ML19322B865

Analysis of Capsule OCI-E,Reactor Vessel Matls Surveillance Program.

ML19322B865
Person / Time
Site:	Oconee
Issue date:	09/30/1977
From:	BABCOCK & WILCOX CO.
To:
References
BAW-1436, NUDOCS 7912060695
Download: ML19322B865 (86)

Text

-

o o

?

e (j C-%) was. ggg _ ----.

g' BAW-1436 [p#

f f eptember S 1977  !

.p1 I/

p/g g1

/

RETURN TO RE0!!!.ATCH CENTimL FILES

! , ROOM 018

- /

~

1 8

1 l

ANALYSIS OF CAPSULE OCI-E I DUKE POWER COMPANY

)

OCONEE NUCIZAR STATION - UNIT 1 l

- Reactor Vessel Materials Surveillance Program -

1

- NOTICE -  !

l T'.8 Ali ACHED FILES AHE OFFICIAL RECORDS OF THE l J1VC'ON OF DOCUvf NT CONTROL THE Y HA\'E BE EN ,

C*"HGED TO YOU FOH A LIMiiED ilME PE HtOD AND '

MU51 81 RETURNED TO THE RECORDS _F ACILIT_Y BHANCH 016 PL E ASE DO NOT SE ND DOCUMENTS

{14 ARGE D CUT THHOUGH.iHE MAIL. HE MOVAL OF ANY PAGtiSI FHOM DOCUME NT FOR HEPRODUCilON MUST l

, B" HE F E HkE D TO FILE PE HSONNE L.

DE ADLINE HE T JRN DATE ,

i RECORDS FACILITY BRANCH t_

I iiETURN 10 hatiuuwa attSilhock&Wilcox ROOM 016 - -. . - .- - -.

7912060f g

l l

BA'4-11. 3 6 September 1977 ANALYSIS OF CAPSULE OCI-E DUKE POWER COMPANY OCONEE NUCLEAR STATION - CNIT 1

- Reactor Vessel Materials Surveilltace Program -

1 by A. L. Lowe. Jr., PE E. T. Chulick H. S. Palme C. L. Whitmarsh C. F. Zur11ppe r

B&W Contract No. 595-7020-51 BABCOCK & WILCOX Power Generation Group Nuclear Power Generation Division P. O. Box 1260

, Lynchburg. Virginia 24505 Babcock &Wilcox

CONTE)ffS Page

1. INTRODUCTION . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1

2. RACNCROUND . . . . . . . . . .. . . . . . . . . . . . . . . . . 2-1

3. SURVEILLANCE PROGRAM DESCRIPTION . . . . . . . . . . . . . . . . 3-1

4. PREIRRADIATION TESTS . . . . ..... . . . . . . . . . . . . . 4-1 4.1. Tensile Tests . . . . .... . . . . . . . . . . . . . . 4-1 4.2. Impact Tests . . . . . .......... . . . . . . . . 4-1

5. POSTIRRADIATION TESTS . . . . . . . . . . . . . . . . . . . . . 5-1 5.1. Thermal Monitors . . . .... . . . . . . . . . . . . . . 5-1 5.2. Chemical Analysis . . .... . . . . . . . . . . . . . . 3-1

5. 3. Tensile Test Results . ...... . . . . . . . . . . . . 5-1 5.4. Charpy V-Notch Impact Test Results . . . . . . . . . . . . 5-2

6. NEUTEON DOSIMETRY . . . . . . . . . . . . . . . . . . . . . . . 6-1 6.1. Introduction . . . . . . . . . . . . . . . . . . . . . . . 6-1 6.2. Analytical Approach . .......... . . . . . . . . 6-2 6.3. Results . . . . . . . . . . . . . . . . . . . . . . . . . 6-3

7. DISCUSSION OF CAPSULE RESULTS . . . . . . . . . . . . . . . . . 7-1 7.1. Preirradiation Property Data . . . . . . . . . . . . . . . 7-1 7.2. Irradiated Property Data . . . . . . . . . . . . . . . . . 7-1 7.2.1. Tensile Properties . . . . . . . . . . . . . . . . 7-1 7.2.2. Impact Properties . . . . . . . . . . . . . . . . 7-1

8. DETERMINATION OF RCPB PRESSURE-TD(PERATURE LIMITS . . . . . . 8-1

9.

SUMMARY

OF RESULTS . . . . . ........... . . . . . . . 9-1

10. SURVEILLANCE CAPS',LE RD007AL SCHII)ULE . . .. . . . . . . . . . 10-1

11. CERTIFICATION . . .. . . . . . .. . . . . . . . . . . . . . . 11-1

12. REFERENCES . . . . . .. . . . . . . . . . . . . . . . . . . . .

12-1

, - 111 - Babcock & WilCC2 f

1 i

l

\

Contents (Cont'd)

Page APPE.NDIXES

,A. Reactor Vessel Surveillance Program - Background Data and Information . . . . . . . . . . . . . . ..... A-1 B. Preirradiation Tensile Data . . . . . . . . . . ..... B-1 C. Preirradiation Cha py Impact Data . . . . . . . ..... C-1 D. Threshold Detector Information . . . . . . . . . ..... D-1 l

List of Tables Table 3-1. Specimens in Surveillance Capsule OCI-E . . . . . . . .....

3-2. Chemistry and Heat Treatment of Surveillance Materials . 3-2 3-3. Chemistry and Heat Treatment of Correlation Material -- Heat .... 3-3 A-!!95-1 A533 Grade B, Class 1 . . . . . . . . . . . . .... 3-4 5-1. Chemistry Data on Unirradiated Oconee 1 RVSP Material .....

5-2. 5-2 Tensile Properties of Capsule OCI-E Base Metal and Weld Metal Irradiated to 1.5 a 1016 avt . . . . . . . . . . . . . . .... 5-3 5-3. Charpy Impact Data for Capsule OCI-E Base Metal Irradiated to 1.5 = 1018 nyt . . . . . . ... . . . . . . . . . . .... 5-4 5-4. Charpy Impact Data for Capsule OCI-E Weld Metal (WF-112)

Irradiated t . 5 = 1018 nyt

. ........ . . . . . .... 5-5 5-5. Charpy Impa Data for Capsule OCI-E Correlation Monitor Material Irradiated to 1.5 = 1018 nvt . . . . . . . . . .... 5-5 6-1. Surveillance Capsule Detectors .

6-2. Flux Adjustment Factor .

. . . . . . . . . . . . .... 6-5

. . . .... . . . . . . . . . .... 6-5 6-3. Dosimeter Activations After Cycle 2 . . . . . . . . . . .... 6-6 6-4. Normalized Flux Spectra, E > 1 MeV . . . . . . . . . . . .... 6-7 6-5. Fast Neutron Fluence . . . . .

6-6.

................. 6-7 Predicted Fast Fluence in Pressure Vessel for 10 EFPY . .... 6-8 7-1. Comparison of Tensile Test Results . . . . . . . . . . . . . . .

7-2. 7-4 observed Vs Predicted Changes in Irradiated Charpy Impact Properties 8-1. . .. . . . .. . . . . . . . . . . . .... 7-5 Data for Preparation of Pressure-Temperature Limit Curves for Oconee Unit 1, Applicable Through 8 EFPY . . . . . . .... 8-4 A-1. Surveillance Program Materials Selection Data for Oconee 1. . . A-3 A-2. Materials and Specimens in Upper Surveillance Capsules OCI-A, OCI-C, and OCI-E . . . ................. A-4 A-3. Materials and Specimens in Imer Surveillance Capsules DCI-B, OCI-D, and OCI-F . . . ................. A-4 B-1. Preirradiation Tensile Properties of Shell Plate Material. ,

Heat C-3265-1 .................... . .... 5-2 B-2. Preirradiation Tensile Properties of Shell Plate Material.

B-3.

HAZ, Heat C-3265-1 . . . . . . . . . . . . . . . . . . . . . . . 5-3 Preirradiation Tensile Properties of Weld Metal Weld Qualification No WF112 . . . . . . . . . . . . . . . . . . . . . 5-4

- iv - hock Wakox ,

I

I.i s t of Tables (Cont'd)

Table Page C-1. Preirradiation Charpy impact Data for Shell Plate Material.

Longitudinal Direction, Heat C-3265-1 . . . . . . . . . ..... C-2 C-2. Preirradiation Charpy Impact Data for Shell Plate Material, Transverse Direction. Heat C-3261-1 . . . . . . . . . . ..... C-3 C-3. Preirradiation Charpy Impact Data for Shell Material, HAZ, Longitudinal Direction, Ileat C-3265-1 . . .. . . . . . ..... C-4 C-4. Preirradiation Charpy Impact Data ior Shell Material HAZ, Transverse Direction. Heat C-3265-1. . . . . . . . . . ..... C-5 C-5. Pretrradiation Charpy Impact Data for Weld Metal, Weld Qualification No. WF 112 . . . . . . . . ... . . . . ..... C-6 D-1. Detector Composition and Shielding . . . . . . . . . . ..... D-2 D-2. Measured Detector Activities After Cycle 1 . . . . . . ..... D-3 D- 3. Measured Detector Activities Af ter Cycle 2 . . . . . . ....- D-4 D-4. Dosimeter Activation Cross Sections . . . .... . . . ..... D-7 List of Figures Figure 3-1. Reactor Vessel Cross Section Showing Surveillance Capsule 1,ocations .

. . . . . . . . . . . . . . . ... . . . . ..... 3-5 5-1. Charpy Impact Data From Irradiated Base Metal, Longitudinal Orientation . . . . . . . . . . . . . . . .... . . . ..... 5-6 5-2. Charpy Impact Data From Irradiated Base Metal, Transver e Orientation . . . . . . . . . . . . . . . . . . . . . . ..... 5-7 5-3. Charpy Impact Data From Irradiated Base Metal Heat-Af fected Zone, Longitudinal Orientation . . . . . ....... ..... 5-8 5-4. Charpy Impact Data From Irradiated Weld Metal . . . . . . . . . . 5-9 5-5. Charpy Impact Data From Irradiated Correlation Monitor Material . . . . . . . . . . . . . . . . .... . . . ..... 5-10 6-1. Predicted Fast Neutron Fluences at Various Locations Through Reactnr Vessel Wall for First 10 EFPY . . ...... . ..... 6-9 7-1. Irradiated Vs Unirradiated Charpy Impact Properties of Base Metal, Longitudinal Orientation . . . . . ... . . . . ..... 7-6 7-2. Irradiated Vs Unitradiated Charpy Impact Properties of Base '

Metal, Transverse Orientation . . . . . . ..... . . ..... 7-7 7-3. Irradiated Vs Unitradiated Charpy impact Properties of Base Metal Heat-Af fected Zone, Longitudinal Orientation . . ..... 7-8 7-4. Irradiated Vs Unirradiated Charpy lapact Properties of Weld Metal . . . . . . . . . . . . . . . ..... . . ..... 7-9 7-5. Irradiated Vs Unirradiated Charpy impact Properties of Correlation Mctitor Material . . . . . . .......... .. 7-10 8-1. Fast Neutron Fluence of Surveillance Capsule Center Compared to Various Locations Through Reactor Vessel Wall for First 10 EFPY . . . . . . . . . . . . . . . . . ..... . . . .... B-5 8-2. Reactor Vessel Pressure-Temperature Limit Curves for Normal

• Operation Heatup Applicable for First 8 EFPY . . . . . . .... 8-6

.,_ Babcock & Wilcox e

a m nen .m ~ ..e O

G

List of Figures (Cont'd) p-Figure Page 8-3. Reactor Vessel Pressure-Temperature Limit Curve for Normal Operation - Cooldown Applicable for First 8 EFPY . . .... .. 8-7 ,

6-4 Reactor Vessel Pressure-Temperature Limit Curve f or Inservice .

Leak and H;drostatic Tests, Applicable for First 8 EFPY ,... 8-8 A-1. Location and Identification of Materials Used in Fabrication of Oconee Unit 1 Reactor Pressure Value . . . . . . ... ... A-5 A-2. Location of Longitudinal Welds in Upper Lower, and Intermediate Shell Courses . . . . . . . .. . . . . . . .. ... .... A-6 C-1. Impact Data From Unitradiated Base Metal A - Longitudinal Orientation . . . . . . . . . .. . . . . . . . . ...... C-7 C-2. Impact Data From Unirradiated Base Metal A - Transverse Urlentation . . . . . . . . . .. . . . . . . . . ....... C-8 C-3. Impact Data From Un'rradiated Base Metal A -- HAZ, Longitudinal Orientation . . . . . . . . . ...... . . . . .... ... C-9 C-4. Impact Data From Unirradiated Base Metal A - HAZ, Transverse Orientation . . . . . . . . . ..... . . . . . . . .. ... C-10 C-). Icpact Data From Unirradiated Weld Metal Longitudanal Orien*ation. . . . . . . . . . ... . . . . . . . ... .... C-11 C-6. Impact Data Unirradiated Correlation Material . . ...

... ' C-12 t

a .

I

- v1 - .

Babcock & Wilcox ,

1

1. INTRODUCTION' This report de* ribes the results of the examination of the second capsult of t he- Duke Power Company's Ocon. e Nuclear Station, Unit ; reactor vessel surveil-Idnce program. The first capsule of t he pro,; ram was re: aved and examarxd af ter the first year ut operatton; the results are reported in BAW-1421.I The ob',ective of the program is to monitor the effects of neutron irradiation en the tensile and impact properties of reactor pressure vessel mate-ials under actual operat ing conditions. The surveillance program for Oconee I was designed and furnished by Babcock 6 Wilcox an dessribea in BAW-10006A.- The program was designed in accordance with E185-66 and was planned to monitor the effects of neutron irradiation on the reactor vessel material for the 40-year design life of the .eactor pressure vessel.

The surveillance program for Oconee I was not designed in accordance with Appendixes G and 11 to 10 CFR 50 since the requirements did not exist at the time the program was designed. Because of this difference, at the time the first capsule was removed and specimens tested, additional tests and evalua-tions were performed to ensure that the requirement s of 10 CFR 50. Appendizes C and 11 were met. The recommendations for the future operation of Oconee 1 after the evaluation of the first capsule complied with these requirements.

