Text

Reactor Vessel Radiation Surveillance Program.
	ML20002E349
Person / Time
Site:	Farley
Issue date:	12/31/1976
From:	Davidson J, Phillips J, Yanichko S; WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	;
Shared Package
ML20002E348	List: ML20002E347; ML20002E349; ML20002E350;
References
	WCAP-8810, NUDOCS 8101270519
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SOUTHERN ALABAMA POWER COMPANY N f JOSEPH M. FARLEY NUCLEAR PLANT

! UNIT NO.1 REACTOR VESSEL RADIATION I SURVEILLANCE PROGRAM j"f 1

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i, J. A. Davidson y J. H. Phillips e i S. E. Yanichko g

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Work Performed Under ALA-106 ,

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WESTINGHOUSE ELECTRIC CORPORATION Q.If Nuclear Energy Systems y P. O. Box 355 ;N;*e Pittsburgh, Pennsylvania 15230 ;

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ABSTRACT

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A pressure-vessel steel surveillance program was developed for the Southern Alabama

l Power Company Joseph M. Farley Unit No.1 Nuclear Reactor to monitor the radiation effects on the reactor pressure vessel material under operating conditions.

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e A description of the program, including the materit. to be tested, specimen and capsule design, and preirradiation test results, is presented.

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TABLE OF CONTENTS g Title Page

- Section 1-1 I 1 PURPOSE AND SCOPE 21 2 SAMPLE PREPARATION

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Pressure Vessel Material 21 l'

2-1.

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2-2. Machining

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I 2-3. Charpy V-Notch Impact Specimens Tensile Specimens 21 se j 2-4. '

Bend Ba. Specimens 2-1

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'g 2-6. %T Compact Tension Specimens 25 r

2-7. Dosimeters 25 W I

2-8. Thermal Monitors 2-5 2-5 h

W 2-9. Capsule Loading Gi l

2-10. Specimen Capsule 2-6 y l Y l

t 3 PREIRRADIATION TESTING 31 g .

j 3-1. Charpy V Notch Tests 3-1 $.

Tensile Tests 3-1 @

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Dropweight Tests 3-2 h.;

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t 3-4. RTNOT Results 3-15 %

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i j 4 POSTIRRADIATION TESTING 4-1 h c \

4-1. Capsule Removal 41 4-1 kh

! 4-2.. Charpy V Notch Impact Tests '

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4-3. Tensile Tests 4-2 a-'

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! 4-4. Fracture Toughness Tests on %T Compact 5 Tension Specimens and Bend Bar Specimens 4-2 @.

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Section Title Page 4-5. Postirradiation Test Equipment 42 l Appendix A JOSEPH M. FARLEY UNIT NO.1 REACTOR PRESSURE VESSEL SURVEILLANCE MATERIAL A1 i i

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! 21 Charpy V-Notch Impact Specimen 2-3 ,

t, 22 Tensile Specimen ,

2-4 Precracked Bend Bar Specimen 2-3 Compact Tension Specimen

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!' 2-4 Irradiation Capsule Assembly 2-9/2 10 _

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Specimen Locations in the Reactor Surveillance f

27 Test Capsules 2 13/2-14 Ik-Ib' t

Preirradiation Charpy V Notch impact Energy for the f 3-1 Joseph M. Farley Unit No.1 Reactor Pressure Vessel

[ Lower Shell Plate B69191 (Longitudinal Direction) 3-5 %-

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Preirradiation Charpy V-Notch Impact Energy for the

,[ 32 Joseph M. Farley Unit No.1 Reactor Pressure Vessel $

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Lower Shell Plate B6919-1 (Transverse Direction)

Preirradiation Charpy V Notch Impact Enargy for the Ny

[ 3-3 Joseph M. Farley Unit No.1 Reactor Pressure Vessel I

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Preirradiation Charpy V-Notch Impact Energy for the g

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[ Joseph M. Farley Unit No.1 Reactor Preesure Vessel r 3-10 Core Region Weld Heat Affected Zone Material Preirradiation Tensile Properties for the Joseph M. Farley E

3-5 Unit No.1 Reactor Pressure Vessel Lower Shell Plate h_

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B6919-1 (Transverse Direction) g W*

3-6 Preirradiation Tensile Properties for the Joseph M. Farley Fi Unit No.1 Reactor Pressure Vessel Lower Shell Plate B6919-1 (Longitudinal Direction) 3-13 y g-Preirradiation Tensile Properties for the Joseph M. Farley 4 s 37 y Unit No.1 Reactor Pressure Vessel Core Region Weld 3-14 l' Metal

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' Q j- 2-1 Type and Number of Specimens in the Joseph M. @

t Farley Unit No.1 Surveillance Test Capsules 2-15 4

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- 3-1 Preirradiatico Charpy V-Notch Impact Data for the M f Joseph M. Farley Unit No.1 Reactor Pressure Vessel Lower Shell Plate B6919-1 (Longitudinal Direction) 3-3 h, l el 3-2 Preirradiation Charpy V-Notch Impact Data for the Joseph M. Farley Unit No.1 Reactor Pressure Vessel i C

l Lower Shell Plate B6919-1 (Transverse Direction) 3-4

-g 3-3 Preirradiation Charpy V-Notch Impact Data for the g.

Joseph M. Farley Unit No.1 Reactor Pressure Vessel .

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i 3-4 Preirradiation Charpy V-Notch Impact Data for the .

