ML19308B550
| ML19308B550 | |
| Person / Time | |
|---|---|
| Site: | Oconee  |
| Issue date: | 05/31/1977 |
| From: | BABCOCK & WILCOX CO. |
| To: | |
| References | |
| BAW-1437, NUDOCS 8001090544 | |
| Download: ML19308B550 (75) | |
Text
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L: - Q .* val?,??.i g',' ,', f,.iGN 5IN ANALYSIS OF CAPSULE OCII-C FROM DUKE POWER C0".PANY OCONEE NUCI EAR STATION. UNIT 2
- Reactor Vessel ;'at.crials Surveillance Progra:2-Babcock &Wilcox soolo906TT g
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BAW-1437
.%y 1977 t-_
ANALYSIS OF CAPSULE OCII-C FROt1 DUKE POWER COMPANY OCO:EE NUCLEAR STATION UNIT 2
- Reactor Vessel-Materials Surveillance Program -
a t by A. L. Lowe. Jr. , PE E. T. Chulick H. S. Palme C. L. Whitmarsh C. F. Zurlippc. l B&W Contract No. 595-7020-61 1 BABCOCK & WILCOX- i 1 Power Generation Group Nuclear' Power Generation Divisien l L
'P.-O. Box 1260 Lynchburg, Virginia 24505 Babcock & Wilcox l' ~
1 1 i' l (
y.; - . o - o.. b Y CO?.7 E!.~I_S_ Page
- 1. =lNTKostCI:05.. . . . . . . . . . . . . . . . . . . . . ..... 'l-1
-. ~ iL\GX;HtT *Cr; . ';. . . . . . . .. . . . . . . . . . . . - ..... 2-1 L . SUnVFILt_iNCE PKtER.V1 DESCRIPTION . . . . . . . . . . . . .... 3-1
- 4. PREIR RAD I AI ION TES TS . .. . . . . . . . . . . . . . . . .... 4-1 4.1. re ,$ g e Tests .. . . ... . .. . . . . . .
' . . .... 4-1
-4.2. .l=pakt Tests . . . . .. . . . . .. . . . . . . . . ....
4-1
-5. POST!RRANIATION TESTS . .. . . . . . . . . . . . . . ....
5-1 5.1. Thensal Monitors . . ... . . . . . . . . . . . . .... 5-1 l 5 . .' . Chinical Analysis .. .. . . . . . . . . . . . . . .... 5-1
- 5. J . Ten s i l e - Tes t Results-. .. . . . . . . . . . . . . .... 5-2 5.4 Cha rpy Y-Notch != pact Test Results . . . . . . . . .... 5-2
- 6. . NEUTRON TOSINETRY J
r
. . ... . . . . . . . . . . . . ... . . 6-1 6.~ l . Int'sdustien... .
.5.2.
. . .. . . . . . . . . . . . . . .... 6-1 Ana4ytical Approach .. . . . . . . . . . . . . . .... 6-2
- 6. 3. -Results . . . . . . - . . .. . . . . . . . . . . . . .... 6-3 7.- DISCUSSION OF CAPSULL RESULTS
. . . . . . . . . . . . . .... 7-1 7.1. Preltradiation Property Data . . . . . . . . . . . .... 7-1 7.2. Irradiated Property Data'. . . . . . . . . . . . . . .... 7-1 7.2.1. Tensile Properties . . . . . . . . . . .. . .... 7-1 7.~2. 2. . lupact Properties' . . . . . . . . . . . . ....
7-1
- 8. DETERMINATION OF RCPB PRESSURE-TDfPERATURE LIMITS . . . .... 8-1
- 9. Sr.TtARY OF RESULTS . . .. . ... . . . . . . . . . . . ....
9-1
- 10. SURVEILLAL:E CAPSULE RD 0 VAL SCHEDULE . . . . . . . . . .... 10-1
- 11. - ' CERTIFICATION . .. . . . . .... . . . . . . . . . . .... 11-1 12.' REFERENCES . .. . . . . . . . ... . . . ... . . . . . ....
12-1 111 - Babcock & Wilcox
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e EL -
l C_ONTENTS NonL'd) Page A PPF'O IXES A. r:eactor Vewel Surveillance Program - Eackground Data . and InformatIon . . . . . . . . . . . . . . . . . . . .. . A-1 E. Prefrradiation Tensile Da t.a . . . . . . . . . . . . ... . B-1 C. Preirradiation Charpy impact Data . . . . . . . . . .. . . C-1 D. Thresh.11d Detector Information . . . . . . . . . . . .. . D-1 List of Tables table 1-1. Speelmens in Surveil,ance Capsule OCII-C . . . . . . . . . .. . 3-2 1- 2. Chemistry and licat Treatment of Surveillance Materials . ... . 3-3 b 1. Chemiatry and lleat Treatment of Correlation Material - liest A-Il95-1, A333 Grade n. Class 1 . . . . . . . . . . . . . ... 3-4 i-1. Chemist ry Data on Unitradiated RVSP Material . . . . . . .. . . 5-2 i-2. Tc:n. i le Properties of Capsule OCII-C Base and Weld Metals Irradiated to 9 . *. 3 1037 n/cm- . . . . . . . . . . . . . .. . 5-3 5 - 1. Ciia r;iy Impact Data for Capsule OCII-C Base Metal Irradiated to 9.43 ' 19 ' n/en' 1
. . . . . . . . . . . . . . . . . . ... . 5-4 ;-4 Charpy impact Data for Capsule OCII-C Weld Metal (WF-209-1A)
Irradiated to 9.43 - 1017 n/cm- . . . . . . . . . . . . .. . . 5-5 i- i . Charpy Irpact D.ata t or Capsule OCII-C Correlation Monitor "aterial Irrad tated to 9.43 1017 n/cm2 , ilSST-PL-02, 1:. .i t 1195-1 . . . . . . . . . . . . . . . .. . . . . . . . .. 5-5 6-1. SarveilJance Capsule Detectors . . . . . . . . . . . . . . ... 6-5 4 .' . Flux Adjustment Factor . . . . . . . . . . . . . . . . . .... 6-5 h - 1. Do.. i r:et e r Ac t i va t ion s . . . . . . . . . . . . . . . . . . .. . 6-6 o-4 Nernalized Flux Spectra. E L MeV . . . . . ... . . . . . . .. 6-6 n- i. Neutron Fluence . . . . . . . . . . . . . .. . . . . . ... . 6-7 n-+>. Predicted Fast Neutron Fluence in Pressure Vessel for 10 EFPY , 6-7 7-1. t'emparison of Tensile Test Results . . . . . . . . . . . ... . 7-3 7-2. Observed Vs Predicted Changes in Irradiated Charpy Impact Propertles . . . . . . . . . . . . . . . . .. . . . . . .... 7-4 8-1. l'a t a for Preparat ion of Pressure-Tenperature Limit Curves for Duke Power Company's Oconee Nuclear Station Unit 2 - Applicable Through Eight Full-Power Years . . . . . . . . ... 8-4 A-1. Surveillance Program Materials Selection Data for oconee 2 . .. A-3 A-2. Materials and Specimens in Upper Surveillance Capsules OCil-A, OC l l-C, and OCII-E . . . . . . . . . . . .
. . . . . . ... . A-4 A- 1. Materials and Specimens in Lower Surveillance Capsules OCII-B, 0C11-9, and OCll-F . . . . . . . . . . . . . . . . . . . .. ..
A-4 11 - 1 . Prei radiation Tensile Properties of Shell Plate Material, Ilear AAW 163 . . . . . . . . . . . . . . . . . . . . . . ... . B-2 R-2. Preirradiat ion Tensile Properties of Shell Plate M tterial - IIA 2, lleat A.W 16 3
. . . . . . . . . . . . . . . . . . . .. . B-3
- iv - Babcock s. Wilcox
r- [
- f. Tables (Cont'd?
Table Page
&-3. Preirradiation Tensile Properties of Shell Plate Material, Heat AWG 164 . .. .. . . . . ................ 3-4 S- 4 . .Preirradiation Tensile Properties of Shell Plate Material -
IIAZ. Ileat AWG 164 . . . . . . .. ............... S-5 S- 5, Preirradiation Tensile Properties of Weld Metal -- Longitudinal, VF-209-1A . . . . . . . . . . . ................ S-6 C-1. Preirradiation Charpy I= pact Data for Shell Course Material - Longit udinal Orienta: ion, Heat AAW 163 . . . . . . . . ..... C-2 C-2. Preirradiation Charpy Icpact Data for Shell Course Material - Transverse Orientation, Heat AAW 163 ............. C-3 C-3. Preirradiation Charpy Icpact Data for Shell Course Material - IIAZ . Longitudinal Orientation, fleat AAW 163 . . . . .. .... C-4 C-4. Preirradiation Charpy impact Data for Shell Course Material - IIAZ. Transverse Orientation, Heat AAW 163 . . . .. . .... C-5 C-5. Pleirradiation Chirpy Impact Data for Shell Course Material - C-6. Longitudinal Orientation. Heat AWG 164 . . . . . . . . ..... c-6 Preirradiation Charpy Icpact Data for Shell Course Material - Transverse Orientation, IIeat AWG 164 ............. C-7 i C-7. Preirrad.stion Charpy Impact Data for Shell Course Material -- IIAZ, Lons;itudinal Orientation, Heat AWG 164 . . . . . ..... C-8 C-8. Preirradiation Charpy Impact Data for Shell Course Material - IIAZ, Transverse Orientation, lient AWG 164 . . . . . . ..... C-9 C-9. Preirradiation Charpy Impact Data for Weld Metal, WF-209-IA .. C-10 D-1. 1ktector Composition and Shielding . .............. D-2 D-2. Oconee 2, Cycle 1 Neutron Dositeters . ............. D-3 List of Figures Figure 1- 1. Reactor Vessel Cross Section Showing Surveillance Capsule Locations . . . . . . . . . . . 3-1. ................ 3-5 Impact Data From Irradiated Base Metal A. Longitudinal orientation . . . . . . . . . . . 5-2. ............... 5-6 Impact Data From Irradiated Base Metal A Transverse Orientation .. . . . . . . . . . 5-3.
............... 5-7 Impact Data From Irradiated Base Metal A - RAZ, Longitudinal Orientation 5-4. . . . . . . . . . . ........ . . . . .... 5-8 Impact Data From Irradiated Weld Metal, Transverse Orientation . 5-9 5-5. Impact Data From Correlation Monitor Material, Transverse Orientation . . . . . . . . . . ...
6-1. ............. 5-10 Fast Neutron Fluence of Surveillance Capsule Center Compared to Various Locations Through Reactor Vessel Wall for First 10 EFPY . . . . . . . . . . . . . ............... 7-1. 6-8 Irradiat- Vs Unirradiated Charpy impact Properties of Base Metal, Longitudinal Orienration 7-2.
............... 7-5 Irradlated Vs Unirradiated Charpy Impact Properties of Base Metal Transverse Orientation . . ............... 7-6
-v- Bcbcock 8. Wilcox
F!aures Ccmt'd), Figure Page 7- 3. I r radiated Vs Unirradiated Charp . Impa c t Properties of Base Metal, ILiZ . . . . . . . . . . . 7-7 7-4 Irradiated Ys Unirradiated Charpy. . . . . . . . of Impact Properties . . Weld Metal, Transverse Orientation . . . . . . . . . . . . . . . . . 7-8 7-5. Irradiated Vs tnirradiated Impac: Properties of Correlation Monitor Material, Longitudiaal Nientation . . . . . . .... . 7-9 8-1. Pred icted Fast Neutron Fluences at Various Locatior.s Through Reactor Vessel Wall f or First 10 EFPY . . . . . . . . . . . . . 8-5 8-2. !!eac t or Vessel Pre ssure- Tempe ra t are Limit Curves for Nornal operation - Ifeat up. Applicable t er First Six Ef fective Full-Power Years . . . . . . . . . . . . . . . . . . . . . . . . . 8-6 8- 1. @ actor Ver, el Pressure-Temperature Limit Curves for Normal Operation ~ Cooldown, Applicable for First Six Effective Full-Pi wer Yea ra . . . . . . . . . . . . . . . . . . . . . .. . . . 8-7 b 4. Seac t.or Vessel Pressure-Temperat :re Limit Curve for Inservice
! eak and Ilydrostat ic Tests, Appl: cable for First Six A-1.
Effective Full-Poser h ars . . .. . . . . . . . . . . . . . . . 8-3 i.eeation and Id"ntification of wateria m Used in Fabrtcation
..i I:eactor Pressure Vessel . . . . . . . . . . . . . . . . . . . A-5 C-1. Impact Data from Unirradiated he Metal A, Longitudinal orientation . . . . . . . . . . . . . . . . . . . . . . . . . . C-11 t.- J . 1:apac t Data From Ur.f rradiated Base !*etal A. Transverse orientation . . . . . . . . . . . . . . . . . . . . . . . . . . C-12 r- 3. Impact D.ita From Unirradiated 2ase Metal A, liAZ , Longitudinal j
e ,- 4 . or!cnt.ition . . . . . . . . . . . . . . . . . . . . . . . . . . C-13 i I. pact Data from Un* irradiated b e Metal A, HAZ, Transverse orlentation . . . . . . . . . . . . . . . . . . . . . . . . . . C- 1.",
-3. 1" pact Data from Unirradiated Case Metal B, Longitudinal orientation . . . . . . . . . . . . . . . . . . . . . . . . . . C-15 C- h. Impact Data From Unirradiated b e Metal B, Transverse Orientatten . . . . . . . . . .. . . . . . . . . . . . . . . . C -16 C-1. Impact Data From Unirradiated b e Metal B, HAZ, Longitudinal Orient.ition . . . . .
. . . . . . . . . . . . . . . . . . . . C-17 c-8. I m; .w t Data From Unirradiated Ecse Metal B, HAZ, Transverse drientation . . . . . . . . . . . . . . . . . . . . C-18 8.-u. Inpact luta From Unirradiated WId Metal, Trap c Trse . . . . . .
orsentation . . . . . . . . . . . . . . . . . . . . . . ... . C-19 _ ,, 3 _ Babcock r. Wilcox
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i L I-4 p 1. INTRODUCTION
~ This.. report" describes the results of the examination of the first capsule of Duke' Power Company's Oconee Nuclear Station. thit 2 reactor vessel surveil-if . lance program. ' The objective of the program is to monitor the ef f ects of neu-tron irradiation on'the tensile end impact properties of reactor pressure ves-sel materials under actual operating conditions. The surveillance progthm for Oconee -2 was designed and furnished by Babcock & Wilcox; it is described in BAW-10006A.I The program was planned to conitor the effects of neutron irrad-lation on the reactor vessel materials for the 40-year design life of the re-actor pressure
- vessel.
i- [ The surveillance program for Oconee 2 was designed in accordance with E185-66 and thus is not in compliance with Appendi es G and H to 10 CFR 50 since the requi renents did not exist at the time the program was designed. Because of this di f f erence, a !dit 'onal . tests and evaluations were required to ensure meeting the requirements of 10 CFR 50, Appendixes C and II. The recocenenda-t ion.s for the future operation of Oconee ? is -luded in this report do ce= ply with these requirecents. 1-1 gekock & Wilcox
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- 2. EACKGROUND 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 reglen of the reactor vessel is the most critical re-
~
gion of the vessel-because it is exposed to neut ron irradiation. The general
~ ef f ee,ts of fast neut ron irradiation on . the mechanical properties of such low-alloy ferritic steels as SA508 Class 2 forgings used in the fabrication of t he' Oconee 2 reactor vessel are well characterized and documented in the lit-erature. The low-alloy ferritic steels used in the beltline region of reactor vessels exhibit an increase.in ultimate and yield strength properties with a cor responding decrease in ductility af ter irradiation. In reactor pressure l
veissel steels, the most serious mechanical property change is the increase Ln temperature for.the transition from brittle to ductile fracture accompanied by ,. a reduct ion in the upper she'If impact strength. AppenJIx G to 10 CFR 50, " Fracture Toughness Requirements," specifies minimum fracture toughness requirements for the ferritic materials of the pressure-
- retaining coeponents of the reactor coolant pressure boundary (RCPB) of water-coolea power reactors and provides specific guidelines for determining the pressure-temperature limitations on operation of the RCPB. The toughness and operational requirements are specified to provide adequate safety margins dur-ing any cendition of: normal operation, including anticipated operational oc-t eurrences and system hydrostatic tests, to which the pressure boundary may be
. subleeted over its service lifetime. Although the requirements of Appendix G to 10 CFR 50 became effective on August 13, 1973, the requirements are appli-cable to all ' boiling' and pressurized water-cooled nuclear power reactors, in-cluding those under construction or in operation on the ef fective date.