The future operating limitations established af ter the evaluation af the second surveillance capsule are also in accordance with the raquiren?nt of 10 CFR 50 Appendixes G and H. The recommended operating period was extended from five to ciRht effective full power years as a result of ti.. second cap-sule evaluation.

~

t 1-1 Babcock 8.Wilcox

2. BACEGROUND 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 te-

.ac to rs.

The beltline region of the reactor vessel is the ost critical region of the vessel bemause it is exposed to neutron irradiat ien. The general et-fects of fast neutron irradiation on the mechanical properties of such low-

. alloy ferritic steels as SA302B, Code Case 1339, used in the faorication of tt.e Oconee I reactor vessel are well characterized and documented in the litera-ture.

The low-allov ferritic steels used in the celtline region of teactor vessels exhibit an increase in ultimate and yield strength properties with a corresponding decrease in ductility after irradiation. In reactor pressure vessel steels, the most serious mechanical property change is the increase in temperature for the transition from brittle to ductile fracture accompanied by a reduction in the upper shelf impact streng:h.

Appendix C 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 prc sure boundary (RCPB) of water- l cooled power reactors and provides specific guidelines for determining the pressure-temperature 11a s tations on operation of the RCPB. The toughness and operatinnal requirements tire specified to provide adequate safety margins during any condition of normal operation, including ant icipated operational occurrences and system hydrostatic tests, to which the pressure boundary may be subjected over its service lifetime. Although the requirements of Appendix C to 10 CFR 50 became effective on August 13, 1973, the requirements are ap-plicable to all bo' ling and prersuriped water-cooled nuclear power reactors, including those under construction or in operation on the ef fective date.

Appendix H to 10 CFR 50, " Reactor Vessel Materials Surveillance Program Re-quirements," defines the material surveillance program required to monitor changes in the f racture toughness properties of ferritic materials in the re-actor vesset beltline region of water-cooled reactors resulting from exposure 2-1 Babcock & Wilcox l

t

to neutron irradiation and the thermal environment. Fracture toughness test data are obtained from material specimens withdrawn periodically from the re-actor vessel. These data will permit determination of the conditions under which the vessel can be operated with adequate safety margins against fracture throughout its service life.

A method for guarding against brittle fracture in reactor preasure vessels is described in Appendix G to the ASME Boiler and Pressure Vessel Code, Section

!!!. This method utilizes fracture mechanics concepts and the reference nil-ductility temperature. RT # * # '" * # #" #' "

N?T*

• weight nil-ductility transition temperature (per ASTM E-208) or tr e tempera-ture that is 60F below that at which the material exhibtts 50 ft ib and 35 mils lateral expansion. The RTNDT ' E **" ' 'I ' * " " * *" "* '

material to a refertnce stress intensity factor curve (K #"'" ' ' 'P~

IR pears in Appendix G of ASME Section III. The K #"'"' '* ' "* * # #""

IR dynamic, static, and crack arrest fracture toughness results obtained from several heats of pressure vess.*1 steel. When a given material is indexed to IR curve, allowable stress intensity factors can be obtained for this ma-the K terial as a function of temperature. Allowable operating limits can then be determined using these a!!owable stress intensity factors.

The RT.DT and, in turn, the operating Limits of a nuclear power plant, can be 3

adjusted to account for the effects of radiation on the properties of the re-wtor vessel materials. The radiation embrittlement and the resultant changes

' in mechanical prcpercies of a given pressure vessel steel can be monitored by

.su r vei l lance program in which a surveillance capsule containing prepared speci-mens 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-f t-Ib temperature, or the increase in the 35 mils of lateral expansion tem-perature, whichever results in the larger temperature shift due to irradiation, is added to the original RT t adjust it f r radiation embrittlement. This NDT adjusted RT is used to index the material to the KIR curve, which, in turn DT is used to set operating limits for the nuclear power plant. These new limits take into account the effects of irradiation on the reactor vessel materials.

• i 2-2 Babcock & VVilcor -

4

l l

l

)

1. %I'kVI.!!.LA.N CE PitiURNt DESCRIPfled The

.u rve i l l .en ce- pro. r.m t er os onee I cecprises eight s 4 r .:111 ance (aps.ules de-igaed to nonitor the effect of neutron aa<1 thernal environment on t he is.n te-r i .e l s of the re.ntor pressure vessel core region. The cat sules, whte:i were inserted int., t :.e react.ir vessel before inita.sl plant startup, w re positioned i n.i f de the re.*itsr vess l bstween the thermal shield .and the vessel w.all at tue locations t ,vn in figure 3-1. Six of the capsules, placed two in c..ch

• htldt r ( tthe ,

.t r e po%itle>tsed near the pe.ek axi.a) and azinuthal neutron flux.

The remaintne, tw., capsules are thermal aging capsules and . ire placed in an

. ires of e*,entially zero neutron flux. BAW-10006A includes a full description on .apsule 1.*c.a t t ens and design. I Capsule OCl-E was removed during the second refueling shutdown of Oconce 1.

This capsule cont.iined Charpy V-notch impact and tensile specimens fabricated ot SA102. c.r

• Ndified Steel, weld metal and car' relation steel. The specimen

.ontained in the tapsule are described in Table 3-1, and the chemistry and heat

• treatment of the wurveillance material in capsule OCl-E are described in Tab!*

l-2.

The capsule also contained longitudinal Charpy V-notch specimens frori correla-tion material obtained f rom plate 02 of the t'SAEC Heavy Section Steel Techno!-

oity Progr.im.

Th!= 12-inch-thick plate of ASTM 533. Grade B. Clas.. I steel was produced by the Luken Ste el Company (hear A-1195-1) and heat-treated by combus-tion Engineering. The chemistry and heat treatment of the correlation m.sterial are described in Table 3-3.

All test sp.rcimens were machined f rom the l#.-thickness locatic, of the plates.

Charpy V-notch and tensile specimens f rom the vessel material were oriented with their longitudinal axes parallel to the principal rolling direction of the plate; the specimens were also oriented transverse to the principal rolling direction. C.apsule OCl-E contained dosimeter wires, described as follows:

3-1 Babcock & Wilcox 4

L

N s,i ne t e r wi re Snielding I'- Al .a l lany Cd-Ag alloy Np-Al alloy Cd-A4 allev 1 Nickel Cd-Ag .allev 0.66: Co-Al . alloy Cd 0.66; Co-Al . alloy None

'Fe None l

Thermal monitors of low-melting eutectic alloys were included in the c apsule. The cutectic alloys and their melting points are as follows:

l Alloy Melting point,J 90: Pb, $1 Ag, SI Sn 558 9 7. 's : Pb, 2.5 Ag 580 j 97.51 Ph. 1.5% Ag, 1.0% Sn 588 1 1.c.id 621 Cadmium 610 I l

Table 3-1. Specimens in Surveillance Capsule OCI-E

)

No. of specimens

• Liter!.nl description Tensile Charpv .

held metal WF-112 4 3 Heat-affected zone (HAZ) I Heat A - C3265-1 (longitud.) 0 8 I Baseline material Heat A - C3265-1 longitud.) 4 8 Heat A - C3265-1 (transverse) 0 4 Cnrrelation HSST plate 02 0 8  %

Total per capsule 8 36 I

9 t

l 1

l 3-2 Babcock & Wilcox j l

i e

Tabic 3-3. Chemistry and Heat Treatment of Correlation Material - Heat A-1195-1. A333 Crade B, Class 1 (HSST Plate 02)

Chemical Analysis (1/4T)("}

Element Wt %

C 0.23 Hn 1.39 P 0.013 S 0.013 Si 0.21

, Ni 0.64 Mo 0.50 Cu 0.17 Heat Treatment

1. Normalized at 1675F t 75b.

2. 1600F 2 75F for 4 h/ vater-quenched.

3. 1225F 2 25F for 4 h/ furnace-cooled.

4. 1125F ! 25F for 40 h/ furnace-cooled.

• ORNL-4463.

I 1

l 1

4 4

W I

I 3-4 Babcock & Wilcox

a Tal, le 3-2. Chemistry and Heat Treatment of Surveillance Materials Chemical Analysts

, Heat "a*cid metal F. l eme n t C-3265-1 WF-ll2 C 0.21 0.075 Mn 1.42 1.'30 P 0.015 0.016 5 0.0?$ 0.006 Si 0.23 0.60 Ni 0.50 0.58 Mc 0.49 0.51 Cu 0.13 0. 32 Keat Treatm.nt ifca t Temp, Time.

No. _ F h Coo'ing C-3265-1 1600-1650 9.75 drine quench 1200-1225 9.5 Brine quench 1100-1150 40.0 Furnacc cooled aF-112 1100-1150 31.0 Furnaca cooled 3-3 Babcock & Wilcox s s.. ., ew a .n -.,e d

Figure 3-1.

Reactor Vessel Cross Section Shwing Surveillance Capsule Locations j Surveillance Capsule Holder Tubes - Capsules OCI-C.

l OCI-3 f

/

/ . ..

/

' e tb \

s

' e l ee e '

s e e e ' '

e e e e ' , ', , -

e s eo ee w

,r e se e e e 4p g__-  ; . - - -

h -y

- e s e /

) ee e

e e se e e 1k

/ e e , ,

g oo /

/,'

f s ce o e /

y N e /

N

//s el : 6 N . /

/

Surveillance Capsule h Holder Tubes - Cap- I Surveillance Capsule l sules OCI-A, OCl-B ,

i  ;

Holder Tube - Capsules Z OCl-E, OCI-F, OCI-G*, *, .

OCI-H*

i

• Thermal aging capsules. {

l 3-5 Babcock & Wilcox l

i

s

4. PREIRRADIATION TLSTS Unirradiated material was evaluated for two purposes: (1) to establish a baseline of data to which irradiated properties data could be referenced, and (2) to determine those materials properties to the extent practical from avail-able material, as required for compliance with Appendixes G and H to 10 CFR 50.

4.1. Tensile Tests Tensile specimens were tabricated from the reactor vessel shell course plate and wcld metal. The substre specimens were 4.25 inches long with a reduced section 1.750 inches long by 0.357 inch in diameter. They were tested on a 20,000-lb-load capacity universal test machine at a crosshead speed of 0.005 inch per minute. Test conditions were in accordance with the applicable re-quirenents of ASTM A370-72. For each material type and/cr condition, six specimens in groups of t.'see were tested at both room temperature and 600F.

An LVDT-type clamp-end screw-on extensoneter was used to determine the 0.2%

yictd point. The tension-compression load cell used had a certified accuracy of hetter than s 0.5% of full scale (10,000 lb). All test data for the pre-Irradiation tensile specimens are given in Appendix B.

4.2. Impact Tests Charpy V-notch impact testswereconductedinaccordancewiththerequirementb of ASTN Standard Methods A370-72 and E23-72 on a remote controlled impact test-er certified to meet Watertown standards. Test specimens were of the Charpy V-notch type, which are 0.394 inch square and 2.165 inches long.

Prior to testing, specimens were temperature-conditioned in a combination re-sistance-heated /carben dioxide-cooled chamber, designed to cover the tempera-ture range from -85 to +550F. The specimen support arm, which is linked to the pneumatic t ransfer mechanism, is instrumented with a contacting thermo-couple allowing instantaneous specieen temperature determinations. Specimens were transferred f rom the conditfor.. g chamber to the test f rame anvil and pre-cisely pretest-positioned with a fully automated, remotely controlled apparatus.

4-1 BaM & Mca f

Trar.sfer times were less than 3 seconds and repeat within 0.1 second. Once the specimen was positionad, the electronic interlock opened, and the pendulum was released from its preset drop height. After failing, the specimen, the hammer pendulum was slowed on its return stroke and raised back to its start Position. Impact test data for the unirradiated baseline reference materials are presented in Appendix C.. Tables C-1 through C-5 contain the basis data which are plotted in Figures C-1 through C-5.

t t

t t

'l '

1 l

l l 4-2 Babcock & Wilcox -

I

5. POSTIRRADIATION TESTS 5.1. Thermal Monitors Surveillaace capsule OCl-E contained three temperature conitor holder tubes, each containing five fusible alloys with different melting points ranging f rom 558 to 621F. All the thermal monitors at 558, 530, and 588F had melted, while these at 610 and 621F remained in their original configuration as initially placed in the capsule. From these data it was concluded that the irradiated specimens had been exposed to a maximum temperature in the range of 588 to less than 610F during the reactor vessel operating period. There appeared to be no significant temperature gradient along the capsule length.

5. 2. Chemical Analysis The unirradiated base metal (Heat 3265-1) and the weld metal (WF-112) were analyzed for nickel, copper, phosphorus and sulfur contents to verify original mill test report data. The results from chemical analyses are reported in Table 5-1.

5. 3. Tenstle Test Results The results of the postirradiation tensile tests are presented in Table 5-2.

Tests were performed on specimens at both room temperature and 580F using the same test procedures and techniques used to test the unirradiated specimens (secticn 4.1). In general, the ultimate 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 is within the range of changes to be expected for the radiation environment to which the specimens were ex-posed.

The results of the preirradiation tensile tests are presented in Appendix B.

5-1 Babcock & Wilcox

5.4. Charpy V-Notch I= pact Test Results The test results from the irradiated Charpy V-notch specimens of the reactor vessel beltline material and the correlation monitor material are preserited in Tables 5-3 and 5-4 and Figures Sr-1 through 5-5. The test procedures and techniques were the same as those used to test the unirradiated specimens (section 4.2). The data show that the material exhibited a sensitivity to ir-radiation 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 g$ en in Ap-pendix C.

Table 5-1. Chemistry Data on Unirradiated Oconee 1 RVSP Material Material type / position, wt %

heat No. Ni Ce P S Base metal /C-3265-1 0.50 0.17 0.007 0.027 Weld metal /WF-112 0.59 0.32 0.016 0.016 i

?

I 5-2 Babcock &Wilcox .

I

Table 5-2. Tensile Properties of Capsule OCI-E Base Metal and Weld Metal Irradiated to 1.5 = 1018 nyt Test Strength, psi Elongation, %

Specimen temp, Yield Uniform ID No.

