ML20073Q409

Forwards Info Re Steam Generator Tube Repair,Requested in Aslab 830322 Order.Westinghouse Proprietary Info Withheld (Ref 10CFR2.790)

ML20073Q409
Person / Time
Site:	Point Beach
Issue date:	04/27/1983
From:	Churchill B SHAW, PITTMAN, POTTS & TROWBRIDGE, WISCONSIN ELECTRIC POWER CO.
To:	Gotchy R, Johnson W, Moore T NRC ATOMIC SAFETY & LICENSING APPEAL PANEL (ASLAP)
References
ISSUANCES-OLA-2, TAC-48752, NUDOCS 8304290099
Download: ML20073Q409 (51)

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wR<TER s DIRECT DIAL sevaeSER (202) 822-1051 Thomas G. Mccre Dr. W. Reed Johnscn Administrative Judge Administrative Judge Chairman, Atomic Safety and Atomic Safety and Licensing Licensing Appeal Board Appeal Board U.S. Nuclear Regulatory

.U.S. Nuclear Regulatory Commission Commission Washington, D.C.

20555 Washington, D.C.

20555 Dr. Reginald L. Gotchy Administrative Judge Atomic Safety and Licensing Appeal Board U.S. Nuclear Regulatory Commission Washington, D.C.

20555 In the Matter of Wisconsin Electric Power Company (Point Beach Nuclear Plant, Unit 1)

Docket No. 50-266 (OLA-2)

Dear Administrative Judges:

On behalf of Wisconsin Electric Power Company

(" Licensee".),

I am herewith submitting the information requested in the Appeal Board's Order of March 22, 1983.

The information is simultaneously being submitted to the NRC Staff, at its request, in conjunction with its independent review of Licensee's application for steam generator repair at Point Beach Nuclear Plant, Unit 1.

8304290099 830427

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PDR ADOCK 05000266 o

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i SHAw, PITTMAN, PoTTs & TROWORIDGE' I

Atomic Safety and Licensing-Appeal Board Page~Two April 27, 1983 Certain of the enclosed information is proprietary to Westinghouse Electric Corporation, and the NRC Staff has been requested to withhold it from public disclosure pursuant to 10 C.F.R. S 2.790.

The Service List distribution will include nonproprietary versions of the information submitted to the Appeal Board.

While Licensee is not objecting to providing the requested information to the Appeal Beard, the request raises'the interest-ing question of whether the Appeal Board has jurisdiction to

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exercise sua sponte review of the merits of this particular application.

In ALAB-719, the Appeal. Board

• affirmed the dis-nissal by the Licensing Board of the sole petition for leave to intervene and request for a hearing on the repair amendment.

The Commission, through issuance of its notice of cpportunity i

for a hearing, 47 Fed. Reg. 30,125 (July 12, 1982) did not find I

of its own accord that a hearing would be required in the public interest, 10 C.F.R. 5 2.104 (a), but rather offered an opportunity for members of the public to request a hearing, 10 C.F.R. S 2.105 (a) and (d).

The denial of the petition, affirmed by ALAE-719 as a result of an appeal by petitioner pursuant to 10 C.F.R. S 2.714a, l

returns the application to the status of one as to which there is no petition for leave to intervene.

Since there is no petition for leave to intervene which has been granted or is pending before the Commission, this is no longer a contested proceeding.

10 C.F. R. S 2. 4 (n).

(The peti-tioner has recently petitioned for discretionary Commission review of ALAB-719, but it is the request for review of the denial of the intervention petition, not the intervention petition.itself, which is pending before the Commission.)

Be-cause aus appeal under the narrow provisions of secition 2.714a does not encompass substantive determinations on the merits of an application, and because we believe Commission regulations do not otherwise provide Appeal Board jurisdiction over uncon-tested proceedings involving operating license amendments, there is a question of whether, and the extent to which, the Appeal Board's sua sponte jurisdiction pursuant to 10 C.F.R. S 2.785 (b) (2) is applicable in this case.

Licensee is not now raquesting a ruling on this question.

.We point it out only to assure that, by submitting the requested information, Licensee does not waive its rights to raise the jurisdictional question.

f ctulysubmitped, R

db.

b k)U, ruc Churchill Delissa A. Ridgway Counsel for Licensee

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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Appeal Board In the Matter.of.

)

)

WISCONSIN ELECTRIC l POWER COMPANY

)

Docket No.-50-266 (OLA-2)

)

(Point Beach Nuclear Plant,

)

Unit 1)

)

CERTIFICATE OF SERVICE i

This is to certify that copies of " Licensee's Responses to Questions in the March 22, 1983 Appeal Board Order" were served, by deposit in the United States mail, first class, postage prepaid, to all those on the attached Service' List, this.27th day of April, 1983.

[

a 8Q Bruce W. Churchill Dated:

April 27, 1983 4

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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION

'Before the Atomic Safety and Licensing Acceal Board In the Matter of

)

)

WISCONSIN ELECTRIC POWER COMPANY

)

Docket No. 50-266-OLA2

)

(Point Beach Nuclear Plant,

)

Unit 1)-

)

SERVICE LIST Thomas S. Moore, Chairman

-Atomic Safety and Licensing Board Atomic Safety and Licensing Fanel Appes.1 Board U.

S. Nuclear Regulatory Ccmmission U.

S. Nuclear Regulatory Washington, D. C.

205E5 Commission Washington, L. C.

10555 Atemic Safety and Licensing Appeal Board Panel Dr. W. Reed Johnson U.

S. Nuclear Regulato:y Commission Atomic Safety and Licensing Washington, D. C.

20555 Appeal Board -

U.-S. Nuclear Regulatory Docketing and' Service Section Commission Office of the Secretary Washington, D. C.

20555

'U. S. Nuclear Regulatory Commission Washington, D. C.

20555 Dr. Reginald L. Gotchy Atomic Safety'and Licensing Stuart A. Treby, Esquire Appeal Board Office of the. Executive Legal Directos U.

