ML20064G999

From kanterella
Jump to navigation Jump to search
Forwards Addl Info Requested in Re Pressure Vessel Fracture Toughness Properties for Facil.Info Incl: Radiat Surveillance Prog & Purchase Background Matl
ML20064G999
Person / Time
Site: North Anna  Dominion icon.png
Issue date: 12/11/1978
From: Stallings C
VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To: Harold Denton
Office of Nuclear Reactor Regulation
References
NUDOCS 7812150277
Download: ML20064G999 (64)
v • d • e


Text

.

,a f-VIRGINIA EI.zCTRIC AND Powsm COMPANY Rucumown.VamarmzA anset December 11, 1978 Mr. Harold R. Denton, Director Serial No. 581B/120677 Office of Nuclear Reactor Regulation T0/DLB:sej Attention: Olan D. Parr, Branch Chief Docket Nos:

50-338 Light Water Reactors Branch No. 3 50-339 Division of Project Management License Nos: NPF-4 U. S. Nuclear Regulatory Comunission CPPR-77 Washington, D. C.

20555

Dear Mr. Denton:

  • ' (

Subject:

Pressure Vessel Fracture Toughness Properties North Anna Power Station, Unit Nos. I and 2 This is in response to your letter of December 6,1977 which requested -

additional information on pressure vessel fracture toughness properties for North Anna Power Station Unit Nos. I and 2.

The information requested is enclosed.

r Very truly yours, A!b!7Y C. M. Stallings Vice President-Power Supply and Production Operations Attachments 1.

North Anna Unit No.1 & 2 Reactor Pressure Vessel Information (pages 1-7).

2.

North Anna Unit No. 1 Reactor Vessel Radiation Surveillance Program (WCAP-8771).

I 3.

North Anna Unit No. 2 Reactor Vessel Radiation Surveillance Program (WCAP-8772).

cc:

Mr. James P. O'Reilly 73'l2f5o237

D

,e e,

North Anna Unit No.1&2 Reactor Pressure Vessel Information 1.

Purchase date of both vessels was May 1,1969 2.

Purchase order for both vessel was placed with De Rotterdamsche Droogdok Maatschappij N.V.

3.

Vessel fabricator - De Rotterdamsche Droogdok Maatschappij N.V.

4.

Vessels fabricated to 1968 Winter Addenda of the ASME Section III Boiler ~

.and Pressure Vessel Code i

i Identification, chemical composition and fracture toughness properties of'

('

reactor vessel. beltline region material for the Unit No.1 vessel are shown gl

~

in Figure 1, and Tables 1 thru,,4 and for the Unit No. 2 vessel in Figure 2 and Tables 5 thru 8.

Surveillance program for the vessels is described in WCAP-8771 and 8722 for Unit No.1 and 2 respectively. Surveillance program for both vessels is not f

'in compliance with Para. II.C.2 of Appendix H to 10CFR50 since some capsules t

are located in regions of the vessel where the neutron flux received by the specimens will be less than that received by the vessel inner surface.

t i

I I

l l

1 1

e f

i l

. i

'r '

SECTION 1 PURPOSE AND SCOPE The purpose of the Virginia Electric and Power Company, North Anna Unit No. 2, sur-veillance program is to obtain information on the effects of radiation on the reactor vessel materials of a reactor during normal operating conditions. Surveillance material is selected as the most limiting material based on surveillance selection procedures which are outlined in ASTM E185 73, Annex A 1. Evaluation of the radiation effects is based on the pre-irradiation testing of Charpy V notch, tensile, and dropweight specimens, and post irradiation testing of Charpy V notch, tensile, and wedge-opening-loading (WOL) specimens.

Curreret reactor pressure vessel material test requirements and acceptance standards use the i

reference nil ductility temperature, RTNDT, as a basis. RTNDT s determined from the dropweight nil-ductility transition temperature, NDTT, or the weak (transverse-oriented) direction 50 ft Ib Charpy V-notch impact temperature (which ever value is greater) as defined by the following equation:

RTNDT = NDTT, if NDTT >T50(35) - 60*F or 50(35) - 604, if T50(35) - 60'F > NDTT RT

=T NDT where i

RT

= Refence niMuctHity ternperature NDT NDTT = Nil-ductility transition temperature as per ASTM E208 T

= 50 ft Ib temperature from transverse-oriented Charpy 50(35)

V-notch impact specimens (or the 35 mil temperature l

if it is greaterIU) l

1. In the case where at leest 35 mits lateret expension is not obtained at the 50 ft Ib temperature, the temperature at whch 35 rmis lateret expension occurs is used.

11

9 An empirical relationship between RT and fracture toughness for reactor vessel steels NDT has been developed and is presented in appendix G, section 111, of the ASME Boiler and Pressure Vessel Code (Protection Against Non Ductile Failure). This relationship can be employed to set allowable pressure-temperature relationships, based on fracture mechanics concepts, for the normal operation of reactors. Appendix G of the ASME Boiler and Pressure Vessel Code defines an acceptable method for calculating these limitations.

It is known that radiation can shift the Charpy impact energy curve to higher tempera-tures. [1.21 Thus, the 50 ft Ib temperature, and correspondingly, the RTNDT, increase with radiation exposure. The extent of the shift ilt the impact energy curve - that is, the radiation embrittlement - is enhanced by certain chemical elements, such as copper, present in reactor vessel steels. [3.41 The 50 ft Ib temperature, and correspondingly the RTNDT, increase with service and can be monitored by a surveillance program which consists of periodically checking irradiated reactor 4

)

vessel surveillance specimens. The surveillance program is based on ASTM E185-73 (Standard Recommended Practice for Surveillance Tests for Nuclear Reactor Vessels). WOL fracture mechanics specimens will be used in addition to the Charpy impact specimens to evaluate the effects of radiation on the fracture toughness of the reactor vessel materials. [5.6.7.8,9,10.11]

1. L. F. Porter. " Radiation Effects in Steet." in Merenals m Nuclear Applications, ASTM STP 276. pp.147-195.

Arnevican Society for Testing and Metenais, Philadelphia.1960.

2. L. E. Steele and J. R. Hawthorne. "New Information on Neutron Emeritttement and Emantttement Relief of Reactor Pressure Vesset Steels." NRL41eo. August 1964.
3. U. Potapovs and J. R. Hewthorne. "The Effect of Residual Element on 550*F liradiation Response of Selected Pressure Vessei Steels and Weidments." NRL4803. September 1968.
4. L. E. Steeie. " Structure sad Composetion Effects on irradiation Sensitmty of Pressure Vessel Stee4s." m /rrad,ation Effects on Structural Alloys for Nucteer Reactor Applications ASTM STP4ae. pp.164-175. Amencen Society for Tester.g and Meterials. Philadetohia.1970.
5. E. Landermen. S. E. Yanschko. and W. S. Hazetton. "An Evaluation of Radiation osmage to Reactor Vessel Steets Us.ng Both Transition Temperature and Fracture M thanacs Approaches." in The Effects of Radiation on struerural Metais. ASTM STP426. pp. 260 277, Amencan Society for Testing and Materials, "hiladespnia,1967.
6. M. J. Mantoine. "8 annel 8nttle Fracture Tests." Trans. Am. Soc. Mech. Eners. e7, Senes o, 293 298 0 965).
7. L. Porse. " Reactor Vesses ossegn Considenng Radiation Effects." Trans. Am. Soc. Meen. Enges. 8e. Senes o.

743 749 0 964).

8. R. E. Johnson " Fracture Mechanics: A Bases for Brittle Fracture Prevention." WAPo-TM405. November 1965.
9. E. T. Wesset and W. H. Pryte. " Investigation of the Applicacelity of the Bianist 8nttle Fracture Test for ootermming Fracture Toughness." WERL484411 August 1965.
10. W. K. Wilson. " Analytic cetermmotion of Stress Intensity Factors for the Menjoine Snttle Fracture Test Specimen," WERL 0029 3. August 19e5.
11. R. E. Johnson end E. J. Poseerb. " Fracture Toughness of Irradsated A302 8 Steel as influenced by Microstructure."

Trans. Amer. Moct. Soc. 9. 390 393 0 966).

12

,~

's Post-irradiation testing of the Charpy impact specimens provides a guide for determining pressure temperature limits on the plant. A temperature shift in the reference temperature will occur in the irradiated Charpy impact specimen test data as a result of radiation exposure at plant temperatures. These data can then be reviewed to verify or establish new pressure-temperature limits of the vessel during start-up and cooldown. This allows a check of the predicted shift in the reference temperature. The post irradiation test results on the WOL specimens provide actual fracture toughness properties for the vessel material. These properties may be used for subsequent evaluation as per the methods outlined in ASME Code, appendix G.

Eight material test capsules are located in the reactor between the thermal shield and the vessel wall and are positioned opposite the center of the core. The test capsules are located in guide tubes attached to the thermal shield. The capsules contain Charpy impact, WOL, f

and tensile specimens from the limiting core region forging. This forging is the reactor vessel intermediate shell plate adjacent to the core region. Charpy impact, WOL, and tensile specimens obtained from the representative core region weld metal, and Charpy impact specimens from the weld material heat-affected zone (HAZ), are also located in the capsules. In addition, dosimeters to measure the integrated neutron flux and thermal monitors to measure temperature are located in each of the eight material test capsules.

The thermal history or heat treatment given to these specimens is similar to the thermal history of the reactor vessel material, except that the post-weld heat treatment received

~

by t.ie specimens has been simulated (appendix A).

l i

t I

i l

10 13

SECTION 2 SAMPLE PREPARATION 21.

PRESSURE VESSEL MATERIAL Reactor vessel material was supplied by the Rotterdam Dockyard Company from intermediate shell forging 04, heat no. 990496/292424. A submerged arc weldment which joined sections of material from this forging and an adjoining lower shell course forging was also supplied by the Rotterdam Dockyard Company. Data un the pressure vessel material are presented in appendix A.

2-2.

MACHINING Test material was obtained from the intermediate shell course forging after the thermal heat treatment was complete and the forging formed. All test specimens were machined from the 1/4 thickness section of the forging after a simulated postweld stress-relieving treatment on the test material was performed. The test specimens represent material taken at laast one forging thickness (as-quenched) from the quenched ends of the forging. Specimens were machined from weld and heat-affected zone (HAZ) material of a stress-relieved weldment which joined sections of the intermediate and lower shell courses. All HAZ specimens were obtained from the weld HAZ of forging 04, heat no. 990496/292424.

(

23.

Charpy V Notch Impact Specimens (Figure 2-1)

Charpy V notch impact specimens from forging 04 were machined in both the tangential orientatien (longitudinal axis of specimen parallel to major working direction) and axial orientation (longitudinal axis of specimen perpendicular to major working direction). The core region weld Charpy impact specimens were machined from the weldment such that the

~

long dimension of the Charpy was normal to the weld direction; the notch was machined such that the direction of crack propagation in the specimen was in the weld direction.

24.

Tensile Specimens (Figure 2 2)

Tensile specimens were machined with the longitudinal axis of the specimen perpendicul6r to the major working direction of the forging. Tensile specimens were also removed from the core region weldment.

2-1

na a

-~,

a-J.<,

a a

As._

"a2_,

w

's e

?

e o

D 2

Y E

n I

a S

E i

m_

n n

qf 8

u a.

koo.

g

-- e g

(

0 0as g

o-a g:

=

66 c

a e

.N m-o 5

G

~~

.,e b

m O

/

a h

\\

s 2

a

o. $.