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3-5 Preirradiation Tensile Properties for the Joseph M. Farley c{. Unit No.1 Reactor Pressure Vessel Lower Shell Plate Nl I B6919-1 and Core Region Weld 3-11 $

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SECTION 1 6 PURPOSE AND SCOPE ('

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The purpose of this program is to monitor the radiation effects on the reactor vessel ^

k mates'als of the Southem Alabama Power Company Joseph M. Farley Unit No.1 under E

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! actual aperating conditions. Evaluation of the radiation effects is based on the_pg . p k irradiation testing _of Charpy V notch, tensile, and dropweight specimens and post- . <

igadiation testing of Charpy V-notch, tensile, precracked bend bar, and compact tension ,

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speciment.

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Current reactor pressure vessel material test requirements and acceptance standards utilize 2 I the reference nil-ductility temperature, RTNDT, as a basis. RT NDT si determined from p' y

L the drop-weight nil-ductility transition temperature (NDTT) per ASTM E208 and the ['

"weaka (11 direction 50 ft Ib Charpy V-notch temperature (or the 35-mil lateral expansion

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h e, temperature if it is creater). RT NDT si defined as the dropweight NDTT or the temperature E 60*F less than the 50 ft ib (or 35-mil) Charpy V-notch temperature, whichever is higher. i N

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RTNDT = NDTT, if NDTT > T50(35) - 60*F w L, ,

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= Reference nil-ductility temperature $

. NDTT = Nil-ductility transition temperature per ASTM E208 k T50(35) = 50 ft Ib temperature from Charpy V-notch specimens oriented in the (

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direction normal to the major rolling. direction (or the 35 mil f.

temperature if it is greater) f f

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2 An empirical relationship between RTNDT and fracture toughness for reactor vessel steels has been developed in Appendix G, ' Protection Against luon-ductile Failure," to Section ill of l the ASME Boiler and Pressure Vessel Code. This relationship can be employed to set allow- I able pressure-temperature limitations for normal operation of reactors ivhich are based on fracture mechanics concepts. Appendix G defines an acceptable method for calculating these

' l limitations.

it is known that radiation can shift the Charpy V-notch impact energy curve to higher tem- _

peratures[,U Thus, the 50 ft Ib temperature and RT NDT i ncrease with radiatioe, exposure.

( The extent of the shift in the impact energy curve, i.e., the radiation embrittlement is i-enhanced by certain chemical elements .(such as coppar) present in reactor vessel steels.L

The 50 ft Ib temperature or RT NDT ncrease i with service can be monitured by a surveillance program which entails the periodic checking of irradiated reactor vessel surveillance specimens.

g adl: The surveillance program is based on ASTM E185-73 (Standard Recommended Practice for Ig Surveillance Tests for Nuclear Reactor Vessels). Compact Tension fracture mechanics specimens

$gp will be used in addition to the Charpy V notch specimens and the precracked bend bar i -

specimens to evaluate the radiation effects on the fracture toughness of the reactor vessel materials.BA7.8,9,10,W M

Posti rradiation testing of the Charpy impact specimens will provide a guide for determining

-$.SM-pressure-temperature limits on the plant. Charpy impact test data will determine the shift

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OPw (1) L F. Porter. "Radiatior Effects in Steel," in American Society fpr Testing Materials in Nucteer Applications, op.147195, American Society for Testing Materials, Philadelphia.1950.

%+314 (2] L E. Steele and J. R. Hawthorne "New Information on Neutron Embrittlemerrt and Embrittlement Relief Qg4.

7v ' to Reactor Pressure Vessel Steels," NRL4160, August,1964.

$$jf [3] U. Potapovs and J. R. Hawthorne, "The Effect of Residual E:ements on 55o*F Irradiation Response of selected Pressure Vessel Steets and Weidments," NRL 6803, September 9,196a.

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[4] L E. Steele, " Structure and Composition Effects on Irradiateori sensitivity of Prenure Vessel stests." ASTM-STP.

484, pp.164175. American society for Testing and Materials, Philacelphia,197o.

..hg [5] E. Landerman, S. E. Yanienko, and W. S. Hazetton, "An Evaluation of Radiation osmage to Reactor Vessel Steels

$4 ~~w' using So.h Transition Temperature and Fracture Mechanics Approaches," American society for Testing and D Materials, STP 426, December,1967.

[6] M. .) Manioine, "Bianial Brittle Fracture Tests," Trans. ASME, Su. D. J. Easic Eng. 87. 293-98 (1965).

,17] L orse, " Reactor Vessel Design Considering Radiation Effects." Trans. ASME. Ser. D. J. Basic Eng.

(,$g

(-j;W 87, 743-49 (1964).

[8] re. E. Johnson, " Frat.2re Mechanics: A Basic for Brittle F-2cture Prevention." WAPD-TM-505, Bettis Atomic Q";;{ Power Laboratcry, November,1965.

4.,64'

[9] E. T. Wessel and W. H. Pryle, " Investigation of the Apolicabilit y of the Biamial Brittle Fracture Test for Determining 2EM4 Fracture Toughness " WERL-8844-11, W6stinghouse Rasearch

aborator,es, Augpest,1965.

.*w [10] W. K. Wilson, " Analytic Determination of Stress intensity Factors for the Manioine Brittle Fracture Test Specirnen," l

".}: g g s WERL-0029 3, Westinghouse Research Laboratories. August,1965. ]

2 .M. :M [11] R. E. Johnson and E. J. Pasierb. " Fracture Toughness of frradiated A302 S Steel as influenced by Microstructure," l

.- -1 Amer. Nuct. Soc. Trans. 9. 390-92 (1966).

