ML20056F298

Ultrasonic Exam Results of Lower Head-to-Bottom Head Weld (2RRV-W01) for McGuire Unit II Jul 1993 Reactor Pressure Vessel Exam

ML20056F298
Person / Time
Site:	Mcguire
Issue date:	08/12/1993
From:	Fairbrother D, Martin C, Stoppelmann H DUKE POWER CO.
To:
Shared Package
ML20056F253	List: ML20056F251 ML20056F294 ML20056F298
References
NUDOCS 9308260330
Download: ML20056F298 (25)

Text

/

ULTRASONIC EXAMINATION RESULTS OF THE LOWER HEAD-TO-BOTTOM HEAD WELD (2RRV-WO1)

FOR THE McGUIRE UNIT II JULY 1993 REACTOR PR %SURE VESSEL EXAMINATION l

Prepared by:

./

~ d Date: Nus,.

/f f /78 C4r C.E. Martin - N IlfServices - BWNT Reviewed by: //fxpo, x

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Date: h2 }?uG 79 H.W. StoppelrMnfi - Level III NDE Services - BWNT Approved by:

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Date:

W/7 /'

D.B. Fairbrother - Manager - NDE Services /BWNT

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l Revision 0 I

CONTENTS SECTION.

INTRODUCTION 1

1 l

SUMMARY

OF THE INDICATION DETECTED IN THE 2

LOWER HEAD-TO BOTTOM HEAD WELD

' FIGURES 3

i I

ACCUSONEX PRINTOUTS 4

REFERENCE DOCUMENTATION 5

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INTRODUCTION During the July 1993 ultrasonic examination of the McGuire Unit II Reactor Pressure Vessel, an OD surface connected, planar flaw was detected adjacent to the Lower Head-to-Bottom Head weld (weld 2RPV-WO1).

Contained in this report is a description of the examination technique used for detection of the indication together with a

(

summary of the examination results used for dispositioning the l

indication.

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SUMMARY

OF THE INDICATION DETECTED IN THE LOWER HEAD TO BOTTOM HEAD WELD j

i The ultrasonic (UT) examination was conducted using the B&W Nuclear j

I Service Company's (BWNS) Automated Data Acquisition and Imaging System ACCUSONEX, together with the Automated Reactor Inspection i

System ARIS.

ARIS served as a delivery tool for positioning the j

ultrasonic transducers in the examination region of interest.

i Data collected from the transducers was routed via.triax cabling (500 i

feet) to the ACCUSONEX acquisition system which was staged inside I

the ARIS control trailer.

Here the data was filtered, amplified, and subsequently digitized for storage onto an optical disk.

The i

disk was then hand carried to the on-site analysis station for transferring, via computer modem, to BWNS Lynchburg for data analysis.

The data was then analyzed in-accordance-with the BWNS procedure ISI-138, Revision 13 (Remote Ultrasonic Examination for

(

Reactor Vessel and Associated Piping Welds Using ACCUSONEX).

A copy of this procedure is provided in Section 5.

r Examination of weld 2RPV-WO1 was performed using the contact (vice

{

immersion),

full-node examination technique.

(A full-node examination was performed due to an Incore Nozzle limiting the ARIS j-range of movement in the circumferential direction.)

The water in i

F the reactor vessel served as the coupling agent between the i

transducer and vessel's cladded surface.

A total of seven transducers were simultaneously utilized for examination of the f

weld, namely two 45* shear waves, two 60 shear waves, two 70o longitudinal waves, and one o degree longitudinal wave.

The i

transducers were housed in a 3X3 spring-loaded matrix referred to as the Contact-Head.

(See figure 1 for an illustration of the l

transducer arrangement as viewed from the back of the Contact-Head (i.e.,

facing the vessel's ID surface)).

The Contact-Head is i

attached to the ARIS Boom for delivery to the vessel's surface (see

(

figures 2 and 3).

Associated with the ARIS tool is a total of

[

eight ranges of motion, also illustrated in figure 3.

By lowering the Mast into the vessel ("Z" motion) contact can be made with the Lower Head-to-Bottom Head weld as illustrated in figure 4.

After

{

contact is made, examination of weld 2RPV-W01 can be performed using only two of the ARIS motions, namely Theta and Alpha.

Theta is a circumferential movement in the horizontal plane, and Alpha is a circumferential movement in the vertical plane.