O

,a O

d y

C $

l e

h

}

=

=

f E

a. "

a o II 22

a e0 e

t Q

f

=

s v

4

~

=

8 w

o o,

B" w

M E

ad.

e w

K 4

C E f

o.

c o a

w I

A "E

w E

W m

se, sao 9

- m

.a s

m w o-a 3

w K

g-

= a q

z 4

  1. 8 I

m w Q

%a w *

- o i,.

o.

yA y

i o

u a

8 4/5 *=

W

=a s/)

- Y w

Z 4 E

.. - e Om

." $ U

('

=

r

-o

o..m.

- - - ~

egg S

de *

  1. m g

a a

U G y 5

o.

~

oo o

I o

o e

e. s N

aggs====

aF g

JL JL d I

C R

/

o l

/

E s.

y y

g w

w g

L.l O

3 E

/

N t-'

=

u

.n o :=

s g

~

K N G S o-o g

- e g g a o.

m Di a

=

=

.: d E

E S S EY

=

Y CG e

Y o m n a

g g.*.

=

mg s d y

E" l

/-

4

\\_/ J !L oB e

w WT Y

d E5 O

BB gg l

l l

t i

l 23

l 2-5.

Wedge Opening Loading Specimens (Figure 2 3)

Wedge opening loading (WOL) test specimens were machined along the axial orientation so that the specimen would be loaded perpendicular to the major working direction of the forging and the simulated crack would propagate along the hoop (tangential) direction. The weld specimens were machined such that the crack will propogate in the weld direction.

All specimens were fatigue precracked according to ASTM E399 70T.

26.

MONITORS 2-7.

Dosimeters Eight capsules of the type shown in figure 24 contain dosimeters of copper, iron, nickel, and aluminum 0.15% cobalt (cadmium-shielded and unshielded; wire, neptunium 237, and uranium 238. The dosimeters are used to measure the integrated flux at specific neutron energy levels.

2-8.

Thermal Monitors The capsules contain two low-melting-point eutectic alloys so that the maximum temperature attained by the test specimens during irradiation can be accurately determined. The thermal monitors are sealed in Pyrex tubes and then inserted in spacers (figure 2 4). The two eutectic alloys and their melting points are as follows:

2.5% Ag, 97.5% Pb Melting point 579'F 1.75% Ag, 0.75% Sn, 97.5% Pb Melting point 590*F 29.

SURVEILLANCE CAPSULES 2 10.

Capsule Preparation The specimens were seal welded into square austenitic stainless steel capsules to prevent cor-rosion of specimen surfaces during irradiation. The capsules were hydrostatically tested in domineralized water to collapse the capsule on the specimens so tht optimum thermal con-ductivity between the specimens and the reactor coolant could be obtained. The capsules were helium leak tested as a final inspection procedure. Finally, the capsules were coded S, T, U, V, W, X, Y, and Z. Fabrication details and testing procedures are listed in figure 2 4.

2 11.

Capsule Loading Upon receipt, the eight test capsules are positioned in the reactor between the thermal shield and the vessel wall at the locations shown in figure 2-4. Each capsule contains 44 Charpy V-notch specimens,4 tensile specimens, and 4 WOL specimens.

2-4

~ '

NOTE S:

l.

@ ALL OVER

2. MOICH DE PlH 10 Bf Exif n[4 D Bf 0.09-0.1%

Bf f AllGUE CRACRING

3. SLOI 10 8[ PARALLEL 10 SURIAC[

'A' WilHIM 0.00N TOIAL

-C-l

4. SLOI IO Bf PERftNDICI;LAR to SURE ACE S 'S' 5 'C' 4.45 WIININ 0.005 1.43
5. DIM.S MARalD IMUS
  • MUSI NOI natl A !. k.0. ItCllDanG l.130-t 0.001 ON OPPOSlif (ACIS m

0 380 0'

0 f

Sil #0lf 2 0.' l

+

9 A

X h

0.279

0. 01 0.439 0*283 y

h I.1 0 43

-M~~~~~@

0.499 1.005 k

0.439

~

0.995 Ei. E E #__#_r _r j h

JL h

"2

.)

0 435

--.-,=_.s,---

. m g __

- h * '.

'il l 0 V. $.

  • y.h DIM
  • Y" -

P d

h

\\

l

-A-l

0. 00.',

4 (0. l l 2 ). 4 Mut-2 8 0.001 0.501 O

0.500-20 THD. CLASS '38' O.499 Df E P (4)

DIM.*I* DIN.*Y*

0.375 DEEP 0.123

0. 0163 0.0231 0.0473 0.0236 SPECIMEN IDEMIIFICAIION (2 ) LOCAll0NS Figure 2 3. Wedge Opening Loading Specimen
u. m i It 1

Ili

\\

.?$ \\k

\\

\\

-h k:S."r-

.g

[@4 E W 1-St.:"

a3

. G ;; -

3 0

33Q bb k

j ig Fa :

5

?Wi

  • NJ.

I!f a

.i-N Nh

/*

i r.=

~a/

,1 p.;

m?- e-it.rt W

N r

e j

i g4 N '\\

N~ b '

3m 11 ds b!

I ilv

,jt ts in d@M g-ji rgq j

N

-e s

~

3%

gNN e

.c u

)

  • NN

)N i}[

ph \\

i 2

75\\

t N.

V N

'i G/*

N3 e

4 i Y7N*

Lj 7:

5

]% dd e

jd 5y~*

pQ i

4g i p p'14Li s.

}.$ ail

!,Ta yu e lDjd I

l The relationship of the test material to the type and number of specimens in each capsule is shown in table 21. Dosimeters of pure cooper, iron, and nickel and cadmium-shielded wires are secured in holes drilled in spacers located in the capsule positions shown in figure 2 4. Each capsule also contains a dosimeter block (figure 2 5) which is located at the center of the capsule. Two cadmium-oxide-shielded capsules, each containing isotopes of either U or Np237 (both 238 99.9 percent pure) are located in the dosimeter block. The double containment afforded by 238 237 and their the dosimeter assembly prevents loss and contamination by the U and Np activation products. The amounts of each are presented in table 2-2. Both of them are held in a 3/8 inch long by 1/4-inch-OD sealed brass tube and stainless steel tube, respectively. 238 and one Each tube is placed in a 1/2 inch-diameter hole in the dosimeter block (one U 237 Np tube per block), and the space around the tube filled with cadmium oxide. After ( placement of this material each hole is blocked with two 1/16-inch thick aluminum spacer discs and an outer 1/8-inch thick steel cover disc welded in place. - The numbering system for the capsule specimens and their locations are shown in figure 2-6. TABLE 21 463 TYPE AND NUMBER OF SPECIMENS IN THE NORTH ANNA UNIT NO. 2 SURVEILLANCE TEST CAPSULES Capsule S,V,W, and Z Capsule T,U,X, and Y Material Charpy Tensile WOL Charpy Tensile WOL Forging 04 (Tangential) 8 8 Forging 04 (Axial) 12 2 4 12 2 Weld Metal 12 2 12 2 4 HAZ 12 12 1 TABLE 2-2 QUANTITY OF ISOTOPES CONTAINED IN THE DOSIMETER BLOCKS lootope Weight (mg) Compound Weight (mg) Np 37 { j~ 237 , NpO2 20 1 U238 12 U038 14.25 27

HAIERIAL N0. 1IEH IlILI SPECliICAIl0h REQ'O. I EIOCK I I 2 C0VER 2 I I s l l l 1 l 3 SPACER 4 232 4 NEPIUNIUH SEALED CAPSULE STALKLESS 1 (0.250 OD a 0.375 LG) SIEEL 230 5 URAN 10H SEALED CAPSULE BRASS 1 (0.250 0D a 0.375 LG) 6 C ADHitM 0Xt 0E AS Rf0'0 2 M { 00 0.06 A ' TYP* l ~ l g ~ I I y

,,,),,,,-

O 6 I Figure 2-5. Dosimeter Block Assembly I g. U V

.I 1 ) 1 l l l 1 ~1 i l e .e e. e. a e e. e. s e e e e e e e a 2 a as 3 J + 3 I I 1 .I 3 i 2 i sers e 5 e e I 1 1 1 3 8.. .s** .s I as e e e e g g, g 9 I l st a a i --): i.f 5 3 5 v 5 1 5 0 5 t i I V E V 4 I i 4 i V V. .I i g i i e i ( T.""T 6 "T"" T =. a. w r-e I -J 4. = s a h 1 V e a l ..a m ., _3 a _a.a t i _\\ _I i F h i .I I A i I. A 3 ,.p T" r .e r r r r -.h _ = .. = r. s j r-me use m m s.. m m s.. wn i i = 1 t y* .( v a ,3,, 3 E ,1 q l J {

  • g.?

e 7" 7" 'F" T" T T 'T"" "T"'

  • I v

r ~ ~~ 5 y .e 't ...=. ),, s 4 s 1 t i i i r. i"' i i.. " b. 7." t.' i. i t. ". " +.

t. t, T

i"' I ..a n ..s 2 i i T.T y ? E E 1 1 h sl I E i s, s .~ .1-. _c z 4 a. .\\ s g c c c L T~ E E

" =

S e1 r, L I 2 _m . _4 y s s s s s S. 1 g s 6 a ? s x r-r- r-I -l a ~ g a a z +; z .m ..Y s m.- ~.. -w-- ri E E e a a t 5 .s i .s a. s_ a. s. -p 5 3 E P b "F" "F"" "F" "F" "F"" "P" "T" "F" _s..a .x ..m . 1. s.i 2 i a a s T 7"~ ~"~ w e, s,n + .e-n- +- + + r-F- + + + n 5-e t 3-2 3" _s, 3 5, 5, 3 .J S 3 4 Ab ] I l$ i. E. '8. { 5 .A 3 5 .E l 1 P-e "w-j e e 't" "P"' 1r" "r"" ""F" "r' "p-" "r=' "F"" 's 3

2...l._

a l ,,C ..t r. I g, ,.r n. .I .I s .a S. 3 .A 3 1 5 g = l 49 f P-1P" "F"' T "#" T "7" i 1 e e "F" ?"" .iL ..f"r 3 3 3 a-3 3 4 3 l .m ..s .c ..=

a...a F

f '= --'. v [ so n s-- r-r- h C.u: 2 i i, 9 4 4 } ..C ed e 9 i - ' I

SECTION 4 POST-IRRADIATION TESTING 4-1. CAPSULE REMOVAL Specimen capsules will be removed from the reactor only during normal refueling periods. Because six of the capsules (namely S, T, U, W, Y, and Z) are located in areas where the lead factor is less than one, capsule reinsertions from these locations to areas exceeding a lead factor of one are in the recommended schedule for capsule removal. Reinsertion of the capsules with low lead factors insures that, at the time of removal, the capsules will have received a neutron fluence greater than the maximum fluence at the vessel wall. Therefore, there will be at least one capsule leading the vessel fluence throughout the life of the vessel. TABLE 4-1 SCHEDULE FOR REMOVAL OF SPECIMEN CAPSULES I Capsule Lead Factor 'I Removal Time Comments V 1.52 End of First Core Cycle Post-irradiation Test X 1.52 10 yrs Post Irradiation Test W 0.82 10 yrs Reinsert in Capsule V Location Y 0.82 10 yrs Reinsert in Capsule X Location W 20 yrs Post Irradiation Test Z 0.54 20 yrs Reinsert in Capsule W Location (originally Capsule V Location) Y 30 yrs Post irradiation Test U 0.82 30 yrs Reinsert in Capsule Y Location (originally Capsule X Location) T 0.54 Standby S 0 41 Standby .. weev.no e.etor by v.h en tow c ui. no.ac. i o. t,= i n nu.no. Each specimen capsule is removed after radiation exposure and transferred to a post-l irradiation test facility for disassembly of the capsule and testing of all specimens within that capsule. l l 4-1 1 i