Appendix it to .10 CFR 50, " Reactor Vessel Matericls Surveillance Program Re-quirements," defines the material surveillance program requ, ired to monitor changes in the fracture toughness properties of ferritic materials in the re-actor vessel beltline ragion of water-cooled reactors resulting f rom exposure 2-1 Babcock & Wilax t- .
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= '-
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- to neutron irradiation and 'the thermal environcent. Fracture toughness test data'are obtained from naterial specimens withdrawn periodically fron the re-
..ctor' vessel. These data will permit detereination of the conditions under
. unien the vessel can be operated with adequate safety margins against fracture throughaat i ts service life.
A nethed for guarding against brit t le f racture' in reactor pressure vessels is described in Appendix G.to the ASME boiler and Pressure Vessel Code, Section III. This met hod utilizes f racture cechanics can'cepts and the reference nil-ducti,11ty temperature, RTNDT' ' * " "* ""
- E'#* '" * # E '
.ctrat nil-ductility transition te=perature (per AST11 E-208) or the te=pera-ture-that 1.s 69F below that at whleh the eaterial exhibits 30 ft-Ib and 35 utia la:eral expansion. The RT NDT o a g ven ma crial is used to index that msterini to a reference stress intensity. factor curve (K 73 curve), wlitch ap-pears in Appendix C of ASME Section III. The K IR curve is a I wer bound of dynario. 3 tatie, and crack arrest fracture toughness results obtained from several heats of; pressure vessel steel. When a given material is indexed to the V; curve, allowable . stress intensity factors can be obtained for this ma-
- criai as a function of temperature. Allowable operating Ifnits can then be art errined using t hese allowable st ress intensity factors.
ihe RT .g :snd. it. turn, the operating limits of a nuclear power plant, can be
.d N>ted to account for the ef fec ts of radiation on the properties of the re-wr vs.w el caterials. 1he radiati..n enbrittlement and the resultant changes in r.schanical properties of a given pressure vessel steel can be monitored by
-ere. !!!an progran in which a surveillance capsule cont.aining prepared speci-w of tLe re.setor vessel caterials i , p'eriodically renoved from the operating
- i. it.sr react or and the specimens tested. The increase in the Charpy V-notch
>"-ft-Ib tenperature, or the increase in the 35 mils-of lateral expansion tem-
;erature, whichever results in the larr,er tecperature shif t due to irradiation, 1
is . hided to'the orie,Inal RT.7 to adjust it for radiation enbrittlement. This adjustec Kr;gg7 is used to index the material to the KIR curve, which, in turn is used to set operating limits for the nuclear power plant. These new linits take into account the ef fects of irradiation on the reacter vessel materials. t
- -2 , Babcock t \Vilcox
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4.
- 3. SURVEILi>5CE PROGRAM DESCRIPTION The surveillance program for Ocence 2 comprises six surveillance capsules de-
. signed to monitor the ef fects of neutron and thermal environment on the mate-riats' 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 thermal ~ shield and the vessel wall at the locations shown in Figure 3-1. The capsules, placed two in each holder ts.be, are positioned near the peak axial and azimuthal neutron flux. KAW-
= 10006A f r.cludes a . full description of capsule locations and design. '
Capsule 0Cll-C was removed during the . first refueling shutdown of Oconee 2. This capcule contained Charpy V-notch impact and tensile specimens fabricated of SA502, Class 2 steel, weld metal, and correlation steel. The specimen con-tained in the capsule as described in Table 3-1, and the chemistry and heat-treatreat of the surveillance naterial are described in Table 3-2. The capsule also contained longitudinal- Charpy V-notch specimens from correla-tion naterial obtained f rom Plate 02 of the USAEC Heavy Section Steel Technol-ogy progran. This 12-inch-thick plate of ASTM 533, Grade B, Class 1 steel was pralnced by the' Luken Steel Company . (heat A-1195-1) and heat-treated by combus-t ion ::ngineering. ' The chemistry and heat treatment of the correlation cate-rfal are described in Table 3-3. All test specimens were machined from the 1/4-thickness location of the plates.
- Charry V-notch and tensile speci= ens from the vessel material were oriented
-i th their longitudinal axes parallel to the principal' rolling direction of the plate; .the speelnens were also oriented transverse to the prit.cipal roll-Ing direction.
Capsule OCII-C centained dosimeter wires, described as follows: 3-1 Babcock a. Wilcox
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p.- .-
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j- , ,
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3~
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. , 7. . * , ,
- \.:, n ~ ,
~ Shielding
'Doaineter dire.
- t'-Al ' al loy -
- ~ Cf-Ag alloy-
;;p-Al' alloy' Cd-Ag' alloy
- Nickel Cd-Ag alloyc j . c: .0.660 Co-Al aIIny Cd
'O.66* Co-A14<111oy' Lne
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- Fe ! :*one Therma 17:or.i tors of ' lot.c-r elt ing eutect ic alley.s: vere incl .2ded in the capsule.
- The cutectic alloys-and. their neiting points are as foliews:
.'-551 t ing point, F
~
~
Alloy *
- - 9 0.' Ph , 5% Ag, 3% Sn 55i 97.5% Pb, 2.5% Ag 58G
..'97.5% Ph -1.5% Ag, 1.02' Sn 5a8 Lead -621 Cadniu= 610 y;
4 Table 3-1. Speci_nens in' Surveil:ance Capsule (CII-C L. of spec imens ,
~
?!aterial desci'.ption- Tensile Charov t.
+
- Weld cetal.J F-209-1A 4- -
8
'lient-af fey ted zone "A" (HAZ) . .
IIcat . AAW-163.. Longitudinal 0 8
- Baseline caterial, plate "A",
= IIeat- AAW-163, ' Longitudinal 4- 8 Transverse 0 4
- -. Correlation,11SST, Plate 02
-(A-1195-1). 0 8 .
4 :: Total per. capsule- '8 36 1 1
*e OA
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Table 3-2.' .. , Che:21stry and **a:- Treat =cer -
- of surieillartcir"F.aterials :
, - *J,e mi ca l A:ia : e s i s . "
~ -
, Erat; 11e.a t Veld retal
' Ele =e.:- A.W-143 . AWC WF-209-1A C 0.24 -0.2L O.067
.Ma ~ 0.61 0.62' 1.5E P. 'O.006 -0.0L0 ~0.020 15 0.012- 0.010. 0.GC5-Sir 0.25- 0.23 0.36 NI _. 0.73. 0.80 0.a'
.h - - 0. 6 2 ' .0.54 0.33 D Cu . 0.04 0.02 0.30
- h n Treat:cy._*. _.
Time, pl No. Tetep . F- h Coolir.2 JJ'-163 4 ;620-1660- 4.0 Cold water quench -
'1570-1610 4.0 Cold water quench
- 240-1280 10.0 Cold water quench 1100-1150 "40.0 Turnace eccled
- A G 164 1620-1660 _ 4. 0 Cold water qu nch --
~%70-1610' 4.0 Cold. water quench
.. *0-1250- 10.0~ Col. water quench 1100-1150 '. 0. 0 72rnace ccoled
... 47-209-1A :110')-1150 33.0 Turnace cooled
.+
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9 E
+
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- Table 3-3. Chemistry-an'd' Heat Treatment of Correlation. _-
~ Material - Heat A-1195-1.. A533 Grade S.
Class 1 (HSST Plate 02) Chemical' Analysis -(1/4T)(#I Element =
*'t -
'C r -1?
3tn : 1.34. P' 0.013'
.S 0.013-
.Si 0.01- '
- N 1 --- 0.t;
... .% : .0.53 - ,
Cu - 0.17 lise t Treat e:en
-1. Lirmattzed at 1673r : 75F.
-2. 1600F '* 75F for 4 ' hiwatequenched.
J. 1225F t -25F f or 4 1.ffurn. ace-cooled.
- 4. Il25F.t 25F for 40 h/ furnace-cooled.
-'(a) enNL-446 3.
- (b)Per plate sectien identif'icaties card.
4 i ' 4 t 6 2 3-4 Babcocka WHcom 1 e i
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Figure 3-1. Reactor Vessel Cross Section Showing Surveillance Capsule Locations j Sur.eillane: C ivsule rolder j , Tubes - Capsule = All-C, 1 OCII-D t / r
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~
- 4. PRELF. RADIATION IESTS Unirradiated material was evaluated for ti.o purposes: (1) to establish a baseline of data to which irradiated propt-ties data could be referenced, and
(.'s to. determine those materials properties to the extent practical from avail-able mat erial, as required for ecepliance with Appendixes C and H to 10 CFR 50. 4.1. -Tensile Tests Tvn e. l l e 1.pecimens we re fabricated fro:n the reactor vessel shell course plate and we ld met al . Ths- subsize specierns were 4.25 inche.' long with a reduced oction I.750. Im het long by 0.35' inch in diameter. They were tested on a St.000-lb-load cap.n ity universal test machine at a crosshead speed of 0.005 inch per minute. Test conditions wre in accordance with the applicable re-
<psiren. nt s of ASIM A370-72.
For each material type and/or condition, six e rec trwns in groups of three were tested at both room te=perature and 570F. An l.VDI-t ype cla=p-end s crr ~on exten.ometer was used to determine the 0.2% vield point. 1he t ension-compression load cell used had a certified accuracy
- het t. r t ita n U.S? of full ucale (10,000 lb). All test data for the pre-1 raJ!ation tensile speeleens are given in' Appendix B.
4... 1--act Test s Charpy V-notch impact t est s were conducted in accordance with the requirements of ASUI Standard MethNs A370-72 and E23-72 on a remote ,;a rolled impact test-er certified to meet Wa t e r t own s tane.a rd s . Test specimens were of the Charpy
. Y-notch t ype, which are 0.39? inch square and 2.165 inches long.
Prior to t e..t ing 'specI= ens were te=:perature-conditioned in a combination re-si tance-heated /carben dioxide-cooled chamber, designed to cover the tempera-ture range from -85 to +550F. The specimen support arm, which is linked to the pneunatie t'ransfer techanism, is instru=ented with a contacting thermo-couple allowing instantaneous specimen temperatirre determinations. Specimens were transferred f ron the conditioning cham!:er to the test frame anvil and pre-elsely pretest-positione.1 with a fully automated, remotely controlled apparatus. 4_; Babcock & Wilcox
+
' e-T Transfer times were less_than 3 seconds and. repeat within 0.1 second. Once the specimen was positioned,-the electronic interlock opened, and the pendulum was released f rom its preset drop height. Af ter failing, the specimen, the
- l. Leser pendu".un was slowed on its return stroke and raised back to its start position. I= pact test data for the unirradiated baseline reference materials are presented in Appendix C. Tables C-L through C-9 contain the basis data which are plotted in Figures C-1 through C-9.
I { 3 I l l , 4_ - BabcocklL VVilcox , i l
1 f. 7
- 5. POSTIRRADIATION TESTS 5.1. Thermal Monitors-Surveillance capsule OCII-C contained three temperature monitor holder tubes, each containing five fusible alloys with different melting points ranging from 558 to 621F.
All the ther=al nonitors at '558, 580, 588, and 610F had =elted, while the moni tor at 621F remained in its original configuration as initially placed in the capsule. From these data it was concluded that the irradiated specimens had been exposed to a maximun temperature in the range of 588 to less than 610F during the reactor vessel operating period. These high temper-atures are not characteristic of operating conditions and probably occurred during reactor startup hot-functional tests. There appeared to be no signif-icant temperature gradient along: the capsule length. 5.2. Chemical An.nlysis two broken impact specimen halves taken at random from each of the two unir-radicited base metals (heats AAW-163 and AW-164) and the weld metal (WF-209-1A) were analyzed for nickel, copper, phosphorus, and sulfur contents to ver-ify original mill test report data. To minimize possible scatter in data due t o surf ace condi t ions, specimens were mechanically cleaned prior to analysis, four sets of copper analysis data were obtained from each of the six specimen halves, and three analyses were performed for the remaining elements (nickel, phosphorus, and sulfur). the following analytical techniques were employed:
- 1. Nickel and copper content B5W Standard Analytical Methodby x-ray XR-2.fluorescence analysis using 2.
phosphorus content by colorimetric absorption measurement us - ing " molybdenum blue complex" ~ solutions in accordance with ASTM E350-72 and B&W Standard Analytical Method P-1.
- 3. . Sulfur content by gravimetric precipitation of EaS0s in ac-cordance with ASTM E350-72.
5-1 Babcock & Wilcox
The data f rom chemical analyses were averaged, and the results are reported in Table.5-1 along with the applicable standard deviation values.
- 5. 3. Tensile Test Results The. results of the postirradiation tensile tests are presented in Tabic 5-2.
Tests were perf orred on specimens at both room te=perature and 570F using the sa:e test procedures and' techniques used to test the unirradiated specimens (section 4.1) . In general, the ultimate yield strength of the material in-cre ased slightly with a corresponding slight decrease in ductility; both ef-fects werc_the result of neutron radiation damage. The type of behavior ob-rerTe.1.and the degree to which the material properties changed is within the , rante of changes to be expected for the radiation environment to which the epecinens were exposed. The r.-sults of the preirradiation tensile tests are presented in Appendix B. 5.'. Charpy V-Notch Impact Test Results The test results from the irradiated Charpy V-notch specimens of the reactor v .--c! beltline material and the corralation monitor material are presented ta r bles 5-3 and 5-4 and Figures 5-1 through 5-5. The test procedures and tecalq 2cs were t he same as those used to test the unirradiated specimens ( ,e r : : en 4. 2 ) . The data show that the caterial exhibited a sensitivity to Irrai:atien within the values predicted f rom its chemical composition and the
- la m o to which it was exposed.