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of the reference temperature with radiation exposure at the plant temperatures. These data can then be reviewed to verify or revise pressure-temperature limits of the vessel during start-up and cooldown (the Charpy specimens are most nearly indicative of the radiation exposure experienced by the vessel). This will altosv a check of the predicted shift in the reference ,

temperature. The postirradiation test results of the compact tension specimens and precracked bend bar specimens will provide actual fracture toughness properties for Joseph M. Farley -

Unit No.1. These properties may be utilized to establish allowable stress intensity factors .

for normal operation per ASME Ccde Apper. dix G methods. -

Six material test capsules, located in the reactor between the neutron shielding pads and the ' '

vessel wall, are positioned opposite the center of the core. The test capsules are located in _

guide tubes attached to the neutron shielding pads. The capsules contain test specimens from a 9-inch-thick p_ late _from the reactor vessel lower shell course adjacent to the core r'egion, representative weld metal and heat-affected zone (HAZ) metal. The thermal history or heat treatment given these specimens is similar to the thermal history of the reactor vessel material _

with the exception that the post weld heat treatment received by the specimens ,has been simulated (sec appendix A}.

The six material test capsules contain Charpy V-notch irapact specimens, precracked bend bar

.b specimens !from the limiting core region Iower shell course plate) tensile specimens, compact tension specimens (from tne limiting core region lower shell course plate of the reactor vesse! '.~

and associated weld metal) and Charpy V notch impact specimens of HAZ metal. Dosimeters -

and thermal monitors to measure the integrated neutron flux and the temperature are also located in each of the six material test capsules.

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SECTION 2 SAMPLE PREPARATION j

PRESSURE VESSEL MATERIA L .!

2-1. '

, l8 Reactor vessel material from _ lower shell plEte 36919 1 and a weldmen joining sections of material from this plate and an adjoining intermediate shell course plate were supplied by 7

Combustion Engineering. Data on this material are presented in appendix A.

2-2. MACHINING e Test material obtained from the lower shell course plate (after thermal heat treatment and C2 forming of the plate) was taken at least one plate thickness (9 inches) from the quenched ,

t ands of the plate. Test specimens were machined from the 1/4 thickness focation of ' he 5 plate after performing a simulated postweld stress-relieving treatment on the material. Speci-N

._ mens were malso ' achined from weld and heat affected zone metal of a stress relieve i p

ment joining sections of the intermediat; and lower shell plates. All heat-affected zone spect-mens were obtained from the weld-heat-affected zone of plate B69191.

2-3. Charpy V-Notch impact Specimens ,

Charpy V notch specimens from plate B6919-1 were machined with the longitudinal axis of '

the specimens both para!!el and normal to the major rolling direction. Charpy V-notch speci-7 mens from the weld and weld heat affected zone metal were machined perpendicular to the l

weld direction with the notch oriented in the direction of the weld (see figure 2-1). l

24. Tensile Specimens f Tensile specimens from plate B6919-1 were machined with the longitudinal axis of the specimens both parallel snd normal to the major rolling direction. Weld specimens were -l oriented normal to the weld direction (see figure 2 2). l l

2-5. Bend Bar Specimens Bend Bar Specimens were machined from plate 86919 1 with the longitudinal axis of the specimen oriented normal to the rolling direction of the plate such that the simulated ,'

crack would propagate in the rolling direction of the plate. All bend bar specimens were fatigue precracked according to ASTM E399 (see figure 2-3).

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g 26. 1/2 Compact Tension Specimens 1 Compact tension test specimens from plate B6919-1 were machined in both the transverse and e , longitudinal orientations. This was done to obtain fracture toughness data both normal and [

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1 were fatigue precracked according to ASTM E399 (see figure 2-4).

2 7. DOSIMETERS -

Six capsules of the type shown in figure 2-5 contain dosimeters of pure copper, iron, nickel l

.l j and afuminum 0.15 wt percent cobalt wire (cadmium shielded and unshielded) and.Cd shielded e , Np 237 and U 23s which will measure the integratt:d flux at specific neutron energy leve(s, !

9 l y 2-8. THERMAL MONITORS J c7T I

The capsules contain two low-melting point entectic alloys to define more accurately the i m

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c a maximum temperature attained by the test specimens during irradiation. The thermal moni- -

gs k tors will be sealed in Pyrex tubes and then inserted in spacers located as shown in figure 2 5. ]

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.g 2.5% Ag, 97.5% Pb Melting point 579 F [

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8 d- 1.75% Ag, 0.75% Sn, 97.5% Pb Melting point 590 F [

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" 2-9. CAPSULE LOADING f E

y The six test capsules coded U, V, W, X, Y, and Z are positioned in the reactor between the ,

neutron shielding pads and the ve:,sel wall at the locations shown in figure 2-5. Each capsule l contains 60 Charpy V-notch specimens, nine tensile specimens, twelve 1/2T compact tension I

specimens and one bend bar.

The relationship of the test material to the type and r :mber of specimens in each capsule (

g is shown in table 2-1.