Ult. Total Red'n of F (YS) (UTS) (UE) (TE)gg area, %

Base Metal - lleat C-1265-1. Longi tudinal AAE 704 RT 67,430 89.510 10.97 22.06 68.3 AAE 721 RT 67,430 87,910 10.51 21.60 69.8 Mean, E 67,430 88,710 10.74 21.83 69.G4 Std dev'n, 6 0 800 3.23 0.23 0.75 AAE 708 580 61,440 88,810 13.35 26.13 AAE 730 67.4 580 61,440 89,910 10.84 21.40 69.8 Mean, i 61,440 89,360 12.35 23.77 68.6 Std dev'n, 6 0 550 1.505 2.365 1.2 Weld Metal - WF-112 AAE 109 RT 77,920 93,910 12.82 24.66 AAE 120 56.3 RT 79,920 95,710 12.32 23.04 57.1 Mean, E 78,920 94,810 12.57 Std dev'n, 6 23.85 56.7 1,000 900 0.25 0.81 0.4 i

AAE 110 583 69,930 91,860 10.12 16.2 44.1 AAE 126 580 69,930 92,410 9.49 l 15.73 45.7 Mean, E 69,930 92,135 9.8L 15.97 44.9  ;

Std dev'n, 6 0 275 0.315 0.235 0.8  !

I" TE values calculated from chart record. ,

i 1

I l

l 1

l 5-3 Babcock & Wilcox

~

l l

z e

Table 5-3. Charpy Impact Data for Capsule OCI-E Base Metal Irradiated to 1.5 = 1038 nvt Test Abs Lateral Shear Specimen temp, energy, expans, fracture.

ID No. F ft-lb 103 in.  %

Heat C-3 65-1. Longitudinal AAE 708 325 139 77.5 100 730 201 131.5 66 100 721 120 110 70.5 55 746 106 92 65 35 709 92 72 57 18 723 73 52 41 8 706 70 66 52 15 753 63 44 36 4 Heat C-3265-1. Transverse AAE 611 203 107 69 100 610 147 74 56 75 630 97 50 40 12 613 79 44 37.5 8 HAZ - Heat C-3265-1, Longitudinal AAE 403 318 86 67 100 423 199 83 54.5 100 439 120 65 54 92 436 100 74 54 85 418 63 70 45 75 452 27 67 46 25 435 13 39 29 20 420 -21 36 26 10 5-4 Babcock & Wilcox ;

Table 3-4 Charpy Impact Data for Capsule OCI-E Weld Metal (WF-112)

Irradiated to 1.5 = 1018 nyt Test Abs Lateral Shear Specimen temp, energy, expans, fracture, ID No. F ft-lb 103 in.  %

AAE 059 315 55 49.5 100 017 198 52.5 45.5 96 003 170 36 35.5 5 ',

041 170 44 34 'a0 044 152 52 49.5 85 048 121 42 30 35 058 89 33 29.5 12 056 62 27 22.5 8 Table 5-5. Charpy Impact Data for Capsule OCI-E Correlation Monitor Material Irradiated to 1.5 a 1038 avt Test Abs Lateral Shear Specimen temp, ID No. F energy, expans, fracture.

ft-lb 103 in.  %

AAE 929 428 104 73.5 100 935 321 111 71 100 966 260 98 69 95 964 197 78 52 65 908 154 57 42.5 30 927 120 50 39 30 971 99 36 30 6 967 62 28 22.5 2 5-5 bM & Mox

~ - ~ . - . . ..

M

Figure 5-1. Charpy Impact Data From Irradiated Base Metal, Longitudinal Orientation s a y y a

• 1- &

o* .M . .

1_

Jn .

n L

• n n e E a a n n M u s 3 g y a a i sg C "

'N $

• I f.

, r.r

._ . - = _ . . . - - _ _ - - - - -..- __ . ..-

I e

,' QT. .

og I I I i t i I f f a a u r i s s a a s y a s

. __.ta. f a_ simp _.

g, . I. , LL dcLcnis.d ,

r, t35 sua M_

pt, . f,, (So s ts) .2D 7 .

(.ug (am _l'sR f t-Ib W - 87.,10f- .U:sst.catimatz) -

IM -

.h IT -

.f.

y w ai -

t.

.! so -

40 -

% tets SA302 Cr 8 Mod _.

20 - tusevassen Lonattsdtnal Fur.we 1 Ar+ta ovr._._,'

luar Ibsese C-3265 1 i . . .

e .

• -w o e ao un no 2no 2o m so :w <n im romaar . t 5-6 Babcock & Wilcox i

n r w r + _-

--.-m--- - .

v.

o Figure 5-2. Charpy Impact n?ta From Irradiated Base Metal, Transverse Orientation D' . . , , , .

~

.e n, .

l

_1 j v .--_---.-------- - - - - - -._ - - - - - - - _ -..

7 v b. -

5 3 5 5 g g 5 5 t 3

.o i 3 -

5

? ----_--- -- - _ - - . . - _ . . . .

T.

e

/ .T*

e

,, , , , a i e i e i i .

p r. 8 s i a a

. 'A ta'gm i a y e 3, , t,, w e A , . ,_ wa in815 sue 93F  !

IV - 5, ($4 er-tal 96F .

(.U2 (,y _107 fr-It,

. IV - 87., .j6.f_ Wa t es t i ma g) i  ;

5 17 -

t 15 -

U y

w F -

t

~~ w -

40 - .

\'

% ness u w c . m he-29 - batovarse Transverse---

Ftmence 1.5E+14 avt Near beta C-3265-1 0 ' e i i . . . ,

m ** 3

• uo 00 no in m m m v.o a Test Tamnarums, F 97 Babcock & Wilcox I

l l

1

i i

a Figure 5-3. Cnarpy Impact Data From Irradiated o.se Metal Heat-Affeeted Zone, Longitudinal Orientation 3, >

> > > r

- n .

e -

,.1 ,, . . . . - - - - . .

.Y 25 -

, - i g, a a e a a a a a 1 a a

.M e e s . e e a i i

~

r . 's - -

t e

• e

. rw .

.'l?' -

~

1 r.

,y, . . . . . . . .

$ )

30 . . . . . . . .

l

. , , i JI A '49927 I I

he A , . r . < n..a g , f,,

in (35 eus 17F 3s,c .. I, (50 es ts) _20F ,

(-U5E tavel ji f t-lb__

. le . ti,, . lor (best cettmate) .

E IM -

2

.: IT ~

.Y Ir .  :

t

~ p -

e no .

%ftenat KAlo? Cr W 20 . Outstavsom MA2. fang Fusence IJE+18 nyt._ _ ,_~

o lisat Ibses a C-32hS-1.__. l m w e o so uo tse in m m m n en in, rim.aw, s I

5-8 Babcock 3. Wilcox .

l

l l

1 i

1 F'gure 5-t.. Charpy impact Data From Irradiated k' eld Metal eI g , y a y g g , s

• • ?$

• j ., . . _ _ . . _ - - - _ - - - . - ..- -. ---- _ -.---- .

7

- 7. -

a a a a a e e a a e a s s 5 5 I 5 5 8 I f

g e- -

.w -

-.g___---____._ ___ ._.

JP "o-r __ f f a f f f f i f e a a u a a e g a , , , j

. . _F A_1*___- il

.< - '., u a.u m a i 73 d 55 mi _ID F 3+f - .,, ISO s o tal I7-F e

as .d fan) % f t-!b

, tv - af,, _t j a r (be_.t es t isu t e ) .

I IM -

~

.; IT -

~

m 1 y -

t.

t

~ y . *

--_----.___________J_ ____________!______..

w- *

%reein Wld NeuJ__ __

g , hisuratsas --

Ftw=ca 1 5E+18 nvt__ .' l nia, m . wr-i i 7 _ _

e > - . . . . . , , .

80 w

• m m o so no rm zes m m en fast femtearues. 8 1

5_9 Babcock 3. Wilcox

l l

l l

l i

Figure 5-5. Charpy Topact Data From Is radiated  !

Correlation Monitor Material try) .

" n -

s.

j w _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

=

Y 2',

w -

• g a f a a a 9 f a # n n 5 I 3 a g 3 5 a 3 a l

d. .

• 427 M i .n -

.?

• • ' . fy.C -

e .

3 . ____.,_- _ _ _- _ _ _ _ _ _ . . _ _ _ _ _ _ _

r.

r J .020 -

?.

.e.e I I i f f f f f f e e I u a u a e s 6 5 SATA Stattury ,

gy 1 ,. .uar A*rermin-A s -

To .35 'ul ll8F Ju . T , (50 av-ts) 138F .

(.U2 tavs) 111 ft-lb 3

1:e0 - IT , _78F (bu t estimate) .

5 l20 -

IL

- N 100 -

427 e ,

5

% Y w M . -

7.,

t g . i I

~

alartesat 113ST PL-02

, Onsastation -- I Ftsence 1.5E+14 nyt naar mmnen A-1195-1 0 '

-w -w o w' s'o

= n e i 12o 160 20o 2e -m 520 w e l Test Temmarums. F  !

l 5-10 Babcock & Wilcox i

I l

's

6. NEUTRON DOSIMETRY 6.1. Introduction A significant aspect of the surveillance program is te provide a correlation between the neutron fluence above 1 MeV and the radiation-induced property changes noted in the surveillance specimen. To permit such a correlation, activation detectors with reaction thresholds in the enargy range of interest were placed in each surveillance capsule. The properties of interest for the detectors are given in Table 6-1.

Because of a long half life of 30 years and an effective energy range of 0.5 MeV or greater, only the measurements of 137 Cs production from fission reac-tions in 237 Np (and possibly 230 U) are directly applicable to scalytical de-terminations of fast neutron (E i 1 MeV) fluence over cycles 1 and 2. The other dosimeters are useful as corroborating data for shorter time intervals and/or highe. energy fluxes. Comparisons of measured data (cycles 1 and 2) to calculated results reauire the use of calculated data previously reported for cycle 1 to augment the cycle 2 analysis presented here.

The energy-dependent nertron flux is not directly available from the activa-tion detectors because the process provides the integrated effect of the neu-tron flux on the target material as a function of both irradiation time and neutron energy. To obtain an accurate estimate of the average neutron flun incident upon the detector, the following parameters must be known: (1) the operating history of the reactor, (2) the energy response of the given detec-tor, and (3) the neutron spectrum at the detector location. Of these param-eters the definition of the neutron spectrum is the most difficult to obtain.

Essentially, the tw' o means available for obtaining the spectrum are iterative I

unfolding of experimental foil data and analytical methods. Due to a lack of sufficient threshold foil detectors satisfying both the threshold energy and ,

the half-life requirements necessary for a surveillance program, iterative un- , ,

folding could not be used; this leaves the specification of the neutron spectrum i to the analytical method. 1 l

6-1 Babcock s.Wilcox

W 6.2. Analvtical Approach Energy-dependent neutron fluxes seen by the detector were determined by dis-crete ordinate solution of the Boltzmann transport equatien. Specifically. '

3 ANISN , a one-dimensional code, and DOT'*, a two-dimensienal code, were used to calculate the flux at the detector position. In both codes the Oconee system was radially modeled from the coretout to the air gap outside the pressure vessel. The model included the core with a time-aieraged radial power distri-bution, core liner, barrel, thermal shield, pressure vessel, and water regions.

Including the internal components enables the analytical cethod to accouot for-tne castortions of the required energy spectrum by attenuation in these com-penents. The ANISN code used the CASK 5 , 22-group neutron cross sectional set with an Sg ordar of angular quadrature and a P3 expansion of the scattering matrix. The problem was run along a radius across the core flats. Azimuthal variations were obtained with a DOT r-theta cniculation that modeled one-eighth of a plan view of the core and included a pin-by-pin, time-averaged power dis-tribution. The' DOT calculation used S6 quadrature and a Pg cross section set g

d(rtved from CASK.

Fluxes calculated with this DOT model must be adjusted to account for the lack of P 3 cross-sectional detail in calculations of anisotropic scattering, a per-turbation caused by the presence of the capsule, and the axial power distribu- -

tion. l The first two items are both energy- and position-dependent. .. P 3fP3 correction factor was obtained by comparing two ANISN 1-D andel calculations in which only the order of scattering was varied. The capsule perturbacion factor was obtained from a conarison of two DOT x-y model calculations - one with a capsule explicitly modeled with SS304 cladding, Al filler regiot., d carbon steel specimens and the other with water in those regions. The effect of axial power distribution was determined from a previous DOT r -z model using an estimated average (axial) over three fuel cycles. The net results from thase para. meter studies was a flux adjustment factor K (Table 6-2), which should be applicable to the appropriate dosimeters in all 177-FA surveillance programs in which the capsules are located at a radius of 211 centimeters from the core center and at 11* from a majce axis. {

The calculation described above provides the neutron flux as a function of

[

energy at the detector position. These calculated data are used in the fol-lowing equations to obtain the calculated activities used for comparison with the experimental values. The basic equation 6 for the activity D (in pCi/g) is '

6-2 Babcock & Wilcox r

CN M -A t -n Fj (1-e g 3)e i(T-t3)

10. It[E n( $( )

O i A 3.7 ($_g) j=1 C = normalizing constant (ratio of measured to calculated flux),

N = Avagadro's number, An =

atomic weight of target material n.

f g= cither weight fraction of target isotope in nth miterial or fission yield of desired isotope, o" (E) = group-averaged cross sections for material n (listed in Table D-4),

(E) = group-averaged fluxes calculated by DOT analysis, F) = f raction of full power during jth time interval t),

)g = decay constant of ich naterial, t) = interval of power history.

T = sum of total irradiation time, i.e., residual time in reactor and wait time betweers reactor shutdown and counting, t

3 = cumulative time fron reactor stattup to end of jth time period, i.e.,

t

1) = k=1 k'

The nornalizing constant C can be obtaiacd by equating the right side of equa-tion 6-1 to the measured activity. With C specified the neutron fluence greater than 1 MeV can be calculated from 15 MeV H 3(E e 1.0 MeV) = C [

E=1

$(E)

[ F)t) j=1 (6-2) where M is the number of irradiatian time intervals; the other va2ues are de-fined above.