Thus Theta is used to scan parallel to the weld axis, and Alpha is used to scan perpendicular to the weld axis.

I I

L 2-1 i

-w.

h Revision 0 Weld 2RPV-WD1 was examined using six axial scans and twelve l

circumferential scans.

(Axial is perpendicular to the weld axis, and circumferential is parallel to the weld axis.)

The indication was detected on circumferential scan Bl.2.1:C-2.

(B1.2.1:C-2 is a scan. identification number (SIN) unique to the ARIS tool.)

The indication was initially detected, at 1/2 Vee, using the 60* and 45* shear wave transducers (transducers number 8 and 9 in figure-i 1).

Both transducer beam directions were pointing clockwise when viewed looking down into the Reactor Vessel (i.e.,

Theta

+

direction).

The following information was recorded with these scans:

I ACCUSONEX printouts are provided in Section 4 I

i i

Angle Peak Amp.

Depth Length Alnha Theta 45 S 16% DAC 5.72" 2.40"2 35.51 to 37.08 129.64 60 S 13% DAC 5.88" 1.74"2 34.85 to 36.36 129.71 j

1.

Noise level length e

2.

Half maximum amplitude length j

To further evaluate the indication, supplemental RF scans were performed using 40

, 50*, and 60* longitudinal wave transducers.

All three transducers were configured such that the beam direction j

again pointed in the clockwise direction.

Relative _to figure 1, I

the 60* was loaded into position 7, the 50* into position 8, and the 40*

into position 9.

The results obtained from these l

transducers supported the results from the 45* and 60* shear wave l

transducers used for the initial detection scans.

i EVALUATION Evaluation of the indication was based on RF data acquired'using the 45 shear wave transducer.

As mentioned

above, the longitudinal wave scans supported the shear wave results; however-

- the shear wave yielded better sizing information.

Sizing of the indication was performed in both the through-wall dimension (TWD) and the length dimension.

The TWD was based on results from the Tip Diffraction Technique,-

and the-length dimension was based on " Noise level length".

That_is, the length where the - indication. blends. with the noise. generated from _ the

- component.

The tip diffraction method yielded a TWD of 0.5" (one-half inch),

and the noise level length method yielded a length of 2.4" (two and four tenths inches).

Furthermore, the indication was determined to be oriented in the axial direction (vertical plane)-

at a mean Theta value of 129.68* (average of 129.64

• and 129. 71*

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.w th an Alpha range of 35.51* to 37.08 Physically, Theta zero coincides.with the Reactor Vessel *W-Axis",

and an Alpha value of 2-2

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36.1* coincides with the centerline of weld 2RPV-WO1.

Therefore summarizing:

the indication is located 129.68* clockwise (looking into the vessel) from the "W-Axis",

it is oriented transversely (relative to the weld axis) and travels from 0.9" above the weld l

centerline to 1.5" below the weld centerline for a total length of 2.4 inches.

It should be noted that the indication was only i

characterized from one scan direction (i.e., Theta + direction) due i

to its location in close proximity to an Incore nozzle which i

prevented scanning from the opposite direction (i.e.,

Theta - ).

Disposition of the indication yields the following results:

OD Surface Connected, Planar Flaw.

[

Oriented Perpendicular to the Weld Axis i

Vessel Thickness @ Indication = 5.85" - 0.157" (clad) = 5.69" i

length (1) 2.4" t

=

TWD (a) l 0.50"

=

a/l 0.50"/2.4" = 0.21

=

a/t%

0.50"/5.69" *100 = 8.8%

=

l 1980 ASME Code,Section XI, IWB 3510-1 acceptance standard allows a maximum 2.8%-a/t for an aspect ratio (a/l) of 0.21.

l Therefore based on the above findings, the indication has been determined to be UNACCEPTABLE.

1 P.

In Section 3, Figure 5 a drawing is provided showing the physical location of the indication.

Also, in Section 5 a copy of the Scan

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Plan drawing utilized for examination of the weld is provided for

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additional information.