4 2. CHARPY V NOTCH IMPACT TESTS The testing of the Charpy impact specimens from the intermediate shell course forging, the weld metal, and HAZ material in each capsule can be done singly at approximately five different temperatures.' The extra specimens can be used to run cuplicate tests at test temperatures of interest. Charpy impact tests are to be conducted in according with ASTM E 23 testing criteria. The initial Charpy specimen from the first capsu8:e removed should be tested at room tem-perature. The impact energy value for this temperature should be compared with the pre-irradiation test data; the testing temperatures for the remaining specimens should then be raised and lowered as needed. The test temperatures of specimens from capsules exposed j to longer irradiation periods should be determined by the test results from the previous capsu'e. 4 3. TENSILE TESTS The tensile specimens for each of the irradiated materials should be tested at test tempera-tures consistant with test temperatures of the WOL specimens, and in accordance with ASTM E 8 and E 21 testing procedures. gN 44. WEDGE OPENING LOADING Kid FRACTURE TOUGHNESS TESTS in light of current requirements of 10 CFR, part 50, ASME Code, Appendix G, the WOL specimens should be tested dynamically to adequately characterize the fracture toughness properties of the reactor vessel up to the initiation of the fracture toughness upper shelf. The WOL specimens for each of the irradiated materials should be tested in accordance with ASTM E399 74 with appropriate modifications necessary for dynamic tests. Testing dyna-3 mically in the fracture toughness ductile-to-brittle transition region and at upper shelf j initiation temperatures results in not only lower bound data but also provides an opportunity for obtaining valid N fracture toughness data upto the onset of upper shelf. This results from non linear cleavage behavior which occurs only in dynamic testing at these temperatures. The load-displacement curve exhibits an unambiguous drop in load at the onset of crack i initiation, thereby eliminating any possible doubt as to the start of crack initiation, as is the case in static loading conditions at these temperatures. Test temperatures which are recommended are characteristic of the upper fracture toughness shelf initiation temperature and lower.

1. P. c. Roccardelle and J. (.. swedlow. A Comaoned Analytrest Emeromontet Frecture Study of the Teo Lee 6ng Theones of ElastocPastre Fracture (J-Integret and Eauuetent Energy). HSST Program Tecnmcel Reoort No. 33 october 1973. WCAP4224 4

42

O 8556-11 ( l m m w ac &m ( STRAIN I Figure 3 7. Typical Tensile Test Stress-Strain Curve 3 13

4 3-3. DROPWEIGHT TESTS l The NDTT was determined for forging 04, the core region weld metal, and HAZ material by 4 dropweight tests (ASTM E 208) performed at The Rotterdam Dockyard Co. The following results were obtained: I Material NDTT (*F) Forging 04 48 + Weld Metal 66 HAZ 48 I i T l ] ) 6 3 14

9956-6 I I I l l l 100 n V U O 90 ULTIMATE TEMSlLE STRENGTH E 80 8 m t ^ V 0 g 0.2% YlELD STRENGTH Ug 60 50 j i 40 80 70 - 60 REDUCTION IN AREA E O O O 50 ac E O O O 40 >C y 30 8 m TOTAL ELONGATION g 0 U D-2 m m m V 10 v V UNIFORM ELONGATION I 0 0 100 200 300 400 500 600 700 TEMPERATURE (OF) Figure 3 5. Preirradiation Tensile Properties for the North Anna Unit No. 2 Reactor Pressure Vessel intermediate Shell Forging 04, t Heat No. 990496/292424 (Axial Orientation) 3-11

9956-1 100 l I I I I l 90 C ULTINATE TENSILE STRENGTH E M 80 y V 70 n a u m 0.2% YlELD STRENGTH O 60 m 50 ] 40 80 m REDUCTION IN AREA 70 O O g 60 O 5 M 50 r

40 e

d 30 h TOTAL ELONGATION -O O 20 V c e a 2 v V 10 UNIFORM ELONGATION i I I i i I o O 100 200 300 400 500 600 700 TEMPERATURE (OF) l l Figure 3-6. Preirradiation Ter.sile Properties for the North Anna Unit No. 2 Reactor Pressure Vessel Core Region Weld Metal 3-12

i 9956-5 i 140 130 120 O 110 O O 100 m o { 80 O O -5. 70 a O 5 60 b 50 w 30 O 20 O 10 I l l l l 0 -300 -200 -100 0 100 200 300 TEMPERATURE (OF) 'l I Figure 3-4. Preirradiation Charpy V Notch impact Energy Curve for the North Anna Unit No. 2 Reactor Pressure I Vessel Core Region Weld Heat Affected Zone Material i 3-9

TABLE 3-5 PREIRRADIATION TENSILE PROPERTIES FOR THE NORTH ANNA UNIT NO. 2 REACTOR PRESSURE VESSEL INTERMEDIATE SHELL FORGING 04 AND CORE REGION WELD METAL Ultimate Test 0.2% Yield Ter.sile Fracture Fracture Uniform Total Reduction Temp Strength Strength Load Stress Elongation Elongation in Area Vessel Material (* F) (psil (psi) (ib) (psi) (%) (%) (%) Forging 04 ROOM 84800 101550 1875 86650 13.1 20.2 56.4 (Axial Orientation) ROOM 85000 102450 1975 S6300 13.3 17.8 40.0 300 77700 95500 3875 144600 10.9 16.8 45.7 300 77600 94800 3300 142250 10.1 16.6 52.7 550 75600 97050 3925 169000 11.4 18.6 52.9 9 550 75900 97200 4020 145000 11.6 17.7 43.8 5 l Weld Metal ROOM 77750 86150 2690 185300 14.2 24.5 70.6 l ROOM 74400 85650 2875 181600 14.2 23.9 67.9 300 67450 76600 2750 155650 10.8 19.6 64.4 300 69500 77600 2475 160800 10.6 223 68.9 l 550 64000 80100 2680 164600 12.7 22 1 66.9 l 550 63700 79950 2725 132200 11.1 20.3 58.2 i t Q )

s 4 TABLE 3-4 x PREIRHADIATION CHARPY V NOTCH IMPACT DATA FOR THE NORTH ANNA UNIT NO. 2 REACTOR PRESSURE VESSEL CORE REGION WELD HEAT AFFECTED ZONE MATERIAL Test Temp (*F) Impact Energy (ft Ib) Sheer (%) Lateral Expansion (mils) 115 5 8 1 115 8 9 3 115 15 15 9 35 51' ~ 57 ,34 35 50 52 32 -35 52 52 32 32 27 30 26 32 62 72 47 32 90 81 59 68 95 100 63 68 94 100 67 68 104 100 69 140 76 100 59 140 92 100 62 140 113 100 70 "212 106 100 66 212 122 100 71 212 77 100 58 1 s A \\ l 1 3-7 ve

i t i c i 'E / .A i 9956-4 i / I' J. 130 s 120 . i r O 110 100 ,1 / 90 O 80 o .) 3 70 _C >. 60 E w5 50 40 30 0 20 l l l l l 0 -300 -200 -100 0' 100 200 300 TEMPERATURE' (CF) f f f Figure 3-3. Preirradiation Charpy V Notch impact Energy Curve for the North Anna Unit No. 2 Reactor Pressure Vessel Core Region Weld Metal 3-8 4 t

9956-3 100 90 80 O O 70 8 g 60 a d O 50 e5 O 5# O 30 O i 20 O 10 I I I l e, -200 -100 0 100 200 300 TEMPERATURE (DF) l l Figure 3-2. Preirradiation Charpy V Notch Impact Energy Curve for the North Anna Unit No. 2 Reactor Pressure Vessel intermediate Shell Forging 04, Heat No. 990496/ 292424 (Axial Orientation) 3-5 i

TABLE 3 3 PREIRRADIATION CHARPY V NOTCH IMPACT DATA FOR THE NORTH ANNA UNIT NO. 2 REACTOR PRESSURE VESSEL CORE REGION WELD METAL Test Temp (*F)

mpact Energy (ft Ib)

Shear (%) Lateral Expansion (mils) -115 11 13 7 115 6 9 1 115 7 12 3 -55 19.5 23 16 -55 18.5 20 12 -55 18 23 14 15 43.5 41 39 -15 30.5 40 27 15 43 42 35 15 76.5 72 59 15 69 70 55 15 77 73 63 68 83 77 64 68 91 85 73 68 88 85 74 140 109 100 87 140 116 100 89 140 108 100 82 l 3-6 l

TABLE 3 2 PREIRRADIATION CHARPY V NOTCil IMPACT DATA FOR THE NORTH ANNA UNIT NO. 2 REACTOR PRESSURE VESSEL INTERMEDIATE SHELL FORGING 04, HEAT NO. 990496/292424 (AXIAL ORIENTATION) i Test Temp (* F) Impact Energy (ft Ib) Shear (%) Lateral Expension(mils) -50 14 <5 6 -50 4 <5 0 50 4 <5 0 15 15 5 7 15 14.5 5 4 ) 15 22 10 14 1 75 53 55 38 75 36.5 40 31 75 31 40 23 120 53 65 ~49 120 44.5 55 46 120 50.5 55 43 j 210 80 100 64 210 72 100 63 ( 210 71 100 63 300 64.5 100 49 300 67 100 64 300 .76 100 63 3-3

9956-2 150 140 130 0 0 120 l10 O O O 100 g g ) 90 p a 80 g !;; 70 5 O' 5 60 50 40 ~ h 30 20 O 10 l l l 0 -300 -200 -100 0 100 200 300 TEMPERATURE (OF) Figure 31.Preirradiation Charpy V Notch impact Energy Curve For the North Anna Unit No. 2 Aesctor Pressure Vessel intermediate Shell Forging 04, Heat No. 990496/ 292424 (Tangential Orientation) 34

SECTION 3 PREIRRADIATION TESTING 3-1. CHARPY V-NOTCH IMPACT TESTS Charpy V-notch impact tests per ASTM E-23 were performed on the vessel intermediate shell course forging 04, heat no. 990496/292424, at various temperatures from 50* to 300*F to obtain a full Charpy V-notch transition curve in both the tangential and axial orientations (tables 3-1 and 3 2 and figures 3-1 and 3 2). Charpy V-notch impact tests were performed on weld metal and HAZ material at various temperatures from -115* to 212*F. The results are reported in tables 3-3 and 3 4, and figures 3-3 and 3 4, respectively. The Charpy impact specimens were tested on a Sontag S! 1 impact machine which is inspected and calibrated every 12 months using Charpy V-notch impact specimens of known energy values. These impact specimens are supplied by the Watertown Arsenal. 3-2. TENSILE TESTS Tensile tests per ASTM E-8 and E-21 were performed on the vessel intermediate shell course forging 04 (in axial orientation) and the weld metal at room temperature,300*F, and 550*F. The results are shown in table 3 5 and figures 3 5 and 36. Tensile tests for the intermediate shell coursa forging and weld metal were performed or an Instron TT-C tensile testing mr. chine using the standard Instron gripping devices. A full stress-strain curve was obtained for each specimen using a Baldwin Lima Hamilton Class B-1 extensometer and chart recorder, the latter calibrated to the extensometer. The method of measuring and co'ntrolling speeds for tensile tests on the Instron TT-C are governed by ASTM A370-68 (Mechanical Testing of Steel Products). The Instron TT-C tensile testing machine and the Baldwin-Lima Hamilton extensometer are calibrated by test equipment which has been certified by the National Bureau of Standards. A typical stress-strain curve is shown in figure 3 7. l l 3-1

r TABLE 3-1 PREIRRADIATION CHARPY V-NOTCH IMPACT DATA FOR THE NORTH ANNA UNIT NO. 2 REACTOR PRESSURE VESSEL INTERMEDIATE SHELL FORGifjG 04, HEAT NO. 990496/292424 (TANGENTIAL ORIENTATION) Test Temp ( F) Impact Energy (ft Ib) Shear (%) I ateral Expansion (mils) -50 33 14 24 -50 16 10 9 50 30 14 22 15 35 15 23 15 36 15 25 15 61.5 28 45 50 68 35 49 50 90 55 66 50 62 30 45 100 37 45 36 100 109 100 78 100 102 88 72 150 102 100 75 150 126 100 83 150 125 100 84 212 113 100 81 212 102 100 75 212 129 100 80 3-2