S. Nuclear Regulatory U. S. Nuclear Regulatory Commission Commission Washington, D. C.

20555 Washington, D. C.

20555 Richard ~G.

Bachmann, Esquire PGter B.

Bloch, Chairman Office of the Executive Legal D2. rectos Atomic Safety and Li. censing U. S. Nuclear Regulatory Commission Board Washington, D. C.

20555 U.

S. Nuclear Regulatory Commission Myron Karman, Esquire Washington, D. C.

20555 Office of the Executive Legal Directos U. S. Nuclear Regulatory Commission Dr. Hugh C. Paxton Washingtori, D. C.

20555 1229 - 41st Street Los Alamos, New Mexico -87544 Peter Anderson Wisconsin'.s Environmental Decade Dr. Jerry R.

Kline 114 North Carroll Street Atomic Safety and Licensing suite 208 Board Madison, Wisconsin 53703 U.

S. Nuclear Regulatory Commission Washington, D.

C.

20555

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bHsconsin Electnc mmcwmr 231 W. MICHIGAN, P.O. BOX 2046, MILWAUKEE, WI 53201 April 27, 1983 Mr. H. R. Denton, Director Office of Nuclear Reactor _ Regulation U. S. NUCLEAR REGULATORY COMMISSION Washington, D. C.

20555 Attention:

Mr. R. A. Clark, Chief Operating Reactors Branch 3 Gentlemen:

DOCKET NO.30-266

. Mu GENIRATOR KEPAIK

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POIITBEACH NUCLEAR PfJd4T, UNIT 1 By Order dated March 22, 1983, the Atomic Safety and Licensing Appeal Board requested certain information related to the repair of the Unit 1 steam generators.

That information has u

also been requested by the NRC Staff and is enclosed herewith.

The enclosed information contains certain information which is proprietary to Westinghouse Electric Corporation.

The proprietary information has been identified by brackets.

In conformance with the requirements of 10 CFR Section 2.790 of the commission's regulations, we are requesting withholding this proprietary material from public disclosure and enclose an affidavit from Westinghouse Electric Corporation in support of that application.

The affidavit sets forth the basis on which the information should be withheld from public disclosure.

l Very truly yours, t

e l

C. W. Fay Vice President-Nuclear Power i

e Enclosure cc:

Service List NRC Resident Inspector i

Westinghouse Water Reactor Ba32

""8 N'mi'" "

Electric Corporation Divisions April 27, 1983 CAW-83-35 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Phillip-Building 7920 Norfolk Avenue Bethesda, MD 20014

Reference:

Wisconsin Elactric Power Company Letter (C. W. Fay to H. P.. Denton), dated April 27, 1983

" Docket 50-266 Point Beach Nuclear Plent Unit i Steam Generator Repair

Dear Mr. Denton:

The proprietary material for which withholding is being requested by the Wisconsin Electric Power Company is further identified in an affidavit signed by the owner of the proprietary information, Westinghouse Electric Corpora-tion.

the affidavit, which accompanies this letter sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses specifically the considerations listed in paragraph (b)(4) of 10CFR Section 2.790 of the Coasussion's regulations.

Accordingly, this letter authorizes the utilization of the accompanying affi-davit in support of the Wisconsin Electric Power Company.

Correspondence with respect to the proprietary aspects of the application for withholding or the Westinghouse affidavit should reference this letter, CAW-83-35, and should be addressed to the undersigned.

Very truly yours, l

LU4W4 W' R. A. Wiesemann, Manager Regulatory and Legislative Affairs

/wpc cc:

E. C. Shomaker, Esq.

l Office of the Executive Legal Director, NRC O

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CAW-83-35 AFFIDAVIT COMMONEALTH OF PENNSYLVANIA:

ss COUNTY OF ALLEGHENY:

Before me, the undersigned authority, personally appeared Robert A. Wiesemann, who, being by me duly sworn according to law, deposes and says that he is authorized to execute this Affidavit on behalf of Westinghouse Electric Corporation (" Westinghouse") and that the averments of fact ret forth in this Affidavit are true and correct to the best of his knowledge, inforaiation, and belief:

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I Robert A. Wiesemann, Manager Regulatory and Legislative Affairs Sworn to and subscribed

.before me this JJ day of MN 1983.

v MEV1&E KISH, NOTAMT ruam WR0puSB080. ALLEGHENY COUNTY W C0WWIS$10N EXP!at? SEPT. 3.1984 Member, Pennsylvania Auxation of Notants,,

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,. CAW-83-35 l

(1)

I am Manager, Regulatory and Legislative Affairs, in the Nuclear Tech-nology Division, of Westinghouse Electric Corporation and as sech, I have been specifically delegated the function of reviewing the proprietary

-information sought to be w.ithheld from public disclosure in connection

(

with nuclear power plant' licensing or rule-making proceedings, and am authorized to apply for its withholding on behalf of the Westinghouse Water Reactor Divisions.

(2)~ I as making this Affidavit in conformance with the provisions of 10CFR

-Section 2.790 of the Cosmiission's regulations and in conjunction with the Westinghouse application for withholding accompanying this Affidavit.

l (3)

I hcVe personal knowledge of the criteria and procedures utilized by 3

Westinghouse Nuclear Energy Systems in designating information as a trade sec;et, privileged or as cor.fidential commercial or financial information.

(4)

Pursuant to the provisions of paragraph (b)(4) of Section 2.790 of the Cosmiission's regulations, the following is furnished for consideration by the Cosmiission in determining whether the information sought to be with-i held from public disclosure should be withheld.

(1)

' The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse.

J (ii)

The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public.

Westinghouse has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when the whether to hole certain types of information in confidence. The application of that system and the substance of that system constitutes Westinghouse policy and provides the rational basis required.

Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:

e

' CAW-83-35 (a) The information reveals the distinguishing aspects of a pro-cess (or component, structure, tool, method, etc.) where prevention of its use by any of Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.

