - _ - - - - - - - - - _ - - - - - -
E o
- e
.Y, .'5 - e e
a e
_ a a e e f f i t I t
. N)
. . B a 5 5 5 . I '
3 %
3 - *
, e 3 _
e, . .
l [
_ e e /
- e /
.. ,y
_ _f____-_-_________.--..___.
, e i
. . c.~
E f I f I f f f f f a
- i N' . . . . . . . . . . .
. _ - -._f,2
- A _%Wew.------._
e +10F *
.., _aner -
' .o 4 5' 'u ; -26F/+t.F e e
.. _- #50 3,.t,3 -12F/+12F e .
. o e C. 't.L tavr.) ._1+20 t t-It> _. e
. 1.- - 8T g, +20F t '
\
! [ US - -
e
- e
') -
e
.i IT ,
3 l l 5V - S -
l T.
l t
- l l -
1 s,- -
- / _
e l
_f_______ _____________________________
L? -
y .
e
%tte sat SA 508. c1 2 Onitofation tongitudiral
~ ~
FLut o None _ , _ _
Hear h.pwe Ak'S- 192 .
, e i . . , , , , f , ,
IE0 -W 0 4 80 in 160 no 2o :so sn m <n itst itsetmarm, F c-15 Babcock & Wilcox l
l l
i
Figure C-6. Charry 1: pact Data From Unirradiated Base
. Metal B. Transverse Orientat ion Im , , , , , .- .
- 75 -
~
yw ____________ ___
Y . .
w M -
" a e e e f g -
9 s e e
.080 , , , , , , , , , , ,,
i . 6
- f .
g .w -
$r .0@ p Y .
- _ - _ _ _ *-. _p-________._________._____ 3 i /
/ -
2 .02C -
e n , , e i . i , ,
,, i 2m , , , , , J CATA SU W sY
+20F ,
3g , 7,,
in (35 au) 18F/+40F
+37F/+62F_ _
160 Icv (50 st-u)
(-U2 (avs) 112 ft-Ib
-RT ,, +20F I . 11 0 5 123 -
e .
E * -
3 110 -
- . l.
.- .Y a 80 - e 3
5 I 1
- w -
/
_________ -/_________________.______________. t
/
y %ftniat SA 538, C1 2 e
. Onst=tation II.ansverse 20 -
Fwtmet None Heat bene ris-192
= ' ' e i e , , . , , ,
0 vn
'80 '40 0 W 80 1.7 IE) D) 243 2 51 Ir. %9 Test Tenetaat n, F C-16 Babcock & Wilcox
Fi gu re C- 7. Charpy I.T. pac t Data From Unirradiated Ease
- v,.tal B. HAZ, Longitudinal Orientation Ifn . . ,
e se 15 - -
W
- $c ___ _ _ _ _ - - - - - _ _ - - - .
e e
as y e e .
- w. 25 .
M.
4 e 3 *
- e e *
$ .rra .
8
- e
- /
O .04 -
e !
el- - - -- __ . - - _ _ _ _.-_ _ _ _ _ - .
5 ,-.
~ /
~
e et 4 .020 -
e I
a l l I f f i f f f . .
IfC . . . . . . . . . . .
CATA SI; W RY y +20F 1r not .
T3 (35 au) -56F/-25F e
160 I o (50 sr-La) -44F/-30F ,
l (_t;5E (aer.) 140 ft-lb _
I 140 - R T,,,, +20F _
E e e *-
5 120 - -
3 S . e
- Iro , '
? ?
e y e i
wm - .
l
, U. .
l I
~~ 60 -
/ -
- /
-~Y~'~~~~~~~~~~~~~~~~~~~~~~~~~~^~~~~~~~~~~
p 40 -
/ .
e %rta:q SA 508. c1 2
- e Onitorario. Longitt:dinal 23 ~
e FLuto 'an,r-Hear u te Ab'S-192
' ' ' ' ' ' i ' ' ' '
0
{ -80 -40 a 40 80 120 160 2'n 243 2iG 529 70 n t Test it = tearuet, F C-17 Babcock & Wilcox g
' 1 1
a
.a i.
Figure C-8. Ch.irpy lepact Data Frem l'nirradiated Base Metal B, HAZ Transverse Orientation IM , , ,
e se 5 T
- 2 ____ --.
4 s,o . - - - _ - __ _ _ _____
- e e e .
. e T e * -
e 25 .
