ML20040B071
ML20040B071 | |
Person / Time | |
---|---|
Site: | Oconee ![]() |
Issue date: | 12/31/1981 |
From: | Ewing J, Lowe A, Pavinich W BABCOCK & WILCOX CO. |
To: | |
Shared Package | |
ML15223A766 | List: |
References | |
BAW-1699, NUDOCS 8201250101 | |
Download: ML20040B071 (100) | |
Text
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1 BAW-1699 December 1981 II 4
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ANALYSIS OF CAPSULE OCII-A FROM llW l
DUKE POWER COMPANY'S OCONEE NUCLEAR STATION, UNIT 2 I,
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- Reactor Vessel Material Surveillance Program -
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BAW-16'39 December 1981 4
i lI lI ANALYSIS OF CAPSULE OCII-A FROM DUKE POWER COMPANY'S OCONEE NUCLEAR STATION, UNIT 2
- Reactor Vessel Material Surveillance Program -
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A. L. Lowe, J r., PE J. W. Ewing
. g W. A. Pavinich g
W. L. Redd J. K. Schmatzer iil I
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I B&W Contract No. 582-7109-71 1
{g BABCOCK 6 WILCOX j
Nuclear Power Group l3 Nuclear Power Generation Division l
P. C. Box 1260 l
Lynchburg, Virginia 24505 i
i Babcock & Wilcox
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!I SU}B1ARY lI j
This report describes the results of the examination of the second capsule of 4
Duke Pcwer Company's Oconee Unit 2 reactor vessel surveillance program. The I
18 capsule was removed and examined after accumulating a fluence of 3.37 x 10 nyt, which is equivalent to approximately 16 effective full-power years (EFPY).
The objective of the program is to monitor the effects of neutron irradiation on the tensile and fracture toughness properties of the reactor pressure ves-sel materials by testing and evaluating tensile, Charpy impact, and compact fracture toughness specimens. The program was designed in accordance with the 1
requirements of Appendix H to 10 CFR 50 and ASTM specification E185-73.
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j The capsule received an average fast fluence of 3.37 x 1
2 10 " n/cm (E > 1 Mev),
lg and the predicted fast fluence for the reactor vessel T/4 location at the end
)E of 3.7 EFTY operation is 9.8 x 10 n/cm (E > 1 Mev).
Based on the calculated 17 2
current fast flux at the vessel inside wall and an 80% load factor, the pro-i jected fast fluence the Oconee Unit 2 reactor prassure vessel will receive in 10 ' n/cm2 1
40 calendar years of operation is 1.20 x (E > 1 Mev).
The results of the tensile tests indicated that the materials exhibitea normal behavior relative to neutron fluence exposure. The Charpy impact results ex-hibited the characteristic behavior of a shift to higher temperature for both j
the 30- and 50-ft-lb transition temperatures as a result of neutron fluence damage and a decrease in upper shelf energy. These results demonstrated that the current techniques used for predicting the change in both the increase in the RT and the decrease in upper shelf properties due to irradiation are ET j E conservative.
f The recommended operating period was extended to 15 EFPY as a result of the lN second capsule evaluation. These new operating liaitations are in accordance with the requirements of 10 CFR 50, Appendix G.
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1.
INTRODUCTION l
This report describes the results of the examination of the second capsule of i
j Duke Power Company's Oconee Nuclear Station, Unit 2 reactor vessel surveillance program. The first capsule from this program was removed and examined after i
f the first year of operation; the results are reported in BAW-1437.2 The objective of the program is to monitor the effects of neution irradiation i
E on the tensile and impact properties of reactor pretsure vessel materials under actual operating conditions. The surveillance program for Oconec 2 was de-I signed and furnished by Babcock & Wilcox; it is described in BAW-10006A.3 The program was planned to monitor the ef fects of neutron irradiation on the reactor vessel materials for the 40 year design life of the reactor pressure tessel.
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1 The surveillance program for Oconee 2 was designed in accordance with E185-66 I
and thus does not comply with Appendixes C and 11 to 10 CFR 50 since the re-l l
quirements did not exist at the time the program was designed.
Because of this j
difference, additional tests and evaluations were required to ensure meeting the requirements of 10 CFR 50, Appendixes G and II.
The recommendations for the future operation of Oconee 2 included in this report do comply with these requirements.
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I 2.
BACKCROUND I
The ability of the reactor pressure vessel to resist fracture is the primary factor in ensuring the safety of the primary system in light water cooled re-actors. The beltline region of the reactor vessel is the most critical region of the vessel because it is exposed to neutron irradiation. The general ef-fects of fast neutron irradiatian on the mechanical properties of such low-alloy ferritic steels as SA508, Class 2 forgings used in the fabrication of the Oconee 2 reactor vessel are well characterized and documented in the lit-erature. The low-alloy ferritic steels used in the beltline region of rcactor vessels exhibit an increase in ultimate and yield strength properties after I
irradiation, with a corresponding decrease in ductility.
The most serious mechanical property change in reactor pressure vessel steels is the increase in temperature for the transition from brittle to ductile fracture accompanied by a reduction in the upper shelf impact strength.
Appendix G to 10 CFR 50, " Fracture Toughness Requirements," specifies minimum fracture toughness requirements for the ferritic materials of the pressure-retaining components of the reactor coolant pressure boundary (RCPB) of water-cooled power reactors and provides specific guidelines for determining the pressure-temperature limitations on operation of the RCPB. The toughness and I
operational requirements are specified to provide adequate safety margins dur-ing any condition of normal operation, including anticipated operational oc-currences and system hydrostatic tests, to which the pressure boundary may be subjected 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 re ntors, in-cluding those under construction or in operation on the effective date.