(b)

It consists of supporting data, including test data, relative to a, process (or component, structure, tool, method, etc.),

the application of which data secures a competitive economic advantage, e.g., by optimization or improved marketability.

(c)

Its use by a competitor would reduce expenditure of resources or improve his competitive position in the design, manufac-ture, shipment, insta11atien, assurance of quality, or licensing a similar product.

l (d)

It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or supp.11ers.

(e)

It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of poten-tial consercial value to Westinghouse.

(f)

It contains patentable ideas, for which patent protection may be desirable.

l (g)

It is not the property.of Westinghouse, but must be treated as proprietary by Westinghouse according to agreements with the owner.

l There are sound policy reasons behind the Westinghouse system which include the following:

(a) The use of information by Westinghouse gives Westinghouse a competitive advantage over its competitors.' It is, there-fare, withheld from disclosure to protect the Westinghouse

L' CAW-83-35 (b) -It is information which is marketable in many ways.

The extent to which such information is available to competitors diminishes the Westinghouse ability to se,11 products and j

services involving the use of the information.

(c)

Use by our competitor would put Westinghouse at a competitive

' disadvantage by reducing his expenditure of resources at our expense.

i (d) Esch component of proprietary information pertinent to a i

particular competitive advantage is potentially as valuable as the total competitive adycntage.

If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westing-house of a competitive advantage.

(e)

Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the competition in those countries.

(f)

The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtain-ing and maintaining a competitive advantage.

j.

(iii) The information is being transmitted to the Commission in confi-dance and, under the provisions of 10CFR Section 2.790, it is to be received in confidence by the Commission.

(iv)

The information sought to be protected is not available in public sources to the best of our knowledge and belief.

(v)

The proprietary information sought to be withheld in this submit-tal is that which is marked in the proprietary version of the document entitled, " Docket 50-266 Point Beach Nuclear Plant Unit l' Steam Generator Repair" from the non-proprietary version of the same report.

i

5-CAW-83-35 This information provides details of equipment design and compre-hensive plant _ data that were developed at significant expense.

This information has substantial comercial value to Westinghouse in connection with competition with other vendors for-service contracts and performance evaluations.

The subject information could only be duplicated by competitors if they were to invest time and effort equivalent to that invested by Westinghouse provided they have the requisite talent and experience.

Public disclosure of this information is likely to cause substan-tial harm to the competitive positfon of Westinghouse because it wou?d simplify design and evaluation tacks without requiring a comensurate investment of time and effort.

Further the dcponent sayeth not, I

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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION Before the Atomic Safety and Licensing Appeal Board

)

In the Matter of

)

)

WISCONSIN ELECTRIC POWER COMPANY

)

Docket No. 50-266 OLA-2

)

(Point Beach Nuclear Plant, Unit 1))

)

LICENSEE'S LESPONSES TO QUESTIONS IN THE MARCH 22, 1983 APPEAL BOARD ORDER QUESTION 1.

The use of " hydraulic" tube expansion in the new steam generators provides for contact between the tube and tubesheet along the full length of the tubesheet holes, and eliminates the crevice between the tube and tubesheet.

This result ostensibly lessens the potential for corrosion.

But, this new design also subjects the tube to several stresses that appear to' converge at the location where the tube emerges from, and is anchored by, the tubesheet.

These stresses are:

(1) residual stress resulting from the expansion process; (2) cyclic stress associated with thermal expansion and contraction during normal operation; (3) hign frequency cyclic stress due to flow-induced I

vibrations of the tubes; and (4) stress resulting from accident loadings (e.g.,

LOCA, steamline break, SSE). The top of the tubesheet also appears to be where the most corrosive environment on the secondary side may be found as a result of sludge deposits collecting there.

l In view of these factors, we request applicant to provide the following information:

1.a) A description of the analyses which lead to the conclusion that tubes of the new, fully expanded tube design steam generator are adequate to withstand the concentration of loading and fatigue at the top of the tubesheet.

RESPONSE

For the Point Beach replacement steam generators, the tubes will be secured to the tubesheet by a hydraulic tube expan-.

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e-sion process extending the full length of the tubesheet.

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purpose of'the hydraulic expansion proces's is-to minimize the crevice'between thn tube and the tubesheet.

An evaluation of tube' stresses at the expansion transition at the top of the tubesheet for transient / normal operation loadings has been performed to verify that the maximum stress intensity range and 4.

cumulative fatigue usage factor were less than the ASME Boiler and Pressure' Vessel Code allowables.

The stresses considered in the AIME Code stress analysis were: 1) normal operating stresses, 2) cyclic stress associated with t.hermal expansion and contraction of the tube during normal' operation, 3) high frequency _ cyclic stress due to flow induced vibration of tubes, and 4) stress resulting from accidental loading ( M., LOCA, steamline break, SSE).

1 Residual stresses are not specifically required to be considered in the Code analysis.

These stresses are recognized by the code wMre they are relevant, such as in fatigue analysis where the cyclic stress range allowables specified by the Code include an allowance for the presence of a mean stress.in addi-tion to cyclic stresses.

For the Code design analysis, residual stresses are not required to be considered because these stresses i

do not affect the load carrying capability of the component.

However, the effects of residual stresses are addressed in cor-rosion testing programs described in the response to Question 1.b).

The transients analyzed for units similar to the Point Beach replacement steam generators are summarized in Table 1.a-l.

Flow induced vibration, while included in tha 1

analysis, is not included.in Table 1.a-1 because tube-stresses resulting from thisl load are negligible.

For each of the lo'ading' conditions, the potential interaction of the flow distribution baffle with the tubesheet is considered.

This interaction-includes the-relatiJe tubesheet-to-flow distribu-tion baffle hole alignment due to radial thermal growth and 1

tubesheet rotation.

The data provided in Table 1.a-1 are for-4 Model 44F replaceuent steam generatars and the results and conclusions are applicable to the Point Beach replacement steam l

generators.

l Several conservative assumptions were made in the tube stress analysis.

These assumptions were:

1.

