ML20197H660

Requests Withholding of WCAP-12763, Steam Generator Tube Collapse Considerations Presentation Matls from Public Disclosure

ML20197H660
Person / Time
Site:	Farley
Issue date:	11/09/1990
From:	Wiesemann R WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	Murley T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML19310C887	List: ML20197H654 ML20197H660 ML20197H665
References
CAW-90-090, CAW-90-90, NUDOCS 9011200034
Download: ML20197H660 (19)

Text

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Nuclear and Advanced Westinghouse Energy Systems

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Electric Corporation Box 355 Pittsburgh Pemsylvania 15230-0355 November 9, 1990 CAW-90-090 Document Control Desk US Nuclear Regulatory Commission i Washington, DC 20555 Attention: Dr. Thomas Murley, Director APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE

Subject:

" Steam Generator Tube Collapse Considerations" (WCAP-12763)

Dear Dr. Murley:

The proprietary information for which withholding is being requested in the-above-referenced letter is further identified in Affidavit CAW-90-090 signed by

-the owner of the proprietary information, Westinghouse . Electric Corporation.

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 with specificity the considerations listed in paragraph. (b)(4) of 10 CFR Section 2.790 of the Commission's regulations.

Accordingly, this-letter authorizes the utilization of the accompanying Affidavit by Alabama Power Company.

. Correspondence with respect to the proprietary aspects of the application. for withholding or the Westinghouse affidavit should reference this letter, CAW-90-090, and should be addressed to the undersigned.

Ver - ruly yours, G -

' l04240Gsut R bert A. Wiesemann, Manager Enclosures Regulatory & Legislative Affairs-cc: C. M.-Holzle, Esq.

Office of the General Counsel, NRC w _

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CAW-90-090 6,FFIDAVIT m COMMONWEALTH 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 set forth in this Affidavit are true and correct to the best of his knowledge, information, and belief:

I W't $ llAlld((JUL Robert A. Wiesemann, Manager.

Regulatory and Legislative Affairs Sworn to and subscribed before me this . /f> g u.-day of M 1990.

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/Ssu Notary Public NOTAAlAL SEAL

. LoARAINEM PiPucANoTARYPUBUC DONRCEV'LLE BCRO ALLEGHENYCoUNTY MY c0MV Ss:ON EXPIAEs DEC 14. m1 Member. Perrsytwe AssccaMn d Nc'.mcs

CAW-90-090 (1) I am Manager, Regulatory and Legislative Affairs, in the Nuclear and Advanced Technology Division, of the Westinghouse Electric Corporation i and as such, I have been specifically delegated the function of reviewing l the proprietary information sought to be withheld from public disclosure in

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connection with nuclear power plant licensing and rulemaking proceedings, and am authorized to apply for its withholding on behalf of the Westinghouse Energy Systems Business Unit.

l-(2) I am making this Affidavit in conformance with the provisions of 10CFR Section 2'.790 of the Commission's regulations and in conjunction with the Westinghouse application-for withholding accompanying this Affidavit.

-(3) I have personal knowledge of the-criteria .and procedures utilized by the ,

Westinghouse Energy Systems Business Unit in designating information as a trade secret, privileged or as confidential commercial or financial infortation. ,

(4) Pursuant to the provisions of paragraph (b)(4) of Section 2.790 of the Commission's regulations, the following is furnished for consideration by the Commission -in determining whether the information sought to be withheld

. from public disclosure -should be withheld.

(i) :The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse, o

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CAW-90-090

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(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 and whether to hold 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:

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

e (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, m

CAW-90-090 (c) Its use by a competitor would reduce his expenditure of.

resources or_ improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or.

licensing a similar product.

(d) It ' reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.

(e) It' reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.

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

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

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

(a) The use of such information by Westinghouse gives Westinghouse a competitive advantage over its competitors, it is, therefore,

, withheld from disclosure to protect the Westinghouse competitive position.

3 CAW-90-090 (b) It is information which is marketable in many ways. The extent to which such information is available to competitors diminishes i the Westinghouse ability to sell products and 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.

(d) 'Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. if competitors acquire components of proprietary information, any one component _ may be -

the key to the entire put thereby depriving Westinghouse of La competitive advantage.

'(e) . Unrestricted d'scle:3re would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the competition of those countries.

(f) The Westinghouse capacity to invest corporate assets in research l

and development depends upon the success in obtaining and maintainingLa competitive advantage.