Appendix H to 10 CFR 50, " Reactor Vessel Material Surveillance Program Require-ments," defines the material surveillance program required to monitor changes in the fracture toughness properties of ferritic materials in the reactor ves-sel beltline region of water-cooled reactors resulting from exposure to neutron I
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iI irradiation and the thermal environment.
Fracture toughness test data are ob-tained from material specimens withdrawn periodically from the reactor vessel.
These data will permit determination of the conditions under which the vessel can be operated with adequate safety margins against fracture throughout its jI service life, i
lg A method for guarding against brittle fracture in reactor pressure vessels is I E described in Appendix G to the ASME Boiler and Pressure Vessel C.:de,Section III.
This method ucilizes fracture mechanics concepts and the reference nil-ductility temperature, RTg. which is defined as the greater of the drop j
weight nil-ductility transition temperature (per ASTM E-?O8) or t he tempera-ture that is 60F helow that at which the material exhibits 50 ft-lb and 35 mils lateral expansion. The RT of a given material is used to index that i
g material to a reference stross intensity factor curve (ItIR ""
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pears in Appendix C cf ASME Section III.
The K IR "
dynamic, static, and crack arrest fracture toughness results obtained from several heats of pressure vessel steel.
When a given material is. Indexed to the K curve, allawable stress intensity factors can be obtained for this ma-g terial as a function of temperature. Allowable operating limits can then be determined using these allowable st ress intensity f actors.
E The RT and, in turn, the operating limits of a nuclear power plant, can be ET adjusted to account for the effects of radiation on the properties of the re-ector vessel materials. The radiat. ion embrittlement ar.d the resultant changes in mechanical properties of a given pressure vessel steel can be monitored by I
a surveillance program in which a surveillance capsule containing prepared specimens of the reactor vessel materials is periodically removed from the operating nuclear reactor and the specimens tested. The increase in the Charpy V-notch 50-ft-Ib temperature, or the increase in the 35 mils of lateral expan-sion temperature, whichever requits ir. the larger temperature shifL due to ir-radiation, is added to the original RT NDT ment.
I in turn, is used to set operating limits for the nuclear power plant. These 1
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,ew limits take into account the effects of irradiation on the reactor vessel i
materials.
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SURVEILLANCE PROGRAM DESCRIPTION The surveillance program for Oconee 2 comprises six surveillance capsules de-I signed to monitor the effects of the neutron and thermal environment on the materials of the reactor pressure vessel core region. The capsules, which were i
4 inserted into tne reactor vessel before initial plant startup, were positioned inside the vessel between the thermal shield and the vessel wall at the loca-tions shown in Figure 3-1.
Two capsules were placed in each holder tube and l
positioned near the peak axial and azimuthal neutron flux.
BAW-10006A includes l
a full description of capsule locations and design.3 After the capsules were 1
i removed from Oconee Unit 2 and were included in the integrated reactor vessel i
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material surveillance program, they were irradiated in the Crystal River Unit 3 reactor. During this period of irradiation Capsule OCII-A was irradiated in site yz as shown in Figure 3-2.
Capsule OCII-A was removed from Crystal River Unit 3 after cycle 2 and an ac-cumulated fluence of approximately 3 x 10 nyt.
This capsule contained Charpy 18 V-notch impact and tensile specimens fabricated of SA508, Class 2 steel, weld metal, and correlation steel.
The specimens contained in the capsule are de-j scribed in Table 3-1, and the chemistry and heat treatment of the surveillance
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=aterial are described in Table 3-2.
lg The capsule contained longitudinal Charpy V-notch specimens from correla-tion material obtained from plate 02 of the USAI'C Heavy Section Steel Technol-j ogy Program. This 12-inch-thick plate of ASTM 533, Grade B, Class I steel was 1
]
produced by the Luken Steel Company (heat A-1193-1) and heat-treated by Combus-3 tion Engineering. The chemistry and heat treatment of the cc,rrelation material 1
i are dencribed in Table 3-3.
All test specimens were machined from the 1/4-thickness location of the shell e
i forging. Charpy V-notch and tensile specimens from the vessel material were oriented with their longitudinal axes parallel to the princ.ipal working direc-tion of the forgings; the specimens were also oriented transverse to the 3-1 Babcock & Wilcox 1
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principal working direction. Capsule OCII-A contained dosimeter wires, de-scribed as follows:
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Dosimeter wire Shielding I
j U-Al alloy Cd-Ag alloy j
Np-Al alloy Cd-Ag alloy I
Nickel Cd-Ag alloy 1
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- 0. 66% Co-Al alloy Cd l
0.66% Co-Al alloy None l
Fe None 4
Thermal monitors of low-ruelting eutectic alloys and pure metals were included
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4 in the capsule. The eutectic alloys and metals and their melting points are i
as follows:
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Alloy Melting point, F us 90*: Pb, 5% Ag, 5% Sn 558 97.5% Pb, 2.5% Ag 580 97.5% Pb, 1.5% Ag, 1.0% Sn 588 Lead 621 Cadmium 610 Table 3-1.
Specimens in Surveillance Capsule OCII-A 1
No. of specimens l
Material description Tensile Charpy Weld metal, WF-209-1A 4
8 lient-af fected zone "A" (HAZ),
He-t AAW-163, Longitudinal 0
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Base metal material, plate "A,"
j Heat AAW-163, Longitudinal 4
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Transverse 0
4 W
Correlation, HSST, plate 02 Heat A-Il95-1 0
8 Total per capsule 8
36 E
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i!I Table 3-2.