Thermal tube stresses are evaluated based on the tem-perature difference between the primary inlet temperature, 4

Thot,.and the subcooled secondary side fluid temperature with no allowance for effects of heat transfer.

2.

The pressure stresses in the tube are calculated using thin wall cylinder equatior.s.

Both the pressure I

stresses and stresses due to the tubesheet/ flow baffle differential growth are calculated for a tube in a thinned condition.

3.

A stress concentration factor is used in the fatigue evaluation.

This factor accounts for any scr&tches or marks which may be present on the tubes' surface.

All the mentioned stresses were combined for each of the transient / operation loadings and the muimum stress intensity was evaluated - results are summarized in Table 1.a-2 for Model 44F replacement steam generators. '

The maximum primary plus secondary stress intensity range occurred between the Lors of Load and the Secondary Leak Test Transients and-was equal to [

] a, c, e KSI.

The maximum

. stress i:itensity range allowed by the ASME Boiler and' Pressure Vessel Code is equal to 79.8 KSI.

Finally, the results of the fatigue evaluation for various transient combinations are shown in Tsble 1.a-3.

Based on anr. lysis, the cumulative fatigue usage factor was calculated to be [

] a, c, e - considerably below i

the 1.0 allowable factgr.

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TABLE 1.a-1 (TRANSIENT

SUMMARY

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PODEL 44F TRANSIENT CYCLES 4,c.e

1) HEATUP/COOLDOWN

2) PLANT LOADING / UNLOADING

3) STEP LOAD DECREASE

4) STEP LOAD INCREASE

5) REACTOR TRIP

6) 40% STEP LOAD DECREASE

7) LOSS OF FLOW

8) LOSS OF LOAD

9) PRIMARY HYDR 0 10)

PRIMARY PRESSURE TEST

11) SECONDARY HYDRO

12) SECONDARY PRESSURE TEST

13) PRIMARY LEAK TEST 1.4) SECONDARY LEAK TEST

15) HOT STANDBY

16) STEADY STATE FLUCTUATIONS 17)' POWER BLACKOUT I

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TABLE 1.a-2 PRIMARY PLUS SECONDARY STRESSES, KSI-HOT LEG SIDE AT THE TOP OF TU3ESHEET MODEL 44F (I I) (dTI d)_ (fA d)

A R

TRANSIENT 4,c,e

1) HEATUP/COOLDOWN

2) PLANTJLOADING/ UNLOADING

3) STEP' LOAD DECREASE 4).ST5PLOADIbREASE

5) REACTOR TRIP

6) ~ 405STEPLOADDECREASE

7) LOSS OF FLOW

8) 1.055 0F LOAD 9)' PRIMARY HYDRO 10), PRIMARY PRESSURE TEST 11)..SECONDARYHYDR0 i2)SECONDARYPRESSURETEST

13) PRIMARY LEAK TEST

14) SECONDARY LEAK TEST l

15) HOT STANDBY

16) STEADY STATE FLUCTUATIONS

17) POWER BLACK 0UT o

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TABLE 1.a-3

. CUMULATIVE FATIGUE USAGE FACTOR - HOT LEG SIDE AT THE TOP OF THE TUBESHEET

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MODEL 44F TRANSIENT lt n

N n

COMBINATION (ksi)

(Cycles)

(Cycles)

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QUESTION 1.-

b).

A description of the analyses and/or tests'that establish the adequacy of-the in-service corrosion resistance of the tubes in-the regions of high stress at.the top of the tubesheet.

RESPONSE

A summary of the development of the hydraulic expansion process, residual stresses resulting from the hydraulic expansion, and the corrosion tests conducted to verify the improved corrosion resistance of thermally treated Inconel 600 tubing is provided in r

Section 2.2.1.3 of the' Point Beach Nuclear Plant' Unit No. 1 Steam Generator Repair Report, August 1982, Amendment 1 (" Repair Report").

These tests and analyses demonstrate the adequacy of the tube expan-sion process and the corrosion resistance of the thermally treated Inconel 600.

Strosses which could occur during accident conditions are of short duratien relative to stresses which are present during normal operation.

Thus, stresses due to accident condi-j.

tions are not considered directly in corrosion testing programs since they would not be present long enough to affect the cor-rosion process.

Stresses resulting from normal operation, f

including residual stresses due to tube expansion, have been addressed in the corrosion testing programs.

The stress cor-rosion tests of hydraulically expanded specimens subjected the tubing to the maximum tensile stresses representative of normal operation.

Additioncl confirmation was obtained from plastically stressed C-ring specimens.

A detailed description of tests conducted to establish 1

the basis for selection of a tube-to-tubesheet expansion process 1

I is provided in Appendix A.

These tests demonstrate that the hydraulic expansion process results in the lowest residual stresses of the available tube expansion processes.

This is

' due,to'the minimal metal deformation and the gentle transition contours'in the expansion transition compared to conventional L

mechanical rolling expansion processes.

Figure 3 of Appendix A l

provides. profiles of a typical hydraulic expansion transition.

Figure 5 of Appendix A provides the results of tests conducted to determine residual stresses resulting from various expansion processes.- These results demonstrate that stresses resulting from hydraulic expansion are only in the order of one-half those from a mechanical rolling process.

The special thermal treatment of'Inconel 600 was specif-1 ically developed to enhance the stress corrosion cracking resis-i tance of the tubing to both primary (" pure" water) and secondary side contaminated environments (caustic and sulphates).

An exten-sive laboratory test program conducted by Westinghouse and others-over the past six to eight years has demonstrated the benefits of this thermal treatment.(1)

A combination of extensive cold work and high temperature is used during testing to accelerate any cracking tendency and reduce the time to crack to several weeks or months compared to many years (if at all) at the 603*F maximum primary water j

temperature of Point Beach.

Accelerating the time to initiate cracking in laboratory specimens by utilizing several high test (1) [

j &, c, e

_.a.._

T temperatures, and then extrapolating the time to cracking at lower temperatures, is an accepted _ technique provided the cracking mechanism remains-the same.