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CAW-90-090 (iii) The information is being transmitted to the Commission in confidence' 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 or available information has not been previously

. employed in the same original manner or method to the best of our_ knowledge and belief.

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l(v) _The-proprietary information sought to be withheld in this submittal is-that which is appropriately marked in " Steam Generator Tube Collapse Considerations" for Joseph M. Farley Unit 2, WCAP-12763,-(Proprietary) being transmitted by-the Alabama Power Company. (APCo) letter and Application for Withholding Proprietary Information from Public Disclosure, Mr. W. G. Hairston 111, APCo, to Document Control Desk, attention Dr ' Thomas Murley, November,1990. The proprietary information as submitted for use by Alabama Power Company for Joseph M. Farley Unit 2 is expectad to be applicable in other-

-licensee submittals in response to.certain NRC requirements for justification ~ of collapse considerations in steam generator tubes.

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CAW-90-090 This information is part or that which will enable Westinghouse to:-

(a)- Provide documentation of the methods for calculating plate

. loads on steam generator tubes.

L (b) Establish applicable testing methods.  !

(c) Establish the use of methodology for combinir.g LOCA and seismic loads.

(d) _ ' Establish effects c.n calculated LOCA peak clad temperature.

(e) Assist the customer to obtain NRC approval.

Further this information has substantial commercial value as follows:

(a) Westinghouse plans to sell the use of similar information.

to its customers for purposes of meeting NRC requirements for licensing documentation.

(b) - Westinghouse can sell . support and defense of the technology to its customers in the licensing process. '

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CAW-90-090 l Public disclosure of this proprietary information is likely to cause substantial harm to the competitive positiro of Westinghouse because it would enhance the ability of competitors to provide similar sleeving services and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.

The development of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Westinghouse effort and the expenditure of a considerable sum of money.

In order for competitors of Westinghouse to duplicate this information, similar technical programs would have to be performed and a significant manpower effort, having the requisite talent and experience, would have to be expended for developing testing and analytical methods and performing tests.

j Further the deponent sayeth not.

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Proprietary Informatit,n Notice 4

Transmitted herewith are proprietary and/or non-proprietary versions of documents furnished to the NRC in connecticn with requests for generic and/c-plant-specific review and approval.

1 In order to conform to the requirements of 10 CFR 2.790 of the Commission's I regulations concerning the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary versions is contained within brackets and where the proprietary information has been l

deleted in the non proprietary versions on the brackets remain, the information i that was contained within brackets and where the proprietary information has l been deleted in the non-proprietary versions only the brackets remain, the information that was contained within the brackets in the proprietary versions having been deleted. The justification for claiming the information so designated as proprietary is indicated in both versions by means of lower case i

letters (a) through (g) contained within parentheses located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower case letters refer to the types of information Westinghouse customarily l- holds in confidence identified in Sections (4)(ii)(a) through (4)(ii)(g) of the l

affidavit accompanying this transmittal pursuant to 10 CFR 2.790(b)(1).

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Copyright Notice The reports transmitted herewith each bear a Westinghouse copyright notice.

The NRC is permitted to make the number of copies of the information contained in these reports which are necessary for its internal use ir, connection with generic and plant-specific reviews and approvals as well as the issuance, dental, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.790 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection not withstanding. With respect to the non-proprietary versions of thes3 reports, the NRC is permitted to make the number of copies bevond those necessary for its internal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose. The NRC is not authorized to male copies for the personal use of members of the public who make use of the NRC public document rooms. Copies made by the NRC mest include the copyright notice in all instances and the proprietary intice if the original was identified as proprietary.

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ATTACHMENT 2 1.0- Assessment of Tube Deformation for J. M. rarley Unit 2 Assessments of the steam generator (SG) tubes for LOCA and SSF loads have been performed to demonstrate that the tubes will not collapse for postulated LOCA events. Section 2.0 demonstrates that the principles of Leak-Before-Break (LBB) apply and that it is not necessary to postulate breaks in the RCS primary loop piping. This revised design basis is permitted by GDC 4. Section 3.0 describes the method that was used to generate time-pressure relationships as input for the structural evaluation of the steam generator tubes. The LOCA loads were generated for the pressurizer surge line and for the accumulator line breaks. Section 4.0 presents the methods used for the analysis of the steam generator tubes and concludes that no tube collapse was calculated for J. M. Farley for the combination of branch line LOCA loads with the SSE loads.

