ML20040B071

Analysis of Capsule OCII-A from Duke Power Co Oconee Nuclear Station,Unit 2,Reactor Vessel Matl Surveillance Program.

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: ML15223A765 ML15223A767 ML20040B068 ML20040B071
References
BAW-1699, NUDOCS 8201250101
Download: ML20040B071 (100)

Text

- _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _______

1 BAW-1699 December 1981 4

II it i

i 1

ANALYSIS OF CAPSULE OCII-A FROM l DUKE POWER COMPANY'S llW OCONEE NUCLEAR STATION, UNIT 2 I,

lg - Reactor Vessel Material Surveillance Program -

ig

}

i i

!B lI l3 ll l

B l 9201250101 920112 Babcock &Wilcox e uto. -on com,. ,

PDR ADOCK 05000269 lI 1

P PDR l _ _ _ . . , _ _ . _ , - _ - _ _ _

BAW- 16'39 December 1981 4

i lI ANALYSIS OF CAPSULE OCII-A FROM DUKE POWER COMPANY'S lI OCONEE NUCLEAR STATION, UNIT 2

- Reactor Vessel Material Surveillance Program -

i i

by I A. L. Lowe, J r. , PE J. W. Ewing

. g W. A. Pavinich g W. L. Redd

. J. K. Schmatzer i

il J

I

!I i

i II I

l

)

i lI i

I B&W Contract No. 582-7109-71 1

BABCOCK 6 WILCOX

{g j Nuclear Power Group l3 Nuclear Power Generation Division l P. C. Box 1260 l Lynchburg, Virginia 24505 i

i Babcock & Wilcox

t

i l

l

!I SU}B1ARY lI j

4 This report describes the results of the examination of the second capsule of Duke Pcwer Company's Oconee Unit 2 reactor vessel surveillance program. The I capsule was removed and examined after accumulating a fluence of 3.37 x 10 18 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 requirements of Appendix H to 10 CFR 501 and ASTM specification E185-73.

i 1

j The capsule received an average fast fluence of 3.37 x 101 " n/cm2 (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 17 n/cm2 (E > 1 Mev). Based on the calculated 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 40 calendar years of operation is 1.20 x 101 ' n/cm2 (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 ET and the decrease in upper shelf properties due to irradiation are jE 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.

lI

!l lg Babcock & Wilcox

!R

1 l

i i

lI i

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.

lI l

I The surveillance program for Oconee 2 was designed in accordance with E185-66 and thus does not comply with Appendixes C and 11 to 10 CFR 50 since the re-1 l

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.

I i

1 4 l i

l 1

1 lI l

l l

Babcock & Wilcox lI l-1

8 I

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 I pressure-temperature limitations on operation of the RCPB. The toughness and 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

2-1 Babcock 8. Wilcox

( ..

l

}

5 i

I irradiation and the thermal environment. Fracture toughness test data are ob-tained from material specimens withdrawn periodically from the reactor vessel.

l These data will permit determination of the conditions under which the vessel can be operated with adequate safety margins against fracture throughout its jI i

service life, lg A method for guarding against brittle fracture in reactor pressure vessels is IE 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 g i material to a reference stross intensity factor curve (ItIR "" ' "P~

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 g curve, allawable stress intensity factors can be obtained for this ma-terial as a function of temperature. Allowable operating limits can then be E

determined using these allowable st ress intensity f actors.

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. The adjusted RT * """ " ' * "

NDT 1R """" " '

I 1

in turn, is used to set operating limits for the nuclear power plant. These i

l ,ew limits take into account the effects of irradiation on the reactor vessel materials.

I 1 l

!I 4

l 2-2 Babcock 8. Wilcox

!I i

1 ___ __ _ _ , _ _ _ - __ _ . _ . _ _ _ __ - _____

I I

f I

i

3. 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

{ 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 18 nyt. This capsule contained Charpy 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

!E lg

=aterial are described in Table 3-2.

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-1 3 tion Engineering. The chemistry and heat treatment of the cc,rrelation material 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 _ . . _ _ _ _ __ _ _ _ . _

I lI lI l principal working direction. Capsule OCII-A contained dosimeter wires, de-scribed as follows:

I l Dosimeter wire Shielding I

j U-Al alloy Cd-Ag alloy j Np-Al alloy Cd-Ag alloy I Nickel Cd-Ag alloy 1

i j 0. 66% Co-Al alloy Cd 0.66% Co-Al alloy None l

l 4

Fe None Thermal monitors of low-ruelting eutectic alloys and pure metals were included 4

in the capsule. The eutectic alloys and metals and their melting points are )

i as follows:

4 l

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 l

• Table 3-1. Specimens in Surveillance Capsule OCII-A 1

No. of specimens Material description Tensile Charpy l

Weld metal, WF-209-1A 4 8 lient-af fected zone "A" (HAZ),

I He-t AAW-163, Longitudinal 0 8 Base metal material, plate "A,"

j Heat AAW-163, Longitudinal 4 8 l Transverse 0 4 W Correlation, HSST, plate 02 Heat A-Il95-1 0 8 Total per capsule 8 36 E

3-2 Babcock & Wilcox l

l

it i

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 Mn P

0.24 0.63 0.006 0.057 1,58 0.020 I S Si Ni 0.012 0.25 0.75 0.005 0.56 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

il i

i II 1

I I

il I l

1 .

l l

l 3-3 Babcock 8. Wilcox i____________________ _ _ _ _ _ _ _ _ _ . . _ _ _ . _ _ . . . _ _ _ _ . . _ _ _ . . _ _ . _ . . _ _ _ _ _

i II I

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

}

i Cu 0.17

}

{

i; eat Trea t men t (b )

i f 1. 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 i

(b)Per plar e section identification ca,rd.

i il i

I

!I i

lI lI i

l it i

i I 3-4 Babcock & Wilcox

!,I .

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' ~~./

L 1, /- ,

,n s ,

. . s -

/ /

e i e o e s I , , , s e e I e e ',

e e e e e y E

w.

y& , _e.

ee g

! e

,____.g.

e e , n_L

'/-)y f'

e i e e

^

\ e . e i o e ~

. . i .. .~ ~;

> e se 'l ,/ '

..I ' s s e o I e e /

s . /

, i s

o .

) ,/lI/ /,\ *

!I /

{

</

// '. Surveillance Capsule j / h Holder Tube -- Cap-sules OC11-A, OCIl-B g l .

Surveillance Capoule lloider Tube - Capsules Z ,

OC11-E, OCII-F

'I I l ll iE 3-5 Babcock & Wilcox

I I 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

' I

,I '

x f i . . . . s e e

/ l is g i I et

• e e _

i r le I le e le el

• ("d z s;i

.i ielei

.e i

+c

. ei .e c

I , )j 3 j Y j% s .

l i le! e e _ s IRRADIATION elet le e e -

SITE OCONEE ej e e e e CAPSULE e o e ~ / O Cl l- A

\ e e e e

'I eQ. y , jl I -

/

, [y .' 7 - __

CAPSULE HOLDER CAPSULF HOLDER TUBE I

I I

I 3-6 Babcock & Wilcox

!I i

i lI i

I 1

1

4. PREIRRADIATION TESTS

!I l

Unirradiated material was evaluated for tm purposes: (1) to establish a base-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 l I properties as required for compliance with Appendixes G and H to 10 CFR 50. l

I l

l 4.1. Tensile Tests b l 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 I

ture and 580F. The tension-coupression load cell used had a certified accuracy 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

g meet Watertown standards. Test specimens were of the Charpy V-notch type, j which were nominally 0.394 inch square and 2.165 inches long.

j

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

! 4-1 Babcock & Wilcox l

--,m - - - - - - - . . - - - - . - - . _ . _ . , - , _ , _ . ,

___,,e,n,-n . .

1

'I I 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.

I II I

I I

I

'I I

i I

I I

il I

I 4-2 Babcock & Wilcox I

\ _

l_.

1 lE 1

1 l

i I

j 5. POSTIRRADIATION TESTS i 5.1. Thermal Monitors 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 I lead factor in the Crystal River Unit 3 radiation site would cause a slight 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.

I i

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

a l

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.

} The results of the preirradiation Charpy V-notch impact test are given in Appendix C.

l

}

!I

Table 5-1. Irradiation Tensile Preperties of Capsule OCII-A Base l Metal and Weld Metal Irradiated to 3.37 x 10 18 nyt l (E 1.0 Mev)

Red'n

I f

Specimen No.

Test temp, F Yield eng , psi Ult.

Elongat ion, %

Unif Total in area, 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

1 i 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

] Std dev'n 70 1,270 0.07 3.04 0.64 lI J

i 5-2 Babcock & Wilcox 4

, -- - . - - - - m -- , ,,

t I

I l Table 5-2. Irradiated Charpy Impact Data for Capsule OCII-A i Irradiated to 3.37 > 10 18 nyt, Base Metal IIAZ, i Longitudinal Orientation 1

1 I

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 51.0 117.0 40.0 77.0 70 50 l- EE-439 75 EE-419 140 127.5 94.0 100

EE-436 224 139.0 93.0 100 EE-424 440 115.0 83.5 100

>I f '

i Table 5-3. ' Irradiated Charpy Impact Data for Capsule OCII-A Irradiated to 3.37 x 10 18 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 I I

I 5-3 Babcock & Wilcox

i I

4 l

i Table 5-4. Irradiated Charpy impact Data for Capsule OCII-A l Irradiated to 3.37 x 10 18 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 ig EE-737 0 41.5 29.0 0 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

i I

l Table 5-5. Irradiated Charpy Inpact Data for Capsule GCII-i.

