NRC 2005-0004, Response to Request for Additional Information Regarding the Point Beach Nuclear Plant License Renewal Application and Request for Withholding of Proprietary Information from Public Disclosure

Response to Request for Additional Information Regarding the Point Beach Nuclear Plant License Renewal Application and Request for Withholding of Proprietary Information from Public Disclosure

ML050400483
Person / Time
Site:	Point Beach
Issue date:	01/28/2005
From:	Koehl D Nuclear Management Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
FOIA/PA-2010-0209, NRC 2005-0004, TAC MC2099, TAC MC2100
Download: ML050400483 (37)

Text

NMC Committed to Nuclear Exce Point Beach Nuclear Plant Operated by Nuclear Management Company, LLC January 28, 2005 NRC 2005-0004 10 CFR 2.390 10 CFR 54 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington DC 20555 Point Beach Nuclear Plant, Units 1 and 2 Dockets 50-266 and 50-301 License Nos. DPR-24 and DPR-27 Response to Request for Additional Information Regarding the Point Beach Nuclear Plant License Renewal Application (TAC Nos. MC2099 and MC2100) and Request for Withholding of Proprietarv Information from Public Disclosure By letter dated February 25, 2004, Nuclear Management Company, LLC (NMC),

submitted the Point Beach Nuclear Plant (PBNP) Units 1 and 2 License Renewal Application (LRA). On November 17, 2004, the Nuclear Regulatory Commission (NRC) requested additional information regarding Time Limited Aging Analysis (TLAA)

(LRA Section 4.3). Enclosure 2 to this letter contains NMC's responses to the staff's questions. contains a Westinghouse proprietary authorization letter, CAW-04-1931; accompanying affidavit; Proprietary Information Notice; and Copyright Notice concerning information proprietary to Westinghouse Electric Company, LLC

("Westinghouse"). (Responses to RAI 4.3.2.2 and 4.3.4.3).

As Enclosure 2 contains information proprietary to Westinghouse, it is supported by an affidavit signed by Westinghouse; the owner of the information. The affidavit 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 2.390 of the Commission's regulations.

Accordingly, it is respectfully requested that the information, which is proprietary to Westinghouse, be withheld from public disclosure in accordance with 10 CFR 2.390 of the Commission's regulations. Correspondence regarding the copyright or proprietary aspects of the items listed above, or the supporting Westinghouse affidavit, should reference CAW-04-1931 and should be addressed to J. A. Gresham, Manager, 6590 Nuclear Road

• Two Rivers, Wisconsin 54241 Telephone: 920.755.2321

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-=t Document Control 'Desk' Page 2' Regulatory Compliance and Plant Licensing, Westinghouse Electric Company, LLC, P.O. Box 355, Pittsburdh,'Penhsylvania'=15230-0355.

contains the non-proprietary version of Enclosure 2.

On December 1,- 2004, the NRC staff 'verbally provided additional time for NMC to respond to this request for additional information in order for further'clarifications to be provided. The clarifications allowed thePBNP. License Renewal project staff to clearly understand the information needed 'and for further analysis to be completed.

This letter-contains no new commitments-and no revisions to existing commitments.

I declare under penalty of perjury that the forgoing is true and correct.0 Executed on January 28, 2005.-

Dennis L. Koehl' Site Vice-President,-Point Beach Nuclear-Plant Nuclear Management'Company,-'LLC Enclosures (3) cc:

(w/o enclosures)

Regional Administrator,-Region 111, USNRC Project Manager,--Point Beach Nuclear Plant, USNRC Resident Inspector,'Point Beach 'Nuclear. Plant, USNRC

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PROPRIETARY INFORMATION NOTICE Transmitted herewith are proprictary and/or non-pioprietary versioms of documents furnished to the NRC in connection with requests for generic and/or plant-specific review and approval.

In order to conform to the requirements of 10 CFR 2.390 of the Commission's regulations concerning the protection of proprietary information so submitted to thc'NRC, the information which is proprietary in the proprietary versions is contained within' brackets-,and vhere the proprietary infornation has been deetcd in the non-proprietary versions. onily the brackets renTain (the information that was contained within the brackets in the proprietary versions having been dektcd).'-The justification for claiming the information so designated as proprietary is indicated in botlivirsions by means of lower case letters (a) through(fl) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such inforation.' These lower case lctters refer to the types of information Westinghouse custornarilyhh6lds in-confidence identified in Sections (4)(ii)(a) through (4)(ii)(f) of the affidavit accompanying thi nsmittal pursuant to 10 CFR 2.390(b)(1).

COPYRIGHT NOTICE The reports transmitted herewith ea'ch bear a Westinghoudsecbcpyright notice. The NRC is permnitted to mak~e thenumb~erof copiesof the inform'ation containeciin these reports which arc nece~ssary for its' internal use iii connection with tcreric and plitf-i~ec'ific reiviews and approv'als as well as the issuance.

denial, amendment, transfer,. renewal, modification, suspension ocatioji, or violation of a license, permit. -order, or regulation subject to the requiremen~ts of 10 CFR 2.390 regarding restrictions on public disclosure to, the extent such informat o~n has been-ide nt'ified as'proprietary by-WestinghoDuse, copyright protection-notwithstanding. With respect tothe non-pruoritar-versions of these r.ports, the NRC is permitted to mhake'the number of ciopieis'beyond th6sieri&cc ssary ror its internal use. which are necessary in order to have one copy available for public ~'iewing in, the apprprae ocefiles 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 p~ur'pose. Copies ma de by the NRC must include the copyright notice in all instanices'and the proprietary notic-e if the original w~as identified as proprietary.

Page 1 of 7

e W St inghouse-

%;estighouse lectricCompany Nudear Services P.O. B= 355 Pittsburgh.Pennsylvania 15230-0355 USA U.S. Nuclear Regulatory Commission Direct tel: (412)374.4643 Document Control Desk Diret Fax: (412) 3744011 Washington, DC 20555-0001 e-mail: greshajazwesringhouse.com Our ref: CAW-04-1931 December 14, 2004

- APPLICATION FOR I OLDING PROPRIETARY INFORMATION FROM PUBJ.IC DISCLOSURE

Subject:

"Response to Point Beach Nuclear Plant. UnitsI arid 2. License Renewal Application (LRA)

Request for Additional Information (RAls)" (Proprietary)

The proprietary information of-r which 'wihholding is being requested in the above-referenced report is further identified in Affidavit CAW-041931 signed by the owner of the proprietary informnation.

Westinghouse Electric Company LLC. The affidavit, which accompanies this letter, sets forth the basis on which the information nay be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)X4) of 10 CFR Section 2.390 of the Conunission's regulations.

Accordingly, this letter authorizes the tilizatioi of the accompanying affidavit by Nuclear Management Company.--.

Correspondence with respect to the proprietary aspects of the application for withholding or the Westinghouse affidavit should reference this letter, CAW-44-193 1. and should be addressed to J. A. Gresham, Manager, Regulatory Compliance and Plant Licensing, Westinghouse Electric Company LLC. P.O. Box 355. Pittsburgh,'Pennsylania 15230-0355.-

-Very truly yours,

- J. A Gresham, Manager Regulatory;Compliance and Plant Licensing Enclosures cc: B. Benney L. Feizollahi A BNFI Group company Page 2 of 7

CAW 1931 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA-S5 COUNTY OF ALLEGHENY:

Before me, the undersigned authority, personally appeared J. A. Gresham, 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 Company LLC (Westinghouse), and that the avermcnts of fact set forth in this Affidavit are true and correct to the best of his knowledge. information, and belief:

'A. Gresham, Manager Regulatory Compliance and Plant Licensing Sworn to and subscribe-before me this day of X

2 2004 Notary Public NoWW Sod

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' Page 3 of 7

2 -CAW.04-1931

( 1) am Manager, Regulatory Compliance and Plant Licensing, in Nuclear Services, Westinghouse Electric Company LLC (Westinghouse). and as such. i have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rule rakitig proceedings. and am authorized to apply for its withholding on behalf of Westinghouse.

(2) 1 am making this Affidavit in conformance with the provisions of 10 CFR Section 2.390 of the Commission's regulations and in conjunction with the Westinghouse "Application for Witthholding" accompanying this Affidavit.