The < salts of the preirradiation Charpy V-notch impact test are given in Ap-pencir i. Table 5-1. Chemistry Data en Unitradiated RVSP Material CYN-
- Composition, wt specimen Material type /
No. heat No. Ni ! 0.01 Cu ! 0.02 P2 0.003 S: 0.004 EE 7bh Base /AAW-163 0.S0 0.05 0.014 0.006 FF 620 Base /AWG-164 0.78 0.07 0.017 0.006 EI 016 We ld /WF-209-I A 0.60 0.34 0.013 0.010 5-2 Babcock & \Vilcox i
a Ta'ble 5-2.~ Tensile Properties 'of Capsule OCII-C Base and Weld Metals Irradiated to 9.43 x 101 ' n /em2 (E > 1 MeV) Strength, psi Elongation , % Test Specimen temp, ~ Yield Uniform Ult. Total Red'n of
.,1D No. F (YS) (UTS) (CE) (TE) area. I Base Metal - 11 eat A.W-163, Longitudinal EE 703 RT 69,150 90,550 10.61 27.64 69.95 EE 713 RT 71,740 93,190 9.54 26.50 70.34 Mean 70,445 91.870 10.08 27.07 70. 15 Std dev'n 1,295 1,'320 0.54- 0.57 0.20 EE 714 580 66,270 90,650 10.46 28.07 72. 04 EE 716 580 72,630 89,910 9.55 28.64 71.03 Mean 69,450 90,280 10.01 28.36 71.54 Std dev'n 3,180 370 0.46 0.29 0.51 Veld Metal - WF-209-1A EE 106 RT 89,280 104,610 11.36 23.57 EE 114 58.4 RT 88,480 103,810 10.79 23.00 59.1 Mean 88,880 104,210 11.08 23.29 58.75 Std dev'n 400 400 0.29 0.29 0.35 EE 116 580 81,060 102,010 10.23 19.29 41.4 EE 122 580 80,060 99,200 21.14 9.52 46.9 .'k an 80,560 100,610 9.88 Std dev'n 20.22 44.15 500 1,410 0.36 0.93 2.75 i
> l 5-3 Babcock a Wilcox
- J c
~
l
. s, l
.l
< Table 5 -3. -Charpy Impact Data for Capsule OCII-C Base Metal Irradiated to 9.43 10 17 n/c d (E > 1 Mev)
~ Test Abs Lateral Shear
.Specinen temp.- energy, expansion, fracture,-
ID No. F ft-lb 103 in. :
~
Base Metal - Heat AAW- 163, 1.cn gi t ud inal EE 708 .319.8 119.5 64 100 725 229.2 135 66.5 100 703 139.1 132 63 100 710' ~40.5 112 71 45 724 19.8 72.5 49.5 14 726 10.6 62.5 41 3 713 0.6 38.5 25 3 738 .-24.6 6.5 3 0 Base Metal - Itent AAW-16 3. T-ansve rse EE 604 140.1 121 71.5 100 601 41.0 86 57.5 35 612 10.0 57 39.5 10 623 -10.2 65 47 12 liAf. - Heat AAW-163, Longitudinal EE Alb 319.4 128 63.5 100 44n 238.5 138 60 100 405 139.3 88.5 52 100 4 33 85.1 75 42 96 404 41 147 69.5 95 420 0.2 11', 63 40 431 -39.5 87.5 56 20 4 32 -79.5 68.5 37 O l I l
)
5-4 Babcock & Wilcox
Table 5-4. Charpy Impact Data for Capsule OCII-C Jeld Metal
-(WF-209-1A) Irradiated to 9.43 = 1017 m/m2
-(E > 1 MeV)
Test Abs Lateral Shear Specicen 'tenp, energy, expansion, fracture, ID No."' F fr-Ib 103 in. __ EE 009 321.2 51.5 37.5 100 028 230.6 54 39.5 99 040 180.0 51.5 35 96 004 139.1 50 32 97 034 114.9 40.5 25.5 75 025 86.0 38 25 55 017 59.8 32.5 24 35 001 40.3 28 20 25 Table 5-5. Charpy Impact Data for Capsule OCII-C Correlation
>kinitor Material Irradiated to 9.43 = 1037 n/cm2 (E > I MeV) IISST-PL-02, Heat A-Il95-1 Test Abs La teral 5;: ear Speci.,en teep, energy, expansion, fracture, ID No. F fr-lb 103 in. %
EE 921 319.0 103.5 57.5 100 958 230.0 193 59 95 941 190.4 92.5 54 E3 912 140.5 65 47.5 40 939 119.6 58 39 28 930 99.7 37 26 18 928 70. - 28 22.5 12 911 39.s 23 19.5 4
]
i i 5-5 Babcock & Wilcox
-\
l
l l l Figure 5-1. Impact Data From Irradiatcd B. se Metal, A Longitudinal orle:-tation g g , w . g 3 ,
.e R .
1 0 $. . _-._ ___ _ . _ _ _ _ _ _ - - - - _ _ - - - - - ~ - _ _ - -
.l*
~:
1 25 - i f f f f E p 3 m R f f
.M . . . i , . . . . . -
r e n . e -
, e.
[" S
. . ra .
e w
. ?"
l ; e i e i i e e i i . . j ,,,.7 CATA St W EY Not available - - - - -
. - ner -
5F ( (35 e 5F
- .c _in (50 er- u ) ,
135 ft-lb L,-tfi (svc) __
.. . - R T,,:, - 5 5.F_( C_e s t es t ) .
,* s v e
,5 tv - -
.'.p - _
r r
~~ .; . _
~ ~
% Tram SA508. C1.2 Omstagarro, longitudinal v
~
Furtace9.43E + 17 nyt - g , g , ANJ-163
, , , . . , , , e , ,
!w -3 0 '.0 80 123 160 27) 243 283 370 %9 (t) itst Triptearm, F 1
l
, . 5-6 Babcock & Wilcox
1 l l Firare 5-2. Impact Data Fro: Irradiated Base !!etal. A 1rensverse Orie- ation )
- n . . . , , - - . . . , , .
.'r .
1 3, . - -. - - a 25 - e, e a a n s e e a e e e i
.T i . . . . . . . . i
-f . '5 : .
5 e
* .M -
p g s -
-_f____._-___-_...,..
/
+
; .7 -
~,
,, p, i e e e n , e i . .
W s . . . . s . . , , , CAta S W Arnt
- !;ot available
.' 'mos In (35 c, -26_F
- .t .ig (50 en uj -17F
(-b2 (avo 121 ft-Ib _
*
- 3- . gy nor -- -34F (Best est) -
,, ~~
/. . . .
~
s
.; Irr -
2 s'
.. y .
t e
~ er -
---/_-___.-___._________..___,.__.__
/
M - p
%rters SA508, cl.2
, Gesterars on Transverse FLutoct 9.43E + 17 avt
- NtAT bet. AAW-163 3
i > . . . . , , , ,
-M -O C 40 63 120 160 20 20 28) '329 50 vy) itst itweearw., F -
l :- 5- 7 Babcock 8. Wilcox 4
Figurc 5- 3. Inpact Lata Pro:a Irradiated hae Metal A. HAZ. I.ongitudinal Orie.tation
.. 7 . .
J
-5 4 t,. .. _ . . - . _ . _ - - . . - - - . . _ . _
3 J ;m . .
. a e a e a t i e n e e 1
.n 1 i p 4 3 1 4 s a u 3 l l
e i
+
e e - 1 o e _ .
. s. . . . . . . _ _ _ ._ .. .. __ . .. - . .. .. _ . ..
l
, .t1:n -
y, q s e a t e t t e v a
*T , . . . . . . . . . .
MIA Zvi.GY
- , s:,t -Not- ~ ~ - - -available- - - - - ~ -
~
I,., (I$ 'ti)
~
;.; 4
, , f 9 fi LS)
I M ft-Ib f s .. #s.5) a
.p gg - - l i n F (lies t. _es t )..
.. _ _ _ . . . g .
e..144 - ty - l 1.- - e
- 0 -
e a h s3 - l
.r. .
=a rt ei g SA508, C1.2 De s tsrafim IIAZ-Lon g i t - '
M - 9.f.3E + 17 nvt - Foutwt I Maar 4mte AAW-163 p e . .. , , , . , . ,
*:U *=3 0 *0 M 120 160 2% 2-1 .' s) 329 -' 0
'1 IEST IEmptearbet, F 5-8 Babcock s. Wilcox
Fis;ure 5 '. . Irpact Da t a Fro .
- r rad f.a ted ***e l d .% t a l .
Trans'.erge sirie-tation 3
- I a -
, _ , s
.. 'e .
e
?
3v . - - - - - - - - - . - - - - - - - - - - - - - - - - - . _ _ .
*I 25 -
p e a e n , , , , , , ,
* . e s : a a . . , a f
,rl .v: -
e I
.W "
- ~ ~ ~ - - - - - - - - - - - - _ _ _ _ ___ _,____3_______~
r Ja
.nY -
.e.
, r/(,
I I I f f f f f f e e P r, . i . . . . . . , , , CA?4 tt W 8Y , f(
- n:;r --Not available - - -
in 05 i4s 1 -- *'r g4 ?n '50si.te, 170F r,.'yl tavs) 51 ft-lb
. 16* - U ,:,, -110 F f i.es t est) e - -
5 5 120 - u
.7 IT -
E' 1 w go - i
- y . .
------.-------,--___s_ _________n______,__
f.0 - ,,
%rtang b* eld Metal 20 - Oestavation -
Fusteet 9.43E + 17 nyt* M41m m a kT-2n9-1A a f . . . . , , , , , ,
*D ~~3 0 ~'s to 120 1sa zn 2a m m un un l Trst Tanteanm, ;
5_9 Babcock & Wilcox
Figure >-5. 1 pact Data Fro:. Corre nion :4oniter .%terial.
'I rans eerse Orientatir.
IT i . . . . . . . . 73 J E ys0 __ _ _ _ _ _ _ _ ._ _ __. _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . 3 Jn - . c
.041 i . . . . . . . i 5
i two - -
- e 5
m
.30 -
T r e
~
2 .Tt> ~ i
- e
~. ~
I r ? , e s , 01
, e e t t l
2M , . . . . . , , , . .
*Aia
. si;WT.f ne T.ci - Not avai_lable - - - - .
Tg (35 u l ll6F
. y-
, 116F To (50 rt-ts) ,
I (-USE(avs) 103 ft-lb___ 14 - RT,, 56T (Best est) - i r l ( j 7, - - 1 l
' I l i 3 ;rr. - -
5 i y . . t., t e - ___ __ _ __ __ ______ ___________________________ j l se0
%rtes ac H5ST-PL-02 Destavarre. L ngitudinal
~
Fwtmet 9.l.3E + 17 nyt - Hast N a ta A-Il95-1 _
, , , , , , t , ,
, g i i ! -80 -40 0 '3 80 120 160 ly' 20 230 520 iG .% f tst itweentunt. 8 5-10 Babcock & Wilcox s. w
w S. NEITKON DOSIMETRY E
-6.1. Introduction.
A significant ~ aspect of the surveillance prograca is to provide a correlation between the neutron fluence above 1 W V and the radiation-induced property changes noted in the surveillance specinen. Io permit such a correlation. activation detectors with re etion thresholds in t he energy range of interest were placsd in each survelIlance capsule. The propertles of interest for the detectors are given in Table 6-1. Because of a long half-life (30 years) and an ef fectIve energy range of cure than 0.5 NY. only ths neasurecients of IU Cs product ion f rem fission react ions in Np aand "t*) are directly applicable to analytical determinations of the l'ast neutron (E
- 1 NV) fluence over Cycle 1. The other dosimeter reac-tfon. are usetut as corrobor.ating dat.: for shorter time intervals and/or high-er energy fluxes. Short-lived isotope activities are representative of reactor
. nditions over the lat ter portion or the irradiation period (fuel cycle)
.ntv. where.as reactions with a high threshold energy do not record a signifi-
, ant part of the total fast flux.
the energy-degcndent neutton flux is not directly available from activatten
~
detectors 5cacuse the precess provides the integrated effect of the neutron flux on the target material as a function of beth irradiation time and neutron eneigv. To obtain an accurate estimate of the average neutron flux incident u; on t he detector , several - paracwtera. mast he kn Nnt the operating history of the reactor t he energy response of the given detector, and the neutron spec-trun at the-detector locatlon. Of these paracieters. the defint' W of the neut ron spectr um is the cost difficult to obtain; essentially. c re eans are availablet iterative unfolding of experimental foil data and analytical esthoda. h ause of a lack of suffielent threshold foil detectors satisfying both the thre8 hold energy and half-life requirenents necessary for.a e-1 Babcock s.Wilcox
y - A 3urveillance program, iterative unfolding could n1t be used. This leaves the precitication of*the neutron spectrum to the analytical method. 6.J. An a Iv t I c a ! _ App r o.i,r,rz Ener.:v-Je;>endent neut ron t' luxes seen by the detector were determined by a dis-
< rete ordinate soliitton of the holtzmann transport ec uat ion . Specifically.
A';:SN '. a one .limensional code , and D)1*. .i two-di=ensional code, were used to calculate the t !iax at t he detector posit ion. In both codes, the Oeonee svstem has modeled radially from th'c core out to t he air gap outside the pressure ves-wel. De model included the core with a t irw-averaged radial power dist ribu-t ien, corr liner. barrel, t hernal shield, pressure vessel, and water agions. Including the internal components is neecssary to account for the distertions si the required energy spectrum by attenuation in these components. The ANISN code used the CASK' .'2-group neut ron cross sect ion set with an S; order of an-r.nlar qu.adrat t.re ana a p t expan .lon of t he scat tering mat rix. The problem was r u. aloag a r.ullu. across the core flats. Aetmuthat variations were obtained wi t h a DOT r-t heta cairulat ten t hat modeled a one-eighth plan view of the core it it nidplanc an.1 included a pin-hy-pin. t i ,e-averaeed power dis;ribution.
~
- >M cal.ulation a-ed i. qu edrature and a p, c r o s., section set derival f rom
' 9: .
F1.nes . a lcul ated wi t h t hi s Di'T rudel mus t he adjusted to account for 2ack of P c r. + - -er t ion de t a i l in calculat ions of anisot ropic scat terina, a pert ur-natian ..eieed b'v the presence of the capsule, and the axial power distribution. The t'rnt t wo i t er.s .i re both energy- and radial l ocat ion-de pendent . A p3/pg corte.t;,m tacter wa3 obtain. J hv conparing two ANISN one-dicension il rnlel cal n!at ions. In which only t he order of scat tering was varied. Se capsule "es t urb it ion t act or was obtained f rom a comparison of two DOT x-y rudel calcu-
- 4. t . ca ." . ene with a s.apsule explicitly nodeled - SS 304 cladding, alumi num
#111er region, and carbon steel specimens - and the other with wat er in t hose reciens. The ef f ect of axial power dist ribut ion was determined f rom burnup calculat iens as a funet ton of axial location for the outer rows of fuel as-ocablies. The net result fre= these parametric studies was a flux adjustment f actor K (Table 6-2) which should be applicable to the appropriate dosimeters
'in .all 177-fuel assembly surveillance programs in which the capsules are lo-cated at a radius of 211 cm fron the core center and 11 degrees from a maior axis.