,} Irradiated to 3.37 x 10 a nst , Weld Metal 1

)

) Absorbed Lateral Shear j Specimen Test temp, energy, expansion, fracture,

) No. F ft-lb 10-3 in.  % i l

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 1  !

j 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

j ' EE-0 33 408 44.5 44.0 100

I lI l

l

' l 5-4 Babcock & Wilcox

I

l l

j table 5-6. Irradiated Charpy Impact Data for Capsule OCII-A Irradiated to 3.37 x 10 18 nvt, Correlation Material iI j Absorbed Lateral Shear

Specimen Test temp, energy, e xpa ns ion , fracture, l No. F f t-lb 10-3 in.  %

EE-930 12.5 9.5 I

72 2 EE-942 110 25.5 21.0 30 l

i EE-903 120 33.0 29.5 15 EE-90 '4 130 27.0 32.0 30 l EE-934 140 62.0 52.5 35 EE-919 204 84.5 64.0 100 EE-914 318 103.0 E4.0 100 EE-935 440 102.5 77.5 100 lI 1

!I

!I 1

lI 1

I I

i lI 1

i 5-5 Babcock & Wilcox

!I

Figure 5-1. Irapact Data From Irradiated Base Metal l (AAW-163) IIAZ, Longitudinal Orientation 100 [ i g j , g g  ;  ; i F

! " 75- *

)

J 3

C ------- - - - - - - - - - - - - - - - - - - - - - - - - - - - -

. 50 l

4  :

d 25-

• 1 i

' I  ! ' I I '  !

0 'F

$ 94 h 93 ,h

.08 g  ;  ; , g g j g l .

, g I 5 .06- -

! 5 O .04 5 .___________________________] .

l y.02 - -

l o

' I I I I I I I I I I I 0 li-200 i i ,' F

j g g  ;  ; g  ;

! DATA SUMFARY N.A.

180" T,,37 _

] Tcy(35 mtE) _l"D-Tcy(50 n -ts) '"D-150 -

IND.

1 Trv (30 FT-ts)

Cy-USE (avc.) 138 FT-Le l

i . 140 RT~*'

l $

f l,

5120-I j

l 100~ -

l

E 1 i

) 7 i - 80- -

I C l

r I

~ 60 - -

)

l

.____._-._n_______-._______________--

~

W MATER ut SA508.CL2 l __.,-______ _..___ _-___ .

20-I FLUENCE 3.37E18 nvt HEAT No. A A w-16 3 I I I I I I I l i i i 0

-80 -40 0 40 80 12C 160 200 240 280 320 360 Te Test TEMPERATURE, F 3-6 Babcock 8. Wilcox

1 1

4 i l Figure 5-2. Impact Data From Irradiated Base Metal )

(AAW-163), Transverse Orientation l l 100 i i j j i j g g l [ l ,

l 71s

! 3

g w ______________________________

l J 25 - -

! g l / 1 1 I I I I I I I

.08 7 l . I i  ! I I I I I I I .i i f l 5 .06 i 1

! t i i w ,% _ e _

3 .____ ______________________.____

1 5 l # .02 - -

. 5

) !

0 I I I  ! I I I I I I I

! # ' ' i I I I i I I l I

= DATA SUTARY

"*A-180" T,g, - l i

i I 160 - CV Tcy(35 mtt)

T (50 rf-ts) + 13 F

+5F Tcy(30 FT-ts) ~8'F

]l l

{W . 140 Cy -USE (avr.)

gr , -g7F

'"0" 0

$120- -

3100" -

I 5 1

1, 80- -

C.

i 1I

~ 00-0-

l MA TERI AL 8' 8 ORIENTATION TRANSVERSE ~~

20~ FLutNet 3.37 slo'8 avt -

HEAT No. AAW-163 0

I I I I I I I l ' ' '

-00 -40 0 40 80 120 160 200 240 280 320 360 400 I

itsT TamerRafuRa, F

, 5_7 Babcock & Wilcox n

I,I 4

lI Figure 5- 3. Impact Data From Irradiated Base Metal (AAW-163) , Longitudinal Orientat ion

~ -

100 g l [ g l l , , ,

l l " 75 -

s I O - - - - _ _ _ _ _ - - - - - - - - - -

g 50 - - _ _ - _

i w M

l 0

4 r

d 25- -

0-

.08 j g  ; , g , i g , ,

.!, f.

) $ .06 -

7 I

2

" 04- .

3 .____

=

,Eir 4 .02 -

5 I I I I I I I I I I I 0

2 ' ' l l 3

I i l I I l l DATA

SUMMARY

4

"*A*

! 180 - T,gy -

Tcy(35 att) -*F l

l 160- T CV (50 n-ts) -6F -

Tcy(30 n-a) -30F l'

! Cy -USE (Avo) 135 FT-La

.s 140 RT -54F -

NDT .

b lI

$120- -

1 >

O 3100- -

i i 5 80- -

0 I

r i 60 -

d

~

MATE R I AL SA508,CL2 R ENTAfl0N LONGlTUDINAT."

l 20- Ftutscg 3.37El8 nvt -

Huf No, A AW-16 3

! I I I I I I I I ' ' '

O 1 -C0 -40 0 40 80 120 160 200 240 280 320 360 400 I Test TemptnATuat, F 5-8 Babcock & Wilcox I

I

_________m_ _ _ _. - - _ . , .,

~

i f .

il i

l Figure 5-4 Impact Data From Irradiated Weld Metal (W F-209- 1 )

t

! IN I I -~ l ~ - l ~~ I i i e i ,

1

! - 7s - _

1 i J

?

yw ---_________________________-_

l 1 w l a 25 -

l I I I I I I I I I I I

! n i

i . ' 08 i i i I i l I i i l I E

e .06-1  ;

} * '

I O .04- ___________.___._.______-,

1 5 ----_________

l,

,E i

i .02-a I I I I I I I I I I I 0

1* '

I

' ' l i I I I I l 1 DATA SUPPARY 90- T,o, "*A- -

Tcy(35 mtt) +126F j

, 80 Tcv (50 rt-ts) _ino _

! Tcy(30 FT-ts) +118F Cy -U5E (ava) __38 FT-Las I

i I C y

70 RT Not 66F -

ll}g i ( 0,-

I  :

i j

j h50_____________________..___________

L w

d 40.- -

l

! C 1

• 30------~~~-* - - - - - - _ - - - - - - - _ - - - -

i l

4 20_

• WELD METAL j . P27ge at i -

j Oettatation 10m coutNsr 3.37E18 nvt -

! HEAT No. wF-209-l j o I i 1 L I i 1 1 i i l

-C0 -4') 0 40 80 120 160 200 240 28J 320 3f4 400 Test Trmetaaruar, F l

5-9 Babcock & Wilcox i

I

l i

I 4

i i

J

' Figure 5-5. Impact Data From Irradiated Correlation Material I (ilSST Plate 02)

I W I I I I I I i i i i 'h f

!g " 75 -

1 J i 5 i n - - - - - - - - - - - - - - - - - -

i

~

50 - - - - - - - - - - - - - - - - - -

I i - . .

! d 25 -

i

• l I I I I I

] 0 I I I I I

'F

.08 j j 1

; y

!E . i ,  ; g g I

g I

' N.0f -

l 4 -

)' "' 04-

3. -- - ______

l 5 i

j i .02 -

]

=

1

\

0 I I I '

I I I I I I I I j 0 'f-2 ' ' ' I I i i l i T I 'h DATA SIMARY 4

i 180- T ,ny N.A. .

I cy(35 MLE) 124F

) .

160- cy T (50 FT-ts) 137F -

l' Tcy(?O FT-ts) Il9F l

4 Cy -USE (Avo) 103 FT-L85 i . 140 RT 77F j y NDT E

.$120-1 E I 3 ,,

'C lg 3100-15 a J

j i 80 - _

l C 1 E T

60 m l - --__.---___---------

40-(-------__-- -

i MATERIAL HSST-PLO2 OattnTAttom L0h G I TU D I N AL'~

FLutset 3.37EI8 nvt j 20-l Hui No. A- 1195-1 I I I I I I I I ' ' ' -

' 0 ' 1

-CD -40 0 40 80 120 160 200 240 280 320 360 440 itsi TEmprRATunt, F 1

i 1 i

5-10 Babcock & Wilcox

I II i

i i

J l

6. NEUTRON DOSIMETRY I

l i

i 6.1. Backgro und l

I As a part of the pressure vessel surveillance program, fluence analysis has three primary objectives: (1) decermination of maximum ilucnce at the pres-

, sure vessel as a function of reactor operation, (2) prediction of pressure 3 vessel fluence in the future, and (3) determination of the test specimen j fluence within the surveillance capsule. Vessel fluence data are used to

} evaluate changes in reference transition temperature and upper shelf energy i

! levels and to establish pressure-temperature operation curves. Test speci-1 men fluence data are used to establish the correlation between changes in material properties and fluence exposure. To provide this information, a model for calculating flux distributions in the reactor is established. The

]

)

i accuracy of calculated fast flux is enhanced by the use of a normalization f actor which utilizes measured activity data obtained from capsule dosimeters.