(3) 1 have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged or as confidential commercial or financial information.

(4)

Pursuant to the provisions of paragraph (b)(4) of Section 2.390 of the Commission's regulations, the following is furnished for consideration by the Cormmission in determining whether the information sought to be withheld from public disclosure should be withheld.

(i)

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

(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 ccrtain 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.

tinder that system, information is held in conridence 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 Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.

Page 4 of 7

3 CAW 1931 (b)

It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, ec.) the application of which data secures a competitive economic advantage, e g.. by optimization or improved marketability.

(c)

Its use by a competitor would reduc hi di f

ve his competitive position in the design. manufacture, shipment, installation, assurance of quality, or licensing a similar product.

(d)

It reveals cost or price information, productioncapacities. budget levels, or conmercial 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.

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.

(b)

It is information that is marketable in rnany ways. The extent to which such information is available to competitors diminishes 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 puzzle, thereby depriving Westinghouse of a competitive advantage.

Page 5 of 7

4 CAW.04-1931 (e)

Unrestricted disclosure 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 and development depends upon the success in obtaining and rmaintaining a competitive advantage.

(iii)

The information is being transmitted to tlhe Coissjion in confidence and. under the provisions of 10 CFR Section 2.390, it is to be received in confidence by the Commission.

(iv)

The information sought lobe 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.

(v)

The proprietary information sought to be withheld in this submittal is that which is appropriately marked in "Response to Point Beach Nucleair Plant, Units I and 2. License Renewal Application (LRA) Request for Additional Information (RAts) '(Proprietary) dated December 2004, being transmitted by the Nuclear Management Company letter and Application for Withholding Proprietary Information from Public Disclosure, to the Document Control Desk. The proprietary information as submitted for use by Westinghouse for the Point Beach Nuclear Plant. Units I and 2 is expected to be applicable for other licensee submittals in response to certain NRC requirements forjustification of continued safe operation of Point Beach Units I and 2.

This information is pan of that which will enable Westinghouse to:

(a) Assess the technical justification for renewing the operating license.

(b) Assist the customer in obtaining 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.

Page 6 of 7

5 CAWNV4-1931 (b)

Westinghouse can sell support and defense of continued safe operation with a renewed license.

(c)

The information requested to be withheld reveals the distinguishing aspects of a methodology which was dcveloped by Westinghouse.

Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it iwould enhance the ability of competitors to provide similar support documentation and licensing dcfnsc 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.

Further the deponent sayeth not.

Page 7 of 7

WESTINGHOUSE NON-PROPRIETARY CLASS 3 ENCLOSURE 3 RESPONSE TO POINT BEACH NUCLEAR PLANT, UNITS I AND 2 REGARDING LICENSE RENEWAL APPLICATION (LRA)

REQUESTS FOR ADDITIONAL INFORMATION (RAIs)

January 2005 This document Is the property of and contains Prorietar Infmto we yWetnhueEeti Company LLC and /or Its subcontractors and suppliers. It is transmitted to you' in confidence and trust, and you agree to treat this document in strict accordance, w ith the terms and conditions of the agreement under which It was provided to you.

Westinghouse Electric Company LLC P.O. Box 355 Pittsburgh, PA 15230-0355 02004 Westinghouse Electric Company LLC All Rights Reserved Page 1 of 27

WESYINGHOUSE NON-PROPRIETARY CLASS 3

'ENCLOSURE 3 RESPONSE TO POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 REGARDING LICENSE RENEWAL APPLICATION (LRA)

REQUESTS FOR ADDITIONAL'INFORMATION (RAls)

The following information is provided in response to the Nuclear Regulatory Commission (NRC) staffs request for additional information (RAI) regarding the Point Beach Nuclear Plant (PBNP) License Renewal Application (LRA).

The NRC staffs questions are restated below, with the Nuclear Management Company (NMC) response following.

4.3 Metal Fatigue NRC Question RAI-4.3.1 -'Reactor Vessel Structural Integrity:

Provide confirmation that the limiting locations of the PBNS reactor vessels evaluated for extended operation correspond to the structures and/or components listed in Table IV.A2 of NUREG-1801, Volume 2, 'for PWR reactor vessels structures and/or components, where cumulative fatigue dam-age/fatigue is the aging effect/mechanism, and which require further evaluation as TLAAs for the period of extended operation.

Alternatively, provide the location in the LRA where this information is shown.

NMC Response:

As noted in the following table, the limiting locations of the PBNP reactor pressure vessel (RPVs) evaluated for extended operation correspond to the structures and/or components listed in Table IV.A2 of NUREG-1 801,volume 2, for PWR reactor vessel structures and/or components, where cumulative fatigue damage/fatigue is the aging effect/mechanism.

Page 2 of 27

WESTINGHOUSE NON-PROPRIETARY CLASS 3 NUREG-1801 RPV Components (1)

Table IV.A2 Component Description Evaluated for Fatigue at PBNP Item A2.1-b Closure Head:

A2.1.1 Dome Yes A2.1-e Closure Head:

A2.1.3

- Stud Assembly Yes A2.2-c Control Rod Drive Head Penetration:

Y A2.2.1 Nozzle Yes A2.2.2 Pressure Housing Yes A2.3-c Nozzles:

A2.3.1

'- Inlet Yes A2.3.2 Outlet Yes A2.3.3 Safety Injection Yes A2.4-a Nozzle Safe Ends:

A2.4.1

- -Inlet Yes (2)

A2.4.2

- Outlet Yes (2)

A2.4.3

- Safety Injection Yes (2)

A2.5-d Vessel shell:

A2.5.1 Upper (nozzle) Shell Yes A2.5.2 Intermediate and Lower Shell Yes A2.5.3 Vessel Flange.

Yes A2.5.4

- -'Bottom Headi Yes A2.8-a Pressure Vessel Support:

A2.8.1 Skirt Support Yes (3)

(1) - Reactor Vessel structures and/or components listed in Table IV.A2 of NUREG-1801, Volume 2, for PWR reactor-vessels'structures and/or components.

(2) - Included with specific nozzle analyses.

(3) - PBNP RPVs are supported off of external support brackets at the nozzle elevation.

NRC Question RAI-4.3.2.1 Reactor Vessel Internals Structural Integrity:

Provide confirmation that the limiting locations of the PBNS reactor vessel internals evaluated for extended operation correspond to the structures and/or components listed in Table IV.B2 of NUREG-1801, Volume 2, for PWR reactor vessel internals structures and/or components, where cumulative fatigue damage/fatigue is the aging effect/mechanism, and which require further evaluation as TLAAs for the period of Page 3 of 27

WESTINGHOUSE NON-PROPRIETARY CLASS 3 extended operation. Altematively, provide the location in the LRA where this information is shown.

NMC Response:

The following table identifies the structures and/or components listed in Table IV.B2 of NUREG-1801, Volume 2, for PWR reactor vessel internals structures and/or components where cumulative fatigue damage/fatigue is the aging effect/mechanism.

The table also identifies whether the component locations for PBNP reactor vessel internals-were evaluated for fatigue for extended operation. The major components have been evaluated for fatigue. Since the6PBNP reactor internals were designed and manufactured prior to the release of Subsection' NG of the ASME Code Section III, fatigue evaluations were not performed nor were they required for all of the locations noted in Table IV.B2 of NUREG-1 801, Volume 2, for PWR reactor vessel internals structures and/or components, where cumulative fatigue damage/fatigue is the aging effect/mechanism.

