JPN-98-040, Forwards Proprietary Rev 0 to GENE-187-30-0598, CRD Bolting Flaw Evaluation for Ja FitzPatrick Nuclear Power Plant, in Response to NRC 980831 RAI Re CR Bolting Relief request.Non- Proprietary Response,Encl.Proprietary Info Withheld

Forwards Proprietary Rev 0 to GENE-187-30-0598, CRD Bolting Flaw Evaluation for Ja FitzPatrick Nuclear Power Plant, in Response to NRC 980831 RAI Re CR Bolting Relief request.Non- Proprietary Response,Encl.Proprietary Info Withheld

ML20151Z573
Person / Time
Site:	FitzPatrick
Issue date:	09/11/1998
From:	James Knubel POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20138L021	List: JPN-98-040, Forwards Proprietary Rev 0 to GENE-187-30-0598, CRD Bolting Flaw Evaluation for Ja FitzPatrick Nuclear Power Plant, in Response to NRC 980831 RAI Re CR Bolting Relief request.Non- Proprietary Response,Encl.Proprietary Info Withheld
References
JPN-98-040, JPN-98-40, TAC-MA1764, NUDOCS 9809220027
Download: ML20151Z573 (14)

Text

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, 1 123 Main Street

• . - White Plains, New York 10601

t. 914 681.6950 914 287.3309 (Fax) s A NewWrkher a.< xa#8 i

& Authority s"N1"a*:,"

September 11.1998 JPN-98-040 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Stop P1-137 Washington, DC 20555

Subject:

James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 Response to Request for Additional Information Regarding Control Rod Boltina Relief Reauest (TAC No. MA1764)

References:

1. JPN 98-017, NYPA Letter to NRC," Request for Approval of Relief Request No. RR-14 to the FitzPatrick Inservice inspection (ISI) Program", dated May 8, l 1998.

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2. NRC Letter to NYPA, " Additional Information Regarding Control Rod Bolting l Relief Request (TAC No MA1764)", dated August 31,1998.

Dear Sir:

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On May 8,1998 the New York Power Authority somitted a request for relief from inservice I inspection requirements for control rod drive bolting at the James A. FitzPatrick Nuclear Power l c Plant (Reference 1). The NRC staff subsequently determined that additional information was (' I required to complete the review of the relief request (Reference 2). Attachment I to this letter I

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provides the required information.

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- E Enclosure 2 to this letter, General Electric Nuclear Energy Report GENE-187-30-0598, Revision 0, "CRD Bolting Flaw Evaluation for James A. FitzPatrick Nuclear Power Plant", dated May 1998, contains General Electric proprietary information. It is supported by an Affidavit signed by g'

J General Electric, the owner of the information. Enclosure 1 is the General Electric authorization M: letter and accompanying Affidavit. The Authority requests that the information wh.h is i l proprietary to General Electric, and is clearly identified in the report, be withheld froit public  ;

E disclosure in accordance with 10 CFR 2.790. ~he Affidavit sets forth the basis on which 'he l

,, information may be withheld from public disclosure by the NRC and addresses with specGicity gcy the considerations listed in 10 CFR 2.790(b)(4). The Authority will forward a copy of a non-5:n.t proprietary version of the General Electric Report to the NRC within 60 days of the date of this C W 1G E Ltr Eric!

, f op pga i p'h 4 fC+ku\i l i

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, letter. Correspondence regarding the proprietary nature of the General Electric Report or the  !

Affidavit should reference DRF #B13-01920-30 and be addressed to David Robare, Technical l Account Manager, Technical Services, General Electric Company,175 Curtner Ave., San Jose, i CA 95125.

Commitments made by the Authority in this letter are listed on Attachmerit 2. If you have any questions, please contact Ms. C. D. Faison.

Very truly yours, kapu /4MH f. bw 8. l 1 J. nubel l Senior Vice President &  !

Chief Nuclear Officer l

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Attachments: as stated

Enclosures:

as stated cc: Regional Administrator 1 U.S. Nuclear Regulatory Commission l 475 Allendale Road .

King of Prussia, PA 19406 Resident inspector's Office  !