Chemistry and lleat Treatment of Surveillance Materials i
l Chemical Analysis Ileat Weld metal i
Element AAW-163 WF-209-1A I
C 0.24 0.057 Mn 0.63 1,58 P
0.006 0.020 I
S 0.012 0.005 Si 0.25 0.56 Ni 0.75 0.48 i
Mo 0.62 0.33 i
Cu 0.04 0.30 lleat Treatment
- Time, Ileat No.
Temp, F h
Cooling AAW-163 1620-1660 4.0 Cold water quench 1570-1610 4.0 Cold water quench 1240-1280 10.0 Cold water quench 1100-1150 40.0 Furnace-cooled KT-209-1 A 1100-1150 33.0 Furnace-cooled I
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1 Table 3-3.
Chemistry and liest Treatment of Correlation l
Material - Eleat A-1195-1, A533 Grade B, Class 1 (HSST Plate 02)
Chemical Analysis (1/4T)("}
l Element yt %
l C
0.23 Mn 1.39 0
0.013 l
S 0.013 I
Li 0.21 i
Ni 0.64 Mo 0.50
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Cu 0.17 i
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Trea t men t (b )
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Normalized at 1575F 1 75F.
2.
1600F 1 75F for 4 h/ water-quenched.
3.
1225F ! 25F for 4 h/ furnace-cnoled.
4.
1125F 1 25F for 40 h/ furnace-cooled.
(" ORNL-4463.
iI (b)Per plar e section identification ca,rd.
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I Figure 3 '.. Reactor Vessel Cross Section Showing Surveillance Capsule Locations at Oconee Unit 2 i
Surst:illance Capsule Holder Tube - Cape.td es OCII-C, OCII-D f'
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Surveillance Capsule h
j Holder Tube -- Cap-
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sules OC11-A, OCIl-B Surveillance Capoule lloider Tube - Capsules Z
OC11-E, OCII-F
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Figure 3-2.
Reactor Vessel Cross Section Showing Location of
'I Oconee Unit 2 Capsule OCII-A in Crystal River Unit 3 Renator I
i SURVElLLANCE CAPSULE HOLDER -
TUBES g
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e SITE OCONEE ej e
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CAPSULE
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7 CAPSULE HOLDER CAPSULF HOLDER TUBE I
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1 4.
PREIRRADIATION TESTS
!I Unirradiated material was evaluated for tm purposes: (1) to establish a base-l j
line of data to which irradiated properties data could be referenced, and (2) j to determine to the ce. tent practical from available material, the material I
properties as required for compliance with Appendixes G and H to 10 CFR 50.
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l 4.1.
Tensile Tests b
rensile specimens were fabricated from the reactor vessel chell course plate I
and we'.d metal. The subsize specimens were 4.25 inches long with a reduced section 1.750 inches long by 0.357 inch in diameter. They were tested on a 55,000-lb load capacity universal test machine at a crosshead speed of 0.050 inch per minute. A four-pole extension device with a strain-gaged extensometer was used to determine the 0.2% yield point. Test conditions were in accordance j
with the applicable requirements of ASTM A370-72.
For each material type and/or condition, six specimens in groups of three were tested at both room tempera-
)E ture and 580F. The tension-coupression load cell used had a certified accuracy I
l of better than 2 0.5% of full scale (25,000 lb).
All test data for the pre-irradiation tensile specimens are given in Appendix B.
'.2.
Impact Tests I
j Charpy V-notch impact tests were conducted in accordance with the requirements of ASTM Standard Methods A370-72 and E23-72 on an impact tester certified to meet Watertown standards. Test specimens were of the Charpy V-notch type,
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which were nominally 0.394 inch square and 2.165 inches long.
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- E Prior to testing, specimens were temperature-controlled in liquid immersion i
j baths capable of covering the temperature range from -85 to +600F.
Specimens were removed from the baths and positioned in the test frame anvil with spe-i cially designed tongs.
The pendulum (hammer) was released manually, allowing the specimens to be broken within 5 seconds after their removal from the tem-perature baths, i
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Impact test data for the unirradiated baseline reference materials are present-ed in Appendir. C.
Tables C-1 through C-4 contain the basic data, which are plotted in Figures C-1 through C-4.
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POSTIRRADIATION TESTS 5.1.
Thermal Monitors i
Surveillance capsule OCII-A contained three temperature monitor holder tubes, each containing five fusible alloys with different melting points ranging from 558 to 621F. All the thermal monitors at 558, 580, 588, and 610F had melted, I
while those at 621F remained in their original configuration as initially i
placed in the capsule except for slight signs of slumping. From these data it was concluded that the irradiated specimens had been exposed to a maximum j
temperature in the range from 610 to less than 621F during the reactor vessel period. This higher (than anticipated) temperature occurred because these capsules were designed for lower lead factor capsule positions. The higher lead factor in the Crystal River Unit 3 radiation site would cause a slight I
increase in gamma heating, which in tt.rn could cause the increase in tempera-ture detected by the thermal monitors. There appeared to be no,significant temperature gradient along the capsule length.
1 5.2.
Tensile Test Results I
The results of the postirradiation tensile tests are presented in Table 5-1.
l Te sts were performed on specimens at both room temperature and 580F using the same test procedures and techniques used to test the unirradiated specimens (section 4.1).
In general, the ultimate strength and yield strength of the material increased slightly with a corresponding slight decrease in ductility; both effects were the result of neutron radiation damage. The type of behavior observed and the degree to which the material properties changed are within i
the range of changes to be expected for the radiation environment to which the specimens were exposed.