- e a e e i $
g A e e e a e i
- * * * * * * =s = i
'e a 'e 5 e
. . e * * * $ v*
g .4r, .
l e
.(r r .
- g .--e _ --- a-__-._.--.__t.______-___--_-_.
~ - ,/
e *
/
2 .t20 -
5 g
e i i , , e i t
= I
- C s i i e i e i e . . . i l CA? A i'JWEY I
1.. ,
. T.n +]3F .
T (35 ,u) - 30F/+21F i
- - 32F/+23F ,
l .. .50 .1o (50 es.ts) e q..y g,,p 125 ft-lb _
gg . at +20F .
I aos ~
e l a g e j
l gg - e ,
t
- e 5 -
.T l Y)
A
- W
." gn . e *
- e
.t i ,
~ 9 . /e * .
/ *
-e ---p.----------------------------------.
0 .
/ .
- / %rtaina. SA 508. C1 2
- Onituration .Iransverse ~
Fu;twa No e kat wa AWS-192
. e _a i e i . . . i , ,
l l
l l
1 l
ricure C- 9. Charp) Ir. pac t D.s t.1 Frem t*nirr..diated ;
'a'e ld "=. t a l '
Im i . . y ,. -
l
. l I M U .
( . i t
I l :s r,0 . - - - - . _ . - _ __ _ _ .- ._ _ _ - _ -..
! O i l :
l T
25 -
n
\
.'180 . . . . . . . . . . .
o t.- .rf .
.. .w .
g e
p e-- - --. - ..
9
- .m -
o .
2 *
, n i , i , i e i i , .
. i .) . . i . . . . . . ,
- - - -*J T A T4WA.~.-~--
1 1 '0F l J not --
i n(35 su 2.+81F/+46F i
g,- .in (50 s t-t:3 8 5 F '+118F _
l
. .gM (a.o 66 f:.Ib v
., 1 0 - ti,,, +25F ,
- . i
! '." i 17 -
l : l
\ .
c.
3 Irc .
1 C -
k I e e .
~~ M - * *
/
t.0 - . - 1 e
n tr e s at -4.e l d Me t a.-l -
1 Oe s t =raen M- *
~
l Ftus=ct Nene 1
' ' +
Heat %me *=T-209-l_B 1 i , , , i ,
p(0 C i
- 86 0 0 80 I
.M . ;60 200 24 41 33 M9 .M l i
I l
l l
m APPENDIX D Threshold Detector Information l
i
P Table D-1 lists the composition of the threshold detectors and the cadalum
~~
thickness used to reduce conpeting thermal reactions. Table D-2 shows the cycle 1 measu ed activity per gram of target material (i.e., per gram of ura-nium, nickel, etc.) corrected for the wait time between irradiation and counting. Activation cross sections for the various materials were flux-weighted with a 235 U spectrum (Table D-3),
i E;
Table D-1. Detector Composition and Shielding Monitors Shielding Reaction 11.87 U-Al Cd-Ag 0.02676" Cd 238U(n . f) 1.61% Np-Al Cd-Ag 0.02676" Cd 237 Np(n f)
Ni Cd-Ag 0.02676" Cd SeNI(n.p)Seco
- I 0.66% Co-Al Cd-0.040" Cd 59Co(n,y)6-co 1 0.66% Co-Al None 59Co(n,y)*0Co Fe None 5" Fe (n . p) S".'5
__ ___ _____--__- _ _ -_ - - - - - --- -- - - - - - - - - - ~
T.shle D-2. Oci nt-e 3 3,c l e 8 m utrou Dosimetern "
PONtfrfud. Nuclide bC(/g of nin t t i>r UCf /g Of
,,yt ,1_ Heartton Nuclide act, act r.a t erial tarje_t}c.d)
AD l_(Sid_cl
- 38U -Al 0.03950 2 "U(n.f)FP M'Z r 0.1265 3.202 31.1 95Nb 0.7513 19.02 185.0 103 Ru 0.1528 3.868 37.5 137 Ca 0.004023 0.1018 0.938 l I Cc 0.1175 2.975 39 l Ce 0.06996 1.771 17.2 10E Ru 0.03338 0.H451 8.20 237 Np-Al 237Np (n, f) FP 0.01483 9kZr 0.1976 2.641 183.0 95Nb 1.118 14.94 1038.0 103 Ru 0.1980 2.646 184.0 137 Cs 0.006103 0.08156 5.66 l4! Ce 0.1286 1.719 119.0 l ' Ce 0.09028 1.206 83.8 106 Ru 0.04265 0.5700 39.6 Ni 0.13304 "Nt (n .p) S"Co 5"Co 65.4 492.0 725.0 GLCo 0.139 1.04 3.99 Co-Al (0.625") 0.02019 59Co(n,y)60Co 60Co 4.11 204.0 3.64+4 i