Mill-annealed Inconel 600 may be susceptible to stress corrosion under certain conditions of extensive cold work (very high stress) in~ pure water or primary coolant containing Li, B, and H.

When extensively strained by 2

cutting the tubing longitudinally and reverse bending over.a-small diameter mandrel, U-bends' of mill annealed tubing can be made-to crack reproducibly in 680*F pure or primary coolant water.

When the tubing was thermally treated and ?. hen reverse bent and tested in this manner, no cracking was ' observed in any heat of tubing afP.er extended exposure.(1)

.Since there was no cracking in the thermally treated specimens at high temperature,

/

it was not possible to determine the time to crack at the lower

(

- operating temperatures.

It is expected however, that the

. hydraulically expanded transitions, being stressed to a much less degree than the reverse U-bends, will not be susceptible to primary water stress corrosion cracking in operation, based upon the laboratory tests demonstrating the effects of the thermal treatment in providing primary water stress corrosion

\\

cracking resistance under conditions more severe than those during operation.

On the secondary side, the normal environment is all volatile treatment (AVT); however, the ingress of contaminantti, (1).It is possible to cause thermally treated Inconel 600 to crack at high temperatures in pure er primary coolant water, but it requires a.iegree of cold work substantially in excess of that used in the laboratory tests.

. l--_

k

. possibly caustic forming, may occur to varying degrees.

The thermally treated Inconel 600 has also been demonstrated in extensive testing to have additional resistance to caustic stress corrosion cracking,-particularly in the highly stressed condition.

Figure 1.b-1 presents the results of one of a j

series offlaboratory experiments comparing the caustic cracking rate _of mill annealed and thermally treated Inconel 600 as a function of temperature and stress.

The environment selected, f

10% caustic, is believed to be one of the more agressive'envir-onments which theoretically could form beneath sludge deposits on the secondary side.

As shown-in Figure 1.b-1, the therrally treated tubing is essentially unaffected at the T Perating Hot temperature of 603'F, even in the overstressed conditions of

/

[

-]alc,e of yield stressi Additional resistance of therm-ally treated over mill annealed material (as used originally in 1

Point Beach Unit 1) is reproducible over a number of tubing heats and laboratory test conditions.

The tests described'above relate to standardized laboratory specimens, selected to accelerate the potential for attack.

Additional tests have been performed on actual transi-tion zones of hydraulically expanded thermally treated tubing.

Production. tubing was expanded into simulated tubesheets, internally pressurized to [

] a, c, e psi hoop stress (representative of the maximum tensile stress to which the tubes are subjected in normal operation) and immersed in 10%

caustic at 600'F and 650'F, such that the solution contacted the OD of the tubing.

Several mechanically expanded and unex-i -

panded specimens (both thermally treated and mill annealed) were included as controls.

At 600*F, the thernally treated specimens-did not crack after about one year of exposure, whereas the two unexpanded mill annealed specimens cracked after 46 days.

Table 1.b-1 presents the detailed test data for the 600*F tests.

At 650*F, the higher temperature produced cracking in the containers holding the test specimens, often resulting in premature termination of the tests.

Even under these conditions, the superiority of the thermally treated and hydraulically expanded specimens was confirmed.

The conclusions from these tests are (1) thermally treated Inconel 600 provides greater resistance to both primary and secondary environments compared to mill-annealed Inconel 600,.

and (2) cold working such as that experienced in the hydraulic expansion process does not negate the benefits of the thermal treatment.

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TABLE 1.b-1 600*F Test Results Description

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QUESTION 1.

c)

An explanation of the extent, if any, that the tube expansion process (i.e.,

cold working) alters the

" corrosion resistant metallurgical structure: of heat treated Inconel 600 (see Repair Report, Section 2.2.1.4).

RESPONSE

As stated in Section 2.2.1.4 of the Repair Report, the increased corrosion resistance of the' thermally treated Inconel 600 is associated with grain boundary precipitate morphology.

The effect of cold working on the metallurgical structure of Inconel 600 would be to elongate the metal grains.

The grain boundary precipitates would not be affected by this process.

Thus, the corrosion resistant metallurgical structure should not be affected except for the residual stresses associated with cold working.

Tests and analyses described in the response to Question 1.b) have demonstrated that stresses from hydraulic expansion are accept.sbly low and that the additional corrosion resistance of thermally treated Inconel 600 is retained.

.=

r_____.-

QUESTION l.

d)

(The replacement steam generators incorporate two design features to reduce the buildup of sludge on th'e tubesheet:

a flow distribution baffle and an improved blowdown system [see Repair Report, Sections 2.2.1.1, 2.2.1.2].)

A discussion of the data relating to sludge buildup that have been obtained from the operation of steam generators incorporating these features.

RESPONSE

Replacement steam generators utilizing the flow dis-tribution baffle and the improved blowdown system have been installed in Surry Units 1 and 2 and in Turkey Point' Unit 3.

Visual examination by fibreoptics and sludge measurement techniques at Surry Unit 2 have shown no significant sludge accumulation after approximately 24 months of operation.

Data from Surry Unit 1 and Turkey Point Unit 3 are not available presently.

Although field verificgtion of the flow distribution baffle and modified blowdown system is limited, the correlation of sludge buildup on the tubesheet with lateral flow velocity has been verified for steam generators without a flow distribu-tion baffle - see Figure 1.d-1.

Based on the computer analysis model, CHARM, low flow velocities (<,[

] a, c, e ft/sec) are predicted off center of l

the tube lane, i.e.,

away from the blowdown intake.

Furthermore, the measured sludge profile height has been correlated with low tube gap velocities - e.pparently, below [

] a, c, e ft/sec sludge particles may settle.

Therefore, to minimize the number l

of tubes exposed to a low crossflow velocity, a flow distribu-l tion baffle and a modified blowdown system have been incorporated 1

1 into the Point Beach replacement steam generators.

The hydrau-lics at the tubesheet with a flow distribution baffle are illustrated in Figure 1.d.-2.