James A. FitzPatrick Nuclear Power Plant

. U.S. Nuclear Regulatory Commission i P.O. Box 136 ,

Lycoming, NY 130931 Mr. Joseph Williams, Project Manager Project Directorate I-1 Division of Reactor Projects - 1/11 U.S. Nuclear Regulator / Commission Mail Stop 14 B2 Washington, DC 20555 Mr. Michael T. Anderson INEEL Research Center ,

2151 North Boulevard i

P.O. Box 1625

. Idaho Falls, Idaho 83415-2209

Enclosure 1 to JPN-98-040 l

Authorization Letter and Proprietary Affidavit in Support of General Electric Nuclear Energy Report GENE-187-30-0598, B13-01920-30, Revision 0, "CRD Bolting Flaw Evaluation for James A. FitzPatrick Nuclear Power Plant",  ;

1 May 1998 l

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New York Power Authority i JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 j DPR-59 4

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GE Nuclear Energy Kh1g gg StructuralMechanics & Materials Enggg. i 175 Curtner Avenue, AVC 747 l San Jose, CA 95125 Phone / Fax:(408) 925-5029/1150 HSM-9804 DRF #B13-01920-30 May 7,1998 l Mr. Robert Penny cc: B.D. Frew

- New York Power Authority, M/S 11C D.O Henry 123 Main Street E.G. Thacker White Plains, NY 10601 D.S. Braden 1

Subject:

CRD Bolting Evaluation - Final Report

Reference:

NYPA Purchase Order No. C95-Z0013, Change Order # 47, Task # 49.

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Dear Bob,

Please find attached a bound copy and one unbound copy of the report titled,"CRD Bolting Flaw Evaluation for James A. FitzPatricF Nuclear Power Plant,". 'Ihe final report has addressed all of the comments received by e-mail and through fed-ex package.

As requested in the comments, we have taken out all of the material related to leakage related examination requirements and have put it into a separate letter report format. 'Ihis letter report

> will be sent under a separate cover early next week.

'Ihis final report contains GE-NE proprietary information which is provided under the New York Power Authority /GE-NE proprietary information agreement. GE-NE customarily maintains this information in confidence and withholds it from public disclosure.

'Ibe attached affidavit identifies that the designated infctmation has been handled and classified as proprietary to GE-NE. Along with the afBdavit this information is suitable for review by the NRC. GE-NE hereby requests that the designated information be withheld from public disclosure in accordance with the provisions of 10 CFR 2.790.

Please feel free to contact me if you need further information on this report. With best regarA, Sincerely, Har Mehta Principal Engineer & Tech leader

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1 General Electric Company 1

AFFIDAVIT 4

I, David J. Robare, being duly sworn, depose and state as follows:

(1) I am Technical Account Manager, Technical 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 hee been authorized to apply for its withholding.

(2) The information sought to be withheld is contained in the GE proprietary report GENE-187-30-0598, DRF#B13-01920-30, CRD Bolting Flow Evaluationfor James A. Fitzpatrick Nuclear Power Plant , Revision 0, dated May 1998, Class III (The GE 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 fir.ancial information obtained from a person and privileged or confidential" (Exemption 4). 'Ihe material for which exunption 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, respectiwly, Critical Mass Energy Proiect v. Nuclear Regulatory Commission. 975F2d871 (DC Cir.1992), and Public Citizen Health Research Gmuo

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

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

a. Information that discloses a process, method, or apparatus, m

• eluding 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. Information which, if used by a competitor, would reduce his expendittne of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product; Affidavit Page i 12/13S3Rm

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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 cmtomer-funded development plans and programs, of potential commercial .

I value to General Electric;

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

He 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.

He information is of a sort customarily held in confxlence by GE, and is in fact so held. He information sought to be withheld has, to the best of my knowledge and helief, 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 tram..dttals to NRC, have been made, or must be made, pursuant to regulatory provis!or,s or proprietary agreements which provide for maintenance of the information in e.ario, ace. 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) De procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal sciatist or other equivalent authority, by the manager of the cognizant marketing function (or his delega:e), and by the Legal Operation, for technical content, competitive efWet, and determination of the accuracy of the proprietary designation. Disclosures outside GE are limited to regulatory bodies, customers, and potential customm, and theb agems, suppliers, and liceaecc.e, and others with a legitimate need for the information, and then only m accordance with appropriate regulatory provisions or proprietar) agreements.

(8) De information identified in paragraph (2), above, is cla<aified as proprietary because it contains detailed results of metallurgical evaluatiors of CRD Cap Screws.

Bis information was developed at significant expense to GE.

(9) Public disclosure of the information sought to be withleid is likely to cause substantial harm to GE's competitive position and foreciose ce reduce the availability

, ,,. m m Amdavit Page 2

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of profit-making opportunities. The information is part of GE's comprehensive i

BWR safety and technolcgy 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 methcds.