NUREG-1801 Reactor Vessel Internals (PWR) -Westinighouse (1)

Table IV.B2 Component Description Evaluated for Fatigue at PBNP Item B2.1-c Upper Internals Assembly:

B2.1.1 Upper Support Plate Yes B2.1.4 Upper Core Plate Yes; B2.1.7 Hold-Down Spring

No (2)

B2.1-h

'Upper Internals Assembly:

B2.1.2 Upper Support Column Yes B2.1 -m Upper. Internals Assembly:

B2.1.6 Fuel Alignment Pins No& (2)

B2.2-c RCCA Guide Tube Assemblies:

B2.2.1 RCCA Guide Tubes Yes' B2.2-f RCCA Guide Tube Assemblies:,-

B2.2.2

- 'RCCA Guide Tube Bolts

'No (2)

B2.2.3 RCCA Guide Tube Support

' Yes

-Pins B2.3-d Core Barrel:

B2.3.1 Core Barrel (CB)

Yes B2.3.2 CB Flange (upper)

Yes B2.3.3 CB Outlet Nozzles Yes B2.3.4

- Thermal Shield Yes B2.4-g Baffle/Former Assembly:

B2.4.1 Baffle and Former Plates

-No (2)

B2.4.2 Baffle/Former Bolts Yes (Bolt Qualification by Testing)

B2.5-d Lower Internal Assembly:

B2.5.1 Lower Core Plate Yes B2.5.4 Lower Support Plate Columns Yes Page 4 of 27

WESTINGHOUSE NON-PROPRIETARY CLASS 3 NUREG-1801 Reactor Vessel Internals (PWR) - Westinghouse (1)

Table IV.B2 Component

Description:

Evaluated for Fatigue at PBNP Item a_-__

_X; B2.5-j Lower Internal Assembly:

B2.5.2

- Fuel Alignment Pins No (2)

B2.5.5

- Lower Support Plate Column No (2)

Bolts B2.5-p Lower Internal Assembly:

B2.5.6

- Radial Keys and Clevis Inserts -Yes B2.5.7

- Clevis Insert Bolts

,No (2)

(1) - Reactor Vessel Internals (PWR) - Westinghouse, structures and/or components listed in Table IV.B2 of NUREG-1 801, Volurmne 2, for PWR reactor vessels structures and/or comiponents.

(2) - Prior to the release of Subsection NG of the ASME Code Section III, the reactor vessel internals were designed to the intent of Subsection NB which was a pressure vessel code;. The' PBNP reactor internals were designed and manufactured prior to the release of Subsection NG of the ASME Code Section 1II.

During this period, several sets of internals were being designed 'and manufactured at or about the same time. The reactor internals designs were segregated as 2-loop, 3-loop and 4-loop,'but plant specific'design packages and calculations were not produced. There were no ASME code'design specifications or stress reports developed for the internal packages manufactured prior to Subsection NG.

Typically, hand calculatiorns where performed for the various subcomponents of the internals. Many of the individual subcomponent calculations were assembled into a single document, which is identified as the Westinghouse 2-Loop Design Manual.

When a particular design feature'w'as changed, new calculations would be performed to demonstrate the adequacy of the changed component. For components that were not changed, the original design and drawings continue to apply.,

Page 5 of 27

WESTINGHOUSE NON-PROPRIETARY CLASS 3 NRC Question RAI-4.3.2.2 Reactor Vessel Internals Structural Integrit :

Provide a summary of 60-year primary-plus-secondary stress intensities and cumulative fatigue usage factors (similar to revised Tables 4.3-1 and 4.3-2 in Appendix A of the LRA for components of the reactor vessel) for the key reactor internal components listed on page 4-41 of the LRA.

NMC Response:

In accordance with discussions with NRC staff on October 28, 2004, it was agreed that a summary of 60-year primary-plus-secondary stress intensities for the PBNP reactor vessel internals need not be provided-since the NRC staff indicated that the summary was not critical for the review and it is not a part of the PBNP current licensing basis (CLB).

A summary of the PBNP key reactor vessel internal component cumulative fatigue evaluation results is included in the following table.

PBNP Key Reactor-Vessel Internals Component Design Basis Cumulative Fatigue Results Cumulative Component Description -

Usage Factor (CUF)

Upper Support Plate Perforations (1

)

Upper Core Plate alignment Pins

-IC (1 )

Upper Core Plate

.lc (2)

Upper Support Column I Base Weld *4(1L RCCA Guide Tube Sheath Weld C

-L 1L Guide Tube Flange Weld (2)

Lower Support Plate/ CB Weld J

bc (2}

CB / Flange Weld c(2 Outlet Nozzle Inner Weld

c(1)

Thermal Shield Flexures -

)-

-i-I::l Thermal Shield Flexure Bolts X:

bi1 Lower Core Plate SC Bolt Holes

-1OC (2)

Lower Support Column Extensions c (2)

Lower Radial Restraint Dowel Pin

-_1 Flexureless Insert I

=,c BRACKETED NUMBERS ARE WESTINGHOUSE PROPRIETARY (1) - WCAP-14459 "Reactor Pressure Vessel and.Internals Evaluations for the Point Beach Units 1 and 2 Power Uprating I Replacement Steam Generator Program,"

April 1996.

(2) - Westinghouse, "Power Uprate Project, Point Beach Nuclear Plant Units I and 2, Volume 1, NSSS Engineering Report," April 2002.

Page 6 of 27

WESTINGHOUSE NON-PROPRIETARY CLASS 3 NRC Question RAI-4.3.3 Control Rod Drive Mechanism Structural Integrity:

Provide a comparison of the CLB set of transient conditions and design cycles, and the revised set of full power uprate transient conditions and design cycles, that were used in the CRDM fatigue' TLAAs to show conformance with the CLB fatigue limits to the end of the period of extended operation.

NMC Response:

The revised set of full power uprate transient conditions and design cycles is PBNP's CLB set of transient conditions and design cycles. These are shown in the revised Table 4.1-8, "Thermal and Loading Cycles" in' Appendix A "FSAR Supplement" of the PBNP LRA. This set of transient conditions and design cycles was used in the evaluation of all PBNP TLAAs that req uired the use of transient conditions and design cycles. This set of transient conditions and design cycles was used in the CRDM fatigue TLAA evaluation to show conformance with the CLB fatigue limits to the end of extended life (EOEL).

NRC Question RAI-4.3.4.1 Steam Generator Structural Integrity:

Provide confirmation that the limiting locations of the PBNS steam generators evaluated for extended operation 'correspond to the structures and/or components listed in Table IV.D1 of NUREG-1 801, Volume62, for PWR reactor vessels structures and/or components, where cumulative fatigue damage/fatigue is the aging effect/mechanism, and which require further evaluation as TLAAs for the period of extended operation.

Alternatively, state the location in the LRA where this information has been provided.

NMC Response:

As noted in the following table, the location's'of the PBNP steam generators evaluated for extended operation correspond to the structures and/or components listed in Table IV.D1 of NUREG-1 801, Volume&2,2-for PWR reactor vessel internals structures and/or components, where cumulative fatigue-damage/fatigu'e is the aging effect/mechanism.

Page 7 of 27

WESTINGHOUSE NON-PROPRIETARY CLASS 3 NUREG-1801 Steam Generator Components (1)

Table Evaluated for Fatigue at PBNP IV.Dt Component Descriptiont Unit 1 (44F)

Unit 2 (A47)

D1.1-a Pressure Boundary and Structural:

D1.1.1 Top Head Yes Yes D1.1.2 Steam Nozzle and Yes Yes Safe End Yes Yes D1.1-b Pressure Boundary and Structural: -

D1.1.3

- Upper and Lower Shell Yes Yes D1.1.4 Transition Cone Yes Yes D1.1.5 FW Nozzle and Yes Yes Safe End Yes (5)

Yes (5)

D1.1.6 FW Impingement Plate and Support' NjA (2)

N/A (2)

D1.1-h Pressure Boundary and Structural:-

D1.1.8

- Lower Head Yes Yes D1.1.9 Primary Nozzles and Yes Yes Safe Ends Yes (4)

Yes (4)

D1.2-d Tube Bundle:

D1.2.1 Tubes and Yes Yes Sleeves N/A (3)

N/A (3)

(1) - Steam Generator structures and/or components listed in Table IV.DI of NUREG-1 801, Volume 2, for PWR reactor vessels structures and/or components.

(2) - The 44F and A47 replacement steam generators are feed ring designs and have no-impingement plates.

(3) - There are no sleeved tubes in the PBNP='steam generators.

(4) - The Unit 1 SGs' safe ends are stainless steel weld buildup, Unit 2 steam generators contain stainless steel safe ends. The Unit I and Unit 2 safe ends were analyzed with the nozzle.

(5) - The Unit l and Unit 2 feedwater nozzles do not have a nozzle-to-piping safe end.

The feedwater nozzles do have a'safe end in the nozzle-to-thermal sleeve.

The thermal sleeve safe ends were included in the nozzle analysis'.