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The results of the preirradiation tensile tests are presented in Appendix B.
1 5.3.
Charpy V-Notch Impact Test Results I
The test results f rom the irradiated Charpy V-notch specimens of the reactor lg vessel beltline material and the correlation monitor material are presented in 5-1 Babcock & Wilcox
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Tables 5-2 through 5-6 and Figures 5-1 through 5-5.
The test procedures and I
techniques were the same as those used to test the unirradiated specimens (section 4.2).
The material exhibited a sensitivity to irradiation within the values predicted from its chemical composition and the fluence to which it was exposed.
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The results of the preirradiation Charpy V-notch impact test are given in Appendix C.
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}!I Table 5-1.
Irradiation Tensile Preperties of Capsule OCII-A Base l
Metal and Weld Metal Irradiated to 3.37 x 10 nyt 18 l
(E 1.0 Mev)
Red'n
- I eng, psi Elongat ion, %
Specimen Test temp, in area, No.
F Yield Ult.
Unif Total f
I lg Base Metal, Longitudinal - Heat AAW-163 19 EE-715 69 72,500 94,400 14.0 27.1 68.9 EE-717 69 70,300 91,900 14.1 30.6 68.9 Mean 69 71,400 93,150 14.05 28.85 68.9 Std dev'n 1,550 1,770 0.07 2.47 0
EE-703 579 63,800 86,300 12.5 24.3 67.3 EE-712 582 65,600 87,500 13.7 25.5 68.9 i
>'ean 580 64,700 86,900 13.57 24.9 68.1 l
Std dev'n 1,270 850 0.74 0.85 1.13 1
1 Weld Metal - WF-209-1 l
EE-115 69 96,300 110,000 14.9 24.9 57.0 ig E E-I l 7 69 99,400 111,300 15.3 17.1 29.8 ig Mean 69 97,850 110,650 15.1 21.0 43.4 Std dev'n 2,190 920 0.28 5.52 19.23 1
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EE-103 584 86,300 103,100 12.1 22.6 43.5 l
EE-121 581 86,200 101,300 12.2 18.3 44.4 Mean 580 86,250 102,200 12.15 20.45 43.95
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Std dev'n 70 1,270 0.07 3.04 0.64 lI J
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Table 5-2.
Irradiated Charpy Impact Data for Capsule OCII-A 18 i
Irradiated to 3.37 > 10 nyt, Base Metal IIAZ, i
Longitudinal Orientation 1
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Absorbed Lateral Chear Specimen Test temp,
- energy, expansion,
- fracture, No.
F ft-lb 10-3 in.
EE-413
-80 62.5 47.0 30 EE-401
-35 78.5 47.0 20 EE-414 0
93.5 63.5 30 I
EE-415 38 51.0 40.0 70 l-EE-439 75 117.0 77.0 50 EE-419 140 127.5 94.0 100 EE-436 224 139.0 93.0 100 EE-424 440 115.0 83.5 100
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' Irradiated Charpy Impact Data for Capsule OCII-A 18 Irradiated to 3.37 x 10 nyt, Base Metal, Ttansverse Orientat-ion 1
Absorbed Lateral Shear Specimen Test temp,
- energy, expansion,
- fracture, j
No.
F ft-lb 10-3 in.
EE-609
-1 41.5 34.0 5
EE-602 38 73.5 51.0 15 EE-614 40 53.5 41.5 5
EE-621 75 97.5 66.5 30 I
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i Table 5-4.
Irradiated Charpy impact Data for Capsule OCII-A 18 Irradiated to 3.37 x 10 nvt, fase Metal, j
j Longitudinal Orientation Absorbed Late ral Shear Specimen Test temp,
- energy, expansion,
- fracture, No.
F ft-lb 10-3 in.
EE-715
-40 22.0 15.5 0
i EE-723
-20 56.5 41.0 10
!f5 EE-737 0
41.5 29.0 0
ig l
EE-704 19 69.0 53.5 30 i
EE-714 75 108.0 72.5 75 EE-707 140 172.0 81.5 85
)
EE-731 224 134.0 87.5 100 EE-739 400 126.0 94.0 100 i
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Table 5-5.
Irradiated Charpy Inpact Data for Capsule GCII-i.
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Irradiated to 3.37 x 10 a nst, Weld Metal 1
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Absorbed Lateral Shear j
Specimen Test temp,
- energy, expansion,
- fracture,
)
No.
F ft-lb 10-3 in.
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EE-013 38 17.5 18.5 10 EE-003 75 19.0 28.5 20 EE-021 110 32.0 33.0 65 i
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EE-039 140 33.0 34.5 90 1
j EE-029 193 42.0 48.0 100 i
j EE-002 226 46.5 44.0 100 I
j EE-005 284 46.0 46.0 100 i
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EE-0 33 408 44.5 44.0 100
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- 7. 9 (+9) 1.8(+9) 2.74(+10)
- 3. 2 (+18) 7.3(+17) j 1
- Flux, Cumulative
- Flux,
- fluence, n/cm -s 2
- 9. 43 (+17) 4.66(+10) 440 EFPD 4
- 2. '.3 (+18) 3.37(+18)
- 1. 96 (+11 )
- Revision 1, January 1976.
- bt e r t.i l 111r n t 4 15 1/4T ront ent,
- 1. lOt l N
- 7. jliF l ?
- 2. /.1F I R 5.U t l 7 li, 69 ih%
- 1. /t >I' I P 7.10117 1%
- 1..we r i t r r um. scam
- 1. 7H l h
- m 4.0 A
- NDI, RIGHT OF TH[ LIMIT CURVE (S).