62 Babcock s,Wilcox
The calculation described in Table 6-2 provide the neutron flux as a function ot e nergy at, ithe detector position. These calculated data are used in the following equatiens to obt ain the calculated activities us. d for comparisen with t he e xperitwntal values. -The basic equation for the activity D (in
. (:i /g) is as foIIcvs:
M - tT-t y) 0 = b.. I f 7: [ F (1 - e qt3)e \g i Ag 3. 7 10- i g n(E)4(E) p, j (6-1) where ' r . narraliting constant, ratio of measured to calculated flux.
. = Ava gad ro*
- numbe r.
Ag = atomic weight of target ctaterial 1 f = either weight fraction of target isoto;e in nth material g or fission yield of desired isotope. (E) =
' group-averated cross sections for material n Elisted in t
Table D-31 ist) = group-averaged fluxes calculated by DOT analysis, c ) = f ract ion of f ull power during jth tic:e interval, t y.
, = decay con-tant of (th material, t
g
=
interval-ef power history, T = sum of tetal i rradiat ion t icie, i.e.. rssidual time in reactor, and wait tic:e between reactor shutdown and counting.
= cumulative time f ren reactor startup to end of jth time 3
1 period. i.e., ; =
,1 t k'
a=1 1he norc:alizing censtant C can be obtained ley equating the right side of e ;uat ion +>-I to t he measured activity. With C speelfled the neutron fluence great er t han i MeV can be calculated from l l' MeV .si
; (!* 1.0 McV) =C [
E=: ((E) [Ft (6-2) j=1 where M is the number of irradiation time intervals; the other values are defined above.
- 6. 3. Results Calculated activities are ccepared to measurements of the dosimeters in Table 6 _3. The ID Cs data show at fast flux'(E - 1 MeV) is somewhat underpredicted 6-3 Babcock & Wilcox
_m._____
r' 010*) by the analytical model described herein (if one assumes that the cal- ~ culat ed flux spect rum is correct). Such agreement is probably within the un-certainty limits of this analysis. Hewever, for censervatism, a flux norm al-1:ation tactor of 1.1 is reco= mended for fast neutren calculat ions near the pre **ure vessel, nie . J ' Kit ac t ivi t ie s cecause ci a short half-life) indicate t h.i t the. core leakage flux over the latter part ei tne cycle was greater than the cycle averace (t he basis for t he . analytical model). An inspection of rel-itive powe r ili s t ritsu t i al in the core shwed that th i s. is true. The 5Mn ae-tivit y ind'eates an overpredit tion ( 201) of more than ' 11eV flux by the ana-Ivtical odel. *he significar.ce of tnis to the fast flux calculation 1a i s r,u re t b v t he tact that .spproximately 50", of the neutrons with E - 1 Ste's are in the t- to 1.)-McV range (Table 6 '.). nie 6"Co activity is the result of Sath of these er : cct s - overprediction of high ener.,y flux (high threshold en-s rga and high leakage flux over the la:ter part of the cycle (short half-IIte). The results of this analysis are consistent with a =revious 177-fuel assechly surv.Illance s pec i- en ana l y.a i s . 2 Future carsule data should add confidence to t he ancilyt ical ;>rocedure and pos8ibly clarify variaticwis in the normalization cew. tant. .Mased on a nor:Lilizat ion constant of 1.1. an average fast flux for Cycle 1 was alculated at the capsule locatson and at the inside surface of the pressure .ss*1 vall. The data (Tahle 6-5) were converted to fast fluence values of e . '. i - 10- n/cm- at the capsule center and 4.94 - ;017 n/cm2 at the pressure vs - -e t t... l l for Cycle I at the full power rating of 2568 L't. Values calcu- . ate I f or the pressure vessel wall (inside surface) refer to the maximum flux, i f i b n,av be totated at a different azimthal and asial position than the sur-veill.nhe capsule. In this analysis the maximum fluence at the pressure ves-sel occurred at an azimuth.il position of 6* frori e wijor axis (capsis te located it 11*) and about *>0 en above the lower active fuel line. 'Ihis is a function of power distribution in the core. The effect of extending the flux range devn te 0.1 PicV was to approximately JM!e tie fitmerece at the capsule and the pressure vessel. Since the same normalization factor is used, additional un-certainty is intr.xluced in this result because nme of the dosimeter reactions is ef fective over this entire energy range, y Babcocks.Wacox
Bsned on the surveillance sa:nple analysis for Cycle 1 and predicted core leakage and f uel burnups - for ' future cycles, pressure vessel fluence was pre-
-dicted up to 10 EFPY (effective full-power years). .These data are listed in Table 6-b and should be cure definitive than the generic design values in ref-erence 9.
Table 6-1. Surveillance Capsule Detectors Threshold 1sotope Detector re.setton energy. MeV half-life
'"Fe(n.p)1'Mn '2.5 303 days
'*Ni(n.p)~"Co >2.3 71.3 days
# ' "L'(n. f ) ! "Cs '1.1 30 years
~"Npfn.f)* WCs >0.5 30 years
* "U (n. f ) 1 ' 'Ru >1.1 39.5 days
' ' 'Np (n . ( ) ' 'Ru >0.5 39.5 days Tab h % .'. Flux Adjustment Factor I:ncrgy Axial power Capsule p fp r_.e n ge . .' teV factor 3 1 perturb'n K
-0.1 1.14 1.22 1.33 1.85
- 41. 5 1.14 1.23 1.26 1.77 1.0 1.14 1.23 1.19 1.67
'2.3 1.14 1.25 1. 03 1.43
'2.5 1.14 1.25 0.99 1.41 6 ,5 ' Bakock & Mcox A
r-- Tabl e' 6-3. Dosimer-r Activations A B ' ceasured . calculated C = A/B activity.(a) activity, no r=aliza tion
' Reaction' .Cl/g- pCE/g constant 5'Fe (n.p) 5'Ma ' 424 493 0.86
~
'"Ni(n.p)5#Co 766 847 0.90 216C(n f)l3'Cs
~
121 1.22 0.99 237Npfn,f)i37Cs 670 5.S1 1.15 2 i +t* (n. f ) ! '* 3Ru 339 38.0 1.02 2'75p(n.f) 0'Ru 183 161 1.14 (a) Average , of . f our dosimeter wi res f rom Table D-2. Table 6-4_ ,_'igreul t zed Fl ux Spectra. E > 1 MeV in water Energy range, near pressure 2350 McV- ve<sel wall fission 12.2-15.0 .0.0016 0.0002 10.0-12.2 0.0064 0.0013 S.18-10.0 0.018 0.0052 6.36-8.18 0.050 0.021 l: 4.96-6.36 0.092 0.051 4.06-4.96 0.078 0.052 3.01-4.06 0.118 0.159 2.46-3.01 0.122 0.132 2.35-2.46 0.039 .0.034 1.83-2.35 0.152 0.178
'l.11-l'83 . 0.278 0.323 1.0 -1.11 0.0457- 0.044
.1.00 1.000 6-6 Babcock & Wilcox M
1n.a
- y. -- ~ ' -
?
Table 6-5. Neut ra Fluence Fast flux. Cycle 1
, n /cm -s (440 EFPD) i
.. Fast fluence, E > 1 MeV Capsule center 2.43+10 9.43+17 Pressure vessel wall (r.ax) 1.39+10 5.28+17 Fluence. E > 0.1 MeV -
Capsule center 4.66+10 1.77+18 Pressure vessel wall (max) 2.67+10 1.02+18 Table 6-6. Predicted Fast Neutron Fluence in Pressure Vessel for 10 EFPY Location in pressure vessel Inside . Outside vall ! T/a 3/4T _ wall Avg fast flux. 1.6+10 9.2+9 2.2+9 9.4+8 n/cm2 -s Fast fluence, n/cm2 5.2+18 2.9+18 6.9+17 3.0+17 l 6-7 Babcock & Wilcox ~ E
Figure 6-1. Fast Neutron Fitence of Surveillance Capsule Center Compared to Varices Locations Through Reactor Vessel Wall for First ;) IFPY 10
- 9. 3 = 10l? nvt 6 -
N. S - 5 pe 5.2 e 10 ! " nvt 4-1 0_ e8
.c ~
f 3 y /, '-
-a .s
~
[ '. ._ C
=
&~
E .c
- s M ..* 2.9 1018 nyt q.N-2 -
go o CN II d 6.9 = 1017 nyt 3/4T '.co" " 3.0 x 1017 nvt D
- 4. e c.-',re 0 ! i I O 2 4 6 o 10 Ti e, IFPY y Babcock s. Wilcox
i
& e-
- 7. DISCUSSION 0. CAPSUI.E HISULTS
- 7. 1. Preirradiation Property Data A review of the unirradiated properties of the reactor vessel core belt re-gi.m indicated no significant deviation from expected properties except in the case of the upper shelf properties of the weld metal. Based on the pre-dieted end-of-service peak neutron fluence value at the 1/4T vessel wall lo-cation and the copper content of this weld, it is predicted that the end-of-service Charpy i.pper shelf energy (USE) will be below 50 f t-Ib. This veld was selected for inclusion in the surveillance program in accordance with the criteria in effect at the time the program was designed for Oconee 2. The a; plicable selection criterion was based on the uctirradiated properties only.
7.2. Irradiated Propertv Data 7.2.1. Tensile Properties Tabic 7-1 cospares irradiated and unirradiated te.sile properties. At both r.*on terparature and %0F, the ultimate and yield strengths increased slight-ly a< a result of irradiation accompanied by a corresponding decrease in duc-tilits. The small change in tensile properties is insignificant relative to the analy..is of the reactor vessel materials at this period in service life. 7.2.2. _Inyct_Prmr t ies lhe Schavior of t he Char;<y V-not ch impact data is snore significant to the cal-culatIcn of the reactor systeia's operating limitations. Table 7-2 compares the eb+erved changes in- irradiated Charpy impact properties with the predicted eban .:6 as shown in Figures 7-1 through 7-5. . The shif t of the data for the weld metal at the 50 f r-lb transition te=pera-ture was larger than the shift that would be predicted according to Regula-tory Guide 1.99. A similar coc:parison of the shift of the correlation cate-rial shows good agreement, while the base metal is significantly smaller. These coeparisons indicate that the estinating curves in Regulatory Guide 1.99 7-1 Babcock s. Wilcox
-_ l
I for low-copper naterials and at low fluence levels are reasonably accurate for predicting the 50 f t-lb transition temperature shif ts. The estimating curves fer high-copper material at low fluence levels are not in ge'd agreement with the observed data because of the abnormally large shif t that develops in low upper shelf naterial as the Charpy curve upper shelf approaches the 50 f t-lb design limit. The increase in the 35-mil lateral expansion transition tecperature is com-pared with the shift in RT ""#" * * '" * "#""*# "' "I * # *" SDT naJe for the 50 ft-lb transition temperature shift. These data show a behav-1 ter similar to that observed from the comparison of the observed and predicted transition data. Again, the significant dif ference is the larger shif t exhib-ited by the weld metal, which is due primarily to the fact that the Charpy curve i.p;'er shelf dropped close to the limit curve. 1herefore, the observed
-targe shift can be at t ributed to the method of mes auring the MLE shif t.
The d.sta f or. the decrease in Charpy CSE with irradiation showed a good compar-I
'isan fur the base met al, having a - lov ennp9r content. The weld metal ' data co pare very well with the predict 2 lue in view of the lack of data for nich-topper-content weldments at low fluence values use ' to develop the esti-na: inn e.rves.
The shi#ts -hown are not in complete agreement with those predicted from Reg-ala t ory Gui.'e 1.99 at the fluence level of this capsule. This indicates that t!e est i ating curves base greater inaccuracies at the very low neutron fluence leve2s (_1 - 103d n/cm?). This inaccuracy is a result of the limited data at the Ie fluence values and of the fact that the cajority of the data used to
.h i f w the curves in Regulatory Guide 1.99 are based on the shift at 30 ft-lb l n ..arAred to the current acquirement of 50 ft-Ib. For most materials the ihiir9 =casured at 50 f t-lb/35 MLE are expected so be higher than those mea-dured at 30 tt-lb. The significan:e of the shif ts at 50 f t-lb and/or 35 MLE is not well understood at present, especially for materials having USEs that approach the 50 f t-lb level and/or the 35 MLE level. Materials with this l
characteristic should be evaluated at transition energy levels lower than 50 [ ft-lb. The design curves for- predicting the shif t at 50 f t-Ib/35 MLE will probably be
-nodified as data become available; until that ti=e, the design curves for pre-dieting the RT s f a given in Regulatory Guide 1.99 are considered NDT 7-2 Babcock & Wilcox
adequate for predicting the RTNDT * ** "" *# ** '" * * ##* not available and will continue to be used to establish the pressure-te=pera-ture operational limitations for the irradiated portions of the reactor vessel. Table 7-1. Comparison of Tensile Tes' suits Room temp test 580F test Unirr Irrad thirr Irrad Base Metal - AAW-161. Longitudinal Fluence, 10U n/cm3 (> 1 MeV) 0 9.43 0 9.43 Cit. tensile strength, ksi 89.2 91.8 83.3 90.3 0.2% yield strength, ksi 67.9 70.4 61.5 69.4 Elongation, Z 28.1 27.1 29.7 28.3 RA, % 69.7 70.1 73.1 71.5 f' g Metal - WF-209-1 A Fluence, 10D n/cm- (s 1 NV) 0 9.43 0 9.43 l'I t . tensile strength, kal 95.2 104.e 29 7 100.6 0.2% yield strength, ksi 81.4 88.9 69.8 89.5 Elongatlon, 2 25.6 23.3 20.6 20.2 K'. . Z 57.9 58.7 48.9 44.1 1 /: / 7-3 Babcock s.Wilcox
~
3 Table 7-2. Observed Vs Predicted Changes in Irradiated Charpy Impact Properties Material Observed Predicted Increase in 50-ft-lb trans temp, F Base material (AAW-163) Longitudinal 7 21 Transverse (b) 21 Base material-(HAZ), longitudinal (b) , 21 Weld metal (kT-209-1 A) 120 100 Correlation material (A-1195-1) 42 48 Increase in 35-MI.E trans temp, F Base neterial (AAW-163) Longitudinal 9 21(c) Transverse (b) - 21(c) Base material (HAZ), longitudinal (b) 21 IC) Weld metal (k'F-209-1A) 12S 100(C) Correlation material (A-1195-1). 55 48 IC) Decrease in Charpy l'SE, f t-lb Ilase material (A.W-163) Longitudinal 20 14 Transverse 12 12 hase material .(P.AZ), ' longitudinal 4 14 Weld tretal (WF-209-1A) 14 16 n rrelation material (A-1195-1) 22 16 ( These values predicted per Regulatory Guide 1.99, Revision 1. Limited and/or large scatter in data prevented a valid comparison. Based on the assumption that MLE as well as 50 fr-lb transition-temperature is used to control the shift in RT NDT* 7-4 Babcoce s.Wilcox ___m---_ _ _ _ - - _ - - - -
Figure 7-1. Irradiated Vs Unirradiated Charpy Impact Proper-ties of Base .' fetal, Loncitudinal Orientation liC , I
** 15 -
9.43 + 17 nvt - j2 .__.._.._____.______________.____________.. y 25 - Unirradiated n . . . . E . . . . . . . . - Unirradiated g % i .m . r
.~ .