Co-Al (0.5") 0.01514 59Co(n,y)60Co Oco 3.14 207.0 3.70+4 Fe 0.15732 5 Fe (n , p) 5"n' S Mn 3.32 21.1 363.0 SBFe(n,y) 597 , 59Fe 4.81 30.6 9.26+3 4
Table D-2. (Cont'd)
Postirrad. Nuclide ;.Ci/g of pCi/g o(
h>ni t or wtz g _, Re.iction Nuclide act.yfi_
a material ( } ta rge,t te d)
AD 2 (Side)
'J 5 219U -Al 0.06034 2 3*U(n. f) ft' Zr 0.2048 3.394 32.9 9 5N'b 1.219 20.20 196.0 103 Ru 0.2425 4.019 39.0 137Cs 0.006677 0.1107 1.07 l ice 0.1864 3.089 30.0 l Ce 0.1091 1.808 17.6 1:6Ru 0.056*.2 0.9350 9.07 737 Np-Al 0.06486 737 Np(n,f)FP "Zr 0.1703 2.626 182.0 D Nb 0.9037 13.93 968.0 103 Ru 0.1707 2.632 183.0 137 Ca 0.005213 0.080'17 5.58 I I Cc 0.05830 ('.8989 62.4 3 'd' Ce 0.08546 1.318 91.5 IC6 Ru 0.03920 0.6044 42.0 SeNi(n.p)S8Co 58 Co 63.4 492.0 726.0 Ni 0.12886 00N1(n.p)60co I0 Co 0.127 0.9H6 3.77 59Co(n,y)c0Co 4.06 207.0 3.69+4 f
i Co-Al (0.625") 0.01963 60Co
[ Co-Al (0~.5") 0.01611 59Co(n,y)60Co 60Co 3.47 215 2.85+4 E
Fe 0.15715 54Fe (n . p ) '% '% 3.32 21.1 363.0
{' 31.8 9.64+3
'd 8Fe(n,y)S9Fe 59Fe 5.00
,, .. _. ... . ~ v 5
e
.s T.thle D-2. - (Corst 'd)
Postirrad.
Monitor _,_, Nuclide pC1/g of al/g of
_ , y t ,_ g ,_, ,,, R,epc t ipn, ,, No.e,1,1 fft ,9c t ,,i>CI rutertal ta rget (CI)
AD 3 (Back) 2 38g.41, o,972gg '
.3aU(n,f)FP 5 %r 0.1770 2.429 23.5 95Nb 0.9962 13.67 133.0 103 Ru
.0.2169 2.976 '28.9 337 Cs 0.005724 0.07854 0.762 l I ce 0.1614 2.215 21.5-I b Ce 0.09712 1.333 12.9 106 Ru 0.04690 0.6435 6.24
, 237 Np-Al 0.07282 237 Np(n. f) Fp 95 %r
' 0.1442 1.9H0 118.0
' Nb 0.8134 11.17 776.0 103 Ru 0.1479 2.031 141.0 137Cs 0.004581 0.06291 4.37 1+1Ce 0.09234 1.268 88.1 l 4 Ce 0.06793 0.9328 64.8 106 Ru 0.03050 0.4188 29.1 N1. 0.12919 'd"NI(n.p)5nco ', n co 46.4 159.0 s
530.0 I' ON!(ri.p) coco 8. Co 0.0939 r 0.727 2.78 l
Co-Al (0.625") 0.02074 59Co(n.3) coco 60 Co 0.606 29.2 5.21+3
- Co-Al (0.5") 0.01596 59Co(n.y)60Co CD Co 2.06 129.0 2.30+4 I
Fe 0.15455 S t.Fe (n . p) 5Mn 5Mn 2.50 16.2 278.0 seFe(n.y)s9pe *a 9 Fe 2.95 19.1 5.78t3 a
s 1able D-2. (Cont'd)
.j Postirrad. Nuclide UCi/g of pCi/g of Monitor _ wt, g Reaction Nuclido act, pCi en. ate ria l target (Cod)
AD 4 (Front _),
2380 -Al 0.06153 2 3P U (n,f)FP '#'.~r 0.2607- 4.237 '41.1 3b Nb 1.447 23.52 228.0 103 Ru
., 0.3125 5.079 49.3 137 Cs 0.008642 'O.1405 1.36 I"I Ce 0.2271 3.691 35.8 I ' Ce 0.1391 2.261 21'.9 106 Ru 0.06755 1.098 10.6 237 Np-Al 237Np(n , f) FP 95 Zr e 0.06730 0.1959 2.911 202.0-b 9S
, Nb- 1.124 16.70 1.16+3 l IO3 Ru 0.2038 3.028 210.0 IJ7 Cs 0.00$885 0.08744 6.07 l '.1Ce 0.1309 1.945 1 35.0 l Ce 0.09055 1.345 93.4 IOC Ru 0.03897 0.5790 40.2 N1 0.13073 8N!(n.p)*a8 Seco 79.24 606.1 894.0 g 60N1(n.p)00 ccCo 0.1696 1.297 4.96 a.