- TEMP, g
- TEMP, 1800 P0:nT Ps s r
- TEMP, Poln1 PSIG F
- 1400 10$100Flh($50Fla ANY 1/2.h PERIOD) cc o
- fluence, data 2
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- 0. 74 AMX 77 SA508 C12 L
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- g Plate Material, Heat AAW 163 ig i
- temp, of area,
- temp, of area, No.
- temp, energy, expansion,
- fracture, No.
- temp, energy, expansion,
- fracture, Nc.
- temp, energy, expansion,
- fracture, No.
- temp, energy, expansion,
- fracture, I
- 75 E
- 75 C
- 50 m
SUMMARY
180-T
+20F I
oy T (35 att)
- "F cy I
FT-LB) "#E CV T (30 FT-ts) -24F cy
. 140-20F
=
C I
5 120 S
I 3 100-E I
J5 80-C I
2 00-l 40-34608.cL2 Martn ut g,,,,,,,,
,,,,,,,,,,,c 20 FLUENCE NONE I
HEAT No. A Aw-163 I
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0
-00
-40 0
40 80 120 160 200 240 280 320 360 400 I
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Impact Dat.t From Unirradiated Weld Metal E
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0 100 I
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g g
g LATA SUTARY 90+ T,37
-20F
]T (35 ntE) +42F I
cy l
+50F 80 Tcv (50 n-La)
T (30 n-ts) cy C -USE (avr.)
67 FT-Le I
y 70 RT
+4F
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Q.--
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u,ATERIAL WELD METAL OniENTAtto=
10-FLUENCE NOME HEAT No.
WF-209-1 I
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-CD
-40 0
40 80 120 160 200 240 280 320 360 400 Test TEMPERATURE, F I
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I APPENDIX D Fluence Analysis Procedures I
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D-1 Babcock & Wilcox l
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1.
Analytical Method Energy-dependent neutron fluxes at the detector locations were determined by a discrete ordinates solution of the Boltzmann transport equation with the two-dimensional code DOT 3.5.'
The Oconee 2 and CR-3 reactors were modeled from the core out to the primary concrete shield in R-theta geometry [ based on a plan view along the core midplane and one-eighth core symmetry in the azimuthal (theta) dimension]. Also included was an explicit model of a surveillance cap-sule assembly in the downcomer region. The reactor model contained the follow-ing regions: core, liner, bypass coolant, core barrel, inlet coolant, thermal shield, inlet coolant (downcomer), pressure vessel, cavity, and concrete shield.
Input data to the code included a pin-by-pin, time-averaged power distribution, CASK 23E 22-group microscopic neutron cross sections 7, Se order of angular quad-rature, and P3 expansion of the scattering cross section matrix. Reactor con-ditions - power distribution, temperature, and pressure - were averaged over the irradiation period. A more detailed description of the calculational pro-cedure (except for capsule modeling) is presented in reference 8.
Because of computer storage limitations, two geometric models were required l
to cover the distance from the core to the primary shield. A boundary source output f rom model A (core to downcomer region) was used as input to model B (thermal' shield to primary shield), which included the capsule assembly.
In those cases where the capsule " shadowed" the maximum flux location in the pres-sure vessel, a model C (model B without a capsule assembly) was used to obtain l
vessel flux unperturbed by the presence of a capsule. For a reactor without surveillance capsules, an additional model A and model C were calculated.
In this way the effect of the specific power distribution in that reactor on ves-sel fluence was accounted for.
Thus, two sets of calculations were required -
I one to determine capsule fluence, which was based on Oconee 2 and CR-3 operat-ing conditions, and a second set to determine vessel fluence, which was based on Cconee 2 operating conditions for cycles 2 through 4.
Flux output from the DOT 3.5 calculations required only an axial distribution correction to provide absolute values.6 An axial shape factor (local: average axial flux ratia) was obtained from fuel burnup distributions in the peripheral fuel ossemblies nearest the capsule location. This procedure assumes that the axial fast flux shape in the capsule and the pressure vessel is the same as the axial power distribution in the closest fuel assembly. This is considered I
D-2 Babcock & Wilcox
i I
l to be a conservative assumption in 177-FA reactor geometry because the axial i
i, shape should tend to flatten as distance from the core increases. This fac-i tor was 1.10 averaged over an elevation corresponding to the capsule length j
in CR-3 and applied to the surveillance capsule; a maximum value of 1.17 was j
applied to the pressure vessel in Oconec 2.
I The calculation described above for CR-3 provides the neutron flux as a func-I tion of energy at the capsule position. These calculated data are used in the following equations to obtain the activities used for comparison with the experimental values. The equation for the calculated activity D (in pCi/g) is as follows:
5 M
-A t
-A (T-T )
=f3.7x
~
)
D 10*
i E n j
1 n
j=1
{
N = Avogadro 's number, J
i A = atomic weight of tare,et material n, j
n f = either weight fraction of target isotope in nth material 4
or fission yield of desired isotope, group-averaged cross sections for material n (listed in o (E)
=
i Table E-3),
4(E) = group-averaged fluxes calculated by DOT 3.5 analysis',
JE F
f raction of full power during jth time interval, t
=
j j,
A
= decay constant of ith material, g
l t = interval of power history, i
j T = sum of total irradiation time, i.e.,
residual time in re-j actor, and wait time between reactor shutdown and counting, I
l T
= cumulative time from reactor startup to end of jth time
[t ig period,
i.e., T
=
k' 13 k=1 l
The normalizing constant C is obtained from the ratio of measured to calculated activities:
I j
D (measured) 15 C=D1 (calculated).
