ML20087L506

Discusses Reactor Vessel Outlet nozzle-to-shell Weld Flaw Indication Evaluation.Requirements of ASME Section XI Satisfied & Reactor Vessel Confirmed for Continued Operation

ML20087L506
Person / Time
Site:	Point Beach
Issue date:	03/23/1984
From:	Fay C WISCONSIN ELECTRIC POWER CO.
To:	Harold Denton, John Miller Office of Nuclear Reactor Regulation
References
TAC-54446, NUDOCS 8403270237
Download: ML20087L506 (6)

Text

-

<sii!!!>

Wisconsin Electnc ne cowa 231 W. MICHIGAN, P.O. BOX 2046, MILWAUKEE, WI 53201 March 23, 1984 Mr. H. R. Denton, Director Office of Nuclear Reactor Regulation U. S. NUCLEAR REGULATORY COMMISSION Washington, D. C. 20555 Attention: Mr. O. R. Miller, Chief Operating Reactors, Branch 3 Gentlemen:

DOCKET 50-266 REACTOR VESSEL OUTLET NOZZLE-TO-SHELL WELD FLAW INDICATION EVALUATION POINT BEACH NUCLEAR PLANT, UNIT 1 In our March 12, 1984 letter we transmitted for your concurrence our procedure for evaluating the acceptability of the flaw indications in the Point Beach Unit 1 reactor vessel outlet nozzle-to-shell welds. In that letter we also advised you that it was our intention to demonstrate that the indications meet,the acceptance criteria of ASME Section XI thereby confirming the continued serviceability of the reactor vessel.

The evaluation consisted of two approaches: experiments by Southwest Research Ins' itute on the Point Beach calibration block and fracture analysic of the flaw indications. The calibration block experiments were undertaken because of the difficulties in sizing the reflectors in the outlet nozzles. Basically, these indications were oriented normal to theThis incident beam and almost provided a high amplitude parallel to the examination surfacc.

response but an insignificant change in metal path. The Code does not outline or specify the technique for sizing this type of reflector. The calibration block experiments in which the position and orientation of the nozzle indications was duplicated with a 3/8" flat-bottomed hole in the calibration block permitted the

' performance of beam spread analyses using the same search unit assembly as was used at Point Beach to more accurately size the reflectors. Based on these studies, beam spread correction factors were determined. The beam spread corrected sizing demonstrates that the two previously unacceptable indications are, in fact, within the limits specified in the Code. Attachment 1 documents the corrected characterization of the reflectors of concern.

8403270237 840323 PDR ADOCK 05000266 P PDR _y_

P 1 1 \

l

E

p '* ~

EMr. H. 1R. Denton

~

March 23, 1984

.Notwith' standing the above_ conclusions, a fracture

..-integrity. evaluation of the flaw indications has now been

-completed by-Southwest Research Institute. The analysis, which. utilized a' linear elastic fracture mechanics approach in.accordance with'ASME Section XI requirements,. chows that

.the' observed flaw indi~ cations, as initially ~ sized, are stable under.all postulated loading ~ conditions. -Additionally, the margins:of. safety..specified in IWB-3612 are met. To aid in your

~~

preliminary review, Attachment 2 is provided which summarizes the ja  ; key' aspects of the flaw indication evaluation.

We expect to forward complete reports on Southwest Research Institute's beam spread analysis and fracture integrity

evaluation by March- 30, 1984. Should'you have further questions

please do not hesitate to contact us.- If you desire, a meeting

= 'can be arranged to discuss the evaluation reports.

It is our present' conclusion.that the requirements of AS;ME Section XI have been satisfied and that the serviceability of tdum Point Beach Unit 1 reactor. vessel is confirmed for continued operation.

Very-truly yours,-

. ,Y hV '

Vice President-Nuclear Power C..W.1 Fay

' Copy to NRC Resident Inspector

-Attachments- ,

w I J 5

.'A' k li - p gg i ri -

i

___._m..

~

-

• Attachment 1 POINT BEACH NUCLEAR PLANT, UNIT 1 REACTOR VESSEL OUTLET NOZZLE REFINED FLAW CHARACTERIZATION

. Southwest'Research Institute's experiments on the calibration g a block corroborated the' original calibration very well. These Kstudies also produced correction factors based on the beam spread' associated with the particular search unit used at

~

Point Beach.- These correction factors, for these indications only, were determined to be 1.035" in length and 0.709" in the

through-wall dimension. Applying'these-factors to the two indications in'the outlet nozzles: yields the following refined

-reflector characterization: ,

Flaw'#2 ._

Flaw #4

~

'"A". outlet. nozzle-to-shell weld. "C" outlet nozzle-to-shell weld.

2a = 0.531" _, __

2a = 0.431"

-1 = 0.9" 1 = 0.255" a/1 = 0.295 a/l = 0.5 (Limit) a/t%' a/t%

Allowable- Measured Allowable Measured

.4. 05. 2.9 6.5 2.4 where-t = 9.125"

'As can be seen, both indications are within Code allowable.