A significant aspect of the surveillance program is to provide a correlation J

) between the neutron fluence above 1 Mev and the radiation-induccd property changes noted in the surveillance specimens. To permit such a correlation, I activation detectors with reaction thresholds in the energy range of interest were placed in each surveillance capsule. The significant properties of the j detectors are given in Table 6-1.

1 Because of a long half-life (30 years) and an effective threshold energy of 0.5 Mev, the measurements of 137 Cs production from fission reactions in 237 2 38 Np (and U) are more directly applicable to analytical determinations of the f ast neutron fluence (E > 1 Mev) for ru1tiple fuel cycles than the other dosimeter reactions. The other dosimeter reactions are useful as cor-roborating data for shorter time intervals and/or higher energy fluxes. Short-lived isotope activities are representative of reactor conditions only over the latter portion of the irradiation period (fuel cycle), whereas reactions with j a thresbald energy higher than 2 or 3 Mev do not record a significant part of the total fast flux.

i 6-1 Babcock & Wilcox i

1 I

J i

!I

{ The cuergy-dependent neutron flux is not directly available from activation detectors because the dosimeters register only the integrated ef fect of the i neutron flux on the target mat erial as a function of both irradiation time

{ and neut ron energy. To obtain an accurate estimate of the average neutron 1

l flux incident upon the detector, sevecal parameters must be known: the oper- ,

I j ating history of the reactor, the energy response of the given detector, and l the neutron spectrum at the detector location. Of these parameters, the definition of the neutron spectrum is the most difficult to obtain. Essen-

): tially two means are available to obtain it: iterative unfolding of experi-I mental dosimeter data and/or analytical methods. Because of a lack of suf fi-1 I c ie nt threshold reaction detectors that satisfy both the threshold energy and half-life requirements of a surveillance program, calculated spectra are used in this analysis.

Neutron transport calculations in two-dimensional geometry are used to calcu-late energy-dependent flux distributions throughout the reactor. Reactor con-l ditions are selected to be representative of an average over the irradiation time period. Geometric details are selected to explicitly represent the sur-veillance capsule assembly and the pressure vessel. The detailed calculational l procedure is described in Appendix D.

6.2. Vessel Fluence l The maximum fluence (E > 1.0 Mev) in the Oconee 2 pressure vessel through cy-cle 4 was determined to be 1.77 (+18) n /cm2based on an average neutron flux of 1.57 (+10) n /cm2 -s for cycles 2, 3, and 4 and 1.39 (+10) n/cm 2 -s for cycle 1 (Tables 6-2 and 6-3). The location o f maximtun f.l uence is a point at the

. cladding / vessel interface at an elevation about 100 cm above the lower active

{ fuel boundary and an azimuthal (pe riphe ral) location of about 12* from a major 1

j axis (across flats diameter). -Pluence data have been extrapolated to 32 EFPY

( of operation based on the premise that ex-core flux ic proportlanal to fast flux that escapes the reactor core (Appendix D). Core escape flux values are available f rom fuel management analyses of future fuel cycles.

Relative fluence as a function of radial location in the pressure vessel is shown in Figure 6-1. Corresponding Icad factors from the cladding interface to T/4, T/2, and 3T/4 are 1.8, 3.7, and 7.9, respectively. Relative fluence as a function of azimuthal angle is shown in Figure 6-2. A peak occurs at I about 12*, which roughly corresponds to a corner of the core and to three 6-2 Babcock & Wilcox

II

] symmetric capsule locations. Two other capsule locations correspond to the azimuthal minimum at about 26 However, it should be noted that the maximum:

i minimum flux ratio is only 1.3, and the data in Figure 6-2 do not account for i flux perturbation by the capsule itself. Fast neutron flux is increased by j approximately 1.25 in the capsule due to dif ferences in scattering and absorp- '

tion cross sections between sr. eel and water, i.e. , model with and without cap-sule.

l

', 6.3. Capsule Fluence I

i j Fluence at the center of the surveillance capsule was calculated to be 3.37 l (+18) n/cm2, about 28% of which was received in Oconee 2 and about 72% in Crystal River 3 (Table 6-4) . These data represent average values in the cap-l sule. Capsule OCII-A was located in an upper holder tube position 11 off axis and approximately 211 cm from the core center for 440 EFPD in Oconee Unit 1 2, cycle 1. (This corresponds to the original 177-fuel assembly holder tube design.) It was then inserted in Crystal River 3 in an upper holder tube po-lW sition 11* of f axis and about 202 cm from the core axis for an additional 338 4

j EFPD. During the latter irradiation period, the capsule was estimated to have been rotated 110 clockwise relative to its original design orientation (key l facing the reactor core).

!I i

1 t'

1 1

6 1

i \

il I

1 1

i I

I 1

i 6-3 Babcock & Wilcox il

. _ _ - - - . - - - - - - . . . - _ - - _ _ . m

i lI II Table 6-1. Surveillance Capsule Detectors l

Effective lower energy limit, Isotope ig lg Detector reaction Mev half-life

} 5"Fe(n p)5"Mn 2.5 312.5 d se Ni(n,p) seCo 70.85 d 2.3 l 238 U(n,f)l37Cs 1.1 30.03 yr l 237 Np(n,f)137Cs 30.03 yr 0.5 j

{

238 U(n,f)1c(Ru 1.1 368.2 d

- 237(Np(n,f)lU6Ru 0.5 369 d i

23a U(n,f)1c3Ru '

1.1 39.43 d 237 Np(n f)l 0 3Ru

0.5 39.43 d 23e U(n,f)l""Ce 1.1 284.4 d f

237 Np(n,f)l""Ce 0.5 284.4 d jh 2 38 ur U(n.f)'5Zr 1.1 64.1 d 237.3p g ,g)9sZr j 0.5 64.1 d

!l l

1 Table 6-2. Pressure Vessel Flux Fast flux, n /cm2 -s "*'

(E > 1 Mev) O v Inside surface Inside surface Cycle (max location) T/4 3T/4 (max location) f Cycle 1("), 1.39(+10) 7. 9 (+9) 1.8(+9) 2.74(+10)

(g 440 EFPD 1g Cycles 2,3,4, 1.57(+10) 8.6(+9) 2.0(+9) 3.34(+10) 921.2 EFPD

!E I

(a) Data from reference 1.

'I

,I 6-4 Babcock & Wilcox l I

t 1

l Table 6-3. Pressure Vessel Fluence Fast fluence, n/cm2 (E > 1 Mev)

Cumula t ive Inside surface )

irradiation time (max location) T/4 3T /4 l End of cycle 1( ) 5.28(+17) 2.9(+17) 6.6(+16)

(440 EFPD)

End of cycle 4 1.77(+18) 9.8(+17) 2. 3(+17) f (1361 EFPD) iI j 15 EFPY 5.83(+18) 3. 2 (+18) 7.3(+17) 1 32 EFPY 1.20(+19) 6.7(+18) 1.5(+18) 1

^ Data from reference 1.

I Table 6-4. Surveillance Capsule Fluence g

iE Flux, Cumulative Flux,

} n/cn -s 2

Fluence fluence, n/cm2 -s 2 2 (E > 1 Mev) n /cm n /cm (E > 0.1 Mev) oc 2, cycle 1 (") , 2.48(+10) 9.43(+17) 9. 43 (+17) 4.66(+10) 4 440 EFPD

!I j

CR-3, cycles IB and 2, 8.34(+10) 2. '.3 (+18) 3.37(+18) 1. 96 (+11 )

338 EFPD (a) Data f rom reference 1.

lI .

I I

'I 6-5 Babcock & Wilcox I

ll lI i

Figure 6-1. Radial Fluence Gradient Through Reactor Vessel

!I n

5 eW

. 5 E E '

l E E 4

e i

g W T, 4 REACTOR VESSEL j BASE METAL 10 Ill b

=

l l ^

' ~

!' T/2 3

5 b

4  %.

a

i. -

i cE

! 3T/4 i

i I

n 2

i a m 2 M E 2 w >

m =

i o

m o i E 5 o =

l l

l

! , , e  ! t t 217.32 220 225 230 235 238

! l i  ;

Distance From Core Axis, em l l

!I 6-6 Babcock & Wilcox

!I

Figure 6-2. Azimuthal Fluence Gradient (E s 1.0 Mev) at Inside Surface of Oconee 2 Reactor Vessel, Cycles 2, 3, and 4 1.15 l.10 -.

^

o 1.05 -

A

~

e L =

M 1.00 -

3 m

?'