- 9.4 3 + 17 nvt
- s. . y,n .
r .. . - _ _ .. i' i e I e - ' g . '220 - ?T = 9F l i
,, r e e e v # i e . .
1 1
~'O . . . . . ,
'V -
N ' t .*.USE = 20 f t-lb - a 14'. - b- ) [ 1:1 - Unirradiated % F, 3:Y - i 1 5
- 7 .r - T - 9.43 + 17 nvt Y,
t
~
AT = 7F E-tl -
%rtesg _ SA508. C1.2 c~.
thaturarros Longitudinal Fustnes See above - e l . . . . . Heat Nwera _ AN4-163
-m -40 0 40 80 123 160 2'M 20 itst Trmaarvet, F 280 520 %Q vn 7_3 Babcock & Wilcox 1
figure 7-2. Irradiated 'Is Unirradiated Charpy := pac t Proper-ties of Base &tal, ~rannerse Orientatio,
.v . . . . g l
n . Unirradiated - 9.43 + 17 nyt 2 c; . - - - . - - - . - - . - - . - - - - - - - - - - - - - - - - - .
.T 2s - -
.2 . . . . . . . . .
i .rs; . L Uni-radiated . 7 I w 9.13 + IT nyt
~
I .. . y- .
, ..--..-p . . - - - - - - . - - - - - - . - - - - - - . . . .
4 / i
.E -
~ l
.,,1 i , i , , , . . , . .
- ' C , . . . . . . . . .
;g- . .
l
.M - -
.g . AUSE = 12 f t-lb _
\ j.
~ "'
, * * ~~T- '
l Unirradiat-d i I J IT - -
. 9.?3 + 17 nyt 7
1 e ' t
- g . -
I i g~ . / .
/ %rtss at SA5C8, cl.2 omisuraf t, Transverse l M "
Ft.mtzca See above - l l Ntaikees _ANJ-163 Yao .e o e a) us 163 rn 24 m m m ,m Tw Tsmnearm, s 7- 6 Babcock & Wilcox b _ __ _ , .--_,y
rir,ure 7-3. Irradiated V:4 t'nirradiated Char; v rpu t Pr:pe r t ie s O f Sr e Nt al . IIAZ
'Y
. . i ,
i 1
. x . 9.4 3 + 17 nyt 2 Vnitradiated 4 ;-
a w a
,Y 5 -
. . . . . . . i .
I
/ s ,. Unter44 tat.,8 . <
be
- w. 3 + 17 nvt *
. 2- .
i 1 ..T- ? e 4
- .T r
- g. i , , * , , . . . , ,
l . n- . s- . hi - l ust-4re-i> , l 9
. 1.' -
i
.a T
Unttradiated c ;~v . t dW
~
9.41 + 17 nyt - 41 l 1 i w K .
.. 1 .t. .
t-7 g --
- li 40 - .
%ttaint M W . C1.2 M- Os:tutars:n h 2 '.eng11 m Fwtmet See above -
l . g , w, .W -163 g ' '
-53 -O C W dc 123 :50 .m 20 28) R: %3 .v71 Itti ItwteatW, i
- 7. ; Babcock 8. Wilcox
i risure 7 ..
- r ra,11.ited Ys l'ni r radi.ited Garpv :. pact Preper-l t ie s 0: We :d "e t.it . Tran4ver.<e orie n t at ion 1.Y1 , , , ,
Unir aJiated .
" '5
- 9.43 + 17 nyt -
?
y rg 3 . Y o N . - n . , , . . . . . . . .
.M , . 1 e ~, . . 1 .
s .%- . . 5 7-Unirradi.ited i
.i j On .0e( -
9---.. .._ . - _._ j !: .
; .'. T
- 128F g , .y .
a . l . 1 i r c.43 + 17 nvt l
, f ' t i t f B t t t 4 a + 4 I s i a a
;r .
f N - 1
,, t w' - .
r O i .* l T ! y
.i 11 -
F
- P -
"US E = 14 f t-lb-
- Unirradiated K ~
l_ = 120F _ k
~
! --- -- ,-------c-------=.---.... l M -
%rteig Veld Metal __. .-
- 4.4 3 + 17 nyt % g,,,, , , -
20 . . Tietoca Se'* above g erar sente k'F-209-1A i e i , i , , , e i i i 80 40 0 % ?Q 120 160 2% .% 3 25) !!" :' " '
.* 9 Ttst itw eearver. F
_3 Babcock a. Wilcox : I
r i ra ri- T- 3. I r ra di a:e.: W l'nirradiated 1 pact Properties. Of Corre-1.1: 1. n ".'- f t .' r Ma t e r i a l , I on g i t ud ir.a l s'r ie n t a t i en ! ./ . . . . . . 1 i l . .g . Unitradiated f if _ __ l
$x .
9.43 + 17 nyt l Y' , . . . Unitradiated -~ r .., .
- u. .
9.43 + 17 nyt 1
. . ._.___e. . _._ . _ _ .._
d k_ ar - ssr E i 1 1 1 i i t i f n e i e i a a u a s a l aOSE
- 22 f t-1b .
Entaradiated M I 4 f A . * . s -
. 9.4 3 + 17 nut -
t A - * "- aT = 42F - l l 1 ( , *.
%ttes s P.55T-PL-02 l
l
Onitavattaa L-- e t rud N 1
~
Fustect see ik've
,
- 1 - e i . . .
Neat hwwe A '195-1__ FC 40 ) O A0 Z 21) N W 160 143 280 's! Test itweesta. F 79 Babcock & Wilcox : i
. - - - , - - ,m , -
- 6. DETERMINATION OF RCPB PRESSI;RE-TEMPERATt RE 1.1MiTS
%e pre %<ure-temperature limits of the reactor coolant pressure boundary (RCPB) of th en.* J hive been established in accordance with the requirements of 10 CFR 50 Appvndix G.
The ectbod-i and criteria employed to establish operating pres-sure and te perature limits are described in topical report BAW-10046.~ Re objectIse of these limits is to prevent nonductile failure during any normal operatinz conditlen, including anticipated operational occurrences and system hydrostatic tests. We loading conditions of interest include the following: t
- 1. Soreal operations, including heatup and coaldown.
- 2. Inservice leak and hydrostatic tests.
- 1. Ikaetor core operation.
Ihe m.nior cenponents of the RCPB have been analyzed in accordance with 10 CFR
.0 , \ppendix i;. TM closure head region, the reactor vessel outlet nozzle, anit t ue be lt line . region have been identi fied as the only regions of the reac-t or ve*.e l , aa.1 consequen t ly of the RCPB, that regulate the pressure-tempera-ture limits. Stace the clo.ure head region is significantly stressed at rel-atIvets 1.% t c:trera t ures (due to mechanical loads resulting from bolt preload).
this region largely coe.trols the pressure-temperature limits of the first sev- 1 eral serviec periods. The reactor vessel outlet nozzle also affects the pres- I iu re- t er,v ra t ure limit curves of the first several s;-vice periods. His is due to t he h! :S lxal st rc<ses at the inside corner of the nozzle, which can be two to three times the membrane stresses of the shell. Af ter the first several i vear- of neutron radiation exposure, the RT # "" # ""* NDT ' als will be high enough that the beltline region of the reactor vessel will l start to control the pressure-temperature limits of the RCPB. For the service period t or which the limit curves are established, the maximum allowable pres- ) sure as a funct ion of fluid temperature is obtained through a point-by-point comparison of the licits imposed by the closure head region, the outlet noz-ale, and the belt line region. The tsaximum allowable pressure is taken to be the lowest of the three calculated pressures. g,1 Babcock a. Wilcox
l 1 l Tho eighth full-power year was selected because it is estimated that the sec-ond surveillance capsule will be withdrawn at the end of the refuelina cycle, which corre'sponds to approximately eight full-power years. ne time di f fer-ance between the withdrawal of the first and second surveillance capsules pro-vides adequate tic:e for. re-establishing the operating pressure and teeperat ure limits for the period of operation between the second and third surveillance capsule withdrawals.. Tha limit curves for Oconee 2 are based on the predicted values of the ad-justed reterence temperatures of all the belt line regia naterials at the end of the eighth f ull-power year. De unirradiated inpact properties were deter-nined for the surveillance beltline region materials in accordance with 10 CFR 50, Appendixes G and H. For the other beltline region and RCP8 naterials, thz tmirradiated impact properties were estinated using the methods described in MW-100*.6P.* The unitradiated impact properties and residual element s of tho beltline region materials are listed in Table A-1. The adjusted reference temperatures are calculated by adding the predicted radiation-induced .*.RT NDT
.ud the uni rradiated RTNDT- e Predicted ARTg is ca!culated using the re-
.pec tive neutron fluence and copper and phosphorus contents. Ee design cur.e. of Regulatory Guide 1.99* were used to predict the radiation-induced DT values as a ftmeti n
- 2. P.T the material's copper and phosphorus content
.c-t neutron fluence. Figure 8-1 illustrates the calculated peak acutron flu-e: ce at several locations through the reactor vessel beltline region wall as f wct ion of exposure time. The support ing information for Figure 8-1 is de-ribed in MW-10100. ' The neutron fluence values of Figure 8-1 are the pre-dict-d f luences, which have been demonstrated (sectica 6) to be conservative.
P.c rc:utron fluences and adjusted RTNDT "" ** * *I" "8 " "# * * # I ~ als at the end of the eighth f ull-power year are listed in Table 8-1. The neutron f1wnees and adjusted RT valms am g m e t. e 1 and W NDT vessel wall locations (T = wall thickness). he assumed RTSDT **I""" hsad region and the outlet nozzle steel forgings is 60F. in accordance with BL*-10046P.* Revision 1. January 1976. I 8-2 Babcocks.Wdcox
Figure 8-2 shcws the reactor vessel's pressure-te:nperature limit curves for normal heatup. This figure also shows the cere criticality limits as required by 10 CTR 50, Appendix C. Figures 8-3 amd 3-1. shcw the vessel's pressure-temperature limit curve for nort:ul coolds a and for heatup during inservice leak and hydrostatic tests, respectively. All pressure-temperature limit curves are applicible up to the ninth effective full-power year. Protection against nonduct ile failure is ensured by maintaining the coolant pressure be-low the upper limit:4 of the pressure-temperature limit curves. The accept-able pressure and temperature combinations for reactor vessel operation are below and to the right of the limit curve. The reactor is not permitted to c* critical unt il the pre sure-temperature combisations are to the right of the criticality linit curve. To establish the pressure-temperature limits for prot ect ion ac.aln se nondwtile f ailure of the ItCP3, t he limits presented in Figures M-2 through 8 *. ::ust be adjusted by the pressure dif ferential between t he poin t of % < .t em pressure measurerent and the pressure on the reactor ves-set controlling the li:si t curves. This is r:ecessary because the reactor ves-el is the mo s t limitine component of the RCP3. J 8-3 Babcock a Wilcox
3 s*. ,ml. ~ P s e. e 4 4- = n *- ,
.a. s ,4
= ., a t
-=E. e. ,
6 .a w 4 *w e
% e.
. P- -. e.
4 .. r a., e. . b - 4 w e F 4*
.M.
C r ..
- e. s >
6.
.c =x. i. *m
- e c +
6 m,r=..e.~.,..
. 1 - - .. .,a 4 e e t
...n
- 7. it =
y . a e>s
*- e . . .
e, e y
, c 2 e. .m. e. e. c. -
u 5
) .s. .
e e e e e w e a ,-
=a y r.s ,f r e e e e e.
.4 . . -* . . . .., .
] 2 -s - y n g.
6
. .s. s, ., =... ,,.= .
l
~ -
- 2 : A 5 ga : * - o+ 9 -e
-6, N . ::N.-:'d
$. " .- _5 d
- s. c.
u.* == s C. 7. > , u .
- u n6 - a
. g o a v p ye., -t y: *t. ;. . 2.: =. : -- .
e : o c u .~' r V ~.
- 2 O
-l ~ ;
I * : - . : 6 .
.r. -
.;. w 2 A *. -:. ~*.-
"d. J i
I
-; ;. Uk ~- * * ,
- ~.'. ! : a .s s
=.r 2., 1 ;. - - ~
_s
- .. b s
G-b . 76
- k. , e .*
- - -6 , e . . . . .
- c. ; ; * , *- ' *
- e e
% ~, .-
w * ' L L. -y'
- s. t s 3,a.
* ~I.
- e . . .* e -4
- a: . - -
,- - s. t _. ,
-.=,
. F. = g T
i , e
. ..e., , 9
, *;. =,
e
.g s.
e, . . : ..
.- ,t, . .
, i a . + n, + - a.
~
.s -. -
u ,- - N N N D. U 4J u - e,.l c,,
,J'. %. .c. o o ,
_, l' : m e m a a s o . - .
- L,. ~
~ . -. .,,
, . N.
t 2 - = . . . I e' .c h a m I % 35 4 3 3 J . 1 E 34 Babcock 8. Wilcox
,~
e j -
- { I i 2 g >
.- -? > C 4 = =
x
- ~ o ~*
m- + . + 5 ~
.= , .
- . n c
? - . ,
_.'. ~ T
=
a n
=
o t:
.,=.
?. : -
_' i-J (' ~~ 7
%- 2~
c'
- - r o- .s a x
x
- ~-
7 _ -n
.1 . . - -
~
e l l ~ I I I I l
- o. c I ! 1 I I I l ' I
- e. no. o. o. e. c. -
.c 4 e e e e .
O. o. N. r. 4 -r -r - - e e % e. o.
- ~ c o o gua/u *(A.y g . 3) gru 4 01 = asum13 tonn.es
-5 Babcock s. Wilcox 1
l 4
Fleure ri .' . ". .
? . . r '. . . c ! Pi . s s oi e- ! cry..-e s t u re I.f r:I t Cur ves for %,rs:ul Oper.it ion -
ib .i t . a.. . ' p;> 1 16 ..h ! c ter first six 1:f tect ive Feil l-P..wc r Ye.t r s 2400-Assumed NT
.. .. ...... NDT, F D 2200 -
. C Beltline region 1/41 112 Belt 1ine region 1/4T 56 2000 -
Closure head region 64 Outlet nozzle 60 y 1800 - Prce** re. Tee:p , a, Po i n.t. . io.1. F [ 1600 - A 450 60 g B 625 146 Appl ic.-ihl e for g C 625 273 lie.it up R.itcu g 1400 - D 2250 J02 up to 100F/h
, E 625 271 ..
