HL-4877, Forwards Core Shroud Stabilizer Design Submittal,Providing Analyses & Conclusions Supporting Implementation of Facility Shroud Repair.Ge Proprietary Info Also Encl.Ge Proprietary Info Withheld Per 10CFR2.790

Forwards Core Shroud Stabilizer Design Submittal,Providing Analyses & Conclusions Supporting Implementation of Facility Shroud Repair.Ge Proprietary Info Also Encl.Ge Proprietary Info Withheld Per 10CFR2.790

ML20086E798
Person / Time
Site:	Hatch
Issue date:	07/03/1995
From:	Beckham J GEORGIA POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML19325F570	List: HL-4877, Forwards Core Shroud Stabilizer Design Submittal,Providing Analyses & Conclusions Supporting Implementation of Facility Shroud Repair.Ge Proprietary Info Also Encl.Ge Proprietary Info Withheld Per 10CFR2.790 ML20086E812 ML20086E814 ML20086E824 ML20086E831 ML20086E847
References
HL-4877, TAC-M90095, NUDOCS 9507120262
Download: ML20086E798 (42)

Text

r- s 1 1

Georgia Power Company 40 inverness Center Parkway Post Office Dox 1295 ~

Birmingham. A!abama 35201

. Telephone 205 877 7279 hice krosIdenWN clear GeorgiaPower Hatch Project te soutten ectx s>stan July 3, 1995 Docket No. 50-366 HL-4877 TAC No. M90095 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555 Edwin I. Hatch Nuclear Plant - Unit 2 Core Shmud Stabilizer Desipt Submittal Gentlemen:

By letter dated July 25,1994, the Nuclear Regulatory Commission (NRC) staffissued Generic Letter (GL) 94-03 concerning core shroud cracking in boiling water reactors. By letter dated August 24,1994, Georgia Power Company (GPC) provided a response to GL 94-03 indicating plans.to implement a permanent repair on the Edwin I. Hatch Nuclear Plant Unit 2 during the Fall 1995 refueling outage. A permanent repair was installed on Unit I during the Fall 1994 refueling outage.

Georgia Power Company has determined that the proposed core shroud repair is not '

included under the American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, definition for repair or replacement. Therefore, pursuant to 10 CFR 50.55a(a)(3), GPC is submitting, for NRC staff review and approval, the design details of the shroud repair as an alternative repair. The enclosed documentation provides the analyses and conclusions supporting implementation of the Unit 2 shroud repair. It should be noted that the Unit 2 design differs from the previously installed Unit I design primarily due to the difference between the units Design Basis Earthquakes and GPC's objectives for increasing design margins on the Unit 2 repair.

In accordance with 10 CFR 50.59, GPC has completed an evaluation ofchanges made to the facility as described in the Final Safety Analysis Report. GPC has concluded that the shroud repair does not involve an unreviewed safety question.

Please be advised that this submittal contains information considered proprietary by the General Electric Company (GE). In accordance with the provisions of 10 CFR 2.790, I GPC requests that the proprietary information be withheld from public disclosure. The proprietary information has been so designated and the required affidavit is enclosed. I i

• 2()6E Il

/Wr. f#/,

7 95o7120262 950703 DR ADOCK 0500 6

}l [g / /

r Georgia Powerkh i

U.S. Nuclear Regulatory Commission Page 2 July 3, 1995 Additionally, GPC is aware of the NRC staff s concerns related to generic items which the staff considers to have been inadequately addressed in previous core shroud repair design submittals. These concerns were expressed during a meeting between the Boiling Water Reactor Vessel Internals Project and the NRC staff held on June 22,1995. The concerns are addressed in the attached documents.

1 Georgia Power Company respectfully requests an elevated priority for NRC staff review to support the planned implementation during the Fall 1995, Unit 2 refueling outage. In order to support the current schedule, approval is required no later than September 20, 1995. GPC is available to meet with the NRC staff or expeditiously supply additional information as requested.

Should you have any questions in this regard, please contact this office.

Sincerely, J. T. Beckham, Jr. I JKB/eb  ;

Enclosure:

Core Shroud Stabilizer Design Submittal- Unit 2 Attachments:

1. GENE-B11-00637-006, Revision 0,"GE Responses to NRC Questions, Hatch Unit 2 Shroud Repair," June 1995

2. GENE Specification 25A5718, Revision 0, " Shroud Repair Hardware, Design l Specificstion," Hatch Unit 2, May 1995

3. GENE Specification 25A5717, Revision 1," Shroud Stabilizers, Code Design l Specification," Hatch Unit 2, June 1995 j i

4. GENE-B11-00637-003, Revision 0,"Edwin I. Hatch Nuclear Plant Unit 2 Shroud Repair Seismic Analysis Report for OBE, DBE & 1/2 SME," May 1995 (Continued on next page.)

HL-4877

y

. Georgia Power d U.S. Nuclear Regulatory Commission Page 3 July 3, 1995 Attachments: (Continued)

5. GENE-B11-00637-002, Revision 0, " Shroud Mechanical Repair Program, Hatch Unit 2, Shroud and Shroud Repair Hardware Stress Analysis," June 1995

6. GENE Specification 25A5721, Revision 0, " Shroud Stabilizers, Stress Report, Hatch Unit 2," June 1995

7. GENE-B11-00637-005, Revision 0, " Safety Evaluation for the Installation of Stabilizers on the Edwin I. Hatch Nuclear Plant Unit 2 Core Shroud"

8. GENE Specification 25A5719, Revision 0," Fabrication of Shroud Stabilizer, Fabrication Specification, Hatch Unit 2," June 1995

9. GENE FDI No. HT2-0121-12900, Revision 0, " Field Disposition Instruction, Hatch Unit 2, Shroud Repair Program," June 1995 cc: Georgia Power Comrumy Mr. H. L. Sumner, Jr., Nuclear Plant General Manager NORMS U. S. Nuclear Regulatorv Commission. Washington. D. C.

Mr. K. Jabbour, Licensing Project Manager - Hatch U. S. Nuclear Remdatorv Commission. Rezion H Mr. S. D. Ebneter, Regional Administrator Mr. B. L. Holbrook, Senior Resident Inspector - Hatch HL-4877

e- ,

l General Electric Company AFFIDAVIT I, George II. Stramback, being duly sworn, depose and state as follows:

(1) I am Project Manager, Licensing Services, General Electric Company ("GE") and have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding.

(2) The information sought to be withheld is contained in the GE proprietary report GENE-B 11-00637-006, GE Response to NRC Questionsfor Hatch Unit 2 Shroud Repair, Revision 0, (GE Proprietary), dated June 1995. The proprietary information is delineated by bars marked in the margin adjacent to the specific material.

(3) In making this application for withholding of proprietary information of which it is the owner, GE relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4), and the Trade Secrets Act,18 USC Sec. 1905, and NRC regulations 10 CFR 9.17(a)(4), 2.790(a)(4), and 2.790(d)(1) for " trade secrets and commercial or financial information obtained from a person and privileged or confidential" (Exemption 4). The material for which exemption from disclosure is here sought is all " confidential commercial infbrmation", and some portions also qualify under the narrower definition of" trade secret", within the meanings assigned to those terms for purposes of FOfA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission. 975F2d871 (DC Cir.1992), and Public Citizen llealth Research Groun

v. FDA,704F2dl280 (DC Cir.1983).

(4) Some examples of categories of information which fit into the definition of proprietary information are:

a. Infom1ation that discloses a process, method, or apparatus, including supporting data and analyses, where prevention ofits use by General Electric's competitors without license from General Electric constitutes a competitive economic advantage over other companies;

b. Infomiation which, if used 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 of a similar product; GllS-95-7-afilaRAll. doc Aflidavit Page 1

r j

c. Information which reveals cost or price information, production capacities, budget levels, or commercial strategies of General Electric, its customers, or its suppliers;

d. Information which reveals aspects of past, present, or future General Electric customer-funded development plans and programs, of potential commercial value to General Electric;

c. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

The information sought to be withheld is considered to be proprietary for the reasons set forth in both paragraphs (4)a. and (4)b., above.

(5) The information sought to be withheld is being submitted to NRC in confidence.

The information is of a sort customarily held in confidence by GE, and is in fact so held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by GE, no public disclosure has been made, and it is not available in public sources. All disclosures to third parties including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following.

(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge. . Access to such documents within GE is limited on a "need to know" basis.

(7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing function (or his delegate), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GE are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitirnate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements.

(8) The information identified in paragraph (2), above, is classified as proprietary because it contains detailed results of analytical models, methods and processes, including computer codes, which GE has developed and applied to perform evaluations of the shroud and a hardware modification for the BWR. This report evaluates a hardware design modification (stabilizer for the shroud horizontal welds) intended to be installed in a reactor to resolve the reactor pressure vessel core shroud GBS-95-7-afilaRAll. doc Affidavit Page 2

r

b. .

weld cracking concern. The development and approval of this design modification utilized systems, components, and models and computer codes that were developed at a significant cost to GE, on the order of several hundred thousand dollars.

The development of the supporting processes was at a significant additional cost to GE, in excess of a million dollars, over and above the large cost of developing the underlying individual proprietary report information.

(9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GE's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of GE's comprehensive BWR safety and technology base, and its commercial value extends beyond the original development cost. The value of the technology base goes beyond the extensive physical database and analytical methodology and includes development I of the expertise to determine and apply the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses ,

done with NRC-approved methods.