Analysis should be performed using the J-Integral or Equivalent Energy Concept.[1,21 Testing i 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 conservative upper shelf fracture toughness properties. Research in this area is currently in progress at Westing-house Research and Development, ASTM E24, NRC, and other places. Use of this technique will be further evaluated as to applicability for surveillance specimen testing. 45. POST IRRADIATION TEST EQUIPMENT i The following minimum equipment is required for the post-irradittion testing operations. Milling machine or special cutoff wheel for opening capsules, and dosimeter m blocks and spa:ers Hot cell tensile testing machine with pin type adapter for testing tensile specimens = Hot cell dynamic WOL testing machine with clevis and appropriate measuring a equipment associated with dynamic testing Hot cell Charpy impact testing machine a Sodium iodide scintillation detector and pulse height analyzer for gamma counting s of the specific activities of the dosimeters i 1

1. P. C. Riccernette end J. L. Swediow, A Cornbined Anetrocel Esperimental fracture Study of the Two Leading Theones of Electuc4*ientrc Frecture (JInterret and Equiemeent Energy), HssT Prosrom Technical Report No. 33, october 1973. WCAP4224
2. T. R. Moger and C. Buchelet, "Emporimental Verification of Lower Bound Kee Values Utilizing the Equivalent Energy Concept," in propress in Flow Growth and Frecture Toupeness Teepar, ASTM STP436, pp. 281296, Amencen Socwty for Testing and Motorials, Philadelphie,1973 l

l 4-3 I i l l

r ' APPENDIX A NORTH ANNA UNIT NO. 2 REACTOR PRESSURE VESSEL SURVEILLANCE MATERIAL The Rotterdam Dockyard Company supplied the Westinghouse Electric Corporation with sections of SA508 Class 2 forging used in the core region of the North Anna Unit No. 2 Reactor Pressure Vessel for the Reactor Vessel Radiation Surveillance Program. The sections of material were removed from the 10-inch intermediate shell course forging 04 of the pressure vessel heat treated as shown in table A-1. The Rotterdam Dockyard Company also supplied a weldment made from sections of forging 04 and adjoining lower shell course forcing 03 using weld wire representative of that used in the original fabrication. The forgings were produced by Rheinstahl Huttenwecke. The heat treatment history and quan-i titative chemical analysis of the pressure vessel surveillance material are presented in tables A 1 and A 2, respectively. TABLE A-1 HEAT TREATMENT HISTORY Temperature Time Material ( F) (hrs) Coolant Intermediate shell forging 1688 - 1697 2 1/2 Water quenched ( 04, Heat No. 990496/ 292424 1220 1229 6.0 Furnace moled to 842*F 1130 1 25 14 3/4 Furnace cooled Weld 1130 25 13 1/2 Furnace cooled I A1

TABLE A-2 QUANTITATIVE CHEMICAL ANALYSIS (WEIGHT. PERCENT) Forging 04, Heat No. 990496/292424 Weld Metal Rotterdam Dockyard Element Westinghouse Analysis [a] Analysis Westinghouse Analysisfal C 0.19 0.19 0.08 S 0.011 0.015 0.011 N2 0.011 0.011 Co 0.003 0.011 <0.002 Cu 0.11 0.09 0.088 Si 0.25 0.21 0.25 Mo 0.60 0.63 0.49 Ni 0.86 0.80 0.084 Mn 0.76 0.67 1.82 Cr 0.35 0.34 0.042 c. V 0.031 0.02 0.002 P 0.018 0.010 0.017 Sn 0.016 0.004 Al 0.023 0.017 0.015

a. All elements not listed we leu than 0.010 mightweent A2 l

e ...,e f C

  • I g

f -g ;~ u t Iw*

  1. 4 >1

'.t !,!n!l t =c t \\ L n., M N ijil r' s . 'M;-g aa}GN Ni! _ fi5i r e-g%p m 9 5,'. ll Ph:rf e e 4-p t/,. N.I\\

  • % i:!
  • E w

\\*\\ / E e_ s N. e - ~ e N u-. 3 I lQ[ EF T, iW-% i, E ~ e -T N l ! I!!I t--- Lga i N tN ':ii g;N e M,:-h

J

' ) L* I N 8 l t' 7

  • ~~~~ N l

%l[ Ni l y h s '!N 4 -:- $2t _.__ ' L ur,.J I!r (ju' il ik[ I I I ". o. 8 l i l! e l m., il> m l l!lglIlhllj]n lj t. unl, i e w'f'g l l l1billii Hiiiiiigjjjjj;tj I l lIl!! 'I f i LUli 'jllI, i n !, ! b' ',Illli/HiflIIIl [, i i ui f f 4

I l The relationship of the test material to the type and number of specimens in each capsule ) is shown in table 2-1. l Dosameters of pure copper, iron, and nickel and cadmium-shielded wires are secured in holes drilled in spacers located in the capsule positions shown in figure 2-4. Each capsule also contains a dosimeter block (figure 2-5) which is located at the center of the capsule. Two cadmium oxide-shielded capsules, each containing isotopes of either U or Np237 (both 238 99.9 percent pure) are located in the dosimeta block. The double containment afforded by 238 237 the dosimeter assembly prevents loss and contamination by the U and Np and their i activation products. The amounts of each are presented in table 2-2. Both of them are held in a 3/8-inch long by 1/4-inch-OD sealed brass tube and stainless steel tube, respectively. Each tube is placed in a 1/2-inch-diameter hole in the dosimeter block (one U238 and one Np tube per block), and the space around the tube filled with cadmium oxide. After f 237 placement of this material each hole is blocked with two 1/16-inch-thick aluminum spacer l discs and an outer 1/8-inch-thick steel cover disc welded in place. i i I The numbering system for the capsule specimens and their locations are shown in figure 24. TABLE 2-1 TYPE AND NUMBER OF SPECIMENS IN THE NORTH ANNA f UNIT NO.1 SURVEILLANCE TEST CAPSULES Capsule S,V,W, and Z Capsule T,U,X, and Y Motorial Charpy Tensile WOL Cherpy Tensiis WOL [ l Forging 03 (Tangential) 8 8 Forging 03 (Axial) 12 2 4 12 2 l Weld Metal 12 2 12 2 4 i l HAZ 12 12 l TABLE 2-2 QUANTITY OF ISOTOPES CONTAINED IN THE DOSIMETER BLOCKS I k i Weight Weight Isotope (mg) Compound (mg) 237 Np 17

  • 1 NpO2 2011 U238 12 U038 14.25 27 l

a D D O Q I 2 4 I I '0 f N E R R S A h 0 L 1 T A S A S RI C E f I L F T N L S A I I E S C N A E A f T T R P S SS B E L U E S L P U A S C) P) G A G D L CL E L 5 D5 A 7 E 7 3 3 E L E S A L 0 E0 E T 7 S D I 3 a x I 2 8 I X HD 3 D O UO 2 O ly H N R IN0 U0 U b K R E U5 I 5 l m C E C T 2 2 H N A D e O V A P s L 0 P E 0 R 0 A s B C S N( U( C A H k E c 1 i 2 3 4 5 6 o 1 l B r i e te m is 3 o 6 D 2 5 -2 e ru 7 iF g P' g Y I T g A 7 A-M i 60 0 w s' g y 8 yV e N0 3 llllll l 1t

.~ ii I.l i.l l,l I.l l.i I.i li a a o m a a i e i I.l i s t .s u E l.l I.l ii i.l l ~ ~ m I.I n* ii I.l 11 .l l 4.l i.i i 1 1, 1 i.l 1 4 F-o p + + t F 4 t t i l. i l 3 i. i. 4 p b.p H.g b.p F.t .p F1 I. q i. !.] f.1 i..l ] i l H.l - p .i

i. i.

p p p p. H.] F t - p t pp b.p F.F F.p H=F p.p !i 1.l a., ~= g .a~ I.1 l.l I..i . p. p p. H.lp b T H.p p.b i. i I.l l

s. p.

F p p + t I.p p.p p p. F.p F.P [ !i I..i. H..i I.1H".p+ t . H I.I. .nn., i. i. l l..; I.l p b o p p F p i. I. i. t .~._iiu..p 9, b.p. 1 P' = p.I p.p p[ 'l l i.l p p p. p - p .1 I.l l !.i. .j l i. . i. ~ l p +H ~ i. i. i. i. t i. t I p. pp p.p pp b.p p.P' p I. l.; H..H. i. ~. H. 4 i.j l. i q

i. i.

~ p + i. + + u i. i. i. p p + l pp r.p p.p pp H.F p.p F.p p.F m.r i. 1 .l p p.p F.p . + p p t 9 H..l 1 l.l l. H.. 3.l p t p + p p p u j p.p p.p p.p p.F .p ..l ].q i..; i..j l l l.l l + p + p p p p p p p + p. i. t + i. i.p H.p pp H.p p.p p.p F.l p pj l1 H..] .j l H.. i.l ~ p p p. + p p + l. p jp p.p F.p pF p.p p.p p.p p.H i i. w. i. [. i. .l l.l l.l l p p p p p p p-p p. i..j ..l l + t p.p p.p p p. + i. l i. i. p.p p.p p.p H.b z pi l.l l..] j; l.l l.) i. H..l - ~ i. 3.j ~. e. p i. i. o p + p p p + t +H j i. H.- !p p.i p.p pp p.p p.p p.p J

i.