Page 8 of 27

WESTINGHOUSE NON-PROPRIETARY CLASS 3 NRC Question RAI-4.3.4.2 Steam Generator Structural Integrity:

Provide a comparison of the CLB set of transient conditions and design cycles, and the revised set of Steam Generator Replacement and Full Power Uprate transient conditions and design cycles, that were used in the Units I and 2 steam generator fatigue TLAAs to show conformance with the' CLB fatigue CUF limit to the end of the period of extended operation. Alternatively, provide clarification stating that the applicable transient conditions and design cycles'are those stated in Table 4.1-8 of Appendix A to the LRA.

NMC Response:

The revised set of full power uprate transient conditions and design cycles is PBNP's CLB set of transient conditions and design cycles. These are shown in the revised Table 4.1-8, "Thermal and Loading Cycles," in Appendix A "FSAR Supplement" of the PBNP LRA. This set of transient conditions and design cycles was used in the evaluation of all PBNP TLAAs that required the use of transient conditions and design cycles. This set of transient conditions and design cycles was used in the steam generator fatigue TLAA evaluation to show conformance with the CLB fatigue limits to the EOEL.

NRC Question RAI-4.3.4.3 Steam Gebnerator Structural Integrity:

List the key Units I and 2 steam generator components, and provide for each a summary of 60-year primary-plus-secondary stress intensities and cumulative fatigue usage factors (similar to revised Table~s4.3-1 a-nd 4.3-2 in Appendix A of the LRA for components of the reactor vessel) for these ecomponents.

NMC Response:

In accordance with discussions with NRC 'staff on October 28, 2004, it was agreed that a summary of 60-year primary-plus-secondary stress intensities for the PBNP steam generators need not be provided since the secondary stress intensities were not critical for the review and are not a part of the PBNP CLB.

A summary of the PBNP key steam generator components fatigue evaluation results is included in the following table.

It should be noted that the design of the steam generators for the two PBNP units are not identical. The Unit 1 steam generators are'an early 80's vintage, incorporating bounding generic analyses based on the Westinghouse 41 series steam generator design. The Unit 2 steam generators a'reb~a mid 90's vintage, incorporating a PBNP specific design analysis.

Page 9of27

WESTINGHOUSE NON-PROPRIETARY CLASS 3 PBNP Key Steam Generator Components Design Basis Cumulative Fatigue Results Component Description Unit 1 (44F)

Unit 2 (A47)

CUF CUF Tube(s) 4t~[f 1

1 Tube-to-Tubesheet Weld

-i c (2)

,C (4)

Primary Chamber, Tubesheet, and Stub Barrel

]bc (1) b,C (4)

Primary Nozzle(s)

-]

c (3)

C (4)

Primary Manway Openings

]

(3)

L..1I(4) (6)

Divider Plate i]

(2)

[

]

(4 Steam Nozzle, Upper Head, Upper Shell

-]

C (3)

]

C (1)

Steam Nozzle Venturi

]b.C(

3

)

Feedwater Nozzle

[

Transition Cone

,c (7) 11 Secondary Manway Opening c (8)

,c Secondary Handholes / Access Openings

i.

i.

b

.C (1L)...1

"~ (4) -

Secondary Inspection Ports LJ C (1)(5)

[

1X(4)

Minor Penetrations DC (3)

,C 4)

BRACKETED NUMBERS ARE WESTINGHOUSE PROPRIETARY (1) - Westinghouse, "Power Upiate Project, Point Beach Nuclear Plant Units 1 and 2, Volume 1, NSSS Engineering Report," April 2002.

(2) - WCAP-14602, Volume 1, "Point Beach~Nuclear Plant Unit 2 Steam Generator Replacement Engineering Report," March 1996.

(3) - Westinghouse WNEP-8393, "Model 44F Replacement Steam Generator Stress Report for Wisconsin Electric Power Company Point Beach Unit 1," Revision 0, October 1983.

(4) - WNEP-9513, "Delta 47-Steam Generator -Stress Report Summary Wisconsin Electric Power Company Point Beach Unit 2," Revision 1,l -December-1996.

(5) - The limiting location shown is the bolts.- These are managed by replacement on a periodic basis. The next limiting location is the manway pad with a CUF of

[ -

b,-c (6) -The bolts and drain hole are qualified for fatigue based on tests. The CUF shown is for the cover and pad knuckle.;

(7) - WTD-EM-79-039, "Model 44F Steam Generator Replacement Units Shell / Cone /

Lower Shell Analysis," April 1, 1979.-

(8) - Westinghouse WNEP-8393, "Model 44F Replacement Steam Generator Stress Report for Wisconsin Electric Power Company Point Beach Unit 1," Revision 0,

-October 1983. - The value for the inside radius of the steam nozzle is bounding.

Page 10 of 27

WESTINGHOUSE NON-PROPRIETARY CLASS 3 NRC Question RAI-4.3.5.1i Pressurizer Structural Inte'rity:

Provide confirmation that the limiting fatigue locations of the PBNS pressurizers evaluated for extended operation correspond to the pressurizer structures and/or components listed in Table IV.C2.5 of NUREG-1 801, Volume 2, for PWR reactor vessels structures and/or components, where cumulative fatigue damage/fatigue is the aging effect/mechanism, and which'recuire' further evaluation as TLAAs for the period of extended operation. Alternatively, state the location in the LRA where this information has been provided.

NMC Response:

As noted in the following table, the fatigue locations of the PBNP pressurizers evaluated for extended operation correspond to the'pressurizer structures and/or components listed in Table IV.C2.5 of NUREG-1 801,Voiume 2, for PWR reactor vessels structures and/or components, where cumulative fatigue damage/fatigue is the aging effect/mechanism.

NUREG-1801 Pressu'rizer Components (1)

Table IV.C2.5 Component Description Evaluated for Fatigue at PBNP Item-C2.5-a Pressurizer:

C2.5.1 Shell and Yes Heads Yes C2.5-d Pressurizer:

C2.5.2

- Spray Line Nozzle Yes C2.5.4

- Spray Head No (2)

C2.5-e Pressurizer:

C2.5.3 Surge line nozzle Yes C2.5-f Pressurizer:

C2.5.5 Thermal -sleeves Yes C2.5.6 Instrument 'penetrations

'Yes C2.5.7 Safe ends Yes C2.5-q

-Pressurizer:

C2.5.10 Heater Sheaths and Yes

- Heater Sleeves Yes C2.5-t Pressurizer:

C2.5.11

- Support keys, No (N/A for PBNP) (3)

- Skirt and Yes ShearLugs

No (N/A for PBNP) (3)

C2.5-w Pressurizer:

C2.5.12 Integral Support No (N/A for PBNP) (3)

Page 11 of 27

WESTINGHOUSE NON-PROPRIETARY CLASS 3 (1) - Pressurizer structures and/or components listed in Table lV.C2.5 of NUREG-1 801, Volume 2, for PWR reactor vessels structures and/or components.

(2) - The spray head is a non-structural or pressure retaining component and is not in the scope of License Renewal.'

(3) - The PBNP pressurizers do not havesupport keys, shear lugs, or integral supports.

NRC Question RAI-4.3.5.2 Pressurizer Structural Integrity:

Provide a comparison of the CLB set of transient conditions and design cycles, and the revised set of Steam Generator Replacement and Full Power Uprate transient conditions and design cycles,'that were used in the Units 1 and 2 pressurizers fatigue TLAAs to show conformance with the CLB fatigue limit to the end of the 'period of extended operation.

NMC Response:

The revised set of full power uprate transient conditions and design cycles is PBNP's CLB set of transient conditions and design cycles. These are shown in the revised Table 4.1-8, "Thermal and Loading Cycles," in'Appendix A, "FSAR Supplement," of the PBNP LRA. This set of transient'conditions and design cycles was used in the evaluation of all PBNP TLAAs that required the' use of transient conditions and design cycles. This set of transient conditions and design cycles was used in the pressurizer fatigue TLAA evaluation to show conformance"with the CLB fatigue limits to the EOEL.

NRC Question RAI-4.3.5.3 Pressurizer Structural Integrity:

Provide clarification that the "plant-specific insurge/outsurge" fatigue analyses are based on the combination of the insurge/outsurge qransient condition and the transients listed in the revised set of Steam Generator Replacement and Full Power Uprate transient conditions.