2.0 Primary Loop. Leak-Before-Break Applicability 2.1 Puroose This section of the report applies to the Farley plant reactor coolant system primary loop piping. It is intended to demonstrate that specific parameters for the Farley plant are enveloped by the generic analyses performed by Westinghouse in WCAP-9558, Revision 2 (Reference 1) and accepted by the NRC, and the applicability of the LBB methodology to the

[ Farley Units 1 and 2 primary loops (Reference 8),

2.2 1 rap _q ,

The current structural design basis for the Reactor Coolant System (RCS)

I primary loop requires that pipe breaks be postulated as defined in the L approved Westinghouse Topical Report WCAP-8082. In addition, protective l measures for the dynamic effects associated with RCS primary loop pipe

! breaks have been incorporated in the Farley plant design. However, L Westinghouse has demonstrated on a generic basis that RCS primary loop l- pipe breaks are highly unlikcly and should not be included in the structural design basis of Westinghouse plants. In order to demonstrate this applicability of the generic evaluations to the Farley plant, Westinghouse has performed a comparison of the loads for the Farley plant which are enveloped by loads used in the generic analyses, a fracture mechanics evaluation, a determination of leak rates from a through-wall crack, fatigue crack growth assessment, thermal aging 14 assessment, and an assessment of margins, o

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Attachment 2 I Page 2 1 2.3 Ob.iectives The conclusions of WCAP-9558, Revision 2 support the elimination of RCS primary loop pipe breaks for the Farley plant. In order to validate l this conclusion the following objectives must be achieved:

a. Demonstrate that Farley plant parameters are enveloped by generic

• Westinghouse studies.

b. Demonstrate that margin exists between the critical crack size and a postulated crack which yields a detectable leak rate.

. c. Demonstrate that there is sufficient margin between the leakage l through a postulated crack and the leak detection capability of L the Farley plant,

d. Demonstrate that fatigue crack growth is negligible.

2.4 Operation and Stability of the Reactor Coolant System The Westinghouse reactor coolant system primary loop has an operating history which demonstrates the inherent stability characteristics of the design. This includes a icw susceptibility to cracking failure from the i effects of corrosion (e.g., intergranular stress corrosion cracking),

L water hammer, or fatigue (low and high cycle). This operating history totals over 500 reactor-years, including five plants each having over 19 1 years of operation and 15 other plants each with over 14 years of I operation.

2.5 Primary looo leak-Before-Break Assessment

. The various considerations of this assessment are itemized below and

! conclusions are drawn.

• I. Westinghouse generic analyses reported in WCAP-9558 Revision 2 (Reference 1) eliminate postulated primary loop pipe breaks as a l- resolution of the A-2 safety issue on asymmetric LOCA loads (Reference 2). We have performed a preliminary assessment of the applicability of our generic efforts to Farley Units 1 and 2 which is summarized as follows: The primary loop parameters of Farley Units 1 and 2 are enveloped by the generic Westinghouse. studies. Specifically, in Reference 1, the maximum reactor coolant loop piping applied moment was 45,600 in-kips, whereas the maximum moment for Farley Units 1 and 2 is about 75 percent of that value, 33,200 in-kips.

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Attachment 2 Page 3 II. In other plant-specific applications where the plant had SA351 CF8A and SA351 CF8M materials, Westinghouse evaluated the thermal aging degradation and succenfully demonstrated Leak-Before-Break by predicting and using end-of-life (40 years) toughness properties. The maximum loads for Farley Units 1 and 2 are similar in magnitude when compared with the maximum loads in these plants.

III. A review of the plant specific heats of the material was performed for Farley Units 1 and 2. The component for maximum thermal aging degradation was identified as the 50 degree elbow at the inlet nozzle of the steam generator in Farley Unit 1. The lowest end-of-life toughness properties were established using the methodology and criteria of Reference 3. The lowest end-of-life toughness for the steam generator inlet elbow is estimated to be:

JI c - 345 in-lb/in2, Tmat - 2.4 and Jmax - 404 in-lb/in2 The maximum faulted loads in this elbow are: Axial Force - 2,100 kips and Bending Moment = 26,000 in-kips. The junction of the hot leg and the reactor pressure vessel outlet nozzle is found to be the highest stressed' location. The maximum faulted loads at this location are Axial

-Force - 1,780 kips and Bending Moment - 33,200 in-kips. The toughness properties for this location are estimated to be:

JIc .750 in-lb/in2 and Tmat - 60.

l IV. . At the steam generator inlet location (location of lowest toughness) the flaw length yielding a leak rate of 10 gpm (termed as the l_ " Leakage Size Flaw") is estimated to be about 4.5 inches. Based on available information it is found that, using elastic-plastic fracture mechanics, the length of the stable flaw exceeds 9 inches. Thus, a margin of a factor of 2 for flaw size is demonstrated. Similarly, at the reactor vessel outlet nozzle junction (highest stressed location) l the leakage size flaw is estimated to be less than 3.5 inches. The length of the stable flaw at this location exceeds 7 inches. Again, a margin of at least a factor of 2 for flaw size is demonstrated.

V. The " maximum allowable" flaw sizes at governing weld locations using the limit load method were calculated. The welds at these L . locations are field welds which are fabricated using either the GTAW or SMAW procedure. However, the calculations are performed by conservatively assuming that the weld at the gcverning location is fabricated by the SAW procedure; therefore, the lowest toughness among the three types of welds was used for the welds at the governing ,

Attachment 2 Page 4 ,

locations. The loads were amplified by the "Z" factor as per Reference

4. The maximum allowable flaw size (based on this approach) at the steam generator inlet nozzle location was found to be at least 20 inches. At the reactor vessel outlet nozzle the corresponding flaw size would exceed 14 inches.

VI. The effects of low and high cycle fatigue on the integrity of the primary piping are negligible. It is feasible to demonstrate that the fatigue crack growth would be negligible.

VII. The Farley Plant RCS pressure boundary leak detection system is consistent with the guidelines of Regulatory Guide 1.45 for detecting leakage of 1 gpm in one hour as documented in FSAR Sections 5.2.7 and NRC Safety Evaluation Report Section 5.6.

Conclusions It is judged that Leak-Before-Break can be demonstrated for Farley Units 1 and 2 primary loops with a high degree of confidence by reasons of:

1. Comparison with A-2 generic LBB studies.

2. Comparison with loads from plant specific analyses which addressed thermal aging.

3. Review of plant specific material chemistry to assess the thermal aging degradation.

4. Estimates of 'the leakage size flaw and the maximum length of stable flaw size based on the elastic-plastic fracture mechanics (J-T) approach. A margin of a. factor of 10 exists between the calculated leak rete from the leakage size flaw and the leak detection capability of 1 gpm. It is feasible to demonstrate a margin of at least 2 between the critical flaw size and the flaw size yielding a leak rate <

of 10 gpm.

5. If . limit load is used as the basis for critical flaw size, the margin for global- stability well exceeds that based on local stability elastic-plastic fracture mechanics evaluation.

6. It is feasible to demonstrate that the fatigue crack growth would be negligible.

The currently ongoing detailed LBB analysis will verify and further support the conclusions of this assessment.

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l Page 5 3.0 LOCA Induced Tube Loads The design and evaluation of steam generator tubes includes the consideration of the effects of postulated LOCA. To demonstrate that the steam generator tubes will not collapse for the postulated LOCA event, breaks in the reactor coolant loo) and its branch lines are analyzed using the MULTIFLEX 1.0 computer code w11ch has been reviewed and approved by the '

NRC (Reference 5). The assessment in Section 2.0 indicates that breaks in the main loop piping of the RCS need not be postulated. Accordingly, breaks in the largest branch lines have been postulated as the cause of the hydraulic loads on the steam generator and the steam generator tubes. The largest branch lines are the cold le'; accumulator line and the pressurizer surge line. These breaks have been analyzed to determine the loads and pressure time histories to be applied to the steam generator structural model. Based on site and location, the surge line and the accumulator line result in the limiting loads in the structural model (Reference 7).

4.0 Steam Generator Tube Integrity Assessment The evaluation of tube integrity for SSE + LOCA for the accumulator line and pressurizer surge line break LOCAs follows the methodology of WCAP-10043 submitted as part of the supporting documentation for the Callaway SER. As the analysis methodology of WCAP-10043 has never been applied to earlier model SG designs such as the 51 Series SG of Farley Unit 2, it is necessary to estimate the bundle response and potential for tube collapse by scaling to the Farley Unit 2 geometry and conditions. The 51 Series SG is more closely represented by the Model D3 SG and, therefore, the basis far the Farley Unit 2 estimate is the Model D3 SG at Watts Bar, analyzed in WCAP-10821 using the methodology of WCAP-10043.