F s Desimeters of aluminum 0.15 percent cobalt, cadmium shielded aluminum 0.15 percent co-

balt, pure copper, iron and nickel wires are secured in holes drilled in spacers located at f capsule positions shown in figure 2-5. Each capsule also contains a dosimeter block (figure t

t 2 6) which will be located at the center of the capsule. Two cadmium-oxide-shielded capsules, each containing isotopes of either U 238 or Np237, are located in the dosimeter block. The doub!e containment afforded by the dosimeter assembly prevents loss and contamination by I the U238 and Np 237 and their activation products. Each dosimeter block contains approximately t

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ir-12 milligrams of U2 sa and 17 milligrams of Np* held in a 3/8 inch-long by 1/4 inch-OD

{, sealed brass tube and stainless steel tube, resactively. Each tube was placed in a 1/2-inch-

-Q 237 tube per block), and the t diameter hole in the dosimeter block (one U 23a and one Np

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around the tube was filled with cadmium oxide. After placement of this material, each ho

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'gs j was blocked with two 1/16 inch-thick aluminum spacer discs and an outer 1/8-inch-thick s-Q- cover disc welded in place.

t ik The numbaring system for the capsule specimens and their locations are shown in figure 2 ( l y.

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2-10. SPECIMEN CAPSULE 1 + wn .

The specimens are seal welded into a square capsule of austenitic stainless to prevent corrc-

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' -sn M] of specimen surfaces during irradiation. The capsules were hydrostatically tested in demine-ized water to collapse the capsule on the specimens optimizing thermal conductivity betwe

_7.XI e.w the specimens and the r.eactor coolant. The capsules were helium leak tested as a final ins e

f' procedure. Fabrication details and testing procedures are listed in figure 2-5.

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54 f 0.052 '4

': , i ITEM TITLE NO REQ'D 5 SP C F ATION ...

~

i BLOCK I ' -

2 2 2 f C OV ER ;;

~'

1 3 4 '-

SPACER 4 N E PTUN IUM 237 SE ALED CAPSULE STAI N LESS ,

(0.250 OD 0.375 LG) STEEL I

[ .

g g

5 URANIU4 238 SEALED CAPSULE (0.250 00 x 0.375 LG) SRASS I

-%T h

g 6 CACMluM 0x10E AS REQ'D .'

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=

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5 i gli_ 11111 3

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L r r

[je E! ilili M b b N M

==g; - -

E525 3 r f C r i !

i;; = - -

g ::: 13 -

=

l -

1

= - -

=

!i y

E m

TABLE 21 g '

N TYPE AND NUMBER OF SPECIMENS IN THE JOSEPH M. FARLEY UNIT NO.1 SURVEILLANCE TEST CAPSUL.ES

\

Capsules U, V, W, X, Y, and Z g,

f. Material Charpy Tensile CT Bend Bar j g h Plate B6919-1 I k:!

(Longitudinal) 15 3 4 -

15 3 4 1 (Transverse)

Weld Metal 15 3 4 ,

- - e HAZ 15 -

j M

k k

b a

k..

i b ~

i _

u

, k

- ; Fe i

W

p ll I

@g m

I l I E.

_e. c a

E; I

E.

w f3 l 2 15 f, t

w$

>:h~

i SECTION 3 .

PREIRRADIATION TESTING

I. _

CHARPY V-NOTCH TESTS 31.

Charpy V notch impact tests were performed en material from the vessel lower shell plate 85919 1 at various test temperatures from 50 to 210*F to obtain a Charpy V notch transition curve in both the longitudinal and transverse orientations (tables 3-1 and 3-2 and figures 3-1 and 3 2). Tests were also performed on weld metal and HAZ metal at various temperatures from 150 to 210*F. The results are reported in tables 3-3 and 3 4 and figures 3-3 and 3-4. .

The specimens were tested on a Sontag SI 1 impact machine which is inspected and calibrated '

every 12 months. Charpy V notch impact specimens of known energy values, supplied by the Watertown Arsenal, are used for the calibration.

^

32. TENSILE TESTS 5

Table 3-5 and figures 3-5,3-6, and 3-7 give results of tensile tests performed on material from 1

the vessel lower shell plate B69191 and from the weld metal. Specimens from the shell plate were tested at room temperature,300*F and 550*F in both longitudinal and transverse directions.

The Instron TT C tensile testing machine was set up with the standard instron gripping devices.

A Baldwin-Lima Hamilton Class B-1 extensometer and chart recorder provided a stress strain Yk t

curve for each specimen. The chart recorder was calibrated to the Class B 1 extensometer.The measurement and control of speeds in the tests conformed to ASTM A370-68 (Mechanical Testing of Steel Products). The Instron TT C and the Baldwin Lima Hamilton extensometer are certified as traceable to tht; National Bureau of Standards. A typical stress-strain curve hgi is shown in figure 3 8.

E 12 isis ra Mme

>-- a FD 31 26

M ~ -

W. .

, Y;. -

1 3-3. DROPWEIGHT TESTS

. The nil-ductility transition temperature (NDTT) was determined for plate B6919-1, the core W) region weld metal anc' the heat affected zone by dropweight tests (ASTM E 208) perforrned

$y' Il at Combustion Engineeneg. The following results were obtained:

>2N NDTT ( F)

Material

ds.

, -M Plate 869191 20 k -60

@ Weld Metal

.D M -10 HAZ f.'

ETh nu EM Jy% -

.2 E

WE Thi#.

r- m W5e

%R n:rL cf.'e W d hk-t .

4a *

L .

6. j

! t Wi cm

s. 4 EN n!#

r.:.m

'sbd$ -

=mu

-w w' BOW,

.m ,

%Lii

$ 5Q ,

2%

-a::~u

(

n+e:1;

' a.

r \..gQ

. 4I-

._ a 7

q ~

3-2

. x w:2. n , _ _.a.

y- .