, g F 625 313 8
-< 1200 - G 2250 342 8
u en g 1000 - g 800 - w Critic.s!- Ity I.imit u.* !! s O M 600 - l. F. A 400 - - m n .. . cr' .b le e a..r. -' eape r.e ui.- . .>d s i. i s..n. ar b 1. cu . .: e ., o.c r 1,c.: .., o= t s a e : . rve c i. n.< I a mt . .. r . .s. . , tr .e i a .: o ,i.e.....r,.ise#,r.. es.: br e .. c., e s., e sne ..: Q gon _ . o i .* r . . .... . ... . e . I e i... r < ... .. ..n e i e ,. , e .., n v.. .l era..n c.4ir..trin,ei;, ts.it . ncv. . ..e .a, . i :: 6. .: se n .... or ... set, e... p....ibi, in tr e nt error. P t i t i I I I f n 0 f,0 120 160 200 240 280 !!O ~1f.0 o x Reactor Vesnel Coolant Temperature. F i i
l'i v. ire 8- 1. L f:. .w1 t8.eemr 'ie, 10,11 w e l !* re.w o re-l er:pe ra t u re 1. i ni t Curve for Norr 1 Oper.ition
! . t o r l'i r s t si x E t' t'eet i ve l'ul l-Powe r Yea rs 2400 A asiu.! rig. I' llon -
E heltline re g t.in 1/ *. I 112 Beltline region 1/41 56 2000 Closuri he.i-! region, 60 - that let nan t e. 60 3 IM00 - I're%ure. Temp. I.O I HI . P.I ._ .F,_ ,, [ 16081 - A 250 70
$ 15 625 119
" Applicable for C 625 205 U 1400 -
D Cooltfown Rates 1120 213 up to 100F/le e E 2250 281 z e
.', 1200 -
j y D t 1000 - e e t g 800 - u o I B
- 600 - g.'
400 - .. . . . . . . .
,. . .. .i ... . ..w. ,, , e . . ,, e : .
W ...,.it. ..e. ,- .. i .. . r. . e u e...t . t 1 . I. t sa
... .. t . r . o . i ~ i i . .. . . .i..
t... i e .- .tif t e e. nt t I t t w.es t he i=,4 t .. t n i. - i e. .... . . .. w e te i.e.... t.- e . . ..e o 200-- A .. . . . i . ,. . ...t e .. I t oi e s., i t . u . . . . . . . . , e t t e t 8. . I n n .. + -.i.-. t -, e.. . o rr in -r e.. x- ee eer. i .
- V i
g O f
~~m f f f f I n
- 40 H0 120 160 200 o 240 280 120 M
Itcactor Vcusel Coolant Temperature. F
'4 Le w') . A
~* .
L- = 4m w g ye
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- 9. ' SUMyaRY OF RESULTS L
The analvsis of the reactor vessel material contained in the first surveillance capsule -(OCll-C) removed f rom the Oconee 2 pressure vessel led to the following conclusions:
- 1. 1he 'capsulp received an average fast fluence'of 9.43 = 1017 n/cm (E ' > l '
SkV) . The predicted fast fluence for the reactor vessel 1/4T location at the end of the first fuel cycle is 3.54 = 10I7 (E > 1 MeV). 2.- The fast fluence of 2. 32 < 10 l' c/cm2 (E > 1 & V) increased the RTg of the pressure vessel core region shell caterials to a maximum of 132F at t he eighth EFPY. 3.
. Based on a ratio of 1.6 between the fas t flux at the surveillance capsule loc... lon to t ha t at the vessel vall and an 80% load factor, the projected
' t'a s t f li,sence t ha t the Oconee 2 reactor pressure vessel will receive in 40 calendar j. cars' operation is 1.06 e 1013 n/cm2 (g 1 3ry),
E. The increase in RTNDT f r the base plate material was less than that pre-dicted by the current ly used design curves of J.RT versus fluence because 5DT of tuaccuracles in the prediction curves r esulting from lack of irradiation data for liu-copper. raterials at low fluences.
- i. The increase in the RT2T f r the weld retal was greater than that predict-
-cd by the currently acted design curves of ART DI because f the inaccura-cies in the nethod of measuring the shif t in t ransition te=perature as the
. data curves approach the specified limit curves.
- 6. The current techniq ues used for predict ing the . chance in Charpy icpact i
upper shelf properties due to irradiation are conservative.
- 7. _ The analysis of the neut ron dosineters demonstrated that the analytical
. techniques used to predict the neutron flux and fluence were accurate.
8.
- The thermal' nonitors . indicated that the caps'ile design was satisfactory for nalntaining the specinens 'within the desired te=perature r..nge.
9-1 Babcock 3. Wilcox w t
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SL2VE!LIX;CE CAPSULE RE. THAL SCHEDC.E Based'on'the postirradiation test results of capsulo OCII-0, the following schedule is recomended for examination of the remaining capsules in the econee 2 reacter vessel sur*;eillance program: Evaluation schedule Est capsule Est date(3) , Capsule fluence, Estimated EFPY data
-ID n/cs- Surface 1/4T availab le OClI-A 3.1 ~ < 19I6 7 8 1980 OCII-n 1.2 10W 18 - 32 1937 OCIl-E '
2.2 a LOD 33 59 19'i7 tC I I-D. Standby. -- -- -
' OCIl-F Standby -- -- -
(a)These' dates do not represen t. the earliest dates that data will be available for the materials that control t he ope ra t i n g l i=.I t a t i ons . Similar material + care in-cleided as part of the HW Integrated Reactor Surveil- ' lance Progran, which wil1 provide necessary fata on .: timely basis. Tr,e earlies t data that these d.ata will' he available is 1980. i b) Capsules contain veld metal speci:wns. f c 3o_3 Babcock s. Wilcox
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CERTIFICATION I The speci::: ens were tested, and the data obtairied from Oconee Nuclear Station, l' nit 2-burveillance capsule OCII-C were evaluated using accepted techniques and established standard methods and procedures in accordance with the re-t quirements.of 10 CFR 50 Appendixes G and }{.
/b
-me EE. !U/evMW A. I. . f.ove, Jr.;7P.E. - ~Date Paoject Technical Manager
- This. report has been reviewed for technical content and accuracy.
K. E. Moore hh2 // <. ; Q-Dye Technical Staff
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- 12. REFERE';CES
' G . ~ .1 Snyder and C. S. - Carter, Reactor Vessel Material Surveillance Fro-gram, BAW-10066Ax._Rev. 3. B.ocock & Wilcox, I.ynchburg, Virginia, January 1975.
- A. l. . Love, Jr., et .
al ._, An tlysis of Capsule OCl-E Frm Ibke Power Company Oconee Unit 1 Reactor Vessel }iaterials Surveillance Program BAW-1436, ILabcock & Wilcox, Lynchburg, Virginia (to be published). I rser's :Lanual for ANISN, a One-31mensional Discrete Ordinates Transp ort Code With Anisot ropic Scat tering, K-lei 9 3 (RSIC-CCC-32), (bion Carbide Corp. , Nuclear Ilivision. Stirch 1967.
rser's Manu.n l for the DdT-IIW Mscrete ordinates Transport Computer Code, E'sM -U!E-1982, December 1969.
CASK Group Coupled Neutren and Gamma-Ray Cross Section D.ita, RS R DLC-2 3 Radiat ion Shielding La:or:r.ation Center. 6 Draft New standard E482-00. Acco=:wnded Practice for Neutron Dosimetry fer Reactor Pressure Vessel Sarwillance," October 10, 1974, W. L. Zi jp, Review of Activatico '1ethMs for the De:ernination of Fast _ Neu-
. tron Spectra, Reactor Cent ru:n Nederland, Petten, ?tir 1965.
H. S. P.alme and 11. W. Behnke, 'kthods of Compliance With Fracture Toughness and Operat lonal Requirements of A;pendix G to 10 CFR 50, SAW-10046P, Bab-cock & Wilcox I.ynchburg, Virginia, October 1975. 9 H. '. Palme, C. . S. . Carter, and C. I.. Whitmarsh , Reacter Vessel Material Sur-veillance Program - Cocipliance With 10 CFR 50, Appendix 3 for Oconee-Class Reactors, BAW-10100A, Babcock & Wilcox. Lynchburg, Virginia, February 1975. 1 0 . 11 . S . Palme, Radiation Embrittlenest Sensitivity of Reactor Pressure Vessel Steel, 8AW-10056A Babcock & Wilcos. Lynchburg, Virginia August 1973. 12-1 Babcock & Wilcox
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I I' ) i-APPENDIX A Reac ter Vessel Surveillance Progran - Back. ground Data and Informat ion a [ 1 1 A-1 Babcock & Wilces Mh hio.gg e. A - . - , -- -.n-
1 Mate rial Select ion ' D.sta The data used to a. elect the caterials for the specimens in the surveillance progr.im,' in accordance .with' E-185-66, are shown in Table A-1. The locations of these :2aterials within the reactor vessel are shown in Figure A-1.
- 2. Defin_iticn of HeIt1ine Region The beltline region of Oconee Nuclear Station, Unit 2 was defined in accord-ance with the data given in BAW-10100A.3 l_. Cafnule_Identificatis rae capsules used in the Geonee 2 surveillance program are identified be'0w by identification number, type, and location.
Capsule Cross Ref erence Data ID No. Tvr a Location OCll-A A Upper OCII-B B Lower LYll-C A Upper OCII-D B Lower LYll-E A. Upper OClI-F B Lower 1._ffee iment per Surve_illance Capsule Ss e Table 4 a-2 and A-3. I l A-2 Babcock & Wilcox
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i- . Table A-2'. ' Materials and Specimens'in Upper Surveillance Capsules OCII-A, i OCII-C, and OCII-E No. of specimens-
-Material description Tensile Charpv Weld metal, WF-209-1A 4 8 HAZ.A. heat'AAW-163, longitudinalI " .O 8 Baseline naterial, plate A, heat AAW-163 Longitudinal. 4 8 Transverse O 4 Correlation, il5ST plate 02 0 8 Total per capsule 8 36 Table A-3. Materials and Spectuens in Lower Surveillance Capsules OC11-B, OCII-D, and OCII-F No. of specimens Material description Tensile Charpv
!!AZ B heat AWC-164, lonsti t udin- 4 10 al Baseline material, plate B hett AWG-164 Longit ud i na l 4 10 i Transverse 0 8 Correlation, ilSST plate 02 0 8
. Total per capsule 8 36 i l
l 34 - Babcock s.Wilcox
e 6 FIRure A-1. Location and Identification of Materials Used in Fabrication of Reactor Pressure Vessel 4 W ( L' ( /~ A!!X 77 (1.ower Nozzle Belt)
. $+ WF-15'. I f N4 163 (Upper Shell) g.-- W F- 2 3 AWG lb*. (lower SheII) 4*-- WF-112 122T293 val (Dutchman)
I a.s Babcock a Wilcox
APPENDIX B Preitradiation Tensile Data l B-1 Babcock & Wilcox
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Table B-1. Preirradiation Tensile Properties of Shell Plate Material, Heat AA*J 163 Test Red'n S pec imen t rength , psi Elongation. 2 temp, of area, No. F Yield' Ult. Uni f . Total longitudinal. EE-701 IUr 70,420 90,680 10.74 27.14 706 69.3 RT 66,790 88,550 11.83 29.28 69.5 709 RT 66,690 88,300 10.88 27.86 71.2 Mean RT 67,970 89,180 11.15 28.09 69.67
-Std dev'n 1,740 300 0.48 0.89 1.13 EE-704 580 61,550. 82.110 9.48 30.7 707. 74.8 380 62,460 S2,110 9.59 30.0 75.7 718 580 60,490 85,630 28.6 9.52 68.8 Mean 580 63,510 83,280 9.53 29.77 73.1 Std dev'n '5 810 1,660 0.05 0.87 3.06 Transverse EE-602 RT 63,010 89,010 10.81 26.79 67.4 604 RT 69,580 90,960 10.81 27.14 66.2 605 RT 65,480 89,820 11.15 25.71 62.3 Mean RT hs,690 89,600 10.92 26.55 65.3 Std dev'n 660 410 0.16 0.61 2.18 EE-601 580 64,480 86,980 10.24 25.7 64.6 60's 580 61,610 85,650 11.15 hob 580 6il,750 25.7 64.6 34,850 10.58 25.7 65.6 Me..n 580 62,280 85,830 10.66 25.7 64.93 Std . lev'n :5 1,590 880 0.38 0 0.47 i
i. B-2 babcock 8. WilC0x >
Table B-2. : Preirradiation Tensile Properties of Shell Plate Material - EAZ, Heat Ar.' 163 ___ Specimen temp, St renRt h, psi Elongation. * " " 9f No. _, F Yield Ult. L'n i f . Total ;
. Longitudinal' EE-401 RT 71,860 92,500 '10.69 403- 23.2 64.6 RT 67,950 90,060 7.5 21.4 )
404 RT 65.1 ) 72.140 92.560 9.9 21.1 '66.1 Mean RT 70,650 91,710 9.36 21.9 65.27 Std dev'n- 1,910 1,170 1.36 0.93 0.62 EE-402 580 67,630 88,510 405 8.09 20.0' 64.4 580 66,750 87,670 7.95 406- 18.6 60.4 580 64,260 66,350 8.05 20.0 65.2 Mean 580 66,210 87,510 Std dev'n 8.03 19.53 63.3 5 .l.430 - 890 0.06 0.66 2.10 Transverse EE-lul RT 69,140 90,180 304 11.79 28.0 71.0 RT 67,270 88.450 11.47 28.2 70.0 { IO's RT 68,380 89,630 11.88 27.9 71.1 Mean RT 68,260 89,420 11.71 28.33 70.7. Std dev'n 770 720 0.18 0.42 0.50 E E- 101 580 59,610 85.150 11.83 27.1 71.5 102 580 61,970 86.300 106 11.83 29.3 73.0 580 63,000 87,110 11.10 26.4 68.5 Mean 580 61,530 86,190 11.59 27.6 71.0 Std dev'n :5 1,420 800 0.34 1.24 1.87 B- Babcock & WilCOX k- _.t,. .. .. . -. .m: _ _ - - - - - - - - - ~ ' - - - - ^' ^
- Table B-3. Preirradia-ion Tensile Properties of Shell Plate Material, Heat AkG 164 m
Test Strengt.n, Psi Elongation. Red'n Specimen teep. g, No. F Yield Ult. 1*n i f . Total %
!.on .;1 t ud in a l FF-703 RT 69,810 90,260 10.88 28.57 71.7 712' RT 68,930 89,500 8.67 26.07 71.1 713' RT 69,680 69,910 9.12 28.57 72.43 Me x. RT 69,470 89,890 9.56 27.74 71.74 Std dev'n 390- 310 0.91 1.18 0.54 FF-702 580 64,220 86,720 10.04 25.0 67.3 704 580 60,830 83,610 10.65 26.4 68.7 708 SSO 61,830 85,270 10.23 25.7 71.1
. Me a 580 62,290 85,200 10.31 25.7 69.03 Std dev'n t5 1,420 1,270 0.25 0.57 L.57
.T_rpu sve rse F F- >n t RT 48,230 88,360 10.82 27.14 69.9
+102 RT 66,790 87,540 10.0 27.86 69.2
~*6 RT d6,270 87,360 11.41 26.43 68.1 "s - i , RT 67,100 87,750 10.74 27.14 69.07 Std Jev'n 630 435 0.58 0.58 0.55 FF 2.03 5SO 90,000 82,840 10.52 26.4 69.3 d "4 580 ol.940 84,080 10.32 26.4 67.5
"#16~ 5M 59,190 82,610 11. 38 28.6 70.4
".... 580 60,390 63,180 10.74 27.13 69.23 A; uv'a :5- 1,150 650 0.46 1.04 1,39 I
l i B-4 Babcock 8. WilCOX l I i l l 1
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h Table B-4. Preirradiatica Tensile Properties of Shell Plate Staterial - HAZ, Heat AWG 164 Specimen tcmp, E'*"EE *P "E*E' "' No. F of area. Yleid Ult. t*n i f . Total *;
!.on gi t ud inal FF-402 RT- 58,650 88,320 8.1 20.0 68.16 408 RT 66,970 88,300 418 7.0 19.6. 68.5 RT 67,410 88,810 6.1 20.4 70.2 flean -RT 64,340 88,480 7.06 20.0 68.95 Std dev'n 4,030 '240 0.82 0.33 0.89 FF-401 581 64.120 83,650 5.98 17.9 64.6 406 580 65,370 84,650 (a) 18.6 407 65.4 580 62,180 82,740 5.88 17.1 60.1 3fean 580 63,890 83,680 5.93 17.87 63.37 Std dev'n 15 1,310 780 0.61 0.05 2.33 j' Transverse
. ,s - F F- 302 RT 66,530 87,170 11.79 27.5 71.0 303 RT 65,630 h7,170 10.83 28.9 71.8 304 RT 66,500 88,530 10.29 30.0 71.9 Stean RT 66.350 87,620 10.99 28.8 71.57 Std dev'n 530 640 0.62 1.02 0.16 FF- 101 580 62,110 84,650 11.03 27.9 70.3 10 5 580 65.360 84,220 (a) 27.9 106 72.0 580 60,080 83,110 10.83 27.9 71.0
>!ea n 580 62.520 83,990 10.93 27.9 71.1 St il dev'n 5 2,170 650 0.10 0 0.70
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' ' s _5 Babcock s. Wilcox
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'T.sble B-$. Preirradiation Tensile Properties of Weld .