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BWNS - NDE SERVICES Ii CUSTOMER: Duke Power Company I

PLANT: McGuire Nuclear Station Unit 2 COMPONENT: Lower Head to Upper Head j

TRANSDUCER: 45 DEG SHEAR (5508-93002)

Conur.ents: 45 degree shear wave indication at 129.64 deg. Theta Length at noise level end points 35.51 to 37.08 deg. Alpha i

SIN: Bl.2.1 SUBSCAN: C-2 ANGLE: 45 BEAM DIR: 90 PATH: #0:/A3213_09.10/

CREATED: Aug 1 09:11:08 1993

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DUKE POWER COMPANY MCGUIRE UNIT 2 1

07/93 SIN 8 /. 2. / : c-1 Channel #

h l

Ind. Type da of

~

A/8[4 36', 5~/

3 7. O P Depth S 72 cR

/2 9.64' Max Amp

/d o Ind.~ Length - 50% DAC

//A 20% DAC

' A/A l

Comments Le,1s 71

=/.57 lee

. 2. Yo ar noibe /eud ns

=

o.sv- " r a d -

-,, a.. - n k v

i Analyst l

7/R[9J V

Acceptance

[cJi3 - 35/O - /

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): 2.V ~.

C =

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ex 7cc es,o n M.A e cher,.Masr c Actual ER#

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Peak More Cursor Measure tM1 Utilities 130.42 128.03 34.10 38.10 3.35 7.46 11 r~.

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l.

McGuire tJucicar Station Unit 2 File No: MCC 1201.01-00-0027 Dev. o l

Subiect:

Evaluation of Reactor Vessel OD Flaw,fPIP 2-N93-0717)

Pa cre :

"23 BV

• 4/hCV Date: F//v/e*2 Ck : d*446)m J Date:

4'-4 -yJ i

f -

<yr-I

! 3.4-1 REACTOR CGOLANT SYSTEM i

3/4.4.9 PRESSURE / TEMPERATURE LIMITS I

REACTOR COOLANT SYSTEM l

t!MITING CONDITION FOR OPERATION i

3.4.9.1 The Reactor Coolant System.(except the pressurizer) temperature and Z

{

pressure shall be ilmited in accordance with the limit lines snown on Figures I

3.4-2, 3.4-3, 3.4-4, and 3.4-5 during heatup, cooldown, criticality inservice leak and hydrostatic testing with*

)

, and j

i Maximum heatup rates as specified in Figures 3.4-2 and 3.4-3 a.

[

b.

Maximum cooldown rates as specified in Figures 3.4 4 and 3.4-5 7

{

A maximum temperature change of less than or equal to 10*F in any c.

j 1-hour period during inservice hydrostatic and leak testing j

operations above the heatup and cooldown limit curves.

i APPLICABILITY: At all times.

ACTION:

l With any of the above limits exceeded, restore the temp sure determine the effects of the out of-limit condition on the structural e

j System remains acceptable for continued operation or be n

i STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and reduce the RCS T, g and pressure to less j

than 200*F and 500 psig, respectively, within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

y 1

SURVEILLANCERNOUIREMENTS 4.4.9.1.1 The Reactor Coolant System temperature and pressure shall be heatup, cooldown, and inservice leak and hydrostatic testing em 4.4.9.1.2 The reactor vessel material irradiation surveillance specimens shall be removed and examined, to determine changes in material properties required by 10 CFR 50, Appendix H in accordance with the schedule in Table

, as 4.4-5.

3.4-2, 3.4-3, 3.4-4, and 3.4-5.The results of these examinations shall be used to u l-a

~

McGUIRE - UNITS 1 and 2 3/4 4-30 Amendment No. 82 (Unit 2)

Amendment No.100 fUnit !)

I i

I l

l l

McGuire tJucj ear St ation Unit 2 File No: MCC 1201.01-00-0027 Rev. O Subiect:

EvaluatJ on of Reactor vesse) OD Flaw.LPIP 2-M93-07171 L% rte :

'2 +

By: 4/FX Date: /'//% e Ck: @ Da4 Date: 6-4 -93 l

l l 3.4-2 2500,'

l

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to 60*F/hr vi*.i6 1

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j t j,,

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~

,6

> Temperature (235"F) for

. 4 i

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50 100 d150 200 250 300 350 400 450 503 l

INDICATED TEMPERATURE (DEG.F) l 1

1 C u a v t s A P P L IC A R L t. F O R M t A T U P M A T E R E A L B A tts i

mATaiUPTos0**tHa*ORTHE 8 8 a v 'C '

Cowt ao tt tNo w ATa n eA L: L O v. t m s a t tL pgntOD UP T O t o E P P Y. CO NT A IN g M A R DIN E of 10's A N D 8 0 P s f G P o n P O S S IB L E COPPEP CO NTE NT: 0.18-tt IN ST R U W E N T ERROR.