The research, development, engineering, analytical and NRC review costs comprise ,

a substantial investment of time and money by GE.

The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial.

GE's competitive advantage will be lost ifits competitors are able to use the results of the GE experience to normalize or verify their own process or if they are able to ,

claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

The value of this information to GE would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive GE of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing these very valuab!c analytical tools.

I GilS-95-7-af11aRAll. doc Affidavit Page 3

STATE OF CALIFORNIA )

) ss:

COUNTY OF SANTA CLARA )

George B. Stramback, being duly swom, deposes and says:

That he has read the foregoing affidavit and the matters stated therein are true and correct to the best of his knowledge, information, and belief.

Executed at San Jose, California, this 9 4 day of bl4d 1995.

6 b4M// 0 hksib

' Georg/B. Strdnback General Electric Company Subscribed and sworn before me this29M day ofb + 1995.

//

"6tary Public, State of Caiifor

- JULfE A. CURTS j..

• COMM. # 974657 E; E Notcry Public - Cofifomia 9

~

3 SANTA CLARA COUNTY My Comm Empires SEP 30.1996 GBS-95-7-afilaRAll. doc Aflidasit Page 4

n l s i General Electric Company AFFIDAVIT I, George H. Stramback, being duly sworn, depose and state as follows:

(1) I am Project Manager, Licensing Services, General Electric Company ("GE") and have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding.

(2) The information sought to be withheld is contained in the GE proprietary report GENE-B l 1-00637-003, Edwin 1. Hatch Nuclear Plant Unit 2 Shroud Repair Seismic Analysis Reportfor OBE, DBE & 1/2 SME, Revision 0, (GE Proprietary), dated May 25,1995. The proprietary information is delineated by bars marked in the margin adjacent to the specific material.

(3) In making this application for withholding of proprietary information of which it is the owner, GE relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4), and the Trade Secrets Act,18 USC Sec.~ 1905, and NRC regulations 10 CFR 9.17(a)(4), 2.790(a)(4), and '

2.790(d)(1) for " trade secrets and commercial or financial information obtained from a person and privileged or confidential" (Exemption 4). The material for which exemption from disclosure is here sought is all " confidential commercial information", and some portions also qualify under the narrower definition of " trade secret", within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Encray Project v. Nuclear Regulatory Commission. 975F2d871 (DC Cir.1992), and Public Citizen Health Research Group -

v. FDA,704F2dl280 (DC Cir.1983).

(4) Some examples of categories of information which fit into the defm' ition of proprietary information are:

a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by General Electric's competitors without license from General Electric constitutes a competitive economic advantage over other companies;

b. Information which, if used 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 of a similar product; i GBS-95-6-aniatch3. doc Affidavit Page 1

< _ J

c. Information which reveals cost or price information, production capacities, budget levels, or commercial strategies of General Electric, its customers, or its suppliers;

d. Information which reveals aspects of past, present, or future General Electric customer-funded development plans and programs, of potential commercial value to General Electric;

c. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

The information sought to be withheld is considered to be proprietary for the reasons set forth in both paragraphs (4)a. and (4)b., above.

(5) The information sought to be withheld is being submitted to NRC in confidence.

The information is of a sort customarily held in confidence by GE, and is in fact so held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by GE, no public disclosure has been made, and it is not available in public sources. All disclosures to third parties including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following.

(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge. Access to such documents within GE is limited on a "need to know" basis.

(7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing function (or his delegate), and by the legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GE are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements.

(8) The information identified in paragraph (2), above, is classified as proprietary because it contains detailed results of analytical models, methods and processes, including computer codes, which GE has developed and applied to perform evaluations of the shroud and a hardware design modification (stabilizer for the shroud horizontal welds) intended to be installed in a reactor to resolve the reactor pressure vessel core shroud weld cracking concern. The development and approval GBS-95-6-af11atch3. doc Affidavit Page 2

of this design modification utilized systems, components, and models and computer codes that were developed at a significant cost to GE, on the order of several hundred thousand dollars.

The development of the evaluation process along with the interpretation and application of the analytical results is derived from the extensive experience database that constitutes a major GE asset.

(9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GE's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of GE's comprehensive BWR safety and technology base, and its commercial value extends beyond the original development cost. The value of the technology base goes beyond the extensive physical database and analytical methodology and includes development of the expenise to determine and apply the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses done with NRC-approved methods.

The research, development, engineering, analytical and NRC review costs comprise a substantial investment of time and money by GE.

The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial.

GE's competitive advantage will be lost if its competitors are able to use the results of the GE experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

The value of this information to GE would be lost if the information were disclosed to the public. Making such inforrnation available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive GE of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing these very valuable analytical tools.

l l

l GilS-95-6-afHatch3. doc Affidavit Page 3

e 1

i STATE OF CALIFORNIA )

) ss:

COUNTY OF SANTA CLARA )

George B. Stramback, being duly sworn, deposes and says:

That he has read the foregoing affidavit and the matters stated therein are true and correct ,

to the best of his knowledge, information, and belief.

Executed at San Jose, California, this 2 6 bday of 7he 1995.

/

Mb0 W l Ge6'rge %Stramback General Electric Company Subscribed and sworn before day meof this hIW85v rk. 1995.

km Y Notary Public, State of California A'Ak H. MARY L KENOALL COMM. # 967864 3 5

2 sT{jfh.

tg, Notory Public - Confomio >

1 K,. SANTA CLARA COUNTY f j My Comm. Empires MAR 26,1997 ) 1

<v -,

I l

l GBS-95-6-alliatch3. doc Affidavit Page 4 l

General Electric Company AFFIDAVIT i

I, George H. Stramback, being duly sworn, depose and state as follows: j l

(1) I am Project Manager, Licensing Services, General Electric Company ("GE") and have been delegated the function of reviewing the informatien described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding. l l

1 (2) The information sought to be withheld is contained in the GE proprietary report i GENE-B11-00637-002, Shroud Mechanical Repair Program Hatch Unit 2 Shroud l and Shroud Repair Hardware Stress ifnalysis, Revision 0, (GE Proprietary Information), dated June 1995. The proprietary information is delineated by bars marked in the margin adjacent to the specific material.

(3) In making this application for withholding of proprietary information of which it is l the owner, GE relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4), and the Trade Secrets Act,18 ,

USC Sec.1905, and NRC regulations 10 CFR 9.17(a)(4), 2.790(a)(4), and l 2.790(d)(1) for " trade secrets and commercial or financial information obtained from l a person and privileged or confidential" (Exemption 4). The material for which exemption from disclosure is here sought is all " confidential commercial 1 information", and some portions also qualify under the narrower definition of" trade secret", within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory i Commission. 975F2d871 (DC Cir.1992), and Public Citizen IJealth Research Groun I

v. FDA,704F2d1280 (DC Cir.1983).  !

(4) Some examples of categories of information which fit into the definition of proprietary information are: 3 i

3

a. Information that discloses a process, method, or apparatus, including supporting j data and analyses, where prevention ofits use by General Electric's competitors '

without license from General Electric constitutes a competitive economic advantage over other companies;

)

b. Infonnation which, if used 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 of a similar product;  ;

1 GBS-95-7-alliamod3. doc Affidavit Page I

o- .

c. Information which reveals cost or price information, production capacities, budget levels, or commercial strategies of General Electric, its customers, or its suppliers;

d. Information which reveals aspects of past, present, or future General Electric customer-funded development plans and programs, of potential commercial value to General Electric;

c. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

The information sought to be withheld is considered to be proprietary for the reasons set forth in both paragraphs (4)a., (4)b. and (4)e., above.

(5) The information sought to be withheld is being submitted to NRC in confidence.

The information is of a sort customarily held in confidence by GE, and is in fact so held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by GE, no public disclosure has been made, and it is not available in public sources. All disclosures to third parties including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following.

(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge. Access to such documents within GE is limited on a "need to know" basis.

(7) The procedure for approval of external release of such a document typically requires review by the stafr manager, project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing function (or his delegate), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GE are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, l and licensees, and others with a legitimate need for the information, and then only m  !

accordance with appropriate regulatory provisions or proprietary agreements.

(8) The information identified in paragraph (2), above, is classified as proprietary because it contains detailed results of a hardware design modification (stabilizers for the shroud horizontal welds) intended to be installed in a reactor to resolve the reactor pressure vessel core shroud weld cracking concern. The development and approval of this design modification utilized system, component, and models and GBS-95-7 afilamod3. doc Affidavit Page 2 l

m

F 4

computer codes that were developed at a significant cost to GE, on the order of several hundred thousand dollars.

The development of the supporting processes was at a significant additional cost to GE, in excess of a million dollars, over and above the large cost of developing the underlying individual proprietary report information.

(9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GE's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of GE's comprehensive llWR safety and technology base, and its commercial value extends beyond the original development cost. The vah:e of the technology base goes beyond the extensive physical database and analytical methodology and includes development of the expertise to determine and apply the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses done with NRC-approved methods.

The research, development, engineering, analytical and NRC review costs comprise a substantial investment of time and money by GE.

The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial.

GE's competitive advantage will be lost ifits competitors are able to use the results of the GE experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

The value of this information to GE would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive GE of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing these very valuable analytical tools.