-l

.. ,~

i h t

TABLE 3-1 ; {

PREIRRADIATION CHARPY V NOTCH IMPACT DATA FOR THE JOSEPH M. FARLEY UNIT NO.1 REACTOR PRESSURE VESSEL 7' l.OWER SHELL PLATE B6919-1 (LONGITUDINAL DIRECTION) h b

U Test Temp (*F) Impact Energy (ftIb) Lateral Expansion (mils) i Shear %

F 50 26 14 18 $,-

}-

50 12 10 8 ,{ -

[ -

~

?

l 50 11 14 8 [

L 0 59 30 44 ,

s O 48 27 37 ,

j r.

0 56 27 38 h l

'l 1

j 40 80 50 . 62

};

' 1s t 44 :

40 52 35 .

X

'[

( 40 68 42 53 80 107 80 70 f-80 100 80 73 80 106 80 71 (

130 135 100 80 f

", 130 140 100 90 f 5 130 145 100 88 !

r 210 131 100 83 j i'

210 129 100 85 -

210 142 100 85 I li f

~

t C

.b

, p 33 g; I

.= '

A , .

.w .

TABLE 3 2 sb

E PREIRRADIATION CHARPY V. NOTCH IMPACT DATA FOR THE 9 JOSEPH M. FARLEY UNIT NO.1 REACTOR PRESSURE VESSEL Q

LOWER SHELL PLATE B69191 (TRANSVERSE DIRECTION)

.$ Test Temp (*F) Impact Energy (ft lb) Shear % l.ateral Expansion (mils)

M

'? 40 12 14 h

L34 7

M. 40 25 14 17

?

rij -40 29 20 18 n l

, '$2 0 ;

30 25 21

?.1

.. t i

i3 0 27 5 25 24 ,

i,?.:

O '35 25 26 I i

3.$ . 40 26 37 26 t l I

-['} 40 37 29 4j 30 l

-g 40 44 52 37 71 .y RT 60 55 50 t.f.

j RT 53 64 45

.jj - ,

c; 3 RT 53 50 46

. j. .' 110 75

'

77 56

i.j 110 69 79

.. 53 *

~-2 i a 11C 72 80 90 52 1

q . . .

210 92 100 71 '

d 210 91 100 72

.a .

.. 210 89 100 68 [

-1.1 '

' . ,Aj

~ .A 1

! l 1

i!

!1

. 34 ! '

i

_ . - . _1

. I 10308-7 200

~

180 -

160 -

140 -

1 120 -

a d

g 100 -

i' a

cs:

W w 80 -

O 60 -

t O .

40 -

O 20 -

. 0 l l !

100 0 100 200 300 TEMPERATURE (OF) i Figure 31. Preirradiation Charpy V. Notch Impact Energy For The Joseph M. Farley Unit No.1 Reactor Pressure Lower Shell Plate B69191. (Longitudinal Orientation) i 35 1 -

~

10306 4

I.i w

200

- .J

.s

+

.. 18 0 -

+ _e 19

+

.. -1 160 -

m 2.

Wig -

140 -

4 2q9.?B MkE.I t _ 120 -

v.

ye b ..a 74y a a

i2,;fE d N , D 100 -

Wh.E

,J ..?,1

~p-q o

ca Y '

1517 5 2nud w 80 -

~~P: T t.W:W Q N.c; %c Pld 60 -

~. h%.9 hiti:1 r A

-:q,. . %. :q . qo -

O 2

'O a,ne3 24

- m, * ~m.

Mrc.T

..e r 20 -

A : . 4, L1 v ay = ;. O IJMik ...c:

0 I ! !

y;j:i~ 14 g -100 0 100 200 300 L:.w riC e-

' TEMPERATURE (CF)

=.T,.*. .r 5 .d, -

,4 tJ

,.:w -j z'- ,. . r 1 w....g .

++ _2,.j,

^ :='1 -

, .. j 3 .:;

e -.

'/ ,* g

. I -

Figure 3 2. Preirr%diation Charpy V Notch Impact Energy For The

~~.~_. -

.foseph M. Farley Unit No.1 Reactor Pressure Vessel Lowel Shell Plate B69191. (Transverse Orientation)

S-36

?

F. .

3-s !

l TA8LE 3 3 PREIRRADIATION CHARPY V. NOTCH IMPACT DATA FOR THE JOSEPH M. FARLEY UNIT NO.1 REACTOR PRESSURE VESSEL CORE REGION WELD METAL iest Temp (*F) Impact Energy (ft Ib) chear (%) Lateral Expansion (mils)

-100 5 15 1 1

14 25 11 100 18 20 11

-100

- 40 53 43 43 60 32 44 40 ,

- 40 75 50 55 10 79 65 54 i

10 86 7:s 63 l

10 80 65 58 72 117 100 82 72 123 100 80 72 113 100 79 I 150 151 100 90 150 144 100 89 118 100 81 150 138 100 88 210 159 100 85 210 151 100 85 210 3-7

-s ,

$p- -

l w .. .

TABLE 34 b PREIRRADIATION CHARPY V NOTCH IMPACT DATA FOR THE

'd JOSEPH M. FARLEY UNIT NO.1 REACTOR PRESSURE VESSEL CORE REGION WELD HTAT AFFECTED ZONE MATERIAL

% g Test Temp (*F) Lateral Expansion (mils)

Impact Energy (ft-lb) Shear (%)

R 150 15 18 7 c.41 22::1 6 150 18 g 11 ,

N -150 58 45 29 ff r 55 54 Q

- s .-,

100 103 im 100 33 32 19 3g 100 67 45 40 i:E

t-n - 75 101 65 62

?%

Cfh 75 110 65 57 1$.$)

80 72 j 20 122 ii]

th

- 20 120 80 '

71 20 84 65 52

-'s 50 141 100 83

"'.T 1.V+ t- 50 142 100 85

%.9 100 74 57 7.1 75 150

e. .-d-I 210 132 100 85

.1hd" 210 170 100 83 E~k_b.

i;c.i2 - -

210 163 100 85

iw:i 11 .

f.4

l tb ~

.! 38

l I

l}

10306-5 t

I I

l l

- 1 200 l

1 i80 -

I60 -

O -

O ,

i 14 0 -

O

120 -

O T

100 -

! E 5

3 80 -

O 60 -

14 0 20 -

0 ! ! !