RT INITI A L: =40'F

{

h0T gy AyygmteifPYt 11d T. 8 0 *

• y 3 /4 T, t 1* F FIGURE 3 A 3 C

M GUIRE LWIT 2 RE ACTOft COOLANT SY3 TEM HE ATlr LtW1TATICPf5 NRC RG 1.M PIEV 2 WLICABLE FOP. THE FIRST 10 E FPY McGUIRE - UNITS 1 and 2 3/4 4-32 Amendment No.

115 (Unit 1)

Amendment No.

97 (Unit 2)

-O 7

Fev. O gcGui t e Nuclear Station Unit 2

Fi]e No: MCC 12.01.

sut1j ect :

Evaluation of peactor Vennel OD Flaw P*

6'//yg?

Ck' D W ' F M 6

Ene : ?S Tiy : rzyff Date:

, 3.4-3 YD ' f i ~ i !!! I 1

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O 50 100 150 200 250 300 350 403 450 500 INDICATED TEMPER ATURE (DEG-F)

M AT E R 8 A L B A sta CURVfB A PP L K A B L E FOR COOLOORN PATts tre TO 164* * /H R FOR THE se rvict P f a lc o CO N T R O L LIN G M A T E R t A L - L o w t m $H E L L COpetP C O N T E N T ; o.i t t s UP TO te EPPY AND CONT AIN E M A P GIN S OF RT INIT I A L: - 30*F 10*P A 8e D 64 PS14 POR PO581sLE 8s D T RT AFTER toRPPY: 1/4T,30*r t N 4 T R U M t es T t m mom s.

hDT 3/4T,81*F 4

FIGURE 3.4-5 P#;GutRE UNIT 2, REACTOR COOL /

f SYSTEM. COOLDOWN LIMITATIONS N R C R G 1.74 R E V 2 APPLtCABLE FOR THE FIRST 10 EFPY

McGuire Nuclear Station Unit 2

File No: MCC 1201.01-00-0027 Rev. O t

Subiect:

Evaluation of Reactor Vessel OD Flaw (PIP 2-M93-0717)

Pace:

W Bv:

MP%

Date: 6N/73 Ck: chhu Date:J'-& L7 13.5 Evaluation 2: Normal and Upset Conditions.

This evaluation considers Normal & Upset Transients.

Table 13.5-1 and its notes describe in summary form the results of this i

evaluation.

Further discussion is given here.

In this evaluation stresses of combined cases of Pressure and Thermal Transient temperature fields (and residual) are used to calculate K

values 1

i PRESSURE and TEMPERATURE STRESS FIELD DETERMINATION Sir. 3 KIa and K

material values are a

function of material Ic temperature, and Ky values are a function of pressure, it is necessary that the appropriate pairings of pressure and temperature are evaluated.

Basic pressure and temperature combinations for Normal and Upset are 2250 psia and 557 'F as shown in Figure 1 of reference [11-4].

(Psia and psig are assumed equivalent herein.)

The effects of uneven heating through the vessel wall are the main subject of this evaluation.

These effects increase in' severity with the rate of heatin'g or cooling, and since it is interior heating that will cause tensile stress on the exterior surface (additive with pressure stress),

heating is evaluated.

Table 13.5-2 shows the temperature changes and rates for those Normal and Upset Transients with an increase in Tcold. (Neither of the two faulted cases have increasing Tcold.

an enveloping case), Instead of performing separate evaluations for each,

" Enveloped Heating Transients",

is created b

taking the maximum temperature rise and the minimum accompanying time.y j

The stress effects of this heating were determined using the computer program TRANS2A, described in section 14.0.

The following summarizes the input.

Section Properties:

SECTION' MATERIAL INSIDE WALL AMBIENT WALL NUMBER NAME TYPE (DIAMETER THICKNESS TEMPERATURE DIVISIONS (IN)

(IN)

(F) 1 LOWRHEAD CMS 176.000 5.690 70.000 20 l

i i

i

i McGuire Nuclear Station Unit 2

File No: MCC 1201.01-00-0027 Rev. O Subiect:

Evaluation of Reactor Vessel OD Flaw (PIP 2-M93-0717)

Pace:

l'7 Bv:

4//A>fL Date: yNyyl Ck: W*nd w Date : 6-N-93 I

The input boundary conditions specified a ramped temperature function at the inside surface beginning at 557

• F at time zero and ending at l

601

'F at eight seconds which is held constant until one hour is reached.