WESTINGHOUSE CLASS 3 R S s \\ VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA UNIT NO.1 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM ( J. A. Davidson J. H. Phillips September 1976 ? g!A APPROVED: J N[.{ gg \\ rigos, ahager Suuctural Materials Engineering I w Work Performed Under Shop Order No. VRA 106 i WESTINGHOUSE ELECTRIC CORPORATION Nuclear Energy Systems P. O. Box 355 Pittsburgh, Pennsylvania 15230

ABSTRACT A pressure vessel steel surveillance prcgram was developed for the Virginia Electric and Power Company, North Anna Unit No.1, to obtain information on the effects of radiation on the reactor vessel material under operating conditions. The program comprises the evaluation of the radiation effects based on comparison with preirradiation testing of a selected group of speci-mens to determine toughness properties of the reactor pressure sessel. Continuous monitoring of these specimens within the reactor pressure vessel provides data on the integrity of the vessel in terms of adequate toughness properties. A description of the surveillance capsules 7 and preirradiated test results is also included. 3

o SECTION 4 POST-IRRADIATION TESTING 41. CAPSULE REMOVAL Specimen capsules will be removed from the reactor only during normal refueling periods. Because six of the capsules (namely S, T, U, W, Y, and Z) are located in areas where the lead factor is less than one, capsule reinsertions from these locations to areas exceeding a lead factor of one are in the recommended schedule for capsule removal. Reinsertion ef the capsules with low lead factors insures that, at the time of removal, the capsules will have received a neutron fluence greater than the maximum fluence at the vessel wall. Therefore, there will be at least one capsule leading the vessel fluence throughout the life of the vessel. TABLE 4-1 SCHEDULE FOR REMOVAL OF SPECIMEN CAPSULES Capsule Lead Factor [a] Removal Time Comments V 1.52 End of First Core Cycle Post-Irradiation Test X 1.52 10 yrs Post-irradiation Test W 0.82 10 yrs Reinsert in Capsule V Location Y 0.82 10 yrs Reinsert in Capsule X Location ( W 20 yrs Post-Irradiation Test Z 0.54 20 yrs Reinsert in Capsule W Location (Originally Capsule V Location) Y 30 yrs Post Irradiation Test U 0.82 30 yrs Reinsert in Capsule Y Location (Originally Capsule X Location) T 0.54 Standby S 0.41 Standby

a. Multiplying factor by evhich the capsule fluence leeds the weasel vuell fluence l

l 4-1

J i g. Each specimen capsule is removed after radiation exposure and transferred to a post-irradiation test facility for disassembly of the capsule and testing of all specimens within that capsule. 4-2. CHARPY V-NOTCH IMPACT TESTS The testing of the Charpy impact specimens from the lower shell course forging, the weld metal, and HAZ material in each capsule can be done singly at approximately five different temperatures. The extra specimens can be used to run duplicate tests at test temperatures of interest. Charpy impact tests are to be conducted in accordance with ASTM E 23 test criteria. The initial Charpy specimen from the first capsule removed should be tested at room tem-perature. The impact energy value for this temperature should be comparec with the preirradiation test data; the testing temperatures for the remaining specimens should then be raised and lowered as needed. The test temperatures of specimens frr m capsules exposed to longer irradiation periods should be determined by the test results from the previous capsule. 4-3. TENSILE TESTS The tensile specimens for each of the irradiated materials should be tested at test tempera-tures consistant with test temperatures of the WOL specimens, and in accordance with ASTM E-8 and E 21 testing procedures. 4-4. WEDGE OPENING LOADING Kid FRACTURE TOUGHNESS TESTS in light of current requirements of 10 CRF, Part 50, ASME Code, appendix G, the WOL specimens should be tested dynhmically to adequately characterize the fracture toughness properties of the reactor vessel up to the initiation of the fracture toughness upper shelf. The WOL specimens for each of the irradiated materials should be tested in accordance with / ASTM E399-74 with appropriate modifications necessary for dynamic tests. Testing dynam-ically in the fracture toughness ductile to-brittle transition region and at upper shelf initiation temperatures results in not only lower bound data but also provides an opportunity for IN racture toughness data up to the onset of upper shelf. This results from obtaining valid f

1. 9. C. Riccaroone end J. L. Swaneew. A Cornbinear Anotyticer Emerunentet Fracture Study of the Two Leeding Theonse of OserveN Fracture 4-intepret and Equivamit Eneryyf, HsST Program Technical Report No. 33.

october.1973. WCAP4224 4-2

8556-11 e r m t m W ae %e 5 5 STRAIN l l \\ ~ \\ 1 Figure 3 7. Typical Tensile Test Stress Strain Curve i 3-13

s' 33. DROPWEIGHT TE3TS The NOTT was determined for forging 03, the core region weld metal, and HAZ material by dropweight tests (ASTM E 208) performed at the Rotterdam Dockyard Co. The following results were obtained: Material NDTT (*F) Forging 03 -12 Weld Metal 12 HAZ 22 j 5 3-14 e n

APPENDIX A NORTH ANNA UNIT NO.1 REACTOR PRESSURE VESSEL i SURVEILLANCE MATERIAL The Rotterdam Dockyard Company supplied the Westinghouse Electric Corporation with sections of SA508 Class 2 forging used in the core region of the North Anna Unit No.1 Reactor Pressure Vessel for the Reactor Vessel Radiation Surveillance Program. The sections of material were removed from a 10 inch lower shell course forging 03 of the pressure vessel heat treated as shown in table A-1. The Rotterdam Dockyard Company also supplied a weldment made from sections of forging 03 and adjoining intermediate shell course forging 04 using weld wire representative of that used in the original fabrication. The forgings were produced by Rheinstahl Huttenwecke. The heat treatment history and quantitative chemical analysis of the pressure vessel surveillance material are presented in tables A-1 and A 2, respectively. TABLE A 1 HEAT TREATMENT HISTORY Temperature Material (* F) Time (hrs) Coolant Lower Shell Forging' 03, 1616 - 1725 2 1/2 Water quenched Heat No. 990400/292332 i 1202 1292 7 1/2 Furnace cooled to 842*F 1130 25 14 3/4 Furnace cooled Weld 1130 25 10 3/4 Furnace cooled ~ A-1

.e TABLE A-2 QUANTITATIVE CHEMICAL ANALYSIS (WEIGHT-PERCENT) Forging 03 Hest Rotterdam Dock-No. 990400/292332 yerd Check Weld Metal Element Westinghouse Analysis *l Analysis Westinghouse Analysis [a] I C 0.20 0.19 0.06 S 0.011 0.014 0.012 N 0.015 0.015 2 Co 0.020, 0.006 Cu 0.16 0.15 0.086 Si 0.26 0.22 0.35 Mo 0.61 0.63 0.49 Ni 0.79 0.80 0.11 Mn 0.68 0.68 1.29 Cr 0.30 0.30 0.025 V 0.037 0.02 0.001 P 0.019 0.010 0.020 Sn 0.017 0.003 Al 0.021 0.009 .. Asi.ienwnes not sist.o n. i.a tren o.oto .gniwe.nt ) i l A-2 l l l

non linear cleavage behavior which occurs only in dynamic testing at these temperatures. The load-displacement curve exhibits an unambiguous drop in load at the onset of crack initiation, thereby eliminating any possible doubt as to the start of crack initiation, as is the case in static loading conditions at these temperatures. Test temperatures which are recommended are characteristic of the upper fracture toughness shelf initiation temperature i and lower. Analysis should be performed using the J Integral or Equivalent Energy Concepth,21. Testing at temperatures characteristic of the fracture toughness upper shelf is not suggested due to l the uncertainty of the point of crack initiation even when dynamic testing is performed. i At these temperatures, static Jge testing appears to be most indicative of conservative upper shelf fracture toughness properties. Research in this area is currently in progress at Westing-l house Research and Development Laboratory, ASTM E24, NRC, and other places. Use of this technique will be further evaluated as to applicability for surveillance specimen testing. P 4-5. POSTIRRADIATION TEST EQUIPMENT I l The following minimum equipment is required for the post-irradiation testing operations. i i Milling machine or special cutoff wheel for opening capsules, and dosimeter e blocks and spacers Hot cell tensile testing machine with pin type adapter for testing tensile a specimens Hot cell dynamic WOL testing machine with clevis and appropriate measuring a equipment associated with dynamic testing t Hot cell Charpy impact testing machine e ( Sodium iodide scintillation detector and pulse height analyzer for gamma a counting of the specific activities of the dosimeters =

1. P. C. Rbccordefle end J. L. swediow, A Combined Anotyticet-Espenonental Frecarre Study of the Two Leading Theonen of EAest,cAertic Freenere (2-intepret and Eeureeient Energy), HsST Program Technical Report No. 33, October,1973, WCAP4224
2. T. R. Meger and C. Suchelet, " Experimental Verification of Lower Sound Kic Values utilizing the Equivoient Energy Concept." In Propreer in FAsw Growth and Freenere Tauenneer Teeting, ASTM STP436, pp. 281-296 l

Amerien Society for Test 6ng and Materials. Phinedelphie,1973 i 4-3

9928-5 100 i l l I l 90 80 ULTIMATE TENSILE STRENGTH a -, 70 N k 60 O 0.2f YlELD STRENGTH 50 40 80 REDET10N IN AREA f 70 rn U 60 h O 0 50-5 a. [ 40 TOTAL ELONGATION z n n E 20 g p v ~ UNIFORM ELONGATION I I O O 100 200 300 400 500 6 00 TEMPERATURE (OF) Figure 3-5. Preirradiation Tensile Properties for the Nor.h Anna Unit No.1 Reactor Pressure Vessel Lower Shell Forging 03, Heat No. 990400/292332 (Axial Orientation) 3-11

9928-6 80 I I I I I ULTIMATE TENSILE STRENGTH r 80 Tg D g O O ~ [0.2%YlELDSTRENGTH = W O M 60 O b ua 50 40 / 80 70 C 2 _ 60 EDUCTION IN AREA $d 50 5 a. 40 ~ UNIFORM ELONGATION d 30 TOTAL ELONGATION Y M n [ c. 20 y n v O O 10 n w v v 0 0 100 200 300 400 500 600 TEMPERATURE (CF) Figure 3-6. Preirradiation Tensile Properties for the North Anna Unit No.1 Reactor Pressure Vessel Core Region Weld Metal 3 12

9928-4 t 220 200 180 ' O 160 ' - O 140 E 9 120 a -5. O >- 100 Ew 80 O 60 O O O 40 O O 20 -O o OI I I I I I I I l o -150 -100 -50 0 50 100 150 200 250 3 00 350 TEMPERATURE (DF) Figure 3-4. Pre:rradiation Charpy V-Notch impact Energy Curve for the North Anna Unit No.1 Reactor Pressure Vessel Core Region Weld Heat Affected Zone Material 3-9

TABLE 3-5 PREIRRADIATION TENSILE PROPERTIES FOR THE NORTH ANNA UNIT NO.1 REACTOR PRESSURE VESSEL LOWER SHELL l FORGING 03 AND CORE REGION W8iLD METAL 02% Ultimate Uniform Reduction Test Yield Tensile Total Elongation in Area Material Temp (*F) Strength (psi) Strength (psi) Elong. (%) (%) (%) Lower Shell Room 70,050 92.404 18.8 11.4 60.7 Forging 03 Room 71,300 92,700 18.8 12.3 57.0 300 64,075 84,600 21.5 13.0 61.0 300 64,750 85,200 23.7 13.7 64.0 550 58,027 87,150 20.6 13.7 52.0 4 550 53,137 85,325 26.0 17.2 57.0 95 Weidment Room 63,150 78,300 18.9 9.6 71.0 Room 65,200 80,400 19.5 9.8 71.0 300 64,300 77,875 19.5 8.6 67.0 300 59,675 73,625 22.5 10.8 68.9 550 60,175 76,850 20.0 9.8 63.0 550 61,650 80,500 18.0 8.9 57.0 I G .J

v \\ s g ( l' ut A f[ TABLE 3-4 PREIRRADIATION CHARPY V NOTCH IMPACT DATA FOR THE NORTH ANNA UNIT NO.1 REACTOR PRESSURE VESSEL CORE REGION WELD HEAT AFFECTED ZONE MATERIAL Test Temp (*F) Impact Energy (ft Ib) Sheer (%) Lateral Expension (mils) 125 21 9-13 4.5 5 2.5 13 5 6 -80 25.5 3 13 -80 76.5 29 47 -80 53 18 21 1 -25 22.5 18 13 -25 47 23 26 25 31.5 13 15 40 104 81 65 40 62 45 43.5 40 78.5 42 46 100 156.5 100 78 100 127.5 79 81 100 125.5 90 73 170 152.5 100 72 170 166 100 82 170 109 100 74 i, > s s 37 i s.