6.3. Results Calculated activities are compared to measurements of the dosimeters in Table 6-3. The 337 Cs data show that fast flux (E > 1 MeV) is somewhat underpredicted

( .10%) by the analytical model described (if one assumes that the calculated flux spectrum is correct). Such an agreement is probably within the uncertain-ty limits of this analysis. However, for conservatism a flux normalization factor of 1.1 is recommended for fast neutron calculations near the pressure 6-3 Babcock a Wilcox

vessel. The 103 Ru activities (because of a short half-lafe) are indicative of the validity of the analytical model representation of the latter part of cycle 2. 5"Mn and 57 Co activities show an overprediction (s25%) of 2 MeV or greater flux by the analytical model. The significance of these results to the fast flux calculation is somewhat lessened since approximately 40% of the neut rons with E greater than 1 MeV are in the I to 2 IteV range (Table 6-4) .

The possibility exists that an inherent overprediction of flux occurs in the analytical model and that measurements of activities from fission wires sre high. It is also possible that the core leakage flux over the latter part of cycle 2 was less chan c'e average for all of cycle 2; this could account for C being less than one since calculated data are based on an average leakage flux and ceasured data for short-lived isotopes are based on the latter part of the cycle.

Future surveillance capsule data should clarify this suppost-tion.

The agreement between C values for long-lived and short-lived dosimeters in-dicates t hat the fission dosimeter normalization constant reported in referenre 2 is in error and lends credence to the sugge sted occurrence of incorrect ac-tivity measurecents after cycle 1. Therefore , cycle 1 fast fluxes have been recalculated with a 1.1 normalizing constant and are presented with cycle 2 data in Table 6-5. Average fast flux over cycles 1 and 2 was extrapolated to a 40 year lif atime with a 0.8 use factor to obtain 21uence at the pressurc vessel wall of 1.7 = 1019 n/cm2 (2568 HWt); this can be compared to a predicted ,

value of 2.9 = 1019 n/cm 2 (based on an estimated power distribution over *hree cycles) reported in reference 7 for a reacto at 2772 MWt. An average lead factar of 1.7 differs from a predicted value 9 of 1.4 primarily because of flux pertubation due to the presence of the capsule (a factor not previously con-sidered).

An extension of the flux range down to 0.1 MeV was calculated based on the same normalization factor (1.1) being applicable. Since no dosimeter reactions l cover that  !

range, additional uncertainty is introduced into the results. The data, which are included in Table 6-5, indicate that fluences are essentia11 f

doubled when the energy range is exten.'ed from 1 to 0.1 Mev.

t 6-4 kbcock & Wilcox j

J e

aue

a Table 6-1. Surveillance Capsule Detectors Threshold Isotope Detector reaction energy. MeV half-life -

5'Fe(n.p)54Mn 2.0 303 days 5'N1(n.p)Seco 2.5 71.3 days

' 3 D(n.f) 3 37Cs 1.5 30 years 237Np(n,f)I37Cs 0.5 30 years 23D(n.f)l03Ru 1.5 39.5 days 237pp (n,gyl03Ru 0.5 39.5 days Table 6-2. Flux Adlustment Factor Energy Axial power Capsule range, MeV factor p3 j,I perturb'n _Y_

>0.1 1.1 1.22 1.40 1.88

>1 1.1 1.23 1.20 1.62

>2 1.1 1.24 1.04 1.42

>2.5 1.1 1.25 0.96 1.32 i

I I

i 6-5 Babcock & Wilcox s

.]

e Table 6-3. Dosimeter Activations fter Cycle 2 A

B - Calculated activity ( }. .C1/g measured *^

activity,(* cycles 1 and 2 normalization Reaction pCi/R Cycle 2 Cycle 1 decayed (C) coastant 5"Fe(n.p)5"Ha 536.5 595.9 397 729.3 0.74 54Ni(n.p)50Co 975.3 1295.5 685 1266 0.77 2 )dC(n.f)337Cs 1.943 1.100 0.638 1.719 1.13 2 37Np(n.f)337Cs 9.32 5.709 3.19 8.799 1.06 23"U(n.f)'03Ru 52.65 53.18 23.4 53.19 0.99 2*'Np(n.f)10iRu 254.3 246.6 106 246.6 1.03

Average of four dosimeter wires from Table D-3.

I Values from reference 2 were multiplied by 1.1 to account for axial power distribution (not included). 54 Mn and 59Co values were converted to gram of target (reference 1 values were per gram of dosimeter).

'c) Cycle 1 values were decayed over the time interval from the end of cycle I to the end of cycle 2 and then added to cycle 2.

Di ,2= D2+De i l

i 6-6 Babcock & Wilcox  ;

i e

a

.s Table 6-4. Normalized Flux Spectra. E > 1 MeV Energy range. In water near 235g MeV pressure vessel vall fission 12.2 - 15.0 0.0015 0.0002 10.0 - 12.2 0.0063 0.0013 8.18 - 10.0 0.0181 0.0052 6.36 - 8.18 0.0499 0.021 4.96 - 6.36 0.0906 0.051 4.06 - 4.96

• 0.0784 0.052 3.01 - 4.06 0.1159 0.159

• 2.46 - 3.01 0.1200 0.132 2.35 - 2.46 0.0389 0.034 1.63 - 2.35 0.1506 0.178 1.11 - 1.83 0.2832 0.323 1.0 - 1.11 0.0466 0.044 ,

Total 1.000 1.000 Table 6-5. Fast Neutron Fluence Cycle 1 Cycle 2 Cycles 1 and 2 309.3 days 291.2 days 600.5 days Lifetime 32 years g ,)

Capsule Center Fast flux (E > 1 MeV) 2.13 + 10 3.76 + 10 -- --

Nvt (E

• 1 Me\) 5.7 + 17 9.5 + 17 1.5 + 18 2.9 + 19 Flux (E > 0.1 MeV) 4.27 + 10 7.53 + 10 Nvt (E > 0.1 MeV) 1.1 +- 18 1.9 + 18 3.0 + 18 5.9 + 19 Pressure Vessel Wall Fast Flux (E > 1 MEV) 1.37 + 10 2.08 + 10 -- --

4 Nvt (E > 1 MeV) 3.7 + 17 5.2 + 17 8.9 + 17 1.7 + 19 Flux (E > 0.1 Mev) 2.78 + 10 4.22 + 10 Nvt (E > 0.1 MeV) 7.4 +17 1.1 + 18 1.8 + 18 3.5 + 19 (a) Extrapolation of average flux in cycles 1 and 2. j l

l l

. 1

~

l 67 Babcock & Wilcox P

+

Table 6-6. Predicted Fast Fluence in Pressure Vessel for 10 ERY(a)

Inside Outside i vall T/4 3 T/4 v311 Average fast flux, n/ce 2-s 1.8 + 10 1.0 + 10 2.4 + 9 1.0 + 9 Fast fluence n/cm2 5.7 + 18 3.2 + 18 7.6 + 7 3.2 + 17 (a)Ti.ese data ar.* based on the hypothesis that pressure vessel fluence is proportional so calculated average core leakage fluxes Tor cycles 3 and 4 combined with specimen analyses for cycles 1 and 2. Subsequent fuel cycles were assumed to be the same as cycle 4.

e a

9 l

l

. l l

l 1

5 I l

l t

. l Y

l 6-8 - hM a m% ' )

~

i

4 Figurr, 6-1. I,redicted F, t ..utrosi j3

' ' ' ' Yff"un !.oi .s t i'."" llit our.fi R ... . g g,"""el L II for First to l'Y (Oct>nt e 1) 6.0 5.6- 5.1 lol" nvt 5.2 -.

p 4.8 -

a 4= 4.4 -

l 4.0 -

~

^

3.6 -

D y 3.2 - 3.2 = late ,,t

,o e g ,c'

- 28 - $4 n

, to*O*

ei 2.4 v s

,I 2.0 ~

m.

E I.6 -

Cgo" 1.O

• U 3

\l N 1.2 -

ce t 0.8 - 7.6 1017 nyt g

o n

• 0.4 ~ 3/4T Location 3.2 1017

p. Outalito Surface

( "6 o 2 3 4 0 6 7 8 9 30 1.!?y

7. DISCUSSIO*i 0F CAPSL1E RESL1TS 7.1. Preirradiation Property Data A review of the unirradiated properties of the reactor vessel core belt region indicated no sigafficant deviation from expected properties except in the case of the upper shelf properties of the wt.d metal. Based on the predicted end-of-service peak neutron fluence value is. 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 the criteria in effect at the time the program was designed for Oconee 1. The applicable selection criterion was based on the u, irradiated properties only.

7_.1._1 r rad ia t ed P rope r tLDa ta 7.1.1. Tensile Properties T.ible 7-1 cc.mpares irradiated and unirradiated tensile properties At both .

room temperature and elevated temperature, the ultimate and yield strength changes in the base metal as a result of irradiation and the corresponding changes in ductilit y are negligible. There appears to be some st rengthening, as indicated by increases in ultimate and yield strength and similar decreases in ductility properties. All changes observed in the base metal are so small as to be considered within experimental error. The changes at both room tem-perature and 580F in the properties of the weld metal are greater than those observed for the base metal, indicating a greater sensitivity of the weld 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 period in service life.

7.2.2. Impact Properties The behavior of the Charpy V-notch impact data is more significant to the cal-culation of the reactor system's, operating limitations. Table 7-2 compares 7-1 Babcock & Wilcox

the observed changes in irradiated Charpy impact properties with the predicted changes as Shown in Figures 7-1 through 7-5.

The 50-f t-lh transition temperature shif t for the weld metal and tne base cetal was in Aood agreement with the shift that would be predicted according to Regu-latory 8;uide 1.99. A similar comparison of the shift of the correlation moni-tor material shows good agreement. The less-than-ideal comparison nay be at-t ributed to the spread in the data of the unirradiated material co=bined with

.i minimum of data points to establish the irradiated curve. Under these con-ditions. the comparison indicates that the estinating curses in RC 1.99 for medium-copper materials and at medium fluence levels are reasonably accurate for predicting the 50-f t-Ib trai.sition temperature shif ts. The estimating curves for high-copper material at medium fluence levels are also in good agreement with the observed data.

The increase in the 35-mil lateral expansion transition temperature is com-pared with the shift in RT NDT curve data in a manner similar to the comparison made for the 50-ft-th transition temperature shift. These data show a behav-for similar to that observed f rom the comparison of the observed and predicted 50-ft-Ib transition data.

The data for t he decrease in Charpy USE with irradiation showed a poor agree-

~

ment with predicted values for both the base matals and the weld metal. Ilow-ever, the weld metal data compares well with the predicted value in view of '

the lack of data for high-copper-content weldments at low to medium iluence q values that were used to develop the estimating curves.

The RT shifts shown are in good agreement with those predicted from Regula-NDT tory Guide 1.99 at the fluence level of this capsule. This good agreement is probably attributable to two factors. First, the fluence level is approaching that for which the bulk of the data used to generate the prediction curves was located. Second, the re-analysis of the chemical composition of the irradiated materials gives a acre reliable value for comparison of observed versus pre-dicted values.

Results from otner capsules indicate that the RTN DT ** * '"# "E ""#"** ***

greater inaccuracies at the very low neutron fluence levels ($1 1016 n/cm2 ).

This inaccuracy is attributed to the limited data at the low fluence values and of the fact that the majority of the data used to define the curves in RG 1.99 are based on the shif t at 30 ft-lb as compared to the current 7-2 Babcock & Wilcox o

require =cr.* of 50 ft-lb.

For most 4terials the shif ts measured at 50 ft- e/

• 35 MLE are expected to be higher '.an those measured at 30 ft-lb. The sig-tf-ic.ince of the shifts c' 50 f t-lb and/ < 35 !!LE is not well understood at pres-ent, especially for materials having USEs that approach the 50 ft-lb level and/or the 35 HLE level. .s 4 1als with this characteristic may have to be evaluated at transitica energy levels lower than 50 ft-lb.

The design curves for predicting the shift at 50 f t-lb/ 35 MLE will probably be ondified as data become available; until that time, the design curves f o r p re-dicting the RT NDT shift as given in Regulatory Guide 1.99 are considered ade-

<;uate for predict!ng the RTNDT " ' ' ** ** "" "# * '"' ' ' " " "'# ' " "

available and will continue to be used to establish the pressure-temperature ep. rational liuitations for the irradiated portions of ti.e reactor vessel.

The 1.ick of good agreement of the change in Charp> ddE is further support or the inaccuracy of the prediction curves at the laver fluence levels. Although the prediction curves are conservative in that they predict a larger drop in upper shelf than is observed for a given fluence and copper content , the con-servat isi can unduly restrict the sperational limitations. These data support the contention that the USE drop curves will have to be codifis-d as rwre re-liable data become available; until that tice the design curves used to predict the decrease in USE are conservative.

4 7-3. Babcock a.Wilcox o

Table 7_-l. C_omparison of Tensile Test Results Room temp test Elevated te=p test L'nirr Irrad 600F 59sF Base Metal - C-3265-1. longitudinal Fluence, 1033 n/cm2 (- 1 MeV) 0 1.5 0 1.5 L'I t . tensile strength, ksi 86.1 88.7 64.4 89.3 0.2% yield strength, kai 64.3 67.4 58.0 61.4 Elongation. % 26.5 21.8 27.3 23.7 RA. I 68.0 69.0 70.0 63.6 a e

- l d Me t a l - WF-l l 2 Fluence. 10 1 '8 n/cm2 (, 1 MeV) 0 1.5 0 1.5 l'I t . tensile strenf,th, ksi 80.5 94.8 80.8 92.1 0.2% yield strength, kai 63.3 78.9 56.4 6 ). 9 Elonnatlon. I 30.9 23.8 24.4 15.9 RA. : 63.0 56.7 61.0 44.9 9

(a)The differences in test temperatures are the result of t he unirr.adiated si'ecimens being tested before removal and evaluation of the capsule t herm.nl runitors.