Based on the computer code analysis, the flow distribution baffle has been decigned with

- the objective of limiting the number of tubes exposed to a sludge settling environment and to limit low crossflow veloci-ties to the center of the tube bundle near the blowdown system.

The correlation of sludge buildup on the tubesheet with lateral flow velocity has also been experimentally verified using a flow visualization model at the Carnegie-Mellon University (see Figure 1.d-3).

The Carnegie-Mellon flow visualization model was composed of 120 tubes in both the hot and cold legs arranged in a 4 x 30 array.

The model included a tubesheet, wrapper wall, and a single tube support plate.

The model did not include a flow distribution baffle.

Sludge particle deposi-tion was simulated ~using particulate material in a working fluid of Refrigerant 113.

This test also confirmed the corre-l lation of measured sludge height with low tube gap velocities.

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QUESTION l '.

e) discussion of the adequacy of eddy current testing, or any other means of in-service inspection, for detecting and assessing steam generator tube degrada-tion taking place in the r?gion where the fully expanded tube emerges from tile tubesheet.

RESPONSE

The routine inspection of steam generator tubing is carried out by using a standard bobbin type probe and a multi-frequency eddy current system which uses multi-parameter tech-niques for minimizing interference signals.

This system has adequate sensitivity for detecting and estimating tube degrada-

~

tion in regions not involving any significant tube deformation.

However, inspection of the tubes in the hydraulic expansion transition region at the-tubesheet-interface requires the reduction of the interference signals from the expansion transi-tion, tubesheet edge and possible sludge. The signals from the tubesheet edge and the sludge are routinely minimized in the l

field by using the multi-frequency eddy current technique for i

signal processing while using the standard bobbin type eddy current probe.

This technique provides adequate sensitivity to detect tube degradation near the top of the tubesheet and at

)

a tube support plate in the absence of significant tube deform-l ation. At the position of the tube expansion transition, the detectability of tube degradation is reduced, since the' standard bobbin type probe is very sensitive to any changes in the inside diameter of the tube.

A change in the inside diameter

.results in a large signal which could interfere with a signal i

from the tube degradation.

However, the standard bobbin type i -

2..._,.

....v..

_. ~-. ~ --

. probe can be used to establish a baseline. signature of the hydraulically expanded region.of the tube.which can then be compared to eddy current signals from subsequent inspections.

It is expected that_ steam generator-. tube degradation in the

' hydraulically expanded region would result in a change in the baseline signature.

If a change from the baseline signature of the expansion transition is observed, this region can be inspected with specialized probes.

There are at least four other probe config-

~

urations capable of enhancing the detection of tube degradation i

in expansion transitions.

One such probe is the cross-wound coil probe.

This probe is insensitive to discontinuities with 360-degrees symmetry, thus minimizing the interference signals from the expansion transition and also the tubesheet edge since they botii possess the 360 degree symmetry.

In the event that the expansion transitions do not possess exact 360 degree

-symmetry, the eddy current signals using the cross-wound coil

' probe at two different frequencies are processed using multi-frequency, multi-parametric techniques for minimizing the remaining signal from the expansion transition.

This technique results in sensitivity for tube degradation at the expansion transition such that a 20% through wall flat bottom drill hole, equivalent in volume to ASME standards, can be detected.

This system is currently being used with success for the inspection of the expansion transition regions of sleeved tubes in steam generators which.have similar hydraulic expansion transitions.

l

A probe. consisting of multiple coils riding along the inside surface of the tube can also be used for the inspection

~

of.such regions.-

Since the coils ride along the tube surface, they are insensitive to the deformation of the tube wall and thus produce clear signals from any significant degradation of-the tubewall.

Such probes have also been used with success for inspecting such regions in the -field.

Another system which is available for the inspection of the expansion transitions uses a three-phase oscillator which drives-three coils, placed 120 degrees apart.

This system also is insensitive to the discontinuities with 360 L

degree symmetry and produces the desired result of minimizing the interference from such factors as the tubesheet edge and the expansion transition.

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f QUESTION 2..

The new steam' generator design has ferritic stainless steel (SA-240 Type 405) tube support plates,. with quatrefoil-type tube passages-(see Repair Report, Section 2.2.1.7 and Figure 2-3).

The method by which the quatrefoil holes are formed is not specified, but their irregular shape suggests that, as a result of the forming process, residual stresses' may exist in the region of these penetrations.

The applicant should describe the analyses and/or tests '

performed that relate to the possibility of stress corrosion in the region'of the quatrefoil penetrations, and indicate whether stresses in the support plates associated with i

normal operation of the new steam generators were included in the analyses or tests.

RESPONSE

Laboratory tests conducted by Westinghouse utilizing l

highly stressed Type 405 stainless steel U-bends exposed to caustic and chloride environments, and heated crevice and model boiler tests utilizing actual broached quatrefoil samples exposed to'the environments which caused tube denting and i

cracking of the carbon steel support plates, as well as litera-ture searches, have verified that Type 405 stainless steel, as fabricated, is not susceptible to stress corrosion cracking in the steam generator operating environment.

The fabrication of the Type 40S stainless steel support plates does not produce significant resiilual stresses.

The plate material is initially strengthened by heat treatment i

and tempered at 1325-1375'F.

While the purpose o'f these heat treatments is to optimize the mech'anical properties and corrosion resistance of the material, the tempering operation also minimizes any residual stresses which may be present in the plate material.

Small holes are then drilled at the required locations for the i

quatrefoil openings.

The quatrefoil openings are produced by -.

broaching; an operation involving multiple shaving, i.e.,

progressively removing small amounts of metal by utilizing a tool with stepped cutting edges, which removes less and less metal with each step.

The residual stresses' caused by the broaching operation have not been analyzed, but are judged, based on the metal removal process, to be low.

Although some general corrosion has been observed at tube support lands in certain a'ccelerated heat transfer corrosion tests, there has been no appearance of stress corrosion cracking indicative of high residual stresses.