(D-2) jg i
With C specified, the neutron fluence greater than 1 Mev can be calculated
+
from i
i t
i D-3 Babcock & Wilcox i
i i
l l
l 15 Mev M
?(E > 1.0 Mev) =C
[
?(E)
[
Ft, (D-3) d E=1 j=>
where M is the number of irradiation time intervals; the other values are de-fined above. The normalization constant for the OCII-A capsule was determined to be 0.93 (Table D-1).
Although this normalization is strictly correct only at the capsule location, it was considered applicable to all locations in the host reactor, CR-3, and in the donor reactor, Oconee 2, because of the simi-larity between the reactors and calculational models (B&W 177-FA reactors have essentially the same configurations and materials.)
In the calculational model, the pressure vessel and the capsule are separated by only 15 cm of water, and I
it is very unlikely that any significant change in accuracy would occur over that distance.
2.
Vessel Fluence Ext rapolat ion For current operation, fluence values in the precsure vessel are calculated I
as described above.
Extrapolation to future operation is required to predict vessel life based on minimum upper shelf energy and for calculation of pres-sure-temperature operation curves. Three time periods are considered: (1) to-date operation for which vessel fluence has been calculated, (2) designed future fuel cycles for which PDQ criticality calculations have been perform (d for fuel management analysis of reload cores, and (3) future fuel cycles for I
which no analyses exist.
Data from time period I are extrapolated through time period 2 hased on the premise that ex-core flux is proportional to the fast flux that escapes the core boundary.
Thus, for the vessel,
&"'X 4'v,x. c v,R g
I e,R where the subscripts are defined as v = vessel, e = core escape, R = reference cycle, and x = a future fue' cycle. Core escape flux is available from PDQ output.
Extrapolation from time periods 2 through 3 is based on the last fuel cycle in 2 having the same relative power distribution as an " equilibrium" I
cycle. Generally, the designed fuel cycles include several cycles into the l
future. Therefore, the last cycle in time period 2 should be representative of an " equilibrium" cycle. Data for Oconee 2 are listed in Table D-2, 1
D-4 Babcock & Wilcox
I
!I 4
- l l
{
This procedure is considered preferable to the alternat ive of assuming that lifetime fluence is based on a single, hypothetical " equilibrium" fuel cycle 1
l because it accounts f or all known power distribut ions.
In addition, it re-doces errorn that may result f rom the selection of a hypothetical " equilibrium" l
cycle.
1 i
l I
l 1
,4 i
j 1
I t
i i
1 i
i
}
1 l
1 4
I i
4 i
I i
1 i
l i
.1 l
D-5 Babcock s. Wilcox I
M M
M M
M M
E i
l Table D-1.
Capsule Normalization Constant Measured act ivity, t.Ci/c (
A B
i At Cycles CR-3 calculated C = A/B f
Cycle 1, 1B and 2, irradiation
- activity, normalization Reaction OC-2 CR-3 only uCi/c cons t ant (c) l l
5"Fe(n,p)S"Mn 1.82(+1) 9.52(+2) 9.34(+2) 1.16(+3) 0.80 l
5'Ni(n.p)5'Co 7
(-4) 1.94(+3) 1.94(+3) 2.49(+3) 0.78 2se U(n,f)137Cs 1.1 4.45 3.35 3.24 1.03 l
j Np(n,f)l37Cs 6.1 2.42(+1) 1.81(+1) 2.09(+1) 0.87 l
277 U(n,f)IU3Ru 1
(-9) 1.17(+2) 1.17(+2) 1.41(+2) 0.84 23e I
T 23'U(n,f)1 'Ru 9
(-1) 2.69(+1) 2.6 (+1) 2.87(+1) 0.91 i
i i
j Np(n,f)1U'Ru 6.2 1.27(+2) 1.21(+2) 1.27(+2) 0.96 237 2 3 ' U(n, f) l"'* Ce 6
(-1) 5.8 (+1) 5.74(+1) 5.67(+1) 1.01 l
237 Np (n, f) 1""Ce 2.6 2.81(+2) 2.78(+2) 3.09(+2) 0.90 l
23sU(n f)S5Zr 7
(-6) 1.04(+2) 1.04(+2) 1.08(+2) 0.96 l
1 237 U(n,f)'5Zr 4
(-5) 6.05(+2) 6.05(+2) 7.13(+2) 0.85 1
W (a) Average of four dosimeter wires from Table E-2.
nicr O
Obtained from A = A2 -Ae where A is the decay constant for the product isotope and l
t is calendar time from EOC-1 in Oconee 2 to EOC-2 in Crystal River 3 (1420 days).
p(
(' Value for irradiation in Crystal River 3 only.
Average of all fission reactions (0.93) i was selected as the normalization constant.
)
I
M M
M M
Table D-2.
Extrapolation of Pressure Vessel Fluence
"""C C"
Cumul.
Core escape
- Time, time, Vessel, flux, Time Cycle pux,n/cm'-s EFPY EFPY n/cm'-s interval Cumulative 1
0.482 (+14) 1.20 1.20 1.39 (+10) 5.28 (+17) 5.28 (+17) 0.560 (+14)(^)
0.76' 2
3 0.612 (+14)(b) 0.79 3.73 1.57 (+10) 1.25 (+18) 1.77 (+18) 4 0'.587 (+14) 0.97; 5
0.428 1.07 4.79 1.14 (+10) 3.84 (+17) 2.16 (+18) 6 0.416 1.10 4.89 1.11 (+10) 3.84 (+17) 2.54 (+!8) 7 0.425 1.15 7.04 1.13 (+10) 4.13 (+17) 2.96 (+18) 8 0.427 1.15 F.19 1.14 (+10)(')
4.14 (+17) 3.37 (+18)
>8 0.427 6.80 15 1.14 (+10)(d) 2.46 (+18) 5.83 (+18)
[
>15 0.427 17 32 1.14 (+10)( }
6.15 (+18) 1.20 (+19)
(^ Weighted.