}

i I

. l 7-4 Babcock & Wcox .

~j.

i i

9'

-f Table 7-2. Observert Vs Predicted Changes in Irradiated C*u r-*, impact Properties '

~

".a t e r i r.1 05servd

• Predicted

!_nc r ea s e in 50-ft-Ib trans te=p, F fuse c:aterial (C-3265-1) longitudinal 33 An Transverse , 12 4S Ifeat-Mfected zone (C-3265-1) 33 48 Weld metal (WF-l'2) 12; 124 Correlatton material f.* 1195-1) t> '. ots

nc rease in 35-MI.E t rans temp, F

, luse m aterial (C-3265-1) longitud!nal ,

38 an Transverse 43 4d lleat-affected zon( (C- 3265-1 ) la 4d(b)

Weld metal (WF-!!2) 109 124 Correlation r:aterial (A-Il95-1) 63 60 Dyj;re.as e in _ Charry l'SE. f t-lb ILese naceri.nl (C-126 5-1) len,:itudinal 3 24 Transverse ND 18 IIcat-affected zone (C- 3265-1 ) 27 19 Weld metal (WF-112) 9 19 Correlat ion matertal (A-1195-1) 19 26 (N

Tnese values predicted per Regulatory Guide 1.99. Revision 1.

' Based on the assumptien that MLE as well temperati..e is used to control the shift in asR50-ft-lb transition

.D..

ND - Not determined.

s t

7-5 . Babcock a.Wilcox m

(

k

Figure 7-!.. Irradiated Vs UMrradiated Charpy Impact Properties of Base N tal, l.ongitudinal Ortentation If0 . > > . i i j.

is .

3 y ,, ___________

1.5 = 10 e avt J 25 . L'nitradiated .

n . . . . . . . . .

. 'A1 , , , , , , . . . .

^

.i.

i .'M -

g i *T = 5dF

; -1. 5 = 1018 nvt

'5 .04c .

e E .-I_____--_____...

t* .n20 -

Unirradiated

,,, i e i i e i i i , .

5 a s a a 3 i a s s 14r -

]V -

AUSE = 3 f t-15

. 140 - -

E T

i 1D -

I100 -

8, an 1.5 = lote nyt e.

1 60 -

( ) AT = 53F

.-_____ ___ ____...____________.._..L.......

40 . *

%Ttetat SA302 Cr 8 Mod f 20 -

Unitradiated hatutatsas Tna tt r ud + u_1_

% See above -

aint amaan C-326 5-1 o . . . . , , , , ,

-80 -# o e so a no 200 m m m m m Test Tammatung, F 7-6 Babcock & Wilcox e

,-v-

Figure 7-2.

Irradiated Vs Unitradiated Charpy Impact Properties of Base Metal. Transverse Orientattoi Im '

75 .

yu - - - - - - - - - - - . . -

= Untreadiated - 1.5

- - =- -1018 -__-._________

3n - ove n'- '

.080 ; -

• 1 . . . ,

E .tv.c -

~

f aT = 43r u el -

1. 5 = 10l * = vt -

O .oec

/y 1

E --------------.-- --__-

j 'u .

a Unitradiated

.7Y) ' ' ' . . e .

y__ '

=

lan-160-140 -

L s -

E 120 -

y 4 -

3 Iro -

' [.'USE=?

t.

F

{" ' AT = 32F %w t , 1.5 = 1018 nyt

=0 - ---- ------------------ -___. _

' Unirradiated %rteza SA107 cr n Pfnd i

Gnatstarsom . Transverse o, _

-m Futac See aboye Neatm,we C-3265-1

['

-=o o e . . .

ao .

120 leo a zoo m 520 Tier Tasnearwas. F wo wri 7-7 1 Babcock & Wilcox  !

1 I

Figure 7-3. Irradiated Vs Uniaradiated Charpy Impact Properties of Base Metal, Heat-Affected Zone. Longitudinal Orientation yn. .

.N thirradiated .

be y ________ _______________________.

Ie 1.5 = lots nyt T

+5 .

p a a A R f E G E e R E 5 a 3 3 3 5 5 5 5 5 inirradiated i .H +-_*

t.T = g g

~

.S r

." ,onc

_ _ =' _10:e _ _nvt_

1.

e ----- . _

t EP

. ope -

.e,e g I f i f f f E R R e a u u y a a s y y a s ,

IF - -

IM -

. le -

I'

& 120 -

3 3 ,,

t,'ni rradia t ed

., 100

f t.USE - 27 ft-lb .

y \f

,,, en --

-ir = ;3r .-. -

t

. ~ 60 - 1.5 = 10:e avt -

- - - _ _ . -- 1

- - - - _ - - . . . - - - . . . - - - - - . . . . . . . . . . . . I

%7talas, SA302 Cr 5 Mod ~

,, . . a.e. ,_ ., ~ , .,

ts t see .d,v.

Riat heare C-1265-1 l 0 a , e . . . i . . '

.ao .* e o

m so 12o les 2ao re suj 520 wi i Test Tevenaw, F l 7-8 Babcock s.Wilcox  ;

}

, Figure 7-4. Irrr.diated Vs 1lnirradiated Charpy impact Properties of 'a' eld Metal 17 '

7., .

J Untrradiated 1

jy ._ ________ _____ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _

1. 5 = I Ol e nyg J 2s .

n . . . . . . . . . .

.'850 . . . . .

s Unirradiated 5U

~

'T = 109F -

- c

,. . w .-___ . . _ -

I j ,

1.5 = 10:e nyg

. .e -

~!

,, g, e s , e i e . . e . .

W . . . . . . . . . . .

IV -

164 .

. IV -

?

7 f 17) -

HIT -

T w Sn -

f.T = 121.F Q <----- > -itJSE = 9 ft-Ib t

y .

's g I!ntrradiated -

%ttesat Weld Metal 1.5 = 1088 nyt hisevassam --

20 -

Fustect See above aga, we WF-112 e i . . . . . . . . . .

-80 -# o 40 so 120 160 2m 2e m m in <n its, Tammansu, F 7-9 Babcock s. Wilcox

l l'!xure 7- 5. Irradiated Ys Unirradiated Charpy Impact Properties of Correlation Monit.*r .u.aterial im , , , , , ,

• • 7 =

,f l'n t e r.adia t ed 3 .z ---- .-.

, 1.5 = 1038 nvt

.Y, 25 -

n = *

• W > . , , , , , , . ,

1 #

p ,rsn Unirradiated

~

.'.T = 63r j  ;

5 1.5 = 10 1 s nyt

.%r -

P I 4 .rg 6 , 1

~

1

, rg I f f f f I f I f a a 2ft' . . . . . . . . . . .

W ,

IM -

IV - t.1SE = 19 it-lb

/ 120 -

" w 3 IT, -

P.

L'nirradiated d 80 -

[- ST = 64r ( )

I

- g . 1.8 = 1038 nvt M - ~

%rteig HSST-PL-02 .

3 , Onstavatsam - l Ftuence See above Ikat Nwqnge Al19%1 _ ,

el ' . . f . . . f , ,  ;

M -W D 40 30 120 27) 2M E) Wo 160 290 W) its? Tewanarung, F l

7_10 Babcock s. Wilcox

8. DETERMINATION OF RCPB PRESSURE-TOHPERATURE LIMITS The pressure-teeperature limits of the reactor coolant pressure boundary (RCPB) of Oconee 1 have been established in accordance with the requirements of 10 CFR

50. Appendix G.

The methods and criteria employed to establish operating pres-sure and temperature limits are described in topical report BAW-10046.8 The objective of these limits is to prevent nondu: tile failure during any normal operating condition, including anticipated operation occurrences and system hy-drostatic tests. The loading conditions of interest include the following:

1. Normal operations, including heatup and cooldown.

2. Inservice leak and hydrostatic tests.

3. Reactor core operation.

The major components of the RCPB have been analyzed in accordance with 10 CFR 50 Appendix C.

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 temperatures (due to mechanical loads resulting from bolt preload), this region largely controls the pressure-temperature limits of the first several service periods.

The reactor vessel outlet nozzle also affects the pressure-temperature limit curves of the first several service periods. This is due to the high local stresses at the inside corner of the nozzle, which can be two to three times the =*=hrane stresses of the shell. After the first several years of neutron radiation exposure, the RTNDT 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 allowabic pressure as a function of fluid temperature is obtained through a point-by-point comparison 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 the three calculated pressures.

s-1 Babcock & Wilcox

-t'

I

~

1 1

The eighth full-power year was selected because the estimated third surveil- I lance c..psule will be withdrawn at the end of the refueling cycle when the i fluene (orresponds to approximately the ninth full-power year. The ti=e dif-ference between the withdrawal of the second and third surveillance capsule provide < adequate time for re-establishing the operating pressure and tempera-ture ll=1ts for the period of operation be tween the third and f ourth surveil-lance capsule withdrawals.

The limit curves for Oconee 1 are based on the predieted values of the adjusted reference temperatures of all the beltline region materials at the end of the sixth full-power year. The unitradiated impact properties were determined for the surveillance beltline region materials in accordance with 10 CFR 50 Ap-pendixes C and 11. For the other beltline region and RCPB materials, the un-Irradiated impact properties were estimated using the methods described in HAW-10046P.8 The unirradiated impact properties and residual elements of the beltline region materials are listed in Table A-1. The adjusted reference temperatures are calculated by adding the predicted radiation-induced SRT NDT

• nd the unirradiated RT NDT. ThepredictedARTNDT is calculated using the respective neutron fluence and copper and phosphorous contents. Figure 8-1 illustrates the calculated peak .'.eutron fluence at several locations through the reactc r vessel beltline region wall and at the center of the surveillance capsules as a function of exposure time. The supporting information for Figure 8-1 is described in BAW-10100.9 The neutron fluence values of Figure 8-1 are the predicted fluences, which have been demonstrated (section 6) to be conser-vative. The design curves of Regulatory Guide 1.99* were used to predict the radiation-induced ART NDT values as a unction of the material's copper and phosphorus content and neutron fluence.

The neutron fluences and adjusted RTNDT ** "'" * ** "* ##8 " ****#

• at the end of the sixth full-power year are listed in Table 8-1. The neutron fluences and adjusted,RTNDT #8 "** '#' 8 **" # * * #" ***** ""

locations (T = wall thickness). The assumed RTNDT f the closure head region and she outlet nozzle steel forgings is 60F, in accordance with BAW-10046P.8 -

f Figure B-2 shows the reactor vessel's pressure-temperature Itait curve for normal heatup. This figure also shows the core criticality limits as required

• Revision 1. January 1976.

8-2 Babcock & Wilcox

by 10 CFR 50, Appendix C. Figures 8-3 and 8-4 show the vessel's pressure-temperature limit curve for normal cooldown and for heatup during inservice leak and hydrostatic tests, respectively. All prassure-temperature limit curves are applicable up to the seventh effective full-;cwer year. Protection against nonductile failure is ensured by maintaining the coolant pressure be-low the upper limits of the pressure-temperature limit curves. The acceptable pressure and temperature combinations for reactor vessel operation are below I

and to the right of the limit curve. The reactor is not permitted to go crit-ical until the pressure-temperature combinations are to the right of the crit-icality limit curve. To establish the pressure-temperature limits for protec-tion against nonductile failure of the RCPB, the limits presented in Figures 8-2 through 8-4 must be adjusted by the pressure dif ferential between the point of system pressure measurement and the pressure on the reactor vessel control-ling the limit curves. This is necessary because the reactor vessel is the most limiting component of the RCPB.

l

s-3 Babcock & Wilcox

/

. . . p e.s

• /

t & 4.

.eD

.. , e I  ;

4 e -

e - - - e o o .

. e.

- ~

e e e

e.

. Iee e .e

<s a

, e-t I we

. e -

s, V. d"'~.

.. = .e. . - e 4

. ~. .e- ~

.e.

e e

< n' g C.  ;

..: . s .

.e.

.a. ~ ~ ~ .e.

+

~ -+ a+ +

L>. . . . . . . . . w + + + me

..%s w

-l

.a. w e

- e w w w e e a e,,, . ,

_=

s

, ..s. .a. .e.

g ,. g. - s. e. a e 4.

. . . a. e.

O ~h.; , -e e s. - - e.

2.,

C.

y,' "p2- f. ' s ~ e . . . .

s

.s. - .a. .e. . ~.

. .e.

et - .e. .e.

r o e e s e .e. e s +

L d:.

p, 'J' 8

• b w

'."'s-'.

as e

s.

.s

• * * * *a **

.e. w .

.* W% .'

me w

'e

w e as. w me w

.e w

,, s =[

y, - a.4' o . . o. o.

o o o o o o

o. o. o. o.

o o.

o o

o. o. .o. .o. o. o. o.

-2 e f w.

o o o o o o

• = 54i

h. F e.I a e - .e o

d ..,4 .a. e. .e. .e. .e. .e.

e

.e.

o. o. o..

. ., .e.

e. e.

• • • l ' . .

wI 3,1 o o o.o .o .o . o o o o o

o. o.

G

.c&.,

=9 e o4 oe.o .eo

{

- w m.

E *=

36 -e  % .o. o

e. o.

e o

e.

o c.

s o

e. e. o.

c o

e. o.

e

,,, , L. aC =. * *w

• s .e.

w w .$

9 w w w w

$ e e p.l w ,

,. . .ei . *. *. *. '. *.

.i -'* - = I.i.

c.

bgc#

>> a h.

• **e r,j .s E o b C o eg t3= , . .

I . .

w kj ~~aL-

' * * ' I I .' *. ~o

s. "[ u . i o _o,t

.e w - .

. . . . . e.

. e. .. . . .

4 ~. l w ~. . .

-..v

~. ~. . . ..

u, n .

.e.

m

=.=,.c.

2

. 2 13 8

g

. .a . .

c 1

- - . . E m . . .

L,.,..L, o.dIu

. L u .L . . . L s

I e - a a .n .e., o, ~ -

= a xa 2 A a } ,A 8 ~ ~ I

.,> e o

o

- g .o ~ . g.~,. , ,, ,,,,,

5 3 333s33 3 a1 i Bis 23

• w- . . ,. . - -

21 - e ~. = ~ ~

.i.