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APPENDIX A e

BASIS OF SELECTION OF TUBE /TUBESHEET EXPANSION PROCESS FOR THE MODEL F STEAM GENERATOR 4

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su-ut-nu-vor BASIS OF SELECTION OF 111BE/TUBESHEET EXPANSION PROCESS FOR THE MODEL F STEAM GENERATOR r

INT.0"UCTION Field experience wit!h operating units in which the tubes utre only part-rolled into the tubesheet indicated that full depth expansion may add margin to

. minimize crevice corrosion.

Various processes have been developed for performing fu.11 depth expansion. Processes used by Westinghouse have included mechanical rolling, explosive expansion and hydraulic expansion. Since late 1978, all units manuf actured at Westinghouse have beer. hydraulically exp anded. This has included some Model D's and E's and all Models 44F, 51F, and F.

The purpose of this report is to surtsnarize the more important reasons for salecting the hydraulic method as the reference process.

,6EERAL REQUIREENTS Any viable process must satisfy concerns relating to:

o technical objectives l

o manuf acturability o

quality control It' is the desire - indeed, the need - to optimize the balance in these various l

requirenents which leads to the selection of one process' over another.

Manufacturability and quality control also impact technical objectives. This 1s because each process has its own characteristics which are to be controlled l

so as not to preclude attainment of the technical objectives.

Process selection is based first and forenost on achievement of the technical objectives.

~

There are four (4) technical aspects which are comon to all tube expansion processes. These relate to:

0587c/73c:5 011483 2 1

i.

o crevice depth o

residual stresses o

joint tightness o

tiibe mabrial aroparties Following a brief description of the expansion processes, these four (4) technical aspects will be discussed and conparisons made between the different

• processes.

PROCESS DESCRIPTION The main steps associated with each expansion process are depicted schematically in Fig.1, and include:

o insertion of tubing o

tack expansion l

o

' tube to tuoesheet welding o

leak testing o

full depth expansion o

final quality control (QC) inspection Following the tubing process (insertion of tubes through support plates and l

tubesheet) the tubes are tack expanded for a depth of about 3/4" on the l

primary side to facilitate tube to tubesheet welding. Tack expansion is accomplished with either mechanical rolling or urethane expansion. The tube to tubesheet weld (actually tube to tubesheet cladding) serves as both a seaf weld, preventing leakage between primary and secondary sides, and a strength weld, supporting the service imposed loads. After leak testing of the welds, the tubes are full depth expanded.

1 For mechanical rolling, expansion is incremental - the rollers being about 1" long - and expansion begirs at the primary side. Mechanical rolling induces considerable redundant metal working - the tube is locally worked as one roller then another passes around the circunference and then the b

transition zone at the edge of each step is rolled out by the~ next roll step.

The hydrailic and explosive processes, by contrast, produce a ri.ther uniform I

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Illustration of major steps involved in full depth tube to tubasheet expansion pr'ocess.

[

b radial expansion which occurs simultaneously along the full thickness of the tubesheet. For all expansion processes, the tube length first decreases as-the tube is expanded out to contact with the tubesheet, and then increases with further expansion due to the tube wall thinning. The net effect is that the tube alongates during mechanical relling (the tube wall thi'os about 2 mils) but shortens during hydraulic or explosive expansion (wall thinning is only a fraction of a mil).

Finally, QC inspections (in-process and final) are performed to verify the l

expansion process and to identify abnormal, conditions. Of the four essential technical aspects previously mentioned, and discussed in the following section, QC inspection usually measures only the crevice depth. Tube to tubesheet joint tightness can be measured by ultrasonic techniques but this is L

not an industry-wide practice. Residual stresses, joint tightness and tube material properties ~are assassei by analytical / experimental laboratory process development and qualification programs prior to production usage.

TECHNICA. REQUIREMENTS The expansion processes which will be consider,ed here include:

Tack Expansion o

mechanical rolling o

urethane expansion Full Depth Expansion o

mechanical rolling o

explosive expansion o

hydraulic expansion CREVICE DEPTH 1

Crevice depth, and the complementary measure, depth of ex::ansion, are defined by the sket' h 1.n Fig. 2.

Depth of expansion, hence crevice depth, is c

controlled by the expansion process parameters. For each process, these i

parameters are optimized tc provide proper balance of conpeting effects.

In 0587c/73c:5 011483 4 4

j Depth Of Espansion d.

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Crevice DerN Crevice Depth = Tubesheet Thickness - Depth Of Espansion Fig. 2.

Definitions of " depth of expansion" and

" crevice depth".

o

^

hydraulic expansion, for example, crevice depth is controlled primarily by mandrel length. Too short a mandrel gives a deeper crevice; too long a mandrel could produce overexpansion in the tube just above the tubesheet.

Hence, mandrel length is optimized so as to minin'* re cre.vice depth while simultaneously precluding the occurrence of overe+ansion. A uIique distribution of crevice depths is obtained for each controlled expansion process. A rather large data base consisting of accurate crevice depth measurenants on production units exists for the hydraulic expansion process.

For earlier units expanded by mechanical rolling and explosive expansion, process control was the basis for controlling crevice depth. A subjective assessment is as follows:

Max. - Crevice Mean Crevice Approx. Transition Process Depth, inch Depth, inch Length, inch Mechanical Roll 1/4 1/8 1/8 - 1/4 j

l Explosive expansion 1/2 1/4 1/2 Hydraalic expansion 1/4 1/8 1/4 A typical hydraalic transition zone profile is shown in Fig. 3.

The transition zone length is usually 0.25 to 0.3 inch long.

Crevice depth data for the hydra;11e expansion process follow a Gaussian distribution. Westinghouse experience in expanding more than 1/4 million tubes by the hydraulic expansion process has demonstrated that crevice depth l

can be controlled to give a mean depth of about 1/8 inch and a maximum depth l

1ess than 1/4 inch.

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E RESIDUAL STRESSES Agreement is unanimous within the industry that residual stresses should be minimized because of their potential role in stress corrosion crackin,. This led Westinghouse to develop alternatives to conventional mecha*hical :olling.

