ML20137J899

Forwards Request for Addl Info Re TR BAW-2251, Demonstration of Aging Effects for Reactor Vessel

ML20137J899
Person / Time
Issue date:	04/02/1997
From:	Kuo P NRC (Affiliation Not Assigned)
To:	Firth D FRAMATOME
References
PROJECT-683 NUDOCS 9704040176
Download: ML20137J899 (5)

Text

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, f< _

4 y ' AprilL2, '1997.

Mr. David J. Firth ,~~

Program Director- >

> 4 ,

" ; Generic License Renewal Program ""

Framatome. Technologies,_Inc.

-P.O. Box 109351 .

t 4 !Lynchburg,' Virginia . 24506-0935

SUBJECT:

. REQUEST FOR ADDITIONAL INFORMATION IN THE CASE OF BAW-2251,. ,

. " DEMONSTRATION OF AGING EFFECTS FOR THE~ REACTOR VESSEL (RAI NOS.

L 18 THROUGH 26)-

[ DearMr.Firthi- 1 The attached is a request for additional information -(RAI)~ pertaining to the i NRC staff's review of BAW-2251,. " Demonstration of Aging Effects for the .

i Reactor Vessel." This request contains RAls number 18 through 26. 'RAls'l i

through.7 and 8'through'17 were forwarded in letters to Mr. Don Cronenberger >

!- dated August 28, 1996 and Mr. Dave Firth dated December 19, 1996, respective- a

~ ly. Your prompt' response to this RAI will' ensure that these aspects of the  ;

staff's review can.be completed.

' :If you have any questions concerning the attached RAI, please contact Robert

j. Prato of py staff at 415-1147.

l'

, Sincerely, L

M JM' Fo A P. T. Kuo, Section Chief License Renewal Project Directorate .

Division of Reactor Program Management

Office of Nuclear Reactor Regulation Project No. 683 ,

Attachment:

RAls- b l

, cc: See attached list JD DISTRIBU"LON w/ attachments: 7RD3 bOS I

'ContrahlNlet RPrato PUBLIC OGC i PDLR R/F ACRS DISTRIBUTION w/ attachments via e-mail: I JBirmingham (JLB4) GBagchi (GXB1) RJohnson (REJ) JDavis (JAD)

LShao'(LCXI)' EJordan (ELJ1) HBrammer (HLB)

MCase --(MJC) LLois (LXL1)- RJones (RCJ)

TMartin (TTM) MBanic (MJB) JFair (JRF)  ;

BElliot (BJE) JStrosnider (JRS2) PDLR Staff DOCUMENT NAME: A:\RAI3.LTR (B&W Owners! Group / AVL Disk)

" To receive a copy of this document, indicate in,fhe box:"C" - Copy without attachment / enclosure OFFICE ME:PDLR- /b & SC:PDLR e (ADE:NRR PD:P%S/ l al RPrato:avi A SNeddrry NAME- PTKuo PCI Ed R GVainas

, DATE: 3 /h /97 9 /3//97 )/31/97- i/V/97  :

[ OF ICIAL RECORD COPY-

6. ynp; . m nnq13 RETURB

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9704040176 970402 T0 R,ECUL/3T0g'[$EqTRAL FILES 97M PDR TOPRP EMVBW - -

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.p - l'a UNITED STATES

. j' NUCLEAR REGULATORY COMMISSION

• WASHINoTON, D.C. 2065 5 0001

%4,,, # -April 2, 1997.

Mr. David J.. Firth Program Director Generic License Renewal Program Framatome' Technologies, Inc.

P.O. Box 10935 Lynchburg, Virginia 24506-0935

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION IN THE CA,SE OF BAW-2251,

" DEMONSTRATION OF AGING EFFECTS FOR THE REACTOR VESSEL (RAI NOS.

18 THROUGH 26)

Dear Mr. Firth:

The attached is a request for additional information (RAI) pertaining to the NRC staff's review of BAW-2251, " Demonstration of Aging Effects for the Reactor Vessel." This request contains RAls number 18 through 26. RAls 1 through 7 and 8 through 17 were forwarded in letters to Mr. Don Cronenberger dated August 28, 1996 and Mr. Dave Firth dated December 19, 1996, respective-ly. Your prompt response to this RAI will ensure that these aspects of the i staff's review can be completed.

If you have any questions concerning the attached RAI, please contact Robert Prato of my staff at 415-1147.

Sincerely, 0 0)v~4 fo g P. T. Kuo, Section Chief License Renewal Project Directorate Division of Reactor Prcgram Management Office of' Nuclear Reactor Regulation Project No. 683

Attachment:

RAls cc: See attached list

l l

Project No. 683 Babcock & Wilcox Owners Group Gener-ic License Renewal Program cc:

Mr. Robert B. Borsum Regional Administrator, Region IV Framatome Technologies U.S. Nuclear Regulatory Commission 1700 Rockville Pike 611 Ryan Plaza Drive, Suite 1000 Suite 525 Arlington, Texas 76011 Rockville, Maryland 20852 Mr. James J. Fisicaro Michael Laggart Director, Licensing Manager, Corporate Licensing Entergy Operations, Inc.

GPU Nuclear Corporation Route 3, Box 137G One Upper Pond Rocd Russelville, Arkansas 72801 Parsippany, New Jersey 07054 Earnest L. Blake, Jr., Esq.