-200 -100 0 100 200 300 TEMPERATURE (OF)

Figure 3-3. Preirradiation Charpy V Notch Impact Energy For The Joseph M. Farley Unit ? o.1 Reactor Pressure Vessel Core Region Weld Metal 39 l

i- i 1

1 -

, j 10308-4 , '

f1 1 200 -

a 4 _

180

O b' 160 - O .

i 140 -

a . O

_ l20 -

3 m ,

A T

+ .

O j i - D., 100 - O a

w 80

{

r 60 -

3 4

O k -

i 40 -

O t

.i

20 -

/

i 7 ' g ,

~) 0 S -200 -iOO O 100 200 300 3 TEMPERATURE (OF) l 3e -

j q.

3 i

-i -

1 i

jl Figure 3 4. : ,

Preirradiation Charpy V Notch Impact Energy For The Joseph M. Farley Unit No.1 Reactor Pressure Vessel

. Core Region Weld Heat Affected Zone Material 3 10 l

s -

I ;j! , i1 J1 Jw

_o

.m p w

a e

m r

A m

)

_ I n%( .

. 5 9 6 5 1 6 1 8 8 7 7, 9 0 1 1 4 6 0 .

d 1 0 9 1 7 8 l 4 7 3 1 9 3 2 1 1 6 5 e '

R 7 7 6 7 6 6 6 6 5 6 6 5 7 7 7 7 6 6 .

. n l

o . .

l E )% 5 8 5 3 6 0 5 9 3 2, 3 9 8 8 2 l

a( 1 7 1 .

t o 7 8 4 4 3 4 5 2 0 1 1 0 5 1 0 0 1 0 .

T 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 2 _

Y n

E o .

l ,

L E R E ,

AA T m

r )% .

F L f i

o( 8 0 6 0 5 7 -

7 2 7 4 0 3 8 1 3 0 2 6

.P n 5 5 3 3 1 3 5 1 2 2 3 2 2 1 0 2 1 0 n 1 1 M U 1 1 1 1 1 1 1 1 1 1 1 1 f. 1 1 1 L

HL s s _. ,

P EL e r

EHA S t

w . .

SSI R

w O el r is 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 0 J ET RE u p 5 0 0 0 0 0 2

0 0 0 0 5 0 0 0 5 0 0 0 0, 0 4, 2, 1, 3, t

c( 8, 2, 0 8 7 0 9 5 2, 7, 6, 1, 1, 8, 6, 7, 7, 7 2 3 3 7 8 5 8 3 6 8 7 E WA a

r 8 7 6 7 5 6 5 7 4 5 5 5 0 9 8 8 61 7 H OM T

F 1 1 1 1 1 1 1 1 1 1 1 1 2 1 1 1 1 L ,

R LL D d a

O F

EE L o

SW S .n.

5 S EN e )h r

3 E u (t m

L. I T VO t c 0 0 0 0 0 0 5 5 0 0 0 0 0 0 0 0 5 0 5 0 0 5 7 5 0 01 5 5 5 0 2 2 2 7 5 E

L REG I

F a

r 5 5 5 4 6 0 2 2 2 2 2 2 8 0 0 7 0 2 3 3 2 2 3 1 7 7 6 8 7 0 2 2 2 2 2 3

.n B ERE A PU R T OS R SE P E R EPR O

)

(

i S

T s

p 0 0 G 0 0 0 0 0 0 5 5 0 7, 8, 0, 8, 4, 5 4 4 8 7 0 0 0 0 0 0 0 0 0 0 0 0 0 5 0, 7, 3, 8 6 0 8 8 5 7 2, 4 0 0 0 0 0 0 5 5 0 5 0 0 4, 2, 4, 6, 2 0 0 3 3 0 8 2,

w

.w L C U 8 8 7 7 8 8 8 9 7 7 8 8 9 9 8 8 8 8 .,

I R -.

SO D N . ,n.

NT l ECA d i

s p

I T EA1 -

l ie( h N Yt IO1 R1 9

9 %e 2 r g

n 0 0 0 0 0 0 0 0 0 0 0 0 6, 1, J. 0, 6, 8, 0 0 0 0 0 0 5 5 0 5 5 0 1

3, 6, 3, 8 8, 0 0 0 0 0 0 5 0 0 0 0 0 2, 8, 0, 9, 4, 1, n

T 6 0S t 4 4 8 9 4 4 6 6 5 5 5 5 5 0 8 9 8 0 6 6 5 5 5 6 8 7 9 0 7 9 7 7 6 7 6 6 -

A O.B I

DN )

F n A . ,.

RT I

(

p '

RN I

m e

EU T ) )

R t ), ) ) ) .

P s e 5 57 00 00 505 0 5 7 7 5 00 00 5 05 0 5 7 7 5 00 00 5 G5 0

=

1 T ( ( 3 3 5 5 ( 1 3 3 5 5 ( ( 3 3 5 5

)

n o )

n it c o e it r

c iD e i

i 1

9 l

a 1

9

- D -

n c l .

t a

1 i 1

s a 9 d 9 t .