U

. 0.95 -

ca 0.90 .-

ai er n

O Q-e,

$ 0.85 ' ' ' ' ' ' i ,

f 0 5 10 15 20 25 30 35 40 45 M

Deviation From Major Axis, negrees I

\

I I

I

7. DISCUSSION OF CAPSULE RESULTS W 7.1. Preirradiation Property Data I A review of the unirradiated properties of the reactor vessel core belt region indicated no significant deviation from expected properties except in the case of the upper shelf properties of the weld metal. Based on the predicted end-I of-service peak neutron fluence value at the 1/4T vessel wall location and the copper content of this weld, it is predicted that the end-of-service Charpy upper shelf energy (USE) will be below 50 ft-lb. This weld was selected for inclusion in the surveillance program in accordance with tl:e criteria in ef-fect at the time the program was designed for oconee Unit 2. The applicable selection criterion was based on the unirradiated properties only.

7.2. Irradiated Property Data 7.2.1. Tensile Properties Table 7-1 compares irradiated and unirradiated tensile properties. At both room temperature and elevated temperature, the ultimate and yield strength changes in the base metal due to irradiation and the corresponding changes in ductility are negligible. There appears to be some strengthening, as indi-I cated by increases in ultimate and yield strength and similar decreases in ductility properties. All changes observed in the base metal are such as to be considered within acceptable limits. The changes in the properties of the weld metal at both. room temperature and 580F are greater than those observed for the base metal, indicating the greater sensitivity of the base metal to irradiation damage. In either case, the changes in tensile properties are insignificant relative to the analysis of the reactor vessel materials at this l period in service life.

7.2.2. Impact Properties The behavior of the Charpy V-notch impact data is more significant than ten-sile properties to the calculation of reactor system operating limitations. ,

l I 7-1 Babcock & Wilcox l

1

!I i

1 lI Table 7-2 compares the observed changes in irradiated Charpy impact properties i

to the predicted changes using Regulatory Guide 1.99, Revision 1.

I i

The 50-ft-lb transition temperature shift for the base metal was in good agree-ment with the shift that would be predicted according to Regulatory Guide 1.99.

i The less-than-ideal comparison may be attributed to the spread in the data of

} the unirradiated material combined with the small number cf data points to es-l l

i tablish the irradiated curve. Under these toaditions, the comparison indicates i

l that the estimated curves in RG 1.99 for medium-copper materials and at low fluence levels are reasonably accurate and conservative for predicting the 50-1 i

f t-lb transition temperature shif ts.

The 30-f t-lb transition t emperature shif t for the base metal does not agrec l as well with the value predicted according to RG 1.99, although these values are expected to exhibit better comparison when it is considered that a major portion of the data used to develop RG 1.99 was taken at the 30- f t-1b tempera-2 ture.

The increase in the 35-mil lateral expansion transition temperature is compared with the shift in RT:DT r cu m data in a manner simHar to the compadson mah for the 50-ft-1b transition temperature shift. These data show a behavior sim-i jg ilar to. that observed from the comparison of the observed and predicted 50-ft-

'E lb transition data.

l All the tranoition temperature measurements for the weld metal agree poorly i

I with the predicted shift. This can be attributed to the chemist.ry of the weld i

i i metal compared to the nominal chemistry of normal weld metal for which the prediction curves were developed. The data do indicate that the prediction 1

! techniques produce conservative values for the predicted values of shift of the weld metal.

The data for the decrease in Charpy USE with irradiation showed a conservative agreement with predicted values for both the base and weld metals. Ilowever, a less-than-ideal comparison of the measured data and the predicted value would not be unexpected in view of the lack of data for medium- to high-copper-con-tent materials at the low to medium fluence values that were used to develop the estimating curves.

I Results from other capsules indicate that the RT@T estimating curves have greater inaccuracies at the very low neutron fluence levels (11x 10 " n/cm ),

1 2 I

7-2 Babcock & Wilcox

l l l

ll l This inaccuracy is attributed to the limited data at the low fluence values and to the fact that the majority of the data used to define the curves in RG 1.99 are based on the <.hift at 30 ft-lb as compared to the current requirement l of 50 ft-lb. For most materials, the shifts measured at 50 ft-lb/35 MLE are expected to be higher than those measured at 30 ft-Ib. The significance of the shifts at 50 f t-lb and/or 35 MLE is not well understood at present, espe-jI cially for materials having USE values that cpproach the 50-f t-lb level and/or i

the 35-MLE level. Materials with this characteristic may have to be evaluated at transition energy levels lower than 50 ft-lb.

I j The design curves for predicting the shift at 50 ft-lb/35 MLE will probably be modified as data become available. Until that time, the design curves for j predicting the RT s as gMn in E 1.% m conhM aMuau for NDT predicting the RT, shift of those materials for which data are not available and will continue to be used to establish the pressure-temperature operational limitations for the irradiated portions of the reactor vessel.

j The less-than-ideal agreement of the change in Charpy USE is further support l cf the inaccuracy of the prediction curves at the lower fluence levels. Al-1 j though the prediction curves are consercative in that they predict a larger l drop in USE than is observed for a given fluence and copper content, the con-servatism can unduly restrict operational limitations. These data support

the contention that the USE drop curves will have to be modified as more re-liable data become available; until that time the design curves used to pre-

, dict the decrease in USE are conservative.

II i

lE 1

I I

7-3 Babcock & Wilcox

l

)

i  ;

l l

!I l Table 7-1. Comparison of Tensile Test Results i

Elevated Room temp test temp test (580F)

Unirr Irrad Unirr Irrad Base Metal - AAW-163, Longitudinal Fluence, 10 1" n/cm2 (> 1 Mev) 0 3.37 0 3.37 Ult tensile strength, ksi 89.2 93.1 83.3 86.9 '

! 0.2% yield strength, ksi 68.0 71.4 61.5 64.7 i Uniform eloagation, % 11.1 14.1 9.5 13.6 I Total clongation, % 28.1 28.8 29.8 24.9  !

" RA, % 69.7 68.9 73.1 68.1 i

I Weld Metal - UF-209-1 Fluence, 10 1e n/cm 2 (> 1 Mev) 0 3.37 0 3.37 Ult tensile strength, ksi 95.2 110.6 89.7 102.2 j 0.2% yield strength, ksi 81.4 97.8 69.8 86.3 4

Uniform elongation, % 10.7 15.1 10.4 12.1 i

i Total elongation, % 25.6 21.0 20.6 20.4 1

i RA, % 57.9 43.4 48.9 43.9 i

'I j

I l

I

!I i

i l

!I i

7-4 Babcock & Wilecx i

I

• I. . _ _ - _ _ _ _ -._ . .. .. . _ - - _ _ . -_ . - - - _ _ _ _ .-_ - __

i ll l

1 Tabic 7-2. Observed Vs Predicted Changes in Irradiated Charpy Impact Properties

!I l

Material Observed Predicted (a) j Increase in 30-ft-lb trans temp, F t

! Base material (AAW-163)

Transverse 17 25

{

j Longitudinal Neg. --

Heat-affected zone (AAW-163) -- --

Weld metal (WF-209-1) 114 186 1 Correlation, HSST plate 02 71 90 Increase in 50-ft-lb trans temp, F i

i Base material (AAW-163)

Transverse 20 25 Longitudinal Neg. --

{l i

jg Heat-affected zone (AAW-163)

Weld metal (WF-209-1)

Ind.

186

$ Correlation, HSST plate 02 53 90 l Increase in 35-MLE trans temp, F Base material (AAW-163)

I Transverse Longitudinal 18 0 --

25 g)

Heat-cffected zone (AAW-163) -- --

i j

Weld metal (WF-209-1) 84 186(b)

Correlation, HSST plate 02 69 90(b) l Decrease in Charpy USE, ft-lb

,I 1

Base material (AAW-163)

Transverse Longitudinal 17 23

, Heat-affected zone (AAW-163) 4 23

=

Wcld metal (WF-209-1) 19 24 Correlation, HSST plate 02 27 25 I

j (a)These values predicted per RG 1.99, Revision 1.

iN

{g (b) Based on the assumption that MLE, as well as 50-ft-lb j transition temperature, is used to control the shift in RT

{g ET .

!u i 7-5 Babcock & Wilcox l 4 _ . _ _ _ _ - _ _ _ _ _ _ _ , _ _ _ . _ _ _ _ - - _ _ _ __ _ _ _ _ _ _ _ _

I I

I

8. DETERMINATION OF RCPB PRESSURE-TEMPERATURE LIMITS I The pressure-temperature limits of the reactor coolant pressure boundary (RCPB)

(RCPB) of Oconee Unit 2 are established in accordance with the requirements of 10 CFR 50, Appendix G. The methods and criteria enployed to establish operat-ing pressure and temperature limits are described in topical report BAW-10046P."

Tiie objective of these limits is to prevent nonductile failure during any nor-

. rcal operating condition, including anticipated operation occurrences and system W hydrostatic tests. The loading conditions of interest include the following:

I

1. Normal operations, including heatup and cooldown.

2. Inservice leak and hydrostatic tests.

3. Reactor core operation.

The major components of the RCPB have been analyzed in accordance with 10 CFR 50, Appendix G. The closure head region, the reactor vessel outlet nozzle, and the beltline region have been identified as the only regions of the reactor vessel (and consequently of the RCPB) that regulate the pressure-temperature limits.