GBS-95-7-aQiamod3. doc Affidavit Page 3 t

i o .,

STATE OF CALIFORNIA )

) ss:

l COUNTY OF SANTA CLARA )

George B. Stramback, being duly swom, deposes and says:

I That he has read the foregoing affidavit and the matters stated therein are true and correct l to the best of his knowledge, infonnation, and belief. i Executed at San Jose, Califomia, this /1 day of oms 1995. l l

p h ht4 Gdfirge B. Stramback General Electric Company Subscribed and swom before me this Ikb day of Ob& 1995.

O li

..'^^] O A I h tsife )q

~

PAULA F, HUSSEY Notary Public, State of Californf y COMM. d1046)20 g 3,-

3,,, we - CovoWo h 4 l

$ ANT A CLARA COUNTV MyComm W *O'C --

f, . ,,,,,

1 I

GBS-95-7-alliamod3. doc Affidavit Page 4 t . . . . . .

r 1

,2 . ;, .

l ,

General Electric Company AFFIDAVIT I, George 11. Stramback, being duly sworn, depose and state as follows:

(1) I am Project Manager, Licensing Services, General Electric Company ("GE") and have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding.

(2) The information sought to be withheld is contained in the GE specification document 25A5721. Hatch 2 Shroud Stablizers Stress Report, Revision 0, (GE Proprietary Information), dated June 6,1995. The proprietary information is delineated by bars masked in the margin adjacent to the specific material.

,(3) In making this application for withholding of proprietary information of which it is the owner, GE relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4), and the Trade Secrets Act,18 USC Sec.1905, and NRC regulations 10 CFR 9.17(a)(4), 2.790(a)(4), and 2.790(d)(1) for " trade secrets and commercial or financial information obtained from a person and privileged or confidential" (Exemption 4). The material for which exemption from disclosure is here sought is all " confidential commercial information", and some portions also qualify under the narrower definition of " trade secret", within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Reculatory Commission. 975F2d871 (DC Cir.1992), and Public Citizen Health Research Group

v. FDA,704F2d1280 (DC Cir.1983).

(4) Some examples of categories of information which fit into the definition of proprietary information are:

a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by General Electric's competitors without license from General Electric constitutes a competitive economic advantage over other companies;

b. Information which, if used 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 of a similar product; GilS-95-7-alliaDRFl. doc Affidavit Page i

F j + ..

c. Information which reveals cost or price information, production capacities, budget levels, or commercial strategies of General Electric, its customers, or its suppliers;

d. Information which reveals aspects of past, present, or future General Electric customer-funded development plans and programs, of potential commercial value to General Electric;

c. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

The information sought to be withheld is considered to be proprietary for the reasons set forth in both paragraphs (4)a. and (4)b., above.

(5) The information sought to be withheld is being submitted to NRC in confidence.

The information is of a sort customarily held in confidence by GE, and is in fact so held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by GE, no public disclosure has been made, and it is not available in public sources. All disclosures to third parties including any required transmittals to NRC, have been inade, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following.

(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge. Access to such documents within GE is limited on a "need to know" basis.

(7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing function (or his delegate), and by the Ixgal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GE are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements.

(8) The information identified in paragraph (2), above, is classified as proprietary because it contains detailed stress report results of analytical models, methods and processes, including computer codes, which are the bases for technical conclusions.

This stress report is part of the supporting information to evaluate a hardware design modification (stabilizer for the shroud horizontal welds) intended to be installed in a reactor to resolve the reactor pressure vessel core shroud weld cracking concern.

GilS-95 7-afHaDRFl. doc Affidavit Page 2

)

r j O ,e This information shows in specific detail the processes, codes and methods employed to perfonn the evaluations summarized in the above identified document.

The development and approval of this design modification utilized systems, components, and models and computer codes that were developed at a significant cost to GE, on the order of several hundred thousand dollars.

The development of the supporting processes was at a significant additional cost to GE, in excess of a million dollars, over and above the large cost of developing the underlying individual proprietary report infonnation.

(9) Public disclosure of the infonnation sought to be withheld is likely to cause substantial harm to GE's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of GE's comprehensive BWR safety and technology base, and its commercial value extends beyond the original development cost. The value of the technology base goes beyond the extensive physical database and analytical methodology and includes development of the expertise to determine and apply the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses done with NRC-approved methods.

The research, development, engineering, analytical and NRC review costs comprise a substantialinvestment of time and money by GE.

The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial.

GE's competitive advantage will be lost if its competitors are able to use the results of the GE experience to normalize or verify their own process or if they are able to

, claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

The value of this infonnation to GE would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive GE of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing these very valuable analytical tools.

GBS-95-7-afHaDRFl. doc Affidavit Page 3 J

R y

,g ,*, o * ,

e STATE OF CALIFORNIA )

) ss:

COUNTY OF SANTA CLARA )

George B. Stramback, being duly sworn, deposes and says:

That he has read the foregoing af6 davit and the matters stated therein are true and correct to the best of his knowledge,information, and belief.

Executed at San Jose, California, this b +k day of 'M 1995.

O b4 u dl. M Geofge B. stramback

~

General Electric Company 1

1 Subscribed and sworn before me this day of / fW 1995.

_ LC "

otary Public, State of Calg

.g .,

JUllE A. CURTS COMM. # 974657 fg

• • 3.' '

g , Notory Public - Cotifornio N u .. SANTA CLARA COUNTY

{

My Comm. Enones SEP 30.1996

=--, , , , , , , , ,

GIIS-95 7-afHaDRFl. doc Affidavit Page 4 1

Enclosure Edwin I. Hatch Nuclear Plant - Unit 2 Core Shroud Stabili:er Desien Submittal

1.0 BACKGROUND

The core shroud is a stainless-steel cylinder within the reactor pressure vessel (RPV) that provides lateral support to the fuel assemblies and provides a partition to separate the upward flow of coolant through the core from the downward recirculation flow. This partition separates the core region from the downcomer annulus, thus providing a floodable region following a postulated recirculation line break. The RPV, core shroud, and other RPV internals are designed to accomplish the following three basic safety functions:

1. Provide a refloodable coolant volume for the reactor core to assure adequate core cooling in the event of a nuclear process barrier breach.

2. Limit deflections and deformation ofinternal safety-related RPV components to assure that control rods and emergency core cooling systems (ECCSs) can perform their safety functions during anticipated operational transients and/or design basis accidents (DBAs).

3. Assure that the safety functions of the core internals are satisfied with respect to safe shutdown of the reactor and proper removal of decay heat.

On July 25,1994, the Nuclear Regulatory Commission issued Generic Letter 94-03 to address the potential for cracking in core shrouds and to request licensees '

to take certain actions. By letter dated August 24,1994, Georgia Power Company (GPC) responded to GL 94-03. GPC stated its support and participation in the BWR Vessel Internals Project and indicated plans to install a permanent pre-emptive repair on both Plant Hatch units. A permanent repair was subsequently installed on Unit I during the Fall 1994 refueling outage. The repair encompassed the entire set of circumferential welds in the core shroud and involved the installation of four tie rod assemblies in the annulus region around the core shroud.

GPC hereby submits the details of the planned repair to the Unit 2 core shroud.

2.0 DESCRIPTION

c The function of the Unit 2 core shroud repair is to structurally replace all horizontal girth (circumferential) welds from the H1 weld at the top of the core shroud to the H8 weld at the bottom of the core shroud. The Unit 2 core shroud contains a total of 9 horizontal girth welds. These welds are labeled H1 through H5, H6A, H6B, H7, and H8. The only significant cracking of BWR core shrouds  ;

has been associated with these welds.

HL-4877 E-1

l l

Enclosurs Core Shroud Stabilizer Desien Submittal l

The core shroud repair is designed to restrain the core shroud head, the top guide i support ring, the core, and the core support plate, and to limit upward displacement of the core shroud to acceptable levels during normal, upset and postulated accident conditions. The modification is an alternative to the <

requirements of the ASME Boiler and Pressure Vessel (B&PV) Code pursuant to 10 CFR 50.55a(a)(3Xi). The repair design provides structural integrity for, and )

takes the place of, all circumferential welds subject to cracking in the core shroud. i The repair is designed for the remaining life of the plant and any possible life extension beyond the current operating license. The repair is also designed to accommodate uprated power conditions corresponding to 105 percent rated power (2558 MWt).  ;

Additionally, the repair is consistent with and meets the criteria developed by the i Boiling Water Reactor Vessel Internals Project and contained in the document entitled "BWR Core Shroud Repair Design Criteria," dated August 18,1994, with one exception. The exception is related to the process of air cooling Type XM-19 1 material from the solution annealing temperature as opposed to water quenching. l Justifications for this exception are presented as part of this enclosure and in {

Attachment 1.

I The core shroud repair design consists of four tie rod stabilizer assemblies installed 90 apart in the core shroud / reactor vessel annulus. Each assembly consists of a tie  !

rod, an upper bracket, upper stabilizers, a lower spring, a middle support assembly, and a collet mount connected by a solid rod. The assemblies, which are designed and fabricated as safety-related components, are used to maintain the alignment of 1 the core shroud assuming all circumferential welds are cracked 360 throughwall.