4

e. ~.r.t ~.

.ft

~ .

o 8 R

4

. .f%.

Z. ~ - ~ ,. . .. .. . . . .

i2 3  : oaooa a as t atals 8-4 Babcock a.Wilcox -

L i

r I

Figure 8-1.

Fast Neutron Fluence of Surveillance Capsule Center Compared to Various Locations Through Reactor Vessel Wall for First 10 ETPY 10 7

9.5 x 1018 5 nvt 9

8 -

7 -

S z

A r.a 6 -

e4 f , 5.3 x 1018

,. ,p nvt S 5 - 9 d

C a

E Y

/

}/

8 C 4 -

c a Qb

get 3x 10I9

)_ 3e wt

/

2 -

o si @' go ge*Sel 3,

7.1 x 1017 nyt Vessel 3/4T Location 0 = 10 7 nyt

" vannel Oetside Surface

' t 3 2 3 l b 6 ) 0 > ic Time, EFPY s-5 Babcock & Wilcox

Figure ti-2.

React or Ve .cl Pi c.ssarc-Ti nce. .it ure 1.i:ni t Curves f or Lorrr.il Ope r.it li n lic.it up Appiicable tor rir..: 6 1.!'rY i

2400 A %s. meil i:1 jpg, l 0

Beltline region 1/!.T 147 Beltline rer.lon 3/41 S2 2000 Closure heaal rer.lon 60 oo Outlet norzte 60 7

c. I800 -

g 1600 -

8 Pressure. Ten;> ,

8- Appiicable for 1400 - Po,1nt

, _ p s,1. ,,,, F lleatup Rate.S e up to 100F/h 2 1200 -

^ '70 70 Y 8 B 625 180

• U C 625 273 T 1000 Criticality

- D 2250 102 3 1. imi t

1. 625 271

$ 800 -

F 625 31 3 C 2250 342 3 8 C 600

{

a A

I 400 - The acceptable pressure-temperature cumbinatione are below and to tk.2 right of the limit curve (s). The limit curves do not in-clude the pressure differential betueen the g.sint of system ca, 200 - pressure measurement and the preneure on tlw reactor vessel re-n ston controlling the limit curve nor do they include any aJJt-g' tional mergin of safety for poselble instrument error.

8 p i l l I l l l

1. 1 P 40 riO 120 160 200 240 880  !!O 160 400 W

=;

o I:eactor Vessel Coolant Ter nerat ure, F O

x

==. .

Elgure b-3.

Reactor Ve>=e1 Pre ==ure-Temper.sture Limit curve for .Lrma!

Operation - CoolJovn Applicable for first 8 EFPY 2400-Assumed RTNDT, F 2200 - g Beltline region 1/4T 147 Beltline region 3/4T 82 2000 - Closure head region 60 Outlet nozzle 60 y 1800 -

Pressure Temp.

g Point gel _ F Applicable for

= 1600 -

A 250 70 Cooldown kates s B up to 100F/h e 625 137 e C 625 205

'0 1400 - D 970 211 E 2250 296

= e

e

- 1200 -

8 u

e e= 1000 -

e e D 6.

o 800 -

U e

a mI 600 -

C 1,...,.i...............

. a ... . 4 . w . i e. .e .

i....... t a i.n ... 4.

400 - i-i . . eir e.. i.: .

m ...e.,.......

as '

A ***'**''**'***'**'n**.'.o'.*****A

• * *

• sw pe

.'.'.'*, ' "i .' " a. . i . i.. i . . ..e

.e 2M - **'********'**"""****"**'

x 1 I i l I f 40 l n

KU I?0 1 t.0 ;un  ::.ti :no $;o O

M Re.netor l'essel Coolant Temper.it ure. F

Figure 8-4 Reactor Vessel Pressure-Temperature Limit Curve for Inservice Leak and Hydrostatic Tests Applicable for First 8 EFPY 2600 Assurned RT

_ _.._ _..._ _NDT, F E Belt line region 1/4T 147 Beltline region 3/4T 82 2200 -

Closure head region 60 0 Outlet norrie 60 2000 -

d

[. I rion -

, Pressure Tenp, 2 Paint , p s i. , ,F_ Applicable for Hr.st up i 160r, - and Cooldown Rates up A 330 T; 70 to 100F/H (:.50F in any 2 0 625 150 30-minute per ~od)

, it.rgj _ C 625 245 j D 2200 268 8 7 L 25W) 272 j 1200 -

Ia

-$ 1000 -

s.  ;

O I

}G MK) ~

\

=

B 600 - C The acceptable pressure-temperature combi-nations are below and to the right of the 400 limit curva(s). The limit curves do not 4 include the pressure differential between

, the point of system pressure measurement and the pressure on the reactor vessel re-

• 00 - 31 a e otrollies the limit curve, nor do they include any additional mergin of safety for poselble instrumeat error.

O 60 100 140 130 220 260 JGO Reacto: Vessel Coolant Tes:pe ra t ure, t l

l l

l l

l 8-8 Babcock & Wilcox n

l 9.

SUMMARY

OF RESULTS The analysis of the reactor vessel material conta!ned in the first surfeillance capsule OCI-E removed from the Oconee 1 pressure vessel led to the following conclusions:

1. The capsule neceived an average fast fluence of 1.5 = 1018 n/cm2 (E > !

McV). The predicted fast fluence for the reactor vessel T/4 location at the end of the second foel cycle is 4 9 = 1037 n/cm2 (E > 1 HeV).

2.

The fast fluence of 1.5 = 1018 n/cm2 (E > 1 HeV) increased the RTSDT "'

the capsule reactor vessel core region shell materials to a max; mum of 124F.

3.

Based on a ratio of 1.6 between the fast flux at the surveillance capsule loc a t ion to t ha t a t the vessel wall and an 80% load factor, the projected fant fluence that the Oconee 1 reactor pressure vessel will receiva,in 40 calendar years' operation is 1.8 = 10I9 n/cm2 (E > I MeV).

4. The increare in the RTNDT tor the base plate material was in good agree-ment with that predicted by the currently used design curves of "RT SDT versas fluence.

5. The increase in the RTNDT f r the weld metal was in good agreement with that predic'.ed by the currently used design curves of /.RT v##8"" l"'

DT ence.

6.

The current techniques used for predicting the change in Charpy impact upper shelf properties due co irradiation are conservative.

7.

The analysis of the neutron dosimeters demonstrated that the .'nalytical techniques used to predict the neutrcn flux and f1 sence were accurate.

8.

The thermal monitors indicated that the capsule design was s atisf actory for maintaining ti.e specimens within the desired temperature range.'

9-1 Babcock & Wilcox E

10. SURVEILLANCE CAPSULE RE.TVAL SCHEDULE Based on the postirradiation test results of capsule OCI-E, the following schedule is recommended for exam iation of the remaining capsules in the Oconee I reactor vessel surveillance program:

Evaluation schedule Capsule Est. capsule Est. EFPD st. date #

fluence, data ID 1019 n/c,2_ Surface 1/4T available OCI-A

• 1.2 18 32 1984 OCI-CI ") 2.3 34 61 1988 OCI-B Standby -- -- --

OCI-D Standby -- -- --

OCI-G "' Standby -- -- --

OCI-H Standby -- -- --

(a) Capsules contain weld metal specimens.

(b) Capsules designsted thermal aging capsules.

" These dates do not represent the earliest dates that data will be available for the materials that control the oper2 ting limi-tations. Similar materials are included as part of the B&W Integrated Reactor surveillance Program, which will provide necessary data on a timely basis. The earliest date that these data vill be available is 1980.

10-1 Babcock s.Wilcox 4

. \

o

11. CERTIFICAT*05 The specimens were tested, and the data obtained from Oconee Nuclear Station, Unit 1 surveillance capsule OCI-E were evaluated using accepted techniques and established standard methods and procedures in accordance with the requirements of 10 CFR 50, Appendixes C and II.

M s2/2

/Ms WFFPZl.

M L. W . Jr.',

275 4it77 bate ~

Project Technical *hanager This report has been reviewed for technical content and accuracy.

K. E. Moore Y 7[

Date Technical Staff 11-1 Babcock & Wilcox

. , . n

12. REFERENCES I A. L. Lowe, Jr. , et al., Analysis of Capsule OCl-E From Duke Power Company Oconee Unit 1 Reactor Vessel Materials Surveillance Program, BAW-1421, Rev.

1 Babcock & Wilcox, Lynchburg, Virginia, September 1975.

2 C. J. Snyder and G. S. Carter, Reactor Vessel Material Surveillance Program, BAW-10006A, Rev. 3, Babcock & Wilcox, Lynchburg, Virginia, January 1975.

3 User's Manual fot ANISN, a One-Dimensional Discrete Ordinates Transport Code With Anisotropic Scattering, K-1693 (RSIC-CCC-82), Union Carbide Corp.,

Nuclear Division, March 1967.

User's Manual for the DOT-llW Discrete Ordinates Transport Computer Code, WANL-DfE-1982 December 1969.

5 CASK Croup Coupled Neutron and Gamma-Ray Cross Section Data, RSIC-DLC-23. Radiation Shielding Information Center.

6 Draf t -- New Standard E482-00, "Recnemended Practice for Neutron Dostmetry for Reactor Pressure Vessel Surveillance," October 10, 1974.

7 H. S. Palme, G. S. Carter, and C. L. Whitmarsh, Reactor Vessel Material Sur-veillance Program - Compliance With 10 CFR 50, Appendix H. for Oconee-Class Reactors, BAW-10100A, Babcock & Wilcox, Lynchburg, Virginia, February 1975.

8 H. S. Palme and H. W. Behnke, Methods of Compliance With Fracture Toughness and Operational Requirements of Appendix C to 10 CFR 50, BAW-10046P, Bab-cock & Wilcox, Lynchburg, Virginia, October 1975.

12-1 Babcock a Wilcox

i I

l l

APPENDIX A Reactor Vessel Surveillance Program -

Background Data and Informat. ion A-1 Babcock 8.Wilcox

1. Material Selection Data The data used to select the materials for the specimens in the surveillance program, in accordance with E-ISS-66, are shown in Table A-1. The locations of these materials within the reactor vessel are shown in Figures A-1 and A-2.

2. Definition of Beltline Region The beltline region of Oconee Nuclear Unit I was defined in accordance with the data given in BAW-10100A. -

3. _ Capsule Identification 1he capsules used in the Oconee Nuclear Unit I surveillance program are identified below by identification number, type, and location.

Capsule Cross Reference Data Number Type  !.ocation OCI-A A Upper OCI-B B Imer

• OCI-C A Upper OCI-D B Imer OCI-E A Upper OCI-F B Lower OCI-G , A Thermal aging OCI-Il B Thermal aging

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

f.

1 I

l A-2 ,

M & WibX I r) l

Table A-1. Surveillake_Prgr.m Materials Select ion Data f or Oconee 1 Distanse - - - - -

Cherry data. C"-

core mid- Transwere.

Material Belttine plane to

• 1D. heet Material ,

region weld center- Long., it-1b 50-ft-Ib. ,7 pg ,,,,,,

No. type location 7N37' , 35 MLE. Est. NDT' J tee. ce e 10F F F f r-l? P. 1 F CW. 1 Stu! NIM AMR 34 SA50(. C1 2 Nossle belt .. IUI gyf -. -- -- -- 0.15 0.006 0.010 --

C-2197-2 SA302 5 Inters shell -- -- 39, 45, 26 -- -- - --

0.10 C-3278-1 0.009 0.010 --

SA102 8 Upper ehe11 --

<to 35, 2y, 51 .- -- -- --

0.12 0.010 0.016 --

C-3265-1 SA302 8 Upper shell -

0 34. 64. 27 -. --

109 20 0.15 0.015 0.015 --

C-2000-1 54302 8 lower she!! ..

• 10 34, 39, 31 -- - -- --

0.11 C-2800-2 0.011 0.017 --

SA302 5 Lower shell --

20 32, 33. 49 -- - 119 20 0.11 0.012 0.037 --

SA-1430 Weld lang seen

-- -. 54, 52, 53 -- -- -- --

0.16 0.011 0.015 --

SA-1493 . Weld tens seaa -- -

41. 35. 40 -- -- -. -.

0.22 0.017 0.010 --

SA-10i) + Weld 14ng seam -- -- 40, 45, 39 -- -- .-

Wp-9 0.21 0.025 0.017 -

Weld Circ seaa -249 -- 46, 43, 45 -- -- .. -- 0.17 0.015 0.012 --

4 8A 1585 Wald Circ seen 61 - 31, 32, 31 -- -- - --

0.25 0.016 0.011 W9-25 Weld Circ seem -40

+123 38, 28, 49 -- -- 82 9 0.29 0.019 0.010 -

$4-1229 Weld Circ sean e123 -- 55, 45, 40 - -- -. ~

0.20 0.021 0.012 -

SA-1135 Weld Circ seam +199 -- 56, 44, 15 -- -- -- --

0.17 0.015 0.013 --

SA-1526 Weld Cire some +245 -- 31, 33, 33 -- -- -- --

0.36 0.016 0.012 -

SA-1494 Weld Circ seem +245 -- 54, 25, 44 -- -- -- --

0.14 0.015 0.012 --

a K

O o

3t* <

P b

ii M

Table A-2. Materials and Specimens in Upper Surveillance Capsules OCI-A, OCI-C and OCI-E No. of specimens Material description Tensile Charpy Weld metal WF-112 4 8 Iteat-affected zone (HAZ)

Heat A - C-3265-1, longitud 0 8 Baseline material Heat A - C-3265-1, longitud 4 8 Feat A - C-3265-1, transverse 0 4 i

Correlation HSST plate 02 0 8 Total per capsule 8 36 Table A-3. Materials and Specimens in Lower Zurveillance Capsules OCI-8, OCI-D and OCI-F No. of specimens Material description Tensile Charpy Heat-affected zone (HAZ)

Heat B - 2800-2, longitud 4 10 Baseline material Heat B - C-2800-2, longitud 4 10 Heat B - C-2800-2, transverse 0 8 Correlation HSST plate 02 0 8 Total per capsule 8 36 v

7 1

5 A-4 Babcock & Wilcox

• i

Figure A-1. Location and Identification of Materials l

Used in Fabrication of Oconee Unit 1 i Reactor Pressure Value I I

l 1

\

l

)

l .  ::t I

ZV2861 e (Nozzle Belt) 7 SA15261 Outlet V SA1494[ Nozzles Only N

, m SA1135

~ -g 3 o2

• 5 0$ NC2197-2 (Inter-mediate Shell)

SA1229 - 61% (ID) 2e W 25 - 39% (OD)

$ 2 C3265-11 upper Shell g m  %

C3278-1/

d5 I

w SA1585 7

g { C2800-ll , Lower Shell

- n A C2800-2) m

~

. Jw W112 122S34VA1 Dutchman e

l A-5 Babcock s. Wilcox l

e  !