/

s

^$ ' .'

q:

Att chment 2

SUMMARY

'OF-SOUTHWEST-RESEARCH INSTITUTE'S FLAW INDICATION EVALUATION POINT BEACII NUCLEAR PLANT, UNIT 1 ,

The'following' summarizes key data,-; assumptions,,-and results'from Southwest Research' Institute.'s evaluation of tha flaw indications,'as originally sized, in the Point' Beach Nuclear Plant,

Unit.1' reactor vessel outlet' nozzle-to-shell. welds:

. l' '. Initial Flaw Characterization:.

Flaw #2* . Flaw-#4 (a) Location: "A" outlet' nozzle-to'-shell weld. "C" outlet nozzle-to-shell weld.

Nozzle azimuth: 104' -Nozzle. azimuth: 3118-Flaw center: '7" deep in weld- Flaw center: 5" deep in' weld.

.(b)- Parameters: 2a ~ = 1.24" >

2a '= l.14" -

s 1 .='l.935" ,

1 = 1.-29" a/l = 0.32 a/1 = 0.'44 (c) ASME Section XI:

a/t% . a/t%

Allowable Measured Allowable Measured 4.3 6.8 3.8 6.2 where-t = 9.125"

• In our March 12, 1984 letter in paragraph 1.a. (1) , the cha.racterization of the proximity indication in-the "A" outlet nozzle contained two clerical errors. It in correctly characterized as follows:

2a = 0.2" l =-0.2" Area < l-2/3% of the larger indication It remains an acceptable indication per the limits of Table IWB-3512.

m' .g *c ^

~

• w' = - * '* * ^

g:

1 s

' zt >

< , , ;Q ;,' q '

~ mm . ^ 9b

" ~

S'

\ i

. ~ .

• ~ --

_ * ;g

~.

"  ; ~

~: '

w

o. ,

w

• 12. iWeld. Material 1Datia:= , < .

hThe' . nozzle-to-shd11 0welds ' are ;primarily; welded .with : Mn-Mo-Ni: wire HT L #8T1554B an'd i

LindeF#80 flux. Lot 8479.;_'The;weldLchemistryi(Wt  %) - is . as . follows :'- .

C^ ~Mn- 7P S Si -Cr Ni' Mo Cu

!0.014 0.40-0.08 1.58; IOLO12 ;0.45- 10. 07. 0. 6 0 -- 0.19 ThefinitiallRTNDT"and Charpylupperishelf,energyrare taken to be 25*F and 65=ft-lbs respectively.:- The.end-of-life fluence.at the. nozzles'is predicted to_be 1.65 x'1016-ln/cm2~. "

This1fluencesvalue is"less than the minimum values used inxRegulatory Guide.1.99.to ~

predict irradiated l material property shifts; therefore :the~unirradiated;RTNDT and Charpy -

u upper shelf energy values.willzbe:used in;the analysis (., Forapurposes:of this-nozzle analysisLonly, the:Charpy. upper. shelf energy, 65 ft-lbs,. has'b_een conservatively: correlated

~to an' irradiated upper shelf. fracture toughness of 149-ksi-din. Thisivalue, K lc = Kla =-

149 ksidin --was used;to provide aiconservative= upper bound on!the ASME' Code Kla-and Klc curves. . Note,that.this conservative' upper shelf fracture toughness was established to expedite completion of the analysis.. We believe a.significantly) higher-v'alue.can be demonstrated butLhave~not chosen to pursue'furtherirefinement=since the current-analysis is, satisfactory with.the conservative value. Hence, this' conservatism?is unique-and

'applicableLonly-to.this analysis.

3< Fatigue Crack' Growth' Analysis:

Fatigue growth'of the flaw indications was analyzed per ASME,Section XI, Appendix A. ,

It was . calculated ~ that the. present : crack' depth of Flaw #2 in the "A" outlet nozzle 2 x;ai =~2 x 0.62 = 1.24" fcould grow to accrack depth 2 x af ='2 x-0.625 = 1.25" by end-offlife. Since'the predicted growth of Flaw #2 is so small,' interaction with

.theLsatellite (proximity) flaw is.not expected to occur. Cumulative crack growth for

.. Flaw #4;in the "C"'ontlet. nozzle was calculated and is insignificant. The final crack size af is used"in the flaw. integrity evaluations.

4. Flaw; Integrity Assessment:.

The acceptance criterion for normal, upset, and test conditions (Levels I and II) is given'by:

Kla/K I>h10L

,~ .

~

g

..jj .

- q ;.

'The. maximum: stresses are ob'tained for both. Flaw #2 and. Flaw #4 during'the cold-hydro test at 3,125 psia. .The acceptance criteria are satisfied for both flaw indications, hence.they.are acceptable.

-Plaw #2 -Flaw #4 Ka

_l = ' 149 =

3.43>r[bi Ka l

149

3.92 > d10 Ki- 43.5 K1 38.02 The acceptance criterion for a flaw to withstand crack initiation under emergency and faulted (Levels III and IV) conditions is given by:

Klc/K1. > di Three accident conditions were considered: loss-of iccol' ant accident; large steam line break (with and without off-site power); and locked. rotor pressure transient-(loss of load).

For'all transients considered it was noted that those with high pressure provided the worst case. It was also noted-that in-the cooldown. transient cases the thermal stresses tended to be compressive at the crack location and, thus, lessen the contribution of the pressure stress. The cold hydro' test condition stresses, therefore, bounded the stresses imposed by the accident conditions.- Hence, the flaws are'still acceptable.

,