The tie rods provide the vertical load carrying capability from the upper bracket to the shroud support collet attachment, supporting the upper stabilizers and lower radial restraint spring. The upper bracket and stabilizers provide an attachment feature to the top of the shroud. The upper spring provides lateral load carrying capability from the core shroud, at the top guide elevation, to the RPV. The collet at the bottom of the assembly provide a detachable connection from the lower radial restraint spring to the shroud support plate. The lower spring provides lateral load carrying capability from the core shroud at the core support plate elevation to the RPV, The middle support provides a limit stop for shroud sections between the intermediate welds and also increases the natural frequencies of the tie rod to reduce the potential for flow-induced vibration.

The vertical locations of the upper stabilizers and lower spring were chosen to provide the maximum horizontal support for the fuel assemblies. The upper springs are at the top guide elevation and the lower springs are at the core support l

lI HL-4877 E-2 l

1 Enclosure Core Shroud Stabilizer Desier Submittal plate elevation. All horizontal support for the fuel assemblies is provided by the top guide and the core support plate. 1 At the top of the shroud, each stabilizer assembly fits into a slot that is machined i partially into the top shroud flangejust below the shroud head. The stabilizer upper bracket is inserted into this slot and extends downward to below weld H3 providing support for the upper stabilizer. The tie rod passes through a hole in the upper bracket and is held against the upper bracket with a nut. The tie rod extends downward approximately 151 in. to the lower spring. At the middle of the tie rod, a support is installed between the tie rod and the RPV to: 1) minimize the potential for vibration, and 2) provide a limit to the potential motion of the shroud  !

between welds H4 and H5. The bottom of the tie rod threads into the lower spring .

which has a clevis at its bottom which is attached to a collet connector with a pin.

The collet connects to the shroud suppon through a hole that is machined in the shroud suppon. j Each cylindrical section of the shroud is prevented from unacceptable motion by i the stabilizers even ifit is assumed that its respective welds contain 360 l throughwall cracks. The motion of the sections above H1, between H1 and H2, and between H2 and H3, is restrained by the upper bracket which contacts the  :

shroud and is radially restrained by the upper stabilizers that contact the RPV. A feature on the upper bracket prevents unacceptable motion of the section between  !

H3 and H4. The section between H4 and H5 is prevented from unacceptable motion by a limit stop which is part of the upper tie rod middle support. The  !

lower spring contacts the shroud such that it prevents unacceptable motion of the section between H5 and H6A. An extension on the lower spring prevents unacceptable motion of the section between H6A and H6B, as well as the section .

between H6B and H7. The section between H7 and H8 is prevented from [

unacceptable motion by a limit stop which is part of the collet assembly. Together, the stabilizer assemblies and the lateral restraints resist both vertical and lateral loads resulting from normal operation and design accident loads, including seismic ,

loads and postulated pipe ruptures.

The stabilizer assemblies are installed with a small vertical preload, which assures that all components are tight after installation and during cold shutdown, and  !

provides approximately 3500 lbs, of axialload on the 3.75 in. diameter tie rods.

The upper bracket, upper stabilizers, lower spring, and collet are fabricated from alloy X-750. The tie rod is fabricated from Type XM-19 stainless steel. The spring, bracket, and tie rod materials have a smaller coefficient of thermal expansion than does the 304L shroud. Thus, the stabilizer assemblies are thermally preloaded when the reactor is at operating conditions. The spring constant of the stabilizers in the vertical direction is designed to provide a total vertical preload at operating conditions which is greater than the net upward applied loads on the l

l l

l HL-4877 E-3 l

l

)

Enclosure  ;

Core ShroudStabilizer Desien Submittal ,

l i

shroud. Thus, if a combination, or all, of welds H1 through H8 were completely -

cracked, the stabilizers would vertically restrain the shroud such that no displacement would occur during normal operation, which minimizes potential leakage through the cracks. A detailed evaluation supporting this conclusion is provided in Attachment 5.

Vertical separation for any of the welds is limited to allowable values. Sections of ,

a cracked shroud may separate under main steam line break (MSLB) loss of coolant accident (LOCA) pressures exceeding the mechanical and thermal preload l in the rods. The shroud sections may also separate under seismic loads with, or l without, simultancous LOCA loads (i.e., tipping). The crack opening due to -

postulated seismic events include rotation at the crack with variation in the amount of opening across the diameter. The maximum crack opening for the design basis earthquake (DBE) combined with a MSLB LOCA is 0.588 in.. The maximum calculated gap about one edge for a DBE with normal operating loads is 0.244 in..

Excessive preload in the rods would be required to prevent any separation for these events. This excessive preload could exacerbate the cracking mechanism.

3.0 STRUCTURAL EVALUATION Core Shroud and Tie Rod Stabilizer Assemblies The stabilizer assemblies are designed to the structural criteria specified in the updated Edwin I. Hatch Nuclear Plant Final Safety Analysis Report (UFSAR).

The seismic analysis model consists of the reactor building, RPV and internals, postulated crack (s) of circumferential weld (s) on the shroud, and the repair hardware. The seismic analyses were performed in accordance with the methods 1 described in the UFSAR. The seismic input used the time history of the motion at the base of the reactor building, with the seismic response computed using the modal superposition time-history method of analysis. The loads and load combinations relevant to the core shroud were included in the design.

i The tie rod stabilizer assemblies were designed using the ASME Code Section III, i 1989 Edition, subsections NB and NG as a guide. The original ASME Code Section III (1968 Edition and addenda through Summer 1970) for the design and construction of the RPV did not contain design requirements for core support structures. The additionalloads placed on the RPV by the stabilizer assemblies were evaluated to the original design code.

The load combinations required by the UFSAR for normal, upset, emergency, and faulted, in addition to consideration of one-half the seismic margins earthquake (1/2 SME), operational condition were evaluated. Also, the 1/2 SME was ,

evaluated using applicable design basis criteria.

HL-4877 E-4

Enclosure Core Shroud Stabilizer Desiat Submittal Ooerational Condition Load Combination j

~

Normal Normal operation pressures + deadweight l

Upset 1 OBE + normal operation pressures + deadweight Upset 2 Thermal transient + normal pressure difference + deadweight Emergency 1 DBE + normal operation pressures + deadweight Emergency 2 MSLB LOCA pressures + deadweight Emergency 3 1/2 SME + normal pressure difference + deadweight l

4 Faulted 1 DBE + MSLB LOCA pressures + deadweight Faulted 2 DBE + RLB LOCA pressures + deadweight l

Faulted 3 1/2 SME + MSLB LOCA pressures + deadweight Faulted 4 1/2 SME + RLB LOCA pressures + deadweight OBE: Operating Basis Earthquake RLB LOCA: Recirculation Line Break

)

A three-dimensional finite element analysis using the ANSYS code was used to calculate the shroud structural response. Design loads were obtained by appropriate combination of the mechanical preload; loads from differential thermal expansions of the shroud and repair hardware; gravity, including buoyancy effect; pressure differences; and seismic loads. For each of the seismic cases, the vertical and horizontal seismic loads, were combined by absolute summation. For each of I

the faulted cases, the pressure loads were also combined by absolute summation.

The MSLB LOCA loads were applied as uniform static upward pressures on the core support plate and the shroud head. The LOCA pressures, together with the gravity loads, were used in the analyses for a MSLB LOCA event. Pressures and gravity loads were combined with the peak DBE and 1/2 SME seismic loads in the analyses for the simultaneous MSLB LOCA and DBE event.

The RLB LOCA produces a spatial and time varying lateral pressure in the I shroud / reactor vessel annulus. The initial acoustic phase of this transient is very  !

abrupt relative to the shroud inertia and frequencies, and does not have a j significant effect on the shroud or stabilizer performance. The remainder of the I transient extends over a relatively long period of time and as such, is considered a l static pressure. This static pressure produces a 25000 lb. lateral load in the shroud  ;

section between welds H4 and H5. This load, combined with gravity and normal operating pressure differences, was used in the analyses for an RLB LOCA event.

The LOCA pressures, operating pressure, and gravity loads were combined with HL-4877 E-5 l

6

, Enclosure  ;

l Core Shroud Stabilizer Desien Submittal r

i the peak DBE seismic loads in the analyses for the simultaneous RLB LOCA and DBE. The evaluations show that the RLB LOCA loads were bounded by the MSLB LOCA loads for the design of the stabilizer assembly. Additional details relative to the RLB LOCA are provided in Attachment 1.

Loads during normal operation are a combination of the mechanical preload, differential expansions between the shroud and repair hardware, and the gravity and pressure loads. With the exception of a few crack combinations, the largest operating loads are obtained if the shroud is uncracked, when applying the mechanical preload, and remains uncracked during operation. Shroud loads and generally, the tie rod loads are smaller if the shroud cracks after the repair hardware is installed. While all crack configurations were investigated to identify the enveloping loads to evaluate structural integrity of the system, configurations producing minimum shroud compression were also analyzed to assure that shroud sections would not separate across cracks and produce leakage flow during normal operation. The separation analyses and structural loads are described in detail in Attachment 5.

The limiting upset loading condition event is the cold feedwater transient which corresponds to a reactor scram with a loss of feedwater pumps. Other upset conditions, such as the loss of feedwater heaters, were considered. However, these conditions were bounded. During the cold feedwater transient, a maximum temperature difference of 133 F between the core shroud and the cooler stabilizer components could exist due to the injection of relatively cold feedwater into the '

core shroud annulus. This event results in an increase in the tensile load on the stabilizer assembly and an increase in the compressive load in the core shroud.