} Avg = 0.587(+14).
(c)Value from 1
10 (1.57 x 10 ) = 1.14 x 10 x
8
(
Cycle 8 assumed to be equilibrium cycle for future operation.
5 1
K 1
1 g
1 x
P
._w X
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I I
APPENDIX E Capsule Dosimetry Data I
I i
I I
I l
l E-1 Babcock & Wilcox
Table E-1 lists the composit ton of the threshold detectors and the cadmiura thicknesses used to reduce competing thermal reactions. Table E-2 shows the capsule OCII-A measured activity per gram of target material (i.e., per gram of uranium, nickel, etc.) corrected for the wait time between irradiation and counting. Activation cross sections for the various materials were flux-23s weighted with a U fission spectrum (Table E-3).
Table E-1.
Detector Composition and Shielding Monitors Shielding Reaction 23e 10.38% U-Al Cd-Ag 0.02676" Cd U(n,f) 1.44% Np-Al Cd-Ag 0.02676" Cd 237 g f)
Ni 100%
Cd-Ag 0.02676" Cd Ni(n.p)S8Co se 0.56% Co-Al Cd-0.040" Cd Co(n,y)Co s9 0.56% Co-Al None 5'Co(n,y)6"Co Fe 100%
None 5"Fe(n p) s"Mn I
I l
l E-2 Babcock & Wilcox
b Tabic E-2.
Capsule OCII-A Dosimeter Activity Measurements Post-irr Nuclide Specific
- Activity, Dostneter
- weight, Radio-
- activity, activity, uCi/g of material
_ e Reaction nuclide
.Ci LCi/v target Dosineter AD-i f
l Co-Al (bare) 0.0154 5'Co(n,3)
Co 20,31 1320 236,000 Co-Al (Cd) 0.0128 5 Co i n, y )
' Co 14.08 1100 196,000 Ni G.1309 5'Ni(n p) 5'Co 151.1 1150 1,700
Ni(n,p)
Co 0.3732 2.85 10.9 0.1513 5'Fe(n.p) 5"Mn 7.5J6 49.7 855 5"Fe(n,y) 5'Fe 22.94 152 45,900
'"i-Al 0.0329
*U(n,F)
'5.r 0.3094 9.40 91.3 7
IU'Ru 0.3003 9.13 88.6 "2u 0.0841 2.56 24.8 l'7 l
Cs 0.01506 0.458 4.44 1'Ce 0.1773 5.39 52.3 i
Np-Al 0.0291 Np(n,F)
Zr 0.2321 7.98 554 23
'5 IRu d
1 "Ru 0.0408 1.40 97.4 l'7 Cs 0.009562 0.329 22.8 1"'Ce 0.1061 3.65 253 i
l Dos imet e r AD-2 Co-Al (bare) 0.0145 5'Co(n,y)
Co 21.07 1453 259,000 Co-Al (Cd) 0.0145 5'Co(n,y)
Co 17.13 1180 211,000 l
Ni 0.1330 5*Ni(n.p) 5'Co 177.0 1331 1,960 l
Ni(n.p)
Co 0.3748 2.82 10.8
'0 Fe 0.1543 5'Fe(n.p) 5'Mn 8.754 56.7 975
' Fe(n 3) 5'Fe 28.11 182 55,200 1
l 2 " U-Al 0.0433 2 "U(n.F)
Zr 0.5158 11.9 116
'5 1U'Ru 0.6173 14.3 138 l
1 'Ru 0.118 2.72 26.5 l
lCs 0.01957 0.452 4.39 l
I"'Ce 0.2738 6.32 61.4 2Np-Al 0.0239 Np(n.F)
Zr 0.2054 8.59 597 2'7
'5 1Ci Ru i
1"'Ru 0.0534 2.23 155 lCs 0.007954 0.333 23.1 (I
lCe 0.09692 4.06 282 I
I E-3 Babcock & Wilcox
4 1
1 i
I Table E-2.
(Cont'd) l 4
l Post-irr Nuclide Specific Ac t ivit y.