Ntal - Longitudinal . WF-209-1A Strength, psi Elo-igat f on . : Speci: men . temp, *fre% ,
*: 3. F Yield 111 t . Unif. Total T EE-102 .RT 81,193 95,140 10.84 26.1 60.7
-105' RT 83,530- 96,640 10'.36 24.6 51.0 120' RT '79,430
-93,800 11.0 26.1 62.1
- ean RT 81,380 95,190 10.73 25.6 57.94 Std.dev'n '1,680 1,160 0.27 0.71 4.94 EI-113 560 69,170 89,210 10.'4 20.7 46.8 ~
lib' 580 .69,730 89,910 10.4 20.7 49.9
'119 ~580 70,610 90,000 10.25- 20.4 50.0 S an - 580 69,840 89,710 10.35 20.6 48.9 Std dev'n 25 596 356 0.07 0.14 1.49 5
- B-6 Babcock 3.Wilcox
- 3.,
y m., w v t. - - - - - $_ ~ i k i 5 a 1 APPENDIX C Preirradiation Charpy Impact Data { I i f i-l I
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- Table C-1.' 'Preirrad'iation Charpy Impact Data for Shell Course Material -- 1.ongitudinal Orientation, Fest ' A.W 16 3 Test
- Absorbed Lateral- Shear Specirun tenp, energy, expansion, fracture, No. F ft-lb 10-3 in. %
EE-72f -361.7 140 53.5 100 720. 361.4' 134 6S 100 732 359.5' 132 71 100 EE-750. 250.9 151 65 100
;742 259.4 145 ~69.5 100 745 260.3 145 74.5 100-EE-734. 200.9 150.5 63 100 706 200.5 154 71 100 702 200.4- 147 74 100 EE-748 141 146 69.5 100 741 140.8 156 68.5 100 746 140.5 153 70.5 100 EE-719 80.2 108 67.5 45 736' 80 122 68.5 65 712- 79.9 99 64 o 35 EF-752 21.3 83 64.5 25 747~ 20.9 81 02.5 12 751 20.8 82 65.5 20 EE-735 0.1 55 40 1 701 0.1 47.5 35.5 --
730 -0.6 60 44.5 -- EE-744 11.8 75 54 6
- 74) .-12.4- 79 57 8
~749 -13.9 16 12.5 <1 EE-705 -40.9 13 9 0 lin -41.1 23 16 0
'724 -41.6 27 20 0 w
c-2 Babcock & Wilcox
l "7' Table C-2. . Preirradiatien Charpy Impact Data for Shell Cour8e Material - Transverse Wientation.
' Heat .W.' 163 -
Test Absorbed Lateral Shear
' Specimen. terp. energy,. expansion, fracture, No. _ f,_ ' t' t - I b 10 ' in.
- EE-626 3nl.1 -122 74 100 625_ Jnl.1- 12'; eS 100
.'615 199.3 '124: 75 100 620 358.7 102 70.5' 100 EE-630 271.4- 127 71 100 0 32 .' 7 8. 8 -' 118 6S 100 637J 276.3 122 69 100 EE-61d .' 02 -124 73 100 606 201- 12S 69.5 100 627 197.4 147 70.5 100 EE-639- 141.'1 134- 73.5 100 633 140.5 114 63.5 92 640 1 39 12S 70.5 100 EE-619 79.9 119 60.5' 100 608- 79.3 104 65 65
- 607 79.8 - 117 67 80
, EE-629 - 20.5 -62 50.5 8 631 20.4 77 60 25 CE-610- 9.9 6; 50 1 625 0.5 44 34 1 616 ~-1.5 61 44 o EE-615 -15 54 40 4 634 ' -15.5 34 23 2 636 -18 36 31 2 EE-64) -38.9 32 24.5 1 n01 -39 24 17 o
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Table C-3. Preirradiation Charpy Icipact Data for Shell ~ Course Material -- HAZ, Longitudinal Orien-tat ion, Heat AA'J 16 3 lest Absorbed Lateral Shear Specinen temp. energy, expansion - f rac ture,- No. F ft-lb 10-' in. 2 EE-434 -361.3 164 68 100 418 .360.8 174 67.5 -100 411 360.3 131 76.5 100 EE-444 281.6 113 74 100 449 280.6 104 74 100 442- .c o u ' 145 os' 100
-EE-407 202 124 57.5 100 425 200.5 86 49 100 435 199.5 185 73.5 100 EE-451 ~140.9 148 75.5 100 446 139 132 75 100 441 133.9 ~129 69.5 100 EE-437 80.6 81 43.5 100 421 80.5 134 73.5 75 402 80.3 158 73 100 EE-443' 41.9 112 62.5 45 44o 41.1 101 62 35 450 41.3 75 54.5 40 EE-438 0.6 97 56 20 418 0.5 128.5 63 40 406 0 114 57 45 EE-447 -39.7 81 53 12 452 -40.0 57 37 3 445 -40.1 53 36.5 8 EE-422 -79.7 90 49 35 426 -80.0 100 53.5 20 423 -82.5 73' 39.5 5 C-4 Babcock a.Wilcox
; o h
j e W h t Table;C-4.. Prcirradiation' Charpy lepact Data for Shell'
. Course Material - KAZ. Transverse Orienta-tion. Heat AAW 163-lea: Absorbed La te ra l Shear Specimen- : e:p . . energy.. e xpan s ien, . fracture, No. 'E A lb 10-3 in. *
~
- EE- 312 361.6 141- 61.5 100 EE-311 360.9: 108 75.5- 100-EE-307 -357.3 122 73 100 EE-321 260.9 120 74.5 100 EE-326 280.3 132 80 100 EE- 130 274.3 114' 75.5 100 EE-303 200.8 - 136 78.5 100 EE-308 200.5 113.5 57 96 EE-317 200.4- 129 76 100 EE-326 14116 -120 68. 100 EE-32 7 140.2- 102' 60.5 SS EE-3 32 140.1 134 75.5 100 EE-319' S0.7 87 65.5 50 EE-301~ .80.6 119- 71 75 EE- 369 80.3 101- 64 55
.EE-324. 31.2 78 54.5 40 EE- 322' 29.7 67 50 10 EE- 32 3 - 29.1 87 56.5 45 EE- 305 0.5 85 57 s EE-310 0.5 67.5 47 10 EE- 318 0 35.5- 26.5 2 EE- 325 - 39.' 5 66 43 8 EE- 329 . -40.1 84 45 15 t
EE-331 .-40.4 33 21.5 4
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EE- 306 -79.3 18 15.5 <1 EE- 304 -74.2 50 31.5 2 EE- 302 - -SI.2 33 22 3 1 A t c-5 Babck & hom a
t. Table.C-5. 'Preirradiation Charpy Impact Data for Shell Course Material - Longitudinal Orientation. Heat .WG 164 Test Absorbed Lateral Shear Specimen temp. energy, expansien, fracture, W. F ft-lb _ _10- 3 in. % FF-72 5 359.n 158 59 100 731 359.4 150 73.5 100 738 359.2 156 71 100 FF-706 2840. 4 161 66.5 100 708 200.4 156 71 100 72 8 2 th). 2 162 62 100 FF-760 140.7 165 68 100 754 140.7 156 72.5 100 732 119.9 154 74.,5 100 FF-707 60.5 140 77 713 65 80 179 64.5 734 100 79.8 162 69.5 100 , FF-703 40.8 103 73.5 730 30 40 124 80 717 55 34.6 106 75.5 45 FF-750 25.5 97 70 753 25 25.4 106 72.5 30 755 25.3 114 78 45 FF-751 11.8 102 69 749 18 12.4 58 44.5 6 756 12.7 93 63.5 15 FF-726 0.5 27- 21.5 1 732 0.5 67 51 6 729_ o 57 44.5 2 FF-759 -40 38 29.5 4 757 -41 14 12 0 , 75 8 -41 18 15 -1 l ( C-6 Wock & MCox
4 c Table'C-6. Preirradiation Charpy Impact Data. for Shell Course Heat
.'taterial - Transverse Orientation.
._ _ A*.*G 164 Test Absorbed Lateral Shear Speelmen. ' teep, energy,- expansion, fracture, No. F~ ft-lb 10' in.
-FF-628 362' 1 la 72.5 100 608 359.6- 128 70.5~ 190 FF-620 204 142.5 71 100 -
'602 200.4 133.5 70.5 100 6 32 200.3 138.5 76 100 FF-n32 139.5 130 70 100 651 -1 39. 3 104 65 100 642 139.0 134 73 100 FF- 621 -80.5 98 61 25 919 80.5 148 81.5 100 629- 80 90 62.5 12 636- 80 -126 73.5 85 611 79.8 151 72.5 100 4
~625 79.6 96 65 60-FF-646 57.2 91 65 55 649 56.1 86 63 35 647 56 52 44.5 30 F F-645 40 88 63.5 40 650 38.7 48 40 14 648 17.8 76 57 25 FF-644 20.5 70 56.5 8 643 20.4 78 61.5 5 641 20.5 FF-630 0.9 49 37.5 2 613 0.5' 24 20.5 0
-612 0 59.5 46 4 1
r c-7 Babcock & Wilcox
- tr . - , , --,-a ,,s
+---- 1 .- e w w
.y. . - .
- ab le C-7. 'Preirradiation Charpy impact Data ' f or Shell '
t Course ' Material - RAZ, Longi t udinal Orien-- ; tation. He.at N.*G 164 '
, ' Test Absorbed Lateral Shear Specinen' tenp.. energy, expansion, fracture,
, p. F (t-Ib 10~3 in. 4 FF-418 161.M 194 71.5 100 4 34 - 361.2 104 61 100 410 359.6 156 78.5 100 FF-449 284.4 116 81 100 454 279.5 122 80.5 100 450 278 128 80.5 100
?F-440 201.2 129.5 53.5 100 442 199.8 153.5 79.5 100 421 199.5 d2.5 43.5 100 FF-455 141 123 76.5 100 452 140.7 127 76.5 100 457 140.5 124 78 100 FF-402 .S0.2 147 73.5 80 431 SO 147 66 100 448 SO 99 53 95 FT-4nd 40.6 92 54 50 453 4 0. 7 ~ 102 70.5 55
- 456 40.8 S7 59 30 FF-404 0 96.5 50 15 409 0 148.2 72 70 441 0 59 33.5 55 FF-459 -39.9 36 28.5 4 ass -40.0 62 44.5 8 451 -40.3 10 9 0 FF-408 -78.8 138 64.5 50 407 -79.2 42 27.5 40 426 -79.6 93 50 55 C-8 Babcock & Wilcok i
L l 7 th le C-8. Preirradiation Charpy in: pact Data for Shell Course Material - HAZ, Trar.sverse Orien-tation. Heat nC 164 _
. hest Absorbed I.a t e ral Shear spec traen temp, fracture.
energy, expansion, ' W. F ft-lb 10~' in. % i j FF- 107 ' 362.8 124- 78.5 100 l 1 310 161.3 137 73.5 100 108 358.1 140 79.5 100 FF-324 '262 113 78 100 11 9 241.1 95 51.5 100 322 278.5 117 78.5 100
- FF- 30 ) 199.9 144.5 78.5 100 305 198.9 154 74.5 . 100 FF-318 140 120 72. y 100 325 113.7 130 79.5 100 12H 1 38.2 118 75 100 FF- 312 30.5 ISS 76.5 100 315 so.4' 141 74.5 100 116 74.8 148 78 100 FF-127- 39.6 118 69.5 70 321 19. 5 82 61.5 30 120 49. 0 62 50.5 15 FF-306 U.5 91 57.5 10 109 0.1 101.5 60 15 302 0 107 69 25 FF- 101 - 19. 6 71 40.5 6-111 -40 82 54.5 14 311 -40.1 117 53 12 FF-126 -77.9 6 3 0 317 -79.6 26 13 1 12 3 -ho.h- 36 19 1 c_9 Babcock s. Wilcox
4 Table C-9. Preirradiation Charpy Impact Data for Weld j _ Metal. WF-209-1A Test Absorbed Lateral Shear Speci:nen . . temp, energy, expansion, fracture, No. F fr-lb 10-3 in . : EE-014 361 67 53.5 100 032 360,6 72 52.5 100 035 360.1 65 49 100 EE-019 202.5 63 49.5 100 036 201.4 67.5 45 100 016 198.2 64 46 100 EE-020 120.9 68 47 . 93 011 120.3 66 47 100
-006 120.3 66 '44 100 EE-031 80.9 60 42.5 75 015 80.2 60 42 80 012 80.2 55 32 60 EE-024 0.5 27 25.5 25 022
-0.5 31 28.5 20 008 -0.5 25 23 15 4
i L L c-10 Babcock &Wilcox
Firure r-1. pact Da ta Fr"- Unirradiated Base .%:.11 is, f.on itu !!nal orient at ion i W . . , , , - - l l 75 . 1 e e
$ y - - - -- - - - ~ ~ - .,-
.- ~ ~ ~ - - ~ .
e 7 e .3 - e . e e, a a e a e i r e a e e 5 5 4 a e s g 5 I g 5
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. FLutact Ncne Neat iluesta _ AAV-163 0 . _ _ i f I t t t I
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?