NMC Response:

The plant-specific insurge/outsurge fatigue analysis was'based on a combination of the actual insurge/outsurge transients and other loadings experienced by the pressurizer and surge line components.' Projections \\were made backward and forward in time to estimate the cumulative fatigue 'usage for these components for the entire 60-year operating life.

Extensive experience with fatigue monitoring has demonstrated that a significant system temperature differential (i.e.- the difference between pressurizer water temperature and RCS hot leg temperature) is required to produce thermal fatigue in the surge line and lower head. This effect occurs'during plant heatups and cooldowns.

Other transients, such as a reactor trip, do not produce stresses above the minimum fatigue threshold.

Page 12 of 27

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Several types of loadings contribute to heatup and cooldown transients, including:

• Internal pressure

• Surge line piping thermal expansion Surge line piping thermal stratification

• Thermal shock, or "insurge/outeurge" temperature transients from flow reversals The FatiguePro software installed at Point Beach Unit 1 and Unit 2 was used to evaluate these effects on the pressurizer locations affected by the insurge/outsurge transients.

FatiguePro computes stresses in various fatigue-sensitive components based on real plant data.' A' computational scheme was devised to compute the water temperature at various zones in the surge line and pressurizer lower head based on available temperatures, flows and other applicable instruments to capture any insurge/outsurge effect that the plant may experience duringoperation.-

The following locations in the surge line and pressurizer lower head are monitored:

• Hot Leg Surge Nozzle,

• Pressurizer Surge Nozzle

• Pressurizer Heater Penetration Weld

• Pressurizer Water Temperature Instrurment Nozzle The pressurizer heater penetration weld was determined to be the bounding location for fatigue usage in the surge line and pressurizer lower head. Plant data was available for Point Beach Units 1 and 2 from 1994 to present. The data was screened for heatup and cooldown transients to be analyzed by FatiguePro software. A cooldown followed by a heatup was assumed to represent a transient cycle.

Fatigue Usaae Proiections Fatigue usage projections were based on theiassumed number of future cycles that the plant will experience. For the purpose of projecting future fatigue usage, the' incremental usage for a cooldown/heatup cycle was assumed to be the average incremental fatigue from the template periods.'

The projected number of pressurizer heatups for the life of the plant is 100 for Unit 1 and 90 'for Unit 2. Using these values and the average incremental fatigue usage for the periods in question, the projected'faigue usage'for each location of interest were computed.

Backward Proiections Because the Point Beach plants may have historically operated at a higher system AT than in the template period of available plant data, a sensitivity analysis was performed to account for the possibly higher average incremental fatigue usage in the earlier time period.

Page 13 of 27

WESTINGHOUSE NON-PROPRIETARY CLASS 3 For the time period beforedata was available, some Point Beach heatups and cooldowns were assumed to be at higher maximum system ATs than current operation for conservatism.

This sensitivity analysis was performed by running simulated data with higher ATs (by lowering the hot leg temperature) to6determine a correlation between maximum AT and an-increased fatigue usage factor.

This analysis demonstrated that increases in fatigue usage were small. The hot leg temperature was reduced by 1 000F (which has the effect of increasing the maximum AT by the same amount). This resulted in a' relatively small increase in average incremental fatigue usage for the highest usage location, the pressurizer heater weld.

In most cases the fatigue usage was dependent primarily on the rate of temperature change of the pressurizer water temperature,- which is not expected to change significantly during operationis with different system ATs and relatively few significant insurges and outsurges.

However, it was considered conservative to assume that on the average, increased system ATs during the earlier time frame' resulted in a maximum 50% increase from the current operation s average incremental fatigue-usage. For Unit 1, 53 RCS cooldown cycles occurred before 1994. For Unitf2, 39 RCS cooldown cycles-occurred before 1994.

Proiections The 50% increase was assumed to apply to6the first 53 cycles for Unit 1 (39 for Unit 2).

In reality, only approximately 100 heatup/6cooldown cyclesfor Unit 1 are expected for the plant based on the frequency of past heatup/cooldown occurrences (90 cycles for Unit 2).

The Point Beach operation during' heatup and cooldown is relatively benign because large system temperature differentials (ATs) do not occur. In most cases the AT is less than 150'F. Therefore, any amount of insurge/outsurge-due to flow reversal is unlikely to contribute significantly to future fatigue usage.

Environmental Effects The fatigue usage projections discussed above were adjusted for the maximum effect of environmental fatigue by multiplying by 15.35. The resulting environmental cumulative usage factors are acceptable (less than 1.0)', and are shown in the LRA Table 4.3.10.2.

NRC Question RAI-4.3.5.4 Pressurizer Structural Integrity:

Provide a description of the "Modified Operating Procedures" (page 4-45) that were used to minimize or eliminate in-surge/out-surge cycling.

Page 14 of 27

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Sty e.l NMC Response:

PBNP follows a water solid heatup/cooldown method for both units. The modified operating procedures set a maximumrallowable AT limit of 21 0F between the RCS hot leg and the pressurizer liquid space. This, ensures that operation of the plant is within the AT limit assumed in the surge line thermal stratification analyses.

NRC Question RAI-4.3.5.5 Pressurizer Structural Intecritv:

List the key Units 1 and 2 pressurizer components, and provide for each a summary of 60-year primary-plus-secondary stress' iften'sities and cumulative fatigue usage factors (similar to revised Tables 4.3-1 and 4.3-2 in Appendix A of the LRA for components of the reactor vessel) for these components.

NMC Response:

In accordance with discussions with NRC staff on October 28, 2004 it was agreed that a summary of 60-year primary-plus-secondary stress intensities for the PBNP pressurizers need not be provided since the stressintensities are not critical for the review and are not a part of the PBNP CLB.

A summary of the PBNP pressurizers key component fatigue evaluation results is included in the following table.

PBNP Key Pressurizer Component Design Basis Cumulative Fatigue Results Component Description CUF Upper Head and Shell Lower Head Perforation (1)

Spray Nozzle (1i)

Surge Nozzle Instrument Nozzle...;

1Ai)

Lower Head Heater Well

[

1 Z(i)

Immersion Heater'

[

C(1)

Support Skirt and Flange b1)

Safety and Relief Nozzle bC (1)

Manway Pad Manway Cover Manway Bolts

[

] c BRACKETED NUMBERS ARE WESTINGHOUSE PROPRIETARY (1) -Westinghouse, "Power Uprate Project, Point Beach Nuclear Plant Units 1 and 2, Volume 1, NSSS Engineering Report," April 2002.

Page 15 of 27

WESTINGHOUSE NON-PROPRIETARY CLASS 3 NRC Question RAI-4.3.7.-Pressurizer Sur-e Line Structural Integrity:

Provide a comparison of the CLB set of transient conditions and design cycles, and the revised set of Steam Generator Replacement and Full Power Uprate transient conditions and design cycles, that were used in-the Units 1 and 2 pressurizer surge line fatigue TLAAs to show conformance with.the CLB fatigue limit to the end of the period of extended operation.

NMC Response:

Westinghouse performed the original PBNP surge line thermal stratification evaluations.

The analysis results are documented in WCAP-13509 and WCAP-13510, "Structural Evaluation of the Point Beach Units 1 & 2 Pressurizer Surge Lines, Considering the Effects of Thermal Stratification," dated October -1992. WCAP-1 3509 is the Westinghouse Proprietary.Class 2 version of the analysis, and WCAP-13510 is the Westinghouse Proprietary Class 3 version. Copies of both WCAP-13509 and--

WCAP-1 3510 were transmitted to the NRC inWisconsin Electric' letter VPNPD-92-360, NRC-92-139, dated November 24, 1992, "Docket 50-266 and 50-301, Completion of the Reporting Requirements for Action Item1.'d of NRC IE Bulletin 88-11." The specific analysis transient conditions'and design cycles are detailed in the noted WCAPs.