Chairman Shaw, Pittman, Potts Board of County Commissioners and Trowbridge of Dauphin County 2300 N. Street, NW Daughin County Courthouse Washington, D.C. 20037 Harrisburg, Pei .isylvania 17120 Regional Administrator, Region I Hr. J. W. Hampton U.S. Nuclear Regulatory Commission Nuclear Generation Vice President 475 Allendale Road Duke Power Company King of Prussia, Pennsylvania 19406 Oconee Nuclear Station MC: ONO IVP B. Gutherman, Manager P.O. Box 1439 Licensing Seneca, South Carolina 29679 Florida Power Corporation (SA2A)  :

Crystal River Energy Complex  ;

Mr. John R. McGaha 15760 W. Powerline Street i Vice President, Operations Support Crystal River, FL 34428-6708 l Entergy Operations, Inc.

P.O. Box 31995 William Dornsife, Acting Director Jacksonville, Mississippi 39286 Bureau of Radiation Protection Pennsylvania Department of Regional Administrator, Region II Environmental Resources i U.S. Nuclear Regulatory Commission P.O. Box 2063 101 Marietta St., N.W. Suite 2900 Harrisburg, Pennsylvania 17120 Atlanta, Georgia 30323 Chairman Mr. R. L. Gill Board of Supervisors 1 GLRP Licensing Coordinator of Londonderry Township l c/o Duke Power Company R.D. #1 Geyers Church Road EC-12R Middletown, Pennsylvania 17057 P.O. Box 1006 Charlotte, North Carolina 28201- Mr. J. E. Burchfield 1006 Compliance Duke Power Company Oconee Nuclear Site P.O. Box 1439 Seneca, South Carolina 29679

_. . _ . ~ __ __. _ _ _ . . . _ .. _ _

1 BAW-2251.'Reauest for Additional Information Nos. 18 throuah 26 Reaardina Anoendix B (BAW-2275)

18. Page 3-1, 3.2 Mechanical Properties of Weld Materials Provide bases for using a yield strength of 71 ksi and young's modulus of 27450 ksi for weld materials in the current low upper-shelf toughness analysis.

Include a discussion for not using a yield strength of 85.1 ksi and [

Young's modulus of 26975 ksi which were used in a similar analysis reported in BAW-2178P.

l 19. Page 6-1, 6.2 Limiting Level C and D Service Loading For Level C and D i conditions, meeting the acceptance criteria of Appendix K has not been l demonstrated for cracks with a depth less than one tenth of the wall i

, thickness (1/10T). Provide a quantitative analysis to demonstrate that i'

the analysis on a crack depth of 1/10T is bounding.

-Reaardina Accendix C (BAW-2274) ,

20. A General Question Demonstrate that PTS is not a concern for this additional class of cracks by providing the fracture toughness margin, i.e., the ratio of applied K to X from a deterministic fracture i mechanicsanalysisusingtheworsk,,transientfromthePTSstudyof1982 (the Extended HPI transient), which has a constant pressure of 2250 psi.

i

21. Page 3-4, 3.3.2 All Transient Conditions What kind of boundary condi-

! tions were applied to the FEM model of the hollow cylinder in the PCRIT )

Code to ensure that the model is simulating a whole vessel for all l loading conditions?

4 22. Page 3-5, 3.3.3 Discontinuity Stresses Provide information regarding the use of RV FEM results from a complete vessel model to estimate i

discontinuity stresses, with which the PCRIT results from a uniform cylinder model are combined to obtain the final stresses for various transients. Confirm that the discontinuity stresses have been consid-ered for pressure, thermal, and external loads. If the direct ratio of the RV FEM stress at the location t.f geometry discontinuity to that at the location of uniform section was applied, report the ratios for all loading conditions analyzed for the complete vessel model. If a bounding value was used to account for the geometry discontinuity, demonstrate that the discontinuity stress used is bounding.

! 23. Page 3-6, 3.5.2 Emergency and Faulted Condition Provide the document no. of the reference that contains stress analysis of the vessel under i= " deadweight + SSE + LOCA + interaction loads."

i

! ATTACHMENT

BAW-2251. Reauest for Additional Information Nos.18 throuah 26 (Cont'd)

Reaardina Anoendix C (BAW-2274) (Cont'd) >

24. Page 4-4, 4.5.2 Circumferential Flaw in Beltline and Nozzle Belt Regions It was mentioned that the Line-Spring Model by Kumar was used for the pressure stress loading, and the Raju-Newman method was used for the thermal stress loading. Confirm that the circumferential flaw is internal, and provide 'he reference of the Raju-Newman method regarding this crack geometry.

25. Page 4-8, 4.9 Determination of K It was stated that the stresses obtainedbyapplyinganequation,*/r,,omReference13byTimoshenkowere verified using computer code ANSYS which modeled both the vessel wall with clad and without the clad. What methodology, Timoshenko equation or ANSYS FEM method, was used in K r calculation for each fatigue group at the three analyzed regions *l d Is the ANSYS model for determining K the same as the RV FEM model mentioned previously? Confirm that 4 RNYe.grence 13 contains the solution for the vessel with clad as indicated in the report, and provide the coefficients of thermal expansion for both the vessel and the clad used in the analysis.

26. Page 5-3, Figure 5-1 Confirm that dK(a) was calculated based on the summation of K,,, Kir, Krdisc , and K,,i,g. Provide justification if it was not.

b

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