9928-3 180 160 140 ) 120 E O* " 100 O C C [ 80 t 2 O u 60 o 40 Q j 20 I I I I I I I I I 0 -150 -100 -50 0 50 100 150 200 250 300 350 t TEMPERATURE (OF) i Figure 3 3. Preirradiation Charpy V-Notch impact Energy Curve for the North Anna Unit No.1 Reactor Pressure Vessel Core Region Weld Metal I 3-8 i

9928-2 200 180 160 140 t 7 120 S b 100 a 80 O 60 d 2 40 20 '~ l I I I I I I I o -150 -100 -50 0 50 I00 150 200 250 300 350 TEMPERATURE (OF) Figure 3-2. Preirradiation Charpy V-Notch Impact Energy Curve for the North Anna Unit No.1 Reactor Pressure Vessel Lower Shell Forging 03, Heat No. 990400/292332 (Axial Orientation) i j 3-5 ., + -

TABLE 3-3 PREIRRADIATION CHARPY V-NOTCH IMPACT DATA FOR THE NORTH ANNA UNIT NO.1 REACTOR PRESSURE VESSEL CORE REGION WELD METAL Test Temp (*F) Impact Energy (ft Ib) Sheer (%) Laterol Expension (mils) 4 110 9 2 ~ 0 9 110 5 0 9 110 13 0 7 0 74.5 48 61 0 58 54 52 0 60 43 70 48 37.5 31 39.5 48 40 51 45 48 42 40 49.5 75 83 79 68 75 81 81 71.5 75 49.5 65 50 175 106 100 89 175 106 100 88 175 94 96 76 250 87 100 79 250 90.5 100 81 250 84.5 100 77 34

i L TABLE 3-2 PREIRRADIATION CHARPY V-NOTCH IMPACT DATA FOR THE NORTH ANNA UNIT NO.1 REACTOR PRESSURE VESSEL LOWER SHELL FORGING 03, HEAT NO. 900400/292332 l (AXIAL ORIENTATION) i 4 Tut Temp (*F) impact Energy (ft Ib) Sheer (%). Laterol Expension (mils) 15 16 10 7 l 15 17 10 10 15 15 10 10 45 31.5 35 21 [ 45 30 35 20 I 45 25 30 21 i 75 47.5 30 37 l 75 43.5 23 37 i 75 47.5 27 33 105 62.5 55 50 105 56.5 55 50 105 59 55 50 150 80.5 90 66 150 84.5 100 72 150 67.5 78 62 210 85 100 69 210 82.5 100 68 210, 86.5 100 73 i i 8 r 3-3

.? 9928-1 4 200 180 160 140 3 ) E 120 Q a c ~ 100 O G h 80 O = 0 60 40 O 20 0 l l 1 I I I I l l a -150 -100 -50 0 50 100 150 200 250 300 350 TEMPERATURE (OF) 1 l Figure 31. Preirradiation Charpy V-Notch Impact Energy Curve for the North Anna Unit No.1 Reactor Fressure Vessel Lower Shell Forging 03, Heat No. 990400/292332 (Tangential Orientation) 34

I SECTION 3 PREIRRADIATION TESTING 31. CHARPY V NOTCH IMPACT TESTS Charpy V-notch impact tests per ASTM E-23 were performed on the vessel lower shell course forging 03, heat no. 990400/292332, at various temperatures from.15 to 212*F to obtain a full Charpy V-notch transition curve in both the tangential and axia'l orientations (tables 3-1 and 3-2 and figures 3-1 and 3 2). Charpy V-notch impact tests were performed on weld metal and HAZ material at various temperatures from -150 to f 250*F. The results are reported in tables 3-3 and 3-4, and figures 3-3 and 3 4 respectively. The Charpy impact specimens were tested on a Sontag Sl 1 impact machine which is inspected and calibrated every 12 months using Charpy V-notch impact specimens of known energy values. These impact specimens are supplied by the Watertown Arsenal. 3 2. TENSILE TESTS Tensile tests per ASTM E-8 and E-21 were performed on the vessel lower shell course forging 03 (in axial orientation) and the weld metal at room temperature,300 F, and 550*F. The results are shown in table 3 5 and figures 3-5 and 34. Tensile tests for the lower shell course forging and weld metal were performed on an instron TT-C tensile testing machine using the standard instron gripping devices. A full stress-strain curve was obtained for each specimen using a Baldwin Lima-Hamilton Class B 1 extonsometer and chart recorder, the latter calibrated to the extensometer. The method of measuring and controlling speeds for tensile tests on the Instron TTC are governed by ASTM A370-68 (Mechanical Testing of Steel Products). The instron TT C tensile testing machine and the Baldwin Lima-Hamilton extensometer are calibrated by test equipment which has been certified by the National Bureau of Standards. A typical stress-strain curve is shown in figure 3 7. 31

TABLE 31 PREIRRADIATION CHARPY V-NOTCH IMPACT DATA FOR THE NORTH ANNA UNIT NO.1 REACTOR PRESSURE VESSEL LOWER SHELL FORGING 03, HEAT NO. 990400/292332 (TANGENTIAL ORIENTATION) Test Temp (*F) Impact Energy (ft Ib) Sheer (%) Lateral Expensson (mils) 15 30 9 21 15 9 0 5 15 19 3 16 40 66 40 48 40 84 45 60 40 78 56 59 s ) 74 96.5 38 67 74 101 64 74 74 73.5 38 59 125 116 77 74 125 115 92 78 125 99 79 67.5 170 143 100 80 170 147 100 83 170 144 100 81 210 126.5 100 88 210 123 100 84.5 210 127 100 80 32

x le0f ES: I.. M ALL OVER

2. NOICH DEPIH IO BE EXIIIIDED BY 0.09-0.156 8Y FAllGUE CRACKING
3. SLOI 10 BE PARALLEL 10 SURI ACE *A" W11HIN 0.00510f at

-C-l

4. SLOI 10 BE PERPENDICULAR 10 SURF ACES *8" 8 *C*

1.45 WiiHIN 0.005 1.43

5. DIM.S MAREED IMUS
  • MUST NOT HAVE A T. R.0. EXCIEDING m

1.130 10.001 ON OPPOSITE FACES m 30 0.755* l.0% +! O. 71 5 " 0.995 7 - SEE NOIE 2 0.1 h A 0.501 0.439 0.279 0 283 h 4 1.005 ~ 0'995 a 0.439 n

L h

^ E -]..-[- _ ~ '~.I b" O 435 h g dim "X" - [ I* $ ~ O lM "Y" P \\ \\ 0.002, l - A-l 4 ( 0. l l 2 )-40UNC-28 0.001 0.501 .500-20 THD. CLASS "38" 0.127 0,qgg DEEP (4) 0.375 DEEP 0.123 0.0463 0.0231 0.0473 0.0236 SPECIMEN IDEMilFICATION (2) LOCAil0NS Figure 2-3. Wedge Opening Loading Specimen ,= 4

9 8 O a = u a o w E a N 3, g w O m sa. + E w w w -a 4 Z and O E p. W O Q O f3 $M u - m a 'o D. see as o d g hg $h p O .,? 4 ** e m M w 0 0 I--b 3 W. Q w - of a !i s >= w .J L.) o u I M-44J 4 "o and m 2 "E E .. m m E O. r w d E." 2 E2 4 mm [ g m

    • a c

m m @ e* 4 6 = ~ 4 ? T T

v. ~+

.E oo o d d I RE o

m. e.

oo .e JL JL N T C o R / l-e / d o g i. s, a, v a o 3 s I :/\\ ~ if e Jk JL 5 '" Q b E, 8 W N ] N s.s. m u o ~ d = = ~. = o = S S* T ET o o ~ ~ q q2 2 sg o o y E" l /' 4 'G t 2 3 ,m u s M$$ 5 E5 O BB f\\ w' U v i wm c l l 2-3

E.

  • )

25. Wedge Opening Loading Specimens (Figure 2 3) Wedge opening loading (WOL) test specimen 2 were machined along the axial orientation so that the specimen would be loaded perpendicular to the major working direction of the forging and the simulated crack would propagete along the hoop (tangential) direction. The weld specimens were machined such that the crack would propogate in the weld direction. All specimens were fatigue procracked according to ASTM E399-707. 2-6. MONITORS 27. Dosimeters Eight capsules of the type shown in figure 24 contain dosimeters of copper, iron, nickel, and aluminum-0.15% cobalt (cadmium-shielded and unshielded) wire, neptunium-237, and uranium-238. The dosimeters are used to measure the integrated flux at specific neutron energy levels. 2-8. Thermal Monitors The capsules contain two low-melting-point eutectic alloys so that the maximum temperature attained by the test specimens during irradiation can be accurately determined. The thermal monitors are sealed in Pyrex tubes and then inserted in spacers (figure 24). The two eutectic alloys and their melting points are as follows: 2.5% Ag, 97.5% Pb Melting point 579'F 4 1.75% Ag, 0.75% Sn, 97.5% Pb Melting point 590*F 29. SURVEILLANCE CAPSULES 2 10. Capsule Properation ) .i .) The specimens were seal-welded into square austenitic stainless steel capsules to prevent cor-rosion of specimen surfaces during irradiation. The capsules were hydrostatically tested in domineralized water to collapse the capsule on the specimens so titat optimum thermal con-ductivity between the specimens and the reactor coolant could be obtained. The capsules were helium leak tested as a final inspection procedure. Finally, the espsules were coded S, T, U, V, W, X, Y, and Z. Fabrication details and testing procedures are listed in figure 24. 2 11. Capsule Loading Upon receipt, the eight test capsules are positioned in the reactor between the thermal shield and the vessel well at the locations shown in figure 24. Each capsule contains 44 Charpy V-notch specimens,- four tensile specimens, and four WOL specimens. i j 24 . ~ ~ ~, -,,.. - - -

SECTION 2 SAMPLE PREPARATION 21. PRESSURE VESSEL MATERIAL Reactor vessel material was supplied by the Rotterdam Dockyard Company from lower shell forging 03, heat no. 990400/292332. A submerged arc weldment which joined sections of material from this forging and adjoining intermediate shell course forging was also supplied by the Rotterdam Dockyard Company. Data on the pressure vessel material are presented in appendix A. 2-2. MACHINING l Test material was obtained from the lower shell course forging after the thermal heat treat-ment was complete and the forging formed. All test specimens were machined from the 1/4-thickness section of the forging after a simulated postweld stress-relieving treatment on the test material was perforrned. The test specimens represent material taken at least one forging thickness (as-quenched) from the quenched ends of the forging. Specimens were machined from weld and heat-affected zone (HAZ) material of a stress-relieved weldment which joined sections of the lower and intermediate shell courses. All HAZ specimens were obtained from the weld HAZ of forging 03, heat no. 990400/292332. 2 3. Charpy V Notch impact Specimens (Figure 2-1) Charpy V-notch impact specimens from forging 03 were machined in both the tangential orientation (longitudinal axis of specimen parallel to major working direction) and axial or!antation (longitudinal axis of specimen perpendicular to major working direction). The core region weld Charpy impact specimens were machined from the weldment such that the long dimension of the Charpy was normal to the weld direction; the notch was machined such that the direction of crack propagation in the specimen was in the weld direction. 24. Tensile Specimens (Figure 2 2) P Tensile specimens were machined with the longitudinal axis of the specimen perpendicular to the major working direction of the forging. Tensile specimens were also removed from l the core region weldment. 2-1

.se 0 [l fl C EP S 3 9 t 3 3 s n G-'O w g [ H C00' I O 1 5 S 0 0 S 09 E 98 53 L i 9 9 s ,33i u 00 R E n 48 0 O 9 V e 08 0 m i L 0 0 c L e ) p S t c ap "6 "4 m \\ 4 4 i -~ / hc toN -V 3 3 y 6 5 0 0, pra 1 g hC 1 5 5 2 0 2 (1 f 6 e 4 i. r \\ 2 2 u g iF G a u 6 4 1 1 3 3 0 0 y