Since the closure head region is significantly stressed at relatively low tem-peratures (due to mechanical loads resulting from bolt preload), this region largely controls the pressure-temperature limits for the first several service periods. The reactor vessel outlet nozzle also affects the pressure-tempera-ture limit curves of the first several service periods. This is due to the I high local stresses at the inside corner of the nozzle, which can be two to three times the membrane stresses of the shell. After the first several years of neutron radiation exposure, the RT of the beltline region materials will be high enough that the beltline region of the reactor vessel will start to control the pressure-temperature limits of the RCPB. For the service period for which the limit curves are established, the maximum allowable pressure as a function of fluid temperature is obtained through a point-by-point compati-son of the limits imposed by the closure head region, the outlet nozzle, and the beltline region. The maximum allowable pressure is taken to be the lowest of three calculated pressures.

8-1 Babcock & Wilcox

I I The limit curves for Oconee Unit 2 are based on the predicted values of the adjusted reference temperatures of all the beltline region materials at the end of the fifteenth full-power year. This year wac selected because it is estimated that the third surveillance capsule will be withdrawn at the end of I the refueling cycle when the estinated fluence corresponds to approximatel:

the eighteenth full-power year. The time dif ference between the withdrawal of the second and third surveillar.cc capsules provides enough time to re-estab-lish the operating pressure and temperature limits for the period of operation beyond the fifteenth full-power year.

The unirradiated impact properties were determined for the surveillance belt-line region materials in accordance with 10 CFR 50, Appendixes G and H. For the other beltline region and RCPB materials for which the measured properties I are not available, the unirradiated impact properties and residual elements (as originally established for the beltline region materials) are listed in Table A-1. The adjusted reference temperatures are calculated by adding the pre-dicted radiation-induced LRTg and the unirradiated RT'DT. The predicted LRT g is calculated using the respective neutron fluence and copper and phos-phorus contents. Figure 8-1 illustrates the calculated peak neutron fluence at several locations through the reactor vessel beltline region wall as a func-tion of-exposure time. The supporting information for Figure 8-1 is described j in section 6. The peutron fluence values of Figure 8-1 are the predicted flu-ences, which have been demonstrated (section 6) to be conservative. The de-E sign curves of Regulatory Guide 1.99* were used to predict the radiation-l induced ART v lues as a function of the material's copper and phosphorus NDT contents and neutron fluence.

The neutron fluences and adjusted RT 'DT values of the beltline region materials at the end of the fifteenth full-power yeat are listed in Table 8-1. The neu-tron fluences and adjusted RT g values are given for the 1/4T and 3/4T vessel f the closure head wall locations (T = wall thickness). The assumed RT 'DT region and the outlet nozzle steel forgings is 60F, in accordance with BAk'- 1004 6 P . "

• Revision 1, January 1976.

I 8-2 Babcock & Wilcox

I 1

Figure 8-2 is the reactor vessel pressure-temperature limit curve for normal l he r. t u p . This figure also shows the core criticality limits as required by j 10 CFR 50, Appendix C. Figures 8-3 and 8-4 show the vessel pressure-tempera-ture limit curve for normal cooldown and for heatup during inservice leak and j hydrostatic tests, respectively. All pressure-temperature limit curves are j

i apclicable up to the sixteenth EFPY. Protection against nonductile tailure j is ensured by maintaining the coolant pressure below the upper limits of the i

pressure-temperature limit curves. The acceptable pressure and temperature combinations for reactor vessel operation are below and to the right of the limit curve. The reactor is not permitted to go critical until the pressure-temperature combinations are to the right of the criticality limit curve. To j establish the pressure-temperature limits for protection against nonductile failure of the RCPB, the limits presented in Figures 8-2 through 8-4 must be

]

ja adjusted by the pressure differential between the point of system pressure

{

measurement and the pressure on the reactor vessel controlling the limit curves.

This is necessary because the reactor vessel is the most limiting component of the RCPB.

I I

I 1

1 l

1

)

!I 8-3 Babcock s.Wilcox L __

1 M M M M M M M M M M M M M M M M M Table 8-1. Data for Preparation of Pressure-Temperature Limit Curves for Oconee Unit 2 - Applicable Through 15 EFPY J

l eliment loc at ion Io<atlon P iil st Ic.n- In bu ed core I'*"

f ra C * I ' t #1 - -- - ' "", '

9

' I' l yg at e+** -l '

Ad tost r 1 RI at

• bt e r t .i l 111r n t Ite l t l i ne to weld ".'E 1/4T l'n i r r (F l -P'. v ) ,-n /i m --I 'm- - -F F f"i 'Ff*)- cal --of--15 F i ri , t 4 15 ront ent , s nnt ent . - - - - - -

He at No. l i {_+ re p lovi ! <*r at i m _f t , re derrees location , I)T N '7  ? I At 1/4T At 1/4I At fla! At 1/*T At f / '41 At 1/ 4 f AMA-77 9450M, rl 2 lower norrie belg -- -- --

(en) 0.n6 n.006 7 4 RlR . 5 '.I' l l lr, 9 gn og AAh-lbl S A '>HM , Cl ' l'ppe r i.be l l -- - --

20 n,04 0.1Hih )..oFIR /.lotl7 21 11 4) ti AA'w - t hs S A wa . (1 2 Icwer 9.le l l -- -- -- -10 0.02 0.010 1. lOt l N 7. jliF l ? 24 l' 18 3 WF-154 be I <! t*pper c i ri um, t e am 121 --

Yes (20) (-) (-) 2. /.1F I R 5.U t l 7 li, 69 ih% 84 WF-25 belj me t it -, i r c um. acam -61 -- Yes -14 (-) (-) 1. /t >I' I P 7.10117 1% 94 lRI 19 WF-ll2 keld 1. .we r i t r r um. scam -24'i --

tes O (-) (-) 1. 7H l h * .5 li 4 11 6 I,

Ngtes: ( ): e s t i ma t e d pe r bas-lfin WF.'

1 1

(-): pe r FL AW-l il i r , t +< t ot cr 19M0 i

I ta): per prn .l tt e rv <.o pte 1.94, Revision 1.

en I

i V il 1

i t

i i

i 1

I l

f W i tu I 7 O

o O

X" 9*

=

l O O

X i

' ~

m m m M W W M M M M M M M M M Figure 8-1. Predicted Fast Neutron Fluences at Various Locations Through Reactor Vessel Wall for First 15 EFPY 6.0 5.83 x 1018 nvt e 5.6 -

5.2 -

E R 4.8 -

E 4.4 -

m A

4.0 -

uJ

~

3.6 -

C 3.2x1018 nvt

= 3.2 -

0 S

- 2.8 -

S 5 2.4 -

?

2.0 -

@ 1/4T LOCATION -

3 1.6 -

z I.2 -

7.3 x 10I7 nyt .

cn 0.8 -

m 3/4T LOCATION

\

7 0 .2 x 1017 nv 0.4 -

4 OUTSIDE SURFACE

e. 0.0 -

. i e i i i e i i i i (8 0 1 2 3 4 5 6 7 EFPY 8 9 10 11 12 13 14 15 X

E E E E E E E E E E E E E E E Figure 8-2. Reactor Vessel Pressure-Temperature Limit Curves for Normal Operation IIeatup, Applicable for First 15 EFPY 2400 THE ACCEPTABLE PRESSURE-IFMPERATURL ASSUMED RI NDI, F COMBINATIONS ARE BELOW AND TO THE D RIGHT OF TH[ LIMIT CURVE (S). THE F 2200 _ BELIllME REGION 1/41 181 LIMil CURVES DO NOT INCLUDE THE BELTLINE REGION 3/4T 89 PRESSURE Dif f E REN il AL BIIWEEN THE POINI 0F SYSTEM PRESSURI ME ASU REME N T CLOSURE HEAD REGION 60 AND THE PRfSSURE ON lHE RfACIOR 2000 -

OUTLET N0z2LE 60 YESSEL REGION CONTROLLING IHE L Itt l i CURVE. OR ANY ADDlif0NAL MARGIN OF SAFETY FOR POSSIBLE INSTRUMENT ERROR.

$ PRESSURE. TEMP, g 1800 -

P0 INT PSiG r

- A 390 60 h _

B 390 170

" 320 C 835 g APPLICABLt FOR w D 2250 410 HEAIUP RAIES UP 450 273 T9 100F/h I 1400

- E

?

g F 1020 380 G 2250 450 g C"C#'

" 1200 -

(

r t i til i

@ 1000 -

a 800 - C C

2

=

000 -

cn

$ 400 -

C O B

g A n

7 200 -

E 8 0 t i I I 8 8 ' 8 I I

• 200 240 280 320 360 400 440 480 40 80 120 160 Reactor Vessel Coolant Temperatule, F

Figure 8-3. Reactor Vessel Pressure-Temperature Limit Curve f o r No rma l Operation - Cooldown, Applicable for First 15 EFPY 2400 ASSUMED RI NDT' 2200 - BELTLINE REGI0n 1/4T 184 BELTLINE REGION 3/4T 89 CLOSURE HEAD REGION 60 2000 -

OUTLET nozzle 60 PRESSURE. TEMP, a 1800 -

P0:nT Ps s r J A 240 70 8 710 24

$ 1600 - C g C 1650 310 g D 2250 333

- 1400 -

2 cc o 4 0 1200 -

l[ APPLICABLE FOR

" COOLDCeN RATES U 1000 - uP 10 Ic0F/h 3

800 E

B 600 THE ACCEPTABLE PRESSURE-TEMPERATURE COMetNATIONS ARE BELOW AND 10 THE RIGHT 400 -

0F THE LIMIT CURVE (S). THE LIMIT CURVES cn DO NOT INCLUDE THE PRESSURE DIFFERENTIAL m BETWEEN INE POINT OF STSTEM PRISSURE g 'O -

C MEASUREMENT AND TNE PRESSURE CN INE REACTOR VESSEL REGION CONTROLLING INE LIMIT CURVE, o A 0 OR ANY ADDITIONAL 64ARClu 0F SAFEiY FOR 7

p , , , , POSSlejE INSTRUMingfliOR.