M 6 u 6 ie e ,

B it U w M l

e s e n g

e n a d -.

s e

t o t a i le L T V Pla (

l P ( W y '- ~

~

's

'o j .. .

4 1030s 3 1

l 81 .

'9; i 3 '

le l l l l l

\G -

n l s<

'$ ~

m

@ 80 -

'g 15 8 o '

t ,_ -

ULTIMATE TENSILE

'A P e STRENGTH '

j 0 60 V

O Cc N

m 4 Y h : -

5 0.2% YlELD STREMGTH 3

5 40 h

g

.T -

h4 80 ma c

REDUCTl0M IN AREA s~; n -

g v n i 9

t

. Cm:

60 -

O

W ;

%. . u

.. cc ,

N E

w hi 40 -

s 3 r UNIFORM ELONGATION II -

E -

TOTAL ELONGATION I g ci D U O

u 20 -

o- .

9

$.q 1 v

m- o

i i i I i 1

,g 0

}i 0 100 200 300 400 500 600

{l TEMPERATURE (OF) as 3).

Figure 3 5. Preirradiation Tensile Properties For The Joseph M. Farley Unit No.1 Reactor r. essure Vessel Lower Shell Plate 86919-1 (Transv:ne Direction) ek 3-12

.y .

.V 9 L.

k ts 10305-2 h r.

I -

PJ i-1 .

P g

i 100 1 I I I I I e 2 i

. 2 ,

0-e r 3 b 80 -

h 8 4 G

~ ULTIMATE TENSILE b i

STRENGTH 0

I

0 60 -

1 ,.

, N o 4 4 b

/ i -

2 C.25 YIELD $12ENGTH f l . . ,-

i 40 : L c

N t

80 T-f G 1 t V 4_

h , p 60 - e O REDUCTION IN AREA i 5 75 '

E w .i: 6 i 40 -

UNIFORM ELONGATION [

[ TOTAL ELONGATION [*

3 2

~

M }2 - m

) Q g

o 20 -

1

^

" O I V _

c I I '

I I i O

. h.

0 100 200 300 400 500 600 1 L

TEMPERATURE (CF) j i

f Figure 3 6. Preirradiation Tensile Properties For The Joseph M. Farley

Unit No.1 Reactor Pressure Vessel Lower Shell Plate

! B6919-1. (Longitudina! Direction) I

! h I

3-13 ;I l

J .

1 .

i 10308-1 1

J l

100

I I I I l

?

- 4 80 -

3 C [ ULTIMATE TEMSILE STRENGTH v G i e V

e i

g 60 -

o.2% YlELD STRENGTH 1,

i e i

T 1 40 1 .