Post-irradiation testing of the Charpy impact specimens provides a guide for determining pressure temperature limits on the plant. A temperature shift in the reference temperature will occur in the irradiated Charpy impact specimen test data as a result of radiation expo-sure at plant temperatures. These data can then be reviewed to verify or establish new pressure temperature limits of the vessel duririg startup and cooldown. This allows a check of the predicted shift in the reference temperature. The post-irradiation test results on the WOL specimens provide actual fracture toughness properties for the vessel material. These properties may be used for subsequent evaluation at per the methods outlined in the ASME Code, appendix G. Eight material test capsules are located in the reactor between the thermal shield and the vessel wall and are positioned opposite the center of the core. The test capsules are located in guide tubes attached to the thermal shield. The capsules contain Charpy impact, WOL, and tensile specimens from the limiting core region forging. This forging is the reactor vessel lower shell forging adjacent to the core region. Charpy impact, WOL, and tensile specimens obtained from the representative core region weld metal, and Charpy impact specimens from the weld material heat-affected zone (HAZ), are also located in the capsules, in addition, dosimeters to measure the integrated neutron flux and thermal monitors to measure tem-perature are located in each of the eight material test capsules. The thermal history or heat treatment given to these specimens is similar to the thermal history of the reactor vessel material, except that the postweld heat treatment received by the specimens has been simulated (appendix A). l i f i i 1-3 f g-

1 SECTION 1 l l PURPOSE AND SCOPE The purpose of the Virginia Electric and Power Company, North Anna Unit No.1, surveil-lance program is to obtain information on the effects of radiation on the reactor vessel materials of a reactor during normal operating conditions. Surveillance material is selected as the most limiting material based on surveillance selection procedures which are outlined in ASTM E185-73, Annex A1. Evaluation of the radiation effects is based on the preirra-diation testing of Charpy V-notch, tensile, and dropweight specimens, and post irradiation testing of Cherpy V-notch, tensile, and wedge-opening-loading (WOL) specimens. Current reactor pressure vessel material test requirements and acceptance standards use the reference nil ductility temperature, RTNDT, as a basis. RT i NDT s determined from the dropweight nil-ductility transition temperature, NDTT, or the 50 ft Ib Charpy V-notch impact temperature trom specimens oriented normal to the major working direction (whichever value is greater) as defined by the following equation: RTNDT = NDTT, if NDTT > T50(35) 60*F or RTNDT = T50(35) - 60 F, if T50(35) 60 F > NDTT where RTNDT = Reference nil ductility temperature NDTT = Nilductility transition temperature as per ASTM E208 50(35) = 50 ft Ib temperature from Charpy V-notch t T impact specimens (or the 35 mil lateral expansion temperature if it is greater N) 1. in th. wher.. i it as n.us ist. ei.mo n on is not cet in.d et the so et in t.mp reture, the torno.r tur. . which 35 mais ist.r 1.mo.n. son occurs is u d. 1 -1

L An empirical relationship between RTNDT and fracture t:ughness for reactor vessei steels has been developed and is presented in appendix G, section lil, of the ASME Boiler and Pressure Vessel Code (Protection Against Non-Ductile Failure) This relationship can be l employed to set allowable pressure-temperature relationships, based on fracture mechanics concepts, for the normal operation of reactors. Appendix G of the ASME Boiler and Pres-l sure Vessel Code defines an acceptable method for esiculating these limitations. it is known that radiation can shift the Charpy impact energy curve to higher tentpera-tures.[1,2] Thus, the 50 ft Ib temperature, and correspondingly, the' RTNDT, increase with radiation exposure. The extent of the shift in the impact energy curve - that is, the radia-tion embrittlement - is enhanced by certain chemical elements, such as copper, present in reactor vessel steels.[3,41 The 50 ft Ib temperature, and correspondingly the RTNDT, increase with semice and can be monitored by a surveillance program which consists of periodically checking irradiated reactor vessel surveillance specimens. The surveillance program is based on ASTM E185 73 (Standard ~ ' Recommended Practice for Su.veillance Tests for Nuclear Reactor Vessels). WOL fracture mechanics specimens will be used in addition to the Charpy impact specimens to evaluate the effects of radiation on the fracture toughness of the reactor vessel materials.[5,6,7,8,9,1_0,11) 1. L. F. Porter, " Radiation Effects in Steef," in A4pearrat in Murdeer Ame/icsoont, ASTM STP.27e, s,p.147196, Amoncen Socesty for Testing and Motonees, Phitedsephia,19eo 2. L. E. Steele and J. R. Howthorne, "Neuw information on Neutron Emerittlement and Ernerittlement Relief of Reector Pressure Vessel Steses," NRL41eo, August 19e4 3.

u. Potopows and J. R. Howthorne "The Effect of Reseduel Element on 560*F trrodiatson Respones of Selected s

Pressure vessel Stee6s and Weedmonts," NRL4e03, Septemeer 19ee 4 L E. Stee6e, " Structure and Composetion Effects on irradiation Seneetwity of Pressure Vsesse Steess," in IrreaWetion ENecte on Structurel AIIeye 96r Nucceer Meneter Appliceoons, ASTM STPsed, pp. le4-175 Amencen Society for Testing and Metenses, Philadelphie,1970 E. Londermen. S. E. Yansehko, and W. S. Heaeston, "An Evolustoon of Rediet6on Damage to Reactor Vessee Steses Usene soth Trenation Temperature and Fracture Mechemco Approaches," in The Efflecte of MacNecon on I 3drucasret 4 Apes 4. ASTM STP42e, pp. 2o0-277, American Society for Testing and Motonels, Philodoephie,19e7 e. M. J. Monicene, "86emiel erittle Frecoure Tests," Trans. Am. Ser. Adece. Enpre. #7, Series D,293 20e (19 eel 7. L Porse, " Reactor Vasese Design Cons 6dering Redistion Effects," Trene. Am. Soc. Miece. Eners. at, Sortes D. i 743 74e (19e4) e. R. E. Johnson, " Fracture Mechemce: A sees for arittle Practure Prevention /' WAPo.TM40s, Novemeer 19e6 9. E. T. Wessee end W. H. Pryle, "invenuestion of the Applicenility of the Siemasi erittle Fracture Test for 1 Ostermeneng Fracture Toughness," WERL4e4411 August 19ee i 10. W. K. W6teos, "Analytse Determmotion of stress Intenesty Factors for the Montoine erittle Fracture Test Specimen /* l WERL402e.3, August 19e5 -11 R. E. Johnson and E. J. Poseere, " Fracture Toughness of irradiated A3024 Steel se influenced by Microstructure /* Trene. Amer. Muct. Soc. 9, 300 3e3 (19 eel l c 1-2 i L

i LIST OF ILLUSTRATIONS Figure Title Page 2-1 Charpy V-Notch impact Specimen 2-2 2-2 Tensile Specimen 2-3 2-3 Wedge Opening Loading Specimen 2-5 2-4 Irradiation Capsule Assembly 26 2-5 Dosimeter Block Assembly 24 2-6 Location of Specimens in the Reactor Surveillance Test Capsules 2-9 i 3-1 Preirradiation Charpy V Notch Impact Energy Curve for the North Anna Unit No.1 Reactor Pressure Vessel Lower Shell Forging 03, Heat No. 990400/292332 (Tangential Orientation) 3-4 3-2 Preirradiation Charpy V-Notch impact Energy Curve for the North Anna Unit No.1 Reector Pressure Vessel Lower Shell Forging 03, Heat No. 990400/292332 (Axial Orientation) 3-5 3-3 Preirradiation Charpy V Notch impact Energy Cuve for the North Anna Unit No.1 Reactor Pressure Vessel Core Region Weld Metal 3-8 3-4 Preirradiation Charpy V-Notch Impact Energy Curve for the North Anna Unit No.1 Reactor Pressure Vessel Core Region Weld Hert; Affected Zone Material 3-9 3 5. Prairradiation Tensile Properties for the North Anna Unit No.1 Reactor Pressure Vessel Lower Shell Forging 03, Heat No. 990400/292332 (Axial Orientation) 3-11 t 3-6. Preirradiation Tensile Properites for the North Anna Unit No.1 Reactor Pressure Vessel Core Region Weld Metal 3-12 37 Typical Tensile Test Stress-Strain Curve 3-13 iv 1 L-

s i s LIST OF TABLES Table Title Page Type and Number of Specimens in the North Anna 21 Unit No.1 Surveillance Test Capsules 27 ~ 22 Quantity of isotopes Contained in the Dosimeter Blocks 2-7 i 31 Preirradiation Charpy V-Notch impact Data for the North Anna Unit No.1 Reactor Pressure Vessel Lower Shell Forging 03, Heat No. 990400/292332 (Tangential J Orientation) 3-2 3-2 Preirradiation Charpy V-Notch impact Data for the North Anna Unit No.1 Reactor Pressure Vessel Lower Shell Forging 03, Heat No. 990400/292332 (Axial Orientation) 33 3-3 Preirradiation Charpy V-Notch Impact Data for the North Anna Unit No.1 Reactor Pressure Vessel Core ] Region Weld Metal 36 3-4 Preirmdiation Charpy V-Notch impact Data for the North Anna Unit No.1 Reactor Pressure Vessel Core Region Weld Heat Affected Zone Material 3-7 35 Preirradiation Tensile Properties for the North Anna No.1 Reactor Pressure Vessel Lower Shell Forging 03 and Core Region Weld Metal 3-10 41 Schedule for Removal of Specimen Capsules 4-1 A1 Heat Treatment History A1 A2 Quantitative Chemical Analysis (Weight-Percent) A-2 1 \\ 4 I l V

TABLE OF CONTENTS Section Title Page 1 PURPOSE AND SCOPE 11 2 SAMPLING PREPARA rl0N 21 21. Pressure Vessel Material 21 2-2. Machining 21 23. Charpy V-Notch impact Specimens 21 2-4. Tensile Specimens 2-1 2-5. Wedge Opening Loading Specimens 2-4 2-6. Monitors 24 27. Dosimeters 24 2-8. Thermal Monitors 24 2-9. Surveillance Capsules 24 2-10. Capsule Preparation 2-4 2 11. Capsule Loading 2-4 3 PREIRRADIATION TESTING 3-1 3-1. Charpy V-Notch Impact Tests 3-1 3-2. Tensile Tests 3-1 3-3. Dropweight Tests 3 14 4 POST IRRADIATION TESTING 41 4-1. Capsule Removal 41 4-2. Charpy V-Notch Impact Tests 4-2 4-3. Tensile Tests 4-2 4-4. Wedge Opening Loading Kid Fracture Toughness Tests 4-2 4-5. Postirradiation Test Equipment 4-3 APPENDlX A NORTH ANNA UNIT NO.1 REACTOR PRESSURE VESSEL SURVEILLANCE MATERIAL A1 iii

l Figure 1 4 Identificati.on and Location of North Anna Unit No.1 i Reactor Vessel Beltline. Material k J N \\ d F o n c-i a c, o f u g / 8 W g 3 d WELD SEAM WOf h J ( i s /~ORG / A/G OV l W 6 3 Yuh a ~2 OoRE as! WCLo SCAM klOV J J foAGidG C3 g } t J / / 3o 4 -Y e l e g

p m TABLE 1 Identification & Chemical Coniposition 6f North Anna Unit No.1 Beltline Region Forgings ChemlCal Co@Osition (WT.%) g.) Com onent Code No. Heat No. Spec. No. C Mn P S Si Ni Cr Mo V Cu Nozzle Shell 05 990286/295213 A508 Class 2 .20 .71 .013 .012 .21 .74 . 39 .64 .03 . li '.21 .75 .010 .019 .21 .82 .33 .64 .05 .1. Inter. Shell 04 990311/298244 990400/292332 619 .68 .009 .014 .22 .80 .30 .63. 02 . li Lower Shell 03 ~ TABLE 2 Fracture Toughness Properties of Unit No. 'l Beltline Region Forgings _, Maximum Inner Wall End-of-Life IO ALEE (FT-LB) T RT NDT Fluenbe NDT NDT USE n/cm R.G. 1.99 W R.G. 1.99 W Component Code No._ 'F