! 40 80 120 160 200 240 280 320 360 h Reactor Vessel Coolant Temperature, F

m m m m m M m' M M M M M M M M M Figure 8-4. Reactor Vessel Pressure-Temperature Limit Curve for I nse rv ice Leak and Hydrostatic Tests, Applicable for First 15 FFPY 2600 -

ASSUMED RTnDT*'

~

BELTLINE REGION 1/4T 181 BELTLINE REGlos 3/4T 89 CLOSURE HEAD REGl0A 60 OUTLET n022LE 2000 -

PRESSURE. TEMP, Poln1 PSIG F s 1800 -

A 220 70 485 170 h 8 C 890 3 3 1600 -

360) APPLIC ABLE FOR HEAT'JP 0

U 0 1650)

AND C00LD0mM RATES UP

$ E 2500 394 oc  ; 1400 - 10$100Flh($50Fla

" ANY 1/2.h PERIOD) cc _

o 0 1200 -

m 1000 -

o 5 800 -

C E

~

THE ACCEPTA8LE PRESSURE-IEMPERATURE COM81 pal 10NS ARE 8EL0w AND 10 InE RIGNT OF THE LIMIT CURVE (S). THE C3 400 - B LIMIT CuRvtS o0 a0T luCLuot int P R r t s >. R t OlFFtREnTIAL h SETWEEn TNE POINT OF SYSTEM PRES!URE MEASUREMEm1 AND h THE PRESSURE On THE REACTOR VESSEL REGlas CONTROLLING g 200 - g inE LIMIT curve, OR Amr AeolilonAt MARGen 0F SAFETT p FOR POSSIBLE INSTRUMEni ERROR.

g 0 i a e i e r i e 60 100 300 340 380 420

{

140 180 220 260 Reactor VeS$el C00lant Temperature, F

t lI r

I i

1

9. SDIMARY OF RESULTS ,.

1 The analysis of the reactor vessel material contained in OCII-A, the second surveillance capsule of the Oconee Unit 2 pressure vessel surveillance pro-gram, led to the following conclusions:

1. The capsule received an average fast fluence of 3.37 x 1018 n/cm 2  !

l (E > 1 Mev). The predicted fast fluence for the reactor vessel T/4 loca-l tion at the end of cycle 4 (1361 EFpD) is 9.8 x 10 17 n/cm 2 (E > 1 Mev).

I

2. The fast fluence of 3.37 x 10" n/cm 2 (E > 1 Mev) increased the RT 'DT f

the capsule reactor vessel core region shell materials a maximum of 114F.

^

3. Based on a comparison of the fast flux in the surveillance capsule to that at the vessel wall and an 80% load factor, the calculated projected maxi-mum fast fluence that the Oconee Unit 2 reactor pressure vessel will re-

! ceive in 40 calendar years of operation is 1.20 x 10 ' n/cm2 (E > 1 Mev). 1 1

l 4. The ' increase in the RT g for the base plate material agreed acceptably

{ with that predicted by the currently used design curves of ART versus ET fluence and demonstrated that the prediction techniques are conservative.

5. The increase in the RT g for the weld metal did not agree well with that predicted by the currently used design curves of ART g versus fluence be-l cause the decrease in upper shelf energy distorted the measured shifts. ,

i 1

! 6. The current techniques used for predicting the change in Charpy impact j upper shelf properties due to irradiation are conservative.

7. The analysis of the neutron dosimeters demonstrated that the analytical j techniques used to predict the neutron flux and fluence were accurate.

8. The thermal monitors indicated that the capsule design was satisfactory i for maintaining the specimens within tne desired temperature range.

l 1

9-1 1

Babcock & Wilcox 6

i

'-_ . _ _ _ _ . . _ _ - - . _ _ _ - _ _ - - _ _ _ _ _ . _ _ - - . - . - - - ~ _ . _ . - _ _ - - - - _ . - _ . . _ - - _ _ - - , . . _ - - _ _ . - _ _ . , _ . _ , . .

l, 1

l l

r 1

l l

}

i

10. SURVEILLANCE CAPSULE REMOVAL SCHEDULE r

I Based on the postirradiation test results of capsule OCII-A, the following schedule is recommended for examination of the remaining capsules in the Oconee 2 reactor vessel surveillance program:

Evaluation schedule 4

8 (a) j Est capsul Estimated EFPY data j Capsule fluence, 2

i- n /cm Surface 1/4T _available OC11-B 1.2 x 10 ' 1 18 32 1987 OCII-EI ) 2.2 x 10 ' 1 33 59 1987

, OCII-D Stand by -- -- --

OCII-F Standby -- -- --

(a)These dates do not represent the earliest dates

that data will be available for the materials that l control the operating limitations. Similar ma-terials are included as part of the B&W Integrated lI Reactor Surveillance Program, which necessary data on a timely basis.

will provide (b) Capsule contains weld metal specimens.

!ll l

I ie i

i 10-1 Babcock & kVilcox

)ll 1 - _ _ _

I i

e i

11. CERTIFICATION 1

1 i

} The specimens were tested, and the data obtained from Oconee Nuclear Station, I

! Unit 2 surveillance capsule OCII-A were evaluated using accepted techniques f and established standard methods and procedures in accordance with the require-l ment, of 10 CFR 50, Appendixes G and 11.

!I I

! r l /fxhL- 425

~

%%/17/

l A. L. Lbwe, Jr., V(E. Date

! Project Tecl.nical Manager i

!I i

} This report has been reviewed for technical content and accuracy.

i i Mc%

J. L. ith i

%. n+y Date l

, Compor ent Engineer ng ]

l i

1 J

l il

!I 1 i l l l 1

i

11-1 Babcock & Wilcox I

I

l r

lI f

I I

I I

l l

lI 1

!I  !

1 l

l i

i l I APPENDIX A Reactor Vessel Surveillance Program -

j Background Data and Information I l I

I .

!I I

4, i

k i

I i

I i

il i,

i A-1 babcock & Wilcox i

I j

!I l.

l. Material Selection Data l

The data used to select the materials for the specimens in the surveillance l program, in accordance with E-185-66, are shown in Table A-1. The locations of these materials within the reactor vessel are shown in Figure A-1.

2. Definition of Beltline Region i

The beltline region of Oconee Nuclear Station, Unit 2 was defined in accord-l ance with the data given in BAW-10100A.5 j 3. Capsule Identification l

l The capsules used in the Oconee 2 surveillance program are identified below by number and type.

f Capsule Cross Reference Data 1

i I ID No. Type l OCII-A I OCII-B II OCII-C I

)"

4 OCII-D II

OCII-E I OCII-F II

4. Specimens per Surveillance Capsule 1 See Tables A-2 and A-3.

4 i

4

)

l

I I

l i

f A-2 Babcock 8. Wilcox

M M M Mi M M M M M' M' M ,

i Table A-1. Surveillance Procram Material Selection Data for Oconee 2 Charpy data, Og,N fore Dmp Tran m m Be l t l ine midplane .

Material ident. weight "- '

r nJ m N nu I ngitudinal 50 NDT' -

T F Heat No. Type lo<ation CL, em NDI' H 10F, ft-lb it-Ib 35 MLE l'$ E F Cu P S Ni AMX 77 SA508 C12 L r norzte -- --

0.06 0.006 0.009 0. 74 g

AAW 163 SA50R C1 2 l' ope r shell --

20 62, 77, 40 -- -- -- --

0.04 0.006 0.012 0.78 AWG 164 SA503 C1 2 Imer shell --

20 82, 83, 90 -- -- -- -- 0.02 0.010 0.010 0.78 WF-154 Weld Cir cum seam 123 --

41, 37, 43 -- -- -- -- 0.20 0.015 0.021 0.59 WF-25 Weld Circum seam -63 -- 38, 28, 49 -- -- -- -- 0.29 0.019 0.010 0.71 Wi-ll? Weld Circum seam -249 -- 35, 40, 30 -- -- -- -- 0.22 0.024 0.006 0.58 WF-209-1A Weld Surv. weld -- -- 29, 30, 32 -- -- -- -

0.34 0.013 0.010 0.48 i

l s

W I

I l

cn i tu

! cr O

O O

X j p 1 &

=

0 O

M

l

!E l

I Table A-2. Materials and Specimens in . Upper Surveillance Capsules OCII-A, OCII-C, and OCl?-E No. of specimOps 1