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

l 'Ihe 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 *o )

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

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

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i i-l ir/iisia m Affidavit Pace 3

STATE OF CALIFORNIA )

) ss:

COUNTY OF SANTA CLARA )

David J. Robare, being duly swom, deposes and says:

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

Executcd at San Jose, Califomia, this 7 th day of May 1998.

David J. Robare General Electric Company Subscribed and swom before me this 7 th day of May 1998.

< acomun natAuw >

enumssennoswe6 .

enemyness-omunes  :

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< . aeromanNYsume h bMbdblic, State of California --

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Aflidavit Page 4 12h3M3Rm

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1 Attachment I to JPN-98-040 l Response to Request for Additional Information Regarding Control Rod

., Bolting Relief Request (TAC No. MA1764) )

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l New York Power Autherity JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR-59

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f, Attachment I to JPN-98-040

-' Page 1 of 4 Response to Request for Additional Information Regarding Control Rod Bolting Relief Request (TAC No. MA1764)

Additional Information Reauired

a. Request for Relief No. RR-14 seeks relief from the additional examination requirements of IWD-2430 for control rod drive (CRD) housing bolts. However, the licensee has not established

' the burden of performing these examinations in accordance with 10 CFR 50.55a(a)(3)(ii). In addition, the licensee has not explained how tha proposed alternative, to replace existing botting with bolts of a new design, will provide an acceptable level of quality and safety,-

Part of the basis for relief explains that replacing all of the remaining 73 CRD mechanism bolts would result in approximately 18 person-rem exposure. It appears that this exposure will be accrued regardless of the time frame necessary to complete the replacement of all the botting.

Furthermore, replacing existing cracked bolts with ones of the new design does not provide reasonable assurance that the structural integrity of the remaining bolts (those in the expanded ]

examination sample) will be maintained. Have crack propagation rates for existing cracked bolts been established? Whtt is the schedule for completion of CRD bolt replacement?

Provide information to support the licensee's position that the additional examinations either create a hardship, or that the proposed alternative will provide an acceptable level of quality and safety.  !

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Response

1. CRD Bolt Replacement Schedule CRD bolts will continue to be inspected and replaced as part of routine maintenance of the l CRD's. All old design boltt will be replaced prior to the end of the third ten year ISI interval, l which began on Septemou 28,1997.

2. Basis of Hardship Normally ten to thirty CRD mechanisms require routine maintenance each refueling outage,  !

during which CRD botting is removed, inspected and replaced. Combining CRD bolt i replacement with routine CRD maintenance will result in icwer total accumule.ted exposures than if the activities were performed independently. Reptarement of the CRD bolts on the remaining 73 CRD mechanisms during the next refueling outage would result in exposure of approdnlately 18 person-rem. Future routine CRD maintenance would result in '

additional exposure beyond that accumulated during the bolt replacement activities.

Linking the bolt replacement with routine CRD maintenance eliminates the double exposure 1

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Attachment I to JPN-98-040  !

Page 2 of 4 i 2

that would be received while removing and installing the bolts twice, once for bolt replacement, then again for routine CRD maintenance. Also, if all remaining bolts ware replaced in the next refueling outage, when routine CRD maintenance was required at a

later date, the new bolts may have become contaminated and would require cleaning F before inspection, resulting in additional exposure that would not be received if the bolts

- were simply replaced during the maintenance activity. Combining CRD bolt replacement j with routine maintenance is expected to result in a reduction of approximately 18 person-1 rem in total accumulated exposure.

3. Flaw Evaluation

Crack propagation rates for existing CRD bolts have not been established since the crack i depth does not appear to be time or heat dependent (Section 2.9 of Reference 1).

j Assurance of the structuralintegrity of the remaining bolts is provided by analysis and testing performed by General Electric (Reference 1) and Structural lntegrity Associsites, ,

d Inc. (Reference 2) on CRD botting with crack like indications. The majority of the crack tips i were blunted, signifying that crack growth had subsided. The metallurgical evaluations ,

showed that the deepest crack found was 0.053" deep while the bolts have a nominal  !

diameter of 1". Many of the CRD bolts had been in service for approximately 20 years. At i j FitzPatrick only six bolts out of the 306 inspected had crack like indications and of the two '

bolts with the longest linear indications (longest was 7/8") the maximum crack depth was only 0.036". Again, the crack tips were blunted signitying crack growth had subsided.

f The General Electric analysis has also shown that even with uniform circumferential

[ cracking of all 8 bolts on a CRD mechanism, up to 0.15"in depth, ASME Code required i structural integrity is maintained. Based on the JAF inspection results, which showed that the deepest crack of 0.036" with a length of 7/8", there is a significant margin. The General 4

Electric Report (Reference 1) is enclosed for your information.