Measurement Techniques Various techniques have been used to assess the residual stresses in the transition zone between the expanded and unexpanded portions of the tube, which is the region of prime interest. Strain gage relaxation and X-ray diffraction techniques have given antiguous results - partly because of difficulty in performing the tests, and partly because of inherent difficulties in interpreting the results. Analytical calculations have been made using finite element analyses. This technique is more suitable for hydraulic expansion than mechanical rolling because it is easier to model the' physics of the process for hydraalic expansion. These techniques - strain gages, X-rays and finite element analyses - attenpt to determine absolute values of residual stresses. The most.widely used technique, however, employs I

stress corrosion cracking (SCC) tests which yield primarily qualitative results although the results, in principle, can be quantitative.

I In SCC tests the tubing is first given a sensitization heat treatment which

r. enders its microstructure sensitive to SCC when strassed and exposed to a srteific corrosive environment. Samples are prepared, say mechanically rolled and hydraJ11cally expanded and exposed to the corrosive environment.

In time, the sanples crack; the longer the time the lower the residual stresses. Thus, l

the test provides a relative comparison of the residual stresses from different expansion processes, or for different parameters o'f a given

process, It further identifies.the location of the highest tensile residual stress (point d'ere crack occurs) and the direction of the stress (normal to the orientation of the crack).

The test can be made quantitative by calibration.

For this purpose, specimens of sensitized tubing are stressed to various known applied stress levels (no residual stresses) and tested in the corrosive environment as depicted in I

~

Fig. 4.

The specimens may be uniaxial tensile specimens but more often are

'C-ring specimens loaded to a given outer fiber stress. The results give a cune (called C-ring curve for C-ring specimens) of applied stress versus time

'o crack initiation. Sanples having only residual stresses are then prepared

..id tested. Using the time to cracking, the C-ring or calibration curve is entered at that time to determine the magnitude of the residual stress as depicted in Fig. 4.

Inconel 600 tubing is tested in polythionic acid at room temperature.

However, in order to sensitize the microstructure it is usually necessary to solution anneal the tubing prior to aging and this gives a coarse grain size.

Thus, the sensitized tubing has mechanical properties different fran the properties of regular steam generator tubing, and for this reason the results are only seni-quantitative.

As an alternative to Inconel 600 in polythionic ccid, tests may be run on stainless steel tubing in boiling magnesiun chloride. This is a n. ore difficult test to perform but has the advutage that the stainless steel tubing has a fine grain size. Since the two types of tubing have similar i

mechanical properties, results on stainless steel should be qualitatively I

applicable to evaluation of Inconel 600.

SCC Results The residual stresses are somewhat higher on tha ID than on the OD. The l

relative magnitude of the residual stresses for various expansion processes are compared in Fig. 5.

The comparison uses polythionic acid test results and the C-ring curve but should be viewed only in a relative qualitative sense.

l The residual stresses in the transition zone are highest for hard (high torque) mechanipal rolling and lowest for hydraulic expansion. Though the l

data are limited, the residual stresses for explosive expansion (called WEXTEX j

by Westinghouse, i.e., Westinghouse Explosive Tube Expansion) are low, similar i

to hydra 11e expans' ion.

A comparison of the ID residual stresses at the tack expansion transition is l

(

also shown in Fig. 5 for sof t (low torque) tack rolling and urethane tack 1

O

'(dr Estahlbh eyarimenthcalibrdion curve.

p-

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IIeat treat tubin'g to respond to SCC in corrosive environment.

Corros,we Erwironment.

Place test-specimens in corrosive environment and load to known stress levels.

l A

Determine times to crack initiation.

Plot calibration curve of acclied 7g stress versus time to cracking.

w time tu crack initiation t

(b)

Test corneneni (expandect sample) n_-

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Prepara sample and place in

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corrosive environmenr.

derr65Ivt Env'itertmerd Determine time t' for residual stress to cause SCC.

Enter calibration curve at time t' to get residual stress crR*

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Fig. 4 Illustration of how residual stresses are determined by stress corrosing cracking tests.

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5 to 15 to 25 30 35 40 45 30 SS 46 Time to Crack Initiation, hrs.

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Fig. 5 Relative comparison of residual stresses for various expansion processes. Though stress and time scales are given to provide perspective, the figure is intended to be more schematic than quantitative.

h

expansion. The residual stresses are much lower for. urethane tack expansion, which is the process currently used at Westinghouse.

Finite Element Ar-psis Results of a finite element analysis of the hydraulic expansion process are sunnarized in Fig. 6.

The locatiors and orientations of the maximun residual tensile stresses are in good agreement with the experimental SCC results.

' JOINT TIGHTNESS All of the full depth expansion processes used by Westinghouse close the gap tietween tube and tubesheet to virtually zero. This minimizes the potential for initiation of OD stress corrosion cracking of tubing within the tubesheet region and enhances the strength of the tube to tubesheet joint. However, the tube to tubesheet weld at the primary side of the tubesheet provides the primary barrier to leakage between primary and secondary sides, and also provides the required structural strength - no credit is taken for frictional resistance along the tube to tubesheet joint.

4.

TUBE MATERIAL PROPERTIES Inconel 600 Westinghouse Steam generator tubing is specially thermally treated to impart enhanced corrosion resistance. Tube expansion must not degrade this corrosion resistance.

Intuitively, it might be reasoned that hydraulic expansion muld have the least effect. Test results confirm that hydraulic expansion does not degrade the inherent corrosion resistance of the thermally treated tubing.

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SUMARY

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At Westinghouse, the current practice is to expand tubes full depth through the tM.:kness of c tubesheet to prov. le. ::d rn. gin against crevice s

corrosion. The present hydraulic' expansion process was selected over the former processes of explosive expansion or mechanical rolling because hydraulic expansion provides optimta balance in the attainment of the technical objectives relating to minimal crevice depth, low residual stresses, a.1d adequate joint tightness.

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