Material descrig on Tensile Charpy l

i Weld metal, WF-209-1A 4 8 l

HAZ A, heat AAW-163, longitudinal 0 8 l

t j Base metal material, plate A, heat AAW-163 Lon gitudinal 4 8 Transverse 0 4 j Correlation, HSST plate 02 0 8

! Total per capsule 8 36 I

.! Table A-3. Materials and Specimens in Lower j Surveillance Capsules OCII-B, OCII-D, and OCII-F

'I No. of specimens Material description Tensile Charpy HAZ B, heat AWG-164, lon gi t udin- 4 10

^1 i

Base metal material, plate B, heat jg AWG-164 lM Longitudinal 4 10 Transverse 0 8 Correlation, HSST plate 02 0 8 Total per capsule 8 36 I

I l lI l

A_4 Babcock a, WHcox i

I E

j I Figure A-1. Location and Identification of Materials Used in Fabrication of Reactor Pressure Vessel i

I I

^'

1'\

\ s

\_

lI k

< r A!C{ 77 (Lower Nozzle Belt) i C+ WF- 154

]

k*

p AAW 163 (Upper Shell) i

! I  ;

C = WF-25

I i

AWG 164 (Lower Shell)

'I 4% WF-112

\

' ' = >"^' < "'<" "">

lI i

lI i

A-5 Babcock & Wilcox

lI l

'I l

ll l

lI

<I lI

,I

'I APPENDIX B Preirradiation Tensile Data

~I lI 1

! l l

'I I

!I

'I l B-1 Babcock & Wilcox lI i

I i

I j Table B-1. Preirradiation Tensile Properties of Shell

g Plate Material, Heat AAW 163 ig i

i Test Strength, psi Elongation. %

ked'n ,

Ig Specimen temp, of area,

!E so. r Yie1d Uit. unif. Tota 1  %

j Lon gi tudina l i

fB EE-701 RT 70,420 90,680 10.74 27.14 69.3

! 706 RT 66,790 88,550 11.83 29.28 68.5 1 709 RT 66,690 88,300 10.88 27.86 71.2 i

j Mean RT 67,970 89,180 11.15 28.09 69.67 i Std dev'n 1,740 300 0.48 0.89 1.13 EE-704 580 61,580 82,110 9.48 30.7 74.8 l 707 580 62,460 82,110 9.59 30.0 75.7 t 718 580 60,490 85,630 9.52 28.6 68.8 Mean 580 61,510 83,280 9.53 29.77 73.1 l Std dev'n 25 810 1,660 0.05 0.87 3.06 i

I il l

Table B-2. Preirradiation Tensile Properties of Weld Metal - Longitudinal, WF-209-1 Specimen renR , Psi Hongation. %

temp, of area, No. F Yield Ult. Unif. Total  %

l

) EE-102 RT 81,190 95,140 10.84 26.1 60.7 i 105 RT 83,530 96,640 10.36 24.6 51.0

]

120 RT 79,430 93,800 11.0 26.1 62.1 i Mean RT 81,380 95,190 10.73 25.6 57.94 l Std dev'n , 1,680 1,160 0.27 0.71 4.94 EE-113 580 69,170 89,210 10.4 20.7 46.8 l 118 580 69,730 89,910 10.4 20.7 49.9 j 119 580 70,610 90,000 10.25 20.4 50.0 Mean 580 69,840 89,710 10.35 20.6 48.9.

Std dev'n !5 596 356 0.07 0.14 1.49 lI 1

j l

! B-2 Babcock & Wilcox

E I ,

I I

I I

I I

I APPENDIX C Preirradiation Charpy Impact Data I

B g

l I

I I

I I c-1 Babcock s Wilcox I

E Table C-1. Preirradiation Charpy Impact Data for Shell I Course Material - Longitudinal Orientation, Heat AAW 163 Absorbed I Shear Test Lateral Specimen temp, energy, expansion, fracture, No. F ft-lb 10-3 in. 7.

I EE-727 720 732 361.7 361.4 359.5 140 134 132 63.5 68 71 100 100 100 I EE-750 742 745 280.9 280.4 280.3 151 145 145 65 69.5 74.5 100 100 100 I EE-734 706 702 200.9 200.5 200.4 150.5 154 147 68 71 74 100 100 100 I EE-748 741 746 141 140.8 140.5 146 156 153 69.5 68.5 70.5 100 100 103 I EE-719 736 712 80.2 80 79.9 108 122 98 67.5 68.5 64 45 65 35 I EE-752 747 751 21.3 20.9 20.8 83 81 82 64.5 62.5 65.5 25 12 20 I EE-735 701 730 0.1 0.1

-0.6 55 47.5 60 40 35.5 44.5 1

I EE-744 743 749 11.8

-12.4

-13.9 75 79 16 54 57 12.5 6

<1 8

I EE-705 716 729

-40.9

-41.1

-41.6 13 23 27 9

16 20 0

0 0

I 1

I I

I I c-2 Babcock & Wilcox I

E I Table C-2. Preirradiation Charpy Impact Data for Sh' ell I Course Material - Transverse Orientation, Heat AAW 163 I Specimen Nc.

Test temp, F

Absorbed energy, ft-lb Lateral expansion, 10-3 in.

Shear fracture, I EE-626 625 615 361.3 361.1 359.3 122 121 124 74 68 75 100 100 100 I 620 EE-630 632 358.7 279.4 278.8 122 127 118 70.5 71 68 100 100 100 I 637 EE-613 606 278.3 202 201 122 124 128 69 73 69.5 100 100 100 I 627 EE-639 633 197.4 141.1 140.5 147 134 119 70.5 73.5 63.5 100 100 92 I 640 EE-619 608 139 79.9 79.8 128 119 109 70.5 60.5 65 100 100 65 I 607 EE-629 631 79.8 20.5 20.4 117 62 77 67 50.5 60 80 25 8

I EE-610 625 616 9.9 0.5

-1.5 65 49 61 50 39 44 1

1 0

I EE-635 634 638

-15

-15.5

-18 54 39 38 40 28 31 4

2 2

I EE-605 603 622

-38.9

-39

-41.1 32 24 18 24.5 17 14 1

0 0

I I

I I

I -

Babcock s. Wilcox l

C-3 I

E I .

Table C-3. Preirradiation Charpy Impact Data for Shell I Course Material - HAZ, Longitudinal Orien-tation, Heat AeW 163 I Specimen No.

Test temp,

_F Absorbed energy, ft-lb Lateral expansion, 10-3 in.

Shear fracture, EE-434 361.8 164 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 74 100 442 280 145 65 100 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 75 100 441 138.9 129 69.5 100 EE-i37 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 45 448 41.1 101 62 35 450 41.8 75 54.5 40 EE-438 0.6 97 56 20 418 0.5 173.5 68 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 -80.6 100 58.5 20 423 -82.5 73 39.5 5 l

l I

I I .

I Babcock a. Wilcox C-4

I Table C-4. Preirradiation Charpy Impact Data for Weld Metal, WF-209-1 Test Absorbed Lateral Shear Specimen expansion, I

temp, energy, fracture, No. F ft-lb 10-3 in.  %

EE-014 361 67 53.5 100 I 032 035 EE-019 360.8 360.1 202.5 72 65 63 52.5 49 100 100 49.5 100 I 036 016 EE-020 201.4 198.2 120.9 67.5 64 68 45 46 47 100 100 98 I 011 006 EE-031 120.3 120.3 80.9 66 66 60 47 44 42.5 100 100 75 I 015 012 EE-024 80.2 80.2 0.5 60 55 42 32 80 60 27 25.5 25 I 022 008

-0.5

-0.5 31 25 28.5 23 20 15 I

I -

I I

I I

I I

I C-5 Babcock s. Wilcox I

I I Figure C-1. Impact Data From Unirradiated Base Metal (AAW-163) HAZ, Longitudinal Orientation 100 i i [ T-3 i j i  ; i i

• • 75 - * -

E g 50 - - - - - - - - - - - - - - - - - . - - - - - - - - - - - - - - - - - - - -

'I m

= o J 25 - -

I 0" I I I I I I I I I I

.08 j j g , ;  ;  ;  ; g 4 i  ;

, 7

!.De ,- s I <

O .04 5

G

.__z-_____'_

r-

=

I f.02 I I I I I I I I I I I 0

200 i i i g g g g g  ;  ; g DATA SUTARY

~

180w T mDT -

I " D-

• Tcv (35 *Lt) 166 TCV (50 ri-ta) luD. . -

, Tcy(30 FT-La) I"8* .

Cy -USE (Avo) e42 FT-te , _

. 140 RT , -50F -

I C .

• 5 120- -

{ * . .

.I a100*l-e l

• i 80 % *

• I

L Y

60l -

I .________-._______ _ ____

MATERIAL DetENTAT!@ LONG. MAZ ~~

$A508,Cl2

~

I 20- FLutscg NOME -

HEAT No. AAW-163 0' I I I I I I I I ' ' i

-C0 -40 0 40 80 120 160 200 240 280 320 360 400 Test itmernarust, F 1

l g C-6 Babcock & Wilcox 3 .

l lI I

!I Figure C-2. Impact Data From Unirradiated Base Metal

[ ( AAW-163) , Transverse Orientation j

3 -

100 g  ; i  ; i

i , , 3

]

! ** 75 -

, C - - - - - - - - - - - - - - - - - - - - - - - -

• 50 - - - - - - -

m

) 5 J 25-j t . 1 I I I I I I i 1 t

O I I I +

f . 08 I I I I I I

I A

=

. ~

• s

'd.06 - .