. D ,

V

= .

N W

/ -  !

a=4 D

m

- c 3 -

+ w D

N rs 8

be b

. se u

• l

. f4 6 x ~ d

r. ,

M 's

• • 1.

.'d L

.t

=~ e,e w e=e 2k .

I E'

5,} N >e as, t

't ,

a, I

g "C  ;

N. -

• . ~J N

O g N e M

• a M

• 8 i

6 f W

a 2

/

ame

~

= -p > m

.e x

l 6.

4

• 4 E D t

i W Y N

/

A-6 Sabcock a witco, ,

I

m_

i l

1 I

l APPENDIX P.

Preirradiation Tenap Dats 4 1

i I

B-1 Babcock & Wilcox

Table B-1. Preirradiation Tensile Properties of Shell Plate Haterial, Heat C-3265-1 Specimen Strength, psi Elongation. : # "

te No. F Yield Ult.

• Unif. Total Longitudinal AA-701 RT 64,000 85,900 15.1 705 28.9 63.0 RT 64.200 85,900 ND 27.8 71.0

'723 RT 64,700 86,300 14.9 22.8 '71.0 Mean, E --

64,300 86,100 15.0 26.5 68.0 Std dev'n --

299.3 213.8 0.163 3.24 4.0 AA-706 600 56,900 83,200 15.1 26.4 68.0 710 600 57,800 85,200 16.3 27.1 68.0 715 600 59,200 84,800 15.0 28.5 73.0 Mean, 5 --

58,000 84,400 15.5 27.3 70.01 Std dev'n --

1,170 1,050 0.72 1.09 2.625 Transverse AA-601 RT 65,100 606 86,600 14.6 26.0 63.0 RT 65,500 86,900 607 14.5 27.8 67.0 RT 64,700 86,100 16.1 24.6 65.0 Mean E --

65,100 86.500 15.1 26.1 65.0 Std dev't. --

400.1 408.1 0.89 1.61 2.1 AA-604 600 57,800 605 84,500 15.7 25.4 63.0 600 57,100 83,400 16.0 f 609 600 26.1 63.0 58,900 84,800 14.2 26.1 67.0 Mean, 5 --

57,900 .

84.200 15.3 25.8 64.0 Std dev'n --

905.2 708.8 0.93 0.37 2.4 i

a-2 Babcock & Wilcox g

. s 1 i

i

Table B-2. Preirradiation Tensile Properties of Shell Plate Material, RAZ, Heat C-3265-1 8

3p,cg ,, Strength, psi

• No.

,3p, F

Elongation. ",

_ Yield Ult. Unif. Total  %

Longitudinal AA-406 RT 60,700 79,500 407 10.9 21.6 64.0 RT 63,709 81,800 409 11.2 22.6 62.0 RT 60.600 79,900 11.5 29.5 64.0 Mean, i

-- 65,600 80,400 11.2 Std dev'n --

1,750 24.6 63.0 1,000 0.29 4.30 1.0 AA-401 600 56,100 81,800 404 12.2 20.0 53.0 600 57,700 81,200 12.3 408 600 56,800 21.0 47.0 80,400 11.1 19.6 Mean, 5.

52.0 56,900 81.100 11.9 Std dev'n 20.2 51.0 795.4 725.4 0.08 0.71 3.2 Transverse AA-302 RT 65,600 301 86.400 14.4 27.8 70.0 RT 64,900 86,300 304 15.2 29.2 71.0 RT 69,000 89,700 14.1 29.0 74.0 Mean, i --

66,500 87,500 Std dev'n 14.7 28.7 72.0 2,210 1,910 0.49 0.75 2.0 AA-?OS 600 57.100 83,600 306 16.3 30.0 72.0 600 55,200 85,300 301 15.8 31.4 68.0 600 58,500 86,800 13.7 25.2 70.0 Me.in 5 --

56,900 85,300 15.3 28.8 Std dev'n --

1,620 1,620 70.0 1.33 3.24 1.9

-l l

l n-3 Babc5k & Wilcox s.

i Taole B-3. Preirradiation Tensile Properties of Weld Metal, Weld Qualification No bT112 Test Red'n Specimen reng , psi Elongation, %

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

Longitudinal 0C1-105 RT 63,900 81,100 17.2 32.1 63.0 108 RT 63,400 80,300 16.9 30.7 64 0 123 RT 62,700 80,100 16.6 30.0 64.0 Mean, x ~63,300 80,500 16.9 30.9 63.0 ,

Std dev'n --

625.8 548.4 0.30 1.0 0.5 0C1-118 600 55,300 79,600 16.8 25.0 63.0 121 600 58,400 82,200 16.9 22.9 60.0 124 600 55,500 80,500 16.7 25.3 60.0 Hean, X 56,400 80,800 16.8 61.0 24.4 Std dev'n --

1763.7 1337.1 0.13 1.32 1.8 -

m 5

e n-4 Babcock & Wiicox  ;

i

APPENDIX C Preirradiation Charpy Impact Data t

c-1 Babcock &Wilcox

i Table C-1. Preirradiation Charpy Impact Data for Shell Plate Materisi, Longitudinal Direct ion, Heat C- 326 5-1 Test Absorbe* Lateral Shear Specimen temp. ener No.

. expansion, f racture".

F f t- 10-3 in.

• AA-715 320 430 72 '100 724 320 150 e4 100 899 320 139 70 100 AA-897 7' 142 73 100 AA-707 137 66 85 747 i 10 143 73 80 898 200 140 72 80 AA ~ . 4 131 125 67 75 749 131 110 62 SO AA-731 67 98 46 45 735 67 88 38 35 744 67 82 60 15 AA-718 33 67 55 10 736 35 57 46 5 752 35 69 66 12 AA-725 0 23 18 0 740 0 61 49 0 895 0 32 2$ 0 AA-896 -30 8 6 0 AA-720 -58 11 7 0 900 -60 8 4 0 t

c-2 Babcock &Wilcox

• b

- w

Table C-2. Preirradiation Charpy Impact Data for Shell Plate Material, Transverse Direction. Heat C- 3261- 1 Test Absorbed Lateral Shear Specimen temp, energy, expansion, fracture, No. F ft-lb 10-3 in.  %

AA-619 320 106 72 100 631 320 108 74 100 700 320 114 72 100 AA-698 261 103 72 100 699 262 113 72 100 AA-623 200 105 63 85 625 200 98 68 696 90 200 110 74 90 AA-614 170 97 81 95 AA-626 129 87 60 85 695 130 85 59 65 AA-628 100 70 56 629 30 .

100 65 52 50 AA-604 66 63 49 10 617 66 65 26 15 620 67 42 34 15 AA-601 30 46 37 2 638 30 36 32 2 697 30 36 30 2 AA-624 -40 10 8 1 632 -40 27 20 0

-C-3 Babcock a. Wilco.

r.

f Table C-3. Preitradiation Charpy Impact Data for Shell Material, HAZ, Longitudinal Direction.

Heat C-3265-1 Test Absorbed Lateral Shear Specimen teop, energy, No.

expansion, fracture.

F ft-Ib 10-3 in.  %

AA-399 320 116 70 100 422 320 114 68 100 428 320 108 74 100 AA-397 262 125 70 100 AA-396 200 117 67 70 401 200 82 58 85 421 200 98 62 90 AA-395 130 108 60 85 427 130 86 57 95 AA-419 67 90 49 85 426 67 90 52 90 442 67 94 52 90 AA-406 30 87 54 55 407 30 71 48 50 414 30 70 52 30 AA-398 20 h3 53 35 AA-411 10 63 45 35 AA-400 220 58 41 15 430 -20 44 30 10 438 -19 35 24 12 AA-429 -59 28 18 5 I

O t.

9 C-4 babcock & WilCOE

Table C-4 Preirradiation Charpy Impact Data for Shell Material, HAZ, Transverse Dirertion, Heat C-3265-1 Test Absorbed Lateral Shear Specimen #

t et:p, energy expansion, fracture, No. F ft-lb 10~3 in.  %

AA-302 320 100 60 100 304 320 92 62 100 596 320 124 65 100 AA-598 260 102 56 '. 00 AA-303 200 85 59 100 307 200 79 56 100 599 200 111 64 90 AA-301 71 78 48 85 309 67 105 59 100 318 67 92 49 85 AA-310 20 51 40 25 313 20 76 52 55 AA-595 10 68 44 55 AA-308 -20 50 33 15 314 -20 56 37 15 597 -20 38 28 10 AA-319 -45 55 36 5 AA-320 -50 35 25 6 AA-312 -79 21 13 2 316 -79 25 14 1 1 l

l C-5 Babcock & Wilcox

. , - , . . , s

Table C-5. Preirradiation Charpy Impact Data for Weld Metal, Weld Qualification No. WF 112 Test Absorbed Lateral Shear Specimen temp, energy, expansion, fracture, No. F ft-lb 10-3 in.  %

OCI-015 320 68 62 100 016 320 65 55 100 AA-296 320 62 52 100 OCI-004 261 64 59 100 AA-300 260 61 54 100 OCI-030 200 71 55 100 035 200 63 59 98

'AA-299 200 63 52 100 AA-298 140 6t. 57 65 OCI-022 66 57 47 65 025 70 60 41 75 034 70 55 47 60 OC1-011 40 41 39 10 036 40 51 48 45 OCI-020 10 30 28 15 031 11 40 38 20 AA-295 -10 37 33 25 OCI-332 -15 33 27 5 OCI-018 -40 19 14 5 023 -40 25 21 5 AA-297 -40 13 14 5 C-6 Babcock & Wilcox -

h

Eigure C-1. Impact Date From Unirradiated Base Metal A - Longitudinal Orientation

^

IT . . . . . . .

75 . .

J 1

W $c _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ , . _ .

J *

=

Y 25 - -

- = *. . . . . . . . .

n

.N . . . . . . . . . . .

i S

• e s .m 5 +

. e

=. .S.C - .

r .-- _ . ____ . _ _ _ _ _ _ _ . _ _ _ _ _ _ . _

af J .020 -

5 y i i i i e i i i , , ,

F . . . . . . . . . . .

• Ata suwn 7 OF lar eet .

1. F In (35 m)

g .fg (50 n-u) 17F ,

(.t2 tml 141 ft-lbs. .

IE - ti e, =

E 120 -

2 e 3 -

Ifr . .

3 W

wm .

• O

- w . .

o

,/ .

_-_____f____________...___..,_

• iserat SA307 n ma

. Onservattom Longs tudinal 20 -

• Fuence None e Haar h e c-3265 1 e

-80 -40 0 40 80 120 160 200 243 280 320 W) Wy1 Test To ,anatume, F l

c-7 Babcock a.Wilcox

-+

Figure C-2. Impact Data From Unirradiated Base Metal A - Transverse Orientation

• TJ . i s a s .. - . . ,

e e e

.e 75 -

J e 1

jsc .____________ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .

3 55 -

n

• N i i i s s s . . . . .

. e . .

$ w w

$ .4c 2 e

$ ,or,c . ,#

5 C

--.----- 9-

-/________.

r

= e

.020 - * .

5 3 e

.. l . . . . . . . . . . .

W i i . . . . . . , . .

CATA $dTWrf y 0F lar mer .

Tc , (35 sta) 48F 64F 16C Icv (So at-u) ,

(-USE (avs) 108 f t-lbs

. le - RT,,, .

b s tn -

e e 2 e 3

.7 Ifl0 -

e -

W 80 -

O.

1 w 3 e

,s .

_ - _ + _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ .

e f g . *

%ftang SA302 B Mod ~ ,

e

• 20 -

heravarrom Transverse None -

11:47 liuseca C-3265-1

• o

.ao -e o e so 120 16o 20o 2e m no yo a Test Toetaanas, F c-8 Babcock a. Wilcox i I

s Figure C-3. Impact Data From Unirraatated base . Metal A -

1:AZ . Longitudinal orientation w w

' ~

I I E y s F I g I g

e

• ee D . .

. e I

A h

y se .____ __e___________.___________._ __.

3

~

e e

J, 25 . .

p, I f f e a e E n e e a

• 5 8 3 5 5 5 5 5 5 g 5 e

E

• o $

i .W -

e ,

i e* _

e e

.". .m: . e a r, e e'

.07 -

5 a

g I l f f f f 9 9 t E a W i . . . . . . . . ,

'ATA SU' WRY I mor -#

1 90 .

In (35 m ) -17F 160 Ic, (50 n -ts) _ -13F ,

(.tg gam 113-ft-1bs 3 140 - 87 , .

5 e

A 120 -

7

• 1 I g

• e 3 IT -

e Y ,a

• w 80 -

e ,

r, a

S t

~ % -

e / .

/

7.----- _ __ ________

60 -

p .

%rtaeas. SA302 B (HAZ)

Onasstanon Longitudinal 20 -

% None -

Star twete . C-32 65-1 i, . . . . . , f , , ,

e#

-*0 0 40 80 120 160 .oo 2e 2m m yo son Test Tesvenaw, F C-9 Babcock & Wilcox

1 l

l 1

i Figure C-4. Impact Data From 1:nirradiated Base Metal A -

HAZ, Transverse Orientation Im , , , . , . . , , l

• i e.

se .,

1

% 1 w

y so -___--__e. - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

5 25 -

e -

e ga n I a a e a a e a e

• 4 4 5 a g 3 a . . .

e I i .v -

. - s, -

1 i .

