ML20199K058

Requests Authorization to Perform Alternative to Requirements of ASME Section Xi,Subarticle IWA-5250.Util Requests That NRC Review & Approve Relief Request IWA-5250 to Support Next Refueling Outage Scheduled for 980411

ML20199K058
Person / Time
Site:	Grand Gulf, Arkansas Nuclear, River Bend, Waterford
Issue date:	01/29/1998
From:	Dewease J ENTERGY OPERATIONS, INC.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
CNRO-98-00004, CNRO-98-4, NUDOCS 9802060093
Download: ML20199K058 (6)

Text

,

,e Ertl*rgy Oper; tion,in).

== ENTERGY Zl:Ulbms Tei(41368 5760 s

Fax 601368 5768 Jerrold O. Dewease h port January 29,19C8 U. S. Nuclear Regulatory Commission Attn.: Document Control D,'sk Mail Stop OP1-17 Washington, DC 20555-0001

Subject:

Entergy Operations, Inc.

Alternative to ASME Code Requirements Arkansas Nuclear One - Units 1 & 2 Grand Gulf Nuclear Station Docket Nos. 50-313 & 50-368 D 7cket No. 50416 License Nos. DPR-51 & NPF-6 License No. NPF-29 i

River Bend Station Waterford 3 Steam Electric Staticn Decket No. 50-458 Docket No. 50-382 License No. NDF-47 License No. NPF-38 CNRO-98/00004 Gen +1emon:

Pursuant to 10CFR50.55a(a)(3), Entergy Operations, Inc. (Entergy) requests authorization to perform an attemative to the requirements of ASME Section XI, Subarticle IWA-5250. Relief Request 1S12-08 (see attachment) proposes an alternative to corrective actions specified in IWA-5250. The relief request applies to Entergy's nuclear units: Arkansas Nuclear One - Units 1 and 2; Grand Gulf Nuclear Station; Waterford 3 Steam Electric Station; and River Bend Station. Similar altematives have been proposed and aoproved for use at Wolf Creek and Callaway nuclear power plants.

Entergy requests the NRC review and approve Relief Request 1S12-08 in order to support the next refueling outage at Grand Gulf which is currently scheduled to begin April 11,1998.

hD k

i lWlylgpA.[A*'hh h

9802c60093 7eo129 PDR ADOCK 05000313 ilh

• P PDR UuJ.LJ1

,j'

Alt

rnetiva to ASME Cods Requirements CNRO-98/00004

-January 29,1998 Page 2 of 3 Should you have any questions regarding this submittal, please contact Guy Davant ~

at (601) 368-5756.

Very truly yours, J40 JGD/SJB/GHD/baa attachment-cc: (see next page)

~

IT

Altern;tiva to ASME Cods Requirements CNRO-98/00004 -

January 29,1998

- Page 3 of S cc:

Mr. C.' M. Dugger (W-GSB-300)-

Mr. J.-J. Hagan (G-ESC 3-VPO)

Mr. C. R. Hutchinson (N-GSB)

- Mr. J. R. McGaha (R-GSB 40)

Mr. J. W. Yelverton (M-ECH-65)

Mr. Ellis W. Merschoff Mr. William D. Reckley Regional AdmirArrator, Region IV NRR Project Manager, ANC-1 & 2

- U. S. Nuclear Regulatory Commission U. S. Nuclear Regulatory Commission 61'l Ryan Plaza Drive, Suite 400 M/S OWFN 13-H-3

' Arlington, TX 76011-8064 Washington, DC 20555 f

Mr. Kriss M. Kennedy-Mr. Jack N. D.nohew NRC Senior Resident inspector NRR Project Manager, Grand Gulf Arkansas 14uclear One U. S. Nucleae Regulatory Commission l

P. O. Box 310 M/S OWFN 13-H-3 London, AR 72847 Washington, DC 20555 Ms. Jennifer Dixon-Herrity Mr. ChandJ P. Patel NRC Senior Resident inspectc,r NRR Pro,iect Manager, Waterford-3 Grand Gulf Nuclear Station U. S. Nuclear Regulatory Commission Route 2 Box 399 M/S OWFN 13-H-3 Port Gibson, MS 39159 Washington, DC 20555 Mr. Jack Keeton Mr. David L. Wigginton NRC Resident inspector NRR Project Manager, RBS Waterford 3 Steam Electric Statior.

U. S. Nuclear Regulatory Commission WMSB 4326 M/S OWFN 13-H-3 P.O.BoxL' Washington, DC 20555 Killona, LA 'iOO6C Mr. George Replogle NRC Senior Resident inspeuor River Bend Station P. O. Box 1051 St. Francisville, LA 70775 l

I

p Sp'

' Altemativa to ASME Coda Requirements CNRO-98/00004

' January 20,1998 a

/Attaohment =

Page 1 of 3 -

4 L

' Rellef Request 1812-08:

System Pawssure Test Corrective Actions

<1,- = Code Requiremt nt

In a letter to Entergy, the NRC authorized the use of the 1992 Edition and portions of tt e 1993 Addenda of the ASME Boiler and Pressure Vessel Code, f

Section XI for the updated inservice inspection program at Entergy's nuclear sites',

ASME Section XI,1992 Edition,1993 Addenda, Subarticle.lWA-5250(a)(2) states that if leakage occurs at a bolted connection during a system pressure test, one -

~ bolt shall be removed, VT-3 examined, and evaluated for degradation in accordance with IWA 3100.