However, the evaluation shows that the stresses in the stabilizer and in the core shroud are less than the ASME Code upset allowable stresses and less than the material yield stress. The required preload is maintained for all normal and upset conditions thus ensuring that no crack separation occurs during normal operation.

An evaluation of the potential effects of radiation on the stabilizer assemblies showed that the flux levels at the stabilizers are low compared to the values which could degrade material properties. Even after 20 years of operation, the maximum fast fluence at the stabilizers will be approximately 2E19, which is well below the value which would cause damage to stainless steel. l 1

'l Seismic Analysis '

To support the shroud repair project, two-dimensional seismic analysis of a coupled mathematical model of the reactor building and RPV was performed to obtain seismic loads in the nuclear steam supply system (NSSS) components. The coupled system model was essentially the same as the model used for the original liL-4877 E-6

Enclosure Core Shroud Stabilizer Desien Submittal RPV, The analysis included the determination of the natural frequencies and mode shapes of the RPV building system, and the dynamic responses to the Unit 2 OBE, DBE,1/2 SME ground motions. Analyses were performed for the complete range of postulated shroud weld joint cracks, as well as for a fully uncracked configuration with shroud repair hardware installed. The seismic analyses were performed using the GE SAP 4G07 computer code which was accepted by the NRC staff for this application. The resulting seismic forces, moments, displacements and accelerations were used as input to the design of the shroud repair hardware and to validate the continued structural integrity of the core support structure and RPV internals. The details of the seismic analysis are presented in Attachment 4.

For each of the three seismic conditions, separate analyses of the shroud repair model were performed for the N-S, E-W, and vertical directions. The horizontal and vertical seismic model essentially corresponded to the original seismic models, I

with the exception of the nuclear core, core shroud, repair hardware, and minor updates. The nuclear core was updated to the projected Cycle 13 configuration.

The seismic models incorporated the stabilizer assembles and the core shroild with postulated 360' throughwall cracks. The stabilizer assemblies were modeled as one equivalent lateral spring at the top guide and at the core plate level and a rotational spring connecting the shroud top flange with the shroud support plate.

The lateral spring accounts for the combined stiffness of the core shroud, stabilizer assembly, and RPV, The rotational spring represents the tie rod assembly. The core shroud is anticipated to respond nonlinearly to horizontal seismic excitations, because the characteristics of the weld crack interface depend on the magnitudes l of the different loads acting on the interface and continually vary between hinge I and roller conditions during the seismic event. Performing linear dynamic analyses l with the weld cracks modeled both as hinges and as rollers, and bounding peak

! seismic loads and displacements will bound the seismic demands for both the repair ,

l hardware and the evaluation of the core shroud.

Each of the horizontal weld cracks is modeled as either hinge or roller. The seismic results for the roller cases are used for the stress analysis only when the 1 non-seismic loads allow the crack to fully separate. This condition only occurs for ,

I load combinations which include the MSLB LOCA. For load combinations, other l than LOCA, local areas of the shroud may temporarily separate under seismic j loading. However, the majority of the shroud remains in contact, allowing the

! jagged intergranular stress corrosion cracking (IGSCC) to transmit shear. This condition is modeled as a hinge in combination with horizontal seismic loading.

During normal operation, the crack surfaces are held fully in contact by the tie l

rods.

i l

l lIL-4877 E-7 l

l

Enclosure Core ShromiStabilizer Desitm Submittal The design spectra for the OBE and DBE are the modified Newmark spectra. The peak acceleration of the free field ground motion is 0.08g and 0.15g for the horizontal components of the OBE and DBE, respectively. The vertical ground accelerations are two-thirds of the corresponding horizontal values. The free field ground motion time history for the OBE and DBE is a synthetic time history having response spectra closely enveloping the design spectra. Due to the different frequency range of response of the reactor building determined from the Plant Hatch seismic margin assessment (SMA), a 1/2 SME case was considered. It was treated as an additional DBE load case. The 1/2 SME has the same peak ground acceleration as the DBE; i.e.,0.15g for the horizontal component and 0.10g for the vertical component. For each of the three seismic conditions, separate analyses of the shroud repair model were performed for seismic motion in the N-S, E-W, and vertical directions.

For each seismic analysis case, example in-stnicture response spectra (IRS) were calculated and compared to the appropriate IRS of record for OBE, DBE, and 1/2 SME to verify that the input time histories and the seismic models were correctly used and that the cracked shroud and/or shroud repair did not have any significant effect on the reactor building seismic response.

In the time history analysis, the effect of uncedainties in material properties and modeling technique for the shroud repair model was accounted for by shining the peak frequency of the input motion. The frequency shin was accomplished by varying the digitization time step for the input motion to maximize the seismic response ofinterest. The range of such peak frequency shiR for the OBE and DDE input motions was -10% to +10%. This is consistent with the UFSAR criterion of a i10% spectrum peak broadening. For the 1/2 SME, the peak frequency shin was performed to meet the intent of the Plant Hatch SMA procedure for broadening and enveloping the IRS for the SME. The procedure for the reactor building SME IRS was to envelope the three spectra; i.e., the lower bound, intermediate, and upper bound soil cases, with the intermediate case SME IRS broadened by 15% in the frequency coordinate, and for the horizontal direction, an additional +10% broadening of the three spectra in the 2 to 3 Hz frequency range.

An enveloping combination of cracked /uncracked welds was analyzed to define the worst-case combination for the core plate and top guide displacements. The stabilizer design is based on the worst-case scenario to ensure control rod insenion and safe shutdown, should this postulated scenario occur. Each cracked weld was postulated to have a 360' throughwall crack. It was concluded that five cracked cases bound the numerous possible combinations of assumed cracked welds, while considering the various plant operating conditions and providing the maximum spring loads for the shroud repair hardware. The five bounding cases are:

HL-4877 E-8

Enclosure Core Shroud Stabilizer Desint Submittal All welds cracked - Weld H1 modeled as a roller, H2 through H8 modeled as hinges.

Weld H4 cracked - H4 modeled as a hinge.

Weld H4 cracked - H4 modeled as a roller.

Weld H8 cracked -

H8 modeled as a hinge.

l Weld H8 cracked -

H8 modeled as a roller.

The limiting loads in the tie rods, upper stabilizers, and lower springs occur with the different assumed shroud cracks. The limiting loads in the tie rods occur when it is assumed that there is a 360 throughwall crack in weld H4. The limiting loads in the radial direction on the upper stabilizers occur when it is assumed there is a 360 throughwall crack in weld H8. The limiting loads in the radial direction on the lower springs occur for the all-welds-cracked case.

In order to ensure that the installation of the stabilizer assembly design does not adversely affect the existing dynamic qualification of the RPV and internals, analyses for the uncracked case were performed with and without the shroud repair in place. It was concluded that seismic loads in the RPV and internal structures are decreased, or at least not significantly increased, by the shroud stabilizer installation. It was also shown that loads in the RPV and intemals are further reduced by the inclusion of the most limiting combination of assumed cracks. This is due to the fact that as shroud rigidity is decreased, the fuel is isolated, and the seismic load is mainly carried by the stabilizer springs and the tie rods.

The maximum deflection of any part of the shroud that is not directly supported by either the upper or lower radial springs is limited to approximately 0.75 in. by mechanical limit stops. These stops do not perform any function unless a section of the shroud, e.g., between H4 and HS, becomes loose and a combined LOCA plus seismic event occurs. If this unlikely scenario occurs, the stops limit the horizontal displacement to approximately 0.75 in., which is equal to one half of the shroud wall thickness. The limit stops do not invalidate the linear seismic analysis, because very little mass is associated with any potential loose and unsupported section of the shroud. A displacement equal to one half of the shroud wall thickness does not result in post event leakage that prevents core cooling, because the shroud sections still overlap each other by one half (0.75 in.) of the shroud wall thickness.

HL-4877 E-9

l Enclosure Core ShrousiStabilizer Desipt Submittal l

l The results of the seismic and stress analyses show that all structural limits are 1 satisfied. Additionally, the predicted worst-case (smallest margin) transient  !

deflections of the core plate is 0.24 in. for a load combination of a DBE, assuming i all welds are cracked. The allowable transient displacement for this emergency event is 1.12 in.. All predicted stress intensities in the lower radial spring meet UFSAR allowables. The predicted worst-case transient deflection of the top guide is 0.23 in. for a DBE, plus MSLB LOCA, assuming weld H8 is cracked and acting as a roller. The allowable transient deflection of the top guide for this faulted  ;

event is 4.0 in.. The stresses in the upper radial stabilizers meet UFSAR j allowables. Neither the upper nor the lower springs will have a permanent l deformation following a DBE emergency event. Further, the maximum transient l deflections are far less than the allowable permanent deformations for emergency l' combination which are 1.4 in. in the upper spring and 0.5 in. in the lower spring for the emergency combination (or 1.87 in and 0.67 in., respectively, for the faulted combination). The predicted deflections of both the top guide and the core plate, for all load combinations, are well within allowables.