  • F Ft-Lb 8.9 x 10l3

~ 137 110 Nozzle Shell 05 2 6* l9 Inter. Shell 04 - 31 17 92 5.7 x 10 214 140 29' 24 Lower Shell 03 -13 38 85 274 175 31 30

  • Estimated per NRC Standard Review Plan Secticn 5.3.2 Material from lower shell forging n used in surveillance program. Material selected per ASTM E185-73

. _3-

  1. *3

D O TABLE 3 ~ Identification & Chemical Comosition uf North Anna Unit No.1 Beltline Region We1d Metal We1<t Wire Flux Chemical Composition (WT%) Weld Weld Location Process

hpe,

., eat No. M Lot No. C M P S

S_i, fti, R

V Cu .37 .30 36 Nizzle Shell to Inter. Sub-Arc Smit 40 25295 Smit 89 1170 .10 1.50_ .37 -.11 .33 S4 Mo 4278 1211 .09 1.49 Shall Seam WO5 Intor. Shell to Lower Sub-Arc Smit 40 25531 Smit 89 1211 .06 1.29.020.012.35.11 .025.49.001 .086 Shell Seam WO4 (Used'on sl/2" of the seam starting at the vessel.inside diameter TABLE 4 Fracture Toughness Properties of Unit No.1 Beltline Region Weld Metal Maximum Inner Wall End-of-Life NDT( F) AUSE ("F)' Weld Wire Flux T RT NOT NOT USE Fluence Weld Seam 3pe_ Heat No, hpe. Lot No. 'F

  • F FT-LB n/cm2 R.G. 1.99 W

R.G. 1.99 W WQ5 . Smit 40 25295 Smit 89 1170 0** 0** 8.9 x 10j8 275 190 8.9 x 1018 +100 111 WOS S4 Mo 4278 1211' 0** 0** l9 WO4 Smit 40 25531 -13 19 102 5.7 x 10 253 160 34 .17 ' ** Estimated per NRC Standard Review Plan Section 5.3.2 Weld wire & flux associated with weld seam WO4 used to fabricate surveillance weld. a

Figure 2 Identification and Location of North Anna Unit No. 2 Reactor Vessel Beltline Material i m, fo&GItJG of, .) J / j u O 2 d k/Ct.b SEAM Wof t I foncias oy. 'i a a u h Q i U e 1 Co&C d WCLb SCAM Woy ~ t J oW 2 /~ottGiA/G 03 h i ( e i x. t u 1 3 O ~ .J i l I I s I i

.m m TABLE 5 Identification & Chemical Cogosition of North Anna Unit No. 2 Beltline Region Forgings a Mat'l Component Code No. Heat No. Spec. No. C M P S Si, N_1 Cr, M V fu Nazzle Shell 05 990598/291396 A508 Class 2 .20 .68 .'010 .013 25 .77 .34 .60 <.01 .08 Inter. Shell 04 990496/292424 .195 .78 .011 .016 .24 .83 .35 .62 .02 .09 Lower Shell 03 990533/297355 .16,'i .66 .013 .017 .15 .83 .34 .59 .01 .i3 i i iABLE 6 Frecture Toughness Pmperties of Unit No. 2 Beltline Region Forgings Maximum Inner Wall End-of-Life NDT AME M -LB) T U NDT NDT USE Fluen e Co m onent Code No. 'F 'F FT-LB_ n/cm R.G. 1.99 W R.G. 1.99 W ~' 18 8.9 x 10 47 74 Nozzle Shell 05 5 9* ^ 5.7 x 1019 155 120 21 18 Inter. Shell 04 -49 75 74 Lc'wer Shell 03 -13 56 8C 275 153 26 26 l

  • Estimated per NRC Standard Review Plan Section 5.3.2

^ Material from Inter. Shell Forging 04 used in surveillance pmgram. Material selected per ASTM E185-73 Annex Para. A1.3.4 (Special Situation) I ,a ~..

~ ~, TABLE 7 f Identification & Chemical Condositiori of North Anna Unit No. 2 Beltline Region Weld Metal Weld Wire Flux Chemical Composition (WT%) Weld Weld Location Process Type Heat No. _ Type. Lot No. C Mn P S Si Ni Cr Mo V Cu N:zzle Shell.to Inter. Sub-Arc S4 Mo 4278 ' Smit 89 1211 .09 1.49 . 33 .11 .37 .18 .51 801 .086 1.58.012.012.43 Shell Seam WO5

  • i Intar. Shell to Lower Sub-Arc S3 Mo 716126 LW320 26

.08 1.82.017.011 .25.084.042.49.002.088 Shell Seam WO4 t cysed o'n %1/2" of the seam starting at the vessel'inside diameter TABLE 8. Fracture Toughness Properties of Unit No. 2 Beltline Region Weld Metal Maximum Inner Wall End-of-Life IN AUSE (FT-L8) T RT NDT NDT NDT USE Fluenge n/cm R.G. 1.99 W R.G. 1.99 W Weld. Seam Type. Heat No._ Type Lot No._

  • F

'F FT-LB 18 8.9 x 10 +100 111 WO5 .S4 Mo 4278 Smit 89 1211 0* 0* 18 150 151 8.9 x 10 WO5 801 0* 0* I9 WO4 S3 Mo 716126 LW 320 26 -67 -48 107 5.7 x 10 222 161 35 18 s

  • Estimated per NRC Standard Review Plan Section 5.3.2 Wald wire & flux associated with weld seam WO4 used to fabricate surveillance weld

',a

WESTINGHOUSE CLASS 3 l m R= ko3 VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA UNIT NO. 2 REACTOR VESSEL RADIATION SURVEILLANCE PROGRAM l' J. A. Davidson J. H. Phillips S. E. Yanichko November,1976 f, h4 APPROVED: J.5 N. Uniragos, MarjagET Structural Materials Engineering Work Performed Under Shop Order No. VGB 106 WESTINGHOUSE' ELECTRIC CORPORATION I Nuclear Energy Systems P. O. Box 355 Pittsburgh, Pennsylvania 15230

r. ~ r. ABSTRACT A pressure vessel steel surveillance program was developed for the Virginia Electric and Power Company, North Anna Unit No. 2, to obtain information on the effects of radiation on the reactor vessel material under operating conditions. The program comprises the evaluation of the radiation effects based on comparison with preirradiation testing of a selected group of specimens to determine toughness properties of the reactor pressure vessel. Continuous monitoring of these specimens within the reactor pressure vessel pro-vides data on the integrity of the vessel in terms of adequate toughness properties. A description of the surveillance capsules and preirradiated test results is also included. II

TABLE OF CONTENTS Section Title Page 1 PURPOSE AND SCOPE 11 2 SAMPLE PREPARATION 21 21. Pressure Vessel Material 21 { 2 2. Machining 21 2 3. Charpy V-Notch Impact Specimens 21 l 2-4. Tensile Specimens 2-1 2-5. Wedge Opening Loading Specimens 2-4 26. Monitors 2-4 27. Dosimeters 2-4 2-8. Thermal Monitors 24 2-9. Surveillance Capsules 2-4 2-10. Capsule Preparation 2-4 2-11. Capsule Loading 2-4 3 PREIRRADIATION TESTING 3-1 3-1. Charpy V-Notch impact Tests 31 ( 3-2. Tensile Tests 31 3-3. Dropweight Tests 3-14 4 POSTlRRADIATION TESTING 4* 4-1. Capsule Removal 4-1 4 2. Charpy V-Notch impact Tests 4-2 4-3. Tensile Tests 4-2 4-4. Wedge Opening Loading Kid Fracture Toughness Tests 4-2 4-5. Postirradiation Test Equipment 4-2 APPENDIX A NORTH ANNA UNIT NO. 2 REACTOR PRESSURE VESSEL SURVElLLANCE MATERIAL A1 i lii

( LIST OF ILLUSTRATIONS Figure Title Page 21 Charpy V-Notch Impact Specimen 2-2 22 Tensile Specimen 23 23 Wedge Opening Loading Specimen 25 24 Irradiation Capsule Assembly 2-6 25 Dosimeter Block Assembly 28 26 Location of Specimens in the Reactor Surveillance Test Capsules 29 31 Preirradiation Charpy V-Notch Impact Energy Curve for the North Anna Unit No. 2 Reactor Pressure Vessel intermediate Shell Forging 04, Heat No. 990496/292424 (Tangential Orientation) 34 32 Preirradiation Charpy V Notch impact Energy Curve for the North Anna Unit No. 2 Reactor Pressure Vessel Intermediate Shell Forging 04, Heat No. 990496/292424 (Axial Orientation) 35 3-3 Preirradiation Charpy V Notch Impact Energy Curve for the North Anna Unit No. 2 Reactor Pressure Vessel Core Region Weld Metal 3-8 / 34 Preirradiation Charpy V-Notch Impact Energy Curve for the North Anna Unit No. 2 Reactor Pressure ] Vessel Core Region Weld Heat Affected Zone Material 3-9 35 Preirradiation Tensile Properties for the North Anna Unit No. 2 Reactor Pressure Vessel Intermediate Shell Forging 04, Heat No. 990496/292424 (Axial Orientation) 3 11 3-6 Preirradiation Tensile Properties for the North Anna Unit No. 2 Reactor Pressure Vessel Core Region Weld Metal 3 12 37 Typical Tensile Test Stress-Strain Curve 3 13 l i i iv I

...s LIST OF TABLES Table Title Page 2-1 Type and Number of Specimens in the North Anna Unit No. 2 Surveillance Test Capsules 2-7 2-2 Quantity of isotopes Contained in the Dosimeter Blocks 2-7 [ 3-1 Preirradiation Charpy V-Notch Impact Data for the North Anna Unit No. 2 Reactor Pressure Vessel 'ntermediate Shell Forging 04, Heat No. 990496/292424 (Tangential Orientation) 32 32 Preirradiation Charpy V-Notch Impact for the North Anna Unit No. 2 Reactor Pressure Vessel Intermediate Shell Forging 04, Heat No. 990496/292424 (Axial Orientation) 3-3 3-3 Preirradiation Charpy V-Notch impact Data for the ~ North Anna Unit No. 2 Reactor Pressure Vessel Core Region Weld Metal 3-6 3-4 Preirradiation Charpy V-Notch impact Data for the North Anna Unit No. 2 Reactor Pressure Vessel Core Region Weld Heat-Affected Zone Material 3-7 3-5 Preirradiation Tensile Properties for the North Anna Unit No. 2 Reactor Pressurc Vessel Intermediate ( Shell Forging 04 and Core Region Weld Metal 3-10 4-1 Schedule for Removal of Specimen Capsules 4-1 A-1 Heat Treatment History A1 l A2 Quantitative Cheraical Analysis (Weight-Percent) A-2 l i G I v __}}

Retrieved from "https://kanterella.com/w/index.php?title=ML20064G999&oldid=4443119"