The accompanying flow was set artificially high to conservatively force the inside metal temperature to respond at the same rate as the fluid.

In view of the conservatisms

employed, the L

effects of not considering the cladding are negligible.

The solution was numerically determined using one time step for each degree of temperature rise for the first 8

seconds, followed by one time step for each second until one minute is reached, followed by one 4

time step for each minute until one hour is reached.

Figure 13.5-1 shows the inside, mid

depth, crack

tip, and outside temperature responses vs. time.

i Since no rotation takes place in the vessel wall, the resulting stress at any point at any time is E

• alpha

• OT / (1-nu) where E is Young's

modulus, alpha is the linear coefficient of thermal expansion, cT is the difference between the temperature at the point and the average 1

through wall temperature at that time, and nu is Poisson's ratio.

This evaluation used the maximum stress at the exterior surface and conservatively assumed it to be all membrane in calculating K.r To find this maximum stress, Figure 13.5-2 shows the plot of the subject OT vs.

time, and Figure 13.5-3 shows the profile of stress through the section at the time (240 seconds) when this oT is maximum.

(Note the near flat slope of the profile near the exterior surface, l

reinforcing the decision to characterize this stress as all membrane in accordance with ASME Section XI Figure A-3200-1.)

l In Figure 13.5-2, this maximum OT is 13.45 "F.

)

Pressure Stress is computed using thin wall theory.

In Table 13.5-1, a stress of 2485 psig is conservatively used.

1 This location is sufficiently remote from significant structural discontinuities to preclude the necessity for stress multipliers.

RESIDUAL STRESS The 10 ksi residual stress (as discussed in section 13.3 bending stress in accordance with ASME Section XI Figure)A-3200-1.is applied as K

DETERMINATION 7

Calculation of KI and comparison to allowables for the appropriate combinations of stresses are shown in summary form in Table 13.5-1.

i I

i

i 1

l l

I l

b McGuire Nuclear Station Unit 2

File No: MCC 1201.01-00-0027 Rev. O I

Subiect:

Evaluation of Reactor Vessel OD Flaw.(PIP 2-M93-0717)

Pace:

26 Hv:

1. Wp Date: 8//r,/92 Ck: thud hu Date: F-/ 6 -f3 i

l

SUMMARY

i t

I Table 13.5-1 shows that the worst comparison of calculated K to l

1 allowable is 62.8%, even with all the conservatisms employed.

i 1

l t

l 4

\\

i 4

i I

McGuire Nuclear Station' Unit 2

File No: MCC 1201.01-00-0027 Rev. o EUhlect:

Evaluation of Reactor Vessel OD Flaw.fPIP 2-M93-0717)

Page:

1.'t By:

MK Date: Nyq/r 2 Ck: cM a.a Date: E-/4-y>

Table 13.5-1 1980 ASME Section XI Appendix A Analysis of Flaw Indications j

Normal & Upset Transients w (Enveloped Heating Tr/Tcold Increasing ansients)

Units inch, kip, "F

Reactor Note Temp J

557.000 Press J

2485.000 a

f 0.500 1

I2 2.400 t

d2 5.690 a/t 0.088 a/l 0.208 RTNDT (3) 10.000 T-RT 547.000 NDT KIa (4) 200.090 K

(4) 200.000 l

Ic om pres (5) 19.29h OT nemb (61 13.450 i

a OT f6h 4.090 a"mresid f7 i

0 o

[8) 23.389 m

ab Pres (5)

O t

oT bend 1.170 i

ab oT 0

ab resid 10.000 ob 10.000 M

(91 1.100 a$

M i9h 0.920 fl0 42.516 I

QY8 (11 1.214 I

K (12 39.723 y

V2K /KIc (13) 0.281 y

V10Ky/hla (13) 0.628 Notes:

(1)

Bounding reactor coolant temperature & pressure for Normal & Upset Transients w/Tcold Increasing; Reference [11-3] & Table 13.5-2 (2)

Crack depth, length & vessel wall thickness, Reference 12-4

& " Characterization of d

Flaw" section 13.2 (3)

Determination of Reference Transition / Nil Ductility Temperature, reference section 13.1 (4) Available fracture toughness based on crack arrest fracture initiation, respectively, for the corresponding crack tip temperacure (ksiVin) as defined in ASME Section XI, Appendix A, Figure A-4200-1 (reference section 4.0).