Evaluation of Shroud Shell and Shroud Support Plate Evaluations of the core shroud shell and core shroud support plate stresses were performed using finite element models and the ANSYS computer code. The effect of the additional loads from the core shroud repair were evaluated for the combined loadings resulting from the specified normal operating, upset, emergency, and faulted conditions. The shroud seismic loads were obtained in terms of spring loads, tie rod moments, and shroud shear / moment diagrams. The maximum values of these parameters required for repair hardware analyses were obtained by scanning the system seismic time-histoiy analysis results. These maximum values occurred at different times and thus did not form a consistent set of parameters required for shroud cylinder analyses. Therefore, the effects of inertial moments combined with the pressure, gravity, and tie rod loads were analyzed separately from the effects of the maximum stabilizer spring loads. As a result of these evaluations, the stresses were shown to be within ASME Code allowable stresses. A detailed discussion is provided in Section 7.0 of Attachment 5.

Flow-Induced Vibration l

The potential for flow-induced vibration was evaluated by calculating the lowest natural frequency of the tie rods and the highest vortex shedding frequency due to the water in the downcomer. The lowest natural frequency is 69 Hz and the maximum vonex shedding frequency is 4.4 Hz. The results show ample margin between excitation frequency and lowest natural frequency compared to the HL-4877 E-10

l Enclosure Core Shroud Stabilizer Desien Submittal standard design goal of a factor of 3. As a result, essentially no flow-induced vibration fatigue of the tie rod stabilizer assembly components will occur.

Loose-Parts Considerations All components of the stabilizer assemblies are locked in place with mechanical devices. Loose pieces cannot occur without the failure of a locking device. The stresses in the stabilizer components during normal plant operation are less than one-third of the normal event allowable stresses.

The stabilizer assemblies are fabricated from stress conosion-resistant material.

Therefore, it is unlikely that a stabilizer component will fail. However, if one stabilizer assembly is postulated to fail during normal plant operation, there would be no consequence to the shroud (even ifit is cracked) or to the other three stabilizer assemblies. The leakage through a cracked shroud may increase slightly, but it would not be detectable. The unit would continue to operate until the next refueling outage, when the broken stabilizer assembly would most likely be detected and repaired. The postulated broken component may fall to the shroud support plate or may be sucked into the recirculation pump. The consequences of the postulated loose stabilizer assembly piece would be consistent with the consequence of other postulated loose pieces.

Installation of the stabilizer assemblies requires the machining of eight slots in the core shroud head flange just below the shroud head, and drilling / machining four holes in the shroud support plate. The flange slots are approximately 2.7 in. x 2.7 in. x 3.5 in. in size, and the support plate holes are a nominal 4.25 in.

in diameter. The slots are cut using the Electric Discharge Machining (EDM) process. Using a Trepan style drill specifically developed and tested for this i

application, the holes in the shroud support plate are drilled to within 0.75 in. of l the underside of the shroud support. The holes will be completed by EDM process which is monitored by a remote camera.

l The above machining activities could generate small objects or debris that may remain in the reactor after the repair is installed. EDM generates swarf, which is very fine particles comprised of carbon, nickel, iron, chromium, etc. (the elements l

contained in the EDM electrode, and the shroud and shroud support material).

These particles are very small(approximately 1-50 microns). Greater than .

95 percent of the swarf generated is collected by the EDM electrode flushing j system. However, when the EDM electrode breaks through the shroud support, 1 the flushing system cannot collect the swarf. This swarf remains in the reactor.

The amount of swarfis very small, representing less than a tenth of one percent of the total generated. Consequently, it is considered insignificant.

HL-4877 E-11

Enclosure Core Shroud Stabili:er Desien Submittal The minute sand-like particles resulting from the EDM process are too fine and small to be caught at one of the fuel spacers. Most likely, these particles will be carried by the cooling flow up through the length of fuel bundles and then discharged from the reactor core through the top of the upper tie plate. They will  ;

eventually be removed from the reactor coolant by the reactor water cleanup (RWCU) system. Therefore, there is no potential for fuel fretting due to the EDM process.

The potential for the particles generated by the repair processes could cause  :

control rod drive (CRD) seal wear and was therefore also evaluated. Because the particles generated are so small, they w:ll most likely be carried by the cooling flow up through the length of fuel bundles and then discharged from the reactor core through the top of the upper tie plate, or by the core bypass flow through the core region and then discharged through the top guide. The particles will eventually be removed from the reactor coolant by the RWCU system. The upward flow direction makes it highly unlikely these particles will be deposited on the top of the core plate so that they can migrate to the bottom of the control rod guide tubes where they could be sucked into the CRD. Therefore, it is very unlikely these particles will have any significant effect on CRD seal wear or adverse effects on CRD operation.

In addition to the CRD seals, the potential for the particles generated by the repair processes adversely affecting the reactor recirculation pump seal performance or ,

life was evaluated. Ideally, the reactor recirculation pump seats should be operated in a clean, air-free environment. This objective is achieved by venting the seals after maintenance and purging the seals during operation. Seal spurge injects 2 gal per minute of clean water into the seal cartridge to keep solids, such as reactor corrosion products, or in this case the EDM byproducts, from reaching the critical components of the seal.

The debris generated by the EDM process during the shroud modification will not increase short- or long-term degradation of the CRD or recirculation pump wear.

1. The amount of material released by the EDM process is small when compared to the corrosion byproducts that are routinely released by the reactor and the ,

carbon steel piping. The shroud stabilizer installation will only slightly l aggravate the normal cleanliness of the reactor water. l

)

2. Solids generated by the EDM process are not as abrasive as the normal reactor corrosion products (iron oxide).

3. The sites of the EDM process are relatively far from the pump seals. The

]

debris will primarily be picked up by the RWCU system. Some debris may llL-4877 E-12

! l

_ . - , . ..,..-- .- _ y , --__- _ . , -r e

i Enclosure Core Shroud Stabilizer Desien Submittal

, eventually reach the recirculation pump casing; however, the seal purge will  :

prevent entry into the seal cavity.

4. The site venting and purging procedures along with the recommendations stated below, will appropriately protect the seals against debris generated by the shroud stabilizer installation. To allow residual trapped air to be L

discharged, all venting should be performed in stages: 1) prior to pump startup, 2) after 3-10 minutes of pump operation. i Based on the above factors, it is believed that the shroud stabilizer installation will not adversely affect the reactor recirculation pump seal performance or life.

l The potential for the particles generated by the installation processes having j adverse effects on instrumentation was also reviewed. Because the remaining  !

particles are expected to be dispersed by the flow throughout the reactor and no l

flow through the instrumentation would tend to draw in these particles, it is not expected that these panicles would be able to migrate into the instrumentation lines in sufficient quantities to cause plugging or other adverse effects. Therefore, it is very unlikely that these particles will have any significant effect on instrumentation. j In summary, the EDM particles and the metal particles generated by the installation of the shroud stabilizers do not represent a concern for fuel fretting and subsequent fuel damage nor do they represent a concern for CRD seal wear, )

reactor recirculation pump seallife, or adverse effects on instrumentation. Field experience from previous repairs has not identified any operational problems due to the particles generated by the installation processes. In the unlikely event that any abnormal results occur from an EDM process, they will be addressed by a separate evaluation at the time of occurrence.

The drilling process for the four holes in the shroud support plate produces marble sized nodules of shroud support material which are captured by a suction system integrated as part of the drilling tool. In the unlikely event that a nodule was not captured, the nodule would come to rest on the shroud support. This area will be 1 inspected with the remote camera and vacuumed with suction equ.~pment to assure complete capture of drilling particles. Complete capture of drilling particles is thus 4 assured. )

Each of the four holes is completed by EDM. Capture of the hole core is assured by a hydraulic collet style clamp on the inside of the EDM actuator. EDM core capture by this method has been proven reliable in past applications. In the l

unlikely event that the approximately 25 lb. hole core were to detach before it was HL-4877 E-13

l l

Enclosurs  !

Core Shroud Stabili:er Desien Submittal l extracted from the hole, it would simply drop to the bottom of the RPV, where it l would then be retrieved. )

4.0 SYSTEM EVALUATIONS Evaluations were performed relative to plant operation with postulated 360 throughwall cracks in the core shroud weld (s) and with the stabilizer assemblies installed. The evaluations demonstrate that required emergency core cooling system (ECCS) performance and control rod insertion would be maintained given the postulated cracks for normal operation and design basis events with the stabilizer assemblies installed. Conditions analyzed include the stabilizer assembly induced leakage, core shroud weld crack leakage, downcomer flow characteristics, lateral displacement of the core shroud, and vertical separation of the core shroud.

The hardware design to repair the shroud for Unit 2 requires machining of four holes through the shroud support plate and eight slots in the shroud head flange. Each of the holes in the shroud support plate will have some clearance, which will allow leakage flow to bypass the core and the steam separation system.

The slots in the shroud head flange will be sufficiently shallow to prevent any leakage from the upper core plenum to the vessel downcomer region. Additional leakage may occur through the weld cracks (H1 through II8) and the existing l

replacement access hole cover. The leakage flows and their performance impact were evaluated for 100% uprated power (corresponding to 105% rated power, 2558 MWt) and 87 to 105% rated core flow.