• F II '

O .04 5

j e ,

i i .02 -

O I I I I I

' I I I I I I 0

1. ' ' ' I I I I i i I l DATA SUPPARY

+20F _

I 180*- T,37 Tey (35 mLE)

-13F T (50 FT-La) -7 F -

i 160- CV i Tcy(30 n-La) -2 8 F I Cy -USE (avo) 133 FT-Le ,

j . 140 gi +20F y not ,

I '. m i g 120- . .

S -

.i 100- .

i -

4 g

! E i w 80 -

• j -

i V

l r .

- 60 t I. _ _ _

l 40- SA508.CL2

hita nt ORttniaTrom TRANSVERSE -~

FL ut=ct MOME 20- .

HEAT No, AAW-163 I I I I I I I I ' ' '

0

-CD -40 0 40 80 120 160 200 240 280 320 360 400 Test TEMPERArust, F c-7 Babcock & Wilcox

I l

Figure C-3. Impact Data From l'nirradiated Base Metal

( AAW- 163) , Longitudinal Orientation 100 l  ; g g  ;  ;  ;  ;

15 -

~

r

• a yg I .

5 d 25 -

I 0 I

I I I I I I I I I

.08 I .

j g g g i j g i g  ; j E -

=

• e e .06 ,

y

• I  :

j.04-G w

I d .02-5

.5 t

i 1 1 I i l l I l l 0

200 i i  ;

i

;  ;  ;  ; g j DATA

SUMMARY

+20F I 180-T oy Tcy(35 att)

I CV FT-LB) "#E

• "F Tcy(30 FT-ts) -24F *

. 140- 20F = "

C I ;

5 120 S

I 3 100- -

I E J5 80- . _

I C

2 1

00- - l 40- -

Martn ut 34608.cL2 g,,,,,,,, ,,,,,,,,,,,c I

20 -

. FLUENCE NONE HEAT No. A Aw-163 0

I I I I I I I I 2 ' '

-00 -40 0 40 80 120 160 200 240 280 320 360 400 I Itsi Tt=*enarupt, F I C-8 Babcock & Wilcox

I Figure C-4. Impact Dat.t From Unirradiated Weld Metal E I i i i e i e i i I I

~ 75 - _

J .

0

So ___-_ ___.__ __ __________

u I  :

5 25~

• I -

0

. ' 08 I

I I

I I

I I

I I

I I

I I

I I

I I

I I

I I

I

.06- -

t .

"g.04- .______ _____________________-

s .02- -

I %

a 5

0 I I I I I I I I I I I I

100 i i i g , g  ;

g g ,  ;

LATA SUTARY 90+ T,37 -20F I ]T (35 l

cy 80 Tcv (50 n-La) ntE) +42F Tcy(30 n-ts)

+50F Cy-USE (avr.) 67 FT-Le I  ?

70 RT +4F

$ Q.--

g i F0_

I S

E .

S f%----__-- ____________-_-____---_-

I a 5

.' 40 -

0 l I i 1 '

30 -- - - - _

20r-I l u,ATERIAL WELD METAL OniENTAtto=

10- FLUENCE NOME -

I 0

-CD -40 I

0 I I 40 80 I I 120 I

160 200 Test TEMPERATURE, F I

240 HEAT No.

280 WF-209-1 320 360 400 I

C-9 Babcock & Wilcox

I I

I I

I I

I I

I APPENDIX D Fluence Analysis Procedures I

I I

I I

,I l

D-1 Babcock & Wilcox l

I I 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-I sel fluence was accounted for. Thus, two sets of calculations were required -

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 I the axial power distribution in the closest fuel assembly. This is considered Babcock & Wilcox l D-2

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 I

The calculation described above for CR-3 provides the neutron flux as a func-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 )

)

1 D

=f3.7x n

10* i E n j=1 j

~

{ N = Avogadro 's number, J

i A = atomic weight of tare,et material n, j n f4 = either weight fraction of target isotope in nth material or fission yield of desired isotope, o (E) = group-averaged cross sections for material n (listed in i Table E-3),

4(E) = group-averaged fluxes calculated by DOT 3.5 analysis',

JE F j

= f raction of full power during jth time interval, t j,

A g

= decay constant of ith material, l t = interval of power history, i j j

T = sum of total irradiation time, i.e., residual time in re-

actor, and wait time between reactor shutdown and counting, I l T = cumulative time from reactor startup to end of jth time ig period , i.e. , T =

[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 (D-2) jg C=D1 (calculated).

i

With C specified, the neutron fluence greater than 1 Mev can be calculated i

+ 1

from l i

i t

i D-3 Babcock & Wilcox i

i i

l l

l 15 Mev M

?(E > 1.0 Mev) =C [  ?(E) [

j=>

Ft, d

(D-3)

E=1 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 I For current operation, fluence values in the precsure vessel are calculated 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 g 4'v,x . c v,R 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

because it accounts f or all known power distribut ions. In addition, it re-l doces errorn that may result f rom the selection of a hypothetical " equilibrium" l cycle.

1 i

l I  :

l l

l 1

,4 i

j 1

I t

i 1

i  !

1 i i

}

l 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 ,

Table D-1. Capsule Normalization Constant Measured act ivity, t.Ci/c (

A B i At Cycles CR-3 calculated C = A/B Cycle 1, 1B and 2, irradiation activity, normalization Reaction OC-2 CR-3 only uCi/c cons t ant (c) 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 1.03 277 j Np(n,f)l37Cs 6.1 2.42(+1) 1.81(+1) 2.09(+1) 0.87 23e U(n,f)IU3Ru

1 (-9) 1.17(+2) 1.17(+2) 1.41(+2) 0.84 I T 23'U(n,f)1 'Ru 9 (-1) 2.69(+1) 2.6 (+1) 2.87(+1) 0.91 i

237 j Np(n,f)1U'Ru 6.2 1.27(+2) 1.21(+2) 1.27(+2) 0.96 2 3 ' U(n , f) l"'* Ce 6 (-1) 5.8 (+1) 5.74(+1) 5.67(+1) 1.01 237 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 1

237 U(n,f)'5Zr 4 (-5) 6.05(+2) 6.05(+2) 7.13(+2) 0.85 1

W ni cr (a) Average of four dosimeter wires from Table E-2.

O Obtained from A = A2 -Ae l where A is the decay constant for the product isotope and

p t is calendar time from EOC-1 in Oconee 2 to EOC-2 in Crystal River 3 (1420 days).

i

( (' Value for irradiation in Crystal River 3 only.

was selected as the normalization constant.

Average of all fission reactions (0.93)

)

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) 2 0.560 (+14)(^) 0.76' 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 8 x (1.57 x 10 ) = 1.14 x 10 10 1

(

Cycle 8 assumed to be equilibrium cycle for future operation.

5 1 K 1 1

g 1 x

P w

X

I I

I I

I I

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-weighted with a 23s 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) se Ni 100% Cd-Ag 0.02676" Cd Ni(n.p)S8Co 0.56% Co-Al Cd-0.040" Cd s9 Co(n,y)Co 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.r7 0.3094 9.40 91.3 IU'Ru 0.3003 9.13 88.6

, "2u 0.0841 2.56 24.8 l'7 Cs l 0.01506 0.458 4.44 1'Ce 0.1773 5.39 52.3 23 '5 i Np-Al 0.0291 Np(n,F) Zr 0.2321 7.98 554 d

IRu -- -- --

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) '0 Co 0.3748 2.82 10.8

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

'5 l 2 " U-Al 0.0433 2 "U(n.F) Zr 0.5158 11.9 116 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 2'7 Np(n.F) '5 Zr 0.2054 8.59 597 1Ci Ru -- -- --

i 1"'Ru 0.0534 2.23 155 lCs 0.007954 0.333 23.1 lCe 0.09692 4.06 282 (I

I I

E-3 Babcock & Wilcox -

l 1

l

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

'2 '0 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

' Fe ( n ,3 ) 53 1 Fe 24.55 163 49,400 2"U-Al 0.0364 23'U(n,F) 'Zr 0.3420 9.40 91.2 103 Ru 1 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) '5 Zr 0.3481 7.18 498 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) 'O Co 29.27 1810 323,000 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 i

""Ce 0.2485 7.12 69.1 1

'Np-Al 0.0486 2Np(n.F) '5 Zr 0.5393 11.1 771

( l03 Ru -- -- --

106 Ru l

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, c 237 23e g 58 s i. pc Mev 3p 31 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 l

1 iI I

i i

l 1

E-5 Babcock & Wilcox l

! 1 1

l 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

l F-1 babcock & Wilcox

i I

1 l H. S. Palme, C. S. Carter, and C. L. Whitmarsh, Reactor Vessel Material 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

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, Radiation Shielding Information Center.

l l 8 C. L. Whitmarsh, Pressure Vessel Fluence Analysis, BAW-1485, Babcock &

Wilcox, Lynchburg, Virginia, June 1978.

s i

l I

l l

F-2 Babcock & Wilcox

._ _