(5) Membrane pressure stress = PD/4t Bending compon(thin wall theory hoop stress in a spherical shell) where D = 176.75".

ent of pressure stress = 0.

(6) Maximum of average minus outside temperature from TRANS2A analysis (microfiche attachment M1 & Figure 13.5-2), used to calculate om (EaoT/(1 p) where E'= 29.9E3, a

= 7.12E-6 in/in/*F,

= 0.3).

conservatively define all transient st.;ss to. be membrane (M is larger than M I" EI computation, reference ASME Section XI Appe1 dix b

A, article A,-3300).

(7) Residual stress is conservatively assumed to be 10 ksi bending per the 1986 edition of the ASME Section XI Appendix E, Table E-2.

(8) Total membrane / bending stress to be used.in K

determination.

Sum of pressure, transient & residual stresses.

y (9) Correction factors for membrane & bending stress as defined in ASME Section XI, I

Appendix A, article A-3300.

See Figures A-3300-3 & A-3300-5 (M = 1.1 & Mb = 0.9).

(10)

Yield stress of material O temperature.

Reference Figure 13.4-7 & ASME Section

III, j

Appendix I,.1977.

(11)

Shape factor for flaw as defined in Figure A-3300-1 of ASME Section XI Appendix A.

(12) Stress intensity factor as defined in ASME Section XI Appendix A, article A-3300.

o M M (a/Q) +

K M

a/Q)

=

ok b< K l

(13)

Rktio E3o allowables; m

dre rdqu/V10 for normal conditions and K7<K

/V2 for emergenr.:y i

a article IWB-N12.

& faulted conditions, ired by ASME Section XI i

~

~

DUKE POWER COMPANY MCGUIRE UNIT 2 07/93

//

T'.'

I sin 8/. 2. I : C-2 Dif: [O 3

Cbannel #

AL antu km As2n or.it vi.n sis Ind. Type 72AA/4A litN/A 3L/.Y5 3 4. 34 Depth

5. E'Y WETA J29. 39 130.02 Max Amp l3 b 'DAC-Ind. Length 50% DAC

'~

N/A

_20% DAC M/A Comments HA<A LEMn = ('36.

- SV.9hh X(81.2 + 5. C) Y 4=lW Two = 5. Tf Ef"- 5. 29 ~ =

0.59"

( rips )

/?LP/fA b17/dTA GH=

YY nao iH/

Werthscn +3y*

/

r Analyst

's.

$/

? ! 21!9 3

(

l t

Acceptance f&>B-3570-/

Calculation

.See e v a I m 4 7;,,

eV

,,}l,, nin u,Th ys "

Disposition UA/BCr EP777MZ E Actual ER#

Reviewer 8 amu [

uud,w.

I u

Page I

of )

^

f I

ATTACHMENT 2 DUKE POWER COMPANY MCGUIRE NUCLEAR STATION FABRICATION AND INSPECTION HISTORY OF WELD WOI I

f e

Fabrication and Inspection History of Weld WOI j

The wcld WOI Index provides a chronological list of fabrication events and inspections performed on this weld by the Rotterdam Dockyard Company (RDM) from 9/23/71 untd l

4/15/74. The associated reports are organized by file number from 664-W01-001 thru j

664-W01-011. Reports for additional heat treatments that this weld underwent as a part of i

larger assemblies or the complete vessel are organized in file numbers 664-WO3-006,664-l WO5-004, and 30664-002. Also included are RDM weld specification and the baseline UT l

examination report.

l IIcat Treatments The following is a summary of all post weld heat treatments for weld WOl(see Attachment I for drawings of the referenced assemblies):

Date Report File Heat Un Rate Hold Temp Hold Time Cooldown Rale i

11/5/71 W01-005 36 "C/hr from 315 C 595-605 "C Oh 28m 30.5 C/hr to 300 C l

U f

U 12/22n1 W01-005 38 C/hr from 315 C $95-610 C Oh 44m 33 C/hr to 300 C O

O 2/28D2 W01-005 43 C/hr from 315 C 595-610 C Oh 40m 35.6 C/hr to 300 C Total for WOI in Assembly 1 = lh 52m f