II. Requested Authorization Entergy requests authorization to perform an attemative to the code-required '

removal and VT-3 visual examination of bolting if leakage occurs during a system L pressure test of Class 1,2, and 3 systems.

Ill. Basis for Requesting Authorization.

Entergy believes the actions specified in IWA,5250(a)(2) are not always the most prudent course of action to determine the condition of the bolting and/or the root cause of the leak.

- A s' ituation which may be encountered involves a leaking joint following complete replacement of bolting materials (studs, bolts, nuts, washers, etc.). When the assen,ated system process piping is pressurized during pl ant start-up, the joint

'- Letter dated December 12,1996 from Mr. Williara D. Beckner, Director - Project Directorate IV-1,

- Office of Nuclear Reactor Regulation, NRC to Mr. Jerrold G. Dewease, Vice President - Operations Support, Fntergy Operations, Inc., " Evaluation of Entergy Operations, Inc. Request for Authorization to

~

Update inservice Inspection Programs to the 1992 and Portions of the 1993 ASME Boiler and Pressure Vessel Code,Section XI for Arkansas Nuclear One, Units 1 and 2, Grand Gulf Nuclear Station, River

Bend Station, and Waterford Steam Electric Station, Unit 0 (TAC Nos. M94472, M94471, M94454, M94473, and M94488)"

J

Alt:rnativa to ASME Code Requirsments CNRO-98/00004 January 29,1998-i Attaohment Page 2 of 3; leaks._ The root cause of this leakage may be due to thermal expansion of the

- piping and bolting materials and subsequent process fluid seepage at the joint -

- gasket. : In such cases, re-torquing the joint botting usually stops the leak.

Removing any of the joint bolting to evaluate for corrosion would be unwarranted in this situation if the bolting material is new. - ASME Section XI Interpretation XI '

z 1-92-01 recognizes this situation as one in which the requirements of IWA-5250(a)(2) do not apply.

Additionally,-lWA-5250(a)(2) does not address other factors which may indicate the condition of mechanical joint bolting. Entergy considers this requirement to g

be unnecessarily prescriptive and restrictive.

4

Other factors which should be considered when evaluating bolting condition at a -

leaking mechanical joint include, but are not limited to.

. Tjoint bolting materials service age of joint bolting materials-

e

. - location of the leakage -

history ofleakage at the joint evidence of corrosion with the joint assembled -

e:

corrosiveness of process fluid e

plant / industry studies of similar bolting materials in a similar environment -

IV. Proposed Altemative Examination-In accordance with 10 CFR 50.55a(a)(3)(l), Entergy proposes the following attemative to the requirementtof IWA-5250(a)(2).

When leakage is identified at bolted connections by visual VT-2 examination during system pressure testing, an evaluation will be performed. The evaluation will determine the susceptibility of the botting to corrosion, assess the potential for

? failure l and identify appropriate corrective actions.- The following factors will be-considered, as necessary,- when evaluating the leakago:

.1) Bolting materials

- 2)' Corrosivenese of the process fluid

3) Leakagelocation

Altem tiva to ASME Coda Requirements CNRO-98/00004-January 29,1998

--Attachment Page 3 of 3-

4) Leakage history at connection

5) Visual evidence of corrosion at connection (connection assembled) c

6) Industry studies and history of similai bolting in similar environment

7) Condition and leakage history of adjacent components Furthermore if the initial evaluation indicates the need for a more in-denth

- evaluation, the actions specified in IWA-5250(a)(2) shall be performed.

i Entergy believes this proposed alternative provides an eqt.lvalent level'of quality and safety when evaluating leakage and bolting material condition at Class 1,2, and 3 bolted connections.

V. _ Conclusion

-10CFR50.55a(a)(3) ctates:

" Proposed attematives to the requirements of (c), (d), (e), (f), (g), and (h) of this section or portions thereof may be used when authorized by the Director -

of the Office of Nuclear Reactor Regulation. _ The applicant shall demonstrate-that:

(i) The proposed alte' matives would provide an acceptable level of quality and safety, or (ii) Compliance with the specified requirements of this section would result in hardship or unusual difficulty without a compensating increase in the level -

of quality and safety."

Entergy believes that the proposed altemative, to use a systematic approach with -

an engineering evaluation, provides an acceptable level of quality and safety.

Therefore, we request the proposed alternative be authorized pursuant to 10CFR50.55a(a)(3)(i).

4

,