The seismic analysis is a non-linear finite element analysis of the SG accounting for gaps that exist between the tube support plates (TSP) and wrapper and between the wrapper and SG shall, Acceleration time histories that bound the response spectra are appli.ed to the steam generator model composed of-primarily beam elements with Oppropriate mass representation.

Groups of tubes are modeled as composite tubes in the U-bend region to allow different response of tubes with d?fferent frequencies. Stiffness matrices supply boundary conditions for the piping connecting to the SG and for the SG supports. TSP impact force time histories are used for evaluation of the TSP's at the wedge locations between the TSP's and wrapper. The Model D3 loads are scaled to Farley Unit 2 spectrum levels by ratio of the energy content of the spectra.

Attachment 2 Page 6 The principal loading of the TSP wedges from the LOCA transients is due to the pressure rarefaction wave traveling through the U-bend. The pressure drop induces a lateral force at the top of the U-bend in the plane of the U-bend. The pressure time history is applied to a finite element model of three different size tubes (large, medium and small radius) and their dynamic response determined for three different support conditions at the top TSP, thereby, maximizing the potential loading on the TSP. The TSP reaction load time history for the three tubes is the basis for determining the reaction load for each tube in the bundle as a function of its frequency. Integrating these individual tube loads provides the total load at the TSP wedge locations.

In addition to the LOCA rarefaction wave, the SG would experience a shaking induced force. The response to this shaking is determined using the seismic model and analysis approach using displacement time histories as input. The LOCA rarefaction and LOCA shaking TSP loads are summed and the sum is combined by the square root of the sum of the squares with the seismic loads.

The TSP deformation that occurs from the wedge location impact forces (i.e., wedge loads) is estimated based on static load tests of a Model D configuration plate (similar in pattern to the 51 Series SG). Increasing loads were applied to the plate to obtain the load-deflection response and to measure final tube diameters after unloading. Knowing final tube diameters and the tube diameter change required to result in collapse by

, external pressure, the maximum allowable wedge load that results in no tube l collapse was determined for the Farley Unit I and 2 steam generators.

Comparing the maximum allowable load to each of the loads produced by the accumulator line break LOCA + SSE and the pressurizer surge line break LOCA

+ SSE,-no tubes are expected to be deformed to a final diameter that would l collapse under external pressure following the LOCA for Farley Units 1 and l

2 (Reference 6).

5.0 Conclusion l

The steam generator tubes in J. M. Farley Unit 2 will not collapse for the postulated LOCA event. The assessments presented in Section 2.0 of this report show that it is not necessary to postulate breaks in the RCS primary loop piping. Loads from the limiting branch line breaks in combination with SSE loads have been analyzed and it has been concluded that tubes are not calculated to collapse from these loads. Based on these facts, it has been demonstrated that the J. M. Farley Unit 2 steam generator tubes will not collapse.

Attachment 2 Page 7 l

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References:

1. WCAP-9558, Rev. 2, " Mechanistic Fracture Evaluation of Reactor Coolant i Pipe Containing a Postulated Circumferential Through-Wall Crack," i Westinghouse Proprietary Class 2, June 1981.

2. USNRC Generic letter 84-04,

Subject:

" Safety Evaluation of Westinghouse Topical Reports Dealing with Elimination of Postulated Pipe Breaks in PWR Primary Main Loops," February 1,1984. i

-3. Witt, F. J., Kim, C. C., " Toughness Criteria for Thermally Aged Cast Stainless Steel," WCAP-10931, Revision 1, Westinghouse Electric Corporation, July 1986, (Westinghouse Proprietary Class 2).

4. ASME Code Section XI, Winter 1985 Addendum, Article IWB-3640.

5. Takeuchi, K. et al, "MULTIFLEX, A FORTRAN IV Computer Program for Analyzing Thermal Hydraulic Structure System Dynamic", WCAP-8709-A (Non-Proprietary), September 1977.

6. NSD-JLH-0136, " Tube Integrity Assessment for LOCA + SSE - Farley Unit I and 2", November 14, 1990.

7. NS-SAT-SAII-90-338, "Model 51 Steam Generator Pressure / Time Histories", November 15, 1990.

8. MT-SMT-096, "Farley Units 1 and 2 Primary Loop, Pressurizer Surge Line, and Accumulator Lines Leak-Before-Break Assessment," November i

14, 1990.

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