The impacts of the leakage flows through the shroud repair holes and the potential weld cracks in the shroud have been evaluated. The results show that at uprated ,

power and 87 to 105% rated core flow, the leakage flow from the shroud repair holes, weld cracks, and replacement access hole covers is predicted equal to about 0.22% of core flow. This leakage flow is sufficiently small that the steam separation system performance, jet pump performance, core monitoring, fuel thermal margin, and fuel cycle length remain adequate. Also, the impact on ECCS performance is sufficiently small to be judged insignificant, and, hence, the licensing basis peak cladding temperature (PCT) for the normal condition with no shroud leakage is applicable. A discussion of these evaluations is provided below.

l As stated previously, installation of the shroud repair mechanism requires 1 machining four holes through the shroud support plate. A collet assembly is installed through each hole to anchor tie rods installed as part of the repair. The leakage flow areas are based on the clearances of the collet assembly parts passing through the shroud support plate. In addition, a total of nine circumferential shroud welds (111 - H8) are considered as potential leakage paths--two above the j top guide support ring, four on the upper shroud between the core support ring IIL-4877 E-14 l

l i

l 1

i Enclosure Core ShroudStabill:er Design Submittal and the top guide support ring, and three between the core support ring and the shroud support plate. It is conservatively assumed that each of these welds develops a complete circumferential crack that opens to 0.001 in..

l The leaka8e flows for 100% uprated power and 105% rated core flow [ increased  !

core flow (ICF)] are summarized in Table 1 below. These leakage flows are based on applicable loss coefficients and reactor internal pressure differences (RIPDs) I across the applicable shroud components. Leakage from the weld cracks above l the top guide support ring is assumed to be two-phase fluid at the core exit quality, i Leakage from the remaining paths below the top guide support ring is considered to be single-phase liquid. All leakage flows bypass the steam separators and .

dryers. The leakage flows below the core support ring also bypass the core. The )

results show that the leakage flows from the repair holes, weld cracks, and the access hole cover result in a combined leakage of about 0.22 percent of core flow i

at 100% uprated power and 87 to 105% rated core flow.

Table 1. Summary of Leakage Flows at Uprated Power and ICF Leakage flow (gpm)

Weld cracks 130 Repair holes in support plate 185 Access hole covers 185 i Leakage-to-core mass flow (%)

Weld cracks 0.045 Repair holes in support plate 0.085 Access hole covers 0.085 The steam portion of the leakage flow will combine with the carryunder from the steam separators, resulting in increased total carryunder in the vessel downcomer region. The impacts of the total leakage on the steam separation system performance, jet pump performance, core monitoring, fuel thermal margin, ECCS l performance and fuel cycle length were evaluated and are summarized in the following subsections. .

Steam Separation System The leakage flow through weld cracks H1 and H2 occurs above the top guide support ring and includes steam flow, which slightly increases the total carryunder in the downcomer by about 0.001% at rated power. The total leakage flow also has the effect of slightly decreasing the flow per separator and slightly increasing the separator inlet quality. Separator performance is based on the applicable separator test data over the operating water level range. The combined effective l HL-4877 E-15

Enclosura Core Shroud Stabilizer Design Submittal carryunder from the separators and the shroud leakage is at uprated power and 87 to 105% rated core flow and is bounded by the design value (0.25 wt.%). The carryover from the separators remains within the design limits so that moisture from the dryer meets the plant performance requirements ofless than 0.1%.

Jet Pumps The increased total carryunder will decrease the subcooling of the flow in the downcomer. This in turn reduces the margin tojet pump cavitation. However, because the total carryunder meets the design condition carryunder value, there is no impact onjet pump performance <:ompared with the design condition.

Core Monitoring The impact of the leakage results in an over prediction of core flow by about 0.19% of core flow. This over prediction is small compared with the core flow measurement uncertainty of 2.5% forjet pump plants used in the Minimum Critical Power Ratio (MCPR) Safety Limit evaluations. Additionally, the decrease in core flow resulting from over prediction results in only about 0.1% decrease in the calculated MCPR. Therefore, it is concluded that the impact is not significant.

Anticipated Abnormal Transients The code used to evaluate performance under anticipated abnormal transients and determine fuel thermal margin includes canyunder as one of the inputs. The effect of the increased carryunder due to leakage results in greater compressibility of the downcomer region and, hence, a reduced maximum vessel pressure. Since this is a favorable effect, the thermal limits are not impacted.

Emergency Core Cooling System The ECCS consists of the high pressure coolant injection (HPCI) system, the automatic depressurization system (ADS), the low pressure coolant injection (LPCI) system, and the core spray (CS) system Leakage through weld cracks H1 and H2 results in slightly increased carryunder that causes the initial core inlet enthalpy to increase slightly, with a corresponding

decrease in the core inlet subcooling. Leakage flows from the repair holes and the weld cracks decreases the time for coolant inventory loss and also increases the time for coolant inventory recovery. The combined effect was conservatively assessed to increase the peak cladding temperature (PCT) for the limiting LOCA by less than 10 F. The current analysis yields a PCT of 1526*F for the RLB LOCA with battery failure (limiting LOCA event). The 10 CFR 50.46 regulatory HL-4877 E-16 i .

l Enclosure Core Shroud Stabilizer Design Submittal .

T limit PCT is 2200*F. Because the maximum potential effect on the' design basis LOCA PCT is very small, the margin of safety is not adversely affected. This impact is sufficiently small to be judged insignificant, and, hence,'the licensing basis PCT for the normal condition with no shroud leakage is applicable. The sequence of events remains essentially unchanged for LOCA events with the shroud repair -

leakage.

Fuel Cycle Length .

The increased carryunder due to leakage flow above the top guide support ring results in a slight increase in the core inlet enthalpy, compared with the no-leakage condition. The combined impact of the reduced core inlet subcooling and the reduced core flow due to the leakage results in a minor effect (~0.5 days) on fuel cycle length and is considered negligible.

Downcomer Flow Characteristics An analysis of the effect of the stabilizer assemblies on the downcomer flow  !

characteristics shows that the small decrease in flow area and small increase in pressure drop will not result in a significant impact.

l The largest additional flow blockage area introduced by inclusion of the shroud hardware is 259 in.2 from the upper stabilizer at the top of the shroud. The original ECCS evaluation is based upon a nominal RPV inside diameter of 218 in.. ,

The as-built dimensions indicate an RPV inside diameter of 220 in.. This as-built 3 diameter provides an additional flow area of 688 in.2 The additional flow blockage area of 259 in.' is much less than the margin in the ECCS evaluation due to the as-built diameter. Therefore, the installation of the stabilizer assemblies should not affect the recirculation flow of the reactor. Additional details are provided in i Attachment 1.

Eptential Lateral Disolacement of the Core Shroud r The results of the seismic and stress analyses have shown that the maximum lateral displacement of the core shroud at the top guide and core plate locations remain below acceptable values under normal operations and design load combinations, 3 such as DBE, MSLB LOCA, and RLB LOCA, assuming 360 throughwall cracks at any circumferential weld location. The lateral displacements of the shroud are  :

limited by the radial restraint springs provided on the stabilizer assemblies. The I motion of the sections above H1, between H1 and H2, and between H2 and H3 are restrained by the upper bracket. The upper bracket contacts the shroud and is radially restrained by the upper stabilizer which contacts the RPV. A feature on the upper bracket prevents unacceptable motion of the section between H3 and HL-4877 E-17 l

l

. . _ . - . _ _ _ . . _ . . . ~ _ . . _ _

i Enclosure Core Shroud Stabill:er Desien Submittal H4. The section between H4 and H5 is prevented from unacceptable motion by a limit stop which is part of the tie rod mid support. The lower spring contacts the shroud such that it prevents unacceptable motion of the sections between H5 and H6A. There is an extension on the lower spring which prevents unacceptable motion of the section between H6A and H6B, as well as the section between H6B and H7. The section between H7 and H8 is prevented from unacceptable motion by a limit stop which is pan of the collect assembly. .

As a result, the maximum displacement of any part of the shroud that is not directly supported by a radial restraint is limited to approximately 0.75 in. by i mechanical limit stops. This value is equal to one-half of the shroud wall thickness. Accordingly, the separated portions of the core shroud will remain overlapped during the worst-case condition of a DBE combined with LOCA conditions.

The predicted worst-case (smallest margin) transient deflections of the core plate  ;

are 0.24 in. for a load combination of a DBE, assuming all welds cracked. The allowable transient displacement for this emergency event is 1.12 in.. - All predicted i stress intensities in the lower radial spring meet UFSAR allowables. The predicted '

worst-case transient displacement of the top guide is 0.23 in. for a DBE, plus MSLB LOCA, assuming weld H8 is cracked and acting as a roller. The allowable i transients deflection of the top guide for this faulted event is 4.0 in.. The stresses in the upper radial stabilizers meet UFSAR allowables. Neither the upper nor the ,

lower springs will have a permanent deformation after a DBE emergency event.

Further, the maximum transient deflections are far less then even the allowable permanent deformation which are 1.4 in. in the upper spring and 0.5 in. in the lower spring for emergency, or 1.87 in. and 0.67 in. respectively for the faulted combination. The predicted deflections of both the top guide and the core plate, for all load combinations, are well within the allowable values.

As a result, the maximum lateral displacement of the core shroud will not result in significant leakage from the core to the downcomer region following a DBA scenario, and the ability to reflood the core will be assured. Additionally, the predicted permanent lateral displacement of the top guide or core plate will not significantly increase control rod insertion times.