O O

6/12D2 WO3-006 28.5 C/hr from 315 C 595-605 C Oh 34m 42 C/hr to 300 C Total for W01 in Assembiv 2 = 2h 26m O

4/4/73 W05-004 33 C/hr from 315 C 595-615 C lh 20m 24.5 C/hr to 300 C O

4/2In3 WO5-004 44 C/hr from 315 C 595-625 C lh 30m 39 C/hr to 300 C 5/25D3 WO3-006 31 "C/hr from 315 C 595-605 C Oh55m 30.5 "C/hr to 300 C j

U 6/9/73 30664-002 18 C/hr from 315 C 595-620 C 1Ih 20m 16 C/hr to 300 C O

O Total for Wo1 in Complete Vessel = 17h 31m F

Fabrication Inspections:

Magnetic Particle examinations were performed on weld WOI in as follows:

Date Surface Examined Ersuhs 11/11/71 WOI inside No indications found 12/15/71 WO1 ground rootpass No indications found 12/28/71 WOI circle seam, outside No indications found 2/22/72 WOI circle seam cavities No indications found 2/24/72 WOI repaired areas inside & outside No indications foend 3/2/72 WOI repaired area on outside No indications found 3/20/74 Complete vessel No indications found Ultrasonic examinations were performed on weld W01 in as follows:

D_ ate Volume Examined E nu.hs 1/3/72 WOI (complete weld)

Some indications found i

2/2/72 W01 (repaired areas)

Cleared 1/3/72 results 1

4/25/73 W01, 3, 4, 5, 6, 7, and W11-W18 No indications found in W01 3/19/74 WOI after hydrostatic test No indications found i

Page1of2 l

i i

Fabrication Inspections (cont'd) l A liquid penetrant examination was performed on weld WOI of the cladding CO2 surface to verify removal of an indication found during the 1/21/72 radiography.

1 Baseline Radiography:

The only radiography for this weld was performed on 1/21/72 and 3/20/72 per report 030772, file number 664-W01-010. This report records all indications observed in the 1/21/72 inspection were cleared with 3/20/72 inspection. A recent review of this original i

inspection film from the area of the weld indication showed no defect to be present.

The dates of this radiography correspond to the fabrication step where only the bottom head and spherical shell were joined together. Therefore indications due to the additional j

fabrication required to make the complete vessel, shipping, or installation would not be present on this film.

I E

Baseline Ultrasonic Examination:

l The baseline ultrasonic examination was performed during the preoperational inservice inspection on 9/8/78. This examination was performed by B&W using an immersion UT technique, with water in the reactor vessel acting as the coupling agent, versus the contact l

technique used for the 10 year ISI. No indications were found during this examination in the area of the current indication.

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OUAllTY CONTROL TRAVELLER REPORT No.

Date :

Project:

DUKE POWER Order no.: 30664 Contract ref.: 54 6-CYQ-116629 BP RDM Drswing no.: 3o73s-1536 a

R.I.S*

ltem po.: 01-02 Inspection by:Llo I

vien Veldjoint bottenheadcap to spherical ring M

N D

Notes Result no.

re.

wty resor Per/Dets 01 23.06-6 MT Surface of weldprep on botten OK 1971 headcap for WOI 23 g,

664-v01-ool 02 Soaking af te r buttering for 1971 breakets OK 06 664 v01-oo3 03 23.06. 6 MT Surfacs cf butterings

' oh na 1971-1o.

Yh 10 664-vol-col o%

Soaking after buttering for UK 1971-10 temp. lifting luge 18 66le -W01-003 05 23.06-6 Mr Surface of ws,1dprep on spherions ox

'or 1971-1c-21 ring for WOI 66f4-WO1 o01 06 rit up 1971-10 21 y-, +

s

(

664-vol-oc!4

'07.

Devintion request 1971 'a 25 664-vol-coa-015 A*adt:

Acovmes:

l OK

M LT : Leak tergHg D

Mbner imperduenons aff : gespans petists testing.

)

ressound and romanined.

PT : h aedne.

U

U e NT : Hoeness suchs.

RT : AndayusMsasesing.

TT : Twskness ensing.

UT : Leansunds essehe.

DIM :Dissensiend spansk.

FT : Ferrtes emedag.

r. e.1 1 C

* !.