Co-Al (0.625") 0.01991 59Co(n.1)60Cu 60Co 0.884 44.4 7.9 3+3 Co-Al (0.5") 0.01560 '9' Co(n,y)60Co G co 4.54 291.0 5.20t4 I
w 2 .
_ -. ~ ~~
e
__a
- ~
v i
I.3Id _ P.-l. . .(.CO.n,t ' d ) -
- Postirrad.
_..h"I.to,r,,_
Neclide pC1/g of' t.C1/g of
~
, w t ,, g . Ih actfon Nucl t ale act, i Cf material (b) t arget(c.d)
- Fe 0.16062 ' Fe ( n . p) *'5tn L"M1 4.20 26.1 449.0 L6Fe(n,y)'Fe I'9Fe 6.85 - 42.6 - 1.29+4
(" AnIslyses perf ormed dt 1ynchburg Research Center and reported in the'mencrandum, J. K. Schmutzer to A. 1,. l. owe,'"Oconee'3 Neutron Dosimetry." 5136-55, August.9, 1976.
(b)This column is the disintegration rate'per gram of wire using the postarradiation weight, "I
Th iri column in t he illnint er, rat ion rat e per y, ram o f t arget nucl ide, . vl a:. , NU , ? I7Np, 5"N1, e o y g , t,1cn, *s ** ye ,
s
- s UYe.
(d)The followtug abundances and weight percents were used to calculate the disintegration rate per gram of target nuclide:
ts t 2 38.J 10.38 wt %; 99.27% isotopic 237Np 1.44'wt %; 100% isotopic NI 100 wt %I ',Rygg g7,77g gygggggg, M,ggg 7g,gg g,gggpgp Co 0.66 wt Zi S9 Co 100% 1sotopte Fe 100 wt %; 5*.Fo 5.82% isotopic, 'UFe d 0.33% isotopic I
e e
So 74
. O F
h
l Table D-3. Dosimeter Act1vation Cross Sections (*
Energy range, C MeV 237 3p 23eg Segt __ 5 Fe 1 13.3 -15.0 2.231 1.073 0.460 0.425 2 10.0 -12.2 2.34 0.981 0.622 0.537 3 8.18 -10.0 2.31 0.991 0.659 0.583 4 6.36 - 8.18 2.09 0.917 0.638 0.572 i 5 4.96 - 6.36 1.54 0.60 0.54 0.473 6 4.06 - 4.96 1.53 0.562 0.403 0.325 i 7 3.01 - 4.06 1.616 0.553 0.264 0.206 8 2.46 - 3.01 1.69 0.550 0.139 0.096 9 2.35 - 2.46 1.695 0.553 0.089 0.0524 10 1.83 - 2.35 1.676 0.535 0.051 0.022 11 1.11 - 1.83 1.593 0.229 0.0128 0.0115 12 0.55 - 1.11 1.217 0.008 0.00048 --
13 0.111 - 0.55 0.1946 0.00013 - -- l 14 0.0033- 0.111 0.0410 - - --
1
("}ENDT/4 values flux-weighted with a fission spectrum.
i D-8 Babcock s.Wilcox
END l i
ER0 PHOTOGRAPHER _x,- _ _ _.
DATE m e z2__
/;$,, ' j..
ili(15:,# 'jI?
- , .Mg .-
1." _
MICROFILM SECTION N AVV PUBLICATIONS AND PAINT 8NG SE RVICE OFFICE BUILDING 157 2. WASHINGTON NAVY YARD W ASHINGTON. D C. 20374