Potential Vertical Seoaration of the Core Shroud The results of the seismic and stress analyses show that the maximum vertical displacements of the repaired core shroud remain well within acceptable values assuming 360* throughwall cracks at any circumferential weld during a DBE combined with LOCA conditions. As previously described, there is no vertical separation during normal events. With the stabilizer assemblies installed, the HL-4877 E-18

q i

Enclosure Core Shroud Stabilizer Desien Submittal maximum vertical separation during the scenario of a MSLB LOCA is limited to 0.214 in. at the H6B weld location. The vertical separation occurs as the main steam LOCA pressure differentials across the shroud head result in the tie rod mechanical and thermal preloads being momentarily exceeded. The separation is ultimately limited by the tie rod stabilizer assemblies.

Vertical separations for cases, including the DBE, and the DBE combined with an MSLB LOCA, were also analyzed. For these cases, the shroud cracks were modeled as hinges in the seismic analyses with the shroud sections separated by a crack assumed to rotate about the shroud center diameter when the tie rod moment was not sufficient to overcome the preload across the cracks, and about an edge of the shroud when the calculated moment would overcome the preload on one side of the shroud. The moment was translated into individual tie rod loads using the rotational stiffness of the tie rod system corresponding to the assumed rotation configuration. The loads from shroud vertical acceleration were added to these loads absolutely to obtain the total seismic loads which were then combined absolutely with the nonseismic loads to obtain the total loads. The results of these analyses show that a maximum momentary separation of 0.244 in, would occur ,

during a DBE from tipping of the core shroud. The largest vertical edge separation of 0.588 in. was calculated for the case of MSLB LOCA combined with a DBE. The vertical displacements are momentary and the shroud will return to rest on the lower portion. After the momentary lift, no significant amount of shroud flow bypass will occur. Shroud bypass leakage is not considered safety. [

significant during the momentary lift during a MSLB LOCA, because there is no loss of coolant from the lower vessel area and the small vertical lift of 0.588 in.

does not adversely affect the safety function or performance of the ECCS for this event. Therefore, the predicted separations would not preclude any of the systems from performing their safety functions.

4.0 MATERIALS AND FABRICATION The stabilizer assemblies are fabricated entirely from Type 316 or 316L stainless steel, Type XM-19 stainless steel and alloy X-750. The upper stabilizers, lower ,

spring, upper nut, upper bracket, and collet are fabricated from alloy X-750 (ni-cr-fe) material. The tie rods are fabricated from XM-19 material. The other components are fabricated from Type 316 or 316L stainless steel. These materials have been used for a number of other components in the BWR environment and have demonstrated good resistance to stress corrosion cracking by laboratory  ;

testing and long-term service experience. Welding is not included in the  ;

fabrication and installation of the tie rod stabilizer assemblies. l The alloy X-750 material was selected for the upper stabilizers, lower spring, upper nut, upper bracket, and collet because of the material's inherent high IIL-4877 E-19 i l

I i

i

Enclosure Core ShroudStabilizer Desien Submittal strength and lower coefficient of thermal expansion than that of the 304L stainless-steel material used for the core shroud. The alloy X-750 material has been heat treated at 1975125*F followed by an air cool and age hardening process after machining to increase its strength. The annealing and age hardening processes used are the same as those used on the improved jet pump beams. As an intergranular attack (IGA) control, IGA testing is performed unless a minimum of 0.030 in, of material is removed after the last exposure to acid pickling or high temperature annealing. This material is certified to ASTM B637, Grade UNS N07750.

The Type XM-19 stainless-steel material used for the tie rods has a carbon content less than 0.04 percent and was solution annealed at 1950 to 1975 F followed by air cooling to a temperature below 800*F within 20 minutes of removal from the furnace. The Type XM-19 material was selected for fabrication of the tie rods because of the material's increased strength as compared to Type 316, the slightly lower coefficient of thermal expansion characteristics, and the material's higher resistance to sensitization. The Type XM-19 material used for the tie rods is procured in the solution annealed condition. The machined threads on the tie rods are not solution annealed after the machining of the threads. It should be noted that the process of air cooling from the solution annealing temperature for Type XM-19 material is not consistent with the Boiling Water Reactor Vessel Internals Project (BWR VIP) guidelines as provided in Section 5.10.7 of(BWR VIP-9401),

"BWR Core Shroud Repair Design Criteria," dated August 18,1994, where water quenching from solution annealing temperature is specified. Air cooling is used, because water quenching results in excessive warping. Air cooling maintains the rod straightness. This deviation from the BWR-VIP guidelines isjustified as described below.

General Electric has been specifying and using XM-19 components in commercial power reactors since the mid 1970, especially for parts which must be stainless steel and nitrided, a process which involves holding the parts in a chlorine-rich nitrogen environment at 1060 to 1100*F for 16 to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Only XM-19 is able to withstand this treatment without becoming sensitized to IGSCC. Types 304, 304L,316, and 316L quickly fail. Therefore, it is not necessary to remove or anneal any "IGSCC prone" layers after machining, because the material is inherently resistant to IGSCC, even in the as-machined condition.

General Electric has access to test results which show that Type XM-19, in BWR environments, is more corroslon and crack resistant in tight crevice situations than 316L or other 300-series stainless. To cite an example, a set of tensile specimens, each bearing a sleeve to create a crevice at the specimen surface, was loaded to  !

120% of yield in 1981; that load was maintained for the duration of the test. I I

Today,14 years later, none of the Type XM 19 specimens have failed or even I

IIL-4877 E-20

Enclosure Core Shroud Stabilizer Deslas Submittal cracked. Thus, concerns about durability of Type XM-19 in threaded joints because ofcrevices are not valid.

The Type XM-19 purchased for, and used to date, in BWRs is specified to be heat treated at 1950 to 2050'F for 15 to 20 minutes per in. thickness minimum, then water quenched to below 800 F in 4 minutes or less. Type XM-19 has experienced no known failures or other problems in approximately 20 years of BWR service. This is considered to be adequate confirmation that the material is acceptable for use.

Type XM-19 represents an excellent material for use in BWR applications. It may safely be used in the as machined condition; it may safety be used in threaded and other highly creviced geometries, and is superior to Types 316,316L, and other alloys for the intended application. Its only drawbacks are cost and availability, since it is not as commonly found in suppliers' inventories as the more traditional materials.

In the middle of 1970 in the interest ofimproving the margin of CRD performance, GE developed and implemented the use of Type XM-19 for piston and index tubes in place of Type 304 stainless steel. The logic was that as a low carbon, high chromium, mildly stabilized (Nb, V), austenitic alloy, Type XM-19 would offer a higher margin of resistance to IGSCC in the nitrided condition than Type 304. >

Nitriding involves heating to ~1100 F for several hours and results in furnace sensitization of 300 series stainless steels. As a side benefit Type XM-19 has a significantly higher strength than Type 304; therefore equivalent components are ,

stressed to a lower fraction ofyield stress in service. Since the late 1970 all control rod drives manufactured by GE have contained Type XM-19 piston and index tubes. This includes all BWR-6s (more than 1500 drives) plus several other BWR-4/5s under constmetion at the time, as well as all replacement drives manufactured since. In total, more than 2000 such CRDs are in service.

By the nature of the CRD design there are numerous crevices, including threaded joints, exposed to the reactor environment. On the average 10 to 20% of the drives at a given plant are refurbished each outage. During this work the drives are disassembled giving ample opportunity for examination and detection of problems. To date, no instances ofintergranular attack oflGSCC of nitrided Type XM-19 have been reported, i

Additional details relative to materials and fabrication are presented in Attachments 1 and 8.

IIL-4877 E-21

l

-l Enclosure I Core Shroud Stabilizer Design Submittal Pre Modification and Post-Modification Inspection 1

The pre-modification inspection plan to support installation of the Unit 2 core shroud repair consists of the following inspections:

1. Enhanced VT-1 inspection of accessible portions of the H9 weld at four Azimuth locations where tie rod stabilizer assemblies are to be installed.

Weld H9 is the circumferential weld connecting the shroud support plate to the RPV wall. An approximate 4 in, length on either side of the installation point ,

o is to be inspected.

2. Enhanced VT-1 examination of the accessible portions of the shroud vertical welds intersecting H4 (inside diameter and outside diameter) for approximately 4 in. above and below weld H4. Additional details are provided in Attachment 1.

The post-modification plan for the core shroud and the stabilizer assembly is currently under evaluation. GPC is participating in the development of the BWR VIP reinspection guidelines. In the interim, a reinspection plan similar to the current Unit 1 plan will be used. The two plans are slightly different due to design and configuration differences. The Unit 2 plan consists of the following:

Stabill:erAssemblies

1. Verify the bolt tightness on all stabilizer assemblies 6uring the first refueling outage, and verify the bolt tightness on each stabilizer assembly once every 10 years thereafter.

2. Perform a VT-3 examination of one stabilizer assembly each regularly scheduled refueling outage. Verify the upper and lower supports are in alignment with the RPV wall and shroud, and the mid-span support is in proper alignment.

Core Plate Wedges Perfonn a VT-3 examination on one of the core plate wedge assemblies at each regularly scheduled refueling outage.

, S_hroud Vertical I

Perform an enhanced VT-1 examination of vertical welds intersecting weld H4 l which have indications > 2 in. long. In addition, perform an enhanced VT-1 of )

one vertical weld at the intersection with weld H4 for 4 in, above and below, l 1

HL-4877 E-22  !

I

e-

] .

Enclosure l Core ShroudStabilizer Desier Submittal i cach refueling outage with future frequencies determined based on the results of the inspections.

t i

i i

F

)

l t

I r

l 4

HL-4877 E-23