Westinghouse evaluated the impacts of the cha'nges in the RCS conditions, thermal design transients,'and a 60-year life 'on the BNP surge line thermal stratification analyses. The impact of changes in the revised RCS conditions, thermal design transients, and the 60-year life extension were factored into determining the ASME stress levels and allowable stress levels for the surge line. This evaluation included a review of the fatigue analysis and the stratification loadings that were transmitted to the pressurizer nozzle from the surge lin'e piping.'^The changes and the percent increases for the thermal design transients wereltabulated and the impact on the cumulative fatigue usage factor was calculated. The forces and moments' that were generated by the stratified conditions in'the'surge line also exist at'the pressurizer.nozzle. The power uprate conditions were reviewed to determine if.the old enveloping loads on the nozzle changed significantly. 'Temperature differences between therhot leg and pressurizer were used.to calculate stratified moments'in t e surge line piping. The difference between the old Thot (hot leg temperature) and the new That was determined and used in the determination of new nozzle'loads.

The results of this evaluation for the pressurizer surge line stratification showed that the power uprate conditions changed the-cumulative fatigue usage factor at the location of highest usage factor by. a negligible amount. ITh-ecalculated change in loadings on the pressurizer nozzle due to stratification for the power uprate conditions was not considered significant. The results presented in WCAP-1 3509 and WCAP-1 3510 remain unchanged.

Page 16 of 27

WESTINGHOUSE NON-PROPRIETARY CLASS 3 NRC Question RAI-4.3.8: "

Pressurizer Spray Header Piping Structural Integrity This section states that: "In view of the lack of margin with the Unit 1 piping system analysis results for end of life extension (EOLE),-additional analysis investigations were pursued. The original 88-08 analysis incorporated simplified analysis techniques and assumptions. It was not clear that the analysis-was in fact conservative. The 88-08 analyses were re-performed using the original temperature monitoring data, :and refined analysis techniques and assumptions." -Provide a detailed description and basis of the "refined analyses techniques and assumptions" that were used in the 88-08 reevaluation to reduce the 60-year CUF of 0.99 for the Unit 1 piping system to a 60-year CUF of 0.277.

NMC Response:

The refined analysis techniques and assumptions referred to evaluation of actual thermocouple data and to the use of special purpose programs for performing piping analysis.- Details are discussed-below.'

In late 1989,'as a response to NRC Bulletin 88-08 issues, PBNP installed two sets of three thermocouples on the horizontal section of the Unit 2 main and auxiliary spray piping to detect leakage and thermal statification in the lines. These thermocouple sets were installed near the tee joining the'auxiliary-pressurizer spray line to one of the main spray lines.' Thermally stratified conditions were discovered during heatup and normal operation. ' Using this data, Sargent and Lundy (S&L) performed a 40-year fatigue calculation for both units for the main 'andauxiliary spray'piping, resulting in a fatigue usage in Unit 1 of 0.66 and 0.30 for Unit 2.- The Unit 1 60-year fatigue usage projections yielded a value of 0.99. The-S&L'calculation included a simplified hand calculation of the thermal stratification stresses.- The simplifying assumptions used may or may not have been conservative. Therefore, a further review was performed.

Review of the S&L work to determine a more accurate fatigue usage for a 60-year design life for Unit 1 consisted of:'

1. Reviewing the thermocouple data to verify the basis for the stresses and contributions to fatigue of the operating parameters. The rationale for extrapolating the results of the data collection sample period to the entire plant life was revisited. Contributions to fatigue usage attributed to heatup, cooldown, auxiliary spray actuation,' and thermal stratification due to valve leakage were identified.

2. Reviewing the simplified hand calculation of the thermal stratification stresses done by S&L. These -calculati6ns contained estimates of global bowing moment, radial local stress, and axial local stress. Some of the simplifying assumptions used were conservative and some were possibly not conservative. The correct thermal stratification stresses are determined using the'Structural Integrity (SI) program TOPBOT.

Page 17 of 27

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Thermal stratification data was collected for a 153-day period, beginning with a plant heatup. The thermocouple data indicated that thermal stratification was present during most of the monitoring period, and the magnitude of the top-to-bottom gradient varied over time. The midlevel temperature, assumed to be indicative of the pipe average temperature also varied over time, not only due to plant heatup but also due to variations in spray demand. Thus, two types of thermal cycling occur: global thermal cycling, based on the-mid-pipe temperature variations, which affects the thermal expansion moments in the pipe and the global stratification bowing effects; and thermal gradient cycling, based on the variation-of the difference between the top and bottom pipe temperatures, which affects the local an'd global thermal stratification stresses.

In the S&L analysis, node point 210 in Unit 1 was identified as the limiting location for stress and fatigue usage. Node point 210 is located on the 3-inch line between the first main spray tee and the reducer before the second tee. Because the calculations affect all locations proportionally, point 210 rem'ained limiting and was used for determining the 60-year fatigue usage.

The thermocouple data was reviewed in detail and temperature cycles were constructed. Two types of cycles were constructed-Type A cycles, which are thermal expansion moment cycles based on the midlevel pipe temperature variations; and Type B, local and global thermal stratification cycles, based on the top-to-bottom thermal gradient magnitude variations.' 'Type Acyclesof less than 100'F and type B cycles of less than 500F are'neglected'because of small ATs. The peaks andvalleysof these cycles were paired according to the ASME method of matching highest peak with lowest valley, second highest peak with second lowest valley, etc. This methodology was conservative because higher cyclic ranges produce exponentially higher fatigue usage.

The most severe top-to-bottom thermal stratification temp'erature profile is determined and modeled in the Si program TOPBOT.- TOPBOT calculates the fixed-end thermal bowing moment and the local peak stresses due'to the nonlinear, non-axisymmetric thermal gradient. The bowing moment was compared to that determined by S&L, and estimated piping stresses were scaled acc6rding to the more accurate moment determined by TOPBOT. These results and the local peak stresses determined by TOPBOT, were then scaled to the varied'cycle' amplitudes to develop stress cycles.

These results were used in a revised fatigue usage calculation to determine the projected fatigue usage for a 60-year plant'life.

The thermal stratification appears to have"been caused by leakage past auxiliary spray isolation valve CV-296. Although leakage pa'st CV-296 may initially be hot as it is taken from the charging syste'm, the flow is sufficientiy small and the valve is far enough away from the main spray tee (i.e., 80 feet) that the leakage mixes with the stagnant fluid in the auxiliary spray line and arrives at the'tee'at roughly containment ambient temperature. Although preventive maintenance, including changes to the internals, was performed on valve CV-296 after measurement 'of the stratification, and it was unlikely that leakage continued to occur to the degree that was measured, for purposes of bounding analysis it was conservatively assumed that the same amount of leakage and consequent magnitude of stratification continued to exist for the remainder of plant life.

Page 18 of 27

WESTINGHOUSE NON-PROPRIETARY CLASS 3 It was also assumed that the 153-day period of data was representative of all plant operation. This assumption was conservative because this 153-day period included a heatup, which contained many more stress cycles than did normal steady state operation. The heatup was also considered representative of a cooldown, because the mean pipe temperature range is the same, and there are multiple main spray and auxiliary spray actuations, as well as numerous variations in spray flow rate. Heatups tend to contain more thermal cycles than cooldowns, as there are typically more procedural steps, hold points, and tests conducted than during cooldown.

To extrapolate the 153-days of cycles to 60 years of operation, the following assumptions were used:

• The cycles are considered to be repeated every 153 days of operation, despite the fact that plant heatups do not occur this often

• The plant was not operated for one month per two years due to refueling outages Using these assumptions, S&L calculated the global thermal stratification fixed-end moment, and the local thermal stratification stresses due to the nonlinear top-to-bottom thermal gradient. These calculations were done by hand and were somewhat simplified in that they assumed that the thermocouple temperatures measured at the outside of the pipe were representative of the inside fluid temperatures, and did not account for the through-wall thermal gradients. A more accurate thermal stratification stress analysis was done using the SI program TOPBOT.

TOPBOT solves the transient thermal and stress response within a pipe subjected to a step or ramp change in boundary temperatures and heat transfer coefficients. Initial temperature conditions are specified, and then a temperature change is applied to either the top or bottom fluid in the pipe, or both. The pipe is considered to be two dimensional at a pipe cross-section (or assumed to be extremely long in the axial direction). Symmetry about the pipe vertical centerline is assumed. The pipe is modeled with rectilinear elements within the R-theta coordinate system. Stresses and temperatures are computed at the center of each element (mean radius and mean angular locations). In addition, temperatures and stresses are computed at the inside and outside surfaces of the pipe, based on the "steady state" temperature distribution between the surface elements and external boundary temperatures. Required thermal input parameters include thermal conductivity and the product of the density and specific heat, modulus of elasticity, coefficient of thermal expansion, and Poisson's ratio.