Dosimeter
- weight, Radio-
- activity, activity, LCi/g of q
material a
Reaction nuclide LCi uCi/g target i
j Dosimeter AD-3 l
Co-Al (bare) 0.0153
Co(n.,)
Co 20.49 1340 239,000 I
Co-Al (Cd) 0.0133 5'Co(n,3)
Co 14.86 1120 200,000 Ni 0.1341 "Ni(n,p)
" Co 150.9 1120 1,660
'0
'2 Ni(n p)
Co 0.3661 2.73 10.4 i
l Fe 0.15'05 5'Fe(n,p) 5"Mn 7.146 47.5 816 1
' Fe ( n,3 )
Fe 24.55 163 49,400 53 2"U-Al 0.0364 23'U(n,F)
'Zr 0.3420 9.40 91.2 103 1
Ru 0.4051 11.1 108
(
10'Ru 0.0850 2.34 22.7 l
13'Cs 0.01483 0.407 3.96 i
1"*Ce 0.1844 5.07 49.2 1
23'Np-Al 0.0485 23'Np(n,F)
Zr 0.3481 7.18 498
'5 1
103 Ru l
1"Ru 0.0706 1.46 101 137 Cs 0.01503 0.3'O 21.5 lCe 0.1645 3.39 236 1
Dosimeter AD-4 Co-Al (bare) 0.0162 5'Co(n,y)
Co 29.27 1810 323,000
'O Co-Al (Cd) 0.0133 5'Co(n,y)
Co 20.50 1540 275,000 i
f Ni 0.1342 5'Ni(n p)
" Co 222.3 1660 2,440
'0
'0 j
Ni(n.p)
Co 0.4599 3.4, 13.1 l
l Fe 0.1520 5"Fe(n.p) 5'Mn 10.27 67.6 1,160 j
'Fe(n,3) 5'Fe 34.66 228 69,100^
23'U-Al 0.0349 23'U(n,F)
'5 Zr 0.4271 12.2 119
'3 Ru 0.4849 13.9 135 1 " Ru 0.121 3.47 33.7 137 Cs 0.01803 0.517 5.02
""Ce 0.2485 7.12 69.1 i
1
'Np-Al 0.0486 2Np(n.F)
Zr 0.5393 11.1 771
'5
(
l03 Ru l
106 Ru 0.109 2.24 156 137 Cs 0.02065 0.425 29.5 l'"Ce 0.2475 5.09 354 I
E-4 Babcock & Wilcox
l i
l l
l l
Table E-3.
Dosimeter Activation Cross Scctions #
i ss sections, Watom Energy range, 237 23e 58 c
Mev 3p g
31 s i. pc 1
12.2 - 15 2.323 1.050 4.830(-1) 4.133(-1) 2 10.0 - 12.2 2.341 9.851(-1) 5.735(-1) 4.728(-1) 3 8.18 - 10.0 2.309 9.935(-1) 5.981(-1) 4.772(-1) 4 6.36 - 8.18 2.093 9.110(-1) 5.921(-1) 4.714(-1) 5 4.96 - 6.36 1.541 5.777(-1) 5.223(-1) 4.321(-1) 6 4.06 - 4.96 1.532 5.454(-1) 4.146(-1) 3.275(-1) l 7
3.01 - 4.06 1.614 5.340(-1) 2.701(-1) 2.193(-1) 8 2.46 - 3.01 1.689 5.272(-1) 1.445(-1) 1.080(-1) 9 2.35 - 2.46 1.695 5.298(-1) 9.154(-2) 5.613(-2) 10 1.83 - 2.35 1.677 5.313(-1) 4.856(-2) 2.940(-2) 11 1.11 - 1.83 1.596 2.608(-1) 1.180(-2) 2.948(-3) 12 0.55 - 1.11 1.241 9.845(-3) 6.770(-4) 6.999(-5) 13 0.111 - 0.55 2.34(-1) 2.432(-4) 1.17 t. (-6) 1.578(-8) 14 0.0033 - 0.111 6.928(-3) 3.616(-5) 1.023(-7) 1.389(-9) 1
(
ENDF/B5 values that have been flux-weighted (over CASK energy groups) 235
]
based on a U fission spectrum in the fast energy range plus a 1/E j
shape in the intermediate energy range.
i 1
4 i
l 1
iI I
i i
l 1'
E-5 Babcock & Wilcox l
1 4
t i
j a
i N
i 1
i i
i i
j j
i l
i 1
4 i
l i
I I
1 1
i a
1 i
APPENDIX F References i
J 1,
i i
i 1,
4 i
r I
I 1
l l
l l
F-1 babcock & Wilcox
i I
l H.
S.
Palme, C. S. Carter, and C. L. Whitmarsh, Reactor Vessel Material 1
i l
Surveillance Program - Compliance with 10 CFR 50, Appendix II, for Oconee-Class Reactors, BAW-10100A, Babcock & Wilcox, Lynchburg, Virginia, February l
1975.
l 2
A.
L. Lowe. J r., et al., Analysis of Capsule from Duke Power Company Oconee l
l Nuclear Stat ion, Unit 2, Reactor Vessel Material Surveillance Program, BAW-1437, Babcock & Wilcox, Lynchburg, Virginia, May 1977.
3 G.
J. Snyder and G.
S.
Carter, Reactor Vessel Material Surveillance Program, BAW-10006A, Rev. 3, Babcock 6 Wilcox, Lynchburg, Virginia, January 1975.
" 11. S. Palme and 11. W.
Behnke, Methods of Compliance With Fracture Toughness and Operational Requirements of Appendix G to 10 CFR 50, BAW-10046P, Babcock 6 Wilcox, Lynchburg, Virginia, October 1975.
5 l
11.
S.
Palme, G.
S. Carter, and C.
L. Whitmarsh, Reactor Vessel Material l
Surveillance Program - Compliance With 10 CFR 50, Appendix 11, for Oconee-l t
Class Reactors, BAW-10100A, Babcock & Wilcox, Lynchburg, Virginia, February l
1975.
' DOT 3.5 - Two-Dimensional Discrete Ordinates Radiation Transport Code (CCC-l 2 76), WANL-TME-1982, Oak Ridge National Laboratory, December 1969.
l 1
7 j
CASK Group Coupled Neutron and Gamma-Ray Cross Sect ion Data, DLC-2 3E, l
Radiation Shielding Information Center.
l 8
C. L. Whitmarsh, Pressure Vessel Fluence Analysis, BAW-1485, Babcock &
Wilcox, Lynchburg, Virginia, June 1978.
s i
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l l
F-2 Babcock & Wilcox
.