=

• e

• "- . 'ho -

,.e .

-- r- + - - - - - - - _ _ _ - _ _ - - - _ - - - - - - - - - - - . .

5 e s *** l 2 . Ws S .

,rq f I E I 1 f I t f n ,

l rn i i i . . . . . . . ,

T.ATA 'A'T!JY T -40r 1V

^

mas .

To (35 sus -17F jw fy(50 at-ts) -14F _ ,

(-t:SE (avs) 100 f t-lbs

. lW1 - Rf , .

b

$W e

S e

100 -

e g _

a e

e w

7 e an -

. , e e -

t *

- 60 -

e e

-/----------------------.....---......

%ttagat SA302 B (HAZ)

( 20 hiestatsan Transverse.

Fustuct None maar ihmeta c-3265-1 c i e . i o'

i .

e'

-so -e so 120 160 200 2e rao 5a sto vyi Tsst Tapermaru, F

/

c-10 Babcock & Wilcox -

b

Figure C-5. Icpact Data Froci Unirradiated ".* eld

!!etal Longitudinal Orientatie:

17 i . . . .

• F, -

e -

5

.3 e yv -________

e 3

Y a M -

e .

e e

r * * ' e s e a e a e .

.f3?? . . . . . . . . . .

5

t. .rf ', -

e e *

• _

e e

.Z

~

e

~ .'W -

8 e .

" - - - - - - - p --______ ___ ___ , _ . _ _ _ , _ _ _ .

. a s f 4 . 7.* - e .

, rm

• e i e e r i e e .

Im = . .

fAfa 91998Y f et '07

^

LV .

73 (35 m )l6F Jg 1, (50 st-ts) 50F _

6/. ft-1bs

(.U2 (aw)

., I t - af est -50F .

t E 120 -

0

IT -

5 7

an -

~

7,

• e

• _t - -

M - *

• e e .

- - _ _ _ _ _ _ .a ,

o

% - e e' .

• %ftens Weld Metal

• thisteratta. Lonzitudical 3,,

, Fwince Mone deat Nwere 17-112 e

' ' i . .

W

-# 0 W 80 120 160 200 2M 25) 3a %9 vn Test Tsetaatume, F c-11 Babcock s. Wilcox t

Figure C-6. Impact Data Unirradiated Correlation Material a s a a v -- -, s 75 .

. e e o

?

• j 2 .. _ ___~___ __ ____..________ _______

=

Y w 25 -

ee

• e q t a a e a . s e e n 5 5 g i 3 3 u r u s e i - e i m -

e e

- m .

i e

" , ".N

,' gs .

- _ _ - . - _ _ ~ _

h #

g . D. -

O

.t.

q f le f f I f f f f a e

24) e e < > . . . . . . ,

TATA Sb912Y 10 Ieet A Ic , (35 m) $$F ,_,

Ig .To (50 st-ta) 74F , ,

(-uq (an) 130 ft-lbs 1

. Irac - RT , NA 1

. e e

5 !?O -

j

.- . i 2

110 h

W w 83 -

7, .

t

~ g .

• e _

4 - 's

, f .

krtasac RSST-PL-02 3, Onsrevarsee Fumeer h* -

o f NaArthsese A-1195-1

. f . . , f ,

-80 -40 o ao so 120 1co 200 2e m 120 yo vn Test femmarums, F C-12 Babcock a. Wilcox I

APPENDIX D Threshold Detector Information D-1 Babcock & Wilcox 1

Table D-1 lists the ccaposition of the threshold detectors and the thickness of cadmium used to reduce competing thermal reactions. Table D-2 shows the cycle 1 ::casured activity per gram of carget material (i.e., per gram of uranium, nickel, etc.) corrected for the wait' time between irradiation and counting. Measurements after cycle 2 are listed in Table D-3. Activation cross sectionc for the various target materials were flux weighted with a 135 U spectrum (Table D-4).

Table D-1. Detector Composition and Shieldig Monitors Shielding Reaction 11.87 U-Al Cd-Ag 0.02676-inch Cd 238 C(n,f) 1.61% Np-Al Cd-Ag 0.02676-inch Cd 237Np(n f) ,

Ni Cd-Ag 0.02676-inch Cd 59 Ni(n.p)59Co 0.66% Co-Al Cd-0.040-inch Cd 59Co(n,y)60Co 0.66% Co-Al None 59Co(n,y)60Co Fe None 5"Fe(n.p)S4Mn 1

u-2 Bsbcock & Wilcox g

1

~

Table D-2. Measured Detector A:tivities After Cycle 1

__ Activity. pCi/g Monitor Nuclide OCI-FD5 OCI-FL6 OCI-7D7 OCI-FD8 239 9 103Ru 8.70 x 101 6.80 101 6.91 x 101 5.06 = 103 137 Cs 6.81 x 10-1 1.15 1.57 1.07 I" Ice 5.69 x 101 4.44 x 101 4.45 = 101 3.32 x 101 144 Ce 3.31 x 101 2.58 101 2.59 x 103 1.97 x 101 103 Ru 4.07 x 102 2.86 x 102 3.02 x 102 2.41 x 102 237 3p 137 Cs 1.09 x 101 7.76 7.95 6.39 I"l Ce 2.68 x 102 1.82 = 102 1.96 102 1.48 x 102 5bga 14h ce 1,32 x 192 1.C3 x 102 1,to , 102 8.94 x 102 seco 9.61 x 102 7.33 x 102 7.27 x 102 59 5.64 d 102 Co(Cd) 60Co 7.43 x 103 5.94 x 103 5.78 = 103 5.17 x 103 50Co 60Co 4.72 = 10" 3.06 x 10" 3.01 x 10 4 2.04 x 10" 5"re 5"Mn 3.92 x 102 2.85 x 102 3.18 x 102 2.82 = 102 t

a 2

D-3 Babcock & Wilcox

Table D-3. Measured Detector Activities After Cvele 2 Post-irradiation Total, Total, Monitor wt, .sg Target, Isotope _uCi uCi/g pCi/cm(a)

Reaction

_n.1 23eU 47.00 95 Zr 2.047-1 4.36 238 U(n f)FP 4.23+1 95Nb 5.017-1 1.07+1 1.04+2 103 Ru 2.424-1 5.16 5.01+1 137 Cs 9.01-3 1.92-1 1.66 141Ce 1.820-1 3.87 3. 76+1 10b Ru 6.685-2 1.42 1.38+1 140Ba 1.400-1 2.98 237 3p 2.89+1 79,44 952r 2.911-1 3.66 2 37 Np(n.f)FP 2.54+2 95Nb 6.330-1 7.97 5.53+2 103 Ru 2.998-1 3.77 2.62+2 137 Cs 9.612-3 1.21-1 8.40 140Ba 1.805-1 2.27

1.58+2 I"3 Ce 2.944-1 3.71 2.58+2 106Ru 6.904-2 8.69-1 6.03+1

( Ni 133.22 58Co 8.706+1 6.54+2 58 Ni(n.p)S8co a.s5+2

{ 60Co 2.313-1 1.74 60 Ni(n.p)6cCo 6.65 Co(Cd) 16.00 60Co 1.002 6.26+1 59Co(n,y)60Co 9.48+3

! Fe 151.66 54Mn 4.673 3.08+1 54 Fe (t. p)S"Mn 5.29+2 59Fe 9.230 6.09+1 58 Fe(n,y)S9Fe 1.85+4 Co 15.70 00Co 7.601 4.84+2 59 Co(n,y)60Co 7.33+4 ED2 23s0 55.74 95Zr 3.268-1 5.86 238 U(n f)FP 5.69+1 95Nb 7.841-1 1.41+1 1.37+2 103 Ru 3.864-1 6.93 6.73+1 137 Cs 1.14-2 2.05-l' 1.99 140 Ba ,

2.464-1 4.42 4.29+1 141 Ce 2.887-1 5.18 5.03+1 106Ru 9.383-2 1.68 1.63+1 D-4 Babcock & Wilcox

I 1

l l

Table D-3. (Cont'd)

Post-irradiation Total. Total, 'arget Monitor wt. mg Isotope pCL uCi/g Reaction 'CCf/gaka) 237 Np 84.2 95Zr 3.590-1 4.26 237 Np (n, f) FP

2. 9 6+2 95Nb 7.387-1 8.77 6.09+2 103Ru 3.677-1 4.37 3.03+2 137 Cs 1.169-2 1.39-1 9.65 140Ba 2.276-1 2.70 1.S8+2 141Ce 3.648-1 4.33 3.01+2 1""Ce 1.898-1 2.25 1.56+2 106Ru 7.255-2 8.62-1 5.99+1 Ni 136.03 SdCo 1.106+2 8.13+2 SSNi(n.p)58Co 1.20+3 60Co 2.301-1 1.69 60Ni(n.p)60Co 6.46 Co(Cd) 13.75 60Co 1.306 6.97+1 59Co(n,y) 60Co 1.06+4 Fe 154.44 5'Mn 5.549 3.59+1 54Fe(n.p)s4Mn 6.17+2 59Fe 1.314+1 8.51+1 seFe(n,y)S9Fe 2.5S+4 Co 16.72 60Co 8.984 5.37+2 59Co (n,y) 60Co 8.14+4 ED3 238U 50.1 95Zr 1.756-1 3.51 23eU(n,f)FP 3.41+1 95Nb 3.507-1 7.00 6.79+1 103Ru 2.043-1 4.08 3.96+1 137Cs 8.91-3 1.78-1 1.72 140Ba 1.369-1 2.73 2.65+1 141Ce 1.615-1 3.22 3.12+1 106Ru 6.272-2 1.25 1.21+1 237 Np 82.66 ~

95Zr 2.323-1 2.81 237Np(n, f)FP 1.95+2

~

95Nb 4.579-1 5.54 3. 85+2 103Ru 2.372-1 2.87 1.99+2 137Cs '

708-3 1.17-1 8.13 140Ba l.' L19-1 2.20 1.53+2 ,

. 1" Ice 2.423-1 2.93 2.03+2 106Ru 4.977-2 6.02-1 4.18+1 NL 128.29 58Co 6,422+1 5.01+2 58 Ni(n.p)Seco 7.39+2 60Co 2.006-1 1.56 60Ni(n.p)60Co 5.96 D-5 hock & Mca

I-Table D-3. (Cont'd)

Post-ircadiation Total, Total, Target.

Monitor wt. eg Isotope pCi 'C1/g

. Reaction pCi/gm(a)

Co(Cd) 20.28 60Co 1.269 6.26+1 59Co(n.y)60Co 9.48+3 Fe 150.81 5"Mn 3. 90 '. 5"Fe(n.p)54Mn

2. 59+1 4.45+2 59Fe 5.308 3.50+1 sere (n,y)S9Fe 1.07+4 Co 16.88 6.529 59Co(n,y)60co 3.87+2 ,..86+4 M

23eu 52.92 952r 2.509-1 4.74 23sU(n f)FP 4.60+1 95Nb 6.527-1 1.23+1 1.19+2 103Ru 2.921-1 5.52 5.36+1 137Cs 1.20-2 2.27-1 2.20 141Ce 2.250-1 4.25 4.12+1 106Ru 1.031-1 1.95 1.89+1 237 5p 69.22 95Zr 2.522-1 3.64 237 Np(n f)FP 2.53+2 95 Nb 5.967-1 8.62 5.99+2 103Ru 2.524-1 3.65 2.53+2 137Cs 1.11-2 1.60-1 1.11+1 141Ce 2.502-1 3.61 2.51+2 344 Ce 1.368-1 1.98 1.38+2 106 Ru

~

6.397-2 9.24-1 6.42+1 Ni 129.20 58Co 8.734+1 6.76+2 58 Ni(n.p)58Co 9.97+2 60Co 2.238-1 1.73 60.ii(n .p) 60Co 6.61 Co(Cd) 20.25 60Co 1.331 6.57+1 59Co(n,y)60Co 9.95+3 Fe 154.27 54Mn 4.978 54 Fe(n.p)S4Mn 5.55+2 3.23+1 59Fe 1.013+1 6.57+1 58Fe(n,y)S9Fe 1.99+4 Co 16.46 60 Co 8.518 5.18+2 59Co(n,y)60Co 7.85+4

• The following abundance and weight percents were used to calculate the disintegration rate per gram of target: '

1. 238U: 10.38 wt %; 99.27% isotopic.

2. 237 Np: 1.44 wt %; 100% isotopic.

3. Ni: 100%; SONi 67.77% isotopic; 6cNi 26.16% isotopic.

4. Co: 0.66 wt %; 59Co 100% isotopic.

5. Fe: 100%; 54Fe 5,82 isotopic; 58Fe 0.33% isotopic.

D-6 Babcock & Wilcox I

e

9 Table D-4. Dosimeter Activation Cross Sections (*

Energy range, C gey 237 3p 2380 58,y t S t. p, 1 13.3 -15.0 2.231 1.073 0.460 0.425 2 10.0 -12.2 2.34 0.981 0.622 0.537 3 8.18 -10.0 2.31 0.991 0.659 0.583 4 6.36 -8.18 2.09 0.917 0.638 0.572 5 4.96 -6.36 1.54 0.60 0.54 0.473 6 4.06 -4.96 1.53 0.562 0.403 0.325 7 3.01 -4.06 1.616 0.553 0.264 0.206 8 2.46 -3.01 1.69 0.550 0.139 0.096 9 2.35 -2.46 1.695 0.553 0.089 0.0524 10 1.83 -2.35 1.676 0.535 0.051 0.022 11 1.11 -1.83 1.593 0.229 0.C128 0.0115 12 0.55 -1.11 1.217 0.008 0.00048 -

13 0.111 -0.55 0.1946 0.00013 - -

14 0.0033 -0.111 0.0410 - - -

(a)ENDF/84 values flux weighted with a fission spectrum .

I l

D-7 Babcock & Wilcox l

. . .. . . . - . . . - . . ~ .---