The thermal boundary conditions are input as internal and external temperature distributions and heat transfer distributions. For most problems, the initial pipe temperature is uniform: a uniform temperature and heat transfer coefficient is specified on the outside of the pipe, and two sets of temperature and heat transfer coefficients are specified inside the pipe, representing the top and bottom temperatures and flow rates. These two sets of temperatures, heat transfer coefficients, and their interface level can be varied linearly by specifying'different values at specific points in time. The pipe temperature distribution is determined using a classical finite difference method.

An energy balance is written for each element of the model.

Page 19 of 27

.:V WESTINGHOUSE NON-PROPRIETARY CLASS 3 For this analysis, the maximum top-to-bottom temperature distribution was modeled.

The stress results were scaled for smaller thermal gradients. The outside pipe temperatures at the top, midlevel and bottom -of the pipe were observed. In order to determine accurate thermcal 'stratification stresses, the insid6'fluid temperatures that produce the temperature distribution measured at the outside of the pipe were determined. This required some trial and error, as it is a function of the hot-cold fluid interface level and the convective heat transfer coefficients at the top and bottom inside surface of the pipe; these coefficients in turn-'depend on the flow rates and temperatures of the fluid levels.

The inside surface forced convection heat transfer coefficients were determined using the following relation for turbulent flow:'

h = 0.023 Re0 8 Pr0 4 k I D where Re = Reynolds number = pVD/p > 4000 for turbulent'flow Pr = Prandtl number = pc k k = thermal'conductivity, BTU-hr-fttF' D = hydraulic diameter = 4A/P A = flow area P = flow perimeter V = flow velocity, ft/sec p = density, Ibm/h3 p = dynamic viscosity, Ibm /ft-sec cp= specific heat at constant pressure After trial and error, the outside pipe temperature distribution was replicated if the inside fluid temperatures were 530 0F at the top (stom'e heat loss from RCS cold leg temperature) and 1 00F at the bottom (containment ambient temperature), the interface level was at 1480-from the top of the pipe;-and the top fluid velocity was 0.516 ft/sec, or 3.1 gpm (bypass flow circulation) and the bottom fluid velocity was 0.19 ft/sec, or 0.3 gpm (leakage past auxiliary spray control valve). The actual valve leakage'and the other parameters may slightly deviate, -but'as long as the temperature distribution is established in the analytical model,fthe stress 'results remain representative.

Thermal Stratification-Analysis Results -

The results of the TOPBOT thermal stratification analysis were the following:

Fixed-End Moment: 108.19 in-kip (for a 300TF top-to-bottom thermal gradient)

Local Stress (maximum location): 40.31 ksi (for a 300OF top-to-bottom thermal gradient)

= 134.37 psV0F Page 20 of 27

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Stress Combinations and Fatigue Usaae The stress and fatigue usage were computedat the limiting location, node point 210.

This location is classified as ASME Code Class 2. ASMESedction l1l, subsection NC-3600, does not provide explicit fatigue usage criteria; however, the approach used by Markl in his fatigue testing of piping components was used, which has been implicitly adopted in NC-3600:

iS (N)0 2 = 280,000 where S =stress range, psi i = stress intensification factor = 1.0 at node point 210 (straight pipe)

N = number of allowable full range",stress-cycles at stress S The stress is the total of the contributions from thermal expansion moments, global thermal stratification, and local thermal stratification.

The thermal expansion moment stress at point 210 was (32113 -'6960) - 25,153 psi.

b

.ss at-d'V as-6 6-

s. "

't's was The global thermal stratification stress calculated was 6960 psi. This stress was obtained by applying-the stratification fixed-end moment to the piping model. A more accurate fixed-end moment was 'calculated'by TOPBOT as described above.

Therefore, the stress adjusted by the ratio of the fixed-end moments is:

Global stratification stress = 6960 (1 08.19/1 04.34) = 7,217 psi.

Per the S&L analysis; the magnitude of stratification at point 210 is 0.93 of that measured at the'instrumented point. The iocal stress is:

= 134.37 psiI0F (300) (.93) 37,489 psi.,

The stress cycles are grouped in the following manner:

1. 1x140=140 cycles of 4550F thermai expansion moment range + 3000F global stratification + 300OF local stratification.

2. 2x140=280 cycles of 1500F thermal expansion moment range + 3000F global

-.I.

I

.IZ r.-.

stratification + 3000F local stratification.

3. 1x140=140 cycles of 1500F thermal expansion moment range + 2300F global stratification + 2300F local stratification.

4. 16x140=2240 cycles of 2300F global stratification + 2300F local stratification.

5. 11 xl40=1540 cycles of 110F global stratification + 110 0F local stratification.

Page 21 of 27

WESTINGHOUSE NON-PROPRIETARY CLASS 3 The thermal expansion moment stress is scaled according to the temperature range of the cycle. For example, for a 1500F range, the thermal expansion range is:

= 25,153 (150/455) = 8,292 psi.

The stratification stresses, global and local, are scaled according to the top-to-bottom gradient. Thus for a 2300F gradient, the global stratification stress is:

= 7,217 (230/300) = 5,533 psi.

and the local stratification stress is:

= 37,489 (230/300) = 28,742 psi.

The number of stress cycles for groups two through five were then converted into an equivalent number of full range stress cycles using equation (2) in paragraph NC-361 1.2 (e) (3) of the ASME'B&PV Code. This relationship was developed by Markl in his fatigue testing of piping components and has been incorporated into the ASME Class 2/3 piping Code:

N = N1 + (S2/Sl)5 N2 + (S31S1)5 N3 + (SdSS) 5 N4 + (S5/S1)5 N5 Thus, following this' approach, 280 cycles of group 2 are equivalent to 70.36 full range cycles, for example. The total equivalent full range cycles of all five groups is 287.

Using Markl's equation, the allowable numberof cycles at a stress of 69,859 psi is 1,034. Therefore, the total fatigue usage is 0.277.

Conclusions The result of the stress and fatigue analysis is that for the limiting location on the Unit 1 pressurizer spray line, the calculated fatigu& u'sage for a 60-year plant design life is 0.277. This is well below the allowable of 1'.0; The stresses conservatively assume that a significant 'amount of leakage past theauxiliary spray control valve continues to exist.

Thus the stress and fatigue usage for this line= is acceptable for a 60-year design life.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 NRC Question RAI-4.3.10.1 Environmental Effects on Fatigue:

For the USAS B31.1 locations, provide a description of the PBNP-specific simplified ASME Section III fatigue analyses that were used to calculate environmentally based cumulative usage factors.

NMC Response:

Charging Nozzle This location was analyzed using actual plant data and the projected number of cycles in the charging nozzle model included in a FatiguePro application for PBNP and the tensile strain-integrated environmental correction factor (Fen) factor.

The PBNP design basis includes several design transients that affect the charging nozzles. The most severe of these events is the loss of charging and loss of letdown with delayed return to service. Examination of actual PBNP Units 1 and 2 plant data since 1994 revealed one actual event that accurately represents this design transient.

This event was used to represent a bounding loss of charging/loss of letdown event.

Normal fatigue usage for this event was computed. An Fen was then determined to account for environmental effects. The resulting total usage was applied over the expected number of occurrences for this event (17 loss of charging and loss of letdown events projected for the 60-year operating life of each units). The incremental fatigue usage attributed to the loss of charging/loss of letdown event was determined to be 0.00695. The bounding transient is depicted in the following figure.