1 */

8888 of cor. 41

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.dh; l

1 WAALITY AIS r 8-N-

,.,4 e _,

QUALITY CONTROL TRAVELLER N EL 6 45-7 7

usee Amust s-N 08

.3.91-9 Heattreatment after welding OK 1971-11 4 circle seem from ID side 05 f/,

664-WO1-005 19 23.06-6 MT Surface of circle seam OK 1971 -11 ~r

' Ma B*

(inside)

IfY 664-WO1-001

-n t

10

5.06-7 MT Rootpass OK Ms 1971-12.e.-

i

..' t.15.s 664-WO1-001

. 4%

an.

11 33.01-9 Heattreatment af ter welding*

OK 1971-12R' i

c'.rcle seam from OD u

L

'22f Q

664-WO1-005

_~Ud l

~- -

1?

2'.'C-7 MT Surface of circle smaa

.:,m

• 'r OK TH ' 1971-12 4 l

MrdB,Y j

L (outside)

- V M ~j 9 P.7 D C@.

N e Vf4 PuE 664-WO1-001 - N. "

[

l

. '2.i4 $

/fi4 1972-0 W k

13 Varint!on report J-o

- f21 w

- n. ;4

• 664-WO1-006-Y48.

J.

A5p @x

,y.. g. M:,g 1972-of ;<;

-...q 2

14 UT 3epth cen,uring defects "4

7 ;.,

.CK M

(

664-W01-007 ~y g rfy; Yei EC{f26,j?

n24p';.:

% 4;F)

(see V4S

' ~. H y,'

,wvs

'~

15 DD1 Dimensional check envities L

rd 1972-02 $

664-WO1-00k b O

E.if 1

o

.vcec.

-..y

+>

t urface of cavities (elears *;y#q.T

• 0E T No y 1 O

16 par 2.3 of V48) c.

eWingE pk 1 :Grl, 23.36-7 M.T c

664-W01-001 '3@dtd3 d

.%?

e 4 c..

y.,w.

~

<. 3 Surface of repatrd. m e('elears;.

iOK "

Pu - 197 17 23.06-7 MT ares 2 par 2.6 of Y48)V :,i Z?Fl*h&yij$

5.y Q

$f ud 664-W01-00i e

86 Y

seam WO1,- ;

M ;4 /19.'. 7 7-

]' ff

.g[d 19i27 ~

18 2?,07-5 RT Radiograph d d d t M N5Ffelds' 4k 1

_ 21 y

664-WO1-010 a c':

• n i~

- N. "

m

}l5 kidMhb kN M Mb 1972 Nib 6$j&bkkt$;$$.$

19 21.07-1 UT Circle seam

~ "h.Wh L.~ G0

~

20 53.01-9 Hest treat ment after repair .,

OT.

1972-02m;,:

,; 28.3 (clears par 2,5 'of 748)

F.

664 W11-005 l

5;+

pups af

~;P 1. :1 C.

~

t.' s.1

.1 Ce

QUAUTY AaB kWTALLARWCAL N QUALITY CONTROL TRAVELLER REPORT NO.

h M

D

• y suasr no 3"#f*

• f C1'diiUE OK H

1972 21

?4.76-8 PT (clears par 2.7 of V48)

Gr 02 6f a m1_nnq

??

2".05-9 UT Cladbonding (clears par 2,7 of OK B1 1972 02 V48) 6 6 4

• .1 3 1 - 0 0 9 23 2?.07-5 HT Rera i rs (c1 ears par 2.8 of V48)

OK 1972 20 664-V01-010 Circle s e mass : WO1,WO 3,WO4, WO 5, WO6, Ye OK vD 1973 24 21.07-2 UT WO7 and Wil thru W18 B1 25-26 60'prohr 664-WO1-011 D

25 21.08-7 UT Circle sea = WO1 af ter Hydro test OK Gr 1974-d3 18 664-WO1-011 26 23.03 4 MT Surface a rter Ilyd ro test OK 1974-O %

04-15 664-vol-001 O

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OVERSIZE DOCUMENT PAGE PULLED SEE APERTURE CARDS NUMBER OF OVERSIZE PAGES FILMED ON APERTURE CARDS 9307%o 33pg -02 j

APERTURr CARD /HARD COPY AVAILABLE FROM RECORDS AND REPORTS MANAGEMENT BRANCH j

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