In addition, a General Electric analysis presented in Service Information Letter 483, Revision 2, "CRD Cap Screw Indications" (Reference 3), has shown that only 3 uniformly distributed bolts out of 8 were needed to support CRD loads and remain within the stress

-limits identified in ASME codes. This is reiterated in Reference 1 which states that "only three uncracked cap screws out of eight cap screws present at a CRD flange, are sufficient to meet Code structural margins".

Thus analysis and metallurgical evaluations have shown the following:

Crackinc ' c+ed at 8 plants showed that the deepest crack was 6 653" deep. The result of : + xamination showed crack blunting with oxide fille 5 crack tips signifying crack g: w " ad subsided. In addition, the crack depths were very similar, regardlee of the are of the plant and/or the heat of material of the bolts.

No crack within the population metallurgically evaluated approached the maximum analytical allowable crack depth.

Attachment I to JPN-98-040

, Page 3 of 4 The linear indications were relatively short, with the longest indication at JAF being 7/8 inch.

Fracture mecharics and Code required area calculations (per IWB-3600) indicate that CRD flange integrity is assured with required ASME Code margins even with 360 degree circumferential cracks up to 0.15 inch depth in all CRD bolts (Reference 1). The flaw acceptance standards which require no analysis, allow a crack depth of 0.071 inch (References 1 and 2). Thus, the maximum allowable flaw depth is 0.15 inch.

Only 3 uniformly distnbuted bolts out of 8 are needed to meet ASME Code margins.

Incidence of cracking found at FitzPatrick was only six of 306 bolts.

Assurance of the structural integrity of the new bolts is provided by an improved design and material selection. The new bolting material (AISI 4340) has higher yield and tensile strengths than the old bolting material (AISI 4140) and is more resistant to stress corrosion cracking. The larger radius in the head to shank region of the new bolts will reduce stress concentrations. Additionally, a new slotted washer design will be utilized in conjunction J

with the new bolting which will facilitate drainage and reduce the likelihood of stress corrosion cracking.

4. Summary Analysis and destructive testing has shown that the CRD bolt cracking does not affect safety since significant margin to ASME Code allowable exists and the population of bolts in which cracking has been noted is relatively small. The proposed a!ternative VT-1 examinations, with augmented surface examination if indications which exceed the acceptance standards of IWB-3517 are found during the VT-1 examination, and replacement of bolts with a new design more resistant to crs.cking, ensun , an acceptable level of quality and safety.

Based on the above information there is no safety significance associated with use of the proposed alternative.

b. The licensee has described a course of action for replacement of existing bolting with bolts of a new design it is stated in the Altemate Examination that "the new design bolts may be re-used after VT-1 and augmented surface examination determines that there are no relevant indications." However, the licensee'4 intentions regarding additional examinations for the new bolting is not clear. Additional examinations for the new bolting should be performed as required by the Code unless otherwise justified. Clarify the scope of this request regarding additional examinations to IWB-2430 and bolts of the new design that do not meet the acceptance criteria of IWB-3517.

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. j Attachment I to JPN-98-040 4 Page 4 of 4

Response

For both old and new design bolting the Authority will continue to do the VT-1 examination as required by the code, with augmented surface examination if indications which exceed the acceptance standards of IWB-3517 are found during the VT-1 examination. If new bolts are reused then a VT-1 examination, with augmented surface examination if indications which exceed the acceptance standards of IWB-3517 are found during the VT-1, will be performed prior to re-installation. The relief request applies to both the old and new bolting. Expanded examinations required by IWB-2430 will not be performed for either the new or old design bolts.

Justificatior. ' not performing the expanded examinations is provided in items 1,2,3, and 4 in response to question a above.

References

1. General Electric Nuclear Energy Report GENE-187-30-0598, B13-01920-30, Revision 0, "CRD Bolting Flaw Evaluation for James A. FitzPatrick Nuclear Power Plant", May 1998.

2. StructuralIntegrity Associates, Inc. Report PCR-97-025," Flaw Acceptar Standards for CRD Bolting Examinations at J. A. FitzPatrick", May 28,1997.

3. General Electric Service Information Letter (SIL) 483, Revision 2, "CRD Cap Screw Indications".

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Attachment 11 to JPN-98-040 Commitment List Commitment Commitment Due Date Number Replace all old design bolts prior to the end of the third 9/27/06 JPN-98-040 -1 ten year ISI interval, which began on September 28, 1997.

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New York Power Authority l JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR-59

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