Transient Instrument Values and Stress for Bounding Transient Charging Notxde - Point Beach Unit 2 CHRGLNOZ

+-

ACOLD A,

StrCHRGLNOZ aD CD E2 T126 -

FT128C 412.286 212.286 12.2861-

-187.714 -

-387.714

-587.714 62.1501 32.1501

-n a

2.15007

°0 CD

-27.8499 CL:

-57.8499 (

-87.8499 MT134C

+

-117.85 0331 Datefrime [12f201200a]

Page 23 of 27 CM

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Using these results, the 60-year projection for these events (no environmental effects) is:

CUF = (17 events) * (0.00695 per event) 0.11815 Environmental Effects - F n Approach The fatigue usage analysis above does not consider environmental effects on the fatigue curve. An environmental fatigue factor (Fen) will be determined based on the equations provided in NUREG/CR-5704. -Using this methodology, the Fen factor is computed as a function of three parameters using the following equation:

Fen = exp (0.935 - T*. Edot*

• O*)

where:

T*=0 T<200C T* = 1 T 2 200C Edot* = 0 Edot > 0.4%/sec Edot* = In (Sdot/0.4) 0.0004 < Edot

• 0.4%/sec Edot* = In (0.0004/0.4)

Edot < 0.0004%/sec; O* = 0.260 DO < 0.05 ppm O* =0.172 DO20.05ppm For'this transient, the factors are:

• Strain range - over the tensile portions of the plant transient

• Strain rate - computed over the range of tensile strain range Temperature - conservatively assumed to be greater than 2000C Dissolved oxygen - conservatively assumed to be less than 0.05 ppm Individual Fen values were integrated over the tensile strain range of the transient stress cycle(s) being analyzed. For the purpose of determining strain rate, the stress changes for eachltime step 'with increasing stress were converted to strain by dividing by the modulus of elasticity for stainless steel (28.3 x 103 ksi) from the ASME fatigue curve.

The strain difference was divided bytheqength of the time step, then converted into units of (% strain/sec).

For each time step, incremental Fen was computed using the equation shown above.

An effective Fen for the entire transient was computed by integrating Fenk for each time step with the strain step associated with that time step over the entire tensile strain range. The effective Fen was computed to be 6.994.

Page 24 of 27

-I I F I We 1

WESTINGHOUSE NON-PROPRIETARY CLASS 3 The environmentally-assisted fatigue of less than'1.0 projected to 60 years of operation is thus:

CUFEAF = 0.11815

• 6.994 = 0.8264 Safety Iniection Nozzle and RHR Tee These two locations were analyzed using design transients and the design number of cycles in a combined ASME Code Section' Ill NB-3600 Class 1 plant-specific piping model of the safety injection (SI) piping system and RHR system.

The SI piping, including the RHR tee, was modeled using the computer program, PIPESTRESS. To perform an ASME Section III Class 1 piping fatigue analysis with PIPESTRESS, thermal transients and thermal expansion cases were defined.

To evaluate the fatigue usage with PIPESTRESS, the thermal expansion cases correspond to the final temperature of each analyzed transient plus the steady state operating'case.

Typically, the governing contribution to-fatigue usage is from thermal stresses, not pressure stresses. Thus, the pressure' for each case will be the operating pressure.

Use of the operating pressure is conservative'when used to calculate pressure stresses.

This will also be conservative when used-in the fatigue evaluation.

Modes of operation were defined. The forces and moments due to differential thermal expansion as analyzed by the piping progra m, PIPESTRESS, were included in the fatigue evaluation.

The number of design transients that will occur through the end of the extended operation (60 years) were assumed to be le.ss than or equal to the design limit (40 years) for each design transient. Seismic loading and thermal anchor movements were considered for fatigue analysis.

The results of the fatigue evaluation show that the fatigue usage at the SI to cold leg branch connection (SI Nozzle) is 0.0013 and the tee from the RHR to the SI (RHR Tee) is 0.0146. Application-of the maximum' possible Fen of 15.35 produced environmentally assisted fatigue usage values of 0.02 and 0.224,'for the SI Nozzle and RHR Tee, respectively result in values well below 1.0.'

NRC Question RAI-4.3.10.2:

Environmental Effects on Fatigue The Pressurizer CUFs are determined based on EPRI MRP-47 methodology. The staff has not endorsed MRP-47. Provide the'e'nvironmentally assisted CUFs for the Pressurizer locations, based on the staff-accepted methodology as stated in Sections 4.3.2.2 and 4.3.3.2 of NUREG-1 800.

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WESTINGHOUSE NON-PROPRIETARY CLASS 3 NMC Response:

The pressurizer components were evaluated for'environmental effects on fatigue per the direction of Applicant Action Item 3.3.1.1-1 of the NRC SER of Westinghouse GTR WCAP-14574-A. Applicant Actionh Item 3.3.1.1-1 of the NRC SER of Westinghouse GTR WCAP-14574-A'does-not specify a requisite method for addressing the environmental effects on fatigue. =NUREG-1800 does not require that pressurizer components be evaluated for environmental effects on fatigue.

Sections 4.3.2.2 and 4.3.3.2 of NUREG-1800 note that formulas for calculating the environmental life correction factors are those contained in NUREG/CR-6583 for carbon and low-alloy steels, and in NUREG/CR-5704 for austenitic stainless steels.

Application of the NUREG/CR-6583 and 5704 formulas for calculating the environmental life correction factors for the' PBNP pressurizers, versus the EPRI MRP47 Appendix B formulas will Snot have an significant impact on the pressurizer components evaluations. This Iis-because the EPRI MRP-47 "Z" factor was not used in the evaluations, the evaluation temperature of 345 OC results in the stainless steel formulas being identical between the'twomethods, and the carbon or low alloy steel formulas were not applied since the dissolved oxygen (DO) values were well below the threshold values.

The environmentally adjusted fatigue (EAF) CUF for the spray nozzle safe end is not affected since this location is stainless'steel and the formulas are identical in the analyzed temperature range.

The 60-year EAF CUF for the surge nozzle safe end is not affected since this location is stainless steel and the formulas are identical in the analyzed temperature range.

The 60-year EAF CUF for the limiting 'carbon steel portion of the surge nozzle was noted to be the same as the design basis value since the DO level is well below the threshold level. Using the NUREG/CR-6583 formula, the design basis value would be multiplied by an Fen of 1.17 as a result of the temperature, regardless of the low DO level. 'The resulting 60-year EAF CUF would be 0.73, which is acceptable since it is less than 1.0.

The 60-year EAF CUF for the carbon steel /low alloy steel junction of the upper head and shell was noted to be the same as the design basis value since the DO level is well below the threshold level. Using the NUREG/CR-6583 formula, the design basis value would be multiplied by an Fen of 1.65 as a'result of the temperature, regardless of the low DO level. The resulting 60-year EAF CUF would be 1.28. Sinrce the evaluation for the carbon steel / low alloy steel junction of the upper head and shell is based on the design transient set, the results of the evauuation are extremely conservative in both-the severity and numbers of transients. Significant reductions in the estimates are possible if adjustments are made to remove the operational transients that are not experienced or practiced at PBNP. The EPRI FatiguePro software program was customized to monitor the carbon steel to low alloy steel junction of the upper head and shell at PBNP.

An analysis was performed based on available template sets of real plant data to Page 26 of 27

WESTINGHOUSE NON-PROPRIETARY CLASS 3 determine the incremental fatigue usage factor for known plant transients. A cumulative usage factor for the operating life of the plant was computed based on the results of real plant data and expected future usage was computed using projections of expected plant cycles. The 60-year CUFjofrthe Unit 1'pressurizer's carbon steel /low alloy steel junction of the upper head and shell bounds the 2 units and is projected to be 0.156.

Applying the maximum environmental fatigue correction factor of 2.53 to the projected CUF-of the carbon steel / low alloy steel junction of the upper head and shell location, results in a conservative 60-year EAF CUF of 0.39. This demonstrates adequate structural integrity, including the effects of environmental conditions, for a projected 60-year operational period.

The 60-year EAF CUF for the safety and relief nozzle safe ends was noted to be the same as the design basis value since the CUF of these components is zero.

The 60-year EAF CUF for the limiting carbon steel portion of the safety and relief nozzle was noted to be the same as the design basis value since the DO level is well below the threshold level. Using the NUREG/CR-65836formula, the design basis value would be multiplied by an Fen of 1.17-as a result of the temperature, regardless of the low DO '

level. The resulting 60-year EAF CUF would be 0.174, which is acceptable since it is less than 1.0.

The environmentally adjusted fatigue (EAF) CUF for the instrument nozzle is not affected since this location is stainless steel and the formulas are identical in the analyzed temperature range.

The environmentally adjusted fatigue (EAF) CUF for the heater well is not affected since this location is stainless steel and the formulas are identical in the analyzed temperature range.

Page 27 of 27