160 2M 241 280 320 M1 (t) itsr it.,tearuet. F c-11 Babcock s. Wilcox
.. ~
Firizr. C-2. I:r. pac t Data Fro . L~ni rraci.it e : Ba-v % 21 . Tr c.sverse Grientat ion I I
- g' 5 y 6 w - w
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e ~ ( r m .ct Nc-2e Heat Enore AG-163 3
-0 -43. 0 40 60 120 160 2 71 241 280 -. EO #6 g Test itw tnarvet. F l c-12 Babcock s. Wilcox
F...r. '-:. '. pact Da t a F re.~. '. :1 r r.id I a t e d Sa me ", t.i l A. i 1 H V. , i.cn .-i t ud i n.il t r ien t.a t ion l I j .r )
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40 I witniat.SA508. C1.2
, Onituratto, I.ongitudinal Foutect NOM ~
Ntar Anete _AAV-163 o .
-m
-so o o so m m 160 itsr it weestung F 2m 2o no m vn c-1I Babcock 8. Wilcox
Fi ct re C- !. . : n,,a c t D.s t a F ro:. Un f r ra.f i..t ed E sse R t :il .'s. it'a. Irr.nsver e Orientation
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, p's st,v;.;T
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- .r . 1,., 150 st ts) -59F/14F ,
,. m ( m ) 126 f t-lb je -50F e
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w 81 - e . e e r t e a 6". - # j - n- -----+-----_-______________-___-__-_______ 40 - / l e / %;t ai n. SA508 C1.2
~
l Seiteration .Irpm.gverse l 3 ! Ftusscg None - l
*igaf Nwege _ AAb'-16 3
-o ' . . . . , , , , , ,
to -4 c M 80 123 160 20 24 281 320 %3 v/t Test itweearvet, F c-11 Babcock t. Wilcox
?!uure t-!.
i :nac : L:sta Fron Unitradiated Base .** e tal 3.
. i t .dinal Orientatiort
*g 5 a W - -
v os *K . e . I i 4 'f, . . . - - ----- - - - - J - - ~ ~ ~ ~ ~ - ~ ~ ~ ~ ~ ~ ~ . e 3
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75 e - E n a f n e i f 1
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- f f f f 1 f a
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r -. .l O. _F m .
.- 7 _
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= Ic. . ernot --.0F -1 e .
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- w- I 9 -
---g------------------------------------
e.c - , I I s igniat __ W OS. C1.2 w- . Ofst=f atim _ Longi tudinal e rougneg None -
- gar biogn _AWG-164 y f ,
-f3 -W 0 4 60 120 160 2M 2 ~) 2iG 329 i'3 #6 Ttst Tewsessant. F c 13 Babcock s. Wila <
FI . .i re C - +s . ! .,9ac t Data Fron L*nirr.liiated Base "etal S. l
'ransv.ar u Orlentation 10 . . i , . . .
's .
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;g i,,' ("A e s.td 9F/52F ,,
. .f l g,g)
.a 138 f t.-lb e
. l e. s;%.. -10F *_ -
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80 - / _
%rtant SA508, cl.2 e Onst=7Atica Transverse 23 -
None - pg,g,c, MtAT NWeta Na'0"Ibb g . . . . , . . i e i
-E0 W 0 < 60 120 If 200 20 280 '20 58 0 47) itst Tenetu.unt, i c-16 Babcock & Wilcox l
i
Fi c _ ri F-7. Irpa; Data Fr m Unirrad a ed Base .9tal B. R\Z.
- 0n ti t udinal Orien:.t:ien 2m > i y . . . .
s . l . e
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05 u: -60F/4 F
. ,3
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! %tti ng SA503, cl.2 Osiotatica 1.on g i t udina l
'M .
' ~
r,st wa Nore j e g., gm, N.'C-164 i 3 i i , , , , , , , W 40 .O O #0 13 ;sa 2 31 N . ga 33 tcg yn 7tst it = tear et, i t_- Babcock & Wilcox
Figure C-8. Inpact Data Fron t'nirradiated Base N tal 3 liAZ . Transverse or itstien
- r 8 s a 3 , w - w w De .3 -
=
. e Y
V <f . - - - . _ . . - . - . _ .-.-___-. -. ------ --.
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e 8 *
,, t e a e e a a e t 5 s g e .
r . . t r[ a g .
/ .'.,y ~
f p....--. ----.------.--- ---.-----------,
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i . . . . . . . . . CA's I. w Rf e -10F
- Tcv .
. f 55 sai -t 3f/~46I i ., 1,, 4 50 es.ts , -eM/16F ,
, .g<,,g g ,o 11. 4 ft-lb *
, --- w
- 10 T * *
, i. . si g, l ." *
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a e
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~
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- 4
~ v . / .
l
- -----,/- - - - - - - - - - - - - - . - - - - - - - - - - - - - - - - . - - - .
V / y . gnug SA509. C1.2 l
'n . Caststatica Transverse .
Ft wage , NN ? war we E-1164 i i . , , , , , rJ 'O ; o 83 120 160 t v) :o is) m < .n Test Itweentwet. F c.;3 Babcock & Wilcox I
Ficare C-9. Igac t Data Trem Unirradiated '.leid Metal, Transvers. Orient ation
*1n i i 4 .
I e e . a 2
- y se - - - . _ . . _ _ - _ . _- - . _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ .
e_ E Y
- s. 28 -
e e e, a e e a t a e e e n
- 5 4 4 3 4 4 s s s u 9
f . v. - s~ ca
=
e e s t O W *
.r t
- -= -*= . . , .
[ - ame. . _ r a-
, , '1",
- a.w e
-4 vy; I f f f f
- 1 0 9 e s
..i 5 = - > > a . i i .
. g-
.I y, _ - - - - -
a2F/100F . 3 (55 %
... ?, , (50 st-u t 50F/6I.F
(-N (a.o _ t 3 f t-lb
! *' - F F e , _ _ _.
3 - s l
- 1
?. ,; tv -
l
- l l
t. I. 3 . i " e e s, .
- e
- -------- /
/
sc - /
%,g Weld ?k tal
, , Ca r t=*at s os .- _
Ftas:t None - I ' ? Maar gnes *T-209-1A
=
I 1 t t N ) C M 80 3 t t t t LN n 2d .'n m m <xt itst Itetent.ne. F c-19 Babcock s. Wilcox
+
h t APPENDIX D 71 resheld Detector Infortutien i r D-1 Babcock & Wilcox
l .' Tchle' D-1 lists the cumposition of 'the threshold detectors and the cadmium thickness used to reduce coc:peting thernal reactions. Table D-2 shows the Cycle .1 measured activity per gram of target material (i.e. , pet gram of ura-nium, nickel, etc.) corrected for the wair time between irradiation and counting. Activation cross sections foc the various materials were flux-wmighted with a 235 U spectrus (Table D-3). i Table D-1. Detector Composition and Shselding mnitors Shielding Reaction 11.87 U-Al Cd-Ag 0.02676" Cd 23eg(,,g) 1.61: Np-Al Cd-Ag 0.02676" Cd 237 Np(n f) Ni Cd-Ag 0.02676" Cd seNi(n.p)S9Co 0.66* Co-Al Cd-0.0!.0" Cd 5?Co(n,y)sDCo 0.66 Co-Al None 5'3Co(n,y)03Co Fs' None 5"Fe(n.p)S4m I D-2 Babcock a. Wilcox
' Table D-2.
iA ongle, Cycle 1.Yeut ron ikwilj;eterpI ") Pos t I trad. Monitor Jt3,g _ Nucilde acl/g of keaction_ Nuc1ide act, LCl material b) .t.Ci/go{C*d)i .- targett Cli-l (Side), I'1*C-Al 0.04844
' "i (n , f ) FP '"'Z r 0.1780 '3.675 35.7
'"' Nb .l.128 21.29 226 IO Ru ~
0.2140 4.418 42.9 137 Cs 0.006329 0.1307.- 1.27 Mice 0.1621 3.346 32.5 l 'Ce 0.1010 2.126 20.6 10cRu 0.08162 1.685 16.4 237Np-Al 237 Np(n.f)FP 0.08617 95 Zr 0.2317. 2.689 187 3kNb 1,.655 19.21 1.33+3 103 Ru 0.2289 2.656 184 3 '7Cs 0.008440 0.09795 6.80 8 '* l ce 0.1651 1.916 133 l 4 Ct. O.II15 1.294 H9.9 106Ru 0.1225 1.422 98.7 N1 0.13126 LB N1 (n . p) 8Co 58Co 68.19 519.5 767
$ 60N1 (n .p)
- 0Co I'"Co 0.1585 1.208 4.62 Co-Al (0.625") 0.01212 '9' Co(n,y)60Co ' 'Co
- in cd 2.807 231.6 4.14+4 P
f Co-Al (0.5") 0.01605 59 Co(n ,y )00Co 6o Co 8 4.624- 288.1 5.14+4 M 4
- 79.hI" D-2,._(conLQ
.1 Niscilde pC1/g cf(U pC1/A e 'd )
l'.
- t i r ra.f . e matie r tol t ergt R. n t ic n Nuclidia_ nei , nl
' Monitor _& n 3.867 25.45 437 5Fe(n .p) S 5tn 5 Nn Fe 0.15195 6.779 44.61 1.35+4 ;
" Fe (n , r ) LFc 59Fe j
- t CD 2 (Side) .
35.0 ~l 2 3eU(n,f)FP 95Z r 0.2268 3.610 23811-Al' O.06283 1.355 21.57 *09 95Nb 0.2757 4.388 42.6 103Ru 137Cs 0.008274 0.1317 1.28 141Ce 0.2071 3.296 32.0-I""Ce 0,1254 1.996 19.4 i 106 Ru 0.09003 -1.433 .13.9 237 Np (n , f) FF 95 Zr 0.1648 2.428 169. 237Np-Al 0.06787 1.327 19.55 1.16+3 95Nb 0.1672 2.464 171 103Ru 137Cs 0.005327 0.07849 5.45 l Ice 0.1149 1.693 118. 14 8.Ce 0.17175 1.057 73.4 100Ru 0.09897 1.458 101. g 531.0 784.
$ 0.I3327 5"N! (n ,p)"Co 58Co 70.77 R N1 0.1569 1.177 4.50 60Ni(n.p)60Co coco p.
s 0.01723 59Co(n,y) coco c0Co 2.607 151.3 .2.70&4 N Co-Al (0.625") E in cd
j-V T.thip_,Il-1.,, friutt,'si)_ Posttrrad. Nuclide
~
Monitor i,C1/g cl CD3/ wt,,g _ Reaction Ntselide .act. bCi' mat erial(b) uti/g target o[#'d) CiD. Co-A1 (0.5") 0.01535 V'Co(n.r)'"Co ' '1 Co 4.483 292.1 '5.22+4- - - Fe 0.15257 . ' Fe(n .p) 5Mn 5 t'Mn '3.980 26.09 448.
'*Fe(n.3)'dfe 'A Fe 7.368 48.29 1.46&4 CD 3 (Frong 2'"U-Al 0.05768 M"U(n . f ) FP 'Z r 0.2250 3.901 37.9 1.76 95ht 1.238 21.46 208. 2.12 103 Ru 0.2719 4. 714 45.8 1.H8 ,
337Cs 0 J08099 0.1404 1.36 1.49
? '
. V' I'* Ice 0.2061 3.573- '34. 7 - 1.99 . 14'.Ce 0.1305 2.262 22.0 1.64
.7 10'Ru 0.07146 1.239 12.0 1.66
?)?gp.gg g,g77gg 237Np(n . f)FP 9 Z r U.2526 3.278 228. 1.62 35Nb 1.675 21.74 1.51+3 1.54 103Ru 0.2597 3.371 234. 1.64 137Cs 0.008720 0.1132 7.86 1.18
- 14. Ice 0.1833 2.379 165.
I 1.70 I"Ce 0.1153 1.496 104. 1.62 w ICERu 0.09836 1.277 88.6 1.30 e= Ni 0.13121 58Ni(n.p) S 9Co 59Co 86.02 655.6 967. 1.78 [ 60 N1 (n .p)0Co coco 0.2166 a 1.651 6 . 11 1.81 M
'p.
Tphly D-2. (Cont't) l'o% t I r ra*l . Nuclido pC1/g of CD3/ Monitor _ yt ,_ g_ -, Reaction Nuclide act, pCi material b) pCi/g targeto[Ced) CD4 Co-Al-(0.625") 0.01007 'U' Co (n ,1 ) 5 OCo 60Co 3.248 322.5 5.76+4 3.49 in Cel Co-Al (0.5") 0.01565 - 'Co(n,y) coco GUCo 5.672 362.4 6.47+4 2.21 Fe 0.15395 5Fe (n .p) L 5tn S a.Nn 4.463 '28.99 498. 1.60 58 Fe(n y)s97, e 5')Fe 8.952 58.15 1.76+4' 2.51 Cl) 4 (Hack)
'238U -Al 0.04700 2 3aU(n , f)FP 95Zr 0.1041 2.215 21.5 95Nb 0.4757 10.12 98.2 y 101Hu 0.117Y /.500 24.1 es 13 /Cu 0.004412 0.0938/ 1.2053 18.lCe 0.08411 1.790- 17.4 l a.i.Ce 0.06506 1.384 13.4 10bRu 0.03494 0.7434 7.22 237 Np-Al 0.08679 237 Np (n ,f )FP 0.1762
'I'd Z r 2.030 141.
95Ni> 1.226 14.13 981. IO3Ru 0.1782 2.053 143. F 137Cs 0.008336 0.09605 765.5 l4 Ice 0.1216 1.401 97.3 [ l 4Ce 0.08039 0.'263 64.3 g
~
106Ru 0.08543 0.9843 68.4 n E- NI 0.I3241 S aN!(n .p) 58Co 58co 48.84 368.9 6.695
'30N1 (n .p)60Co 60Co 0.1204 0.9093 3.48 naser m = ~ - . . .
TabIe D-2. (cont'3Q
.Powtitrad. Nuclide LC1/g of Monitor
^
wLL ' Reaction Nuclide act, pCi material ') UCI/g o[C'd) target Co-Al (0.625") 0.02026 ' *co (n , y )"Co
'#co 1.867-an cd 92.15 1.65+4 Co-Al (0.5") 0.01558 Co (n . ) ) 6 0Co "Co 2.557 164.1 2.93+4 Fe 0.15463 '^ Fe (n . p) *' 5!n . 52.Mn 2.806 18.15 423.75 56 Fe(n , y )"* 3Fe 59Fe 3.573 23.11 7.'Od&3 (a) Analyses performed at Lynchburg Research Center and reported eo n in theJ.mK.m Schmot-um, ra d zur to A. L. Lowe, "Oconce 2 Neut ron Dosimetry " 5135-55, August 10, 1976.
(h)1his colunst . is the disintegration rate per gram of wire using-the postirradiation weight. . (* Thin column in the disintegration rate per gram of target nuellde , v i z . , 2 3 "U , 2 "Np . . "N 1, 6DNi, 59Co. 4 Fe, Supe, (d)1he following abundances and weight percents were used to calculate theegration disint rate per gram of target nuclide: 2 '"11 . . 10.3H wt %; 99.27% isotopic 2 "Np 1.44 wt %; 100% isotopic N1 100 wt %; 5"Ni 67.77% inotopic, " Ni 26.16 INotopic. Co 0.66 wt %; 59Co 100% isotopic cn Fe 100 wt %; 54Fe 5.82% !sotopic, "Fe 0.33% isotopic an O R w f* n t
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