I 1 -

t l 80 a

'.! ~

9

- 60 -

. w E REDUCTION IN AREA

=

40 -

UNIFORM ELONGATION i s

~~~

d TOTAL ELONGATION 2

8 c1 20 - O d g -

n v n e

a 9 4 I I I 0 I l !

0 100 200 300 400 500 600 TEMPERATURE (OF)

I Figure 3-7. Preirradiation Tensile Properties For The Joseph M. Farley Unit' No.1 Reactor Pressure Vessel Core Region Weld Metal i

3 14 !

i I

. g1 f

~.

C?

r>e . !

,. 3 l

Q*

e.

n.~

, C*..

R"'.,

e:!

t-:

.i Qr

. G.

1. R.-

~

w

- w

. . rp

. La l

Ri

, h.*.

4 P 8

IW e e r

F

?

- A ,? >-

4

. vw

, .= ,

1 ,

1 b.

t. .c .

> .l ic l ~ :x

.< , o .e4-tu (f

2 y . c4.- c U le x e

. ::C_.

-- ( w

- I,, e cm t

- . w m .

m ) l,'.

1 w

m *

~

fr n : t. ..

u

. u

'A ! . D. J H e tr.-

IL t p:. -

co.

a n .;

< f: ,

3 .- .

.es

u. L.

,. -r.

t i.

'r. .

l '>

q,.

4

. O l.

SS3HIS l l 1

I >a

(

!'l ir ,

e 1r-P 3 15

?

. . ,r .

\

t .

SEC' lion 4

_ ; POSTIRRADIATION TESTING i

.}

I 4-1. CAPSULE REMOVAL

>I

Specimen capsules will be removed from the reactor only during normal refueling periods.

., The recommended capsule removal schedule is as follows:

Multiplying Factor By Capsule Which,the Capsule Leads ,

I identification Vessel Maximum Exposure Removal Time U 2.6 End of First Core Cycle i W 2.0 10 Years i

l Y 2.0 20 Years

~j 30 Years

Z 2.0 V 2.6 Standby 2.6 Standby X

Each specimen capsule, removed after radiation exposure, will be transferred to a post-irradiation test facility for capsule disassembly and testing of all specimens.

4-2. CHARPY V. NOTCH IMPACT TESTS The testing of the Charpy impact specimens from the lower shell plate, the weld metal, and HAZ metal in each capsule can be done singly at approximately fiva different temperatures.

The extra specimer.s can be used to run duplicate tests at temperatures of interest.

The initial Charpy specimen from the first capsule should be tested at room temperature.

The test value for this temperature should be compared with preirradiation test data. The temperatures for the remaining specimens should then be appropriately higher or lower. For ;

succeeding tests after longcr irradiation periods, the test temperature in each case should be chosen in the light of results from the revious :apsule.

l l 4-1

> AW ~

.M .

}E -

43. TENSILE TESTS

-d.

a f The tensile and fracture toughness soecimens,for each of the irradiated materials should be tested at room temperature, 300 F, and 550 F.

h 6

4-4.

FRACTURE TOUGHni?SS TESTS ON 1/2T COMPACT TENSION SPECIMEN AND BEND BAR SPECR.MNS g in light of current requirements of 10 CFFi, Part 50, ASME Code, appendix G,1/2T compact Q tension (CT) specimens should be tested dynamically to adequately characterize the fracture

(

6 toughness properties.of the reactor vessel up to the initiation of the fracture toughness upper

shelf. The CT specimens for each of the irradiated materials should be tested in aceo:tfaces

$4 with ASMT E399-74 with appropriate modifications necessary for dynamic tests. Testing -

.$g dynamically in the fracture toughness ductile to brittle transitiors region and at upper shelf

% initiation temperatures results not only in lower bound data but also provides an opportunity for obtaining valid N fracture toughness data up to the onset of upper shelf. This results from Q{h fy nonlinsar cleavage behavior which occurs only in dynamic' testing at these temperatures. The load-displacement curve exhibits a claar drop in load at the onset of cnck initiation, thereby (jf eliminating any possible doubt as to the start of crack initiation, as is the case in static loading l QII conditions at these temperatures. Recommended test temperatures are equal to or lower than '

7 those characteristic of the upper fracture toughness shelf initiation temperature.

h Analysis should be performed using the J Integral or Equivalent Energy Concept [1,21. Testing

at temperatures characteristic of the fracture toughness upper shelf is not suggested due to the '

uncertainty of the point of crack initiation even when dynamic testing is performed. At these

@ temperatures, static Jge testing appears to be most indicative of conservstive upper shelf fractura -

$ toughness properties. Research in this area is currently being conducted by Westinghouse Resear: ' ;

.- 6,

.. m and Develcpment Laboratory, ASTM E24, NRC, and others. Use of this technique will' be furthe ;

% evaluated as it app!ies to surveil lance specimen testing. The precracked bend bars will be used l

h* ~f to obtain additional toughness data at a temperature indicated by the toughness results of the

)

compact tension testing.

ul w .

d,e 4-5. POSTlRRADIATION TEST EQUIPMENT li The following minimum equipment is required for the postirradiation testing operations: -

_il

$ a Milling machine or special cutoff wheel for opening capsules, dosimeter blocks

i. . .l and spacers

1. " Fracture Toughness." American socaty for Test.w and Materials, STP 514. septemt>er,1972.

's 2. T. R. Ma9er, "Emperimental verificaten of Lome Bund K. Values utilizing the Equivalent Energy

A Concept." HsST semi-Annual informanon Meeting. Pacer N$ 23, April,1972.

42 n- ___ _

W.. *

. . . . t

[ j il e Hot-cell tensile testing machirie with:

11 oin type adapter for testing tensile specimens * :J

2) three-point loading assembly for testing tha bend bar specimens m Hot-cell dynamic CT testing machine with clevis and appropriate measuring

. i, equipment associated with dynamic testing .. '

s Hot cell Charpy impact testing machine I '

a Sodium iodide scintillation detector and pulse-height analyzer for gamma counting of the specific activities of the desi:neters t' .

? .

i ,

I i

5 t,

rch ner 1;

i 4-3 I

7 *' m o

7 u

r.

. .?.,-

9 7@

! g APPENDIX A 4

JOSEPH M. FARLEY UNIT NO.1 REACTOR @

PRESSURE VESSEL SURVEILLANCE MATERIAL $

For the Reactor Vessel Radiation Surveillance Program, Combustion Engin - t Westinghouse with sections of SA533 Grade Bfrom Class 1 plate the 9-inch low. used in the the Joseph M. Farley Unit No.1 Reactor Pressure Vessel, specifically,B69191 /

er shelt plate B6919-1. Also supplied was a weldment made fromd sections of pl and adjoining intermediate shell plate B6903-2, using weld wire representati -L in the original fabrication. The plates were produced byt ialLukens is as Steel Compa treatment history and chemical analysis of the pressure vessel surveillance ma er follows:

i Heat Treatment ,

j 1550*/1650*F . 4 hr - Water Quenched Lower Shell Plate B6919-1 1225*F : 25'F 4 hr 1150*F 2 25*F 40 hr . Furnace Cooled to 600*F l

. o l 1150 F 25 F 16 hr . Furnace Cooled ;

Weidment 6

A-1 l

1 - - - - -- g

s -

, _. l I

TABLE A 1

{ CHEMICAL COMPOSITION (WT %)

Plate B69191 Weld Metal I'*'"I Combustion Engineering Westinghouse Westinghouse 1 Analysis Analysis Analysis i -

y C 0.20 -

0.13 l S 0.015 0.013 a 0.009 f N2 -

0.003 0.005 Co 0.008 2 0.016 0.018 l Cu 0.14 0.10 0.14 i

? -

Si 0.18 0.28 027 3 Mo 0.56 0.51 0.50 1'

j Ni 0.55 0.56 0.19 t-3 Mn 1.39 1.40 1.06 I ,

Cr -

0.13 0.063 V

{ -

<0.001 0.003 i

P 0.015 l 0.015 0.016 I

t- Sn -

0.008 0.005 Al 0.025 -

0.009

'All elements not listed are less than 0.010 weight %.

I A2

-. 4 ._ ,-; -- -- - - - - -

ML20002E349

Text

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